ORNL-2157.txt

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A ; This decument contains e

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* o S P : OAK RIDGE NATIONAL LABORATORY S o R 1N
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' ORNL-2157, Parts 1-5
’ o o . . -84 Reactors-—Specwl Features of Aircraft Reactors

< _‘"
. o _ This document consists of 330 pages.
_ " ' ‘ : | Copy Ql/of 237 copies. Series A,

t ) . = Contract No, W=7405-eng-26
i .

| AiééR}AF'f NUCLEAR-PRdPULSl-ON PROJECT
- | QUARTERLY PROGRESS REPORT
" ) | For Period Ending September 10, 1956
| o W. H. Jordan, Director

- o ‘ .' o _‘-f | L S. J. Cromer, Co-Director
A. J. Miller, Assistant Director

l.-’ -
- DATE ISSUED

DEC 13 1956

 

 

 

o ’ - T UNION CARBIDE NUCLEAR COMPANY ,
- A Divlslon of Union Carbide dnd Carbon Corporation

. o _ Post Office Box X~
5 o , -~ Oak Ridge, Tnnnessae

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[ED DATA

Y .defined in the Atomic
Energy Act of 1954, lts transmimel B ure of its contents
in any manner to an unauthorized pers®

      
         
  
   

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Reports previously issued in this series are as follows:
- Period Er_l.d_i'n_/g November 30, 1949 o
Period Ending February 28,1950

. ORNL-528

ORNL-768
ORNL-858
ORNL-919
ANP-60

 ANP-65

ORNL-1154
ORNL-1170
ORNL-1227
ORNL-~1294
ORNL-1375
ORNL-1439
ORNL-1515
ORNL-1556
ORNL-1609
ORNL-1649
ORNL-1692
ORNL-1729
ORNL-1771

 

~ ORNL-629

v

ORNL-1816”
ORNL-1864"
ORNL-1896V

ORNL-1947
ORNL.-2012

- ORNL-2061

ORNL.-2106

Vv’

 

Period Ending May 31, 1950
Period Ending August 31, 1950

~ Period Ending December 10, 1950

Period Ending March 10, 1951
Period Ending June 10, 1951
Period Ending September 10, 1951
Period Ending December 10, 1951
Period Ending March 10, 1952
Period Ending June 10, 1952
Period Ending September 10, 1952
Period Ending December 10, 1952
Period Ending March 10, 1953
Period Ending June 10, 1953
Period Ending September 10, 1953
Period Ending December 10, 1953
Period Ending March 10, 1954
Period Ending June 10, 1954
Period Ending Septémber 10, 1954
Period Ending December 10, 1954
Period Ending March 10, 1955
Period Ending June 10, 1955
Period Ending September 10, 1955
Period Ending December 10, 1955
Period Ending March 10, 1956 |
Period Ending June 10, 1956.

43

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ORNL-2157, Parts 1-5

C-84 - Reactors=Special Features of Airéruft Reactors

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9. E.P
100. G.C.
101. J.C
102, C. E

Whitman

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ilson

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Willioms
W
Winters

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130-131.
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137-139.
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121-123. Central Research Library

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FOREWORD

This quarterly progress report of the Aircraft Nuclear Propulsion Project at ORNL
records the technical progress of the research on circulating-fuel reactors and other ANP
research ot the Laboratory under its Contract W-7405-eng-26.. The report is divided into
five major parts: 1. Aircraft Reactor Engineering, 2. Chemistry, 3. Metallurgy, 4. Heat
Transfer and Physical Properties, Radiation Damage, Fuel Recovery and Reprocessing,
and Critical Experiments, and 5. Reactor Shielding.

The ANP Project is comprised of about 550 technical and scientific personnel
engaged in many phases of research directed toward the achievement of nuclear pro-
pulsion of aircraft. A considerable portion of this research is performed in support of
the work of other organizations participating in the national ANP effort. However, the
bulk of the ANP research at ORNL is directed toward the development of o circulating-
fuel type of reactor.

The design, construction, and operation of the Aircraft Reactor Test (ART), with
the cooperation of the Pratt & Whitney Aircraft Division, are the specific objectives of
the project. The ART is to be a power plant system that will include a 60-Mw circu-
lating-fuel reflector-moderator reactor and adequate means for heat disposal. Operation
of the system will be for the purpose of determining feasibility and for studying the
problems associated with the design, construction, and operation of a high-power circu-

lating-fuel reflector-moderated aircraft reactor system.

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CONTENTS
FOREWORD ......ooocceeerreessesesessssscssesssessassssssesessessesesssssesseneesossnes SRR e vii
< SUMMARY ...c.cccenrn e S, ]
PART 1. AIRCRAFT REACTOR ENGINEERING

1.1. AIRCRAFT REACTOR TEST DESIGN .....covevrrmeiiieirrnirnnnenae eeeerereeeeesreeeaereertebtenaereseteanerneas 17

Status of ART Design....ccoeccerererreenincmniniinsnienan, ettt ettt a et et A s s bt eb e be A eras b es s re e sn s s saees 17

Reactor Assembly ......ccmvereevernienriinniicicsnsinineenaes e e areseaas eeeveseaenenetereeetansaaasntnes 17

Reflector-Moderator Cooling Circuit .. ivirmmsmmriincreerericnisssitsssissssssss s 17

Heat Exchangers .......ccccvcennniccricnnncinnsennisnnns eeutetebee st eaearte st seaRent s a et et s e et s e st e e eae w17

Fuel Pumps and Expansion Tank.........cccoiininiiiiitneseneseiese s s 17

Fuel Recovery Tank......ccoovvrrerrrenccencicninnencans eeestetetes et e bttt esensere R e bt eseanatares et 17

Applied Mechanics and Stress Analysis ............. e tai R AR AR 18

Reaetor SUPPOIt ...ttt ssree st mass s st st ne s a s st s s nesans e es e eaneenes 18

NaK Piping Inside Reactor Cell................... rerereenens eeteuereearebatesaeaereseserasstetebetasastasesaeaeatassstteearare 18

Fill-and-Drain Tank and SUPPOTT .......ccviireierenrictcnerreeecces st sss s snenssssssasssasssssases 21

Core-Shell Low-Frequency Thermal-Cycling Test................. eveseerensereettebatesatate s et assesensranesneater 23

 Core Hydrodynamics ...t et b b s s sa st 26

- 1.2. ART PHYSICS ..ot eteeeeeeseattaeatestasssas et h A a e R e R e e e b e bR e s b ettt nb s s 29

Comparison of Basic Gamma-Ray Data with Bulk Shielding Reactor Gamma

5 Heating MeaSUIEMENTS .....ccoruiccrrrecresiecesircenseenseraseressesssssssassansses s s ssas s esb st ses s assssasssss st sesssnsueres 29

- Gamma-Ray Heating in ART ........... eevensrereneaenns errenenene ieereereeeebeeetesstesetateseeteteaner s et st eresnsasa bt b 31

GAMMA-Ray SoUrce SHENGHh ..ioviiecerrcererreecerseeneseserssessesiasesssssesssssssssssssresssssss s smsbessssssassasassasess 31

Gamma-Ray Heating ......ccovvcvnveimemminniininnn, enurpibine s ssatssnssesens reebeeeetebeertete e e asae e s ssanens 32

1.3. ART INSTRUMENTS AND CONTROLS ........ccccruveneen. ceeeteeetetiessasabenebete e e e At e s e aeanre e asesneneneater st bebetn 35
Reflector-Moderator Temperature Control Simulation........ocecumrecmmiveimiiisnniinssiiesns e 35

Instrument Development ......ccccevvcrreinrnciniesscnerineanen. ereeeeeatetesatasasararaserarasababesetassataeaebeteae et e nsaseasaneen 35

Fuel-Expansion-Tank Level |ndlcafor ............................................................................................ 35

Systems for Testing Liquid-Level-Sensing Devnces in Flowmg Liquids..ccoecnrnrncneenccenencnnas 36

High-Temperature Turbine-Type Flowmeter ... nsisssesssses 36

High-Temperature Pressure Transmitter ....cooveeencicccnnninnannenecs evererearaeserbasastetar s easnarasasn 37

Tests of Heliarc-Welded Inconel-Sheathed Thermocouples in Sodium ......covuviiniiiiniinininiinnnn, . 37

Thermocouple Data Reductlon Coiesessarassasasaissseesseissssatesnestasstereseseeastsesrsssssnats et . 37

1.4. COMPONENT DEVELOPMENT AND TESTING.........._ ........................................................................ 39

Pump Development T eSS iorimrersimseeressesssesssssssssasessens pebset e e sae st s b a s et an s reen rerrrsensnaseserenes .39

Bearing and Seal Tests ........... eeeeerereresr et ata et e e s s ses s bR e s srnees revetereeesae e e et e sbearaeaten 39

.- Test of Gas Attenuation by Seals seiesibsireresseietiressntse s et R s et et be s e st R et bR SRR E b e R e s e P SRR SRS 00 39
; Fuel Pump Development Tests....ccocininninnniiine. ettt es rrnersessessoseseasiiibababea s sr s s s ansens 40
‘ Fuel Pump Endurance Tests .irennnnnnnessnisiessesesascnsannnssncnns sesssemssssessinessarsssnsssisasnenias 42
e ~ Sodium Pump Development Tests .c..rireimnrmsnisscsserssssssisssssnsssssssrsesssersens reensveesereststesesesesteneneas 43
| Primary and Auxiliary NaK Pump Development Tests covnnnrrrrenens et st 43
l Heat Exchanger and Rodiator Development TEStS ............ccuwrmmmmimsssssessesssssssssmssssssssssssssssssesessseees 43
N Intermediate Heat Exchanger Tests................. eerst bbb se st sba st reen et sensiaes 43
Small Heat Exchanger Tests....cooeevreninenininnicsieiiiesinnenens rereereerenns evreerrereetesetesaaaransens 47

Cold-Trap Evaluation in Heat Exchanger Test Loops ...t 51

 

 
 

 

 

1.5.

1.6.

1.7.

2.1.

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Water Flow Tests of Aluminum North-Head Mockup..... et st 52
Dump Valve Development Tests ....ooeovecncccenncnne Cemeeeeneseedaess s iviseeeeenesieessirenns 53
Outer Core Shell Thermal Stability Test ..... 55
Auxiliary Component Development ........ccocooneerccncrrcnnnnene rereterenetarneserer s re s ea sy tassesansenors reesesnssaessansen 56
High-Frequency Thermal-Cycling Apparatus ......ccouiremvinermrssinssisinssississiistssetsssesesssessseseseser - 56
Cold Traps and Plugging Indicators ......ccvvccnverecnnrisninensennen: cnsnsnsanes reerneesearieraras rereresnes - 57
Liquid-Metal-Vapor Condensers ..........ccceemeercennnersnnesennieescsiseessseresesesssssssssssesssssasesasaesssssnssassenss 60
Zirconium Fluoride Vapor Trap ....cccocccinvmmnnncnininnsnenincnneressssnsssencsenes resreniossarsasensesasninsssrseare 61
PROCUREMENT AND CONSTRUCTION .....oouvireereisassesssesiessms s sasessssssssressssesssseressnnes reeverensee S 64
ART FEITIY wecverresrrseneeseessssssessssssesssssassasssssesessassssssssssssssseassssssesisresssesessassssssssssssescsseneers v 64
ART-ETU Reoctor Construcflon reesratsnsaeae st gttt re s s st etseresnaresanaseeres prerruenansnanes ereresrereensennnessesesesesss 18
ETU Facility oot crcrssscesiessss e snessssesnssesssesens eeseenereranannees veresersnsenensrnitearsassnons 78
ART-ETU Reactor Fabncahon and Assembly ...... ebensamsarersaeasasptste et sbssrantitsrsasnassrerenansne resesseennene 18
“ART Cell COMPONENES ..cuecvireenecerversssesiisesensaeressassssssesesssnssossssesssassons rreerersesseenes 79
ART-ETU Reactor Component Procurement .........ouo.vccciveesecnorssnnaens eeiereretsae sttt r s s sen e sasbene 79
' Fuel-to-NaK Heat Exchangers ..........cuerenncninrcsteeneernsssseescsessssssesensens ereretensnenesbeseserenns 79
Sodium-to-NaK Heat Exchangers ...t reversensaressasesrsasans 79
Core Shells ..ottt st sassseresasseneesesassessonsrenasserasnens treeaneateneemaeneanaanastenasresens 79
Beryllium Reflector-Moderator ...........ciinmicnincncnemssmsssssessess 80
Boron Layers.......ooinieiecnnicienienisiaons reemsaseeesanresereasessass eeernerse et e rmensanssrreas rene 80
Pressure Shell... e sssesesesessssassnesees eereseesasasneneaes aerneseanisaenen eenenens 80
FRCONEL ettt et se e e ra s s s s s e n e nsase e SRRt s et b s b SRR bR 80
IN-PILE LOOP DEVELOPMENT AND TESTS.........ccooosmmmsercesomssrsssmsssssssssssssssssasessnes 1
Operation of In-Pile Loop No. 5....coeuereneneneen. s a1 R4 e s RS 81
In-Pile Loop No. 6 e ceeeeenne reteeeereneennanens eeeseeesesarerteresnerrrensettenresenesaasereastsbteraanes 81
Horizontal-Shaft Sump Pump for In-Pile Loops .......cccovreeerereeercrnrenennnnnens st snsisnsnenesesgassons 81
ADVANCED REACTOR DESIGN .......covceeeireiricrrssressessesaeseennes . ....... st snses veerenenes 82
Graphite and Beryllium Oxide Moderated Circulating-Fluoride-Fuel Reactor ...........cuiveuiureccernee .. 82
Hydrlde-Moderated Circulating-Fuel Reactor - 84

PART 2. CHEMISTRY |

PHASE EQUILIBRIUM STUDIES ....cvroooovvornecmecnssecnscssssesnmessesns s eeeesrsrersssneesmssessssnseneess 8T
The System LiF-RbF ..o veeneenens eeenenesene e sneanes tmserarannasonssaggresgnessienin 87
The System LiF-UF, ..o s emee e esesern e ren s eneemrasaeee peemunsaesaonsasin 87
The System RbF-UF, ............... et erdsbe s anrn s bbb e siee e siens 88
The System NaF-RbF-UF, .........cccrmmrsmreercsreersene eesensn e snsisn st 90
The System NaF-RbF-ZeF [ -UF  ....ccccoeeorresnncesessnsessressssss s ssmsssssssssssssssssssssssssssissssssses — 91

The System RbF-CaF, o, rrerrenesreanareeraeetisaeses dersrsenses reeterteesrnneserarsarenaasans 91
The Alkali Fluoride~Cerous Flucride Systems ........ccocovccreviunnecirnnnnne ,f’;’{ ........................................... 92
The System CSF-BeF  .......cooommmmmmererrcessmmmasnermeessemsnsnesssescecs ersurssn st RRSRS R ORARSO 92
The System NaF-LiF-BeF , c.c ettt sss s st s s 93
The System NaF-RBFBEF 5 oottt - 94
The System KCI-ZrCl, .....ovmrmmersssssssmsnssisssssomsssssssssississsnsssssmssssssmsssssmssssosssssssssssssmssosassssirs 9

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2.2

2.3.

2.4.

2.5.

2.6.

 

 

CHEMICAL REACTIONS IN MOLTEN SALTS ....... et bRt er bR n R A bbb aee e s eas 96
Equilibrium Reduction of N|F by H, in NoF ZIF firermssssrmossssscssssesmonssrasstesmsasssnenassssssssecs weevrersaers 96
Stability of N{F2 in NaF-ZrF4 Mixtures ......ccocevevvrnane. veveerrrenens ........ 97
Activity of Chromium in Chromium-Nickel Alloys ... 100
Solubility and Stability of Structural Metal Flucrides in Vanous Fluoride Mixtures ......co......... 100
Reduction of UF , by Structural Metals ...ttt 103
Solubility Defermlnatlons by Measurement of Electromotive Forces of
Concentration Cells .................. eeebsteasuenireanrsstetebontaeataeaasesstenenasbeberteastiseranansenseseseesasnsasans reerorsnenesnsanne 106
Activities in the RbF-ZsF | System e 108
Activities in the KF-Z:-F4 SYSEEIM ettt r e et e st ene e sbe e s ae et e st s ar et e aa b e e aeten 109
Solubility of LaF ; and of CeF ; in Molten Fluoride Mixtures ..o, 110
Solubility of Xenon in Fused Salts ...ttt se st sse s sesn st e e snesane 113
Molten Salt Polarography .......c.ccveiniecrncieecerrcnenannns e etreeeeterebeteieattereretetehes sttt bebaresesear st sebetetn e 113
Activities from EMF Measurements on Solid Solutions of Salts....c.cccoooiiiieicenniciriccceeenis 116
PHYSICAL PROPERTIES OF MOLTEN MATERIALS ...t s s 118
Surface Tension and Density of NaF-ZrF ; Mixtures ...ccoovrierrrireinnieennen e evenreneenererereas 118
Surface Tensions of Molten SAlts .t es 118
Fluorozirconate MIXBUTES .......coccoiiiereeieee e es st sttt sare e s e se s s ses e s et 118
Uranium Tetrafluoride .......coveeeeeerccrerreecrceereene, e eeterteetenrerearteintteare e e et et ereesteretensan et e e reneereereans 119
"PRODUCTION OF PURIFIED FLUORIDE MIXTURES ....ocmmeeseicirecteeeseesras s eeenesenssestasenseensseans 120
Laboratory-Scale Purification Operations .........cceereiorinieseneesnssessensascessssssscssesssssmmessesesssssssanss 120
Quality Control of Raw Materials and Products ..........coomeeeeeerererssesnersssessmmsensesssssmsssesssssseesssssssanns 120
Pilot-Scale Purification Operations ... iesiossvesessesssssssssssssssssssessasssessssssssssessssnses 120
Production-Scale Operations .......cccceeeveeemiresieinneesseesstessssssesssessrsssssssassssemssssssssnsassssessssesessansasesnsasas 120
" Batching and Dispensing Operations .................................................................................................... 121
Preparation of ZrF from ZrCI ................................................................ 121
SPECiGl SEIVICES uueurenirirerrrrnisereiereeesenesessnsisssessesssenes eeeeeeeresseseesemesesesa e sseesssseeroen rreeteesbsasasaen 121
Filling, Draining, and Somplmg Opefuhons eteeieetereesnesreseeraebeveeae it s et ebea e st a e e b et e s e te s ea s s g eraaes 121
Special Enriched Fuel for In-Pile Loops .. ssa s sessaas s s 122
Shield Mockup Core Materials ..o rieieeiecrreraeisetesiesesenaesivesssssisssssessesassssessssssssssssssssnssnsans e 122
Experimental Preparation of Various Fluorides 122
COMPATIBILITY OF MATERIALS AT HIGH TEMF’ERATURES ceveressresas s anbeneaas — 123 .
Penetration of Graphite by Molten Fluarldes ...... etvre e e s e e e bt s srasasae spess ettt reneranes .. 123
Hydrogen Pressure of the NGOH-NI Reactlon et seeeaane rersieiaeeensasntsasasisanserne . wieeirensissentsessennsenenes 123
'Physical Properties of Elastomers Exposed to Attack by Liquid Mefals ...... eenssmsssnsssresssnerenensene 124
‘Determlnahon of Oxygen in NuK .............. Laressssaninssisd s anssssanass sesssssmmenssssanamaen - eresstseiss e sasntrass Y
ANALYTICAL CHEMISTRY .o covooesressoesssssssasssossssssssssssssss st s ssssssasessssssssississssessses 125
Determination of Rare-Earth Elemenfs in Fluonde Fuels ................................. et 125
Spectrophotometric Determination of Cerium .........oovueveeeeerercrerereresnesessisestiessscnensenens reerreenenes 125
Spectrophotometric Determination of Lanthanum ... 125

xi

 

 
 

 

 

 

 

3.1

3.2,

3.3,
 Nickel-Molybdenum Alicy Development eeeteeeessessees e e et se s easesasseresaaes et eaesresbenens eeeraresnssreressesnans

xii

Determination of Tantalum in Fluoride SIS ... iiiiieciecittinnnitscsssiesesssssesrssssssseissssssnsaesansersans

Determination of Oxygen in NaK ... ssesssssesssesnes reveneeneenes eree

- Sampler for Alkali Metals ... res s eseos

Determination of Oxygen, Nitrogen, and Carbon in Metallic Lithium ......ccccccvvvnncennncee.

Determination of Oxygen ..o ceiiccenrenienenessenisnsereresesaesssasassnsssssosssessassenes retenraensaensnras

Determination of Nitrogen and Carbon.........cccvviccniinnniisnninnsisnisecerserenenens vrenssensassases S
Detection of Traces of NaK in Air ........ccu.......... et e R s RS cresrrrnssesaas

Compatibility of Fluoride Salts and Alkali Metals with Pump Lubricants ...ccconiivcieinirciennne.

Determination of Chromium in Trioctylphosphine Oxide Extracts with

DiphenYI CGrbaZide F e R e E e e R e ettt r e et e ra R et blasLuRiT IRl adtisRtIRsaIIa .--oocooaun---hl----n---; -------------------- ]

ANP Service Laboratory .....ccooeerreecencvesneeenenencenenes reeereeere e e arasenes eevnsnassasesasssasns reeseerranenes ereeranane

DYNAMIC CORROSION ..covervtrrrvsmerssesssensssisesnsnisesnes e

Forced-Circulation Loop Tests .....eciecrnininierenieccnsessinisassssssessseresesessensnes cevesusesreasseasanessasrannans

Fuel Mixtures in Inconel and Hastelloy B ........cccoviiiincicnnnnnnns reeeereeaas eresereserrrerasersneasssrassanens '

Sedium in Inconel and Stainless Steel...........ocvveriinnnnnnnn. aseereentn s rn s stessassnast e sa s rus b S RA S st 008
Sodium-Beryllium-Inconel Compatibility ....c.ccccoeimveverirninirinnenininnns ceeeesstetes st sreranenans R

Thermal-Convection Loop Tests .c....cciiieeecicerecncresncsnenns ereerrerereanaeeeanesineesaseses ceeesseisenaerainine
Sodium-Beryllium-Hastelloy B Compatibility ............... vernrrenns ceereteresesesesatentanesserenenasenesierasaesrens ‘

Sodium-Beryllium-Inconel Compatibility ......cccriniiveimsinniinniinccisn s s
Fuel Mixtures in Hastelloy and Special Nickel-Molybdenum Alloy Loops............. rersnsinnensoserae
Special NaF-Z¢F ,-UF Mixtures in Inconel Loops ... irrrnnc st snenssesesssssssssesssees
Screening Tests of Special Fuel Mixtures ... e

INCOnE] SHrAinN-Cycling TestS.vneinriirenerereeeseseissesssstsisassssstsnssesssassessssasssassesessssesisabesesntseessssensassasnens

GENERAL CORROSION STUDIES ........covetirenrcerennnereresesesessssasssssssssasssenes
_ Tests of Inconel Tube-to-Header Joints with Recrystallized Welds in NaK and

IN FIUOride FUuel ...t v st s s sase e s e b s b a s sm e st et
Tests of Coast Metals Brazing Alloy No. 53 in NaK and in Fluoride Fuel .cccovvvrennnee Lovrssreraseass
Tests of Haynes Brazing Alloy No. 40 in NaK and in Flucride Fuel ........cccoveciicnnnnennnenesennncn.
Niobium in Sodium and in Lithium ..coee........ rveeesertet et tetaanbebens s s aneas _
Vapor-Zone Attack in Inconel-Fluoride Fuel Systems ......oomicieieiinsivciniinninsnsinsniesnssnsssennenes

Carburization of Various Alloys by Molten Sodium ..ot srores

Sodium-Beryllium-Inconel Compatibility in a Static System (Test No. 2) ..cccocvrmrernccsinrerecenenne. |

Effect of Diffusion Cold Traps on Mass Transfer in Inconel-Sedium ‘
Thermal-Convection Loops .....uiircicnireneninnssessssesess s sns st cssssensesssesesssassssessnasasarens seesssnnes

Thermal-Convection Loop Tests of Various Structural Materials and Lithium crasssensesrrassssasastrsssens

Tests of Intermetallic Compounds in Sedium and in Fluoride Fuel cerveeeesieseeseeress s e asene e st eaneeteens

FABRICATION RESEARCH ..ccovieserseereensssepsssesissesesesssssessenessse st s sesssssssssssssnss s

Battelle Memorial Institute ALIOYs ........cccceeeeenreserernreneerensnsesisiisesesstssnensssssesesesssasssssasnnsssssssssssns

152
153

1}

iy
 

 

 

 

T

wr)

3.4.

3.5.

- 3.6.

3.7.

 

International Nickel Company AlloySs ..c.ccooviveerioicrnneiecct s e reetrennine . 176
ORINL AHOYS .ottt ee s st atenesnsssessrasanessesesasssasasassestessonsn eteeereenesnnresnssessensisersases 170
Hastelloy W Seamless Tubing.......... ettreeteeabiarebereeete ey e re ke ehe b e s b et e be e e e neetRe et s s e bt en e e sate e bR s EsRe e 177
Consumable-Electrode Experiments.........ccovcmicnnninnissnssss 177
Shield Plug for ART PUMPS c.ccoveserrecsereecmsressneesssssesssssesisssssssssssssssssssssssssssssssssssssesssss s s w177
Gamma-Ray Shielding Matenal .................................................... trvevesranererrenensanes eserssenrerssmansaaraents 177
THEIMAl SHIEIA .ooveeeeecre ettt e ersas b e ss s s an st s e eas seasrae st bsbsseneasane 180
Neutron Shield Material for the ART ..ot eeerraereeaebetere e e enseannns 180
Ceramic B, C Tiles oot reveestessasesseressesereseiateteeeaaasaeaetere et st sRsebesnebase s 180
Clad Copper-B4C COIMELS ..c.cevereeericrs e snecoressre e b s bbb srm R bbb bbbt b 180
Lithium-Magnesium ATLOYS 1ooveoresees s iees st v sssss s sses s ssse s es e bes e s aR SRR 180
Tubular Control Rods ......cccverunen. esessssansssssrmssesnees rerietebetastasbet st e b sa st e R s be et ere e b st R bR R 181
Seamless Tubular Fuel Elements ceereeeeneeenee e ieerb b st e e R e ba A b R b 182
Niobium Fabrlcahon Studies ..ccerrrneenene reerueseessteseereetesbeteaseaaaeanseatraeaaaeet e e eRaee et e s e e naseesanenbarbe b sorsnaeas 183
Evaluation of Arc-Melted Niobium .....c.ooiviiriicrscinceneeese et st sasees 183
Nb-UO, Compacts.......crerrsmreernnens eeteesveseeeeneesaaettsasienerent e et b st e oReR st e e e A eR TR RO A ae e ehaas et e eseREatese et saeranen 183
Nb-U Alloys ........................................................................................................ 183
Fobrlcahon of Hydrides ..o 183
WELDING AND BRAZING INVESTIGATIONS ...oooooos oo esecmemenessssessssssmseesesssssssmssesssnssess s 184
Fabrication of Primary NaK Pump VOIUBES wvvvvvvereeseeesceesesesessseeeeeeesssesesesesesesasesssessssssossssmssesnesenssess 184
Fabrication of Primary NaK Pump Impellers ... 192
Shrinkage of Inconel Core Shell Welds .........occcveciiniiicicniice e 192
Examinations of NaK-to-Air Radiators After Service ... 195
Fabrication of Cermet Valve Components .......ccoiicrinniiiniininccinine st sesersssnssssesssssssessns 197
Continuous Production of Sintered Brazing Alloy Rings ..oveeeeceicciicee eeraeeeanas 200
Fabrication of Heat Exchangers and Radigtors ... 201
Studies of the Aging Characteristics of Hastelloy B, 202
MECHANICAL PROPERTIES STUDIES .....oocovv R e e 205
Relaxation Tésting of Inconel.......... eeeeseeee it taen et r s s sttt R et e R R st enees reeneesestetennanssasasanane 205
Effects of Fused-Salt Corrosion on the Siress-Rupture Properties of Hastelloy X .....ccccovvnnnneee. 206
| Creep-Rupture Testing of an 80% Mg-—20% Li Alloy .......................................................................... 208
CERAMIC RESEARCH ............. rerereesenas eeeesesesssiereniensasains rrrerreeneanes rerenrens pevseseaessrenaden s sanesnaces vereeeennieee 211
Rare-Carth Mutenals ................ reeeeeeene 211
Fuel and Moderahng Matermls ettt eninn s snsaaes . ..... eesaeeensenetaanens 21
Component Fabrication............. eeesennenaes ~.....;._.’ ...... rereseesieessenestesananens s reveresessssvenesosssinssenns 211
High-Temperature X-Ray lefructlon Vesresiasesioieans Vivassasemsnsissasaseenitieattisas s st e et b e o sRsiabara bt an s 211
Flyoride Fuel |nvest|gcmons...'.r_......_...'..'.'.'........' ..... s ' ............ 211
NONDESTRUCTIVE TESTING STUDIES ....cooemrrrenrerennnisnsssessesssneseennee eeesenes s reevemmeeeessaens 212
Eddy-Current Inspection of Small-Diameter Tublng SO UT 212
Inspection of Tubing by the Ultrasonic Method ....... eveerestetetebererats et sersRoReanbed s At st et st et s en s s e se e s ben 212

xiii

 

 
 

 

 

 

3.8.

4.1.

4.2.

xXiv

 

 

 

 

VLUHLE -~

 

218

INSPECHON OF PiPE .rveruriierercerieseensesesenessssnssessssessasmesssemsssnssasssssesss s e ssons besassastassssassans shemssassssssessases
Inspection of Thin Sheet .....cccvvninieessininennncrcssnneesenseresierssssesensas peesmeresressaseerserensenens revirane yereesenannas
INSPECTION OF MATERIALS AND COMPONENTS.......cccccvieeenvnernnens ........ ereenneieenennesnes |
Weld Inspections ..ceeeuercemrreetnrscncn s sissascnanes ceeeeseeeensenenneeretase s ss s R benaes
Material INSPECHONS ...ttt s s b sasn s s san et s s eeeresssesessenesesentans
InSpection of COMPONENES .....ccerueerinessnsesessssacsssssessssssesssssssiess erereseesnetenaereinertesasatenaseaneres veereereens -
Fluorescent-Penetrant Inspection of Tubmg..,...; ...... .......... ersss s e e st R0
Welder Qualification Program st ersesersssees eeeereesiseraiin st rsssmsmiernas e egedisens
PART 4. HEAT TRANSFER AND PHYSICAL PROPERTlES RADIATION DAMAGE S
FUEL RECOVERY AND REPROCESS\NG CRITICAL EXPERIMENTS o
HEAT TRANSFER AND PHYSICAL PROPERTIES ......cconnrenececenranrenne rveereraeestsrsaeneasrerensies
ART Fuel-to-NaK Heat Exchanger e eatesesesssitesstesriaseRsRe e SeTT TS LA S RS A s s e bR R Rt
- ART Hydrodynamics
- Core Hydrodynamics......... ©uomesesesessreberthebs et ettt A A SRS R SRS A SR RS eE PR SRR SRR RO RO SHSS e ASE SR E SR SRRSO RS RO S RS
Sodium Flow in Reflector Cooling Sysfern Ceeeeeteesessesstssesssstsesstssesteesteesessstesrerararasantsassassenenasnetes
Fluid Flow Visuglization ........cceecincreiinnniinne et s evereeresreeneveienenssnes
ART Core Heat Transfer STUAY ........occcoiereeeieenennieennrincssssessossssssesssnsessisisisesmsssssssssassssssssasssssnsasnssasens
Thermal-C'ycling Experiment .....cccccveeveiicenas Heeresesenesseseheserasaetatat et at Rt e s e s e et ensaenaseapnase L RO RS e SaR RO RESS
Shield MOCKUP COre STUGY «.c..vvuemmeeemsssssssremenssssssssssssssssssssssasessssas sesssssssssssenssssssssssasasssssasassesssssssssessesses
Heat Capaity ...occceenecicnecerecrcsnecrsnes e oot sesesnsosssssastshssssss sanssasessesisssnssossanasssssasssnsesns eeeveenensarasns
Viscosity and Density ... rcrmnrenerecemesecesssisssisstiss s e _
Thermal CondUctivity ....ccccveeenirienii et b s s e e s s sbr s ab e e st s s me s s b s as s e e s sanenansasessene
RADIATION DAMAGE ... scverseseessassse e s s s ssasstssesass st ssssasesessassasasasesssbesasssrsssnosssasssssss
Examination of Disassembled MTR In-Pile Loops Nos. 3, 4, and 5 eeseeses e seesassssmaneeren
Investigation of Materials Removed from MTR In-Pile Loops Nos. 3 and 5.
Creep and Stress-Corrosion Tests of Inconel ...
MTR Stress-Corrosion ApParatus ......co..ccerernsssemmissssssssorssrersstssssssssssssssenssessssns reeeeensasrtessnseresassaasens
Effect of Radiation on Static Corrosion of Structural Materials by Fused | : -
SAIt FUBIS ..o reecverenertree e e reesarsee e sessassessasaessesesobabes a0 beonisbes e Re s R R P e Sa SRS SRS S e O R Ra ST RS S s an S e e s R e84
Holdup of Fission Gases by Charcoal Traps ....c.c.wiciennineeincnsesessissessessasesssesssssssssssnsess
ART Off-Gas System lnstrumentation AnGlySis ..ot sesssssssnssasessenss
LITR Vertical In-Pile Loop.............. reveeressseeneesesertirnretsenesaetasrabant s et A st e e nboabasansienebenaensasnas yoeresreasssensins
Design Calculations for LITR Vertical In-Pile Loop eereeseseseressasbe b s R b se st R e et s asrsri e e e et s s e snnne
ART Reactivity Transients Associated with Fluctuations in Xenon-Removal 7
EFFICIency ettt ettt as st ne resentraseenesanenannsnes resrenreneesense
Use of Zr?5 as a Fission Monitor .....ommeeccersnmmissasssessecsesssecesnssses reremerans rerenerensasares rsassssasaraes
" Fast-Neutron Detectors .......ccccuueen.n.
Effects of Radiation on Electronic Components .....cvuvceenneininsinrenensinsnsissrsssisessseraesens emesaaeeresensnes

Xy
 

 

 

"

-l

4.3.

4.4.

5.1,

5.2.

5.3.

5.4.

 

Irradiation Effects on Thermal-Neutron Shield Materials .........memcrssmimmmssismsssnssrsnssssssss 256
. Irradiations of Stressed Shielding Materials .......ccccvuvicirrerereccciiniic s 258
FUEL RECOVERY AND REPROCESSING ...t creensstseneraesssssisssssassssssssssssssnsasassanesens 261
Volatility Pilot Plant Design and Construction ... sesssssessnsens 261
Engineering Development ...t sss st sesi s s st sssesa s 261
Chemical Development .........ccocoevviiincconninnnn s 262
CRITICAL EXPERIMENTS ..ottt cnininenns e ereteesserr b et er e bt ren et e esasasarenaaan 264
Reflector-Moderated Reactor EXPeriments ......c.cooeeiriesieieneicierecentisinnncsseesisisisessssssessssesssssssssssassnas 264
Experiments at Room Temperature ........coccewreiiimemiiisinisissisesssssssssassssssssssssssnssssananssssssassscsenins 264
Experiments at Elevated Temperature ...t 264
PART 5. REACTOR SHIELDING
SHIELDING THEORY ...ooiireecemrcreteimisesseseneseseresssssasssaesasesassssss sessssenssstsssassesesessesssstsssssssss sresasasesssasasesenes 271
A Monte Carlo Study of the Gamma-Ray Energy Flux, Dose Rate, and Buildup
Factors in a Lead-Water Slab Shield of Finite Thickness ..o 271
LID TANK SHIELDING FACILITY ottt resensnsestsesssan s ssssessasasssstes st sssssannssesens 278
Study of Advanced Shielding Materigls ...t 278
BULK SHIELDING FACILEITY oot srerescnesissesssssastesssssnsssssssssasasssssssesssnsssanessasesasasessessans 291
The Fission-Product Gamma-Ray Energy Spectrum ...t 291
Gamma-Ray Streaming Through the NaK Pipes that Penetrate the ART Shield......coccoveevnnenenne 296
SHIELD MOCKUP CORE ...ttt eeestsess s e sessess st ssssensssasssnassnssbsrssssnssasssnsassasans sassssss 299
Selection of Caleulational Models ... s s snens 299
Caleulational Procedure ... i st s s bt sesesessssssssststssasssasassss s sa s asas e s sasessbesesas 301
Calculations for SMC Configurations Using UO,-Stainless Steel Fuel Elements ........ccccoinennennec 301
Caleulations for SMC Configurations Using Uranium-Aluminum Fuel Elements .......ccoovviinnenee 305
Core GammA ROYS c.ocvveveeeeemrerireiensieencsemsine s sssressasass s s ssss st ess s ss s st sas s s ssssssssess saessassebasants 309
Neutron-Capture Gamma Rays from the Core Shells rerame s ae RS e s 310
Neutron-Capture Gamma Rays from the Reflector .........umoreereessmnmerserissssnssssssssrsesssssssins 310
Gamma Rays from the Heat Exchanger Region ..., 310
Neutron-Ccpture Gomma Rays from the Pressure Shell and Shield woeeeeecicreeeenenrereereeeesereeneas . 310
Neutron-Capture Gamma Rays from the Air and Crew Compartment ......c..coceeeemsiemsreresisssssees 310
Design STatUs ccvcereireeececreeenssnscssnscscssnessssnsssesessens reesmensesirelassesetensarsestssabene et seane s et tes s s st shsasenpereaee 310

Xv

 

 
 

 

 

 

 
 

 

.«3}

 
  

SEC

ANP PROJECT QUARTERLY PROGRESS REPORT

| SUMMARY

PART 1. AIRCRAFT REACTOR ENGINEERING

1.1.i Aircraft Reactor Test

Most of the detail drawings of the reactor-core,
heat-exchanger, pump, and pressure-shell assembly

have been completed, and design work on the re-

actor shield and the interior of the reactor cell is
proceeding, A review of the detail drawings indi-
cated that the pressure drop through the reflector-
moderator cooling circuit would probably be higher
than originally estimated, and therefore larger and
higher-speed motors are being ordered for the
sodiumepump drives. A re-examination is being
made of the tolerances and clecrances specified
for the fuel-to-NaK heat exchangers in an effort
to ease the fabrication problems,

Tests of a fuel pump in a high-temperature test
rig have indicated that the cavitation limit is
somewhat lower than anticipated.. However, the
increased cavitation suppression head required for
satisfactory operation can be accommodated by
increasing the fuel expansion tank pressure by

about 10 psi. In tests of a fuel pump and expansion

tank assembly no objectionable aeration or pressure
fluctuations were observed in the system over the
entire speed range when the fuel levetl in the fuel
expansion tank was varied from 1/2 to 3in.

The preliminary design of the fuel-recovery tank

was completed, The tank has been designed so

that no electrical, cooling, or other lines will need -
to be connected to it at the time it is removed from _

the cell. Lo

Studles of the reactor supporf structure were come

pleted The complete reactor and shield assembly

is to be suspended from an overhead bridge-like
structure. The weight of the reactor will be frans-

mitted to the bridge structure through the four pump

barrels. To cllow for displacement of the reactor
as a result of thermal expansion of the NaK lines, -
the lines are to be precut so that their exponsion’
will move the reactor to the desired operating posis

tion. Several bends have been added to the NaK
lines to allow for differential expansion between
the lines.

  

The stresses to be expected in the fill-ond-drain

“tank were also evaluated. The support of the fill-

and-drain tank assembly is to be accomplished by
means of a nitrogen cylinder located beneath the
tank.

A series of low-frequency thermal-cycling tests
of the core shells is under woy., The thermal
cycle being used in the tests simulates that to be
‘expected in the reactor. In the initial test the
model was subjected to 300 cycles, a factor of 10
more than expected in operation of the ART,
Visual and dimensional inspection revealed no
gross failures. Metallurgical examinations are
being made.

Additional core flow studies were made on the
full-scale model of the ART core. Heat transfer
coefficients were calculated for the 0,010-in.-thick
fuel layer odjacent to the outer and inner core
walls for the core flow pattern established by
using a swirl-type header with and without guide
vanes,

1.2, ART Physics

The results of experiments performed to determine
the rate of gamma heating in vorious target ma-
terials near the Bulk Shielding Facility (BSF)
reactor were compared with results calculated by
the method used for determining similar ART gamma
heating rates. The good agreement between the
experimental ond the calculated heat generation

~ values for samples near the BSF indicates that the
“basic gamma-ray data and the Prott & Whitney

gamma-ray deposition code used for the ART cal-
culations should give representative results for
the gomma-ray heat- deposihon in various parts of

the ART,

- Two-dimensional gamma-ray hecfmg calculations

' for the ART have been performed by using a code

developed for the IBM-704 computer by Kneidler
and Wenzel ot Pratt & Whitney. Source strengths

~ for - the calculations were obtained -from two-

dimensional multigroup neutron calculations and
from critical experiments. The total gamma-ray
deposition (excluding that caused by sources in the
heat exchanger region) in the reactor was found to

 
 

 

 

ANP PROJECT PROGRESS REPORT

be 3.86 Mw, which amounts to approximately 84%

of the total source. This means that about 16% of
the gamma-ray source energy escapes.

1.3, ART Instruments and Controls

The ART simulator was used for a study of tem-
perature control of the reflector-moderator and the
temperature responses of the reflector-moderator
cooling system. A control system was evolved
which will hold the mean reflector-moderator tem-
perature constant to within +15°F for step changes
between zero, 50, and 100% design power and be-
tween 1200 and 1425°F mecan fuel temperature (of
50% and zero power).

Tests were continved on a helwm-bubbler type
of level indicator for the fuel expansion tank, and
three test systems are being constructed for study-
ing other liquid-level-sensing devices under con-
trolled conditions in flowing liquids. Additional

tests of a high-temperature turbine type of flow- -

meter have indicated that these units can be ex-
pected to operate satisfactorily at 1500°F for
3000 hr. A unit with a 3‘/2-in.-dia housing for oper-
ation in o system with flow rates up to 1400 gpm
is being fabricated,

Evaluation tests of several types of high-
temperature pressure transmitters were continuved.
The effects of out-of-pile aging on the calibration
of thermocouples are being studied, and plans are
being made for in-pile tests, Life tests of two
synchronized high-speed mercury-jet switches for
scanning thermocouple signals are under way.

1.4, Component Development and T_esiifig

Evaluation studies of lubricating and cooling

fluids for pumps were continued. Tests are being

made in order to determine the compatibility of
various lubricants and reactor process fluids and to

evaluate elastomers for seal application.

Preliminary test data were obtained which indi-
cated that fission-gas leakage from the sumps of
the ART fuel pumps would be attenuated by a
factor of at least 104 from the expansion tank to
the shaft seal and by another factor of at least
104 between the seal and the iubricating oil
reservoir, . These data are being analyzed to de-
termine the radiation dose to be expected at the
oil reservoirs,

Developmental test worl: on the. fuel pump im-
peller continued. Modifications were made to the
slinger impeller and to the seal plate over the

centrifuge cup to improve performance. An en-
durance test of a fuel pump was terminated after
approximately 2000 hr of satisfactory operation.
The pump was disassembled, examined, and re-
assembled for further high-temperature testing.
More than 400 additional hours of satisfactory

operation had been accumulated by the end. of the

report period.
Water testing of on ART sodlum pump continued,

Various configurations of the flow passages above

the centrifuge were tested, and a configuration was
found which gave stable operation of the pump in
a performance-acceptance loop.

Water tests were also performed .on lnconel ports
for the NaK pumps. Data for the Inconel parts
were found to be satisfactory when compared with
data obtained previously with a brass impeller.

York NaK-to-air radiator No. 9, the first 500-kw
radiator of the revised design to be tested, was
found to be in satisfactory condition after comple-
tion of a program of 30 thermal cycles.. Endurance
testing of a heat exchanger at ART fuel-to-NaK
heat exchanger design temperature and flow condi-
tions with a heat load of 400 kw was initiated,

Two 4-in.~dia circulating cold traps equipped with
stainless steel cooling coils were placed in opera-
tion with no difficulties from oxide plugging. A
procedure for startup of the ART cold traps was
developed.

Water tests of the aluminum mockup of the ART
fuel system components of the north head were
continved.  Fabrication of a second aluminum
north head is under way in which the sodium sys-
tem components will be mocked up.

Four ART prototype dump valves received from
vendors were fested, but none met performance
specifications for ART operation. Assembly and
and welding procedures ore being analyzed, and
seat materials are being tested in fused salts.

Apparatus and techniques are being developed
for investigating the effect of high-frequency
thermal oscillations and the resultant thermal
fatigue stresses on the ART core. A high-frequency
pulse pump has been demgned for dynamic mixing
of hot and cold fluids to provide a thermal-oscilla-
tion frequency range of 1 to 10 cps and a surface
amplitude of about 100°F.

Various types of cold traps and pluggmg mdn-
cators are being evaluated in test stands. The
test stands are also equipped with Argonne samplers
and Mine Safety Appliances samplers -so that

-}
 

 

L}

)

chemical analyses of the NaK being circulated
can be used to calibrate the plugging indicators,

Liquid-metal-vapor condensers are being - de-
veloped for use in helium purge systems in the

ART sodium and NaK circuits and the NaK dump -

tanks, High~temperature chemical absorbents and
thermal-precipitation traps arte being investigated
in the development of a zirconium fluoride vapor

. trap. - The most promising system consists of a
. bed of hot alumina which reacts with gaseous
- ZrF  to form solid ZrO, and AlF,,

1.5, Procurement and Constructren

Work on the building aodditions, building altera-

tions, and cell instaliation for the ART has been -

completed, except for installation of six circvit
breakers and the completion of modifications to the

~ cell floor structure. A rigid leak test of the re-

actor cell was made; a zero leakage rate was ob-
tained with a detector sensitive to 5 micron 3 /hr.,
Since the contract specification allowed 32 micron

" #3/hr, the wvessel satisfactorily met the leak

tightness requirement. The installation of aquxiliary
services piping was completed, except for re-
placement of 15 incorrect valves in- the nitrogen
system, The electrical control centers and the
spectrometer room electrical and air-conditioning
equipment were completed. Work on the diesel
generators was delayed because one of the gen-
erators and three control panels were damaged in
shipment. - Two ‘of the four main blowers were
installed, -
Program and design planning for d:sossembly of

the ART were initiated. Procedures and tools

necessary for removing the reactor are’ belng de-
veloped, and. a new, Iarge hot-cell facnllty is bemg

~planned.. -
s Constructlon work started on the Engineermg

Test  Unit (ETU) facility, Details of the main

"NaK piping were determined and the design of the
air duct for the NaK-ts-air radiators was prepared.’
~The. study of the reactor assembly problem was

continued, -and genera! assembly procedures were

,prepured. ‘Nearly .all the components and spare

parts . needed for three reactors have been ordered.

) '_Fabncotion of the north head for the ETU reactor
‘is 40% complete._ The lower half of the beryllium
reflector-moderator has been contouted, und dnllmg'

of the coolant holes was started.
The York Corp, ond Black, Sivalls & Bryson,
Inc., are developing techniques for fabricating

PERIOD ENDING SEPTEMBER 10, 1956

the fuel-to-NaK heat exchangers, The sodium-to-
NaK ' heat exchangers are being fabricated and
should be available soon. -The Hydrospin machine
has been enlarged to produce the various shells
required for the reactors, and the outer core shells
for the ETU are being fabricated. '
Acceptable samples of finished boron carbide
tiles were received from the Norfon Company. An

‘order has been placed for forgings for the reactor

pressure shell.

1.6. In-Pile Loop Development and Tests
Examination of in-pile loop No. 5, which was
inserted in the MTR but could not be filled with

fuel, revealed that a plug had formed in the fill
line that was a mixture of fuel and zirconium oxide.

~ Inepile loop No. 6, which is to be inserted in the

MTR soon, is to operate at a maximum loop tem-
perature of 1600°F with the fuel mixture NaF-ZrF ,-

(53.5-40-6.5 mole %), o 300°F temperature
differential across the nose coil, and a power
density of about 0.75 kw/cm3, A horizontal-shaft

sump pump, identical to that in in-pile loop No. 6,

is operating satisfactorily in o test loop. It will
have accumulated 1150 hr of operation by the time
in-pile loop No. 6 is inserted in the MTR,

1.7. Advanced Reactor Design

" Reactor designs are being investigated in which
the fuel cools the moderator, in order to eliminate
the sodium circuit used for cooling the moderator
in reactors of the ART type. One of these designs
depends on moderation by beryllium oxide in the
reflector and some additional moderation by graphite
in the core.’ Thermal stresses dictate the use of
the - graphite in the core.  The relatively poor
moderating properties of graphite are not too ob-

~ jectionable ‘becouse of the relatively large core

size (24 in. in digmeter) required by the total
power and the allowable power densities. On the
other hand, the use of a very good moderatfor, such
as zirconium or yHrium hydride, in the core would
make the reactor independent of moderation by the
reflector, and metals could be used in the reflector.
Metals require few cooling holes and are good
gammastay and fast-neutron shields, With presently
assumed materials properties, the hydride-moderated
reactors would be better than the graphite~beryllium
oxide moderated reactors; on the other hand, the
properties of graphite and beryllium oxide are
somewhat better known.

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

PART 2. CHEMISTRY
2.1, Phase Equilibrium Studies

Phase equilibrium studies of fuel system com=

ponents were continved. A re-examination of the
LiF-RbF system has revealed the compound
LiF-RbF, and a revised equilibrium diagram has
been prepared. Further examination of the LiF-UF,
system has confirmed the identity of the previously
arbitrarily designated compound 3LiF:UF,. There
is evidence that the compound is metastable.

Studies of the RbF-UF 4 System progressed suf-
ficiently for a tentative equilibrium diagram to be
prepared. The boundary curves, compatibility
triangles, peritectic temperatures, and eutectic
temperatures for the NaF-RbF-UF , ternary system
are now well-established. Detailed examination of
the four-component system NaF-RbF-ZrF 4-UF 4 has
established several mixtures which are of possnble
interest as reactor fuels. '

In studies of the RbF-CaF, system it has not
been possible to get reproducible data for compo-
sitions . containing more than 15 mole % CaF..
There is evidence of a compound with a cubic
structure, which is presumed to be RbF-Con
Examinations of the systems LiF-CeF,, NaF-CeF,,
and RbF.CeF 3 were continued, but, as yet, 1here
is insufficient data for the preparation of equilibrium
diagrams. The NaF-CeF, system was shown to
have a binary eutectic, whlch melts at 775 1 10°C,

A study of the CsF-BeF, system was started
in order to complete the study of the alkali fluoride—
beryllium fluoride systems. This study is incom-
plete, but, as in other alkali fluoride~beryllium
fluoride systems, liquidus temperatures diminish
rapidly in this system with increasing BeF, con-
tent in the 33'6 to 50 mole % BeF , region.

Study of the NaF-LiF-BeF, system progressed
sufficiently for a plot of the boundary curves, com-
patibility triangles, and pertinent temperatures to
be prepared, although the composition of one of the
ternary compounds is not completely established.
The ternary system NaF-RbF-BeF, requires addi-
tional study before the analogous diagram con be
constructed.

Lowemelting chloride systems, which are poten-
t_nolly useful as heat transfer mediums, are being
investigated.  The present investigation of the
KCI-ZrF, system covers the composition range of
10 to 75 mole % ZrF,. It will be of interest to
compare phase equnl:brwm behavior in these

MCI-ZrCl, systems with behavior in the analogous
MF-ZrF systems, which have been thoroughly
stuched ot ORNL. :

22. Chemical Reocflons in Molien Salts -

: lnveshgatlon of the equnllbrlum reduction of
NlF by H, in NaF-ZrF , was continued. A related
mveshgohon of the stobzllty of NiF, in the solvent
was made that indicated the essential validity of
the equilibration data which have been obtained at
500, 575, and 625°C. The data obtained to _dofe
indicate that the activity coefficient for NiF,
NaF-ZrF, (53-47 mole %) must be about 2750 ot
600°C, The _corresponding value for FeF, in this
solvent was previously found to be 3_28 In the
investigation of the stability of NiF, in the solvent,
a number of peculiar effects were observed that
appear to be important in mechanisms of cotrosion
of nickel by molten salts. Additional study of
these phenomena is to be undertaken,

The study of the activity of chromium in chromium-
nickel alloys was completed, and the values ob-
tained at 750 and 965°C are presented. The
studies of the solubility and stability of the
structural metal fluorides in fluoride mixtures
were extended to include CrF » FeF,, and NiF,
the solvents LiF-ZrF, (52-48 mole %), NoF-ZrF
(59-41 mole %), and KF—ZrF (52-48 mole %). In
addition, the solubility of CrF in the NeF-LiF-KF
eutectic was determined. The solubility of CrF,
in these solvents voried markedly with the solvent
used. In all tests with FeF, at 600°C the solu-
bility increased with excess Eer present and the
zirconium-to-sodium ratios tended to decrease with
increasing excess FeF,, ‘The solubility of NiF,
in NaF-ZrF, and KF-ZrF, was found to be the
same and nearly independent of the amount of NiF
added. However, the solubility in LiF-ZrF, is a
function of the amount of NiF, added at both 600
and 800°C, :

Studies of the reducflon of UF by structural
metals were extended to include the reaction media
RbF-ZrF, (60-40 mole %) and L|FQBeF (48-52
mole %). The equilibrium chromium concentrahons
in the RbF-ZrF, mixture at 800°C were found to
be lower than those in any other alkali fluoride—
ZrF, mixture studied, The equilibrium iron con-
centrohons were found to be virtually independent

‘of the reaction medium and in cll cases were

higher at 600 than ot 800°C. This behavior is in
marked contrast to that observed for the Cr -UF
 

 

b

¥

system, which exhibits strong dependence on the
reaction medium,

The solubilities of the structural metal fluor:des
NiF,, FeF,, and CGrF, in molten NaF-ZrF, (53-47
mole %) that were determined by using concentra-
tion cells and an emf method are presented. The
results fall within the range of the results obtamed
by filtration,

- Vapor pressure data were used fo calculute the

| effects of concentration and temperature on activi-

ties in the RbF-ZrF and KF-ZrF, systems.
Activities of KF were very slightly hlgher than the
activities of RbF ot correspondmg compositions
and temperatures, since the larger rubidium ion
has less affinity for the fluoride and RbF is g
better complexmg ‘agent than KF., The ZeF,
activity was found to decrease with mcreasmg

~ ionic radius of the alkali cation.

A systematic invesfigation of the solubilities of
rare-earth fluorides in ART-type fuels was initiated,
In the initial studies, LaF, and CeF, were used in
the solvents NaF-KF-LiF (11.5-42-46.5 mole %)}

and NaF-ZrF (50-50 mole %). The results obtained

thus for show the solubilities to be comfortably
high from the standpoint of reactor operation.

. Apparatus is being assembled for the determination

of the solubility of xenon in fused salts, -
Polarography was utilized for quantitative analy-
ses of structural metal jons in the NaF-KF<LiF
eutectic. Half-wave potentials were identified for
Ni*t Fe***, and Cr**?, and solubility determina-
tions were made by following the plot of wave
height vs concentration to the. solubnhty limit ond
beyond. : :

Meosurements of actwrtles in the system AgCI-'

NaCl were made by an emf method as a trial ‘study

with ¢ solid electrolyte of a technique which may
"be applicable to fluoride systems and also as a-
~ demonstration of some metastable solid solution -
behavior. The___feossbi_hty of making equ:hbnum S
' : . ‘mixture prepared with extreme ‘care fo .assure that

* the mixture was oxide-free, - Precautions were con-
sidered desirable after it was learned that the plug’

measurements with a solid electrolyte was demone
strated. 1f such. measurements on solid fluorides

prove to be feasible, @ ceramic container would no =
longer be required for cells containing fluondes, -
‘medsurements ' could be made on solids by using
cells deslgned to avoid transference “potentials,

and compositions with lnconvemently hlgh meltlng
points. could be studled : o -

PERIOD ENDING SEPTEMBER 10, 1956

2.3. Physical Properties of Molten Materials

Difficulties caused by partial plugging of the
capillary tips in the apparatus for applying the
maximum bubble-pressure method to determinations
of surface tensions of NaF-ZrF, mixtures have
prevented the accumulation of accurate data. The
equipment is belng modified to minimize contamina-
tion,

Density determinations were made on NaF.ZrF,

 (53-47 mole %) ot 600, 700, and 800°C, The values

presented are believed to be accurate to within 2%,

The sessile-drop technique was used for addi-
tional measurements of surface tensions of NaF-.
ZrF, (53-47 mole %) mixtures. . Experiments de-
signed to study the effects on surface tension of
change of composition as a function of temperature,

- pressure of the confining gas, and time of equili-

bration are under way. Experiments are also under
way to determine the wetting properties and surface
tensions of UF, and of UF, with small additions
of UO, on graphite.

_ 274. Production of Purified Fluoride Mixtures

The use of copper-lined stainless steel reactor
vessels has proved to be successful in all but one
instance. It has been found that if NaF-KF-LiF-
UF, mixtures are frozen in the copper-lined vessels,
the copper liner will rupture when the mixture is
remelted. Therefore any large-scale production of
such mixtures would require continuous operation
of the production facility so that the mixture could
be maintained molten at all times.

- - A dust filter is being installed so that the con-
- version of about 2000 Ib of :low-hafnium ZrCl, to
ZrF, can be started. It is hoped that the filter
~will prevent Iurge-scale loss of the product during
" the conversion and permit efficient- -operation,

MTR in-pile loop No. 6 was filled- with a fuel

which prevented the. filling of m-pile loop No. 5

may have resulted from oxnde contammatlon of the"

fuel mixture. e
~Data are presented on. the kmds nnd quuntutres

:of matermls produced durmg the quarter. ,

 

 
 

 

 

ANP PROJECT PROGRESS REPORT

2.5. Compatibility of Materials at
High Temperatures -

The diffusion of uranium fluoride~containing
mixtures into graphite was studied by obtaining

counts, with an alpha counter, en samples ob-

tained by filing and scraping, in a uniform manner,
the surfaces of the graphite specimens, . After 10 hr
of exposure to NaF-ZrF ‘-UF 4 (53.5-40-6.5 mole %)
at 600°C, there was no evidence of penetration,
and, after 50 and 250 hr, only slight increases in
alpha count were found. A graphite sample ex-
posed at 800°C, however, showed a sufficient
increase in the alpha coynt after 10 hr of exposure
to indicate the presence of the salt |n5:de the
test spec:men.

. An apparatus was developed for measuring the
equilibrium pressure of hydrogen from the reaction
of NaOH and nickel, The new apparatus permits
measurements . of the pressure inside a sealed
quartz envelope containing the NaOH-Ni. system.
The measurements can be made after the apparatus
has cocled to rcom temperature. Results of pre-
liminary experiments are presented,

Equipment has been assembled for testing
elastomers under tension in NaK at 185 to 190°C,
Screening tests are under way.

In the search for a reliable and practical method
for analyzing for oxygen in NaK, a comparison was
made of the amalgam and the butyl bromide methods.
The values obtained by the two methods disagreed
radically, and therefore further studies of these
methods cre under way.

2.6. Analytical Chemistry

Tiron (disodium-1,2-dihydroxybenzene-3,5-disul-
fonate) was applied to the spectrophotometric de-
termination of cerium in NaF-KF-LiF-UF, and
NaF-ZrF -UF ,. A colored complex, with its ab-
sorption ' maximum at 500 my, results from the
redaction between cerium and Tiron in a slightly
basic solution.  Uranium, zirconium, iron, and
fluoride interfere; however, their interference is
readily . eliminated by various means, including
extraction.  Sodium alizarin sulfonate is being
investigated os a possible reagent for the spec-
trophotometric determination of lanthanum in
fluoride salt mixtures. '

The minimum concentration of tantalum that can
be determined by the tannin-pyrogallol method in
NaF-ZrF «UF, was found to be approximately
1000 pg. | '

The buffering action, which was observed when
the aqueous solution from the dissolution of NaK
in butyl bromide was titrated with acid, was found

" to be attributable to the presence of 7 to 20 ppm

of aluminum in the alloy. Although these small
concentrations will produce buffering, they cannot
account for the high oxide concentrations that
have been found in NaK by the butyl bromide

method. A modified metal reactor was constructed

so that the amalgamation procedure for the de-

termination of oxygen in NaK can be used.

An apparatus for sumplmg alkali metals at hlgh
temperatute was modified by substituting a glass-
tosmetal Wilson seal for a packing gland seal to
facilitate complete removal of oxygen from the
apparatus and by replacing the Teflon-packed gate
valve with a Teflon-packed Jamesbury valve. This
valve is a reliable, vacuum-tight, ball-type valve,
which is completely opened or closed by a one-
quarter turn of the handle,

Methods for the determination of oxygen, carbon,
and nitrogen in lithium were studied. The last
two elements were analyzed by dissolving the
sample in water fo convert nitrides and carbides to
emmonia and acetylene, respectively, which were
then determined by conventional methods,

An all-metallic apparatus for the determmahon of
oxygen in lithium by amolgamation was constructed
which will facilitate transfer of samples without
atmospheric contamination. A modification of the
butyl bromide method for oxygen was studied for
application to lithium. The metal is dissolved in
an ethereal solution of n<buty| iodide in the presence
of iodine. |t is postulated that the lithium forms
the neutral iodide salt, that lithium carbide and
nitride are oxidized, and that the oxide remains
unchanged by the dissolution of the metal,

The experimental investigation of means for the
introduction of reproducible concentrations of
sodium into air was continved with limited success,
Stable aerosols were not formed when helium which
had been equilibrated with 20 to 100 ug of sodium
per liter of gas was mixed with 10 times its volume
of air. Therefore, a test facility, which is essen-
tially a mockup of the ART radiators, is being
fabricated to test the instruments which have
been developed for the detecflon of traces of NaK
in air,

A device was constructed which will permit
visual observation of compatibility tests of molten
 

-«

K}l

 

 

fluoride salts and alkali 'meffdls-:fwifh'npump- lubri-

cants,  Dowtherm-A at 200°F. was: found to be

-essentially inert to sodium at 1100°F; however,

under identical temperature conditions, Cellulube-
150 underwent a vigorous, exothermic reaction,

An investigation was initiated to study the ex-
“traction of ART fuel components, corrosion prod-
uets, and fission products from -acidic solutions
into soluhons of trlalkylphosphlne oxides i in orgamc'j ‘_
e _‘_solvents. The _extractant was. Opplled to evaluate .
~ the defermlnatlon of chromwm by extracting it with
the phosphme oxide and then anolyzmg the extract

d:recfly by the dlphenyl carbozlde method

PART 3. ME TALLURGY

3.1, Dyncmlc Corrosron

- Two Inconel forced-crrculctlon Ioops were oper-

mole %) at maximum fuel temperatures of about

| - 1500 and 1650°F in order to compare the corrosion’
properties of this fuel mixture with those of the
. commonly . used mixture NaF-ZrF ‘-UF (50-46-4

mole %). - Chromium was found in a crystalline

deposit in the cold zone of the ioop operated at
“about -1500°F, but no crystals were found in the

loop operated at about 1650°F, Deposits of this

~ type have not been found in loops operated with |

NaF-ZrF -UF (50-46-4 mole %). The depth of

| hot-leg aflock which was the same for both loops,
7 mils, was mrdwuy between the average depths of -
~ hot-leg oflack found at 1500 (4.5 mils) and 1650°F

(9 mils) in Ioops operated w:th NaF-ZrF -UF
(50-46-4 mole %). .

Examination of a Hastelloy B forced-cnrculctlon

loop that opercfed with NaF- KF-LIF-UF (H.2-4|-; :

"45.3-2.5 ‘mole %) verified' ‘the exceflent corrosion
resistance of Hastelloy B to_fluoride fuel mixtures
- 'as observed’ prevrously in thermaloconvechon loops.:
~“An Inconel forced-circulation loop’ that had. been
o operated with' NaF-ZrF ~UF, (50-46-4. mole %) in -

" an endurance fest was also exummed. 'Operation

. had been terminated at 8300 hr because of a.pump-
- drive farlure. -The loop was gas-fired, and it opers

. _;-ated with @ maximum fuel temperature of 1450°F,

e femperafure drop of 200°F, -a maximum wall

temperoture of 1550°F, and a Reynolds number of -

-10,000. . “The hot-leg surface was roughened, ond -

. there’ was_heavy mtergrunulur void formation toa

depth of 25 mils. The cold-leg surfoce was rough-

" ened, but no metallic deposits were found. Another.

PERIOD ENDING SEPTEMBER 10, 1956

~ Inconel loop that had operated with the same fuel

mixture for 3000 hr with maximum and minimum

fluid temperatures of 1600 ond 1300°F, a Reynolds

number of 5750, and a meximum wall temperature
of 1700°F was also examined, The maximum attack

.was to a depth of 14 mils, and no deposits were

found in the cold zones that could be attributed to

_mass transfer.

It was found that i mcreasmg ‘the temperafure drop

from 300 to 400°F in Inconel loops circulating
. sodl_um at the same maximum fluid temperature
_appreciably increased the amount of mass transfer,

The mass transfer was also appreciably increased
by increasing the maximum fluid temperature from

1350 to 1500°F. For type 316 stainless steel

loops circulating sodium, the effect of increasing

the maximum fluid temperatures from 1500 to 1650°F

~ was  slight. A type 310 stainless steel loop
--_"uted with the fuel mixture NaF-ZrF -UF (56-39-5

showed substanticlly larger deposits than those

- found in a type 316 stainless steel loop operated

under similar conditions, ,

~ Two Inconel forced-circulation loops with beryl.
fium inserts in the hot legs were operated with
sodium. As observed previously, alloying occurred

_where the beryllium and Inconel were in contact,
- but in areas where there was a positive separation

between the beryllium and the Inconel no alloying
was found. Similar results were obtained with
Hastelloy B thermal-convection loops containing

_beryllium inserts, A series of five Inconel thermal-

convection loops with berylhum inserts operated

with sodium ot temperatures varying from 1200 to

1500°F showed that attack of the beryllium ine
creased with temperature, There was some alloy
formation in all the loops operated at temperatures -

ubove 1300°F.

" Examination of a Hastel!oy X thermal-convechon

Ioop operated with NaF-ZrF +UF, (50-46-4 mole %)

" at 1500°F showed heavy. mtergranulor subsurface-
void formation to a depth of 35 mils. In contrast,

a Hastelloy W loop and a nickel-molybdenum alloy

- (85% Ni-15% Mo) loop operated with NaF-KF-LiF- -
UF, (. 2-41+45,3-2.5 ‘mole %) under similar con-
' ;dmons showed ‘maximum depths of attack in the
~ hot leg of 0.5 and 2 mils, respectively, '

- Inconel loops operated with NaF-Z¢F -UF' (50- ’

- 46-4 mole %) prepared with especrally pure ZrF

showed 2163 mils less attack than that found'

-~ with fhe standard mixture, ‘A similor reduction in
. attack was found in loops operated with fuel te-
_ claimed from experiments,

 
 

 

 

 

 

ANP PROJECT PROGRESS REPORT

- Additional tests were completed in the screening
tests of special fuel mixtures in Inconel thermal-
convection loops, . The results obtained to date ore
presented. : S

Data obtained in cyclic strain tests of Inconel in
a ‘helium atmosphere are also presented. At all
temperatures studied (1200, 1400, and 1600°F), o
2-hr hold time appears to produce cracks in fewer
cycl_es-than are required to produce cracks in tests
with a 4-hr hold time. Cracks initiate at approxi-
mately the same time in tests at 1400 and I600°F
but they propogute faster at 1600°F.

3.2. Genercl Corrosion Studies

The recrystallized portion of resistance-welded
Inconel . tube-to-header joints fabricated by the
**flange-during-welding method’’
The Glenn L. Martin Co. was found to have gocd
corrosion resistance to NaF-ZrF UF, (50-46-4
mole %) and to NaK {56-44 wt %) ot 1500°F dwing
a 100«hr test, The maximum attack on the joints
tested in NoK was 0.5 mil, and those tested in the
fuel showed 1.5 mil of attack on the header portion
and less than 0.5 mil of attack on the recrystallized
portion of the worked tube,

Seesaw corrosion tests of Coast Metals brazing
alloy No. 53 (81% Ni~8% Cr—4% B—-4% Si-3% Fe)
in NaF-ZrF -UF  (53.5-40-6.5 mole %) and in NaK
(56-44 wt %) showed the same type of boron deple-
tion as that found previcusly with Coast Metals
brazing alloy No. 52. Boron was depleted from
the specimen tested in NaK for 100 hr at 1500°F
to a depth of 2.5 mils and from the specimen tested
in the fuel mixture to a depth of 1.5 mils,

Haynes brazing alloy No. 40 (nominal compo-
sition, in wt %: Cr, 13.5; Fe, 4.5; Si, 3.9; B, 3.2;
Co, 1.0; C, 0.42; Mn, 0.25; Ni, balance) showed
good corrosion resistance to NaK (56-44 wt %) but
only fair resistance (3 mils of heavy attack) to
NaF-ZrF ,-UF (50-46~4 mole %) during 100-hr
tests at 1500"1:

No mass transfer and very little corrosion occurred
in tests of niobium in sodium and in lithium in

seesaw apparatus. In these tests for 100-hr periods -

the hot-zone temperature was 1535°F and the cold-
zone temperature was 995°F.

Long-duration tests of static NaF-ZrF -UF,
(56-40-4 mole %) in Inconel in which temperature
gradient was used to make the more volatile ZrF
crystallize above the fused-salt bath revealed that
no attack occurred in the vapor zone and that the

and supplied by

attack- in the bath region was less than' that nor-

~ mally observed in fused-salt Inconel tests.

Of the four alloys subjected to carburization by

‘exposure for 100 hr ot 1500°F to static sodium

containing various amounts of added graphite, the
alloys that experienced the greatest amount of
carbon pickup were, in descending order, type 430
stainless steel, type 316 stdainless steel, type 310
stainless steel, and Hastelloy B. The alloys that
suffered the greatest amount of carbon penetration
were, in descending order, type 430 stainless steel,
type 316 stainless steel, Hastelloy B and type

- 310 stainless steel.

It was demonsirated that chromiume-plating Inconel
specimens prior to testing them in contact with
beryllium while immersed in sodium at 1300°F was
effective in reducing the extent of alloying. A
minimum of 5 mils of chromium plate will be neces-
sary to ensure that all the chromium is not con-
sumed by an alloying reaction with beryllium,

Two Inconel thermal-convection loops were
operated ‘with sodium in fests conducted to de-
termine the effect of a diffusion cold trap on the
amount of mass transfer into the cold leg of the
loop. Corrosion and mass transfer were found to be
less in the loop with the diffusion cold trap than
in a standard loop with no trap.

Seventeen thermal-convection loops constructed
of various grades of the stainless steels ond
Inconel were operated with lithium. None of the
stainless steels had satisfactory resistance to
mass transfer at hot-zone temperatures of 1500°F
and Inconel was not satlsfactory at a temperature
of 1300°F.

- The intermetallic compounds NiAi NiAl + 5% Ni,
NiAl + 4% Zr, and MoAl were corrosion tested in
static sodium and in NaF.ZrF +UF, (53.5-40-6.5
mole %) for 100 hr at 1500°F The compounds
were not attacked by the sodium, but they were
severely attacked by the fused sait.- '

3.3.

Work on the development of Ni-Mo base alloys
was continued, with emphasis on the fabrication
of seamless tubing of the various alloys for cor-
rosion evaluation, Tube blanks of the various
alloys have been successfully extruded, and re-
duction of the tube blanks to tubing, by redrawing
at the Superior Tube Co., has proceeded satis-
factorily, However, the large materia| losses that
now occur during the conditioning of the extruded

Fnfiricution Reseurch‘A =
 

 

 

»

__thereby demonstrated. : S
~ “The fabrication of expenmentol composmons of'_:_
o ’tungsten “carbide ‘with nicke! alloy binders for use
~in the ART. pump lmpeller “shield. plug was cone
-"_,tmued. It was established -that o compos ition of -
- 25-10-30% Hastelloy C with tungsten carbide had -

- optimum - properties, Hot-pressed ‘models of the

. .gamma-ray . “shield were - smcessl’ully brazed fo

- Inconel ‘plate,: Specnmens of Zr0,, were fabricafed -
. _and tested -for ‘use ‘in.the thermal ‘shield of ihe ;

_ _\_'sh ield plug. und were: founcl 1o be sohsfactory. :

A sample of the ‘type 400 ‘stainless sfeel clod
'Cu-B C neutron shield material, which was fabri-
cated by -the Allegheny Ludlum Steel Corp,, was

tube by machining prior to the tube-reducmg opera-

tion are not satisfactory,

The reduction of tube blanks of the special
alloys received from Battelle Memorial Institute
and the International Nickel Company, which were
extruded during the previous quarter, was only
partly successful, ~ The alloys which contained

" moderately high carbon contents of 0.12 to 0.25%

tended to crack badly during cold reduction. Metal-
lographic examination revealed stringers of carbides
in the microstructure; the cracks tendecl to propagote
along the stringers,

Forged billets of four new olloys ‘were received

from Battelle Memorial Institute. “These billets
~were extruded without difficulty, and the extrusions
‘were sent fo Superlor Tube Co. for reductxon to

tubing.

The preparation of tubing of the ternary alloys
based on Ni-17% Mo alloys with additions of Ti,
Al, W, Nb, Cr, Fe, ¥, or C was continved. These
alloys are being investigated to establish the upper
limit, from a corrosion viewpoint, of the third
element needed to improve the strength of the
binary alloy. In addition, cast billets of three of

the ORNL INOR-designated nickel-molybdenum

base alloys were prepared and extruded to produce
tubing for corrosion tests.

- Tubing redrawn from the ternary _al_loys prepared

previously was received and fabricated into thermal-

- convection - loops. ~ Acceptable ylelds of tubing

were obtained from cll the alloys.

Twenty-five feet of seamless- Hcsfelloy W tubing
0,187 in. in dmmeter was redrawn from  the first
successful extrusuon, reporfed prevuously, and the
feasibility of producing’ seamless tubing- of thls,
: 'operated for 1356 hr. Small cracks were found in
- several of the tube-to-support plote joints, and
-evndence of. cruckmg ‘was observed in several
ube-to-sump plate |o:nts. The results of the ex-
- amination confirmed previous observations of the
relcmonshlp between ‘the incidence of fracture and

type of alloy for heat’ exchunger apphcotlons was

evaluated, The sample plate was satisfactory,

PERIOD ENDING SEPTEMBER 10, 1956

except for thickness deviations and surface rough-
ness. '

"Encouraging results were obtained in the fobrl-
cation of tubular contrel rods by co-extrusion, The

proposed rods utilize a core of 70% nickel-30%

Lindsoy oxide, Thermal conductivity data were
obtained for the core material in the temperature
range 165 to 554°C, '

Simulated fuel element tubing redrawn from three-

~ ply tube blanks extruded previously was received
- and evaluated.” The redrawn tubing containing fine

oxide particles showed tensile fractures in the
core, while the tubing containing coarse oxide
‘porti@:les showed excessive stringering of the
oxide particles in the core, The study of fiow
patterns in three-ply extrusions was continued,
with the effects of sectional cores and cores with
tapered ends being investigated, A 29-deg taper
in the ends of the core reduced the lengths of the

_end defects,

An evaluation of arc-melted nlobwm is bemg
carried out by comparing its properties with those
of . powder-base wrought material. An alloy of 80%
Nb—-20% U is also being fabricated into strip for
mechanical testing and cladding experiments,

3.4. Welding-oml Brazing Investigations

_Experiments were conducted to determine the
approximate weld shrinkage to be expected in the
fabrication of the volutes and impellers for the
ART primary NaK pumps and in the fabrication of
the Inconel core shells. Procedures for obtaining

~ the desired dimensions have been established,

‘A 500-kw high-conductivity-fin NaK-to-air radi-

g ctor (design now obsolete) was examined that had

the presence of support members or plates.
A modified’ procedure for fobncatmg ‘cermet-valve

* components was developed and soflsfoctory disks

ond - seats ‘can now be- obtained, *The optimum

',iemperoture for consistent bonding was determined
- for each cermet composition ond was found to vory |
"sllghfly with the cermet type. '

“Equipment and ‘procedures for - the conhnuous

‘production - of sintered brazing alloy rings have

been developed, ond an experimental pilot plant

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

has been built in which a production rate of 8000
rings per hour has been attained. [t is felt that
production of this magnitude will definitely estab-
lish a satisfactory cost for ring production,

Experiments were conducted to determine the
most satisfactory method for preplacing an adequate
supply of braze metal at each joint in the assembly
of multitube heat exchangers, The most successful
method of initiating capillary flow was found to be
the metalespray primer technique.

Hardness values were determined for cold-rolled
Hastelloy B sheet after aging freatments for various
times at 1200°F., A microstructure correlation is
being made to study the effect of aging on the
physical properties of Hastelloy B.

3.5, Mei:finnicnl Properties Studies

" Equipment has been designed and constructed
with which it is possible to determine the relaxa-
tion behavior of metals at high temperatures. Infor-
mation of this type is required in order to analyze
the -residual plastic strain induced by cyclic
thermal stresses. Data on the reloxation charac-
teristics of Inconel at 1300 and 1500°F are pre-
sented,

Tests were run to determine the effect of the
fuel mixture NaF-ZrF +UF (50-46-4 mole %) on
the creep strength of Hastelloy X. The results in-
dicate that Hastelioy X is subject to such severe
corrosive attack by the fuel that o drastic reduc-
tion of its creep strength occurs at temperatures
above 1500°F,

Creep-rupture testing of an 80% Mg-20% L.i ailoy
at 200°F is under way. The data obtained to date
are presented in the form of design curves, with
stress as the ordinate and times to various elonga-
tions and rupture as the abscissa,

3.6. ‘Ceramic Research

A study of the fabrication and physical properties
of europium oxide was initiated, Numerous shapes
of nickel—trare earth oxide cermets, Al O, and
CaF, were produced for use in a high-temperature
critical experiment. Test pieces were produced
of high-density, high-purity beryllium oxide and of
zirconium carbide, . A hydriding furnace is being
built for producing zirconium end yttrium hydrides.
A die is being designed for fabricating a l/4-in.-OD
beryllium oxide rod with five longitudinal holes.

A high-temperature x-ray diffraction study of
Hastelloy B ot 500, 600, and 700°C showed only

10

a slight shift to larger d values with increasing
temperature. No inversions were found, and a
face-centered cubic pattern was obtamed at all
temperatures, ' o

7. Nondesfiucfiv_e Testing Sfudié_s,

* Much time was devoted during the quarter to the

routine inspection of approximately 17,000 ft of -

pipe and tubing for the reactor construction pro-
gram, Both encircling-coil eddy-current and
immersed-ultrasound methods of inspection were
used. Metallographic examinations were made of
regions that showed defects in order to correlate
the defect types and sizes with the data presenfed
by the inspection instruments. .

An inspection method for the . detection of small
laminar defects in sheet material was developed.
After consideration and rejection of such tech-
niques as ultrasonic resonance, conventional

pulse-echo vltrasound, and tfransmission attenua- -
tion of ultrasound by using transducers on opposite
sides of the sheet, a new vultrasonic method was

conceived and is presently being investigated.
This method requires a pulse of ultrasound of 5 to
20 psec duration tuned to such a frequency that
the sheet thickness is an exact multiple of the
half wavelength, Under these conditions a rever-
beration, or ringing, of the ultrasound between
the two sheet surfaces is obtained, and the presence
of laminations is detected by a decrease in the
ringing. The reflectoscope is being modified for
application to this sheet inspection technique,

3.8. Inspection of Materials and Components

Inspection of 3183 critical welds resulted in an
11% rejection rate for porosity, cracks, misalign-

ment, lack of fusion, and lack of penetration.
Inspections by radiographic, dye-penetrant, ultra-

sonic, and visual techniques were completed on
over 13,000 ft of tubing, and various emounts of
pipe,. plate, sheet, rod, and other Inconel material
were inspected. The rejection rate for pipe and
tubing has averaged approximately 10%, except
for two lots which were completely rejected, one
for oxidation of the inner surface and the other for
defects. Rejection rates for other shapes were not
greater than 5% and averaged much less.

Sixty feet of Inconel W and 208 ft. of Hastelloy . B ”

tubing were also inspected. The Inconel W tubing

was found to be acceptable, but over 65% of the

Hastelloy B tubing was rejected. However, be-
cause of the immediate need for the Hastelloy B

\fl
 

 

»

o3

tubing, it was reworked and all ‘but the grossest
defects were accepted, . Inspection of fabricated
components included thermal-convection loops,
dished Inconel heads, and small heat exchanger
_unlts. : -

Fluorescent-penetrant mspechon eqmpmenf was

instatled, and tests are under way to compare this

type of inspection with the dye-penetrcmt process

it will replace. It appears that hlgher quallty

tubing con be ensured by the greater sensmv:ty

| " of the new method

PART 4. HEAT TRANSFER_AND PHYSICAL

PROPERTIES, RADIATION DAMAGE, FUEL
RECOVERY AND REPROCESSING,
CRITICAL EXPERIMENTS -

4.1. Hect Trunsfer and Physicul Properhes

The fluid fncflon characteristics of 60-deg stag- - "
,gered spacers and 60-deg inclined spacers placed

clternately in a model of the ART fuel-to-NaK heat
exchanger were determined, and the data are pre-

‘- sented. A full-scale model of - the present ART
fuel-to-NaK heat exchanger, which contains more-
_tubes and a somewhat different spacer configura-

tion than that of the previcus design, has been
assemb!ed and pressure drop data are to be ob-

. tained.

A series of screens hove been placed in the
northern hemisphere of an ART core medel for the

purpose of .stabilizing the fluid flow, and the .
screens have apparently eliminated the reverse

flow on -the outer core shell wall for the case of

- straight-through  flow. Pressure drop measure-

ments were made for this system.

The static pressure dlsmbutsons in the feflector_'
and island cooling annuli were calculated with the
aid  of - information prevuously ‘obtained from the
- experimental ‘study ‘of ‘ the -flow dlstnbuflon ina

-:model of the reflector onnulus. s T

. Heat transfer expenments were. conducted on the_
" ART core mode! with a vaned entrance. system-'
cnder ‘uniform volume heat -source’ conditions.
~ Mean, ‘'uncooled wall temperature measurements are -

-;.presenfed ‘and’ ‘compared with predicted tempera- -
" tures for an. zdealrzed system. ‘Large temperatuyre
5 asymmetrles, as well as- large temperature fluctua:

. tions, were recorded which stem from hydrodynamic
asymmetries and flow instabilities, respectively.
These temperature fields are compared to those

obtained previously for the swirl flow system.

- position product.  Considerable deposits of Cs
~ and - Sr8? were found in the off-gas line, and smal
~amounts of these fission products were present in
- in the charcoal adsorber trap. -lron, chromium, and
nickel analyses of the fuel confirmed the metallo-
“graphic observahon that Imle corrosnon hud oc-

PERIOD ENDING SEPTEMBER 10, 1956

Six thermal cycling experiments have been con-
ducted in a simple heated-tube flow system in
which cylic thermal stresses are generated in the

tube. . Some preliminory and specific information
- ‘on. corrosion rates and tube strength has been

obtained, ,
The enthalpies and heot capacmes of a zirco-

nium-base fuel with additives to simulate fission

products were determined; no significant effects

of the additives were found, Viscosity measure-

ments were made on ‘@ thorium-bearing fluoride

.mixture. A preliminary thermal conductwuty corre-

lation expression has been developed which satis-

~factorily relates the conductlvmes of a number of
fused salt mleures.

4.2. Radiation Damage

Dlsassembly and examination of MTR in-pile
Ioop No. 3 has been completed, except for metal-

lographic examination of the pump impeller. Cor-

rosion of the nose coil varied from 1 mil near the
mlet to @ maximum of 3 mils near the outlet. No
mass-transferred crystalllne deposits were found.

- Diassembly of in-pile loop No. 4 has been started,

but as yet no results are available. In-pile loop
No. 5, which was inserted in the MTR but could

not be filled, was received for disassembly, Ex-
amination of the fill line revealed the probable

location of the plug which prevented filling. Fuel
samples have been obtained from varicus sections
of the system for chemical analyses.

- Chemical analyses of materials from in-pile loop
No. 3 have indicated thot the amber-colored mate-

rial found in the pump was probably an oil-decom-
137

curred, S
A series of burst tests of lnconel tubing ot a
stress ‘of 2000 psi in helium has been essentially

_completed out-of-pile and.in the LITR at a temper-
" ature of 1500°F, A decrease in the rupture life of
“the ~0.010-in.-wall, 0.191-in.-ID tubing that was
‘used for these tests appears to have resulted from
" irradiation. - - The tubing stock used gave many

indications of defects during ultrasonic nonde-

- structive testing, however, and therefore carefully

inspected tubing stock of the type to be used in

1

 
 

ANP PROJECT PkOGRESS REPORT

ART NaK-to-air radiators has been obtained for
further burst tests in helium and in fuel mixtures.

Irradiation of static corrosion capsules in the
MTR has been continued, Two Inconel capsules
containing the fuel mixture NaF.-ZrF -UF, (53.5-
40-6.5 mole %) were irradiated ot 6 kw/cm3 for
over nine weeks. One capsule released fission-
product ‘activity after a temperature excursion re-
sulting from the failure of a thermocouple, but
contemination control by freezing the fuel made it

unnecessary to interrupt MTR operation. Two-

other capsules are being irradiated at 6 kw /m3,
Five Inconel capsules have been opened, ond
samples have been submitted for chemical analysis
and metallographic examination. Hastelloy B
capsules have been filled with the fuel mixture
NaF-KF-LiF-UF and are being prepared for bench
tests and MTR irradiations.

‘A series of charcoal traps of different geometries
but with the same amount of charcoal are being
tested in radiokrypton holdup experiments. The
results of the tests will provide information needed
for the evaluation of the effects of the shape of the
trap on the holdup efficiency.

The cause of failure of the pump in the LITR
vertical in-pile loop was found to be o loose shear
pin which lodged in the impeller, threw it out of
balance, and bent the shaft until it rubbed against
the impeller housing. [n the future, internal loop
parts are to be joined by welding. Thermal mockup
and bearing tests have been run on a modified
shaft which is shorter and stiffer and which has
improved bearings, Pump performance tests were
run with water to optimize impeller clearances, and
the results were interpreted in terms of fuel
Reynolds number in a loop with the optimized
pump. A full-scale bench-test pump is being as-
sembled, and in-pile loop parts are being modified
to conform with the improved pump design.

A new electronic analog circuit has been de-
veloped for simulation of the thermal behavior of
the LITR vertical in-pile loop. A series of calcu-
letions has been made which illustrates the be-
havior of the loop when circulating the fuel mixture
NaF-ZrF ,-UF , (63-25-12-mole %) in position C-46
of the LITR. -

Electronic analog simulation was used to inves-
tigate the possibility that fluctuations in the
efficiency of removal of xenon from the fuel by the
pumps might cause troublesome reactivity tran-
sients in the ART. It was found that no transient

12

exists which has a period that is short enough to

cause concern for the controllability of the ART.

The fission yield of 65-day Zr?5 was determined
in enriched U,04 and in the fuel mixture (No. 44)
NaF-ZrF ,-UF (53.5-40-6.5 mole %). For use in
fission momtormg, the yield found for Ze95 s
0.0664 +0.0013 atom/fission.

Characteristic curves have been obtained for a
Philco surface-barrier transistor before and after a
series of “irradiations in hole 5IN of the ORNL
Graphite Reactor. Although the coflector charac-
teristic showed considerable change, the grounded-
base characteristic. family of curves remained
practically unchanged after irradiation to an inte-
grated dose of about 1072 nvt,, A model is pro-
posed to explain why transistors of different con-
struction show various sensmvmes to surfoce
conditions,

Metallographic examination of a scmple of stain-~

less-steel-clad CaB _ .Fe that had been irradiated
in the LITR showeg that the clad-core interface
had retained its continuity.” There was on evidence
of gross cracking in the core or of deformation of
the specimen. Additional somples are being irra-
diated. Irradiation tests were started in the LITR
on stainless-steel-clad boron nitride~nickel cermet
specimens at room temperature -and at 1600°F.
Experimental equipment was developed and shipped
to the MTR for the irradiation of neutron shield
materials at elevated temperatures. The first
material to be tested will be a stainlessisteel-clad
copper—boron carbide cermet. The temperature of
irradiation will be 1600°F. Boron nitride powder
samples were obtained and analyzed, and irradic-
tions of pressed bodies of this materna[ are
planned.

Creep and compression tests on copper—boron
carbide compacts have been [made at - elevated
temperatures.  High-temperature deformation was
found to strengthen the material. = Equipment for
testing stressed material during |rrad|ahon in the
LITR is being prepared.

4.3. Fuel Recovery and Reprocessing

The fused salt—fluoride volatility pilot plant is
essentially complete and shakedown tests have
started, Reliability tests were run on the freeze
valve to be used in the molten salt lines. In 120
tests with 20 psi N,, one test with 100 psi N,, and
one test with 100 psi Freon, no leckage was

113
 

 

dertected.

was designed which allows the tempercture of the
piping to be above 525°C while the temperature of
the connector to the copper cable is below 85°C.
In order to maintain the electrical -insulating gus-

kets in the freeze valve vent lines at or below

140°C, copper cooling fins were installed on the
vent lines, A molten-salt sampler was tested and
found to be satisfactory for pilot plant use,

‘The decomposition of the complex UF ¢*3NaF to
a nonvolatile product was studied over the temper-

ature range 245 to 355°C in order to evaluate the
effect of the decomposition on operation of the NaF

absorption-desorption step of the volatility process.

- At temperatures above 250°C the decomposition

rate was of sufficient magnitude to seriously in-

 crease the uranium retention on the NaF bed if the

temperature and sweep gas.flow rate specified by
the process flowsheet were not precnsely con-
trolled.

- The probable decomposmon reaction is

UF ,-3NF —> UF ;-xNaF + 0.5F,

Data obtained on the temperature dependence of

~ the reaction were successfully fitted by the ex-

pression _
log r = 6.09 — (5.22 x 10%/T) ,

where r is the decomposition rate and T the abso-

lute temperature. The activation energy for the

decomposition reaction was calculated to be

+23.9-'k¢c|/mp|e_ of UF ;.3NaF.

4.4, _Cri_ticdl_ Experiments N

‘Additional robmétempe'roture experiments ‘were

" made on a reconstructed critical assembly that
represents -the cnrculahng-fuel reflector-moderated

‘teactor. . The fuel" ‘region consists of alternate

“Tamince of Teflon and enriched uranium foil, It
-'wus found in experiments in which the outermost
loyer of _the beryllium reflector was replaced with -

stainless steel that beryllium is 2.75 times more

“effective than stainless steel in this region.

The reactivity coefficients of several materials

of engineering interest were evaluated at various
points ulong the radrus ‘at the mid-plane of the
reactor.  The results are presented as the change

in reactivity introduced by filling @ void with the
material. In the assembly being studied one end

A suitable stub for. the ‘junction of |
electrical cable to self-resistance-heated piping

PERIOD ENDING SEPTEMBER 10, 1956

duct is thicker than the other so that the effect of
end-duct thickness can be measured, :

Equipment is being fabricated for an elevated-
temperature critical experiment for investigating
the design features and nucleor characteristics of

the circulating-fuel reflector-moderated reactor

being designed by Pratt & Whitney Aircraft, The
design of the assembly is quite similar to that of
the ART high-temperature critical assembly which
was tested previously.

. PART 5. REACTQ‘R SHIELD‘NG
5.1, Shielding Theory

The results of calculations of gamma-ray energy
flux, dose rate, and buildup factors in a lead-water
shield of finite thickness are presented. A Monte
Carlo method was used for the calculations, which
included 1., 3+, and 6-Mev photons incident along a
normal ond at on angle of 60 deg. Comparisons
are made with data from the results of the moments-
method solution and from an earlier Monte Carlo
calculation for 3-Mev photons normally incident
upon a one-region finite lead shield.

The results of the calculations for radiation
incident at 60 deg indicate that the practice of
using only normal incidence data for shield designs
can lead to a poor approximation. This problem
becomes most acute when the number of mean free
paths across the shield is small or when the angular
distribution is such that a large portion of the
radiation is not normal or nearly normal to the
slab. _
5.2. Lid Tank Shielding Focility

The ‘,;tudies of udvcmced shielding materials
were continued with mockups consisting of a beryl-

- livm moderator region, a lead or depleted~uranium
‘gamma-ay shield, and a lithium hydride and oil
~ neutron shield. In some mockups a boral sheet was

inserted outside the beryllium layer to prevent
thermal neutrens from ‘entering - the gamma-ray
shield material. :

Data obtained thus for indicate a strong influence
of the placement of the various materials on.the
gamma-ray dose rate, The thermal-neutron traverses
for the various configurations, however, show the
flux to be independent of the order of the lithium
hydride and lead, except for the difference at the
beginning of each traverse caused by variations
in the amount of oil trapped between the various

13

 
 

 

 

slabs. For a similar crrangement containing de-

depleted uranium, moving the uranium out of the.

intense neutron field redyced the thermal-neutron
flux. The results of fast-neutron dose rate traverses
are being analyzed, and studies of similar cone
figurations are being continued,

53, Bulk Shielding Facility
The gamma-ray energy spectrum and time decay
characteristics of the fission products of U235
as related to circulating-fuel reactors are being
investigated, Based on the preliminary energy-
spectrum dota obtained thus far, an integration of

the spectra between 0.28 and 5.0 Mev and between
1.25 ond 1600 sec gave a total of 2.81 photons

emitted per fission with a total energy of 3.22 Mev .
. per fission, These values are to be compured with

data obtained from experiments on time decay
characteristics which gave 2,92 photons per
fission and 3.23 Mev per fission, All these values
carry an estimated error of about $25%.
Experiments are under way in an investigation of
the effect on the dose rate outside the ART lead

14

shield of gamma-ray streaming through the NaK-
filled pipes. Measurements made beyond a duct
placed through a mockup of the ART shield af an
angle of 51 deg 30 min are reported, Some measure-
ments made beyond straight-through penetrations
mocking up portions of the south-head ducts are
being studied. Measurements are now being made
in order to determine the effects of various com-
ponents in the mockups and of adding patches to
the shietding, o L

5.4, Shield Mockup Core

Nuciear calculations were made in order to pro-
vide information for detailed design of the reactor
core-reflector-island region of the Shield Mockup
Core (SMC). The SMC, c 5-Mw fixed-fuel reactor,
is being designed for shielding studies of the
circulating-fuel reflector-moderated ‘reactor, - The
leakage fluxes of the SMC, with the exception of
the fission-product radiation from the heat ex-
changer, are to be the same as those from the
reactor now being designed by Pratt & Whitney
Aircraft and designated as the PWAR-1. =
 

 

ta

)

Part 1

AIRCRAFT REACTOR ENGINEERING

S. J. Cromer

 
 

 

 

™

 

 
 

 

 

A

LU

‘ molntenonce

1.1. AIRCRAFT REACTOR TEST DESIGN
A. P. Fraas

STATUS OF ART DESIGN
Reactor Assembly

Most of the detail drawings of the reactor-core,

heat-exchanger, pump, and pressure-shell assembly -

have been completed, and the drawings have been

‘approved for procurement. Design work on the

reactor shield and the cell interior is proceeding
concurrenfly with the construction of a one-half-
scale model of the reactor- pressure-shell, NaK-

~ manifold, and lead-shield assembly, and a one-

sixth-scale model of the reactor and cell, These
models are proving most helpful in visualizing
interference problems in both assembly and in

cedures for assembly, operation, and servicing is

- continuing, with major attention being given to

vnusual contingencies.

