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ALRCRAFT NUCLEAR PROPULSION PROJECT

QUARTERLY PROGRESS HEPORT
FOR PERIOD ENDIND DECEMBER 10, 1950

  
    
 
   

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This document
National Defen

applicable Felg

ORNL 919

This document consists of 290 pages.
Copy /O of 192. Series A

Contract No. W-7405, eng 26

AIRCRAFT NUCLEAR PROPULSION PROJECT

QUARTERLY PROGRESS REPORT
for Period Ending December 10, 1950

R. C. Briant
Director, ANP Division

Edited by:

C. B. Ellis and W. B. Cottrell

DATE ISSUED FEB 26 1951

OAK RIDGE NATIONAL LABORATORY
operated by
CARBIDE AND CARBON CHEMICALS DIVISION
Union Carbide and Carbon Corporation
Post O0ffice Box P
0ak Ridge, Tennessee

 
~ ORNL 919
Reactors

INTERNAL DISTRIBUTION Progress Report
1. G. T. Felbeck (C&CCD) 26. A. H. Snell 46. R. M. Jones
2-3. Chemistry Library 27. A. Hollaender 47. E. C. Miller
4. Physics Library 28. F. L. Steahly 48. D. S. Billington
5. Biology Library 29, K. Z. Morgan 49. E. P. Blizard
6. Health Physics Library 30. D. W. Cardwell 50. W. K. Ergen
7. Metallurgy Library 31. M. T. Kelley 51. C. E. Clifford
8-9. Training School Library 32, W. H. Pennington 52. M. L. Nelson
10-13. Central Files 33. C. E. VWinters 53. G. H. Clewett
14. C. E. Center 34. J. A. Lane 54. C. P. Keim
15. C. E. Larson 35. M. M. Mann 55. 0, Sisman
16. W. B. Humes (K-25) 36. G. E. Boyd 56. A. D. Callihan
17. W. D. Lavers (Y-12) 37. R.  W. Stoughton - 57. R. S. Livingston
18. A. M. Weinberg 38. F. R. Bruce 58. W. D. Manly
19. J. A. Swartout 39. H. W. Savage 59. J. L. Meem
20. E. D. Shipley 40, W, K. Eister 60. C. D. Susano
21. E. J. Murphy 41. A. S. Householder 61. W. B, Cottrell
22. F. C. VonderLage 42, C. B. Graham -~ 62. A. S. Kitzes
23. R.  C. Briant 43. R. N. Lyon 63-72. ANP Library
24. C. B. Ellis 44. R. E. Engberg 73-78. Central Files (0.P.)
25. E. H. Taylor 45. W. R. Gall

EXTERNAL DISTRIBUTION

79-83. Air Force Engineering Office, Oak Ridge
84-95. Argonne National Laboratory
96-103. Atomic Energy Commission, Washington
10%4. Atomic Energy Commission, Wilmington
105. Battelle Memorial Institute
106- 109. Brookhaven National Laboratory
110. Bureau of Aeronautics
111. Bureau of Ships
112-117. Carbide and Carbon Chemicals Division (Y-12)
118. Chicago Patent Group
'119. Chief of Naval Research
120-123. Du Pont Company
124-127. General Electric Company, Richland
128. Hanford Operations Office
129-132. Idaho Operations Office
133, Towa State College
134-137. Knolls Atomic Power Laboratory
138-140. Los Alamos
141. Massachusetts Institute of Technology (Kaufmann)
142-143. National Advisory Committee for Aeronautics, Cleveland
144. National Advisory Committee for Aeronautics, Washington
145- 163. NEPA Project
164. New Operations Office
165-166. New York Operations Office
167-169. North American Aviation, Inc.
170. Office of Naval Research
171. Patent Branch, Washington
172-186. Technical Information Service, Oak Ridge
187-188. University of California Radiation Laboratory
189-192. Westinghouse Electric Corporation

 
TABLE OF CONTENTS

SUMMARY
PART I. RESEARCH CONTRIBUTING TO THE ARE

1. DESIGN OF THE AIRCRAFT REACTOR EXPERIMENT

Core design
Coolant circuit design

Building design for the ARE

2. REACTOR PHYSICS

Introduction .
Bare Reactor Criticality Calculations
Theory and Assumptions
Results and conclusions
Reflected Reactor Criticality Calculations
Spherical reflected reactor.
Reflector saving computations
Future program
Kinetics of Liquid-fuel Reactors
Perturbation calculations
Thermal relaxation time for fuel rods
Background Problems
Effect on cross sections of atomic motion
Adjoint fluxes and perturbation theory
Cylindrical multigroup calculations
Calculations for the Critical Experiment

3, CRITICAL EXPERIMENTS

4. NUCLEAR MEASUREMENTS

Mechanical Velocity Selector

Molybdenum Cross-section Measurements

Intermediate Xenon Cross-Section Measurements
Practicality of preparing large xenon sources

The 5-Mev Van de Graaff Accelerator

5. SHIELDING RESEARCH

The ANP Shielding Board

Divided Shields for More Conservative Ground Specifications
Shield specifications
Performance of divided shields

Page
12 -

21

22
35
36

37

38
40
41
42
46
62
67
67
70
70
75
80
80
85
86
86

88

91

92
92
94
94
97

98

99

101
101
102
s

TABLE OF CONTENTS (Cont” d)

Page
5. SHIELDING RESEARCH (Cont’d)

Lid Tank 104
Iron and borated water 105
Boron carbide and water - 127
Measurements of fast neutrons in the Lid Tank using a sulfur

threshold detector 141

Liquid-metal- Duct Test in the Thermal Column 142

Shield Calculations 145
Analysis of Lid Tank data 145
Analysis of Lid Tank data on boron carbide and water 146
Theory of neutron attenuation 150
Shield calculation methods 151
Ducting difference equations 151
Gamma activity due to fission fragments 152

New Bulk Shield Testing Facility 153

Mock-up of Unit Shield , 155

6. EXPERIMENTAL ENGINEERING 156
Corrosion tests; harps 157
Figure-eight loop 161
Calibration loop - 168
Seal test device “ 168
Bearing tests 168
Air tests 168
Insulation tests 168
Pumps 171
Purification . | 173
Disposal and cleaning facilities | 176
Safety 178

7. LIQUID-METAL AND HEAT-TRANSFER RESEARCH 179

Experimental Lithium Heat Transfer 181

Heat-transfer Coefficients 182

Mean Conductance Data Using NaOH 184

Boiling Liquid Metals . 186

Natural Convection in Liquid-fuel Elements 187

Theoretical Thermal Entrance Analyses . 187

Fluid Flow and Heat Transfer in Noncircular Ducts 189

4
TABLE OF CONTENTS (Cont’d)

7. LIQUID-METAL AND HEAT-TRANSFER RESEARCH (Cont’d)

Physical Properties
Specific heat
Thermal conductivity
Viscosity
Density
Development of Components for Experimental Heat-transfer Systems
Pumps
Flow measuring devices
Liquid Metals In-pile Experiments

8. METALLURGY

Static Corrosion Testing
Materials in lead
Stainless steels in lithium
Metallic elements in sodium
Metals in uranium-aluminum alloy
Dynamic Corrosion Testing
Thermal convection loop
Fuel-element Fabrication
Static corrosion tests
Welding Laboratory
Welding of molybdenum
Welding of niobium
Fabrication of thermal convection loop
Creep-Rupture Laboratory

9. RADIATION DAMAGE

Y-12 Cyclotron Experiments

In-pile Creep

Corrosion Experiment of North American Aviation, Inc.
Creep Experiments of Purdue University

Properties of Metals

Other Activities

Page

190
192
196
198
199
199
199
201
202

204

205
206
209
210
211
211
213
213
214
218
218
220
220
220

222

223
225
231
231
232
232
TABLE OF CONTENTS (Cont’d)

10. CHEMISTRY OF LIQUID FUELS

Suspensions of Uranium Compounds in Sodium Hydroxide
Preliminary observation of the suspensions
Settling rate measurements
Identity of the uranium compound

Low-melting Fluoride Systems
Experimental methods
Properties of the fluoride systems

Corrosion Test of Metals

Corrosion Test of Ceramics and Fission Products

11. ANALYTICAL CHEMISTRY

Determination of Oxygen in Sodium

Preparation of standard samples

Time study of the Pepkowitz and Judd method
Microanalysis of Silicon Carbide
Flame Photometric Analysis of Lithium
Stability of a Silicone 0Oil in Contact with Sodium
Service Analyses

Sodium analysis

Analysis of boron carbide

Analysis of ferrous and nonferrous alloys

Analysis of uranium compounds
Summary of Service Analyses

PART I1. LONGER RANGE ACTIVITIES

12. VAPOR-CYCLE REACTORS

Mercury-vapor Compressor Jet
Gaseous Power Cycle
Sodium-vapor Compressor Jet
Corrosion Experimentation
Sodium-vapor corrosion; no irradiation
Corrosion tests with irradiation
Investigation of Mechanical Properties at High Temperatures

Page

233

235
235
236
240
242
242
243
247
252

255

256
256
257
257
259
259
260
260
261
261
261
261

263

264
266
267
267
267
268
269
TABLE OF CONTENTS (Cont’d)

Page

13. CIRCULATING-FUEL REACTORS 270
Reactor 271

Reactor coolant 272

Shield 272

Engine 273

Plane configuration 274

Ground facilities 275

14. SUPERCRITICAL WATER REACTOR 276
15. SUPERSONIC TUG-TOW SYSTEM 278
16. LITHIUM-ISOTOPE SEPARATION ) 280
Operation of 48-compartment electroexchanger 281
72-compartment electroexchanger 282
Alternative type electroexchanger 283

Analyses of production problems 283
APPENDIXES 286

A. Report of Nuclear Development Associates, Inc. 287

B. List of Reports Issued 288
Fig.
Fig.
Fig.
Fig.
Fig.
Fig.
Fig.
Fig.
Fig.
Fig.

Fig.

Fig.

Fig.

DN N e e e b e e b e e
N ~ O 00 NN N B W N

N N DN N N N

.4-2.14
.15
.16
.17
.18
.19
.20

.21

.22
.23
.24

LIST OF FIGURES

Helical Coil Arrangement (Reactor Core)

Helical Coil Arrangement (Dome and Fuel Tanks)
Hair-pin'("U") Tube Fuel Element Design (Reactor Core)
Hair -pin ("U") Tube Fuel Element Design (Support Details)
Parallel Tube Fuel Element Design (General Arrangement)
Annular Fuel Tube Arrangement (General Assembly)

Fuel Pin Arrangement (Typical Pin Assembly)

Individual Fuel Pin Arrangement (General Assembly)

Heat Transfer Coefficient for NaOH at 1350°

Xe * Cd Cross Sections * Thermal Flux Distribution

Temperature Coe fficient [(Ak/k)/°F] at Mean Core
Temperature = 1286°F

Reactivity Effects as a Function of Reactor Spectrum
Production Spectra

Flux Distribution

Power Production

Production Spectrum

Production Spectrum

Leakage Spectrum

Thermal Response in the Absence of Delayed Neutrons of
Liquid Fuel Following a Step Change in Reactivity of
103 (i.e., 13¢); Departure of Average Temperature
from Operating Level vs. Time

Perturbation Theory Results: Time Response of Integrated
Flux:-Following a Step Change in Reactivity of 10°3
(~13¢.) in Absence of Delayed Neutrons

(v/Ao) and (V/A;) vs. Time in Fuel Rod
Average Temperature in Fuel Rod vs. Time

Broadening of a Sharp Resonance by a Doppler Effect in a
Material of Atomic Weight 135

24
25
26
27
28
29
30
31
32
43
47

48
49-59
63
65
66
68
69
73

T4

77
78
84

A=
Fig.
Fig.

Fig.

Fig.
Fig.

Fig.

Fig.

Fig.
Fig.

Fig.
Fig.

Fig.

Fig.
Fig.

i v oo oo oo o

.25

.10
11

.12

.13

.14

.15

.16
.17
.18
.19
.20
.21

Importance Function, 2.79-ft Reactor
Preliminary Total Cross-section Curve of Molybdenum

Experiment 10, Comparison of Gamma Attenuation of H,0;
* 0,6% Boron; Fe-H,0 + 0.6% Boron; Fe—BaC -H,0
* 0.6% Boron

Experiment 10, Comparison of Solid Fe and Solid Fe + B,C

Experiment 10, Neutron Attenuation in H,0 Following
Various Thicknesses of Fe

Experiment 10, Neutron Attenuation in H,0 Behind Fe + B,C

Experiment 10', Effect of Removing Fe Slabs Adjacent to
Source

Schematic Drawing of Lid Tank with 24 Iron and 5 B,C Slabs

Experiment 10", Comparison of 87.5% Fe - H,0 * 0.6% Boron
and 50% Fe - H,0 + 0.6% Boron

Experiment 10", Attenuation of 50% Fe - H,0 * 0.6% Boron

Experiment 10“, Schematic Drawing of Lid Tankwith 16-7/8-in.

Iron Slabs (50% Fe-H,0Q)
Experiment 10, Dosimeter Center Line Measurements

Experiment 11, Comparison of Thermal-neutron and Dosimeter
Centerline Measurements for Borated Water

Experiment 11, Attenuation of B4C - H,0 and 0.5% Boron;
Fast-neutron Dosimeter Centerline Measurements

Experiment 11, Attenuation of B,C - H,0 and 0.5% Boron;
Thermal-neutron and Centerline Measurements

Experiment 11, Attenuation of B,C - H,0 + 0.5% Boron;
Centerline Gamma Measurements

Experiment 11 Neutron Attenuation of B,C, 25 in. BF; Counter
in B,C Jacket; Centerline Measurements

Neutron Centerline Measurements from End of Duct
Radial Neutron Flux from Center of Duct

Pb and H,0, Experiment 8, Z = 140

Pb and H,0, Experiment 8, Z = 130

120

1

Pb and H,G, Experiment 8, Z

Reactor Graid

87
93
108

110

112

115
117

119
120

122

124

125
131

133

135

137

139

143
144
147
148
149
154
Fig.
Fig.

Fig.
Fig.
Fig.

Fig.
Fig.
Fig.

Fig.
Fig.

Fig.

Fig.
Fig.
Fig.
Fig.

Fig.
Fig.
Fig.
Fig.

Fig.

~3

|

~N O N N YN YN O N O O O

~N~N g N~
~N AN D B W

H WY O N N T s W N e

o]

Low-carbon Iron Harp After Failure

"Figure 8" Loop Installation

Test Section for Corrosion Tests in "Figure 8" Loop
Test Section for Self-welding Tests in "Figure 8" Loop
Test Section for Stress-Corrosion Tests in "Figure 8" Loop
Details of Parts to be Tested in "Figure 8" Test Section
Calibration Loop

Seal-testing Device

Insulation Testing Device

Centrifugal Pump Modification

Pump Test Stand

Sodium Disposal Unit

Sectional View of Proposed Test Section for Determining
Heat Transfer Coefficient of Liquid Sodium

Proposed Apparatus for Determination of Heat Transfer
Coefficient for Molten NaOH

Apparatus Arrangement for Boiling Liquid Metals
Bunsen Ice Calorimeter Assembly

Closed Bunsen Ice Calorimeter

Enthalpy of Liquid Lithium

Apparatus for the Measurement of the Thermal Conductivity
of Liquid Metals.

Viscosity Measuring Tube
Lithium Metering and E. M. Pump Loop

Corrosion Specimens in Molten U-Al Alloy; Four Hours at

1000°C

a. Stainless Steel and UO, Held 100 hr at 1100°C (Mag 250%)
Stainless ?teel and Molybdenum Held 100 hr at 1100°C

(Mag 250%

a. Stainless Steel and Beryllium Held 100 hr at 800°C
(Mag 250%) _

b. Stainless Steel and Columbium Held 100 hr at 1100°C
(Mag 250%)

162
163
164
165
166
167
169
170
172
174
175
177
183

185

188
193
194
195
197

200
203
212

216

217

10
_—
Fig. 8.4 Columbium and UO, Held 100 hr at 1000°C (Mag 250%) 219
Fig. 9.1 Rotating Target 224
Fig. 9.2 Stationary Target 226
Fig. 9.3 Cantilever Beam Type In-pile Creep Apparatus 228
Fig. 9.4 Cantilever Beam Type In-pile Creep Apparatus (with Furnace) 229
Fig. 10.1 Apparatus for Studying Uranium Suspended in Molten 238
Sodium Hydroxide
Fig. 10.2 Vacuum Distillation of Sodium Hydroxide : 241
Fig. 10.3 Filtration of Low-melting Systems 244
Fig. 10.4 Phase Diagram of UF, - LiF Binary System 245
Fig. 10.5 347 Stainless Steel (Unexposed) (Mag 100%) 249
Fig. 10.6 347 Stainless Steel (24 hr Exposure) (Mag 100%) 250
Fig. 10.7 347 Stainless Steel (135 hr Exposure) (Mag 100%) 251
Fig. 16.1 Vertical Exchange Apparatus (Diagrammatic) 284

H y
SUMMARY

The Aircraft Nuclear Propulsion Project at the Oak Ridge National Labora-
tory has expanded considerably during the past quarter. There are now 236
technical people engaged in all phases of the research work. Thirteen divi-

sions of the Laboratory are represented on this Project.

Design of the Aircraft Reactor Experiment (ARE) is continuing, with
particular emphasis on shielding, control, and fuel material. The NEPA
Division, Fairchild Engine and Aircraft Corporation, has loaned 14 men to

assist in the design and construction of the ARE.

Three major facilities have been completed: the Shielding Reactor, the
ANP Critical Facility, and the 86-in. cyclotron, which will be used part-time
for radiation damage studies. Also, work has begun on increasing the power
level of the MTR mock-up so that it may serve as a Low Intensity Training

Reactor (LITR) and may also be used part-time for radiation damage studies.

ANP SHIELDING BOARD

One of the major activities of the members of the ANP Project at the
Laboratory this quarter was the work on the joint ORNL-NEPA Shielding Board.
During the summer the ANP Technical Advisory Board was able to carry shielding
analysis only through the stage of the 1dealized shield which contains no
ducts, pumps, heat exchangers, structural elements, or other mechanical
arrangements. The ANP Shielding Board was assembled and asked to carry the
work through the next stage and to estimate minimum weights for fully engi-
neered shields containing all the necessary mechanical features. The work of
this Board culminated in the publication of a summary report of 215 pages,
ANP-53, on Oct. 16, 1950. The following quotation from this report lists its

conclusions on shield weights:

"Weights quoted in this section include all shields, structure
within shields, reactor, reflector, ducts, pumps, and heat exchangers.

1. The weight of a divided shield for the assumed standard condi-
tions (3.0-ft-diameter right circular cylindrical reactor core,
200,000 kw output, sodium in both primary and secondary coolant
circuits, and 1 r/hr at the crew compartment) was estimated to

be 98,000 1lb.

. o
The weight of one possible unit shield was found to be over
200,000 1b for the same conditions as above, but with the addi-
tional ground tolerance requirement that the radiation at the
surface of the shield not exceed 0.8 r in the first 8 hr after
shutdown. This latter requirement nearly doubled the weight of
the shield because of the necessity of keeping the sodium in the
secondary circuit from becoming radioactive.

Substantial savings in shield weight can be effected both by
rounding the corners of the square cylindrical core and by
reducing the core diameter. Further savings can be effected
through the use of coolants less inclined to become radioactive
with such an inconvenient half-life as that of sodium (14.8 hr).
Estimated weights for two representative cases are as follows:

a. A 30-in.-diameter reactor core with ellipsoidal ends
and Li’ and Li as coolants gave a unit shield weight of
116,000 1b. For the same conditions but a 36-in.-
diameter core the shield weight was found tobe 122,000
1b.

b. A 30-in.-diameter reactor core with rounded ends and
lead or bismuth as the primary coolant gave a unit

shield weight of 105,000 1b.

The weight of a unit shield for a circulating-fuel reactor not
having a large amount of sodium in the primary circuit would
probably not be much greater than that for the Li”-Li case cited
above unless -— as seems likely -~ the heat exchanger volume must

be increased because of a lower heat transfer coefficient for the

fused salt carrying the fuel. If sodium hydroxide were used in
the primary circuit, a shield weight increase of the order of
25,000 1b would be required to satisfy the ground tolerance
condition. In any case, sodium could not be used in the second-
ary circuit with any circulating—fuel reactor because of acti-
vation from delayed neutrons in the intermediate exchanger.

If the ground tolerance for the Na-Na cooled reactor is relaxed
to 1 r/hr at the shield surface 15 min after shutdown, the unit
shield weight for a 30-in.-diameter core with ellipsoidal ends
is estimated to be 148,000 lb.

A divided shield near to the lower limit in weight has been de-
signed using somewhat relaxed design specifications. For a 2.5~
ft-square cylindrical reactor, 65 ft separation, 150 mw power,
3 to 4 man crew compartment, Li’-Li system, the total weight was

66,450 1b.™

13
For shields which are similar to ideal shields the principal sources of
uncertainty in these quoted weights are in the accuracy of each of the follow-
ing: (1) the Lid Tank detector calibration (since the Lid Tank measurements
form the basis for the computed weights), (2) the effective cross section for
boron, (3) the capture gamma-ray intensity from structural metals, and (4) the
effective oxygen scattering cross section. For shields with large deviations
from the ideal shape, such as a unit shield for a Na-Na system, geometrical
factors become the largest source of uncertainty. Shielding Reactor measure-
ments are required to remove this uncertainty. For systems using sodium as a
secondary coolant, the effective activation cross section of sodium is another
major uncertainty. All these points of uncertainty will be attacked experi-
mentally in the near future. It is to be noted that all the above calculations
are based on Lid Tank measurements with lead and borated water. There is
some theoretical evidence that no other materials will be found togive weights
which are significantly smaller. Other materials, however, will soon be

checked by Lid Tank measurements.

ANP CONTROL BOARD

A parallel effort to the ANP Shielding Board was the work on the joint
ORNL-NEPA Control Board.(!) This group studied the control properties of the
solid-fuel aircraft reactor design developed by Oak Ridge National Laboratory
which has been described in previous reportsfz)The Control Board found that
although this solid-fuel reactor would have a negative short-time coefficient
of reactivity owing to temperature changes of the fuel elements alone, 1t would
have a positive temperature coefficient of reactivity over time 1inter vals of
10 sec or longer owing to temperature changes in the moderator. The Control
Board also believed that this reactor would require 10 or more absorber control
rods. The mechanical actuation system and the heat-removal system would be
somewhat complex. Subsequent reactivity calculations, as described 1n the
Reactor Physics section of this report, have shown that the positive tempera-
ture coefficient difficulties can be greatly diminished by a decrease in the
fraction of moderator in the core so as to raise the mean energy for fission.
However, it is clear that the control of a solid-fuel high-temperature reactor

will not be simple.

(1) Interim Report of the ANP Control Board, NEPA-ORNL, ANP-54, (Nov. 1, 1950).

(2) Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending August 31, 1950,
ORNL-858 (Dec. 4, 1950); Preliminary Feasibility Report for the Aircraft Reactor Experiment,
Oak Ridge National Laboratory, Y-12 Site, Y-F§-15, (July 4, 1950).

14
ARE DESIGN

Largely as a result of the above findings of the ANP Control Board, a
decision was made to suspend work temporarily on the solid-fuel ARE design,
while attempting to develop a suitable liquid-fuel arrangement. It is to be
expected that a "stationary" liquid-fuel system can be designed which will be
almost self-regulating, and which will require only one or two control rods.
The self-regulating feature will arise from the strong negative temperature
coefficient of reactivity since increases in reactor temperature will expand
the fuel-bearing liquid and force some of the uranium outside the active core,
thus leading to a compensating decrease in power. The most promising fuel
mixture to date is a solution of UF, in NaF, with possible admixtures of other
fluorides to lower the melting point to a convenient range. Design 1is now
underway on such a liquid-fuel aircraft reactor and ARE prototype and, as de-
scribed in this report, research is also in progress on the properties of other

possible fuel liquids.

STUDIES ON ALTERNATIVE LINES

In addition to design and research toward a "stationary" liquid-fuel
liquid -metal-cooled aircraft reactor and ARE, a certain amount of work is being
continued by the Laboratory on various alternative lines which show promise
for future aircraft reactors. These studies are described in Part IT of this

report. They include:

1. Calculations on vapor-cycle systems by North American Aviation,

Inc., together with research on the corrosive effects of Na
vapor.
2. A complete design study including reactor, engine, and aircraft

analysis for a nuclear XB-52 plane powered by a circulating-fuel
reactor based on a suspension of uranium-bearing particles in
NaQOH, carried out by the H. K. Ferguson Company.

3. A detailed analysis of the supercritical water-cooled reactor by
Nuclear Development Associates, Inc.

4. Research on the most economical method of separating the isotope
Li7 in quantity for possible use as a high-per formance coolant.

5. Studies on the Supersonic Tug-Tow system.

15
SUMMARY OF RESEARCH RESULTS

From the body of research now underway on all the problems mentioned
above, some of the other tangible results obtained during the last quarter are

listed below:

Physics

1. The Reactor Physics Group has performed extensive computations
of criticality and thermal xenon coefficients in bare reactors,
and similar calculations for reflected reactors are in progress.
Graphs of the spectra of several solid-fuel reactors are in-
cluded. The determination of the kinetic response of some
liquid-fuel stationary-moderator reactor designs has been begun.

2. For the Na—BeO—stainless steel reactors studied, having moder-
ator volume fractions lying in the range 40 to 70%, the following
points are clear:

a. The temperature coefficient of reactivity with solid
fuel decreases rapidly as the mean energy of fission
of the reactor is increased by cutting out moderator.
With a moderator volume fraction less than about 60%,
the temperature coefficient of the solid-fuel reactor
becomes negative.

b. The reactivity effect due to thermal expansion of the
moderator amounts to 1 to 3% in k in going from 183 to

1286°F.

c. The reactivity effect due to maximum transient xenon
decreases from 11% for the near thermal reactor to less
than the equilibrium xenon when the mean energy of
fission 1s raised to 100 ev.

d. For a core containing 52% moderator, using a 6.5-1n.
reflector which was made of 75% Be(O, the reflector
savings came out about equal to the reflector thickness.

3. A preliminary analysis of the fuel kinetics of the NaF-UF,
reactor shows the existence of oscillation in the reactor power
following a change in reactivity, which arises from the coup-
ling between fuel displacement and neutron flux. However,
this oscillation damps out in a few seconds, and in any event
the preliminary result was obtained without taking account of
delayed neutrons. In a more exact calculation which 1s to
follow, any power oscillations are expected to be negligible.

16
4.

A resonance at approximately 49 ev was found fromthe preliminary
neutron cross-section measurements on molybdenum at Columbia.

Nuclear Development Associates, Inc. have computed, asa function
of temperature, both the xenon cross section averaged over a
Maxwellian distribution and the temperature derivative of this
quantity.

Preliminary studies show that it should be possible to prepare a
1000-curie source of Xe!35 if a measurement of the intermediate
xenon cross section proves advisable. The most efficient source
for preparing such a quantity of Xe!3® would probably be the
Homogeneous Reactor Experiment,

Shielding

7.

10.

11.

Heat
12.

Since the close of the ANP Shielding Board, a new divided shield
for a Na-Na reactor with 30-in. rounded-end core has been com-
puted to weigh 127,000 1lb 1f the ground radiation condition 1is
made to be 1 r/hr at a point 50 ft from the reactor with the
latter operating at one-tenth power. Thus, for a moderate in-
crease 1n welght over the purely air-safe divided shields of the
Shielding Board, one might gain the ability to inspect the air-
craft engines, operating one at a time, without special ground
shielding.

A Lid Tank test of a shield proposed by the ANP Technical Advisory
Board has been completed. The design consisted simply of 55 cm
of iron next to the core, followed by 35 cm of borated water.
The results show this arrangement to be impractical, since inter-
mediate-energy neutrons penetrate the iron to such an extent
that their capture gammas cannot be suppressed in the water.

Preliminary results are listed on Lid Tank measurements of
neutron and gamma attenuation in slabs of solid B,C followed by
borated water. First analysis of the data indicates a relaxation
length of 6.2 cm for B,C of theoretical density, 2.5 g/cc. How-
ever, more precise determinations are 1n progress.

A sulfur threshold detector has been put into use in the Lid
Tank for measurement of fast-neutron intensity. The effective
threshold of this instrument is about 3 Mev.

Analysis of recently published Pb-H,0 and Fe-H,0 measurements: in
the LLid Tank indicates effective neutron absorption cross sec-
tions for these experiments of 3.4 barns for Pb and 2.0 barns

for Fe.
Transfer

Theoretical analyses of heat transfer have been completed on
three situations which approximate the entrance conditions 1in-
volved in current ARE core designs. Heat transfer within a wide
range of duct shapes has bLeen analyzed.

17
13.

14.