Reflector-Moderator Cooling Circyit

A careful review of the detail drawings indicated
that the pressure drop through the reflector-modera-
tor cooling, circuit was likely to be substantially
higher than that originally estimated. In order to
provide for this probable increase in pressure drop,
larger and higher-speed motors are being ordered
for the sodium-pump drives. Equipment is being
assembled for water flow tests on the crucial ele-
ments of the sodium circuit, in porticular, the
beryllium cooling holes and annuli and the mani-

fold ‘at the inlet. to the beryllium. - ‘Preliminary .
‘calculations indicate' that the increased pressure
drop -will ‘not present any serious problems’ from -
~the stress sfondpomt but the effects are bemg
,checked : S :

Heat Exchangers i

A greot deal of work and experlmentohon by 1he_",- D
heat exchanger vendors has revealed that it will.
be extremely cosfly in ‘time and money ‘to’ meet S
"the close dimensional . folerances - specrfled A
review’ of the design indicates that ‘certain’ toier- -

ances may be- reloxed with only minor decreases
in performance
vendors it is believed that compromises may be
reached that will permit production of heat ex-
changers that will give acceptable performance}

The preparation of written pro-.

By working closeiy ‘with -the

Fuel Pumps ond Expansion Tank

Further test work on the fuel pump prepared for
the high-temperature test performance rigs has
indicated that the cavitation limit is somewhat
lower than had been anticipated from the early
water tests. This condition could be improved
substantially, as indicated by the early water
tests, through the use of a modified impeller vane,
which would be somewhat more, but not unreason-
ably, difficult to fabricate. Indications are, how-
ever, that the increased cavitation suppression
head required to obtain satisfactory operation with
the present impellers is not excessive and can be
accommodated readily by increasing the fuel ex-
pansion tank pressuwe by about 10 psi. This
problem is still being investigated (see Chap. 1.4,
**Component Development and Testing’’).

Tests on the north-head fuel pump and expansion
tank assembly have shown that satisfactory op-
eration can be obtained with fuel levels in the fuel
expansion tank running from ]/2 to 3 in. with no
objectionable aeration or pressure fluctuations in
the system over the entire speed range. Test work
is being continued with this setup in an effort to
reduce the pressure difference between the fuel
pump inlet and the expansion tank, and to improve
its performance with one pump out or with one
pump at a substantially different speed from the
other. Reduction of the pressure differential be-
tween the pump inlet and the expansion tank is

important - because this pressure differential will

largely determine the stresses in the upper and
lower decks of the north head. It is believed that

 rather- snioll modifications in the details of the

bleed passages coupling the pump inlet with the

" exponslon tank will permit a very marked reduction
in this _pressure dlfferentral

Fuel Recovery Tonk

The prehmmory desrgn of the fuel recovery tonk

" has been completed. The ‘tank has been designed
“on the premise that no electrical, cooling, or other
lines need be connected to ithe tonk at the time
~that it is removed from the cell. . This means that
" the - afterheat being generated in the fuel must

balance the heat losses to convection and radi-

ation to hold the temperature in the fuel between

17

 
 

 

 

ANP PROJECT PROGRESS REPORT

its melting point and 1600°F. Because of the
sharp increase in radiation heat losses with tem-
perature this provides a fairly wide latitude. In
addition, the tank has been designed so that the
amount of air circulating up between the four 8-in.-
dia fuel tanks and the lead shield will be con-
trollable, and hence the air flow through this
stack-type of thermal-convection cooling system
can be varied. It is believed that in this way it

~.will be possible to hold the fuel temperature in the

tank within the desired limits throughout a period
of approximately 30 days, starting about 10 days
after shutdown. Removal of the fuel from the cell
to the reprocessing facility within 10 days of time
‘of shutdown has been considered to be very im-
portant as a demonstration of the practicality of
such an operation and, hence, the possibilities for

operating with a low fuel inventory in a complete

operational system.

APPLIED MECHANICS AND STRESS ANALYSIS

R. V. Meghreblian
Reactor Support

The complete reactor and shield assembly is to
be suspended from an overhead bridgelike struc-
ture, as shown in Fig. 1.1.1. Although the shield
components will be in close contact with the outer
shell of the reactor, the weight of these members
will not be carried by the reactor. A separate
system of tension members is to be provided for
this purpose, and these members will transmit the
shield loads directly to the bridge structure. The
weight of the reactor will be transmitted to the
bridge structure through the four pump barrels.
The attachment of the individual barrels to the
bridge will allow horizontal motion of the barrels
in order to accommodate the relative thermal
growth between the reactor {at operating tempera-
ture) and the bridge (ot room temperature).  Verti-
cal motlon of the barrels will be restrained through
Fabreeka! pads, which are sufficiently resilient to
distribute the load fairly uniformly between the
four barrels. The bridge will be fixed ot each end

 

Irabreeka is o material composed of layers of tightly
twisted, closely woven, lightweight cotton duck
Ihoroughl impregnated with a special rubber com-
pound. This material is manufactured by the Fabreeka
Products Company, Inc., Boston, Mass,

18

to a flexible column consisting of 1.0-in.-thick
steel plates, 28 in. wide and 123 in. long. The
load carried by each column will be approximately
45,000 Ib. This value is somewhere between one-
fourth and one-half the load required to cripple the
column. A precise figure cannot be given since
the "‘end condition’ of the top of the column is
not well-defined.

The principal function of the flemble co!umns is
to allow the NaK lines, which remove the heat
from the reactor, complete freedom when expanding

from room temperature to the operating temperature -

of the reactor (Fig. 1.1.2). During full-power op-
eration, the upper row of NaK lines will be at
1070°F and the lower row at 1500°F. This will
result in horizontal growth of 0.75 in. in the upper
lines and 1.125 in. in the lower. If the reactor
were mounted in the cold condition precisely over
the center line of the column bases, these expan-
sions would translate and rotate the reactor out of
the neutral position and thereby cause bending
stresses in the columns (Fig. 1.1.22). The present
plan is to precut the NoK lines short by the
amounts mentioned above so that at room tempera-
twe the reactor will be located 1.0 in. off the
neutral position. This displacement will be toward
the cell wall through which the NaK lines enter
(Fig. 1.1.2¢c). As the reactor heats up, the NaK
tines will expand and move the reactor back to the
neutral position and thus remove the bending loads
on the columns and the axial loads in the pipes

(Fig. 1.1.25).

NaK Piping Inside Reactor Cell

The flexible columns described above will allow
for gross expansions of the NaK lines, but they
do not provide for differential expansion between
the lines of any one row. In order to provide some
margin for operational incidents and accidents?

and greater freedom in controlling NaK tempera-

tures, some additional flexibility has been intro-
duced into the piping inside the cell. This has
been accomplished by the addition of several
bends in each line. The bends will accommodate
300 to 400°F temperature differences between

adjacent lines. A horizontal view of the NaK

 

2For example, the failure of the pump in any given
NaK circuit would cause o change in the temperature
of the pipes, and this in turn would produce thermal
deformations in the piping relative to the plpmg in the
other NaK circuits.

'y
 

CONSDEMTHAN
ORNL—LR—-DWG 16123

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

REACTOR CELL™|

 

 

 

 

 

 

 

 

- \—Nal( LINES ‘.’

121t 8%/, in.

 

 

 

 

 

-

   

b 5 ft 9% in.——

H._ " \— L WATER SHIELD CONTAINER
Bemmmgeee] N FLEXIBLE COLUMN

 

 

   

 

 

 

5ft 05 in,

l Y

Fig. 1.1.1. ART Support Structure,

 

  

 

 

61

9661 ‘0l ¥IGWILJIS ONIANI QOI¥3d

 

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

 

~NoK LINE

N

 

TTIhNH

 

AT -TEMPERATURE, WITHOUT
PRECUTTING OF NaK LINES

(o}

  

\cowmas

 

© SONPIDENTA. .
.~ ORNL-LR-DWG 16124

PUMP BARRELS

REACTOR

 

77 77777 N ' %

AT TEMPERATURE, WITH
PRECUTTING OF NaK LINES

(&)

POSITION OF REACTOR AT ROOM TEMPERATURE,
WITH NoK LINES PRECUT

{c)

Fig. 1.1.2, Movement of Renctor‘ Cau#ed by Expansion of NaK Lines,

piping inside the test cell is presented in Fig.
1.1.3. This layout was selected on the basis of a
hand calculation for which the method of Spielvogel3
was used. A more exacting analysis, which in-
cludes a careful study of the manifolding system,
is now under way at the M. W. Kellogg Co. These
calculations are being carried out with the aid of
fast computing machines.

A general schematic diagram of a complete NaK
circuit is shown in Fig. 1.1.4. The essential
feature of this system is that the. NaK lines are
rigidly attached at only one point, the reactor cell

wall. The ends of each circuit, which terminate

at the reactor and at the pump-radiator assembly,
have considerable freedom to expand. The only

constraints imposed on individual lines of a given -

circuit arise from the differential thermal growth
between the various lines as a result of differences
in length and irregularities in temperature control.
The NaK piping external to the reactor cell was
- described previously.4 ‘

 

35. W. Spielvogel, Piping Stress Calculations Simpli-
fied, 4th ed., Loke Success, New York, 1951.

4R, V. Meghreblian, ANP Quar. Prog. Rep. March 10,
1956, ORNL-2061, p 22, esp Fig. 1.7.

20

The principal stresses in the NaK lines will be
due to the therma!l effects described above. Each
time the reactor system passes from one operating
condition to another, there will be a redistribution
of the thermal stresses in the various lines. The
stresses produced by relative thermal growths do

not persist for all time, however. Relaxation tests

(see Chap. 3.5, ‘“Mechanical Properties Studies'’)
indicate that stresses due to a fixed strain, such
as that present here, decay rapidly. This decay
takes place as more and more of the initial strain
on the member is transformed to plastic deformo-

tion. The decay is sufficiently rapid that the

plastic strain developed in a matter of minutes
after loading is a significant fraction of that which
would be developed in several hours. Thus with
each change in operating conditions, the thermally
loaded members will be subjected to some plastic

deformation. An important criterion for design,

then, is the strain-cycle life to which a given
member can be subjected.

The strain-cycle criterion was used as a basis’

for the design of the NaK lines. The objective of

various analyses now under way, therefore, is the

determination of the plastic strain that will be
 

 

 

o

developed with each chonge in operating condi-
tions. The results will then be compared with
strain-cycling data by using a Coffin® type of cor-
relation. Some data for Inconel are presently
available,® and additional information is now being
collected ot the University of Alabama and at

ORNL.

 

L. F. Coffin, Jr., ""A Study of the Effects of Cyclic
Thermal Stresses on a Ductile Metal," Trans. Am Soc.

Mech. Engrs. 76, 931 (1954).

6pratt & Whitney Aircraft, Nuclear Propulsion Program
Engineering Progress Report No. 18, PWAC-554, Oct, 1,
1955-Dec. 31, 1955, p 63, Fig. 21.

 

 

 

 

PERIOD ENDING SEPTEMBER 10, 1956

Fill-and-Drain Tank and Support

- On the basis of the present design and the op-
erational philosophy of the ART system, there is
some reason to believe that the safest position for
the fuel during the normal progress of the experi-
ment will be in the reactor proper, because in the
event of trouble it will be easier to dump the fuel
into the fill-and-drain tank than to pressurize the
fuel back into the reactor. It has been suggested
that this may also be the safest place for the fuel
during certain off-design situations, such as the
one-pump-out condition, partly for the same reason

SESREP
ORNL-LR-DWG 16022

CELL WALL

 

8-in. PIPE ‘
. _ i

Fig. 1.1.3. NaK Piping in the Reactor Cell,

21

 
 

 

 

 

 

- ANP PROJECT PROGRESS REPORT -

 

 

1

|_—THERMAL SLEEVE

~p————— {070° F NoK

. ORNL-LR=-DWG 16125

  
 
    

_CONSTANT LOAD ,
SPRING HANGERS )

MOTION ALLOWED

 

 

 

 

 

 

REACTOR

 

 

 

 

 

 

 

 

 

 

    
   

— FLEXIBLE
COLUMNS

 

 

 

 

————— 1500 °F NaK

 

- : RADIATOR AND
PUMP ASSEMBLY

A MOTION ALLOWED BY FLEXIBLE COLUMNS.
B MOTION ALLOWED BY RADIATOR-PUMP ASSEMBLY.
C FIXED POINT IN NaK PIPING.

Fig. 1.1.4. Over-cll Suspension and Constraints on NaK Piping System,

and partly because of the increased probability of
a failure duwring the switching operations associ-
ated with an emergency condition. A study is now
under way to determine the validity of these
premises. In addition, there are several off-design
conditions and accidents in which it will be
clearly unsafe to leave the fuel in the reactor; for
example, if there is an internal leak between the
fuel and sodium (or NaK) systems, a chemical
reaction might occur which could produce metallic
vranium.” Under certain operating conditions, the
vranivm could be deposited throughout the hot
portions of the fuel circuit. In order to control
these reactions and to minimize their effects, it
would be necessary to remove the fuel from the
reactor. '

A fuel fill-and-drain tank for receiving the fuel
is incorporated in the ART system design (Fig.
1.1.5). The tank will be an Inconel cylinder 38 in.

in diameter and 38 in. in length, which will be

kept at a steady-state temperature of 1100°F by

means of a continuous flow of NaK from two in-
dependent external circuits. These circuits have

 

7W. B. Cottrell et al,, Airc.raft Reactor Test Hazards
Summary Report, p 38, ORNL-1835 (Jan. 19, 1955).

22

SURPIBENTA
ORNL~LR-DWG 16126

NaK SUPPLY—1 j,
(1100°F)

     
  
  
  
 
 
  
 

FUEL DRAIN

'/ (1200°F)

~~— INNER CYLINDER
{He BLANKET AT 45 psi )

BAFFLES {FOR SUPPORT
AND FLOW DIVISION)

SUPPORT CYLINDER z
] (5000 1b UPLOAD) .

Fig. 1.1.5. Fuel Fill-and-Drain Tank. U
 

A i ot b e

w)

several functions. First, they will remove the
decay afterheat from the fuel following a dump;
second, they will keep the fuel above its freezing
point at long times after the dump has occurred
and the decay heat has been dissipated; and, third,
they will maintain the tank at a sufficiently high
temperature to minimize thermal shock effects
when a dump takes place.

The design heat load for the tank is ]75 Mw

even though it is estimated that the decay heat

at the instant of shutdown after operation at 60

~ Mw for one month would be 3.6 Mw. This reduced

operational requirement is based. on the premise

" that the fuel will be cooled,in th:e‘ Vreoctor"to
1200°F before a dump and that @ minimum of 8 sec -
‘will elapse after the coolant rod ‘has been fully

inserted and before the dump valves are opened.

~ Each of the two NaK cooling circuits will be

capable of extracting the 1.75 Mw of decay heat
under these conditions; however, it is expected
that under one-circuit operation, the fuel and metal
temperatures at certain locations within the tank

. may be as high-as 2000°F. Such short-time effects

will ‘not be encountered during o normal dump.
During normal operation, both circuits will be in
use, and the maximum metal temperature is not

~expected to exceed 1600°F.

The primary structural loads within the tonk will
be due to the NaK pumping pressure of 40 psi.

- Since the tank is to serve as an ever-safe deposi-
tory for the fuel, it is necessary that it be capable .

of surviving some 2000 hr of continuous operation

at temperature and possibly several fast dumps.

The principal structural requirements of the tank,

then, - are based on. creep-rupture ond thermol-:_____};

shock cons lderotrons

Analyses show thot the most sensmve areas in -
this design are at the - joints’ between the tube_,f-
header sheets ond the mner cyllnder The dis-"
continuity stresses in the'se areas are of the order_ o
of 6000 psi. This is a relotlvely hlgh stress level,
but it is- expected thot these stresses er be_‘f,:: _
_'_'Q_fi'tlon (1200°F tsothermol) to- full .power: some 30 "
times.” One cycle consists of 8 hr ot idle ‘and’
16 hr at tuil power. In- possmg from the idle condi-
- tion_to full power - the temperature dnstributlon in

markedly reduced os the ‘metal-in_ that region de- -
forms in- creep. Slnce “the extent to which" dlS-_‘f.'»_
conhnmty stresses can relax is not known, a test
__program’is under way to study this effect by means ..
~of tube-burst tests. Fmolly, because -of ‘the im-~ *
'portunce of thls component to the over—o“ sofety,
of the experiment, it'is planned to test the entire

tank assembly by using one of the NaK circuits.
The test should demonstrate the adequacy of the

PERIOD ENDING SEPTEMBER 10, 1956

design in withstanding the creep and thermal shock
effects mentioned above.

The support of the fill-and-drain tank assembly
will be accomplished by means of a nitrogen
cylinder located beneath the tank (Fig. 1.1.6).
Although the tank will be attached rigidly to the
reactor pressure shell through the fuel drain lines,
the major portion of the tank weight will not be
allowed to bear on the shell. The total weight of

~the tank, including the fuel, will be 6000 Ib, of

which 5000 b will be carried by the nitrogen
cylinder. The weight of the tank without the fuel

" will be 4000 Ib. This support arrangement intro-

duces some complication in the stability aspects

~ of the system, but calculations show that the

proposed distribution of lever arms, structural
stiffness of the drain lines, and the weights are
‘well within the stability limits of the system.

In the event of the failure of the nitrogen cylin-
der, all the tank weight will be exerted upon the
drain line and reactor pressure shell, This situa-
tion will produce high local stresses in the shell
at the points where the two drain lines will be
attached and will add an additional 1000 Ib to the
load in each pump barrel. Although this will re-
duce the design safety margin in these com-
ponents, the system can support the tank in this
manner. The purpose of the nitrogen cylinder,

- then, is to reduce the operating loads in the

barrels and in the pressure shell.

Core-Shell Low-Frequency Thermal-Cycling Test

A one-fourth-scale model of the core shells is

. presently being subjected to the long-time tempera-
ture variations expected in the reactor as a result
- “of chonges in power level (see Chap. 1.4, ‘‘Com-
"f'_-ponent Development and Testing’’). Such tempera-
~ ture ‘variations: produce thermal . distortions in the:

shells, “and - these effects. are among the most
severe condntlons ‘imposed on the shells.” The

"'present operating :plan for “the ART anticipates

that ‘the reacfor will be Cycled from the idle condi-

the “core shells will change from a unn‘orm prof:le‘-

" to a linear drstnbutlon, ‘with a maximum grodlent_

' of about 300°F through the wall.

The cyclic conditions described above have been
simulated in the one-fourth-scale model of the

23

 
 

 

 

ANP PROJECT PROGRESS REPORT

SUPPORT BRIDGE

 

 

 

  
  

DUMP VALVE

 

 

 

  
  

(( |
N

aK SUPPLY

ATTACHMENT ————— o=~
TO FLOOR SES

Fig. 1.1.6. Fuel Fill-and-Drain Tank Support,

24

: CONFIDENTIAL )
ORNL- LR~DWG 16127

C‘
 

 

outer core shell by means of a dual NaK circuit
which heats the. inner surface of the shell and
cools the outer surface. The test cycle selected
was | hr at power and 1 hr |sothermal at 1200°F.
At the power condition the shell is exposed to a
maximum temperature gradient of--300°F through

" the wall. The stress buildup and relaxation with

each change in the operating conditions is illus-
trated in Fig. 1.1.7. The solid-line curve gives
the stress-time variation expected in the actual
shells during the operation of the reactor. The
broken-line curve gives the stress hme hlstory for

the ‘model.

seerrT

 

 

 

 

 

 

 

 

 

ORNL ~LR-DWG 16128
T 17 1T b 1 1 |
l"""" f\‘ MODEL
NS
REACTOR
o [ £
& 0 —1
A
7
\/
|
V y
b
0 2 4 6 8 10
TIME {hr)

Flg. 1.1.7. Stress.Time History of Model Used-
for Core-Shell Thermal-Cycling Tests in Com-

ith Stress- tory E ted in the - =
panson Al "es T_We .I'_irs ory :_cpec”e rn € an upprecnable fraction of the applied deformation

owill have been converted ‘to_plastic’ strain. On
”__r_)thas basis it was urgued that the cycle times for
_ ¢ ‘relax ' it the- ‘model test’ could be reduced to.the 1-hr inter-
was ¥ound fhut because of the very ropad decoy of
©the stress it was not.necessary to design the core-
S "'she“ test for fhe 8- to 16-hr Cycle time: pianned_'}-i_-'f
" for the reactor. - “The 2hr- Cycle time was:found to

 
 
 

7 be suffrcrem‘ to. develop a substan‘hal fraction of -
< the plastic “deformation. expected in the- octual"‘
- shells, A" typlcol set of: relaxoflon ‘curves’ for
~ “Inconel,. "as. obtained" from various- sources, iso
'_';,presenteds,ln Flg 11,8 The curves lobeled;:‘-f
'l}f"ORNL" were obtained from ‘measurements -made -

" by the ‘Metallurgy DIVISIOI‘I of ORNL; the curves

labeled *‘University of Michigan’’ are based on

measurements made at the Engineering Research -

 

PERIOD ENDING SEPTEMBER 10, 1956

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

SONEDENTIET"
. ORNL~LR~-DWG (6329
25,000 E | !
INITIAL STRESS (24,000 psi)|
20,000
\ L —UNIVERSITY OF MICHIGAN
15,000 ~—
o ‘-——L\-k INITIAL STRESS (12,000 psi)
e N \v FROM CREEP DATA
B W\ —~—
10,000 Rt RNL e = ORNL
S SSemea [
—_—i _-:_:-___.:_::___-_.-. . ——
"'z"“—‘-'r'-':-‘:‘.-;_-:;:;::“:::_
L —— ]
UNIVERSITY OF MICHIGAN—
5000
0
0 04 02 03 04 05 06 07

TIME (hr)

Fig. 1.1.8. Relaxation Data for Inconel at

1300°F,

Institute of the University of Michigan; and the
curves labeled **From Creep Data’’ were synthe-
sized from creep curves.

It is evident from these curves that within a

‘matter - of minutes after the start of the test, be-

cause of the extremely rapid decay of the stress,

vals bemg used, since opproxrmotely 70 to 90% of
the - total strain’ deveioped in 810 16 hr could be

_ioch|eved in L hr. . In this way it was possible to
" reduce - the totol model ?est time - by a chtor of
Cabow 0

" The flrst of rhese Iow frequency thermul-cycllng

tests has been completed “The ‘model was sub-
|ected to. 300 full cycles a factor of 10 more ‘than
) expected in:operation of the ART, and appeors to

have survived without gross fcnlure It is believed

“that the factor of 10 increase in the number of

test cycles, in comparison with the cycles ex-

" pected in the ART, is not an overly conservative

25

 
 

 

 

ANP PROJECT PROGRESS REPORT

margin for tests of so vital a structural component.
A complete metallurgical examination of the shell

-has not yet been completed. [t should be men-

tioned that the test program was interrupted after
57 cycles because of a leak in the external cir-
cuits and the specimen was allowed to cool to
room temperature. At the completion of repairs,
the model was brought to temperature again, and
the test was resumed.

CORE HYDRODYNAMICS

W. J. Stelzman W. T. Furgerson

Further core flow studies were made on the full.-
scale model of the ART core by using the tech-
niques described previously.® The configuration

.sG. D. Whitman, W. J. Stelzman, and W. T. Furgerson,
ANP Quar. Prog. Rep. Dec. 10, 1955, ORNL-2012, p 23.

 

 

EXPANSION JOINT

FUEL PUMP VOLUTE

FUEL FLOW ' l

TN
'

 

 

 

 

 

 

 

 

 

tested consisted of swirl-type header No. 2 with «
simulated conical island expansion joint (Fig.
1.1.9). Tests were run with and without inlet
guide vanes.. The inlet-guide-vane system tested
consisted of a set of 14 paired vanes designed by
G. F. Wislicenus? to induce radial mixing in the
fuel annulus. The direction of rotation of the
vortexes was to be alternately clockwise and
counterclockwise around the annulus. = -
The flow pattern produced by the blades agreed
with the design pattern in approximately the upper

“third of the core annulus, but the individual vortex

cores rapidly degenerated into a region of low-
velocity random motion at the equator. As the flow
accelerated in the portion of the core belew the

 

9Consulfant.

A
ORNL~LR-DWG 16130

FUEL PUMP VOLUTE

 

 

 

- OUTER CORE SHELL

ISLAND WALL

Fig. 1.1.9. Swirl-Type Core Header No. 2.

26

xp
 

equator it plcked up a pattern of rotation centered
around the core axis and was, in general, typical
of the flow obtained in this region with other inlet
configurations.

A lack of circumferential symmetry was noted in
the flow pattern produced by these guide vanes.

Previous configurations have not been perfectly -

symmetrical in this respect, but this system ap-
pears to be worse than the others. It is possible

that the vanes themselves were not symmetrical.

They are complex in shape and have no suitable
reference surface to use in effecting their align-
ment.  This analysis of the flow is based on
observations of injected dye, and there are no

quantitative data on this system. The decidedly

three-dimensional character of the flow made the
use of claw-probe traverses impractical.

Complete probe traverses were made of the same
core-header combination without inlet guide vanes.
Circumferential  asymmetry of the axial velocity
profiles, as indicated by the maximum deviation of
the reading of either. of two probes at each station
from the average, was as follows:

Maximum Deviation

‘ Station h . (%)
8 10
7 10
6 20
5 20

PERIOD ENDING SEPTEMBER 10, 1956

22

- A W M
=]

- The profiles given by the two probes at each sta-
‘tion were, in general, similar in shape.

In an effort to obtain information on the heat
conduction to be expected through the core shells,
local heat transfer coefficients were calculated by

Reynolds number analogy from the probe survey

data extrapolated to the walls by means of uni-
versal velocity distribution equations.'®  The
local coefficients thus obtained were then inte-
grated from the wall out into the stream to obtain
the film conductivity. The integration was carried
out for an arbitrary distance of 0.010 in. from the
wall. This dimension was chosen because further
depth did not result in an appreciable change in
the film conductivity. The fuel assumed for these
caleculations was the mixture (No. 30) NaF- ZrF -
UF (50-46-4 mole %). The heat transfer coefhcu-
enfs for the 0.010-in.-thick fluid layer adjacent to
the walls, as obtained for the swirl-type header
No. 2 without inlet guide vones and with GS-2
guide wvanes (Fig. 1.1.10), are presented in

~ Table 1.1.1.

 

‘°w H. McAdams, Heat Transfer, 3rd ed., p 209,
McGraw-Hill, N. Y., 1954,

TABLE 1.1.1. HEAT TRANSFER COEFFICIENTS FOR 0.010-in.-THICK FUEL LAYER ADJACENT TO OUTER
' AND INNER CORE WALLS OBTAINED FOR SWIRL-TYPE HEADER NO. 2
WITH AND WITHOUT INLET GUIDE VANES

Fuel mixture: NcF-ZrF4-UF4 (50-46-4 mole %)

 

Heat Transfer Coefficients (Btu/hr-ft 20 F)

 

 

GS-2 Guide Vanes

 

 

Sfufiofi ' ' No Guide Vanes -
' | Outer Wall '~ Inner Wall Outer Wall - lnper Wall
8 4986 -2 5411 , 4913
7 4631 2469 2418 2066
6 3800 3182 2651 2526
5 3394 3243 2601 : 22
4 3032 4265 2559 2348
3 4959 : 4865 2494 2880
2 5678 5310 5794 3706
1 6193 5660 Not available - Not available

 

 
 

 

 

 

 
 
  

Fig. 1.1.10. Guide Vane
of ART Core.

28

-ANP PROJECT PROGRESS REPORT

‘UNCLASSIFIED.
_ PHOTO'25581

and Baffle-Plate Arrangement (GS-2) for Distribution of Flow in Plastic Model
 

 

 

PERIOD ENDING SEPTEMBER 10, 1956

1.2, ART PHYSICS
A. M. Perry

COMPARISON OF BASIC GAMMA-RAY DATA WITH
BULK SHIELDING REACTOR GAMMA
HEATING MEASUREMENTS

H. W. Bertini C. M. Copenhaver

The series of experiments performed to determine
the rate of gamma heating in various target mate-
rials near the Bulk Shielding Facility (BSF) re-
actor ! afforded an opportunity for a direct verifica-
tion of the basic gamma-ray data and the method
used for the ART heat-deposition calculations.?

Such a comparison is of considerable interest,

since the agreement between the calculated results

and the results for the BSF experiments tends to

confirm the methods and data used in the ART
calculations, :
Experimental heat-generahon rates were measured

in aluminum, iron, ‘and lead samples at several

distances along the center line of the reactor, as
illustrated by Fig. 1.2.1. In the experiments,
measurements were made of the transient tempera-
tures of the samples during heating in the gamma-
ray field and during cooling after removal from the
gamma-ray field,

The calculated heat-generation rates indicate
that the energy deposition in the target metals is
due to the following principal radiation components:
1. core gamma rays,

2. reflector gamma rays (water capture gamma
rays),
3. target gamma and beta rays.
Since on Oracle code for the calcuiahon of

gomma heating in reactors of rectangular geom-

etry3+4 was available, the heat generation rate in
the target samples caused by gamma rays emanating
from the core was determined. The heat generation

 

1k, T. Binford, E, S, Bettis, and J, T. Howe, Gamma
Heating Measurements in_the Bulk Sb:eld’mg Reactor.
ORNL CF~56-3-72 (March 7, 1956). :

2I-i. V. Bertini et al,, Basic Gamma-Ray Data for
ART Heat - Depos:tzon, Calculauons. - ORNL-2113
(Sepi'. 17, 1956). = =

3. C. Claiborne and T. B. Fowler, Tbe Calculatzon

of Gamma Heating in Reactors of Rectangulozd Geometry.
ORNL CF-56-7-97 (July 20, 1956) - '

41. B. Fowler and H. C. Clcuborne, Operating Manualx

for Oracle Code No. 243: The Calculation of Gamma
Heating in Reactors of Rectangular Geometry, ORNL
CF-56-8-126 (Aug. 20, 1956).

- HEAT GENERATION (w/g OF MATERIAL, NORMALIZED TO A REACTOR POWER OF 1 Mw)

in the target caused by gamma rays from the ith
energy group, H(V), for one volume element of the
core is

K N/{V) E, P
Hz(V) = . e-x Ba — s
4 [r (V) + (V)12 P

 

UNCLASSIFIED
ORNL-LR-DWG {6282

=0 EXPERIMENTIAL (REF {)
-==® CALCULATED

05 |

0.2

Od

ALUMINUM

0o
O
o

o
O
r

 

0.014
0 2 4 6 8 10

DISTANCE FROM CORE FACE (in.)

Fige 1.2.1. Gomma-Ray Hecting in Metals at
Various Distonces from the Core Face of the
BSF Reactor,

29

 
 

 

 

ANP PROJECT PROGRESS REPORT

where

a constant which normalizes the thermal

flux to 1 Mw and converts Mev/sec to

watts, .

N,(V) = number of gamma rays per second in the

ith energy group for the particular vol-
ume element, V,

E. = average energy of the ith energy group,
= radii of core and reflector, respectively,
x = pr(V} + pr(V), where p_r (V) and
p,r (V) are the number of mean free paths
in a volume element, V, of the core and

reflector, respectively,

B_ = A(e™ ~ 1) + 1, the energy absorption
buildup factor (the constants A and a
were evaluated for the materials of in-
terest in the report ORNL-2113),2

‘pg = energy absorption cross section for
gamma rays for material being irradi-
ated, '

p = density of material being irradiated.

~
It

-~
0
-

~
-
{

Appropriate B, data were then calculated by
using an equivalent atomic number and interpo-
lating the available buildup factor datas for point
isotropic sources in homogeneous mixtures, Since
the effective Z of the core is 9.9 and the effective
Z of the reflector is 7.5, the homogeneous mixture
approximation is valid, ond an equivalent Z of 9
was used.

The BSF reactor is a thermal reactor, and there-
fore the significant gamma radiation results from
thermal neutron fission or capture. Thus, the
spectrum of prompt gamma rays plus U235 capture
gamma rays is 8.8 e~ 1-01E photons/Mev-fission.2
For 5500 kw/hr of reactor operation per day,$ the
average ~daily fission-product decay gamma-ray
energy would be about 85% of the saturation value
and the resulting decay gamma-ray spectrum is
8.92 ¢~1.33F photons/Mev.fission. The gamma-
ray energies per thermal neutron capture in alumi-
num for the various intervals from 3 to 10 Mev were
obtained by summing up the individual contributions
for each interval from a paper by Kinsey, Bar-
tholomew, and Walker.” An additional energy of

 

5H. Goldstein and J. E. Wilkins, Jr., Calculation of
the Penetration of Gamma Rays, Final Report, NYO-
3075 (June 30, 1954).

Se., B. Johnson, private communication (1956) to
the authors,

7B, B, Kinsey, G. .A. Bartholomew and W, H. Walker,
Phys. Rev, 83, 519 (1951).

30

1.80 Mev was added to the third group (1.5 to
3.0 Mev) to account for the decay gamma-ray energy
resulting from Al28 (half life, 2.3 min). Thermal-
neutron capture by water gives rise to a single
gamma ray, 2,23 Mev, per capture by hydrogen,

The heating in the target sample by capture
gamma rays emanating from the water reflector was
obtained by using disk sources 2 in. thick and
40 ¢m in radius. An average flux in the disk was
obtained from the qverage flux along the disk
center line, determined experimentally, ® as a func-
tion of distance from the reactor north fece, multi-
plied by 0.8 to account for the decrease in flux in
the disk radially. -

In computing the heating in the target sample
from self-absorption of gamma and beta rays re-
sulting from thermal-neutron capture in the targets,
local perturbations in flux caused by the presence
of the target sample were not considered. The only
significant beta-ray source in the target samples
was the 2,87-Mev beta ray resulting from the decay
of Al?8, This beta-ray source was assumed to be
entirely obsorbed in the aluminum somple. The
fraction of gamma rays absorbed in the target
sample was obtained from an expression by Storm
et al.® based on the straight-ahead scattering ap-
proximation for an infinite cylinder containing a
vniform source distribution. Since the expression
neglects multiple scattering effects, the energy
absorbed in each cylinder was taken as the mean
of the values calculated by using the energy-
absorption and the total-absorption cross sections.

A comparison of the experimental and calculated
results is shown in Fig. 1.2.1. The rounding of the
calculated curves close to the reactor surface
results from the peaking of the thermal fiux curve
in the reflector region. The magnitudes of the ex-
perimental and calculated values agree quite well
where the Z of the target metal is close to the Z of
the surrounding mixture, Z = 9, but the agreement
becomes worse as the difference in the atomic
numbers increases. For materials having similar
Z's, the equivalent Z method should yield good
results, as confirmed by the aluminum sample.
However, for a heavy Z material following a light
Z material, the large, low-energy spectrum built up
in the light Z material is highly absorbed in the
first mean free path of the heavy Z material, ond o

 

Sm. L. Storm, H. Hurwitz, Jr., and G. M. Roe, Gamma-
Ray Absorption Distributions, for Plane, Spherical,
and Cylindrical Geometrics, KAPL-783 (July 24, 1952).
 

 

buildup factor based upon @ homogeneous mixture
will yield a low result that becomes less and less
accurate as the depth of penetration in the light
material increases. This light-heavy sequence
phenomenon has also been indicated by Monte
Carlo calculations of gomma-ray energy deposition
for water-lead slabs. ?

In view of the good agreement between the exper-

imental ond calculated heat-generation values for

- samples near the BSF reactor, it seems reasonable

to conclude that the basic gamma-ray data given
elsewhere,? ysed in conjunction with the Pratt &
Whitney gamma-ray deposition code, should give
representative results for the gamma-ray heat de-

_position in various parts of the ART.

GAMMA-RAY HEATING IN ART

R. B. Stevenson

Gamma-Ray Source Strength

The gamma-ray source strengths of the island,
core shells, fuel, and reflector-moderator of the
ART were determined so that a calculation of the
heating due to these gamma rays could be carried
out. The prompt, decay, and nonfission capture
gamma rays in the fuel, the gamma rays caused by
inelastic neutron scattering in the fuel, and the

caopture gamma rays in the Inconel core shells,

beryllium, and the sodium coolant were taken into
account in the calculation of the gamma-ray source
strengths., |

The capture gamma rays and the gamma rays
caused by inelastic scattering were computed by
using neutron fluxes determined by a Curtiss-Wright
two-dimensional neutron calculation for the ART.

The gomma rays resulting from inelastic scattering

were assumed to be emitted with an energy of
1 Mev. The distribution of the prompt and decay
gamma rays in the fuel was calculated from the
fission power distribution determined from. the

Curtiss-Wright two-dimensional neutron calculation -

and from an analysis of the hlgh- emperature criti-
cal experiment, 10 '

In both the Curflss-erght two-dnmens:onol cal-
culation and the determination of the gamma-ray
source strength, the ART was divided intc a large
number - of regions, the materials in each region
being conmdered to be homogeneous for ease of

 

9s. Auslender, ANP Quar. Prog. Rep. Marcb 10, 1956.
ORNL-2061, p 223.

wA. M. Perty, Fission Power Distribution in the ART,
ORNL CF-56-1-172 (Jon. 25, 1956).

PERIOD ENDING SEPTEMBER 10, 1956

computation. The basic data for the calculation of
the gamma-ray sources were taken from a report by
H. W. Bertini et al. 11

The results of these calculations are shown in
Figs. 1.2.2, 1.2.3, and 1.2.4. The distribution of
the gamma-ray sources in the island, fuel, and
reflector-moderator is shown in Fig. 1.2.2. The
values given, in w/em3, are the sums over the
seven energy groups actually used. The distribu-
tion of the gamma-ray source strengthin the Inconel
core shells as a function of the distance from the
equatorial plane is given in Fig. 1.2.3, while
Fig. 1.2.4 gives the source strength in the region
adjacent to the control rod (a combination of sedium
and Inconel) as a function of distance from the
equatorial plane. The values given in these plots
are felt to be fairly accurate, although there may
be some discrepancies near the Inconel core shells
due to the approximations in geometry used in the
neutron flux calculations.

There is some ambiguity in the source strength
values assigned to the Inconel core shells, and the
values may be in error by as much as 20%. This
ambiguity arises from the fact that use was made
of two different two-dimensional neutron calcula-
tions in determining the core shell neutron absorp-
tions. One neutron calculation was for the high-
temperature critical experiment, which contained
no sodium, while the second was for the ART with
four times more sodium than will be present, The
resulting neutron absorptions in the core shells are
quite different for these two cases, so an appropri-
ate average was used in calculating the values
given in Fig. 1.2.3.

The gamma-ray source strength in the core shells
represents the absorption of approximately. 3.1% of

_the total number of neutrons. This is substantially

lower than the figure of 7.5% which was obtained
for o spherical mockup of the ART, However, re-
cent calculations!? of the Pratt & Whitney reactor,
with both spherical and cylindrical geometry, have
shown that for a cylindrical mockup the percentage
of neutrons absorbed in the core shells is less
than that for a spherical mockup by about a factor
of 2. Thus the figure of 3.1% obtained from the

two-dimensional neutron calculations does not

 

My, W. Bertini et al,, Basic Gamma-Ra Data for
ART Heat Deposition Calculations, RNL-2113
(SGPf- ‘7' 1956)0

12¢, Wagner, Pratt & Whitney Aircraft, private
communication to R. B, Stevenson.

31

 
 

 

 

-ANP PROJECT PROGRESS REPORT

ShSRES
ORNL-LR-DWG t6288

 

0 !5 0.05

25 "'
¢
¢
P

0.25%

“‘ 0.20
$
0

.'A
/ )

oy

S e e X
U)\
I
>
Z
o

45

i \
0.30
0.35
0.40
0.45
. 0.50
1
: 0.55
060
.65

CONTROL ROD

 

 

 

  

ALL VALUES IN wattsfem®

 

 

REFLECTOR — MODERATOR

 

al
1

Fig. 1.2.2, Gamma-Rey Soutce Strength in ART,

seem to be inconsistent with figures obtained for a
spherical mockup of the ART.

The total gamma-ray source strength in the ART
was found to be 4.6 Mw, of which approximately
4.1 Mw comes from the core, and 0.28 Mw comes
from the Inconel core shells. The remainder comes
from the copture gamma rays in the island and the
reflector-moderator.

Gamma-Ray Heating

The heating resulting from the gamma-ray sources
described above has been obtained by using a two-
dimensional gamma-ray heating calculation. The
calculation, which was carried out on an IBM-704
computer, utilized a program developed by Kniedler
and Wenzel of Pratt & Whitney Aircraft. The

program uses a buildup factor approach to account

32

for the scattered gamma-ray photons. In the case
described here, the buildup factor for water was
used throughout the reactor. -
The sources were computed at a number of points
in the reactor, the volume associated with each
source point being approximately a cube 4 ¢cm on @
side, Each source was divided into seven energy
groups. The heating was determined at @ large
number of points for each energy group for _all
source points within a specified distance from the
heating point. This distance was taken as 40 cm.
Thus, a point at the outside of the reflector on the
equatorial plane ‘‘sees’’ the entire fuel system
source on the equator. | '
The results are shown in Figs. 1.2,5, 1.2.6, and
1.2.7. The heating in the island, core, and re-
flector-moderator are shown in Fig. 1.2,5. The
 

SESRET
ORNL—~LR—DWG 16283

S 0 O, N @
o O O O

o

CORE SHELL

W
o

SOURCE STRENGTH {(w/em®)
n
o

o

 

Q

0 10 20 30 40 50
DISTANCE ABOVE EQUATORIAL PLANE (cm)

~ Fig. 1.2.3. | Gcmma-Rfiy- Source Strength in

Inconel Core Shells as o Function of Distance
Above Equatorial Plane, |

PERIOD ENDING SEPTEMBER 10, 1956

RS
ORNL-LR-DWG 16284

04

SOURCE STRENGTH {w/cm3)
T o o
N o

©

 

o
0 ) 20 30 40

DISTANCE ABOVE EQUATORIAL PLANE (cm)

Fig, 1.2.4, Gaomma-Ray Source Strength in
Region Adjacent to Control Rod as o Function of
Distance Above Equatorial Plane,

SECRET
ORNL- LR~ DWG 16267

 

. * CONTROL. ROD
T T 7T LT T LTl T I T I I I T T 7T T ITIT T 77 oI o727 T 2ol e o222

 

 

 

 

      

- REFLECTOR —MODERATOR

ALL VALUES IN watts /cm3

 

Figs 1.2,5. Gamma-Ray Heating in the ART,

33

 
 

 

 

ANP PROJECT PROGRESS REPORT

values given, in w/cm3, are the sums over the
seven groups used. The heating in the Inconel
core shells as a function of distance from the
equatorial plane is given in Fig. 1.2,6. The heating
in the region adjacent to the control rod is shown
in ‘Fig. 1.2.7. The heating caused by the gamma
rays from the control rod is not included in these
results. The heating from the heat exchanger
region was reported previously!3 and is included
in the outer regions of the reflector.

 

13H. W, IBertini, ANP Quar. Prog. Rep, June 10,
1956, ORNL-2106, p 28.

TENGT
ORNL-LR-DWG 16285

o
o

-]
o

INNER CORE SHELL

5

QUTER CORE SHELL

GAMMA=RAY HEATING {w/em®)
o
o

 

0 10 20 30 40 50
DISTANCE ABOVE EQUATORIAL PLANE {cm)

Fige 1.2.6. Gamma-Ray Heating in Inconel Core

Shells as a Function of Distance Above Equatorial
. Plane, . '

34

The uncertainty in the source strength of the
Inconel core shells does not have much of an effect
on the heating values, since the source of gamma
rays in the core is so much greater than that in the
core shells, A test calculation has shown that a
50% change in the source strength of the core
shells will change the heating values at a point
close to the core shells by less than 5%.

The total gamma-ray energy deposition (excluding
that caused by sources in- the heat exchanger
region) in the reactor was found to be 3.86 Mw,
which amounts to approximotely 84% of the total

“source. This means that about 16% of the gamma-

ray source energy escapes,

PEeREF
ORNL-LR-DWG 16286

 

o

 

]

\

T~

N

 

o

 

N
7

 

GAMMA-RAY HEATING (w/em®)

 

 

 

 

 

 

 

o

10 20 30 40 50
DISTANCE ABOVE EQUATORIAL PLANE (cm)

o

Fig. 1,2,7, Gamma-Ray Heating in Area Adjacent
to Control Rod as a Function of Distance Above
Equatorial Plane,
 

 

 

 

PERIOD ENDING SEPTEMBER 10, 1956

1.3. ART INSTRUMENTS AND CONTROLS

E. R. Monn

REFLECTOR-MODERATOR TEMPERATURE
CONTROL SIMULATION

J. M. Eastman! F. P. Green
E. R. Mann

The ART simulator was used for a study of a

_reflectorfmoderator temperature control. system and

the temperature responses of the cooling system

~ which will circulate sodium through the reflector-

moderator and through sodium-to-NaK heat ex-

-changers. The NaK is cooled in external NaK-to-
. air radiators. With the reactor held at design-point

conditions, abrupt closure of the louvers of one of

the two NaK-to-air radiators caused an increase of

approximately 100°F in the maximum sodium tem-
perature (ot the core outlet) and in the mean
beryllium temperature. The response was essen-
tially first order, with a time constant of approxi-
mately 2‘/3 min. A 50% reduction in power from
design-point conditions caused o decrease of
about 150°F in the maximum sodium temperature
ond in the mean beryllium temperature, with a time
constant of approximately 3 min. For operation at
50% power, an abrupt change of the mean fuel
temperature from 1425 to 1250°F caused a de-
crease of about 160°F in the maximum sodium and
the mean beryllium temperatures, with, again,
about @ 3-min time constant.

The mean reflector-moderator temperature was
computed in the process of simulation and was
used as the main criterion for evaluating the tem-
perature control system. ‘A control system was
evolved which performed satisfactorily and will be
used for the ART. The control method consisted
in (1) holding constant a signal composed of the

‘maximum  sodium temperature biased with the

temperature differential in the NaK across the
radiators and (2) causing the louvers to open when

‘this signal exceeded its set value by more than

2°F and to close when the signal became less
than the set value by more than 2°F.  This ar-
rangement provided a proportionality characteristic
adjusted to cause the maximum sodium tempera-
ture to be held at 1200°F for zero reactor power
and 1250°F for design-point power. These are

 

10n assignment from Bendix Products Division.

C. S. Walker

the values calculated (ond built into the simula-
tor) for holding the mean beryllium temperature
constant at 1200°F. The louver actuating speed
was adjusted to give full stroke in 30 sec.

This control system was operated on the simu-
lator with.step changes between 0, 50, and 100%
design power and between 1200 and 1425°F mean
fuel temperature (at 50% and at zero power). For
these disturbances the mean moderator tempera-
ture was held constant to within £15°F. The
system was not stable in the sense that the
louvers remained fixed for steady-state conditions.
In order to maintain the desired temperatures, the
control system caused the louvers to shift posi-
tions at 30- to 60-sec intervals. When the tem-
perature control system hardware has been as-
sembled, it will be checked on the simulator,

INSTRUMENT DEVELOPMENT
R. G. Affel

Fuel-Expansion-Tank Level Indicator
R. F. Hyland

Tests were continued on a helium-bubbler type
of level indicator, described previously,? for the
ART fuel expansion tank. In these preliminary
tests made to check the operation of the bubbler,
two bubbler tubes are used and the fuel is static.
One test rig was shut down after 3000 hr of opera-
tion at a fuel temperature of 1150°F; neither
bubbler tube had clogged. Another rig has com-
pleted 2686 hr at 1500°F, and another has com-

pleted 2352 hr at 1500°F. In both these rigs one

of the two bubbler tubes has become clogged.
Examination of a bubbler tube from the rig that
was terminated showed that o thin film had de-
posited on the inner surface. A chemical analysis
of a 10-mg sample of the film showed it to contain
7 mg ZrO, and about 1.8 mg O,. - Since the helium
supply of the building contains an average of
1.5. ppm O,, sufficient oxygen was present in
1000 liters of helium (the approximate flow for a
75-hr period) to account for the entire 1.8 mg of
O,, if complete reaction is assumed. Analyses

 

2 .
R. F. Hyland, ANP Quar. Prog: Rep. June 10, 1956,
ORNL-2106, p 43.

35

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

of the contents of two bubbler tubes that plugged
in a previous test at 1500°F, during which the
system was accidentally contaminated with oxy-
gen, also showed high ZrO2 contents. It is ap-
parent therefore that the cause of bubbler-tube
clogging is oxygen contamination and the subse-
quent formation of Z0,, a high-melting-point
(3000°C) solid, rather than ZrF ,-vapor deposits
or fuel, as anticipated. It is planned to use either
a NaK ‘scrubber or the incandescent-titanium-
sponge method to purify the helium used in further
tests.

Systems for Testing Liquid-Level-Sensing
Devices in Flowing Liquids

“H. J. Metz

Three test systems are being constructed for
studying the operational characteristics of various
liquid-level-sensing devices under controlled con-
ditions in flowing liquids. The test system con-
sists of pressure vessels at both ends of a motor-
driven teeter-totter. Liquid will flow from the
rising vessel to the vessel going down and the
liquid levels in the vessels will remain constant.
A gear-motor drive is provided which can rock the
vessels at a rate of 0 to 10 cpm. A helium atmos-
phere at a pressure of about 1 to 3 psig is pro-
vided in a closed circuit over the vessels to pre-
vent contamination of the fluid., The desired
temperatures are obtained and maintained with
heaters and variable autotransformers,

Each vessel has six spark-plug probes con-
nected to level-indicating lamps, and the liquid
levels in the vessels can be adjusted from 1 to
10 in. Two level-sensing devices can be tested
in each vessel at various levels. The tests will
be of long duration to assure that the level-sensing
devices will work for 2000 to 3000 hr with the

surface of the fluid in constant motion.

High-Temperature Turbine-Type Flowmeter
G. H. Burger

The first turbine-type flowmeter was designed
and built in October 1955 for measuring flow in
l-in.-ID tubing. Since then numerous tests have
been made in order to determine the operating

characteristics and the life of the units. The
initial design has been modified as test data have
indicated the need for changes and three succes-
sive models have been built.?

The test results obtained thus far have been
disappointing, particularly the life test results.
As stated above, four flowmeters have been tested,
but none has even approached the required 3000 hr
of operation. The brevity of the test, however, in
all cases except the first, has been due to system
failures, such as pipe leaks and pump difficulties.

Performance tests of the ‘vnits have been quite

satisfactory. The units have ‘been tested in both
NaK and fused-salt fuel mixtures with excellent
results. The accuracy of the measurements in the
NaK system have been well within 1%, as checked
against a magnetic flowmeter and venturi. . The
accuracy of the units in the fused-salt mixtures
could not be precisely determined because of the
absence of a venturi in the system. The salt sys-
tem also had a very low maximum flow rate which
was out of the design range of the flowmeter. The
bearing material and the bearing design of the unit
tested in the fused salt appeared to be satis-
factory, since no evidence of bearing-sticking or
self-welding was evident from the operation of the
units or from metallurgical examination of the
units after removal from the system. The turbine
blades and the turbine bedy appeared to be un-
marked by the fuel or the NaK, Even though the
tests in fuel and in NaK have been.relatively
short, in the neighborhood of 300 hr, the units
have operated at temperatures up to 1500°F, and
the indications are that the units can be expected
to operate satisfactorily for 3000 hr.
- A fifth model of the flowmeter has been fabri-
cated which incorporates improved bearings. - The
water test and calibration of the unit have been
completed, and it appears to be quite satisfactory.
A similar turbine-type flowmeter with a 3}§-in.-dic
housing has been designed for operation in a sys-
tem with flow rates up to 1400 gpm and is now
being fabricated. This unit will be evaluated in
a loop for testing NaK pumps. It will be cali-
brated against a venturi installed in the system.

 

3
G. H. Burger, ANP Quar. Prog. Rep. ]une 10, 1956,
ORNL-2106, p 43.
 

 

 

High-Temperature Pressure Transmitter
W. R. Miller

Pressure transmitters of several types obtained
from five different vendors have been tested at
temperatures  between room temperature and
1400°F. Four modefs were of the pneumatic force-
balance type which employs a bellows or dia-
phragm as the sensing element. The all-welded
“bellows appears to be the most sotusfactory Six
of the transmitters tested produced an average
accuracy of +0.37% full range if heid at a constant
temperature, but zero shifts as large as $3% full
scale were observed between room temperature and
1400°F. S -
 _Five .units were tested whlch utnllzed a dio-
,'Vphrogm-lsolated NaK fllled tube system in which
‘NeK hydrostottcolly ‘transmits the dlaphragm
-sensor motion to a 3 to 15psi transmltter-mdlco- -
tor.. These units are available in ronges from 0 to -
50 ‘and 0 o 200 psi w:th tube ‘system CQpl”Gry
|engths ‘of up to 50 ft.. The units tested were

random selectlons from a lot of 25 and were found

. to have an average accuracy of iO 43% full scale,
with- ‘zero shifts between room temperature and
1400°F which averaged 11.7% full scale.

" Tests have been performed on three units which
employ the tube-system type of sensor but termi-

nate in a four-legged unbonded stram-gage bridge. i _
_modification of its input-damping network.

Results to date show .an average -accuracy of

10.25 full scale at a constant temperature, w:th»-"':__
repeatedly (after intervals ranging from one to

zero shifts of +1.7% between room temperature and
1400°F. Although these units are the most ac-
cwate and promising to date,

span- shifts of £2% full scale when the tempero-'_
ture of the strain-gage housing is elevated o’
150°F.  One more unit of this type remains to be

tested as well as two units whuch uttllze a two-- 3

!egged strom-gage brudge. ) | o
| S5 J-\fl“‘\"‘ifi*&h“wfliflt ‘!Fr-- ¥
i =,“ g‘;;.,,m w\-’h s_‘“&'h (&2

Tests of Heliurc-Welded lnconel Sheothed
Thermocouples in Sodwm '

. T DeLorenzo o

 

Tests have been made in‘an uttempt to determlne S

the effect of @ hellorc-welded !nconel sheath on

the calibration of a thermocouple. The data indi- f{ 4J T. D,Lo,emo and W. R. Miller, ANP Quar. Pro
f Rep. June 10, 1956, ORNL-2106, p 42, Fig. 1:3.1.

F
N [).

cate that the weldment produces no noticeable |

o , .

tests show that |
temperature variations at the - strain gage - produce__. )

i,

Mg

 

 

B g AL ST

2l

 

n,fufi‘ ]

~ purged switch showed indications of plugging.
- The marked improvement produced by the nitrogen
.~ purge is illustrated in Fig. 1.3.1.
: ‘*:teSting"of‘the_syst_em ‘both switches will be purged. }

PERIOD ENDING SEPTEMBER 10, 1956

 

deviation of the cahbratlon from the normal
changes induced by aging effects over the interval
of the tests. Equipment is being fabricated and
assembled for aging tests of such welds in static
sodium at temperatures above 1000°F for 3000 hr.
A preliminary investigation of possible in-pile
calibration tests is under way. For these tests
sodium vapor pressure would be used as a tem-
perature standard. The instrumentation require-
ments and the system design are being studied.

Thermocouple Data Reduction
J. T. DeLorenzo |

Life tests were begun on two 5ynchron|zed h:gh-
speed mercury-jet switches, described prev;ously,

- which operate at 1800 rpm and scan 80 thermo-
' couple signals per revolution.

The input signal
to the transmitter switch consists of a voltage-

'dlwder network with 79 one-millivolt increments
i and one 0.5-v tap, which triggers the sweep of a

17-in. oscilloscope. The ‘oscilloscope monitors
the common line between the switches. The input
signal produces o stair-stepped trace that ranges
on the oscilloscope from 1 to 79 mv in 79 steps.
It was found that any ten consecutive output points

~on the receiver switch could be recorded reliably

on a standard . 12-point recorder after a slight
Initially, the life tests had to be terminated

three weeks) because of plugging in the jet
nozzles. After each case of plugging, the switches
were disassembled and examined.
-heavy black scum was found on the surface of the

An each case, a

" mercwy ond adhering to the various surfaces of

“the mercury pool. A chemical ahclysistevealed
that the scum was primarily mercurous oxide.
~ in an ottempt to eliminate the formation of this

- oxide, one switch was continuously purged with

“dry nitrogen. After approximately 20 days the un-

In further life

 

ey, .
-

&

37

nt_g-\vgew e Rt ik ek 4.,

wm&wwf*

 

 

 

 

 

 
 

 

 

ANP PROJECT PROGRESS REPORT

"UNCLASSIFIED

 

 

Fig. 1.3.1. Nozzles of Two Mercury-Jet Switches Showing Clean Condition of Switch Thot Was
Continvously Purged with Dry Nitrogen in Comparison with Mercurous Oxide Scum Formation on Switch

That Was Not Purged.

38

o

(
 

 

 

 

PERIOD ENDING SEPTEMBER 10, 1956

1.4. COMPONENT DEVELOPMENT AND TESTING
| H. W. Savage -

PUMP DEVELOPMENT TESTS
E. R. Dytko! A. G. Grindell
. Bearing and Seal Tests
W. L. Snapp! W. K. Stair?
A petroleum !ubncont for the ART reoctor pumps®
has been specified on the basis of present infor-
mation, but the search for better lubricating and

- cooling fluids for pumps has’ continued. .
program for determmmg the compatibility of vari-

. ous lubricating fluids with the reactor process
o fluids. (sodlum ond fuel) has been ‘undertaken. The"
first tests. were conducted with Dowtherm A ond

Cellulube 150 heated to 200°F under -a helium

- atmosphere, The heated Iubrlcont was odmtfled to
- g container of sodium at a temperature of 1100°F,

which was alsé hefium blanketed.  The reaction of

the Cellulube 150 was vigorous, enoUgh to be near
: exploswe proportions, while there was’ very little
reaction between the molten sodium and Dowtherm -

A.  These: results ‘were- ‘sufficient to preclude
-~ further Cellulube teshng.

‘A loaded journal-bearing test was conducted with .

Dowtherm A as the lubricant and with an applied
bearing load of 250 Ib. This test was terminated
after 659 hr of trouble-free opercmon " The total
measured seal Iéokdge was 1 cm® at the upper
seal and 12 cm? at the lower seol The process
sides of both seals were found to have the same
type of gummy carbon deposit as that found previ-

ously in tests in which Dowtherm A was used as
the lubricant.® There was ‘some visible evndencefl
- of bearmg ‘rubbing, which was opporenfly mmor*
since no deleterious effects could be found in
either. the beonng operoflon or.in- the drlve-mofor* :

power trace. ‘ :
“An mveshgohon of the possnble effects of in-

creasing the present reoctor-pump (fuel ond sodium) -

rotary-element _operating speed range "is to be
- made.
_operating speeds ‘and increased bearing loads.

* Similar fesfs are plonned for the NaK pump rotory

'element

 

On osugnment from Prufi & Whnney Alrcroff E _
Consultonf from the Umversny of Tennessee. -

3W L. Snapp and W. K, Stair, ANP Quar Prog Rep
June 10, 1956, ORNL-2106, p46

A test

~ material.

Tests will be storted soon with’ hlgher -

Excessive axial motion was found in the original
NaK pump thrust-bearing design. A change to a
duplex pair of angular-contact ball bearings
mounted face-to-face has provided the desired
axial rigidity without sacrificing transverse flexi-
bility.

A seal design proposed for the NaK pump by a
vendor is now being tested. In general, it has
been found that the vendor’s seal operates with
modest leakage rates during the early part of a

~.test but that subsequent operation always results
_in much higher leakage rates.’
‘seal, similar to that to be used in the reactor

A bellows type of

has been tested for NaK pump applica-
Further tests

pumps, 4
tion, with inconclusive results.

‘should result in a modification of this type of sea!

which will give acceptable performance. Modified
D_urometollic seals® have been received and are to
be tested soon.

Testing of elastomers for seal application has
been impeded by delays encountered in obtaining
test samples of recommended materials. A test
was conducted, however, in which Cellulube 150
and Buna-N base elastomers were used. Since

Cellulube 150 requires the use of butyl rubber

elastomers, the results, as expected, were poor
with respect to strength and hardness of the seal
The Buna-N O-rings did, however, main-
tain a satisfactory seal throughout the 800 hr of
test operation at 220°F. It is evident that more

~ compatibility data are necessary for a realistic

evaluation of elostomers, since the unsatisfactory
performance observed thus far in tests at 220°F is

" in contrast to' satisfactory results reporfed for
" tests of elastomers at 150°F in other fluids.5

Tesfs of Gos Aflenuotlon by Sea!s
S M. DeComp, Jr. V. K. Staie

Tests have been concluded that were designed

“to determine whether there would be fission-gas
" leakage - from the sumps of the ART fuel pumps
~into the 7;'Iubricat_ing oil catch basins and thence

 

) ‘W L. Snopp end W. K, Stolr, ANP Quar Prog Rep.

 March 10, 1956, ORNL-2061, p 43.

SUCON Fluids and Lubricants. Carbide and Carbon
Chemicals Co., F-6500D, 1955.

39

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

into the lubricant reservoirs external to the reac- -
tor cell. Attenuation of the fission-gas concentra-..

tion by a factor of 10'® from the fuel expansion
tank to the lubricant reservoir was considered to
be necessary to eliminate biological shielding of
the lubricant systems. Preliminary test data in-
dicate attenuation by a factor of at least 104 from
the expansion tank to the pump shaft seal and by
another factor of at least 104 between the seal and
the reservoir. These data are being analyzed to
determine the | upper limit to be expected for the
radiation dose from the lubrlcahng-orl tank.

Fuel Pump Development Tests

S AW Simon !
M. E. Lackey "~ G. Samuels

Developmental test work on the fuel pump im-
peller’ continued, with water belng used as the
pumped fluid. Efforts are under way to reduce the
maognitude of the system pressure fluctuations, to
produce the desired gas pressure gradient down
the shaft annulus in the direction of the purge gas
fiow, to determine the fuel flow through the centri-
fuge, to reduce the pressure level of the system to
meet the stress limitations of the north-head deck
plate, and to obtain cavitation data. Some of
these problems are, as yet, only partly solved.

Part of the work on the reduction of the magni-
tude of the pressure fluctuations has been carried
on in conjunction with the efforts to obtain the
desired gas pressure gradient down the pump shaft
into the expansion tank and out into the off-gas
line. These two problems are interrelated -be-
cause both are affected by the upper seal face of
the slinger impeller, area 1 in Fig. 1.4.1. The
increase in the pressure fluctuations that was
experienced  duwing operation with high liquid
levels in the expansion tank can be attributed to
the priming and unpriming of the slinger impeller
or perhaps to a hunting of the liquid level on the
face of this impeller. As changes were made to
the slinger impeller in attempts to remove the
pressure fluctuations, the changes .in the gas pres-
‘sure gradient were noted. With the original slinger
impeller design, complete with 12 blades on the
upper. surface, both the fluctuations and the re-
verse pressure gradient were at their maximums.
The reverse pressure gradient indicated that the
upper seal face of the slinger impeller could act
as an effective gas pump and cause gas pressure
in the expansion tank to build up to greater than

40

~ the gas sipply pressure, with the gas pressure in

the oil-leakage catch basin in the rotary element
being the lowest pressure in the system. This
condition would be especially undesirable if the
pump stopped abruptly or even if there were a
sudden reduction in the pump speed, beccuse the
gas pressures would be such that fuel might be
forced into the seal region of the rotary element.

The upper seal face of the slinger impelier has
been modified so that it is a smooth disk, and the
axial clearance, as well as the radlal clearance,
of the impeller is kept as small as possrble This
configuration keeps the amount of liquid in and
around the impeller face to a minimum, regardless
of the liquid level in the expansion tank. Opero-
tion with this modlfied impeller produced a normal
gas pressure drop in the direction of flow, with the
oil-catch-basin gas pressure being ot an lnter-
mediate level and acbove the expansion tank gas
pressure. Further modlflcahons of this region are
being tested in an attempt to provide a pressure
equalizing volume that may be necessary for
alieviating any possible unequal liquid level dis-
tribution around the pump shaft. These tests will
provide a basic shape for further development
tests in the twin-pump aluminum-north-head water
tests, where the effects of unequal pump speeds
can be noted. ‘

A ‘second trouble area with respect to pressure
fluctuations was found to exist at the lower face
of the slinger, that is, the degassing. face of the
slinger impeller, which is area 2 of Fig. 1.4.1.
The first mocllficahon for correchng thls situation
consisted in simply rémoving the vanes, or blades,
from the impeller face. The resulting large clear-

ance rendered the degassing feature of the centri-

fuge ineffective. The shape that appears to
function acceptably consists of a smooth dlsk that
maintains the previous clearances. ‘

Mcdifications were also made to the seal plate
over the centrifuge cup, area 3 of Fig. 1.4.1, in an
effort to reduce the centrifuge leakage flow. A
lip was added that extended vertically downward
from the inner surface of the seal plate to the top
of the centrifuge cup. This resulted in operation
that was, in general, poorer than before. The
pressure fluctuations were greater and the degas-
sing time was longer ‘because the greatly in-
creased velocity of the liquid discharging into the
already restricted flow area only further restricted
the centrifuge cup inlet flow. When this. lip was

dp
 

 

PERIOD ENDING SEPTEMBER 10, 1956

SECRTT -
ORNL-LR-DWG 16015

_—
_—

.
-

EXPANSION VOLUME

 

 

 

 

 

 

 

 

 

 

 

~

% — < =S

EXPANSION
VOLUME

| ®/ , ' L— SEAL PLATE
fi[ — . : r_- -‘]/———
3

i 3

Y

 

 

 

 

— SLINGER IMPELLER

 

—MIXING CHAMBER

 

N R SRR

 

A L1 h

 

 

 

 

 

CENTRIFUGE SEAL VANE
CENTRIFUGE BLADE — ‘
CENTRIFUGE

 

 

BAFFLE —1 | ' : ~~ CENTRIFUGE HOLE

 

 

 

 

 

 

 

 

 

 

 

.PUMP | ' ‘ ' ' -
" piscHARGe | FUEL PUMP 1 , ; /

 

 

 

 

N

= v - ' : / . PUMP SUCTION - ' \

I

 

 

 

Fig. 1.4.1. 'Dfugrmfi ;f ..'Iu_npeller Region of ARf Fuel Pdmb. o |

41

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

removed and a similar one placed on the outer
surface of the seal plate, the operation of the pump
was improved. In locating this lip on the outer
surface, the axial clearance from the lower edge
of the lip to the top of the centrifuge cup was
made identical to that existing between the top of
the centrifuge seal vanes and the lower side of the
seal plate. While this modification increased the
resistance of the flow passage from the centrifuge
discharge holes, area 4 of Fig. 1.4.1, to the
centrifuge inlet, it did not reduce the available
head of the seal vanes on the top of the centrifuge
cup and thus did not affect the discharge pressure
of the centrifuge in any way. This reduced flow
has less effect upon the character of the flow at
the centrifuge inlet, and pressure fluctuations are
therefore less.

Another series of tests was performed to deter-
mine the amount of flow that passes through the
shaft feed holes, area 5 of Fig. 1.4.1, the ex-
pansion tank, and the centrifuge. An indirect
method was used to measure the flow, since it is
virtually impossible to measure it directly. The
internal xenon-removal feed flow up the shaft was
blocked, and data were token while known xenon-
removal feed flows were added to the expansion
tank through on external circuit. The pressure
difference between the expansion tank and the
pump suction was plotted vs the external circuit
flow for a constant main loop flow and a constant
pump speed. This pressure difference was equated
to the difference determined with the unblocked
shaft. The shaft flow was computed from this
pressure difference and was found to be approxi-
mately 11.5 gpm.

The reduction of the fuel system pressure level
will be attempted by two methods. The first will
involve reducing the size of the centrifuge dis-
charge holes in an attempt to reduce the head de-
veloped by the centrifuge. The second will in-
volve increasing the internal diameter of the
centrifuge cup to reduce the centrifuge head.
Results of incomplete tests performed on each of
these modifications indicated that each will allow
a reduction of the pressure level to obtain an
acceptable stress in the north-head upper and
lower decks. The final selection of one of these
modifications will be determined by future tests
concerning the stability of the pressure level as
a function of the liquid level in the expansion tank.

42

The results from some initial test runs indicated
that a potential problem of cavitation exists with
the present impeller blade desngn Values of
Thoma's cavitation parameter,® o, are about 0.71
for 2400 rpm and 0.51 for 2700 rpm. An impeller
with new blades having an improved entrance
angle was tested for cavitation, and o was found
to be about 0.54 for 2400 rpm and 0.41 for 2700
rpm. Investigations of the cavitation characteris-
tics of the present impeller design will be con-
tinved.

Fuel Pump Endurance Tests
S. M. DeCamp, Jr.

An endurance test of an ART fuel (MF-2) pump'
was started’ on April 10, 1956, and was terminated
on July 2, 1956, after 1964 hr of operation. Op-
eration during this period was satisfactory.
Troubles experienced earlier® with gas flow down
the pump shaft and with removal of oil from the
lower catch basin drain did not recur. Removal
of the pump from the pump barrel was relatively
easy, and disassembly of the pump was accom-
plished by heating the pump impeller and as-
sociated equipment and removing the parts while
they were still hot.

The test was terminated when the operating
sounds of the pump changed and the source of the
sounds could not be definitely located. Examina-
tion of the components after disassemb|y gave no
indication of trouble in the pump. It was reas-
sembled and installed in the cold mechanical
shakedown stand, where it was run for 100 hr. At
the end of this period the pump was operating

“satisfactorily, with no leakage detectable at the

lower seal.

The pump and a new hydraulic motor were then
reinstalled in the high-temperature endurance stand
for further testing. At the end of 400 hr, the pump
was still operating satisfactorily, with no sign of
lower seal leakage. Operation of this pump'is
continuing.

 

A, H. Church, Centrifugal Pumps and Blowers, p 82,
Wiley, New York 1944,

s, M. DeCamp, ANP Quar. Prog. Rep. ]une 10, 1956
ORNL-2106, p 48.

8. M. DeCamp, ANP Quar. Prog. Rep. March 10,
1956, ORNL-2061, p 48. i

-

C
 

 

 

Sodium Pump Development Tests
S. M. DeCamp, Jr.

Water testing of an ART sodium pump in a per-
formance-acceptance loop, initiated previously,?
was continved. The tests conducted previously
were aimed at solving the problems of bypass flow
around the main impeller and ingassing of the main
fluid circuit. The bypass flow was from pump
discharge through a pressure breakdown labyrinth
to the expansion tank, and it was returned to pump
suction. Bypass flows as high as 10 gpm were
possible without ingassing with this bypass flow
circuit. Examination of the reactor north-head
geometry and thermal stress conditions, however,
indicated that bypass flow of this type was un-
desirable because of the low sodium temperature
(1050°F) in the expansion tank. A preferred solu-
tion is to bring hot sodium (1250°F) from directly

below the lower deck of the sodium expansion

tank to cool the top lid of the sodium tank and then
to return this flow to pump discharge by means of
a centrifuge. ,

Various conflgurcflons of the flow passages

above the centrifuge have been tested.  The most

recently tested configuration is shown in Fig.
1.4.2, and performance curves obtained for the
configuration are presented in Fig. 1.4.3.

It became apparent in earlier tests that the size
of the slot in the side of the pump barrel was in-
adequate to handle the return flow of liquid to the
centrifuge. In order to handle the required flow, a
second slot was cut into the barrel below the
original slot. The addition of this slot increased
the maximum obtainable flow into the centrifuge

tion of the loop

anury and Auxdlary NuK Pump
Developmeni Testsm

H C Young J.G. Teague

7 The flrst Inconel stahonary assembly - for an
ART primary  NaK- pump (PK-P), consisting of
volute and pump tank, was received. This as- . -
serrbly was welded into a hlgh-temperature test-fif"

 

‘95, M. DeCamp, ANP Quar ng Rep ]une 10, 1956 -'

ORNL-2106, p 50.

mThe pr‘imary NaK pumps are to be used to clrculo'fe. :

NaK through the ART fuel-to-NaK heat exchanger sys-
tem; the auxiliary pumps are to be used to circu-
late NaK through the ART sodium-to-NaK heat exchanger
system.

PERIOD ENDING SEPTEMBER 10, 1956

loop made up of approximately 60 ft of 3}’2- and
4-in. IPS Inconel pipe. A heat exchanger and the
necessary instrumentation were added to the loop,
and water tests were conducted.

Initial tests indicated that the pump would not
prime because of excessive quantities of gas
having been trapped in the discharge pipe during
the loop filling operation. A ¥ .-in.-dia hole was
therefore drilled through the top of the discharge
pipe inside the pump tank to permit the trapped
gas to be vented during filling. After the pump
was primed, a continuous leckage of 2 to 3 gpm
flowed through this vent to the pump tank. A
baffle and a deflector were developed and in-
stalled to control the splashing caused by this
high-velocity leakage. The pump primed satis-
factorily after the vent was installed, and such a
vent will be mcorporuted into all future PK-P
pumps. .

- The water test performance data obtained on
these Inconel pump parts agree very well with

“initial data obtained with the original brass im-

peller and aluminum volute;!! however, a slightly
higher efficiency was obtained in these tests. The
cavitation performance of the Inconel impeller was
not so good as that of the brass impeller; however,
the cavitation characteristics are still considered
to be satisfactory.

Weld shrinkage tests of pump volute halves have
shown that the shrinkage can be controlled to

- within oacceptable tolerances (see Chap. 3.4,

*Welding and Brazing Investigations’). Several
pairs of Inconel volute halves, with the volute
passage profile mochmed hcve been received from

the- vendor
ond, at the same time, caused more stable opera- -

HEAT EXCHANGER AND RADIATOR
DEVELOPMENT TESTS

E. R. Dytko

R. E.'rMacP'l-ners'on' J. C. Amos

| -lnt_ermedidik'erHecit,Eié_hanger Tests _
J.w Cooke ! H. C. Hopkins !

A summary of intermediate heat exchanger test

- stand operations ‘during this quarter is presented -

in Table 1.4.1.  York radiator No. 9 completed a
" - program of 30 thermal cycles and was removed from
" test.stand A for metallographic examination. The

 

My, c. Young, J. G. Teague, and M. E. Lackey, ANP
Quar. Prog. Rep. June 10, 1956, ORNL-2106, p 51.

43

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

SEORER
ORNL-LR-DWG 18132

      
  

 

SURGE TANK
LIQUID LEVEL

 

 

  

=
N

©——

 

  

(B —>= ——= HELIUM PURGE GAS

"""" — -~ GAS SEPARATED FROM LIQUID
(©)—-—= —-—= LIQUID FLOW THROUGH CENTRIFUGE

(©)—> — MAINFLOW
@——-— et C+ B

Fig. 1.4.2. Diagram of Impeller and Centrifuge Regions of ART Sodium Pump.

44
 

 

PERIOD ENDING SEPTEMBER 10, 1956

SECRTY
ORNL-LR-DWG 16133

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

140
120 |— : 3400 rpm wh—g
: o ) —
3200 rpm =g 0~
100 .
. S —. |
o _ - 3000 rpm deg | -5
g 80 .
w 2700 rpm =—game_|
; .-‘."“o-
2 60 7 2400 rpm e
- o= .-..__
40 : 2100 7PM —gr— ~e=6 z
| S
- - u-
1800 rpm -o 8 —egg 3
150C rpm g4 0
20 T eete- &
0 : :
0 50 100 {50 200 250 300 350 400 450 500 550 600
: : FL.OW (gpm)
Fig. 1.4.3. Water Test Performance Characteristics of ART Sodium Pump.
TABLE 1.4,1. SUMMARY OF INTERMEDIATE HEAT EXCHANGER TEST STAND OPERATION
Hours of
Test Unit* Nonisothermal Total Hours of Number of Reason for
Operatién Operation Thermal Cycles Termination
Test Stand A
York radiator No. 9 {revised ' 695 1283 30 Test completed
"~ design) ' ' : '
Circulating cold trap Ne, 2 B , 1430 ! ' Test completed

(4 in. in diameter)

NaK screen filter o . ' 264 ' Test completed

Test Sfund B
Biock,_Sivalls & Bryson heat 1008 - 1398 ' 11 : Test completed

exchangers Nos. 1 and 2
{(type 1HE-3) _

Cumbriage radiators Nos. 1 and 2 1466 , 2045 . 14 Test completed
{modification 3) ' ' o

Circulofing cold trap No. 5 o 1260 o Test completed

(4 in. in diometer)

 

 

*Includes only units tested during this quarter,

45

 
 

 

 

ANP PROJECT PROGRESS REPORT

radiator performance data and details of the
thermal-cycling program were reported previ-
ously. 2 This radiator was the first 500-kw radia-
tor of the revised design!3 to be tested. Post-
operational examination showed no evidence of
cracks in the radiator tubes. The available infor-
mation on maximum mass transfer and total in-
crease in NaK pressure drop for all radiators
tested, including York No. 9, is summarized in
Table 1.4.2, At the completion of York radiator
No. 9 testing, test stand A was dismantled to
provide floor space for the Engineering Test Unit
(ETU). Analysis of the contents of the NaK
screen filter tested in this stand revealed no
significant amount of metal particles or foreign
material. |

Testing of Cambridge radiators Nos. 1 and 2 and
Black, Sivalls & Bryson heat exchangers Nos. 1
and 2, type IHE-3, described previously, '2 was

completed in test stand B. These test units are

currently undergoing metallographic examination.
The data on NaK pressure-drop increases and
maximum mass-transfer deposits in the Cambridge
radiators are given in Table 1.4.2. These radia-
tors were of the original 500-kw design modified
by removing the side plates and slitting the sup-

 

12 w. Cooke and H. C. Hopkins, ANP Quar. Prog.
Rep. June 10, 1956, ORNL-2106, p 56.

13, W. Cooke, H. C. Hopkins, and L. R. Enstice,
ANP Quar. Prog. Rep. March 10, 1956, ORNL-2061, p 52.

- port plates and base plates. Me'tdllbg'rfq'phié__”ex»

amination revealed cracks in the radiator tubes at
the rigid base plate and support plates. With the
exception of the degree of increase in NaK pres-
suwre drop, the performance data obtained on these
radiators and heat exchangers were in substantial
agreement with data obtained from units previously
tested. The NaK pressure drop in Black, Sivalls
& Bryson heat exchanger No. 2, where heat was
transferred from fuel to NaK, did not increase

‘throughout the test, whereas, in heat exchanger

No. 1, where the heat was transferred from NaK to
fuel, there was a total NaK pressure-drop increase
of approximately 240%. This pressure-drop in-
crease is attributed to mass-transfer deposits and
has been observed as a function of time in all test
units in which NaK has been cooled.

~ York radiators Nos. 11 and 12 (revised design)
have been installed in test stand B, and tests will
start as soon as Black, Sivalls & Bryson heat
exchangers Nos. 1 and 2, type IHE-8, previously
described, 14 have been received and installed in
the stand. Construction of test stand C is essen-
tially complete, with the exception of the installa-
tion of the ART test radiator being supplied by
York Corp. The delivery of this radiator has been
delayed by fabricational difficulties.

-

e

 

4R, D. Peak e al., ANP Quar. Prog. Rep. Dec. 10,
1955, ORNL-2012, p 42, '

TABLE 1.4.2. SUMMARY OF DATA OBTAINED ON NoK PRESSURE DROP INCREASES AND MASS
TRANSFER BUILDUP IN TEST RADIATORS

 

 

 

 

Maximum Pressure
Test Total Hours of Maximum NaK Temperature Drop Maximum Thickness
Test Unit Stand Nonisothermal Temperature Differential Increase - ©f Mass Transfer
Operation (°F) ©F) (%) Deposit {in.)*
PWA-1 and 2 IHE-B 585 1500 500 50 0.004
ORNL-3 SHE-B 295 1500 230 30 €.0035
York-4 SHE-B 748 1500 265 30 0.003
York-9 IHE-A 695 1500 400 118 0.006
Cambridge 1 and 2 |HE-B 1466 1450 210 140 0.005

 

*Maximum mass ftransfer occurred in air inlet side (cold side) of the radiator in each instance. The Jensify of the

mass transfer and the length of radiator tube over which mass transfer was found were greater in the units which ex-

hibited the greater pressure drop increases.

46
 

Small Heat Exchanger Tests
L. H. Devlin! ~ J. G. Turner!

A summary of small heat exchanger test stand
operation during this. quarter is presented in
Table 1.4.3. Testing of York radiator No. 7 and
Process Engineering Corp. heat exchanger No. 1,
type SHE-2, is continuing in test stand B. The

radiator NaK pressure drop increased approxi-

mately 100% during the first 970 hr of nonisother-
mal operation and remained essentially unchanged
thereafter. This latter period of operation has

been at a maximum NaK temperature of 1500°F, -

with a radiator NaK temperature drop of 100°F for
216 hr and 430°F for 144 hr.

The radiator was initially placed on power op-
eration in a series of steps designed to simulate
temperatwe conditions which will be experienced
by the ART radiators as the reactor is taken to

PERIOD ENDING SEPTEMBER 10, 1956

full power. Radiator NaK temperatures, operating
times, plugging-indicator maximum break tempera-
ture, and increase in radiator NaK pressure drop
for these steps are given in Table 1.4.4, During
the test, cold-trap flow was maintained at 0.25 gpm
(equivalent to rate of ART cold-trap system flow)
and an outlet temperature of 300°F.

‘Testing of the ‘Process Engineering Corp. small
heat exchanger No. 1, type SHE-7, in test stand C

" was interrupted by the failure of York radiator

No. 5, shown in Fig. 1.4.4. This was the last

‘medified radiator of the original 500-kw design to

be tested. The shift in elevation of the l/!é-in.
support plates graphically illustrates the necessity
of eliminating the rigid support plates and base
plates from the radiator fin matrix in order to
minimize thermal stresses. This radiator was
replaced by York radiator No. 8 (revised design),
and test operations were resumed.

TABLE 1.4.3. SUMMARY OF SMALL HEAT EXCHANGER TEST STAND OPERATION

 

 

- (4 in_.ri'n diameter, modifif:ation-l) 7 :

Hours of Number of
Test Unit* Nonisothermal Total Ho.urs** Thgrmol Status of Test
‘ Operation of Operation Cycles
Test Stand B
Process Engineering Corp. heat 1330 2165 13!’2 Test continving
exchanger No. 1 (type SHE-2) * :
York radiator No. 7 (revised - 1330 2165 - 133’2 Test continuing
design)
Circulating cold trap No. 6 (4 in. 2165 Test continuing
in diameter, modification 1)
) Test Stand C
Process Engineering Corp. heat ' 356 598 12 Test continving
'exchung'érrrlflor. 1 (type SHE-7) -
' Yor:k f_adi'afor No. 5 . 223 1265 22}'2 Terminated because of
(modification 2) - ' _ radiator failure
York ri_n.iki‘c'né_rr No. 8 {revised - 239 300 7}'2 Test continuing
. design) _ . '
Circfildfing céid_ trap No. 4 298 : Replaced by cold frc.:lp
(4 in. in diameter) - with stainless steel
- _ cooling coil
Circulating cold trap No. 7

300 , Test continuing

 

*Includes only units tested during this quarter.

**For tests in progress the operating hours are shown as

of August 15, 1956.

47

 
 

 

ANP PROJECT PROGRESS REPORT

The type SHE-7 heat exchanger, shown in Fig. 1.4.6. Endurance testing of thls heat exchanger is

1.4.5, was designed to operate at ART intermedi- currently under way.

ate heat exchanger design temperature and flow A summary of the operatmg condmons and the
conditions with a heat load of 400 kw. The actual  corrosion of the fuel side of the heat'exchangers
operating conditions obtained are shown in Fig. tested thus far is presented in Table 1.4.5.

TABLE 1.4.4. STEP INCREASES IN OPERATING CONDITIONS FOR YORK RADIATOR NO. 7

Plugging indicator maximum break temperature: 300°F

 

 

Radiator Inlef Radiator Outlet Hours at Increase in :
Temperature Temperature Temperature Pressure Drop . Increase
cF F) Indicated (%) (%/he)
1195 1125 20 0 0
1230 1160 25 0 0
1250 1070 17 0 - ' 0
1290 1070 72 14 0,192
1355 . 1070 30 3 0.100
1430 1070 49 14 0.285

1500 1070 98 35 0.365

 

TABLE 1.4.5. SUMMARY OF OPERATING CONDITIONS AND CORROSION FOUND ON
FUEL SIDE OF TEST HEAT EXCHANGERS

 

 

 

Maximum
Test Total Hours of Operation at Yarious Maximum Depth of
Unit Stand Hours of Fuel Temperatures Corrosion Remarks
Operation  400_1500°F 1500-1400°F 1400-1200°F {mils)
ORNL-1, SHE-A 1557 928 606 23 7
type SHE-1
ORNL-1, IHE-B 1129 965 164 12 _ Area where failure
type IHE-3* occurred excluded
in depth of cor-
rosion determina-
tion -
ORNL.-2, IHE-B 1129 965 164 4 Examination of end
type IHE-3 _ sections is not
yet completed
ORNL-1, - SHE-B 2071 214 412 1445 6
type SHE-2

 

*This heat exchanger transferred heat from NaK to fuel; therefore the NaK temperatures and the tube wall tempera-
tures were higher than the fuel temperatures, This will not be the case in the ART intermediate heat exchanger.

48
 

PERIOD ENDING SEPTEMBER 10, 1956

 

49

&8
wl
3
<

45
gz
5o

5 After Failure in Test Operation,

York Radiator No,

1.4.4,

ige

F

 

 
 

 

 
0S

 

   

 

 

TCONEGELLLM.
ORNL-LR-DWG 15289

 

MATERIAL: INCONEL ’ i
: 25 TUBES {0.1875-in. OD, 0.025-in. WALLS)

NaK OUTLET

 

NaK INLET

SECTION A-A

    
    
 
  
 
   
   

NoX HEADER

{0 in. {REF}

HEAT EXCHANGER TUBES

FUEL HEADER

HEAT EXCHANGER SHELL

 

   

FUEL OUTLET

SETS COMB SPACERS INLET

SPACER WIRE Q.031 x 0055 in

66in.

Fig. 1.4.5. Small Fuel-to-NoK Heat Exchanger Type SHE-7.

 

LAODIY SSIII0Vd LI23r0dd NV

 
 

 

 

R
ORNL-LR~DWG 46134

442 kw -

1471°F Nak-TO-AIR | 1046°F
328 psi| RADIATOR  [35 pgi

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

APy = 7.8 psi , Y
' 25 psi
) . ; NaK
J 354.9Pm| oime
ReNCIK = 450,000 i
] 874 psi
APNOK = 54.6 psi )
. AT = 425°F o Y
1471°F 1046°F
32.8 psi FUEL—-TO—-NgK . 874 psi-
1594°F HEAT EXCHANGER 1247°F
56.6 psi - 25 psi |
APFUEL== 31.6 psi
. 25 psi

 

Repyg = 3580

. FUEL
] 975 gom| oive

 

 

 

 

 

 

 

 

_ 442 kw Y
1594°F | REsISTANCE | 1247°F
56.6 psi HEATER

 

 

 

Fig. 1.4.6, Operating Conditions for Tests of
Process Engineering Corp, Small Heat Exchanger
No. 1, Type SHE-7. Conditions given are approxi-
mately the design conditions for the ART,

Cold-Trap Evaluation in Heat Exchanger
"~ Test Loops

F. A. Anderson 13 J. C. Amos
Two 4-in.-dia circulating cold traps equipped

with stainless steel cooling coils have been

placed in operation with no difficulties from oxide

‘plugging. The procedure listed below, which has

been proposed for - ART cold—trap startup, was
followed in each case: -
‘1. Operate with maximum cold-trap Nak flow und

no cooling until main NaK system reaches 1200°F.

2. Turn qir cooling to maximum and maintain the
cold-trap inlet temperature above 1he pluggmg in-
dicator break temperature.

3. Reduce the cold-trap outlet temperature by'

reducing the cold-trap NaK flow as the break tem-
perature drops.

4. When cold-trap outlet temperature reaches '

300°F or less, transfer to water cooling and in-

 

YSConsultant from the University of Mississippi.

PERIOD ENDING SEPTEMBER 10, 1956

crease the cold-trap NaK flow to design rate. Ad-
just water flow ‘to obtain desired cold-trap tem-
peratures.

During ART operation it may be necessary, for
maintenance purposes, to cool the reactor to ap-

- proximately 300°F and subsequently reheat it. A

test was therefore run on the NaK loop of SHE
test stand C to determine whether such operation
would seriously upset the oxide balance or other-
wise cause cold-trap operating difficulties. With
a system temperature of 1100°F and the cold trap
operating at approximately 0.75 gpm at a NaK out-
let temperature of 300°F and an oxide saturation
temperature of 270°F, the main loop was cooled to
300°F and reheated without disturbing the cold-
trap NaK or coolant flows. No adverse effect on
the cold trap was noted, and at no time during the
course of the test did the oxide saturation tem-
perature rise above its initial level. The results
of this test are plotted in Fig. 1.4.7.

In order to obtain a comparison of actual and
predicted over-all heat transfer coefficients for the
cold-trap circuit, a series of tests were run on the
40-in. economizer and on the 4-in. circulating cold
trap on test stand C with water as the coolant
passing through the slightly flattened, copper
cooling coil wound around the cold trap. Experi-
mental over-all heat transfer coefficients were ob-
tained that ranged from 9.9 to 20.9, in comparison

UNCLASSIFIED
ORNL-LR-DWG 16135

1200

1000

_ 8OO

o MAIN SYSTEM NaK
o TEMPERATURE

@

2 600

x

Y COLD TRAP NoK OUTLET
2 a00 TEMPERATURE
L= -

 

200 '
PLUGGING INDICATOR =
. BREAK 12 &
=
0 o o
TRAP FLOW 5
: 08
8 z
o
06 g
0 10 20 30 a0 50 60

OPERATING TIME {hr} -

Fig. l..4.7. Cold'Trdp. Operation During Cooling
and Reheating of Main NaK System in SHE Test
Stand C,

51

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

with predicted values of ~20 Btu/hr-ft2.°F. The
coefficients increased significantly with increased
NaK flow (range, 0.45 to 1.5 gpm), but they were
essentially unaffected by a 30% variation in the
water flow rate. A plot of the over-all coefficient,
U, vs the logarithm of the NaK flow rate, x (in
gpm), yields a straight line, which can be repre-
sented by the following equation:

log x = 0.049U - 0.855 .

The over-all coefficients obtained with air used
as the coolant ranged from 7.2 to 10.7, in compari-
son to a predicted value of ~ 10 Btu/hr-ft2.°F, and
appeared to be, within experimental error, almost
independent of NaK flow rates in the range from
0.50 to 1.0 gpm. For both air and water cooling,
the over-all coefficients were based on the inside
circumferential area of the packed section of the
cold trap.

In the economizer unit the over-all coefficients
(based on the outside area of the inner pipe)
ranged from 458 to 664, in comparison with an
average calculated valve of 656 Btu/hr-ft2.°F, and
increased as the NaK flow rate increased from
0.50 to 1.5 gpm.

Since the ART cold traps will be equipped with
%-in. stainless steel cooling coils, the tests re-
ported above were repeated for an identical 4-in.
cold trap wound with the ART type of coil. With
water as the coolant, the over-all coefficients
ranged from 17.9 to 29.0 Btu/hr-f12.°F as the NaK
flow rate was increased from 0.43 to 1.25 gpm.
These values for the coefficient are approximately
50% greater than the values obtained for the trap
wound with the copper coil. |t is thought that the
higher values are a result of better contact of the
stainless steel cooling coil with the cold-trap
shell. With air as the coolant, the effect of the
better contact was again noticeable, the over-all
coefficients for the new trap being somewhat better
than those obtained for the original copper coil
trap.

WATER FLOW TESTS OF ALUMINUM
NORTH-HEAD MOCKUP

E. R. Dytko
R. E. MacPherson R. Curry!
~D. R. Ward

It was originally planned that the ART fuel sys-
tem components of the north head would be fabri-

52

cated and water-tested and then that the sodium
system components would be added for testing.
The desire to extend the fuel circuit testing and
the urgent need for starting the sodium circuit
testing have led, however, to the decision to build

a second oluminum north-head mockup for water

tests of the sodium system components. The

aluminum parts for the sodium-circuit water tests.

are about 75% completed.

Water tests that simulated reactor .fuel system
filling revealed that the addition of water into the
bottom of the system at a rate of 1 gpm with both
pumps running at 200 rpm or less caused only
minor ingassing. Most of the gas in the system
apparently vented upward into the surge tank
during filling. At pump speeds of 400 rpm, how-

-ever, the trapped gas was mixed thoroughly into

the liquid being pumped, and extensive ingassing
was observed.

In another set of experiments, water was drained
from the simulated fuel system while both pumps
were running at various matched speeds from 800
to 3000 rpm. It was found, in all cases, that the
liquid level dropped to the floor of the fuel surge
tank before ingassing could be observed. During
this drop in liquid level the pump discharge and
suction pressures dropped in unison, and the
pumping heads remained essentially constant,
When the ingassing occurred, the pump suction
pressure was within 1 psi of the surge tank pres-
sure. Final confirmation of satisfactory perform-
ance, with water, of the twin fuel pumps in the

north-head configuration will be sought when a

final impeller design has been established, based
on single pump tests.

Work has progressed on solving the problem of
pressure and flow surging in the test system. As

reported previously, 16 these pressure fluctuations

were believed to be due partly to the hydraulic
characteristics of the pump centrifuge section and
partly to the unfavorable pump inlet configuration.
In an attempt to separate these two effects, the
external circuit was rebuilt to provide a 10-dia
straight inlet to the pump suction. To further
idealize the pump inlet, straightening vanes were
provided in the inlet section. Subsequent opera-
tions demonstrated a 75% reduction in pressure
and flow fluctuations, and it is felt that some

 

ME. R, Dytko et al., ANP Quar. Prog. Rep. June 10,
1956, ORNL-2106, p 22. :
 

 

portion of the remaining instability will be re-
lieved by the final impeller design.

The sensitivity of the loop to pump inlet geome-
try has demonstrated the desirability of mocking
up, as a later step in the fuel circuit mockup test
program, the actual reactor fuel pump inlet region.
As a further refinement to the test unit, provisions
are being made for the later addition of a full-
scale aluminum core mockup that is to be tested
in conjunction with the mockup of the reactqr pump
inlet region. T

DUMP VALVYE DEVELOPMENT TESTS
E. R. Dytko

R. E. MacPherson L. P. Carpenter
M. H. Cooper!

Four ART prototype dump valves have been re-
ceived from vendors, but none have met the per-
formance specifications for ART operation. . _In
all cases the seat leakage rate was excessive, 17
Other difficulties encountered have included oxida-
tion of the flame-plated stem in air and fcnlure of
the seat-ring braze joint. The cause of the braze
joint failure has been corrected, but a satisfcéfery
oxidation-resistant flame-plated stem has nqt yet
been tested. 18

 

71, P. Carpenter, J. W. Kingsley, and J. J. Milich,
ANP Quar. Prog. Rep. March 10, 1956, ORNL-2061, p 60.

18 p, Carpenter, ANP Quar. Prog. Rep. June 10,
1956, ORNL-2106, p 62.

PERIOD ENDING SEPTEMBER 10, 1956

The seat materials of the valves tested thus far
were Kennametals 152B, 151A, and 162B (see
Chap. 3.4, ‘'Welding and Brazing Investigations,"’
for compositions of these cermets). These Kenna-
metals showed the greatest- resistance to solid-
phase bonding in screening tests. '

The exterior of the valve body ‘will operate in
air at a high tempe_rafure for an-extended period of
time. Therefore a ‘stem plohng which will not
oxidize is required. A Stellite:6 weld overlay on
the stem at the stem guides and an aluminum oxide
flame plate. are ‘to-be’ performance fested.

The excess;ve seat Ieakoge of the first four
prototype duinp’ ‘valves was caused by welding
distortion and. mlsahgnmen'r of the assembly prior
to final welding. - The assembly and welding pro-
cedures have’ therefore been altered in an attempt
to attain satlsfactory alignment. The seat ma-
terials were satisfactorily tested for leakage, with
water and air, ‘prior to assembly in the valves. In
each case: “the, -leakage with water and air in-
creased after the ‘valves had: been assembled and
welded. The test results on the four prototype
dump valves are-summarized in Table 1.4.6.

Several - rngs are being constructed for testing

. valve seat. motermls in contact with fused salts.