Equipment has been designed for the measurement of thermal
conductivity, specific heat, and other high-temperature properties
of various liquid metals and molten salts. The specific heat of

lithium between 550 and 900°C has been measured as 1.0 t 10%.

Both a type 316 and type 347 stainless steel convection harp
containing liquid sodium have now been operated almost 800 hr at
1500°F without failure.

Metallurgy

15.

16.

17.

18.

19.

20.

Static corrosion measurements have been made on numerous ma-
terials in Pb at 1800°F. Tungsten and Armco iron showed no
evidence of any attacky Cb, Mo, Ta, Ni, Be, Zr, and several
stainless steels are attacked in varying degrees; 18-8 molybdenum
type alloy shows encouraging preliminary results.

The attack on stainless steels by Li at 1800°F is found to be
largely intergranular in nature with a considerable amount of
decarburization and phase transformation in some cases.

Substances so far found to have good resistance to Na at 1800°F
include types 316 and 347 stainless steel and Ni. Fair resis-
tance under these conditions is shown by Co, Mo, Ta, Alloy N-155,
Inconel, and Inconel X.

Molten U-Al alloy has been suggested as a possible liquid fuel.
Materials to contain this liquid at 1800°F have not yet been
found. Ti, Fe, Zr, Co, and Mo are severely attacked.

Preliminary tests at 2000°F for 100 hr on the compatibility of
materials suggested for fuel elements show moderate reactions
occurring between type 316 stainless steel and both U0, and Mo.

Between type 316 stainless steel and Be, in the .above circum-
stances, there was a strong reaction. With BeQO and 316 stainless
steel no reaction was observed.

Radiation Damage

21.

22.

Radiation damage experiments on various metals exposed to a
neutron flux of 1.2 X 102 at 500°F and 1000 psi pressure for as
much as four months have shown no significant changes in shape,
electrical resistivity, permeability, or hardness of the metals.

A creep test is now underway on type 316 stainless steel in the
Oak Ridge reactor at 1050°F under a stress of 4000 psi. Pre-
liminary results show at least a transient increase in creep rate
under these conditions, although possible spurious effects
caused by gamma and neutron heating of the specimen have not yet
been excluded.

18
Liquid Fuels

23.

24.

25.

Attempts have been made to find a liquid fuel consisting of a
uranium compound dissolved or suspended in NaOH. All uranium
materials tested to date have been found to react rapidly with
NaOH to form a finely divided compound which is probably sodium.
uranate. The possibility of developing stable suspensions in
this system is being studied.

The equilibrium diagram of the molten salt system LiF-UF, has
been established, and study of such systems as NaF-KF-UF, is
underway.

Preliminary corrosion studies of numerous metals in NaF-UF, at
1300°F for 160 hr show hastelloy C, Inconel, and Mo to be the
least attacked.

Alternative Systems

26.

27.

28.

29.

Neither the helium turbojet system nor the two mercury vapor
compressor -jet systems so far studied by North American Aviation,
Inc. looks desirable for use in nuclear aircraft, However, the
sodium vapor compressor-jet system still shows great promise if
high reactor temperatures can be attained.

Corrosion tests of samples in sodium vapor at 1650°F for 20 hr
show no attack on type 316 stainless steel or on AWG graphite.
These experiments will be carried to higher temperatures.

Nuclear Development Associates, Inc. has pointed out that esti-
mates contained in the Lexington report indicate the possibility
of a reactor shield for the tug-tow system weighing no more than
about 15,000 1b. Such a weight, if substantiated, would permit
the use of extremely small, easily built aircraft.

With the 48-compartment electroexchanger for separating Li
isotopes, samples of 94.5% Li’ have been obtained. A separation
factor of approximately 1.05 per stage is found. Large-scale
production appears to be feasible if this material turns out to
be really needed. The principal economic factors which are still
under study are the large power costs and the very large quanti-
ties of mercury which would be required with the present arrange-
ment .

19
| Part I

RESEARCH CONTRIBUTING TO THE ARE

20
1. DESIGN OF THE AIRCRAFT REACTOR EXPERIMENT

21
1. DESIGN OF THE AIRCRAFT REACTOR EXPERIMENT

R. C. Briant R. W. Schroeder
A. P. Fraas J. H. Wyld

ANP Division

The reactor designs now being studied for possible use in the Aircraft
Reactor Experiment all involve some arrangement of a quiescent liquid fuel and
the use of liquid sodium as both the primary and secondary coolant.: Five such
core designs are presented. The coolant circuit specifications which have

been adopted are discussed.-

Core Design. Several core designs utilizing a quiescent liquid fuel are
being considered for possible use in the Aircraft Reactor Experiment.: The
liquid which is believed most likely to serve for this purpose is a solution
of UF, in NaF, with the probable addition of one or more other fluorides to
reduce the melting point of the solution.: The simplest method of containing
the liquid fuel appears to be in tubes.: The fuel-container wall thickness
should be kept down to the order of 0.010 in. so as to minimize the uranium
inventory necessary to overcome neutron absorption in the container metal.
The combined effects of changes in thermal gradients and pressure gradients
under transient conditions would probably be serious for any container shape

other than a round tube.

A number of different types of construction have been proposed. The five
designs shown in Figs. 1.1 to 1.8 are representative of the more promising of

these proposals. The major features of these may be sketched as follows:

1.- Helical coil arrangement, Figs. 1.1 and 1.2.: This design con-
tains fuel inside tubing which is wound in helical form between
cylindrical shells of moderator.: The arrangement permits the
use of long fuel containers, thereby minimizing the number of
welds required."

2. Hairpin ("U") tube fuel-element design, Figs. 1.3 and 1.4. This
configuration features about 12,000 closely spaced U-tube fuel
elements, with inlet and outlet headers both located at the top
of the reactor for convenience of access and to simplify thermal
expansion problems. - Any desired degree of compartmentalization
may be achieved by dividing the headers.:

3. Parallel-tube fuel-element design, Fig.: 1.5.: This arrangement
is essentially the same as the U-tube type except that the tubes

22
are interconnected by headers at both top and bottom to facili-
tate filling and draining.-

4. Annular fuel-tube arrangement, Fig. 1.6. This design consists
of moderator cylinders surrounded successively bycoolant annuli,
fuel annuli, and coolant annuli. Approximately three hundred
such concentric elements are housed in a moderator matrix.:

5. Individual fuel-pin arrangement, Figs. 1.7 and 1.8.° In this
case the fuel is contained in individual noninterconnected tubes
with void volumes provided for fuel expansion and fission product
gas accumulation. The design has the disadvantage that the fuel
could not be drained from the core and replaced by means of a
purely liquid-flow operation.-

The first three of these are similar in that they all involve about 30,000 ft
of 0.100-in.-0.D., 0.080-in.=-1.D.  tubing. The fourth type, having the fuel in
the annulus between concentric tubes, requires less total footage of tubing

but presents a more complex set of fabrication problems.-

In comparing these various designs, the fuel element i1tself is, of course,
the item requiring the most study.- The nature of the heat transfer through
the liquid fuel to the tube wall presents a major question. As many pieces of
data as possible have been assembled on the physical properties of substances
which might be involved in this problem, and these are shown inTables 1.1 and
1.2 and Fig. 1.9. When necessary, thermal conductivities have been calculated
by Bridgmann’s formula for nonmetallic liquids [Proc. Natl. Acad. Sci. 9,
341 (1923)]; specific heats have been calculated by Kopp’s rule [Lange, N. A.,
Handbook of Chemistry, Tth ed, Handbook Publishers, Sandusky, Ohio (1949)]."

The temperature drop in the liquid=fuel tube from the center to the wall
is very high i1f computed on the basis that thermal conduction is the sole heat
transfer mechanism. However, the effects of thermal convection and of any
change in the properties of the liquid due to radiation would be expected to
reduce this temperature differential. 1In any case, the coolant tubes will

have to be rather small to minimize trouble from this source.

It is expected that the core diameter will fall between 30 and 36 in. and
the volume percentage of moderator between 50 and 60%. The core shape will be
a right circular cylinder with the corners rounded off to give ellipsoidal

ends.

23
   
  

   

 

 

 

          

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28
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TABLE 1.¢

Physical Data of Fluorides in the Fused State

 

FORMULA

M.P. (°C)

B.P. (°C)

SPECIFIC GRAVITY

SPECIFIC HEAT*
(Btu/1lb °F)

VISCOSITY
(centipoises)

THERMAL CONDUCTIVITY
(Btu/hr ft °F)

 

LiF

NaF

KF

BeF
RbF

CsF

AlF;
CaF,
Mgk,
BaF,
AlF;. 3NaF

 

870 (1)

980 - 997(1)

ggo (1)

800 (1)
760¢1)

684¢1)

1040¢1?
1360¢1)
1396¢1)
1280¢1)

 

 

1676<1)

1700¢1)

1500¢1)

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1410¢1)

1250¢1)
(1,4)
(1.4)

9939¢1)
2137¢(H)

 

1.798 - 0.000437 (¢t - 850) at
°c(?)

1.789 at 868.5°C(3)

1.753 at 944°c®)

1.713 at 1029°C(®)

1.672 at 1117°c®)

1.629 at 1208°C(3)

1599 at 1270°CC3)

1.942 - 0.000564 (¢t - 1000)
at tOC(2)
1.936 at 1010°C®)
1.887 at 1097°c(®)
1.859 at 1147°CC3)
1.810 at 1234°C®)
1.766 at 1313°C®)
1.714 at 1405°C(3)
1662 at 1497°C)
1.634 at 1546°C(3)

1.878 - 0.000669 (t - 900) at
¢°c(®
1.986 for the solid**

2. 873 - 0.000967 (t - 825) at
toC(2 )

3.611 - 0.001234 (t - 700) at
t°c(?)

3.07(1) for the solid**

3.18(1) for the solid**

2.9 to 3.2¢!) for the solid**

4831 for the solid**

0.578

0.357

0 258

0.319
0. 144

0.099

197
.281
. 353
125
-352

o O O o ©

 

 

~] (estimated from

NaBr and NaCl)

 

~1.5 (calculated)(s)

~] (estimated from

LiF and KF)

.7 (calculated)(s)

~1 (calculated)(s)

 

* Specific heat calculated by Kopp’'s rule for liquids.
** For LiF, NaF, and KF, (Cp liq)/(Cp solid) ~0 6 to 0 75;- for CsF, (Cp liq)/(C_ solid) ~ 1(?).
(1) Handbook of Chemistry and Physics, Ed. C. D ‘Hodgman, 31st ed , Chemical Rubber Publishing Co , Cleveland, Ohio, 1949. -

(2) Jager, F. M., Z. anorg. Chemn., 101, 1 (1917). (3)
(4) Lange, N. A, Handbook of Chemistry, 7th ed., Handbook Publishers, Sandusky, Ohio, 1949.

International Critical Tables, McGraw-Hill, New York, 1929.

(5) Manson, S. V., Theoretical Equations for Estimating Thermal Conductivity of Liquids, ORNL, Y-12 Site, Y-F8.6 (Nov. 13, 1950).
¥e

TABLE 1.1

Physical Data for Hydroxides In the Fused State

 

THERMAL

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

FORMULA| M.P. (°C) B.P. (°C) SPECIFIC GRAVITY ngflfig Eg?T VISCOSITY (centipoises) (3’ fg:g%i?i:{gg
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350°C | 400°C |450°C [500°C |550°C|s00°C
NaOH 318. 4(¥) 1390¢*)  f2.11 - 0.00063¢ for ¢ between 340 and 440°C(})|0 50(5) 4.0 2.8 2.2 1.8 |15 2.23("
1.746 [1 - 2.74 x 107%(¢ - 400)]39) 0.47 by Kopp’s 1(®)
rule for
liquids¢®)
1.786 at 320°c(?%®)
1.771 at 350°c(?®)
1 746 at 400°c(?@)
1.722 at 450°¢c(?®)
11.90 at 320°C(2®)
189 at 340°C(20)
1.88 at 360°C(20)
1 87 at 380°C(2D)
1.86 at 400°c¢2b)
1.85 at 420°C(2®)
1.84 at 440°C(20)
110 1b/£¢* at m.p. (S .
KOH (360 4 £ 0.7*) |1320-1324(*)12.25  0.001¢ for ¢t from 380 to 440°C(D) 0.34 by Kopp's 23 | 17 [ 13 [10 ]| 08} If the com-
rule for p;esglblllty
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times tgat of
1 717 at 400°c(?¢) NaOH, the
thermal con-
1 695 at 450°¢¢2¢) ductivity of
'1.673 at 500°¢(2¢) KOH would be
| about 0.65,
1.651 at 550°C(2€) (thermal con-
on(2¢) ductivity of
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Bridgmann’s
equation and
items b and d
of Y-F8-6)
(1) International Critical Tables, McGraw-Hill, New York, 1929.
(2) Landolt-Bornstein Tables, J. Springer. Berlin:

(3)
(4)
(5)
(6)
(7)

(a) Eq. II, Part I, Table 81%, p. 223.
(b) Eq. I, Table 81F, £=:177.
(¢) Eq 'II, Part I, Table 81F, p. 219 -

Arndt and Ploetz, Z. phys. Chem. 121, 439 (1926)

Handbook of Chemistry and Physics, Ed. C. D. Hodgman, 3lst ed. ; Chemical Rubber Publishing Co ;, Cleveland, Ohio, 1949.
Lange, N. A ; Handbook of Chemistry. Tth ed., Handbook Publishers, Sandusky, Ohio, 1949.

NEPA. Report IC-50-4-20 (April, 1950).

Battelle Memorial Institute. telephone conversation.
If the various problems associated with reactors of this type can be
solved satisfactorily, designs 1, 2, 3, and 4 would afford great flexibility in
loading and unloading the fuel. The entire core could be fabricated and
assembled within the shield and the liquid metal coolant put in the system and
the various pumps brought up to operating levels and temperatures before the
reactor contains any fuel. The fuel could then be added slowly and might have
its concentration varied over a fairly wide range., The initial start-up

problems thus seem much less difficult than with solid-fuel reactors.

By compartmentalization and the use of various concentrations of uranium
in the fuel solution, a considerable variation of fuel distribution across the
reactor core could be obtained, if desired for adjustment of the flux pattern.
Another potential advantage of the liquid fuel for an aircraft reactor is that
most of the radioactivity could be drained from the system at one step, thus

greatly simplifying the ground handling problems if a divided shield is used.

Coolant Circuit Design. It is expected, as indicated in previous reports,
that sodium will be both the primary and the secondary coolant for the ARE.
Coolant circuit design work currently is in progress with particular emphasis
on the components which affect the design of the building. Fundamental coolant

circuit specifications' which have been adopted at the present time are:

1. The fluid circuit will simulate the airplane system as far as
practicable within the shield if a unit shield is adopted. The
secondary circuit outside the shield is to be designed primarily
for safety and convenience. Flow rates are to be as required to
remove power at the rate developed by the ARE.-

2. Final heat rejection is to be by transfer from the secondary
circuit to air.

3. Multiple primary coolant pumps will be used.
4. Multiple intermediate heat exchangers will be used.

5. The ABRE is expected to contain two completely independent sec-
ondary systems with separate pipes, pumps, radiators, radiator
blowers, and accessories.

6. Two independent power sources of auxiliary power will serve the
secondary loops and will transmit power to the primary pumps so
as to maintain circulation in the event of failure of one power
source.

35
In addition to the above features, which are similar to those expected in
the final aircraft system, several special items are planned for the ARE for

protection of personnel and reactor under experimental conditions. These
include the following:

7. Dump systems for control of liquid-metal fires will be provided.

8. A pressure shell will be constructed about the reactor-shield

assembly, to contain materials that might escape from the shield
in the event of an accident.

9. The pump room will be isolated and shielded to protect personnel
from radioactive gases in the event of partial failure.

Building Design for the ARE.. The Test Facility building design now pro-
posed consists of a steel, concrete, and masonry fire-resistant structure B30
ft wide by 90 ft long. The building is expected to contain a crane bay
approximately 42 ft high, so that large pieces of the reactor or of the shield
may be lifted out of the assembly. A basement level is expected to house the

liquid-metal pumps and the disassembly and decontamination rooms.

36
MmO Om >

2. REACTOR PHYSICS

N. M. Smith, Jr., Chairman
ANP Physics Group, Physics Division

Introduction

Bare Reactor Criticality Calculations
Reflected Reactor Criticality Calculations
Kinetics of Liquid-fuel Reactors
Background Problems

Calculations for the Critical Experiment

37

38
40
46
70
80
86
2. REACTOR PHYSICS

N. M. Smith, Jr., Chairman

ANP Physics Group, Physics Division

A. INTRODUCTION

The principal efforts of the ANP Reactor Physics Group during the last
quarter have been directed toward two general types of calculations: compu-
tations of criticality and thermal xenon coefficients in both bare and re-
flected reactors, and calculations on the kinetic response of some liquid-fuel

statlonary-moderator reactor designs.

The criticality calculations summarized in the following sections are
gpplicable in principle to both solid- and liquid-fuel reactors of the types
discussed for the ARE. This arises from the fact that both types of ARE
designs use approximately the same volume percent of the same moderator, BeO,
and use the same coolant, Na. The difference between the criticality calcu-
lations for the oxygen in the U0, of the solid-fuel designs and the fluorine
in the NaF-UF, of the liquid-fuel designs is likely to lie within the un-

certainty limits of the present results.

These calculations deal with homogeneous reactors, both bare and re-
flected. The principal effect which must be added to the picture in order to
apply the answers in detail to designs of the ARE type is the one brought
about by the heterogeneous nature of the core, i.e., the effect of self-
shielding caused by lumping. Thus the present calculation of the critical
mass of uranium may be in error from 50 to 100%. The effects of perturbations,
however, are obtained more accurately. The application of methods of calcu-
lating "effective homogeneous" cross sections to replace the atomic cross
sections now used should reduce the pre-experimental errors of critical mass
to around 30%. Experience with calculating results of critical assemblies
should ultimately permit making calculations on anew reactor within 10 to 15%.
In spite of these uncertainties, the present refinement of the calculation is
valuable in guiding design in a semiquantitative sense and for gaining an
understanding of the overall field of intermediate reactors through exploratory

calculations,

38
It 1s to be noted that the "bare reactor" calculations described 1in
Section B really represent the first approximation to an actual reflected ARE
reactor, since the effect of a reflector was simulated by modification of
the core diameter to allow for estimated reflector savings. The so-called
"reflected reactor" calculations (Section C) represent a second, and more
difficult, approximation in which the core-plus-reflector combination 1is

treated explicitly as a two-region problem.

In addition to the calculations of critical masses and associated core
fluxes, neutron lifetimes, and control rod effects, particular attention has
been paid to that contribution to the temperature coefficient of reactivity
which arises from the shift of the thermal base across the xenon absorption
curve. This parameter is vitally important for reactor control. For those
ARE designs which involve a large volume percentage of moderator, the positive
xenon coe fficient 1s so large that it yields a net positive temperature
-coefficient for the reactor as awhole. This coefficient is, however, associat-
ed with the temperature of the moderator. It is quite clear that the design
to be used for the aircraft reactor must be arranged to have either a total
temperature coefficient of reactivity which is negative, or else a thermal
response time of the moderator which is so long that slow-acting servo con-

trols are adequate and safe.

As described previously in this report, the liquid-fuel designs for the
ARE offer the attractive possibility of gaining a rapidly acting negative
temperature coefficient of reactivity through expansion of the fuel to points
outside the active core. This allows control by means of arelatively sluggish
temperature-sensing device in place of a neutron-flux-—--sensing receptor. How-
ever, the problem of the kinetics of a "stationary" liquid-fuel reactor 1is
intricate. It requires study not only of the nuclear response of the system
to a change in reactivity, but also an investigation of the time-temperature-
density transients of the liquid fuel, the coolant, and the moderator. Studies
of maximum stresses in the fuel tubes, etc., under emergency conditions are
also required. This large body of analyses is now being undertaken. Some
initial perturbation-theory results are presented in Section D on the flux
behavior and the kinetics of the liquid fuel for the simplified case where
delayed neutrons are neglected. These perturbation theory calculations
apply to infinitesimal effects and serve to determine designs having stable
‘kinetics. Numerical integration of more exact equations are required for

analysis of the stresses resulting from large perturbations.

39
In Section E some of the background problems are listed which contribute
to the broader calculations. They include studies on the Doppler effect, on
adjoint functions, and on cylindrical multigroup age equations. Further work
is in progress, as mentioned in Section F, in preparation for interpreting the

results from forthcoming ANP critical experiments.

B. BARE-REACTOR CRITICALITY CALCULATIONS(!)

Four solid-fuel reactors having the same heat-transfer area but covering
a range of mean neutron energy from epithermal to epiresonance were considered.
The shift of spectrum was made by changing the moderator volume. Thus analysis

has been directed toward:

1. The establishment of a rough lower limit to the median energy for
fission of aBeO reactor such that xenon poisoning does not create
a control problem.

2. The calculation of pile parameters such as critical mass, neutron
life, core flux, and control effectiveness.

At the high operating temperature of the aircraft reactor the kT energy
associated with the peakiof the thermal neutron distribution is above the
energy of the peak of the xenon cross section (if the Breit-Wigner fit 1is
assumed). Thus the xenon creates a positive component of the overall tempera-
ture coefficient (Ak/kT). 1If the overall coefficient is positive an unstable
condition results which could present a very difficult control problem. How-
ever, if the moderator temperature increases slowly as the reactor power is in-w-
creased, the velocities of the thermal neutrons will then change slowly and
the resultant slow change in reactivity could probably be handled. This
temperature behavior of the moderator is being studied. The secondaryaims of
the calculations were to obtain rough estimates of critical mass, potential
reactivity changes that must be offset by shim control rods, mean neutron
lifetimes, mean xenon lifetime for absorption, total core flux, and the

effectiveness of B4C as a control rod material.

The calculations were made by reducing reflected reactors to equivalent
bare reactors and using bare reactor computational methods. In computing the
temperature coefficients several effects, such as the Doppler effect, effects
of heterogefiéityflgand“effect of the spatial variation of temperature, jwere
neglected. Work is in progress to set up a method of accounting for these

omissions.

(1) Except for minor editorial changes, this section is the same as tfie following report; Webster, J. We.
(NEPA), and Macauley, B. T. (USAF), Results of Some Bare Calculations of Critical Mass and Reactiv-
ity Effects, 0ak Ridge National Laboratory, Y-12 Site, y-F10-22 (Dec. 1, 1950).

40
Theory and Assumptions. The four reactors considered constitute a
family, each with about the same heat transfer area. In each case the fuel
was UO, of density 10.9 g/cc, the moderator was BeO of density 2.8 g/cc, the
coolant was sodium of density 0.75 g/cc, the structural material was type 347
stainless steel of density 7.67 g/cc, and the control material whenpresent was
B,C of density 2.5 g/cc. These densities are quoted for the operating temper-

ature of the aircraft reactor. The dimensions and compositions were as

follows:

 

VOLUME FRACTION

 

DIAMETER OF EQUIVALENT
NO.
REACTOR NO SPHERICAL CORE (ft) BeO STAINLESS STEEL Na

 

1 2.604 0.4700 0.08723 0.3986
2 2.690 0.5200 0.07900 0.3610
3 2.790 0.5700 0.07077 0.32341
4 3.000 0.7005 0.05020 0.2227

 

 

 

 

 

 

 

The core in each case was arbitrarily taken to have a reflector equivalent to
10 cm of additional core mixture. For a good reflecting material the conse-
quent savings from the reflector should be about equal to the reflector thick-
ness. (This assumption is verified in the first IBM calculation of a reactor
reflected with BeO, as discussed in Section C.) The extrapolation distance

was assumed to be 2 cm for all energy groups.

The reactor core at its operating temperature is taken to have a uniform
temperature of 1286°F. The cold reactor temperature was assumed to be 183°F,
which is approximately the melting point of the coolant. The power was
200 megawatts, The detailed data used in the xenon calculations are given

elsewhere. (2)

The solution of the pile equations as used in these calculations is also

given elsewhere.(3) However, the relation

(2) Webster, J. W. (NEPA), Xe Effect in an Epi-thermal Reactor, Oak Ridge National Laboratory, Y-12
Site, Y-F10-17 (Octe 10, 1950).

(3) Nielsen, M. Jo (USAF), Bare Pile Adjoint Solution, pp. 4-6, Oak Ridge National Laboratory, Y-12
Site’ Y‘F].O'lg (OCto 279 1950)& i

41
 

 

nv(u) =
(u) tor
was substituted for
nv(u) = q(u)
(o

where absorption was large.

The method of handling the xenon effect is described in an earlier
report.f*) Figure 2.1, taken from this report, illustrates the origin of the
positive temperature coefficient of reactivity contributed by xenon. The
operating temperature and the shape of the xenon resonance are such that the
effective absorption of xenon decreases with an increase in the temperature of

the thermal base.

Results and Conclusions. The following results and conclusions are dis-
cussed in terms of the median energy for fission (MEF) of the reactors. Table
2.1 relates the MEF to the size and composition of the reactor. It is to be
remembered that these calculations were made for solid-fuel reactors and there-
fore do not include thermal coefficient effects arising from fuel expansion,

as in the NaF-UF, designs.

1. With maximum transient xenon present, the temperature coefficient
(neglecting any contribution by the Doppler effect) decreases
rather rapidly with increasing MEF of the reactor and becomes
negative at a MEF of approximately 3 ev. It therefore appears
from these rough calculations that, even with short time lags in
moderator temperature rise, the solid-fuel reactor could be de-
signed to have acceptable control characteristics. This con-
trollable reactor would have a core diameter equal to or less
than approximately 2.9 ft, with 10 cm reflector savings and a
moderator percentage less than approximately 60%.

2. The number of moles of B,C necessary to provide shim control
remains roughly constant as the MEF of a reactor increases;
therefore the necessary number of control rods for an inter-
mediate reactor should be no more than for a thermal reactor.

(4) Webster, op. cit.
42
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FIGURE 2.1

43
¥

TABLE 2.1

tiiscellaneous Reactivity Results for Reactors of

Various Core Sizes and 1¢ cizx Reflector Savings

 

 

 

 

 

 

 

 

MEDIAN AVERAGE Sk/k FOR
ENERGY | NEUTEON LIFETIME COMPLETE B,C TO
CORE VOLUME FRACTION CRITICAL FOR TOTAL FLUX | LFETINE OF Xe FOR REMOVAL PROVIDE
DIAMETER - MASS FISSION AT 200 Mw v * ABSORPTION OF COOLANT | SHIM CONTROL
r 2 {zec)
(ft) Be SS 347 Na (1b) (ev) [neutrons/(cm )(sec)] . ‘ (hr) (moles’
2.604 0.4700 0.02723 { 0.3986 ~125 321 6.5 x 1015 .4 X 10°° 300 -0.076 ~9
2.69 0.5200 0.07900 t 0.3610 ~100 114 6.0 x 103 .2 x 10°° 40 ~6
2.79 0.5700 0.07077 0.3234 ~75 27 6.0 x 105 5x 10°° 5 ~5
3.00 0.7005 0.0502 0.2227 ~55 0.4 3.1 x 105 .0 x 10°° 1 -0.002 ~7

 

 

 

 

 

 

 

 

 

 

 

 
The change in reactivity due to the rise of thermal base in the
clean reactor from cold to operating temperature is small (<1%)
for those reactors having =EF greater than ~0.4 ev.

The change in reactivity caused by equilibrium xenon decreases
only slightly as the MEF increases up to approximately 20 ev, at
which point it falls off rapidly. (This behavior arises, of
course, because the equilibrium xenon concentration increases as
the burn-out decreases in the less thermal reactors, and the
lecrease in effectiveness of the xenon tends to be offset by
this increase 1n quantity. Ultimately, however, as the reactor
spectrum becomes faster the xenon concentration approaches a
meximum value determined from the equation

Rate of natural decay = rate of formation

The decrease in xenon effectiveness is no longer offset by in-
crease in quantity and the reactivity effect drops off rapidly.)

The reactivity effect due to thermal expansion rises approxi-
mately linearly with log MEF and ranges from about 1 to 3% in k
for the reactors studied here. [In the more thermal reactors
considered, the moderator atoms are at a high concentration and
the core size is large. Consequently, the neutron leakage is
small. Upon expansion, with the resultant change 1n distance
between scattering nuclei, the chance that a neutron will escape
from the reactor remains small since the neutron still sees, on
the average, many nuclei in its path. The reactivity change 1is
consequently small. (In the limiting case of infinite size a
uniform change in density obviously causes no change 1in re-
activity.) For the smaller reactor with less moderator, it
follows by reverse argument that the reactivity effect due to’
expansion 1is larger. ]

The reactivity effect due to maximum transient xenon is very
large (~11%) for the near thermal reactor considered and drops
approximately linearly with log MEF, until it becomes no greater
than that due to equilibrium xenon at MEF = 100 ev. The concen-
tration of the xenon at its maximum transient value is essen-
tially independent of the reactor spectrum. It is determined
primarily from the concentration of the i1odine at shutdown, and
the equilibrium iodine concentration is a function only of

reactor power density. Since the power density is large for
these reactors, the effect on k of the after-shutdown xenon 1s
large for the near thermal cases. As the mean neutron energy

gets above the xenon resonance, the effect drops off. Apparently,
when the reactor MEF is about 100 ev, the equilibrium xenon
concentration is determined primarily by xenon decay, and there
is no appreciable rise in concentration after shutdown.