In addition ‘to the Kennametals that are being
tested, tests will be made on"molybdenum, copper,
tungsten, and titanium carbide. For these tests
the seat ring is brazed to fhe bottom of the valve

TABLE 1.4.6. SUMMARY OF PROTOTYPE ‘DUMP YALVE TESTS IN THE FUEL MIXTURE
(No. 30) NaF-ZrF - UF4 '(50-46-4 mole %) AT 1200°F

 

 

: - _Pressure 7 S
Profofyp_e ' _ Seat Leokage - 'Dif_ferenfial_ Total _
Valve No, - . (cm /hr) o Aeross Seat Stem Thrust o Reme;ks
- L s () T
_]V :‘."7'-'-‘_ 0.59 C 6 - 1200 _ Leak rate _E;;teosed with time .
- 0.9 90 1200 : Removed,tadetermine c&iuse for sticking of stem
2 - , .27.1_3 6 1200 Valve remc;{red when sedt ring pulled loose
3 055 6 700 Valve would not open with 225015 stem pull
L ' and would not fully close
4 R P - -6 - 1050 Spring-loaded valve operator used
35 ' 50 1050 After being closed for two weeks, seat leakage

increased ropidly on opening and reseating

 

53

 
 

 

 

ANP PROJECT PROGRESS REPORT

barrel, and the plug is rigidly mounted on the stem,
which passes through an O-ring seal to a hydraulic

operator. The seat ring and plug are shown in
Fig. 1.4.8. Salt from a sump is forced through a

dip line and against the seat and plug by helium
pressure. The leakage collects in the valve body
and drains through an overflow line into a tared
receiver. To ensure the removal of gas pockets
from the dip line, the plug is not seated in the
ring until after the salt flows through the over-
flow line. The stem thrust, the pressure dif-
ferential across the seat, and the leakage are
accurately measured. ,
The results of seat materials test No. 1, in
which the seat ring was Kennametal 162B and the
plug was Kennametal 152B, are summarized in

 

Fig. 1.4.9. The valve had been cycled 15 times
as of August 15, with the maximum leakage being
2.0 cm3/hr. The stem force is 750 Ib; the seat
pressure differential, 80 psi; and the test tempera-

‘ture, 1200°F. After 1000 hr in test, the valve will

be left closed for 500 hr to test for long-range
self-welding. , ,

Seat materials test No. 2 was terminated 24 hr
after its start because of excessive leakage.” The
seat ring material was Kennametal 152B and the

plug material was Kennametal 151A. The leak

rate was 38 em3/hr at 25-psi differential pressure
across the seat and a 750-Ib stem thrust. The
high leakage was caused by misalignment of the
stem. '

UNCLASSIFIED
PHOTO 28937

 

 

 

IT—II_'._l_IIITT“II]I]I‘l‘lj-l‘ [
o 1 2 3

INCHES

Fig. 1.4.8. Kennametal Seat Ring and Plug.
 

 

PERIOD ENDING SEPTEMBER 10, 1956

GECREA-
ORNL—LR-DWG 16t36

VALVE OPENED AND CLOSED; NUMBER INDICATES FORCE REQUIRED TO OPEN (Ib)

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

o oo 0 Soowl 28
3 23 ® oRRB Qo
vy v ¥ Y ¥ ¥ Y. ¥
24 _
28
RING MATERIAL: KENNAMETAL 162 B
2.0 PLUG MATERIAL: KENNAMETAL 1528
TEST TEMPERATURE : 1200°F LEAK RATE—
i.6 1 R 1C0
- DIFFERENTIAL PRESSURE
n{ ACROSS SEAT
£ L 7
- S S DA Dt BoveBoeflioe e bractioelom Lol Lo dios e ol ke Ben e Ben iy et
w12 &g bt t 7 3
<« Pl : K E
@ i § Z
¥ . : T L
< o
w w
. /\ i
1000 0.8 50 5§
— i
= o
= \1 \/ 2
E m
> o
T 750 0.4 —e e—8—g® *—0—0—=8 e—o—p—8— 25 0
- \ 5
= ISTEM THRUST ud
W w
l_
w
500 0 0
25 30 35 40

 

OPERATING PERIOD (days)

Fig. 1.4.9. Results of Valve Seat Materials Test (No. 1) in the Fuel Mixture (No. 30) NoF-Z¢F-UF

(50-46-4 mole %).

OUTER CORE SHELL THERMAL
STABILITY TEST

E. R. Dytko

R. E. MacPherson J. C. Amos
L. H. Devlin

The 300 thermal cycles scheduled for the outer
core shell thermal stability test were completed.
The model was disassembled, and the core shell
was removed for examination. Dye-penetrant and
x-ray inspections revealed no surface cracks or
internal flaws. '

Measurements were carefully made of the inside

surface of the shell by using the same procedure
as that used for obtaining measurements after
57 cycles.!? The previous measurements showed
the shell to be slightly elliptical, and after com-
pletion of 300 cycles the elliptical deformation
was even more pronounced.

 

5. Curry and A. M, Smith, ANP Quar. Prog. Rep.
June 10, 1956, ORNL-2106, p 65.

The core shell was subjected to sodium tempera-
tures of 1200°F or higher for 1050 hr. A thermal
cycle comprised 1 hr with a temperature dif-
ferential and 1 hr of isothermal operation at
1250°F. The transient time was 10 min. Bursting
pressures of 24 and 16 psi existed across the
shell at the bottom and top, respectively.

The temperatures of the sodium flowing inside
and outside the shell were measured by thermo-
couples located on the pipe walls at their respec-
tive entrances and exits to the model assembly.
During the temperature differential phase of a
‘cycle these temperatures were approximately as
follows:

Inner sodium entrance 1600°F
Inner sodium exit 1235°F
Outer sodium exit ' 1085°F
Outer sodium entrance 980°F

The temperature differential fromr the inner stream
to the outer stream, which were flowing counter-

55

 
 

 

 

ANP PROJECT PROGRESS REPORT

current to one another, was 515°F at the bottom
and 255°F at the top.

The core shell will now be subjected to a creep-
buckling test at o temperature of 1500°F ond an
external pressure of 52 psi. The time to buckle
and the type of failure will be observed. A new
shell will be installed in the thermal stability test
apparatus, and a second test program, which will
be identical to the one just completed, will be
initiated. '

AUXILIARY COMPONENT DEVELOPMENT
J. J. Keyes

High-Frequency Thermal-Cycling Apparatus
W. J. Stelzman J. M Trum_mel20

Apparatus and techniques are being developed
for investigating the effect of high-frequency ther-
mal oscillations and the resultant thermal fatigue
stresses on the ART core. Based on volume heat

 

 

2°Consu|fant from the University of lowa.

  
  

HOT FLUID

 COOL FLUID

FLOW, q\

  
   
 
 
  

=3

N _
" VELOCITY, v, \

\-.

HOT FLUID TEMPERATURE

]_-'
AT (AT gax

COOL FLUID TEMPERATURE

 

 

 

 

 

TEMPERATURE

 

 

TIME ~—————

sowce data,?! a frequency range of from 1 to
‘ ‘ g

10 cps and a surface temperature amplitude of
about 100°F have been suggested for simulating
possible ART conditions. Dynamic mixing of hot
and cold fluids appears to be a feasible method
for achieving this frequency range and amplitude.
A schematic diagram of e high-frequency pulse
pump cutrently being developed is shown in Fig.
1.4.10. Alternate slugs of hot and cold fluid are
dynamically mixed at the ‘T’ joining the legs of
the system by pneumatically pulsing the level in
two static pots joined to the hot and cold legs and
connected to a pair of pistons operating in phase
opposition. This design produces nearly sinus-
oidal temperature variations in the common line,
as well as steady flow, for convenience from the

“point of view of data analysis.

The amplitude of the temperature oscillations
generated depends on many factors, but it can be

 

MEAN GAS VOLUME, @,
MEAN GAS PRESSURE, £

21N, D. Greene et al., ANP Quar. Prog. Rep. June 10,
1956, ORNL-2106, p 222. :

.

GOMNFITTETHRLLL
ORNL-LR-DWG 16137

EQUIVALENT DAMPING LENGTH = /o

PIPE FRICTION FACTOR = '

GAS COMPRESSION EXPONENT =4, P-Q¥=C
ANGULAR ROTATION OF CRANK SHAFT = w
FREQUENCY OF PUMP PULSES = F

Fig. 1.4.10. Schematic Diagram of High-Frequency Pulse Pump and Outlet Flvid Temperature

Variation.

56

"
 

 

 

 

made to approach, as a theoretical limit,

A - (THot Leg — TCold Leg) = ATmax
max _ 2 | . 2 *

From energy-balance - consideration, it can be
shown that the efficiency of the pulse pump, de-
fined as the ratio of output amplitude to the maxi-
mum amplitude, is given by the equation

 

 

PERIOD ENDING SEPTEMBER 10, 1956

tially maximum efficiency up to 7.5 cps, with only
a slight drop in efficiency at 10 cps.

CoLD TVR_APS AND PLUGGING INDICATORS
R. D. Peak?

The first cold trap evaluation test stand, de-

" scribed previously, 22 has been shut down and dis-

 

 

2 2\ ~1/2
l 4y 2 2 4 2
— [pyy | wp [L + I[—
A AT wV 2 dq 2 dy
A AT "2 R U | ‘
max max 7 P°7Td$ gckpofld%
40, 490,

angular rotation of crank shaft,

piston displacement,

flow rate of fluid,

equivalent damping length,

= pipe friction factor,

density of fluid,

velocity of fluid,

pulse chamber diameter,

pipe diameter,

mean gas pressure,

mean gas volume,

mean height of fluid in pulse chamber,

length of pipe from pulse chamber to T,

= gravitational correction factor,

= gas compressron exponent (P-0% = a con-
stant) : : :

un u

I

o

a0 . PN NG ~e g <E
i

- A grcph of the - relahonshlp, for the Spec1flc'_‘ "
conditions indicated ‘for. water, s’ presented m:'-';
Fig. 1.4.11. The circled pomfs are test data ob- .
tained on a mockup of the ‘pump; _water was. usedi_-f " The first #ype -used seven lengths of tubing in
as the test fluid. “There is good agreement’ of the

test dqfa with the calculatlons for adiabatic com- - - ID,-instead of a plugging disk. The stand operated

- pressmn of the gas (k= 1 4) The plateais in “the for 14 days ‘during which time the plugglng indico-
- “data occurs for AT/ AT
to- the partncular plston drsPIucement Ve136x -
| ']0" #3, “used in the ‘tests, Obvaously, toin-
, 'crease the umplfiude at low and at Jmgh frequenCy,

'|t is’ necessory oniy to mcreuse fhe plston dls--".
plucement ElEs L e
_Based an: sahsfacfory ogreement of the culcula-'—_f'

ci4s

tions with water test data, a pulse pump to be used

with  the fuel mixture (No. 70) NaF-ZrF UF4
(56-39-5 mole %) has been designed for substan-

 

mantled. A third test stand is being assembled
to test the plugging indicator and cold trap planned
for use in the 70-gal NaK systems of the ART.
These units are shown in Figs. 1.4.12 and 1.4.13.
The plugging indicator is cooled by air in the
outer jacket, and the cold trap is cooled by air or
water flowing through the stainless steel tubing
wound tightly around the packed length of the

trap. The stand will have instruments to measure

the air and water flows under various operating
conditions in order to determine utility require-
ments for the ART. The stand also has provisions
for connecting both Argonne samplers and Mine
Safety Appliances samplers in order to calibrate
the plugging indicator against chemical analysis

B r-of the sodium oxide in the NaK.

The second test stand, described prevnously

‘" has. been used to calibrate two different types of
-plugging - mdlcators against chemicol analysis.

porullel ‘the 1ubmg bemg 2in. long and 0.055 in..

=1, which corresponds - tor was run-46 times and- 17 samples  were taken -
- for - chemlcai anulysas with the Argonne sampier -
"?Datu for -one 14 he. penod under - constunt condi-
" tions” showed . that, with ‘a cold- trap outlet tem-

perature of: 780°F (equlvalent to 350 ppm’ -0,),

'-"7._the plugglng mdmutor breok temperoture was .

 

22J J. Milich, ANP Quar Prog Rep Dec 10 1955,
ORNL- 2012, p 60.

23R, D, Peak, ANP Quar. Prog. Rep. June 10, 1956,
ORNL-2106, p 63.

57

 
 

 

 

ANP PROJECT PROGRESS REPORT

CONFIDENTIAL
ORNL~LR—DWG 16138

 

TEST FLUID:

PISTON VOLUME:
MEAN GAS VOLUME, Qp:

CONDITIONS FOR TEST AND FOR CALCULATION:

FLOW, g, IN EACH BRANCH:

MEAN GAS PRESSURE, FPp:

PULSE CHAMBER DIAMETER, d: .. 0446 ft
PIPE DIAMETER, d: 0.0875 ft -
MEAN HEIGHT OF FLUID IN PULSE CHAMBER, L: 0.25 ft

LENGTH OF PIPE FROM PULSE CHAMBER TO TEE, /: 4.5ft

3 ADDITIONAL ASSUMPTIONS FOR CALCULATIONS: FLOW FRICTION FACTOR,
{x 10°) f’=0.03; EQUIVALENT LENGTH OF PIPE, /, = 58 DIAMETERS

WATER

0.0478 ft3/sec
.36 x 4073 £t3
22.8 x10°3 ft3
29.5 x 144 psfa

 

-
N

' CURVES ARE FROM CALCULATIONS

 

 

/ IA\ POINTS ARE FROM EXPERIMENTAL DATA
1.0 L ,
/ \ |
/v \\ AT
0.8 17 \ AT yax

 

 

0.6 | 47

 

\\‘/— ADIABATIC COMPRESSION (4 =1.4) -
| G ,

\\O\\i>

X

 

 

1/

 

TEMPERATURE RATIO, AT/A T yax, PER UNIT
VOLUME OF PISTON DISPLACEMENT, V (ft3)

 

ISOTHERMAL COMPRESSION {4 =1)

 

Vi

 

 

 

 

0 l
0 2 4

6 8 10 12

FREQUENCY (cycles /sec)

Fig. 1.4.11. Response of Pulse Pump as a Function of Frequency.

615°F (equivalent to 135 ppm O,), while chemical
analysis of the NaK gave only 82 ppm O,. This
type of plugging indicator was considered un-
satisfactory, because the characteristic break in
NaK flow was not obtained for oxide saturation
temperctures below 400°F. This made interpreta-
tion of low oxide contents virtually impossible.

The second type. of plugging indicator used o
plugging disk with 18 holes 0.31 in. in diameter

and one hole 0.049 in. in diameter. The stand

58

operated for 25 days, during which time the plug-
ging indicator was run 68 times and 47 samples
were taken for chemical analysis with two Argonne
samplers. Averaged data for a 24-hr period under
constant conditions showed that, with the cold-
trap outlet temperatwre at 900°F (equivalent to
470 ppm 0,), the plugging indicator breck. tem-
perature was 640°F (equivalent to 160 ppm O,),
while the chemical analysis of the NaK showed
170 ppm O,. Performance of this plugging indica-
tor was, in general, satisfactory. S
65

 

 

 

 

ADONPIDENEL
ORNL-LR-DWG 16139

   
        
 

THERMOGOUPLE

0.034~in~DIA HOLE
0.050-in—DIA HOLE

3jg=in. SCH 40
INCONEL PIPE

 

Fig. 1.4.12, ART Plugging Indicator,

9561 ‘0l ¥3dWILdIS ONIGNZ QOld3d

 

 

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

Y% —in-DIA. HOLES

STAINLESS STEEL TUBING

Va~in. INCONEL PLATE,
ROLL TO 8%-in. OD

YORK DEMISTER PACKING

AIR OR WATER IN \{
NaK OUT —\ ©

BENEHBENT I
ORNL—-LR-DWG 16140

THERMOGOUPLE WELL

   
   
  
   

Fig. 1.4.13. ART Circulating Cold Trap for 70-gal NaK System,

LIQUID-METAL-VAPOR CONDENSERS

M. H. Cooper

The specification of bleed gas purges through
ART sodium and NaK pumps requires that con-
densers be provided for removing liquid metal
vapor from the exit helium streams. in addition,
the NaK dump tanks must be provided with similar
condensers capable of removing the NaK vapor
from the helium exhausted by an emergency NaK
dump.

Development work on such condensers is under
way, and the dimensions and design parameters of
the units currently being investigated are given
below:

For Na For NaK
Vapor Vapor

Condenser diameter %-in.-IPS pipe  l-in. tubing

Condenser length 60 in. 36 in.
 Helium flow rate 500 liters/day 1.7 cfm

Inlet temperature 1200°F 1200°F

Outlet temperature 200°F 600°F

60

Schematic test layouts are shown in Figs. 1.4.14
and 1.4.15. For the NaoK test, helium at 1200°F,
saturated with NaK vapor, was periodically vented
from the 15-ft3 sump at a flow of 1.7 c¢fm and
exhausted to the atmosphere through the con-
denser. Two condensers were tested, the second
of which was refined by adding external fins
(2}’2-in.-dia Inconel, eight per inch) and internal
Demister packing. -

The first test, for which the conditions were
those tabulated above except that the outlet tem-
perature was 300°F, was terminated after the
second cycle because of entrained NaK in the
control panel. The second test, for which fins
were added to the condenser and the outlet tem-
perature was 100°F, was terminated after 70
cycles because of entrained NaK in the control
panel. A larger trap downstream of the condenser
was used in the second test and may have been
responsible for the improved performance.

The test results obtained to date indicate that
the NaK condenser design is inadequate for ART
 

 

 

g e

He {4.7 cfm}

 

 

 

 

 

— W =t — NaK

 

 

 

 

 

 

 

{(1200°F)

  

He (500 l'ifers/doy)—-- ‘3/ in. PIPE
. B :

CONDENSER

 

 

1-in. TUBING

symMP o o  TRAP.

   

PERIOD ENDING SEPTEMBER 10, 1956
UNGLASSIFIED
ORNL=-LR~DWG 16141

THROTTLE
VALVE

   

 

EXHAUST TO
ATMOSPHERE

   
 

 

ROTAMETER

Fig. 1.4._14.i.|l.nlyout of Nc_l(i-Vupor-Condensgf'.l_'e;t Apparatus,

UNCLASSIFIED
ORNL—LR—DWG 16142

)__,,—;—60 in-—"""”-l

 

 

 

 

 

 

 

EXHAUST
. _
AL
WET-TEST
GAS METER
TRAP

Fig. 1.4.15, L‘.cl:you_t of Sod ium-Vupor-Condenser Test Apparatus,

application. A reliable ‘gas‘-'lic';uid:' separ'a'tér_ must’
be used to remove condensed NaK entrained in the
helium stream. - e Lo

B For fhé 56dium"'fe§f; helibm' at a flow of 500

liters/day was bubbled through a small sump con-
taining sodium ot 1200°F and vented to the atmos-
phere through the sodium condenser. The sodium
condenser plugged 10 hr after flow commenced. A
sodium oxide plug had formed at the Swagelok
joint between the sump and the condenser inlet

_tube. Therefore the test equipment is being re-
‘designed to eliminate the mechanical joint.

ZIRCONIUM FLUORIDE VAPOR TRAP
F. A. Anderson M. H. Cooper

High-temperature chemical absorbents and ther-
mal-precipitation traps are being investigated in

the program for the development of a zirconium

fluoride vapor trap. The most promising system
consists of a bed of hot aluming which reacts with

gaseous ZrF , to form solid ZrO, and AIF,. All

61

 
 

 

 

 

 

ANP PROJECT PROGRESS REPORT

the thermal-precipitation traps tested have formed
ZrF , plugs at the initial cold section.

The second test with an alumina bed, initicted
previously,24 was terminated after 450 hr of opero-
tion because of flow stoppage. X-ray diffraction
analysis of the packing showed that the Al,O,
had reacted with the ZrF , according to the follow-
ing equation:

3ZeF, + 2A1,0, —> 3210, + 4AIF,

Wet analysis of the packing showed that the reac-

tion was about 90% complete. The wet analysis

also showed that the plug was caused by thermal

precipitation of ZrF ,, in the cooled outlet section
of the test piece, ofter the Al,0, had been

depleted.

A shell-and-tube prototype trap2® was packed
with 4- to 8-mesh Al,O, pellets. The schematic
flowsheet for the test is shown in Fig. 1.4.16. -

 

24
M. H. Cooper, ANP r. Prog. Rep. June 10, 1956,
ORNL-2106, p 63. Qua & Reb

25), J. Milich and J. W. Kingsley, ANP Quar. Prog.
Rep. Dec. 10, 1956, QRNL-20'I2, p 60.

Helium was bubbled through the sump containing
the fuel mixture (No. 30) NaF-ZrF -UF , (50-46-4
mole %) at 1500°F at o rate of 5000 liters/day and
exhausted through the trap, which was heated to
1350°F. This test, in which the fuel temperature
was 100°F higher than the temperature anticipated
in the ART, operated for 690 hr before fiig pres-
sure required to maintain the ART design flow
rate started to increase. Examination of the pack-
ing after termination of the test indicated that
the flow stoppage had been caused by thermal
precipitation of ZrF, after the Al O, had been
consumed.

The test of UF, pellets as a high-temperature
absorbent for ZrF, was unsuccessful. The ZrF,
was. not removed from the helium, and plugs of
thermally precipitated ZrF, were formed in the
outlet of the test piece. |

A thermal trap was modified by the addition of
four water-cooled baffles, as shown in Fig. 1.4.17.
Helium, ot the ART design flow rate of 5000
liters/day, was bubbled through the fuel (No. 30)
at 1400°F and passed through the trap. This test
was terminated after 231 hr as a resvit of a ZrF
plug, shown in Fig. 1.4.18, at the first baffle.

VEORCY
ORNL-LR~DWG 16143

TRAP PACKED WITH Al;,0,

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

RISER
He {5000 liters /day) (1350 —1400°F}
i —W_—-_\
‘ e 40 in, ———fa— {2 in. —
1350-1300°F 1300 -80C°F
Iz — -3 =
_ .O —_— . = - —_—— - =__
- — - _ P WET-TEST
| _ - = GAS METER

 

 

 

FUEL
(1500°F) .

 

 

 

 

 

 

WATER TRAPS

Figs 1.4.16. Schematic FI_&wsheet for Tests of Al, 0, Trap for ZrF, chér. '

62
 

{
i

 

 

 

 

 

~

PERIOD ENDING SEPTEMBER 10, 1956

SEGRET
ORNL-LR-DWG 46144

COOLING
WATER IN

f

WATER COOLING
coiL

THERMAL-TRAP
TUBE BUNDLE

 

  
 

   
 

 
 

BAFFLES

' - COOLING
‘ WATER OUT
5 in. 5 in. 5in. 5 in. 5in, I 50 in. =j

 
 

£
i

 

 

Fige 1.4.17. Modified Thermal Trap for ZrF ¢ Yapor,

 

 

Fig. 1.4.18. Pfug of ZtF, at First Baffle of Modified Thermal Trap. Deposit broken through to

show thickness. (Secret with caption)

63

 
 

 

 

ANP PROJECT PROGRESS REPORT

1.5. PROCUREMENT AND CONSTRUCTION
W. F. Boudreau

ART FACILITY
F. R. McQuilkin

Construction work on the contract portion of the
ART facility in Building 7503 is in the final
stages. Work on the building edditions, building
alterations, and cell installation (package 1) has
been completed, with the exception of installation
of the last six circuit breakers (which were delayed
by the Westinghouse strike) and completion of two
modifications to the cell floor structure (which
were ordered late in the contract period).

Work performed during the quarter included com-
pletion and testing of the cell; installation of
mechanical items, such as the penthouse coolers,
a 3-ton hoist, the air duct liner and insulation;
electrical testing; painting; grading and paving;
and miscellaneous cleanup jobs. The contractor
vacated his site office on July 20, 1936.

In order to ascertain the leck tightness of the
reactor cell, the 24-ft-dia tank was completely
assembled (with temporary test plates over all
nozzles) and subjected to a rigid leak test. Fol-
lowing a satisfactory pressure rise test of 50-hr
duration, the annulus between the two vessels was
flooded with 21,000 13 of circulating helium for
the leak-rate measurement, By using standerd
leaks of 14.7 and 28.24 micron ft3/hr on the sys-
tem, it was established that the two Model 24-102
Consolidated Engineering Company leak detectors
could detect a system leakage rate of 5 micren
f13/hr or larger. Inasmuch as a zero leak rate
was measured and the contract specification
allowed 32 micron f3/hr, the vessel satisfoctorily
met the leak tighiness requirements.

Work on the installation of auxiliary services
piping (package A) has been completed, with the

exception of replacement of 15 valves in the nitro-

gen system. Erroneous vendor information led to
the installation of incorrect vaives for the service.

They were rejected when pressure and leck testing -

revealed the error and their inadequacy. Other-
wise, work performed during the quarter included
completion and testing of the contractorsestablished

portions of the lube oil, hydraulic drive oif, cooling

water, gir, nitrogen, helium, ond vent piping
systems,

The work which includes the installation of the
diesel generators and facility, electrical control

64

centers, and spectrometer room electrical air con-
ditioning equipment (package 2) was at the 91.8%
completion point on September 1, 1956. All work
has been completed on this contract, with the ex-
ception of the installation and testing of the diesel
generators.  This work has been delayed because
one of the deisel-generator units and three control
panels were damaged in shipment, The damaged
components were returned to the manufacturer for
repair and are scheduled for reshipment early in
October 1956, Special vibration testing of the
units will be performed during the field tests to
ascertain that all damage has been accounted for.
Two other items of work were negotiated with the
package 2 contractor for execution while he com-
pletes his other work. The contractor was given
an order to remove the shield blocks now occupying
three sides of the ARE pit to nearby outside
storage so that this building space can be used.
To provide compliete access into the high bay
through the north door and to provide greater flexi-
bility in the use of the building cranes in unloading,
the contractor was also given an order to remove
the parapet wall above the floor level at an-eleva-
tion of 852 ft around the ARE pits. The roof plugs
for the pits will provide the floor surface over
them,

Design work continued and installation work was
started on package 3, which concerns the installa-
tion of process piping, process equipment, etc.
Completion of the design is currently scheduled for
October 1, 1956. During the quorter two of the

four main blowers were installed.

Some of the construction work may be seen in

Figs. 1.5.1 through 1.5.13.

Program and design planning for disassembly of
the ART were initiated. The problem is being
handled in two basic parts; the first being that of
removal of the reactor from the cell, and the second
that of transporting and disassembling it in hot-
cell facilities. Procedures and tools necessary
for removing the reactor are being developed. A
new, large hot-cell facility is being planned. It is
proposed that the new facility be constructed in
conjunction with the construction of a group of
smaller examination cells. The tentative location
of these cells is on the north side of the ORNL
Area adjacent to the Bethel Valley Road. Transfer
 

 

o
w

 

 

s —— A

 

 

 

' UNCLASSIFIED
PHOTO 177233

 

Fig. 1.5.1. View Looking South Across the 7503 Cell. In the foreground is the top of the cell inner vessel. At the top right is
the concrete penthouse structure which will house the primary coolant pumps and their drive motors. To the left of the penthouse
are the roof plugs which will cover the special equipment room immediately below.

9561 ‘01 ¥39WILd3S ONIANI 4oly¥3d

 
 

 

 

 

AT

UNCLASSIFIED
PHOTOC 17960

 

o . . , % 3
b . - 4

Fig. 1.5.2. View Looking South from the North End of the Original 7503 Building. This view, showing the fop of the call water

tank in place as it will be located during ART operation, was photographed during the vacuum-testing operction on the inner vessel,
which is enclosed by the water tank. | ) __ - S

LY¥0dIY SSITY20V¥d LI3r0¥d ANV

 
 

O~
~I

 

 

UNCLASSIFIED
PHOTO 17961

 

"."'i“g.'l_"‘l.5.3., Vievlv_: Lodkihg. N.orf'hl, Acfoss the Cell While the Water Tank Top Is in Place. At the lower left are the penthouse
structure and the area in which the main coolant pumps and motors will be located.

9561 ‘0l ¥33IWILd3IS ONIANT QOId3d

 

 

 

 
 

   

¥ ¥
e ??‘i”?}a

 

Fig. 1.5.4. View Looking Norfh Across the Opened 7503 Cell Tunk. The rectangular plafes shown on ihe floor sfructure of the
inner. vessel will support the reucfor. The circular plates shown will support the shielded fuel fill-and-drain tank and fuel removal
tank. The 24-m.-d|o nozzles through the sude wall of the vessel will provide for the penetration of the insfrumentaflon and control
lines, lube oil lines, hydrouhc drive imes, cooling water lines, gas lines, and heater leads to the cell equipment.

LA0dIY $SSIVO0V¥d LDIFr0dd dNY

i

 
   

 

69

         

 

 

UNCLASSIFIED
PHOTO 18080

 

Fig. 1.5.5. View Looking North Across the Completed and Painted 7503 Cell Tanks. The floor 'structure was removed from the
inner vessel at the time this photograph was taken and therefore the fluid distribution weirs located in the bottom of the vessel may
be seen. The scalloped ring shown outside of the weirs is the support for the floor structure. At the exireme bottom of the photo-
graph are the NaK piping sleeves with their expansion joints which penetrate the two cell tanks. To the right of these sleeves are
portions of three of the 24-in.-dia spectrometer tubes.

9661 ‘0l ¥IIGWILJIS ONIANI QOId3d

 

 

 
0L

 

' UNCLASSIFIED '
| .-, PHOTO 18032

 

 

 

 

Fig. 1.5.6. Vnew Lookmg Soufhwest Towutd the Modified Bui|ding. In th_e_'fkbr‘egr‘oondl cré _tHe two top héad's.fo'r ‘:the‘ cell tanks
whnch hove been removed to storoge outsude the facility. o : ‘ . ‘

LY0d3IY SSIAO0Vd LIIr0dd dNV

 
                 

 

1L

UNCLASSIFIED
PHOTO 18033

 

 

 

 

 

 

Fig. 1.5.7. Vie\.n'Looking. Southwest Toward the Diesel-Génerafir House. In the lower left is the substation for the 13.8-kv
purchased-power supply from the TVA system. The gas cylinders in the lower center of the photograph will be used for storage of
nitrogen during ART operation.

 

9561 ‘0L ¥39WILJIS ONIANI GoiId3d

 

 
ZL

 

 

 

 

 

 

 

 

 

- Fig. 1.5.8. View Looking West Inside the Diesel-Generator House Which Shows a Stage of the Insfallofion Work on the. Five
Diesel-Generator Units for the Auxiliary Power System,

13043y SSIFH00dd LI23r0odd dNY

 
 

 

 

 

 

 

        

UNCLASSIFIED
PHOTO 12966

 

Flg. 'I 5 9, View 'Tuken in the Switchhouse Which Shows the Primory Switchgear and Instruments Associated With the Receipt
and Dlstribuflon of the: Two Power Supplies to the Facility. The switchgear and instruments on the left will serve the i incoming
purchused-power from TVA. Those on the right are for the diesel-generator power.

~4
W

9561 ‘0L ¥IIWILJIS ONIANI AOI¥3d

 

 

 
vL

 

 

T TR
UNCLASSIFIED
“PHOTO 18233

 

- Fig, 1.5.10. View Taken Inside the Blower House Which Shows the Supply End of the Main Cooling Air Duct and Installation
Work on Two of the Four 82,000-cfm Blowers. On both sides of the main duct are the 10,000-cfm blowers which will supply air into
the annulus between the building concrete and the insulated steel duct for the purpose of preventing overheating of the building
structure, At the lower left is the opening for the ramp which leads down to the radiator pit beneath the cooling air duct.

-

130d3Y SSIIO0Vd LI3rQ0dd dNV

. .
¢ a roa ¢ " » L . e o

 
74

R e e et

 

 

UNCLASSIFIED
~ PHOTO 18152

 

 

 

 

 

 

 

 

 

Fi.g;'Jll.S..i.l.\ ViewLooking -Wésf and from Inside the Cooling Air Duct. This view shows the 16-gage stainless steel facing over
the insulated steel liner of the duct. The supply end of the duct is the opening shown in the center of the photograph.

9561 ‘01 ¥3GW3ILJ3S INIANI QOI¥3d

 
9L

 

 

 

UNCLASSIFIED
PHOTO 18153

 

 

 

 

 

Fig. 1.5.12. View Taken Inside the Main Air Duct from the Base of the Discharge Stack. This photograph shows the portion of
the duct in which the NaK-to-air radiator banks will hang. At the right center are the exterior facing of the cell water tank ond the
sleeves through which the NaK pipe lines will pass. Above the duct, fhrough the opening shown, is the penthouse. Below the duct,
through the opening in the floor, is the radiator pit.

LI0dTY $53490dd LI3r0dd ANV

 
 

LL

 

           

UNCLASSIFIED
PHOTO 17734

e e

 

Fig. 1.5 '|3. View Showing the Southwest Corner of the Auxilmry Equipment Room. This room, which was used during ARE
operation as fhe heat exchanger pit, is being modified to serve as the auxiliary equipment room ond will house the lubricating oil
pumping system ond the hydraulic drive ‘equipment.

 

9661 ‘01 ¥IGWILJIS ONIANT aold3d

 

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

of the rodioactivé reactor from Building 7503 to
the disassembly cell would be by a shielded *‘low
boy’’ carrier and from the larger cell to the smaller

cells by means of a conveyor in a shielded transfer -

ond storage facility.

ART-ETU REACTOR CONSTRUCT'ON
M. Bender - G.D, Whltman
. ETU Facility |
To J. BO”esl » V. Jo Ke“eghun
P,A.Gnadt = A, M. Smith

Construction work has started on the ETU facility.
Structural steel is being fobricated, and erection of
the stand from which the NaK pumps and piping
are to be supported has begun. The H. K. Furgerson

Company is procuring steel which will be installed
under the track floor to support the weight of the
stand. Designs for this work are complete. Design

of the control room enclosure is complete except
for minor details. -

The ‘details of the main NaK piping have been
determined, and piping layouts are being reviewed.
The economizers and the plug indicators for the
NaK cold-trap loops are being fabricated.

The design of the ait duct for the NaK-to-air
radiators is being reviewed. The air duct is to be
supported from the stand which supports the NaK
system. A variable-speed blower will send air
through the duct and across the radiators. The
air will be discharged outside the building.

A second layout of the piping for auxiliary
services has been completed. The auxiliary

services include instrument air, plant air, water, .

helium, lube oil, and hydraulic oil. Changes in
the system requirements have caused delays in

the start of construction of these systems. Neces-

sary changes have been made in the existing gas
pipingto the furnaces which will be used to provide
the heat load for the ETU.

The basic electrical system has been de.fugned |

A normal-supply substation of 1500-kva capacity
will be used, This power will be distributed by
means - of a 450-v switchgear unit, Individual
circuit breakers in this switchgear unit will supply

_ power to six wound-rotor NaK motors; to a process-
air-blower motor; to a motor control center for the
‘reactor pump drive motors; fo motor control centers

for "the lubrication pump drive motors; fo trans-

 

" Ton assignment from Pratt & Whitney Aircraft.

78

former banks which will, in turn, supply power for
controls and instrumentation, valve actuators, and
control - motors; and to two sets of transformer
banks which will, in turn, supply power to fhe
heating circuits.

An emergency electrical supply will also be
available which will consist of @ 300-kw capacity,

' 480-v, 3-phase, diesel-generator - unit.  Upon

failure of the normal supply this emergency unit

~will supply power through automatic switching

devices for an orderly shutdown of the equipment.

 The spare lube oil systems drive motors, instru-

ments and controls, valve actuators, and some

~ necessary heaters in the system will continue to
. function until it is felt that the experiment can be

safely stopped or restarted. The drives for the
fuel, sodium, and NaK pumps are not connected to

-this emergency supply because of its limited
- capacity. - Preheat and additional auxiliary heat

required during the course of the experiment will
be supplied by means of ceramic clamshell and
tubular heaters designed for high-temperature
applications.  The 1500-kva substation, 460-v
switchgear unit, and the emergency diesel gener-
ctor have been installed,

ART-ETU Reactor Fabrication and Assembly

R. Cordova C.K. McGlothlan

The study of the reactor assembly problem has

continued, and general assembly procedures have

been prepared and reviewed. A sequence of opera-
tions has been outlined which covers the assembly
of the reflector-moderator, nérth head, heat ex-
changers, island, and the various shells into a
complete reactor. Detailed instruction sheets are
being prepared from this general outline,. which
will serve as the assembly manual for the craftsmen
who will assemble the reactor. - 7

The more detailed review of the reactor assembly
problem has indicated the need for many jigs and
assembly fixtures, and design of such tools is
proceeding. A device for checking the concen-
tricity of the reflector-moderator components about
the polar axis during assembly has been designed.

“Weld ‘shrinkage tests on geometries pecuhor to
the reactor design are continuing (see Chap. 3.4,
“Welding ond Brazing - Investigations® ). It is

‘necessary that the weld shrinkage be predictable

to within a few thousandths of an inch if the close
tolerances designed into the reactor are to be held.
 

 

 

"

This is porhcuiarly true in the fitup of the thin
_lnconel shells surroundlng the reflector-moderator.-

- Nearly all the component parts and spare parts

-for three reactors have been ordered. Fabrication

of the north - head for the ETU reactor has con-
tinved. wflhout dlfflculty and is approxumately 40%

| complete. -

A gas-drying facllxty is. being desngned for ine
stallation in the Y-12 foundry, This equipment

“will be used to supply the controlled atmospheres -

necessary for proper heat treutment of the north
head and snmllar weldments, ‘

 The lower half of the berylhum reflector-moderator

has been contoured, and drilling of the coolant
holes has been started. A rough dimensional check
of the contoured piece of beryllium mdlcated that
it was acceptable,

The water-flow tests to be run on the first re-

- actor assembled have been defined, and design

work is practically complete on the equipment for

the first of these tests, which is to be run on the

beryllium reflector-moderator,  Four water-flow
tests are scheduled for the critical sodium circuits
to determine flow distribution and pressure drops
in the reflector-moderator and island sodium
passages. Dummy ‘parts will be used in some
cases in order to expedlte these tests.

The construction work on the reactor assembly

and inspection area was completed. This area
has been air conditioned and a 20-ton bridge crane
has been reinstalled over the building main floor
for lifting-access to the assembly area through a

hatch equipped with removable covers. Space has

also been made available for a degreaser, a vacuum-

drying chullty, and a storage area. for reactor parts.

ART Cell Components
‘ A M Smlth -

Desugn work on various ART ceII components',
'.__.':cnd mstallatlon loyouts is contlnumg. Loyout
_ . .drawings of the lead shield, the top support plate
- form, the fuel flll-and-dram tank, piping manifolds, -
~_auxiliary piping, and the junction panel and junc-
tion panel expansion joint have been completed,
- and- fabrication and procurement are under way.
" These. items constitute approximately 5% of the -
* ‘equipment ‘needed - for the cell,
-hydraulic piping layout ms|de the cell is approxi-
.mately 90% complete, and the lube oil plplng layout

is about 40% complete.

Design of ‘the

PERIOD ENDING SEPTEMBER 10, 1958

ART-ETU REACTOR COMPONENT
~ PROCUREMENT )
W. R. Osborn. J. Zasler
- The fabrication of the fuel-to-NaK and the sodium-
to-NaK  heat exchangers still appears to be one
of the most difficult procurement problems. Al-
though considerable progress has been made, much

remains to be done in refining the design ond in

developing machining, tube bending, welding, and

. brazing techniques for both these heat exchangers.

Fuel-to-NoK Heut Exchangers

The York Corp. has ordered all the toolmg
necessary for production of the fuel-to-NaK heat
exchangers based on the use of siretch-formed
tubes and of channels made from premachined
parts, which will be welded and stress-relieved

-as afinal operation, Black, Sivalls & Brysen, Inc.,

feel that the channels must be machined after

‘welding to hold the required tolerances, and they
are providing special tooling for their large boring
~mill for this purpose. They are also developing

at least two different means of forming tubes to

avoid some of the uncertainties of stretch-forming,
 Work is being carried out at ORNL, as well as at
- the plants of the two vendors, on the various

welding and brazing problems, and some design
changes are being studied for overcoming existing

difficulties with close tolerances, accessibility

for welding, and distortion problems. -

Sodium-to;Na'K Heat Exchangers

Minor redesign of the sodium-to-NaK heat ex-
changers and major improvement of the vendors’

- tooling have -been found necessary to permit fabris
~ cation of this unit ‘within the required tolerances.
“This work is now in progress, and should result in’
' the dellvery of the sodium-to-NaK heat exchongers
"~ for the ETU durmg the next quarter,

Core Shells

The modlflcatlon of the Hydrospin machme to

- increase the size of the shell_s that- can be pro-
" duced has been completed, and a number of reason-
‘ably successful efforts have been made to spin

the |s|und core . ‘shell.  The mandrels for the outer

_core . shell are ‘essentially complete, and both
these shells should be available for shipment to
ORNL - in September. The mandrels for all the

remaining shells are in various stages of manu-
facture and will be completed. during the next
quarter,

79

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

Beryllium ReflectorsModerator

- The Brush Beryllium Co. is now machining both
* the north beryllium hemisphere for the ETU (to be
shipped about October 5, 1956) and o hemisphere
for the ART., The major difficulties in the fabri-

cation of these units seem to have been overcome,

Boron Layers

Acceptable samples of finished boren carblde
tiles have been received from the Nerton Company,
and no major difficulties are expected with pro-
duction of these components. Sample sheets of
the boronecopper cermet have been received from
Allegheny-Ludlum which are acceptable except
for surface finish. Some difficulties are antici-
pated in forming and finishing this material.

80

Pressure Shell

An order has been placed for forgings for the
reactor pressure shell with the Ladish Co. These
forgings will be muchmed by fhe Allls-Cholmers
Mfg. Co.

|ncane|

The availability of adequate supplies of Inconel
in the proper forms that are acceptable according
toe ART processing and inspection standards cons
tinves to be a major problem and a limitation on
the speed of fabrication for many items. A great
deal of work is being done with the International
Nickel Company and the Superior Tube Co. toward
the solution of this problem.
 

W "

 

 

PERIOD ENDING SEPTEMBER 10, 1956

1.6. IN-PILE LOOP DEVELOPMENT AND TESTS

H. W. Savage

OPERATION OF IN-PILE LOOP NO. 5

VVC. C. Bolte! B. L. Greenstreet
J. A. Conlin D. M. Haines!
R. A. Dreisbach! M. M. Yarosh

‘In-pile loop No. 5 was inserted in the MTR on
June 11, 1956, but it could not be filled. It was
therefore returned to ORNL for disassembly and
inspection. The fill system consisted of a fill
tank, a vent line, and a fill line to the pump sump.
In this system the fluid flows by gravity into the
preheated loop when a plug of frozen fuel mixture
is melted out of the fill line. Upon disassembly

‘and analysis a section of the fill line was found

to have ¢ plug formed of a 50-50 mixture of fuel
and zirconium oxide. Subsequent examination of
the initial batch of fuel showed it to contain 0.5%
zirconium oxide. It was also determined that
repeated melting and freezing of the fuel salt could
cause the zirconium oxide to separate out and
settle in the bottom of the fill tank; however, it

has not been possible to positively establish that

this was the mechanism of formation of the oxide
plug in the fill lines. Another consideration was
the possibility of accidental exposure of the
molten salt to moisture. However, the evacuation
and purging of the system with inert gas, which
were done prior to filling the tank, plus the evacu-
ation before and purging during the attempted

melt-out preclude the presence of. air-or moisture .
in the system. - Close control was held on loop
cleaning during assembly. Future. |oops will be
filled ‘with salt” produced under stricter ‘control,
-and every precautlon will be. conflnued to prevent
“air and moisture from - entermg the - loop ‘prior ‘to -
and after. charging of - the fill tank. - Increased -
o efforts will be -made to-minimize - thermal cyclmg
‘durmg mmol f;llmg und melt-out of the flll system.'

 

10“ dé?ignmefif from Pran & 'w”"i'l;ne'y"Aircrdft, e e

D. B. Trauger

IN-PILE LOOP NO. 6

In-pile loop No. 6 is nearing completion and is
scheduled for insertion in the MTR on September 3.
This loop is similar to previous loops, with the
principal variations being the mstallahoh of a
modified (Mark 11) horizontal-shaft sump pump,
lmproved electrical hermetic seals between the

" nose and intermediate, or bearing housing, region,

and the use of fuel in which the zirconium oxide
content has been minimized. The planned oper-
ating conditions for this loop are a 1600°F
maximum loop temperature with the fuel mixture
(No. 44) NaF-ZrF -UF, (53.5-40-6.5 mole %), a
300°F temperature daf?erenhal across the nose
coil, and a power density of about 0.75 kw/em3.
These conditions are to be maintained over two
MTR operating cycles to give approximately 700 hr
of operation with the reactor at full power.

HORIZONTAL-SHAFT SUMP PUMP FOR IN-PILE
LOOPS

J.A. Cpnfin 'D. M. Haines
- W.S. Karn!

A prototype (Mark 11) horizontal-shaft sump pump
for in-pile loops was modified to increase the
radial clearance between the pump housing and
shaft slinger from 6 to 60 mils in an effort to
prevent shaft seizure such as that which occurred
previously because of ZrF, vapor deposits.2 This

_-pump, which is identical to that installed in loop
‘No. :6,-is now being tested and should have ac-
,cumulated 1150 hr of operation by the MTR in-
: semon date for foop Ne. 6. The’ pump temperature

in_toop-No. 6-will be hlgher than in previous tests,

- buf it will be 50 to 75°F lower than the _temperature
of the test pump now operating. ~It should therefore’
~have @  lower rate . of zircomum fluoride ac-
i c’Urfiuldtion than the test pump wnll have. :

 

2w. S, Kam, ANP. Quar. Prog. Rep. ]une zo, 1956.

ORNL-2106; P 76.

81

 
 

 

 

 

 

ANP PROJECT PROGRESS REPORT

1.7. ADVANCED REACTOR DESIGN

W. K. Ergen

GRAPHITE AND BERYLLIUM OXIDE
MODERATED CIRCULATING-FLUORIDE-FUEL
REACTOR

W. K. Ergen

The complexities involved in the use of sodium

to cool the beryllium reflector-moderator of the

ART have directed thinking toward a reactor design
in which the moderator is cooled by the circulating
fuel. A new design now being studied is based

~ on a multitubular core, similar to that used for the
ARE,. in which fuel passages would penetrate o

moderator block or, altematively, moderator rods

‘would be immersed in the circulating fuel.

If the fuel is to accept heat from the moderator,

~ the moderator must be at o higher temperature than

the fuel. This eliminates  beryllium metal, as
used in the ART, from consideration! because of
its poor mechanical properties at temperatures
above the fuel temperature of interest, that is,
1600°F. Therefore graphite, beryllium oxide, and
the hydrides of zirconium and yttrium are being
considered as possible moderators.

The required volume of the multitubular reactor
core will, of course, depend largely on the desired
total reactor power and the permissible power
density in the fuel. A 24-in.-dia sphere has a
volume of 1.2 x 105 em3, and, if it is assumed,

optimistically, that 80% of this volume could be

occupied by fuel, about 105 cm? of fuel volume
would be available, Thus, in a 24-in.-dia multi-
tubular core, 100 Mw of power could be generated
with a power density of 1 kw/cm3, which is
approximately twice the average power density of
the ART, and 300 Mw would require 3 kw/cm3.
It is clear that in many cases of interest, therefore,
larger, rather than smaller, core volumes may be
necessary. Thus far, this study has been limited
to an evaluation of 24-in.-dia spherical cores, but
the general conclusions are applicable to larger
cores.

The rate of heat generation in the moderator and
the resulting thermal stresses would be of prime

 

Vit would, conceivably, be possible to use beryllium
in some reglons of the reactor, such as the cutside of
the reflector, and to cool these regions with the rela-
tively cold fuel coming from the heat exchanger.

82

A. M. Perry

importance. The gamma-ray heat sources? have
been shown to be prompt gamma rays plus U235
capture gamma rays (9 Mev), fission-frogment-decay
gamma rays (1 Mev, based on 20% of the fuel being
inside the core), gamma rays resulting from cap-
tures in material other than U235 (5 Mev, depending
strongly on the *‘thermal utilization” of the
reactor), and gamma rays resulting from inelastic
scatfering (1 Mev). This adds up to about 8% of

the fission energy. The density of the fuel was

considered to be 3.2 g/cm3, which is approximately
the density of the fuel mixture (No. 30) NaF-
ZrF -UF, (50-46-4 mole %), Hence, the radius
ussumed for the core would correspond to 2 or 3
reloxation lengths of the gamma rays, even if
aliowance were made for the moderator, which
would occupy some of the core and would have a
somewhat lower density than that of the fuel ‘Most
of the gamma rays generated at the center of the
core would be absorbed in the core. At points
closer to the surface, the power generotion by
gamma-ray absorption would be less.

The gamma-ray absorption would be distributed
between the fuel and the moderator approximately
according to the densities. In a volume filled to
80% by fuel of density 3.2 and to 20% by graphite
of density 1.8, the graphite would contribute about
'/ to the density. Beryllium oxide, of density
2 8, if it occupied the same volume fraction in
the saome fuel, would contribute about ¥ to the
density. With a heat generation rate in the fuel
of 1 kw/cm3 and a fuel volume fraction of 80%,
0.8 kw/cm® would be generated, on the average,
over the core; 8% of this, or 64 w/cm3, would be
gamma-ray energy, of which ‘é or 1’1'5 would be
absorbed in the moderator, depending on whether
the moderator is graphite or beryllium oxide. Since
the moderator would occupy Y% of the volume, the
power density in the graphite would be 40 w/cm?,
and, in the bery!lium oxide, 60 w/ecm3.

 

2H. W, Bertini et al., Basic Gammé-Ray Data for ART
Heat Deposition Calculanons, ORNL-2113 (Sept. 17,
1956).

3ANP Physical Properties Group, Physical Properties
Charts for Some Reactor Fuels, Coolants, and Miscel-
laneous Materials, 4th ed., ORNL CF-54-6-188 (June 21,
1954).

at
- e i e e

M e gk e AL et

 

 

 

n

would shatter,”

The heating as a result of neutron moderafion
should, of course, be added to the gamma-ray
heating. However, the neutron heating in the core
would depend strongly on the design of the reactor,
and, therefore, for the purpose of this discussion,
it has been assumed. that it is compensated by
the loss of gamma rays from the core. _

The permissible stress for graphite has been

assumed to -be 2000 psi, and, for beryllium oxide,4
1000 psi.

The stress data® and the power density
data given above were used to obtain the following
maximum dimensions for berylllum oxide and

graphite bodtes in the proposed mulhtubular core:

Flat Plates, Rods,
‘Thickness ~ Diameter 7
{in) {in.)
Beryilium oxide 0.16 | 0.26_ '
Graphite - 5.6 2.1

As may be seen, if beryllium oxide were used,
the structures in the core would have to be rather

. delicate. - Graphlte, on the other hand, would allow
rather rugged construction, in spite -of nuclear

considerations, such as self-shielding, which
would require o somewhat finer structure than that
indicated by the stress limitations. -Furthermore,
the cladding that would be required around be-

ryllium oxide bodies would add appreciably to the
~ poisoning of the core, whereas the cladding around
graphite, with its small surface-area-to-volume
" ratio, would odd negligible poisoning.
With the advantages of graphite in mind, ‘a three- -
group reactor calculation based on the mulmubular:
. core was performed on the Oracle. Accordmg to
~ this. CG|CU|Gfl°n.,G 24-in. -dia .'.phere Cm*fllfllflg' :‘fluorade-fuel reactor with graphite as a moderator
- 80 vol %, -of the fuel mixture - (No, 30)- NaF-ZrF .
: _"'UF (50-46-4 mole %) ‘and 20 vol % of: graphne,?'
' "W";'would be somewhat supetcritical.
~lation poison was consndered to be ‘present in the
_'_form of a motertal havmg a macroscoplc obsorpflon

 

Af substanha"y higher sh’esses, bery!llum cxide -
. The consequences of such shattering. -
- are, however, not quite clear and could be determined

Concelvably such experiments "-‘of fast neutrons, whlch ‘on one hand mlght flnd

only by experiments, -

'-could make berylHum oxide appear more am'acfive fhun

it is now considered to be,

Sk, A Field, Temperature Gradient and Thermal

Stresses in Heat Generatmg Bodies, ORNL, CF-54-5-196
(May 21, 1954).

assumed in the Oracle calculation.

In this calcu- " outer part of the reflector.

 

PERIOD ENDING SEPTEMBER 10, 1956

cross section of 10% of the macroscopic absorption
cross section of the fuel, both absorption cross
sections being averaged over the core. This
amount of poisoning is far in excess of that which
could be caused by the cladding of the graphite.
The results of the calculation indicated that there
would be no need from the criticality viewpoint
to use beryllium oxide in the core and that graphite
would be an acceptable moderator. Since the
reactor power and the permissible power density
prescribe a relatively large core volume, the

~ somewhat poorer moderating properties of the

graphite appear to be adequate.

~An_essentially infinite graphite reflector was
Because of
the relatively poor moderating properties of the

‘graphite, such a reflector would have to be rather

thick. Also, graphite has a relatively small macro-

‘scopic removal cross section for fast neutrons,

that is, 0.073 cm=1 compared with 0.14 cm=! for
beryllium oxide, 0.17 cm=1! for nicke! and copper,

~ and 0.096 cm~ for zirconium metal.® The small

removal cross section would result in large neutron
leakage and intense sodium activation in the heat
exchanger. On the other hand, the power density
in the reflector would be smaller than that in the
core, and there would be no need to use graphite.
An *‘essentially infinite'® beryllium oxide reflector
would be thinner than an ‘‘essentially infinite"’
graphite reflector, because of the lower age in
beryllium oxide, and in any geometry, except plane
geometry, the beryllium oxide reflector would give
greater reactivity because of the larger solid angle

~which the core would subtend, as viewed from the
“location of @ neutron ‘which had slowed down to
thermal or any other given Iethargy.

‘Thus, there emerges a picture of a circulating-

on the msnde, ‘where' the heat generation is large,
and beryllitm oxide as @ moderator in ot least the
‘The “design of the
interface between -the beryllium - ox;de and the

" graphite remains to be established.

“Even at the outsnde of the reflector fl'lere would

' ;be a certain amount of heof genercmon, and, hence,
- fuel passages would “be required for ‘cooling.
- Fissions in these passages would act as sources

 

‘ 6Ct:il'npufecl from data given by G. T, Chapman and
C. L. Sterrs, Effective Neutron Removal Cross Sections
for Shielding, ORNL -1843, p 22 ond 26 (Aug. 31, 1955).

83

 
 

 

 

 

 

their way back to the core and increase the re-

activity, but which, on the other hand, would also

leak out and increase the shielding requirements
and the sodium activation. Therefore it would
probably be advantageous to poison some of the
outer fuel passages, but the radius at which the
poisoning should start also has not yet been
- established. -

Both graphite and beryllium oxide are poor
gamma-ray shields. 1t may therefore be.advan-
tageous to introduce some heavy material into the
moderator, at the expense of reducing the moder-
ating effectiveness. The heavy material would
protect the outer layer from gamma-ray heating and

would make it possible to use beryllium oxide
rather than graphite farther into the core and thus

gain back the lost moderation.

 HYDRIDE-MODERATED CIRCULATING-FUEL
' "REACTOR '

A. M, Perry

An alternative design now being studied includes
the use of a somewhat more effective moderator
in the core, such as the hydrides of zirconium or
ytirium, than the graphite and beryllium oxide
moderators discussed above. The hydride moder-
ators would reduce the dependence of the reactor

on its reflector for moderation and would introduce
the possibility of using a heavy material, such
as nickel or .copper, for the reflector. These
materials have unusually large removal cross

sections for fast neutrons, as given above, and

are far more effective as gamma-ray absorbers
than the moderating materials usually used as
reflectors. Thus a significant portion of the
gamma-ray shield might be placed immediately
around the core, where its volume would be greatly

reduced. While the density of radiation heating’
in a metal reflector would be greater than in a

beryllium oxide reflector, the allowable stress
would be much greater also, so that fewer cooling
holes would be required for a nickel than for a
beryllium oxide reflector, In the case of copper,
the conductivity is so high that cooling might be
required only at the inner and outer surfaces of
an 8-in.-thick reflector.

Preliminary calculations indicate that a few

thousand pounds in shield weight might be saved

by using a dense metal reflector. The effect of
capture gamma rays in the reflector, as well as
the activation of NaK in the heat exchanger,
remain to be determined. Further calculations of
fuel concentrations in reactors with metal re-
flectors are also required.
 

sl e reim e .

 

o m—————

Part 2 -
* CHEMISTRY

w.‘ R. Grimes

 
 

 

 

 
 

 

* »

2.1. PHASE EQUILIBRIUM STUDIES!

C. J. Barton

R. E. Moore

R. E. Thoma

H. Insley, Consultant

Phase equilibrium studies with a variety of
binary, ternary, and quaternary systems have been
continved. Each of the several methods described
in previous reports of this series has been applied
to several of the systems, Although earlier studies
in this laboratory and elsewhere did not disclose
its existence, careful examination of the LiF-RbF
system has revealed the compound LiF:RbF.
Careful re-examination of the LiF-UF , system has
yielded evidence which strongly indicates that the
compound 3LiF:UF, is metastable.

Study of the RbF-UF , system has progressed to
a point such that o tentative diagram for this
complex system can be presented. The boundary
curves, compatibility triangles, peritectic temper-
atures, and eutectic temperatures for the NaF-RbF-
UF, ternary system are now well established.
The four-component system NaF-RbF-ZrF +UF,
has been examined in some detail. Several mixtures
which are of possible interest for reactor fuels
have been established.

Study of the NaF-LiF-BeF , system has progressed
sufficiently for a plot of the boundary curves,
compatibility triangles, and pertinent temperatures
to be presented, although the composition of one
of the ternary compounds is not completely es-
tablished.  The ternary system NaF-RbF-BeF,
requires additional study before the analogous
diagram can be constructed.,

THE SYSTEM LiF-RbF
L. M. Bratcher

A recently published? equilibrium diagram for the
system LiF-CsF showed the existence of an
incongruently melting compound LiF.CsF that
was analogous to the binary compound previously
observed3 on petrographic examination of mixtures
containing LiF and RbF. Earlier thermal analysis

 

1 The petrographic examinations reported here were
performed - by G. D. White, Metallurgy Division, and
T. No McVay and H. insley, Consultants. The xway
examinations were performed by R. E. Thoma and
B. A. Soderberg, Chemlistry Division, :

2|, M. Bratcher, ANP Quar. Prog. Rep. June 10, 1956,
0RNL'2]06; P 90, Figo 2:147,

3L. M. Bratcher et als, ANP Quar. Prog, Rep. June 10,
1954, ORNL-1729, p 44.

studies43 failed to reveal the existence of the
LiF-RbF compound. The system was re-examined
recently by the thermal analysis technique, and
selected compositions were examined petro-
graphically and by x-ray diffraction. A revised
equilibrium diagram for the system, based on the
results of these examinations, is shown in

‘Fig. 2.1.1. The shape of the liquidus curves gives

evidence of complex formation. However, the
difference between the melting point of the LiF-
LiF.RbF eutectic and the incongruent melting point
of the LiF-RbF compound (about 5°C) is less than
the uncertainty of the temperature measurements.
Petrographic examination of a mixture containing
50 mole % LiF showed that the birefringent
complex was the predominant phase present;
excess LiF and the complex were found in a
mixture containing 70 mole % LiF.

THE SYSTEM LiF-UF,

B. A. Soderberg R. E. Moore
H. Davis®

The discussion of the system LiF-UF which
was given previously,”’ mentioned a phase con-
taining 20 to 25 mole % UF, whose formula and
stability characteristics were unknown., Recently,
thermal gradient quenching studies were made in
an attempt to completely characterize this com-
pound,

The evidence from these and other quenched
samples from this system indicates that there is a
metastable compound in the LiF-UF, system
whose formula is 3LiF-UF,. In the quenched
samples the compound has the appearance of
quench growth. The crystals are sufficiently well
formed to give a good uniaxial-negative inter-
ference figure and a refractive index of approxi-
mately 1.514, but they have the characteristic
mottled texture and wavy extinction of quench

 

4J. P. Blakely, L. M. Bratcher, and Cs J+ Barton,

ANP Quar. Prog. Rep. Dec. 10, 1951, ORNL-1170, p 85.

5E. P, Dergunov, Doklady Akad. Nauk 5.5.5.R. 58,
1369 (1947)« '

50n assignment from Pratt & Whitney Aircraft,

7R, E. Moore, ANP Quar. Prog. Rep. Dec. 10, 1955,
ORNL-2012, p 77,

87

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

SOMRB Nt
ORNL-LR-DWG 16145

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

S00
4
— -—
800 T
) ‘\ . . ‘f(—
\ -~
T0C
§ i
w /
a
2 600 - h
<
o . .
Wl
o
- - .
500 , N /
_—“-____’_' . s L ———-hé-—u —————
400
' s
[
L
-
300
ROF 10 20 30 40 50 60 70 80 90 LiF
LiF (mole %) ‘ '
Fig. 2.1.1. The Sysfem LiF-RbF.
growths. X-ray diffraction patterns of these  Since the quenched samples of 25 mole % UF ,-75

samples ore distinct and unique in the LiF-UF
" system.

Large quenched samples (~10 mg) contained
3LiF-UF 4 95 quench growth.  Medium-sized
(~5mg)or very small samples (2 to 3 mg) quenched
in platinum tubes allowed the liquid to freeze to
material with a glassy appearance and isotropic
optical character, but even in these samples x-ray
analysis showed the glassy material to be largely
3LiF.UF «+ There is no observable difference,
optically or by x-ray pattern, in the 3LiF.UF 4
which occurs below the liquidus and that from
quenches which were attempted well above the
accepted liquidus. Under no conditions has
3LiF:UF, been observed as homogeneous, well-
crystallized moterial free of quench growth texture.
The inconsistency of the appearance, petro-
graphically, of the 3LiF.UF , which occurs as a
glass (isotropic, no optic figures) and that which
occurs as quench growth, with texture gradation,
is evidence of the metastability of the compound.

88

mole % LiF are more consistently single-phase
than those of any other composition, the formula
of the compound is considered to be 3LiF-UF‘.

THE SYSTEM RbF-UF4

H. A. Friedman R. E. Moore

Quenching studies on the system RbF-UF, are
not complete, but considerable progress has
made. A tentative phase diagram, based on thermal
analysis studies and petrographic and x-ray dif-
fraction examinations of quenched and slowly
cooled samples, is presented in Fig. 2.1.2. The
formulas of RbF-UF4,' 2RbF-3UF‘, RbF-3UF4,
and RbF.6UF, were established on the basis of
nearly single-phase material that was found in
quenched samples of the corresponding compo-
sitions which had been equilibrated just below
the solidus temperatures. The recently obtained
thermal analysis data are given in Table 2.1.1,
and the results of quenching studies are presented

in Table 2.1.2.

een
 

 

600

TEMPERATURE (°C)

PERIOD ENDING SEPTEMBER 10, 1956

fecRes
ORNL—LR-~OWG 16146

 

1100

]

 

1000

g

 

200

 

 

 

800

 

 

 

 

 

700

 

 

 

500

 

 

3RbF - UF,

 

 

 

 

 

 

2RbF- UF,
7RbF-BUF,4
!

 

. RbF - UF4
2RbF -3UF,

 

 

RbF - 3UF,

 

 

 

RbF- 6UF,

 

 

 

 

400

RbF . 10

20

TABLE 2.1.1. THERMAL ANALYSIS DATA FOR RE:I_'-'-_IV.IF4 MIXTURES

. 30 . 40

50 60

UF, {mole 7o)

Figs 2,1.2, The System RbF-UF‘l.

70 80

90 UF,

 

CLos0

UF,
. Ceontent L

oo (mole %) -

Casa

s

osg

- 57 R

| Liguidus
'{ém_berhf“’e. - o '

ceetey

90

g

. Solidus

Tempérgturp- e

e[

cess |

~UF,

60

.75
80

 

-»-;_1.~C6r_lfent :
. /:':_‘r,‘(m'o'le' %y -

625
66.7 .

. :85;8': :;-.; -

, Liqui&us

-+ . Temperature

o) =

740
740
802

825

937

Solidus
Temperature
€O
/1

745 -
-743

o

 
 

 

 

ANP PROJECT PROGRESS REPORT

TABLE 2.1.2. RESULTS OF QUENCHING STUDIES OF RbF-UF, MIXTURES

 

 

| Ct:an:nf TeLl::’:iil::fe Primary TZ':.:::;:, Phases™ Below T:;::m | #h“‘e"_ Below
(mole %) ‘ (°c) Phase* (°C) Teonsition °C) Solidus
333 o <572 2RbF-UF, >745  2RbF-UF, |
46.2 712 RbF-UF, - | | 683 7RbF-6UF4,'R'bF‘-UF4_ |
50 727 ..RbF-UFa 727 RbF-UF . B
51 736 RbF-UF, 706 RbF-UF,, [u.m.]
53 rp3 RbF-UF, 709 RbF-UF 4, [usm.]
53 725 RbF-UF, 710 RbF-UF ,, 2RbF-3UF
53 721 RbF-UF 715 RbF-UF ,, 2RbF-3UF
57 737 RbF-6UF, 724 2RbF-3UF ,, Q.G. 714 2RbF-UF ,, RbF-UF
- 60 2772 (RbF-6UF )
60 >770 (RbF+6UF ) 731 RbF-3UF , Q.G. 720 2RbF3UF ,, [ vem.]
62.5 723 2RbF-3UF,, RbF-3UF,
[Q.G.)
6647 >752 (RbF-6UF ) 733 RbF:3UF , Q.G. 709 RbF3UF , 2RbF-3UF4
6647 >757 (RbF6UF ) 725 RbF:3UF,, Q.G. 717 RbF-3UF ,, 2RbF-3UF
66:7 >758 (RbF-6UF ) 745 RbF:3UF ,, Q:G. 714 2RbF-3UF ;, RbF-3UF
75 >755 (RbF-6UF ) 714 RbF-3UF
75 >753 (RbF-6UF ) 721 RbF3UF
80 >760 (RbF+6UF ) 724 RbF-6UF ,, RbF-3UF
85.8 >824 RbF-6UF

 

*The phase present in greatest quantity is given first,

Phases found in very small or trace amounts are in brackets,

Where a phase is enclosed in parentheses it is the phase found below the indicated temperature, but it is not neces~
sarily the primary phase. Quench growth is indicated by Q-G., vnidentified material by vem.

THE SYSTEM NoF-RbF-UF,

B. A. Soderberg
V. Frechette$

H. A. Friedman
H. Davis$

An intensive study of the phase-equilibrium
characteristics of the system NaF-RbF-UF  has
been made. Equilibrium dota were derived from
petrographic and x-ray diffraction examination of

quenched ond slowly cooled melts. The boundary

| curves, compatibility triangles, and the peritectic
5 and eutectic temperatures characteristic of the

90

system are shown in Figs. 2.1.3 and 2.1 4. It
is noteworthy thot there is extensive solid so-
lution along the join 7NaF. 6UF4-7RbF-6UF
The extent of the miscibility gap is not yet de-
termined. This solid solution effecr is general
among the compounds 7MF-6UF ,, where MF is an
alkali fluoride. There is one ternary compound in
the system, with the formula NaFRbF-. UF ,. The
compound has a subsolidus existence at its compo-
sition below 530°C. The system contains the
following five ternary eutectics:
 

 

 

-~ UF,,

- Composition (mole %) 7 ' Melting Point

 

) NaoF RbF UF, (°c)
25 25 50 630
26 27 47 .. 620
. 33 30 a7 535
47 33 20 | 470

wooB 9 e

o o
ORNL-LR-DWG 16147
- UFg |

      
    
 

{wwe-SOLID SOLUTION)
NaF - 2UF,

e e i i o

-
—fi__‘
-

- BNaF -3UF,
2NoF - UF,,

i .t
——

——
-"--.__
-

-
-
P
- -
-

-
"
-
-
———

-
-
-
-
-

 

Fig; 2.1.3. B'o'ur:llaary’ Curves and Compatibility
Triangles in the System NaF-RbF-UF".

PRI
* ORANL-LR-DWS 16148
UF,

=\ RbF - 6UF,

 

 

and Eutectic Temperature in the System NaF-RbF-

‘Fig. 2.1.4. Primary Phase Fields and Peritectic

PERIOD ENDING SEPTEMBER 10, 1956

The * boundary curve between the 2RbF-UF,
and NaF-.RbF.UF, primary phase fields has its
highest liquidus temperature at the intersection
of the boundary curve and the extension of
the 7RbF-6'UF“-?\lc:F-RbF-UF4 Alkemade line.
(“‘Alkemade Line: In a ternary phase diagram a
straight line connecting the composition points of
two primary phases whose areas are adjacent and
the intersection of which forms a boundary
curve, "B)

- The subsohdus compound NaF. UF has no
primary phase field in the equilibrium diagram,
At liquidus temperatures. which will permit an

NaF.2UF  primary phase field, RbF-UF , compounds

. are the primary phases.

THE SYSTEM NdF-RbF-ZrF4-UF4

H. A. Friedman B. A. Soderberg
I H. Davis

The system NaF-RbF-ZrF +UF, offers several
mixtures which may be sohsfcctory fuels for a
cuculohng-fuel reactor.? Quenching studies of
this system were continued in order to establish
the primary phase fields of the uranium-containing
compounds.  Primary phase determinations were
also made on slowly cooled melts. These studies
provide a basis for indicating the potential fuel
mixtures from which the high-uranium-content
compounds might segregate.

The results shown in Figs. 2.1.5 and 2.1.6 are
for the ternary compositions NaF-RbF-ZrF, re-
calculated from the quaternary data. Where both
the primary and secondary phases contain uranium,

‘the secondary phase is shown as the second
~ listing,

THE SYSTEM RbF-CuF
L. M Bratcher :

Apphcotlon of fhermcl anolysns to mlxtures of

R ff_"RbF and Cc:F2 has “consistently failed to yield

"7 reproducible data for compositions “containing more
than 15 mole % CaF ..

. than ‘normal - cooling rates nor the subsmuhon of
Covery thin thermocouple ‘shields served to improve
. the _data ,obtomed. ,

Nelther the use of slower

The fmlure of. fhermocouple

 

8E. Me Levin, He F- McMurd;e, and’ F. P. Hall Phase
Diagrams ~ for Ceramists, American Ceramic Socaefy,
Columbus, Ohio, 1956,

;9H- A. Friedman and H. Davis, ANP Quar. Prog.
Rep. June 10, 1956, ORNL-2106, p 86.

N

 
 

 

ANP PROJECT PROGRESS REPORT

2rF, T ORNL-LR-DWS 18149

   
 
 
   
 

A= TNaF - 6Zr{U)F,
6= RbF-3UF,
€ = RbF-UF,
D« SRYF- 42¢ (UIF,
£ = RbF - 6UF,
F=7RLF-6UF, - 75
LIQUIOUS TEMPERATURES 70
PLOTTED IN°C

 

 

 

 

 

mp7 A.‘.b.‘

Fig. 2.1.5. Primary end Secondary Uranium-
Containing Phases in the System NaF-RbF-Z¢F -
UF, wnh 4 mole % UF,.

shlelds after a few experiments suggests that the
mixture is extremely corrosive; this may be due to
the presence of some sulfate in the CaF .. '
Examinations of slowly cooled melts by x-ray
diffraction and by the petrographic microscope
reveal a compound with a cubic structure, which is
presumably RbF.CaF, (ref 10). Since free CaF,
has been observed in mixtures containing 40 to
45 mole % CaF,, it seems likely that RbF -CoF,

melts incongruently.

THE ALKALI FLUORIDE-CEROUS
FLUORIDE SYSTEMS

L. M. Bratcher

Examinations of the systems LiF-CeF,, NaF-
CeF,, and RbF-CeF , were continued; however, the
data do not yet perrmt construction of satisfactory
equilibrium diagrams for these systems. Com-
parisons of data obtained with CeF; prepared at

ORNL with dats obtained - with commercially

prepared CeF , strongly suggest that the commercial
material contums a small amount of water. It

appears that treatment of the MF-CeF , mixture .

with NH HF, will suffice to eliminate difficulty
from this source. The NaF-CeF, system has been
shown to have a binary eutectic, which melts at
775 £ 10°C, between CeF, and a binary compound
believed to have the composition. NaF-CeF_s.

 

10y, L. W. Ludekens and A J. E. Welch, Acta Cryst.
5, 841 (1952).

92

A'w TNGF: 6Zr () F

" 8 = RbF-3UF, ‘ Zrhy
€ = RbF - UFy

| O =5ROF-4Zr{UIF,
E = AbF-6UFy
LIQUIDUS . TEMPERATURES
PLOTTED IN °C

  
  
 
 

 

Fig. 2.1.6. Primary and Secondary Uranium-
Containing Phases in the System NaF-RbF-ZrF,.
UF, with 7 mole % UF,.

THE SYSTEM'CsF-Bch
L. M. Bratcher

Study of the system CsF-BeF, was started .

recently in order to complete the study of the
alkali fluoride—beryllium fluoride systems and
to obtain additional data on the effect of varying
ion size on phase relationships. Thermal analysis
data have been obtained with mixtures at § mole %
intervals in the range 5 to 50 mole % BeF,, but
study of the slowly cooled melts is mcornpfefe.
There is a eutectic, presumably between CsF and
3CstBeF2, which melts at 605 + 5°C and con-
tains about 14 mole % Ber. Melting point re-
lationships of the compound 3CsF:BeF ., have not
been definitely established. The compound
2CsF-BeF , undoubtedly melts congruently, proba-

bly at a femperoture befow that of the analogous
compound 2RbF. BeF (800°C). As in the other
“alkali fluorlde-berylhum fluoride systems, liquidus

temperatures diminish rapidly in this system with
increasing BeF, content in the 33',5 to 50 mole %
BeF, region. The solidus temperature for mixtures
containing 35 to 45 mole % BeF, appears to be

about 450°C, which is very close to the repc;rfet:!I P

mcongruent meltmg point of RbF.BeF,,.

 

HL. M. Bratcher, R« E. Meadows, and Re Jo Sheil,
ANP Quar. Prog. Rep. March 10, 1956, ORNL+2061, p 74.

ORNL-LR-DWG 16180

RLF
 

 

THE SYSTEM NaF-LiF-BeF,

R. E. Meadows

A dmgrom of the half of the system NaF-LiF-
BeF2 represented by the traungle NaF-LiF-
NaF.BeF_ was published recently.12 The quench-
ing study of the LiF-NaF.BeF,-BeF, triangle is
now sufficiently complete to permtt construchon of
a tentative diagram for the entire system, as shown
in Fig. 2.1.7. Binary eutectics, ternary eutectics,
and -peritectic points are indicated in the diagram,

along with the corresponding temperatures. Sub-

solidus compounds are shown in parentheses. The
temperatures following all the compound formulas
are the liquidus temperatures for congruently

'melhng compounds or the upper sfob|||ty limits

for other compounds. :

Difficulties in the petrographic identification of
the phases in the system, as mentioned previously,
are also present in the work on the LiF-NaF.BeF .-
BeF, portion of the system. Consequently it has
been necessary, in some cases, to prepare and
examine petrographically and by x-ray diffraction o
great many samples in order to obtain a single
result, _

Examination of slowly cooled samples, as well
as quenched ternary samples, shows that a ternary
compound containing more than 50 mole % BeF
exists in this system. Its identity has not been
completely established, but it is indicated on the

 

12n. E. Meadows, ANP Quar. Prog. Rep. June 10,
1956, ORNL'2]06J p 88, Fig' 2616

TS
ORNL-LA-0WS 16451
£= EUTECTIC ’ S
T= TERNARY EUTECTIC Befy
TC= TEANARY COMPOUND ) oo Bez
P= PERITECTIC : :

  
  
  

SUBSOLIDUS COMPOUNDS SHOWN
IN PARENTHESES .
LIQUIDUS TEMPERATURES ARE IN °C

 

LiF i
844 _ £649

Fig. 2.1.7.. The System NaF-LiF-BeF,,

PERIOD ENDING SEPTEMBER 10, 1956

diagram by the symbol ““TC"" at 60 mole % BeF ,-20
mole % LiF-20 mole % NaF. The possibility that
it may be an unreported binary compound has been
eliminated by x-ray diffraction examination of
equilibrated and quenched samples containing
66.7 mole % BeF,-33.3 mole % NaF and 66.7
mole % BeF2—33 3 mole % LiF.

The join 2LiF.-BeF o"2NaF.LiF.2BeF, was
established as a quasu-bmary system f:y the
examination of equilibrated and quenched samples
at 38 mole % BeF2—35 mole % LiF-27 mole %
NaF. The liquidus is 334°C and the primary
phase is 2LiF.BeF,. The phasesbelow the solidus
(315°C) are 2LiF- BeF ‘and 2NaF.LiF. 2BeF2 An

extrapolation places t 2he binary eutectic at 38.5

‘mole % BeF 2~30.5 mole % LiF-31 mole % NaF.
The join is not a true binary system because the
primary phase field of LiF overhangs the compound
2LiF-BeF, very slightly.13  The triangle LiF-
2LiF.BeF -2NaF «L.iF:2BeF, is a compatibility
triangle; no phases other than the three mentioned
have been found in samples within the triangle.

. Examinations of quenches of a composition con-

taining 50 mole % BeF2-38 mole % LiF-12 mole %
NaF, which is on the join 2LiF.BeF,-TC, show
that the liquidus is above 354°C ond that the
primary phase is 2LiF.BeF The solidus is
280°C, and the secondary ilase is the ternary
compound TC. The absence of a third phase
indicates that the composition is on or near the
binary system 2LiF.BeF -TC, At the composition
50 mole % BeF ,~12 mofe % LiF~38 mole % NaF
the liquidus is 344°C, and the primary phase is
NaF.BeF,. Below the solidus (280°C) the three
phases NaF-Ber,'ZLiF-Ber, and the ternary
phase TC coexist. At the composition 50 mole %
BeF ,~25 mole % LiF-25 mole % NaF the liquidus-
was found to be 276°C, with 2LiF. BeF, as the
primary phase.” The same three phases, NaF BeF2
2LiF.BeF,, and the ternary phase TC, were found
to be present below the solidus (271°C). The
results of these experiments show that _2LiF-_BeF2,'
NaF.BeF,, and the ternary phase TC form a com-
patibility triangle. However, the join 2LiF-BeF2-

" NaF.BeF is not a true binary system, even if the
~slight overhang of 2LiF.BeF, by the primary

phase field of LiF is disregarded. Quenching

 experiments with the composition 45 mole %

BeF2-20,mole % LiF~35 mole % NaF, which is on
the join, show that, down fo 300°C, 2LiF-BeF,

 

]3J- L+ Speirs, Thesis, Michigan State College, 1952,

93

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

and 2NaoF.LiF 2BeF, are present.  Thus the
primary  phase fueld of 2NaF-.LiF 2Bef, cuts
across the join, There must be a pentecflc point
below 300°C in the triangle 2LiF. Ber-NaF Ber-
TC, where 2LiF. BeF,, 2NaF.LiF. 2Be|:2 NaF.BeF,
are in equilibrium, Locahons of the phase bounda-
ries, the peritectic point, and the ternary eutectic
were deduced from the evidence given above and

_are considered to be tentative.

THE SYSTEM No F-RbF-Ber

L. M. Bratcher

Study of the ternory system NaF-RbF- BeF, by
thermal analysis supplemented by petrographic and
x-ray examination of the slowly cooled melts was
continued. Cooling curves obtained with compo-
sitions within the triongle NaF-RbF-3RbF. BeF,,
prevuously reported to comprise a compahblllty
trlangle, indicated the existence of a ternary
eutectic having a melting point of approximately
625°C. Additional data will be required in order
to determine the eutectic composition, but the
available data show that it will be close to the
NaF-RbF eutectic (74 mole % RbF, melting point
675°C). An unidentified crystalline phase, believed
to be a ternary compound, has been observed in a
number of siowly cooled melts containing 45 to
60 mole % BeF,; this phase is usually associated
with an isotropic phase believed to be glass. The
cooling curves for the mixture NaF-RbF-BeF
(35-15-50 mole %)} showed only one thermal effect

‘at about 240°C, and the cooled melt was found to

be predominantly glass. The occurrence of glass
in ternary mixtures having BeF, concentrations
that produce only crystalline compounds in the
corresponding binary alkali fluoride—beryllium
fluoride systems was noted previously in the
NaF-KF-BeF, system.'5  This interesting phe-
nomenon may be due either to the difference in
ion sizes of the alkali ions involved or to the
lowering of solidus temperatures in the ternary
system to the point where the high viscosity of the
melt inhibits crystallization. Recent petrographic
and x-ray diffraction studies of slowly cooled
melts have confirmed the existence of the ternary
compound 2NaF-RbF-_BeF2; it also appears that

 

"L. M. Bratcher, ANP Quar. Prog. Rep. June 10,
1956, ORNL-2106, p 89

5., M. Brotcher et al, ANP Quar, Prog. Rep.
Dec. 10, 1955, ORNL-2012, p 84.

94

the compounds 2RbF. BeF,, 2NaF-RbF BeF o and
NaF.BeF, comprise a compahbnhty trmngle.
The ~Nof.BeF -2NaF-RbF-BeF, side of this
triangle passes through the minimum melting-point
mixture (480 5°C) on the 2NaF BeF ,-2RbF.BeF,

|°|no

THE SYSTEM ‘KC'I-ZrCI‘
R. J. Sheil

Interest in zirconium chloride—~alkali chloride
systems stems from several sources. Low-melting-
point chloride systems are potentially useful
as heat transfer media. It is possible that fused-
salt systems containing Z:Cl, will be of interest
in the ZrCl, volatility method proposed for
processing o? metallic zirconium-uranium fuel
elements.'® [n addition it is of interest to com-
pare phase equilibrium behavior in these MCl-ZrCl
systems with behavior in the analogous MF-ZrF
systems, which have been thoroughly siudied at
ORNL.

The alkali chloride—zirconium chloride systems
have received relatively little attention. A
Russian investigation of the NaCl-ZrCl, system is
on record.'? Preliminary data obtained at ORNL
indicate serious errors in the published data; the
study being made here is not, however, sufficiently
complete to permit a diagram to be drawn.

Some data on the KCI-ZrCl, and NaCl- ZCl,
systems in the ZrC|4-r|ch parts of the systems
were obtained by investigators at Columbia Uni-
versity.18 The present investigation of the KCl-
ZrCl, system covers the composition range 10 to
75 mole % ZrCl4. The data obtained are in
reasonably good agreement with the Columbia data
in the limited area where the two investigations
overlap. The thermal analysis data obtained
from cooling curves with mixtures contained in
sealed glass, quartz, or nickel tubes fitted with
thermocouple wells ore shown in Fig. 2.1.8. In
addition to the use of the thermal analysis tech-
nique, visual observation and quenching techniques
were applied to the study of several mixtures in the

 

V6Chem. Tech., Semiann, Prog, Rep. March 31, 1956,

17N, A, Bolozersky and O. A, Kucherenko, Zbur,
Prikad. Khim, 13, 1551 (1940).

184, H. Kellogg, L. J. Howell, and R, C. Sommer,
NaCl-ZrC|4, KCl-CrC|4, and NaCI-KCl-ZrCl4. Summary
Report, NYO-3108 (April 7, 1955).

"
 

e et et —

 

e o

PERIOD ENDING SEPTEMBER 10, 1956

UNCLASSIFIED
CRNL-LR-DWG 16152

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

800
- r .
600 - - 3
—-——‘
5 < \
s 5
500
w =
o S |
E el
<
i 3 ———
& 400 gL \ =
= = /"
Lt =
- @ W I
Q T
S ? | /
300 = ~
g ©
o —
O
G |
J ~ - i
200 -.”— '-—.-.- v o' —V — - ——-.——-}——-—- _____
100 '
KCI 10 20 30 40 50 60 70 80 20 2rCl,

ZrGly (mole %)

Fig. 2.1.8. Tentative Diagram of the System KCI-Z+Cl .

system, and most of the slowly cooled melts used
for thermal analysis were examined petrographi-
cally. As shown by Fig. 2.1.8, there is a eutectic
between KC! and the compound 2KCl.Z:Cl & which
melts congruently ot 790 + 5°C. The eutechc
melts at 600 1 5°C and contains about 23 mole %
ZrCl,. There is also another compound, tentatively
ass:gned the formula 7KCl.6ZrCl 4+ Which appears
to melt incongruently to give 2KCI.ZrCl . and
liquid at 565 + 5°C. The choice of the compound
formula was based partly upon thermal analysis
dota, which indicate thot the compound contains
less than 50 mole % ZrCl 4+ and partly upon the

“results of petrographic excmmahon of slowly

cooled and quenched samples in this region.
More careful study, particularly of quenched
samples, will be required to eliminate uncertainty
in regard to the compound composition. Thermal
effects on cooling curves at temperatures ranging
from about 218 to 225°C were noted with compo-
sitions containing more than 33'6 mole % ZrCl,.

Since this is reasonably close to the reportedi8
“melting point of the eutectic at 65 mole % ZrCl
(234.5°C), the appearance of the thermal effect
ot a temperature close to this value with compo-
sitions containing less than 50 mole % ZrCl
‘might be taken as an indication of incomplete
reaction of liquid with solid 2KC|-ZrC|4 at the
peritectic point. However, the magnitude of the
thermal effect on cooling curves, coupled with
the fact that no liquid was visible in mixtures
containing less than 50 mole % ZrCl, ot temper-
atures below 250°C, indicates the likelihood that
a solid phase transition occurs at a temperature so
close to the solidus temperature that the two
effects cannot be differentiated either by thermal
analysis or differential thermal anclysis. Petro-
graphic examination of a quenched sample con-
taining 55 mole % ZrCl, showed that, above 237°C,
well-crystallized phoses were present but that at
219°C and lower temperatures the samples con-
tained only microcrystalline aggregates.

95

 
 

 

 

ANP PROJECT PROGRESS REPORT

2.2, CHEMICAL REACTIONS IN MOLTEN SALTS

F. F. Blankenship
R. F. Newton

EQUILIBRIUM REDUCTION OF NIF, BY H,
IN NaF-ZrF,

C. M. Blood
Investigation of the equilibrium
~ NiF,(d) + H,(g) == Ni(s) + 2HF(g)
with NaF-ZrF, (53-47 mole %) as the reaction

medium was continued. Since the system was
observed to behave rather peculiarly at high
temperatures, this study was interrupted while
reloted investigations of the stability of NiF,

L. G. Overholser
G. M. Watson

in this solvent were conducted. These experiments
have served to indicate the essential validity -of
the equilibration data previously obtained! at
625°C and of the data obtained at 500 and 575°C,
as reported below. The experimental data obtained
by using sintered-copper filters and the general
technique described previously are shown in
Table 2.2.1. The mean of 18 values previously

. obtained for the equilibrium constant at 625°C

was 3.0 x 10=4 atmosphere. -

]C. M. Blood, ANP Quar. Prog. Rep. June 10, 1956,
ORNL-2106, p 100.

 

TABLE 2.2.1. EQUILIBRIUM RATIOS FOR THE REACTION
NiF,(d) + H,(g) = Ni(s) + 2HF(g)

 

 

Ni in Melt Pressure of H,* Pressure of HF K ** % 10-4
(ppm) (atm) {atm) x
Measurements Made at 575°C

315 0.0266 0.434 1432
345 0.0257 0.453 1.25
365 : 0.0248 0.472 1.45
385 0.0245 0.477 1.42
420 0.0243 0.482 1434
435 040241 0486 1.32
180 0.0301 0.358 1.38
165 0.0297 0.368 1.62
185 0.0293 04375 1.52
220 0.0290 ' 0.382 1.35
150 0.0309 0,342 ‘ 1.49

Av 1.40  0.09

Measurements Made at 550°C .

175 0.0318
205 0,0309
205 ~ 0.0306
200 | 040306

0.321 1.09
0.341 1.08
0.347 112

0.347 1.15

 

*Initial H2-H_e mixture was 4.69% H2.

"'*Kx = PE!F/XNIszfiz' where x is mole fraction and P is pressure in atmospheres,

96

+
 

 

o

o

-"obtamed for NlF

From the data obtained to date it appears that
the activity coefficient for NiF, must be about
2750 at 600°C, with NiF ,(s) taken as the standard
in this -
solvent was 3.28 (ref 2). The very large apparent
is considered to be

state. The correspondmg figure for Fer

activity coefficient for NiF,

. quite surprising.

' STABlLlTY OF NiF
J G. Eversole

lN NGF-ZrF MIXTURES
C M Bload

The very hlgh apparenf achvnty caeff:cnents
in molten NaF-ZrF, prompted |
" .a careful exammahon of the behavior of such
" solutions under a variety of conditions. "
- this ‘examination -has served to substantiate the’

While

validity of the equilibration data obtained, a number

- of rather peculiar effects have been observed. ‘
- An initial mveshgahon was performed to dls-.
close whether any yncerfainties in the chemical -
X analyses were obscuring the real picture. A total
of ten standard samples prepared by adding ,

weighed quantities of NiF, with or without added

'CuF to powdered NaF-Zer mixture were prepared
‘and submmed for analys:s. Analytical results for

these samples’ covered the range from 125 to

2000 ppm nickel and were, with a single exception, .
within 5% of the theoretical value; the single

exception was the most dilute sample, for which

‘the analytical result differed from the *‘true”
- value by 22%.

In addition the reproducibility of
the chemical analyses was tested by submitting,
in duplicate, finely ground samples of filtrates
from solutions of NiF,in NaF-ZrF ., The duplicate
samples were submmed under glfferent sample

" numbers, with one to two weeks being allowed

“between the time - the. sample and its duplicate

~were submitted. The data presented belaw show
i excel!ent agreement in every case' ' O

Nlckel Found (ppm)

Inltlal Sumple s Dupllcafe'Sampl_é s

Cess s

1040 1040

oems. o s
005 1015
1030 - -7 1025

40 40

 

ZC. Me Blood and G. M. Wotson, ANP Quar. Prog.

Rep. March 10, 1956, 0RNL-206'|, P 89.

50 hr of sparging with helium at 625°C.

~cant quantity of NiF,

PERIOD ENDING SEPTEMBER 10, 1956

40 35
40 35
260 260
1165 . ms

- 1070 1050
1030 1010
1030 1030

It seems clear that uncertainties in analyses for

-NiF, in the samples submitted is not a contributor
. to the surprising observations.

It has been observed thot NiF, dissolved in
NaF-ZrF, mixtures is stable in containers of
A-nicke! under continuous sparging with helium

- up to temperatures of 625°C. At temperatures

of 700°C ond above, .however, the concentration

- of dissolved nickel decreases continvously. The
. results of a typlcal experiment are presented in

Table 2.2.2.

concentration .

It may be observed .that the N|F2
remained - constant during about
During
equilibration experiments it is necessary to sparge
the system with helium for about 30 min when

~ samples are taken. In view of this stability under

long sparging, however, it appears that no signifi-

is lost during the short

treatment required in equilibration experiments.
The disappearance of NiF, from solution at the

‘higher temperatures is not well understood. The

presence of significant concentrations of reducing
impurities, such as hydrogen, methane, or oil
vapor on the helium, can apparently be eliminated
from consideration. The influent helium stream
was passed over a bed of copper oxide at 500°C,

- a magnesium perchlorate drying tower, and a liquid
‘nitrogen cold trap without affecting the experi-

~ mental results. - Moreover, the effluent helium was
‘carefully ‘monitored by . passage through KOH
solution. The amount of HF recovered was only a
few per cent of that recoverable fram the NiF

lost from solution. ‘And fmally, even though the

. NiF, is apparently lost from the soluhon, passage

of - fiydrogen gas through the system hberatas

'essenhally the stoichiometric quantity of HF,

_ Thermodynamlc_ considerations indicate that the

: ethbrwm pressure for decomposition -of NiF,
‘into Ni® and F,is neglxglbly small,

It is possible,
however, that a volatile subfluoride of nickel may
be responsible for this peculiar behavior. [f the
reaction

NiF ,(d) + Ni{s)== 2NiF(g)

97

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

TABLE 2.2.2. STABILITY OF NiF, IN NaF-ZrF ; (53-47 mole %)* UNDER
CONTINUOUS SPARGING WITH HELIUM ‘ -

 

Filter . - - ' Found in Filtrate (ppm)

 

 

Temperature Time at Temperature
_(°C) ' _ (hr) Material - - Ni  Fe " Cr
625 . 24 Ni IE 225 ' 25
48-50 NP 1050 255 25
Ni 1030 240 25
‘cw - 1000 225 - 30
Ni 1025 250 30
Cu. 015 245 .25
Ni 1040 - 250 20
NN . . 35 145 - - 20

80 . 65

 

*Initial charge: 1 kg of NaF-ZrF , containing 260 ppm Ni and 220 ppm Fe. Additive: 1000 ppm Ni as NiF,,

can occur, it may be possible to explain the
phenomenon observed. No concrete evidence for
existence of this compound has been found..

A large number of equilibration measurements
have been ‘performed by using copper filtering
tubes. During the course of the sampling procedure
the copper filter tube becomes plated with nickel.
It has, accordingly, been necessary to determine
whether the concentration of nickel in solution is
changed by the introduction of the copper tube.
The results listed in Table 2.2.2, for samples

taken in the interval of 48 to 50 hr indicate that

nickel plating of the copper tubes does not change
the nickel concentration in solution. In addition,
a copper tube was immersed in the solution for
24 hr lmmedlately following the 48- to 50-hr period,
and a massive nickel deposit, which contained
large dendritic nickel crystals, was obtained
(Fig. '2.2.1). The weight of the deposit was
determined to be approximately 15 g; the weight
of nickel in solution just before immersion of the
copper tube was only about 0.4 g. The nickel
concentration of the solution immediately after
the heavy nickel plating occurred was determined
to be 960 ppm, as compared with a value of 1030
ppm obtained by averaging the results for all the
'samples taken during the 48- to 50-hr period.

The formation of a nickel plate on the copper |

tubing could be explained by simple concentration
cell action; however, such a mechanism would

98

cease to function after a very thin film of nickel .

was laid down unless a copper-nickel alloy were
formed.  Careful examination of the massive
deposit- shown in Fig. 2.2.1 revealed essentially
no copper.on the outer surface.
force for this plating process were the temperature
difference between the nickel reactor wall and
copper tube, which was cooled by influent helium
gas, similar deposits of nickel should have ap-
peared on the nickel sampling tubes, but no such
deposits have ever been observed. In all these
experiments the copper tubing was inserted through
a stainless steel Swagelok fitting at the top of
the reactor, where the temperature was about
200°C.  Accordingly a low-temperature copper-
steel-nickel junction was in electrical contact in
series with a high-temperature copper-nickel
junction through the molten salt. It appears
likely that the resulting thermoelectric potential
may have been the driving force for the plating
process. Although experiments are planned to
supply explanations for some of these phenomena,
it can be concluded that the use of copper filters
introduces no significant error into the results
presented.

In an effort to obtain concrete evidence for the
formation of ‘a volatile nickel subfluoride, two
platinum crucibles were loaded with known weights
of nickel fluoride and three nickel crucibles were
charged with mixtuwes of metallic nickel powder

if the driving -

-
L A Al Sy

 

 

 

 

 

PERIOD ENDING SEPTEMBER 10, 1956

SXT UNCLASSIFIED
R PHOTC 26802

Fig.2.2.1. Massiie Nickel DepoSit with Lafge Dendritic Crystals on Copper Tube That Was Immersed
for 24 _hr in an NaF-Z¢F, (53-47 mole %) Solution Containing NiF,.

-und mckel fluornde. The five crucibles were.
enclosed ‘in a nickel ‘vessel fitted with ‘a copper
spiral coil to serve as a condenser. Cool helium -
gas was passed "through ‘the copper -coil cone
~f|nuous|y at high rates. A flowing" atmosphere -of
“helium gas was also: conhnuousiy passed over the
samples _as ‘the ‘temperature of the “system was -
brought to 800°C and held constant for a period of

3 hr. - The samples were allowed to cool and were

reweighed. The material condensed on the copper -

coil was curefu“y collected and " submitted for

_petrogrophlc -and . xeray : d:ffrachon ‘examination,
After having been weighed, the Ioaded crucibles -

were heated again in the same manner for. an
additional 19 hr. The results of the experlment
are summarized in Table 2.2.3.