45
7. The temperature coefficient due to thermal base plus expansion
is roughly constant with change of MEF (~2.5 X 10°% units).

8. In the range of reactors studied, the temperature coefficient
due to equilibrium xenon *+ expansion t+ thermal base change was
calculated to be always negative.

9. The effect of the sodium coolant in these reactors as opposed to
void is to decrease the uranium requirement.

10. Other pertinent results are shown in Table 2.1 and Figs. 2.2
through 2.14, which are self-explanatory.

C. REFLECTED REACTOR CRITICALITY CALCUIATIONS

D. K. Holmes and O. A. Schulze

The large amount of time required to make hand calculations for a re-
flected reactor made desirable the setting up of a machine method of calcula-
tions. Accordingly, the G.E. multigroup method was adopted with modifications
and set up on IBM machines by the uranium Control Department (F. C. Uffelman)
at Y-12. The corresponding step has already been taken by the IBM division
at G.E., and they are now able to make calculations for two complete 13-group
reflected reactors per day, which 1s contrasted with 10 computer-days needed

for the same calculation by hand.

In conversation and correspondence with the theoretical section of the
Physics Division of KAPL at Schenectady the essentials of the KAPL technique
for adapting the method for IBM calculation were obtained.(%) However, a
slight variation of their present procedure has actually been used here,
although the variation is in reality equivalent to a method, called "scheme
b," which has been discussed in a KAPL report.{(®) Whereas KAPL obtains the
relation between the average flux over a lethargy group and the values of the
flux at the group limits by an iterative, self-consistency method, the present
method assumes that the flux is linear over a group. Thus, the present method
avoids a certain amount of calculation time which would be devoted to repeat-

ing groups. It has been decided that the calculation time on the IBM would

(5) Letter from Hurwitz and Ehrlich to Ne Me Smith, Aug. 14, 1950.

(6) Tonks, L., Analyses o[ Errors in Method for Computing Critical Masses of Intermediate Piles Which
Arise from a Spatially Discontinuous Source Distribution and Other Factoryg, GE-LT-2 (1947)

46
+|

£ fop) x 10°

v
(8
1

TEMP COEF

O

FIGURE 2.2 TEMP COEFF.(

 

 

100
REACTOR MEDIAN ENERGY FOR FISSION (EM)

Ak

- /,p) AT MEAN CORE TEMPERATURE =1286°F

81201 ON ONIMVYO
DRAW'NlG NO. 10219

 

A 1.0 10 100 1000
REACTOR MEDIAN ENERGY FOR FISSION (EV)

FIGURE 2.3 REACTIVITY EFFECTS AS A FUNGCTION
OF REACTOR SPECTRUM

48
 

13

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19
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FIGURE 2.4 PRODUCTION SPECTRUM
0S
FRACTION PRODUCTION PER UNIT uX 102

Th

19

o.l

7

16

1.0

15

    

           

4 13 12 i 1o 9 8 7 6 5

| I | |
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FIGURE 2.5 PRODUCTION SPECTRUM

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FIGURE 2.6 PRODUCTION SPECTRUM

       

3

 

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52

   

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FIGURE 2.7 PRODUCTION SPECTRUM
€S

X IC)2

FRACTION PRODUCTION PER UNIT u

12

Th

19

18
i

0.l

17

16
'
10

5 14 13 12 1l 10 9 8 7 6 5 4
i i i 1 1
10 102 ToXe 104 0%
FIGURE 2.8 PRDUCTION SPECTRUM

3

2
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10

  

 

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54

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FIGURE 2.9 PRODUCTION SPECTRUM

10

1.0

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a0l X 1 LINN ¥3d NOILLONAOYd NOILOVYHI

55

10

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107 E(ev.)

102 103 104 105 108
FIGURE 2.10 PRODUCTION SPECTRUM

10

1.0

o.1
96

FRACTION PRODUCTION PER UNIT u x102

o

o

Th

19

18

\7

16
1.0

15 14 13 12
l
10
FIGURE 2.1l

i
{
102

10

9
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8

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|
104

6

5

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PRODUCTION SPECTRUM

  
  

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57

 

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19

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FIGURE2.12 PRODUGTION SPECTRUM
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7 E(ev)

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FIGURE2.13 PRODUCGCTION SPECTRUM

10

8 17 16 16 14 13 12
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58
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FIGURE 2.14
not increase linearly with the number of groups, and that, to increase the
validity of the general method, 1t would be acceptable to use 32 groups over

the energy range.

Table 2.2 presents the results of a comparison of the two methods for the
first six groups of a 13-group reflected reactor. The physical quantities
tabulated are the total leakage of neutrons in a group (E), the total number
of neutrons slowing down from a group (D), and the total number of neutrons
absorbed in a group (A4), all normalized to one neutron produced throughout the
reactor. A more complete discussion of the linear approximation has been
made.(7) A report on the detailed procedure which has now been adopted for

multigroup calculations on spherical reflected reactors has been issued.(®)

The complete reactor calculation involves computations to obtain each of

the following:

1. Average macroscopic cross sections for the core and reflector
from basic cross-section data.

2. Various constants for each group from the average cross sections,

3. Auxiliary functions needed for fitting the boundary conditions
in each group.

4. Fluxes and slowing-down densities over the reactor for each
group.

5. Net neutron leakage, absorption, and total fissions 1n each group.

6. Power distribution and the k__., for the reactor.

The setup of all these steps on the IBM machines has been very laborious,
requiring, for example, the wiring of 16 electronic control panels and seven
supplementary boards. However, the preparation period 1is now essentially
completed so that spherical reactors of the UQ,--BeO--Na--stainless steel type
may be calculated on a production basis. In addition, partial preparation has

been made for the following calculations:
1. Adjoint functions for reflected reactors.
2. Reactors of the UF, -BeF,-NaF liquid-fuel types.
3. Multiregion spherical reactors.

(7) Holmes, D. K., Results of a Test of the Linear Approximation Multigroup Calculations, 0Oak Ridge
National Laboratory, Y-12 Site, Y-F10-20 (Nove. 6, 1950).

(8) Holmes, D. K.. and Schulze, 0. A., IBM Multigroup Numzrical Procedures, oOak Ridge National Labora-
tory, Y-12 SitE, Y-F10-21 (NOV. 28, 1950)0

<0
TABLE 2.2

 

 

 

SUMMARY OF TEST OF LINEAR METHOD
GROUP CORE REFLECTOR
NO. G.E. METHOD L LINEAR METHOD G.E. ' METHOD LINEAR METHOD
I E. 0.00484387 0.00484380 E' 0.00293864 0.0029387
D. 0.0180802 0.0180804 D' 0.00190526 0.0019051
A. 0.0000768409 0.00007675 A" 3.81052 % 10°8 2.858 X 10°8
II E 0.0161205 0.0164871 E' 0.00778762 0.00797926
D 0.103625 0.103246 D’ 0.0102378 0.0104133
A 0.000335745 0.000345174 A' 1.02378 x 1077 1.23184 X 10°7
ITI E  0.0924723 0.0926290 E' 0.0535788 0.0533433
D 0.453414 0.452860 D' 0.0491315 0.0496958
A 0.00275676 0.00275551 A' 9.82630 X 10°7 9.01638 X 10°7
IV E  0.0483388 0.0480298 E' 0.0237181 | 0.0243173
D 0.702250 0.701981 D’ 0.0737526 0.0734080
A 0.00283709 0.00284091 A' 7.37526 % 10"7 1.23104 X 10°°
v E  0.0452667 0.0453560 E' 0.0285398 0.0277470
D 0.783166 0.782790 D' 0.0904788 0.0910144
A 0.00382968 0.00382328 A' 9.04788 %X 10”7 8.22112 X 10°7
VI E  0.121633 0.122265 E' 0.126826 0.121989
D 0.636685 0.635922 D' 0.0851816 0.0912816
A 0.0249708 0.0246217 A' 5.11090 % 10°° 5.46888 X 1076

 

 

 

 

 

61
Spherical Reflected Reactor. The first spherical reflected reactor

computed by the IBM group had the following basic properties-

1. Core composition:- 1.96% UO29 52.0% BeQ, 7.9% stainless steel,
36.1% Na.

2. Core diameter: 2.68 ft.
3. BReflector composition: 75% BeO, 20% Na, 5% stainless steel.

4. BReflector thickness: 6.5 in.

The results of the calculation, including the neutron fluxes in the various
groups, the power distribution, the fissions per unit lethargy interval, and
the neutron leakage as a function of lethargy, are presented graphically. The
k_.. at the operating temperature is 1.043, while the temperature coefficient,
computed on the change of thermal base alone, is *4.6 x 10°® per degree

Fahrenheit.

The flux distribution in each group, all plotted with the same vertical
scale, is shown in Fig. 2.15. Table 2.3 gives the energy limits of the various
groups. The flux is large in groups 1 through 9, into which are thrown the
fission neutrons. It is interesting to note the build-up in the lower groups

of the flux in the reflector.

This build up explains the up turn of the power distribution curve (Fig.
2.16) near the reflector in the lower energy groups the net neutron current
at the core-reflector inter face is from the reflector into the core. The
neutrons having nearly thermal energies cause fissioning before they have
penetrated far into the core. Figure 2.16 also compares the power distribution
of the reflected reactor with that of the equivalent bare reactor, normalized
to the same number of fissions in the core volume. The reflector effectively
flattens the power distribution, thus increasing the importance of the outer
regions. (It must be remembered in examining Fig. 2.16 that we are treating
a spherical reactor, so that the regions at greater radii are more important

than the curve would indicate.)

That the peak in power distribution near the reflector is actually due to
near -thermal fissions is shown by the number of fissions per unit lethargy.
This is shown (Fig. 2.17) for three regions of the reactor. The outstanding
feature of these curves is the peak, near thermal, for the region near the

reflector. This result also indicates that the xenon effect, which from the

62
10

20

. ot
DRAWING NO. 10232

30 O [o) 20 30 O 10 20 30 0
SPAGE NUMBER (I Space=2.05cm)

FIGURE2.1I5FLUX DISTRIBUTION

63

REACTOR 80l

0196 UOg
52 BeO
079 S.S.
361 Na

.75 BeO

.05 S.S.
.20 Na

10 20

 

30
TABLE 2.3

Energy and Lethargy Intervals ysed in IBM Multigroup Calculations

 

 

 

GROUP NO. LETHARGY INTERVALS* ENERGY INTERVALS (ev)
1 0.0 - 0.5 107 - 6.07 x 10°
2 0.5 - 1.0 6.07 X 10% - 3.68 x 10
3 1.0 - 1.5 3.68 x 10% - 2,23 x 108
4 1.5 - 2.0 2.23 x 10 - 1.35 x 108
5 2.0 - 2.5 1.35 x 106 - 0.82 x 106
6 2.5 - 3.0 0.82 x 10° - 0.498 x 10°¢
7 3.0 - 3.5 0.498 x 10° - 0.302 x 106
8 3.5 - 4.0 0.302 x 10° - 0.183 x 10°
9 4.0 - 7.0 0.183 x 10 - 9120

10 7.0 - 10.0 9120 - 450
11 10.0 - 11.4 450 - 105
12 11.4 - 12.6 105 - 33

13 12.6 - 13.4 33 - 15

14 13.4 - 13.8 15 - 10

15 13.8 ~ 14.6 10 - 4.5
16 14.6 - 15.8 4.5 - 1.36
17 15.8 - 16.2 1.36 - 0.91
18 16.2 - 16.6 0.91 - 0.61
19 16.6 - 17.0 0.61 - 0.41
20 17.0 - 17.4 0.41 - 0.275
21 17.4 - 17.6 0.275 - 0.225
22 17.6 - 17.8 0.225 - 0.185
23 17.8 - 18.0 0.185 - 0.150
24 18.0 - 18.2 0.150 - 0.126
25 18.2 - 18.4 0.126 - 0.101
26 18.4 - 18.6 0.101 - 0.085
27 18.6 - 18.8 0.085 - 0.068
28 18.8 - 19.0 0.068 - 0.055
29 19.0 - 19.2 0.055 - 0.045
30 19.2 - 19.4 0.045 - 0.038
31 19.4 - 19.6 0.038 - 0.0305

 

 

 

 

* Lethargy, u, is defined by u = in 107/E, whevre F -+ energy in electron volts.

64
NUMBER OF NEUTRONS PRODUCED PER C.C. OF CORE

 

8 10 12 14 16 18 20

SPACE NUMBER
(I SPACE = 2.05 ¢m)

FIGURE2.16 POWER PRODUCTION

65

 

DRAWING NQ 10233

22 24 26 28
99

FRACTION PRODUCTION PER UNIT

15 14 13 12 I 0 9 8 7 6 5 4
| | | | |

10 102 103 104 103

FIGURE2.17 PRODUCTION SPECTRUM

b o il

DRAWING NO.

  

T
10234

 
total production spectrum (Fig. 2.18) would be large for the reflected reactor,
will actually be most important out near the reflector. Thus the total xenon
effect will be greater in the reflected reactor than in the bare (because of
the larger number of thermal fissions), but this 1s mitigated by the fact that in
the reflected reactor the xenon 1s produced out near the reflector, where

neutron absorptions are less important to the overall reactivity.

The interesting feature of the curve (Fig. 2.19) of the fraction of the
neutrons which escape from the reactor in each group is the very high leakage
in group 4. This 1s a real effect resulting from a dip in the Be(Q scattering

cross section in that energy range.

Reflector Saving Computations. In order to check the reflector saving
which has been assumed in the calculation of equivalent bare reactors, three
bare reactors having the same composition as the core of the reflected reactor

were computed with the following results

(1) 2.68 ft plus 10 cm shell for reflector savings plus 2 cm shell

for extrapolation distance.’keH = 0.943.

(2) 2.68 ft plus 12 cm shell for reflector savings plus 2 cm shell
for extrapolation distance k_,, -~ 0.982

(3) 2.68 ft plus 15 cm shell for reflector savings plus 2 cm shell
for extrapolation distance k_, = 11035 Evidently the reflector
savings for this reactor is about 15 c¢m, nearly equal to the
reflector thickness.

The second reactor computed by the IBM group is the same as the first
except for the total amount of uranium, which is 100 lb as compared to 120 lb
for the first reactor (corresponding to the 1.96% UO, listed on p. 62).The k_,,
for the second reactor 1s 1.0065 the two values allow the following pre-

liminary estimate of the relationship between the k_,, and the mass of uranium

 

Sk oM
eff - 0.18 v
keff Ml

Future Program. The immediate program for the IBM group is the investi-

gation of UO, ~BeO -Na -stainless steel reactors. with the following goals:

67
89

FRACTION PRODUCTION PER UNIT u

.08

.06

03

02

Ol

 

€201 'ON SNIMVY¥Q

Th i9 18 IT 16 15 14 13 12 H 10 9 8 7 6 5 &4 3 2 | 0 u
l | | | | | | i
0.1 10 10 102 103 104 103 108 107 E(ev.)

FIGURE 2.18 PRODUCTION SPECTRUM
69

NUMBERXLEAKING PER UNIT u

.07

06

.05

.02

.0l

 

5 14 13 12 Il 10 9 8 7 6 5
10 102 |c'>3 10
FIGURE2.19 LEAKAGE SPECTRUM

|
10%

   
    

9¢201 ON ONIMVYHQ

   

0 u
I
107 E(e.v.)
1. Establish the optimum reflector thickness.

2. Obtain the relationship between the change in uranium mass and
the change in k__,
3. Establish the optimum core size and moderator percentage for
meeting the requirements that the solid-fuel reactor be
a- Large enough to allow sufficient heat extraction,

b. Small enough to keep the shield weights within
required limits, and

o

Fast enough that the xenon parts of the tempera-
ture coefficients are tolerable, and that the
reactor has optimum median energy of fission-,
production spectrum from control requirements.

4. Estimate the xenon effect on k_ and temperature coefficient

of reactivity.

ff

5. Perform calculations which will permit an estimate of the errors
involved in the approximations of the multigroup method.

D. KINETICS OF LIQUID-FUEL REACTORS

Perturbation Calculations(?’ (N. M. Smith, Jr.). Perturbation calcu-
lations of the kinetic response of a liquid-fuel reactor have been initiated.
Such calculations are useful in searching for instabilities, for surveying a
broad field for desirable characteristics, and for general self-education into
reactor kinetics. Space does not permit the detailed description of the
calculations, but these may be obtained from the reference. Furthermore, the
physical data on heat transfer, expansion coefficients, and heat capacity of
the NaF-BeF, UF, systems are not available at the time of the preparation of
this report. Hence the results described must be considered largely as

i1llustrative.

The standard method of calculation was followed as used by Nordheim in
his first calculations of xenon effects. The present reactor is complicated by
having reactivity coefficients associated with the temperature and density of
the moderator, the density of the liquid fuel, and, in the high-spectrum

reactors., a coefficient associated with the density of the coolant. The

(9) This describes the beginning of a seriesz of exploratory perturbation calculations to be made on
reactor kinetics, The parts will be published separately, with very limited distribution as the
calculations proceed and later collected together in one publication with more extensive distri-
bution, The work described here will be included in Y-F10-30 (ANPmemorandumat Oak Ridge National
Laboratory, Y-12 Site), Perturbation Equations for the Kinetic Response of a Liquid-fuel Reactor,
by Ne Me Smith, Jre, Te Rubin, Me Je Nlelsen, and R, R. Coveyou. Copies of any ANP Physics Group
report may be obtained by an authorized person by request from the ANP Library.

70
thermal relaxation time of the liquid fuel and containing tube is small, with
estimates varying, from one design to another, from 0.3 to 0.09 sec:; that of
the moderator 1s larger, with estimates from5 sec to many tens of seconds. The
situation is further complicated by the possibility of neutron - -flux thermal
elastic oscillations of the tube and liquid fuel systems inasingle tube, and,
since the tubes may be connected by a header at top and bottom, of space--
modal —neutron flux elastic oscillations. Still further complications arise
from the movement of fuel and fission poisons from regions of one importance

to another.

Obviously, no attempt 1s made to consider all these effects simultaneously
at first, but, through a series of increasingly realistic approximations, they
are added one by one. The reference report contains equations for reactor
impedance., 1including contributions of delayed neutrons, fuel response,
moderator response, xenon poisoning, and temperature coefficient and provides
for reactor admittance calculations for control rod and for coolant temperature

perturbations .

The chief interest 1s in the characteristics of a three medium system,
with coefficients of reactivity associated with the temperatures of the
moderator and of fuel only. The following constants were calculated, esti-

mated, or assumed for the 70 lb uranium--~70% BeQ 3-ft bare reactor;

FUEL MODERATOR

Thermal relaxation time (sec) 0.25 5.0
Total heat capacity (cal/°C) 2.5 x 10* 2.4 x 10°
Temperature rise rate in absence

of coolant (°C/sec) 1800 12
Reactivity coefficient (per °C) 3.3 x 10°°% 3.6 X 105
Cooling capacity (cal/°C/sec) 3.1 % 10°
Integrated flux (neutrons/cm?/sec) 3.1 x 10153
Neutron lifetime (sec) 2.5 %x 10°5

For these constants the temperature increase over the operating tempera-
ture of the liquid fuel following a step change in reactivity of 10°3 and in
the absence of delayed neutrons and xenon effects is shown to be, in degrees

centigrade,

71
91'(t) = 16.22 + 13.87e-0-3597% _ 30,10e0-4422% ;o5 (48.54t+ 0.002255)

This expression 1s plotted in Fig. 2.20.

The accompanying integrated excess flux is shown to be given by

D = 2.70 x 1013 + 2,02 x 1013e70-3597t + 9 59 x 1015e-0-4422¢% 43n (48.54t - 0.01818)

and is plotted in Fig. 2.21.

Both 0, ' and ¢’ are inversely proportional to the neutron lifetime. It
is therefore advantageous to have as long a lifetime as possible. The life-
time in a reflected reactor, particularly in a moderating reflector, will be
considerably increased over the corresponding value in the "equivalent" bare
reactor (i.e., a bare reactor of the same composition as the core of the
reflected reactor but increased in size until i1ts reactivity 1is that of the
reflected reactor). The lifetime also is greater in the reactors of lower
median fissioning energy. As the reactor is made more nearly thermal, however,
the xenon fission poison introduces a positive temperature coefficient of

reactivity assoclated with the moderator.

This positive coefficient is tolerable or not, depending on the thermal
relaxation time of the moderator (i.e., the average time required to each
equilibrium after a thermal transient). If this time 1is short compared to the
response time of the liquid fuel level-~temperature-sensing control system,
then a positive coefficient of reactivity associated with the moderator 1is not
allowable. However, if this relaxation time 1s long, a temperature-sensing
servo control system will have adequate time to act. Such a reactor would
not be self-regulating in the broad sense but only within a time defined by
the moderator thermal relaxation time. Indeed, it is true that there 1s no
meaning to a "completely self regulating reactor." The so-called "self-
regulating features" designed in systems are merely characteristics which
allow the requirements of the external servo-control systems to be relaxed to

such an extent that control is simple and safe.

72
70

'y}
o

O
O

DRAWING NO. 10237

 

0 O 0 O 0 O 0 O O O 0
n e} < < M N o N -— -—_—

3AVYOILNID S334930

- 3SI1¥ 3YNLVY3IdW3L

73

IN SECONDS

FIGURE 2.20

TIME
 

Q w0 w O w0 o
0 N o - -

-0l X 98s/,wd/SNOYLNIN “13IAIT

74

SESEPT
DRAWING NO 10238

IN SECONDS
FIGURE 2.21

TIME

-2.0

Q
0
'

0
N
1

XN14 ONILvY3d0 WOdd4d 3d4NnLly¥vd3d
It 1s not yet clear what combination of fissioning spectrum, neutron
lifetime, xenon effect, etc., will yield an optimum system for control. The

discovery of these specifications is the objective of the present course of

calculations. It is apparent, however, that the optimum exists and may well
be in the direction of the epithermal reactor. Other considerations, such as
shield size and weight, may, of course, affect these specifications. The

possibility of compensation of a positive xenon coefficient by means of an

external mechanical - -thermal system exists.
The perturbation calculations are being extended to include:

. An average group of delayed neutrons.
. Xenon polsoning.
Response to ramp changes of control rod or of entrance coolant.

Coolant loop and ramp changes of power demand.

g AW N -

. Mechanical elastic response of the liquid-fuel system.

The exact integration of the equations of motion for large perturbations

is the subject of a separate investigation.

Thermal Relaxation Time for Fuel Rods (T. Rubin, NEPA). An important
quantity in determining the kinetic response of a reactor is the time it takes
for a fuel rod to reach temperature equilibrium after having been subjected to
a change in flux. This quantity is determined by considering the following

heat conduction problem:

A cylinder has finite radius and an initial temperature v = (; heat is
produced for time t > (0 at the constant rate A0 per unit volume per unit time;
the surface of the cylinder is kept at constant temperature t = 0. The

solution given(19) ig

 

2 . .2 -katt
v _a r 2 E . % J. (ra_) (1)
A0 4K aK agJ(aan) .
n = 1

(10) Carslaw, He S., and Jaeger, J« Co, Conduction of Heat in Solids, pe. 277, Oxford, Toronto, 1947.

75
where v = temperature of cylinder at radius r and time ¢
K = thermal conductivity
a = radius of cylinder

k= diffusivity of material; k = K/pc, where p = density
and ¢ = specific heat

a, = roots of the equation J, (aa ) = 0

The temperature of a cylindrical fuel rod, when subjected to a step in-
crease in power, due to an increase in flux, and whose walls are kept at con-
stant temperature by coolant flow, would respond in the manner given by Eq.
(1). Note that changing the boundary condition to v(t = Q) = a > (O and
v(r = a) = b > (0, where a and b are constants, introduces only slight modifi-

cations into the given solution of the problem.

Included in this report are plots, Fig. 2.22, of v/A0 for a given r as
a function of time for a fuel rod whose radius is 0.13 cm and which consists
of a mixture of U0, and BeO. Also included is a plot, Fig. 2.23, of v/A0 as a

function of time, where

The time constant for the fuel rod is computed from the following equation:

w

{ [D(w) - T(t)]¢t dt
J
0

 

J [v(®) - v(t)] dt

The time constant was found to be 0.09 sec. Note that KO does not depend on

the magnitude of 4 .

76
Ll

10"

>

103

RADIUS OF FUEL ROD, .I3cms.
FUEL ROD MATERIAL, UO,~BeO

10" ¢ 1072

Flez.22 (¥ ) AND

 

102 107!

6€£201 'ON ONIMVHA
81

)xIO2
0o
|

>|<

o
2
>

     

© 0v20I1 'ON

0 25 2 50 10
tx |10 —sec

FIGURE 2.23 AVERAGE TEMPERATURE IN FUEL ROD VS, TIME
In an attempt to determine the time of response of the moderator to a
sudden power 1increase, the following analogous heat conduction problem has

been solved:

Two concentric cylinders of radius @ and b (b > a), respectively, are
initially at temperature v = 0. A sudden uniform and steady source of A,
units of heat per cubic centimeter per second is initiated at time t = 0 in

the interior cylinder. The following boundary conditions are imposed on the

solution:
(1) At r = a, v, = v, and K, dv /dr = K dv, /dr
(2) At r = b, dv,/dr = 0
The symbols have previously been defined. It is necessary only to note that

subscript 1 refers to fuel and subscript 2 to moderator.

Note that boundary condition (1) means that, in this approximation, the
fuel rod is assumed to be in direct contact with the moderator and that con-
tact resistance is neglected. Therefore the resultant calculated time constant
is expected to be shorter than the actual case, in which there is coolant

flowing between moderator and fuel rod.

The fuel rod=moderator system considered in the problem was a typical
"cell" of the reactor. Boundary condition (2) is a statement that the solution

in this cell repeats itself in the other units.

The solution obtained 1is

2
v K1k2a t

 

A K1k2a2 + szl(lfl - a?

-a 2t -
an a

1 - e n
+ zklkzsiz ————— Ji(ea,) |y (ma,) Yol - ¥ (na,) gy

n =0 n - 2

79
where the a ’'s are the roots of the following equation:

K.k} J (uu){J, (na) Y (eu) - J (eu) ¥, (nu)}
t K RAE o J (uu){J, (eu) Y (nu) - Y (eu) J (nu)} = 0

and where

a b a .
Cw TTw R
% @ k

Numerical work will soon be in progress to determine the quantities that

were obtained for the fuel rod.

E. BACKGROUND PROBLEMS

Effect on Cross Sections of Atomic Motion* (R. K. Coveyou, Mathematics
Panel). 1In this and succeeding studies methods will be set down for taking
into account the effect on various diffusion phenomena of the finite velocity
of the atoms of the medium in which neutrons are diffusing. This first study
will concern itself with the effect of this motion on values of nuclear cross

sections.

Consider a neutron of speed s in collision with an atom of speed t,
selected out of a distribution isotropic in direction. Then the relative

speed of neutron and atom 1is given by

r2 = g2 + 9 \st + t2

* To appear later as part of another report.

eo
where r is the relative speed and A is the cosine of the angle between the
velocities. Denoting by P(X) the probability of the event described by X, we

have

By the hypothesis of isotropy, A is uniformly distributed on (-1, 1). Hence,

 

 

) 1 x2 - 92 - t2
Pelrs x) =51 95t ' s-tgag s bt (3)
or
x dx
Py(x g r g %+ dx) o s - t] g xg |s + t] (4)
ST

Now suppose the distribution of atomic speeds is Maxwellian, i.e.,

4'63 2,2
PlZ s t< Z +dz) =——172%2"B"2" {4z (5)
T

where 52 = M/2kT, M is the atomic mass 6 k is Boltzmann's constant, and T is

the absolute temperature.

81
Then the distribution of relative speeds "seen'" by a neutron of speed S

is in this case given by

x dx rfis 2.2
2st J77 ¢ (6a)

 

F(x £ r & x+dx)=

where the integration is with respect to t and the limits are given by

 

|s - t| € x < |s *+ t|. But this is the condition that x#, s, and t form a
triangle, hence .jis symmetric and can be written Is - x| £t < ‘s + x|. Thus,
|2+
x dx |- 2,2
P(x g rg x + dx):zfir——- (282t dt)e Rt
Jm s |
| 2-s]
= fx_dx e-,BZ(Jc-s)2 e-,82(15+.<3)2 . (6b)
7S

Suppose now that the cross section for a given process is o(s), as a function
of relative speed. Let o*(s) be the effective cross section for this process

as a function of neutron speed. 1lhen,
o

or(s) = (BAT s) | x o(x) |e Blxme)? L "’°52(x+8)2} ax (7)

0

82
If

 

o(s) = (1/v absorption)

then
s

B
T050 2 2
og*(s) = —_— e % du
s N7 |
0

 

 

0
= erf (Ls)

Since erf (Z) —> 1 as s —> ®, this states that the cross section at high

neutron volocities is unaffected, as would be expected. If

g(s) = 508(3 - s,) {sharp resonance}

then

2 e_fiz(s+so)2

 

-B2(s-s,)
fiaoso e o

The effective cross-section curves have been calculated for the case of a _
material of atomic weight 135 and a sharp resonance at 0.0863 ev, at tempera-
tures 300, 1000, and 1500°C (see Fig. 2.24). We note that the maximum of
the effective curve always lies below the resonance energy, though the dis-

placement is negligible in the cases considered.