At the end of 3 hr the N:F n- the platinum
crucibles was discolored - Ond a considerable
amount of magnetic material, later  identified as

Ni% was present. . No apprecioble change in

welght of the platinum occurred. Slight deposits

- on the copper. condenser tube were tentatively

| identified as NiF,, Ni, and CuF2

- The data may be mterpreted as. p0551bly sup-

‘ porhng the reversible reaction:

NiF (s) + N;(s)-—->2N|F(g)

‘ The magnitude of the effect noted ‘was, however,

so small that there is no conclusive proof of such
a mechanism. No explanation can be offered for
the partial reduction of NiF, in the platinum
crucibles,

99

 
 

 

 

 

 

»
ANP PROJECT PROGRESS REPORT

TABLE 2.2.3, WEIGHT LOSSES OF NiF ,(s)
ON HEATING TO 800°C

" Crucible  Weight Loss of Initial Nfl:2 (%)

 

 

Material ‘After 3 he After 22 hr Total
Platinum 1.78 0.18 1.96
Platinum 234 0.38 7 272
Nickel 3.58 0,93 4,51
Nickel 3.06 0,78  3.84
Nickel 3411 0.86 3.97

 

All these phenomena will, presumably, be
important in mechanisms of corrosion of nickel by
molten salts. It is anticipated that considerable
additional study of these problems will be under-
taken.

ACTIVITY OF CHROMIUM
‘IN CHROMIUM-NICKEL ALLOYS

M. B. Panish

The determlnatnon of the activity of chromium
in chromium-nickel alloys has been completed
with the use of the electrode concentration cells
and the techniques previously described.34 The
values obtained for the activity of chromium in
various chromium-nickel alloys at 750 and 965°C
during this investigation are shown in Fig., 2.2.2.
The curve obtained by Grube and Flad> at 1100°C
is shown for comparison. From these data it
appears that at concentrations below about 25 at. %

the activity of chromium is somewhat lower than

its atomic fraction in the alloy.

The curve obtained at 750°C in this investigation
is consistent with the expected behavior of a
system which has an immiscibility gap. However,
both the Grube and Flad curve and the data ob-
tained at 965°C in this investigation seem to yield
implausibly fow chromium activities for the
chromium-rich solid sclutions.

 

M. B. Panish, ANP Quar, Prog. Rep. March 10, 1956,
ORNL.-2061, p 92.

‘M- B. Panish, ANP Quar, Prog. Rep. June 10, 1956,
ORNL-2106, p 93,

5G, Grube and M. Flad, Z. Elektrochem. 48, 377
(1942).

100

UNGLASSIFIED

 

 

 

 

 

 

 

     

 

 

 

 

 

 

 

 

 

 

. . ORNL=LR—-DWG 16153
100 o
TWO-PHASE
REGION AT 750°¢" | /
. 7
lo
075 TWO-PHASE REGION —
p—t—ar 965°C e ¥
i- ‘L/ -
> C
': 7.
= 050 ‘ /
- P
[ & /'_ ;
< / TWO-PHASE REGION
) |=t— aT t100°¢ ——|
0.25
S
o 750°C
/ A965°C i
/7 e 1100°C (GRUBE AND FLAD)
o o
Ni 25 50 75 - Cr

CHROMIUM (ot. %)

Fig. 2.2.2, Activity of Chromiufi in :Nilckel-
Chromium Alloys,

SOLUBILITY AND STABILITY OF STRUCTURAL ,

METAL FLUORIDES IN VARIOUS
FLUORIDE MIXTURES

J. D. Redman

Data gathered from studies on the stability and
solubility of CrF,, FeF,, and NiF, in NoF-ZrF
(53-47 mole %) at 600 and 800°C were reported
previously.®  The results obtained at 600°C
demonstrated that the solubility of the structural
metal fluoride was a function of the excess of
structural metal fluoride present. It is believed
that the change in solubility is caused by the
presence of a solid phase, probably MF2-ZrF4, and
the resulting change of the NaF concentration in
the melt. The studies have been extended to
include the solvents LlF-ZrF (52-48 mole %),
NaF-ZrF (59-41 mole %), ond KF-ZrF (52-48
mole %), anci additional studies have been mcde on
the solubility of C!'l"'3 in the NaF-LiF<KF eutectic.

Data for the solubility of CrF, at 600°C in the
three additional solvents mentloned above, as
presented in Tables 2.2.4, 2.2.5, and 2.2.6, show
that the solubility varies markedly with the solvent

 

6). D. Redman, ANP Quar. Prog. Rep, June 10, 1956,
ORNL-2106, p 101,
 

 

 

PERIOD ENDING SEPTEMBER 10, 1956

TABLE 2.2.4. son.usu.nv AND STABILITY OF CtF, IN MOLTEN LiF-ZrF,
(5248 mole %) AT 600 AND soo°

 

Conditions of Equilibration -

Found in Filtrate

 

 

 

 

 

 

0,70

Temperature _ Cr*f, Zr=to-Cr - Zr - L Zeto=L i cett . Total Cr
(°c) (wt %) - Mole Ratio - (wt %) - {wt %) Mole Ratio* {wt %) (wt %)
600 25 10 4444 4.15 0.81 131, 159

25 10 44.8 409 0.82 125 154
50 49 432 4.33 0,76 2.0 2.3
5.0 49 -43.0 432 0.76 2.4 2.6
100 2.2 40,4’ 435 0.71 46 5.8
10.0 2.2 402 4.28 0.72 4.6 5.8
10,0 2.2 40.2 4.24 0,72 5.0 5.9
800 100 22 386 3.2 0.89 64 9.5
10,0 22 38.7 3.49 0485 641 945
20 0.9 3066 2445 0.95 12.8 19.1
20 0.9 3006 343 0.75 14.1 19.0
_ *Ratio of Zr to Li was 0.92 in charge material,
TABLE 2.2.5. sor.usu.m AND STABILITY OF CrF, IN MOLTEN NaF-ZrF,
e ‘ (59-41 mole %) AT 600 AND soo"c
'Conditions of Eqbilibraflen Found in Filtrate
' Tombera’ture' - crtt ) 'Zr-'to.-Cr'_  Zr Na Zr'-td-N.u cett Total Cr
°cy (wt %) Mole Ratio.  (wt %) (wt %) Mole Ratio* (wt %) (wt %)
600 L5 15392 4 069 RRY 1434
s s s a2 0.69 1,03 1.35
60 35 3600 0137 0:66 4.7 5.4
&0 35 364 1342 0469 46~ 5.4
100 20 340 120 071 B 89
800 - 120 s 329 124 067 88 10:3
20w 323 2 068 93 . 13
.20 -0 08 266 . 9.8 068 V75 1943
L2008 2T 96 s

19.3

 

*Ratio of Zr to Ne was 0.70 in charge fnntéridl.

101

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

TABLE 2.2.6.- SOLUBILITY AND STABILITY OF CrF, IN MOLTEN KF-ZrF,
(52-48 mole %) AT 600 AND soo°c

 

Conditions of Equilibration

" Found in Filtrate

 

 

 

Temperature Cr'H' Zr-to-Cr Zr K . Zreto=K ottt . -Total Cr
- (°0) (Wt %) Mole Ratio (wt %) (wt %) Mole Ratio* (wt %) (wt %)
. 600 5.0 42 35.8 17.0 0.90 48 448
5.0 42 34.2 168 -~ 0.88 4.3 4.8
10 1.9 345 16:3 091 69 6.8
10 1.9 348 163 0.92 7.1 7.0
- 800 10 1.9 32.8 150 0.94 9.4 9.7
| 10 1.9 33.1 156 0.91 0.2 9.7
20 o8 2640 13.4 0.84 175 194
200 0.8 2641 12.8 0.88 158 1945

 

*Ratio of Zr to K was 0.92 in charge material.

employed. The solubility in NaF-ZrF, (59-41
mole %), as shown in Table 2.2.5, is higher than
that previously found for NaF-ZrF 4 (53-47 mole %)
and substantiates the belief that the solubility
increases with increasing NaF concentrations.
Only in the case of LiF-ZrF, was a sufficient
excess of CrF2 present to affect the zirconium-to-
alkali ratio. An examination of Table 2.2.4 reveals
a decrease. in this ratio os the excess of CrF

present increases. The solute is too soluble at
800°C in all the solvents studied to show any
similar effect, if indeed it does exist. Divalent
chromium appears to be the stable valence state
in these solvents, although there is some indication
that a smaller fraction of the chromium is present
as Ce** in the LiF-ZrF , mixture than in the other
two solvents.

Comparable data for the solubility of FeF
in these solvents are presented in Tables 2.2.7
and 2.2.8. It may be seen that in all tests at
600°C the solubility increased with excess FeF,
present and that the zirconium-to-sodium ratios
tended to decrease with increasing excess FeF‘.2
The solubility of FeF, at 600°C is higher in the
NaF-ZrF , mixture than in either the LiF-ZrF,
mixture or the KF-ZrF, mixture. However, if the
solubility in these two solvents is compared with
the solubility in NaF-ZrF (53-47 mole %), it will
be found that with 1 wt% FeF added the solubility
in the NaF-ZrF , mixture is |ower than in any of the

102

others and that for a 5 wt % addition all three give
essentially the same values. The increase of the
NaF content in the NaF-ZrF, mixtures causes an
increase in the solubility of FeF, which is most
pronounced at the lower iron concenfrat:ons. The
data indicate that divalency is the stable form of
iron in these solvents, There is a slight indication
that Fe** is not so stable in NaF-ZtF, (59-41
mole %) as in the other solvents studied, mcludmg
NaF-ZrF , (53-47 mole %).

Comparable data on the solubility of NiF,
these solvents are presented in Tables 2 2 9,
2.2.10, and 2.2.11. From the data it is apparent
that the solubility of NiF, in NaF-Z:F, and
KF-ZrF, is nearly the same and independent of
the amount of NiF, added. However, the solubility
of NiF, in the LiF-ZrF, mixture is a function of
the amount of NiF, added at both 600 and 800°C.
Values found previously - in NaF-ZrF,  (53-47
mole %) are quite similar to those found for the
LiF-ZrF, mixture and smaller than those given
for NaFZZeF, (59-41 mole %) at 600°C with 1.5
wt % NiF, additions, but they are essentially
equal with 5 wt % NiF, present.

Data were reported prevmusly7 for the solubility

of CrF3 in NaF-LiF-KF (11.5-46.5-42 mole %) .

at 600 and 800°C. Although the results in all

 

7). Do Redman, ANP Quar. Prog. Rep. Dec. 10, 1955,
ORNL-2012, p 88

4

.
 

 

%

PERIOD ENDING SEPTEMBER 10, 1956

TABLE 2.2.7. SOLUBILITY AND STABILITY OF FeF, IN MOLTEN NoF-ZrF

(5941 mole %) AT 600 AND soo°c

 

Conditions of Equilibration

. Found in Filtrate

 

 

Temperature Fett Zr=to-Fe Zr Na Zr-to=Na F Fett Total Fe
(°c) (wt %)  Mole Ratio  (wt%) (Wt %)  Mole Ratio*  (wt %)  (wt %) (wt %)
600 3.0 7.9 38.3 14.5 0.67 4501 1.8 1.8

3.0 7.9 383 1446 0466 46.0 1.6

5.0 4.7 37.4 15.2 0.62 44,1 3.6 3.8

5.0 4.7 37.5 153 0,62 4441 2.9 3.8

10.0 2.1 37.6 14.8 0e64 3.7

10,0 24 37.6 15.9 0.60 44.9 3.5 3.7

800 10.0 24 34 125 0:67 43.5 8.0 102
100 24 40 119 0:72 4401 7.7 9.7

100 20 33.9 11.8 0.72 443 7.2 9.4

20 0.8 31.2 11.5 068 43.8 12.0 14.8

20 0.8 30 1.2 0.68 42,9 1149 1640

 

*Ratio of Zr to Na was 0,70 in charge material,

cases indicated that trivalent chromium was stable
in this solvent, wide variations in the values for
the solubility were found. It has been established
that these variations resulted from errors in
analysis, and, following the elimination of the
analytical difficulties, the experiments were
rerun. The results of the reruns, which are
presented in Table 2.2.12, are believed to represent
the solublllty under these condmons. B

. REDUCT!ON OF UF BY STRUCTURAL METALS' o

J D Redman

. Addmonal studles on the reduchon of UF by '
- Cr% or Fe® have been made in various reachon
" media by using the filtration method. Data obtained

- from .earlier studies on these reactions with the
- solvents. NuF-ZrF P (50-50 mole %, 53-47 mole %, -

~ and - 59-41 -mole %), LiF-KF (52-48 mole %),

 KF-ZrF < (52-48 mole %), Li F-NaF-ZrF , (55-22-23 .
" mole. %), -and NaF-L:F-KF (11.5-46. 5-42 mole %)
_.as the reaction media were presented prevnously., .

~ More recently - RbF-ZrF‘ (6040 mole. %) ond

 

J. D. Redman, ANP Quar. Prog. Rep. June 10, 1956,
ORNL-21 0§, p 94,

LiF-BeF, (48-52 mole %) have been studied as the
reaction media.

Preliminary data for the reaction of UF, with
Cr® at 600 and 800°C in the solvent Rblé-ZrF 4
(60-40 mole %) are given in Table 2.2.13. Ap-
proximately 2 g of Cr% was reacted with UF,
(9.2 wt %, 4.0 mole %) dissolved in apprommateiy
40 g of the RbF-ZrF , mixture contained in nickel
in these tests, Slm:1ar studies with Fe? in place
of Cr® yielded the results reported in Table 2.2.14.

The 800°C data shown in Table 2.2.13 are not

satisfactorily precise. However, .it appears that

- the ‘equilibrium chromium concentrations existing
. in the RbF-ZrF, mixture are lower than those
- i,noted in any of the other alkali fluorlde—ZrF‘

mixtures studied. The lowest values found previ-
: ously were those measured in the KF-ZrF 4 (52-48

mole %) mixture, where roughly- 1000 ppm of

" chromium was found at both 600 and 800°C. Valuves
for other mixtures are: Li F-ZrF (52-48 mole %),
12900 and 3900 ppm at 600 and 800°C respechvely,
NaF-ZrF 4
600 and 800°C, respectively.:

(53-47 mole %), 1700 and 2100 ppm at .
Thus it appears
that the fluoride ion activity, as measured by the
equilibrium chromium concentration, decreased,

103

 
 

 

 

 

‘ANP PROJECT PROGRESS REPORT

TABLE 2.2.8. SOLUBILITY AND STABILITY OF FeF,
IN LiF-ZrF, (5248 mole %) AND KF-ZrF,
(52-48 mole %) AT 600 AND aoo°c

 

Conditions of Equilibration Found in Filtrate

 

 

{wt %)
Temperature Fo't Fett Total
(°c) (wt %) ° Fe
Solvent: LiF-ZrF‘ _
600 1.0 0.59 0.63
' 'I 0 0.48 0.43
5.0 1.0 1.1
5.0 1.3 1.1
800 5.0 4.8 5.1
) 5.0 4.9 5.0
10 9.0 94
10 87 9.4
18 1546 17.5
18 14.6 172
Solvent; KF-ZrF 4
600 1.0 1.0 1.1
1.0 1.0 1.2
5.0 1.3 1.5
5.0 144 1.6
10 2.3 3.3
'800 5-0 4.5 500
5.0 449 4.9
10 8.3 9.3
10 8.3 9.0
18 8.8 102
18 : 9.1 9.7

 

in the order listed, for the atkali fluoride-ZrF"
mixtures containing Rb, K, Na, and Li.

The data presented in Table 2.2.14 suggest that
equnhbnurfi for the Fe%-UF, system is attained
rather slowly in RbF-ZrF 4+ and, although it is
believed that equilibrium was attained after 5 hr
of heating, longer periods will be studied to
establish this point. If it is assumed that the
values for the 5-hr heating period are equilibrium
values, a comparison of these with the iron values
obtained for the other alkali fluoride—ZrF , mixtures
shows that the iron concentrations at both temper-
atures studied are virtually independent of the
reaction medium used and in all cases are higher

104

'TABLE 2.2.9. SOLUBILITY OF NiF, IN MOLTEN
LiF-ZeF, (52-48 mole %) AT 600 AND 800°C

 

 

 

Conditions of Equilibration - Total Ni
Temperature Nitt - Found in Filtrate
°C)y - (wr %) - (wt %)
600 1.5 0,12
1.5 015
5.0 0.33
5.0 023
80 0.33
8.0 - 0.49
800 1.5 1.3
1.5 ' 13
5.0 2.2

5.0 ) 2.3

 

at 600 than at 800°C. This behavior is in marked
contrast to that observed for the Cr%UF 4 System,
which exhibits a marked dependence on the re-
action medium employed No satisfactory expla-
nation can be offered for this indifference of the
Fe -UF system toward changing fluoride ion
actsvmes.

Studies were also made of the reaction of Cr?
with UF, in molten LiF- BeF, (48-52 mole %).
The data are reported in Table 2.2.15. 1In these
tests approximately 2 g of Cr® was reacted with
UF, (1.5 wt %, 1.5 mole %) dissolved in approxi-
mately 30 g of the LiF-BeF, mixture contained in
nickel. The data presented in Table 2.2.15 show
that the Cr -UF system is strongly temperature
dependent over 1he range studied, and consequently
the attack by UF, on chromium-containing alloys
in this system might be expected to be serious.
The only solvent previously studied with which a
comparison can be made is LlF-ZrF (52-48
mole %). This mixture yielded chromwm values

of 2900 and 3900 ppm at 600 and 800°C, values

which are considerably higher than those reported

for the LiF-BeF, mixture.

. Several experlments were performed in which
5 wt % Cr, either as CrF, or GF,, was added to
the LiF-BeF. mixture and equilibrated for 5 hr at

800°C. In the case of the CrF, addition, all the

o

e
 

 

PERIOD ENDING SEPTEMBER 10, 1956

TABLE 2.2.10. SOLUBILITY OF NiF, IN MOLTEN NaF-ZrF, (59-41 mole %) AT 600 AND 800°C

 

 

 

 

 

 

 

 

 

 

 

 

 

 

3 Conditions of Equilibration ‘Found in Filtrate
Temperature Ni++ Zr-to-Ni _ Zr Na Zr=to=Na Total Ni
(°C)  (wt %) ~ Mole Ratie (wt %) {wt %) Mole Ratio* (wt %)
=
600 1.5 , 17 40.3 15.5 0.66 0.38
1.5 17 40,2 15.0 0.68 0.36
- ' ’ - 5.0 4.7 4042 14.9 0.68 0.36
5.0 4.7 40.3 15.2 0.67 0.32
800 1.5 17 39.9 15.1 0.67 0.98
1.5 17 40.1 14.9 0.68 0.87
5.0 : 4,7 - 404 15.1 0.68 0.94
5.0 4.7 40.0 15.2 067 1.03
*Ratio of Zr to Na was 0.69 in charge material,
TABLE 2,2.11: SOLUBILITY OF NlF2 IN MOLTE‘N KF-ZrF4 (52-48 mole %) AT 600 AND 800°C
~ Conditions of Equilibration : Found in Filtrate
* Temperature Nitt Zr=to=Ni Zr K Zr-to-K Total Ni
(°C) (wt %) Mole Ratio (wt %) (wt %) Mole Ratio* {wt %)
T 600 1.5 17 38.3 18.6 0.89 0.35
1.5 17 37.9 18.7 0487 0.33
5.0 446 37.9 18.9 0.86 0.21
: 5.0 46 38.1 18.9 0.87 0.34
800 1.5 17 38.5 18.1 0:91 0.98
'|.-_5 17 _ 38.4 18.4 0.90 0.9
50 46 - 385 18.4 0.90 1.0
50 46 385 18.4 0.90 0.90
*Raho of Z!‘ io K was 0.92 in charge muferscl. B o o 7
TABLE 2.2 12. son.uerrY oF CrF IN NuF-LiF-KF (11.5-46.5-42 mole %) AT 600 AND 800°C
Cond |'hons of Equihbruhon ' Founde:ltrute - :_"Cofidjifibl}s ‘oquuilil':rqt'ior'l "--F.c-»'imd ifi F‘i“-rd.';er .’ _-
. - ' Temperaiure_ . C +++ . T:‘St‘." C’ P 'T_"f“i Ni - .","':V-TB""-'PVOTQ:ffireV | Cr‘{"_Hb - : :Totqj Cr - Total Ni -
| ©o . wmw (wf %) . Aepm). 7 {OC) Cwt%) o wt %) o (ppm)
. 600 . 10 - o.zo_-' Cooowest o 80 5.0 3.0 - 65
| | 5.0 023 195 | 10 36 25
o’ 5.0 0.27 20 10 4.6 235

 

105

 
 

 

 

ANP PROJECT PROGRESS REPORT

TABLE 2.2.13. DATA FOR THE REACTION OF UF, WITH cr® IN MOLTEN
RbF+ZrF , (60-40 mole %) AT 600 AND 800°C

 

Conditions of Equilibration

Present in Filtrate

 

 

 

Temperature Time Total U Total Cr* Total Ni
(°C) (hr) (wt %) (ppm) (ppm)
600 3 649 550 45
3 649 570 50
5 648 630 60
5 7.0 650 65

800 3 7.1 520 55
3 7.1 370 70
5 7.1 1100 70
5 71

630 55

 

*Blank of 150 ppm of Cr ot 800°C.

TABLE 2.2.14. DATA FOR THE REACTION OF UF WiTH Fe? IN MOLTEN
RbF-ZrF (60-40 mole %) AT 600 AND 800°C

 

Conditions of Equilibration

 

Present in Filtrate

 

 

Temperature Time Total U Total Fe* Total Ni
(°c) (hr) (wt %) (ppm) (ppm)
600 3 7.1 300 45

3 7.0 320 40
5 7.1 530 70
5 7.2 540 75
800 3 7.3 100 35
3 7.3 110 140
5 7.2 420 75
5 7.2 360 65

 

 

 

*Blank of 150 ppm of Fe at 800°C,

chromium was soluble and at least 80% of it was
present as Cr**. In the case of the CrF. addition,
all the chromium was soluble, but only about 20%
was reduced to Cr** by interaction with the nickel
container. A fair balance was observed between
the quantities of Cr** and Ni** formed. It appears
that Cr** is the stable valence state for chromium
in the LiF-BeF, mixture and that Cr*** is reduced
to Cr** by nickel in this solvent. This behavior
is in agreement with that noted for chromium in the
various aikali fluoride-ZrF4 mixtures studied.

106

SOLUBILITY DETERMINATIONS BY
MEASUREMENT OF ELECTROMOTIVE FORCES
OF CONCENTRATION CELLS
L. E. Topol

The solubilities of the structural metal fluorides
NiF,, FeF,, and CrF, in molten NoF-ZtF, (53-47

mole %) that were determined by using concen-

tration cells ond an emf method? are presenied in

 

L. E. Topol, ANP Quar. Prog. Rep. June 10, 1956,
ORNL-2106, p 103 !

¥)
 

 

T AL s b e

 

&

TABLE 2. 2.15. DATA FOR THE REACTION OF UF,
WITH Cr® IN MOLTEN LIF-BeF (48-52 mole %)

 

Temperature of Present in Filtrate

 

 

Equilibration Total U Total Cr*  Total Ni
0
("C) (wt %) {(ppm) (ppm)
550 7.4 930 370
7.8 920 - 310
650 8.1 1470 240
7.9 980 320
8.0 1450 310
8.0 1300 280
800 8.3 2260 270
8.3 1900 230
- 8.4 1850 330

8.3 2220 270

 

*Blank of 450 ppm of Cr at 806°C,

Fig. 2.2.3. The procedure con5|sfed in measuring
the emf of cells of the type

MlMFz(c]), NaF-ZrF | |NaF-ZrF4, MFz(cz)IM

where M is the structural metal, as a function of
temperature. The temperature at which the solute
becomes soluble on heating or precipitates out on
cooling is readily determined by a discontinuity
in the plot. of voltage against temperature, The
cells were contained in crucibles of aluminag,

platinum, or the metal M, and measurements were -

The temperature
Electrical contact

made in a helium atmosphere.

range was 525 to 750°C.

between the half cells was effected by using a
ZrO bridge lmpregncted with the solvent NaF-

ZrF

Further efforts to determine the formula. of
the safurahng phase, previously ‘thought to be
MF.zoZrF‘, were unsuccessful. The solubilities
shown in Fig. 2.2.3, in general, fall within the
range reported by Redman!® for solubility samples

 

104, D, Redman, ANP Quér. Prog. Rep. June 10, 1956,
ORNL-2106, p 101,

PERIOD ENDING SEPTEMBER 10, 1956

OB
ORNL—LR-DWG 16154

0.5

0.2

0.05

0.02

0.01

SOLUBILITY (mole fraction}

0.005

0.002

 

0.001
800 750 700 650 600 550 500

TEMPERATURE (°C)

Fig. 2.2.3. Solubility of Structural Metal
Fluorides in NaF-ZrF, (53-47 mole %),

obtained by filtration. However, in order to be
consistent with the filtration data, the emf values
should correspond to filtration values extrapolated
to correspond to a negligible amount of precipitate.
An adequate explanation for the emf values being
higher than the extrapolated filtration values has

_not been found

‘From the FeF, solubility values listed previ-

. ously!V and the more recent values of 0.57 wt %,

or 0.62 mole %, ot 598°C, and 5.00 wt %, or 5 37
mole %, - at 684°C, the solubility of FeF,
NaF-ZrF (53-47 mole %) can be expressed by

| 36.0 x 10°
|OgN’="-—------————+ 6.87;

4576 T

where N is the mole fraction of the solute and T is
in degrees Kelvin. These dote yield a heat of

 

N, E Topol ANP Quar Prog. Rep. June 10, 1956,
ORNL-2106, p 105, Table 2.2.11.

107

 
 

 

 

 

 

 

ANP PROJECT PROGRESS REPORT

ll!

solution of 36.0 kcal and an *‘ideal’’ melting point

of 873°C for FeF ,.ZrF .

 

Solubility data obtained for CrF, in NaF-ZrF,
(53-47 mole %) are presented below:
Temperature CI'F2 Found
(°c) wt % mole %
554 0.42 0.47
565 0.65 0.73
575 0.90 1.00
592 1.07 1.19
-592 1.14 1.28
606 1.50 1468 -
609 ‘ 2.04 2.28
634 3.57 3.99
644 4.45 4.98
657 6440 7.13
660 . 7423 ‘ 8.06
710 20.28 22.20
These data can be expressed as
' 38 4 x 103
logN = =« o —— + 7.86 ,
4.576 T

and they yield a heat of solution of 38.4 kcal and
a melting point of 795°C.

The solubilities of FeF’2 CrF,, and NiF,, as
well as those of the corresponding fluorozrrconate
complexes (MF ,-ZrF ), were determined in an
effort to find a quasi binary. As can be seen in
Fig. 2.2.3, there is no detectable difference in
solubility between each metal fluoride and its
cotrresponding fluorozirconate, except for chromium,
and even for chromium the difference is small and
may be due to experimental error. This means that
the liquidus temperatures along the two joins
(MF ,-7NaF-6ZrF, ond MF,.-ZrF 7NaF.6Z¢F )
in tfi ternary phase diagram are almost the same.
The most recent evidence from quenched prepa-
rations suggests that only in the case of FeF
is FeF,-ZrF, the saturating phase and that a
solid ternary phase is precipitated in the other
systems.

Similar concentration cefls containing CuF,
(with copper crucibles and electrodes) have
frequently yielded indeterminate results. Although
CuF, in the presence of metallic copper in NaF-
ZrF melts might be expected to reduce com-
pletely to CuF, the cells have behaved as though

108

both CuF and Cqu were present, with the mono-
valent salt predominating at higher temperatures.

A vessel fabricated of boron nitride (obtained
from the Carborundum Co.) was charged with an
NaF«KF-LiF (eutectic) mixture and heated for
three days at temperatures from 550 to 770°C in a
helium atmosphere. The resistance between a
nickel wire immersed in the melt and the metal
platform supporting the crucible remained high
(107 ohms at 650°C, 4 x 10# ochms at 770°C) and
was reproducible during the run. Examination of a
cross section of the cooled container revealed
penetration by the melt of about l/ in.,- but no
petrographic evidence of contammaflon of the
melt wos found. Further tests will be made to
study the stability of this ceramic to fused
fluorides.

ACTIVITIES IN THE RI.':I"'-ZI'F4 SYSTEM
S. Cantor S. D. Christian

The vapor pressure data previously reported?!?
for the RbF-ZrF, and the KF-ZrF, systems have
been processed to yield some information on how
the activities change with the concentration and
temperature. Since ZrF , was found to be the only
volatile constituent, the activities of ZrF, based
on liquid standard states were obtained from the
appropriate log p vs 1/T equation, presented
previously,'2 by calculating the pressure at a
given temperature and dividing by the pressure of
pure, liquid ZrF .

The only value available for the vapor pressure
of pure, liquid ZrF, was obtained at the melting
point, 912°C. To obtain vapor pressures for pure,
liquid ZrF at lower temperatures the following
relation was used:

P, AH, /1
log — = —_ -1,
P, 230R (Tm T)

 

where
AH_ = heat of melting in cal/mole,
T, = melting point in degrees Kelvin,
P_ = sublimation pressure,
P, = supercooled liquid wvapor pressure,

R = gas constant.
If a tentatively estimated value of 13.5 keal/mole
is adopted for the heat of fusion of ZrF ,, P, for

 

125, Cantor, ANP Quar. Prog. Rep, June 10, 1956,
 

4

ZrF, VAPOR PRESSURE (mm Hg)

ZrF4 is found to be 2.64, 25.9, and 164 mm at
600, 700, and 800°C, respectively. At 912°C
the vapor pressure of pure ZrF, is 905 mm. A
plot of the vapor pressure of Zrl4-'4 from solutions
of RbF in ZrF, at various temperatures is shown
in Fig. 2.2.4, and the activities of ZrF, obtained
from these values are given in Table 2.2.16.

A curve-fitting process showed that the activities
of ZrF, at 912°C could be expressed by the

following analytical equation:

| AzeF 4
L

XZrF 4

x2 4
RbF
AONPITERT s
. ORNL-LR-DWG 16455

1000

100

  
 

 

0.000 L

. ZeFy {mole fraction). . L

Fig. 2.2,4. Vapor Pressure vs Cémposfiion m

the Rb F-ZtF, System,

 and

SO0 e s e02 1_}_,; .04 - 08 -0 08 _-’:'4.07

PERIOD ENDING SEPTEMBER 10, 1956

where a is the activity and X is the mole fraction
of the species indicated by the subscript. By
the use of the Gibbs-Duhem equation in the form

2zr F

| ARbF
-i-XRden X

 

 

XZrF4 dln

XZrF4

the analytical equation

ARbF

 

log
XRbF

X2 4
ZrF‘

was obtained. Fromthe two analytical expressions,
the activities of ZrF , and RbF shown in Fig. 2.2.5
were then calculated.

Rational activities of RbF were not calculated
for lower temperatures because of uncertainties
which developed when the curve-fitting procedure
was attempted. There were insufficient data to
avoid large uncertainties from extrapolation into
regions where X, . approached unity.

ACTIVITIES IN THE.KF-ZI'F‘ SYSTEM
S. Cantor S. D. Christian -

- The ZrF, activities for the KF-ZrF, system
were calculated, by the same method as that used
for the RbF-ZrF , system, from the vapor pressures
shown in Fig. 5.2.6.' "The results are tabulsted
in Table 2.2.17 and plotted in Fig. 2.2.7.

The expressions used in getting rational ac-

. tivities of KF at 912°C were

o '5aZfI-:4 o
©4YZrF, ' '
o e s j _ .

 

a

log ——
KF : - . :

=~ 1237 + 7.74 erF4 .

ZfF4

 

From these equations the curve for a_ . vs compo-
sition, as shown in Fig. 2.2.7, was calculated.

109

 
 

 

 

ANP PROJECT PROGRESS REPORT

TABLE 2.2.16. ACTIVITIES OF ZrF‘ IN RbF-Z:F, SOLUTIONS

 

 

 

Composition Activity

(mole fraction ZrF ) At 600°C At 700°C At 800°C At 912°C
0816 0.293 0,398 0,507 0632
0:746 0:260 0:219 0382 0.446
0,700 0:209 0.230 04253 0,273
0.651 04194 0.192 " 0.192 0190
0:600 0,131 0119 0.112 0104
0.548 0.0687 0,0595 0.0536 0.0482
0.504 0.0581 0.0390 0.0338 0,0270
0.452 0.0146 0.0123 0.0109 0.00920
0.400 0.00345 0.00316 0.00305 0.00288
0350 0.000645 0,000780 0.000839

0.000710

 

 

As a consequence of strong complexing, the

activities decrecse extremely rapidly with compo-
sition. Activities of KF are very slightly higher
than the activities of RbF at corresponding compo-
sitions and temperatures, since the larger rubidium
ion has less affinity for the fluoride and RbF is
therefore a better complexing agent than KF.

An interesting comparison is found in the plot
of aZrF in alkali fluoride solutions against ionic

radius of the alkali metal. If the activity of
ZrF, is arbitrarily chosen as that ot 50 mole %,
the plot shown in Fig. 2.2.8 is obtained at 912°C.
The ZrF | activity decreases with increasing ionic
radius o? the alkali cation. The activities in
LiF-ZrF , {ref 13) and NaF-ZrF (ref 14) solutions
were obtcmed by dividing the vapor pressure of the

solution by 905 mm, the vapor pressure for pure

ZrF

 

3R, E. Moore, ANP Quar, Prog. Rep. June 10, 1955,
ORNL" 896: P 800

14K, A, Sense et al.,, Vapor Pressures of the Sodium

Fluoride=Zirconium Fluoruf System and Derived Infor-
mation, BMI-1064, p 22 (Jon. 9, 1956).

110

SOLUBILITY OF LeF, AND OF CeF, IN
MOLTEN FLUORIDE MIXTURES

W. T. Ward

A systematic investigation of solubilities of
some rare-earth fluorides in ART-type fuels was
initiated. Two introductory experiments related
to this investigation were reported previously.}5¢16
Cost and availability considerations indicated
lanthanum and cerivm fluorides to be logical
starting materials for the investigation. Mixtures
considered to be the least difficult from the stand-
point of chemical analysis were  investigated
first. The solvents NaF-KF-LiF (11.5-42-46.5
mole %) and NaF-ZrF (50-50 mole %) have been
studied.

The experimental procedure consisted in mixing
a given quantity of the rare-earth fluoride with
previously purified solvent to make an initial

 

15e, L. Daley and W. T. Ward, ANP Quar. Prog.
Rep. ]une 100 1956’ ORNL.2106} P 108.

16¢, M. Blood, ANP Quar. Prog. Rep. June 10, 1956,
0RNL-2106. p 105 ,

Vo
 

 

e the” s s 4 o

 

N

"

.

GONFIDENPIRL
ORNL-LR-DWG 16156
0.5

0.2

0.05

c.o2
0.0t

0.005

a, ACTIVITY

0.002

0.004

0.000%

0.0002

0.0004

— IDEAL
0.00005 o i,

0.00002

 

0.00004
0 0.2 0.4 0.6 0.8 1.0
ZrF, {mole fraction)

Fig, 225, Activities of ZiF, and RbF ot
912°C vs Composition in the RbF-ZrF, System,

charge of 1 kg'. - The mixture was purified in a

_nicke! reactor by hydrofluorination and hydrogenation -
at 800°C, Then, with continuous helium sparging,

the temperature of the system was increased to
850°C, and the mixture was stirred for approxis

mately 30 min. The temperature was then lowered.
to 800°C, and, cfter approximately 60 min, a
 filtrate was obtained with a nickel filtering tube,

The temperature was then further lowered, and
filtrates were obtsined in ¢ similar manner at
700, 600, and 550 or 500°C depending on the
liquidus temperature of the solvent used. In order
to avoid cold spots in the reactor and pronounced

PERIOD ENDING SEPTEMBER 10, 1956

OONTITEN I
ORNL-LR-DWG {6157
1000

100

Zrf, VAPOR PRESSURE (mm Hg)

o X

0.0

 

0.004
o 0.2 0.4 0.6 0.8 1.0
ZrFy {mole fraction)

Fige 2.2.6. VYapor Pressure vs Composition in
the KF-ZrF ¢ System,

temperature gradients, two independently operated
Calrod heaters were installed at the bottom and top
sections of the reactor. Prior to obtaining each

~ filtrate, the temperatures of the auxiliary heaters
“were adjusted until temperature-depth surveys in

the sclution revealed temperature differences no
greater than 10 deg. In attempting to ascertain
whether the solubility of the additive was a function
of the amount added, the filtration procedure was
repeated after the initial quantity of rare-earth
fluoride had been doubled by the proper addition
of more rare-earth fluoride. The system was

‘cooled overnight prior to the additions in order to

avoid possibilities of contamination and the losses
inherent in high-temperature additions. The

11

 
 

 

ANP PROJECT PROGRESS REPORT

results of several experiments are summerized in
Tables 2.2.18 through 2.2.21. _

The results presented in Table 2.2.18 indicate
that the solubility of CeF, in the NaF-KF-LiF
eutectic mixture is at least 6§ wt % at 500°C.
When the survey of other ART fuels is completed
this experiment will be repeated with greater
additions of CeF,. The dota in Table 2.2.19
indicate that the solubility of LaF, in the NaF-
KF-LiF eutectic mixture is at least 12.0 wt %

- (4.12 mole %) at 500°C. Petrographic and x-ray

diffraction examination of the second 800°C

filtrate revealed traces of KLaF 4 The NalaF 4
compound was not detected.

The data in Table 2.2.20 indicate that, within
the apparent analytical -precision, the solubility
of CeF 3in NaF-ZrF 4 (50-50 mole %) is independent
of the amount of CeF, added. Petrographic and
x-ray examination of the reactor heel revealed
relatively large quantities of pure CeF,, and only
traces of an unidentified phase that exhibited high
birefringence.  Free CeF, was present in the

800°C filtrates.

TABLE 2.2,17. ACTIVITIES OF Z¢F, IN l(F-ZrF4 SOLUTIONS

 

 

 

CompbsitiOn : Aéfi""”
(mole fraction ZrF ) At 600°C At 700°C At 800°C. At 912°C

00659 , 0.201 0.221 O o.240 0,260
0.59 , 0.155 0.151 0.149 0.146
06550 | 0118 0.0998  0.0872 00765
0,499 0.0739 0.0557 0.0443  0.0358

' 0.476 0.0508 0.0395 | 0.0321 0.0264
0453 - 040243 0.0200 0:0174 0.0150
0:404 0.00432 0.00436 0.00441 0.00450
0.349 0.000688 0,000811 0,000880 0.00105

 

TABLE 2.2.18. SOLUBILITY OF CeF; IN NaF-KF-LiF (11.5-42-46.5 mole. %)

 

~ CeF, Added

 

 

 

Temperature Ce Found in Filtrate (wt %)
wt % Ce mole % Ce (°c) Colorimetrically By Titration
3.00 : '0.914 800 " 3.2
| ‘ 700 3.0
600 3.1
500 3.1
613 o4 800 6:2 -‘ 6ot
“ 700 &3 63
600 6.1 | 5.7 .
500 63 o 642

 

112

Yo
 

J‘

[ 1)

WONPTOEIL
ORNL-LR-DWG 16158
0.5

0.2

0.05

0.02
0.04

0.005

a, ACTIVITY

0.002

0.001

0.0005

0.0002

0.0001

= = — |DEAL
0.00005 s ZrF,

KF

0.00002

 

0.0000t
0 0.2 0.4 0.6 0.8 1.0
ZrF, {mole fraction)

Fig. 2.2.7. Activities of ZrF, and KF at 912°C
in the KF. ZrF System. |

Petrogrophi'c orid _x-roy_ diffraction examinations

of the reactor heel from the ex'perlmenrts reported

in' Table 2.2.21 ‘showed large amounts of free
Lan, the compound INaF. 6ZfF4, and a trace of
an unidentified compound The third 800°C filtrate
showed a small amount of free LaF ‘and a large
amount . of the compound INaF. 6ZrF "From the
stondpomt of reactor operoflon, the solublhhes
are comfortably high,
study " the feasnblhty of exchange and fractional

crystall_lzon_on of the Lc::F3 from  the “solution ot
the lower temperatures in order to provide infore -

mation for use in fuel reprocessing.

Efforts will be made to

PERIOD ENDING SEPTEMBER 10, 1956

GONPOTRTIE
ORNL-LR-DWG 16159

 

0095

0075

 

a, ACTMITY

Ne* ‘\
0050
N

 

\K+
. N
+

0025 \Rb

08 08 10 (2 14 16
R* CRYSTAL RADII (A)

 

 

 

 

 

 

 

Fig, 2.2,8, Activity of ZrF, ot 50 mole % in
Alkali Fluoride Binary Systems at 912°C vs Size
of Alkali Cation,

SOLUBILITY OF XENON IN FUSED SALTS
R. F. Newton

The components of apparatus for the determi-
nation of the solubility of xenon in fused salts
have been constructed and assembly is nearly
complete. The radioactive isotope Xel33 is to
be used for these studies. Saturation will be
accomplished in a slowly rocking, nearly hori-
zontal, tube about half full of melt. Small samples
of the saturated melt will then be counted with a
scintillation counter.  The sensitivity of the
counter is such that a solubility of 10~7 mole/cm®
should be determinable to within perhaps 20%,
and a solubility as low as 10=1° mole/cm3 can
probably be detected.

Xenon-133 is particularly suitable for this de-
termination because it decays to a stable daughter,
Cs 33, and hence the results will not be falsified
by the counting of o radiocactive daughter which
is. readily soluble in the melt. The xenon is
readily obtainable in quantities more than suf-
ficient for the purpose from the waste gas from

the recovery of iodine from spent fuel slugs.

MOLTEN SALT POLAROGRAPHY

The use of polarography in the analysis of
molten fluorides has been developed sufficiently
for it to be possible to carry out quantitative

113

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

analyses ‘of structurdl metal ions in the NaF-KF-
LiF eutectic. Half-wave potentials have been
identified for Ni**, Fe***, and Cr***, and solu-
bility determinations were made by following the
plot of wave height vs concentration to the solu-
bility limit and beyond.

The successful determinations were made with a
“*bubble cathode’’ of the type employed by Ludwig

and Carter.'? This type of electrode was origi-
nally designed by Lyalikov and Karmazinl® for
use in molten KNO,.. :

 

7 Carter Laboratory Progress Report for Dec. 1955,
Subcontract No. 530, Pasadena, California,

18y, s, Lyolikov and V. l¢ Karmazin, Zavodskaya

Lab. 14, 144 (1948).

TABLE 2.2.19. SOLUBILITY OF LaF, IN NaF-KF-LIF (11.5-42-46.5 mole %)

 

 

 

- LaFy Added Temperature La Found in Filtrate
wt % La mole % Lo (°C) (wt %)

5.99 1.91 800 ' 642
700 5.8

600 6.0

500 ' 5.9

12,01 4012 800 122
700 12.3

600 12.4

500 12.6

 

TABLE 2.2.20. SOLUBILITY OF CeF, IN NaF-ZrF‘ (50-50 mole %)

 

 

 

 

Cer Added Temperature Ce Found in Filtrate (wt %)
wt % Ce mole % Ce (°C) Colorimetrically By Titration

3.00 2.28 800 2.8 2.9
700 2.8 2.6

600 2.7 26

6.00 4467 800 5.4 5.3
700 3.9 3.8

600 2.8 2.7

550 2.6 ‘ 2+4

12.0 9.73 800 5.9
' 700 3.9

600 3.0

550 2.8

 

114

% n
 

P

g

 

PERIOD ENDING SEPTEMBER 10, 1956

TABLE 2,2,21. SOLUBILITY OF LeF, IN__lfluFoZrF‘ (50-50 mole %)

 

LaF, Added

 

 

‘ K Temperefure ‘ La Found in Filtrate
wt % La - mole % La_ - (°C) - (wt %)
3.0 230 800 ' | 3.1

| | 700 3.2

600 o 3.3

ss0 33

60 470 800 6.0
| | 700 | 4.2
600 33

550 2.8

120 | 9481 800 o - 5.7
| | 700 42

600 34

550 | 2.9

 

The elec?rode'wds a thin wire of noble metal

which barely protruded from a tube through which
" inert gas was passed. - When the electrode was
- immersed in a melt, the escaping bubbles caused
~ ‘an intermittent confact of the wire with the melt,
. and ‘thus provided o periodic renewal of the difs
‘fusion layer which’ opproxlmated the behavnor at g

dropping-metal electrode.

Because of . the high conductlvny of molten.
alkali fluorndes, the ‘polarographic circuit. was -
. .designed to operote at currents as high as 100 ma.
~ Suitable bubble rafes (30/min) required o recorder

' “'with'a much faster response than -that commonly _-
e found - “on - aqueous - polurographs.- To aftain this
Bt ,-‘.-_J;'speed ‘a‘recotder with a l-sec full-scale ‘response
7 wes equnpped with d ratchet which allowed the
" recorder pen :to: travel only in- the dlrectlon of -
f."-"_'increasmg current. AT - S
"7 The dipping cathode was a sllver wire fused into -
s __f'_r-'a .thin: Morgamte tube, :which was, in ‘turn,. sure

L fl_'fr‘ounded by a -3-in. nickel tube. - The Morganite

- tube limited the effective electrode area to a very.

- small valve: ‘A nickel crucible ‘served both as a

" container -ond as the anode.  The entire cell

iussembly was - placed in a:helium-filled chamber
that was supplied with’ electrical leads and glove

parts and equipped with o furnace.

" Because the anode was nickel, the polarographic

‘wave for Ni **in NaF-KF-LiF (11.5-42-46.5 mole %)

.voccurrec‘l between 0 and 0.1 v. The diffusion

current (that ‘is, the height of the polarographic
wo've) was a linear function of the concentration.
- The results of the determination of the solubility

- of NiF, at 550°C are shown in Fig. 2.2.9, which

is a plot of diffusion current vs amount of
NiF, added to 33.4 g of the NaF-KF-LiF mixture.

The break represents the saturation concentration

 at 0.42wt % NIF

It had - been" anhcnpated thof once saturation

) ;was ‘reached the “diffusion current would remain

constant. The large increase in the current above
saturation is- probably explained by the presence

of colloidal particles of NiF, in the saturated

melt; the NiF, particles impede the establishment
of o concentruhon ‘gradient around - the micro-
e!ectrode., As cbserved visually, the melt was

clear “until saturated, while further eddmon of'-

"-NlF caused turbidity. ‘

Similar diffusion current-concentrohon curves
v_for NiF, in NaF-KF-LiF (11.5-42:46.5 mole %)
at 500°é gave o solubllity of 0.146 wt % NiF,.
In-tests * with ‘CrF,, -a ‘reproducible half-wave
potential of 0.9 v was found, but no solubility
measurements were attempted.

115

 
 

 

ANP PROJECT PROGRESS REPORT

CONTTOENFL
ORNL~LR~DWG 16160

90 —— /

 

 

 

 

 

 

 

80
70

. I

E

8 60

i /

£ 50 :

w

m /

[rs

p

o

= 40

&

o

L 30 {

n
o
e

 

042 W% NiF,

e

 

-
o

/i————-

o 25 50 75 100 125 150 175 200
NiF, ADDED (ing)

 

 

 

 

 

 

 

 

 

 

Fig. 2.2.9. Diffusion Current vs Nin Concen-
tration in 3304 g Of NGF‘KF'LIF (1105‘42'4605
mole %) at 550°C,

The haif-wave potential for the reaction
Fet**— Fe** + e
was 0.4 v. The change in diffusion current upon
the addition of FeF; to 26.6 g of NaF-KF-LiF
euvtectic at 550°C is shown in Fig. 2.2.10. The
NiF, diffusion current remained very low during
thls process until saturation was appreached, and

then it increased.rapidly, probably because of the
reaction

Ni +2FeF 45— 2FeF, + NiF,,  F° = =24 keal .

Apparently - the dissolved FeF, was so highly
complexed that reaction could not occur in spite
of the favorable free energy, but the presence of
excess FeF, changed the situation. Saturation
occurred at 0.19 wt % FeF,.

Much time was devoted to an unsuccessful
attempt to obtain reproducible results for NiF
in NaF-ZrF ;. The failure may be related to the
peculiar behcv:or reported above for NlF and
nickel.

116

ORNL—~LR -~DWG 16161
. 90 —— - — e /
80 : — /
70

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

s ~
Feo|— : ] _
% - - Fe>* DIFFUSION CURRENT‘/ -
g 50 —— ‘
o ’ .
5 ) .
8 40
W
a
30 ' /
0.19 wt % ) /
| /
20 /
/I 4/' ‘
10 /' i+* oiFFusion 7
CURRENT -4
' .
N
0 o —
0 10 20 30 40 50 60 70

FeF, ADDED (mg)

Fig. 2.2.10. Diffusion Currents Obhinéd from
Adding FeF3 to 26,6 g of NaF- KF-LIF (11.5-42-

4605 mole %) at SSOOCO

ACTIVITIES FROM EMF MEASUREMENTS
- ON SOLID SOLUTIONS OF SALTS

M. B. Panish

Measurements of activities in the system AgCl-
NaCl by an emf method have been of interest both
as a trial study with a solid elecfrolyte of o
technique which may be applicable ‘to fluoride
systems and also as o demonstration of some
metastable solid solution behavior. The cell being
investigated was comprised of ‘Agf% and CI'J
electrodes, with AgCi-NaCl as the elecfrolyte,
so that the cell reaction was the formation of
AgCl from the elements. The activity of AgCl
was determined from the emf by comparison with
pure AgCl. The eiectrolyte concentration was
varied from 3 mole % to pure AgCl, and the temper-
ature range was extended from the liquid into the

L
 

 

 

I.

 

solid region, that is, to temperatures as low as

250°C.

- The feasibility of makmg equulsbrlum mecsure--_'
ments with a solid electrolyte, as reported by
. earlier workers,’? 'was demonstrated. This is

probably easier’ with the Ag* ion.in a chloride
system, but, with a_ svitable amplifier, measure-
ments on solid fluorides might also be feasible.
If so;, a ceramic container’ would no longer be.a

problem for cells containing fluorides, measure-
ments could be made on solids by using cells
- designed to cvord ‘transference potentials, and
- compositions - with™ mconvenlently hlgh meltmg

pomts could be studied.”
“In the liquid state, the AgCI-NcCI system ‘showed

a slight posifive deviafion from’ ideality. From the

phase dragrum, a continuous series of solid so-

Jutions was expected below the solidus. However,
the activity of AgCl in the solid so!uhon was not
a monotonic function of concentration, The AgCl-

rlch soluhons showed marked posmve devrahon

 

.lqu Wachter, J. Am. -Cbem. Soc. 54,'919 (1932).

 

   

  

PERIOD ENDING SEPTEMBER 10, 1956

‘from Raoult's law and gave an activity vs concen-
-tration curve which rose well above, and pre-

sumably did not connect with, the curve for NaCl-

rich' compositions. The NaClerich compositions

showed - a.slight pegative deviation, 'if any, from
Raoult’s law. There was apparently a region in the
middle where the two curves overlapped without

_ ‘cbrmecfing, however, the ‘experimental uncertainty
_in this region was large. Obviously the solution

wrfh the higher AgCl activity was unstable with
respect to the lower one, but no transition from

‘the higher curve to the lower was noted. Also an

achwty-concenfraflon relcmon of thls type requires

‘that * the 'stable condition is separation into
‘c.oniugate solutions; no evidence of such phase

separation was found by petrographic or x-ray

fe_xaminotions.' The \m‘etostable_' condition was
- praduced by quenching from the liquid and was

maintained, in spite ‘of annealing, because of the
slowness of diffusion in the solid. These findings

“agree qualitatively with those found for the

same solid solution by Wachter,1? who worked at a

lower temperature and used different electrodes.

117

 
 

 

 

 

 

ANP PROJECT PROGRESS REPORT

2.3, PHYSICAL PROPERTIES OF MOLTEN MATERIALS

F. F. Blankensh i-p

SURFACE TENSION AND DENSITY
OF NaF-ZrF, MIXTURES.

F. W. Miles

Attempts to apply the maximum bubble-pressure
method! to determinations of surface tensions of
NaF-ZrF mixtures have been accompanied by
dlfficulfles caused by partial plugging of the
capillary tips. Therefore the values that have
been obtained are not considered to be sufficiently
accurate to report. Additional equipment is being
added to the gas-purification train, and the bubble-
pressure apparutus is being simplified in an effort
fo minimize contamination of the system and
consequent pluggmg of the tips.

The results of several densuty determmutlons
on NaF-ZrF (53-47 mole %) are now available.
The density data (in g/ml) are summarized below:

 

At 600°C At700°Cc At 800°C
3.049 2.916 2.805
3.014 2,948 2.813
3.051 2.895 2.821

- 3.039 2.946 2.833
3.003
3.020 o T

Av 3,03 % 0.02 2.93 T 0.02 2.82 t 0.01

The average values may be represented by the
equation

d = 3651 - (1.4 x 10~3) T ,

where T is temperature in °C.

The data were obtained on several different
batches of NaF-ZrF mixture of the same compo-
sition. |t is believed, however, that the density
values are probably correct to within 2%. The
calculated densities? corresponding to this compo-
sition are higher by approximately 5% than the

 

VE. W. Miles, ANP Quar. Prog. Rep. March 10, 1956,
ORNL-2061, p 106.

25, |, Cohen and T. N, Jones, A Summary of Density
Measurements on Molten Fluoride Mixtures and a Corre-
lation Useful for Predicting Densities of Fluoride
Mgu:tures of Known Compositions, ORNL-1702 (May 14,
1954).

118

G. M. Watson

average values listed, Efforts are being made to
improve the precision of the density ond surfuce
tension measurements. :

SURFACE TENSIONS OF MOLTEN SALTS
S. Langer
Flucrozirconate Mixtures

Additional measurements of the surface tensions
of NaF-ZrF, (53-47 mole %) mixtures have been
made by USmg the sessile-drop technique, pre-
viously described.3+4  Recent experiments at
600°C tend to substantiate the results reported

previously. Some data have been obtained.at

700°C, and o preliminary experiment at 800°C
has been run. The data accumulated to date are
summarized in Table 2.3.1. Data from the previous
report4 are included and corrected by the use of
more precise densities calculated from the equation

d = 3.651 - (1.04 x 10-3) T(°C) ,

reported above.

The “‘best’’ values for the surface tension of
these mixtures are 127 * 6 dynes/cm at 600°C
and 119 t 4 dynes/cm at 700°C, where the error
range is the root-mean-square deviation. It should
be noted that there appears to be a small difference
(within experimental error) between the data
obtained in helium and in hydrogen atmospheres at
600°C. The average surface tension in helium is
128 dynes/cm, while the average in hydrogen is
122 dynes/cm. More experimental work will be
needed to demonstrate the effect of the atmosphere.

The first evidence of a large change of surface
tension with composition is shown by the data
at 800°C. The large loss of weight due to vapori-
zation of ZrF‘ from the sessile drop and the
concomitant change of composition probably ex-
plain the increase in the surface tension of the
melt. Experiments designed to study the effects
of change of composition as a function of temper-
ature, pressure of the confining gas, and time of

equilibration are under way. The results will

 

3S. Langer, ANP Quar., Prog. Rep. March 10, 1956,
ORNL-2061, p 105.

45, Langer, ANP Quar. Prog. Rep. June 10, 1956,
ORNL<2106, p 118-

v
 

U D et i My e St Pt L S ek

L]

AW

o b

 

PERIOD ENDING SEPTEMBER 10, 1956

TABLE 2.3.1. SURFACE TENSIONS OF NuF-ZrF (53-47 mole %) MIXTURES CBTAINED
. BY THE SESSILE-DROP TECHNIQUE

 

Temperature Sompie Weight Pressure Weight Loss

b

Compositidn Nurfiber of -Surface Tension

 

(°C) )] ~ {mm Hg) {%) - (mole % Z?F4)"' .Photographs Taken (dynes/em)
600 04502 240 3 46,0 6 132 £ 6
0307 250 3.7 458 6 125 £ 4.3
05978 . 410 ) 2 126 £ 4.6
03154 . 350 43 45.6 6 122 1 2,6°
700 07758 260 173 | 40.8 5 118 1 0.8
04632 . 51s 129 400 4 16145
- 04487 515 9.0 . 414 4 122t 1.4
S sIs 0 1S 4l 3 123 1 0.6
800 - 03513 530 - 354 29.8 s 138 1 7

 

) aCalculated from the weight loss of the scmple by assummg thot only ZrF4 is volatilized.

_ ‘ation,

'bSumples meusured on. graphite p!oques under a hellum otmosphere. The error range is the root-mean-square devi-

cThls #cmpie was meusured under hydrogen ruther tl'mn helium.

permlt the estlmahon of -the surface fensions over
‘@ range of composmons from | smgle sample. :

o " In the prevuous reports the penetrahon of APC.
- graphite by NaF-ZrF (53-47 mole %) mixtures was
described. This behavtor is surprising in view of
- the nonwemng properties -of this mixture on C-18.

. ‘;grqphlte, as evidenced by the surface-tensson :

meosurements reported here, Several. prellmmary

‘ "expenments in which APC grcphlte was used as
the -supporting plaque have therefore been carried
- out,. While the’ experimental conditions are some-
. what - different : from those of the penetruhon'i -
. “studies, the sessile drops do not appear to have
" either wet ‘or penetrated the plaque, - The reason -
“for the discrepancy between these data and the
_prev:ous observahons is not mmedmtely upparent. '

 

SH. J. Botram and G, F. Schenck, ANP Quar. Prog.

' Rep. ]une 10, 1956. 0RNL~2106, P 'I25o L

' Uraniurh Téfraflucride

Several preliminary experiments have been

. carried out to study the wetting properties and

‘surface tensions of UF, and of UF, with small
. additions of vo,.
- viously® indicated that melts containing 4%, or
- _more, UO would penetrate graphite. This obser-
* vation wus borne out by preliminary observations
_of the surface tension. No values for the surface
~ tensions are available as yet, but the preliminary
'expen_ments have shown that, even though UF
“with about 1% UO, did not immediately wet C-18

Visual observation had pre-

or APC. graphlte, after o period of time, during

- .which small amounts of air leaked into the system,
" the contact gngle receded and the UF 4-UO, melt

started. t6 wet the plaques. These studies will

- be contlnued

 

6R, J. Sheil, ANP Quar. Prog. Rep. June 10, 1956,

'ORNL-ztoa. P 91. -

119

 
 

 

 

ANP PROJECT PROGRESS REPORT

2.4, PRODUCTION OF PURiFIEDM FLUORIDE MIXTURES

G. J. Nessle
LABORATORY-SCALE PURIFICATION
OPERATIONS
W. T. Ward

The standard hydroflucrination-hydrogenation
process was used, with appropriate meodifications
to laboratory-scale equipment, to prepare a number
of especially pure flucride mixtures requested for
various' reseurch programs, Several of these
compositions were dispensed directly into con-
tainers furnished by the requester to minimize
atmospheric-contamination and handling.

QUALITY CONTROL OF RAW MATERIALS
AND PRODUCTS

W. T. Ward

. The processing characteristics of an NaF-ZrF
mixture supplied by a commercial vendor were
determined, and the product was found to meet
specifications. Samples of all production batches
were obtained for petrographic and x-ray diffraction
analysis to assure quality control.

PILOT-SCALE PURIFICATION OPERATIONS

C. R. Croft J. Truitt
J. P. Blakely

The pllot-scoie purification facility processed
53 batches totaling 835 Ib of various fluoride
compositions for use in smallsscale corrosion
testing, phase equilibrivm studies, and physical
property studies. The demand for special compo-
sitions continved at a fairly constant level, and
thus efficient operation and maintenance of these
facilities at about 30% capacity was possuble
without accumulatmg a backloeg.

The use of copper-lined stainless steel reactor
vessels in these small size units has proved
successful in all but one case. Mixtures con-
taining NaF, LiF, KF, and UF cannot be frozen
in the copper liners and then remelted In three
attempts to process NaF-KF-LiF- UF (11.2-41-
45.32.5 mole %) the copper liners ruptured when
the melt was frozen and then remelted in the
reactor vessel. This means that any full-scale
operation of the 250-1b facility for the preparation
of such mixtures would necessitate a 24-hr-day
7-day-week basis of operation so that the mixture
could be maintained molten at all times.

120

G. M. Watson

PRODUCTION-SCALE OPERATIONS
J. E. Eorgan J. P. Blakely

The production-scale facility processed 19
batches totaling about 4550 1b of purified fluoride
melts. The use of copper-lined stainless steel
reactor vessels continued to prove very satis-
factory in this facility. One reactor vessel has
now produced sixteen 250-Ib batches, and it. shows

" no signs of deterioration. Product quality has

remained excellent throughout the equipment life;
however, some difficulties have been encountered
because of the high oxygen and water content of

the poor-grade hydrogen now being supplied in

this area. This difficulty has apparently been
overcome by installation of catalyst trains and

- cold traps of higher capacity than previously used

on the hydrogen purification system,

The production facility was shut down during
July because the relatively low consumption rate
resulted in a shortage of receivers for storage of
purified product. The present disposition of sixty-
four 250-Ib containers is the following: -

Held by Pratt & Whitney 23 |

To be shipped to Pratt & Whitney (August) - 8
ART High-Temperature Critical Expenmenf _' 3
storage }
Stock inventory (full) ' 20
Ernpt} containers on hand _ 10
Total . : &4

Forty new 250-1b storage containers are presently
on order. Delivery of these containers is expected
to begin in September, and reactivation of the
production facility is scheduled accordingly. It
is extremely unlikely, however, that the shortage
of such containers will be alleviated before
January. Accordingly production of purified ma-
terials will probably be 8,000 to 10,000 Ib behind
presently scheduled demands by the end of this
calendar year.

An order for 30,000 Ib of NeZrF, has been
placed with a commercial vendor. It is estimated
that this amount should be sufficient to maintain
operation for FY 1957. Delivery of 4000 lb of
this material per month is expected to begin in

. w
 

 

ne

August 1956. Each shipment will be analyzed
thoroughly to ascertain that specifications of purity
have been met. : '

BATCHING AND DISPENSING OPERATIONS
F. A. Doss - J. P. Blakely
The batching and dispensing facility dispensed
128 batches totaling approximately 8025 Ib of
processed fluorides in batch sizes ranging from

-1 to 250 Ib.. This is a sizeable increase over the

amount normally handled because of large ship-

~ments to Pratt & Whitney. As a result, the stock
inventory of processed fluorides decreased ap-

preciably. A material balonce for the quarter is
given in Table 2.4,1.

The main consumers of the processed fluoride
mixtures and their allotments during the quarter
are given below: -

- ORNL-ANP groups

For chemical and physical properties - 258 1b
- studies :

For experimental engineering tests _ 2680
For metallurgical studies and fuel 593

reprocessing development

Pratt & Whitney Aircraft 3207

Other cofiirccfors, including BMI, NRL, 200
and WADC

Salvage and reprocessing 1087

Total _ 8025 b

PERIOD ENDING SEPTEMBER 10, 1956

PREPARATION OF ZrF‘ FROM Zl'(:l4
J. E. Eorgan J. P. Blakely

A dust filter for the ZrCl -conversion unit is
presently being installed. tonversion of some
2000 Ib of low-hafnium ZrCI4 to ZrF, will begin
as soon as installation of the filter is complete.
It is hoped that the new dust filter will prevent
large-scale loss of the product during the con-
version and permit efficient operation. The con-
version unit will be operated on a 24-hour-day
five-day-week basis until the present supply of

~ZrCl, is all converted to ZrF,. The recent

demand for 1200 Ib of low-hafium NaF-ZrF,
(50-50 mole %) for the Pratt & Whitney high-
temperature critical test has made it imperative
that this unit be operated as soon as possible.

In order to meet known demands in FY 1957 for
fluoride compositions requiring low-hafnium ZrF ,
an order for 7000 Ib of low-hafnium ZrCI4 has
recently been initiated. At least half this order
has been requested for delivery before December
1956, and the remainder is to be delivered some-

time in the last half of FY 1957.

SPECIAL SERVICES

F. A. Doss J. P. Blakely
J. Truitt N. V. Smith

Filling, Draining, and Sampling Operations

The filling, draining, and sampling operations
were at a normal level during the quarter. Ap-
proximately 2700 |b of processed fluorides and
1000 b of liquid metals were used to charge

TABLE 2.4.1. MATERIAL BALANCE FOR FLUORIDE MIXTURE PRODUCTION AND USE

 

 Material (Ib) -

 

 

 

 

 

. " - Mixture No. 30,  Mixture No. 31, Mixture No. 108, . Mixture No. 71, Special

" NoF-ZF (UF, - NaFZiF, NaF-ZrF,UF, - NaF-ZrF pPecial ol

(50464 mole %) (50-50 mole'%)  (56-37.56.5 mole %)  (54.145.9 male %)
Onhand at beginning of 595 2,073 | 1,302 9,325

quarter "~ T A | |

Produced dur-ing:iqq‘m_e'r' | 655 0 2673 1232° 835 5395
Tewl - eg0s . 2,07 e wm o 20w 4o
Dispensed duringquarter - 4,161~ 833 1679 242 - 1,10 8,025
Onhand of end of quarter 2,444 1,240 994 990 1,027 6,69

 

12]

 
 

 

 

ANP PROJECT PROGRESS REPORT

engineering tests in charge sizes ranging from
1 to 500 Ib. Electric power has been installed
at the liquid metals disposal quarry to facilitate
the disposal of bulk quantities of NaK and Na.

Special Enriched Fuel for In-Pile Loops

Inpile loop No. 5 failed to operate at the MTR
because of inability to move the fuel from the sump
tank to the pump (see Chap. 1.6, ‘““‘In-Pile Loop
Development and Tests"). Later investigations
showed that the fill line between the pump and
the sump tank was plugged. This plug was found
to be nearly 50% ZrO s and subsequent analysis
of the original batch showed evidence of oxides
and oxyfluorides. Examination of two other unused
batches showed oxides and oxyfluorides to be
present in undesirable quantities. As a result
- of these findings, a new batch of the fuel mixture
(No. 44) NoF-ZtF, -UF, (53.5-40-6.5 mole %) was
processed with extreme care for use in in-pile
loop No. 6. The batren carrier was prepared and
processed first to determine whether oxides of
zirconium were persisting in the batch. Since
the carrier was found to be oxide-free, the U235.
enriched UF, was added to the batch, and the
processing was repeated. When the batch was
shown fo be oxide-free, a fuel charge was trans-
ferred from the batch into a loading can. A sample
was token during this operation to reverify the
oxide purity of the charge. The charge was then
loaded into the loop, with a sample being taken
as the charge moved from the loading can into
the loop. This sample was also verified as being
oxide-free. Therefore, it can be assumed that
the charge that went into the sump tank of in-pile
loop No. 6 was in satisfactory condition.

Spectrographic anclysis of the plug in the fill
line of in-pile loop No. 5 revealed sufficient
amounts of foreign impurities, such as aluminum,
silicon, calcium, and zinc, for there to be
reasonable doubt as to whether the oxide content
of the original charge was the actual cause of the
plug formation. However, to remove at least one
uncertainty in future loop operation, extreme care
will be taken to assure the delivery of pure
material into the loop sump tanks.

Shield Mockup Core Materials

Fabrication of two *‘orange slices’ which simu-
late sections of the SMC has been completed (see
Chap. 5.4, “*Shield Mockup Core’). Attempts will

122

be made to pour into these sections the molten
salt NaF-ZrF -UF ,-KF (61.7-16.4-1.4-20.5 mole %),
which wiil be used to simulate the fuel and NaK
coolant in the heat exchanger region. Critical
measurements of the ‘‘orange slices’ have been
made and any changes in these dimensions as a
result of the pouring operation will be noted. One
section is fabricated from l/‘-ln type 310 stainless
steel plate and the other from é-m type 310
stainless steel plate. After pouring and cooling,
the **orange slices’ will be sectioned thoroughly
to determine the possibility of void formation in
the frozen salt and whether any segregation of
the uranium occurred in the freezing process,

EXPERIMENTAL PREPARATION OF VARIOUS
FLUORIDES

B. J. Sturm L. G. Overholser

Additional quantities of the several structural
metal fluorides have been prepared by the methods
employed previously. Increasing interest in the
use of rare-earth fluorides for various experimental
purposes required the preparation of several such
fluorides in substantial quantities. Continuved use
has been made of chemical, x-ray, and petrographic
examinations to establish the identity and purity
of the products.

Several batches of CrF, were prepared by
hydrogen reduction of (NH4)3CrF6, and additional
N:F was obtained by hydroflucrination of NiCl,.
Three pounds of K,FeF, was synthesized from
aqueous solutions of FeCf and KF. Four pounds
of CeF; was prepared by the interaction of an
aqueous HF solution with Ce,(CO,),, followed
by washing and drying. A small batch of YF,
was prepared in a similar manner, with YCI, bemg
used as the starting material. Approximately
‘é Ib of NdF, was synthesized from Nd,(CO,)

by using an aqueous HF solution. Also a smafl'

batch of PrF_ was prepared by the same general
method by starting with PrCl,. A small quantity
of K,TeF, was synthesized by the addition of
aqueous K% to an aqueous mixture of TeO, and
HF. A sample of CaF, was treated wifl'\'NH‘F HF
and ignited at 1125°C. A somple of K, ZrF was
freed of oxide by hydrofluorination at 780°C
Recently, work was started on the preparation of
molybdenum fluorides by using the reduction of
MoF and separation of the reduction products by
condensatlon at various temperatures.

'y
 

IR 1 a4 . -+ s S bt e . sl g i

o

"

PERIOD ENDING SEPTEMBER 10, 1956

2.5, COMPATIBILITY OF MATERIALS AT HIGH TEMPERATURE
F. Kertesz

PENETRATION OF GRAPHITE BY
MOLTEN FLUORIDES

H. J. Buttram - G. F. Schenck!
The study of the behavior of graphite exposed to

molten fluoride fuel mixtures was continved. Pre-

viously reported preliminary results? showed that
APC graphite was completely penetrated by NaF-
ZrF, (53-47 mole %) in 1 hr ot 600°C, while
NaF-ZrF 4-UF, (53.5-40-6.5 mole %) had not de-
tectably penetrated the .graphite after 10 hr. In an
attempt to investigate this behavior more thoroughly,
an alpha counter was employed to follow the dif-
fusion of the urdnium-fluoride-confaining mixture
into the graphite. After exposure of 1’ X / x ¥ in,
graphite specimens to the fluoride mlxture at 606°C
counts were taken on various samples obtained by
filing and scraping the surfaces in @ .uniform
manner, ' : R

The petrographic microscope was used for the
examination of samples, as in the previously
described experiment,? and it was found that after
10 hr of exposure to NaF-ZrF .UF, (53.5-40-6.5
mole %) there was no evidence of penetrahon into
the graphite specimens. After 50 and 250 hr, only
slight increases in alpha count were found, while
no penetration could be detected by microscopic

examination, A graphite sample was then exposed -

to the NaF-ZrF -UF, mixture at 800°C and it was
observed that after 10 hr there was a sufficient
increase in the alpha count to indicate the presence
of the su!f msude the tesf specimen. o _

HYDROGEN PRESSURE OF THE
NaOH-NI REACTION o

H J. Bufl’ram 5"' F. A. Knox

| The equ:llbrwm pressure of hydrogen from the
'reachon of NaOH and mckel was. studled earher in -
Can: apporatus in which a manometer . was connected :
to a quartz. envelope contommg the NaOH ‘in‘a
metal - crucible,-- More recently an: opparatus was

developed ‘which. ‘permits - measurements of the

f.pressure inside a -sealed quartz ‘envelope - con-_'
_tummg the NaOH-Nl system. The measurements

 

lOn oks.sierrm_lent /fro-m' Pratt & Whitne} Aircfuh.
2H. Buttram and G. F. Schenck, ANP Quar. Prog. Rep.
June 10, 1956, ORNL-2106, p 125

can be made after the apparatus has cooled to room
temperature.

In a series of experlments, 10-g somples of NaOH
were sealed in l/.‘,,x 3 in, tubes of nickel, and each
tube was sealed in a quartz jacket. Three sizes
of quartz jackets, approximately 25, 42, and 73 ml,
respectively, were used. The specimens were
soaked at 800°C for periods of 3 to 729 hr, and
after cooling to room temperature the pressure of
H, in each was determined by breaking the quartz
jacket inside a calibrated volume in which the
pressure could be ascertained.

The quantities of hydrogen recovered from the
capsules are shown in Table 2.5.1, and the pres-
sure (extrapolated to 800°C) of hydrogen in each
capsule is plotted in Fig. 2.5.1. It is apparent

~ TABLE 2.5.1. HYDROGEN COLLECTED FROM
REACTION OF NaOH WITH Ni® AT 800°C

 

Hydrogen Collected {cm3+atm)

 

 

 

 

Exposure
Time From 25ml From 42-ml From 73-ml
(hr) Capsule Capsule Capsule
3 11.3 16.2 21.8
27 37.2 42,0 50.6
81 47.5 54,5 56.0
243 23.0 32.4 47.4
79 0.61 'I.l_ 3.8
. . chmsétfleo
.- ORNL-LR~DWE 16162
O
o
2 1600
5
~B 1200 :
£ o0
] ..
5
&% 400
'
&
0 : = _
S 10 C 100 ' 1000

TIME (hr}

" Fig. 2.5.1. -»Hicliregen".Pres;ures Developed in

Quartz-Jacketed Nickel Capsules Containing
NaOH After Various Periods of Exposure at 800°C,

123

 
 

 

 

 

 

ANP PROJECT PROGRESS REPORT

from the data that in these experiments the hydrogen
pressure, as calculated, ond the amount of hydrogen
recovered passes through a relatively sharp maxi-
mum at cbout 100 hr and drops to a negligible
value. The amounts of hydrogen recovered seem
to be somewhat dependent on the volume of the
system. ‘ :
. These results must be considered somewhat sur-
prising in that they indicate that after about 100 hr
the rate of evolution of hydrogen from the system
become slow with respect to the rate of loss of
hydrogen by diffusion from the quartz envelope.
Whether this phenomenon is due to @ much reduced
rate of reaction or whether the character of the
reaction changes abruptly is not known at present,
Further studies of this situation will be conducted.

- PHYSICAL PROPERTIES OF ELASTOMERS
EXPOSED TO ATTACK BY LIQUID METALS

D. Zucker

Equipment has been assembled for the testing of
elastomers exposed to NaK. In the tests, fresh
samples of various elastomer materials are heated
slowly under tension to 185 to 190°C in NoK in a
dry box coentaining a high-purity helium atmos-
phere. Two samples of G-E silicone 81576 failed
below 165°C, both samples of silicone 81577
failed shortly after reaching 180 to 185°C, ond one
sample of silicone 81578 failed cfter a few hours
at 185 to 190°C. A second specimen of silicone
81578 was still intact after 25 days of exposure.
Two of three Du Pont materials showed promise;
sample SR-5550 failed at 150 to 160°C, but SR-5570
remained intact for nine days at 185 to 190°C,
Sample 5806, a silicone-resin-treated fabric, was
still intact after 25 days.

124

DETERMINATION OF OXY GEN IN NaK
Eo Eo KetChen Go Fo SChean

A source of NaK of known oxygen content is
needed in connection with studies on the solu-
bility of structural metals in NoK, and thus «
reliable and practical method for analyzing for
oxygen is required. A study has been made of
oxygen determinations by the amalgam and the
butyl bromide methods in an attempt to evaluate
the methods and establish whether or not any corre-
lation exists between the results obtained. The
amalgam method, in which NaK is removed from the
oxides by mercury, appeared to be the most adapt-
able, since it could be carried cut in the vacuum
drybox at the time samples of NaK were loaded into
the capsules used in the solubility studies. In
this method, samples of NaK which had been passed

through a 10-p sintered-glass filter, as well as

samples from the same lot that had not been
filtered, were analyzed. The results showed about
10 and 25 ppm of oxygen for the filtered and un-

filtered samples, respectively, The precision was:

good in all cases. To test the butyl bromide
method, samples of the filtered NaK were removed
from the drybox and reacted with dry, purified
n-butyl bromide in a glass vacuum system. When
the water extracts were titrated with 0.1 N HCI to
a pH of 6.5, values of 80 and 130 ppm of oxygen
were obtained. Thus the values obtained by the
butyl bromide method are completely out of line
with those obtained by the amalgam method.” It was
observed, however, that the aqueous extracts
behaved as if they were buffered, and consequently
these results cannot be considered at all reliable.
The cause of this buffering action is being studied
(see Chap. 2.6, *‘Analytical Chemistry’’),

*
 

 

 

 

W

“aw

PERrOD'ENDlNG SEPTEMBER 10, 1956

2.6, ANALYTICAL CHEMISTRY
J. C White

DETERMINATION OF RARE-EARTH

ELEMENTS:IN FLUORIDE FUELS

A. S. Meyer, Jr. B. L. McDowell -
The procedure for the determlnoflon of lanthanum

in samples of NoF-ZrF +UF, by precipitation as
the oxalate ! has been opphed with good precision

~.dnd accuracy to the determination’ of lanthanum-in
virtually - pure LoFa. “The same . procedure, how-
‘ever, did not yield very precise or accurate resylts-
- for the determination of cerium in samples of CeF,,
- or ‘in NaF-ZrF; -UF ~that ‘contained CeF,, Ac-

cordingly, spectrophotometnc methods for the de-

termination of cerium are being investigated. The
“procedure for the determination of fanthanum is, as -
reported earlier, } quite lengthy, and therefore spec--
~trophotometric methods for the defermmoflon of :
lanthanum are also. being studied. '

Speetrophotometric Deiermlnoflon of Cerium -
A spectrophofomefnc merhod2 for the determi-

nation  of cerium with Tiron (dlsodium-'l 2-drhy-‘
. droxybenzene-3, 5-d|sulfonote) was opphed o the
- - determination of ¢erium in samples of NoF-LlF-KF '
~and ‘NaF-ZrF, ‘4 ‘which contained CeF " The re-.
- action. of cerium with Tiron yields a colored com-
plex which exhibits an -~ absorption - moxlmum at
500 mp: This reaction is selective for cerium in
- that colored reaction products are. not formed with-
~any ‘other lanthanide element,2- ‘The .color is de-
- veloped in solutions.of high ionic strength at a pH
~_greater. than 8. ‘Although the .color. developmenr is
" instantanecus with cerium(lV), ot least 6 hr is: -
required __for complete " color - development with. -
- cerium(lIl), " For - soluhons of cerium(1il) su!fote,; o
- the coefficient of variation of this method is less. ,;_._ferences.
. - than 2% for the ronge of cerium concentrohon of - '
. 2to 50 pg/ml. - I, : 1
" Several . elements thot are normoily present in
'_"ffiuorlde foels” mterfere with  this determmation.'
-~ lron"and uranium react with Tiren to form colored*
'f,_;products that obsorb in this region.- erconium o
. reacts to" form a colorless comp!ex ond ‘thereby
";r'consumes reogent, whale fluorlde destroys the 5

 

TA. 5. Meyer; Jr., and B. L. McDowell, ANP Quar.

Prog. Rep. June 10, 1956, ORNL-2106, p 126.
2B, Sarma, J. Sci. Ind. Research 14B, 538 (1955),

cerlum(lll) Tlron product. Fluoride is normally
removed by volatilization during the dissolution of
the somple with sulfuric acid. The metallic inter-
ferences can be removed by extraction with tri-
octylphosphine oxide (TOPO) (see subsequent sec-
tion, ‘‘Extraction of ART Fuel Components with
Trioetytlphosphirie Oxide,’’ this chapter).

- Spectrophotometric Determination of Lenthanum
 The method of Rinehart3 for the spectrophoto-

B 'ihetric:deferminofion of the rare earths and yttrium
‘with sodium glizarin sulfonate (alizarin red-S) is

being investigated. The complex which is formed

by the reaction of alizarin red-S and lanthanum in

an acetate-buffered solution exhibits an absorption

-maximum “at 535 mu. This system conforms to
“Beer's law for concentrations of lanthanum from
21020 pg/ml. For solutions of lanthanum chloride
" the coefficient of variation of the method is 2%.

" All the rare-earth elements form complexes which

- possess similar molar absorbance indices. There-

fore, samples which contain both lanthanum and
cerium can be analyzed by determining the total

moler concentration of rare-earth elements by the

alizarin red-S method, determining the cerium by
the Tiron method, ond cnlculatmg the lanthanum by

_ dlfference.‘ o

Most ions which would normally occur in sulfate

" solutions of the samples of NaF-ZrF ,-UF, and

rare-earth elements interfere serlously wrth this

‘method. Some of these interfering ions are UOz'H

Fe***, AI***, Z:0**, F~, and 50,=~. The ex-
troctlon of the solutlon ‘with TOPO is being tested

":cs a method for fhe ehmmoflon of these inter-

- _‘-oErékulunlofl_oF TANTALUM
IN FLUORIDE SALTS
JP. Young J. R, French

- The determination ‘of tantalum in NoF-ZrF ‘-UF

was studied further in order to ascertain the mini-

“mum concentration of tantalum that can be found by
- the present method, The determination of tantalum

in the range 900 to 1500 ppm in fluoride solts was

 

3R. W. Rinehart, Anal. Chem. 26, 1820 (1954).

125

 
 

 

 

ANP PROJECT PROGRESS REPORT

discussed previously.4¢5 The sample of approxi- .

mately 1 g of salt is carefully digested with dilute
H,SO, in order to hydrolyze any TaF to Ta 05,
then the solution is evaporated to. dryness. The
residue is fused with potassium pyrosulfate, fol-
lowing which the melt is dissolved in a solution of
ammonium oxalate. If the sample consists of alkali
fluorides only, the tantalum is determined.colori-
metrically in - the solution of ammonium oxalate
with pyrogallol.8 The absorbance of the complex
is measured ot 330 mu. If the sample contains
either ZtF, or UF,, the tantalum is precipitated
with tcmnm at a pH of 5 in a hot solution,. The
precipitate is filtered, washed, and then ignited to
Ta,04. This oxide is fused with potassium pyro-
sulfate, the melt is dissolved in ammonium oxalate,
and the tantalum is determined with pyrogallol.
An investigation was made of the minimum con-
centration of tantalum that could be recovered from
synthetic samples by means of the tannin precipi-
tation. It was found that approximately 100 ug of
tantalum was lost in the precipitation step. This
amount of tantalum appeared to be fairly constant
for the separation of tantalum from vranium, zirco-
nium, or a mixture of these two elements., The
separation of tantalum from uranium or zirconium
seems to be practical only for amounts of tantalum
greater than 1000 ug.

DETERMINATION OF OXYGEN IN NaK
A. S. Meyer, Jr. T. W. Gilbert, Jr.

A comparison of the methods for the determina-
tion of oxygen in highly purified NaK has shown
that significantly higher oxide concentrations are
indicated when the butyl bromide method? is used
‘than when the amalgamation procedure® is used
(see Chap. 2.5, *“Compatibility of ngh—Temperuture
‘Materials®’). It was noted that the usual strong-
base—strong-acid titration curves were obtained
when the oxide residues from the amalgamations

 

4. c White, - Determination of Small Amounts of
Tantalum in NaF-LiF<KF and In NaF-L:F-KF-UF4,
ORNL CF-56-1-49 (Jan. 10, 1956). -

5). P. Young and J. R. French ANP Quar. Prog. Rep.
Marcb 10, 1956, ORNL-2061, p 211

J I Dmmn, Anal, Chem. 25, 1803 (1953).

73, C. White, W. J. Ross, and R. Rowan, Jr., Anal
Chem. 26, 210 (1954).

8. p. Pepkownz and W. C. Judd, Anal Cbem. 22,
1283 (1950).

126

- were titrated and that a buffering action occurred

in the titration when the NaK was reacted with

’-bufyl bromide. Since no buffering was observed in

the titration of the oxide in the butyl bromide de-
termination. of oxygen in metallic sodium,’ an
investigation is being carried out to determine the
the source of the buffering action and to evaluate
such action as a source of error in the bufyl bro-
mide methed. :

. No ewdence was found for the formanon of weak
organic acids during the reaction of NaK with butyl
bromide, and it was therefore postulated that: the
buffering could be caused by the presence of traces
of metallic contaminants in the NaK. [f the me-
tallic ions formed weakly dissociated hydroxides,
they could produce the observed buffering effect.

Spectrographic analyses of NaK samples showed
gluminum and silicon to be the principal contami-
nants. The aluminum content was found to vary
from 7 to 20 ppm, and it is known that 20 ppm is
sufficient to cause buffering comparable to that
observed in the titration of oxygen in NaK.: The
aluminum cannot, however, account for the extenf
to which the oxide concentrations found by using
the butyl bromide methed are higher than those
obtained by using the amalgamation procedure. |

Experiments are being continued to determine
whether other contaminants which might contribute
to the high oxide titration are present in NaK.
Also, tests of the amalgamation procedure will be
carried out in metal apparatus which was designed
by ‘Nuclear Development Associates.!® The amal-
gam ‘is filtered through a 10-x filter in this appa-
ratus, and alkali metal oxide should be detected
which may be entrained in the amalgam durmg the
vsual anclytical procedure.

SAMPLER FOR ALKALI METALS

A device that was developed by the Mine Safety
Appliances Co.!! for obtaining samples ‘of alkali
metals was adapted for sampling dynamrc and
static systems at operating temperatures.  The
original sampler can be used only when a specml
chamber, cennected to a ~source of mert gos, IS

 

A, s, Meyer, Jr., G. Goldberg, and W."J. Ross, ANP
Quar. Prog. Rep. Dec. 10, 1954, ORNL-2012, p 186.

0. D. Sax, private communication to A. S. Meyer, Jr.

1R, E. Lee and S: L. Walters, Technique of Samplmg
and Analyzing Hot Flowing NaK Alloys, NP 1527, qu 1,
1950, Mine Safety Appliances Co.

.y
 

 

"

AS

incorporated in the part of the system from which

the sample is to be removed. The modified sam-

pler'2 can be attached at any point in a system
at which a Jamesbury valve can be connected to
provide access to the alkali metal. A Jamesbury
valve is a reliable, compact, vacuum-tight, ball.
type valve which is completely opened or closed,
by a one-quarter ‘turn of the handle. ' In the modi-

fied sampler, a glass-to-metal Wilson seal was

vsed in place of the original packlng-g!and seal

" to facilitate the compIete removal of oxygen from

the sompling opparatus, and a Teflon-packed
Jomesbury valve, as described above, was used in
place of the Teflon-packed gate valve.!3 The
modified sampler is similar to ¢ sampler used
earlier by the Metallurgy Division at ORNL,14

- A sample is obtained with the modified apparatus
by placing a nickel bucket in the sampler and then
connecting the sampler to the sampling port. The
sampler and the sampling port are freed of air by

~ clternately evacuating . and purging with dry,

oxygen-free helium. The bucket is lowered into

the alkali metal through the sampling port. The

filled bucket is . withdrawn and transferred to a
pyrex receiver, where it is sealed off under vacuum.
There is apparently negligible contamination of the
alkali metal during sempling. Concentrations of
oxygen as low as 20-ppm have been determined in
samples of sodium that were taken by using this
sampler,

‘ DETERMINAT'ON OF OXYGEN, NIfROGEN,
AND CARBON IN METALLIC LlTHlUM
. A, S. Meyer, Jr. :
R E Feafhers - TOW, G:Iberf
G Go|dberg

Sensmve methods for the determinahon of oxy-
gen,’ mtrogen, and carbon in metallic lithium are
- :bemg mveshgated since traces of these elements
" _are believed to -influence the rate of corrosion of
 structural metals by molten lithium. Results of the
determmahons w:ll be useful in: the search for

 

B ‘20 Goldberg, As. Meyer, Ir., and J. C. White, The
o Samplmg of Alkali Metal Systems with tbe Modz[aed MSA
Sampler, 0RNL-2147 (Aug. 21,:1956). '

13
Jametbury vaive, type D.22, ccrbon sl'eel und staln-
Iess steel, Jcmesbur:/ Co,, 62 Millbrook St., Worchester,
Mass.

e, E Hoffman, ANP Quar. Prog. Rep. Marcb 10,
1956, ORNL-2061, p 129, Fig. 5.20.

PERIOD ENDING SEPTEMBER 10, 1956

structural metals that are compatible with metallic

fithium at high temperatures.

Determinction of Oxygen

An apparatus for the determination of oxygen in
metallic lithium by the amalgamation method has
been assembled. In this apparatus, which was
designed by Nuclear Development Associates,!5
the sample of lithium is amalgamated in o heated
stainless steel reaction chamber in order to limit

- explosion hazards and to eliminate possible oxide

contamination of the lithium by contact with glass-
ware. The lithium oxide is separated from the
amalgam by filtration through a nickel micrometallic
filter of 10-u porosity, which is welded into the
bottom of the amalgamation vessel, After the oxide
is washed free of amalgam with mercury, it is
dissolved in water and titrated with a standardized
solution of acid. This titration must be corrected
by subtracting the volume of titrant that is required
to titrate the Li N and Li,C, which accompany the
oxide.

The apparatus, which is shown in Fig. 2.6.1,
was modified by the addition of a special sample
port to the amalgamation chamber, A 3/ -in. Jomes-
bury valve was welded to the upper opemng of the
amalgamation vessel, and the valve was fitted with
a 29/42 standard-foper joint fabricated from stain-.
less steel. When the joint is coupled to a sample
chamber that is olso fitted with a Jamesbury valve
and ¢ standard-taper joint, the intervening volume
can be evacuated and purged with an inert gas.
The sample can then be transferred without atmos-

- pheric contamination.

Investigations are also being carried out to find
o method for the determination of oxygen in me-
tallic - Iuthmm by titration of the oxide, after con-

- version of the metalllc lithium to a neutral lithium

halide. The butyl bromide method,” which was

ift':rund to be more convenient for the determination

of oxygen in sodium than the amalgamation pro-

.cedure, cannot be applied to the analysis of lithium,
" The reaction between lithium and n-butyl bromide
is too slow for analytical application and may
yield lithium alkyls which react with water to form

hydroxides. It was found that the lithium alkyls

“could be ‘eliminated effectively by the addition of
- an excess of elemental halogen. The lithium metal

 

. V5ANP Reactor Development Quqrterly Progress Re-
port, October 1, 1955 through December 31, 1955,
NDA-20 (Jan. 23, 1956) p 14.

127

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

UNCLASSIFIED
ORNL-LR- DWG 16163

 

! 29,5 STANDARD-TAPER
STAINLESS STEEL JOINT

 

 

 

 

A 3/ -in. JAMESBURY
/i VALVE

T0 %4 - in. BELLOWS VALVE @
ANT‘( RESERVOIR \

1/, -in. SWAGELOK > &
FITTING '

 

VESSEL

' Py
A
7 , I ] .—RrEACTION

 

 

3g-in, SWAGELOK
: FITTING
MICROMETALLIC FILTER

 

Figs 2.6.1. Reactor for the Determination of
Oxygen in Lithium by an Amalgamation Metheod,

is dissolved much more rapidly in the presence of
ether, in a manner analogous to the Grignard re-
action. The most rapid dissolution was accom-
plished with solutions of n-butyl iodide in diethyl
ether, lodine is added to the solution to destroy
organolithium compounds. Samples of lithium of
approximately 2 g will dissolve in '/2 hr. The
following reactions are postulated for the dissolu-
tion of lithium:

2Li +2C Hgl —> 2Lil + CH,,
2Li + CHyl +(GH;),0 —> LiCH,-(C,Hg),0 + Lil
LiC Hg(C,H,),0 + 1, —> Lil + C Hyl + (C,Hy),0

- The halogenation procedure may offer additional
advantages in that the free iodine is postulated to

128

oxidize lithium nitride and carbide and thus elimi-
nate their contribution to the alkalimetric titration.
Thus the oxide is determined by direct titration,
and the uncertainties inherent in the determination
by difference are eliminated. These postulated
oxidation reactions, which are thermodynamically
feasible, have not yet been demonstrated with pure
Li,N and Li,C,, but they are supported 'by an-
alyses of lithium samples in which the oxide
titration is significantly less than that required for
the titration of the nitrides and carbides.

In the procedure that is now being tested, a 2-g
sample of metallic lithium is added to a solution
that contains 100 ml of n-butyl iodide, 200 ml of
diethyl ether, and 15 g of iodine. After the dis-
solution of the sample is completed, the ether is
volatilized and the organic solvent is extracted
with water, The aqueous phase is titrated poten-

tiometrically with a 0.01 N solution of hydrochloric -

acid to a pH of 7. The iodine is removed from an'

aliquot of the titrated solution and the lithium

iodide in the aliquot is titrated with a solution of-.

silver nitrate to determine the sample weight.
Concentrations of oxygen in lithium as low as
200 ppm have been determined by this procedure.

Determination of Nitrogen and Carbon

The method proposed by Nuclear Development
Associates 15 for the determination of nitrogen and
carbon in metallic lithium is being evaluated. In
this procedure the sample is dissolved in water to

‘convert the nitrides and carbides into ammonia and

acetylene, respectively. Ammonia is absorbed from
the exit gases by a boric acid scrubber and is
determined spectrophotometrically. After the mois-
ture is removed, the acetylene is passed over
heated copper oxide to convert it to carbon dioxide,
which is absorbed on ascarite and weighed. A
modification of this procedure in which the acety-
lene is absorbed in a solution of AgNO, is being
tested as a possible means for improving the
sensitivity of the determination of carbide. The
acetylene is absorbed in a 35% solution of AgNO,,.
When the solution is diluted twentyfold, the acety-

lene is precipitated and weighed as the compound -
 Ag,C,-AgNO;. This procedure offers a fivefold

increase in the sensitivity of the determination and
is selective for acetylene.

.
 

 

L

tion of a radiator leak.

DETECTION OF TRACES OF NcK IN AIR
A. S. Meyer, Jr. J. P. Young

Two methods are being studied for the detection
of traces of NaK in air. Functional descriptions
of the methods and the operational requirements of
the necessary apparatus were presented previ-
ously. 16 The electronic and optical components of
the instrument designed for the photometric de-

tection of microgram quantities of NaK in air have

been fabricated. The detailed engineering drawings

of the instrument for the detection of submicrogram
quantities of NaK in air by observation -of the -

sodium resonance radiation have been prepared,

Experiments were' continued in an effort to find a
means for introducing reproducible concentrations

of sodium into air.'7 When helium which contained

20 to 100 pug of sodium vapor per liter was mixed

with 10 times its volume of air, at least 95% of the
sodium was removed from the air stream before it
had traversed a 3-in. length ‘of transfer line. The
stability of such suspensions of sodium oxide was
not improved by varying the tempercture of the
transfer line from 150 to 900°C or by altering the
design of the mixing chamber to increase the
velocities of the streams of helium and air. |

Since it was not possible to prepare mixtures of
sodium oxide in gir that could be transferred by a
jet of helium, @ small test facility will be fabri-
cated which will more closely simulate the condi-

constructed for the ejection of minute quantities of

NaK at a metered rate. This jet of NaK will be
injected directly into an air duct, The air will be
passed through the duct at velocities comparable |
. to those required for: the cooling of NaK radiators

in the ART, and the air will be’ heated electflcolly,
This test =
‘facility should prov;de samples for definitive tests ' by usmg the pot furnace; g heating mantle is used

_' of the opplscoblhty of the proposed Ieok detectors. e

to temperatures as high .as 1400°F.

COMPATIB!LITY OF FLUOR!DE SALTS AND -.

ALKAL! METALS WiTH PUMP LUBR!CANTS
' A S Meyer, Jro - G Goldberg

 

R h.fieyer.‘ ’Jr.;. et Ihl “ANP -Q‘fiar‘.-"'Prorg. 'Rep._

Maer 10. 19561 0RNL'206‘, P 207

VA S, Me yer, Jr., and J. P. Young, ANP Quar Prog
Rep. June 10. 1956, ORNL-2106, p 128.

An apparatus ‘is being -

PERIOD ENDING SEPTEMBER 10, 1956

and molten fluoride salts with the lubricants which
are proposed for use in ART pumps. These ex-
periments dre designed to simulate the conditions
that would exist if leaks developed in the seals
of the fuel or coolant pumps. It is necessary to

determine whether a hazardous condition would

result from an exothermic reaction between the
molten materials and the lubricants.

The reactions between samples of Dowtherm-A
(diphenyl oxide) and of Cellulube-150 (tricresyl

- phosphate) -at 200°F with sodium maintained at
" 1100°F were carried out in the apparatus 18 which

was developed for the determination of oxygen in

- sodium by the addition of molten sodium to butyl
bromide. Approximately 100 ml of each of these
lubricants was placed in the glass reaction vessel

and heated to the desired temperature with o
heating mantle. The temperature of the oil was
measured by placing a thermometer within the re-
action vessel, After the transfer line was flushed
with sodium at 1100°F, approximately 5 g of the

‘metal was allowed to drop into the lubricant. The

reaction with the Cellulube was quite vigorous
and exothermic, while there was very little reaction
between the molten sodium and the Dowtherm-A
sample.

In order to provide a versatile apparatus for
testing the compatibility of additional lubricants
with both alkali metals and fused fluoride salts,
the test apparatus shown in Fig. 2.6.2 was de-
signed and constructed, In operation, the melt pot
is half-filled with either alkali metal or fluoride
salt. After the oddition of the oil sample to the
reaction vessel, the apparatus is assembled as

~ pictured.. The apparatus is then evacuated and the
“system s pressunzed with helium at 1 to 2 psi.
_The material in the melt pot is heated to 1100°F

to raise the temperature of the oil to 200°F.
‘When the desired temperatures are reached, the

~ helium outlet valve is opened, and the helium inlet
~ regulator is set ot 1 to 2 psi to maintain a helium
.. flush of the opporatus.
~ provided to relieve possible pressure surges after
» Tests are bemg conducted at elevated tempero-'-
_'itures to oscertam the compot:blhty of olkoll metols

(The helium outlet is.

the addition of the molten material to the oil.) By

means of the manual gear drive the hollow plunger
" is lowered to displace the melt which flows through

 

mA. L Meyer, Jr., G. Goldberg, and W. J. Ross, ANP
Quar. Prog Rep. Dec. 10, 1955, ORNL-2012, p 188,
Fig. 9.1.

129

 
 

 

 

ANP PROJECT PROGRESS REFORT

UNCLASSIFIED
ORNL— LR—DWG 16164

 

GEAR DRIVE -
THERMOCOUPLE
WILSON SEAL TO VARIAC
THERMOCOUPLE—_ - BT
. | |
N
- TO VALVE | TO VALVE

- HELIUM OUTLET - = Gl HELIUM INLET

VACUUM QUTLET

 

FLANGE PLATE i .
. . TEFLON GASKET

"
1)
7,
i)
It
Hal|
I

 

 

 

 

   

PLUNGER

 

h
b
1
\ ”= FLANGE PLATE

 

 

|
|| |=—— 6-in. GLASS
PIPE
, :

 

MELT POT

 

OVERFLOW PIPE

 

 

 

 

 

 

S J FURNACE SUPPORT
ill‘“

 

 

 

 

 

 

 

   

INSULATED POT
FURNACE
3-in. GLASS PIPE

OIL UNDER TEST

 

Fi-g:.tv-2.6.2. Apparatus for Testing the Compatibility of Alkali Metals and Molten Flruoi-id'e‘SaAlts with

| Pump Lubricants.

130

-
 

 

 

 

the overflow tube and drops into the heated lubri-
cant, Any reaction that occurs is-observed visually
through a }-m.-thlck plastic pipe (not shown),
which encloses the entire apparatus. Temperature
changes are recorded on a fast-chart-drive, one-
point recorder.

EXTR-ACTION OF ART FUEL COMPONENTS
WITH TRIOCTYLPHOSPHINE OXIDE

A, S. Meyer, Jr. W, J. Ross

“An investigation has been initiated of the ex-
traction of ART fuel components, corrosion prod-
ucts, and fission products from acidic solutions
into solutions of trialkylphosphine oxides in an
organic solvent. Trioctylphosphine oxide (TOPO),
which has been recommended for the extraction of
vranium from leach liquor,'® has been found to
offer two methods for supplementing the existing
analytical methods. Traces of structural metals,
iron, and chromium can be readily concentrated by
extraction from solutions of the fluoride fuels or
alkali metals, and thus the sensitivity of the meas-
urement of the accumulation of corrosion products
can be improved. The extraction may also be used
toisolate certain elements, particularly rare earths,
so that sensitive and convenient methods may be
applied to their determination. ,

Initial results show that TOPO extracts metal
ions from acidic solutions into organic solvents,
such as cyclohexane or Varsol, through the forme-

~ tion of definite complexes. The extent to which

the extraction of Cré*, Fe2*, and Zr4* occurs can

. be made quantitative, or the extraction of the ele-

ments can be completely prevented by varying such
experimental conditions as type and .concentration

 of the acid that is used as a solvent for the fluo-
The results are practically -

ride salt mixture,
independent of equnl:braflon periods longer than

‘5 min and of ratios of volumes of the aqueous and
- organic. phases. A rapid method of isolating and

concentrating the components is fherefore avml-

“able. .

Experimental resulis have shown fhat 0.5 mmoles
of TOPO in cyclohexane extracts quunhtatwely, in

“a single equilibration, (1) 10 mg of Cré* from 1 to
-7 MHCland 1 to 7 M H,S0,, approximately 9 mg
~ from 8 fo 12 M H PO cmd 'l M HN03, cnd lesser

 

"C A. Bloke, K. B. Brown, and C. F. Coleman,
Solvent Extractions of Uranium (and Vanadium) from
Acid Liquors with Trialkylphosphine Oxides, ORNL-
1964 (Nov. 4, 1955).

and S

PERIOD ENDING SEPTEMBER 10, 1956

amounts from other concentrations of these acids;
(2) 10 mg of Fe*** from 4 to 7 M HCl and lesser
amounts from 1 M HCI, but none from HNO,, H 290,
or H,PO,; (3) 10 mg of Zr*t from 1 to 10 M HCI
, but none from H3PO4 or HNO.; 3i and
(4). mllhgram amounts of U0, * from all four ocnds.
Trioctylphosphine oxide does not extract Nit**
Crt*t, La***, or the rare earths. Ferrous ion is
pamally extracted from HCl because of slow oxi-
dation to Fet**,

Since the excess TOPO remains in the organic
phase, the elements which are not extracted can be
determined by the usual methods . Thus, the rare-
earth elements can be separated from the fuel
constituents which would interfere with their spec-

" trophotometric determination. Preliminary studies

indicate that other typical fission products, such
as barium and rubidium, also remain in the aqueous
solution,

‘While the extracted ions can be stripped from the
organic phase for subsequent spectrophotometric
determination in the aqueous solution, more con-
venient determinations of iron and chromium can be
carried out by direct measurement of the absorbance

of the extract. The molar absorbance indices for

the absorption maxima of the Fe*** complex that is

extracted from 2 to 9 M HCI is 8000 ot 365 myu and
7500 ot 317 my. These values are comparable to
the molar absorbance index of the iron ortho-
phenanthroline complex (11,400 at 508 mp).20
Extracts of hexavalent chromium exhibit absorp-
tion maxima at 366 mu and 285 my, with indices of
1300 and 1700, respectively. A more sensitive
spectrophotometric determination of chromium may
be obtained by developing the color with diphenyl

carbazide in the organic phase. A survey of the

extractability of other elements which may be
present in experimental fuels is now being carried

out,

DETERMINATION OF CHROMIUM IN TRIOCTYL-
' PHOSPHINE OXIDE EXTRACTS WITH
DIPHENYL CARBAZIDE

_ C. K. Mann?! |
The diphenyl carbazide method has been applied

_to the spectrophotometric determination of hexa-
- yalent chromium after its extraction into solutions

 

20G, F. Smith and F. P, Richter, Pbenatbroline and

| Substituted Phenanthroline Indicators, Their Preparation,

Properties, and Applications to Analysis, G. Frederick
Smith Chemical Co., Columbus, Ohio, 1944, p 78.

21Research participant, University of Texas.

131

 
 

 

 

 

 

of trioctylphosphine oxide (TOPO) in organic sol-
vents. When a 0.25% solution of diphenyl carbazide

in ethano!l is added to solutions of hexavalent

chromium which have been extracted from 2 M
solutions of sulfuric or hydrochloric acid, ¢ violet
color is developed which is similar to that formed
in an aqueous solution. The organic solutions
exhibit an absorbance maximum at the same wave-
length as that observed in the aqueous solution
(550 mp). When ethanol is used as a diluent, the
molar absorbance index is identical to that ob-
tained in aqueous solutions. Since the chromium
may be concentrated at least twentyfold by ex-
tracting the aqueous solutions with small volumes
of TOPO solution, the method affords a significant

‘increase in sensitivity, The method may also offer

a means for the elimination of the interference of
vanadium, which is not extracted from solutions of
sulfuric acid.

It is proposed to apply this procedure to the de-
termination of hexavalent chromium in a sulfuric
acid solution of fluoride salts or in acid solutions
of alkali metals in 2 M H.‘,SO4 or HCi. The solu-
tion is equilibrated with 5 ml of a 0.1 M solution
of TOPO in benzene. An aliquot of the organic
phase is transferred immediately to a 25-ml volu-
metric flask, Two mitliliters of a 0.25% solution of

132

diphenyl carbazide is odded, and the solution is
diluted to volume with ethanol. “After 1 hr the
absorbance of the solution is measured at 550 my.
Tests are currently being made to evolucte this
particular application. :

ANP SERVICE LABORATORY
W. F. Vaughan ‘

A total of 1427 samples was analyzed, which
involved 5247 reported results, an average of 3.7

per sample. A breakdown of the work is given
below: - -

 

Number Number of

of - Reported

Samples -~ Results
Reactor Chemistry 738 _ 2615
Experimental Engineering 626 2370
Metallurgy 28 30
WADC 27 189
Miscellaneous 8 43
1427 5247

The program of the analysis of special samples
from the Wright Air Development Command (WADC)

was completed.
 

 

 

 

&

 

' Part 3
METALLURGY

W. D. Manly

 

 
 

 

-4

 

 
 

 

 

”

. 3.1. DYNAMIC CORROSION
" JH.DeVan

FORCED_'-CIR'C'UL'ATIONfLOQIP'T_EST'S
| J. H. DeVan -
Fuel Mlxtures in. Inconel und Hasfe"oy B.

" Two Inconel forced-cnrculohon loops were' ex-

~ amined following-1000 hr of operation with the fuel .
. mixture {No.’ 70) NaF- ZrF -UF, (56- 39-5 mole %).

The purpose “of these tests was fo compare the
corrosion properties of this fuel mixture with those

of the similar, more commonly used, -fuel mixture

(No. 30) NaF-ZrF ,-UF (50-46-4 mole %) : The

lower ZrF4 content of maxture No. 70 gives a fuel

. with a considerably lower vapor pressure at reac-
tor operahng temperatures than that of muxture o

- No. 30.

The conditions of operahon for these two Ioops,'

7425-14 and 7425-15, are given in Table 3.1.1.
Operation of the Ioops, including the cleaning
procedue, followed the same pattern as that
normally employed for tests with fuel mixture
No. 30. Visual examination of loop 7425-14, which
was operated with @ maximum fluid temperature of
about 1500°F, showed the presence of few very
small metallic crystuls in the top section of the
cooling coil and ulong the fluid—inert gas inter-

face in the pump. These crystals wer_e_ found by

sééctermphic examination to be >5% Cr, >5% Zr,
2% Ni, and 0.2% Fe. Although the fuel undoubtedly

icompnsed a large portion of the crystals, the
- presence of chromium indicates possible mass

transfer. However, no crystals could be found in
loop 7425-]5 which operated with the same tem-

. -perature gradient but a higher maximum fiuid tem-

perature of about 1650°F,

The .maximum hot-leg attack observed metallo-

. graphically was to a depth of 7 mils and was ap-

proximately the same for both loops. The attack
occurred as light to moderate void formation, as

~ shown in Fig. 3.1.1..

The similarity in the attack in the two loops

(7425-14 and -15);, which operated at different

-maximum fluid temperatures, is thought to have

limited significance. For loops operated with
fuel No. 30 under similar conditions, the average

- hot-leg attack is 4.5 mils at a maximum fluid
temperature of 1500°F and 9 mils at a moximum

fluid temperature of 1650°F. The value of 7 mils

for the hot-leg attack at both temperatures in the

tests with fuel No. 70 is felt to be within the
range of possible experimental deviations and
does not represent an abnormally large difference

- from either of the values obtained for fuel No. 30.

TABLE 311 CONDITIONS OF OPERATION OF THREE FORCED—CIRCULATION
LOOPS WITH FUEL MIXTURES

Operating time: ]000 hr

Temperuture gradienf 200°F o L
- Ratie of hot-leg’ ,s_u_rf_ace' to _loop volume: ‘2.08 in.”

 

e Op"“f'flg iéd'na'ir'tibfi;ij? - —_—

Loop Number

 

 

   

   

o 77425_14 7425-15 7425-13
'._.:-VLoc;p mataria] Inconel i !ncbnel o ,'H‘o‘.strelrloY‘B
' "'-'::;:i.Fuel rmxiure 7 el B N°-70* - : NO 70* No. 107**
{Maxsmum fuel miinxwre te-mpernfun.a‘,‘rm F 1525 - .71643 , | ‘-"505
o ‘:MQXImt;I'I:l fube woli temperuture, F T ]595 | ‘.‘712 ‘555 :
- V'JL’-‘:,:-':‘Reynolds number of fuel ‘7 : 10,900 :' o 95_00.. - . 10,000
;--;Q-'_V.,.;;:,Ve|ocfly of fuel fps L | 48 L S 26 4 |

 

*Composition: ‘NaF-ZrF -UF , (56.39-5 mole %).

**Composition: NaF-KF-LiF-UF (”;2-43-45.3-2.‘5 rfiole %).
4

135

 
 

 

 

ANP PROJECT PROGRESS REPORT

 

 

 

 

 

Fig. 3.1.1. Hot-Leg Surface of Inconel Forced-
Circulation Loop 7425-14 After Operation with

the Fuel Mixture (No. 70) NaF<ZrF -UF  (56-39-5
mole %) for 1000 hr at o Maximum Fue1 Tenper-
ature of 1525°F, 250X. Reduced 34%. {Secret-
with—caption)

A sample of the very thin metallic layer which
was deposited in the cold legs of both loops is
shown in Fig. 3.1.2. The layer was barely visible
at a magnification of 500X, and thus may be seen
to be quite thin. It was not possible to analyze
this layer. Deposited layers of this type have not
been found in loops operated with fuel No. 30.

Analyses of fuel samples taken before and after
operation of these loops are presented in Table
3.1.2. The ancalyses indicate that the chromium
content increased during the tests, but the final
chromium values were somewhat lower than those
usually found in the fuel mixture (No. 30) NaF-
ZrF UF, (50-46-4 mole %), when tested under
comparable conditions.

A Hastelloy B forced-circulation loop (7425-13)
was also examined following 1000 hr of operation.
This loop circulated the fuel mixture (No. 107)
NaF-KF-LiF-UF, (11.2-41-45.3-2.5 mole %) under
the conditions given in' Table 3.1.1. The loop was
constructed of 3’2-in.-OD, 0.035-in.-wall tubing and
was identical in configuration and manner of
heating to the standard loops presently employed
for Inconel corrosion tests. !

The results of the examination verified the
excellent corrosion resistance of Hastelloy B to

 

16. M. Adamson and R. S. Crouse, ANP Quar. Prog.
Rep. June 10, 1955, ORNL-1896, p 86.

136

Fig. 3.1.2. Thin Layer Deposited in Cold Leg
of Inconel Forced-Circulation Loop 7425-15 After
Operation with the Fuel Mixture (No. 70) NaF-
ZF -UF, (56-39-5 mole %) for 1000 hr. 500X,
Reduced %O 5%. {Secrotwith~eaphon)- L

TABLE 3.1.2. URANIUM AND IMPURITY ANALYSES
OF FUEL MIXTURE (NO. 70) NaF-ZrF -UF,
(56-39-5 MOLE %) BEFORE AND AFTER
CIRCULATION IN INCONEL LOOPS
7425-14 AND -15

 

Uranium Impurities

Loop No. Sample Taken Content
(wr %) Ni Cr Fe

Found (ppm)

 

7425-14  Before filling 11.2 15 85 50
After draining 11.4 30 290 90

7425-15 Before filling 1.1 95 90 35
After draining 11.6 4* _255 110

 

*This value is in doubt.

fluoride fuel mixtures as observed previously in
thermal-convection loops. Very little attack was
found in the examination. Some pitting, to @ maxi-
mum depth of 1.5 mils, was observed along both

the hot and cold legs, but the surfaces were quite

similar to those of as-received tubing, as illus-

trated in Fig. 3.1.3. No deposits were found

metallographically either in the hot or the cold
legs of the loop. A visual examination of the cold
leg revealed small metallic-appearing deposits in
 

 

 

e

&

 

 

 

 

Fig, 3;1.3. : w(a) _As-Receieed Haste‘ley B Tubing

Used in the Construction of Loop 7425-13, (b) Hot-
Leg Surface of Hastelloy B Loop 7425-13 After

Circulating the Fuel Mixture (No. 107) NoF-KF-

LiF-UF (11.2-41-45-3-2-5 mole %) for 1000 hr ot -
‘a Maxlmum Fuel Temperuture of 3505°F. - Nofe"
increased amount of seécond phase in (b) 250X _

Reduced 34%. eSeem-vmh-evmm%- i

one - areu, _but from the metal!ographlc examma-'
~ tion, . these appear to. have been fluorldes rafher'
thon metallic crystals. = . o T

Chemical analyses of the fuel before ond ofter

‘operation (Table 3.1.3). shoWed a small increase m"_jf
‘chromium - content “that possibly reflects small -

residual quanhhes “of chromlum im the' tubmg or m"-_"_'
the welds. There was. also a small increase in the .
iron ~content and g small decrease in- the’ mcke!fi_‘
_content. A molybdenum analysis was not obtained -

for the fuel before the test, but the level of this
impurity after the test indicates that the molyb-

-on : ._fluonde fuei . corljos_ten

PERIOD ENDING SEPTEMBER 10, 1956

“TABLE 3.1.3.. URANIUM AND IMPURITY ANALYSES
OF FUEL MIXTURE (NO, 107} NaF-KF-LiF-UF, -
(11.2-41-45.3.2.5 MOLE %) BEFORE AND AFTER
CIRCULATION IN HASTELLOY B LOOP 7425-13

 

Uranium

Content
(wt %) Ni Cr Fe Mo

impurities Found (ppm)

 

Sample Taken

 

- Before filling 12.5 215 35 60
After draining 13.2 80 195 W05 7

 

denum content was probably not affected to a
measurable degree during the test.

The operation of an Inconel forced-circulation
loop -as an endurance test with the fuel mixture
{No. ~30) NaF-ZrF‘-UFAt (50-46-4 mole %) was
terminated because of a pump-drive failure after
8300 hr. A leck developed in the cold leg while
‘the loop was being heated to re-establish operation
'oflef repair of the pump drive. The loop {4935-6)
was gas-fired and operated with a maximum fuel
temperature of 1450°F, a temperatute drop of
200°F, a maximum wall temperature of 1550°F, and
‘a Reynolds number of 10,000.  Metallographic ex-
amination showed roughening of the hot-leg sur-
face and heavy intergranular void formation to a
depth of 25 mils. The cold-leg surface was also
roughened, but there was no evidence of mass-
transferred deposits.

Another forced-circulation loop (7425-9) was
examined that had operated for 3000 hr. This loop,
which was heclted by electrical res:stance, also
_cireulated . fuel No. 30.. “The maximum and mini-
~mum - fluid temperatures employed were - 1600 and
1300°F ond the Reynolds number was- 5750, " The
‘maximum | wafl ‘temperature - -was - 1700°F The
“maximum afiack in this loop, " whlch was fo a depth
“of 14 mils, occurred in the area of hlghest wall
temperature “Visual ond metallographlc examma-
hons failed to reveal deposns in the cold zones

hut could be qmtbuted to mass ‘transfer.

- The chromwm bunldup inthe fuels durmg the
 course. of operahon “of both: these loops (4935-6-
“and. 7425-9) was comparable to that found -in fuels .
used in . 1000- hr tests.  These’ fundmgs substanti-
ate - prevnous observahonsz on the effect of time
. In - the ‘earlier .repc‘m

 

2(3. M. Adamson and R. S. Crouse, ANP Quar. Prog
Rep. Sept. 10, 1955, ORNL-1947, p 95.

137

 
 

 

 

 

ANP PROJECT PROGRESS REFPORT

the rate of attack noted in the test period from 50
to 1000 hr was 3 to 4 miis per 1000 hr. [t would
appear on the basis of these tests of longer dura-
tion, however, that the rate of attack decreased for

the longer operating times, although the operating -

conditions of the loops are unfortunately not simi-
lar enough to provide a comprehensive study of
attack for operating periods beyond 1000 hr.

Sodium in Incone! and Stainless Steel |

An Inconel forced-circulation loop (7426-11) was
examined which had circulated sodium at a maxi-
mum temperature of 1350°F for 1000 hr. Only
scattered metal deposits were found in cold-zone
sections. The loop was operated with a fluid tem-
perature drop of 300°F, and it contained a bypass

cold trap operated at 300°F. The deposits reached

a maximum thickness of 3 mils, as shown in
Fig. 3.1.4, and were found by analysis to be pre-
domingantly nickel. The hot leg of the loop showed
slight intergi'anul'ar attack to a depth of 0.5 mil.

A relatively heavy deposit, also predominantly
nickel, was found in Inconel loop 7426-12 which
circulated sodium ot a maximum temperature of
1500°F, with a temperature gradient of 400°F, for
1000 hr. The loop included a cold trap identical
to that used in loop 7426-11. The deposits were
found primarily in the economizer section of this
loop, and they reached thicknesses of 20 mils, as

shown in Fig. 3.1.5. There was intergranular -

FOARE

    
  

Y
+
4

 

CEFETETE

 

 

 

Fig. 3.1.4. Deposits Found in Inconel Fofced-
Circulation Loop 7426-11 Which Circulated Sodium
at ¢ Maximum Temperature of 1350°F for 1000 hr.

250X. Reduced 32%., {Eonfidemtielwithrteption)

138

 

Fige 3.1.5. Deposits Found in Inconel Forced-
Circulation Loop 7426-12 Which Circulated Sodium
at a Maximum Temperature of 1500°F for 1000 hr,

250X, Reduced 18%. (ceonfidentiat-witirception)

attack in the hot zone to a depth of 2 mils. The
deposits were scraped from the loop walls and
were found to weigh 21 g. This weight value is to
be compared with a value of 15 g reported previ-
ously for the deposits found in a loop operated for
the same length of time with the same maximum
fluid temperature but with a 300°F temperature

drop.® It thus appears that the increase in tem- |

perature drop from 300 to 400°F appreciably in-
creased the amount of mass transfer. Also it can
be seen that the amount of mass transfer was con-
siderably more in these loops operated at 1500°F
than in the loop operated at 1350°F,

A type 316 stainless steel forced-circulation
loop (7426-14) that had operated for 1000 hr with
sodium at a maximum temperature of 1650°F and a
temperature drop of 300°F was also examined.
Very little increase in mass transfer was found in
comparison with that found in a similar loop op-
erated previously at a maximum temperature of

 

3). H. DeVan, ANP Quar. Prog. ‘Rep. Dec. 10, 1955,
ORNL-2012, p 106.
 

e e s e 1 8

a

1500°F .4 Deposits appeared as occasional
clusters of metal crystals 3 to 5 mils thick and
comprised a total weight of 1.7 g. An analysis of
the deposits showed 5.5% Ni, 72.4% Cr, and
22.0% Fe. The hot leg was attacked intergranu-
larly to a depth of 2 mils, and small voids ap-
peared to a depth of 5 mils. A loop fabricated of
type 310 stainless steel (7426-13) and operated
under similar conditions but at a maximum tem-
perature of 1500°F showed substantidlly more ex-
tensive deposits than did the type 316 stainless
steel loop. The total mass of the deposits was
only 225 g, although the deposit thickness
reached 9 mils. Hot-leg attack in this loop was
quite severe, with intergranular attack and void
formation reaching to a depth of 13 mils, as shown

in Fig. 3.1.6.

 

4J H. DeVan, E. A. Kovacevich, and R, §. Crouse,
ANI;7 Quar. Prog -Rep. Marcb 10, 1956, ORNL.-2061,
p 117,

 

 

Circulated Sodiuvm ot a Maximum Temperature of
1500°F for 1000 hr, 250X. Reduced 17%.
Gonfidentiahwith-caption]

PERIOD ENDING SEPTEMBER 10, 1956

Sodium-Beryllium-Inconel Compatibility

Two Inconel forced-circulation loops with beryl-
lium inserts in the hot legs were examined that
had operated with sodium at maximum temperatures
of 1250 and 1300°F. The beryllium insert was
machined to the form of a hollow cylinder 2'/2 in.
long, and it contained an Inconel island that was
concentric within the insert. The insert was
separated from the loop wall by a distance of
0.040 in. along 1 in. of its length and by 0.094 in.
along another 1 in. of its length; the remaining
1’2 in. of the insert was in direct contact with the
Inconel loop wall. The Inconel island was spaced
l/4 in. from the inside surface of the insert. The
beryllium insert tested ot 1300°F for 1000 hr
showed scattered void formation to a depth of
7 mils. Alloying between the beryllium and Inconel
in contact with it produced a brittle layer approxi-
mately 3.5 mils thick. In areas where a positive
separation between the beryllium and Inconel was
maintained, no attack or alloying was seen on the
Inconel. In the loop operated at 1250°F for
1000 hr, voids formed in the beryllium insert to a
depth of 5 mils. Although no alloy formation be-
tween beryllium and Inconel was found, even in
areas of direct contact, this result is considered
to be anomalous, since alloying has been seen to
occur even at temperatures as low as 1200°F be-
tween Inconel and beryllium in direct contact in
sodium. Both the loops operated with cold traps
maintained ot 300°F. No mass-transferred par-

ticles or layers were produced in either loop.

THERMAL-CONVECTION LQOP» TESTS
ELA. Kovacevuch 4. H. DeVan
Sodwm-Beryllium-Hasfelloy B Compchblhfy

Two Hastelloy B ‘thermal-convection loops with

“beryllium inserts in the hot legs were examined

- which had circulated sodium at 1200 and 1300°F
. for 1000 hr. The’ msert cons:sfed of a hollow

';cyhnder 6 in. Iong, with a tube. wall thickness of
--0.080 in. The spacing between the Hastelloy B

and the beryllium was 0.020 + 0.005 in. Holes

; ~ were drilled at the top and the bottom of the in-
F ig.3 1.6. Hof-Leg Afluck ln Type 310 Sfainless i
Steel - Forced-C:rculnhon ‘Loop (7426-13) Which .

sert to permit sodium flow between the beryllium

and the :Hcf,s;e“oy"_Bf ond to eliminate stagnant
sodium areas in the system. The results of ex-

amination of these loops (899 and 900) are pre-
sented in Table 3.1.4.

139

 
orl

TABLE 3.1.4. ‘RESULTS OF EXAMINATIONS OF HASTELLOY B AND INCONEL THERMAL-CONVECTION LOOPS WITH BERYLLIUM

INSERTS IN THE HOT LEGS AFTER CIRCULATING SODIUM AT YARIOUS TEMPERATURES FOR 1000 hr

 

Depth of Attack (mils) and Corrosion Results

 

 

 

 

Hastelloy B Inconel.
Loop Loop Loop Loop Loop Loop
No. 899 No. 912 No. 913 No. 914 No. 915 No. 916
(1200°F*) - (1300°F*) (1200°F *) (1250°F*) (1300°F*) (1400°F*) (1500°F*)
Hot leg 1 1 ] 1 1 _ 1
Cold leg 1 Surface roughened Surface roughened Surface roughened Surface roughened  Surface roughened
Sleeve |
Top section 2 Surface roughened  Surface pitted Surfoce pified 1 0.5
Ni-Be alloy Ni-Be alloy
formed formed
Center section 1.5 Surface pitted 1 Surface pitted 1 Surface pitted; -
' Ni-Be alloy -
formed
Bottom section 1.5 Surface roughened Surface roughened 1 1 ]
1-mil layer of 3.5-mil layer of Ni-Be alloy
Ni-Be alloy Ni-Be alloy formed
formed formed
Beryllium insert
Inner surface top 1 2 Surface pitted 1 1 2
section :
Inner surface center 2 2 2 ] 1 2
section
Inner surface bottom 1 3 2 3 2 2
section
Outer surface top 3 2 5 6 9 13
section
Outer surface center 3 2.5 T 5 8 10
section ' , l
'Outer surface bottom 1 3 3 6 8 8
 section s
*Maximum fluid temperature,
. “oa " . ' w .

L3043 SSJH90¥d 103r0dd dNV

 
 

3

 

The maximum ottack on the beryllium was in
the form of small voids on the outside surface of
the insert. No alloying occurred between the
Hastelloy B and the beryllium across the 0.020-in.
separation in either loop (899 and 900}, but there
was alloy formation where the beryllium came in
direct contact with the Hastelloy B. It appeared
as a homogeneous layer 1 mil thick in loop 899
and as a two-phase layer 3 mils thick in loop 900.
The insert removed from loop 900 had a gray scale
on the outside surface, but no such scale was
present on the insert removed from loop 899. The
x-ray diffraction pattern identified the scale as
Be,yNig. No cold leg deposits were found in
either loop.

Sodium-Beryllium-Inconel Compatibility

A series of five Inconel thermal-convection loops
with beryllium inserts in the hot legs, similar to
those used in the Hastelloy B loops, were op-
erated with sodium for 1000 hr ot maximum fluid
temperatures  of 1200, 1250, 1300, 1400, ond
1500°F. Metaliographic examination of the Inconel
adjacent to the beryllium insert, which was spdced
0.020 % 0.005 in. from the Inconel, showed some

alloy formation in the loops operated at tempera-.

tures of 1300°F and aobove. Nickel-beryllium
alloying typical of that observed is shown in
Fig. 3.1.7. The attack on the beryllium in-
creased, as shown in Table 3.1.4, with increased
fluid temperatures. The deepest attack occurred
on the outside surface of the insert. In all in-
stances, a gray scale, accompanied by a black
deposit, shown'in Fig. 3.1.8, was observed on the

outside and inside surfaces of the insert after the
test.. - These deposnts were also identified by

x-ray diffraction as Be,yyNig.: The thicknesses of

the nickel- bery!lwm alloy formations ranged from
1" mil in loop 914 to a maximum depth of 3.5 mils -
in loop 915; the 3.5-mil formation . is. shown in

Fig. 3. 1.7.. The alloy formed in loop 915 was

reported to. consist of two phases; however, m’,."

the other loops only one phase was ‘observed. No

metallic deposits were found in the cold. iegs of .

these loops.  Chemical analyses showed the

sedium drained from these -loops to contam ap-“'
| proxtmate|y 300 ppm of berylllum.,_ B '

Fuel Mlxiures in Hastelloy and Specnal
Nickel-Molybdenum Alloy Laops

Hastelloy X thermal-convection loop _983 was
operated for 1000 hr with the fuel mixture (No. 30)

PERIOD ENDING SEPTEMBER 10, 1956

 

Fig. 3.1.7. Nickel-Beryllivm Alloy Formed in
Inconel Thermal-Convection Loop 915 Which
Contained a Beryllium Insert in the Hot Leg and
VWas Operated with Sodium at a Maximum Temper-
ature of 1400°F, 250X. Reduced 32.5%. (Eonfi~
I ’l ] Ill Ic ) !

NaF-ZrF -UF, (50-46-4 mole %) at a maximum
temperature of 1500°F to confirm corrosion results
reported previously® for Hastelloy X thermal-
convection loops operated with this fuel No. 30.
The maximum hot-leg attack observed in this loop
was to a depth of 35 mils, shown in Fig. 3.1.9,
and it occurred as heavy intergranular subsurface-
void formation. Metalliographic examination also
revealed light surface roughening in the cold leg,
along with numerous small- to medium-sized metal-
lic crystals, shown in Fig. 3.1.10. Macrescopic
examination of. this |aop also revealed metallic
,crystals in the trap-area.  Chemical analysis: of
the fluorides token from the trap show::d 6.25% U
and the following impurmes. 422 ppm Ni, 16.6%

'Cr,"and 170 ppm Fe.

Hastelloy W thermal- convecflon loop 897 was

- operated with the fuel mixture (No. 107) NaF-KF-

LiF-UF, (11.2-41-45.3-2.5 mole %) for 1000 hr
at 1500°F The maximum attack was to a depth of

0.5 mil in the hot leg, and it appeared as surface
' -'_'roughenlng and pitting, as shown.in Fig. 3.1.11.
The cold leg showed light surface roughenmg but

_no evidence of metal deposits.

A mckel-molybdenum alloy (85% Ni—lS% Mo)
loop was similarly operated with the fuel mixture
(No. 107) for 1000 hr at a maxlmum temperature of

 

5J. H. DeVan and E. A. Kovacevich, ANP Quar. Prog.
Rep. June 10, 1956, ORNL-2106, p 137.

141

 
 

 

 

"ANP PROJECT PROGRESS REPORT

UNCLASSIFIED
T-9944

 

Fig. 3.1.8. Beryllium Insert and Inconel Sleeve from Inconel Thermal-Convection Loop 913 Showing
the Scale Formed During 1000 hr of Operation with Sodium at a Maximum Temperature of 1250°F. 250X,
Reduced 7%. (Confidential with caption) -

 

UNCLASSIFIED
Q 10237

 

 

  

Fig. 3.1.9. Maximum Hot-Leg Attack in Hastelloy Fig. 3.1.10. Smcl_l-to-Mediuhi-Sized Crystals
X Thermal-Convection Loop 983. 1000 Reduced Formed .in Cold Leg of Hastelloy X Thermal-
32.5%. Convection Loop 983, 250X. Reduced 32.5%.

142

ol

3 ¢
 

e e LS Ml L1 o g e ol

 

.

Be

 

Figo 3.1.11. Maximum Hot-Leg Attack in
Hostelloy W Thermal-Convection Loop 897 Oper-
ated for 1000 hr with the Fuel Mixture (No. 107)
NaF-KF-LiF-UF, (11.2-41-45.3-2.5 mole %) ot a
Maximum Temperuture of 1500°F, 250X, Reduced

33%. (Seeretwith-eaption)

1500°F. This loop (986) also showed good cor-
rosion resistance. A few scattered areas revealed
attack to a depth of 2 mils, as shown in Fig.
3.1.12, but, in general, the surfaces appeared to
be unaffected over most of the hottest portions of
the loop. The cold leg had light to moderate sur-
face roughening.

Special NaF-ZrF -UF Mixtures in Inconel Loops

Tests were conducted in standard Inconel ther-
mal-convection loops 937 and 988 to determine the

- corrosion behavior of a special batch (EE 808) of
the fuel mixture (No..30) NoF-Z¢F -UF , (50-46-4
‘mole %) that was prepared with especmlly pure

ZiF .  These loops were operated with maximum
fluid temperutures of 1500°F for 500 hr. -

The maximum hot-leg attack of these loops, as

- reported in Table 3.1.5, was 2 to 3 mils less than
the attack observed in standard loops 860 and 861,

which ‘operated with standard fuel No. 30 (batch
EE 526P{2) containing the commercial-grade zirco-
nium fluoride nprmclly used in the production of

fuel No. 30. As shown in Table 3.1.6, lower

- concentrations of chromlum were found cfter the

tests in the fuels prepared with especially pure

PERIOD ENDING SEPTEMBER 10, 1956

 

Fig. 3.1.12, Maximum Hot-Leg Attack in Nickel-
Molybdenum Alloy (85% Ni=15% Mo) Lcop 986
Operated for 1000 hr with the Fuel Mixture (No,
107) NaF-KF-LiF-UF, (11.2-41-45.3-2,5 mole %)
at a Maximum Temperature of 1500°F. 250X.
Reduced 32%. eeret-with-eaption)}—

TABLE 3.1.5. METALLOGRAPHIC EXAMINATION OF
STANDARD INCONEL THERMAL-CONVECTION
LOOPS OPERATED WITH STANDARD FUEL NO. 30,
'RECLAIMED FUEL NO.30, AND FUEL NO. 30
PREPARED WITH ESPECIALLY PURE ZF,

Test duration: 500 hr
Maximum fluid temperature: 1500°F

 

 

Maximum
Loop No. Fluid Circulated Hot-Leg
Attack (mils)
987 Fuel No. 30 (prepared _ 7
‘with especially pure .. -
_ ZrF ) | |
988 Fuel No. 30 (prepared 7
' with especially _puré
R
858  Reclaimed Fuel No.30 8.5
859 .. Reclaimed Fuel No. 30 5
984  Reclaimed Fuel No.30 6
860 Standard Fuel No. 30 10.5
861  Standard Fuel No. 30 9

 

143

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

TABLE 3.1.6. CHEMICAL ANALYSES OF THE
~ NaF-ZrF -UF , (50-46-4 MOLE %) MIXTURES '
CIRCULATED IN INCONEL THERMAL.

CONVECTION LOOPS 987, 988, 860, AND 861

TABLE 3.1.7. CHEMICAL ANALYSES OF THE =~
‘NaF-ZrF -UF , (50-46-4 MOLE %) MIXTURES
RECLAIMED AND CIRCULATED IN INCONEL -
THERMAL-CONVECTION LOOPS

 

Impurities Found

 

o Uranium
‘Sample Taken From Content (ppm)
{wt %) Ni Cr Fe
Batch EE 808* 878 40 45 60
Batch EE 526Pf2 8.71 20 90 50
Loop 987
Before operation 8.99 115 90 85
~ After operation 8.98 25 625 70
- Loop 988 _ :
Before operation - 9.89 - 50 65 155
* After operation.  8.96 20 720 80
 Loop 860 |
 Before operdfion 8.00 <2 80 95
After operation 8.50 15 1060 80
Loop 861

Before operation 8.70 345** 150 105
After operation 8.79 35 820 100

 

*In addition to the vranium, the analysis showed 38.1
wt % Zr, 11.2 wt % Na, and 42.0 wt % F,

**Filled at the same time as loop 860; value in doubt.

ZrF4 than in the standard fuels, as would be ex-
pected in view of the lower attacks observed.

Another series of three standard Inconel thermal-
convection loops, 858, 859, and 984, was operated
for 500 hr at 1500°F with reclaimed fuel No. 30.
The purpose of these tests was to determine the
difference, if any, in corrosion results from loops
operated with standard fuel and with fuel reclaimed
after use in other experimental tests. In all cases
the reclaimed fuel had been used in Inconel sys-
tems. The reclamation procedure, which was de-
scribed previously,® consisted of the equilibration
of the fuel with zirconium metal chips and the use
of the standard hydrofluorination-hydrogenation
treatment for the removal of impurities.

'Che_micol analyses of the fuel before and after
the treatment are presented in Table 3.1.7. As

 

6F. L. Daley and J. Truitt, ANP Quar. Prog. Rep.
Dec. 10, 1955, ORNL-2012, p 93.

144

Impurities Found

 

Uranium
Sample Taken Content (ppm)
{wt %) Ni Cr Fe
Before reclaiming 8.66 - 2 3075 120
After reclaiming 8.21 220 110 135

Before circulation in 8.36 250 170 95
loop 858 '

After circulation in 8.95 15 670 120
loop 858 S ‘

Before circulation in  7.57 2% 120 115
foop 859 o o

After circulation in 8.84 .30 620 100
loop 859

Before circulation in 8.13 285 155 65
loop 984

After circulation in 8.79 20 820 100
locp 984

 

*This value is in doubt.

may be seen, o definite reduction in chromium
content was effected in the purification treatment.

The summary of the corrosion results, presented
in Table 3.1.5, indicates that a reduction of maxi-
mum hot-leg attack, 2 to 3 mils, occurred in loops
operated with reclaimed fuel that was similar to
the reduction obtained with the fuel prepared with
especially pure ZrF,. This reduction in attack
may be attributed to the removal of impurities from
the fuel through the corrosion processes that oc-
curred during previous operation. It does not ap-
pear to be due to any effect on the level of chro-
mium stemming from prior use of the fuel, since,
as may be seen in Table 3.1.7, the reclamation -
procedure used resulted in a chromium content in
the reclaimed fuel comparable to that of a newly
prepared fuel. However, it should be noted that
any increase in U*** content in these fuels that
resulted from changes in the procedure used fto
prepare the fuels would cause a significant de-
crease in attack. This possibility unfortunately
is difficult to evaluate through conventional chemi-
cal analyses.
 

 

 

 

ga

Screening Tests of Special Fuel Mixtures

A screening program is currently under way in
which the corrosion properties of numerous fluo-
ride mixtures are being evaluated in standard
Inconel thermal-convection loops. Test periods of
500 hr and maximum fluid tenmperatures of 1500°F
are being used. Standard loop filling procedures
are followed, but, since limited amounts of the
fuels are available, the standard precleaning steps
which utilize an extra fill are omitted.

‘The results of the tests that have been com-
pleted are presented in Table 3.1.8 with the data
grouped according to the variable investigated.
Group 1 consists of the zirconium-base fluoride
mixtyre in which the alkali metal fluoride com-
ponents (KF, LiF, RbF, and NaF) were inter-
changed, while the polyvalent fluoride components
(ZrF , and UF 4)'remained constant. The mixtures
are all similar to the fuel mixture (No. 30) NaF-
ZrF UF, (50-46-4 mole %). In these fuel sys-
tems it was desired to establish the corrosion
properties accompanying the various alkali flue-
rides in combination with ZrF, and UF,. Fuels
of the same type but with slightly higher alkali
metal fluoride contents cre also included in this
group.

In group 2 are listed fluoride mixtures which are
complexed with sodium and lithium and which
contain various amounts of ZrF, These fuels
were tested to determine the level of ZrF, below
which such mixtures show corrosion behavior
typical of the fluoride mixture (No. 12) NaF-KF-
LiF (11.5-42-46.5 mole %) rather than the hlgher
ZrF 4-content fuels such as fuel No. 30.

The fuels listed in group 3 were investigated to
determine what effect an increase of the UF
content in a zirconium-base mixture would have on
the corrosion of Inconel. This list of fuels is in-

- complete, with several _mixtures remalnmg to be

evaluated. :

The fuels in group 4 were tested to determlne
whether mass transfer in systems contalnlng BeF,
is a function of the ratio of NaF to BeF '

~ In the group 1 tests involving KF; LIF,‘ RbF,
and NaF systems in combination with ZrF,, the
lowering of the ZrF , content from 46 to 40 mole %

produced no significant effect on the attack in

NaF-containing systems and increased the attack

from 1 to 2 mils in the KF- and RbF-containing
systems. The system containing LiF showed
quite anomalous behavior in that the attack de-

PERIOD ENDING SEPTEMBER 10, 1956

creased murkedly wuth the decrease in ZrF,
content.

The loops operated with fuels containing more
than 50 mole % of the alkali-metal fluoride, with
the exception of the KF-containing system, showed
traces of metallic crystals in the cold leg. In
analyzing the systems from the standpoint of inter-
changing the alkali-metal fluoride component while

‘maintaining the over-all composition constant

(specifically MF-ZrF -UF,, 50-46-4 mole % and
56-40-4 mole %), it wos observed that the KF-
containing system gave the least attack. At the
begiming of the tests, it was expected that the
LiF system would yield the least corrosion; how-
ever, as seen from the data, it produced the maxi-
mum attack.

In order to obtain further corrosion data for the
Kl"'-ZrF“-'UF4 (50-46-4 mole %) mixture, which
showed somewhat less attack than the other sys-
tems, two additional Inconel loops were operated
under the same conditions but for a period of
1500 hr. This increase of operating time from
500 to 1500 hr was accompanied by an increase in
maximum attack of only 2 mils.

In the group 2 tests, little change in hot-leg
attack accompanied the ZrF ,-content decreases.
However, in contrast to results obtained in stand-
ard loops operated with the NaF-ZrF -UF, (50-
46-4 mole %) mixture (fuel No. 30), ewdences of
metallic layers and/or evidences of metallic
crystals were observed metallographically. The
presence of LiF in such mixtures, even those with
46 mole % ZrF‘, did not affect the corrosion

‘appreciably.

As may be seen in Table 3.1.8, an appreciable

~ increase in attack was observed in the tests of
- fuel mixtures containing 26 mole % UF (group 3).

Also, metallographic exammaflon showed metallic

deposits, possibly uranium, in the hot legs of

these loops. However, the increase from 5 to
12 mole % UF, did not seem to affect the corro-
sion. No hot-leg deposits were noted in the loops
operating with the fuel mixture containing the

- lower concentrations of uranium fluoride.

In the bulk of the tests of special fuel mixtures,
especially those in group 4, metallic layers and/or
metallic deposits were noted in the cold legs.
However, because these tests were conducted for
a relatively short time, 500 hr, it is not possible
to accurately assess the mass-transfer properties.
It should be noted that increasing the LiF content

145

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

TABLE 3.1.8; RESULTS OF METALLOGRAPHIC EXAMINATION OF INCONEL THERMAL-CONVECTION
LOOPS OPERATED WiTH SPECIAL FLUORIDE FUEL MIXTURES AT A HOT.LEG
TEMPERATURE OF 1500°F FOR 500 hr

 

Metallographic Results

 

 

Loop Fuel Mixture Fuel Mixture
No. ~ Code and Composition Hot-Leg Cold-Leg
| Designation (mole %) Attack (mils) Appearance
Group 1
839 95 50 RbF-46 ZrF -4 UF4 9 Metallic layer
840 95 " 1" " 9 " "
932 WR-5 56 RbF-40 ZrF4-4 UF4 10 Layer and crystals
933 _ WR-5 " " " N n " "
845 94 50 KF-46 ZrF4—4 UF4 6.5 Metallic layer
846 94 " " " " 7"
930 WR-4 56 KF-40 ZrF ,~4 UF, " "
93] WR-4 " " " " "
847 93 50 LiF-46 ZrF ,—4 UF, 17.5 Layer and crystals
848 93 ‘ ” " " 19 " noooon
928 WR-3 56 LiF-40 ZrF ,—4 UF, 10 " " "
929 WR°3 " " 1" ]0 ” " n"
924 WR-1 56 NaF -39 ZrF 45 UF 7 Layer and crystals
925 WR-1 " " " 9 ’ " " "
926 WR-2 56 NaF 40 ZrF ,—4 UF, 9 vt oon
927 . wR'2 " " " 7 144 " fr
860 30 50 NaF-46 ZrF ;-4 UF, 10.5 No layer or crystals
861 30 " " " 9 noonoon "
948 WR-13 75 LiF-21 ZrF -4 UF, 14.5 Layer and crystals
049 WR-13 " "t 1" 11 " " n
Group 2
950 WR-14 25 LiF-25 NaF-46 ZrF ,—4 UF 12 Layer and crystals
951 WR-]4 1" " " " 1 n 1" "
952 WR-15 20 LiF-36 NaF-40 ZrF4-—4 UF4 10 Layer and crystals
953 WR-15 1r 1" 1" " n " " " :
954 WR-16 25 LiF—41 NaF =30 ZrF4—4 UF, 8.5 Metallic layer
955 WR-16 " " " " 15 Crystals
841 82 55 LiF-20 NaF-21 ZrF4—4 UF, 14 Metallic layer
842 82 " " " " 12-5 1" "
843 9N 35 LiF-53 NoF -8 ZrF4—4 UF, 15 Metallic crystals-
. 844 21 "o " " " 10 " layer
Group 3
934 WR-6 60 NaF-40 ZrF4 5 - Layer and fine crystals -
) 935 WR-6 " " 5 " "noon "o
936 WR-7 60 NaF-35 ZrF -5 UF, . 10.5 Layer and crystals
937 WR-7 " " " 10 1" " "

146
 

 

e

PERIOD ENDING SEPTEMBER 10, 1956

- TABLE 3.1.8 (continued)

 

Fuel Mixture Fuel Mixture

Metallographic Results

 

 

Loop 7
No.’ Code and Cdmpo:ifion Hot-Leg Cold-leg
Designation - {mole %) Attack (mils) Appearance
Group 3
942 WR-10 74 NoF-26 UF, 15 - Layer and crystals
943 WR-10 " " 18 ' " " "

. 944 WR-11 74 LiF-26 UF, 21 Layer and crystals
045 WR-11 . ” " ) 20 1" " "
938 WR-8 . 60 NaF-28 ZrF4-12 UF4 10 Metallic layer
939 WR'S 1" 7 " 13 11" 1"

940 WR-9 60 NaF-14 Z¢F —-26 UF4 11.5 ; Layer | :

941 WR-9 " " 9 Layer and crystals
Group 4

853 - 92 49.5 NaF-48 BeF,-2.5 UF 12 Metallic layer

954 92 - " oon " n "

851 75 67 LiF=-30.5 BeF,-2.5 UF4 17 Metallic layer

852 - 75 " hai ‘7 " 1"

849 79 15 LiF =55 NaF-27.5 BeF,-2.5 UF , ' 1 Layer and crystals
350 79 ) " " - " " 10 " " "
956 WR-17 50 NaF 47 BeF,-3 UF4 8 Layer and crystals
957 WR-I7 LA 1" 9 " " "

958  WR-18 50 LiF 47 BeF 5~3 UF,, 20 Metallic layer

960 WR-19 25 LiF-25 NaF -47 BeF2—3 UF4 13 Metallic layer

961 . WR-19 " " - 12 " "

962 WR-20 66 NaF =31 BeF2—3 UF‘ 9 Metallic layer and crystals

. 963 ’ WR'20 1" " 9 1" 1" " 7"

- 964 WR-21 733 NaF 33 LlF—3'| BeF2-3 UF4 13 Metallic layer

965 CWR21- . "o

. 15 1" n

 

‘in the bery?liUm-tofifdining syvsvtems 'appeol;'s “to
‘also increase the maximum hot-leg attack.  dn

these systems, replacmg the NaF \_mth LsF in-

'-,creased the aflack ‘almost twofold

- |NCONEL STRAIN CYCLING TESTS
: 3. H DeVon '

Inconel have oontmued To date, tests have been

~ made at 1200, 1400 and 1600°F in a helium
o ,afmosphere Two bend-test ossemblles are used
to induce alternate compressive and tensile strains

ranging from 0.2 to 1.0%, as measured in the outer
fibers of the test specimens. Time periods em-
ployed between strain reversals are from 3’2 to 2 hr.

.Méd'lfvlrcahons in the test dsse-mBheS, discussed

.' [m;wim.usly,7 have apparently eliminated the speci-

men warpage encountered in earlier tests. The

results of tests ‘completed after these modifica-
- tions were made are presented in Table 3.1.9. At

. dll  temperatures, a 2-hr hold time appears to
. produce cracks in fewer cycles than are required

Tests '° Sf“dY the effect of CYCI"-' strains °"‘-' . “to produce cracks in tests with a /2 -hr hold time.

Cracks initiate at qpproxlmutely the same time in

- tests at 1400°F and at 1600°F, but they propagate’
faster at 1600°F. :

 

7). C. Amos et al., ANP Quar. Prog. Rep. March 10,

' 1956, ORNL-2061, p 60.

147

 
 

 

 

 

ANP PROJECT PROGRESS REFPORT

TABLE 3.1.9. RESULTS OF INCONEL.STRAIN-C_\'CLING TESTS

 

‘Number

 

, Cycle . Number MfikimufiDath | A‘veraéebepth'
Test No.  Strain Time of _ of Cracks ' of Cracks of Cracks
@ ) Cycles . (mils) (mils) Per Y in.
o : fesf 'lemperufur'e:V.IGOOO»F
K4 1 - h 150 15 015 73
L4 1 h . 220 .3 uas | 81
e 2 - 'so . es o2 . 28
w0 2 60 s 025 35
N 0.4 2 100 0.25 - 0.125 4
M 0.4 2 150 0325 0.125 5
' | | Test tempernturé; 1400°F
1A7 1 b 60 0
1B7 o Y 62 0 |
m R Y 103 0.75 0.5
m 1 % 17 1 0.5 |
L2 1 ), 150 1.75 0.5 78
1H2 1 Y 150 s 0.25 92
K2 1 Y 200 15 0.5 30
162 1 % 200 1 0.75 25
D5 1 2 50 0
1H7 1 2 50 0.75 0.25 14
1C5 1 2 60 0.5 0.25 10
167 1 2 60 0.5
281 0.4 Y 100 0.5
C2A1 0.4 Y 100 0.5
21 0.2 A 150
2L1 0.2 Y 100 . 0.25 0125 5
2A2 0.2 Yo - 300 025 O 0.125 15
282 0.2 Y 200
, | Test femperutl;re: 1200° F
w3 1 2 50 Badly cracked
- 1BS ] 2 50 0
K3 1 2 60 Badly cracked
A5 1 2 60 0
261 1 ), 300 3.5 175 81
2Pl ) 200 375 5 79

 

148
 

s e g e et S+

wh

PERIOD ENDING SEPTEMBER 10, 1956

/3.2, GENERAL CORROSION STUDIES |
E. E. Hoffman

TESTS OF INCONEL TUBE-TO-HEADER JOINTS
WITH RECRYSTALLIZED WELDS IN NoK
AND IN FLUORIDE FUEL

D. H. Jansen

Welded iInconel tube-to-header joints fubricatedr

by The Glenn L. Martin Co, were corrosion tested
in NaK (56-44 wt %) and in the fuel mixture (No. 30)
NaF-ZrF +UF, (50-46-4 mole %) in seesaw appa-
ratus at '|500°F for 100 hr. The test specimens
were resistance welded by using the ‘‘flange-
during-welding method.'’! These welds are made

 

l.!. J. Mueller, C. Shaeffer, and C. L. Lawrence,
Special Welding Techniques, Summary Report, The Glenn
L, Martm Co. Jcn. 1956).

by forging o preflared tube info a drilled header
plate; the edge of the header hole is heated to a
plastic state, which allows the tube to be forged
into position, as shown in Fi,gs. 3.2.1 ond 3.2.2.
The bond is achieved by recrystallization of the
tube-to-header interface. The purpose of the tests
was to determine the corrosion resistance of this
recrystallized area. It was found that the NaK
attacked the specimen to a maximum depth of 0.5
mil in the recrystaliized area and the header plate
(Fig. 3.2.3). The fused salt attacked the header
plate to a maximum depth of 1.5 mils in the form
of small, subsurface voids, but this attack dropped

'to a maximum of about 0.5 mil in the worked tube

and in the recrystallized area (Figs. 3,2.4 and
3.2.5).

 

Fig. 3 2 1 As-Received Tube-io-Header Joint Welded by the "Flange—Durmg-Weldmg Mefhod " The
recrystalhzed area and the worked porflon of the tube can be seen at the top of the header plate. Etched
with aqua regia. 12X,

149

 
 

;
;
i
i
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i
i
i
i
1

 

 

 

 

i

 

}
i
i
|

 

ANP PROJECT PROGRESS REPORT

 

Fig. 3.2.2. Enlargement of Recrystallized Area Shown in Fig. 3.2.1. Etched with glyceria regia.
100X. Reduced 4%.

 

 

 

 

Fig. 3.2.3. Recrystallized Area of Welded Tube-to-l'leuder Joint Exposed to NaK (56-44 wt %) far
100 hr at 1500°F in Seesaw Apparatus. Etched with glyceria regia. 100X. Reduced 4%. (Confidertial-

i ? .
150
 

 

e "PERIOD ENDING SEPTEMBER 10, 1956

 

 

    

 

 

 

|
b 2
;
i
. - ‘Fig. 3.24. --R-e:c‘r'yz.stdllizeid J'Are»o'f-‘df' 'Welded incofiel Tube-to-Header Jérint' Expi:Séc! .tor the F;:el
Mixture (No. 30) NuF-ZrF4-UF‘ (50-46-4 mole %) for 100 hr ot 1500°F in Seesaw Apparatus. Etched
with ,g}y;eriq regia. - 'IOOX ~Reduced 3%. ._(Sec[et with caption) A -
\:J Fig. 3.2.5. Enlargement _(SOOX) of Fig. 3.2.4; Note decrease in attack by the NuF-ZrF4-UF‘ (50-
46-4 mole %) in the worked tube area. Etched with glyceria regia. Reduced 1.5%. (Secret with caption)
151

 

 
 

 

 

ANP PROJECT PROGRESS REPORT

TESTS OF COAST METALS BRAZING ALLOY
NO. 53 IN NaK AND IN FLUORIDE FUEL

D. Hc Jan sen

It was established in previous corrosion tests
that boron is leached from the surface of Coast
Metals brazing alloy No. 52 (89% Ni-5% Si-4% B-
2% Fe) when it is exposed to the fluoride fuels at
1500°F, and it was therefore decided to test a
substitute alloy which might not be as susceptible
to boron leaching but would still maintain good
corrosion resistance to the fluoride fuels. Buttons
of Coast Metals No. 53 (81% Ni-8% Cr—4% B-
4% Si-3% Fe) were corrosion tested in the fuel
mixture {No, 44)NaF-ZrF -UF , (53.5-40-6.5mole %)
and in NaK {56-44 wt %)} in the hope that the chro-
‘mium would tie up the boron and retard its removal
from the surface of the alloy. The same type of
depleted crea as found in Coast Metals No. 52
oppeared, however, in the No. 53 alloy after both

the test in NaK and- the test in the fluoride fuel.
The No, 53 alloy is shown in Fig. 3.2.6 after ex-
posure to the: fuel mixture at temperatures of 1400
and 1500°F for 100 hr. A spectrographic traverse
on the alloy button tested in NaK at 1500°F showed
that, between the surface and a depth of 6 to 7 mils,
the concentration of boron was less than one-third
its value in the interior, Chemical analyses of
the NaK and the fuel mixture have not yet been
completed, All the test systems had surface-to-
volume ratios of 0,83 in.~'. This ratio was cal-
culated by using only the surface areas of the
alloy buttons; the area of the capsule wall that
was covered with the test bath was not included.
Tests will be conducted later in which the surface-
to-volume ratio will be altered to determine whether
it has any effect on the rate of depletion. Results
of the tests described above are given in Table
3.2.1.

 

 

 

 

 

 

 

 

Fig. 3.2.6. Coast Metals Brazing Alloy No. 53 After Exposure to Fuel Mixture (No. 44) NaF.ZsF,-
UF, (53.5-40-6.5 mole %) for 100 hr at (a) 1400°F and (b) 1500°F. Note the increase in the depth of the
depleted area in the specimen tested at 1500°F. Etched with oxalic acid. 150X. Reduced 11.1%.

s Hrcoption)

152
 

 

agw

4

TESTS OF HAYNES BRAZING ALLOY NO. 40
IN NaK AND IN FLUORIDE FUEL

Dn Hc Jansen

Inconel tube-to-header joints brazed with Haynes
brazing alloy No. 40 (nominal composition, in wt %:
Cr, 13.5; Fe, 4.5; Si, 3.9; B, 3.2; Co, 1.0; C, 0.42;
Mn, 0.25; Ni, balance) were corrosion tested in the
fuel mixture (No. 30) NaF-ZrF 4UF ;(50-46-4 mole %)
and in NaK (56-44 wt %) in seesaw apparatus. The

PERIOD ENDING SEPTEMBER 10, 1956

test periods were 100 hr, and the hot zone was
maintained at 1500°F. The specimens were re-
tained in the hot zones of the capsules during the
tests.

The alloy showed good corrosion resistance to
NaK but only fair resistance to the fuel mixture,
There was attack to a depth of 2.5 to 3 mils on
the specimen exposed to the fuel mixtwre, as shown
in Fig. 3.2.7. A considerable amount of porosity
was found in the braze fillets.

TABLE 3.2.1. RESULTS OF EXPOSURE OF BUTTONS OF COAST METALS BRAZING ALLOY NO. 53 TO
NaK AND TO FLUORIDE FUEL FOR 100 hr AT 1400 AND 1500°F

 

Depth of Boron-

 

Temperature
Bath ©F) Depleted Area Metallographic Notes
(mils)

NaK (56-44 wt %) 1400 0.8 No apparent attack

1500 2.5 Small subsurface voids in depleted area
NuF-ZrF4-U F4 1400 0 Small intermittent subsurface voids

{53.5-40-6.5 mole %) along edge to maximum depth of 0.5mil
1500 1.5 Small intermittent stringers along edge

to a moximum depth of 1.5 mils

 

 

Fig. 3. 2 7. H&'ynes Brazing Alloy No. 40 After Exfiosure for "I‘OO hr at 1500°F to (2) the Fuel Mixture
{No. 30) NaF-ZrF -UF (50-46-4 mole %) and to (b) NaK (56-44 wt %) in Seesaw Apparatus. Unetched.
100X. Reduced 'I'I%. (-Seefe+-m-t-h-ea-ph-en)-

153

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

NIGBIUM IN SODIUM AND IN LITHIUM
R. Carlander?

Seesaw apparatus was used for corrosion tests
of niobium in sodium ond in lithium in order to
determine the corrosion resistance of this material
in a dynamic system. Two niobium capsules, 15 in.
long, were filled to 40% of their volume, one with
sodium and the other with lithium, and sealed under
en inert atmosphere. The capsule filled with
sodium was enclosed in an Inconel container,
ond the capsule filled with lithium was enclosed
in a Hastelloy B container. The space between
both the lithium- and the sodium-filled capsules
and their containers was then filled to 40% of its
volume with sodium in order to cobtain good heat
transfer between the capsule and the container
(Fig. 3.2.8). The capsules were tested in a tilting
furnace at 1 cpm for 100 hr with the hot zone at
1535°F and the cold zone ot 995°F,

No mass transfer was found in the niobium cap-
sules, but some dissimilar-metal mass transfer,
detectable only at high magnifications, was found
on the Hastelloy B and the Inconel container walls
(Fig. 3.2.9). The attack in both tests was very

 

20n assignment from Pratt & Whitney Aircraft.

small. Photomicrographs of the niobium capsule
that contained lithium are shown in Fig. 3.2.10.
The results of metallographic examination of the
test specimens are given in Table 3.2.2,

SLCRET™
ORNL -LR-DWG 15484

  
  

{-in. SCH. 40 - ..
HASTELLOY 8 OR INCONEL

Fig. 3.2.8. Schematic Diagram of Capsule and
Container for Corrosion Tests in Seesaw Apparatus
of Niobium in Lithium and in Sodium.

TABLE 3,2,2, RESULTS OF TESTS OF NIOBIUM IN LITHIUM AND IN SODIUM IN SEESAW APPARATUS

Test duration: 100 hr
Hot-zone temperature: 1535°F
Cold-zone temperature: 995°F

 

 

Test Material Bath Attack Metallographic Notes
No. {mils) 7
1 Niobium capsule Lithium on inner surface 2 Intergranular penetration found on both the
inner and the outer surfaces -
Sodium on outer surface 2
Hastelloy B container Sodium on inner surfoce 1.5 Attack in form of voids; small amount of
dissimilar-metal moss transfer detectable
at 500X; precipitate found on inner
surface '

2 Niobium capsule Sodium on inner surface None Cracks found in one area of the outer surs
face; not clearly evident that the cracks -
were due to sodium attack

Sodium on outer surface 1.5
Inconel container Sodium on inner surface None  Surface roughened; small amount of

dissimilar-metal mass transfer crystals

detectable ot SOOX

 

154
 

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{
1

PERIOD ENDING SEPTEMBER 10, 1956

 

. Fig. 3.2.9. Inner Surfaces of (a) Hastelioy B and (b) Inconel Containers for Niobium Capsules After
Exposure to Sodium for 100 hr ot a Hot-Zone Temperature of 1535°F in Seesaw Apparatus. Note the
dissimilar-metal mass-transfer layer on both the Inconel and the Hastelloy B. (4) Etched with 20 em?

of 10% chromic acid and 30 c¢m® of HCl. () Etched with 10% oxalic acid. 500X. Reduced 11%. (Secrot~

 

 

 

| u Fig. 3.2.10. Niobium Capsule That Was Exposed (2) to Lithium on Its Inner Surface and (%) to Sodium
5 on lts Outer Surface in Seesaw Apparatus for 100 hr ot a Hot-Zone Temperature of 1535°F. Etched with
HF-HNO -H 0 (25-25-50 vol %). 250X. Reduced 11%. 4{Seeret-with.captiom)

155

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

VAPOR.ZONE ATTACK IN INCONEL.FLUORIDE
FUEL SYSTEMS

R. Carlander

In several creep experiments on Inconel in the
fuel mixture (No. 44) NaF-ZrF -UF , (53.5-40-6.5
mole %) extremely heavy attack on portions of the

system in the regions above the bath level has

been detected. This type of attack was also noted
in experimental tests of a ZrF -vapor trap fabri-
cated from type 316 stainless steel. A further
investigation has now been made in a test con-
ducted in a manner to approximate roughly the con-
ditions in the creep experimenis where the heavy
attack on Inconel was originally noted.

Evacuated Inconel capsules were exposed to
static NaF-ZrF UF, (53.5-40-6.5 mole %) at
1600°F for 500 and 1000 hr. Two capsules, 12 in.
long, were used in which '/2-in.-diu Inconel thermo-
couple probes extended to within 1 in. of the
bottom. The level of the fluoride mixture bath and
the height of the heated zone was 4 in. A tempera-
ture gradient was maintained over the rest of the
system, Examinations of the capsules after the
tests revealed that the ZrF, vapor had deposited
around the walls of the capsule approximately 1 in.
acbove the original bath level, where the tempera-
ture was approximately 1100°C, A hard layer that
reached a maximum thickness of I/4 in. and con-
sisted of the fiuorides (NaF, ZrF4, and UF ) end
small particles of NiO, Fe, and Cr was found in
a region ‘/.‘, in, above and below the original bath
level. The capsule tested for 1000 hr is shown in
Fig. 3.2.11. Chemical ancalysis of the bath and
of the thermocouple probe revealed that chromium,
as could be expected, was the main element leached
out of the Inconel in both tests. A sample of the
flucride mixture taken from the bottom of the bath
was richer in chromium than a sample taken from
the top. - The results of the chemical analyses are
given in Table 3,2.3.

Metallographic examination revecled that the
portion of the capsule wall around which the vapor
deposited was unattacked in both capsules. The

heaviest attack occurred in the bath zone of both.

capsules. The attack reached a maximum depth
of 4 mils in the 500-hr test and 5 mils in the
1000-hr test (Fig. 3.2.12). The depletion of ZrF‘
from the bath actually decreased the ottack from
that normally observed, and this attack, by itself,
could not have caused the failures in the creep
tests,

156

UNCLASSIFIED
890

Y-18

 

ZsF 4 DEPOSIT

Fig. 3.2.11. Inconel Capsule and Probe Follow-
ing Exposure to the Fuel Mixture (No. 44) NaF-
ZrF -UF, (53.5-40-6.5 mole %) for 1000 hr at
1600°F. <{Secret-withcuption)

CARBURIZATION OF VARIOUS ALLOYS
BY MOLTEN SODIUM '

E. E. Hoffman R. Carlander

The carburization and decarburization of various
alloys by molten sodium at elevated temperatures
has often been observed. The carburization of
stainless steel by sodium contaminated with carbon
results in precipitation at the grain boundaries of
a complex iron-chromium carbide. This precipita-
tion depletes the adjacent grain boundory areas of
chromium, and thus forms areas of a different
‘*stainless’’ composition. Also, the grain bound-
aries become susceptible to intergranular cotrosion.
Hence, an effort was made to study the various
factors that affect the carburization of various
alloys by sodium. '

a
 

 

 

LK

¥

PERIOD ENDING SEPTEMBER 10, 1956

TABLE 3.2.3. RESULTS OF CHEMICAL ANALYSES FOR NICKEL, IRON, AND CHROMIUM OF THE
NoF-ZrF ,~UF , (53.5-40-6.5 MOLE %) BATH AND THE INCONEL CAPSULE FOLLOWING A

TEST FOR 1000 hr AT 1600°F

 

 

Location of Sample Nickel Iron Chromium
: {wt %) (wt %) {wt %)
Inconel Probe
Vapor zone 75.00 7.39 16.08
Bath zone 75.80 7.42 14.37
Top of the fused-salt bath <0.01 0.048 0.002
Bottom of the fused-salt bath 0.32 0.61 3.95

 

 

Fig.”3k2 i2..:- Bufh Zone of lnconel Ccpsule Exposed to NaF-ZrF, -UF (53.5-40-6. 5 mole %) for
1000 hr at 1600°F Etched with 10% oxalic acid. 500X. Reduced 2%. (Ssemt—m-%h—eepfl-en)

157

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

Specimens (2 x 1 x '4; in.) of Hastelloy B and
types 310, 316, and 430 stainless steel were

measured, polished, and weighed prior to testing.

The specimens were loaded into capsules of the
same material, filled with sodium plus 1, 5, and
10 wt % additions of graphite, and sealed under
vacuum. The capsules were then tested in static

sodium at 1500°F for 100 hr. Upon completion of
the tests, the specimens were removed from the
capsules, cleaned, and weighed, and portions of
each were analyzed chemically and metallo-
graphically. The results are presented in Tables
3.2.4 and 3.2,5. Of the four materials tested,
Hastelloy B and type 310 stainless steel (Fig.

TABLE 3.2.4. CARBON ANALYSIS, TOTAL CARBON PICKUP, AND TOTAL WEIGHT CHANGE OF VARIOUS
ALLOYS AFTER EXPOSURE TO SODIUM AND GRAPHITE FOR 100 hr AT 1500°F

 

Capsule Nominal Composition

Graphite Added

Carbon Found in Total Carbon Total

 

Weigh
Material (wt %) to Sodium 3-mil Surface Cut - Pickup :;gst Cl:mnge
ateri eci
¢ ° (wt %) (wt %) {mg) pec men
(mg)
Hastelloy B 25 Cr—20 Mo- 1 0.680 2.6 258
0.25C (mox)-bcl Ni 5 0.577 16.4 456.8
10 .18 224 - 411
Type 310 25 Cr=20 Ni- 1 0.388 0.5 39.3
stainless steel 0.25 C (max)=bal Fe 5 0.994 8.8 47.2
10 1.43 No analysis 55.5
Type 316 17 Cr=12Ni-0.10 C 1 0.606 10.1 48.1
stainless steel (max)=bal Fe 5 1.09 0.7 90.6
10 2,35 50.2 119.3
Type 430 16 Cr=0.10 C (max)- 1 0.267 10.7 - 36,3
stainless steel bal Fe 5 1.30 . 252 | 81.3
10 1.47 55.4 153.4

 

TABLE 3.2.5. METALLOGRAPHICALLY OBSERVED DEPTH OF CARBURIZATION OF VARIOUS ALLOYS
EXPOSED TO STATIC SODIUM FOR 100 hr AT 1500°F

 

Graphite Added

Depth of Carburization {mils)

 

 

to Sodium Type 310 Type 316 - Type 430
Hastelloy B
{wt %) aste oy- Stainless Steel Stainless Steel Stainless Steel

1 5 0-10 4
{varied over
- the surface)

5 8 10 10

10 8 13 24

 

158
 

 

 

|

 

|
;;

»

W

3.2.13) exhibited the least amount of carburiza-
tion, while types 316 and 430 stainless steel
(Fig. 3.2.14) were heavily carburized,

The results can be exploined by the use of
generally accepted - facts about carburization,
Nickel, for example, decteases the rate of pene-
tration of carbon in steel. Therefore, comporing
the three stainless steels, type 430 stainless
steel (less than 0.5 wt % Ni) was carburized to
the deepest extent, while type 310 stainless steel
(20 wt % Ni) experienced the smallest depth of
carbon penetration. Also, the austenitic lattice
dissolves more carbon than the ferritic lattice;
hence, although the depth of carburization was

PERIOD ENDING SEPTEMBER 10, 1956

less in the type 310 stainless steel (austenitic),
the amount of carbon in @ 3-mil surface cut was
approximately the same as that in a similar cut of
type 430 stainless steel (ferritic)y Type 316
stainless steel, which has less nickel and chromium
than type 310 stainless steel, experienced a greater
depth of carbon penetration and had a greater
carbon concentration at the surface. Because of
its -nickel content, Hastelloy B was not expected
to carburize as heavily as it actually did. How-
ever, the presence of molybdenum (a strong carbide
former) in stainless steels renders them more

‘susceptible to carburization, and this may explain

the carburization of Hastelloy B, which also con-
tains molybdenum.

 

| Flg. 3.2,13. 'Husf'ellloy B (@) ond Tfi:e 310 S.f'a'inles's‘: Sieei (b) After Exposure to Static Sodium with
10 wt % Graphite Added for 100 hr ot 1500°F, {z) Etched with phosphoric acid plus H202. (b) Etched
with glyceria regia. 250X. Reduced 10.5%. {Secret—with-captian)

159

 
 

 

 

 

 

 

ANP PROJECT PROGRESS REPORT

s,

e Tyt
I

)

o

 

 

- -
i
¥ Tl

 

 

 

 

 

 

" & St
P " * %
S Mg
% - 2 ok y
K R :
:‘ i ’i-\ I i i -
- 2 3 . . ; L
¢ i e m* T
b ey o
Y & s
8

 

 

 

 

 

 

 

 

Fig. 3.2.14. Type 316 Stainless Steel (a) and Type 430 Stainless Steel (b) After Exposfire to Static
Sodium with 10 wt % Graphite Added for 100 hr ot 1500°F. Etched with glyceria regia. 250X. Reduced

11%. €Serrerwith-eaption)

SODIUM-BERYLLIUM-INCONEL COMPATIBILITY
IN ASTATIC SYSTEM (TEST NO. 2)

The test procedure and the results of sodium-
beryllium-Inconel compatibility test No. 1 were
given in the previous report.®> The results of the
first test indicated that a chromium plate on the
Inconel surface in contact with the beryllium would
reduce the extent of the alloying reaction which
occurs when Inconel and beryllium are in direct
contact in sodium., The thin chromium platings
(1 and 2 mils) used in the first test were entirely
consumed by alloying during the 1000-hr test period
at 1300°F. Therefore o second test was conducted
in which heavier platings were used. A beryllium
oxide specimen was alsc included in this test to
determine whether a reaction would occur between

 

3E. E, Hoffman, ANP Quar. Prog. Rep. June 10, 1956,

160

beryllium oxide and beryllium or beryllium oxide
and Inconel. *

The results of this test are given in Table 3.2.6.
The presence of a chromium plate *‘diffusion
barrier’’ on the Inconel specimen resulted in the
formation of an alloy layer (principally Be,Cr)
which was approximately one-third the thickness
of the alloy layer which forms when Inconel is

‘placed in direct contact with beryllium under the

conditions of this test.

Analyses of the test resuhs indicate that a
minimum of 5 mils of chromium plate will be neces-
sary to ensute that all the chromium is not con-
sumed by the alloying reaction with beryllium. The
Be,Cr phase was identified by x-ray analysis.

Thls phase has a Vickers hardness of 2440, as

compared with hardnesses of 1495 for Be,Nig,
the phase which forms when Inconel and beryllium
are in direct contact, and 180 for Inconel (Figs.
 

PERIOD ENDING SEPTEMBER 10, 1956

TABLE 3,2,6. RESULTS OF METALLOGRAPHIC EXAMINATION OF INTERFACES OF SPECIMENS
FROM SODIUM. BERYLLlUM-INCONEL COMPATIBILITY TEST NO. 2

" Test duration: 1000 hr
~ Test temperature: 1300°F
Contact pressure between specimens: 500 psi

 

Irnterche_

Results of Metallographic Examination

 

Inconel vs berylli‘urm, ‘
direct contact {standard)

Inconel plus 4-mil chromium

Alloy formation (Be2lNi5 and BeNi) 25 mils deep along interface; in
carlier test 24 mils of alloy formed; 4 to 5 mils of Inconel consumed by

alloying reaction

Alloys formed between chromium plate and beryllium to a depth of 8 mils,

 

 

- plate vs beryllium

which consisted of 7 mils of Be,Cr and 1 mil of Be Cr(?); 0 to 2 mils of

chromium plate remained after the test; 2 to 3 mils of Inconel consumed,

probably by alloying with the chromium plate.

‘Inconel plus 6-mil chromium

‘plate vs beryllium

Alloys formed between chromium plate and beI-yllium to a depth of 9 mils,
which consisted of 8 mils of Be Cr and 1 mil of Be Cr(?), 3 mils of

" chromium plate remained after the test; upproxlmfeiy 2 mils of Inconel

consumed by alloying with the chromium plate

Beryllium vs beryllium oxide

Surface of beryllium oxide discolored slightly; specimens were easily

separated and no bonding was evident

Beryllium oxide vs Inconel

Neither the beryllium oxide nor the Inconel surface affected by test

 

3.2.15 and 3.2.16). No reactions were anticipated

between beryllium oxide and beryllium or beryllium

oxide and Inconel, and none were found to have

occurred during this test,

EFFECT OF DIFFUSION COLD TRAPS ON
MASS TRANSFER IN INCONEL-SODIUM

 

: The purpose of opercflon of 1I1e two Inconel- =
sodwm thermal-convechon Ioops dtscussed here;{,;
“was _to “determine - ‘whether fhe presence of o dif-:
- fusion - ‘cold’ trap wouId affect the extent of mass-.f_ o
: r’,transfer observed in the cold legs. The Ioops were . ..
loaded wnh sodlum ‘which had :been preirected by

design, with no cold trap. :
Although many " thermal-convection loop tests
conducted on Inconel-sodium systems in .the past

E. E Hof{man

have failed to show any mass transfer, it has been
demonstrated that mass transfer will occur in such
systems if the hot-zone temperature is in excess of
1500°F and if a steep temperature gradient is
induced in the cold leg by directing an air blast
on a small area near the bottom of the cold leg.

~ These loops (Nos. 28 and 29) were therefore

. o = “operated for 1000 hr at hot- and cold-zone tempera- .
TH,ERMAL CONVEC“ON LOOPS L7 tures “of 1600 and 990°F, respectively, - The loops
“were - carefully: sinpped of sodium after the tests,
The mass-transfer deposits found in the cold zones
of these |oops are shown iin. Fig. 3.2.17. Although
‘;only smull “amounts of mass-fransfer crystals were

:found in these Ioops, it is apparent thot- slightly

more - deposmon occurred in the loop. which had no

_ ;dlffuswn cold trap than in'the loop ‘which mcluded
' i-"cold trapping -for several months to lower the oxy:- a cold trap. -The results of. specfrograph:c analy-
7 gen. ccntent., One loop was - prowdec! WIfl'I a dlf-,_’f-'
~fusion cold trap near the- boftom of the" cold:leg ln"‘"-_.]r-‘.

. order - to reduce ‘the sodium - oxide «content during

o operoflon, ‘while the other loop was of standard"ii

- sis of the metallic crystals recovered from these
loops ‘are presented in Table 3 2.7 -along with. the
*onalysis’ of- the !nconel plpe prior to fest. More
~ tests will be reqmred 1o establtsh whether the
" differences in-chromium and iron confents of the

crystals from the two loops are significant. The
mass-transfer crystals found in a thermal-convection

161

 
 

 

 

 

 

ANP PROJECT PROGRESS REFORT

 

 

Fig. 3.2.15. Alloying Between Chromium-Plated Inconel and Beryllium During 1000-hr Compatibility
Test in Static Sodium at 1300°F. Dark area between Be ,Cr and beryllium is due to separation of speci-
mens after the test. Unetched. 300X. (Secrei-wfl-lfl:up-rren)

 

 

 

 

 

 

Flg. 3.2.16. Enlargement of Area Shown in Fig. 3.2.15. Note extreme hardness of Be,Cr phase as

compared with the hardness of the Inconel and the chromium plate. Unetched. 500X. (Seere!—wflh'
ceptiom~

162
 

 

 

LY ]

T UNCLASSIFIED |

 

PERIOD ENDING SEPTEMBER 10, 1956

S UNCLASSIFIED §
Y-18622

Fig. 3.2.17. Mass-Transfer Deposits in Cold Zones of Inconel Thermal-Convection Loops Which
Circulated Sodium at @ Hot-Zone Temperature of 1600°F and a Cold-Zone Temperature of 990°F. (a)
Cold-zone section of loop which had no diffusion cold trap {loop No. 29). (&) Cold-zone section of loop

which had a diffusion cold trap (loop No. 28). {Cerfidertiatwith~caption)

TABLE 3.2,7. ANALYSES OF MASS-TRANSFER CRYSTALS FROM INCONEL-SODIUM
THERMAL-CONVECTION LOOPS

Operating period: 1000 hr
Hot-zone temperature: 1600°F
Cold-zone temperature: 990°F

 

Material Analyzed

Chemical Analysis (wt %)

 

 

Ni Cr Fe Cu Mn
Crystals from loop No. 28 which included a dif- 84.4 10.6 4.6 0.33 0.07
fusion cold trap
Crystals from loop No. 29 which had no cold trap 77.4 13.3 9.1 0.06 0.08
Inconel pipe {as-received) 73.15 14.81 6.62 : 0.40

 

loop operated previously at a hot-zone temperature

of 1500°F had a much lower iron concentration.*

The results of metallographic examination of speci-_

mens from similar locations in the loops are given
in Table 3.2.8. ‘Examinations of these results
indicate that the loop with no cold trap (No. 29)
had slightly heavier attack in the hot leg (Fig.
3. 2.18) than the loop with a cold trap (No. 28).

 

4E. E. Hoffman, ANP Quar, Prog. Rep. Sept. 10, 1955,
ORNL-1947, p 113.

The deposits found on.the cold-leg surfaces of
these Ioops may be seen in Flg. 3. 2 19

THERMAL-CDNVECTION LOOP TESTS OF
VARIOUS STRUCTURAL MATERJALS
AND LITHIUM

E. E. Hoffman

There has been considerable interest in the
possible application of lithium as a reactor coolant
and, in particular, as a coolant for a solid-fuel-
element aircraft reactor, The corrosion problems

163

 
 

 

ANP PROJECT PROGRESS REPORT

TABLE 3.2,8. RESULTS OF METALLOGRAPHIC EXAMINATION OF VARIOUS SECTIONS
FROM INCONEL-SODIUM THERMAL-CONVECT!ON LOOPS

 

 

 

' Operating Metallographic Reéu'lts, _ L

Location of Temperature Loos No. 28 — Loop No. 20

- Specimen ('QF) oop MNo. - L pp - 29

- , . - {Diffusion cold trap) {No cold trap)
pr'qf hot leg 1600 Grain boundary voids to a deptfi 1 to 2 mils of grain boundary 'aflack |

_ 7 of less than 1 mil ' o
Top of cold leg 1350 S!ighfly irregular surface Slightly irregular surface
‘Bofiom’of cold leg 990 Two-phase surface layer 0.5 mil Two=-phase surface |ayer O.b mlt

s thick thick

 

 

 

 

 

 

 

 

 

 

 

 

 

Fig. 3.2.18. Hot-Leg Surfaces of Inconel-Sodium Thermal-Convection Loops Operated for 1000 hr
with a Hot-Zone Temperature of 1600°F and a Cold-Zone Temperature of 990°F. (a) Hot-leg specimen
from loop No. 28 which included a diffusion cold trap. ' (b) Hot-leg specimen from loop No. 29 which had
no cold trap. 500X. Reduced 4%. (Cenfidential-witheaption)

encountered in attempting to contcin lithium in
high-temperature systems are much more difficult
than those found with liquid sodium at high tem-
peratures. Analyses of o large number of static
tests have indicated that nickel and nickel-base
alloys are very heavily aftacked even by static,
isothermal lithium, and thus many of the commer-
cially available high~temperature alloys, are elimi-
ncted from consideration. Pure iron has shown

164

good resistance to lithium, and therefore con-
siderable effort has been directed towards tests
on iron-base alloys, in particular, the stamless
steels. The results of thermal-convection Ioop
tests of various materials are summarized in Table

3.2.9. The important information contained in

this table is whether or not the loop plugged, the
weight of metal crystols recovered from the cold
leg of the loop, and the depth of attack in the hot
 

 

 

.

PERIOD ENDING SEPTEMBER 10, 1956

 

Fig. 3.2.19, Cold-Leg Surfaces of Inconel-Sodium Thermal-Convection Loops Showing Mass-Transfer
Deposits. The decreased width of the scratch ot the edge of the specimens indicates that this deposit
is quite hard (possibly a carbide) as compared with Inconel or nickel plate. (a) Cold-leg specimen from
loop No. 28 which included a diffusion cold trap. (&) Cold-leg specimen from loop No. 29 which had no
cold trap. Etched with aqua regia. 1000X. Reduced 6%. {Cenfidentiatwithraprion—

leg. All the loops were constructed of Y%-in,
sched-40 pipe (0.84 in. OD, 0.622 in. ID).

The {ow-sodium-content lithium used in these
tests was received from the vendor packed in gas-
tight helium-filled containers, The loops were

foaded with lithium and. inert-gas arc-welded.

Inconel (nommal ‘composition, in wt %: Nl, 77;
Cr, 15; Fe, 7) showed extensive mass transfer and
_hot-leg attack, even at hot-leg temperatures as
low as.1300°F." This result is typical of nickel-
base -alloys exposed to Ilfhlum under these con-

dmons. S

The six type 3]6 stainless steel (nommal com- .
position, in wt %: Fe, 68; Cr, 17; Ni, 12; Mo, 2)"

loops tested showed less mass transfer and less
of a ‘tendency to plug in the three tests conducted

with’ hot-leg temperatures of approximately 1500°F -

than in the other tests conducted at hot-zone tem-
peratures of approximately 1400°F and 1300°F
(Figs. 3.2.20 aond 3.2.21). The loop (No. 25)

operated with a hot-zone temperature of 1400°F

had more mass-transfer crystals in the cold leg
than did the loop (No. 16) operated with a hot-zone
temperature of 1310°F; however, loop No. 25
operated over seven times longer before it plugged
than did loop No. 16. The hot- and cold-leg
surfaces of loop No. 25 after the test are shown in

Fig. 3.2,22, and the hot- and cold-leg surfaces of

loop No. 16, which plugged in 290 hr, are shown in

Fig. 3.2.23. The formation of carbide crystals on
 the surfaces of lithium loops has been noted in

almost all tests conducted with stainless steels,
and it usually occurs in that section of the loop
which is at a temperature of approximately 1300°F.

The carbide deposits have been found in the hot

legs of some loops and in the cold legs of others.

‘Type 321 stainless steel (nominal composition,
in wt %: Fe, 70; Cr, 18; Ni, 10; Ti, 0.5) when
tested at a hot-leg temperature of 1500°F com-
pletely plugged with crystals in 204 hr, Type 347
stainless steel (nominal composition, in wt %:
Fe, 70; Cr, 18; Ni, 10; Nb, 1) showed results
similar to those for type 321 stainless steel, with

165

 
 

 

 

ANP PROJECT PROGRESS REPORT

TABLE 3.2.9. RESULTS OF THERMAL-CONVECTION LOOP TESTS OF VARIOUS ALLOYS

EXPOSED TO CIRCULATING LITHIUM

 

Mefalldgrnphic Results

 

 

 

System Temperatures (°F) Test . —
~ Material ~ Duration Hot-Leg _ R
' Hot Leg Cold Leg Differential ~ (hr) . Aftack ) Col_d Lg,g :
S , o : {mils) o
Inconel 1300 1200 100 " 1000 16 151 g of crystals .
Stainless steel .
Type 316 1500 1292 208 500 13 0.5 g of crystals
| 1490 1220 270 000 L5 0.1 g of crystals
1472 . 1335 117 1000 . 3 . 0.1 g of erystals
1400 1130 270 2150* 23 4.7 g of crystals
1310 1058 252 290* 2 0.8 g of crystals
1292 1094 198 1000 15 0.25 g of crystals
Type 321 1500 1220 280 204* 0.5 1.0 g of crystals
1310 980 330 1230* 3.0 0.7 g of crystals
Type 347 1500 1112 388 280* 2 1.5 g of crystals
1310 1060 250 1000 3.0 1.3 g of crystals
1000 618 382 1000 1 Crystals 0.2 mil thick
1000 618 382 3000 1.5 Crystals 0.3 mil thick
Type 430 1500 1220 280 1500 4.0 L0 g of crystals
Type 446 1500 1166 334 864* 1.0 9 g of crystals
(84 mils of attack)
1500 1200 300 700* 16 6.8 g of crystals
(84 mils of attack)
1292 1058 234 1500 25 0.23 g of crystals

{65 mils of attack)

 

*Loop plugged with crystals.

166
 

 

 

UNCLASSIFIED
Y-18640

   

(o) NS 'N;-CH__

 

(&)

 

Fig. 3.2.20. Type 316 Stainless Steel Thermal-
Convection Loop (No. 25) Which Circulated Lithium
for 2150 hr at a Hot-Zone Temperature of 1400°F
and a Cold-Zone Temperature of 1130°F, (a) Speci-
men from hot zone, (&) Specimen from celd zone.

plugging occurring in 280 hr when a hot-leg tem-
perature of 1500°F was used. A 3000-hr test of a
type 347 stainless steel loop with a hot-leg tem-
perature of 1000°F indicated that this material
would be satisfactory for operation in this tempera-
ture range for very long periods of time if a small
amount of attack and mass transfer could be
tolerated. .

A type 430 stainless steel (nominal composmon,
in wt %: Fe, 83; Cr, 16; C, 0.10) loop operated
at a hot-leg temperature of 1500°F (cold leg tem-
perature, 1220°F) for 1500 hr without plugging, but
it showed considerable mass transfer (Fig. 3.2.24).
Two type 446 stainless steel (nominal composition,

in wt %: Fe, 74; Cr, 25 C, 0.35):loops tested at

a hot-leg temperature of 1500°F (cold leg tempera-
ture, ]200°F) completely plugged in less ‘than =
900 hr and had 6.8 and 9 g of metal crystals in

PERIOD ENDING SEPTEMBER 10, 1956

UNCLASSIFIED
Y-18373

 

()

Fig. 3.2.21. Type 316 Stainless Steel Thermal-
Convection Loop (No. 16) Which Circulated Lithium
for 290 hr at a Hot-Zone Temperature of 1310°F
and a Cold-Zone Temperature of 1058°F. (a) Speci-
men from hot zone, (b) Specimen from cold zone.

(Eorfidentiai-with-eaption)

metals are transferred in appreciable quantities.
If the alloy contains nickel, the mass-transfer
crystals are richer in nickel than is the container

- alloy, . In iron-chromium alloy loops, the crystals
jdepoSited ‘are -richer in iron than is the container

the cold legs (Fig. 3.2.25). The wall-crystal inter-

face on the coldsleg surface is shown in Fig. 3.2.26,
The bulk of the crystals are ferrite (composition,:
in 'wt %: Fe, 90; Cr, ), while those attached to
the surface of the type 446 stainless steel wall

have been identified by x-ray analysis as Cr,,C,. -
The relative hardnesses of these two phases may ;

be noted in the illustration.

The “results of -chemical analysis of the mcss-:"

transfer crystals from the nickel-iron-chromium and
iron-chromium _ alloy loops show that all three

L

_alloy.  Of the three metals - nickel, iron, and
"__‘rchromlum ~ chromium has the least tendency to
“mass’ transfer, but-it is still found to be present
to the extent of 5 to 10% in the mass-transfer
crystals found in stainless steel=lithium tests. .

. The data obtained in these thermal-convection

f’loop tests indicate clearly: that further work needs

" to be ‘done to resolve the" apparent discrepancies
-which exist between the results for type 316 stain-

less steel and ‘other austenitic stainless steels
which contain little or no molybdenum, Also,
types 317 and 318 stainless steels, which contain

167

 
 

 

 

 

 

ANP PROJECT PROGRESS REPORT

227 VHN

25-g LOAD

14543 VHN

 

'266 VHN

 

 

 

 

 

Fig. 3.2.22. Hot- (2) ond Cold-Leg (b) Surfaces of Type 316 Stainless Steel Loop No. 25 Which
Circulated Lithium (See Fig. 3.2.20). Hot-leg surface attacked along grain-boundary carbides. Two-
phase mass-transfer crystal (chromium carbide, VHN, 1513) plus attached crystal containing iron, nickel,
and chromium are shown on cold-leg surface. Etched with glyceria regia. 500X. Reduced 13%. (Gent

dential-witirception

 

 

 

 

 

 

 

 

 

Fig. 3.2.23. Hot- (2) and Cold-Leg (b) Surfaces of Type 316 Stainless Steel Loop No. 16 Which
Circulated Lithium (See Fig. 3.2.21). Crystal on hot-leg surface is chromium carbide. Note mass-'

transfer crystals on cold-leg surface. Etched with glyceria regia. (Genfidentielwith-coption)

168

 
 

 

L

UNCLASSIFIED
“Y-18087

 

 

lal_..

 

 

(b)

Fig. 3.2.24. Type 430 Stainless Steel Thermal-
Convection Loop (No. 14) which Circulated Lithium
for 1500 hr at @ Hot-Zone Temperature of 1500°F
and a Cold-Zone Temperature of 1220°F. () Speci-
men from hot zone., () Specimen from cold zone,

more molybdenum than ls'present in type 316 stain-

less steel, will be tested to determine the effect

-of molybdenum on the mass-transfer tendency of -
stainless steels in contact ‘with molten lithium.

The ferrmc stainless steel test results are not

encouraging, especially those for type 446 stain- -
less steel, which suffers very deep mtergranular.
attack ‘because of. preferential- ‘attack’ of the grain-
boundary carbides. Further tests are- planned for -
type 430 stamless steel. “The superior corroslon',_t
resistance of tfype 430 stainless - steel, in com-
parison - wrth type 446 stainless steel, can be
attributed to the lower carbon ‘content of the ‘type . .
430 stainless steel (0 10% C maximum). Type 446

stainless steel has a maximum of 0.35% C.

'The data obtained in static and thermal-conVectian, -

loop tests ore summarized in Fig. 3,2.27, which
shows that the materials tested thus far are un-

PERIOD ENDING SEPTEMBER 10, 1956

satisfactory as containers for lithium at hot-leg

"vtemperatures of ‘approximately 1500°F in flowing

systems. Tests are presently belng conducted on

~niobium, - molybdenum, and zirconium in dynamic

systems, and it is believed that these refractory
metals, especnally molybdenum, will withstand the

. cotrosive action of lithium under these conditions,
- but, as yet, this has not been demonstrafed. ' The

effect of various. amounts of nitride cantammahon
on the amount of mass transfer will be studied
further. Experiments will be conducted in order to

~ determine the maximum. temperature (in the range
~ from 1000 to 1300°F) ot which a stainless steel-
'dynamic lithium system may be operated, since

these alloys at present seem to be satisfactory at
1000°F -and, in general, unsatisfactory at 1300°F.

" The temperature limits indicated in Fig. 3.2, 27 are

merely estimates arrived at by evaluating the data

4. 'obtained thus far, and they may. be in error by as
| much as 100°F.

TESTS OF lNTERMETALLIC COMPOUNDS IN
SODIUM AND IN FLUORIDE FUEL

W, H. Cook

The mtermetall:c compounds NlAI NiAl + 5% NI,
NiAl + 4% Zr, and MoAl have been screen tested
for 100 hr in static sodium and in static NaF- ~ZrF -
UF4 (53.5-40-6.5 mole %) at 1500°F. No attack
was found on the specimens tested in the sodium.
Those tested in the fused salt were severely
attacked. . The test results, which were obtained
by metallographlc comparisons of untested and
tested specimens, are presented in Table 3.2.10.
Chemical analyses of the test medlums are bemg
made, .