For other cross-section curves o*(s) can be computed by numerical in-
tegration. Existing measurements of cross sections made at room temperature,
such as the xenon absorption resonance, can be taken and Eq. (7) canbewritten

as a matrix. Inversion of the matrix by machine computational methods will,

83
 

30

2.2

DRA

Q 10241

Q

3.2 3.4 3.6 3.8 40 4.2 44 4.6 48
NEUTRON SPEED IN km/sec

3.0

FIGURE 2.24
in principle, give the true cross section. The xenon resonance, for instance,
may be somewhat narrower than that given in the published curves. A narrower
xenon resonance would produce a larger positive temperature coefficient of

reactivity.

Adjoint Fluxes and perturbation Theory (M. J. Nielsen, USAF). For each
of the family of bare reactors discussed above, the differential equation
adjoint to the assumed reactor equation was solved numerically. 1In solving
the adjoint equation, the probability that the ultimate fate of a neutron in
the reactor is to cause a fission is computed as a function of energy. This
permits the assignment to various neutron groups of an "importance to t he
reactor" parameter. In addition, using first-order perturbation theory, the
change in reactivity resulting from small changes in the pile parameters such
as size and fuel content can be computed. (Since most perturbation calculations
for the bare reactor involve approximately the same labor as an additional
reactor calculation, this use of the adjoint function has so far been very
limited.) The adjoint solution also yields a numerical check on the determi-
nation of k_.,. Calculations of the reactors in the preceding section were

checked in this manner.

The lifetimes given above were computed in the standard manner as per-

turbation calculations by assuming that for k_,, = 1 + Ak, the neutron density

as a function of time is given by

(Bk/ 1yt
n(t) =n(0)e-

where | is the lifetime. Then

on(t) Ak 1 Ak

3t Lt =TT e

i 1 Ak .
where » = nv, the neutron flux. The quantity — is then regarded as a
v

perturbation on the absorption cross section of the pile and the corresponding

reactivity change 1s computed, giving

85
Ak
Ak = (constant)-—r

The constant is in the form of a definite integral which can be evaluated

numerically, giving the lifetime directly.

As part of a computational handbook the derivation of the bare-pile
adjoint equation and equations for all types of perturbations as integral
equations and as difference equations are assembled in report Y.-F10-18,¢(!1)
Other references are given.(12-!4) A compilation of methods planmned for
multiregion piles is in preparation. The slowing down density, the importance
function, and the statistical weight for the 2.79-ft reactor are shown in
Fig. 2.25 as a function of energy. The slowing-down density is the number of
neutrons per unit of lethargy slowing down in unit time. The importance
function 1is the net increase in neutron inventory resulting from the introduc-
tion of one neutron. The curve labeled "statistical weight" is the product

of the other two curves. The ordinate scale is for the importance function.

Cylindrical Multigroup Calculations (M. C. Edlund and T. Rubin). The
multigroup age equations have been transformed into difference equations in
cylindrical geometry.('5) Hand calculations are now in progress to check the
method prior to setting wup the problem for the IBM. A satisfactory approxi-
mation to the correct boundary conditions for this geometry has not yet been

found.

F. CALCULATIONS FOR THE CRITICAL EXPERIMENT

It is planned that the ANP Physics Group will devote considerable time in
the future to calculations pertaining to experiments with the critical as-
semblies. The principal activity will be devoted to interpretation of results
from the critical experiments, guiding the experiments so that the results
will be useful and interpretable, and using the experiments to check present
calculational techniques. Some effort has already been made toward selecting
the first few experiments to be of sufficiently simple geometry that a reliable
comparison can be made using the IBM calculation.

(11) Nielsen, Me Jo, Bare Pile Adjoint Solution, 0oak Ridge National Laboratory, Y-12 Site, Y-F10-18
(Octe 27, 1950).

(12) Goertzel, G., “Variational Der ivation of Adjoint Function and Perturbation Equation#’ Append ix,
pe 13, in Numerical Integration of Criticality Equations and Use of Perturbation Theory for Bare
Intermediate Reactors,by As. R. Gruber, Oak Ridge National Laboratory, Y-12 Site, TAB-96 (Aug. 15,
1950).

(13) Nordheim, L. W., and Soodak, H., Application of the Rayleigh-Schroedinger Perturbation Method
to the Theory of Piles, Chicago CP-1638 (June 15, 1944).

(14) VWigner, E. P., Effect of Small Perturbations on Pile Period, Chicago CP-G-3048 (June 13, 1945).

(15) Edlund, M. C., Numerical Integration of the Multigroup Reactor Equations in Cylindrical Geometry,
Oak Ridge National Laboratory, Vv-12 Site, Y-F5-22 (Sept. 27, 1950).

86
 

0
0 < < "
030NCONLN! NOMIN3N ¥3d AHOLN3ANI

o
"

87

a_—

DRAWING NO. 10231

   
  

NOYL1N3N NI 3SV3Y¥ONI

8 17 16 5 14 13 12 1t 10

9

Th

E (e.v.)

10 10 10 10 10 10 10

1.0

o.1

FIGURE 2.25IMPORTANCE FUNCTION -2.79 FT. REACTOR
3. CRITICAL EXPERIMENTS

88
3. CRITICAL EXPERIMENTS

A. D Callihan, Physics Division, and J. F. Coneybear, NEPA

Preparations for joint ORNL-NEPA critical experiments have continued.
The installation of equipment in the laboratory is almost complete, the major
exceptions being the control and safety rods and the personnel shield. An
allotment of uranium has been received and 1s in the process of fabrication.
Some graphite has been received, and the beryllium is being machined. At the
time of this writing it was expected that materials and equipment would be

ready for the initial experiments about Jan. 1, 1951.

Conferences between ORNL and NEPA personnel have resulted in slight modi-
fications in the initial program described in NEPA-1522. Owing to changes in
the ARE design and a desire by the theoretical groups to check the accuracy
of calculations, it is likely that the first assembly will be a simple lattice
rather than one patterned after a particular ARE reactor. The initial experi-
ments will be done with the moderator and reflector of graphite or beryllium,
depending upon the materials on hand at the time the uranium is delivered.
The median energy for fissions in the ARE design now considered is between
1 and 100 ev, probably closer to the former. This change has necessitated

138

consideration of Xe danger-coefficient measurements.

The assembly tables which carry the reactor halves have been installed
and are being tested. The aluminum honeycomb structure has been assembled on
the tables and fixed into position in an iron framework. The mechanism for
inserting a neutron source has been operated. Part of the support for a per-
sonnel shield for use during assembly operations has been placed. The con-
trol and safety rods have been built, but have not been installed in the

assembly because of a delay in obtaining the supporting structure.

The actuating circuits for the table drive and for the control and safety
rods have been assembled. Eight radiation-level-measuring circuits, used for
safety and for operation, have also been installed. This equipment is being

tested.

An allotment of 75 kg of enriched uranium has been made and 1is now being
fabricated into fuel disks 0 01 in. thick. Beryllium is being cast and finish-

machined by the Brush Beryllium Corporation. The delivery schedule calls for

89
500 lb to be shipped during December, 1950, and the total of approximately
2 tons to be delivered by March 1, 1951.- The normal density of AGOT graphite
(~1.6 g/cm®) is being machined by the Y-12 Shop, and partial delivery has been

made. Analysis of this graphite indicates acceptable purity.

Samples of the special high-density graphite (~2.0 g/cm3) have been re-
ceived but difficulties with the dies have delayed delivery. Analysis after

fabrication of test samples indicates acceptable purity.-

Since sodium is being considered as the coolant for the ARE it is neces-
sary to investigate its effect in the critical experiments. It may be built
into these experiments as NaF or as Na metal.  Satisfactory samples of NaF,
formed by either hot or warm pressing, have been received from two vendors.
One vendor is considering canning sodium in type 304 stainless steel, 0.008 in.:
‘thick, and it is believed that the cans may be filled to at least 95% of their

-volumes.-

Stainless steel, type 310, for the heavy-metal reflector studies, is now
being fabricated in the Y-12 Shop, and spectrographic analyses have been made
of the steel.:

Conferences with the NEPA Materials Section concerning dangerecoefficient
samples have continued. It appears now that samples will be available by the

time they are needed.-

90
4. NUCLEAR MEASUREMENTS

91
4. NUCLEAR MEASUREMENTS

The cross-section curve of molybdenum, being determined at Columbia
University, indicates a resonance at 49 ev. Consideration is being given to
the measurement of the xenon cross section at intermediate energies, and
consequently the practicality of the preparation of large xenon sources has
been investigated. The chopper time-of-flight velocity selector for operation
in this intermediate energy range is still under construction. The 5-Mev Van

de Graaff accelerator may be in operation in three or four months.

MECHANICAL VELOCITY SELECTOR

G. Pawlicki and E. C. Smith, Physics Division

The chopper time-of-flight velocity selector for operation in the neutron
energy region up to several thousand electron volts is still in the process of
design and construction. Design-of the rotor is complete, and construction
has started in the shops. Construction of the rather elaborate electronic
equipment is proceeding satisfactorily in?ihe Instrument Shop.: The BF,-filled
ionization chamber to be used for countfng the transmitted neutrons is under

preliminary test.:

MOLYBDENUM CROSS-SECTION MEASUREMENTS

Columbia University*

The cross section of molybdenum is being measured at Columbia University
by Prof. W. W, Havens. This measurement is of great interest to reactor
engineers since molybdenum has not only exceptional high-temperature strength
but also a relatively high resistance to liquid-metal corrosion. The nuclear
requirement for aircraft reactors, however, is that the nucleus should have a
reasonably low intermediate absorption cross section. A preliminary total
cross-section curve, Fig. 4.1, of molybdenum indicates a resonance at 49 ev,
with a height of a few barns. Higher resolution work is in progress, and more

data are expected shortly.:

-

® Work performed under contract with the AEC New York Operations Office, :

92
05

04

 

2
o
5
= w
o 0.3 u
n
2 2
= x
> o
<
@
-
0.2
0.l

o 20 40 60 80 100 120 140 160 180 200
NEUTRON TIME OF FLIGHT MICROSECONDS PER METER

FIG.4.! PRELIMINARY TOTAL CROSS-SECTION CURVE OF MOLYBDENUM
(Measured at Columbia University by Professor WW.Havens)

93
INTERMEDIATE XENON CROSS-SECTION MEASUREMENTS

During the past quarter some study has been given to the desirability and
practicality of measuring the xenon cross section at somewhat higher energies
than those reached in the previous ORNL work. Some rough calculations indicate
that a danger-coefficient measurement might be possible if 1000 curies of
xenon were available.{!) A study was therefore made by the Chemistry Division
on the feasibility of preparing a xenon source of this magnitude. As described
in the following section, it appears that such an operation would be possible,
although it would require work on a rather large scale. However, some theo-
retical reasons have been advanced(?’ for believing that the contribution of
possible higher resonances to the resonance integral of xenon will turn out to
be negligible in the region of interest for the ARE, compared to the tail of
the known xenon absorption band. The matter is, therefore, being considered
further before embarking upon such a large experimental program as would be

required for complete exploration of this spectrum.

Practicality of Preparing Large Xenon Sources.(3) A xenon source of
sufficiently high level may probably be realized from each of the following

three reactors:

1. Low Intensity Training Reactor (LITR)
2. Hot Pilot Plant

3. Homogeneous Reactor Experiment (HRE)

The parameters pertinent to the production of xenon are summarized in Table 4.1

for each of these reactors.:

Of the three methods for the production of xenon, the minimum yield, in
curies, from the Hot Pilot Plant would be lower than desired, whereas that
from the LITR should be adequate, and the yield from the HRE should be more
than ample. In regard to availability, the production—essentially the col-
lection—of xenon could probably be arranged to fit in with the current pro-

gram of the Hot Pilot Plant. Neither the LITR nor HRE is now in operation,

(1) Weinberg, A. M., Resonance Absorption by Xe, CF 50-8-.116 (Aug. 29, 1950).

(2) Arfken, G, B,, Jr., and Welton, T. A., Estimates of Effects of Higher Levels an the Resonance
Integrals of Xel35, CF 50-12.25 (Decs 6, 1950). "

(3) From a letter of Dec. 15, 1950, from G. W, Parker to C, B, Ellis re high-level sources of 9.2-hr
xenony CF 50.12.45,

94
O
I

TABLE 4.1

High-level Sources of 9.2-hr Xenon

 

 

 

AMOUNT AT
ESTIMATED POWER; PORTION HALFLIVES MAXIMUM YIELD

REACTOR Ckw) AVAILABLE TO PROCESS SATURATION (curies)
Hot Pilot Plant
{Clinton Pile) 0:2/sleg 100 siugs 2° 1000 200
Low intensity
Training Reactor 300 1 of 18 clements: O E 1000 750
Homogenecus
Kea.tor Experiment 1000 ALl A 40,000 20,000

 

 

 

 

 

 

#* Type of process for LITR element:
“*+ Type of process for HRE off-gas:

* Type of process for Hot Pilot Plant slugs:

(1) dissolution 1in HNO, ; (2) buik gas fiactionation

7% - ;< : . - \ 5 :
(1) melcing of unit in vacuum; (2} minor gas fractionation

(1) collectron of 24-hr gas sample; (2) minor gas fractionation

 
although preparations for placing the LITR in operation are proceeding rapidly.
The HRE is expected to operate at low level in mid-1951, and full-level
operation may possibly be reached by the end of 1951.

(1) Low Intensity Trawning Reactor. While the operation of the Low
Intensity Training Reactor (LITR) has not yet been definitely
settled with the Safeguards Committee, preparations for placing
the reactor in operation are proceeding rapidly.- A shield
estimated to weigh at least 8 tons is necessary to safely handle
a green fuel assembly of the size and activity required to pro-
duce the xenon.  Since such a shield would be generally useful
for other purposes, it has already been designed and is expected
to be built soon.

It is anticipated that the element would be removed from the
reactor after a short cooling period and transported quickly to
the laboratory working area, where the xenon could be collected
and purified by a relatively small charcoal bed and fractiona-
ting column. With a flux of 5 X 10!? neutrons per second the
time after shutdown for maximum xenon growth has been calcu-
lated*’ as 4.2 hr, which is adequate for the required operations.

The amount of xenon available would be about 1 kilocurie at
300 kw at whatever flux at which the LITR may be operated.  The
processing time after the 4 hr growth period could be short, and
the yield might therefore be as high as 80 to 90%.-

(2) Hot Pilot Plant. The equipment (scrubbers, condenser, etc).:
now being installed in the Hot Pilot Plant for the absorption
of the 10 year krypton of mass 85 from the off-gas stream of
the dissolver operation will also concentrate xenon, although
the latter is cooled to a stable gas in the process. The Pilot
Plant may be mentioned as a possible source of 9.2-hr xenon if
and when the problem of handling large quantities of oxides of
nitrogen and foreign gases can be overcome. Conceivably, a
fair portion of a 100-slug charge could be dissolved in 4 to 6
hr, and i1f the time for fractionation were not excessive, vyields
of 200 or more curies might be expected.

(3) Homogeneous Reactor. The most attractive prospect for future
large-scale xenon work is the Homogeneous Reactor Experiment
(HRE), from which a continuous flow of hydrogen, oxygen, and
water vapor will sweep out kilocurie quantities of xenon and
krypton. The off-gas treatment system now proposed 1includes
a large charcoal bed through which 40,000 curies of xenon will
eventual ly be discharged after around 100 days of holdup for
decay. In order to obtain similar quantities of xenon in a
portable container, 1t would be necessary only to pass the off
gas through a shielded charcoal cold trap and collect samples
of the mixed gases which could then be purified. The equilibrium
xenon activity of 40,000 curies will give off heat at the rate
of 500 watts, thereby introducing am extra requirement for
handling."

(4) Lane, Jo:A.; Xenon Poisoning at Varying Power Levels, CF-49-12-83 (Dec, 14, 1949j.

96
THE 3-MEV VAN DE GRAAFF ACCELERATOR

W. M. Good, Physics Division, and Conway Snyder, NEPA

A 5-Mev Van de Graaff accelerator, constructed by the High Voltage Engi-
neering Corporation of Cambridge, Mass., has been purchased by NEPA. The
final design has been approved, and delivery is expected in the immediate
future. - Upon receipt the accelerator will be erected in the Y-12 area. It is
expected that a period of three months will be required to place it in opera-
tion, following which research will begin on shielding, radiation damage, and

other problems in nuclear research pertinent to the nuclear propulsion of

aircraft.-

97
5. SHIELDING RESEARCH

98
5. SHIELDING RESEARCH

E. P. Blizard, Physics Division

During the past quarter the Shielding Groups at ORNL and NEPA collabo-
rated in an effort to apply the available information to the actual design of
several realistic aircraft shields. There resulted a number of configurations
which were light enough to be highly interesting and sufficiently well founded
in experiment to lend considerable confidence as to their adequacy. Iron and
water shields were in#estigated in the Lid Tank, but to a limited extent only,
since the shield weights appear to be excessive for these materials. The
shielding properties of B4C are currently being measured, and several inter-
esting effects have been discovered with this material. Measurements using
the liquid-metal duct are being made, but as yet no conclusions can be
reached regarding i1ts effectiveness in the shield. Completion of the new Bulk
Shielding Facility has been delayed by the building contractor. The first
test, probably of an ideal unit shield, will commence sometime in March. Some
previous Lid Tank experiments on lead, iron, and water have now been analyzed
éccording to simple theory. A more general approach to 5s5hield optimizing is
being explored, with the hope of carrying out this procedure theoretically

as well as experimentally.
THE ANP SHIELDING BOARD*

This Board, a joint effort of NEPA and ORNL, was convened from September
5 to October 16, primarily to demonstrate the best possible weight estimates
of pertinent, fully engineered shields for the subsonic aircraft. In addi-
tion, any conclusions which could be inferred from the work were to be re-

ported, especially 1f doing so would aid in allocation of effort for the ANP.

The first designs were based on standard specifications which were given
to the Board at its inception for a Na - Na fixed-fuel cycle: 3 ft diameter,
3-ft-long cylindrical reactor, 200 megawatts power, 50 ft reactor-crew sepa-
rations, 1 r/hr flight tolerance. The tolerance for ground servicing was

left to the discretion of the Board, and after discussions with engineers

* ORNL membership included E. P. Blitard, Chairman, C. E. Clifford, A. P. Fraes, Kiu .E. Keyes, R. V.
Schroeder, T. A. Welton, and J. H. Wyld.

99
0.1 r/hr was adopted., Since the unit shields for this set of specifica-
tions appeared to weigh over 200,000 lb, competitive designs were investi-
gated., It was then demonstrated in some detail that for a 30-in.-diameter
nearly spherical-core with lead or bismuth as the primary coolant, other
conditions remaining the same, the weight of reactor, shield, structure within
shield, reflectors, ducts, pump, and heat exchangers was approximately 105,000
1b. Other designs, with somewhat greater weights and lithium as the coolant,
were also completed. In view of the fact that neither lead nor lithium could
surely be used as a coolant because of their corrosive properties, the Na-Na
cycle was re-examined for a smaller core more nearly spherical in size,
Furthermore, the ground tolerance was relaxed to 1 r/hr at the shield surface
15 min after shutdown. For this configuration the weight was estimated to be

148,000 1b.

The divided shield for the Na-Na cycle was, on the other hand, not
excessive in weight, being estimated to be 98,000 1b for the standard condi-
tions given above. In view of this fact, i1t appeared interesting to investi-
gate what might be a lower limit for shield weight; consequently a divided
shield for a 2.5-ft-square cylindrical reactor, 65 ft separation, 3 to 4 man

crew compartment, Li’-Li system was designed. The total weight was 66,450 lb.

General conclusions of the Shielding Board were that a serious weight
penalty must be paid for using the Na-Na system in the unit shield. The
same conclusion applies for a unit shield circulating-fuel system using NaOH
as the fuel vehicle and Na as the secondary coolant. On the other hand, a di-
vided shield for either of these systems will probably be below 100,000 1b in
weight, In comparing unit and divided shields 1t was pointed out that the
divided shield weighs at least 30,000 1b less than the unit shield in most
cases of significance. Furthermore, the frontal area of a divided shield will
always be appreciably less than that of a unit shield. The obvious dis-
advantage of the divided shield, the extreme radiation intensity near the
reactor during and after operation, appears not to be prohibitive since the
outer section of the reactor shield consists of a rather thick layer of gaso-
line which can be replaced while on the ground by a much denser material,
for example, steel shot in o0il, It is also pointed out that both unit- and
divided-shield aircraft will require facilities for handling very radioactive
components during reactor replacement, and, if long flights and low burn-up
become necessary, this will be a frequent operation. The complete findings

of the Board are reported in ANP-53,

100
DIVIDED SHIELDS FOR MORE CONSERVATIVE GROUND SPECIFICATIONS (1)

T. R. Mitchell*

It appeared desirable to initiate an investigation of shields which would
satisfy ground tolerance specifications intermediate to those of ground-safe
unit shields and the air-safe divided shields described in the Shielding Board
Report.(z) Such shields would necessarily be of the divided type but would re-
tain some heavy gamma-ray shielding material completely around the reactor.

The tolerance conditions chosen for this study were as follows:

(a) 1 r/hr outside the crew compartment at 50 ft from the reactor
center, with reactor operating at one-tenth power (20,000 kw).

(b) Y% r/hr in the crew compartment with reactor operating at full
power (200,000 kw); rear of crew compartment to be located 50 ft
from center of reactor.

These specifications are not so stringent as those for the ground-safe unit
shields defined in ANP-53, which require that the radiation intensity at
50 ft be no more than 1 r/hr when operating at full power. They are con-
siderably more conservative, however, than the specifications for the air-safe
divided shields of ANP-53, which limited only the radiation level inside the
crew compartment while the radiation intensity outside the compartment was
allowed to be extremely high. The radiation specifications being considered
here would allow maintenance personnel to remain at the engines for a short
time without auxiliary shielding with the engines running, one at a time, at

about half power.

Shield Specifications. The weights and sizes of the divided shields
investigated in this study are given in Table 5.1 along with similar data for
unit shields around the same reactors. Qf the weights listed for the divided

shields, 22,000 1b is attributed to the crew shield.

# ©n loan from NEPA to the ANP Division for assistance in designing the ARE.

(1) Except for minor editorial changes, this section is the same as the following report: Mitchell,
T. Re, Divided Shields for More Conservative Ground Specifications, Oak Ridge National Laboratory,
Y'lz Sitet Y'F15°4 (NOVQ 20, 1950).

(2) Report of the ANP Shielding Board, NEPA-ORNE, ANP-53 (Oct. 16, 1950).

101
TABLE 5.1

Weights and Sizes of Shields

 

 

 

Na-Na COOLED Na-Na COOLED . Li’ Li COOLED
REACTOR WITH REACTOR WITH REACTOR WITH
ELLIPSOIDAL - ELLIPSOIDAL ELLIPSDIDAL
ENDS; 30-IN.- ENDS; 36-1in.- ENDS; 36-in. -
DIAMETER CORE DIAMETER CORE DIAMETER CORE
SHIELD
SHIELD | SHIELD SHIELD | SHIELD SHIELD SHIELD
WEIGHT | DIAMETER | WEIGHT |DIAMETER | WEIGHT | DIAMETER
(1b) (in.) (1b) (in.) (1b) (in.)
Divided Shield 127,000 130 136,000 132 111,800 132
Ground-safe unit shield, (3)
1 r/hr at 50 ft 148,000 145 160,000 147 122,000 147
Ground-safe unit shield,(a)
Y% r/hr at 50 ft 179,000 155 199, 000 157 148,000 157

 

 

 

 

 

 

 

The radiation intensity at the surface of the 1 r/hr unit shields, 15 min
after reactor shutdown, is approximately 1 r/hr.(3) This will be greater by
more than a factor of 10 at the surface of the reactor portions of these

divided shields.

Performance of Divided Shields. To satisfy the first of the two radiation
tolerance conditdons stated above, it is necessary simply to remove a one-tenth
attenuation thickness of neutron and gamma-ray shielding materials from the
outer layers of the existing ground-safe shield designs (ANP-53) which were
designed to give 1 r/hr at 50 ft when operating at full power (200,000 kw).
The second condition is then met by designing a crew shield for a factor-of-40
total attenuation of neutrons plus gamma rays. Thus, while the weights given
here for the shielding around the reactor were obtained by perturbation of
previously reported shield weights, the crew shield weight was computed in-

dependently,

(3) ANB-53, op. ctit., p. 53

102
Specifiéallyg an equivalent of 7.7 in. of water and 1.71 in. of lead were
removed from the outer portions of the ground-safe shields (1 r/hr at 50 ft
at 200,000 kw) for the following three reactors:

1. Na-Na cooled reactor having a 30-in.-diameter core with ellip-
soidal ends.‘?

2. Na-Na cooled reactor having a 36-in.-diameter core with ellip-
soidal ends (design similar to 1).

3., Li’-Li cooled reactor having a 36-in.-diameter core with ellip-
soidal ends.‘®

Assuming that the neutron shielding outside the last lead layer consists of
gasoline contained in plastic, this removes 43,000, 45,600, and 32,000 1b,
respectively, from the unit shields (1 r/hr at 50 ft) for the reactors listed
abofe. Using the weights which have been previously(5) computed for these
unit shields, the shielding around the reactor for the three divided shields

being considered will weigh 105,000, 114,400, and 89,800 1lb, respectively.

Decreasing the thickness of the reactor shielding materials as described
in the preceding paragraph increases the neutron and gamma-ray leakage from
the reactor shield by a factor of 10, but allows the relative leakage of
gamma rays and neutrons to remain the same, 1.e,, 75% gamma rays and 25%
neutrons in terms of radiation dosage per unit time. The crew shield, ;however,
is designed so that the allowable 4 r/hr received by a crew member will be
caused 50% by neutrons and 50% by gamma rays. Therefore the crew shield must
attenuate the gamma rays by a factor of 60 and the neutrons by a factor of 20
in order to give the required total attenuation factor of 40. The required
thicknesses of shielding materials around the crew compartment (sides, rear,
front) were calculated using the same methods developed by W. B. Thomson
and H. E. Stern in computing the crew shields for the air-safe shields de -
scribed in: ANP-53. In this case, of the neutrons which penetrate into the
crew compartment, 60% were allowed to come through the sides, 10% through the
rear face, and 30% through the front of the compartment. These percentages

v . . - .
were chosen rather arbitrarily, but even an optimum choice of these parameters

(4) ANP-53, op. cit., Fig. 5.
(5) ANP°531 op. Citov Fig°v4°
(6) - ANP-53, op. ctt., p. 53.

103
would not greatly lower the crew shield weight. The calculations indicated
that 5, 15.2, and 1.7 in. of plastic were required at the sides, rear, and
front of the crew compartment, respectively. The plastic was assumed to be
polyethylene, (CH,), , having a density of 0.93 g/cc. All the gamma rays
entering the crew compartment were assumed to penetrate through the rear face,
because the plastic at the sides and at the front was more than enough to
stop all but a negligible portion of the scattered gamma radiation. Therefore
lead is used only at the rear of the crew compartment, where a 2.3-in. thick-
ness is required in addition to the 15.2 in. of plastic which also affords

some gamma shielding.

The crew compartment of this study was that used(7) for the divided
shield for the Na-Na cooled reactor. Based upon this configuration and the
thicknesses of shielding materials indicated above, the weight of the crew

shield was calculated to be 22,000 1lb, 1.e.,

Plastic 13,100 1b

Lead 8,200

Aluminum Structure (Estimated) 700
Total 22,000 1b

Thus the total divided shield weights for the three cases being considered
are 127,000, 136,000, and 111,800 lb, respectively.

If polyethylene, (CH2)n, is used in the outer portion of the reactor
shield instead of gasoline, the shield weights tabulated on p.102 of this
report will remain very nearly the same. The resultant shields, however,
would be about 8 in. smaller in diameter. This would require, roughly, an
additional 30,000 1b of the plastic per shield, at a cost of approximately
$3.50 per pound.

LID TANK
C. E. Clifford E. P. Blizard
J. D. Flynn T. V. Blosser

Physics Division.

L. H. Ballweg, USAF

A shield design consisting of iron next to the core, followed by well-

borated water, was tested in the Lid Tank. Such a design, proposed by the

(7)  ANP-53, op. cit., p. 68,
104
(8)

TAE, was attractive because of its simplicity and low weight, although
there was a recognized cuestion regarding its effectiveness upon intermediate-
energy neutrons. These latest Lid Tank measurements have shown that the
intermediate-energy neutrons can penetrate the iron to such an extent that
supprression of their capture gammas is virtually impossible. 1In another
experiment tle attenuation of neutrons anc ga.mas in solid - T anc water was

measure <,

iros 2.7 Lorntel 4~ter. The Technical Advisory Foard included in its
report a shiel” desipn which was very attractive from the standpoint of
simplicity and ease of construction, which exhibitecd the smallest overall
diameter ¢’ any recent cdesign, anc¢ which appeared to be in contention on a
weight buocis as well. Therefore it was decicded that this shield should be
mocked-ur and tested (ixnt. 1¢)€9) during the waiting veriod between promise

and celivery of the I, C slabs.

The TAF slhield design consists simrly of 55 cm of KFe next to the core
followed by 35 cm of well-borated H,0.(19) The shielding Group at CENL was
strongly of the orinion that the design violated the principle that gamma-
attenuating material could not ne lumped next to the core because of the great
importance of secondary gammas produced at the outside edge of this material

from neutron capture or inelastic scatterinz. The TAP thesis was that

1. Tle iron would attenuate fast neutrons adequately to eliminate
the latter, and on this there was agreement.,

2. Sufficient boron could be located outside the iron to suppress
the captures, and on this there was not agreement.

It should be rointed out that on item 1 the TADIl was able to make a calculation
which proved to be reasonalkly close to the experiment, but on item 2 the
information required for acetailed calculation included capture cross sections
in the intermediate-energy range whicl were comrletely unavailable, as well

as some estimate of neutron-energy spectra which would likewise have been very
difficult to obtain.

(8) Report of the Technical Advisory Board to the Technical Committee of the Aircraft Nuclear Propul-
siton Program, *WP-57 (-ug. 4., 1950).

(9) Blizard 7. P., end Clifford, C. F., Measurements of an [ron and Forated Water Shield in the Lid
Tank, CF-50-12-47 (Tec. 14, 1950).

(10) ANP. 53, op. cit., p. 29.

105
Accordingly a shield mock-up was measured in which the first section
consisted nearly entirely of iron; a second, transition, region was made up of
varying thicknesses of B,C; and the third region was the usual borated water.
The iron samples used in the experiment were in the form of rectangular plates
7/8 in. thick by 56% in, high by 66% in. wide, the total number available for
the experiment being 24, These plates were placed in the tank adjacent to the
source, four at a time. When they were placed in the Lid Tank it was found
that since the plates were not perfectly flat there remained, on the average,
a 1/8-in. water gap between plates. Thermal-neutron measurements, gamma
measurements, and fast-neutron dosimeter measurements were taken in the water

behind the iron.

A layer of B,C 1/8 in. thick, in the form of plastic 50% B,C by volume,
was fastened to the rear face of the last plate in each configuration to re-
duce capture gamma production in the iron at this most sensitive position.
After the addition of 24 iron slabs the gamma level was still very high; -
therefore a layer of B,C (density 1.8 g/cc) in the form of 1-in. slabs was
added up to a thickness of 5 in.

Since the 1-in. B,C slabs were not watertight it was necessary to place
them in a 15-in. by 6-ft by 5-ft dry tank with 1/8-in. steel walls. 1In order
to avoid the introduction of a large void, the dry tank was so placed that 1t
also contained the outer slabs of Fe. At this point it became clear that the
gamma level could not be reduced sufficiently by a reasonable further addition

of B,C, and the design was considered inadequate.

Since the walls of the dry tank are bowed by the water pressure, 1t was
difficult to determine the exact amount of water displaced. For this reason
two measurements were carried out (Table 5.2 last column, Table 5.3 first
column) with and without water in the dry tank. The effect of the addition

of B,C slabs is best inferred by comparison with the "no-water" condition,

Figure 5.1 and Table 5.2 give the gamma attenuation data for the fore-
going experiments, and include as well a comparison of gammas in pure and
borated water. Figure 5.2 and Table 5.3 show more truly the effect of the
B,C slabs.

Figure 5.3 and Tables 5.4 and 5.5 show the attenuation of neutrons in
H,0 behind the various thicknesses of Fe, demonstrating the well-known su-
periority of Fe over H,0 which accounted for the thinness of the TAB shield

design.

106
Figure 5.4 and Table 5.6 show neutron attenuation behind Fe and B,C,
and from this figure the effect of the latter on the emergent neutrons 1is

evident,

The reason that the Fe cannot be lumped next to the core now becomes
evident. Examination of Figs. 5.3 and 5.4 shows that the neutrons emerging
must be predominantly of energy in the neighborhood of 1 Mev since the neutron
attenuation next to the Fe is very great, by both H,0 and B,C. This is not
unexpected since Fe possesses a "window" in this neighborhood just below the
.inelastic scattering region and above that of great absorption. Hence these
intermediate-energy neutrons were penetrating the iron in such large numbers

that the suppression of their capture gammas 1s essentially impossible.

To prove that the gammas observed at the outer side of the shield mock-up
were produced as supposed, the experiment was extended by remeasuring the
gammas as slab after slab of Fe was removed (and, of course, replaced by borated
water) from the source side of the innermost (Fe) region. Since H,0 does not
possess this low-energy window, it would be expected that the removal of the
iron would result initially in a reduction of gamma intensity at the outer side
of the shield. Such was observed to be the case throughout the removal of
eight slabs, as is shown in Fig. 5.5 and Table 5.7. A schematic drawing of

the slab arrangement is shown in Fig. 5.6.

The investigation of the Fe-H,0 system was extended by a measurement of
neutrons and gammas behind a region of 50% Fe by volume, and then by a measure-
ment of the relative effectiveness of the iron as compared to water at various
positions within the mixed region. This latter amounts merely to a determi-
nation of the I described in the last quarterly report(ll) as the important
parameter for shield optimization. The results are shown in Figs. 5.7 and 5.8,
which correspond to Tables 5.8 and 5.9, respectively. A schematic drawing of
the slab arrangement is shown in Fig. 5.9. Dosimeter measurements of the
Fe-B,C system are shown in Fig. 5.10 and Table 5.10. Since some difficulties
were encountered in the electronic circuits during these measurements, they

are probably not accurate to better than *30%.

(11) Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending August. 31, 1950,
ORNL-858, p. 17 (Dec. 4, 1950).

107
80l

ROENTGENS 7 HOUR

20

J 4 X o RO

Pure Ho0

HoO +0.6%
4 7/8-1in.
8 7/8-1in.

12 7/8-in.
19 7/8-in.

24 7/8-in.

24 7/8-in.

24 7/8-in.

30

KEY

Boron
Fe Slabs in Borated Water
Fe Slabs in Borated Water
Fe Slabs in Borated Water
Fe Slabs in Borated Water
Fe Slabs in Borated Water
Fe Slabs, 4 1l-in. B4C Slabs
Fe Slabs, 5 1l-in. B4C Sl abs

40 50 60

70

DWG. 9978

FIG. 5.1
EXPERIMENT 10
COMPARISON OF GAMMA ATTENUATION OF Hx05+ 0.6 %

BORON; Fe-H,0 +0.6%BORON,; Fe-B,C-H,0+0.6%BORON
GAMMA ¢ MEASUREMENTS
10'2 ION GCHAMBER 10'© ION CHAMBER NORMALIZED
GM TUBE NORMALIZED

 

80 90 100 Ito 120 130 140 150 160
Z cm FROM SOURCE
601

Ccomparison of Gamma Attenuation by nzo; H,0 + 0.6% B; Fe-H,0 + 0.6% B: Fe-B,C-H,0 + 0.6% B

TABLE 5.2

(10*2 and 10!° Ton Chambers Normalized, GM Tube Normalized)

 

 

 

 

(roentgens/hour)
- 24 Fe SLABS,
DISTANCE . . . | ¥ATER TN
CHOM SOURCE. | NO SLABS 4 Fe SLABS 8 Fe SLABS 12 Fe SLABS 19 Fe SLABS*|*DRY” TANK
Com) 10°r1.c | 0°%rc | 10°1c | 100r1c | 10°1c | 10%1c | 1001c | 10%1c | w0?1c | 10%1c
30 3.99 3.98 2.48
40 2.10 2.14 1.21 0.686
45 3.81 3.95
46
50 1.22 1.20 0.671 0.335
55 2.19 2.29
60 0.721 0.703 0.397 0.191
65 1.27 1.26
70 0 447 0.431 0.241 - 0.111
75 0.765 0.782
80 0.273 0.273 0.151 0.070
83.5 0.0227 0.00637
85 0.474 0.485
90 0.174 0.175 0.097 0.041 0.0434 0.0170 0.00466
95 0.299 0.305
100 | 0.117 0.113 0.064 0.0608 0.0278 0.0103 0.00302
105 0.181 0.205
110 0.0713 0.0771 0.0413 0.0407 0.0186 0.00655 0.00159
115 0.123 0.126 0.00120
120 & 0.051 0.0516 0.0276 0.0275 0.0133 0.00445
125 0.076 0.087 |
130 0.0359 0.0355 0.0187 0.00852 0.00295
135 0.063 | 0.058
140 0.0249 0.0245 0.0133 0.00575 0.00206
150 0.00917 0.00394 0.00142 -
160

 

 

 

 

 

 

 

 

 

 

 

* 7/8 in. thick.
Oll
ROENTGENS / HOUR

53.66cm. OF ——
SOLID Fe

0.25cm. OF WATER
BETWEEN EACH
Fe SLABS

0.25 cm. OF WATER
BETWEEN EACH SLAB

| I/8" Fe

20

FIG.5.2
EXPERIMENT 10

COMPARISON OF SOLID Fe &
SOLID Fe + B,C

GAMMA ¢ MEASUREMENTS

- III
OCCUPY 6" OF SPACE. |" EXTRA

é" STEEL
IS TAKEN UP IN 50 % AL & 50% AIR

——— TANK WALL

HpoO WITH 0.6% BORON

TANK SIDE (
BY WATER
REAGHES B4C

4 -B4C SLABS I"THICK EACH
WITH Icm.OF AL AND AIR

STEEL) BOWED IN
0.25 cm. OF AIR E UNTIL IT
BETWEEN EAGH
Fe SLAB INSIDE

TANK

Fe

" L“
L' sreeL /f L STEEL TANK WALT

NK WALL

0 24 7/8-1n.
o 24 7/8-1n.
A 24 7/8-1n,

0.25 cm. OF AIR BETWEEN
EACH SLAB IN TANK

16.0 cm.OF AIR

%" STEEL TANK WALL

30 80 90

Z cm FROM SOURCE

40 50 60 70 100 1o i20 130 140

DWG. !

KEY
Fe Slabs, 5 l-in. B4C Slabs
Fe Slabs, 4 1-in. B4C Slabs
Fe Slabs (air in tank)

 

150 160 70
TABLE 5.3

Effect on Gamma Intensity of B,C Added Outside Fe Region

(10'2 Ton Chamber)

 

z

(roentgens/hour)

 

 

 

DISTANCE FROM N%4w£%E§L?§§b§Y“ 24 Fe SLABS,* 24 Fe SLABS,*
SOURCE (cm)’ TANK 4 B,C SLABS** 5 B,C SLABS**
80 0.0058
81 0.0257
81.2 0.00452
90 0.00995 10.00302 0.00260
100 0.00516 0.00170 0.00154
105 0.1329
110 0.00293
120 0.00180
130 0.00107

 

 

 

 

* 7/8 in. thick.

** 1 in. thick.

111

 
2l
RELATIVE THERMAL NEUTRON FLUX

35 SOLID
Fe

3
‘g TYBOR

DWG 9975

F1G. 8.3
EXPERIMENT 10

NEUTRON ATTENUATION IN Hx0O
FOLLOWING VéR‘l_OUS THICKNESSES
OF Fe

THERMAL NEUTRON &
25"BF; COUNTER 8 BF,
COUNTER NORMALIZED
Fe ADDED AS 7/8"SLABS CLEARANCE BETWEEN

SLABS WAS~.25cm & WAS FILLED WITH
BORON IN ALL CASES

H20 WITH 0.6 % BORON
WITH 0.6 % BORON

«25 cm. Hp0 1" STEEL STEEL TANK WALL
BETWEEN EACH 8 waLL
al'souo Fe

O WITH 0.6 % BORON

§ STEEL TANK wALL Hp0 WITH & STEEL TANK WALL
0,6% BORON

O WITH 0.6 % BORON

Ho O WITH 0.6 % BORON

No slabs
4 7/8-1in.
8 7/8-1n.

KEY

Fe Slabs
Fe Slabs

12 7/8-in. Fe Slabs
19 7/8-in. Fe Slabs

24 7/8-in. Fe Slabs (water in tank)

 

70 80 90 100 1o 120 140 150 160 170

Z cm FROM SOURCE
TABLE 5.4

Neutron Attenuation by Solid Fe-H,0 + 0.6% B

[8-in. BF, Counter Normalized to 25-in. BF; Counter (i.e., X 35.7)]

 

(counts/minute)

 

 

 

 

DISTANCE. FROM 12 Fe SLABS*
SOURCE (cm) NO SLABS | 4 Fe SLABS* |8 Fe SLABS* RUN 1 RUN 2
40 15,200,000 12, 900, 000 23,000, 000 62, 900,000
40.6 65, 600, 000
45 34,900,000
50 3,380, 000 2,450,000 2,221,000 7,990,000 8,340,000
55 1,785,000
60 803, 000 515,000 356,000 494,000
70 205, 000 123,000 73,100 67,700
80 55,500 31,700 17,974 8,380
90 10,900 2,920

 

 

 

 

 

 

* 7/8 in. thick.

113

 
TABLE 5.5

Neutron Attenuation by Solid Fe-H,0 *0.6% B

(25-1in. BF; Counter)

 

(counts/minute)

 

 

 

 

. 24 Fe SLABS,*
DISTANCE. FROM WATER. IN
SOURCE (cm) NO SLABS 4 Fe SLABS* 8 Fe SLABS* 12 Fe SLABS* | 19 Fe SLABS* “DRY” TANK
60 740,247 482,082 373,855 489,661
70 199,753 119,707 80,119 65,766
80 54,668 32,274 19,104 12,745
82.1 3,685
85.9 17,439
90 16,038 9,032 4,991 2,970 1,754 5,176
100 4,692 2,731 1,486 814 346 349
110 1,455 791 430 234 89 45
120 469 264 138 71 26
130 148 81 46
140 54

 

 

* 7/8 in. thick.

 

 

114

 
RELATIVE THERMAL NEUTRON FLUX

70

80

DWG 9976

FIG. 5.4
EXPERIMENT 10
NEUTRON ATTENUATION IN H,0
BEHIND Fe + B4C

THERMAL NEUTRON
¢ MEASUREMENTS
25" BF3 COUNTER

KEY
A 24 7/8-in. Fe Slabs (air in tank)
A 24 7/8-in. Fe Slabs (water in tank)

o 24 7/8-in. Fe Slabs, 5 l-ain. B4C Slabs

 

90 |00 110 120 130 140
Z cm. FROM SOURGE

1S
TABLE 5.6

Neutron Attenuation of B,C Behind Fe

(25-in. BF, Counter)

 

 

 

 

 

 

(counts/minute)
24 Fe SLABS* "24 Fe SLABS,* 5 B,C SLABS**
DISTANCE FROM| WATER IN | NO WATER IN
SOURCE (cm) “DRY” TANK “DRY” TANK RIN 1 RUN 2
84.7 267,022
84.8 269,051
85.9 17,439
90 5,176 1,507,812 87,448 84,191
100 349.3 37,140 2,351 2,062
110 45.1 1,326 122.8
120 125.14 19.9

 

 

 

 

 

 

* 7/8 in. thick.

** 1 in. thick.

116
ROENTGENS / HR

DWG. 9982

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

5 |
| lFess |
ab— L) EXPERIMENT 10
:\o EFFECT OF REMOVING Fe
X \\ SLABS ADJAGENT TO
\ SOURCE
: GAMMA ¢ MEASUREMENTS G M
, \ TUBE NORMALIZED
I [ [ I I
\ KEY
o 24 7/8-in. Fe Slabs, 4 l-in. B4C Sl abs
e e 24 7/8-in. Fe Slabs, 5 l-in. B,C Slabs
A 22 7/8-in. Fe Slabs, 5 1l-in. B4C Slabs
A 20 7/8-in. Fe Slabs, 5 1l-in. B,C Slabs
-3 o X 19 7/8-in. Fe Slabs, 5 l-in. B4C Slabs
. N8 A v s S o
\\ in. Fe abs, in. By abs
8 ¥ 16 7/8-in. Fe Slabs, 5 1-in. B,C Slabs
: \
X
6 .
5 \
J \\
4 AN ?\
°
\ \
3 \VI\ \
AR
NN N
2 AN ‘
WL
N\
I‘ o
4 3 \
10 ) ,
A\
9 A\
8 . J
70 80 90 100 IO 120 130 K40 150 160 170 180

Z cm FROM SOURCE

17

 
8ll

TABLE 5.7

Effect of Removing Fe Slabs Adjacent to Source on Gamma Attenuation

(GM TubegNormalized)

 

DISTANCE FROM

(roentgen/hour)

 

16 Fe SLABS,*

17 Fe SLABS,*

18 Fe SLABS,*

19 Fe SLABS,*

20 Fe SLABS, *

22 Fe SLABS,*

24 Fe SLABS,*

24 Fe SLABS, *

 

SOURCE (cm) |5 B,C SLABS**|5 B,C SLABS** | 5 B,C SLABS**| 5 B,C SLABS**| 5 B,C SLABS** |5 B,C SLABS** | 4 B,C SLABS**|5 B,C SLABS**

83.4 0.00397

84 0.00332
90 0.00335 0.00246
100 0.00168 0.00150
110 0.00215 0.000227 0.000263 0.000285 0.000338 0.000422 0.00107 0.000894
120 0.000140 0.000154 0.000162 0.000187 0.000216 '06000271 0.000634 0.000585
130 0.000102 0.000102 07900110 0.000122 0.000138 0.000180 0.000419 0.000382
140 0.000283 0.000256
150 0.000193 0.000177
160 0.000133 0.000122

 

 

 

 

 

 

 

 

 

* 7/8 in. thick.

** 1 in.

thick.
Sinbiy
. FIG.5.6 DWG. 9983
—12 EXPERIMENT 10

3 SCHEMATIC DRAWING OF LID TANK
-4 WITH 24 IRON AND 5 ByC SLABS

5

6

ETC. WATER LEVEL IN LID TANK

 

 

 

 

 

 

 

w
P NI -

 

16 Fe SLABS HARD AGAINST EACH OTHER
A’ |1 & SOURCE PLATE, SLABS ARE %"THIGK EA.

[T~ 8 Fe SLABS -'g" THICK EA. & 5 B4 C SLABS
4T

1" THICK EA.~ INSIDE INNER TANK. INNER
TANK IS DRY.

Fe

 

 

WATER IN LID TANK
184G CONTAINS 0.6% BORON

S

 

BoGC SLAB |
B4C SLAB 2
B4C SLAB 3
-B4C SLAB 4
+-B4GC SLAB 5

 

 

b
X

/—INNER TANK -g THICK STAINLESS
STEEL SIDES

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

!Lf

 

 

 

 

 

 

 

119
RELATIVE THERMAL NEUTRON FLUX

Os

Ou

|
10

50

A No iron
16 7/8-1n. Fe Slars (50% Fe-H20)
19 7/8-1n. Fe Slabs (87.5% Fe)

0

60

 

KEY

70

DWG. 9977

FIG. 5.7 )
EXPERIMENT 10

COMPARISON OF 87.5% Fe—Ho0 + 0.6%
BORON & 50% Fe-H,0 + 0.6% BORON

THERMAL NEUTRON ¢
MEASUREMENTS 25"
BF3 COUNTER

80 90 |00 1[e/ 120 130
Z cm. FROM SOURCE
120

 

140
‘e

TABLE

5.8

Comparison of Neutron Attenuation by 87.5% Fe-H,0 T 0.6%B vs.
50% Fe-H,0 T 0.6% B

(25-1in. BF, Counter)

 

(counts/minute)

 

 

 

DISTANCE. FROM NO. SLABS, 19 Fe SLABS* 16 Fe SLABS?
SOURCE (cm) 100% H,0 87.5% Fe-H,0 50% Fe-H,0
60 740,247
70 199,753
80 54,668
82.1 3685
86.4 8541
90 16,038 1754 3904
100 4,692 346.2 572
110 1,455 88.8 137
120 469 26.43 38
130 148
140 54

 

 

 

 

% 7/8 in. thick.

121

 
DWG. 9979

! FIG.5.8 |,
EXPERIMENT 10

ATTENUATION OF 50°/oFe—H20+O.6 > BORON
GAMMA ¢ MEASUREMENTS
10'2 ION CHAMBER AND
GM TUBE

10

O

ROENTGENS /7 HR.

1O

 

70 80 90 100 11O 120 130 140 150
Z cm. FROM SOURCE

KEY
4 7/8-1in. Fe Slabs (50% Fe—HQO) o Slab No. 12 Out

A

A 8 7/8-in. Fe Slabs (50% Fe-Hzo) Slab No. 7 Out 12 7/8'in- Fe Slabs
X

\Y

(50% Fe-Hg0)

O

16 7/8-in. Fe Slabs (Water in Tank) All Slabs In
16 7/8-1n. Fe Slabs (50% Fe-HzO) B Slab No. 1 Out

122
gcl

TABLE 5.9

Gamma Attenuation of 50% Fe-H,0 + 0.6%B

(10'? and 10'° Ton Chambers Normalized., GM Tube Normalized: 50% Fe-H,0)

 

DISTANCE FROM

(roentgens/hour)

 

12 Fe SLABS*

 

 

SOURCE (cm) | 4 Fe SLABS* | 8 Fe SLABS* | SLAB 12 OUT | SLAB 7 OUT |ALL SLABS IN | SLAB 1 OUT | 16 Fe SLABS®
84 0.00166
90 0.00128
100 0.124 0.0267 0.00698 000660 0.00464 0.00385 0-00084
110 0.0849 00140 0.00459 0 00429 0.00314 0.00266 0.00051
120 0 0570 00120 0 00308 0.00291 0.00208 0.00179 0.00038

 

 

 

 

 

 

 

 

* 7/8 in. thick.
Fi LID TANK

 

DWG. 9980

WATER LEVEL IN
BOTH TANKS

H,0 WITH 0.6 % BORON

INNER TANK

P i e
Wl gy

,~

 

 

APART

7“

Fe SLABS
8

le
8
SPAGED

8-

— LLLLLLLLL T PPl PP PP 77777 7777777777777

 

Y\\\\\\\\\\\\\\\\\\\\\\\\\\\\\\\\\NW\\\\\\\\

 

(WLLLLLLLZZ PP P 7 77077 7777707777777 7777777777

E\\\\\\\\\\.\\\\\\\\\\\\\\\\\\\\\\\\\\\\\\

 

HOZZ 77777 777 7 7 777 77 7777777777777 7 7777 7 77777

 

 

HZ 77777777 77 777 777777777777 7777 77777777777

 

L LTI PP 7P 777777777777 77777777 72727777727,

 

M P77 7T 7 7 777 777777 77 7 777 72 7 27272 77 7 7272772727277

 

 

 

APART

7“

Fe SLABS
8

7
8-3
SPACGED

LLLLEO 7T PP P77 7777 777 77777777777 777 777 777 727

 

| VL LL Z TP PP P77 777777 7777777777 7 77 777727777277

 

 

HZ Z 777 707777777777 7777777777 77 7 777 777 77777

 

 

 

 

Nl L L7 P 7P 777777 77777777777 77772727 77 7 7777772

 

 

WL LT 77777 7 7 777777777 777777777 77777777

 

I\\\\\\\\\\\\\\\\N\\\\\\\\\\\\\\\\\\\\\\\

LI P77 777 7777777777 777 777 777 7777777727271

 

 

 

\w\\\\\\\\\\\\\\\\\\\\\\\\\\\\\\\\\\\\\L_

 

 

 

VLLLLL L LIl PPl 7772727772727 A

I

 

 

SLAB

S

~ 0
|0 2

(NOT TO SCALE)
FIG. 5.9

EXPERIMENT 10"

SCHEMATIC DRAWING OF LID TANK
IRON SLABS (50% Fe-Hg0)

12 4

LID TANK
WITH 16 7/8~IN.

 

 
Gel

MREP /7 HR

FIG.5.10
EXPERIMENT 10

DOSIMETER ¢ MEASUREMENTS

1w
33 OF Ho0 WITH

0.6 % BORON

14" OF 50% Fe H WITH 0.6 % BORON

w e L L
L"STEEL s STEEL

KEY
No Fe
4 1/8-in. Fe Slabs
24 7/8-in. Fe Slabs, 5 l-in. B,C Slabs
16 7/8-in. Fe Slabs (50% Fe-H90)

IO 20 30 40 50 60 70 80 90 100 110
Z cm. FROM SOURGCE

120

130

DWG. !! b

 

140
TABLE 5.10

Dosimeter Center3line Measurements

 

 

9¢1

 

 

 

 

 

 

 

 

 

 

 

(millirep/hour)
DISTANCE FROM NO SLABS 4 Fe SLABS* 24 Fe SLABS,* 5 B C SLABS** | 50% Fe-H,0, 16 Fe SLABS®
SOURCE (em) '™ RUN 1 RON 2 RUN 1 RUN 2 RON 1 RUN 2 RUN 1 RUN 2
30 494 496
40 107.3 109.1
50 24.2 25 3
60 12.13 6.67 6.38
70 3.05 3.35 1.81 1.87
79.2 0.878 0.8096
80 1.12 1.06 0.498 0.593
81.5 0.0620
81.9 0.0845
85 0.148 0.0554
90 0.29 0.36 0.0389 0.0222
95 0.0126
* 7/8 in. thick.

** ] in, thick.
Boron carbide and Water. Attenuation of neutrons and gammas 1in solid B,C
followed by H,0 containing 0.5% boron by weight is currently being measured in
the Lid Tank (Expt.ll)fhah Quantities of particular interest are the variation
of the biological dose with the thickness of B,C at the outer side of the
B,C-H,0 shield, and the activation to be expected in a coolant (particularly

sodium) located at various depths of B,C.

The B,C samples are in the form of one 20-in.-thick block, 5 by 4 ft,
supplied by KAPL, and seventeen l-in. slabs, 5 by 4 ft, supplied by NEPA. The
20-in. block contains B,C of density 1.95 g/cc plus 0.135 g/cc of occluded
H,0. The NEPA slabs have an average density of B,C of 1.8 g/cc and are less
than 0.5% water by weight.

The amount of B,C in each test is given in Tables 5.11 and 5.12. The
following quantities, which were measured in the borated water behind various

thicknesses of B,C, are shown in the corresponding tables and figures:

 

FIGURE TABLE
QUANTITY MEASURED DETECTOR NO. NO.

 

Comparison of fast-neutron dosim- |25-in. BF, proportional counter 5.11 5.13
eter and thermal-neutron counter| vs. proton recoil dosimeter
measurements for borated water

Fast-neutron dosimeter center- Proton recoil dosimeter 5.12 5.14
line measurements

Thermal-neutron centerline meas- | 25-in. BF; proportional counter 5.13 5.15
urements

Gamma dose centerline measure- 1010 and 1012 ion chamber 5.14 5.16
ments

Neutron absorption by boron - 25-in. BF3 proportional counter 5.15 5.17
‘within solid B,C wi thin §4C jacket

 

 

 

 

 

 

Since the 1-in. slabs were not watertight they were placed in a 15-in. by
6-ft by 5-ft dry tank with 1/8-in. steel walls. The remaining space in the
dry tank was occupied by an aluminum tank full of water. The amount of Fe,
Al, and occluded water in each configuration measured is given in Table 5.11.
Some water always unavoidably remained between the source and the samples and
also between the dry tank and the 20-in. block, since none of the surfaces are

perfectly flat. The amount of this water is indicated in Tables 5.11 and 5.12.

(11ay Blizard, E. P., and Clifford, C. E., Measurements of an Iron and Borated Water Shieldin thelid
Tank, Experiment 11, CF-50-12-48 (Dec. 15, 1950).