LAl the mtermetall:c compounds were found to
have a common phase that may have been ‘an oxide

added by the fabricator for strengthening.d - This
phase constituted less than 5% of the su'rfac'e‘ of

any -sectioned intermetallic ”co'mp'ouni:l it was

-homageneously distributed in-all specimens’ with
the exception of the N|A| + 5% Ni compound, in
~-which some ‘segregation was apparent. 'In addition

to this common phase, all ‘the intermetallic com-

' ,pounds had one major phase, except the NiAl + 4%
'Zr - compound, which contamed two -additional
‘phases.' _
',bon.c"le_d‘ globular  particles, and the three minor

The major phase was - present as well-

 

G, Sterh,‘ Refractory Type Materials for High Teme

.berature Application, NP-4527 (August 1953), p 139-140.

169

 
 

 

 

 

ANP PROJEC:I' PROGRESS REPORT

 

B CRYSTALS

e

Fig. 3.2.25.

B UNCLASSIFIED 8
g Y-14932

§ FLL poT E

 

 

Type 446 Stainless Steel Thermal-Convection Loop Which Circulated Lithium for

864 hr. (Eonfidentiot-withTaption

170

 
 

| _ - "PERIOD ENDING SEPTEMBER 10, 1956

 

 

 

 

 

 

 

 

”
Fig. 3.2.26. Cold-Leg Surface of Type 446 Stainless Steel Loop Which Circulated Lithium (See
Fig. 3.2.25). Very hard phase (VHN, ~1100) is Cr,,C,, while soft crystal attached to it is a ferrite
crystal containing 90% Fe and 6% Cr. Efched wnth glyceria regia. (Een{-rd'en+ru-|-m+'l-r-eu-pt-oon) '
TABLE 3.2.10. SUMMARY OF THE RESULTS OF EXPOSURE OF INTERMETALLIC COMPOUNDS
FOR 100 hr IN STATIC NflF-Zl‘F4-UF4 (53.5—40-6.5 MOLE %) AT 1500°F
1 Hic o Attack (mils)
ntermeta € — — — Metallographic Results
'CP""PQUHJ : Maximum Average Minimum :
MoAl - L o C SR Quanfltaflve measurements could not be made on the
' ' ST T - Mo Al compound becnuse of its extreme brittleness,
but visual examination indicated that n hud been
: _ ~attacked by ihe fused snlt 7
L N:A! S 20 o 15 8 The 6tfack pro_duced alternate zones (bands) of
' L : 7 R - B R - various degrees of porosify"pqrallei‘ to the surfaces
L e B L L S ~ of the specimen.
: : NiAD+ 5% NTRE '7,-'32 e 3 o The cnd of the specimen where 1he maxamurn qflack
o b i o - - was found did not appear to ‘be bonded as weli as
* Ve m | L e - :th_e re_mgmder of the_sp:ec:men '
1‘ | NiAl + 4% Zr 74 61 8 The attack was severe to the depths indicated, but it
) S ' was most severe along the edges fer an average

depth of 6 mils

 

171

 
 

 

ANP PROJECT PROGRESS REPORT

UNCLASSIFIED
ORNL-LR-DWG {2080A

 

IRON

 

LOW-ALLOY STEELS

 

 

 

 

FERRITIC (Fe —Cr)
STAINLESS STEELS

  

 

T FRRRRTRFERTTTETTR

 

AUSTENITIC (Fe—Ni—Cr)
STAINLESS STEELS

 

NICKEL

 

 

N R TITRER

 

NICKEL -BASE ALLOYS

T I imamarars™

 

(INCONEL)

 

 

 

REFRACTORY METALS
{Mo, Nb, Ta, Zr, Ti, W, Va)

PRECIOUS METALS
Y POOR RESISTANCE
(Ag, Au, Pt) VER lOO EIS S (IJ

 

PREDICTED

 

 

 

 

 

 

 

NN STATIC SYSTEMS

RESN DYNAMIC SYSTEMS

 

Fig. 3.2.27. Corrosion Resistance of Various Metals and Alloys in Lithium. Bars indicate approxi-
mate temperatures below which a system might be operated for 1000 hr with less than 0.005 in. of ettack

or container surface removal. (Cenfidentiel-withcaption)

phases were located in the few open spaces be-
tween the particles of the major phase. The
NiAl + 5% Ni specimens were the only ones in
which potosity was found, and it was negligible in
these,

_For the NiAl-base intermetallic compounds the
tused-salt attack was probably due to the aluminum.
The attack began intergranularly and advanced into
the grains. The differences in the extent of the
fused salt attack on the NiAl-base intermetallic
compounds was probably caused by differences in
the accessibility of the NiAl to the fused salt,
In the NiAl + 5% Ni specimen it was noted that the
attack was most severe where the ‘‘oxide’’ phase
was concentrated. It is assumed that oxide quickly

172

corroded away and allowed the fused salt to attack
the intermetallic compound.

The severe attack in the case of the NiAl + 4%
Zr specimen can probably be explained by the
removal of both the common phase and the zir-
conium, These corrosion tests indicate that the
NiAl compound in the NiAl-base intermetallic com-
pound was attacked when it was exposed to NaF-

ZrF UF,, (53.5-40-6.5 mole %) at 1500°F.

The poor corrosion resistance of these inter-
metallic compounds in the fused salt obviates
further tests in this medium, However, the results
of the sodium tests are encouraging enough to
warrant more severe corrosion tests in sodium.
 

 

X

PERIOD ENDING SEPTEMBER 10, 1956

3.3. FABRICATION RESEARCH
J. H. Coobs

NICKEL-MOLYBDENUM ALLOY
DEVELOPMENT

T. K. Roche

The search was continued for a superior nickel-
molybdenum-base alloy with the corrosion re-
sistance of Hastelloy B and elevated-temperature
strength at least equal to that of Hastelloy X.
Seamiess tubing has been fabricated from the
various alloys under consideration to provide
material for evaluation in thermal-convection
loops. The extrusion techniques described previ-
ously! have been successfully used to produce
tube blanks of the variety of alloy compositions
being studied.

The nickel-molybdenum-base-alloy tube blanks
submitted to the Superior Tube Co. for redrawing
responded satisfactorily, in general, with the
several exceptions being notably the high-carbon-
content compositions. The processing procedure
used for the as-extruded, 1.5-in.-OD, 0.25-in.-wall
tube blanks is to (1) condition the tube blanks
by machining (resultant average weight loss, 50%);
(2) reduce to 0.875-in. OD and 0.095-in. wall
tubing; (3) anneal at 2050°F in a dry hydrogen
atmosphere and water quench; (4) rod draw and
sink to final size, with intermediate anneals; and
(5) finish by sand blasting and rotary polishing.
An average yield of 70% is being obtained from
the blanks after the initial conditioning operation.
It has been found in the processing of these
alloys that lighter cold reductions per pass must
be taken than in the case of stainless steel or
Inconel. A 20% reduction in area per pass is
considered to be optimum. ' '

H. Inouye

Battelle Memorial Institute Alloys

Ten tube blanks of three different alloys were
extruded and submitted to Superior Tube. Co. for
redrawing in order to assist Battelle Memorial
Institute “with nickel-molybdenum alloy develop-
ment work - under. way there.! The results of
processing these extrusions to 0.500-in.-0D,
0.035-in.-wall tubing are presented in Table 3.3.1.

 

H. Inofiye and T. K. Roche, ANP Quar. Prog. Rep.
June 10, 1956, ORNL-2106, p 163.

The response of these compositions to cold tube-
forming techniques was only moderately suc-
cessful. The difficulties encountered are illus-
trated in Fig. 3.3.1. Metallographic examination
of the failure found in the tube blank of alloy
B-2898 (20% Mo-1% Nb-2% Ti-0.80% Mn-0.12%
C-bal Ni) revealed the stringering of second-phase
particles (that is, carbides), as shown in Fig. 3.3.2,
to be the probable source of the difficulty.

All attempts to produce seamless tubing of alloy
B-2899 (20% Mo-1% Nb-0.80% Mn-0.20% C-bal
Ni) failed. Two of the four extrusions of this
alloy developed longitudinal cracks along their
lengths, as shown in Fig. 3.3.1, while on the
draw bench. The other extrusions, also shown
in Fig. 3.3.1, failed during the initial tube-reducing
operation, Examination of the cracks again
revealed that stringering of the numerous precipi-
tates was the source of failure.

The results indicate that carbon in amounts of
0.20% or greater in the nickel-molybdenum alloys
being ~ investigated will make improbable the
success of the tube-forming process and that
0.12% carbon is marginal in this respect.

UNCLASSIFIED
Y-19445

 

Heat B-2899, Extrusion 64
Alloy Composition: 20% Mo—1% Nb-0.80% Mn- 0.20%.C-Bal Ni

 

Heat B- 2899, Extrusion 60
Aloy Composition: Same as Above

 

Heat B-2898, Extrusion 107
Alloy Compaosition: 20% Mo-{% Nb-2% Ti~0.80% Mn-0.12%C - Bal Ni

Fig. 3.3.1. Examples of Failures Which Occurred
During Cold Reduction of Tube Blanks Fabricated
from Battelle Memorial Institute Nickel-Molybdenum
Alloys.

173

 
 

 

 

 

Vil

 

TABLE 3.3.1. RESULTS OF REDRAWING OF BATTELLE MEMORIAL INSTITUTE; INTERNATIONAL NICKEL COMPANY,
AND ORNL SPECIAL NICKEL MOLYBDENUM ALLOYS TO 0.500:in,-0D, 0,035-In.~WALL TUBING

 

 

All Nominal C 1 Number of Total Length "d'*' _ bor of L
\ oy ominal Composition Extruded Tube of Tubing - Yie . | N':’m Fr : l OOZ:* Remarks
O (wt %) Blanks Received (%) Being Fabricate
Battelie Memorial Institute Alloys
B-2897 20 Mo=1 Nb-? Ti-0.80 Mn-0.12 C—-bal Ni 2 9f2in, 40 1
B-2898 20 Mo~1 Nb-2 Ti-0.80 Mn—0,12 C—bal Ni 4 19 ft 6 in. 2 One tube blank split on
’ ; ' tube reducer

B-2899 20 Mo=1 Nb~0.80 Mn..0,20 C=bal Ni 4 0 0 0 " Two tube blanks split
on tube reducer; two
split on draw bench

International Nickel Company Alloys

T-23011 15 Mo=5 Cr-3 Nb-3 W-0,5 Al~0.02 C~bal Ni 2 11 #t 11 in. 43 1

T-23012 17 Mo-0.5 A1-0,02 C~bal Ni 1 12 ft 69 1

T-23013 15 Mo=3 Nb-3 W_0,5 A1-0,02 C~bal Ni 3 27 ft 6 in, 76 2

T-23014 15 Mo=1,5 Ti=1 Al—-0,02 C~bal Ni 1 7 ft4in, 80 0 Both blanks failed on
tube reducer

ORNL Alloys

30-1 17 Mo=3 Cr=0.06 C=bal Ni 3 30 ft 10 in, 76 2

37A-1 20 Mo-3 Cr-0.02 C~bal Ni 1 8 ft 2 in. - 69 1

30-2 17 Mo~5 Cr=0.06 C—bal Ni 3 36 # 3 in. 81 '3

43A-3 20 Mo~7 Cr—0.02 C~bal Ni 1 8 ft 4 in. 73 . 1

30-4 21 #1 6 in. 81 2

17 Mo=10 Cr-0.06 C-bal Ni 2

 

*Percentage yield of tubing from conditioned tube blank,
**Approximately 10 ft of tubing is required to fabricate a thermal-convection loop,

1A0dIY SS3¥I0Ad LIO3r0dd dNV

 
 

av

The compositions of new alloys that have been
received from Battelle Memorial Institute for
evaluation are presented in Table 3.3.2. These
alloys were received in the form of forged ex-
trusion b:llets, swaged impact-tensile specimens,
and rolled creep-rupture specimens, The impact-

PERIOD ENDING SEPTEMBER 10, 1956

tensile specimens were sent to Rensselaer
Polytechnic Institute for weldability studies, and
the creep-rupture specimens will be tested at
ORNL in the fuel mixture (No. 107) NaF- KF-LlF-
UF, (l] .2-41-45.3-2.5 mole %). Extrusion experi-

mem‘s were conducted at ORNL on two forged

 

- Fig. 3.3.2. Longitudmal Section of Tube Blank of Alloy B.2898, Showing Crack Propugnhon Along
Carb:de Sf;ingers Thai Occurred Durlng Tube Reducmg. : Etched wuh chrome regm. IOOX

TAB LE'f_3'.V3'..'2. - COMPOSITION OF NEW -A'LLofs, RECEIVED FROM
 BATTELLE MEM'oR_l_Al: _ms‘r_rr_uTE FOR _EYALUATION c

 

Nommul Composiflon {wt %)

 

 

Alloy No.. Llmon ‘ , | . —
D e T e W e e R W
- B35 20 o 0-.1.2: . ".‘,'-0.'50":'.:.:}" e 7 . . Bal
-6;32})6»" - 20 o '“'iz,z-’ .: _ :.'6{36' : 2 7 7 Bl
[B-327'7T-’;f’ Coa00 a2 080 i 2 T

| 2 7 . Bal

B8-3278 20 0.12 - 0.80

 

175

 
 

 

ANP PROJECT PROGRESS REPORT

billets of each composition to investigate the
feasibility of producing seamless tubing of the
alloys. The tube blanks were extruded at 2150°F
at a ratio of 7:1. All compositions were fabricated
successfully and sent to Superior Tube Co. for
redrawing.

International Nickel Company Alloys

Tube blanks which were fabricated from air-

melted nickel-molybdenum base alloys supplied by

the International Nickel Company have been
returned by Superior Tube Co. after having been
processed to 0.500-in.-OD, 0.035-in.-wall tubing.
The results of the processing are given in
Table 3.3.1. '

For the most part the limited data indicate that
these ' International Nickel Company alloys are
amenable to tube-producing practice, with one
exception. Alloy T-23015, which contained 0.25%
C, could not be successfully reduced to tubing.

Although the failures from these particular tube
blanks were not examined metallographically, it
is felt that carbide stringers, similar to those
shown in Fig. 3.3.2, were again the cause of the

difficulty., Thermal-convection loops are to be -

constructed from the various lengths of tubing
received.

ORNL Alloys

Extrusion experiments outlined previously! for
producing seamless tubing of ternary alloys con-
taining nickel, 15 to 20% molybdenum, and a third
element were continved, with the intention of
determining the amount of a particular strengthening
element which can be tolerated without serious
effect on the corrosion resistance of the base
composition to fuel No.  107. The tube blanks
which were fabricated during the quarter and
submitted to Superior Tube Co. for redrawing are
listed in Table 3.3.3. The extrusion of these

TABLE 3.3.3. ORNL ALLOYS EXTRUDED FOR TUBING FABRICATION
AND CORROSION EVALUATION

 

Alloy No. Nominal Composition (wt %)

Number of Extruded

 

' Tube Blanks
30-6 17 Mo—=0.06 C~7 Cr—bal Ni 2
30-7 17 Mo~0.06 C=2 Al-bal Ni 3
30-17 17 Mo—0.06 C—4 Al-bal Ni 2
30-8 17 Mo~0.06 C~2 Ti~bal Ni 3
30-18 17 Mo=0.06 C—4 Ti-bal Ni 3
30-9 17 Mo~0.06 C—2 W-bal Ni 3
30-19 17 M0~0.06 C~4 W=bal Ni 3
30-10 17 Mo=0.06 C=2 V~bal Ni 3
30-20 17 Mo—0,06 C—4 V—bal Ni 3
3011 17 Mo=0.06 C—4 Fe—bal Ni 3
30-12 17 Mo—0.06 C—3 Nb—ba! Ni 3
30-21 17 Mo=0.06 C—5 Nb—bal Ni .
30-13 (INOR-3) 16 Mo=1.5 Ti—1 Al-0.06 C—bal Ni 3
30-14 (INOR-4) 16 Mo~1.5 Ti=2 Al..0.06 C—-ba! Ni 3
30-15 (INOR-5) 15 Mo~2 Nb—2 W~0.06 C—bal Ni *
30-16 (INOR-6) 16 Mo—5 Cra1.5 Ti=1 Al-0.06 C~bal Ni 3
30-22 (INOR-7) 16 Mo—=6 Cr~1 Nb—1 Al-0.08 C~bal Ni *

 

*Extrusion not yet made,

176
 

 

 

o

alloys was carried out at 2150°F at a ratio of 7:1.

The extrusion techniques described previously

were used, and no difficulty was encountered with
these ternary alloys, except the 79% Ni-17%
Mo—4% Al alloy, which offered relatively high
resistance to plastic deformation. This is an
indication of the potency of aluminum in increasing
the elevated-temperature strength of the base
composition,

In addition to the ternary alloys, 36-1b vacuum
induction melts of the new alloys INOR-3, -4,
and -6 were prepared and extruded to tube blanks.
The compositions of these alloys and the results
of the extrusions are given in Table 3.3.3. The
purposes of this series of experiments were to
determine the extrudability of these afloys and to
gain a supply of tubing for corrosion testing, As
indicated in Table 3.3.3, three billets of eoch
alloy were successfully extruded. These ex-
trusions were carried out at 2150°F at a ratio
of 7:1.

The quantities of tubing obtmned from extruded
tube blanks previously submitted to Superior Tube
Co. are listed in Table 3.3.1. In general the
redrawing of the chromium-bearing alloys with
intermediate carbon contents (0.06% or less)
proved to be satisfactory. The percentage yields
of 0.500-in,-0D, 0.035-in.-wali tubing from con-
ditioned tube blanks, as well -as the total length
of tubing of each composition, are also shown
in Table 3.3.1. This tubing is to be fabricated

into thermal-convection loops for corrosion testing.

Hastelloy W Seamless Tubing

Results of processing a tube blank of Hastelloy

W, which was successfully extruded at ORNL, to

0.187-in.-0D, 0.025-in.swall tubing indicate the
feasibility of producing small-diameter seamless -
~ tubing of Hastelloy-type alloys for heat exchanger

application. Ayield of 25 ft of tubing was obtained

from this extrusion. Longitudinal and transverse 7
sections of the tube wall are shown in F:g. 3.3.3.

Consumable-Elechode Expenments

Electrodes of mckel and four nickel-base ulloys,' ‘_
-~ 83% Ni-17% Mo, 76% N|—l7% o-7% Cr, Hastelloyv"_
‘B, and Hostelloy W, which were submitted to
Battelle Memorial Institute for consumable-elecs

trode arc melting, have been melted and returned
to ORNL for evaluation. Suitable test specimens
will be fabricated from the ingots to determine

PERIOD ENDING SEPTEMBER 10, 1956

whether arc-melting is mstrumental in |mprovmg
the properhes of these alloys.

SHIELD PLUG FOR ART PUMPS
J. H. Coobs J. P. Page
Gamma-Ray Shielding Material

The thermal conductivities of the six tungsten
carbide—constantan specimens described in the
previous report? were determined, and the data
are presented in Table 3.3.4.

A plot of these data as functions of the volume
percentages of tungsten carbide, constantan, and
pores is presented in Fig. 3.3.4. Examination of
this graph shows the thermal conductivity of this
system to be primarily a function of the tungsten
carbide content, with only minor effects resulting
from variations in porosity. These specimens
bracket quite completely the composition range at
a bulk density of 12.0 g/cm?® ond indicate fairly

 conclusively that the specifications for the gamma-

ray shielding materia! (density, 12.0 g/cm3 ~ mini-
mum; conductivity, 0.10 cal/cm.sec:°C —~ maximum)
cannot be met by the tungsten carbide—constontan
material.

A search for a lower conductivity nickel-base

~alloy for use as a binder resulted in the selection

of Hastelloy C. This alloy has a reported conduc-
tivity of 0.03 cal/em.sec-°C, as compared to

 

25, H. Coobs and J. P, Page, ANP Quar, Prog.
Rep, ]une 10, 1956, ORNL-2106, p 173.

TABLE 3.3.4. THERMAL CONDUCTIVITY OF
EXPERIMENTAL COMPOSITES OF
TUNGSTEN CARBIDE AND CONSTANTAN

 

: Composition
Densify o (fi' %)
(g/em®y —

Thermal Conductivity

—_— e {eal .sec®C
wC _Constanfun (cal/em see )

 

153 57 4o

N 7 29 012
0 85 15 012
| 1:1.1'.'55. w6 | 0.12
1296 79 21 0.13

1267 93 7 0.15

 

177

 
 

 

 

ANP PROJECT PROGRESS REPORT

 

 

Fig. 3.3.3. Longitudinal (a) aond Transverse (b) Sections of 0.187-in.-OD, 0.025-in.-Wall Seamless
Hastelloy W Tubing.

178

 
 

 

ge

 

UNCLASSIFIED
ORNL-LR-DWG 15034

   

 

~ TUNGSTEN {0 20 30 40
CARBIDE
CONSTANTAN {(vol %)

- Fig. 3.3.4. Thermal Conductivity (cal/cm-sec.°C)

of Hot-Pressed Tungsten Carbide~Constantan
Gamma-Ray Shielding Material as a Function of

Composition and Porosity.

0.06 cal/cm-sec:°C for constantan. Also, pre-

alloyed powder is commercially available at a
reasonable price (approximately $5/1b).

Four tungsten carbide~Hastelloy C thermal-
conductivity specimens were hot pressed, ma-
chined, and tested. The results are shown in
Table 3.3.5.

These data are plotted in Fig. 3.3.5, again as
functions of the three composition parameters:
tungsten carbide, Hastelloy C, and porosity. It is
evident that the specifications for the gamma-ray
shielding material are satisfactorily fulfilled by
this material.

The 70 wt % tungsten carbide~30 wt % Hastelloy
C specimen was considered to be the easiest to
hot press ond was tentatively chosen as the
material for the ART shield plug. Two models of
the shield plug have been hot pressed from this
70 wt % tungsten carbide materlal that hove the
followmg dimensions and densi :ty

Model'l  Model 2 : _

' lnslde dwmeter, ine . 1495 ‘_727‘.3762_ _
Outside dlameter, in. 2.229 3-860 7 |
| . ‘He-ghf..m,- o 1aas 1740
Densityygfen® S0 a5

These models are not true MII'IIO'I'UI'eS of the -
plugs, but they have the same ratio of wall-surfoce-i_
area to- cross-secflonoi area as. the shle!d plugs,'._-'

this ratio is on important variable in the hot-
pressing operation. Neither of these models was

PERIOD ENDING SEPTEMBER 10, 1956

TABLE 3.3.5. THERMAL CONDUCTIVITY OF
EXPERIMENTAL COMPOSITES OF
TUNGSTEN CARBIDE AND HASTELLOY C

 

, Composition
Density Thermal Conductivity
3 (wt %) 0
(g/em?y —M8M ———— (cal/cmesec+"C)
WC  Hoastelloy C

 

11.80 92 8 0.08
1.99 81 19 0.07
13.45 90 10 0.13
1190 70 30 0.07

 

UNCLASSIFIED
ORNL-LR-DWG 45030

  

 

TUNGSTEN 10 20 30 40
R
CARBIDE HASTELLOY G (vol %)

Fig. 3.3.5. Thermal Conductivity (cal/cm.sec-°C)

~ of Hot-Pressed Tungsted Carbide~Hastelloy C

Gamma-Ray Shielding Material as a Function of
Composition and Poresity.

_difficult to press, although an alumina coating
-~ on the grcphlte die is required to prevent a
- graphite-nickel reaction at the. pressing temper-

oture. Unless severe temperature gradients are
~ ~encountered in the larger die to be used for the

'ART plug, it should be possnble to hot press the
~ plug as one piece.

‘A secondary specification is that the material

" be readily brazed to Inconel. The 70 wt % tungsten

cnrblde material sahsfactorlly meets this re-

- quirement. - Copper readlly wets the’ composate,

and the coefficients of expansion of the composite

~and of Inconel appear to be similar. Several small

fest pieces. and the two models have been suc-
cessfully copper-brazed to Inconel plate.

179

 
 

 

 

ANP PROJECT PROGRESS REPORT

Thermal Shield

Three specimens of ZrO, were fabricated at
ORNL and sent to Bottelle KAemorial Institute for
thermal conductivity determinations. Results for
the first two specimens tested are presented in

Teble 3.3.6.

TABLE 3,3.6. THERMAL CONDUCTIVITY
OF ZrO, SPECIMENS

 

Thermal Conductivity

 

 

Test (cal/cmesecs®C)
Tem(p: (r:t;fure Specimen with Specimen with
Density of Density of
3.52 g/cm’ 3.08 g/em®
20 0.00251 0.00163
100 0.00244 0.00158
200 0.00235 0.00153
300 0.00225 0.00146
400 0.00218 0.00141
500 0.00211 0.00134
600 0.00203 0.00129
700 0.00196 0.00122
800 0.00192 0.00115

 

NEUTRON SHIELD MATERIAL FOR THE ART
J. H. Coocbs M. R. D’Amore3
Cercmic B4C Tiles

The configuration of the neutron shield for the

- ART was described previously.# The contract

for fabrication of the ceramic B,C tiles for the
ART was awarded to the Norton Company during
the quarter. The tiles will be fabricated by hot-
pressing high-purity B ,C powder to a minimum
guaranteed density of 1.7 g of boron per cubic
centimeter,

Clad Copper-B «C Cermets

The Allegheny Ludlum Steel Corp. is being
considered as a potential supplier of the 0.100-in.-

 

30n assignment from Pratt & Whitney Aircraft,

“M. R. D*Amore and J. H. Coobs, ANP Quar. Prog,
Rep. March 10, 1956, ORNL-2061, p 151.

180

thick type 430 stainless steel—clad copper-B4C

(6.6 wt % B C) neutron shield material. A plate
of this material that was fabricated by Allegheny
Ludlum Steel Corp. was evaluated, The plate was
approximately 7% x 23 in. and 0.100 in. thick,

The evaluation included examination of the-

surface finish, radiography, bend fests, tensile
tests, thickness uniformity measurements, core
density measurements, microstructure examinations,
and investigation of the integrity of the clad«to-
core bonding. No pinholes were evident in the
10-mil-thick cladding stripped from the core
material, although the surface finish on the
cladding was rough. No.segregation of the B,C
or cracking of the core was observed, and the
clad was well-bonded to the core material. The
density of the copper-B 4C cermet, as measured by
the water-displacement method, was 97.9% of
theoretical. Small specimens that had been cut
parallel and transverse to the rolling direction
were bent to a radius of curvature of %, and
27, in., respectively, before fracture occurred,
The plate tapered along the length from a maximum
thickness of 0.099 in. to a minimum thickness of
0.092 in. The dispersion of B C particles in the
copper matrix was excellent. The room-temper-
ature tensile strength data for the material are
presented in Table 3.3.7.

The tensile specimens were punched out with a
blanking die and were 5 in. long with a 2-in, gage
length. The as-punched tensile specimen showed
no elongation and failed at the yield point of the
material, Specimen 4 was prepared by stripping
the cladding from the copper-B 4C core prior to
punching the specimen. :

LITHIUM-MAGNESIUM ALLOYS
R. E. McDonald?
Efforts to develop a 20% Li-80% Mg alloy for

shielding application were continued, with the
possibility of roll cladding the clloy being investi-
gated further. As reported earlier,® 25 aluminum
forms a brittle intermetallic compound with the
alloy, and, in an effort to overcome this difficulty,
nickel foil was used as a diffusion barrier; some
bonding resulted. Work will be continved with
barrier materials during the next quarter.

 

50n assignment from Pratt & Whitney Aircra.fl.—

6R. E. McDonald and C. F, Leitten, Jr., ANP Quar,
Prog, Rep. June 10, 1956, ORNL-2106, p 173.
 

ke

<

 

g

 

 

PERIOD ENDING SEPTEMBER 10, 1956

TABLE 3.3.7. ROOM-TEMPERATURE STRENGTH OF TYPE 430 STAINLESS
STEEL CLAD COPPER-B,C CERMETS

 

Specimen

ield t
Vield Strength at o sile Strength  Elongation

 

No. Orientation Condition 0.2 % C')ffsef (psi) % in 2 in.)
(psi)

1 Trunsvefse torolling  As punched 24,500 24,500 0.0
direction

2 Transverse torolling  Annealed 20 min at 20,000 31,200 5.6
direction 1650°F

3 Parallel to rolling Annealed 20 min at 19,000 34,000 7.5
direction 1650°F '

4 Core material unclad; Annealed 20 min at 15,100 21,400 3.5
transverse to rolling 1650°F
direction

 

Creep and stress-rupture data obtained for this
lithium-magnesium alloy are reported in Chap, 3.5,
*‘Mechanical Properties Studies.”” Further work is
being done on the powder-metallurgy approach to
the preparation of this shield material. Lithium
orthosilicate is being tested as o substitute for
lithium oxide, since it is less reactive in air and
water and easier to handle. However, it has only
one-half the lithium content of lithium oxide.

TUBULAR CONTROL RODS

M. R. D’Amore

The feasibility study of extrusion of tubular
control rods, described previously,” was con-

tinved, A simulated control rod billet consushng ,
of a type 316 stainless -steel can with a core
 composed of 34 vol % AI,‘,O3 (to simulate Lindsay
“oxide) m nickel powder was exfruded at 2150°F -
to a1} -m.-OD /-m.-lD “tube. " The ‘billet was -
.deSIgne to determme the feasubtllty of extrudmg;
o 'thm (obout 0. OSO-m.-thlck) inner and outer claddmgr’
on - control -rod cores ‘and of prepormg cores by -

-fampmg the Ioose powder mixiure into the blllet{'r'
L followmg ranges' I to 3 e 1to 200 3 ‘and 52 to

200 ..

CGHS. : '_ ‘,*;,'7.

Evaluation- of the extruded tube showed flmt the_ -' .
'cloddmg thlckness was . sotlsfoctory. ‘However, -

'tompmg of the 1oose powder does not appear to be'
, L 7 T —,'I-kg somple of the - Lmdsay omde .was converted

" into latger particles by pressing the as+eceived

 

-7y, H. Coobs, R. E. McDonaH, and M. R. D'Amore,
AI;JP3 Quar, Prog, Rep. March 10, 1956, ORNL-2061,
p 163.

promising as a core-preparation method., Trans-
verse sections of the extruded tube showed the
core cross section to be fairly uniform, except in
one area. The nonuniform area can be atiributed
to the sintering and concomitant shrinkage of the
core during the preheating operation, which can
cause longitudinal cracks in the core as a result
of shrinkage around the inner cladding layer or the
formation of void areas between the core and the
outer cladding layer. The void areas can, in turn,

~ cause buckling of the outer cladding layer during

extrusion,
A Hastelloy X billet containing a hot-pressed
core of 30 wt % Lindsay oxide=70 wt % nickel

| ~has been prepared for extrusion, Tensile speci-
~.mens are being machined from plates of 30 wt %
~ Lindsay oxlde-—70 wt % nickel cermet clad with

inconel, which were fabncated by hot rolling and

5‘_usmg the p:cture-frame techmque. - The specimens
- will be used to investigate the effect of the
2 Lmdsuy oxide particle size on the elevated-
- ‘temperature ‘tensile strength of the cermet. The

oxide particle sizes Beung mvestlgofed are in the

-As-received Lmdsay oxtde is @ mixture of very
fine -particles that are about 1 to 3 g in size. A

powder to a low-density compact and sintering
at 2460°F in air. The sintered compacts were then

181

 
 

 

 

ANP PROJECT PROGRESS REPORT

crushed ond ground in @ micropulverizer to attain
the desired particle size.
- Thermal conductivity data were obtained for a
30 wt % Lindsay oxide~70 wt % nickel specimen
at temperatures between 165 and 554°C, and the
results are shown in graph form in Fig. 3.3.6.
-The slopes and intercepts of the two lines were
calculated from test results by using the least-
squares methed.

" SEAMLESS TUBULAR FUEL ELEMENTS
M. R. D’Amore

Two three-ply blanks containing cores of 30

vel % Al,O; (to simulate UQ,) dispersed in type

302B stainless steel powder were redrawn from
1.0-in.-0D, 0.125-in.-wail tubing to 0.187-in.-OD,
0.015-in.-wall tubing at the Superior Tube Co.
The inner and outer layers of the three-ply
composites were type 316 stainless steel. The
finished tubing and samples taken at various
~'stages in the reduction schedule have been re-
ceived and evaluated.

The first extruded tube, for which the core was
prepared by hot pressing a mixture of <325 mesh
particle size Al .05 and stainless powder mixture,
exhibited tenslle g-actures in the core after 30%
total reduction. The finished tubing had severe
fractures in the core,

The second extruded tube blank was prepared by
tamping the powder mixture (=105 +325 mesh
Al,O, in type 302B stainless steel) into the

0.40

billet can. Examination of the redrawn tubing did
not reveal any tensile fractures in the core. How-
ever, the relatively large particles of A|203 were
fractured, and they tended to form stringers that
finally resulted in longitudinal cracking of the core
as the total reduction values increased. Radio-
graphs of the finished tubing revealed the cere
fractures in tube No. 1 but did not show the
longitudinal cracks in the core of tube No. 2.

It is evident from the examinations of the re-
drawn’ tubing that the core containing the larger
Alzo3 particles was stronger than the core with
the fine Al o particle dispersion. Tamping of a
loose powdzer mixture into a billet can therefore
appears to be a satisfactory method of preparing
cores of 30 vol % Al,O, dispersed in 70 vol % of
type 302B stainless steel

The study of flow patterns in three-ply extruded
tubes containing cermet cores was continued with
the extrusion of two three-ply billets to 1Y-in.-OD,
3{‘-in.'lD tubes at 2150°F. The billet can material
was type 316 stainless steel, and the cores were
fabricated by hot pressing @ powder mixture of
type 304 stainless steel and 30 vol % Al,O,.
A 29-deg taper was machined into the ends of the
core of one billet in an effort to eliminate end
defects in the extruded tube and thus to improve
material recovery.  This taper resulted in a slight
reduction in the length of the defect.

The second billet contained a core made in three
sections to determine whether nonuniform areas

NPT TR
ORNL-LR-DWG 15029

 

0.09

 

/

 

o
9

 

'-._\

 

[

X (cal/cm-sec-"C)

o
o
o

 

8
+

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

0.03
0 50 100 150 200 250 300 350

400 450 500 550 €00 650 700 750 800

TEMPERATURE (°C)

Fig. 3.3.6. Thermal Conductivity of a 70% Nickel-30% Lindsay Oxide Specimen.

182
 

 

 

*

would appear at the interfaces between core
sections, The extruded tube is being sectioned
for examination,

NIOBIUM FABRICATION STUDIES

V. M. Kolba®
1. P. Page

7 Evaluation of Ar-c-Melli'e'd 'N‘i_éb'ium

H. Inocuye

The physical and mechanical properties of
arc-melted niobium are being investigated in order
to compare this material with niobium prepared by
conventional powder-metallurgy techniques. Macro-
etching of a section of an arc-melted, cast, and
cold-worked ingot has revealed the large-grained,

‘severely distorted structure. that is usually ob-

tained through such a fabrication process. An

ottempt to cold roll a section of as-wreceived

y.in. plate was terminated after two 3% passes
when some end cracking was noticed.

The cracked ends were cut off, and the re-
maining material was annealed 1 hr ot 1100°C at @
vacuum of 2 x 10=5 mm Hg. This 0.485-in.-thick
section was then successfully cold rolled to
10-mil strip, a reduction in thickness of about
98%. The hardness increased rapidly from 160
VHN (2.5-kg load) to 190 VHN in the first few
passes, and then it increased very siowly to
206 at 98% reduction in thickness. Samples were
taken at several stages in the reduction, split
into thirds, and annealed for Y hr at 850, 1050,
and 1150°C in high vacuum, 'fi'\ese samples will
be examined metallographically to determine the
amount of reduction necessary to produce fme-
grained materloi after annealung. -

Nb-UO Compocts e

 

%on as’si'game'm from The Glenn L;- Mdrtih‘Co."_'_:‘ s

PERIOD ENDING SEPTEMBER 10, 1958

somewhat contaminated, The analytical results
are presented in Table 3.3.8. The Nb-3-C batch
was chosen to make five 70 wt % Nb-30 wt % UO,
compacts for sintering and reaction studies. The
UO, used had previously been high-fired in a
hydrogen atmosphere. The compacts were 82%
dense, as compacted, and had good ‘‘green®
strength. These compacts are being sent to
GE-ANPD for sintering at 2000°C. Attempts to
secure high-purity powder from commercial sources
have been unsuccessful,

Nb-U Alloys

A finger melt of an 80% Nb~20% U alloy was
made. Segregation noted in the initial casting was

~ reduced by subsequent remelting. The hardness of

the as-cast sample was R 86. Attempts to cold
roll the as-cast material proved to be unsuccessful.
A sample of the alloy was then encapsulated and
hot rolled successfully at 1050°C to a reduction
of 85%. Metallographic examination of the hot-
rolled structure revealed that the as-cast structure
had not been broken up during the rolling process.
The as«olled hardness was R 95.

Vacuum-annealing tests are bemg conducted on -
the hot-rolled alloy at 1100, 1200, and 1300°C
to determine the recrystallization temperature of
the clloy. Further rolling studies will then be
made.

FABRICATION OF HYDRIDES
R. E. McDonald

A survey of available literature on the fabri-

 cation of hydrides has been completed, and @
- hydriding furnace and purification system is being
_ Two bc:tches of moblum powder made from :
o wroughf sheet were ana lyzed and were found to be _

built ‘(see Chap. 3.6, ‘‘Ceramic Research),

- Yttrium and zirconium and their alloys will be

-hydnded and fabricoted, and the compatibility
of various barrier materials will be studied.

TABLE 3.3.8. IMPURITIES FOUND BY ANALYSES OF TWO SAMPLES OF NIOBIUM POWDER

 

lmpurifies (wt %) .

 

 

""’S:dm-prorle:‘f L = - - Hordness

.c 0, Ny - H, ~ (VHN)

NB2-C 0042 029 0018 s2x 107t 292
Nb+3-C 0.103 0.18 0.019 5.8 x 1074 179

 

183

 
 

 

 

 

 

ANP PROJECT PROGRESS REPORT

3.4. WELDING AND BRAZING |NVESTIGATIONS
' P. Patriarca

FABRICATION OF PRIMARY NaK
PUMP VOLUTES

P. Patriarca E. J_; Wilson .

Preliminary experiments designed to determine
the approximate weld shrinkage to be expected in
the fabrication of primary NaK pump volutes were
conducted, and the results were reported.! (The
primary NaK pumps are the pumps for circulating
NaK in the ART fuelsto-NaK heat exchange
system,) The information obtained from these
tests has now been used to successfully fabricate
three Inconel volutes, one by the metallic-arc
welding procedure and two_ by inert-arc weldmg
procedures.

The shrinkage of volute-No, 1, fabricated by

‘using the metallicearc, (coated-electrode) welding -

procedure, compared favorably with that of simu-
.lated test components, -However, radiographic
inspection of the weld revealed porosity to an
‘unacceptable degree. The exclusive use of inert-
arc welding was therefore specified for the
~ joining of future volute halves in order to ensure
~acceptable quality,

 

1p. Patriarca and G, M. Slaughter, ANP Quar, Prog,
Rep. ]une 10, 1956, ORNL-2106, p 176.

UNCLASSIFIED
ORNL—LR—DWG 16166

 

 

 

 

' PASS ELECTRODE  ELECTRODE

NUMBER ~ SIZE {in) - MATERIAL  CURRENT (amp)
™ ¥ in. INCO 62 . 90
2 33 in. INCO 62 140
3 g in. INCO 62 170
4-8 g in. INCO 62 190

* ROOT PASS: 6 TACKS, EACH APPROXIMATELY Zin. IN LENGTH,
FOLLOWED 8Y TIE-INS BETWEEN TACKS.

Fig. 3.4.1, Procedure and Joint Design for the
Inert-Arc Welding of Primary NaK Pump Volute
Weld Shrinkage Test Pieces (No, 1). Test pieces
rotated in horizontal position.

184

It was recognized that the magnitiude of the weld
shrinkage would differ sugmflcantly for the two

welding processes, and, since the existing data

were not adequate for predlchng the dimensional
changes, an additional shrmkage test was con-
ducted, © Two test pieces were machined from
2-in. Incone! plate, as described previously,!
and welded - in accordance with the procedure
described in Fig. 3.4.1. Micrometer measurements
were made at four radial sections, as described in

Fig. 3.4.2, prior to and after each subsequent_

operation. The completed test piece is shown in
Fig. 3.4.3. The results of the micrometer measure-
ments are summarized in Table 3.4.1.

These data provided a first approximation of the
shrinkage to be expected when welding of volute
Ne. 2 was undertaken. - A 0.125-in. shrinkage
allowance was incorporated into the joint design
and, after machining, the volute halves were
scribed as shown in Fig. 3.4.4. . Micrometer

UNCLASSIFIED
ORNL - LR~ DWG 16167

 

 

 

 

 

 

Rl

Fig. 3.4.2. Details of Micrometer Measurements
on Welded Primary NaK Pump Veolute Weld
Shrinkage Test Pieces, Measurements made at
four radial sections (A through D) at 90-deg
intervals at positions 1, 2, and 3 (see Table
3.4.1). o : ‘

S
 

 

 

 

 

 

e

PERIOD ENDING SEPTEMBER 10, 1956

UNCLASSIFIED
PHOTO 17788

 

 

| F|g. 3-4.3.,‘. Completed Inert-Arc-Welded Primary NaK Pump Volute Weld Shrinkage Test Piece,

measurements were made before and after each

operation at each intersection shown,

‘The volute halves are shown dséembled for
‘welding in Fig, 3.4.5. The welding procedure
used ‘was. essenflally fhe same ds that shown in

 
 

: "except'f fl\at .as ‘many’ as six addrtionai
passes were - requnred to compiete the weld at the -
‘volute . exit  where ‘the section ‘was abouf 1 ing
fhlck" The complefedvolufe is shown in Flg. 3 4.6

 

 

The results ~_:of fhe mlcrcmefer measurzments are
summarlzed in Tnble 34 2. It may be noted that
‘the’ shrlnkuge was somewhat less than that cleslred
‘The velute ‘entrance dlmensmn ns consldered to be
“the mosf“cnhcci dlmensmn in-the  pump, and,
after an analysus of the daia, ‘a shrmkoge ul-

  

lowance of 0.115 in. was selected for incorporation
into the joint design of volute No, 3. '
The volute halves for volute No, 3 were ma-

-chined, assembled, welded and annealed, with the

IWeidmg procedure agam being similar to that

~given on Fig, 3.4.1. Micrometer measurements
~were made before - ond after each important oper-
aflon, und the data are summcnzed in Table 3.4.3.

It may be noted that adjusting the shrinkage
uflowunce to-0.115 in, achieved a close approach.
~to the dimensional requirements ~of the -volute
entrance.- A further refinement 'is “planned for
~volute No. 4 in that a shnnkage allowance of
0.110'in, will be mcorporated into the joint design,
Any shrmkuge of ‘the ‘volute entrance to less than
70,650 in, will be’ prevented by the use of Inconel
spocers where needed, -

185

 
 

 

ANP PROJECT PROGRESS REPORT

UNCLASSIFIED
ORNL-LR—-DWG 16t68

180° 160°

VOLUTE ENTRANCE,

A
RADIAL POSITION C J ‘

60°

  
 
 
  

EXTERNAL BEVEL EDGE,
RADIAL POSITION A _ —

VOLUTE CENTERLINE,
RADIAL POSITION B

340° 0°

l— T>
i (D)

 

 

 

 

I e
0 =i

SECTION A-A

Fig. 3.4.4, Details of Micrometer Measurement on Volute No. 2.

186
 

i
i
i
i
i
i
!
!
i
i

 

PERIOD ENDING SEPTEMBER 10, 1956

UNCLASSIFIED
PHOTO 17583

 

 

 

A

UNCLASSIFIED
PHOTO 17879

      

 

 

Fig. 3.4.6. Yolute No, 2 After Inert-Arc Welding and Annealing,

187

 

i
§
i

 
 

 

 

- ANP PROJECT PROGRESS REPORT

TABLE 3.4.1. INERT-ARC WELDED PRIMARY NaK PUMP YOLUTE WELD

SHRINKAGE DATA FOR TEST NO. 1

 

Dimensions (in,)

 

 

Position Before After After - Net
Welding -Welding Change Annealing** Change
Alr 0.366 0.244 0.122 0.240 0.126
B1 0.382 0.254 0.128 0,248 0.134
C 0.388 0.264 0.124 0.259 0.129
D1 0.378 0.249 0.129 0.244 0.133
A 2% 0.377 0.259 0.118 0.254 0.122
B2 0.381 0.260 0.121 0.254 0.127
c2 0.388 0.268 0.120 0.264 0.124
D2 0.384 0.263 0.121 0.257 0.127
A3 0.468 0.349 0.119 0.346 0.122
B3 0.469 0.348 0.121 0.343 0.126
C3 0.469 0.349 0.120 0.345 0.124
D3 0.469 0.348 0.121 0.344 0.125

 

*Add 3.317-in, micrometer correction to readings shown at positions 1 and 2,

**Annealed at 1950°F for 2 hr and furncce cooled at a rate of 500°F /hr.

TABLE 3.4.2. RESULTS OF MICROMETER MEASUREMENTS OF WELDED YOLUTE NO, 2

 

Angular Position

Dimensions (in.)

 

 

188

Before After After Total b
(deg) Welding Welding Annealing? Shrinkage Deviation
At Radial Position AC _

0 0.754 0.646 0.648 0.106 -0.019
200 0.765 0.649 0.647 0.118 '-—o.o_oj
220 0.771 0.655 0.652 0.119 -0.006
240 0.758 0.641 0.639 0.119 ~0.006
260 0.763 0.642 0.641 0.122 -0.003
280 0.763 0.639 0.639 0.124 ~0.001
300 0.777 0.654 0.654 0.123 -0.002
320 0.778 0.653 0.654 0.124 -0.001
340 0.773 0.644 0.647 0.126 +0.001
 

PERIOD ENDING SEPTEMBER 10, 1956

TAB LE 304.2. (con'I flued)

 

Dimensions (in.)

 

Angular Position : l
(deg) Before After After Tota b

Welding Welding Annealing? Shrinkage Deviation

 

T

 

At Radial Position B¢

0 0.804  0.690 0.689 0.115 ~0.010
20 0.771 0.665 0.665 0.106 ~0.019
40 0.802 0.696 0.694 0.108 ~0.017
60 0.787 0,675 0.674 0.113 -0.,012
80 0.778 0.664 0.662 0.016 -0.009
100 0.774 0,659 0.656 0.118 ~0.007
120 0,771 0,655 0.653 c.118 ~0.007
140 0.779 0.664 0.661 s -0.007
160 0.773 0.656 0.654 0.119 ~0.006
180 0.766 0,650 0.649 0.117 ~0.008
200 0.779 0.665 0663 0.116 ~0,009
220 0.777 . 0.668 0.663 0.111 -0.014
240 0.764 04649 0.648 0.116 ~0.009
260 077 0,654 0,652 0.119 ~0.006
280 0,776 04659 0.657 0.119 ~0.006
300 0.780 0.664 0.663 0.117 ~0.008
320 0.784 0,668 0.667 0.117 ~0.008
340 0.794 0.674 0,673 0.121 ~0.004

Af Radicl Position Cd‘

-0 - oo 0.780 - - 04675 EERE 0.674 0.106 +0.014
200 . 0781 . 0478 0677 . 0.104 +0.017
40 0784 0681 - 0680 . -0J04  +0.020
60 - 0785 . 0679 0674 ~eamn - +0.014
80 . o785 0678 . 0679 . 006 +0.019

100 - o784 0675 0475 - 0109 +0.015
120 . 0783 D674 . 0674 009 - +0.014

Mo - 0781 0673 0673 - 0108 . - +0.013
160 0780 0673 0472 Cod0s 0 +0.012
180 o781 0673 0673 0108 - +0.013

200 0781 . 0675 0673 - 0108 . +0.013
220 0782 0476 04674 . 0008 +0.014

240 0782 0476 0.674 0.108 +0.014

189

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

TABLE 3.4.2. (continued)

 

Dimensions (in,)

 

Angular Position

 

(deg) Befo:'e ' Afte.r Afte'r . - T.otal Deviation?
Welding Welding Annealing Shrinkage
At Radial Position C?
260 0.782 0.675 0.674 - 0,108 7 - +0.014
280 0.782 0.674 0.674 0.108 +0.014 .
300 0.782 0.674 0.673 0.109 : -.1-70.013
320 0.781 0.673 0.673 0.108 +0.013
340 0.781 0.673 | 0.674 0.107 +0.014

2Annealed at 1950°F for 2 hr and furnace cooled at a rate of 500°F /hr. o

bAt radiol positions A and B the deviation is given as the deviation from the ideal change of 0.125 in. At radial
position C the deviation is given as the deviation from the desired dimension of 0.660 X 0.010 in.

€Add 3,317-in. micrometer correction to readings shown, |

Dimensions shown are absolute micrometer measurements.,

TABLE 3.4.3. RESULTS OF MICROMETER MEASUREMENTS OF WELDED YOLUTE NO. 3

 

Dimensions (in.)

 

Angular Position

 

(deg) Befo.re Afte'r Af‘l'e.f . T.ofol Dev iotionb
Welding Welding Annealing Shrinkage
At Radial Position AS
0 0.741 0.597 0.597 0.144 +0.029
180 0.723 0.604 0,602 0.121 +0,006
200 0.723 0.604 0.601 0.122 +0.007
220 0.719 0.598 0.596 0.123 +0.008
240 0.716 0.593 0.593 0.123 +0.008
260 0.719 0.594 0.593 0.126 +0,011
280 0.725 0.597 0.596 0.129 +0.014
300 0.737 0.604 0.606 0.131 +0.016
320 0,730 0.594 0.599 0.131 +0.016
340 0.733 0.594 0.596 0.137 +0,022.
At Raodial Pesition B
0 0.741 0.616 0.608 0.133 +0.018
20 0.755 0.651 0.645 0.110 ~0.005
40 0.780 0.678 0.673 0.107 ~0.008
60 0.770 0.666 0662 0.108 ~0.007
80 0.738 0.629 0.625 0.113 —0.002

190 :
 

 

 

L)

TABL E 3-4. 3. (Cfln'i nued)

PERIOD ENDING SEPTEMBER 10, 1956

 

Angular Position

Dimensions (in,)

 

 

 

(deg) Before  After After Total Deviation®
Welding Welding Annealing® Shrinkage
At Radial Position B
100 0.738 0.626 0.624 0.114 ~0.001
120 0.734 0.619 0.618 0.116 - +0.001
140 0.728 0.614 0.610 0.118 +0.003
160 0.728 0.611 0.610 0118 +0.003
180 0.728 - 0.610 0.609 0.119 +0.004
200 0.723 0.606 0.603 0.123 +0.008
220 0.725 0.608 0,605 0.125 +0.010
240 0.728 0.609 04606 0.128 +0.013
260 0.728 0.608 0.606 0.128 +0.013
280 0.733 0.609 0.606 0.127 +0.012
300 0.735 0,609 0.607 0.128 +0.013
320 0.735 0.607 0.603 0.132 +0,017
340 0.735 0.606 0.603 0.132 +0.017
o At Radial .Position c? _
0 0,773 0.663 0.656 0.117 ~0.004
20 0.774 0.670 0.662 0.112 +0.002
40 0.775 0.673 0.669 0.112 +0.002
60 0775 0.672 0.671 0.104 +0.009
80 - 0775 0.671 0.671 0.104 +0.011
100 To774 | 0.68 0,668 0.106 +0.008
120 S o774 osss 0.666 0108 +0.006
140" 0774 - 0.665 0,665 0109 - +0.005
16 0774 - 0.665 - 0.665 0.109 +0.005
10 0774 0665 . 0.663 0.11 +0.003
200 0774 0665 - 0.663 011 40,003
o200 oam o 0.665 0.662 L0.M2 . 40,002
'?,;jf;;r'?&‘l R °774 - ir,:'.:_'jé:'-éefviiji;” | 0.661 03 +0.001
Lo as0s o togm o oses 0.659 01s - -0.001
280 e 0774 0,663 o 058 0016 ~0.002

191

 
 

 

~ ANP PROJECT PROGRESS REPORT

TABLE 3.4.3. (continued)

 

Dimensions (in.)

 

Angular Position

 

(deg) Before After After . Total De-viétio;b
Welding Welding Annealing Shrinkage
At Radial Position C%
300 \ 0.773 0.660 0.654 0.119 -0.006
320 0.773 0.659 0.653 0.120 ~0.007
340 0.773 0.660 0,653 0.120 ~0.007

 

“Annealed at 1950°F for 2 hr and furnace cooled at a rate of 500°F/hr,
bay radial positions A and B the deviation is given as the deviation from the ideal change of 0,115 in. At radial
position C the deviation is given as the deviation from the desired dimension of 0,660 £ 0.010 in.

€Add 3.317-in, micrometer correction to readings shown,

Dimensions shown are absolute micrometer measurements,

FABRICATION OF PRIMARY NaK
. PUMP IMPELLERS

P. Patriarca G. M. Slaughter

Four primary NaK pump impellers have been
fabricated by furnace brozing with Coast Metals
brazing alloy No. 52. Since slow heating and
cooling rates are required to minimize distortion
and braze cracking, each impeller brazing cycle
entails o furnace time of 10 to 12 hr. As can be
seen in Fig. 3.4.7, which shows a typical impeller,
the vanes were inert-arc-welded to the housing,
where accessibility permitted, to achieve ad-
ditional reinforcement,

SHRINKAGE OF INCONEL CORE
SHELL WELDS

P. Patriarca A. E. Goldman

Preliminary transverse weld shrinkage tests of
Inconel core shell welds were discussed in the
previous report.2 Further experiments have now
been conducted to study the shrinkage for other
thicknesses of plate, The procedure for the
shrinkage tests of welds of l/s-in.-_thick plates is
described below, and the results are compared
with the results obtained for other thicknesses of
Inconel plate. ,

The initial phase of the test consisted of the
inert-arc welding of two Y%-in. Inconel sheets,
each 6 x 20 in. A 50-deg bevel! with a I'Sz'i“'

 

2p, Patriarce and A. E. Goldman, ANP Quar. Prog.
Rep. June 10, 1956, ORNL<2106, p 184.

192

land was machined on one long edge of each

sheet, The sheets were assembled as shown in
Fig. 3.4.8 and held against a flat horizontal plate
by means of C-clamps. The edges of the assembly
were sealed with tape to prevent air leakage,
since only the torch gas was used to supply

 

Fig. 3.4.7.
Impeller in Which the Vanes Were Both Inert-Arc-
Welded and Furnace-Brazed,

Completed Primary NoK Pump
 

 

 

40

49

UNCLASSIFIED
ORNL-LR-0WG 6469

  
 

Y% x 6 x 20-in,
INCONEL
SHEETS

    
  

HELIUM
INLET TUBE

Fig. 3.4.8. Joint Design for Shrinkage Test of

| Inconel Plate Welds.

backing gas and weld coverage for this test,
A l,é--in.-long tack weld was placed near each end
of the root, and two other z-in.-long tack welds
were equally spaced along the root. The root
pass was made, without filler metal, and, after
wire brushing and cleaning the weld with acetone,
a final pass was made for which 1,16--in. inco No. 62
filler wire was used.

A second test was made that was similar to the
first, except that backing gas was used to elimi-
nate oxidation of the underside of the weld, Neo
shrinkage measurements were taken on these

plates, since the test plates were made solely

to aid the welding operator in adapting his
technique.
Based upon the plate tests, two tests were made

~with Y%-in. Inconel sheets, each 6 x 33 ihr.,-th_af_

were bent and welded into - lO‘é-in.—dia ‘hoops.

A 50-deg bevel with a 'z'z-in. land was machined.
on one long edge of each sheet, The first hoop
- was placed ‘on the flat, horizontal bed of the
~welding" positioner, ‘and the second hoop was .
placed above it.  The two hoops were aligned by
- placing a large perforated steel band around the
“root girth and tightening the band with C-clamps,
‘as shown- in Fig. 3.4.9. Transverse micrometer
measurements were then taken at 4-in. intervals -

around - the circumference, = Eleven _Z-ih;-l_bng

' tdcrkl_".jw'gids',_f,we'r_é', equally spaced around the
circumference, about 3 in, apart., - Helium backing

gas was applied through a copper cone, No
dressing of the land or feathering of the weld
beads was permitted prior to deposition of the
root pass, ‘

PERIOD ENDING SEPTEMBER 10, 1956

After the tack welds were made, the steel band
was removed, micrometer measurements were
again taken, and the assembly was enclosed as
shown in Fig. 3.4.10. The interior was purged

UNCLASSIFIED
ORNL.-LR-DWG {6470

   
  

"¢" CLAMPS

TEST PIECE

  

PERFORATED
STEEL BAND

O

 
 
  

TEST PIECE

Fig, 3.4.9. Method of Assembling Hoops for
Welding.

UNCLASSIFIED
ORNt=LR-DWG 1671

SHEET METAL COVER ELIUM INLET TUBE

TAPED IN PLACE

     

TEST

TACK ASSEMBLY

WELDS

NCONEL DISK

<

TO
WELDING
POSITIONER

Fig, 3.4.10. Hoop Welding Test Assembly After
Tack We’dil'lg.

193

 
 

 

 

e R P e et et e

ANP PROJECT PROGRESS REPORT

for 10 min, after which the root pass was made.
After wire brushing and cleaning the weld with
acetone, the final pass was made, with '4 -in,
Inco No. 62 filler wire again being used. Féinal
micrometer measurements were taken after the
disks were removed from the ends of the assembly.

Two additional tests were then carried out in an
identical manner with Y-in, Inconel sheets, each
6 x 70 in., bent and werded into 22-in.~dia hoops.
These hoops were aligned with C-—clamps, rather
than with a steel band, The pertinent welding
data for the four hoops are listed in Table 3.4.4,
and the micrometer measurements are given in
Table 3.4.5.

The deposited welds were clean and had slight
concavity on the inside surfaces. It was found

that short tack welds enabled the welding operator
to fuse the tack welds into the root pass more
readily. For Y-in. Inconel sheets, inert-arc-
welded under the conditions utilized in this test,
the transverse shrinkage to be expected for loz-in.
hoops is 0.050 £ 0.010 in. and, for 22«in, hoops,
0.041 £ 0.010 in. The circumferential shrinkage
was less than measurable for both hoop sizes.

A summary of the weld shrinkage results for
Y . 1/8-, V., and ¥%-in. Inconel sheet is presented
in Table 3.4.6. It should be noted that the data
are pertinent only under the conditions utilized
for these tests, Any change in the variables may
cause significant variations in the actual weld
shrinkage encountered.

TABLE 3.4.4. WELDING DATA FOR HOOP WELD SHRINKAGE TESTS

 

 

Current Time Torch Gas Backing Gas Filler Rod
(amp) (min) (H3/hr) (3 /hr) (in.)
Hoop Nc. 1
(10% in. dia)
Tack weld 100 8 25 15
Fusion weld 100 13 25 10
Final weld 80 25 25 | 10 110
Hoop No, 2
(10% in. dia)
Tack weld 75 7 25 10
Fusion weld 80 n 25 10
Final weld 80 26 25 10 115.5
Hoop No. 3
(22 in, dia)
Tack weld 100 13 26 17
Fusion weld 90-95 18 26 17
Final weld 75--80 47 26 17 212.5
- Hoop Ne. 4
(22 in. dia)
Tack weld 80 13 26 20
Fusion weld 100 15 26 20
Final weld 80 47 26 20 209.5

 

194
 

 

PERIOD ENDING SEPTEMBER 10, 1956

TABLE 3.4.5. MICROMETER MEASUREMENTS OF HOOP WELDS

 

Tronsver#e Shrinkage (in.)

 

 

. : Deviation,
Average Maximum Minimuym )
. Station to Station

Hoop No, 1

(10.5 in, dia) 0.051 0.056 0.044 0.007
Hoop No, 2 ‘
- (10,5 in. dia) 0.050 0.054 0.045 ' 0.009
Hoop No_;. 3 ‘ ‘ -

(22 in, dia) 0.041 ' 0.048 0.036 0.010
Hoop No. 4 -

 

" Neote: Longitudinal shrinkage was less than measurable because of the small diameters of the hoops.

TABLE 3.4.6. SUMMARY OF HOOP WELD SHRINKAGE DATA

 

Inconel

. Transverse Shrinkage {in,)

Recommended

 

 

Sheet H.oog""-r Greatest Design Allowance Longitudinal
. Diameter ) . .. Shrinkage
Thickness o Average Maximum Minimum Deviation, for Transverse inu/in.)
(in.) (l»n.) - Station to Station Shrinkage (in.) (in./in.
14 6 45 0.024 0.026 0.018 0.005 : 10.024 1 0.010 0.00045
% 105 0.051  0.056  0.044 0.009 0.051 £0.010  Not measurable
% 22 0.041  0.045.  0.036 0.010 0,041 £0.010  Not measurable
% 44 0.120  0.138 0.111 0.011 0.120 £ 0.015 0.002-0.003
¥ 52 0de8 0184 0.155 0.014 0.168%0.020  0.002-0.003

 

*Vertical dimension of all hoops was 12 in,
EXAMINATIONS OF NaK-TO-AIR RADIATORS

AFTER SERYVICE

o P. Patriarca
A. E. Goldman ~ G. M. Slaughter

A 500-kw high-conductivity-fin NdK-fo-efr-roJi-

ator, designated York HCF Radiator No, 4, was’

removed from a test stand on March 3, 1956, after
1356 hr of service. This radiator was tested in
the temperature range of 1000 to 1600°F, with a
temperature differential imposed on the system
during 748 hr of the test.

This radiator differed from radiators tested
previously in that the side plates were removed,
the horizontal spacer plates were cut, and the
base plate was sliced through the middle parallel

" to the air flow. These modifications were among
- those suggested in previous reports,3+4

 

3. 4. Gray and P. Patriarea, Metall&grapbic Ex~
amination of ORNL Radiator No. 1 and York Radiator
No. 1 Failures, ORNL CF-55-10-129 (Oct. 31, 1955).

Q. J. Gray and P, Patriarca, Metallographic Ex-
amination of PWA HCF Radiator No, 2, ORNL CF-56-3-
47 (March 12, 1956).

195

 
 

 

 

 

 

ANP PROJECT PROGRESS REPORT

Twenty-four specimens were cut from the radi-
ator as shown in Fig, 3.4.11. Each specimen
contained a portion of three tubes, with each tube
joined to 15 or more fins, The specimens were
cross sectioned to expose two opposed joint
areas, mounted, and examined at 80X.

A total of 2757 joint areas were examined,
The percentages of fin-to-tube adherence and the
degree of oxidation of the fin collars were noted.

UNCLASSIFIED
Y-18442

The results of this examination in comparison
with the results of examinations of five radiators
operated previously are presented in Table 3.4.7.
Small cracks were found in several of the tube-to-
support plate joints, and evidence of cracking was
observed in several tube<to-sump plate joints.
The relationship between the incidence of fracture
and the presence of support members or plates
was previously demonstrated,3+4

UNCLASSIFIED
Y-18443

 

Fig. 3.4.11. Air Inlet and Exit Sides of York HCF Radiator No. 4 Showing Location of Speclmens
Cut for Metallogrephic Examination After Service for 1356 hr.

196

+*
 

 

PERIOD ENDING SEPTEMBER 10, 1956

TABLE 3.4.7, SUMMARY OF RESULTS OF METALLOGRAPHIC EXAMINATION OF YORK
HCF RADIATOR NO. 4 AND FIVE PREVIOUSLY OPERATED RADIATORS

 

Radiator Designation

 

 

York York Yeork PWA ORNL ORNL
No. 4 No. 3 No. .I No. 2 No. 3 No. ]
Number of hours of service in the 1356 361 152 1199 716 608
temperature range 1000 to 1600°F
Number of joint areas examined 2757 2684 3847 3210 2282 4150
Percentage of joint areas having 75 to . 60.7 90.4 67.4 100 87.7 91.8
100% adherence
Percentage of joint areas having 50 to 18.8 5.9 13.0 3.5 4,2
74% cdherence
Percentage of joint areas having 25 to 5.5 1.1 7.3 1.4 1.4
49% adherence
Percentage of joint areas having ( to 24% 15.0 2.6 12.3 7.4 2.6
adherence '
Percentage of joint areas having non- 61.9 84.5 75.4 100 12.1 59.3
oxidiZed copper flns :
Percenfage of ioinf areas having slightly 19.5 7.5 22.0 ' 2.5 26.1
oxidized copper fins
Percentage of idinf .are_as having oxidized 18.6 8.0 2.6 - 85.4 20.6

copper fins

 

FABRICATION OF CERMET YALVE
COMPONENTS

G. M. Slaughter

A description of the high-temperature cermet
bonding “procedure wused - in the fabrication of

P. Patriarca

severa| valve components was presented previ-

ously.5 At the. relahvely high : temperatures
required for bondmg (approximately 1350°C), the

. cermets lose strength ‘to such an extent that
o ‘occaslonally sngmfucnnt dlstorhon or warpuge of
the cermet body is observed , : _
A modlflcuhon in the fabrication procedure was

therefore . developed which has permitted _ the

successful - fabrication of disks and" seats forr
four different test valve assemblies, The minimum

tempercture for consastent bondmg was determlned

 

5p. Patriarca, A. E. Goldman, and G. M. Slaughter,

ANP3 Quar, Prog. Rep. March 10, 1956, ORNL-2061,
P 143.

for each cermet composition and was found to vary
slightly with the cermet type, as shown in
Table 3.4.8. Occasionally these optimum temper-
atures were found to vary slightly with different
components of the same general composition.
This  observation . can probably be related to
variations in the compacting and sintering pro-
cedure by the original manufacturer of the cermet
bodies. Accordingly, each component was treated
as an individual problem, - ‘

- Observation ports were machined in the nickel

- transition Iayer to permit o visual determination of

the initiation of bonding. - This visual observation

: was - found to be an essential addition to the
precise control and measurement of temperatures

for the consistent prevention of distortion of the

cermets, These ports are clearly visible as black

spots when the assembly in the furnace hot zone
is viewed with dark glasses, When the liquid

reaction product forms, the ports are filled, and

197

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

they disappear from view, with simultaneous
formation of a fillet, which can also be detected
by careful observation through dark glasses,
Cooling rates of approximately 1000°C/hr were
used to minimize stresses during cooling.
Exploded views of the two steps employed in the
fabrication of the valve disks are shown in

Fig. 3.4.12. The cermet-to-nickel subassembly
shown in step 2 was copper-brazed to the Inconel
shank by conventional dry-hydrogen techniques,
Cooling rates of approximately 400°C/hr were
used to prevent cracking. The stages used in the
fabrication of the valve seat components are

shown in Fig. 3.4.13.

TABLE 3.4.8. MINIMUM TEMPERATURES FOR CONSISTENT BONDING
OF CERMET VALVE COMPONENTS

 

 

 

Cermet Cermet Composition Bonding Temperature
Trade Designation (wt %) (°C)
K-151A 70 TiC=10 NbToTiC3—20 Ni 1350
K-152B 64 TiC-6 NbTuTiC3-30 Ni 1350
- K-162B 64 TiC-6 NbTaTiC3—25 Ni-5 Mo 1340
STEP 1 UNCLASSIFIED

Y-18799

 

it COMPLETED
VALVE DIsk LAVER WITH MACHIN CERMET-TO-NICKEL
LAYER WITH MACHINED y
OBSERVATION PORTS SUB-ASSEMBL
ONE INCH
atady
UNCLASSIFIED
STEP 2 LASSIF

  

 

 

CERMET-TO-NICKEL INCONEL COMPLETED VALVE
SUB-ASSEMBLY COPPER BRAZING SHANK DISK ASSEMBLY
TR
FENARE

Fig. 3.4.12, Steps in the Fabrication of e Cermet Valve Disk Assembly.

198
 

| 7 PERIOD ENDING SEPTEMBER 10, 1956

  

        

    

 

|

L . |
UNCLASSIFIED

Y-18852

STEP 1

|  CERMET " NICKEL TRANSITION COMPLETED

| ' VALVE SEAT \ 'LAYER WITH MACHINED CERMET-TO-NICKEL

, . OBSERVATION PORTS SUB-ASSEMBLY

|  EEE

| ' | S UNCLASSIFIED

i - Y-18353

; €

|

|

I

|

| . CERMET-TO~- - - \ -

: " NICKEL COPPER INCONEL COMPLETED

. | _ 'BRAZING SHIM " HOUSING VALVE SEAT

, , SUB-ASSEMBLY L ' B ASSEMBLY

e

Fig. 3.4.13. Steps in the Fabrication of a Cermet Yalve Seat Assembly.

199

|
|
i
I
i
i
1

 
 

 

 

 

i
|
!
|
|
|

 

ANP PROJECT PROGRESS REPORT

CONTINUOUS PRODUCTION OF SINTERED
BRAZING ALLOY RINGS

P. Patriarca R. E. Clausing

The use of sintered Coast Metals brazing alloy
No. 52 rings for the brazing of ART radiators is
dependent upon the achievement of production
rates adequate to reduce the cost per ring to a
value which con compete with the present means
.of preplacing the brazing clloy. An experimental
pilot plant has been built which has cttained a
production rate of more than 8000 rings per hour,
In ‘the first production run, 33,000 rings were
made in 5 hr, It is felt that production of this
magnitude will definitely establish a satisfactory
cost for ring production.

Graphite molds are used, as shown inFig. 3.4.14,
for the production of the brazing alloy rings. The

 

LOADED
BRAZING EMPTY MOLD
ALLOY HOLD |
POWDER |

depth of cut in the mold is adjusted to yield rings
that weigh 4.5 g per 100 rings. The size of the
ring is based on the quantity of alloy required to
produce an acceptable tubesto-fin joint. The molds
are loaded flush with brazing alloy powder and
stacked four high, with each mold serving as a lid
for the one below it. | '
The furnace used in the pilot plant is shown
schematically in Fig. 3.4.15. This furnace is
essentially a two section unit, The first section
serves as a preheater, and the second section
adjusts the temperature to the narrow range in
which satisfactory rings are produced and main-
tains this temperature long enough to allow the
entire mass of the molds to reach this temperature.,
A continuous stainless steel ribbon conveyor belt
runs through the furnace and is driven by @ vari-
able-speed drive. When this equipment is operated

UNCLASSIFIED
Y-18084

 

" BRAZING ALLOY
RINGS AS UNLOADED

MOLD
AFTER

LOADED
AND SINTERING

COVERED

MOLD

| Fig. 3.4.14. Steps in the .Production of Sintered Rings of Coast Metals Brazing Alloy Neo, 52.

200
 

 

PERIOD ENDING SEPTEMBER 10, 1956

at a belt speed of 5 in./min, 9000 rings can be ~ FABRICATION OF HEAT EXCHANGERS

produced in 1 hr; however, the control of the : AND RADIATORS
furnace temperature becomes quite critical at this P. Patriarca
speed, The optimum operating speed for this A. E. Goldman ‘G. M. Slaughter

equipment is 4 in./min, at which speed 7200 |
rings/hr or approximately 50,000 per 8-hr day One of the problems associated with the as-

~ are produced. sembly of multitube heat exchangers involves

A thermal cycle which produces sahsfoctory preplacement of an adequate supply of braze metal
rings is shown in Fig. 3.4.16. This graph was  at each joint. A heat exchanger job sample on

~ .reproduced from a recording of the output of @  which the Coast Metal$ brazing alloy No. 52 was
. thermocouple attached to a set of molds passing  preplaced by the metal-spray process is shown
- through the furnace at 4 in./min. Proper control  in Fig., 3.4,17. Several tubes, one of which is
~ of the temperature permits the easy removal of the  indicated by an arrow, failed to form a braze

sintered rings from the molds through the use of  fillet, the alloy having been ‘‘stolen’ by the
a vibrating tool. The rings may be grit blasted or  adjacent tubes.

tumbled in brazing alloy powder, if necessary, to

remove any traces of carbon or small surface Several experiments were conducted which
imperfections. |f the temperature is carefully indicated that intimate contact between the tube,
controlled ({within a 5°C range), rings with no  tube sheet, and brazing alloy was desirable during
rejects and very good surface finish will be  heating to the brazing temperature. Once the
produced., Rings of satisfactory -quality are initial capillary flow was established, each tube

produced if the temperature is controlled within @ drew sufficient alioy from the metal-sprayed tube

range of about 10°C, : sheet to form a satisfactory fillet.

UNCLASSIFIED
CRNL-LR-DWG 16172

o o 220V, 1 PHASE, 10 A—— 220V, 1 PHASE, 10 A—=] HEAT BARRIER
HELIUM A / HELIUM
\ n i (

THERMOCOUPLE TO e THERMOCQUPLE TO
CONTROLLER o -0- - CONTROLLER
; fl SIL-0-CEL '”SU'-“T"’N\(/ L KANTHAL WINDINGS (
" Jo00000000Q0000O0Q .
7;.:,o-opaouoooooooooooooo-

 

 

 

 

 

 

 

 

3 ~in. -DIA CERAMIC MUFF LE

 

 

 

 

 

 

 

 

 

   

: GRAPHITE MOLDS - 3-1n. STAINLESS STEEL CHANNEL 20 fl LONG _

- VARIABLE SPEED MOTOR -

 

 

\'/4 in. ALUMINUM HOUSING, 18~in-DIA, 36 in. LONG

smm.ess STEEL commuous RiBBON . "fj' .

Flg. 3 4 15 Schemoflc qugrqm of Furnuce for Conflnuous Produchon of Smtered ngs of Coasi
Metals Bruzing Alloy No. 52 e : : . .

201

 
 

 

 

 

 

- ANP PROJECT PROGRESS REPORT

UNCLASSIFIED
ORNL-LR~DWG 16173

 

1200

- PREHEAT FURNACE

e

|
I

-
|
SINTERING FURNACE ——l

 

1000

U

 

800

P

 

 

600

TEMPERATURE (°C)

 

400

 

200

 

- ———.—._..._.._._._...._____..r--..—.._-—-—l-——-————-—-‘!-—-——_- .

 

 

 

 

 

 

 

 

————
s

 

 

S {0

e

O T T T S S S e s e e ——

20
TIME {min)

25 30 35 4C 45

Fig. 3.4.16. Sintering Cycle for the Continuous Production of Brazing Alloy Rings at a Belt Speed

of 4 in./min.

 

UNCLASSIFIED
9

Fig. 3.4.17. Héat Exchanger Job Sample,

202

“sheet during the ‘‘priming’’ operation,

The most successful method of initiating
capillary flow was found to be the metal-spray
“primer’’ technique. As each row of tubes is
assembled into the previously metal-sprayed tube
sheet, the tubes are attached to the tube sheet
with @ small amount of metal-sprayed brazing ailoy.
A tungsten sheet is used as a mask to prevent
accumulation of alloy on the tubes, In a modifi-
cation of this method, sintered brazing alioy
rings are attached to the previously sprayed tube
Single
rings are shown attached to the tube and sprayed
tube sheet in Fig. 3.4.18, and multiple rings are
shown attached in Fig. 3.4.19.

STUDIES OF THE AGING CHARACTERISTICS
OF HASTELLOY B
R. E. Clausing |

Preliminary physical property tests on Hastelloy B
indicate that residual stresses induced by cold

‘9
 

!
1

 

 

 

 

 

PERIOD ENDING SEPTEMBER 10, 1956

 

 

 

 

 

 

 

 

 

Fig. 3.4.18. Sintered Rings and Metal-Sprayed Coast Metals Brazing Alloy No. 52 on Heat Exchanger
Tube Sheet.

 

Fig. 3.4.19. Multiple Sintered Rings and Metal-Sprayed Coast Metals Brozing Alloy No. 52 on Heat
Exchanger Tube Sheet. |

203

 
 

 

 

ANP PROJECT PROGRESS REPORT

rolling or heat treatment may have pronounced
effects on the rate of elevated-temperature precipi-
tation in the alloy. Hardness values for cold-
rolled sheet, originally l/8 in. thick, have been
determined after aging treatments for various
times at 1200°F. The results are shown in

Fig. 3.4.20. Tensile tests show the ductility to be
much lower in specimens aged from the stressed
condition than in those annealed prior to oging,
A microstructure correlation is presently being
made as a means of studying this effect,

UNCLASSIFIED
ORNL—LR-DWG (6174

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

550
ROLLED TO GO% REDUCTION IN AREA
" i
.500 // 507
/'—’-—‘
/.--‘/...—-—-"'-—-" 40%, /___/—-—-.
1 //j? : //
450 - = — 130 %0 1— '
s /// / S
-8 400 ] A :
a A
x w1 / /
: /]
LT | A
350 /,/ Ve
300 -",///
_—-—“/
250
i 2 5 10 20 50 100 - 200 500

AGING TIME

AT 1200°F (hr)

Fig. 3.4.20. Aging Characteristics of Hastelloy B Sheet Containing Residual Stresses from Cold

Rolling.

204

4
 

 

 

e

. Q02
. RECORDER =

PERIOD ENDING SEPTEMBER 10, 1956

3.5. MECHANICAL PROPERTIES STUDIES
D. A. Douglas

RELAXATION TESTING OF INCONEL
C. R. Kennedy"

The evaluation studies of the creep and creep-
rupture characteristics of metals under constant
load and in various environments have been con-
tinved, and, recently, equipment was designed and
constructed for the study of the decay, or relaxe-
tion, of stresses in metals under constant strain.
This phenomenen is time and temperature dependent

-and. assumes particular importance in cases of

thermally induced cyclic stresses, ~For example,
if a structural member is held rigidly at both ends,
and the temperature is raised and then lowered
rapidly, tensile and compressive loads will result,
It is essential in the stress analysis of this situa-
tion to know how much the stress hos decreased,
that is, relaxed, in each condition before the
reverse load is applied, '

 

10n assignment from Pratt & Whitney Aircraft,

   
   
 
 
       
  
    
 

-
1
[ -
I

SPEC
I

b
FURNACE'—'_/: B}
. o i .

L L

| $R4 LDAD CELL

  
  

‘There have been many different machines and
methods used by previous experimenters to de-
termine the relaxation characteristics of metals;
however, most of the work has been in temperature
ranges well below the annealing temperature of the
material tested. Extrapolation of these results
to higher temperature does not give reliable values.
Therefore it was necessary that new experimental
techniques be devised to achieve the accuracy
desired for temperatures in the range of 1200 to
1800°F. The equipment designed for the tests is
shown in Fig. 3.5.1. The test is run by setting the
extensometer for the desired strain and then applying
and automatically controlling the stress to produce
and maintain this strain. The positions of the
electric contacts are established by micrometer

"screws on a 20 to 1 magnification extensometer

similar to the Westinghouse extensometer.2 When

 

. HYDRAULIC JACK

ELECTRONIC
RELAYS

2, Boyd, The Relaxation of Copper at Normal and
Elevated Temperatures, ASTM Proceedings, Vol. 37,

UNCLASSIFIED
ORNL—LR—-DWG 16175

® = SOLENOID VALVE

SOLENOID VALVES
ARE NORMALLY OPEN

LOAD

LOAD
E

 
 
 

LE VALVES

LOAD

  

“ACCUMULATOR

Fig. 3.5.1. Equipment for Relaxation Testing of Metals in the Temperature Range 1200 to 1800°F.

205

 
 

 

 

ANP PROJECT PROGRESS REPORT

the extensometer makes contact the 4-v grid bias

is cut out, and the grid-glow relay opens the

proper solenoid valves to control the hydroulic

_ pressure and thus the stress in the specimen. The
furnace is shunt-wound and is controlled by a
‘Leeds & Northrup model H Speedomax recorder and

DAT 60 controller which will control to $2°F with
a temperature gradient over the 6-in. gage length of
less than 5°F,

Test results of relaxation testing of Inconel at

1300 and 1500°F are shown. in Figs. 3.5.2 and

3.5.3. To produce a plot on which all the relaxa-
tion characteristics can be shown, the first 0.1-hr
period is plotted linearly and the remaining time is
shown on a semilog plot. The initial stresses ob-

tained at both temperatures agree with the moduli
of elasticity of Inconel at the respective tempera-
tures. The plots shown in Figs, 3.5.2 and 3.5.3 are
averages of from 5 to 10 tests, and the material was
tested with varying degrees of plastic strain at the
test temperatures. This information can be used
to determine the residual stresses in structural
members which have undergone o thermal cycle.
The data can only be considered valid for thermal
cycles which produce strains within the range
investigated. ' Further testing is now in progress to
determine ihe relaxation characteristics at higher
strains and various temperatures to ‘establish a
complete understanding of the relaxation charac-
teristics of Inconel,

UNCLASSIFIED
ORNL-LR-DWG 16176

 

 

30,000
e NN
_ ~0.24% STRAIN, INITIAL STRESS 28,500 psi
: ey I
27,000
R I
"~ 0.2% STRAIN, INITIAL STRESS 27,600 psi

 

o N
' r\ 045% STRAIN, INITIAL STRESS 27,000 psi

L

24,000

 

b |
\ |1 04% STRAIN, INITIAL STRESS 24,000 psi

 

18,000 M
y \
\§

 

15,000

STRESS (psi)

 

12,000 N\

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

\\
© 9000 = ~
"-n-..______’:::
\‘-0.05°/o STRAIN, INITIAL STRESS 42,000 psi \.___.
6000 [
-..\
e
_-“-_"'-'--._

3000

o

0 0.02 Q.04 0.06 0.08 0.4 0.2 0.5 { 2 5 10 20 50 . 100

TEST PERIOD (hr)

Fig. 3.5.2. Relaxation Characteristics of As-Received Inconel ot 1300°F When Stressed to Produce

a Constant Strain.

206
 

 

 

4

' STRESS (psi)

PERIOD ENDING SEPTEMBER 10, 1956

UNCLASSIFIED
ORNL-LR-DWG 18177

 

30,000

 

27,000

 

24,000

 

24,000

 

18,000

 

15,000 .
' |~ 0.4% STRAIN, INITIAL STRESS 17,000 ps!

 

12,000

- 0.05% STRAIN, lNITlAL.S'I_'RESS 10,000 psi

 

9000
N

7N

 

 

6000

 

 

 

3000

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

0 Q.02 0.04 0.06 0.08 oX | 0.2

0.5 { 2 5 LY 2 5 100

TEST PERIOD (hr)

Fig. 3.5.3. Relaxation Characteristics of As-Received Inconel at 1500°F When Stressed to Produce

a Constant Strain.

EFFECTS OF FUSED.SALT conn"os:ondn

THE STRESS-RUPTURE PROPERTIES .

OF HASTELLOY x
C. R. Kennedy |

CAs senes of tests was run o determme the effecff ,
of the fuel mlxture (No. 30) NaF-ZrF, +UF, (50-46-4;;_
mole- %) on’ the creep strength of Hasfelloy Xo
This alloy, which has the- ‘nominal. composition
22% Cr, 19% Fe, 9% Mo, end- the balance Ni, has

‘demonstrated reasonably good Creep. sfrengfh ond o
~ duchllfy at high temperatures -in air;. however, as
-may be seen in Fig. 3.5.4, the fuel environment -
has deleterlous effecis on ‘its creep “strength af

~temperatures above 1500°F, Since the fuel leaches
chromwm, ‘the observed reduction in creep strength

was  expected over. the entire temperature range;

’however, at 1500°F the amount of corrosion is
?_rsmall as seen in Fig. 3.5.5, and does not appear
- to affect the creep strength. In creep tests at

1650 and 1800°F the expected intergranular corro-
sion occurred as may be seen in Figs. 3.5.6 and

_'j3.5.7, and there was a serious decrease in creep
- strength, From these fests it appears that Hastelloy
X ‘should not be used where it will be in contact
'wnh the fuel mixture (No, 30) NaF-ZrF +UF,
© (50-46-4 mole %) at service temperctures above

1500°F.

207

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

 SEIRETs
OR_NL' LR-DWG 16478

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

20,000
\\
S —t—
] 1
--'-\
10,000 ] : : ~ . :
' " T4~ 1500°F
[ “a ~J_ _
Pt N : ~T®
\._ .
T &, \%\ |
E \ ~i' ‘\\.
0 s b~— N ™~ ~l.i_| 1650°F
& 5000 ~=x <
@ Lo —— . . .
o N '
\'\ S"'fl-..__ 1
Y
\\ ’ \
e :
- C)
"-._____l-B-OOF
'fi._l.-.-
2009
| IN AR
® IN FUEL MiXTURE (NO. 30) NofF-ZrF,-UF,
(50-46-4 mole %)
1000 L— . : |
10 20 50 100 200 500 1000 2000 5000 10,000

TIME TO RUPTURE (hr)

Fig. 3.5.4. Comparisen of the Stress-Rupture Properties of Hastelloy X in Air and the Fuel Mixture
(No. 30) NaF-Z¢F 4-UF 4 (50-46-4 mole %) at 1500, 1650, and 1800°F.

 CREEP-RUPTURE TESTING OF AN
80% Mg~20% Li ALLOY

~J. R. Weir C. W. Dollins
Creep-rupture testing of an 80% Mg-20% Li alioy

was initiated. The results of creep-rupture testing
of this alloy at 200°F in air are summarized in

‘Fig. 3.5.8, in which the stress is plotted vs time
to 0.2, 0.5, 1, 2, 5, 10, and 20% elongation and

rupture.  The specimens tested at stresses below
400 psi had not deformed to any measurable extent

in 500 hr.

208

" The test specimens were given a chemical sur-

face treatment® before the tests to improve their
oxidation resistance., The treatment consisted of
a nitric acid dip, followed by a rinse in water and
immersion in phosphoric acid until the reaction
ceased, followed by an acetone rinse. No oxidation

“was observed after 500 hr of testing in air. Tests

will be run on a few untreated specimens. for com-
parison, :

 

3R. E. McDonald and C. F. Lemen, Jr., ANP Quar.
Prog. Rep. June 10, 1956. ORNL-2106, p 173-
 

 

PERIOD ENDING SEPTEMBER 10, 1956

 

Fig. 3.5.5. Surfaces of the Stressed and Unstressed Portions of a Hastelloy X Specimen After Creep-
Rupture Testing in the Fuel Mixture (No. 30) NaF.Z(F 4-UF 4 (80-46-4 mole %) at 1500°F, 100X, (Secret
with caption) '

 

Fig. 3.5.6. Surfaces of the Stressed and Unstressed Portions of a Hastelloy X Specimen After
Creep-Rupture Testing in the Fuel Mixture (No. 30) NaF-ZrF ,-UF - (50-46-4 ‘mole %) at 1650°F. 100X.
209

 
 

 

 

 

"ANP PROJECT PROGRESS REPORT

Fig. 3.5.7. Surfaces of the Stressed and Unstressed Portions of a Hastelloy X Specimen After Creep-

100X, (Serrer

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Rupture Testing in the Fuel Mixture (No. 30) NaF.Z¢F 4-UF . (50-46-4 mole %) ot 1800°F.
UNCLASSIFIED
CRNL—LR-DWG 16179
2000
0.2 ¥ ELONGATION
1000 1— 0.5% 1 1% 2 %, 5% 10% 20% | | RUPTURE
- o P ~T— ] ™ N
. ™~ -"\?\\ [ — MR N
‘--._____-. -\\ -\___K -.._‘__-
T —
-’-; — [t
& s00 - ""'--..\
-'-.,-‘."
§ \\\___
5
200
100
2 5 10 20 50 100 200 500 1000

Fig. 3.5.8. Summary of Creep-Rupture Data for the Surface-Treated 80% Mg-20% Li Alldy Tested

at 200°F in Air.

- 210

TEST PERIOD (hr)

.

%

O

“
 

 

 

 

 

&

o

PERIOD ENDING SEPTEMBER 10, 1956

3.6. CERAMIC RESEARCH
L. M. Doney

RARE-EARTH MATERIALS

) Co E. Curtis Jo Ao Griffin
L. M. Doney

A ;study of the properties of europium oxide was
initiated, Physical properties, sintering charac-

teristics, and fabrication techniques will be in-

vestigated.

Numerous shapes of mckel—rare-earth oxide
cermets, Al,O0,, and CaF, are being produced,
These materials are to be used in the fabrication
of a high-temperature critical experiment to be
performed by ORNL for Pratt & Whitney Aircraft,

FUEL AND MODERATING MATERIALS
R. L. Hamner R. A. Potter
The production of test pieces of high- dens:ty,

high-purity beryllium oxide by hot pressing was

continued. Densities of 97 to 98% of theoretical
have been obtained, These pieces will be tested
for corrosion resistance in molten sodium, - Facili-
ties are being prepared for the fabrication of large
pieces.

A small batch of zirconium carblde was made by
the McKenna-Menstruum process.! . This material
is being processed for hot pressing. Physical
properties tests of the fabricated pieces will be
mfldfio

A hydriding furnace is belng built- from a desum
obtained from the General Electric Co.. The

 

'P. M. McKenna, Metals Progress 36, 152 (1939).

furnace will be used to produce both zirconium and
ytirium hydrides for use in physical properties
investigations and in cladding studies.

COMPONENT FABRICATION
C. E. Curtis

The problem of die design for the production of
a / in.-OD beryllium oxide rod with five longi-
tud:nal holes is being studied. This material is
desired for use in a Calrod type of thermocouple.

HIGH-TEMPERATURE X-RAY DIFFRACTION
G. D, White
High-temperature x-ray diffraction plots were

obtained for samples of Hastelloy B at 500, 600,
and 700°C., These plots were taken to determine

~whether an inversion occurs in the vicinity of

600°C. - At all temperatures a face-centered cubic
pattern was obtained., The only difference in the
patterns at the three temperatures was a slight
shift to larger d values with increasing temperature

because of thermal expansion.

FLUORIDE FUEL INVESTIGATIONS
G. D. White T. N. McVay, Consultant

Microscopic examinations were made of approxi-
mately 600 samples of fluoride salts. The samples
included routine test mixtures from experimental
engineering materials and component test stands,
quenches for equilibrium diagram determinations,

~and batches from various research experiments.

211

 
 

 

“ANP PROJECT- PROGRESS REPORT

a ,L?j ¥ 5%" 25 3.7. NONDESTRUCTIVE TESTING STUDIES

 

Ro Bo OI iver

EDDY-CURRENT INSPECTION OF
SMALL.DIAMETER TUBING

J. W. Allen
The results of eddy-current inspection of 14,540 ft

of CX-900 Inconel tubing are presented in Table

3.7.1. The undersize and oversize figures include
rejections on both out-of-tolerance wall thicknesses
and out-of-tolerance diameters, since the two appear
inseparably in the cyclograph type of eddy-current

" readout.’  The discontinuity indications were due

to cracks, gouges, and the pickup of foreign metal
on the inside of the tube.
The presence of foreign metal embedded in the

. inside wall of 3/ -m.-OD 0.025-in.-wall, CX-900

Inconel tubing has been detected in recent sh:p-
ments by the encircling-coil eddy-current tesf.
Subsequent longitudinal macrosections of two such
areas that occutred in one 42-in. length of tubing
are shown in Fig. 3.7.1, along with the eddy-
current signal trace produced by this tube. The
defective areas are indicated by the sharp down-
ward spikes in the trace. The spikes are similar
to the spikes which result from short longitudinal
crocks, except that they are in the direction of
increased wall thickness rather than in the direc-
tion of decreased wall thickness. Stains of the
type shown in Fig. 3.7.1 are yellow to red in color
and are sometimes, but not always, present in the

 

1
J. W. Allen, ANP Quar. Prog. Rep, June 10, 1956,
ORNL-2106, p 214.

areas that have picked up foreign metal, Trans-
verse microsections through several of the defec-
tive areas have failed to reveal that the defects
have perceptible depth.  Therefore the foreign
metal must be highly permeable to have produced
such large eddy-current indications as those shown
in Fig, 3.7.1, which suggests that the foreign metal
consists of bits of steel picked up from the mandrel

over which the tubing was drawn. Further metallo-

graphic study and microspark spectrographlc analy-
sis of these areas is planned

INSPECTION OF TUBING BY THE = -
ULTRASONIC.METHOD_ o

R. W. McClung

Large quantities of tubing are now being in-
spected by the immersed ultrasonic technique.
During this quarter 4224 pieces of 9&6-in.-0D,
0.025-in.-wall, 42-in.-long, CX-900 Inconel tubing
were inspected; 4143 lengths were accepted for
critical use. The rejection rate was slightly less
than 2%, and thus it appears that hlgh-quahty
tubing is being received. All the rejected lengths
contained only very small, single defects. Attempts
to metallographically examine some of the defects
were unsuccessful because of the difficulty of
precisely marking the defect and the tendency of
the metal to smear over small defects during
polishing. Further metallographic studies are
planned.

A total of 2360 ft of 0.242-in.-ID, 0.080-in.-wall,

thermocouple~well tubing in random lengths from

TABLE 3.7.1. RESULTS OF EDDY-CURRENT INSPECTION OF 14,540 f+ OF CX-900 INCONEL TUBING

 

Number of Pieces

 

Total Number Total Number Number of Number of
Tubing Size of Pieces of Pieces Pieces Pieces with Indications
Inspected Rejected Undersize Oversize of Dirscontirnuities
% -in- OD, 0.025in. 3003 167 71 57 39
wall, 42-in. lengths
0.229-in. OD, 0.025-in. 229 3 3 0 o
wall, 78«in. lengths
0.242~in. ID, 0.035in. 236 6. 6 0 0

wall, 132-in. lengths

 

912

 

 

 

 

 

 
 

at

   
  
  

PERIOD ENDING SEPTEMBER 10, 1956

UNCLASSIFIED
ORNL-LR-DWG {4681A

INSTRUMENTATION: CYCLOGRAPH AND ENCIRCLING COIL.
FREQUENCY: 200 kc

INCONEL TUBING: %-in.-OD, 0.025-in,-WALL, 42-in. LENGTHS

Fig. 3.7.1. Eddy-Current Signal Trace of Defective 46-|n.~OD, 0.025-in.-Wall, Inconel Tubing and

Macrosections of Two Areas Detected.

10 to 12 ft was mspected ulfrosomcally. Each of

28 lengths was found to have ‘one short defect.
One length had @ series of short defects throughout

the tube. Since the minimym usable length of this
tubmg was 10 £, only about one-half the rejected

lengths were usable after removal of the defective -
'areo. Thus the rejection rate was: approxlmatelyv»'
6%, Metallographm sectioning -of o few of the
defectwe areas disclosed o small gouge on the
outside of one tube’ that was npproxlmutely 0.005m.’>.‘
‘deep, 0, 012 in.- wide, and 32 in. long,-and on the
outside ‘of another’ tube there .was a small crack :

upproxlmately 0.003 in; deep. Agam, trouble was

encountered in precisely locating the defects, and

there was some smearing durmg metcllogrcphlc
polishing.

44

 

Over 1000 ft of 3/-ln.-OD 10.035-in.-wall, Hastel-
on C. tubing .in rundom lengths. was inspected
ultrasonlcaily. The rejection rate for this tubing
was approxlmately 30%. Metallographic dectioning
disclosed the presence of the ‘cracks and other
discontinuities -illustrated in Flgs. 3.7.2 through

- 3.7.5. A photomacrograph of the inside surface

along the weld of weld-drawn Hastelloy C tubing

‘is_shown in Flg. 3.7.2. There are tiny tears along
_the -weld to parent metal interface, and there is an
-:-;FGpparenf lack of fusion. An internal crack which
~did not extend either to the inside or to the out-
side surface of the tube. wall is shown in Fig.

3.7.3. The crack appears to be in the heat-affected
zone adjacent to the weld. An unetched sample of
Hastelloy C tubing that has two c¢racks is shown

213

0o<

 
 

 

 

 

 

ANP PROJECT PROGRESS REPORT

 

Fig. .7.2. lnside Surface of Weld-Druwn Hastel-

loy C Tubing. _

 

 

in Fig. 3.7.42. One crack is approximately 0,015
in. deep, and the shallower one is about 0.005 in.
deep. Etching of this sample, as shown in Fig.
3.7.4b, disclosed that the cracks were actually
about 0.020 in. and 0.015 in. deep.