127
The experiment (Table 5.17) reversing the position of the boron slabs was
carried out to determine the effect of the relative position of occluded water
in the 20-in. block. Tt will be noted that lower neutron intensities at the
shield exterior (about 140 cm) are obtained with the wet boron preceding the
dry. This is tentatively explained qualitatively as being due to the relative
effectiveness of scattering and absorption in a shield. The difference between
wet and dry boron is that the former has a larger fraction of effective ab-
sorber, since hydrogen degrades neutrons so greatly in elastic collisions.
The effect of reversing the position of the boron slabs can then be thought of
as being due to an interchange of scatterer and absorber between two regions
of the shield, one next to the source and the other intermediate between source
and outer side of the shield. A scatterer is important in a shield only if it
is introduced where neutrons are already collimated, so that deflection seri-
ously affects the probability of its escape. Thus the dry boron, which, com-
paratively speaking, is a scatterer, is less effective in the relatively iso-

tropic flux next to the source.

The measurements with the 25-in. BF, counter enclosed in a jacket of B,C
were taken to determine the activation of a 1/v absorber in neutrons of the
spectrum existing behind a solid B,C shield. Measurements in the water fol-
lowing the B,C will also be taken since these attenuation data give a rough

indication of the spectrum.

128
TABLE 5.11

Description of Materials in B4C-H20 Experiment

 

 

THICKNESS OF |
THICKNESS H2O OCCLUDED .
OF B4C AT IN B4C.OR, THICKNESS THICKNESS
DENSITY 2.5 | BETWEEN SLABS OF Fe OF Al
EXPERIMENT (cm) (cm) (cm) (cm)
20-1in. B4C block 39.62 7.7 1.27
20-1n. B,C block followed by 4 47.03 8.7* 1.91 1.58**
1-in. slabs
20-1n. B,C block followed by 10 57.86 8.7* 1.91 3.46%**
1-in. slabs
20-in. B,C block followed by 12 61.65 8.7" 1.91 3.78%*x+
1-in. slabs

 

 

 

 

 

* About 1 cm of H20 between 20-in. block and tank of 1-in. slabs.

** 4 slabs + 1 aluminum tank containgng 0.315 cm of aluminum.

#** 10 slabs and 1 aluminum tank containing 0.315 cm of aluminum.

**+%* 19 glabs but no tank.

129
TABLE 5.12

Description of NEPA 1-in. B4C Slabs

 

 

_ THICKNESS OF B,C
SLAB IDENTIFICATION | WEIGHT OF DENSITY OF B,C. OF DENSITY 2.5

POSITION NO. B,C (1b) (g/cc) (cm)

1 17 191 1.84 1.7

2 19 190 1.83 1.86

3 16 191 1.84 1.87

4 15 185 1.78 . 1.81

5 4 177 1.70 1.73

11* | 13 197 1.89 1.92

6 20 200 1.92 1.95

7 10 189 1.82 1.85

12* 1 193 1.86 1.87

8 6 172 1.65 1.68

9 9 176 1.69 1.72

10 18 194 1.87 1.90

 

 

 

 

 

* Positions 11 and 12 between slabs as shown
Position ] is adjacent to the source
Aluminum face platés-are 0.062 in. thick; 2 per slab = 0.315 cm/slab
Volume of B,C slabs = 5 x 4 x 1/12 = 20/12 = 1.67 cu ft
Description of 20-in. B,C block:

Density of dry B,C, 1.95 g/cc

Thickness, 20-in. B,C + two )%-in. Fe face plates

Occluded H,0 = 6.48% of B,C + H,0 by weight

g/cm’ of B,C = 20 x 2.54 x 1.95 = 99.06

g/cn? of occluded H,0 = 99.06 x 0.648/0.9352 = 6.90 g/cm’

HéO between block and source, 0.8 cm

Total H,0 = 6.9 + 0.8 = 7.7 g/cm’

Equivalent centimeters of B,C at density 2.50 = 99.06/2.50 = 39.62

 

130
M REP/HR FROM DOSIMETER MEASUREMENTS

40

50

EXPERIMENT 11
COMPARISON OF THERMAL NEUTRON & DOSIMETER

CENTER LINE MEASUREMENTS FOR BORATED WATER

25" BF COUNTER & DOSIMETER

70 80 90 100 1o 120
% CMS FROM SOURCE
FIGURE 5.1I

131

 

130

RELATIVE THERMAL NEUTRON FLUX FROM 25" BFy COUNTER
TABLE 5.13

Comparison of Fast-neutron Dosimeter and Thermal-neutron Counter

Measurements for Borated Hzo, No Slabs

 

FAST-NEUTRON. DOSIMETER,

H,0 + 0.5% B

(millirep/hour)

THERMAL-NEUTRON COUNTER,
H,0 + 0.6% B

(counts/minute)

 

 

 

DISTANCE FROM 8-in. BF; COUNTER |25-in. BF; COUNTER
SOURCE (cm) RUN 1 RUN 2 NORMALIZED TO 25 in.

40 107 91.5 15,230, 000
45 54.9 51.6
50 28.61 27.3 3,380,000
55 15.1 14.7
60 8.1 7.88 803,000 740,000
65 4.19 4.25
70 2.35 2.26 205,000 200,000
80 0.70 0.67 55,500 54,700
90 0.25 0.22 10,900 16,000
100 0.071 4,700
110 0.0241 1,460
120 469
130 148
140 54

 

 

 

 

 

132

 
M REP/HR

40

e 10700

PERIMENT 1|
TION OF B,C-H,0 & 0.5% BORON

FAST NEUTRON DOSIMETER GENTER LINE MEASUREMENTS

KEY
12 1-in. B4C Slabs
No B4C
1 20-in. B4C Block, 12 l-in. B4C Slabs
1 20-1n. B4C Block

 

50 60 70 80 20 100 Ho 120
# CMS FROM SOURCE

FIGURE 5.12

133

130
Neutron Attenuation of B,C-H,0 + 0.5% b

TABLE 5.14

(Fast-neutron Dosimeter Centerline Measurements)

 

.(millinep/hour)

 

1 20-in. B,C SLAB

 

 

 

DISTANCE. FROM 2 1-in. 1 20-in. B,C SLAB,
SOURCE (cm) B,C SLABS RUN 1 RUN 2 RUN 3 12" 1-n. B,C SLABS

40.5 128.3

45 72.1

50 36.9

53.5 30.2 18.9

55 19.6

60 9.90 5.81 6.47 6.29

65 5.23

70 © 2.78 1.43 1.49 1.31

80 0.814 5 0.262 0.326

90 0.265 0.078

90.5 0.340

95 0.145

100 0.087 0.023 0.069

110 0.0154

115 0.0063

120 0.00183*

 

 

 

 

 

 

* Poor statistics.

134

 
RELATIVE THERMAL NEUTRON FLUX

C/MIN.

 

60

DWG. IOIOI

EXPERIMENT 11
ATTENUATION OF B,C-H,0 & 0.5% BORON
THERMAL NEUTRON AND GENTER LINE MEASUREMENTS

25" BF; COUNTER

KEY

. B4C Block, 12 1l-1in. B4C Sl abs
. B4C Block, 10 I-in. B4C Slabs

. B4C B].OC](
. B4C Block, 4 l-in. B4C Slabs

70 80 90 100 1o 120 130
Z CMS FROM SOURCE
FIGURE 5.13

135
TABLE 5.15

Thermal-neutron Attenuation by B4C and HQO + 0.5% B

(25-1n, BF;, Counter in H,0 Behind Solid B4C)

 

 

 

 

 

 

 

 

 

 

 

(counts/minute)
EXPT. 10 EXPT. 11. 1 20-in. B,C SLAB
DISTANCE . FROM NO SLABS* NO SLABS*

SOURCE (cm) (borated water)| (borated water) RUN 1 RUN 2 RUN 3
58.7 1,339, 000** 1,270,000** 1,261,000* *
60 1,280,000+ * 1,252,000 ** 1,181,000**
70 199,753 199,357 224,000 233,000 213,000
80 54,668 53,640 40,600 42,000 38,400
90 16,038 15,178 9,030 9,400 8,975

100 4,692 2,070 2, 400 2,250
110 1,455 705 672 612
120 469 182
130 148
140 54
1 20-in. B,C SLAB, 1 20-in. B,C SLAB, 1 20-in. BC SAB,  [of,poi™ D40
4 1-in. B,C SLABS 10 i-in. B,C SLABS 12 1-in. B,C SLABS B4C'SLAB
RN 1 RUN 2 RN 1 - RUN 2 BARE Cd-COVERED BARE

96. 4 31,400 31,700 49,800 5,350

96.7 5,990 5,840 .

97.4 19,535
100 3,615 3,630 13,700 13,700 35,000 3,107 17,431
105 4,560 9,970 810 |
110 882 867 1,860 1,870 3,100 268 2,922
115 827 1,180
120 230 232 392 394 541 663
125 194
130 67.6 68 99.8 126 175
135 54.9
140 19.7 28 30 32.7 |- 48

 

 

 

 

 

 

 

 

* 0.6% boron was used in this experiment.

** Not reliable owing to counting loss at high counting rate.

136
L€l
MR/ HR

EXPERIMENT I
TION OF B4C-H20 + 0.5% BOF'\’ON

CENTER LINE GAMMA MEASUREMENTS

10'CION CHAMBER & 10'2
ION CHAMBER NORMALIZED

KEY

No B4C

12 1-in. B4C S1 abs

1 20-1in. B4C Block

1 20-in. B4C Block, 4 1-in. B,C Slabs
{1 20-in. B4C Block, 10 1-in. B,C Slabs
1 20-in. B4C Block, 12 l-in. B4C Slabs

 

40 50 60 70 80 20 100 o 120 130 140 150 160
Z CMS FROM SOURCE

FIGURE 5.14
TABLE 5.16

Gamma Attenuation by B4C and H,0 + 0.5% B

(10*° and 10'% Ion Chambers Normalized)

 

(milliroentgens/hour)

 

 

. - . 1 20-in. B,C| 1 20-in.B,C |1 20-in. B,C
DISTANCE FROM 2 1-in. 1 20-in. | SLAB, 4 l-in.| SLAB,10 1-in.[SLAB, 12 1-in.
SOURCE ( cm) NO SLABS | B,C SLABS | B,C SLAB B,C SLABS B,C SLABS | B,C SLABS

42.5 2993

50 2440 1941

55.3 330

60 1380 1152 262

70 825 690 176

80 506 400 119

90 321 255 74.8

92.8 31.4
93 33.8

94.4 46.1
100 200 163 51.3 34.1 25.1 26.0
110 124 105 34.2 22.5 17.1 17.9
115 14.6
120 83 70 23.4 15.0 11.9 11.9
130 53 15.9 10.5 8.4 8.4
140 38 11.2 7.6 5.9 6.0
150 23 7.8 5.0 4.0 3.8
157.7 3.9

 

 

 

 

 

 

 

138
COUNTS PER MINUTE

20

 

30

EWG. 10436

FIGURE 5.15
EXPERIMENT 1i

NEUTRON ATTENUATION OF B, G
25in. BFy COUNTER IN B,C JACKET
CENTERLINE MEASUREMENTS

ECTED
FOR Al Fe 8 H,0

NOT CORRECTED
FOR Al Fe, & Ho0

40 50 60 70 80 90 100
cm. OF B,C OF DENSITY = 2.3

-139-
TABLE 5.17

Hleasuremnents of B4C Attenuation

(25-in. BF; Counter in B,C Jacket Pehind Solid B,C Slabs)

 

 

0¥%1

THICKNESS
OF B,C OF THICKNESS THICKNESS THICKNESS COUNTS/MIN WITH
DENSITY 2. 3* OF Al** OF Fe** OF HZO** 25-IN. BF3
EXPERIMENT (cm) (cm) (cm) (cm) COUNTER
3 in. B,C in jacket 6.30 1.27 0 0.5 7,673,680
1 20-1in. B4C slab 43.1 1.27 1.7
1 20-1n. B,C slab 49.4 1.27 1.46 10.4 1,400.8
in. B,C in jacket
1 20-1in. B,C slab 59.43 2.85 1.46 10.4 323.4
1-in. B4C slabs
3 in. B,C in jacket
1 20-in. B,C slab ' 73.47 5.05 2.095 12.9 29.53
12 1-in. B4C slabs
3 in. B,C in jacket
1 20-in. B,C slab 83.50 6.625 2.095 12.9 7.96
17 1-in. B4C slabs
"3 in. B,C 1in jacket

 

 

 

 

 

 

* Includes B,C in jacket between source and counter.

** If these materials are treated as impurities, corrections should be made using the following relaxation lengths: Al, 12 cm;

Fe, 12 cm; H)0, 4.3 cm.

Description of B4C counter jacket:

Outside dimensions of B,C, 40 in. high by 16 in. wide by 6-1/8 in., rectangular
Inside dimensions of B4C,,25 in. high by 12 in. wide by 3-1/8 in., rectangular
B,C thickness between detector and source, 3 in,

¥-1in., aluminum can over B,C exterior, 1/8-in. aluminum can inside B,C

Density of B,C, 1.90 g/cc

Water content, 1.4% by weight
Measurements of Fast Neutrons in the Lid Tank Using a Sulfur Threshold
Detector (H. E. Hungerford, Physics Division). During the past quarter Lid
Tank Measurements of fast neutrons using a sulfur threshold detector were com-
pleted. The reaction used was the S3?(n,p)P3? reaction. Samples of ammonium
sulfate powder were exposed to fast-neutron flux in the Lid Tank in appropriate
containers. After exposure the radioactive phosphorus was separated from the
inert material by precipitation as ammonium phosphomolybdate. Details of the
separation and preliminary results have already been reported in the preceding

quarterly report.(12)

The purpose of these measurements has been twofold: (1) to test the
feasibility of using such a detector for measuring fast-neutron attenuation,
and (2) to obtain independent measurements of the fast neutrons along the

centerline of the Lid Tank.

Up until the time of the present measurements workers had not had much
success 1in the use of sulfur detectors.(l3) In the present experiments a
technique was developed and refined until a fair amount of reproducibility was
possible. Losses during chemical separation were cut down to less than 2%,
and the counters were standardized using P?? from the Isotopes Division, which
guaranteed that the activity specification was within 3%. This calibration
gives flux values about 15% lower than values of the phosphorus standardiza-

tion previously reported.

It should be pointed out that although the actual reaction threshold
is about 1 Mev, the effective threshold is nearly 3 Mev, so that, for all
practical purposes, the values reported here and previously give the flux for
neutrons of energies of 3 Mev and greater. In Table 5.18 and Fig. 5.18 (p.147)
are presented the most reliable points obtained. Measurements were taken in
the 33% Pb—borated water shield mock-up. Their accuracy is within 25% for

points far from the source, and within 10% for those near the source.

(12) ORNL-858, op. cit., p. 35.
(13) See Teschek, R. F., Radiouctive Threshold Detectors for Neutrons, MDDC-360 (Sept. 17, 1946);

Bopp, C. D., and Sisman, O., The Neutron Flux Spectrum and Fast and Epithermal Flux in Hole 19
of the ORNL Reactor, ORNL-525 (July 28, 1950).

141
TABLE 5.18

Measurement of Fast Neutrons in the Lid Tank

 

 

DISTANCE FROM SOURCE

SHUTTER POSITION (cm) FAST FLUX, nv
Open 4.7 1.10 x 10°
Open 12.0 3.14 x 10°
Open 19.6 1.02 x 10°
Open 30.3 2.32 x 10*
Open 42.0 7.38 x 103
Closed 4.1 7.21 x 104
Closed 18.8 5.54 x 103
Closed 34.2 6.40 x 102

 

 

 

 

 

LIQUID-METAL DUCT TEST IN THE THERMAL COLUMN

R. H. Lewis, NEPA . M. K. Hullings, Physics Division
L. J. Frankwitz, KAPL M. C. Marney, Physics Division

The liquid-metal duct test described in the last quarterly report(14)
is now underway. This experiment is designed primarily to determine the re-
duction in efficiency of a reactor shield when perforated with liquid-metal
ducts, with special emphasis on the activation to be expected in a secondary
coolant such as sodium. Although KAPL is supplying the duct mock-up, the de-
sign is chosen to be of interest to ANP as well as to the Submarine Inter-
mediate Reactor (KAPL project). The sodium in the duct will be simulated by

aluminum powder of proper density.

The first leg of the duct has been measured with air in the 6-in. duct as
well as the 1-in. annulus. Figures 5.16 and 5.17 show part of the data.
Figure 5.16 is the centerline measurement taken in water beyond the end of the
duct along its axis. Figure 5.17 shows the traverses on the lucite fo1l-

mounting plates. These plates are mounted perpendicular to the axis of the

(14) ORNL-858, op. cit.

142
 

| | | —

CENTER LINE EXPERIMENT A

 

0 °-BARE
& - CADMIUM
® —THERMAL

| |11

|

|

 

 

10 \

- A g —>
| [T

| L]

|
|

]
7
|

 

10 \

L1

 

LT

|

 

 

 

 

 

| l | |
10 20 30 40
—— CMS FROM END OF DUCT —>

FIGURE 5.16 NEUTRON CENTER LINE MEASURE"
MENTS FROM END OF DUCT

 

o

14 3
o

 

10

T

|

°=DISC 1
*— DISC 1O
a- DISCIO
A-DISCI¥

[ [ L1

|

 

10

A\

e

-

\

L L]

|

|

 

 

 

N\

 

|

 

 

EXPERIMENT ‘A’ SUMMARY

 

 

 

 

 

 

 

 

BARE |INDIUM
90°r—270°
| | |
20 ~<——DUCT | 20 35
10 O 10
(90°) ~<—————— CMS FROM CENTER OF DUCT — > (270°)

FIGURES.17 RADIAL NEUTRON FLUX FROM CENTER OF DUCT

144
duct at 10-cm intervals. A volume integral of slow-neutron flux obtained from
these curves will indicate the total number of neutrons which have come through
the duct. A second duct with one right-angle bend has been received, and meas-

urements with air in the duct have been taken but are not yet complete.

The effect of varying the shielding material surrounding the ducts will

soon be measured using the B,C jackets now on hand.

A mapping of the flux distribution over the source box has been made with
1-cm? gold foils. The source box contains twenty-four X-10 slugs and provides
a secondary source of approximately 107 fast neutrons/(cm?)(sec) from the

thermally induced fissions.

The aluminum-filled ducts are expected to arrive from KAPL by December 5.
A third duct is now being constructed which will have two right-angle bends.
This configuration will complete the types of duct to be measured in this

experiment as presently planned.
SHIELD CALCULATIONS

Analysis of Lid Tank Data (S. Podgor, NEPA). The analysis of Pb-H,0 and
Fe-H,0 data mentioned in the last quarterly report(ls) was completed and 1is
being issued as a separate report.(16) "Effective” fast-neutron cross sections

for the metals based on the "one-collision" theory of shielding were obtained.

The procedure consisted in obtaining a good empirical fit to the 100% H,O
data and finding the proper metal cross section that would reproduce the
metal-water attenuation data. Analytically this amounted to assuming the fol-

lowing form for the attenuation of a point source of neutrons in water:

Ae~®T + Be AT
P(r) = —m—m—— —
Amr?

(15) ORNL-858, op. cit.

(16) Podgor. S., Analysis of Lid Tank Neutron Data for Lead and Iron, ORNL-895 (Jan. 23, 1951).

145
where r is the distance from the source and A, B, a, and 'S are constants for
the water. The constants were adjusted so that, on integration of the point
kernel over the disk fission source, a fit was obtained to the water attenua-
tion data to within 2%. A constant attenuation for the metal was then inserted
to fit the metal-water data. It was possible to do this to within about 5%.

The values obtained were 3.4 barns for Pb and 2.0 barns for Fe.

The curves in Figs. 5.18 through 5.20 show the experimental and calculated
variation of thermal-neutron flux, at three different distances from the source
plate, as successive lead slabs are added. The random variations in the
experimental values are probably ascribable to statistical and positioning
errors. Two different values for the effective lead cross section are assumed
(3.4 and 3.5 barns) and it is clearly seen that no large variation in this

quantity is possible.

Analysis of Lid Tank Data on Boron Carbide and water (R. Zirkind, Reactor
Technology School student from Bureau of Aeronautics, U. S. Navy). A prelimi-
nary analysis of the Lid Tank data on the attenuation of a 20-in. block of

B,C in H,0 (Expt. 11) was made in accordance with the method described in
ORNL-895.

The 20-in. block of B,C had entrapped about 6% H,0 (by weight). To cor-
rect for this, it was assumed that the dry block had its density decreased
corresponding to the water content, namely'pB4C = 1.95. An effective absorp-
tion (or removal) cross section of 1.2 barns was taken as a reasonable guess,
and the effect of the B,C on the thermal flux in the water was calculated
beyond the B,C. The results of the calculation are given in Table 5.19.

TABLE 5.19

comparison of Theoretical and Experimental Attenuation Data

 

 

DISTANCE. FROM OBSERVED CALCULATED APPROXIMATE
SOURCE (cm) FLUX FLUX ERROR (%)
90 9000 7955 15
100 2260 2251 0.5
110 658 650 1
120* 182 190 4

 

 

 

 

 

 

* Data incomplete.

146
¢ COUNTS/MIN

DWG. l!!SO

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

400r I I I T
o Pb & H?_O EXPERIMENT 8
Z=140
380 COMPARISON OF EXPERIMENTAL VALUES
AT Z = 140 WITH ANALYTICAL FIT FOR
TWO VALUES OF Pb GROSS SECTION
360 e EXPERIMENTAL VALUES
ANALYTICAL o (Pb)= 3.4 BARNS
. ———— ANALYTICAL o (Pb)= 3.5 BARNS
340 [EACH sLAB IS I" THIGK]
\\\. FI1G.5.18
\ ®
\ \
320 \\
\:\.
300 \\ N
\
N\
\
280 2o
\ \
\
\ ®
260 N '\
\
N
\ \
240 \-\ T
N |
N\
220 S
N\
2005 2 3 6 8 10 12 4

NO. OF Pb SLABS
147

 
¢ COUNTS / MIN

1200

1100

1000

900

800

700

650

sl

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

DWG. 10061
| | I l
Pb & HyO EXPERIMENT 8
Z =130
® EXPERIMENTAL POINTS
——— o = 3.5 ANALYTICAL
o = 3.9 ANALYTICAL
FIG. 5.19
\‘
\
\
\®
\
\
@
\ N
\ °
\
\
\
\\\ . ®
\
N\
\\ M .
\
N\
\
N\
N
N\
4 6 8 10 12 14

NO. OF Pb SLABS
148

 
¢ COUNTS / MIN.

DWG. 10062

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

l I l |

3600} Pb & HZO EXPERIMENT 8
2 =120
3500
® EXPERIMENTAL POINTS
——— o = 3.5 ANALYTICAL

o = 3.4 ANALYTICAL
3400 FIG. 5.2 0
3300
3200

v A
3100 \ ® .
\
\
\
3000 \
\
\
\
\
2900 S
\ °

\ Y

\
2800 \

\
\
\\
2700 . .<\<
\ b
\ o o
2600 \
\
\
\
2500 A
0 4 6 8 10 12 14

NO. OF Pb SLABS
149

 
The agreement for Z > 100 cm (or 50 cm of H,0 following the B,C block),
where Z is the distance from source to counter, is excellent; however, for
Z £ 100, the results become progressively poorer. In fact, for Z = 70 cm, or
20 cm of H,0, the error is 100%. This is readily understandable since the B,C
block produces a large build-up of degraded neutrons, which gives rise to a

nonequilibrium region in the initial thickness of the water.

On the basis of the above removal cross section, B,C of theoretical
density, 2.5 g/cc, would exhibit a relaxation length of 6.2 cm. Subsequent
measurements on B,C indicate that the simple picture used for the above anal-
ysis may not be completely correct, and consequently the results may not be

really meaningful. Work is continuing on experiments to settle this point.

Theory of Neutron Attenuation (F. H. Murray, Physics Division). Several
methods are being explored for calculating the neutron attenuation inmaterials
such as B,C with variable cross sections. In previous calculations for mate-
rials with constant cross sections, two terms of the scattering contribution
to the Boltzmann equation were usually sufficient, and this result appears to
justify a simplified approach by means of a small number of differential equa-
tions. Another method based on results for constant cross sections 1s also

being studied.

The general formulas used in the calculations for mixtures of Pb and H,0
were applied to a mixture of equal volumes of Fe and H,0. A brief analysis of
scattering at low energies, £ < 2 Mev, where inelastic scattering in Fe is
relatively unimportant, and with iron considered as a pure scatterer and water
as a pure absorber, indicated an appreciably higher attenuation than was
present at higher energies. From the analysis based on the assumption of
forward scattering in Fe, relaxation lengths for the high-energy flux were

obtained as follows:

 

 

Z (cm) A (cm) z (cm) A (cm)
35 5.3 85 6.2
45 5.6 95 6.3
55 5.8 105 6.3
65 5.9 115 6.4
75 6.1

 

 

 

 

 

 

 

150
The experimental attenuation between 86 and 120 cm is approximately the
same as the calculated high-energy attenuation, but intermediate values show

some differences.

Shield Calculation Methods (A. Simon and T. Welton, Physics Division).
An attempt is being made to set up a flexible and simple method for predicting
shield behavior. Much of the required work was done by the Shielding Group of
the TAB, and this work is being extended so that checks can be made with the
complete lead-water and iron-water Lid Tank data. In this way, it is hoped to

obtain a simple and believable phenomenological shield theory.

A perturbation theory for shields has been set up which allows the cal-
culation of the relaxation length defined by Blizard. This should facilitate

the optimization procedure considerably.

The

solution for the neutron flux at all points of a finite or infinite medium

pucting pifference Equations (D. W. Whitcombe, Mathematics Panel).

which contains a duct cannot be found analytically even for the simplest
geometries. However, numerical solutions can be found by replacing the dif-
ferential equations by the corresponding difference equations. In these com-
putations the medium will be replaced by a grid of area (or volume) elements,
and it is desirable that the geometry of the problem correspond to a natural

grid.

To 1llustrate the procedure consider the drawing

 

el

il

 

il

/1

Void

i

 

 

 

Wil

 

 

11

il

 

 

 

where the shaded material represents concrete, the column of 1’s refers tg a
unit collimated beam source of thermal neutrons, and the column of 0’s indi-

cates that the return current on the outside face of the concrete 1s zero.

151
The diffusion equation

is replaced by

opp - ¢ = 0, o =4 + k*R?
where h is the lattice spacing, p refers to an arbitrary cell, and g is the
sum of the four neighbors of ¢p (i.e., above, below, and on the two sfaes). To
apply these difference equations it is considered that the preceding drawing

is a unit cell that is repeated above and below. The results are shown as

follows:

 

VR

[/ 1LV f11
1 /)/42/8/6/7 Air Void ///1/91/2// 0

 

 

 

 

 

 

 

//////
J11] 170001 ////
1 o 5496 0.3013 0.1483 0
1070 771V /7T

 

 

Because of the symmetry there are only five equatidns and five unknowns. When
the lattice spacing is taken smaller there will be many more unknowns, and the
Fairchild machine will be used to solve the linear equations. A report will
soon be issued which will contain an appendix of problems similar to the above

but where the Fairchild machine obtained phe solutions,

Gamma Activity Due to Fission Fragments (W. K. Ergen, Physics Division).
A refinement of existing calculations(!'7) is needed to determine the gamma
activity due to fission fragments in all cases where servicing has to be car-

ried out in the neighborhood of a shutdown reactor which is enclosed in less

(17) Shielding Board Report, ANP-53, op. cit., Appendix I.

152
than a full ground-safe shield (e.g., reactor in a divided shield, or a par-
tially disassembled unit shield). On the basis of existing literature,(lw
refined estimates of the fission-fragment gamma activity can be made now, but
an experimental check in the swimming pool is still required, and has been

proposed.(lg)

Very short-lived gamma activities of fission-fragment decay gammas are of
interest in the case of circulating-fuel, or homogeneous, reactors. It has
been suggested that they be investigated by means of the pneumatic tube in the
ORNL pile. This facility could also measure the Bremsstrahlung from Li®

betas and settle the question of whether Li® emits a gamma upon deca ,(20)
q g p y

A brief calculation was made as a perturbation on the shield for a
reactor using sodium as both the primary and intermediate coolant, (21) The
perturbation consists in omitting the shutdown tolerance requirements, re-
taining only the running tolerance 'specifications. This saves 5% in. of
lead from the gamma shield around the heat exchanger and 11 in. of B,C in

the neutron shield between the reactor and heat exchanger.

NEW BULK SHIELD TESTING FACILITY

W. M. Breazeale and J. L. Meem, Physics Division

As of December 1 the installation of the reactor and associated equip-
ment was about half completed. Figure 5.21 is a picture of the reactor grid
which will hold the fuel elements. The three cylindrical electromagnets shown
near the top of the picture will support the safety and control rods. The
contractor had not finished the building or installed the overhead traveling

crane but hoped to be finished with this work by December 15.