Another crack is shown in Fig. 3.7.5¢ and & in
the etched and unetched conditions: It may again
be noted that the crack, although it is about
0.015 in. deep, does not extend to either surface.
Again, difficulty was encountered in the preparation
of metallographic sections because the soft matrix
smeared over the defects during the polishing
operation. However, a greater degree of success
was achieved in locating the defects in this ma-
terial than in the Inconel tubing, This can probably
be attributed to the larger size of the defects that
seem to regularly occur in Hastelloy tubing,

 

~ Fig. 3.7.3. lnternal Cruck in Heat-Affected Zone Adjacent to Weld in Weld- Drawn Hastelloy C Tubmg.

Etched with chrome regia. 100X,

214
 

 

PERIOD ENDING SEPTEMBER 10, 1956

»

 

"

 

 

 

 

in Hastelloy C Tube. (a) Unetched. (b) Etched with chrome regia.

Intergranular Cracks

ig. 3.7.4.
Reduced 13%

F

100X

 

 

215

 

 
 

ANP PROJECT PROGRESS REPORT

LAY

4

 

 

 

 

 

 

 

 

 

ith chrome regia. 100X,

(6) Etched w

() Unetched.

Fig. 3.7.5. A Crack in Hastelloy C Tubing.

Reduced 13.5%.

Ces

N
"
@

216

 
 

 

e

e

Over 1000 ft of %-in.-OD, 0.035-in.-wali, CX-900
Inconel tubing in random lengths was examined
ultrasonically, and a total of approximately 25 ft
was rejected. The few defects detected seemed to
be very small. In general, the quality of this
tubing was quite high, as indicated by the rejection
rate of less than 2.5%. Metallographic examing-

tions also failed to disclose any defects of appre-

ciable size. Of the 229 pieces of 0.229-in.-0D,
0.025-in.=walf, 78-in.-long, CX-900 Inconet tubing
also examined ultrasonically, 31 pieces were re-
jected,

An attempt is being made to compare the de-
tection of defects by ultrasound and by other in-
spection methods, and, for the larger defects,
correlations are normally possible. However, for
many of the very small defects, particularly those
on the inside of tubing, ultrasound seems to be
the only feasible detection methed. It is hoped
that, despite the inherent difficulties, metallo-
graphic examinations will continue to provide
additional information concerning the nature of the
defects.

INSPECTION OF PIPE
Jo Ko Whi'e Ro Bo :Oliver

The pipe inspection facility was installed in
April, and about 6000 ft of pipe ranging in size
from "3’ in. IPS to 7 in. OD has been inspected;
most of this material has been Inconel. As a
result of this experience only minor changes in
the details of the inspection technique previously
described? have been made. Because of the ex-
cessive camber in large pipe and the end whip
frequently experienced in small pipe, the present

system of using the pipe as a transducer guide

appears to be the only practicel method for main-

taining alignment, In this -system a hand-held -

search-tube positioner is used with a series of
interchangeable side plates cut to fit each plpe
size. The positioner is translated along the pipe
by hand, and @ moter-driven string is used as @
speed guide (Fig. 3.7.6). The positioner (Fig. 3.7.%)
con . be adjusted to change the transducer-to-pipe
distance (delay), the rotation of the transducer in
the plane perpendicular to the sound beam, and
the angle of incidence of the beam upon the outside

surface of the pipe. The angle of incidence de-

 

2), K. White, ANP Quar, Prog. Rep. June 10, 1956,
ORNL-2106, p 216.

644

PERIOD ENDING SEPTEMBER 10, 1956

 

Fig. 3.7.6. Facility for the Inspection of Pipe
by the Immersed Ultrasonic Method.

termines the depth of inspection of the pipe wall.,
With materials of the same acoustic impedance as
Inconel and approximately the same wall-thickness-
to-diameter ratio as sched-40 pipe, the usaoble
range of angles is from 10 to 30 deg. Incident
angles of less than 10 deg cause interference

- because of reverberations across the pipe wall.

Angles greater than 30 deg set up Rayleigh (surface)
_waves3 and occentuate scraiches selectively more
than deeper defects. Within these limits the inci-
dent angles used are chosen to balance the indi-

_cations from 5% notches on the mslde and outslde

surfaces. Internal defects show up strongly when
angles of about 10 to 17 deg are used, External

,_defects show up strongest ‘when angles of about

,23 to 90 deg are used.
" Rotation of the transducer in the plane perpen-
dtcular to the 'sound beam is sometimes necessary

 

3E, G. Cook and H. E. Van Valkenburg, ASTM Bull
198, May 1954.

217

¢oo

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

UNCLASSIFIED
PHOTO 17936

     

. TV - ‘
mfr"\:‘\"‘" :‘B“ }www.w}

Fig. 3.7.7. Ultrasound Search Tube Mounted for
the Inspection of Pipe.

because the transducers presently available do
not emit energy uniformly. Apparently, uniform
crystal damping is difficult to achieve and poorly
damped areas, or ‘‘hot spots,’”’ are common. The
purchase of a square transducer is contemplated to
minimize this “hot spot’” effect and to achieve
higher coverage per revolution and hence greater
scanning speeds.

Heretofore, experience had been limited to in-
spections made at a frequency of 5 Mc, but for
thick-walled pipe, 2.25 Mc appears to be a promis-
ing frequency, as evidenced by some recent success.
More observations will be made ot this frequency
in the near future,

It has been observed that, in general, pipe quality
has improved since the inspection program was
initiated. Several entire pipe orders were rejected

218

~ during the early stages of the inspection program.

During the lost few months the special core exer-
cised in preparing the pipe, crating it for protection
during shipment, etc., have resulted in marked
improvement in quality, as evidenced by numerous
pipe orders inspected without detection of a single
defect,

INSPECTION OF THIN SHEET

J. W, Allen R. W, McClung-
R. B. Oliver o

Sheet material has extensive application in power
reactors, and, in many cases, formed sheet will

function both as a container for fluids and as a

heat transfer medium between two fluids. Since
high heat fluxes and thermal stresses are involved,
the presence of laminar defects is very undesirable.
The problem of detecting such defects in sheet
has, currently, only one solution, which, in turn,
poses a very difficult mechanical problem. This
situation has motivated an investigation to develop
a similor and betfer inspection method. Liquid

penetrant, radiographic, and eddy-current methods .

cannot be used because of the unfavorable defect
orientation. The ultrasonic resonance method is
not an adequate approach, since it is not capable
of resolvingsmall defects and it inherently requires
slow contact scanning of the sheet. The conven-
tional pulse-echo vitrasonic inspection, even with
a pulse length as short as 1 usec, is not practical
for sections thinner than 0,20 in. The transmission-
attenuation technique is capable of detecting small
laminations in thin sections, but, since it requires
critical alignment of two transducers on opposite
sides of the sheet, it is a very difficult method to
apply to the inspection of large or nonplanar areas.

A new ultrasonic method has been proposed, and
preliminary experiments have given promising
results, The new method requires a pulse of ultra-
sound having o duration of 5 to 20 psec, with the
sound tuned to such a frequency that the sheet
thickness is an exact multiple of the half wave-
length,  With these conditions the ultrasound

reverberates, or rings, between the two sheet sur-

faces for a period of time that is several times
greater than the pulse duration, When a lamination
exists in the sheet the ringing is decreased, both
in duration and amplitude, as a function of the
area of the lamination relative to the transducer
area. To test this method, flat-bottom holes with

GCid  0o7

C
 

 

several different areas were milled into one side
of a sheet of Inconel to various depths. In experi-
ments with existing equipment most of these
reference defects were located. To properly in-
strument this test method, the reflectoscope has
been drastically altered. The pulse repetition was
increased from 60 to 500 pulses per second,

PERIOD ENDING SEPTEMBER 10, 1956

appropriate filters were added to the circuits, a
variable-sweep delay circuit was added to permit
immersed scanning, and external connections were
provided for synchronization signals. This last
chonge will permit the addition of gated alarm
circuits and various data presentation and recording
units in the near future,

219

644 008

 
 

 

 

 
 

 

ne

 

PERIOD ENDING SEPTEMBER 10, 1956

 

material was not available, this material was re-

worked by centerless grinding in order to remove

most of the surface imperfections. A waiver was
made on the porosity, which resulted in rejection
of only large-size porosity defects,

INSPECTION OF COMPONENTS
R. L. Heest_ands =

Sixty thermal-convection loops were received,
and the welds were inspected by the dye-penetrant

‘method. Thirty-six of these loops were found to

have welds which showed numerous indications of
cracks, pinholes, and other defects. The remaining
loops were accepted for use. The defective loops
were repaired and reinspected prior to acceptance.

Thirty-eight pieces of Inconel plate ‘were in-
spected prior to shipment for fabrication into
dished heads, and then they wete re-examined for
manufacturing defects upon return. Several were
found to have numerous pits that were apparently
caused by a foreign material becoming embedded
in the surfaces during the pressing operation.
Four heads were found to have cracks running from
the edge inward, which appeared to be as deep as
Y in. These cracks are to be repaired, and the
areas will be reinspected prior to use.

Four small heat exchanger units were received
and inspected, They were found to be acceptable
for use in test operation,

 

10on assignment from Pratt & Whitney Aircraft,

644 009

.FLUORESCENT-PENETRANT INSPECTION
OF TUBING

G. Tolson

The installation of the fluorescent-penetrant
equipment to be used in the inspection of small-
diameter, thin-walled tubing was completed, Tests
are now being performed to compare the fluorescent-
penetrant with the dye-penetrant type of inspection

used heretofore. Indications are that it will be

possible to ensure higher quality tubing with the
new method because of its higher sensitivity., The

fluorescent penetrant has revealed small pinholes
“and tight laplike defects which were not dis-

cernible by the dye-penetrant method. Exploration
of some of these areas by polishing showed them
to be as deep as 0.002 in. in some cases. How-
ever, the greater sensitivity of the new method will
present some problems, initially, because ex-
perience will be required to distinguish indications
of superficial imperfections from indications of
true defects.

WELDER QUALIFICATION PROGRAM

A, E. Goldman

- The number of welders qualified under ORNL
welding specifications for ANP work is now 42,
An estimate of the additional qualified welders

" that will be required is now being prepared, and as

soon as these requirements are known, qualifica-
tion tests will be given. A qualification program
is now under way at the Paducah installation, but,
to date, no welders have qualified.

 

221

 

 
 

 

 
 

 

 

 

Part 4

HEAT TRANSFER AND PHYSICAI. PROPERTIES

H. F. Poppendiek

RADIATION DAMAGE

G. W. Keitholtz
'FUEL RECOVERY AND REPROCESSING
R. B. Lindaver

CRITICAL EXPERIMENTS

A. D. Callihan

 
 

"
"

P

 
R L LA e A e e P

 

 

 

 

0

4.1. HEAT TRANSFER AND PHYSICAL PROPERTIES
H. F. Poppendiek

ART FUEL-TO-NcK HEAT EXCHANGER
J. L. Wantland - S, |, Cohen

The fluid friction characteristics of the ART
fuel-to-NaK heat exchanger were determined with
six 60-deg staggered spacers and six 60-deg in-

clined spacers alternately placed at 6-in. inter-

vals.! These data are presented in Fig. 4.1.1.

A full-scale model of the present ART fuel-to-
NaK heat exchanger (which contains more tubes
and @ somewhat different spacer configuration than
the previous system) has been assembled, Pres-
sure drop data will soon be obtained for it. The
friction characteristics of this heat exchanger have
been predicted by using the technique previously
described.?2 The results indicate that the experi-
mental data to be obtained should lie very near the
curve in Fig. 4.1.1 for the staggered and inclined
spacers, :

 

5. L. Wantland, Transverse Pressure Di ference
Across Staggered and Inclined Spacers in the ART
fguel;to-NaK Heat Excbanger, ORNL CF-56-6-'|43 {June

56

25, L. Wantland, A Method of Correlating Experimental
Fluid Friction Data for Tube Bundles of Different

‘Size and Tube Bundle to Shell Wall Spacing, RNL

CF-56-4-162 (April 5, 1956).

woTREY
ORNL—LR—DWG 16180

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

O
* N~ l—r= 0.475 #3039
« R SL
,EJ0.0S' : - - H‘IW ~==
- — : ,
w . *025 7\ ——
| reomemon T |
2000 - 5000 - - -10,000

REYNOLDS NUMBER, My, 1"

Fig. 4.'|.'|.. Fricfi'o'n 'Churucterisflcs of 60-Jeg
Staggered Spacers and ‘60-deg Inclined Spacers in
the ART Fuel-to-NaK Heat Exchanger,

Isothermal friction measurements for the delta-
array heat exchanger have nearly been completed.
Heat transfer experimentation will commence in the
near future.

ART HYDRODYNAMICS

C. M. Copenhaver F. E. Lynch
Go Lo MU“er3

Core Hydrodynamics

A new method for stabilizing the flow in the ART
core was studied. |t consisted of placing a number
of screens in the northern hemisphere of the core,
as shown in Fig. 4.1.2. Qualitative velocity pro-
files, which were determined by the phospho-
rescent-particle flowevisualization technique, are
shown in Fig. 4.1.3. The regions of flow sepa-
ration on the outer core shell walls, which existed
in a system without screens, apparently were
eliminated because of the local turbulence gener-
ated by the screens.

The over-all friction loss in the core was deter-
mined and is plotted in Fig. 4.1.4 in terms of a
loss coefficient and the Reynolds number, The
loss coefficient of this system was also estimated
on the basis of screen friction data available in
the literature. The two loss coefficients were in

 

30n assignment from Pratt & Whitney Aircraft.

SELRET
ORNL—L.R—DOWG {6181
STRAIGHT-THROUGH FLOW

 
 
 
   

~ ALL SCREENS
‘/ ARE 20 mesh/in. .
00108-in. WIRE DIA =5 b
- 0.385 SOLIDITY 2.24in. )
3 2.98in.

 

 

 

818in.

 

 

Fig. 4.1.2. Schematic Diagram of Screen Arronge-
ment in 10/44-Scale Model of ART Core.

225

 
 

 

 

 

ANP PROJECT_PROGRESS-REPORT

| SEGRET
ORNL-LR-DWG 416482

    
  
 

CORE ENTRANCE

 

SCREEN POSITIONS

 

THICK BOUNDARY
LAYER
OBSERVED AT A MIDPLANE
REYNOLDS NUMBER OF 6000
CORE EXIT |

Fig.r 4.1.3. Qualitative Velox":il'y Profiles for

* Straight-Through Flow in Model Shown in Fig. 4.1.2.

general ogréemént. In an actual reactor system the
screens would probably be replaced with perforated
plates having an equivalent pressure drop. A rough
“estimate of the amount of Inconel that might be
required for such plates in an ART core is about
7 kg. ,

Flow studies are to be started soon on a straight,
anmular core with a diverging entrance and con-
verging exit. Screens will be placed in the en-
trance. The advantage of this arrangement is that
" the screens will be in a relatively low neutron-flux
region and thus will yield less reactor poisoning
than they would in the system described above.

226

SeREFS
ORNL-LR-DWG 16183

 

ED IN ART

n

ESTIMATED FROM TURE DATA |

FRICTION LOSS COEFFICIENT, &

n

104 2 5 40
REYNOLDS NUMBER AT MIDPLANE

Fig. 4.1.4. Friction Loss Cocfficient in ART
Core with Screens as a Function of the Mid-Plane
Reynolds Number.

Sodium Flow in Reflector Cooling System

The pertinent data for the flow distribution in the
reflector cooling system for concentric and ec-
centric annulus conditions were presented in detail
in the preceding report.4 The static pressure dis-
tributions for the reflector and island annuli were
not included, however, and they are presented here.
The experimental studies indicate that the pressure
losses through the annuli can be expressed as

fL pV2
AP _(2t + nKs+Ka) 2% .
where |
{ = friction factor in a smooth duct having an
equivolent diameter of 27,
t = annulus thickness,
L = annulus length,
n = number of spacer rows,
K, = resistance coefficient per spacer row,
K, = diffusion coefficient (1.0 for reflector an-
nulus and 0.5 for island annulus), o
P = fluid density,
V = mean velocity in annulus,
g = acceleration of gravity.

 

4C. M. Copenhaver, F. E. Lynch, and G. L. Muller,

&
o

.

o
 

 

 

 

n

The results of static pressure distribution calcule-
tions are presented in Fig. 4.1.5. :

Fluid Flow Visualizction Stu&y |

A new attack on the problem of photographing
phosphorescent particles used in-the flow-visuali-
zation technique has been initiated. The following
factors are being investigated: (1) -the matching of
the phosphorescence emission spectrum and the

~ film sensitivity spectrum, (2) optimization of the
~ optical system of the camerg, -and (3) optimization
~of the film development process. - If satisfactory

‘photographs can be obtained, the qualitative phos-
~ phorescent-particle -flow-visualization method can
 become « quanhtatlve method. |

'ART CORE HEAT TRANSFER STUDY
- N, D. Greene - H. F. Poppendiek
~G. L. Muller . G. M. Winn

The results of an exPenmental and unolytlcul._ -

study -of the temperature structure in an uncooled

‘ART core with a swirl entrance system were pre-
~sented in the prevsous report.5 - Experimental mean
" ‘and transient temperature fields in the outer and
‘inner “core shell walls and within the circulating

electrolyte were obfomed from the ART voiume-
heat-source model, ' -
Recently « study pfr the temperature'_Stru_cture in

" the ART core With a vaned entrance was come

pleted. The vane system, described previously,®

~ eliminated the flow separation region on the island

in the northern hemisphere. Six complete power
runs ‘were nidde for fh’e case of bofh pumps. in

'=runged from below to above dessgn flow conditions. o
= -Three power runs were ‘also made for the slmulcted
- case of ‘‘one pump off,"’ Heat bolances were agcun
L rwnhin +4% of being perfect,
. The.inean, uncooled wall and flund temperature’
- profiles . obtained in'the ART core with- the vaned
= entrance are presented iin Fig. 4.1,6. ‘The asym-
"_r'f_.;fQ_metfles in-the outer core shell and island (inner)
- ~core "shell ‘wall temperatures ‘can be explained on
. the baosis of hydrodynamlc flow usyrnmetrles. Note
. that the high island core shell wall temperatures
"",?'—';found for the prevnous swarl-entrcmce case «are no

 

- SN, D, Greene, et al,, ANP Quar Prog. Rep. june 10,

1956. 0RNL-2106 p 222.

6G. D. Whitman, W. J. Stelzman, ond W, Furgerson,
ANP Quar. Prog. Rep. March 10, 1955, 0RNL-206|, p 24.

PERIOD ENDING SEPTEMBER 10, 1956

BECRE T
ORNL-LR-DWG 16184

N =~y
= o

o
(=]

REFLECTOR ANNULUS

bH
o

ANNULUS

STATIC PRESSURE (psi)

-8 8

3

 

o‘ .

.20 16 {2 8 4 0O -4 -8 -42 -6 -20
Z, AXIAL DISTANCE FROM CORE EQUATOR (in.}

Fig. 4.1.5, Static Pressure Distributions for ART

- Reflector and Island Annuli.

‘ “BEoRME
ORNL-LR-DWG 16185
QUTER CORE SHELL
IDEALIZED ART (THEORETICAL})
I.SLAND CORE SHELL .

MIXED MEAN FLUID ~

HELICAL Re 134,000
(VANED ENTRANCE)
W= W (UNIFORM)
Pr=4.0

 

0. 2 4 8 B 10 12 14 6 8
' AXIAL DISTANCE FROM INLET {in.)

" Fig. 4.1.6. Mean, Uncooled Outer Core Shell and
. "'Islund (Inner) Core Shell Wall. Temperature Meas-
~-urements for the Half-Scule ART Core Model with
@ Umform Vo!ume Heut Source. .

*|onger present because the vane system confumed
. a deflector ring which prccflcully eliminated flow
~separation ‘on the island core ‘shell wall. “The wall
‘temperature solution for -an idealized ART (a par-
'allel-plutes Sysfem7) is also plotted in Flg. 4.1.6;

 

7H. F. Poppendiek und L. D Pulmer, Forced Convec-
tion Heat Transfer Between Pargllel Plates and in
Annuli with Volume Heat Sources Within the Fluids,
ORNL-1701 (May 11, 1954).

227

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

this predicted uncooled-wall temperature profile
lies between the island and outer core shell wall
temperature measurements. Typical experimental
transient wall and fluid temperature measurements
are shown in Fig. 4.1.7. The results are expressed
in terms of the total temperature fluctuation divided

by the axial temperature rise of the fluid gomg_

through the core.

SEOREY
ORNL—LR—-DWG 16186

U 2
P~ A~ = 0.25

g Bty

 

 

 

 

 

ISLAND SHELL AT EQUATOR

 

 

be—{ §8C—

 

T\

— At
— = 0.8
Bim

AV _r

CORE SHELL AT EQUATOR

 

 

 

 

 

1A
(AU NN | e

MID-STREAM BEYOND EXIT

 

 

 

 

 

 

 

 

 

 

 

 

 

0 1.0 2.0 3.0
TIME (sec)

HELICAL Re = 131,000
. - W = W {UNIFORM) -
. Afs = TOTAL CHANGE IN WALL TEMPERATURE
Aty = AXIAL FLUID TEMPERATURE RISE IN GOING THROUGH CORE

Fig. 4.1.7. ‘Tfl:n.sient ART Surface and Fluid

Temperatures (Uniform Yolume Heat Source).

The mean, uncooled outer and island core shell
wall temperatures (except the island core shell wall
in the northern hemisphere) for the vaned system
were greater than those found previously for the

" swirl flow system; this results from the helical

Reynolds number for the vaned system being signif-

- icantly lower than that for the swirl system. The

228

temperature fluctuations of the island and outer core
shell walls in the northern hemisphere were rela-
tively lower for the vaned system than for the swirl
system; conversely, the temperature fluctuations
in the southern hemisphere were relatively larger

for the vaned system _than' for- the swirl system.
- The fluid temperature fluctuations beyond the core
‘exit were also larger for the vaned system than for

the swirl system..
~ The frequency spectrums of the temperature fluc-

tugtions for the vaned system were very similar to
the spectrums observed in the swirl system. The
‘peripheral asymmetries were greater in the vaned
system than in the swirl system; this characteristic'

was also observed in the correspondmg hydrody-

) nomlc Structure,

"The uncooled wall temperature rises obove the
mixed mean temperature in the actual ART system,
where the radial volume heat source distribution

will be approximately a hyperbolic cosine function,

were previously shown to be greater then twice
those for a uniform power density distribution such
as that which existed in the volume heat source
experiment. Consequently, temperatures as high
as about 1850°F may occur in the vaned system if
the core shell walls are not cooled properly, Some
interpretations of the wall temperature fluctuations
in the outer core shell, island core shell, and heat
exchanger walls in the actual reactor system have
been described previously. For example, in the
event that a momentary flow stagnation exists
adjacent to the core shell wall, the fuel-Inconel
interface temperature fluctuation will be only about
one-fourth as large as the fluctuation in the tem-
perature of the fuel at some distance from the wall;
this reduction in the temperature rise -occurs be-
cause of the relatively good heat transfer for the
Inconel when the fuel is momentarily stagnant,
However, when a high-velocity turbulent eddy of
lower or higher temperature level suddenly wipes

the Inconel surface, calculations have shown that

the relative heat transfer to or from the Inconel is
then poor, and consequently the fuel-Inconel inter-
face temperatures are nearly as large as those that
would exist if the wall were insulated,

THERMAL-CYCLING EXPERIMENT
H. W. Hoffman D. P. Gregory®

The experimental system designed to investigate
the effect of thermal cycling at an intermediate

 

80n assignment from Pratt & Whitney Aircraft.

5%

M,'
 

frequency range ( 1/4 to 2 cps) on metals in the

. presence of reactor fuel mixtures has been suc-
' cessfully operated, The data obtained at two
.. temperature levels with NaF-ZrF +UF (50-46-4

mole %) flowing turbulently through 12 in,-0OD,
: 0.035-in,-wall |nconel tubes are summarized in
5 Table 4.1.1. The corrosion attack on the metal in
both the heater and test sections during these runs
is shown in Figs. 4.1.8 through 4,1.12,

PERIOD ENDING SEPTEMBER 10, 1956

In the ART core, the high differential temperature
cycling will be experienced by the Inconel core
shell surfaces and the heat exchanger tubes as a
result of the hydrodynamic instabilities that will
exist in the system. These temperature fluctua-
tions have been simulated in a bench-scale experi-
ment by subjecting the salt flowing through an
electrically heated tube to cyclic heating. The
experimental system is shown in Fig. 4.1.13. As

TABLE 4.1.1. PRELIMINARY THERMAL-CYCLING DATA FOR INCONEL TUBES EXPOSED TO
THE FUEL MIXTURE (NO. 30} NaF-ZrF ,-UF, (50-46~4 MOLE %)

 

Heoter Section

Test Section

 

 

 

Mean
Duration  Inlet Fluid Inside Surface Average inside Surface Average Cause of
Run (hr) Temperature . Temperature (0 F) Depth of Temperature ‘o F) Depth of Termination
(°F) - Attack Attack of Test
Average Fluctuation (mis) Average Fluctuation (mils)
ET-A 16 1265 1455 tis0 1273 17 0.5 Oxidized electrode
4 1265 1546 +225 1287 112
. ET-B 240 1275 1415 *170 1330 116 1 Stopped to alter
test conditions
ET-C 4 1580 1755 189 1613 136 2 Melted electrode
: ET-D 14 1580 1845 1240 1590 183 4 Tube break
ET-E 5 1615 1905 1235 1635 143 0.5 Tube break

 

 

duced 31%. (Secret-with-eeptiony——

 

 

. Fig. 4 1.8. Corrosuon Results of Thermal Cyclmg of Inconel Tubes Exposed to the. Fuel Mlxiure‘
(No. 30) NuF-ZrF ‘UF, (50-46-4 mole %). Run ET-A: mean fluid temperature, 1265°F; heater inside
surface temperafure, 1500 1+ 200°F; test section msude surface temperature, 1280 + 10°F, 250X, Re-

229

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

the fluid moves downstream, the surface of the test
section is exposed to periodic temperature fluctua-
tions imposed on the fluid in the heater. Because
of the poor thermal diffusivity of Inconel, the major
effect of this thermal. cycling is confined to a
region close to the metal-fuel interface. In this
region the combination of the high coefficient of

 

 

CTREN

 

EEEPEBRET

thermal expansion and the high modulus of elas-
ticity of Inconel, with its poor thermal conductivity,
may result in thermal stresses which are of suffi-
cient magnitude to cause eventual damage to the
metal by fatigue. These thermal-cycling experi-
ments are being made in order to defermme the
extent of fatigue damage. |

 

Fig. 4.1.9. Corrosion Results of Thermal Cycling of Inconel Tubes Exposed to the Fuel Mixture
(No. 30) NaF-ZrF -UF (50-46-4 mole %). Run ET-B: mean fluid temperature, 1275°F; heater inside
surface temperature, 14]5 t 170°F; test section inside surface temperature, 1330+ 1°7F. 250X. Re-

. duced 31%. {Secretwith-ceptiond

INCHE!

TEFEIAELE

 

 

 

 

¥
L

 

 

Fig. 4.1.10. Corrosion Results of Thermal Cycling of Inconel Tubes Exposed to the Fuel Mixture
(No. 30) NaF-Z(F ,-UF, (50-46-4 mole %). Run ET-C: mean fluid temperature, 1580°F; heater inside
surface temperature, l755:|: 190°F; test section inside surface temperature, 16]3137°F 250X. Re-

duced 31%. 4beTretwith-eaption)

230
 

 

PERIOD ENDING SEPTEMBER 10, 1956

The use of the ““floating’’ electrodes previously  supported from above by thin, flexible rods. The
described? has been temporarily abandoned be-  power connections are made through copper braids
cause of continuing failures at the electrodes. clamped to the electrodes. The bellows provided
The modified electrodes (Fig. 4.1.13), fabricated to relieve the stresses of linear thermal expansion
of Inconel, are welded or brazed to the tube and  has been relocated so that the forces are applied

- “along the axis of the bellows,

9H. W. Hoffman and D. P. Gregory, AfiP éwr. Prog. While the thermal cycling data presented here
Rep. June 10, 1956, ORNL-2106, p 227. - are preliminary, the rate of corrosive attack on the

 

 

-Fig. 4. l 'll Corros:on Results of Thermal Cycling of Inconel Tubes Exposed to the Fuel Mixture
(No. 30) NaF-ZrF -UF, '(50-46-4 mole %). Run ET-D: mean fluid temperature, 1580°F; heater inside
surface temperature, 1755 1 190°F; test section inside surface femperature, 1590 £ 83°F, 250X. Re-

duced 30%. {Sem#-wn'frcupfrefi)—-

 

Fug. 4.1 12 Corrokion Results of Thermal Cyclmg of |nconel Tubes Exposecf 1o fhe Fuel Mnxture
(Ne. 30) NaF.IF -UF, (50-46-4 mole %). Run ET-E: mean fluid temperature, 1615°F; heater inside
surface temperature, 3905:!:235" F; test section inside surface temperature, 1635 + 43°F, 250X. Re-

duced 31%. «(Serret-withcoption)~

231

 
Zee

 

   

UNCLASSIFIED g
PHOTO 26932

 

[
AR TN I
SRR

AT

EERALE LA
IAESREEANEREE)

 

T

wt

LAOJIN §SFJ00¥d LI3IF0dd ANV

 
 

 

 

inconel for the runs ET-C, ET-D, and ET-E is note-

worthy. Whether this is to be attributed to the very
large fuel volumes and the small areas of heated
surface, to the high temperatures at the metal-fluid
interface, or to thermal cycling of this interface has
not yet been established. An experiment is now in
progress to establish a base corrosion rate for this
specific system under isothermal conditions with
interface temperatures in the range of 1600 to
1800°F. It is to be noted that both runs ET-D and
ET-E failed due to breaks in the Inconel tube at
the outlet end of the heater, immediately adjacent
to the electrode. The effect on the tube of the
constraint imposed by the electrode is not known,
Experiments are planned in which the electrode
will contact only .one-half of the tube circumference
at the electrode position. Alternative approaches
to thermal cycling experiments that eliminate the
need for electrodes are being investigated,

SHIELD MOCKUP CORE STUDY
Lo. Co Pdlmer

The recently initiated study of the temperature
structure in the Shield Mockup Core is continuing.
In particular, the temperature distributions in the
beryllium reflector and the various Inconel shells
are being determined, An effort is being made to
explore the possibility of eliminating the cooling
system for the beryllium reflector by the reduction
of certain high thermal resistances in the system.

 

105, |. Cohen, W. D. Powers, and N. D. Greene,
A szstcal Property Summary for ANP Fluonde sztures,
ORNL-2150 (Aug. 3, 1956) _ o

PERIOD ENDING SEPTEMBER 10, 195§

HEAT CAPACITY
W. D. Powers

A study was conducted to determine the influence
of fission products in a fluoride fuel mixture on the
enthalpy and heat capacity in the liquid state,
Quantities of RbF, BaF,, and LaF, were added to
a zirconium-base fuel to sumu!ate an ANP fuel
heavily laden with fission products resulting from
1000 hr of exposure to a neutron flux of 1015
neutrons/cm2.sec.  The ‘results of the measure-
ments, presented in Table 4.1.2, indicated that,
within expenmental error, the change in heat
copoclty as a result of the addition of simulated
fission products was negligible. This result had
previously been predicted on the basis of the
established heat capacity correlation expression.
The enthalpies of the fuel with the additives were
about 10 cal/g less than those of the fuel alone.
Earlier enthalpy and heat capacity measurements
on the fuel without additives are also given in
Table 4.1.2 to illustrate the reproducibility of the
data.

VISCOSITY AND DENSITY
S. I. Cohen

A summary of the density, viscosity, heat ca-
pacity, thermal conductivity, electric conductivity,
and surface tension measurements that have been
made at ORNL during the past several years was
prepared and issued.'® Viscosity measurements

were made on LiF-BeF,-ThF -UF, (71-16-12-1

~ mole %), which is a mlxture of possntle interest in

TABLE 4.] 2. COMPAR!SON OF ENTHALPIES AND HEAT CAPACITIES OF FUEL MIXTURES
~ WITH AND WITHOUT SIMULATED FISSION-PROPUCT ADDITIVIES

 

 

 

 

Entholpy, = Hypoc : Heat Cupdcity
: {eal/g) : _ - {cal/g:°C)
Material — T

- A T A - At At At A
-600°C 700°C 800°C  600°C 700°C 800°C
_NaF-ZfF ~UF, (50-46-4 mole % o | |
Data obtained in 1955 1710 197.2 222.6 0266  0.258 - 0.249
' Dota obtained in 1956 - 1681 1952 - 221.3 0.276 0.266 0.256
NaF-ZrF -UF ,-RbF-BoF ,-LaF, 158.5 186.0  212.0 0.283 0.268 0.253

(47.6-43.7-3.8-1.1-1.0-2.8 mole %)

233

 
 

 

 

~the fused-salt power reactor program.

 

ANP PROJECT PROGRESS REPORT

The vis-
cosities were found to vary from about 13 centi-
poises at 600°C to 4.8 centipoises at 800°C,

THERMAL CONDUCTIVITY
W. R. Gambill

A tentative correlation has been developed for
the prediction of the thermal conductivities of fused
salts and salt mixtures at their melting points,!!
For this correlation the total conductive heat trans-
mission of the fused salt is divided into the two
distinct contributions, One is the atomic or lattice
portion, which arises from the short-range atomic
or moleculor order present in liquids, in general;
and the other is the ionic portion, which arises
from a drift of ions, and subsequent energy transfer,
between atoms.

 

Ny, R, Gambill, Prediciion of the Thermal Conduo
thty of Fused Salts, ORNL CF-56-8-61 (Aug. 10, 1956)

The atomic portion of the total conductivity,
(k,),, in Btu/hr-ft2(°F/ft), may be correlated by-
the expression

T!/z 4/5
(k,), = 1043 — .
F9/5

 

T, = melting temperature or Ilqmdus tempera-
ture (for mixtures), °K
P, = density of fused salt ut T, g/em?,
M = molecular weight for pure salts = Fx.M,
for salt mixtures,
x. = mole fraction of ith component, |
M, = molecular weight of ith component.
The ionic portion of the total conductivity is
plotted in Fig. 4.1.14. With the exception of NaOH,

Saener
ORNL-LR-DWG 16187

 

 

 

 

 

 

 

NONELECTROLYTES
100
NaOH
90 = - "
N=§ X%; N;, WHERE x; = MOLE FRACTION OF /™" COMPONENT -
N; = NUMBER OF IONS PER MOLECULE
80 — A=100 —

. WHERE &, = ATOMIC PORTION OF THERMAL CONDUCTIVITY
T

 

#,= TOTAL THERMAL CONDUCTIVITY

 

\
80 , 14
\ 30
50 35
\ a4
104

40

 

 

4 (%)

SALT MIXTURE NO.

COMPOSITION
" NoF-KF~LIF (11,5-42-46.5 mole %)
NoF=KF-LiF-UF (10.9-43.5-44.5-1.1mole %)}
NaF—Zl;F"-UF"(SO—46-4 mole %)}
NoF -~ BeF, (57-43 male %)
NoF-ZrF - UF,{53.56-40~6.5 mole %)
LiF ~RbF (43~ 57 mole %)

 

 

 

~N

e N

 

SALT MIXTURES > - HTS
NO. 42 AND 104 ‘\‘(

 

 

 

 

 

 

 

 

 

 

 

 

NO. 35
20 . ,
NO. {4 \\ NO. 44
NO. —g\
10 30
Q
o [ 2 3 4 5 6 7 8
N

Fig. 4.1,14. Thermal Conductivities of Fused Salts at Their Melting Temperatures and Snlt Mlxtures

at Their Liquidus Temperatures.

234
 

 

the data lie along @ smooth curve, The deviation

of NaOH is believed to be caused by the intense

hydrogen bonding (between OH radicals), which.

effectively reduces the ionic component of the
conductivity to a negligible value.

PERIOD ENDING SEPTEMBER 10, 1956

~ A relation for predicting the temperature de-
pendence of the thermal conductivities of fused
salts is currently being studied. [t can not satis-
factorily be evaluated, however, until more tem-
perature-dependent conductivity data are available,

235

 
 

 

ANP PROJECT PROGRESS REPORT

4.2. RADIATION DAMAGE
G. W. Keilholtz

EXAMINATION OF DISASSEMBLED MTR IN-PILE
LOOPS NOS. 3, 4, AND 5

A, E. Richt
c. E”is Ro No Ramsey
E. J. Manthos E. D. Sims
W. B. Parsley R. M. Wallace

Examination of MTR in-pile loop No. 3, described
previously,! was completed, except for metallo-
graphic examination of the pump impeller, Corro-
sion of the nose coil was found to vary from 1 mil
near the inlet to a maximum of 3 mils of inter-
granular attack near the outlet, Fig. 4.2.1. Com-
parisons of the inner wall of the nose-coil tubing

 

. P.LCarpenfer et al, ANP Quar. Prog. Rep, Dec, 10,
1955, ORNL-2012, p 27.

UNCLASSIFIED
ORNL~-LR~-DWG 16188

     

264 260
1455°F 1395°F
2 mils
261
263 . £ mil
262
1Y% mis
269
2 mils
273
270 1490°F
1435°F 3 mils
2' mils
' ' 271
{390-1490°F
2 mils
274
1415°F
OUTLET g-amis

“‘_Fig. 4.2.1. Diagram of Sectioned Nose Coil of
MTR In-Pile Loop No. 3 Showing Locations from
Which Metallographic Samples Were Taken, the

'Maximum Temperature Which Occurred at the Par-

ticular Location During Operation, and the Depth
of Corrosive Attack Found by Subsequent Metallo-
graphic Examination of the Samples.

236

785 (01

on the compression side of the coil with the
opposite inner wall on the tension side showed no
differences ‘in depth of attéck, The depths of
attack found on the straight sections of the fuel
tube are compared in Fig. 4.2.2. Photomicrographs
of samples from the nose coil and the straight
sections are presented in Figs. 4.2.3 through
4.2.7. No mass-transferred crystals were found on
any of the specimens, |

Disassembly of MTR in-pile loop No. 4 has been
started, but as yet no results are available. Opera-
tion of this loop was described previously.?2

In-pile loop No. 5, which was inserted in the
MTR but could not be filled, was shipped to ORNL
for disassembly after the section behind the pump
motor was removed at the MTR., During the attempt
to fill the loop the fuel-circulating pump seemed to
be chattering, some of the thermocouples were not
functioning properly, leakage occurred throudh the
pump bulkhead, and a fill-line Calrod heater failed
after it had been raised to higher than normal
operating temperatures.

 

2¢, C. Bolta et al., ANP Quar. Prog, Rep. June 10,
1956, ORNL-21064, p 75.

CONFIDENTIAL
ORNL-LR-DWG 16189

TO NOSE COIL

DISTANCE FROM PUMP (in.)

35 30 25 21t 7 12

s T FUEL LINE

- T =1 1 I
SAMPLENO. 309 308 307 306 305 304
SLIGHT ATTACK FROM SAMPLE NO. 304 TO 309. VARIED
FROM O TO 0.5 mil :

 

0
. =
: =2
FROM NOSE COIL o
DISTANCE FROM PUMP (in.)
42 37 3 26 22 a2 FUEL LINE °
] | | 1 i 1 !
1 I | | 1 i
SAMPLE NO. 284 ' 287 291 295 296 303

DEPTH OF ATTACK (mils)
3 2 i § i i

Fig. 4,2.2. Results of Metallographic Examina-
tions of Straight Sections of Fuel Tubing from In-
Pile LOOP Noo 3.

 

 
 

|
|

 

 

!
j
1

)
i
5
}
i
!
i
|
i
:

 

 

 

 

 

 

 

 

 

Fig. 4.2.3. Sample No. 264, Taken Near Inlet End of Nose Coil of MTR In-Pile Loop No. 3, Showing
Intergranular Corrosion to a Depth of 2 mils. 250X, {Cenfidemiai-with-ception
u Fig. 4.2.4. Sample No. 273, ‘Taken Near Outlet End of Nose Coil of MTR In-Pile Loop No. 3, Showing
Intergranular Corrosion to a Depth of 3 mils. 250X, “FSonfidentictwith option~
237
e %35 002

 

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

 

Fig. 4.2.5. Somple No. 284 Taken from Straight Section of Fuel Tubing Near Outlet of Nose Coil -
at a Point 42 in. from the Pump of MTR In-Pile Loop No. 3. Intergranular corrosion to a depth of 3 mils

may be seen. 250X. {Confrdentivtwittrraption)

 

Fig. 4.2.6. Sample No. 303 Taken from Straight Section of Fuel Tubing Between the Nose Coil Outlet

‘and the Pump at o Point 12 in. from the Pump of MTR In-Pile Loop No. 3. Intergranular corrosion to a

depth of 1 mil moy be seen. 250X. {(Confidentielwith-caption)-

238
 

 

 

"

PERIOD ENDING SEPTEMBER 10, 1956.

L

 

Fig. '4 2.7. Snmple No. 304 Taken from Straight Section of Fuel Tubing Between Pump and Nose
Coil Inlet at a Point 12 in. from the Pump of MTR In-Pile Locp No. 3. Intergranular corrosion to a depth

of 0.5 mil may be seen. 250X {Confidentiehwith-capton)

The inlet fuel line to the nose coil was found to
be full of fuel at a point 24.!/2 in, from the face of
the pump, and a shallow layer of fuel was found in
the outlet line from the nose coil at the same dis-
tance from the pump. The fuel in the inlet and
outlet lines at a point 2234 in. from the face of the
pump is shown in Fig. 4.2.8.

A radiograph of the fill tank, fill line, and vent

line, Fig. 4.2.9, showed the fuel in the fill line to

have frozen solid for the first 4/ in frorn the fill
tank, and, from fhut pomf on, there were - cavities
in the fuel. - The vent line appedred to be clear,
except for a few particles and a deposit close to
the pump. The fuel in the fill line at the point

of maximum - fuel porosafy, 1.6 in. from the pump, -
is’ shown in- Fig. 42,10,  The plug probably -
r_:occurred in the fill line ot this point, since the
fuel found here was dlfferent in appearance from
that. found closer to the fuel tank where no’ porosity
.. was observed.,The fuel m the flll tonk is shown in
- Fig. 42,11, The pump sump was' devoid of fuel.

Thirteen- fuel samples ‘were obtcnned for_chemical
analysis at various focations around the Ioop.

The fill-line Calrod heater failure was located
and is shown in Fig. 4.2.12. The pump chattering

slight, if any, corrosion occurred.

was probably caused by rubbing of a wiper ring
behind the slingers on the slinger shaft, Fig. 4.2.13.
All the thermocouples in the nose-coil region
appeared to be in operating condition. Thermo-
couple No. 28, which was at the rear of the fill
line, could not be found. The cause of the leakage
through the pump bulkhead could not be determined,
since the damage of the glass seals may have
~ occurred during disassembly,3

'INVES‘TI-GA'HON OF MATERIALS REMOVYED
"FROM MTR IN-PILE LOOPS NOS.3 AND 5

R. P, Shields

Chemical analyses of fuel and other materiais

. from MTR in-pile loop No. 3 have been made. A

~partially completed :study of the results has indi-
cated that, on the basis of the chemical analysis
“of the fuel for iron, chromium, and nickel, very
The amber-
colored material found in the pump region was
* probably an oil-decomposition product, This con-
- clusion is based on the fact that the material is

 

3¢, Ellis et al., Examination of ANP In<Pile Loop 5,
ORNL CF-56-7-48 (July 16, 1956).

239

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

 

 

 

-+ UNCLASSIFIED
. PHOTO w7917

Fig. 4.2.8. Section Through Straight Sections of Fuel Tubing Between Nose Coil and Pump at a Point
22?’/‘ in. from Pump of MTR In-Pile Loop No. 5. The small-diameter tube that is full of fuel is the inlet
line, and the small-diameter tube that has a shallow layer of fuel is the outlet line.

soluble in acetone, is waxlike in consistency,
and has a fairly broad melting point of 225 1 10°C,
Considerable deposits of Cs'37 and 537 were
found inside the copper lines leading to the fission-
gas-adsorber trap. Samples of carbon taken from
the trap showed no evidence of oil deposits and had
low radicactivity after they had been flushed with
heliums The fission products Cs'37 and S¢87
were present in the carbon in small amounts.

~Samples of fuel from MTR in-pile loop No. 5
were taken for analysis. The analysis of the fuel
may give some indication as fo why it was not
possible to fill this loop successfully, '

240

-3

Gl

W

CREEP AND STRESS-CORROSION TESTS
OF INCONEL

J. C. Wilson

N. Eo Hinkle
J. C. Zukas

A series of burst tests of Inconel tubing at a
stress of 2000 psi in helium has been essentially
completed out-of-pile and in the LITR at tempera-
tures of 1450, 1500, and 1550°F. The scatter of
the limited amount of data obtained at 1450 and
1550°F prevents any quantitative interpretations of
effects of irradiation at those temperatures. Ten
tests have been completed at 1500°F, however, and

W. W. Davis

v
2
i

¥
 

 

 

a

UNCLASSIFIED
. RMG-1427

 

 

 

 

Fig. 4.2.9. Radiograph of Fill Tank, Fill Line,

and Vent Line of MTR In-Pile Loop No. 5. {Cen=

‘two more tests are under way. Specimens irradigted

inHB-3 of the LITR broke at times varying from 113

to 560 hr, with the arithmetic mean time for rupture

being 260 hr. The out-of-pile test specimens
fractured at times varying from 350 to 1270 hr

-3
&

- W
D

PERIOD ENDING SEPTEMBER 10, 1956

(with two specimens still in test at 407 and 490 hr),
with a mean rupture time of at least 760 hr, Thus,

‘a decrease in rupture life of this thin-walled

(0.010~in.»wall, 0.191-in.-ID)} tubing appears to

have resulted from irradiation.

As was reported earlier,# the tubing stock used
gave many indications of defects during ultrasonic
nondestructive testing, but attempts to find the
indicated defects metallographically were seldom
successful, The presence of defects in the tubing
may have influenced the results of the tests in the
LITR, . Carefully inspected tubing stock of the
type to be used in ART NaK-to-air radiators has
been received for burst tests in helium and in fused
salt fuel mixtures, Additional temperature con-
trollers and pneumatic gaging equipment have been
received with which to expedite the tests.

Another test of the stress-corrosion apparatus
designed for operation in HB-3 of the LITR was
made. The fuel mixture (No. 46) NaF-ZrF ~UF,
(62,512,525 mole %) was used in an efforf to
cttain the desired maximum temperature of 1500°F,
but the highest temperature reached was only
1200°F. In order to assure a maximum temperature
of 1500°F during the next irradiation an improved
heater is to be used and the number of conductive
fins will be reduced. Improved instrumentation has
been built to record continuously the test variables
and to reduce the fission-gas hazard. Work con-
tinues on improved safety systems for the high-
pressure circuit in order to adapt the apparatus for
use in the MTR. Furnace arrangements for the

| ~ MTR irradiations are also being tested.’

Two ':fensile-c:r.eep b'ench' tests of Inconel bar
were made at 1500°F and a stress of 1500 psi in

" helium ‘that duplicated the tensile-creep test made
~in HB-3 of the MTR. Both bench specimens
~ fractured out of gage at 542 and 614 hr. Elonga-

tions in the bench tests were 1.0 and 1.9%, re-
spectively, Preliminary examination suggested
that local overheating outside the gage length had
caused premature fracture. The in-pile test showed
1.7% elongcmon ofter 760 hr. -

 

e

4w :W. ‘Davis, J. C. Zukus, and J. C. Wilson, ANP

' Quar. Prog, Rep. March 10, 1956, ORNL-2061, p 187.

5J. C. Wilson et al., ANP Quar, Prog. Rep. Sept. 10,
1955, 0RNL-1947. P 165.

241

oy

 
 

 

 

 

e

e bk

[—

ANP PROJECT PROGRESS REPORT

 

Fig. 4.2.10. Sectioned Fill Line, 3 in. Long, from Point ot Which Maximum Fuel Porosity Is Shown
in Fig. 4.2.9, 1.6 in. frem Pump. 2.4X. Reduced 10%. i i f i

 

Fig. 4.2.11. Cap from Fill Tenk of MTR In-Pile Loop No. 5§ with Fuel in lt. The hole in the cap
is the vent line entrance. (Genfidential with-eaption)—

242

~3
G
&
)
¢
-3
 

 

 

 

 

o

3

 

Fié. 4.2.12, Fuyel Fill-Line Calrod Heater That
Failed During Attempt to Fill MTR In-Pile Loop
No. 5. (Confidentiel-with-caption)*

 

:PERIOD ENDING SEPTEMBER 10, 1956

MTAR:ASTRESS-CORROSION APPARATUS

Jo CQ Wilson

C. D. Baumann W. E. Brundage
'Final assembly of equipment for MTR stress-
corrosion experiments has been held up by a delay
in shipment of the proper high-temperature, high-
pressure hermetic seals. Molybdenum wire heaters
for the apparatus have been successfully fabricated

‘inside o tubular ceramic form. The fuel mixture

NaF-ZrF ,-UF ,, containing sufficient uranium to
give the heating desired, has been received and
loaded into the Inconel specimen tube.

EFFECi’ OF RADIATION ON STATIC CORROSION

OF STRUCTURAL MATERIALS BY FUSED
SALT FUELS

Ho L- Hemphlll
The study of the effect of radiation on the cot-

W. E. Browning

rosion of structural materials by fused salt fuels

has been continued. Two static Inconel capsules
containing the fuel mixture (No, 44) NaF-ZrF -UF
(53.5-40-6.5 mole %) were irradiated in the MTR
for over nine weeks at approximately 6 kw/cm3,
The capsules were at a temperature of 1570°F
during the irradiation. Near the end of the test

period, a thermocouple failure led to a temperature

UNCLASSIFIED
- RMG-1437

 

INCHES

Fig. 4.2.13. Abraded ‘Region on Pymp Slinger Shaft That Was Caused by Rubbing of a Wiper Ring
Behind the Slingers During Attempt to Fill MTR In-Pile Loop No. 5. (Confidential-withreuprion)—~

=3
G
&
<
&
oo

243

 
 

 

 

ANP PROJECT PROGRESS REPORT

excursion, os. yet unevaluated, and one capsule
released fission-product activity inte the cooling
air off-gas. line. By increasing the flow of cooling-
air sufficiently to freeze the fuel, the radiation
level at the top of the reactor due to the off-gas
line .activity was reduced from 200 r/hr to 100
mr/hr  without interruption of reactor operation,
The radicactivity was confined to the offsgas
system and there was no contamination of the air
of the building., The irradiation of the other cap-
sule was completed without incident, A second
pair of Inconel capsules containing the same fuel
mixture (No. 44) is being irradiated at 1500°F, and
the planned exposure period is again nine weeks.
Five Inconel capsules containing NaF-ZrF -UF
fuel mixtures that were previously irradiated have
been opened, and samples have been submitted
for chemical analysis and metallographic examina-
tion, Hastelloy B capsules have been filled with

an NoF-KF-LiF-UF, fuel mixture containing
5.8 mole % UF ; and are being prepared for bench
tests and MTR irradiation. These capsiles are
nickel-chromium plated on the outside surface to
provide protection from atmospheric oxidation,
Some .improvements in design of the MTR-ORNL-
2B irradiation facility have been accomplished.
The capsule, shown in Fig. 4.2.14, is prepared in
the same way as before; the test capsule is the
part in the center of the photograph with the ther-
mocouples attached. The separate part to the left
is the sleeve which slips over the capsule and
defines the cooling-air annulus. The three small
flattened tubes carry the thermocouple wires into
the tube at the right. The complete length of the
tube at the right can be seen in Fig. 4.2.15 with
a compression spring on the end. The thrust of
the spring keeps the air annulus sleeve in place in
a tapered seat in the outer tube shown at the top

UNCLASSIFIED
PHOTO 17339

Il'l'l'l"l!'lllll'l!l'l‘ll!l!lul

 

 

 

Fig. 4.2.14. Inconel Capsule and Associated Equipment for MTR lrradiation.

UNCLASSIFIED
PHOTO 17340

 

 

 

 

 

244

=3

)

ot
(

i
&
 

 

-

_ leads.

of Fig. 4.2.15. The outer tube has two parts, a
5.8-in,«OD tube for the capsule and a 3.8-|n.-OD
gir-return tube. Both parts are welded to / in.-OD
air tubes which extend to the top of the reactor
tank; the air-in tube carries the thermocouple
Each capsule is jaocketed individually in
this way. This assembly can be irradiated either
in the B.4 facility or in one of the standard A
positions in the MTR.

HOLDUP OF FISSION GASES BY
CHARCOAL TRAPS
| D. E. Guss®
A series of charcoal traps of different geometries
but with the same amount of charcoal are being

W. E. Browning

tested in radiokrypton holdup experiments, Nitro-

gen purge gas is used for these tests at various

‘temperatures below room temperature. The results
of these tests will aid in extending the relation-
‘ships reported earlier’ to the evaluation of traps

of d lfferenf shapes.

ART OFF-GAS SYSTEM INSTRUMENTATION
| ANALYSIS
W. E. Browning R. P. Shields

Preparations are being made for invésfig’btihg

the resolution, sensitivity, and general performance
of scintillation gamma-ray spectrometers of the

type to be used on the ART off-gas system. The
purpose of the tests will be fo determine the
limitations which might result from the rather

" drastic departure which must be made from con-

ventional spectrometer geometry because of the
high intensity of the source.

LITR VERTICAL IN-PILE LOOP
W. E. Browning

D. E. Guss H. E. Robertson’
M. F. Osborne R. P, Shields

‘The cause of the failure of the pump in the LITR

been determined. A screw came loose and lodged

in the impeller and thus threw the impeller out of
“balance ‘enough to bend the shaoft., The bent shaft

rubbed .against fhe ‘impef]._iéf housing ond stalled the

 

80n ussrgnment from the United States Air Force.
7w, E.. ‘Browning 'and C. C. Bolta, ANP Quar, Prog.
Rep. March 10, 1956, ORNL-2061, p 193.

8w. E. Browning et al.,, ANP Quar, Prog. Rep, June 10,
1956, ORNL-2106, p 238.

=3
&1

PERIOD ENDING SEPTEMBER 10, 1956

motor. The screw served as a shear pin to prevent
the impeller from unscrewing from the shaft, The
shaft was peened around the screw to hold it in
place, It is not known why the peening failed to
hold the screw. This method of fastening various
internal parts will be replaced by welding in future
loop fabrication,

Tests have been continued on a modified pump
for use in future loops. The modified motor and
pump assemblies are shown in Fig. 4.2.16. The
modifications include precision bearings, a shorter
and stiffer shaft, a greater clearance around the
impeller, and an enlarged pump exhaust port, An
improved radiation-resistant grease has become

available for use in future operation.

High-temperature tests of the new shaft have led
to a design in which the highest bearing-part tem-
perature will be 105°C, even when the temperature
of the shaft where it enters the fuel is as high as
950°C. These temperatures are for a stationary
shaft; the bearing temperature for a rotating shaft
will be lower because of heat transfer by the
rolling balls, The over-all shaft overhang is only
4 in., of which ]3/ in. is in the fuel, Longitudinal

- conductive heat tronsfer is suppressed by use of a

. clearance

hollow shaft and the four rows of radial holes

visible in Fig. 4.2.16a. Radial heat transfer from
the shaft to the lower bearing is decreased by a
set of longitudinal grooves which reduce the area
of contact between the shaft and the lower bearing.
This distributes the heat flux more equally between
the two bearings and thus reduces the temperature
of the inner race of the lower bearing.

Bearing tests have been run in helium under
simulated conditions of temperature and heat flux
with both the new and old radiation-resistant

lubricants, The satisfactory performance obtained

under these conditions indicates that the bearing
is sufficient for the design thermal
gradient through the bearing and that the bearing

_ _ R ' temperatures are not so high that the grease will
- vertical in-pile loop, previously described,® has -

be damaged in the absence of radiation.

be “made on the - impeller clearance.

2

In the design of the pump o compromise had to
A large
clearance was desired to ease the shaft alignment

‘problem, but the clearance must not be so large

as to allow leakage that would greatly reduce pump
efficiency. Performance tests have been run to
optimize this design choice. Impeller clearances
have been increased from a range of 0.010 to 0.020
in. at various positions to 0.050 to 0.070 in. This

245

()
<=

 
ANP PROJECT PROGRESS REPORT

 

 

 

  
  

 

 

UNCLASSIFIED
PHOTO WIM °
¥
¥
®
-
'-.
-
- UNCLASSIMISD
MTO 15237
i
=
i -
i
i
i
|
; o
»
UNCLALSIFIED
PHOTO W33
-
>
-
;
(e} -
-
-

Flg _4.2.16.7 Modified Motor and Pump for Use in LITR Vertical In-Pile Loop. (a) Impeller and
‘shaft components. (b) Pump assemblies. (c) Motor parts. - :

246

s

¥

“j
L

~3
)
o

 
 

 

 

.

~ance tests were run with water in a plastic model . T 7/

- at the design point, so the pump is still operated at
‘about the same speed,

_were converted to valves for fused salt fuel by

' ‘UF (63-25-12 mole %). A plot of the estlmdted

PERIOD ENDING SEPTEMBER 10, 1956

increase caused- @ measurable but unimportant

e :
. S - ORNL—LR-DWG 16191
decrease in pump efficiency. The pump perform- 9000 -

 

 

 

of the pump. The test results are presented in ' = 7
. . - | | /
Fig. 4.2.17, which shows the pump pressure as a |

function of water flow before and after the increase 5000 |— _
in clearance for various pump speeds. The resist- r'/

ance curve for the test loop is also shown. Both
port than was formerly used. The results of the | /,/
2000

 

 

 

 

sets of curves are for a pump having a larger exit
two changes are approximately equal and opposite

 

- REYNOLDS NUMBER

“The pressure vs flow data obtamed with water -

 

 

 

 

 

 

 

 

 

 

using recently obtained values for the viscosity 1000 L

and density of the fuel mixture (No. 41) NaF-ZrF - .~ 1000 2000 - 5000 10,000
. PUMP SPEED (rpm)

 

Reynol&s number for the fuel in the LITR vertical _ | |
in-pile loop as a function of pump speed -is pre- Fig. 4.2.18. Estimated Reynolds Number of the

sented in Fig. 4.2.18. L : Fuel Mixture (No. 41) NaoF-ZrF «UF, (63-25-12

A full-scale Inconel pump -has been ussembled mole %) in LITR Vertical In-Pile Loop as a Func-
that is identical with the pumps to be used in the tion of Pump Speed

UNCLASSIFIED
ORNL-LR-DWG 16190

———0.050TO 0.07Qin. IMPELLER CLEARANCE
0.010 TO 0020 in. IMPELLER CLEARANCE

‘if HEAD {psi) |

- LOOP RESISTANCE ~

N
- . B i ‘“ .
——

 

o e TRl . 03 - 04 - 08 - 06 or
o : '*VFLOW(gprn) C ‘ ) ' '

Fig.42 17. Performcnce Curves with Wcter for Modlfied Pump for LITR Vemcul In-Pile Loop
Before and After |ncreasing Impeller Clearances.

247

 
 

 

 

"ANP PROJECT PROGRESS REPORT

Inconel in-pile loops. It will be filled with de-
pleted fuel and run in the laboratory under con-
ditions as nearly like those which will exist in
the reactor aos possible. The existing parts for
in-pile loops are being modified to incorporate
these recent changes. B

DESIGN CALCULATIONS FOR LITR
VERTICAL IN-PILE LOOP

M. T. Robinson J. F. | Krause?

An extensive series of design calculations for
the LITR vertical in-pile loop was previously re-
ported.'® [n an appendix to that report, a method
was presented for making such calculations by
_electronic analog simulation.

ential equations describing the heat balance in the
loop, followed by analog computation, supported by

. some desk-machine computation. The correct solu-

tions were then found :by'ir_lterpolation among a
femily of analog computor results. A much-
improved method of analog computation has now

been devised which eliminates the interpolation -

procedure and the desk-machine work, as well as
substantially shortens the amount of analog com-
putation required to obtain correct solutions.

The new circuit has been used to make calcu-
lations of the behavior of the Mark VIl Joop!®
when circulating the fuel mixture (No. 41) NaF-
ZrF‘-UF4 (63-25-12 mole %) in position C-46 of
the LITR. The flux data previously reported!! for
that position were used. The physical property

 

90n loan from Pratt & Whitney.

10M. T. Robinson and D. F. Weekes, Design Calcula-
tions for a Miniature High-Temperature In-Pile Circulat-
ing Fuel Loop, ORNL-1808 ( pt: 19, 1955) with **An
Appendix on Analog Simulation’ by £ R. Mann, F. P.
Greene, and R, S. Stone.

Um, 1. Robinson, Solid State Semiann. Prog. Rep.
Feb, 28, 1954, ORNL-1677, p 27.

~ tip at the mid-plane of the LITR.
“assumed to be depressed 50% by the presence of

_ The method was
based on partial analytical solution of the differ-

data used for air were the same as those used for
the previous calculations.1? The physical property
data for the fuel mixture are given below: |
Viscosity
Specific heat
Thermal conductivity .
Density

4,3 centipoises
1.0 joule/g°C
0.14 w/em°C
3.38 g/cm’

The loop was considered to be located with its
The flux was

the loop. The initial air temperature, To, was
always taken as 30°C. A Reynolds number of
4000 was assumed for the fuel, '

Some results of the calculations are summarized
in Tables 4.2.1 and 4.2.2 and in Figs. 4.2.19 and
4.2.20. = Typical profiles of the mixed-mean fuel
temperature, T,, the fuel-metal interface tempera-
ture, T,, and metal-air interface temperature, T,

‘are shown in Fig. 4.2.19, and profiles of the air
~ temperature, T, and of the heat transfer function,

y, are shown in Fig, 4.2.20. The discontinuities
that occur in the plots of T,, Ty T, end y result
from the use of two separate cooling-air streams,
Local heat flow would be expected to remove these
discontinuities in the actual loop. The very dif-
ferent shapes of T, and of T, and T, are notable.
This feature is important in analyzing experimental
data, since T, is the only quantity for which a
direct measurement can be obtained. Data on the
effect of the initial fuel temperature (T9) on other
quantities are presented in Table 4,2,1. As was
realized before, 19 AT, is very little affected by
changes in T?. It should be noted, however, that "
appreciable changes in cooling-air requirements
are associated with changes in T9 Table 4.2.2
itlustrates the effect of fuel flow rate on other
quantities, The important points to notice are the
close similarity of the values AT, and AT, ex-
cept at the lowest fuel Reynolds numbers; the

TABLE 4.2.1. EFFECT OF INITIAL FUEL TEMPERATURE ON COOLING-AIR REQUIREMENTS

 

 

 

 

. Air
T}’ °c) Reynolds 73 (°0) T3 (°C) AT, (C) AT, (°Q) AT, °C)
Number
750 34,500 659 641 87 81 83
800 31,500 708 690 88 83 85
850 29,750 756 733 89 85 8
248
w3 i
 

 

 

w

-

PERIOD ENDING SEPTEMBER 10, 1956

TABLE 4.2.2. EFFECTS OF FUEL FLOW RATE ON COOLING-AIR REQUIREMENTS FOR AN -
INITIAL FUEL TEMPERATURE, T?, OF 800°C

 

~ Fuel - Air

 

- Reynolds Reyfio_l_ds . T‘,’(°c') | _ | Tg (°c) ATI (°C) AT, (°0 = A1, {°O
7  MNumber | Number '
2,000 45000 477 450 163 134 139
13,000 .~ 34000 650 632 14 04 106
4000 31,500 708 690 88 83 85
6000 29,50  7:1 732 59 57 58
8000 - 29,000 764 74 M 43 45

10,000 29,000 . - 770 - 752 - 3 35 36

 

:860
- aeo
" g20 |

© 860

. “TEMPERATURE {*C)

" 740
L re0

100 b

SR
ORNL-LR-DWG 15244

' FUEL COMPOSITION: NGF-Zrf,-UF, (63-25-12 mole %)
" FUEL REYNOLDS NUMBER: 4000 - ‘ :
_ AIR REYNOLDS NUMBER: 31,500 ,.

FLUX: 50% OF LITR FLUX

LITR POSITION: C-46

MIXED~- MEAN FUEL TEMPERATURE, %

FUEL ~METAL INTERFACE

o~
o
Q

SN
e
Lo

w
2
g
=
i

- AIR INTERFACE

 

680 Lt

2077 2 - 40 .. 80. . _ 80 100 120 t40 . . 180 180 200
e R T T L . DISTANCE FROM FUEL ENTRANCE (cm)

- Fig42.l9 ACclcfi'luie.d Tempéraiure Pr-ofilersu for LITR Vertical In-Pile Loop (Mark Vill).

249

=%

)

et
3
b
h":fi

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

ORNL-LR-DWG 15245

73

FUEL COMPQSITION: NoF-ZrF, -UF, (63-25-12 mote %)

FUEL REYNOLDS NUMBER: 4000
AIR REYNOLDS NUMBER: 31,500
FLUX: 50 % OF LITR FLUX

LITR POSITION: C-46

500

400

300

TEMPERATURE (°C)

t00

0 20 40 60 80 100

 

69

HEAT-TRANSFER FUNCTION, y 1 s

n
-

o
o~
RATE OF HEAT TRANSFER (w/cm)

53

L
@

COOLING-AIR TEMPERATURE, 7,

45

4

37
120 140 160 180 200 .

DISTANCE FROM FUEL ENTRANCE (cm)

Fig. 4.2.20. Calculated Air-Temperature and Hect-Transfer-Function Profiles for LITR Verticel

In-Pile Loop (Mark VIII),

small difference between T? and To, and the large
difference between T® and T W especmlly at lower
flow rates. The first point suggests that experi-
mental values of AT, may be used without correc-
tion to characterize filel “AT". The second point
results from the fairly small heat extraction rate,
around 30 w/cm? or less. The third point is most
important in assessing corrosion results.

ART RE‘ACTIVITY TRANSIENTS ASSOCIATED
- WITH FLUCTUATIONS IN XENON-REMOVAL
EFFICIENCY

M. T. Robinson J. F. Krause

The possibility has been investigated that fluc-
tuations in efficiency of the ART xenon-removal
 equipment might lead to troublesome transients in
reactivity. The calculations previously reported 12

250

3}
)
&

A

. -;

have been extended for this purpose. Since only
short times are of interest here, the equations
may be simplified to

(1) P = May -,
(2) y = fi)tl(x —azy)—Agyt—Bf -Agy '

where _
x = Xe'35 poisoning in the fuel,

y = *‘‘equivalent poisoning’’ in the gas phase,

= RTS, the product of the universal gas con-
stant, R, times the absolute temperatute,
T, times the solubility coefficient of xenon
in the fuel, S, ‘

2y T Rob:nson,' A Tf.veorettcal Study of Xel33

Poisoning Kinetics in Fluid-fueled, Gas<sparged Nu-
clear Reactors, ORNL-1924 (Feb. 6, 1956).

®q

 

prh
91 ]
 

 

 

 

- .

B = ratio of volume of fuel to vqlumé of'gas
~ phase, '

)l/ = rate constant for transfer of xenon from
fuel to gas,

)Lg = rate constant for removal of xenon from
system by off-gas flow,

Equations 1 and 2 were solved with an electronic
analog computer by using the block diagram shown
in Fig. 4.2.21, The output of amplifier 1 is a

PERIOD ENDING SEPTEMBER 10, 1956

vonl‘tagbe proportional to x. This voltage is inte-
grated in amplifier 2, the initial condition on the
integration being set by the potentiometer ‘‘0.2x "',
A voltage proportional to x is fed to the recorder
and to one input of the summing amplifier 4. The
output of amplifier 1 is also fed through the
potentiometer ‘‘a,B'' to one input of the summer-
integrator, amplifier 3. The output of amplifier 3
is fed back through the potentiometer ‘A *' to @

second input of amplifier 3. Thi‘s umpiiffer thus

UNCLASSIFIED
ORNL-LR-DWG 15241

 

1 RECORDER

 

 

~ +400V

 

 

Q.2 as yO

 

 

 

 

 

{
1| 4

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Capy X

 

. Fig;l 4.221 Block chgrum of cln V E'eétrbl;ic Ali&lfig 'rsri'm'hilcto'r for the Eqaafidhs':'c"= t\f(azy—x)

and y = Bhl(x-azy) -A e

89

. ‘u‘&
)

251

 
 

- the previous report, 12

 

 

ANP PROJECT PROGRESS REPORT

performs the operation
@ B% + aAy o= ~a,y

and then integrates y, subject to the initial con-
dition. set by the potentiometer *‘0.2 ", A
voltage proportional to a.y is fed from amplifier 3
to the second input of amplifier 4. The output of
amplifier 4, proportional to (a,y - x), is fed
through ‘the potentiometer ‘A’ to the input of
- amplifier 1. In analyzing Fig. 4.2.21, it should be

- noted that, in addition to performing the appropriate

mathematical operation, each amplifier inverts the
phase of the input voltage.

Scale factors for the computation were selected
such that all constants had their true values. The
scale chosen for x and y was such that 1% poison-
ing was represented by 20 v. The recorder could
~ be used with either 20 v (1%) or 100 v (5%) full-
scale settings. Real time was used throughout the
computation, The data employed were taken from
12 gybsequent modifications in
ART design being considered more or less irrele-
vant. It was assumed that the reactor had reached
xenon equilibrium under conditions specified by
selected values of A, and A , which allowed cal-
culation of x, and y, from the steady-state rela-
tions given before.12 At the start of a calculation,
either A, or A, wos assumed to change very
rapidly to a new value, and the change in x was
plotted during an interval of time up to 2 min. The
results obtained are collected in Table 4.2.3. For
each case the initial slope, —(dx/dt),, ond the
initial period, x /(dx/dt), are given, Some typical
results are plotted in Fig. 4.2.22.

Examination of Table 4.2.3 shows that the
initial period is essentially independent of A_ and
of changes in A , however large. The period de-
creases with increasing A/, whether the disturbance
is a change in A, or in Age In no case is the period
sufficiently short to cause concern for the con-
trollability of the ART. The compensating effects
of the delayed neutrons and of the negative tem-
perature coefficient of the reactor are sufficient to
overcome fluctuating behavior of xenon-removal
equipment,

USE OF Zr?5 AS A FISSION MONITOR
| W. A, Brooksbank, Jr.

"It has been found that the use of Cs'37 as a |

monitor for Uzs‘5 fission in fluoride fuels would
not be entirely satisfactory because of possible

252

¥
@0

h

' SECTET
. ORNL—LR-DWG 15242
0.8

o7

o
m .

{GAS FLOW RATE = 1135 STP liters/day)
Ag = 0.005 sec™!
As £ < 0) = 0.001 sec™

Xe'* POISONING (%)
o
B

o
w

02

0.1

= ¢ ggc™}

 

O 5 10 45 20 25 30 35 40 45 50 55 60 65
" TIME (sec)

Fig. 4.2.22. Change in Xe '35 Poisoning in ART

Following a Change in Phase-Transfer Rate Con-
stant, .

loss of its parent, 3.9-min Xe'37, As a substitute,
65-day Zr?5 has been suggested. In order to cali-
brate the use of this isotope, two samples of en-
riched uranium were irradiated, as solutions, for
2.375 days in hole 10 of the ORNL Graphite Reac-
tor, One solution was prepared from U, 0, the
other from fHluoride fuel composition. After allowmg
the samples to decay for ten days, radiochemical
determinations were made of Cs'%7 and of Zr95,
From these results, the fission yield of Zr?5

calculated to be 0,0664 + 0,0013 atom/fission,
If the fission yield is calcuvlated from the radio-

chemical Zr?3 analyses and the cobalt activation

neutron flux observed from a monitor irradiated with

the samples, the result obtained is 0.0632 £ 0,002]

atom/fission, The discrepancy between the two

results is removed if the cobalt activation flux is
corrected for the cadmium ratio prevailing in hole
10 (about 30), The fission yield recommended for
use in fission monitoring with Zr95 is 0.0664 *
0.0013 atom/fission.

FAST-NEUTRON DETECTORS
D. Binder J. F. Krause

An investigation is being made of methods for

measuring the fast neutron flux of reactors (> 500 ev)

bish
ok
 

 

-

effects with- energy.
fabricated to measure the fast flux-by the method
of Hurst et al.1® The shield will be used-in con- "
junction - ‘with Pu23%, Np237 “and U"'38 fonis In
‘order to compensate for the htgh f:ssion cross sece
-at low energles & 1.5cm path of
B0 1o lhe foil was required. To check for leckage, o

PERIOD ENDING SEPTEMBER 10, 1956

TABLE 4.2.3. INITIAL INCREASES IN ART REACTIVITY FOLLOWING
FLUCTUATIONS IN XENON-REMOVAL EFFICIENCY

 

 

 

 

%) Al(sec“) )‘g(sec-l)* Initial Slope Initial Period
o7 t<0 >0 <0 t>0 (%/sec)  ({sec)
3.98 1 1 0 0.025 1.44 3
0.5 0.5 0 0.025 0.54 3
0.1 S0 0 0.025 0.35 n
0.05 0.05 0 0.025 019 21
0.01 0.01 0 0.025 0.037 106
0.005 0.005 0 0.025 0.019 206
0.74 0,001 0.01  0.025 0.025 0.0069 106
0.001 0.1 0.025 0.025 0.063 12
0.001 R 0.025 0.025 0.27 3
0.75 0.001 0.01 0.015 0.015 ~0,0072 105
| ~0.001 0.1 ~0.015 0.015 0.067 1
0001 1 0.015 0,015 0.27 3
0.82 0.001 0.01 0.005 0.005 0.0075 109
10,001 0.1 0.005 0.005 0.071 12
0.001 B 0.005 0.005 0.27 3
0,97 0,001 0.01 0.002 0.002 0.0095 102
S 0001 01 0.002 0.002 0.080 12
0,001 1 - 0.002 0.002 0.26 4
3.98 0.001 0.01 0 0.038 104
0.001 0.1 0 | 0.35 | n

o001 1

 

| "und for determmmg 1he varaohon of Jrradwhon‘f

A boron  shield has™ ‘been -

~ of fl'se boron-covered to cudmlum-covered achvmes

tion of Pu239

various resonance detectors were irradiated inside

 

136, s. Hurst et al., Rev. Sci. Instr, 27, 153 (1956).

| - *i..The' ih;r_élhg.jas‘:f[o.fi rqte isu g(STP f_iieg\i-rs'/da)-f') !f-,_2427_ X r'_IO5 Ag(sec"‘);

- the Bm shield and- msude cadmlum. The ratios

for cobalt and manganese detectors were 60 and
20, respechvely. For no shielding at the resonance

fenergy, o ratio of cbout 2 would have been ob-

erved. - Therefore, 135~ and 340-ev neutrons

. are sirongly absorbed by the shield.

 

¥4p, J. Hu es, Pile Neutron Research, p 138, Addi-
son-Wesley, Cambridge, Mass., 1953,

253

018

 
 

 

 

ANP PROJECT PROGRESS REPORT

A photoneutron source was used as part of the
study of the variation of displacements with neutron
energy before the exact effectiveness of gamma

rays in displacing atoms was learned. Depending

on the energies, a neutron is 100 to 1000 times as
effective as @ gamma ray. This is a sufficient
ratio for reactor or accelerator irradiations but
not for photoneutron sources. The antimony-
beryllium source which was used produced de-
tectable effects in n-type germanium with initial
concentrations of 10'3 donors/em®, but the ratio
of gamma rays fo neutrons was on the order of
105, which made the gamma rays about 100 times
as effective as the neutrons.

EFFECTS OF RADIATION ON ELECTRONIC
COMPONENTS

J. C. Pigg C. C. Robinson!’

In the previous report!® the behavior of the
characteristic curves of transistor and semicon-
ductor diodes was compared. The emitter charac-
teristic of the tronsistor was compared with the
forward characteristic of the diode., The collector
characteristic of the transistor was compared with
the reverse characteristic of the diode. Correla-
tions were drawn between Co%? gamma irradiation
and reactor neutron effects on the characteristics
of these devices. It was noted, however, that the
barrier change resulting from Co%% gamma irradic-
tion showed properties not noted previously with
reactor irradiation,

Electrical annealing, that is, removal of charged
interstitial atoms by the strong electric field
present in the barrier, was noted. It was pointed
out that because of this effect, the bias applied
to a semiconductor component might affect the
change observed when the component was operating
in @ radiation field.

Exposures of Philco surface-barrier transistors
have been made in the ORNL Graphite Reactor,
but, thus far, attempts to measure the amplification
characteristics of these surface-barrier fransistors
under irradiation have been unsuccessful. How-
ever, the photo-emfs developed across the collector
and emitter barriers during reactor irradiation have
been measured, and ¢ series of before and after
measurements . of collector characteristics have

 

 

150, assignment from Wright Air Development Center.
6Jn c. Pl dnd C. c. Roblnsofl. ANP Quafo Prog.

Rep. June 10. 1956. ORNL-2104, p 243.

254

-3
G

been made. Since a sharp barrier is desirable for
an emitter and a diffuse barrier is desirable for a
collector, it would be expected that there might be
some difference in the behavior of these barriers
in ¢ radiation field. The photo-emfs developed
across such barriers during irradiation in hole 5IN
are shown as a function of integrated dose in
Fig. 4.2.23. There is a considerable difference in
the photo-emfs observed until after the integrated
dose has reached about 1014 nvt, at whlch time

‘the barriers become comparable.

Wil

The change in collector characteristic with the
emitter circuit open, as observed after successive
irradiations, is shown in Fig. 4.2.24. As reported
previously, this behavior compares with the char-
acteristic of a diode biased in the reverse direction.
After the third exposure, the sample was allowed to
anneal for 18 hr at room temperature. The charac-
teristic taken after the 18-hr annea! showed that
the saturation current dropped back toward its
original value and that the leakage current, as
indicated by the slope of the curve, increased.

The data agree with previous observations of
annealing due to the electrostatic field of the
barrier in that the saturation current returns toward
its initial value, The change in slope or leakage
current can be considered to be due to the same
mechanism. Since the base material in this tran-
sistor is n-type and the surface of all germanium
crystals is p-type, there is a p-n junction between
the bulk material and its surface. In the case of a
p-n=p junction device, a p-type channe! occurs
along the surface through which current may flow
without crossing a barrier, as illustrated in Fig.
4.2.25. It would be expected that, as a pn-p
device was irradiated and the n-type material
changed toward p-type, the p-type surfoce layer
would become thicker and more p-type. Both
results would lower the resistance of this channel
and lower the leakage resistance. It would be
expected that an #n-p-n device, however, would
require the current to cross a barrier in going from
one n-type region to another, and thus no channel
would be present. Consequently a p-n-p unit
should be more sensitive to surface conditions
than would an n-p-n unit, )

The change in the grounded-base characterlshc
family of curves after an integrated dose of 9.96 x
1012 nvt, is shown in Fig. 4.2.26. Instecd of an
increase in the collector characteristic, there is a
decrease, and there is no detectable change in

019

A
- gt

PERIOD ENDING SEPTEMBER 10, 1956

UNCLASSIFIED
ORNL-LR-DWG 45064

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

8
s
¢ \
P . N, -
7 : 2 2 2 2 e
A — =—"-—=—.=-.—-_______
N o
6
5
- ® EMITTER - BASE
- A COLLECTOR - BASE
&4
|
o
-
o
a
3
2
1
0
0 0.5 1.0 1.5 2.0 2.5 30 3.5 4.0 a5 5.0

Fig. 4.2.23., Photo-emfs
Integrated Neutron Flux.

INTEGRATED NEUTRON FLUX (10" nyt)

UNCLASSIFIED

NUMBERS INDICATE EXPOSURE

R \RRADATIONS

3 = AFTE

COLLECTOR CURR T ma)

ORNL-LR-DWG 152914

HOURS OF ANNEALING

TER THIRD IRRADIATION

-2 _ AETER 1 IRRADATION

IRR
o L .

0 02 04 06 08 . {0
~ COLLECTOR BIAS (voits) -

 

1.2 14

Fig. 4.2.24. C-hungeé‘ in Collector Current Char-
acteristic of @ Philco Type-L5106 Surface-Barrier
Transistor as a Result of a Series of 2-min Ex-

posures at a Flux of 8.3 x 10! nv .

0

&
&

e

of a Philco Type-L.5104 Surface-Barrier Transistor as a Function of

UNCLASSIFIED
ORNL ~LR~DWG 15290

 

 

 

 

 

 

 

  

 

 

 

 

 

QL Lttt L ORI RN NN Y]

 

Fig. 4.2.25. [llustration of Surface Leakage
Mechanism, ‘ '

slope. The separation of the curves for different
emitter currents is about the same as it was before
bombardment, A circuit designed with normal
engineering tolerances would still be operating
providing there ‘had been no violent fluctuations of
behavior during the bombardment, The smoothness
of the transition is illustrated in Fig. 4.2.27, which
shows the collector characteristic at 1-v bias at

255

 
 

 

 

ANP PROJECT PROGRESS REPORT

UNCLASSIFIED
ORNL-LR-DWG 15292

/g™ 5.85 ma

w

H

L]

BEFORE IRRADIATION
<=—====AFTER IRRADIATION

. COLLECTOR CURRENT {(ma)
™

 

0 0.2 04 0.6 0.8 10 1.2 1.4
‘ COLLECTOR BIAS (volts)

Fig. 4.2.26. Change in Collector Characteristic
of a Philco Type-L5106 Surface-Barrier Transistor
as a Result of Irradiation to an Integrated Dose of
9.96 x 1012 nvt ,

different emitter currents as a function of the inte-
grated dose. These curves are the results of
measurements before and after irradiation, however,
and may not be indicative of behavior in the
presence of the radiation field.

IRRADIATION EFFECTS ON THERMAL.-
NEUTRON SHIELD MATERIALS

J. G, Mergan
R. M. Carroll M. T. Morgan
P. E. Reagan

The thermal-neutron shield for the ART should
have a high absorption cross section, it should be
compatible with adjacent reactor materials, and it
should retain its bulk structure at elevated tem-

peratures under irradiation. Gas release from the

material should be at @ minimum. For temperatures
up to 1600°F, two cermet shield materials, COB‘-
Fe and BN-Ni, clad with stainless steel are being
tested. These were fabricated!? by mixing the
powdered components for 2 hr in an offset rotory
blender, cold pressing the mixed powders at 33 tsi

 

175, H. Coobs, private communication to J. G. Morgan.

256

ok

&>

UNCLASSIFIED

-3
()é 107°) ORNL-LR-DWG 15063

TRANSISTOR: PHILCO TYPE L-510C
3 NO. A-1944

BIAS: { volt

ROOM TEMPERATURE

/p, COLLECTOR CURRENT (ma)

 

0 14
0 1 2 3 4 5 6 (x10')

nvt, INTEGRATED NEUTRON FLUX _(neufrons/cmZ)

'Fig. 4.2.27. Changes in the Collector Current
of a Philco Type-L5106 Surface-Barrier Transistor

as a Function of Irradiation.

ina2x 2'/4-in. steel die, sintering the compacts
for 1 hr at 2000°F in o hydrogen atmosphere,
coining the compacts at 33 tsi to 0,250 in. thick-
ness in a 2 x 2'/4-in. steel die, encapsulating the
coined compacts in type 304 stainless steel
picture frames, hot rolling the composites ot
2000°F to a total reduction in thickness of 7:1,
and machining 0.1875 x 0.500 x 0.050-in.-thick
specimens from the rolled composite sheet. The
powder particle sizes and the compositions of the
cermet core materials are given in Toble 4.2.4,
A CaB,-Fe sample was irradiated in helium in the
LITR to a thermal-neutron dosage of 5.2 x 1017 nvt,
The sample, sealed in quartz, was cocled by the
LITR water. Metallographic examination!® of the
specimens after irradiation (Fig. 4.2.28) showed
that the clad-core interface had retained its con-
tinvity, The core material was clad to ensure
compatibility with the Inconel to be used as the
structural material of the reactor, Many CaB

crystals may be seen in the core matrix, ulihougtn
some were removed in sample preparation. The
samples shown in Fig. 4.2.28 are shown again in-
Fig. 4.2.29 at a lower magnification so that the
cladding on both faces of the sample may be seen.

 

IaA. E. Richt, private communication to J. G. Morgan.

it
o
o
b
 

 

 

n

0

°"

PERIOD ENDING SEPTEMBER 10, 1956

TABLE 4.2.4. PARTICLE SIZES AND COMPOSITIONS OF CERMET SHIELD MATERIALS

 

L Particle Size
Cermet

Weight of Component Volume of Component

 

Component
P ®) (%) (%)
CaBj-Fe CaB, | 44 to 100 7.6 21.0
Fe 147 92.4 79.0
BN-Ni- BN 10 10.3 30
o Ni T 89.7 70

 

 

 

Fig. 4.2.28. Type 304 Stainfess Sfeej Clad -CcB -Fe Cermet Before and After liradiation in the

There is no evidence of grb_ss cracking in the core
or of deformation of the specimen, Because of the

inhomogeneity of the core, hardness tests were’

inconclusive; however, a slight increase in hard-
ness was observed, . Lo

~ Samples of BN-Ni are being irradicted in the
“LITR for six weeks and will be examined. The

specimens are being irradiated in the same facility

“as that used for the CaBé-Fe specimens, _In addi-

ae,I
G
o

N LiITR toa Thermal-Neutron Dose of 5.2 x 109 nvt. (“), finirradiated. (b) Irradiated. 250X. Reduced 27%.

tion, a hid"l-temperu_t'ure irradiation of the clad

BN-Ni material is under way in C-48 of the LITR,
The sample was sealed under vacuum in an Inconel
capsule. A plctinum resistance heater is being
used to control the temperature during the irradia-

- tion at 1600°F. Gas formed as a result of the

&

irradiation will be removed and analyzed.
Six samples of copper—boron carbide were pre-
pared for MTR irradiation. The capsules will be

257

T
&

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

 

 

Fig. 4.2.29. Type 304 Stainless Steel Clad CaB,-Fe Cermet Spec.imens" Sfiown in Fig. 4.2.28 at @
Lower Magnification. Both cladding-cermet interfaces may be seen. () Unirradiated. (b) lrradiated.

75X. Reduced 27%. '

inserted in an *’A’’ piece and irradiated to a
thermal-neutron dosage of 3 x 1020 nvt while the
samples are held at a temperature of 1600°F. The
instrumentation for the experiment has been con-
structed and shipped to the site. The test will be
operated to attain a maximum boron burnup equiva-
lent to that expected in the ART.!? The samples
are 0.1875 x 0.500 x 0.102 in. in thickness and
clad on two faces.2? The composition is 6.6 wt %
B ,C, =325 mesh, clad with a 3-mil copper diffusion
barrier and 7 mils of type 430 stainless steel.
The samples are mounted in two capsules, with
three samples per capsule, as shown in Fig. 42.30.

After irradiation, the samples will be returned to

ORNL for postirradiation examination.
Another promising neutron shield material is hot-
pressed boron nitride. Boron contents of up to

 

194, M, Perry, private communication to J. G. Morgan,

20y, R, D*Amore, private communication to J. G.
Morgan.

- 258

-3
O

I

44 wt % can be obtained with a density of 2,1 g/cm3
in a molded and self-bonded structure. Boron
nitride has a crystalline structure similar to that
of graphite and directionalism is present in all
physical properties,?! especially thermal expan-
sion. Pure (98% and above) boron nitride powder
was obtained from two sources, and complete
analyses were made. Bodies from these powders
will be fabricated for irradiation studies.

IRRADIATIONS OF STRESSED SHIELD ING
MATERIALS

Jo Co Wilson ' o
W. E. Brundage W. W. Davis

Testing equipment is nearly completed and cali-
brated for the irradiation of a 1 wt % boron (B'%)—
stainless steel alloy under stress in the LITR,

 

21k, M. Taylor, Ind. Eng. Chem. 47, 206 (1955).

o
)
W
 

 

-

PERIOD ENDING SEPTEMBER 10, 1956

UNCLASSIFIED
PHQOTO 18123

 

 

 

UNCLASSIFIED
PHOTO 18124

 

 

{c)

 

()

 

Fig. 4.2.30. Equipment for the lrradiation of Clad Cermet Shielding Materials in the LITR. (a) Spec-
imens - in support-grid assembly with thermocouples attached. (b} Inner capsule in which the support-
grid assembly will be weld-sealed in @ helium atmosphere. The ceramic hermetic seal through which
the thermocouple leads are brought out may be seen at the right. (c) Electrical heater made of platinum
coils in an .alumina form and stainless steel can in which mner capsule will be inserted. (d) Aluminum
container for assembly..

259

(o)
)

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

Creep in compression of the specimens will be
measured during the tests by a Bourdons-tube de-
flectometer,

Short-time compression fests on copper—boron
carbide cylinders were run at elevated temperc-
tures. A special setup in an Instron testing
machine permitted load-deformation curves to be
recorded directly, Yield stresses were determined

260

 

for copper and copper plus 20 and 40 wt % B,C at
1300°F. Creep tests were run at the same tem-
perature at a stress of 500 psi. Unloading and
reloading of short-time creep specimens improved
resistance to deformation., This suggests that
hot pressing or hot working improves the sirength

of the materials. Creep tests were also run at

1450 and 1500°F at a stress of 50 psi,

/r’{ %‘}
Ack
 

 

 

PERIOD ENDING SEPTEMBER 10, 1956

4,3, FUEL RECOVERY AND REPROCESSING

R. B.

Lindauer

D. E. Ferguson W, K. Eister H. E. Goeller

VOLATILITY PILOT PLANT DESIGN AND
CONSTRUCTION

F. N. Browder

W. H. Carr W. H. Lewis
R, P. Milford

Construction of the fused salt—fluoride volatility
pilot plant is essentially complete and shakedown
tests have been started. Studies have indicated
that cooling the UF, to 0°C while the product
cylinder is still connected to the recovery cold
traps will significantly reduce the UF, holdup in
the cold traps after each run. Equipment is there-
fore being designed and procured for circulating
a blast of cold dry air or nitrogen through the
annulus between the product cylinder and its
heater after most of the UF, has drained as a

liquid into the cylinder. Freon 12 will be used

in a direct-expansion extended-surface coil to
chill the coolant gas.

TABLE 4.3.1. DIAMETER MEASUREMENTS

" A project engin€ering report covering the design
and construction of the pilot plant is being pre-
pared.

ENGINEERING DEVELOPMENT
J. T. Leng S. H. Stainker

Engineering development work has been chiefly
concerned with demonstrations of the reliability
of freeze valves to be used in the molten salt
lines. A freeze valve of the type to be used in the
pifot plant showed no leakage in 120 pressure
tests with 20-psi N, for 20 min. This valve also
showed no leakage that could be detected with a
halide leak detector in a test with 100-psi N, and
in a test with Freon. Diometer measurements
showed slight distortion of the pipe after 99 pres-
sure tests (Table 4.3.1).

A stub for the junction of electric cable to self-
resistance-heated salt transfer lines was designed
which allows the temperature of the piping to be

OF PIPE IN A MOLTEN SALT FREEZE VALVE

BEFORE AND AFTER PRESSURE TESTS

Tests carried out in elliptical loop, 9 in. on long diameter, with a 3«in., radius of curvature

at the ends, made of %-in. sched-40 Inconel pipe; 2-in. layer of insulation around valve

Heating cycle: 200 to 650°C

Heating current through pipe: 250 amp

 

 

 

Molten salt: NoF-ZrF4-UF4 (56-40-4 mole %)
Orientation of - Outside Diameter of Pipe {in.)
Point on Loo Measurement with. - — -
- P - Respect to Plane Be&_'"f o After 70 ~ After 99
‘of Loop _ Teshng : . Prgs;ure Tesisr_ - Pressure Tests
“9o'clock”  Parallel . 0.840 0842 0.845
v .~ Transverse 0.840 0.840 : 0.835
“go’clock® Parallel 0,780 . 0.780 0.780
Transverse 0.880 0.884 0.888
“3e0’clock"” " Parallel 1 0.850 ~ 0.850 0.850
Transverse 0.840 » 0.840 0.840

 

261

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

above 525°C while the temperature of the connector
to the copper cable is below 85°C, The stub con-
sists of 6 in. of sched-40 Inconel pipe welded to
the salt transfer line, with 6 in. of 1-in. nickel
bar butt-welded to the Inconel pipe. The nickel
bar is covered with a 2-in. layer of Superex insula-
tion to within 2 in. of the Inconel-nickel junction,

The temperature of electrical insulating gaskets
in the freeze valve vent lines must be maintained
at or below 140°C, It was found experimentally
that the temperature of the piping at the location
of the gaskets would be 300°C but that two 6-in.-
dia copper fins spaced 1 in. apart and silver-
soldered to the vent lines just above the vent

furnace would reduce the temperature at the gasket

location to the required 140°C,

A device for sampling the molten salt (Fig. 4.3.1)
was tested and found to be suitable for pilot plant
use., A minimum sample of 2 g was desired, and it
was found to be possible to obtain a 2.9-g sample.
The sample cup used for the tests was fabricated
of low-density graphite, and some difficulty was
experienced in removing the salt from the pores of
the graphite. A high-density-graphite sample -cup
will therefore be tested in an attempt to improve
recovery of the sample.

CHEMICAL DEVELOPMENT

M. R. Bennett G. l. Cathers
R. L, Jolley

A study of the decomposition of the UF («3NaF
complex at temperatures of 245°C and h|d1er has
confirmed the belief that uranium retention ort the
NaF bed will be excessive in the NoF desorption
step if the temperature and sweep gas flow rate
are not properly controlled. The retention results
from decomposition of the UF6-3NaF complex to a
complex of NaF with UF,, which is not volatile,
Maximum decomposition rotes of about 0.01, 0.09
and 0.5% per minute would be incurred at 250, 300
and 350°C, respectively, in the absence of fluorine
if all the vranium were in the form of the solid
complex UF_,.3NaF. Under optimum conditions,
UF, desorption from the NaF bed! competes favor-
ably with the decomposition effect, and the final
uranium retention is small.

The presence of fluorine appears to be essential
in inhibiting the decomposition reaction and,

 

1H, K. Jackson et al., ANP Quar. Prog. Rep. March

262

UNCLASSIFIED
ORNL~-LR-DWG 15857

SET SCREWS

ROD (¥g-in. DIA,
23/, in. LONG)

CONNECTOR TO OUT-
SIDE OF SHIELD

   
 
  
    
 
  

= PIN VISE

WIRE

ROD (%—m DiA,
7/gin. LONG)

0.272-in. ID—
HIGH-DENSITY GRAPHITE
ROD ( %gin. DIA , 2in.LONG} ~er

AT T T

HIGH-DENSITY GRAPHITE  |* Hi€
ROD ( Ya=in. DIA, Y in. LONG) —pmtfi?

 

ROD {¥g-in. DIA, t in. LONG) SAMPLE CUP

Fig. 4.3.1. Molten-Salt Sampler for Volatility
Pilot Plant,

possibly, in promoting refluorination of the non-
volatile compound formed by the decomposition.
The possibility of uranium retention by the decom-
position mechanism in the absorption step at 100°C
appears to be insignificant, even over extended
periods of time.

The dependence of the rate of decomposition of
UF ;«3NaF was determined in a series of runs over
the temperature range 245 to 355°C (Fig. 4.3. 2)
The reaction involved is probably

UF ;-3NaF (solid) —> UF ;+xNaF (solid) +
+ 0.5F, {gas)

x'fil
 

 

 

.4

 

- FRACTIONAL DECOMPOSITION RATE, 7 {min™')
3

" UNCLASSIFIED
. é; 1077) ORNL-LR-DWG 16495

50

log r = 6.09— (5.22 x (0%T)

N
Q0

wn

 

S5 e AT 18 1.9

© 20 {(x107%)
rir S

Fig. 43.2, Dependence of Rate of Decompo-
sition of UF ;«3NaF on Tempemiure,

S The dependence of the decomposctlon rufe on tem-
- percture is gwen by the expressnon SR
Iogr = 6.09 - (5.22 X, 103/'1') v _i
o ",where r.is fbe fracflonnl decomposmon rote in -
reciprocal minutes and T is the cbsolute fempera- -
~.ture. - The rate was’ calculnfed ‘on the bas:s of an-

1 -_absorphon cupocuty of 1.33 g of UF ‘pet gram of
 NaF,  The energy of activation was calculcfed as - 'a

. 423.9 keal/mole of UF, o3NaF comp!ex.., Atist
C possibly slgmflccnt thaf fl'IlS energy change is "_}-

PERIOD ENDING SEPTEMBER 10, 1956

opproximately the same as the enthalpy change of
+23.2 kcal/mole involved in the volcatlllzahon of
UF from the UF ' 3NaF complex.