(18) Gillette, P.R., Final Report, Production Test No. 105-53-P, Measurement of Slug Decay, HW-17781
(May 15, 1950); Ascoli, G., and Sisman, O., Absorption of Radiation from an X" Slug by Lead.
ORNL-53 (May, 1948); Bernstein, S., Talbott, F. L., Leslie, J. K., and Stanford, C. P., teld of
Photoneutrons from U235 Fission Products in Be, CNL-38 (Feb. 20, 1948); Bernstein, S., Preston,
We M., Wolfe, G., and Slattery, R. 'E., “Yield of Photoneutrons from U235 fFission Products in
Heavy Water,” Phys. Rev. 71, 573 (1947),

(19) Ergen, We Ko, Measurements of Fission Fragment Decay Gammas in the Swimming Pool, Oak Ridge
National Laboratory, Y-12 Site, Y-F20-3 (Nov. 28, 1950).

(20) Hornyak, W. F., and Lauritsen, T., “The Beta-Decay of B12 and Ligg" Phys" Rev. 77, 160 (1950).

(21) ANP°539 Op, Cit., Secs 3040

153
 

FIG.5.2I REACTOR GRID Photo 7290

 

154
Since it is expected that ordinary filtered water will be used in the
“pool " at least at first, the fuel elements are being allodized, i.e., treated
with a solution of dichromate salts, to inhibit corrosion of the aluminum.

This work is under the supervision of J. L. English and A. R. Olsen.

Design work at NEPA for a mock-up of the reactor part of a divided shield
is completed and procurement has started. This mock-up conforms closely to
the divided shield described in the Shielding Board Report.(22)

A polonium-beryllium source and nuclear plates are being used to investi-
gate the degradation of the neutron fission spectrum when neutrons are col-
limated under water. If, as seems probable from work elsewhere, the charge in
the spectrum is small, then the problem of providing an underwater collimated

beam of neutrons for the proton recoil counter is solved.

A steel water tank 6 ft in diameter and 8 ft high has been installed in
one corner of the ool” room; it will be equipped with a known source and used

for standardizing chambers and counters.

MOCK-UP OF UNIT SHIELD

A. S. Kitzes, Reactor Technology Division

Preliminary arrangements have been completed for fabrication of a mock-up
of the unit shield which is to be tested in the“swimming pool.” The nuclear
requirements of the shield have now been well enough established to permit

the detailed engineering design to be completed.

The mock-up will be similar to the unit shield described in the Shielding
Board Beport(zs) except that it will correspond to a shield for a 27-in.-
diameter spherical reactor. Moving out from the reactor there will be, in
order, an approximately 7%-in. air space, 3 in. of iron, and nine-1-in. layers

of lead separated by varying thicknesses of borated water.

It is believed that the mock-up can be completed by Mar. 1, 1951, at
which time it will be required to fit the current program of the “swimming

pool.”

‘ (22) ANP°539 op. Cit.:w Pe 64.
(23) ANP-53, op. cit., Appendix A,

155
6. EXPERIMENTAL ENGINEERING

156
6. EXPERIMENTAL ENGINEERING

H. W. Savage, ANP Division

Implementation of the Experimental Engineering Laboratory in Building
9201-3 at Y-12 has continued throughout the past quarter. Operation of some
liquid-metal systems has been possible since mid-September with continuous
round-the-clock operation after October 23. All experiments so far performed
have tested the compatibility of materials with hot liquid sodium, but prep-
arations are underway for using liquid lead and liquid lithium, and preliminary
consideration 1s being given to systems for sodium hydroxide and for mixed
sodium and lithium fluorides. Effort is now shifting toward more complex

systems comparable to those anticipated in an ARE installation.

Equipment has been designed and is now being fabricated or installed for
inserting test samples of materials 1n either dynamic or static hot liquid
metal for corrosion, erosion, self-welding, and stress-corrosion tests. In
certain equipment it will also be possible to move exposed materials with
relation to each other, and also to develop adequate pressure, temperature,
and flow measuring devices. Pumps, valves, flanges, joints, 1insulation,
instruments, and purification equipment are being developed, and the techniques
of pretreating materials and postexperimental examination of used items are

being learned.

An adequate power source for electric heating of loops 1s 1n process of
installation, hoods for adequate ventilation are being fabricated, electro-
magnetic pumps are on order, tubing for loops 1s on order, and a small machine

shop for experimental fabrication is being installed.

Considerable attention is being given to the hazards involved in handling
the coolants proposed, and the Safety Committee, formed during this quarter,
is 1in process of developing adequate protective clothing, fire-fighting

equipment, and disposal and cleaning equipment.

Corrosion Tests: Harps (W. C. Tunnell).: Corrosion effects of liquid
sodium, flowing because of convection, on various metals are being examined in

"harps." Each harp is run to failure or for 1000 hr minimum, and the changes

157
in the metal and the sodium are determined by metallographic and spectrographic

analysis.

Operation of the harps filled with liquid sodium has been performed for
the ANP Metallurgy Group under mutually agreed-upon specifications, and has
required the development of techniques of cleaning, welding, installation,
loading, instrumentation, bressurizingfi and analysis of used materials. Much
of the equipment was obtained from the Metallurgy Group and revised in Y-12
shops to fit local needs. Ten harps were placed in operation for a 1000-hr
continuous test on Oct 23, 1950, and seven others have been placed in opera-
tion since. Table 6.1 summarizes the types tested and the characteristics of

operation through November 30.

In preparing the first harps for use, it was found that some of the
welding had been done without shielding the arc with inert gas and that acid
cleaning attached slag inclusions in these welds and resulted in many leaks.
Rewelding with heliarc remedied these defects, and cleaning was suspended
later since it was suspected to be also attacking the primary metal. It has
been necessary to revise the design somewhat, and a procedure was developed
to assure consistent treatment, vacuum tightness, and noninclusion of unwanted
metals.  Degreasing is accomplished with trisodium phosphate; oxide removal
appears best accomplished with sodium. since acid pickling often damages the
original material although inhibitors may alleviate this situation, and water

is removed by a methanol and ether wash and subsequent heating under vacuum.

Samples of the sodium used in each harp are taken at the start and termi-
nation of each run and submitted for oxygen and spectrographic analysis.
Table 6.2 summarizes the determinations made through November 30.  The tech-
nique used is patterned after practices at General Electric Co., a sample
being drawn into a double-bulbed tube to minimize oxygen contamination. In
general the results indicate more oxygen present than can exist in solution in
the metal.: This may be due to incomplete filtration; consequently the advan-

tages of a 5- instead of a 10-micron filter will be determined.

It appears to be particularly important in preparing any structure for
test to have full control of all details of the fabricating process to assure

a complete history of the treatment and types of materials included.

158
TABLE 6.1

Harp Operations Summary (Liquid Sodium)

 

 

 

HARP OPERATING TOTAL TIME

NO. MATERIAL TEMPERATURE - TO 12/1/50 REMARKS
{°F) (hr)

1 }316 stainless steel 1350 764

2 316 stainless steel 1350 850

3 316 stainless steel 1500 767

4 |[316 stainless steel 1500 625 Failed in middle of hot leg

5 |304 stainless steel 1500 349 Failed in top of hot leg be-
low cup

6 L-605 alloy 1500 688

7 347 stainless steel 1500 795

8 |Low-carbon iron® 1200 579.3 Failed in middle of hot leg;
excesslve warpage and oxi-
dation outside

9 Nickel 1500 0 Failed; did not reach operat-
1ng temperature

10 |Low-carbon iron 1200 25.4 Failed in side of bottom cup

11 | Nickel 1500 13 Failed in top plate of bottom
cup '

12 | Nickel 1500 164.8 Heater failure, appeared to
be leak; has been refilled
and reoperated

13 |Nickel 1500 223.5 Failure in bottom plate of
bottom cup

14 | 347 stainless steel 1350 725

15 |347 stainless steel 1350 894

16. |304 stainless steel 1350 850

17 1304 stainless steel 1350 890

 

 

 

 

 

* See Fig. 6.1

159

 
091

TABLE 6.2

Sodium Samples from Harps

 

 

02 IN Na (wt. %)

 

 

SPECTROGRAPHIC ANALYSIS

 

 

 

 

 

 

 

 

 

 

 

 

 

 

HARP AT FILLING AT TERMINATION ~ (ug/g of Na) HARP MATERIAL
o Upper Lower Upper Lower &;—?— Ni Cr Fe Mo
1 0.092 0.082 316 stainless steel
2 0.56 0.13 316 stainless steel
3 0.067 0.087 316 stainless steel
4 0.086 *0.025 0.012 *<10 <40 <20 50 <4 316 stainless steel
**0.039 0.032 **<10 <35 <20 35 <5
5 0.032 304 stainless steel
6 0.038 0.023 L-605 alloy
7 0.020 0.013 347 stainless steel
8 0.021 0.017 Low-carbon iron
9 0.033 0.018 Nickel
10 0.047 0.025 0:025 0.025 Low-carbon iron
11 0.033 0.063 P (pereical i) Nickel
12 0.044 0.012 0.016 0.009 Nickel
13 0.027 0.026 8500 (vertical leg) Nickel
3300 (bottom leg)
14 0.037 0.090 347 stainless steel
15 0.025 0.019 347 stainless steel
16 0.037 0.051 304 stainless steel
17 0.052 0.016 304 stainless steel

 

* Harp 4 is still in operation; this sample was taken after 12 hr of operation.

** This sample was taken after 14 hr of operation.
Harp operation has been relatively trouble-free except for the final
failures and plugging of gas lines by condensation of the sodium. The latter
is serious 1in that it prevents release of the system pressure when a leak
occurs, thereby increasing the hazard.. No complete solution to the problem

has been found.

Failures appear to have occurred at welds exclusively—some completely
unexpected because of invisible tube joints. Complete metallographic examina-
tion 1is being performed in the Metallurgy Laboratory. One of the more extreme

failures (harp 8) is illustrated in Fig. 6.1.

In addition to the harps, two other convection loops are being operated.:
These are different from harps in that the liquid metal can be drained from
the loop and the liquid is held in position by inert gas pressure without
requiring valves. Testing of the device for holding the liquid level automat-

ically is one of the objectives of operating this loop.

Figure-eight Loop (W. C. Tunnell). The figure-eight loop is designed
to receive test sections for dynamic corrosion and erosion tests, self-welding
tests, and stress-corrosion tests. One such loop has been delivered and 1is
being installed (Fig. 6.2). Seamless tubing was unavailable; consequently

satisfactory operation at very high temperatures is not anticipated.

An a-c electromagnetic pump patterned after a General Electric pump was
designed for use with this loop. As a part of this problem it was desirable
to join copper connectors to the stainless steel pump cell without using
soldering alloys and fluxes. A satisfactory joint was made using a type 316
stainless steel rod and building a boss on the stainless steel from the rod to
a thickness of about 1/8 in. and heliarc-welding the boss to the copper. The

joint was cooled very slowly to prevent cracking.:

Test sections for the figure-eight loop are shown inFigs. 6.3 through 6.6.
Figure 6.3 shows the method of holding test samples (part 4) ina high-velocity
stream for corrosion and erosion studies. Figure 6.4 shows the method of
inserting two abutting pieces in the stream with the attachments for applying
force to aggravate self-welding by holding the abutting surfaces together under
known pressure. Figure 6.5 shows a test piece in the stream subjected to a
tensile force and corrosion and erosion effects simultaneously. Figure 6.6

shows details of the foregoing test sections and another modification (item 3)

!

161
IPHOTO NO. 62132)f
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FIGURE6.l LOW CARBON IRON HARP AFTER FAILURE

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191

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FIGURE6.6 DETAILS OF PARTS TO BE TESTED

IN

"FIGURE 8" TEST SECTION
in which two pieces can be moved axially one inside the other, so that while

in the stream welding tendencies ofoscillating parts can be investigated.

Calibration Loop (W. G. Cobb). An isothermal loop with electromagnetic
pump, sump tank, and expansion tank has been designed for the ready insertion
of devices for measuring pressures and flow rates (Fig. 6.7). Two venturi

meters calibrated on water will be used as standards.

Seal Test Device (J. F. Haines). A device for testing materials to be
used as seals is shown in Fig. 6.8. It consists of a sump for molten metal, a
vertical overhung shaft, and a holder. One element of the sealing materials
is a ring (9) held in the holder and sealed by gaskets to prevent passage of
the ligquid from outside the holder to the i1nside. The other element of the
seal 1s a cup (8) attached to the end of the shaft so that it can be rotated;
it bears on the stationary ring (9) to form the seal. This type of seal 1is
quite common in other branches of engineering practice, and the effectiveness
of the seal in liquid sodium will be determined during operation by visual

observation of the well around the shaft through sight glasses.

Bearing Tests (J. F. Haines). A thrust bearing adaptable to the seal

test device has been designed, and a radial journal bearing is being designed."

Air Tests (H. Turner). An 800 cfm blower has been set up for the test-
ing of models of the heat exchanger and for testing fuel element configura-
tions for head losses and flow characteristics. Auxiliaries being built for
this setup include entrance and exit conduits, entrance orifices for measuring
and regulating flow, and a multiple manometer board for pressure determinations.
Models for test are being designed, the configurations for which are based on

current ARE design concepts.

Insulation Tests (R. T. Schomer). Because of the volume of thermal
insulation anticipated for the liquid-metal circulating systems. a number of
different types have been subjected to hot liquid sodium to determine the most
practical type.- Present evaluation indicates the use of a combination insula-
tion composed of an inner layer of either Superex or Hi-Temp No. 19 thick
enough to bring the interface temperature down to 1200°F, followed by a layer
of lead-mill or steel-mill slag wool thick enough togive the desired heat loss

from the pipe. For small pipe (about 2 in.) at 1800°F these thicknesses will
   
 
 

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SEAL

 

 

 

 

 

 

TESTING DEVICE
be about 2% and 3 in., respectively. The basic difficulty with most insulating
materials is reaction with the sodium. Table 6.3 summarizes the tests made

using the equipment shown in Fig. 6.9.
TABLE 6.3

Insulation Tests

 

 

TYPE PRIMARY CONSTITUENT | PIPE TEMPERATURE RESULTS
(°F)
Johns-Manville RF 300 Thermoflex| Asbestos 1372 Disintegrated
85% Magnesia Mg0 , 1372 Reacted with Na
Philip-Carey Hi-Temp No. 19 Diatomaceous silica 1100 Gradual hardening; max.
penetration 7 1n.
Johns-Manville Superex Diatomaceous silica 1100 Gradual hardening; max.
penetration % in.
Eagle-Picher DE-85 Block Diatomaceous earth 1100 Gradual hardening; max.
penetration % 1in.
Baldwin-Hill Neoblock Lead-mill slag 1100 Caked and crumbly;
penetration 7/8 in.
Owens-Selvins Kaylo Block Lead-silica ' 1100 Caked and crumbly; pene-
tration 1 in.
Friedrick & Derrick Glass Wool Glass wool 1100 Fused to steel; penetra-
tion Y% in.
Baldwin-Hill Black Rockwool Lead-mill slag Not heated Penetration less than Y% in.
Philip-Carey Gray Wool Blanket Lead-mill slag Not heated Pe;pgration less than
4 1N,

Philip-Carey White Wool Blanket | Blast furnace slag 1
Asbestos ard a sbestos Not heated Violent reaction

Johns-Manville Bauroc Blanket Steel-mill slag Not heated Stuck to metal; penetra-
tion less than Y in.

 

 

 

 

 

 

Tests of screening of various meshes were also made to determine penetra-
tion of a jet of sodium. Single layers of stainless steel screening of 30,
100, and 200 mesh and galvanized screening of 16 mesh did not stop a jet of
sodium, whereas a double layer of 200;mesh stainless steel screening was

effective.-

Pumps (J. F. Haines, W. G. Cobb, and A. P. Fraas).: A program of pump
development has awaited the definition of types desired as well as the determi-

nation of seal and bearing materials. Progress on the latter items has been

171
"PHOTO NO. 62080
“UNCLASSIFIED

 
discussed above.-

Both centrifugal and axial flow pumps are now being considered in ARE
design, and a program of testing is being set up.” In this connection a cen-
trifugal pump has been modified, as shown in Fig. 6.10, to have vanes on the
back face of the impeller and a heat dam in the shaft. Initial trial of this
pump with water failed for mechanical reasons, and the structure is being re-

paired.” The test stand is shown in Fig 6.11.

Flectromagnetic pumps have been considered adequate for the small-scale
test loops and one has been fabricated, some are on order, and one new one has
been received. Pump development has been extensively surveyed at Allis-Chal-
mers Corporation, Milwaukee, Wisconsin, and proposals have been submitted by
them. The so-called "centrifugal electromagnetic" pump is being considered as

a design possibly suitable for ARE use.-

It has been agreed with NEPA to set up a loop by July 1, 1951, capable of
testing a 1600-gpm pump to be supplied by them.-

Purification (P. L. Hill, USAF).: The sodium.presently being used in the
convection harp tests and to be used in future figure-eight tests is received
in dry pack form and is freed of oxygen by filtering at about 250 to 300°F in
the process of filling the test apparatus.- According to the experience of
G.E., this low-temperature filtration removes oxygen to a residual concentra-
tion of approximately 10 ppm. The filter medium is a sodium oxide cake
deposited on a 1l0-micron pore size Micrometallic 316 stainless steel filter.-
No attempt 1is made to remove trace elements from the sodium as received from

DuPont, a typical analysis of which is

Na 99.9% S 15-20 ppm
B 1-2 ppm Mg 1 ppm

Ca 110-350 ppm Ni 2-4 ppm
Cl 190 ppm P 10-50 ppm
Fe 5-10 ppm

The lithium will be freed of oxygen and nitrogen by a filtering process
the same as that used for sodium. The lithium available at present contains

approximately 1% impurities (primarily sodium).-

Purification methods for lead,potassium, sodium and potassium hydroxides,
and mixed fluorides are being studied with no decision as to the most suit-

able method having been reached.-

173
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174
i i R R i

"PHOTO NO. 10406

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The inert atmosphere used over the present sodium-test units has been
lamp-grade argon. which is considered oxygen- free.  Gas-purification systems
will be in operation shortly to ensure the purity of both argon and helium for

use as inert atmospheres.-

The purification procedure will consist inpassing the gas over hot copper
turnings to remove the oxygen, through a cold trap to remove water vapor, and
then over hot titanium turnings to remove residual oxygen and the nitrogen.-
In order to prevent impure gas from passing to the test units in case of power
failure, a NaK scrubber column is installed after the copper-titanium units,
through which the gas will bubble and have the oxygen removed.: A second
column containing lithium or lithium and NaK will be used to remove the

residual nitrogen.:

Disposal and Cleaning Facilities (R. Devenish). Small valves and
similar pieces of equipment are freed of sodium by immersion in a mixture of
butyl and methyl alcohols, followed by water washing to be certain of re-
moving the last traces of sodium. Larger items, such as the convection harps,
are freed of sodium by heating above the melting point of the sodium and
pouring as much as possible of the metal into a container.: The partially
cleaned harp is then blown with steam and washed with water or immersed 1in

alcohol and washed with water.

A sodium disposal unit, Fig.  6.12, with a batch capacity of approximately
100 1b of sodium has been designed. The unit is based upon the disposal
methods used at Schenectady by General Electric in which molten sodium 1s

jetted under water through a steam nozzle.:

The Liquid Metals Safety Committee has recommended a central cleaning and
disposal facility for all groups in the Y-12 Area. The facility will be used
for removing small guantities of sodium from pieces of equipment and for

fumeless disposal of waste sodium.-

The cleaning facility will consist of a graveled area upon which will be
erected a steel barrier to protect operators and equipment from splashing or
from flying pieces in case of explosion. Cleaning will be performed by
alcohol immersion, steam blowing, or water washing or a combination of all

methods. "

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FIGURE s.12 SODIUM DISPOSAL UNIT

177
Waste sodium and sodium-contaminated insulation in small quantities will
be disposed of in an abandoned quarry 1f 1t proves impractical to use the

sodium melt unit.

Safety (P. L. Hill, USAF). Experiments performed by various groups with
methods of extinguishing liquid-metal fires have indicated that Pyrene G-1
powder is suitable on sodium, potassium, lithium. and calcium fires; Ansul
Metal-X 1is suitable on sodium and potassium fires but 1s guestionable on
lithium fires; zirconium silicate 1is questionable on lithium fires; rust-free
iron filings may be suitable on all metal fires. several other materials

warrant further tests.

Tests of hot sodium upon protective clothing are still inconclusive but
preliminary results indicate that chrome leather may be suitable. Further

tests are being carried out.
7. LIQUID-METAL AND HEAT-TRANSFER RESEARCH

179
7. LIQUID-METAL AND HEAT-TRANSFER RESEARCH

R. N. Lyon and H. F. Poppendiek
Reactor Technology Division

During the past quarter notable progress has been made in heat-transfer
research, in the determination of the physical properties of liquids and
solids at high temperatures, and in the development of a loop for circulating

hot liquid metals through the ORNL reactor.

In the heat-transfer work, final stages have been reached in the con-
struction of the general lithium heat-transfer system and in the high conduct-
ance —entrance conductance sodium system. Design of a boiling-liquid-metal
heat-transfer system is near enough to completion to start construction, and
preliminary designs of a sodium hydroxide heat-transfer system and a system
to determine the natural convection in liquid-fuel elements have been com-
pleted. Theoretical analyses of heat transfer have been completed on three
postulated systems which approximate some of the entrance conditions involved
in current ARE core proposals, and a wide rangé‘of duct shapes have been
analyzed with regard to velocity distribution and heat transfer with liquid

metals.

A new physical-properties laboratory has been equipped with benches,
lights, and other services. The specific heat of lithium has been found to be
1.0 £ 10% between 550 and 900°C, and redesign of the specific heat equipment
to give improved accuracy has been completed. Detailed design of equipment
for the thermal conductivity of liquids is complete, as is the overall design
of a high-temperature viscosimeter. Consideration has been given to the prob-
lem of measuring the densities of high-temperature liquids. Lialson has been
established with the Office of Air Research and the Materials Laboratory of
the Air Materiel Command, USAF, who have agreed to begin physical-property
work in their own laboratories and to contract with other research laboratories

for determination of physical properties of materials of interest to the ANP.

As a result of a decision at the last meeting of the Committee on Basic
Properties of Liquid Metals, arrangements were made for the rapid dissemination
of reportable liquid-metal data. Also, organization of a second edition of
the "Liquid Metals Handbook" has started. As with the first edition, the
second edition is to be a cooperative effort with other sites, but with the

overall editorial responsibility at ORNL.

180
EXPERIMENTAL LITHIUM HEAT TRANSFER

C. P. Coughlen

As a result of work this quarter, the instruments for the lithium heat-
transfer system are now 90% complete, i1nitial design of the heat exchanger
has been completed, and the lithium melting tank has been charged with 100 1b
of lithium metal. It was found that portions of the system had to be rebuilt
because of the penetration into the stainless steel of silver from the brazing
metal used in fastening Calrods to the equipment. The rebuilding has been
completed. Enough materials have been tested with lithium fires to show the
superiority of graphite and Pyrene -1 powder as extinguishing agents for
lithium fires. The instrument panels have been i1nstalled and most of the
instruments have been mounted. The majority of the electrical connections re-

main to be made.

In charging 100 1b of lithium into the melt tank, degreasing was found
to be ineffective, presumably because of the large amounts of o1l either
dissolved or absorbed by the lithium slugs. It was necessary to pump the
charge at 1100°F at 5 psi absolute pressure for an extended time before the

o1l was removed.

A test heat-exchanger design has been roughed out, and fabrication will
be started in the near future. The design is a simple annulus with mountings
such that no interior tube supports are necessary. Uniformity of flow 1is
also furthered by the entrance and exit designs which are essentially annular
orifices. Tt is planned to test the system and calibrate the flow meters

before the first test heat exchanger is installed.

During one of the preliminary water tests of the system, a large crack
was found in the weld of the pipe that formed the catch tank. After removal
of this tank and careful study with the assistance of the Y-12 metallurgisté,
the crack was found to be caused by penetration of silver into the stainless
steel. This silver was introduced by the braze used for mounting Calrod
heaters on the outer side of the catch tank. A new catch tank was fabricated
and calibrated, and parts of the remainder of the system were replaced. The
Calrod heaters are now mounted by tack welding small saddles across the heaters
at short intervals and depositing stainless steel by spraying to provide good
thermal contact. The Calrods have been remounted by this method at all

important points.

181
Prerequisite to the developing of an adequate fire control for the several
hundred pounds of molten lithium to be circulated in this system, practical
experience with extinguishing materials had to be obtained. As a result of
tests made during the. quarter, the only materials considered reasonably satis-
factory are powdered graphite and Pyrene G-1 powder, which is largely graphite
powder. Sodium- and calcium-containing compounds were reduced to the free
metal, which posed a serious disposal problem. Magnesium oxide plaster was
rapidly penetrated by burning lithium, while sand and zirconium silicate re-
acted energetically to form a slag which rapidly attacked stainless steel and

frequently resulted in perforation of the stainless steel test pans.

Graphite appears to react with lithium to form the carbide and blanket
the metal. The carbide reacts with water to form acetylene in a relatively

qulet process.

HEAT- TRANSFER COEFFICIENTS

W. B. Harrison

Construction is about 95% complete on a system designed for two primary
purposes: (1) Determination of the highest values of heat-transfer coefficient
which are practically attainable with liquid sodium, and (2) exploration of
the change in local coefficients resulting from thermal effects in an entrance
region. It is expected that heat~transfer coefficients between 200,000 and
500,000 Btu/hr per square foot per degree Fahrenheit will be achieved with

reasonable accuracy.

An assembly of the test section is indicated in Fig. 7.1. Liquid sodium
at 250°F will be pumped through the 1/32-in.-diameter hole in the center of a
copper disk. In order to promote radial heat flow, the disk will be insulated
on each side and cooled by water in the tube around its periphery. Obser-
vations of disk temperature at various radial locations will be made after
steady operation is achieved. From these data and the thermal conductivity
of the copper, heat flux and surface temperature at the copper-sodium inter -
face may be computed. Sodium temperature will be measured so as to permit
computation of the heat-transfer coefficient. The coefficients computed in
this way will be average values over the disk thickness. Disk thickness will

be varied from 1/16 to 1/4 in. in order to explore the entrance effects.

182
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FOR DETERMINING HEAT TRANSFER COEFFICIENT
OF LIQUID SODIUM

183
MEAN CONDUCTANCE DATA USING NaOH

H. W. Hoffman

A tentative design has been made for an apparatus for the determination
of the heat-transfer coefficient for a molten sodium hydroxide system under
forced convection at temperatures from the melting point (604°F) to 1000°F
and possibly higher. The test section will consist of an electrically heated
tube, 0.251in. I.D., 0.375 in. O.D., 25 in. long, with appropriately positioned
thermocouples for the measurement of surface temperature. It has been decided
to construct the system of "L" nickel. The components of the system and their

relative positions are indicated in Fig. 7.2.

The design of a system for the measurement of heat-transfer coefficients
can be divided into five major categories; namely, materials of construction,
methods of handling and moving the fluid through the system, the design of
the test section, methods of introducing and removing heat from the system,

and measuring and control instruments.

In this particular investigation both corrosive and high-temperature
conditions exist. The handling of materials at high temperatures does not pose
too difficult a problem since numerous metals and metal alloys exist which will
withstand temperatures well in excess of those expected in this investigation
(604 to 1000°F)., However, the number of materials able towithstand the corrosive
action of molten sodium. hydroxide is very limited. In a preliminary test at
Battelle Memorial Institute regarding the feasibility of using various materials
for containing molten sodium hydroxide, it was found that silver was best with
a corrosion rate of approximately 0.1 mg/cm?/month, with nickel and graphite
next at approximately 1.0 mg/cm?/month. These three are the only materials

recommended.

Silver will soften at temperatures in the vicinity of 1000°F and must be
used in the form of silver-clad steel tubes. Graphite has the serious dis-
advantage of having very poor mechanical properties, thus necessitating heavy
awkward sections. Nickel possesses none of the disadvantages of the above two,
and it has been decided to use "L" nickel for at least the test section of

the apparatus.

184
G 8l

 

 

 

 

 

 

 

 

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METER

APPARATUS TO BE CONSTRUCTED FROM "L" NICKEL

FIGURE 7.2 PROPOSED APPARATUS FOR DETERMINATION OF HEAT TRANSFER
COEFFICIENT FOR MOLTEN NaOH

08201 'ON 9NIMVYAQ
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The molten sodium hydroxide will be circulated through the system by a
centrifugal sump pump. Since the pump will be in the molten sodium hydroxide,
1t is necessary that the entire casing, in addition to the impeller, be con-
structed of nickel. It is planned to use a pump capable of handling 10 gpm

at a discharge pressure of 15 psi.

The sodium hydroxide will be melted with the nickel sump tank and main-
tained at a temperature of about 800°F. Since it is suspected that Na,CO,
contributes greatly to the corrosive action of NaOH, it will be necessary to
purify the commercial caustic. To prevent formation of Na,0 in the melt, a
positive pressure of "forming" gas (10% H,, 90% N,) will be maintained over

the melt in the sump.