The decomposmon data were obtained with 4- to
9-g samples of NuF (Harshaw grcde, classified to
12-20 mesh) held in a U-tube (/ in.~dia stainless
steel tubing), through which gaseous UF, was
passed at atmospheric pressure. An oil bath was
used for manually controlling the temperature to
13°C during the course of each experiment, The
runs were terminated by removing the oil bath and
cooling the sample rapidly, The UF ¢*3NaF was
formed at the beginning by saturating the NaF at a
high UF ; flow rate, after which it was decreased to
a rate of 0,1 to 1 g/min for the remaining time, The
lengths of the runs at various temperatures were
adjusted to obtain a uranium(V) content in the
final product of 1 to 10%. The excess UF, still

_absorbed on the NaF at the end of each test was
not desorbed because of the difficulty of achieving

this without changing the uranium(V) content, The
umeunt of UF ¢*3NaF complex affected by the re-
action was determined from uranium(V) and

uranium(Vl) analyses. The temperature was not

increased to above 355°C, since at the higher
temperatures it would have been difficult to main-
tain saturation of the NaF with the UF . without

the use of a pressurized system,

In prellmmary work on the decomposition reaction

- it was determined that a uranium(V) content of 20

to 26% represented a limit which could not be

~exceeded in one cycle of saturation of the NaF

with UF at 100°C followed by heating as a closed

- system fo 350 10 400°C. Usuadlly in these runs the
 NaF: weight increase corresponded closely to the
" estimate based on the assumption that the decom-
“position product is a complex of UF, with NoF,
Xeray - crystallography ‘data mdlcated that the
'-'UFsoxNaF complex in the decomposmon product
has an orthorhombic structure with ce!l dimensions

ag = 4904, by = 547 &, and ¢ = 3.87 A The
x-ray ‘patterns of y- and B-UF; were not observed

in the material.

263

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

4.4, CRITICAL EXPERIMENTS
A. D- C(I“ihan

REFLECTOR-MODERATED REACTOR
EXPERIMENTS

D. Scott
E. Demski
D. E. McCarty W. J. Fader!

"W, C. Tunnell D. A, Harvey!
E. V. Sandin!

Jo Jo Lynn

Experiments ot Room Temperatute

One of the room-temperature critical assembiies

representing the reflector-moderated reactor with
circulating fuel has been reconstructed in order to
extend the earlier studies in this program.2 The
fuel region contains olternate lamince of Teflon
and enriched uranium foil. In this particular varia-
tion, assembly CA-22.2, one of the fuel end ducts
is somewhat thicker than the other to allow fUrfher
investigation of end-duct thickness as a variable,>
The array was critical with a loading of 23.4 kg
of U235 with about 0.5% excess reactivity. The
estimated critical mass of 22.5 kg is in good agree-
ment with the value® of 23.3 + 1,0 kg, which was
based on the excess reactivity observed in the
previous loading of 28.35 kg in the same geometry.

Reflector Evaluation. = The effect on reactivity
of replacing a part of the outermost layer of the
beryllium reflector with stainless steel was meas-
ured. This peripheral layer, which is approximately
cylindrical and is separated by 8/ in. of beryllium
from the outer core shell, is 2% m. thick and 19%
in. long and is centered about the mid-plane. Re-
moval of 27.3% of the beryllium in this layer gave
a 45.9¢ loss in reactivity, whereas the addition of
stainless steel increased the reactivity 16.7¢. The
beryllium is, therefore, 2.75 times more effective
than stainless steel in this region.

Reactivity Coefficients. — The reactivity co-
efficients of a number of materials of engineering
interest were evaluated at various positions along

 

10n assignment from Pratt & Whitney Aircraft.

2. D. Callihan et al., ANP Quar. Prog. Rep. Sept. 10,
1955, ORNL-1947, p 58. ,

3p. Scott, J. J. Lynn, E. V. Sandin, S. Snyder, Three
Region Reflector Moderated Critical Assembly with End
Ducts — Experimental Results with CA-22, Enlarged End
Duct Modzfzcanon. ORNL CF-56-1-96; see also A. D.
Calllhcn, op. cit,, p 59.

264

the radius at the mid-plape of the reactor, and the
results are given in Figs. 4.4.1 and 4.4.2, in each
case the value given is the change in reactivity
introduced by filling a void with the material.

An etror appeared in the results of the beryllium

reactivity coefficient measurements in the compact- |

core reflector-moderated reactor experiments re-
ported previously.4 The error resulted from a
mistake in dimensioning a sample, and the cor-
rected data are shown in Fig. 4.4.3.

Experiments at Elevated Temperature

A number of design features and nuclear charac-
teristics of the circulating-fuel reflector-moderated

reactor being designed by Pratt & Whitey Aircraft -

are to be investigated in a critical experiment to
be performed at ORNL late in 1956. The equipment
and the experimental program are being prepared in
close collaboration with Pratt & Whitney Aircraft
personnel, who are fabricating most of the com-
ponents for assembly at ORNL, The experiment
will be performed at about 1300°F, and a molten
U235.enriched NaF-ZrF ,-UF , mixture will be used
as the fuel. This salt mixture will not be circu-
lated, but it will be transferred pneumatically from
the storage reservoir to the annulor fuel region.
The purposes of the experiment are fo determine
the critical uranium concentration, the effectiveness
of control materiais, and an over-all temperature
coefficient. A limited number of neutron-flux and
fission-rate distributions will be measured with
appropriate detectors.

The design of this experiment follows quite
closely that of the ART high-temperature critig:al
test performed last year and reported previously,?
although the dimensions and shape of the core

region will be somewhat different, The central -

beryllium island of the critical assembly (Fig.
4,4.4) will be a right circular cylinder (8.0-in.-dia)
enclosed in a Hastelloy X shell (8.5-|n.-OD Zmine=
thick)y A 2.94-in.-dia axial hole in the lsiand
lined with a 0.12-in.-thick Hastelloy X thimble

(2.88-in.«0D), will accommodate the control"

assembly. The poison section of the control rod,
which will be enclosed in Inconel, will be an

 

4A. D, Callihan et al, ANP Quar. Pfog. Rep. Marcb
10. 19569 0RN|—‘206], P 68' F'gc 3-6.

¥ ) H.oy -, »
w [*J
 

 

-

PERIOD ENDING SEPTEMBER 10, 1956

“SeerrT

ORNL-LR-DWG 16196

 

40

1o LiF, 2.689 OF Li, 2% x1%¢ x 0.20 in.

« ‘ ® 20% Li—80% Mg ALLOY, 0.328 g OF Li, 2% x 17 x 0.02 in.
8 16

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

SAMPLE POSITION; RADIAL DISTANCE FROM AXIS {in.)

»
§ A INCONEL, 56.28¢, 2 x1%¢ x»O.lOB in.

w & SODIUM, 122 g, 2Va x 2% x1in,

g

s

<

o

W
SO

s

g FUEL REGION BERYLLIUM REFLECTOR

T} : - -
@

L

a

a

= < :

> -

5 = DIVIDE SCALE BY 100
E‘ : .
Z f\ ‘
2 /
5 /
4 .
,/\\
10 15 20 25

Fig. 4, 4.1. Reactivity Changes Effected by the Addition of Vanous Materials to the RMR Critical

 

 

 

 

 

    

 

 

 

 

 

 

 

   

 

 

 

  
 

 

 
 

 

 

 

 

 

 

Assembly CA-22-2
Boener
8 ORNL~-LR-DWG 16197
e DT 4 U(9347% UZ3%), 1.31g, 2% x 76 x 0.001 in.
7 \ - o A U(93.17% UZ33),43.81q, 4 x 24 x 0.004 in.
' e & BERYLLIUM, 254g, 273 x 27 x¥in.
: % _ o BERYLLIUM, 1259, 2% x1%g x1in. |
o L ) .
- B
i
-t
Q.
= .
o
“ g — , :
S - ' FUEL REGION © " BERYLLIUM REFLECTOR .
E - T
& .a
'>: 7.-. -
=3 \
-_ § | owioe scALE A\
W .| BY 100 -\
® N
z 2 \ .
PLoga o AN ‘><DIVIDESCALEBY100. ‘
N ~OIVIDE SCALE BY10 \n
o — ! ) L

o 5. . a0 s 20
B "SAMPLEPOSITION RADIAL DISTANCE FROM AXIS {in.)

Fig. 4.4.2. Reactivity Changes Effected I:y the Addition of Vurious Materials to the RMR Critical

Assembly CA.22.2,

25

265

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

- AroRET
ORNL-LR~DWG 16198
40 T T
' o URANIUM, 27.47g OF U233 45/5x 27 x 0.008 in.
_ e BERYLLIUM, 125,27 x 17 x tin.
2 35 *
S
i’- ISLAND FUEL REGION BERYLLIUM REFLECTOR -
-t
s 30 —
<l
wn
W
0 /
2 25
<l
a
o
8 / /\
Q
& 20 I—
w
a
z / \
> |,
NN | /l
a DIVIDE SCALE \ DIVIDE SCALE BY 10
& BY 10 /
Zz 10 AN
2 \_J/ \\
< .
o
5 :
i ) \..,
0
o : 5 10 15 20 25

SAMPLE POSITION; RADIAL DISTANCE FROM AXIS (in.)

Fig. 4.4.3. Reactivity Changes Effected by the Addition of Various Materials in the Compact-Core

RMR Critical Assembly.

annulus (2.46-in.-0D, 1.97-in.-ID} made of a mix-
ture of 30% rare earth oxides (samarium and
gadolinium) and 70% nickel. The outer core shell,
also of Hastelloy X, will have an inside diameter
varying from 14.85 to 21.05 in. The wall thickness
will be 0,16 in. at the center, increasing to 0.25 in.
at the top and bottom. The bottom of the fuel
section will be flared to represent the region in the
power reactor where the fuel will be directed
radially ocutward and upward into the heat ex-
changers.

The reflector will be built up of 13 annular slabs
of beryllium with outside diameters varying from
22,5 in. at the top to 48.0 in. ot the mid-plane and
32.0 in. at the bottom. The central holes will be
‘tapered to conform approximately with the shape
of the outer core shell. The beryllium refiector,
therefore, will be slightly over 13 in. thick for @
distance of 8 in. on each side of the mid-plane and
will decrease to 3.3 in. at the top and 6.5 in. at
the bottom. To mock up the coolant passages,
832 vertical holes, 0.25 in. in diameter, will be

266

drilled through the reflector, and 125 holes, also
0.25 in. in diameter, will be drilled through the
island. The coolent which would normally occupy
these holes was not mocked up because of opera-
tional and safety considerations.

Fast-neutron leakage from the entrance and exit
ducts of the power reactor is to be reduced by
suppressing the fission rates in these regions. In
current designs, this suppression is to be accom-
plished by shielding the fuel from neutrons re-
flected by the beryllium. In the critical experiment,
shields composed of a mixture of boron. carbide
and copper will be placed in the beryllium regions
at the bottom of the reactor adjacent to both the
core shells. In similar locations at the top of the
reactor, cylindrical shelis of B19, 0.1 in. thick, in
Hastelloy X annular.containers, can be inserted
vertically to a position 6 in. below the top of the

_ beryllium.

The volume of the reflector, island, and fuel
region will be 36.7 3, and the volume of the fuel
region alone will be 5.3 3,

C
 

 

575, || 481n..

'PERIOD ENDING SEPTEMBER 10, 1956

Eenee
ORNL-LR-DWG 16199

  
    
  
 

ROD THIMBLE

62.07 In.

 

 

STAINLESS STEEL

      
 
 

[ HELIUM

| ATMOSPHERE -

   

3/8 in,

FUEL ANNULUS

 
   
    

BERYLLIUM
REFLECTOR

FUEL ANNULUS -

   
 
     
   
 
   

. 18.0in. 4.

  

DETECTOR
THIMBLES

 

STAINLESS STEEL

‘8-Cu

Fig. 4.4.4. Schematic Diagram of High-Temperature RMR Critical Assembly.

267

 
 

 

 
 

‘.)

5.1. SHIELDING THEORY

A MONTE CARLO STUDY OF THE GAMMA-RAY
ENERGY FLUX, DOSE RATE, AND BUILDUP
FACTORS IN'A LEAD-WATER SLAB
SHIELD OF FINITE THICKNESS

S. Auslender!

The gamma-ray energy flux, dose rate, and buildup
factors in a lead-water shield of finite thickness
have been calculated by a Monte Carlo method.?
The calculation included 1-, 3-, ond 6-Mev photons
incident on the slab both along a normal and at an
angle of 60 deg. {Calculations for an angle of
75Y, deg were also performed, but they are not
included here since the attenuation was quite
large in spite of the large buildup fczctor. For the
same reason the results for 1-Mev photons incident
at an angle of 60 deg were also omitted,)  The
buildup factors for energy and dose obtained in this
calculation were compared with those obtained by
the use of the moments method® for monoenergetic,
plane monodirectional sources normally . incident
upon a semi-infinite, homogeneous medium,

The thicknesses of the lead-water shield in

‘centimeters and in-mean free paths for the various

incident gamma-ray . energies . are given in Table
5.1.1, and the tissue dose equivalents for the
incident fluxes (no shield present) are given in

Table 5.]-20

 

IOn asmgnment from Pratt & Whitney Alrcraft.

2g, Auslender, ANP Quar. Prog. Rep, March 10, -1956.
ORNL-2061, p 223.

3H, Goldstein -and 'J. E. Wilkins, Jr., C'alculatzons of

the Peretration of Gamma Rays. Fmal Report. NYO-
.3075 (June 30, ]954) _ . 7 o

TABLE 5.1 'I. NORMAL THICKNESSES OF A
LEAD-WATER SLAB SHIELD

 

Thn:kness -

 

 

 

e Mecm Free Paths (mfp) .
. Region™ o o
’ sre For 'I-Mev For 3Mev For 6-Mev- o
Phofons - Photons Phctons
P, ss"  _9.0739 S 5';'5_'54?: 5.878 _
; Hzol, 3581 2502 1382 . 096
Total  47.40  11.630 7.035 6.874

 

 

TABLE 5.1.2. TISSUE DOSE EQUIVALENT FOR
INCIDENT FLUX

 

 

Photon Energy ( mr /hr
Do/q5
1
(Mev) . y/cmzo sec
1 _ - 0.001923
3 0.004368
6 0.007206

 

The results of the calculations for 1-Mev photons
are presented in Figs. 5.1.1 through 5.1.4, those
for 3-Mev photons in Figs. 5.1.5 through 5,1.10,
and those for 6-Mev photons in Figs. 5,1.11 through
5.1.16.

The normalized energy fluxes for the 1-, 3-, and
6-Mev energy groups are plotted in Figs. 5.1.1,
5.1.5, and 5.1.11, respectively, as functions of
the normal thickness of the shield (in centimeters).
The uncollided energy flux, normalized to unity at
the initial boundary, is also plotted. The third
curve in each figure is the energy buildup factor,
Bp, which is the ratio of the two energy flux
curves:

bg/E oby ¢k

BE =

e—t/h . qusle—t/h

where , _,

5 .qSE(Mev/cm?osec) = energy flux at the point of
_ - . interest in the shield,

Ed(Mev/y) ‘= initial photon energy, -

-'lc;Sl(y'S/CmZ-Sec')' :number flux  incident von

. _sh:eld o
’dlstance befween the mmul '

- t{em)
.~ _boundary and the :point of
e ‘--.,m?erest in the shleld
Alem) = relaxahon Iength
e“t/A= :'uncolllded energy flux at

o ‘the pomf of mteresf

In a s:mllar manner the normuhzed dose rates
and dose buildup factors, B, for the three energies

271

 

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

are plotted in Figs. 5.1.3, 5.1.7, 5.1.9, 5.1.13, and
5.1.15. Here

D/¢y D

 

B = =
’ Doe =lse¢e G/A/‘#l Doe-t sec 8/A
where
D(mr/hr) = dose rate (tissue) at the point of

interest,

Dy(mr/hr). = dose rate (tissue) of the uncollided
incident photons at the initial
boundary.

The various energy buildup factors as functions
of oblique thickness, in mean free paths, are pre-
sented in Figs. 5.1.2, 51.6, oand 5.1.12. Corre-
sponding dose buildup factors are given in Figs.
5. 1.4, 5.1.8, 5.1.10, 5.1.14, and 51.16, Data from
the resuvits of the moments method solutiond are
also plotted for purposes of comparison, although

it must be remembered that those calculations were
for infinite homogeneous media and are not directly
comparable to the present calculations for a finite
two-region slab. The calculations for lead do agree
reasonably well for the first few reloxation lengths
where the effects stemming from the dissimilarity
of the slabs should be least.

The dose rate and dose buildup factors resulting
from an earlier Monte Carlo calculation for 3-Mev
photons normally incident upon a one-region 8-mfp-~
thick lead shield® are also presented (Figs. 5.1.17
and 5.1.18) as an example of a case intermediate
between the Monte Corlo calculation for a two-
region finite shield oand the moments method solu-
tion for the one-region semi-infinite shield. A
comparison of the dose rate in the one-region
finite shield with the dose rate in the one-region
semi-infinite shield (calculated from buildup fac-
tors reported in ref 3) shows good agreement. The
characteristic dip near the final boundary of the
finite lead shield is apporent.

The dose buildup factors for the 3-Mev incident
photons in the two-region and one-region finite
shields (Figs. 5.1.8 end 5.1.18) are identical to
within 2 mfp of the lead-water interface. It is
apparent that at this energy and angle of incidence
the back scattering is important in lead for a
distance of about 4 em (Ap, = 2,05 em at 3 Mev).

The scattering in the water tends to compensate

 

4s, Auslender, unpublished work.

272

G644

for the finite thickness of the lead-water slab,

It can also be concluded that the back-scattered
flux at that interface is about 12% of the total flux
and about 20% of the scattered flux. The dif-
ferences for the finite and the semi-infinite one-
region shields (Fig. 5.1.18) are particlly due to
differences in the cross-section data used for the
calculations.  An extensive discussion of the
errors in the calculations from the moments method
is included in ref 3, Consistent errors may exist
that will bias the answers obtained with that
method. The Monte Carlo data is much less sub-
ject to bias thon to statistical fluctuation. (The
large fluctuation of the data in Fig. 51.9 may
indicate ‘a relatively large error, but this is stlll
probably less than 10%.)

Since the Monte Carlo calculations for the two-
region finite shield and the moments method solu-
tion for the one-region semi-infinite shield were in
general agreement, the striking differences in the
buildup factors near the lead-water interface and
near the final boundary should be enlightening, In
Figs. 5.1.2 and 5.1.4, for example, the energy and
dose buildup factors for 1-Mev photons in a finite
two-tegion shield increase sharply in water to a
valve about half way between those for semi-
infinite one-region shields of water and lead. At
this energy the finite water region is 2‘/2 mfp thick,
and, at first, the buildup increoses almost at the
same rate as it does initially in the semi-infinite
water shield. This confirms that at this energy,
which is below the minimum in the total cross
section, the uncollided radiation dominates in the
penetration of the lead, as it should.

In Figs. 51.6 and 51.8 the increase of the
buildup factors in the water is very small, since,
for the primary 3-Mev photons for both lead and
water, the major portion of the total cross section
is attributable to scattering. Hence, the difference
in the behavior of the flux in water and in lead is
due to that small portion of the flux which is
absorbed in the lead but is scattered in the water,
For 6-Mev photons (Figs. 5.1.12 and 5.1.14) the
absorption cross section of lead is no longer so
small that the absorption in it is almost negligible.
In this case the buildup factor increases rather
abruptly in the water, almost to the buildup factor
for 6'/ mfp of water calculated by using the
moments method,

It is evident from the results of the calculations
for radiation incident at 60 deg (Figs. 5.1.9, 5.1.10,

11
 

Part 5

REACTOR SHIELDING

E. P, Blizard

 
 

 

 

 

 
 

5.1.15, and 5.1,16) that the practice of using only
normal incidence data for shield designs can lead
to a poor approximation. This problem becomes
most acute when the number of meon free paths
across the shield is small or when the angular
distribution is such that a large portion of the

UNCLASSIFIED -
2=-01-059—104

 

EO = INITIAL PHOTON ENERGY
= INCIDENT NUMBER FLUX
& = ENERGY FLUX
Apy = 4.37cm
Ay, 0 = 14.09cm
By = ENERGY BUILOUP FACTOR

‘o 10 20 30 a0 50 60
1. NORMAL THICKNESS (cm}

Fig. 5.1.1. Gamma-Ray Energy Flux and Energy
Buildup Factor as a Function of the Normal Thick-
ness (Centimeters) of a Finite Lead-Water Slab’
Shield: 1-Mev Normally Incident Photons.

iy

s
boma

PERIOD ENDING SEPTEMBER 10, 1956

radiation is not normal or nearly normal to the
slab. Figure 5.1.16, which shows the flux de-
pression near the initial boundary and the large
buildup factor into the slab, is an excellent ex-
ample of streaming (or short-circuiting).

UNCLASSIFIED

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

. 201~ 059-102
L |
— MONTE CARLO, FINITE SLAB
© H,0] MOMENTS METHQD, =
SEMI- INFINITE SLAB,
NORMAL INCIDENCE
2 8 Pb ) (Ref.: NYO-3075)
<
[+ 4
e
W
£ 0 °
o
> Pb H,0 =
o 2
2
3 /
§ 5 / Ldpeor = 1.63 mfp
o /
z ] .
K /
© § 1
»
2 ° /'
./

 

 

 

 

 

 

 

 

 

 

0 2 a4 6 8 10 12 14 6
fof» OBLIQUE THICKNESS (mfp}

Fig. 5.1.2. Gamma-Ray Energy Buildup Facter
as a Function of the Oblique Thickness (Mzan Free
Paths) of a Finite Lead-Water Slab Shield: 1-Mev
Normally Incident Photons.

273

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

UNCLASSIFED
2-01—059—103

O = DOSE RATE

O = INITIAL DOSE RATE
INCIDENT NUMBER FLUX
Do/, = 0001323 (e /he)/y/eme-sec)
App ™ £.37cm
My,0 = $4.09 cm

&, = DOSE BUILDUP FACTOR

 

0 10 20 30 40 50 €0
7, NORMAL THICKNESS (cm)

Figo 50‘03. 7

Gamma-Ray Dose Rate and Dose

Buildup Factor as a Function of the Normal Thick-
ness (Centimeters) of a Finite Lead-Water Slab

Shield: 1-Mev Normally Incident Photons.

UNCLASSIFIED
2-01-059-104

—— MONTE CARLO, FINITE

o H,0] MOMENTS METHOD,
SEMI- INFINITE SLAB
NORMAL INCIDENCE

® Pb ) (Ref.: NYO-3075)

8,, DOSE BUIL JUP FACTOR

 

0 2 4 6 8 0 {2 14 46
#o”» OBLIQUE THICKNESS (mfp)

Fig. 5.1.4. Gamma-Ray Dose Buildup Factor as
a Function of Obligue Thickness (Mean Free Paths)
of a Finite Lead-Water Slab Shield: 1-Mev Normally

Incident Photons.

274

UNCLASSIFIED
2-01-059-105

= {NITIAL

¢, =INCIDENT NUMBER FLUX
¢ = ENERGY FLUX

kpp= 2.05¢cm

Aun=25.91cm

8 = ENERGY BUILDUP FACTOR

102

1073

 

0 10 20 30 40 30 60
f, NORMAL THICKNESS {cm)

Fig. 5.1.5. Gamma-Ray Emergy Flux and Energy
Buildup Factor as a Function of the Normal Thick-
ness (Centimeters) of a Finite Lead-Water Slab
Shield: 3-Mev Normally Incident Photons.

UNCLASSIFIED
2-01-059-106

— MONTE CARLO, FINITE SLAB

© H,0] MOMENTS METHOD,
SEMI—INFINITE SLAB,
NORMAL INCIDENCE

¢ Pb | ‘(Ref.: NYO—3075)

B, ENERGY BUILDUP FACTOR

 

o -2 4 6 8 10 12 14 16
por+ OBLIQUE THICKNESS (mfp)

Fig. 5.1.6. Gamma-Ray Energy Buildup Factor

as a Function of the Oblique Thickness (Mean Free
Paths) of a Finite Lead-Water Slab Shield: 3-Mev

Normally Incident Photons.

014
 

 

 

UNCLASSIFIED
2-0¢-059-107

O = INITIAL DOSE RATE
¢; = INCIDENT NUMBER
= 0.004368 (mr,
Apy = 2.05¢m
Ayyo = 25.91cm
8, = DOSE BUILOUP FACTOR

 

0 10 20 30 40 30 €0
' #, NORMAL, THICKNESS {cm}

Fig. 5.1.7. Gomma-Ray Dose Reate and Dose
Bmldup Factor as a Function of the Normal Thick-
ness (Centimeters) of a Finite Lead-Water Slab
Shield: 3-Mev Normally Incident Photons.

- o : UNCLASSIFIED
2-01~059-108

i i = MONTE CARLO, FINITE SLAB

- . o ° HZO MOMENTS METHOD,
SEMI- INFINITE SLAB

NORMAL INCIDENCE

2 ® Pb. ) (Ref.: NYO-3075)

S

8;, DOSE BUILDUP FACTOR

 

o _‘-aj_'_;'-4 6 B, 10 12 e
L p.,r oauouETmcxNEssfmfp) Lo

a Function of Oblique Thickness (Mean Free Paths)
of a Finite Lead-Water Slab Shield: 3-Mev Normully
incident Photons,

015

- Ba4

PERIOD ENDING SEPTEMBER 10, 1956

UNCLASSIFIED
2-01-059-109

1072

-3

10 10
D = DOSE RATE
Dg= INITIAL DOSE RATE
4, = INCIDENT NUMBER FLUX
105 0y /%, = 6.004368 {mr/hr)fly scm?- w0
App= 2.05¢cm
ngO' 25.9t cm
8,= DOSE BUILDUP FACTOR
10" o5
10° w08
o/%,
10° w07
—24/)
/ /4,[
0w '0-8

. w0 ®
0 {0 20 30 40 50 60

£, NORMAL THICKNESS (cm)

 

- Fig. 5.1.9. Gamma-Ray Dose Rate and Dose
Buildup Factor as ¢ Function of the Normal Thick-
ness (Centimeters) of a Finite Lead-Water Slab
Shield: 3-Mev Photons Incident at a 6§0-deg Angle.

UNCLASSIFIED
2-01-059—110
102

© H,0 | MOMENTS METHOD,
- SEMI-INFINITE SLAB,
5 NORMAL INCIDENCE
® Pb ] (Ref.: NYO-3075)

S N

8, DOSE BUILDUP FACTOR

o

 

1 -
o 2 4 € . 8 o 2 14 16
#ot, OBLIQUE THICKNESS (mfp)

Fig. 5. 1.8., Gnmmu-Ruy Dose Buildup Fuctor us‘ -Fig. 5.1.10.  Gamma-Ray Dose Buildup Factor as

¢ Function of Obliqgue Thickness (Mean Free
Paths) of a Finite Lead-Water Slab Shield: 3-Mev
Photons Incident at ¢ 60-deg Angle.

275

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

I.IICLAgSl:lED
107 gl 10
5 5
2 2
- £4 = INITIAL PHOTON ENERGY '
¢, = INCIDENT NUMBER FLUX 5
& = ENERGY FLUX
2 Ay, = 1.97em 2
108 =L Mo = 3596 cm - 1ot
F= 5, = ENERGY BUILDUP FACTOR g 5

   

0 10 20 30 40 50 60
7, NORMAL THICKNESS (cm)

Fig. 5.1.11. Gamma-Rey Energy Flux and Energy
Buildup Factor as a Function of the Normal Thick-
ness (Centimeters) of ¢ Finite Lead-Water Slob
Shield: &-Mev Normally Incident Photons.

UNCLASSIFIED
2-04-059—-142

—= MONTE CARLO, FINITE SLAB

© H,0] MOMENTS METHOD,
SEMi— INFINITE SLAB),
NORMAL INCIDENCE

* Pb (Ref.: NYO—-3075)

8., ENERGY BUILDUP FACTOR

 

0 2 4 6 8 10 12 14 16
RBo’ s OBLIQUE THICKNESS (mfp)

Fig. 5.1.12. Gamma-Ray Energy Buildup Factor
as a Function of the Oblique Thickness (Mean
Free Paths) of a Finite Lead-Water Slab Shield:
6-Mev Normally Incident Photons.

276

UNGLASSIFIED
2-01-059-43

_f/y¢!

&= DOSE RATE

Op =INFIAL DOSE RATE _

#, =INCIDENT NUMSER FLUX 1078
4, /$, =0.007206 (me/hr)/(y 7em®. sec)
Apy=1.97cm
Hgo = 3596 cm

8,= DOSE BUILDUP FACTOR

A
1077

 

0 10 20 30 40 50 60
#, NORMAL THICKNESS (cm}

Fig. 5.1.13. Gomma-Ray Dose Rate and Dose
Buildup Factor s a Function of the Normal Thick-
ness (Centimeters) of a Finite Lead-Water Slab
Shield: 6-Mev Normally Incident Photons.

UNCLASSIFIED
2-01—-059~1{14

== MONTE CARLC, FINITE

o H,0 MOMENTS METHOD,
SEMi-INFINITE SLAB,
NORMAL INCIDENCE

® Pb J (Ref.. NYO-3075)

8,, DOSE BUILDUP FACTOR

 

o 2 4 € 8 10 12 14 16
#of, OBLIQUE THICKNESS (mfp)

Fig. 5.1.14. Gamma-Ray Dose Buildup Factor as
a Function of Oblique Thickness (Mean Free Paths)
of a Finite Lead-Water Slab Sl-ueld 6-Nbv Normully
Incident Photons.

N
Y
oy

e

"\
 

Fig. 5.1.16. Gamma-Ray

 

UNCLASSIFIED
2-01<059-115
1072

D = DOSE RATE
Do = INITIAL DOSE RATE
7 = INCIDENT NUMBER FLUX 1073
By /¢y = 0.007208 tme/hri/iy femP.sec) 5
Apy = 1.97cm
)LHao = 35.96 ¢m
& = DOSE BUILDUP FACTOR

 

o 10 20 30 40
% NORMAL THICKNESS {cm)

50 60

Fig. 5.1.15. Gamma-Ray Dose Rate and Dose
Buildup Facfor as a Function of the Normal Thick-
ness (Centimeters) of a. Finite Lead-Water Slab
Shield: 6-Mev Photons Incident at a §0-deg Angle.

UNCLASSIFIED
2-01~-059-116

== MONTE CARLQ, FINITE SLAB

9 Hy0f MOMENTS METHOD,
SEMI-INFINITE SLAB,
NORMAL INCIDENCE

® Pb | (Ref.: NYO-3075)

8-, DOSE BUILDUP FACTOR

 

0o 2 4 6 8 0. 12 14 6.
" pory OBLIQUE THICKNESS (mfp} " =~ =

as a Function of Oblique Thickness (Mean Free
Paths) of a Finite Lead-Water Slab Shield: 6-Mev
Photons Incident at a 60-deg Angle.

  

PERIOD ENDING SEPTEMBER 10, 1956

UNCL ASSIFIED
2-01-058-#17

O = DOSE RATE

fp = INITIAL DOSE RATE
$, = INCIDENT NUMBER FLUX

D°/¢f =0.004368 (mr/hr)/(y!cmz- sec)
Apy=2.05¢cm

® MONTE CARLO, FINITE SLAB
& MOMENTS METHOD, SEMI-INFiNfTE
SLAB (REF: NYQ-3075)

 

2 3 ) ] 5 7 8
Po7 OBLIQUE THICKNESS (mfp)

Fig. 5.1.17. Gamma-Ray Dose Rate as a Func-
tion of Normal Thickness (Mecan Free Paths) of a
Finite Lead Slab Shield: 3-Mev Normally Incident
Photons.

UNCLASSIFIED
2 . 2-04-059-118

® MONTE FINITE SLAB
© MOMENTS METHOD,
SEMI- SLAB (Ref.: NYO-3075)

8-, DOSE BUILDUP FACTOR

  
  
 

2 6 8 10 © 19 6
Hols THICKNESS {mfp)

     

DoseBuildupFuctor Flg. 5.1.18. Gummu-'Rcly'Dose- Buildup Factor as

a Function of Oblique Thickness (Mean Free Paths)
of a Finite Lead Slab Shield: 3-Mev Normally
Incident Photons. '

277

ffiffifigfi’

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

5.2. LID TANK SHIELDING FACILITY
R. W. Peelle

STUDY OF ADVANCED SHIELDING MATERIALS

V. R. Burrus! J. M. Miller
J. R, Smolen? D. R. Otis?
P. B. Hemmig!

The studies of advanced shielding materials45
have now included tests on mockups consisting of

 

10n assignment from U, S. Air Force.
On assignment from Pratt & Whitney Aircraft,
On assignment from Convair, San Diego.

“These experiments have been designed to be of
particular interest to GE-ANPD,

5R. W, Peblle et al, ANP Quar. Prog. Rep. June 10,
1956, ORNL.-210¢4, p 269.

a beryllium moderator region, a lead or depleted
uranium gamma-ray shield, and a lithium hydride
and oil neutron shield. In some tests a boral sheet
was inserted outside the beryllium layer to prevent
thermal neutrons  from entering the gamma-ray
shield. The specific configurations tested are
summarized in Table 5,2.1. The compositions and
dimensions of all the materials except the beryl-
lium aond boral were previously reported. The
metallic beryllium slab is 48 x 49 x 4 in, and the
boral sheet is 48 x 48 x |n. The latter has an
average density of 2.6 g/?: ‘and contains about
20 wt % boron, or 0.66 g of boron per square centi-
meter.

TABLE 5.2.1. SUMMARY OF THE CONFIGURATIONS USED FOR L TSF
MOCKUP TESTS OF ADVANCED SHIELDING MATERIALS

 

Configuration No.

Composition

 

6%-3 4 in. of Be in oil

4 in. of Be +

12 in. of LiH in oil

4 in. of Be + 24 in. of LiH in oil

69f'| 2

4 in. of Be + 3 in, of Pb in oil

4 in. of Be + 3 in. of Pb + 1 ft of LiH in oil
4 in. of Be + 3 in. of Pb + 2 ft of LiH in oil
4 in. of Be +3 in. of Pb + 3 ft of LiH in oil

69-13

4 in. of Be + % in. of boral +3 in. of Pb in oil

4 in. of Be + % in. of boral +3 in. of Pb+ 1 f1 of LiH in oil
4 in. of Be + }’2 in. of boral +3 in. of Pb + 2 ft of LiH in oil
4 in. of Be + % in. of boral +3 in. of Pb+ 3 fi'of LiH in _oi_l

69-14

4 in. of Be + 3 in, of U”s* in oil

4 in. of Be +3 in. of U238 4+ 1 f1 of LiH in oil
4 in. of Be +3 in. of U238 + 2 ft of LiH in oil
4 in. of Be +3 in. of U238 4+ 3 ft of LiH in oil

69-15

4 in. of Be + z In. of boral + 3 in. of U238 in oil

4 in. of Be + g in. of boral +3 in. of Uz’38 4+ 1 f1 of LiH in oil
4 in. of Be + z in. of boral + 3 in. of U238 + 2 £t of LiH in oil
4 in. of Be + ,'2 in. of boral + 3 in. of U238 -|-3, #t of LiH in oil

 

*Uranium depleted to 0,24 wt % in U235.

278
 

 

 

The materials were placed within the oil medium
as close as possible to the Lid Tank converter
plate, The thermal-neutron fluxes quoted from the
data are equal to the neutron density times
2200 m/sec. The gamma-ray tissue dose rate
measurements were obtained with an anthracene
scintillation detector calibrated against a stand-
ardized radium source.

The gamma-ray dose rates beyond the various

configurations “are shown in Figs. 5.2.1 through
5.25. It is not yet clear to what extent these

results were influenced by the thin loyers of oil
between the slabs in the configuraflons. Neutrons

slowed down in these oil layers may give rise to

secondary production- in the adjoining slabs of

materials, as well as in the oil itself. The average -

total thickness of these layers was 3 ¢m for each

conflguratlon, and tests to study thelr effects are

being planned. ‘ : :
Since there was' a noticeable decrease in the

 gamma-ray dose rate whenever lithium hydride was

added to the configuration immediately behind the

gamma-ray shield material, an investigation was

performed to determine how much of this reduction

could be attributed to the elimination of oil capture -

gamma rays. Measurements were taken beyond a
configuration in which there was a I-ft layer of oil
between a lead gamma-ray shield (followed by
boral) and the lithium hydride. The gamma-ray
dose rate was considerably increased (Fig. 5.2.6),

PERIOD ENDING SEPTEMBER 10, 1956

ond a curve of the difference between the two
curves should largely represent the attenuation
curve of cil capture gamma rays.

There was a strong reduction in the gamma-ray
dose rate when the depleted uranium slab was
moved to a region of relatively low neutron flux
(Fig. 5.2.5). The preliminary interpretation of this
reduction is that fissions caused by intermediate
and fast neutrons were largely eliminated by the
additional neutron shielding.

The thermal-neutron trcverses for the various

configurations are shown in Figs. 5.2.7 through
- 5.2.11, Only the traverse in oil was checked

against gold-foil measurements,  Figure 5.2.9

shows that the thermal-neutron flux is independent

of the order of the lithium hydride and lead, except
for the difference at the beginning of each traverse
caused by variations in the amount of oil trapped
between the various slabs. In Fig, 5.2.11, how-
ever, for a similar arrangement containing depleted
uranium, moving the uranium out of the intense
neutron field reduced the thermal-neutron flux.
This seems to be the result of the intermediate or

" fast fissions discussed above in connection with

the gamma-ray measurements.
Studies are being continued on configurations

similor to those reported here. In addition, fast-

neutron dose rate traverses will be made available
in a future report.

279

 

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

SR

404

OIL ONLY
10*

5 4 in. OF Be + OIL

4in. OF Be + 4 ft OF LiH + OIL

102

4 in, OF'_Be + 2 ft OF LiH + OIL

GAMMA-RAY TISSUE DOSE RATE (ergs/g-hrew)'

10

5

2

BASED ON SOURCE PLATE POWER OF 5.24 w

4

5

2

107!
20 40 60 80 100 120 140

Zg, DISTANCE FROM SOURCE PLATE (cm)

Fig. 5.2.1. Gamma-Ray Tissue Dose Rate Traverses Beyond Configuration 69-3.

280

2-04-057-69-275

 

 

iy

460
 

 

GAMMA-RAY TISSUE DOSE RATE (ergs/g-hr-w)

PERIOD ENDING SEPTEMBER 10, 1956

“<ECRET
4 2-04-057-€9-276
10

BASED ON SOURCE PLATE POWER OF 5.24 w

N

OIL ONLY

-
Q
~n

o

4in. OF Be + 3 in. OF Pb+ OIL

N

S

4 in, OF Be + 3in.
OF Pb+1ft OF LiH+ OIL

 

5
2
~ 4in. OF Be + 3in.
OF Pb+ 2 ft OF LiH + OIL
\ ,
5
4 in. OF Be + 3 in.
OF Pb+ 3 ft OF LiH + OIL
2
10741 _ — ~ .
20 .. a0 - 80 80 - 100 120 140 160

Zgo, DISTANCE FROM SOURCE PLATE {cm)

Fig. 522 Gamma-Ray Tissve Do_se Rate Traverses Beyond Configuration.69-12.

281

 
 

 

 

ANP PROJECT PROGRESS REPORT

282

GAMMA=~RAY TISSUE DOSE RATE (ergs/g-hr-w)

 

AREOREPr
2—-01—-057—-69-277

BASED ON SOURCE PLATE POWER OF 5.24w

\\OIL ONLY

N

N

4in.OF Be+ Yo in. OF
BORAL + 3in. OF Pb+ OIL

4in.OF Be + Y4 in. OF
BORAL +3in.OF Pb+1 ft OF LiH +OIL

4in. OF Be +'% in. OF +3in. OF Pb+ 2t OF LiH+ OIL

4in. OF Be + Y4 in. OF BORAL + 3 in. OF Pb-+ 3 ft OF LiH+OIL

20 40 60 80 100 120 140 160
Zgq, DISTANCE FROM SOURCE PLATE {cm)

Fig. 5.2.3. Gamma-Ray Tissve Dose Rate Traverses Beyond Configuration 69-13.
 

 

 

 

GAMMA-RAY TISSUE DOSE RATE (ergs/g-hr-w)

10

1071

PERIOD ENDING SEPTEMBER 10, 1956

SESRET
2-01-057-69-279

OH. ONLY BASED ON A SOURCE PLATE POWEROF 5.2 w

4in. OF Be + 3in. OF U + OIL

4in.OFBe+ 3in.OF U+ 2 ft OF LiH + OIL

4in.OF Be +3in. OF U + { ft OF LiH + OIL

4in. OF Be + 3in.OF U+ 3 ft OF LiH + OIL

 

20 40 - 60 80 100 120 140 160

2., DISTANCE FROM SOURCE PLATE (cm)

Fig. 5.2.4. Gamma-Ray Tissue Dose Rate Traverses Beyond Configuration 69-14.

283

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

284

SECNET
3 2-04-057-69-278
10

BASED ON SOURCE PLATE POWER OF 5.24 w

102

OiL ONLY

4 in. OF Be + ' in.OF
BORAL + 3in. OF U + OIL

10

4in. OF Be + Y in. OF BORAL
+3in. OF U +1ft OF LiH + OIL

5 4in. OF Be + Y% in. OF BORAL
+3in. OF U+ 2 ft OF LiH+ OIL

GAMMA-RAY TISSUE DOSE RATE (ergs/g-hr-w)

4 in. OF Be + %% in. OF BORAL
+3in. OF U+ 3 ft OF LiH+ OIL

107

S 4 in. OF Be + 4 ft OF LiH + ¥ in. OF
BORAL + 3 in. OF U + 1 ft OF LiH + OIL

 

102
20 40 60 80 100 120 140 160
Zgg, DISTANCE FROM SOURCE PLATE (cm)

Fig. 5.2.5. Gamma-Ray Tissue Dose Rate Traverses Beyond Configuration 69-15.
 

 

GAMMA - RAY TISSUE DOSE RATE (ergs/g-he-w)

10

10

20

PERIOD ENDING SEPTEMBER 10, 1956

SEOReS
2-0t-057-69- 285

BASED ON SOURCE PLATE POWER OF 5.24 w

8,3in.OF Pb+ Y2 in.OF BORAL + {ft OF OIL +
1ft OF LiH +0IL

A, 3in.OF Pb+Y2in.OF BORAL + 111 OF LiH+OIL
(CAPTURE GAMMA-RAYS SUPPRESSED)

 

40

- 60 ‘80 400 120 140 160

| Zgo. DISTANCE FROM SOURCE PLATE (cm)

Fig. 5.2.6. Aflenuatién of Gamme Rays Produced by Neutrons Captured in Oil Beyond 3 in. of Lead.

285

 
 

 

 

ANP PROJECT PROGRESS REPORT

nvg, THERMAL - NEUTRON FLUX (neutrons/cm® sec w

286

C

SECREY
2—0{—057-69-280

“w

o

BASED ON SOURCE PLATE POWER OF 5.24 w

Be-+1ft OF LiH+ OIL

OIL ONLY

4in. OF Be + 2t OF LiH + OIL

 

-

O 10 20 30 40 50 60 70 - 80 S0 100 1o . 120 130 140 150 te0 {70
Zgq, DISTANCE FROM SOURCE PLATE (tm)

Fig. 5.2.7. Thermal-Neutron Flux Traverses Beyond Configuration 69-3.
 

 

 

 

ljl

N

- .
o obl‘. n o

n

_..
o
N

n

Ay, THERMAL=NEUTRON FLUX (neutrons /cm®-sec-w)
o

o

S

o

PERIOD ENDING SEPTEMBER 10, 1956

SESRER
2—-01—-057—-69-284

BASED ON SOURCE PLATE POWER OF 5.24 w

in. OF Be + 3in.OF Pb + OIL

OIL ONLY—

4in. OF Be + 3 in. OF Pb+ { ft OF LiH 4 OIL

15 | 4inOF Be + 3 in. OF Pb + 2 ft OF LiH -+ OIL

 

4in. OF Be + 3 in. OF Pb + 3 ff OF LiH + OIL

10 20 30 40 SO 60 70 8O 90 100 10 120 130 140 150
.. s 7 gy, DISTANCE FROM SOURCE PLATE (cm)

Fig. 5.2.8. Thermal-Neutron Flux Traverses Beyond Configuration 69-12.

 

160

 

170

287

 
 

 

 

ANP PROJECT PROGRESS REPORT

S

2-01-057-69-282

BASED ON SOURCE PLATE POWER OF 524 w

w)

4 4in. OF Be + % in, OF BORAL
10 + 3 in. OF Pb + OIL

OIL ONLY

5 4 in. OF Be + % in. OF BORAL
+3in. OF Pb + 1 ft OF LiH + OIL

10 4in. OF Be + 1 ft OF LiH + % in. OF
BORAL + 3 in, OF Pb + 1 ft OF LiH + OIL

Vo, THERMAL— NEUTRON FLUX (neutrons/cm? - s
N

2 4 in. OF Be + ¥ in. OF BORAL
+3in. OF Pb + 2 ft OF LiH + OIL

 

-3
10
0 0 20 30 40 50 60 70 80 90 100 10 120 430 140 450 160 170

250, DISTANCE FROM SOURCE PLATE (cm)

Fig. 5.2.9. Thermal-Neutron Flux Traverses Beyond Configuration 69-13.

288
 

 

avy , THERMAL-NEUTRON FLUX (neutrons/cm? sec-w)

10|

PERIOD ENDING SEPTEMBER 10, 1956

2-01-057-69-283

BASED ON SOURCE PLATE POWER OF 5.24 w

— 4in, OF Be + 3in. OF U + OIL
" Ol ONLY——e\ ' -

4in. OF Be + 3in. OF U+ 4 ft OF LiH + OIL:

17 |4in OF Be+31n. OF U+ 2 OF LIH+ OIL

4in. OF Beé + 3in. OF U4 3t OF LiH+ OIL

40 20 30 4D . SO 60 _TO 80 90 100 40 120 130 140 150

2o + DISTANCE FROM SOURCE PLATE {cm)

Fig. 5.2.10. Thermal-Neutron Flux Traverses Beyond Configuration 69-14.

 

160

470

289

 
 

 

 

ANP PROJECT PROGRESS REPORT

| SRR
2-01-057-69-~284

5 ; - - 4in. OF Be + /2 in.OF BORAL + 3in. OF U +OIL

AQ

=~ OIL ONLY

BASED ON SOURCE PLATE POWER OF 524 w

Lo . N

Ulobl N

4in.OF Be + > in.OF BORAL + 3in.OF U
+ 1t OF LiH ¢+ OIL

Cro

- NEUTRON FLUX ( neutrons/u:,m2 -gsec-w)
w»

o

o

> THERMAL

o

- 4in.OF Be + )2 in.OF BORAL + 3in OF U + 2 ft OF LiH +OIL

1 4in.OF Be +1 ft OF Lit + Y5 in.OF L+ '3in, OF U + 11 OF LiH + OIL

5 4in.OF Be + Y5 in.OF BORAL + 3in. OF U + 31t QF LiH + OIL

 

250 DISTANCE FROM SOURCE PLATE {cm)

Fig. 5.2.11, Thermal-Reutron Flux Traverses Beyond Configuration 69-15.

290

o 10 20 . 30 40 50 60 70 80 90 100 1o 120 130 140 150 160 1707
 

 

 

4

 

PERIOD ENDING SEPTEMBER 10, 1956 fiéé’

5.3. BULK SHIELDING FACILITY
F. C. Maienschein

THE FISSION-PRODUCT GAMMA-RAY
ENERGY SPECTRUM

W. Zobel T. A. Love

The importance of obtaining information about
the gamma-ray energy spectrum and time decay
characteristics of the fission products of U235 as
related to circulating-fuel reactors has been re-
peatedly emphasized.'=3 The preliminary results
on the time decay characteristics (Phase | of the
experiment) were presented in a previous report,3
and equally preliminary results on the energy
spectrum (Phase |l) of these fission-product gamma
rays at different lengths of time after flssuon are
presented here.

The equipment used in the expernment has been
described elsewhere.4  Samples of enriched
vranium weighing ebout 2, 7, 15, and 32 mg, re-
spectively, were irradiated in the ORNL Graphite
Reactor for periods varying between 1 and 64 sec,
depending on the run. The weight cnd bombarding
time were chosen so that the counting rate during
a run did not exceed a manageable value. The
energy calibration of the spectrometer was carried

out in the usual manner, with Na22, Y88 qand
ThC” sources. Sources of Hg203, Cs137, Na22,.
Zn65, Cob9, and Na24, whose absolute source

strengths were determined with the aid of the
high-pressure ion chamber of the ORNL Radio-
isotopes Control Laboratory, were used for the
calibration of the efficiency of the spectrometer
as a function of photon energy. The efficiency
thus obtained “was the total effncuency “of the

spectrometer, ~rather _than the peak effu.nency,_
which is sometimes used - To. correct the data
for counting losses, the average count rate over -
The countlng" '

532 between 0.28 and 5.0 Mev and between 1.25
-~ and 1600 sec gave a total of 2,81 photons emitted
* per fission with a. total energy of 3.22 Mev per
fission. These values carry. an estimated error
of about 125%, -They should be compared with
the values obtained in Phase | of the experi-
ments,3 which, when corrected for differences
3w, Zobel T. A. Love, and R.- W. Peelle, ANP Quar.

the counting period was. used,: _

 

‘R. W. Peelle, T. A Love. ond F. C M:cil.enScHelln, :
"ANP - Quar. Prog Rep ]une 10 (1955, - 0RNL-1896,_ :

p 203.

2R W, Peelie, W. Zobel und T A. Love, ANP Quar.'

Prog. Rep. Dec. 10, 1955, 0RNL-20|2 p 223,

Prog. Rep Marcb 10, 1956, ORNL-2061. P 250. .

41. A Love, R W. Peelle, and F. C. Molenschem,
Electronic Instrumentation for a Multzple-Crystal Gam-
ma-Ray Spectrometer, ORNL.-1929 (Oct. 3, 1955).

 

losses for a given count rate were determined
experimentally,

The determination of the number of fissions
taking place in a sample depends on the length
of time the sample is exposed to the thermal
neutrons and the intensity of the flux. In an
experiment of this type, where the sample is moved
into and out of the flux, the bombarding times
and fluxes used in the calculations must be the
‘‘effective’ bombarding times and neutron flux.
The effective thermal-neutron flux was obtained
by exposing bare and cadmium-covered gold foils
for a nominal 32 sec at the time of each run and
computing the flux from the cadmium difference.
The effective bombarding time was computed from
the activities of gold foils, which were also
exposed for the nominal bombarding times used
in the experiment. It was found that the flux
differed from the previously used value by about
27%, Also, the bombarding times differed some-
what from the earlier ones.

The energy spectra obtained at 1.7, 6.2, ]07
40, 70, 100, 250, 700, 1000, and 1550 sec after
fission are shown in Figs. 5.3.1 and 5.3.2. They

have been separated to avoid any confusion which

‘might arise owing to the scatter of the points
‘belonging to adjacent curves. It should be pointed

out that the peaks merely represent an attempt to

‘'show some of the structure in the curves. The

error on the points has not yet been computed,
so that these peaks are not necessarily final.
They do appear on successives curves, however,
which lends some credence to their existence.
No attempt has yet been made to assign these

- peaks to specific |sotopes

An integration of the spectra in Figs. 5.3.1 and

discovered in the evaluation, were 2,92 (1£25%)

“photons per fission and 3,23 (125%) Mev per
- fission. The agreement between the two phases

is good; this is also shown in Figs. 5.3.3 and

291

 
 

 

 

-----

ANP PROJECT PROGRESS REPORT

Fig. 5.3.1. Fission-Product Photon Energy Spectrum at 1.7, 10.7, 70, 250, and 1000 sec After

Fission.

292

PHOTON INTENSITY {photons/Mev-fission -sec)

5,
N o N

5,

o

16°

167

* COMPTON SPECTROMETER
** PAIR SPECTROMETER

10.7 sec

250 sec

1.0 : 20 30
PHOTON ENERGY (Mev}

o2

kv

| S
g

ek

40

UNCLASSIFIED
2-0{-058-0-57

 

50

-4)
 

 

 

 

W

' 100 sec*

PHOTON INTENSITY ( photons /Mev - fission -sec) -

* COMPTON SPECTROMETER

2o PAIR

o 100 200 . 30
- ' . " PHOTON ENERGY (Mev)

PERIOD ENDING SEPTEMBER 10, 1956

40

UNCLASSIFIED
2—-01-058—0~-58

 

5.0

Fig.5.3.2. Fission-Product Photon Energy Spectrum ot 6.2, 40, 100, 700, and 1550 sec After Fission.

293

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

UNGL ASSIFIED
- 2-01—058—0—59

10.7 sec

PHOTON INTENSITY ( photons /Mev-fission:sec)

 

0 1 2 3 4 5 6
PHOTON ENERGY (Mev)

Fig. 5.3.3. Fission-Product Photon Energy Spectrum from Phase !l of Experiment with Superimposed
Spectrum from Phase |. ' |

294
6i4 (21

)
 

 

 

PERIOD ENDING SEPTEMBER 10, 1956

5.3.4, where cross plots of the data from one here. The values obtained in the LTSF experiment
phase of this experiment are shown with the were 4.2 (£20%) photons per fission and 4.8
appropriate points from the other phase. Another  (£20%) Mev per fission.

comparison may be made with the spectra obtained

in the experiment carried out at the LTSF.2 It

should be kept in mind, however, that there are Further refinement of the evaluation of the data
systematic differences in the evaluation of the is contemplated. It is expected that such re-
data in that experiment and the one described evaluation will show better agreement between

UNCLASSIFIED
2—01—058—0-60

1072

DECAY RATE { photons/fission-sec)

o PHASE | OF E!
° PHASE Il OF

10—3

 

1074 .
1 2 5 10 2 5 102 2 5 10 2 5 10

r, TIME AFTER FISSION (sec)

Fig.‘ 5.3.4. Decay Rote of Fission-Product Gamma Rays from Phase | of Experiment for a 0.28- to
5.0-Mev Energy Range with Superimposed Points from Phase 1.

295

 
 

 

 

ANP PROJECT PROGRESS REPORT

the different experiments and lead to a more defi-
nite answer.

GAMMA-RAY STREAMING THROUGH THE NaK
PIPES THAT PENETRATE THE ART SHIELD

T. V. Blosser D. K. Trubey

A mockup experiment®+é to investigate the in-
crease in the dose rate outside the ART lead
shield caused by gamma-ray streaming through the
NaK-filled pipes that penetrate the shield has
now included measurements beyond a duct placed
through a mockup of the ART shield at an angle of

 

sT. V. Blasser, ANP Quar, Prog. Rep. March 10, 1956,
ORNL-2061, p 249. , :

b1, Blosser and D. K. Trubey, ANP Quar. Prog.
Rep. June 10, 1956, ORNL-2106, p 274.

 
 
 
   
      
   
   
   
  

4.25-in, THICK LEAD

ALUMINUM-FILLED
DucCT

TWO INCONEL PIPES
' AIR

51 deg 30 min (Fig. 5.3.5). The duct consists of
two concentric Inconel pipes (separated by a thin
air ‘layer) filled with aluminum turnings and fitted
with a tungsten collar at the top. The axis of
the duct in the horizontal plane was at a 45-deg
angle from the x axis. For this particular arrange-
ment a specific coordinate system using x’ and
y” axes was adopted to designate the points of
measurement. The x’, y’ axes were at an angle
of 45 deg to the x, y axes as shown in Fig. 5.3.6.

. Measurements were made beyond the ducts along

several x’ traverses (labeled A through E in
Fig. 5.3.6) for two z distances (10.5 and 20,5 cm)
from the lead shield.

The ratio of the dose rate beyond the water-filled
hole (that is, with the duct not in position) to the

UNCLASSIFIED
ORNL-LR-DWG 16200

M LINER

TUNGSTEN COLLAR

51deg 30min

Fig. 5.3.5. Mockup of an ART Duct Penetrating the Lead Shield ot an Angle of 51 deg 30 min.

296 GLs

#}
 

‘dose rate beyond the solid lead shield is plotied
in Fig. 5.3.7, as well as the actual dose rate
beyond the solid lead. (For convenience -in
plotting, the curve actually represents twice the
measured dose rate.) The ratios of the dose rafes
beyond the aluminum-filled duct to the dose rate

.40

20

PERIOD ENDING SEPTEMBER 10, 1956

beyond the solid lead are shawn in Figs. 5.3.8
and 5.3.9. Again the actual dose rates beyond
the solid lead are plotted on the lower portions
of the figures.

The reduction in dése due to tilting the duct is
about a factor of 4.

This indicates that the

UNCLASSIFIED
ORNL-LR—-DWG 16868

 

DISTANCE ALONG y AXIS (cm)

- 10

-20.

 

-0 L—

=40 © ~10

 

O POINTS OF MEASUREMENTS

 

40

10

0 20 30

 

 

" 'DISTANCE ALONG ¥ AXIS (cm)

= Fig. 536Coordmcrte Sysiemfor ART 51-deg 30-min _Dfic_f Mockup Tests. ;

297

Fa

4

g

 
 

 

 

 

 

.ANP PROJECT PROGRESS REPORT -

ENET .
ORNL—LR~DWG 16204

S

oC,y'=0

oD, y/=Tcm

S E, y'= 4icm

AA y= -141em
AB, y'==Tcm

TO DOSE RATE
BEYOND SOLID LEAD SHIELD
W

. RATIO OF DOSE RATE
BEYOND WATER-FILLED HOLE .
N

 

-/hr)
o -

DOSE RATE (x2) BEYOND

SOLID LEAD SHIELD {r,

 

X’ {cm)

Fig. 5.3.7. | :qumu;Ruy Dose Rates Beyénd a
Water-Filled Hole Penetrating ¢ Mockup of the
ART Shield at an Angle of 51 deg 30 min.

HSSRTT
ORNL—LR—-DWG 16202

10
P o C,y/=0
e ® D, y’~Tcm
Fws 5 8 E,y/=14.4¢m
n:§t_-'1 A A y'=—{41cm
gee ABy=-T7
B
30
3g 2
o
£2%
cPm
w 1

 

7x10':

DOSE RATE (x2) BEYOND
SOLID LEAD SHIELD (r/hr)

 

x'(em)

Fig, 5.3.8. Gamma-Ray Dose Rates Beyond

Aluminum-Filled Ducts Penetrating a Mockup of

the ART Lead Shield ot an Angle of 51 deg 30 min
(z = 10,5 cm).

208 644

-20 -0 0 10 20 30 40 SO

mockup gamma-ray source has an approximate
distribution proportional to cos?® 6. This is in
agreement (within' the uncertainty) wnh the esti-
mate of the distribution reported previously.b It
must be remembered, however, that the source
distribution of the ART will probably not corre-
spond to that of the mockup source, even though
the angle of penetration is the same as that of
the mockup reported here. That is, the fuel in

the heat exchanger is directly in line with the
" axis of the duct in the ART, but the fBurce of
“the mockup is believed to be a cos2. @ distribution
about ‘the normal to-the lead. For this reason it-
“would probably be more realisfic to apply the

previously reporfed2 results - for the -straight-
through  penetrations, which mocked . up ‘portions
of the north-head ducts.

Some measurements have been made beyond -
. straight-through penetrations mocking up portions

of the south-head ducts, and the results will be
published later. The effects of various com-
ponients in the mockups and the effects of patches
are now being investigated,

b '
ORNL—LR—DWG 16203

 

 

Wwg 10
<d 3
|-.'.’u:,_E C,y'=0"
§=_,-lcn D, y=7cm
woo 5 RE, y/=144cm
mog
Qal AA y=~-141cm
2 ;
Fo AB, yr=-Tem
b53
8o
oa® 2 '
—p. 0
a8z
30
> 0
]
oo !
=3
&
@3
oY
x5
b
&Y
w o
g3
—20 -$Q ] 10 20 30 40 50
x* {cm) -

Fige 5.3.9. Gamma-Ray Dose Rates Beyond
Aluminum-Filled Ducts Penetrating a Mockup of
the ART Lead Shield at an Angle of 51 deg 30 min
(z = 20.5 cm), '

 

 

 

-}

-]
 

 

 

e

o e,

.

PERIOD ENDING SEPTEMBER 10, 1956

. - 5.4, SHIELD MOCKUP CORE

‘ C. E. Clifford
The design of the Shield Mockup Core (SMC),’

a 5-Mw fixed-fuel reactor proposed for shielding
studies of the circulating-fuel reflector-moderated
reactor (CFRMR), has been gltered somewhat as
dictated by nuclear and shielding calculations per-
formed by the Engineering Physics Group at
Pratt & Whitney Aircraft.2: Other changes also
became necessary for engineering or ecbnomicul
reasons.

The SMC has l:een des;gned so that the leakage

fluxes, with the exception of the fission-product

radiation from the heat exchanger, will be the

..same as those from s CFRMR (PWAR-1) now

contemplated for aircraft propulsion.. The core

will consist of 204 modifie_d'M_TR-type fuel plates

 

lc. E. Clllford and L. B. Holland ANP Quar. Prog.
Rep. June 10, 1956, 0RNL-2ID6, P 279

Calculohons performed by F. C. Mornman and J. B,
Dee, Pratt & Whitney Alrcraft.

L.. B. Holland

placed radially around the island reflector. The
shape of the plates will be such that when they
are all in position the geometry of the CFRMR
will be simulated. Numerous calculations have
been performed to compare the SMC with the
CFRMR, and each deviation from the CFRMR
design was considered. A summary of the calcu-
lations which [ed to the current SMC des;gn
(Flg. 5.4.1) is given below.

SE LECTION ‘OF CALCULATIONAL MODELS

The nonregular geometry of the CFRMR core,
which is somewhere between a sphere and o
cylinder, necessitated the use of several cy-
lindrical and - spherical calculational models to
determine the upper and lower limits of the critical
mass. The validity of the nuclear models was

checked by using them in calculations of the

critical mass of the ART high-temperature critical
experiment. As shown in Table 5.4.1, the results

'» - TABLE 5.4.1. COMPARISON OF CALCULATED AND EXPERIMENTAL CRITICAL MASS AND VOLUME
OF THE ART HIGH-TEMPERATURE CRITICAL EXPERIMENT

 

*

Critical Mass Core Voluhe

 

ond acwal outer radivs of reflector, core rodlus de- ‘

' e iermmecl by welghflng ‘that volume of the core which
:l;’ffidev:otes frorn ‘a cyllncler** by a cosune funcflon and

oddmg to the cylmdrlcol core. volume (cnhcol mass -

_ 'defermmed with uctual core volume)

- {kg) (liters)
E_xperinlentcl 18.2
_Calculated
Spherical models
Equatorial crosscut’ , S : 176 64
‘Equal core volume, ocfuol micl-plone tl'uckrless of lslcmd, . 18.6* _ 95.2
and actual mld-plane outer dmmeter of feflecfor
- Equal volumos for all reglons - o 21.8 o 96.3
Cylindncal models ) :
Equutonal crosscuf S . LT 29.6
Equal core volume, acfuul msd-plane tlnckness of island 22.8
. and actual mld-plane outer diameter of reflector -
Equal importance:’ octual mtd-plone tl'm:kness ol |slond - . 18.7*

 

- *Models used in SMC calculutions.
**That is, region 5 in Fig. 5.4.7.

299

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

SUPPCRT ASSEMBLY

  

REMOVABLE

  
 
 

IONIZATION
CHAMBER

REMOVABLE PLUG —

. SECTION B-B

 

101234567
el v

SCALE IN INCHES

LEAD LIFTING HOOK

REMOVABLE PLUG

LEAD

 

 

 

 

 

 

 

 

    
  

 

 

 

 

   

CONTROL ROD DRIVE

 

   

 

  

 

-
2-01-059-79

 

 

 

 

 

—
FUEL PLATE SPACERS GRAPHITE ISLAND
S
::\‘\ “ N :\
N R COOLING NN
NN LINE NN
' NN
SRR
-.\\\‘\\
SN
CONTROL, RN e
L ‘\\‘\\\\\‘_ S
CHAMBER AR FyEL PLATES
. N \\\\\\.\\\\ 0 \\‘ A
P2 SECTION A-A
Y
IR
N 101234567
g COOLING LINE I
§ SCALE N INCHES
“fi‘
§ GRAPHITE ISLAND
N FUEL
BERYLLIUM
BORON LAYER
INCONEL
SALT REGION
INCONEL
BORON LAYER :
INCONEL (CFR PRESSURE SHELL)
INCONEL
NEUTRON SHIELD
CONTAINER (ALUMINUM)
' 2 3 4

 

SCALE IN FEET

Fig. 5.4.1. The Shield Mockup Core (SMC).

300

)
 

 

 

of both a spherical and a eylindrical model were

in agreement with the experimental critical mass.
Both a continuous slowing-down model and a
Goertzel-Selengut slowing-down model were used
to account for the energy degradation of neutrons
in the light water used as the coolant in the core.
The agreement was good enough to permit the use
of the continuous slowing-down model for para-
metric studies to determine a design point.

CALCULATIONAL PROCEDURE

The calculational procedure first inciuded a
determination of the fuel loading for a critical
condition of a given core configuration by a

multigroup machine calculation, Parametric studies

were then carried out to match the core power
distribution and neutron-absorption distribution in

the beryllium of the SMC with those in the CFRMR.

Once these distributions were in agreement the
contributions of the various radiation sources in
both the SMC and CFRMR to the total dose in the
crew compartment were calculated,”  The SMC
configuration which gave the proper dose contri-
butions was then selected.

In the method used to determine whether the
doses from the gamma-ray sources in the SMC

and the CFRMR were matched, both reactors were

divided into 35 spherical shells, and the dose rate
from each material of each shell at a point 50 ft
from the reactor was calculated. The total contri-
bution from each material in a particular region

was then obtained by adding the dose rate from .

that material in each shell in that region, The
calculations were based on the neutron absorptions

from the critical calculations and published data®
on capture gamma rays. .In the first calculations
an average capture gamma-ray energy was assumed,
but: in later colculatlons a . capture gamma-ray
spectrum was approxlmated by using the ap-

propriate number of 2-, ;6= 7. and  9-Mev

photons indicated for a parhcu!or material.  Since
the operating temperature of the CFRMR will be
~ approximately 1300° F higher than that of the SMC,
. and ' therefore the thermal-neutron "cross section -
will be much smaller, it was necessary fo match’
‘the neutron absorptions, rather than the thermal-
neutron flux, in the beryllium and other materials

 

3P. S. Mittleman and R. A. Liedtke, Nucleénics
13(5), 50 (1955).

tolerated.

PERIOD ENDING SEPTEMBER 10, 1956

of the two reactors to obtain the proper capture
gamma-ray distribution,

The fission neutrons which contribute to the
neutron dose in the crew compartment and thus
are of concern to the shield designer are those
born at high energies. (Fast neutrons thermalized
in the core or reflector will not pass through the
boral ¢urtain outside the reflector. They con-
tribute to the dose in the crew compartment through
the capture gamma rays which they produce in the
core-reflector region.) It is assumed that if the
removal cross sections of the materials in the SMC
core are very nearly the same for fast neutrons as
the removal cross section of the fused salt fuel in
the CFRMR, that if the normalized core power
distributions are matched, and that if the regions
outside the core are correctly mocked up in the
SMC, the number and energy spectrum of neutrons
leaving the SMC shield should be the same as
those leaving the CFRMR shield.

CALCULATIONS FOR SMC CONFIGURATIONS
USING UO,-STAINLESS STEEL FUEL
ELEMENTS

It was assumed originally that the CFRMR
leakage fluxes could best be mocked up in the
SMC with UO_—_stainless steel fuel elements
cooled with D20. The results of early calcu-
tations confirmed that a better match of the core

power distribution and the neutron absorptions in

the beryllium could be obtained if D20 rather than
H O were used (Figs. 5.4.2 and 5.4.3).4 This
resulf was to be expected, since the moderation
of neutrons by D20 is closer to that of the fused

~ salts than is the moderation by H20. However,
. the use of a closed D_O system in an H,O storage
pool at the TSF or the BSF posed many handling

problems and a safety problem that could not be
It became apparent from a plot of the
critical mass of the SMC for all mixtures of D,0

~and H 0 (Flg 5.4.4) that a supercrmcal condition
--could arise almost instantaneously if H20 were

inadvertently pumped into the D,O-cooled core

‘as a result of component failure or mishandling.

It was therefore decided that H20 should be used
as the coolant in the SMC.

 

4All curves are normalized to one fission per cubic
centimeter in the core region.

301

 
 

 

 

ANP PROJECT PROGRESS REFORT

orcren

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

7 2-0t-059-98 -
3.0x407
FUEL ELEMENTS: U0,—STAINLESS STEEL WITH
UNIFORM FUEL DISTRIBUTION.
N ’0-7 /’\
_ / N

©
O
k SMC, Be + Al, 100% D,0 IN VOIDS
3 .
W
& 20x107 _
& /\ \ PBW CFRMR, Be + No |
%
=
o :
5 ' N s SMC, Be ONLY, {00 % D,0 IN VOIDS
& 1.5%407
& / i
S ‘
£ | / PBW CFRMR, Be ONLY
o e ] \
QO .
L72]
3 #/ ><
2 1.0x410-7 3\
o / - K\
m .
- .
=
W
z /

0.5x10-7 \\

SMC, Be + Al, 100 % H,0 IN VOIDS \
! k
~ SMC, Be ONLY, 100 % H,0 IN VOIDS _ '
x10-7 -
35 40 45 50 55 60 65 70

SPACE POINTS IN REFLECTOR (SPHERICAL MODEL)

Fig. 5.4.2. Neutron Absorptions in Reflector Region of an SMC Using UO,.Stainless Steel Fuel

Elements and Hzo or DZO Coolant.

Because of the excessive moderation of neutrons
in the core by H20 and the consequent reduction
of neutron absorptions in the reflector of the SMC,
it was necessary to devise some means of reducing
the amount of H,O in the core. A parametric study

~ was therefore carried out to determine the effect

of substituting aluminum (by means of wedges
between the fuel plates) for some of the H,LO. As
shown in Fig. 5.4.5, the combination which re-
sulted in a neutron absorption distribution in the
SMC reflector that most nearly matched that in
the CFRMR reflector was 25% H,O and 75%

302

aluminum in the spaces between the fuel plates.
[The calculations were for absorptions in the
beryllium, aluminum, and H20 (ref 5) in the SMC
reflector and for beryllium and sodium in the
CFRMR reflector.] The agreement for the region
near the core-reflector interface is poor, but this
difference can be minimized by adjusting the
amount of aluminum ‘in the SMC reflector.

 

5At the time these calculations were performed it was
felt that it would be necessary to cool the beryllium
with H,0. ' :

A}

v}
 

 

 

PERIOD ENDING SEPTEMBER 10, 1956

ST
2-0t-059-91

 

 

2.4

 

P&W GFRMR

 

 

2.0

 

/
/
/

 

o

/[

 

 

)
/
/

 

/7
/

 

H,0, CONTINUOUS SLOWING DOWN

 

/
1
|

 

/
\\/

 

N

 

POWER DENSITY (NORMALIZED TO 1 fission/cm>)

.

Ho,0, G-5 SLOWING DOWN

 

o
»

 

 

 

 

 

 

 

 

 

 

 

15 20 25 30

35 40 45 50 55

SPACE POINTS IN CORE (SPHERICAL MODEL)

Fig. 5. 4.3. Power Distribution in an SMC Usnng UO,-Stainless Steel Fuel Elements and H 0.D0,0

Coolant.

In order to further investigate the effect of
substituting aluminum for H,O in" the core, the
power distribution in the core was determined for -
each H o O-aluminum_ combination (see Fig. 5.4.6).
Once ugam ‘the power distribution for the 25%

H,0-75% aluminum’ combination” most nearly ap-

be attributed to-the fcct that the rodml arrangement

of the SMC “fuel plates will cause areduction -in

the uranium density with increasing radlus, where-
as the homogeneous fused salts in the CFRMR

‘will give a uniform uranium density. |f the uranium

loading in the SMC fuel plates is varied so that

only the outer regions of some of the plates
_contain fuel, as in Fig. 5.4,7, for example, the
fuel density in the core will become more nearly
- constant. For calculational purposes the core was
- proached that of the. CFRMR, . but the agreement =
was net good ‘enough. It was felt that this could -

divided into five .concentric regions and the fuel

':"‘Ioadmg in_ the outer three .regions was varied.
- As shown in Fig. 5.4.8, the power density which
- most nearly approached that of the CFRMR was

for @ 1:1:1:2:3 fuel distribution. A comparison
of neutron absorptions in the reflector of the

303

 
 

 

LANP PROJECT PROGRESS REPORT

 

 

 

 

 

 

SECRET
2-01~059-92
Ha0 (%)
{00 90 80 70 60 50 40 30 20 10 o
9
g 8
=
S
>
o
<
E7
o /
o
g
T
4 ' /
= 6
2 /
T
O .
5
.4

 

 

 

 

 

 

 

 

 

 

 

 

0O 10 .20. 30 40 50 60 70 80 90 400
0,0 (%)

Fig. 5.4.4. Critical Mass of an SMC Using uo,-
Stainless Steel Fuel Elemenfs uqd D20-H 20 Coolant.

SEoREP
_7 2-01-059-TTA

 

"1 SMC, 12.5% H0, B7.5% Al
-P & W CFRMR
- SMC, 25% H,0, 75% Al

2x10°7

 

TN N

TV 7
VI
i

~
N

—_—

07

 

 

 

 

SMC, 50% H,0, 50% Al

| SMC, 100% H,0

o L

3 40 45. 50 85 60 65 70
SPACE POINTS IN REFLECTOR [SPHERICAL MODEL)

NEUTRON ABSORPTION (FRACTION/cmOF REFLECTOR) ‘

 

 

 

 

 

 

Fig. 5.4.5. Fraction of Neutron Absorption in

. Entire Reflector Regien (Be + Al + H,0) for

Various Core Configurations of en SMC Using
UO,-Stainless Steel Fuel Elements and H,0
Coolant. ' '

orCRTT
2-04-059-76A

i
@©

n
>

SMC, H,0
SMC, 50% H,0, 50% Al
SMC, 25% Ho0, 75% Al

ro
o

SMC, 12.5% H,0, 87.5% Al ‘

POWER DENSITY
(NORMALIZED TO { FISSION/¢m’)
N

-~
n

o
@

8 W CFRMR

o
>

 

o

12 16 20 24 28 32
SPACE POINTS IN CORE (SPHERICAL MODEL)

Fig. 5.4.6. Power Distribution in Various Core
Configurations of an SMC Using UOz-»Stuinless
Steel Fuel Elements and H,0 Coolant.

SMC using this configuration with those in the
reflector of the CFRMR also showed gocd agree-
ment.

On the basis of the above results, the first
shielding calculations, that is, the first calcu-
lations of contributions to the gamma-ray dose
by the various regions of the reactor, were per-
formed for an SMC using UQ_~stainless steel fuel
elements with a 1:1:1:2:3 fuel distribution and
H,0 as the coolant filling 25% of the voids
between the fuel plates. It immediately became
apparent that for the lower operating temperature
of the SMC the capture gamma-ray production in
the Inconel core shells was too high. Since the
shell at the core-reflector interface was the im-

portant contributor, succeeding calculations were

made for an SMC with a half-thickness Inconel
shell at that interface. The results of these
calculations are shown in Table 5.4.2 for two
different shield configurations,

The effect that the thinner shell had on the
power distribution was then calculated. As can
be seen in Fig. 5.4.8, the results were in closer
agreement with the power distribution for the
CFRMR than any of the other SMC configurations.

ol

“w

)]
 

 

-

 

 

 

 

r

 

 

 

NOTE:

REGIONS 4 THROUGH 5 INDICATE THEORETICAL
DIVISIONS OF THE CORE VOLUME FOR NUCLEAR
CALCULATIONS S

      
 

PERIOD ENDING SEPTEMBER 10, 1956

S EGRE
2-04-059-100

 

 

 

 

 

II

[~
L/ u-a1

NOT TO SCALE

Fig. 5.4.7. Proéosed Uranium Loading of SMC Fuel Plates.

When ' the - neutron absorptions in the reflector
regions also corresponded (Fig. 5.4.9), it was felt
that the best configuration for an SMC using UO -

 stainless steel fuel elements had been attained.
At the same time, it was realized (Table 5.4.2)

that the capture gamma-ray dose rate from the

stainless - steel in the fuel elements was ex-
cessive, even though the first calculations were

made for a single capture- gamma-ray energy for
each element rather than for an energy spectrum.
It was then decided to convert to uranium-aluminum

fuel elements and to repeat all the calculations

) usihg the spectrum of capture gammaeray energies

previously indicated for the dose rate calculations.

'CALCULATIONS FOR SMC CONF IGURATIONS
USING URANIUM-ALUMINUM FUEL ELEMENTS
The use of uranium and aluminum in the core
will necessitate a thicker fuel plate design;

however, the thickness of the aluminum wedges
between the plates will be adjusted so that the

305

 
 

 

 

ANP PROJECT PROGRESS REPORT

amount of H,O in the core region will remain
the same and it will continue to be referred to
as a 25% H,0-75% aluminum case.

The power distribution in an SMC configuration
using vranium-aluminum fuel elements is shown
in Fig. 5.4.10 for three different fuel distributions.
The 1:1:1:2:3 distribution again gave the best
agreement with the CFRMR. This distribution was

then used in the calculations of the neutron ab-
sorptions in the reflector region (Fig. 5.4.11). A
comparison of Fig. 5.4.11 with Fig. 5.4.9 shows
that the substitution of aluminum for stainless
steel in the core will not appreciably change the
number of neutron absorptions in the reflector.

In view of the agreement in the nuclear calcu-
lations, shielding calculations were then performed

'TABLE 5.4,2. COMPARISON OF CALCULATED GAMMA-RAY DOSE RATES? FROM CORE-REFLECTOR
- REGIONS OF THE CFRMR AND THE smc? usING UO,—STAINLESS STEEL FUEL ELEMENTS |

 

Dose Rate at 50 ft (r/hr)

 

32-cm Alkylbenzene

5.in,-Pb=43.cm-Alkylbenzene

 

 

 

 

 

 

 

 

 

 

 

! * ' -
Regwn Material Reactor Shield (No Pb) Reactor Shield .
SMC CFRMR SMC ‘ CFRMR
Isla n_d-co;'e shell Inconel® 332 96.17 0.2330 0.0544
H,0¢ 11.37 0.00079 |
Sodium 0.1377 0.0002 ‘
Subtotal 343.0 96.0 0,2338 0.0546
Core Stainless steel 1.785 1.8969
Hzo 11.53 0.0169
Aluminum 417.9 0.442
Fuel 3,385 3979 6.456 7.588
Subtotal 5,599 3979 8.8118 7.588
Corereflector shell Inconel 1,570 1454 0.8100 0.7714
H20 0.54 0.0166 .
Sodium 2.017 0.0029
Subtotal 1,571 1456 0.8266 0.7743
Reflector Berylliuvm 4,238 3868 4,272 3.883
Aluminum 1,708 1.735
Sodium 262.8 0.382
Subtotal 5,946 4131 6.007 4.265
Total 13,459 9662 15.879 12.682

 

'aAn average capture Qummo-roy energy was assumed.
bSMC configuration includes UO —stainless steel fuel plates with 1:1:1:2:3 fuel distribution, 25% H,0 and 75% alu-
minum in core voids, a half-thlckness Inconel shell at the core-reflector interface, and a ful|-fh|cl<ness Inconel shelt

at the island-core interface.

€A comparison of the capture gamma-ray production in the individual elements of Hastelloy X, which will be used
in the CFRMR, and Inconel which is more available and inexpensive for the SMC, showed that there was sufficlent

agreement to justify using Inconel in all the calculations.

dSmull amount of water in the tolerance region between the fuel plates and the Inconel shell.

306

&)

)
 

PERIOD ENDING SEPTEMBER 10, 1956

o - cconen
: : 2-01-059-~-93

¥
Py

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

3.6 R ;
g W ;
1 ;
\ - FUEL :
3.2 \ DISTRIBUTION :
Ty — om0 _-:
\ \ — —00:2:2 :
= ' . \ ‘\ — e wm:]:1:3:3 :
o~ 28V \ . — w—{{::2:3 F /
| nE '._‘ \ -—ere ma{:]:2:2:3 .
- < ‘f}.\ A : \ ] eeesessener §24:1:2:3 WITH ST
. < -._\'-,\‘ \ Y , _ HALF-THICKNESS [ s j
@ [N \ : INCONEL CORE SHELL ; / ;o
- N : ' s /
= 24 N : 7y
o Wl W g / 7
- . \ ."’..\}\ ‘\‘ \}“ . ‘-.. Z /fl'... /
o 'o. . ' ) - ) * ..
w .o.\.-:n \ ..-./ *
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q - ., “ > [
- s "N \. N/
. § b, .'.‘v\ :.’ / /..-. 2 *
2 \\s...\ T . AT
-'.: 1 6 <b.:'.?\\ —— / '/ .
.m \ ..'o u - oo, ' - s
E O . b - ‘ .\‘.. 7 7/ .
e — : e P . 7 -
E:J- P, :...""c s a e o e o e >( '---4—..4" ”
= . . " e, e — e lt
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. v e, "*0d b, -
~ = P& W CFRMR
o V . poss ..."'--‘.
0.8 - N s
S SPACE BETWEEN ELEMENTS |
- ' |- " CONTAINS 25 %5 H,0, 75% Al
' o4 L — L || | A .
- 17 19 21 23 25 27 29 39 33 35 37
° T o o . SPACE POINTS IN CORE (CYLINDRICAL MODEL) '

F|g. 5.4. 8 Power Distribution for Vurious Fuel Distributions in an SMC Using UO -Stmnless Steel
Fuel Elements and H O Coolant.

THReRT.
2-01-059—99

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

’ e e ':é'Mc:B + Al |
= R - | ——SMC, Be - o7
g 2_5 e —— < L .7 7
. - w / R P & W CFRMR, Be +Na | SMC: STAINLESS STEEL-~UQz FUEL PLATES
g N . - WITH 14:1:2:3 FUEL DISTRIBUTION,
= 2.0 - - — -~ 25% HpO AND 75 % Al IN CORE vOIDS, —
w o f o \ N HALF ~THICKNESS INCONEL SHELL AT
m SMC, Be ONLY~, | . : CORE-REFLECTOR INTERFACE .-
Z 15— . _— -
5 | \ \\ 7 S 0
E 0 e —— —— - \ L
2. : - \<p 8 W CFRNR, Bo ony | —— N :
* - S - - - . : ' \\
8 ost — eda R o~
40 a5 50 55 ) 65 70 75 80 85 90
| SPACE POINTS IN REFLECTOR {CYLINDRICAL MODEL)
- ' Fig. 5.4.9. Neutron Absorphons in Beryllmm Reflector Region of an SMC Using UO 5-Stainless Steel

Fuel Elements and H O Coolant.
307

 

 
 

 

ANP PROJECT PROGRESS REPORT

-POWER DENSITY (NORMALIZED TO { fission/cm?)

GEeRET
- 2-04—-059-90

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

3.6
FUEL
3.2 DISTRIBUTION ‘
-— e wwe {:1:1:2:2
— e {:{:1:3:3 -
. . P S 1_‘1:’:2:3
2.8 —t—1 /
' SMC: U—Al FUEL PLATES, HALF-THICKNESS INCONEL SHELL /
AT CORE-REFLECTOR INTERFACE, Al SHELL AT CORE- .
ISLAND INTERFACE, AND 75 %, Ai—25 % H20 IN VOIDS / )
24 : _ .
\\ ' i / ,/
\ : / -
2.0 "\\\\ ' /‘./ ,'.’/
\. e, |a aset © ¢ ’
.. \ \\.\ ; P
. .'- -v’ ””
. \w . = ____’ / »
1.6 - /
. n — . ._7/.—
[~ L7
[*
1.2 - ‘ -...,,-.‘.-' :-..-- f{\/
. Sy
T~ - ~——P & W CFRMR
'-o--.__._- o
0.8° [ " =
0.4

 

 

 

 

 

 

 

 

 

 

 

17 19 2t 23 25 27 29 31 33 35 37
SPACE POINTS IN CORE (CYLINDRICAL MODEL) \‘

Fig. 5.4.10. Comparison of Calculoted Power Distributions in the SMC and the P & W CFRMR.

L
2-04-059-97

 

3520077

 

3.0x 4077
SMC: U—AL FUEL PLATES WITH £ 4:4: 2; 3 FUEL OISTRIBUTION, HALF-
THICKNESS INCONEL SHELL AT CORE-REFLECTOR INTERFACE, Al
SHELL AT CORE-ISLAND INTERFACE, AND 75% Al=-25% H0 IN VOIS

N |
1524077 \kff‘ F & W CFRUR, Be +} No |
""" = . % K SMC, Be + Al
_q - -

_ NS
P & W CFRMR, Be WLY/,%QK .

—)
SMC,, Be ONLY \

40 45 ' %0 55 60 65 70 75 80 8
SPACE POINTS IN REFLECTOR (CYLINDRICAL MODEL)

 

2.5x0°7

 

20x40°7

 

 

 

 

 

\
/

 

 

NEUTRON ABSORPTION (FRACTION/em® OF REFLECTOR)
o
o
-
aI
-~

 

 

 

 

 

 

 

 

 

 

 

Fig. 5.4.11. Comparison of Calculated Neutron Absorption in Reflector Region of the SMC and the

P & W CFRMR. :

308

C

"

&)

r

o)
 

 

for an SMC using this configuration, that is, using
vranium-aluminum fuel elements, a 1:1:1:2:3 fuel
distribution, H,O as the coolant filling 25% of
the voids between the fuel pldtes, and a half-
thickness Inconel shell at the core-reflector inter-
face. In addition, a full-thickness aluminum shell
was substituted for the inconel shell at the core-
island interface, since the contribution from that
shell had been too high in the caleulations for
the earlier configurations. (The total contribution

PERIOD ENDING SEPTEMBER 10, 1956

from this shell was only on the order .of 1%,
however.) The results of the shielding calcu-
lations, which are for two shielding configurations,
are presented in Table 5.4,3 and discussed below.

Core Gamma Rays _

The gamma rays from the fused salts in the
CFRMR will be mocked up in the SMC by fission
gamma rays from the uranium plus capture gamma
rays from the aluminum,

TABLE 5.4.3, COMPARISON OF CALCULATED GAMMA-RAY DOSE RATES? FROM CORE-REFLECTOR
REGIONS OF CFRMR AND SMCb USING URANIUM-ALUMINUM FUEL ELEMENTS

 

Dose Rate at 50 ft (v /hr)

 

32.cm Alkylbenzene

5-in.-Pb—43-¢:m-A|kylbenzene

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Region Material Reactor SHie_ld (No Pb) Reactor Shield
SMC CFRMR SMC CFRMR
{sland-core shell Inconel€ 101 - 0.0806
Aluminum 31.6 0.0384 '
H,0% 2.0 0.0021
Sodium 1 0.0018
Subtotal 33.6 102 0.0405 0.0824
Core Uranium 1141 1525 1.823 2,399
Aluminum 360.9 0.509
H,0 66.9 0.097
" Subtotal 15468.5 1525 2,429 2.399
Core-reflector shell . inconel 1938 1453 1.516 1.141
H,0 13.9 0.020
Sodium 13.01 0.020
Subtotal 1951.9. 1466.01 1.536 1.161
| ' Reflector Beryllivm - 3745 4125 4.493 4,959
' - Aluminum - 1894 2.666
Sodium | 1634 2,548
o Su_bt_qtralr 563? 5759 7.159 7.507
Total 9193 8852 1117 1115

 

aThe gcmma-ray spectrum wos opproxsmcted by using the uppropnate number of 2., 4-, 6-, 7-, and 9-Mev phofons ine

.i';i'dlcufed for a purticular maferml.

SMC ‘configuration mcludes U-Al fuel plufes wlth 'I 1:1:2:3 fuel dlsfnbuhon, 25% H20 and 75% aluminum in core
_vouds, a half-thlckness Inconel shell at core-reflector interface, and an aluminum shell at island-core interface.

CA comparison of: the copture gumma-ray production in the individual elements of Hastelloy X, which will be used
in the CFRMR, and {nconel, which is more availoble and inexpensive for the SMC, showed that there was sufficient

agreement to justify using Inconel in all the calculations.

.dSmaII amount of water in the tolerance region between the fuel plates and the Inconel shell.

309

 
 

 

 

| Neutron-Capture Gamma Rays from the Core Shells

The capture gamma rays from neutron absorptions
in the two Hastelloy X shells at the island-core
and core-reflector interfaces of the CFRMR will
be mocked up by neutron absorptions in the half-
thickness Inconel shell and the full-thickness
aluminum shell of the SMC, respectively. The
calculations indicated that the dose from the half-
thickness Inconel shell was still too high, but
from the viewpoint of fabrication it was not
- feasible to reduce the shell thickness further.

Neutron-Capture Gamma Rays from the Reflector

‘The beryllium reflector of the SMC will be
almost an exact copy of the CFRMR beryllium

reflector, but this will not ensure the proper

mockup of the gamma-ray sources in the reflector
region, since the temperature difference again
will affect the gamma-ray source through the
thermal-neutron cross section. The thermal-
neutron flux required to produce the correct number
of absorptions in the reflector will be much lower
in the SMC. As mentioned above, the flux in the
SMC beryllium and the power distribution in the
core can be adjusted by varying the amount of
moderation in the core region (see Fig. 5.4.5).
The total gamma.ray dose rate from the SMC
reflector will be adjusted by the amount and
distribution of aluminum used to mock up the
sodium coolant in the CFRMR reflector (see Figs.
5.4.9 and 5.4,11),

Gamma Reays from the Heat Exchanger Region

In the CFRMR the most important source of
gamma rays from regions outside the reflector will
be from fission-product gamma rays from the
circulating fuel in the heat exchanger. Since
the SMC will use fixed fuel the fission-product
source will not be mocked up, although the heat
exchanger region will be (that is, with Inconel,

N

NaK, and fused salts), so that the proper attenu-

ation of core radiation will be effected. Gamma -

rays resulting from neutron captures in the heat
exchanger region will be reduced to an insig-
nificant level, as far as the crew compartment
dose is concerned, by boron curtains on each side
of the heat exchanger in the CFRMR. These boron
curtains will be mocked up in the SMC.

Neutron-Capture Gamma Rays from the Pressure

Shell and Shield

Gamma rays resulting from neutron captures in
the Inconel pressure shell and the lead and water
shield will also contribute to the crew compartment
dose. The boron curtain outside the heat ex-
changer will reduce the number of capture gamma
rays from the outer Inconel shell. The relative
importance of the capture gamma-ray sources in
the lead and water is essentially a function of
the lead thickness. For the CFRMR lead shield
now contemplated the total contributicn of capture
gamma rays from the lead to the dose will be less
than 5%,

Neutron-Capture Gamma Rays from the Air
and Crew Compartment

The gamma-ray dose in the crew compartment
resulting from neutron capture in air and the crew
compartment materials is a function of the
spectrum of neutrons leaving the reactor shield.
Since this spectrum will be mocked up in the SMC
these capture gamma-ray sources should be correct.

DESIGN STATUS

At this point the nuclear calculations are con-
sidered to be final enough to allow complete
detailed design of the reactor core~reflector-island
region. When the mechanical designs are com-
pleted, final nuclear calculations will again be
made, both by Pratt & Whitney and ORNL, to
check the nuclear model with the mechanical
design.

 

. i !
310 Sffl =

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OFFICIAL. USE ONLY

AT

THE OAK RIDGE NATIONAL LABORATORY

October 1, 1956

 

 

 

 

 

 

THE AIRCRAFT NUCLEAR PROPULSION PROJEGT

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

|
\
ANP PROJECT DIRECTOR W. H. JORDAN RD b
CO-DIRECTOR 5. J. CROMER RD
ASSISTANT DIRECTOR A. J. MILLER RD
D. HILYER, SEC. RD
REPORTS
PRATT & WHITNEY AIRCRAFT A W. SAVOLAINEN ARE
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S. J. CROMER, DIRECTOR RD G CoMMIT ‘ W. H. JORDAN .
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W. K. ERGEN ;
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W. R. GRIMES .
G. W. KEILHOLTZ ;
ASS ;
ISTANT TO DIRECTOR ¥ D MANLY :
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A M. PERR |
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H. W. SAVAGE
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J. A SWARTOUT
A. M. WEINBERG |
|
i
b
EXPERIMENTAL ENGINEERING CHEMISTRY
H. . SAVAGE ARE ! W, R, GRIMES, STAFF ASSISTANT MC "
i
|
| . .
b 1
}
i ,
: NOTE: THIS CHART SHOWS ONLY THE LINES OF TECHNICAL COORDINATION OF THE ANP PROJECT. ]
POWER PLANT ENGINEERING THE VARIOUS INDIVIDUALS AND GROUPS OF PEOPLE LISTED IN THIS AND THE FOLLOWING CHARTS | APPLIED NUCLEAR PHYSICS T
AP, FRans ARE ARE ENGAGED EITHER WHOLLY OR PART TIME ON RESEARCH AND DESIGN WHICH IS COORDINATED £ P. BLIZARD, STAFF ASSISTANT ap .
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GLM  GLENN L, MARTIN COMPANY
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|
|
3an
T Y Y T T —— Y TR~ TR e W -t - " [N ———

D AR

 

 

 
 

 

QFFICIAL USE ONLY

 

e B g

 

 

THE AIRCRAFT NUCLEAR PROPULSION PROJECT

THE OAK RIDGE NATIONAL LABORATORY

CCTOBER {,1956

 

 

AIRCRAFT REACTOR ENGINEERING DIVISION

$. J, CROMER, DIRECTOR
A. B. LIVINGSTON, $EC, ARE

RD

 

 

 

 

ASSISTANT TO DIRECTOR

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

CAROLINA

COLLEGE

 

P. H. PITKANEN, UNIVERSITY OF SOUTH

J. M. TRUMMELL, STATE UNIVERSITY OF IOWA
W, K. STAIR, UNIVERSITY OF TENNESSEE
G. F. WISLICENUS, PENNSYLVANIA STATE

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

R, 5. CARLSMITH ARE
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i
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R. L. MAXWELL, UNIVERSITY OF TENNESSEE

 

312

 

  

 

 

 

 

 

 

 

1

 

 

 

 

 

 

 
P

ol il v o il o e i AN G 4 kO ki e e o i i

it N e gl ok bR i st it s il . el i Gid A i

el it i i,

 

 

OFFICIAL USE ONLY

THE AIRCRAFT NUCLEAR PROPULSION PROJEGT
AT

OCTOBER 4, 1956

 

 

SUPPORTING RESEARCH

 

 

 

 

 

 

THE OAK RIDGE NATIONAL LABORATORY

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

W, H, JORDAN
A. J. MILLER
STAFF ASSISTANT
W. K. ERGEN ARE
STAFF ASSISTANT STAFF ASSISTANT STAFF ASSISTANT STAFF ASSISTANT STAFF ASSISTANT STAFF ASSISTANT
E. P. BLIZARD AP W. R. GRIMES c W. D. MANLY M G, W. KEILHOLTZ s R. B. LINDAUER cT M. F. POPPENDIEK REE
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SHIELDING THEORY TOWER SHIELDING FACILITY SPECIAL PROBLEMS A. 5. MEYER, JR. AC GENERAL CORROSION 3 B VANCLEVE M .
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