Three tentative designs for the measurement of the flow rate are being
considered. Two involve the use of strain gauges in association with an
orifice plate. Both of these require calibration because of the variable
temperature in the system. The third method uses a volumetrically calibrated

catch tank with a magnetic (or radioactive) float.,

Several alternatives present themselves in the design of the testsection,
The following three were considered:

1. Parallel channels of rectangular cross section

2. Concentric channels of circular cross section

3. Electrically heated tube of circular cross section
While all three systems will give information of value, stress has been placed
on the third system, primarily because of the simplicity of its construction,
the ease of calculation of results, and the elimination of the need fora

supplementary system for adding heat, Preliminary calculations indicate that

a 15-kw electrical system will be necessary.

Currently, consideration is being given to removing heat by radiation

from the system as a whole.

BOILING LIQUID METALS
W. S. Farmer

During the past quarter it has been possible to select an arrangement

of apparatus for determining heat-transfer coefficients to boiling liquid

186
metals. The overall design i1s 75% complete while detail drawings are 30%
complete at this time. A work order for construction will have been 1ssued

by the first of December.

The apparatus arrangement finally selected consists of a 12-in. vertical
single-tube boiler. Heat 1is supplied by radiation from a graphite tube, which
will be i1nside and concentric with the molybdenum boiler tube. The metallic
vapors produced in the boiler will be condensed in a vertical tubular header
consisting of several banks of finned tubes, across which air will be drawn.

The design is shown schematically in Fig. 7.3.

Initial experiments will be conducted on natural convection boiling
outside a l-in.-diameter 6-in.-long tube. The flux will be variable up to
10% Btu/hr per square foot with existing variable-voltage power equipment.

Graphite temperatures up to 5000°F will be required.

NATURAL CONVECTION IN LIQUID-FUEL ELEMENTS

P. C. Zmola

Natural convection will be required in the liquid-fuel elements now under
consideration i1f excessive temperatures are to be avoided at the center of
the liquid., Preliminary design has been completed for equipment to determine
the extent and the effectiveness of natural convection within the fuel ele-
ments. FElectric current will be passed through the liquid in a simulated
fuel element. to generate heat within the liquid itself, and heat will be re-

moved from the surface of the fuel element at fluxes which will exceed 10°

Btu/hr per square foot.

THEORETICAL THERMAL ENTRANCE ANALYSES

H. F. Poppendiek

Three new sets of analytical solutions have been developed for inclusion
in a collection of analytical solutions which is now being compiled. The
collection is to consist of solutions for postulated systems which approximate
actual heat-transfer systems now under consideration in the ANP reactor

development.

187
UNCLASSIFIED
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FIGURE7.3 APPARATUS ARRANGEMENT FOR
BOILING LIQUID METALS

188
An analytical turbulent flow solution has been developed for a system
in which (1) a fluid is flowing uniformly (slug flow) between two parallel
plates, (2) the initial fluid and wall temperatures are uniform, (3) the
fluid suddenly flows over surfaces which are emitting a uniform heat flux to
the fluid, and (4) the eddy diffusivity varies linearly with distance from
the duct wall (the total thermal diffusivity becomes the thermal molecular
diffusivity at the wall). This analysis is in the process of being prepared

as a technical report,

Analytical solutions for temperature distributions have been derived for
parallel-plate and pipe-duct systems in which (1) the fluid is characterized
by a turbulent velocity profile (a power-loss relation) and a high thermal
molecular diffusivity compared to the thermal eddy diffusivity, (2) the initial
fluid and wall temperatures are uniform, and (3) the fluid suddenly flows over
surfaces which are at some new uniform temperature. Local heat flow and
convective conductance solutions have also been developed. These analyses

are being prepared as a technical report.(l)

An analytical thermal entrance solution has been obtained for a system
in which (1) a fluid flowing uniformly between parallel plates is characterized
by a high thermal diffusivity compared to the thermal eddy diffusivity and by
a symmetrical volume heat source whose strength is a function of distance
between the parallel plates, (2) the initial fluid and wall temperatures are
uniform, (3) the fluid suddenly flows over surfaces through which heat is lost
to a heat-transfer fluid on the other side (the heat-transfer fluid is at a
uniform temperature and the duct-wall resistance is not the controlling re-
sistance in the thermal circuit). It is felt that solutions of this type may

be useful in describing heat transfer in reactor liquid-fuel elements.

FLUID FLOW AND HEAT TRANSFER IN NONCIRCULAR DUCTS

H. C. Claiborne

Theoretical investigations of fluid flow and heat transfer in noncircular

ducts has progressed to a point where a status report(z) is being prepared.

(1) Poppendiek, H. F., Heat Transfer, ORNL-913, to be issued.

(2) Ibid.

189
The report will be released before the end of the coming quarter and will con-
tain the following topics: (1) analytical solutions for velocity distribution
for laminar or streamline flow in rectangular, equilateral triangular, right
isosceles triangular, eiliptical, and any circle sector ducts; (2) analytical
solutions for the resulting temperature distribution when heat is transferred
to a fluic¢ flowing with a souare wclccity wave (slug flow) through the same
type of ducts mentioned above with the duct walls at a constant zemperature,
(3) analytical solutions for the resulting temperature distribution when leat
is transferred to a fiuid flowing with a square velocity wave through rec-
tangular, equilateral traiangular, any right triangular, elliptical, and 60°
circle sector (a general solution has not keen obtainable) ducts for the case

of constant leat flux from the cduct walls.

Average Musselt’s moduli for the case of constant flux were found to he
6, 4, 3, 2, and 0.000386 for rectangular, equilateral triangular, 45°, 30°,
and 1° right triangular ducts, respectively, as compared with the value of 8

for a circular pipe.

It can be shown that the slug flow solutions will be reasonable approxi-
mations for the case of turbulent flow of liquid metals for Reynolds’ moduli
below about 10,000, and that heat transfer will be underestimated by these

solutions at higher Reynolds’ modula.

Other more complex problems involving noncircular ducts under consider-
ation are entrance solution, effects of finite wall thickness for the case
of constant flux tothe outer side of the. duct wall, heat transfer to a fluid
flowing with an established laminar velocity distribution, and some possible

approximations for the case of turbulent flow.

PHYSICAL PROPERTIES

A..BR. Frithsen, USAF

During the past quarter a laboratory was established for the measure-
ment of physical properties of liquid metals, liquid salts, liquid caustics,
structural materials, insulators, and other materials that may be utilized 1in
the design of an aircraft reactor. The design of apparatus for the measure-

ment of specific heats, thermal conductivities, viscosities, and densities

190
has been initiated. At the time of this writing 1t was anticipated that the
specific heat apparatus would be completed and ready for routine measurements
with an accuracy of the order of %5% by Jan. 1, 1951. It is expected that the
thermal conductivity apparatus and the viscosity apparatus will be capable of
accuracies of about #10%, and will be completed by Apr. 1, 1951; and that the
density apparatus will be capable of accuracies of about 2% and will be com-
pleted about Apr. 15, 1951. An approximate value for the specific heat of’
lithium has been obtained between 550 and 900°C.

In order to avoid unnecessary duplication of effort, personnel of the
physical properties laboratory are coordinating their efforts with other
agencies also involved in this type of work. As a result of a conference
with personnel from the Office of Air Research and the Materials Laboratory
of the Air Materiel Command, USAF, these agencies of the USAF will in the
near future begin work on the measurement of liquid metals, liquid salts, and
other materials at their own laboratories at Wright-Patterson Air Force Base,
Dayton, Ohio, and also by contracting with various universities and research
organizations. In addition these agencies will contract directly for the
development of physical-property-measurement techniques and apparatus in an

attempt to develop methods with extremely high accuracies.

A survey shows that the following data already have been obtained by

other agencies and are considered acceptable for ARE design studies:

MEASUREMENT AGENCY TEMPERATURE (°C)

Specific heat Burcau of Standards

Sodium 100 to 800

Potassium 100 to 700

Sodium-potassium eutectic 100 to 700

Beryllium 0 to 900
Viscosity Naval Research Laboratory

Sodium 100 to 700

Potassium 100 to 700

Sodium-potassium alloys 100 to 700
Viscosity Critical Tables

Lead : 441 to 844
Density " NEPA

Bismuth m.p. to 1000

Lead m.p. to 1000

Lithium m.p. to 1000

Lead-bismuth eutectic m.p. to 1000

191
(Continued)

MEASUREMENT AGENCY TEMPERATURE (°C)
Density Naval Research Laboratory
Sodium 100 to 700
Potassium 100 to 700
Sodium-potassium alloys 100 to 700

In addition to the quantities listed above, it 1s deemed necessary to
determine the specific heats, thermal conductivities, viscosities, and den-
sities of all materials of interest in the design of an aircraft reactor,
and, in addition, 1t will be necessary eventually to extend the above infor-

mation to 1000°C in each case.

Ssecific Heat (R. F. Redmond, L. F. Basel, J. Lones, A. Bates,* D,
Smith,* W. H. Bowman,* D. James,* and J. Roarty*). The specific heat project
has progressed to the stage where the method of determination has been selected
and the apparatus has been designed, constructed, and used to determine the
specific heat of lithium over the temperature range 550 to 900°C. At present

the apparatus, a Bunsen ice calorimeter (Figs. 7.4 and 7.5) is being redesigned.

The enthalpy values for lithium, as obtained experimentally in the
temperature range 500 to 1000°C, are shown in Fig. 7.6. A careful statistical
analysis of these points in this figure indicates that the best curve through

them is a straight line, and, inasmuch as the specific heat is defined as

oH
P absodute '5'?

p

C

then the slope of this straight line, which is 0.964, is the specific heat for
lithium over the temperature range. However, the calorimeter was not cali-
brated, and subsequent experiments with metals of known specific heats showed
the calorimeter to be 5 to 8% low. Calibration of this calorimeter by electri-
cal means evidenced excessive heat leakage. Since minor alterations did not
materially decrease this leakage, the calorimeter has been redesigned with
emphasis given to the improvement of the rate of heat transfer from the heat
capsule to the container. The new calorimeter together with refimements in
technique will, it is hoped, substantially eliminate the previously encountered
experimental errors.

*From MIT Practice School.

192
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193
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TEMPERATURE , (°C)

ENTHALPY OF LIQUID LITHIUM

FIGURE 7.6
Thermal conductivity (L. F. Basel and M. Tobias). As a result of a
survey of possible methods for the measurement of thermal conductivities of
liquid metals from their melting point to 1000°C, a steady-state linear flow
system, similar to one used by Dr. Deem of the Battelle Memorial Institute,

has been selected. The design of this equipment is now nearing completion,

A schematic diagram of the Deem apparatus is shown in Fig. 7.7a. Heat
flows from the upper heating plate down through the liquid, through the Armco-
iron standard, to the copper cooling chamber. The heating plate is movable
so that the liquid-specimen thickness can be varied. Thermocouples are lo-
cated in the Armco-iron standard and in the metal surfaces above and below the
liquid specimen, The resistance to heat flow between the two specimen thermo-
couples can be considered to be the sum of the specimen resistance, x/kA,

and of the fixed resistance of the two liquid-solid interfaces, Ff.

It 1s clear that

AT /Q = Ry + x/kA

where

>
~3
i

temperature difference between specimen thermocouples

Q = total heat flow through specimen

Rf = fixed resistances at liquid-metal interfaces
x = specimen thickness
A = specimen cross-sectional area
k = specimen thermal conductivity

By making several runs at different specimen thicknesses but at the same mean
temperature, a plot of AAT;/Q vs. x can be constructed; the slope of the line

through the points is the reciprocal of the thermal conductivity.

Tests of an unsteady-state method were made at a temperature of approxi-
mately 250°C on Armco ingot iron using the apparatus shown schematically 1in

Fig. 7.7b. The block was heated initially to ¢ constant temperature and,

196
UNCLASSIFIED
DRAWING NO. 10284

 

 

AUXILIARY HEATER/

000

<—FURNACE

 

 

 

 

 

THERMOGCOUPLES

 

 

B. UNSTEADY STATE APPARATUS

 

HEATER
LIQUID SPECIMEN

 

 
  

 

ARMGCO IRON STANDARD

 

4—=— 040" AIR GAPS

 

 

 

 

 

 

 

 

 

" T~—_WATER GOOLER

A. DEEM APPARATUS
STEADY STATE

FIGURE 7.7 APPARATUS FOR THE MEASUREMENT
OF THE THERMAL CONDUCTIVITY OF
LIQUID METALS

197
at a given time, the auxiliary heater was turned on. Temperatures were re-
corded at each of the thermocouple positions every few seconds. Ideally, a,
the thermal diffusivity, can be calculated from the temperature, time, and
oT 9%t
36 dx?

flow. However, the results were inconclusive because much accuracy was lost

distance measurements, since 1t is given by a = for unidirectional

in taking first and second derivatives. It is believed that this source of

error can be reduced by improved design to give greater precision.

Several other types of unsteady-state systems have been considered, but
in most of these systems it is difficult to maintain the proper boundary con-
ditions. Radial heat flow methods cannot be used with liquids because of heat
transfer by convection. To maintain a stable system, heat must be conducted
unidirectionally from top to bottom of the liquid specimen. A steady state
system, similar to that used by Van Dusen of the National Bureau of Standards
for metals, has been adapted for use with liquid metals at KAPL. In this
apparatus, heat is first conducted through the sample and then through a metal
of known thermal conductivity. In the liquid-metal system a thin-walled con-
tainer holds the sample and a correction is applied for the heat conduction
down the wall. This apparatus cannot be used for liquids of low thermal con-
ductivity since the wall correction would be large and the final determination
would be inaccurate. Since there is some interest in materials of low con-
ductivity, and since this apparatus appears to be complicated, the method was

rejected.

Viscosity. (S. I. Kaplan). Several methods of measuring viscosity were
considered, including capillary-flow, falling-ball, and rotational viscosim-
eters. A literature search was made and numerous manufactures were polled on
the availability of suitable commercial units, but none were found which would
fulfill, without extensive modifications, the requirements peculiar to this
project, namely, corrosion resistance, easy cleaning, ready adaptability to
various ligquids, and the ability to yield results of useful accuracy after

brief preparation.

The initial method finally decided upon will employ a falling-ball
viscosimeter. Preliminary design of the unit has been completed and fabri-
cation is now in progress. The apparatus consists essentially of a stainless
steel tube mounted vertically in a furnace and filled with the liquid being

tested. A radioactive ball, of a diameter small compared with that of the

198
tube, is dropped into the liquid. The time the ball takes to fall a measured
distance through the liquid 1s determined by externally mounted detectors.
The viscosity can then be calculated in absolute units from the known rela-
tienships given by Stokes” law. The dimensions of the instrument are ample to

ensure precise velocity measurements (Fig. 7.8).

As a possible secondary means of measurement,,a modification of a com-
mercial viscosimeter employing concentric rotating cylinders is being con-
sidered, and correspondence with the manufacturers on adaptation problems 1is
being conducted. A third method, involving measured flow through a capillary,
has been used by the Naval Research Laboratory for liquid metals. However,
the apparatus 1is more cumbersome than those prdposed, and 1s more difficult to

change over from one liguid to another.

Density. Methods for the determination of the density of high-temperature
systems are currently being investigated. Possibilities include the weighing
of a suspended bob in the liquid, and determination of the pressure required
to force a gas bubble from the mouth of a submerged tube against a known depth

of liquid.

DEVELOPMENT OF COMPONENTS FOR EXPERIMENTAL HEAT-TRANSFER SYSTEMS

A. R. Frithsen and M. Richardson

A completely sealed rotary pump has been developed for use with experi-
mental heat-transfer systems. Although the previously reported hydraulic
bearings have proved adequate, new bearings which will permit increased flow
are being fabricated. Four electromagnetic pumps have been received from

G.E. and three Flowrators from Fischer and Porter Co.

Pumps. The main efforts during this quarter have been devoted to con-
tinuing the development of a completely sealed rotary type pump since this
type appears to show the greatest flexibility for utilization in experimental
systems. The hydraulic bearings discussed in detail in the previous quarterly

report(3) were tested in the bearing test system and operation proved to be

('3) Frithsen, A. R., “Pump Development for FExperimental Systems,” The Aircraft Nuclear PropulsionProj~
ect Quarterly Progress Report for Period Ending August 31, 1950, ORNL-858, 56 (Dec. 4, 1950).

199
UNGLASSIFIED
DRAWING NO. 10285

DROPP!NG TUBE

/———- FILLER PIPE

INERT GAS M

INLET \

DROPPING

UNIT ——————_

 

 

 

 

 

BALL - ALIGNING
GUIDE

 

 

 

 

 

/\/\.
S\

 

 

 

 

 

N

\——-'MAGNE TIC VALVE

 

 

 

 

 

DAY AN
W

FIGURE7.8 VISCOSITY MEASURING TUBE
200

 

 
satisfactory. Several runs of %4 to 3 hr duration each were made; however,
testing had to be terminated before long-duration runs could be made because

of misalignment in the bearing test system,

A new hydraulic bearing has been designed and fabricated, which is
similar to those previously described except that two rows of longitudinal
pressure slots are used and flow of liguid from each end of these slots is now
possible. This latter feature was incorporated owing to thé fact that work
at Allis-Chalmers Manufacturing Co. on similar type bearings indicates that

performance is greatly improved if the flow through the bearing is increased.

The bearing test system has been modifiéed so as to test these new bearings,
and initial tests indicate that, altbough more flow is required in the new
type bearings, they will operate at a much lower pressure than the original
bearings (1.e., 10 psi as compared with 45 psi). An adapter is now being
fabricated which will make it possible to use the same bearing test system to
test both types of bearings. Thus it will be possible to obtain a quick com-
parison of performance of the two types so as to determine the best compromise

between efficient operation and simple construction.

The design of a pump using hydraulic bearings and having a capacity of
12 gpm at 230 ft is almost complete. The system for testing this pump up to

a temperature of 400°F is in the process of being designed.

Four G.E. electromagnetic pumps have been received and one has been
transferred to the Experimental Engineering Section of the ANP Division for use
in a liquid-metal corrosion testing system, The other three pumps will be

used as the need arises.

Flow Measuring Devices. Three Flowrators (rotameters) have been received
from the Fischer and Porter Co. which were designed to meet the following

conditions:

Fluid Lithium or sodium

Operating pressure 100 ps1

Operating temperature 100°F

Flow 2 Flowrators for 3.3 to 33.0 gpm

1 Flowrator for 7.5 to 75.0 gpm

One of the 3.3 to 33.0-gpm Flowrators will be incorporated in the pump

test system to test its operation in sodium and sodium-potassium alloys.

201
LIQUID METALS IN-PILE EXPERIMENTS

R. F. Rattistella C. D. Baumann
O. Sisman

The loop for circulating liquid lithium through the pile and most of
the auxiliary equipment have been built, Final assembly and testing of this
equipment cannot be carried out until the electromagnetic pump and electro-

magnetic flow meter have been tested and calibrated.

The first calibrating system for the pump and flow meter sprang a leak
at a weld about 4 hr after start-up. No damage was caused by the resulting

fire, but the system had to be completely dismantled and rebuilt.

A second calibrating system is now ready for operation, a sketrh of which
is shown in Fig. 7.9. About 2 1lb of lithium will be charged into the charging
tank, and the entire system will be evacuated before the charging tank heaters
are turned on. The molten lithium will be kept under vacuum at 1000°F for
several hours to drive off the o1l in which the lithium is shipped. The
portion of the system between the charging and collector tanks; including
three traps, will be heated next, and the molten lithium will be transferred
through a micrometallic filter into the metering tank by helium pressure.
To calibrate the flow meter, the lithium will be moved from the metering tank
to the charging tank by gas pressure, The flow rate will be regulated by
regulating the gas pressure at the collector tank, and will be measured by
recording the time required to fill up the charging tank between the two
probes. Three different size pipes, one with a flat section, are provided for
the flow meter to determine which will give the largest signal with the

electromagnetic flow meter.

After the flow meter has been calibrated at several temperatures, the
lithium will be circulated around the closed system by the pump and the flow

at various temperatures will be checked by the flow meter.

If operation of the pump and flow meter is satisfactory, the lithium will
be circulated through the system on a 24-hr basis for two weeks or until a
leak occurs. TIf the system holds up for two weeks at 1000°F it will be con-
sidered safe to operate the in-pile loop for one week at the same tempera-
ture. The in-pile loop will then be tested by circulating lithium through

it for 24 hr, and if no leaks occur it will be inserted in the pile.

202
 

€02

 

 

FAqlile 10286

Vacuu elium
Vacuum Hel i1um

 

 

 

 

 

 

 

 

) = Charging Tank
Metering d yrob B
and A N robe
Surge Tank |[l&£ Contacts &——— Micro-Metallic Filter
_ 2
o ;
3 ~ \ Vacuoump Helium
T Probe
Contact

 

 

 

ClCollector

Tank
Heltm ’MJ

— J

 

 

 

 

 

 

   
  
 

Vacuum . W__Metering
‘\\\\——- MeteringR\“—‘giZ:zzzg Station
EM Station
Pump Cell (Duplicate of
Dump Tank Pump Cell)

 

 

FIGURE 7.9
LITHIUM METEF 4G AHD E.M. PUMP LOOP
8. METALLURGY

204
8. METALLURGY

E. C. Miller, Metallurgy Division

One of the most i1mportant requirements of any structure is that the
various members be mutually compatible. Very little information is available
on the high-temperature compatibility of the various materials now being con-
sidered for use in the ABE. - Consequently, emphasis has been placed on both
static and dynamic corrosion tests of such materials. The static corrosion of
metals, particularly the stainless steels, by sodium is being extensively
studied; 1in general, the attack is much milder than similar tests with lithium.
Straight iron-chromium alloys or ferritic grades of stainless steel appear to
be more resistant than other metals to attack by lead. Similar studies of the
corrosive properties of lithium and tests of various metals in a uranium-

aluminum alloy are reported.:

Dynamic corrosion tests in thermal convection loops are underway but no
results are yet available. The laboratory for the fabrication of fuel elements
will not be completed before February; nevertheless, it is in operation and
some fuel elements have been pressed and extruded.  Static corrosion tests
were performed on combinations of materials pertinent to this program. The
Welding Laboratory and the Creep-Rupture Laboratory are still being set up,

although preliminary experiments have been undertaken by both groups.:

STATIC CORROSION TESTING

J. E. Cunningham

This initial phase of the liquid-metal corrosion testing program 1s de-
signed to sort out potential structural materials for use in contact with
various coolants. Liquid-metal coolants now being considered for use in the
proposed ARE include lithium, sodium, and lead. Because of the contribution
this last heat-transfer-fluid would make toward lessening shield weight, the
activity on corrosion has been shifted to lead. Testing of materials 1in

sodium and lithium is being continued, but at a reduced effort.:

Two important observations made concerning liquid-metal corrosion work

in the past six months are: (1) The depth of attack or penetration in liquid-

205
metal media as measured by metallographic examination 1s a more reliable
criterion for evaluating the corrosion behavior of materials (particularly
alloys) than are weight change data: and (2) in many cases, mass transfer
effects prevent use of an inert or third material in the test system,
and tests must be conducted in tubing containing only the test coolant.
On the other hand, information obtained in this manner may be of little value

in an application requiring the presence of more than one solid metal.:

Materials in Lead. Some of the metallic elements which in the preliminary
screening tests showed promise of containing lead were examined metal lographi-
cally. Tungsten and Armco iron showed no evidence of any form of attack.
The refractory elements columbium, molybdenum, and tantalum, which showed an
appreciable weight gain after test, were characterized by formation of films
or coatings. Columbium and molybdenum exhibited a uniform, duplex type of
coating which was rather soft or spongy in nature, while tantalum showed a
thin, brittle type of film formation. FElements like nickel, which essentially
dissolved during the test, and beryllium and tantalum, which lost appreciable
weight, were not examined. FExamination of the graphite-melted zirconium,

which exhibited a slight weight gain, is incomplete.

Forty-hour tests were performed on several stainless steels (Table 8.1)
to determine their corrosion behavior in contact with 1000°C (1832“F) lead.
Both the austenitic and straight ferritic type were represented, as well as
some extra-low-carbon grades. Wherever possible. tubing served as the test
capsule; otherwise a test specimen plus a controlled amount of lead ("Doe Run™
grade, 99.99% Pb, as supplied by the National Lead Company) were contained 1in
an evacuated capsule of relatively inert Armco iron.  Change in thickness and
range of maximum penetration were determined by metallographic examination.
Al]l specimens were cut diagonally and mounted with protection bars at the ex-
posed surface to reduce edge-rounding during polishing. Since thickness
measurements were made before and after the test, the recorded value gives the
total thickness loss resulting from attack on two sides of the test specimen.-
Penetration refers to the maximum depth of corrosion reached beyond the
attacked surface from one side of the specimen only. Six fields were measured

and the range was reported.

Results indicate that, with the exception of 316 stainless steel alloy,

the austenitic type alloys were characterized by an intergranular type of

206
L0G

TABLE 8.1

Static Corrosion pata on Stainless Steels
in 1000°C Lead for 40 hr

 

METALLOGRAPHIC DATA

 

 

RANGE OF
STAINLESS SURFACE - VOLUME THICKNESS MAXIMUM REMARKS
STEEL ALLOY |  CARDON RATIO WEIGHT CHANGE|  ""CHANGE | PENETRATION
° (cm?/cm3) (mg/ cm®) (em) cm)

304 LC 0.033 1.0 - 14. 62 No Metallography

304 LC 0.033 1.0 -11. 44 -0.004 0.004-0 012 Intergranular type of attack; slight amount of phase
transformation ¥ — a in attacked grain boundary reglon

304 LC 0 033 1.0 -1.03 +).001? 0.004-0.014 Same

304 VILC 0.006 0.7 -10. 86 -0.006 0.015-0.024 Grain boundary attack accompanied by phase transformation
v = a; some pitting or hole formation noted in attacked
grain boundaries

304 VLC 0.006 0.7 -4.81 -0 004 0.021.0.027 Same

405 0.078 1.0 -9. 17 -0.003 0.001-0.003 Light uniform attack

405 0.078 1.0 -13.83 +0.001 0.002-0.003 Light uniform attack

410 ELC 0.041 1.0 +1.21 +0.002 0.003-0.005 Irregular type of attack; formation of blue-gray com-
pound in attacked area, pearlitic type of transformation
immediately adjacent to lead-affected zone; parent metal
shcwed an ocicular structure

410 ELC 0.041 1.0 +2.71 Same

410 ELC 0.041 1.0 +2. 27 Same

430 0.085 1.0 2. 10 0:000 0.002-0.004 Intergranular type of attack which left surface rough in
appearance; pearlitic type of transformation found in
attacked grain boundary region; parent metal showed
martensitic type structure

430 0.085 1.0 3 31 0.000 0.002-0.007 Same

430 ELC 0.8 -5.30 -0.002 0.004-0.009 Irregular type of attack which left surface rough in
appearance; blue-gray constituent formed in lead-affected
zone

430 ELC 0.8 -7.43 -0.002 0.002-0.006 Same

430 ELC 0.8 -7.71 -0.001 0.006-0.008 Same

446 0.084 1.0 3.73 -0.002 0.006-0.013 Irregular type of attack which resulted in globular holes

 

 

 

 

 

 

in lead-affected zone; decarburized to a depth of 0.018 cm
 

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attack. Type 304 stainless steel (18-8) showed a phase transformation from
gamma to alpha in the attacked grain boundary region.  The straight iron-
chromium alloys showed én irregular type of attack with formation of an un-
identified blue-gray constituent or globular holes in the lead-affected zone

In general, the straight iron-chromium or ferritic grades appeared to be more
resistant to attack by lead on the basis of depth of attack and change 1in
weight for short exposures. Numerous cases were encountered, however, 1in
which straight iron-chromium grade tubing collapsed during the test, presumably
because of the high temperature and the pressure differential (1 atm) between

the inside and outside of the test tube.-

Additional work is in progress on 18-8 molybdenum type alloy to check the
encouragingly good resistance shown by this high-strength alloy during initial
testing.- Longer (400-hr) tests have been completed, but metallographic data

and interpretation of results are not yet available.

Stainless Steels in Lithium. The study of the corrosion behavior of the
commercial stainless steel alloys in 1000°C (1832°F) lithium was largely
completed and reported last month. The facts revealed by X-ray and metallo-

graphic examination are summarized below:

1. Attack was largely intergranular in nature, but some solution was
noted.

2. In the 18-8 type of stainless steel alloy, areas adjacent to
attacked grain boundaries and similar exposed surfaces had under-
gone a phase transformation from austenite to ferrite, presumably
as a result of solution or diffusion of an austenite former, such
as carbon or nickel, into the lithium."

3. The structure in immediate contact with molten lithium was
decarburized.

4. PBoth the ferritic and austenitic grades showed grain growth with
increasing time. The grain coarsening was largely confined to the
parent metal with little or no effect on the lithium-affected
zone,

5. In general, the straight iron-chromium alloys (ferritic grades)
appear somewhat more resistant to 1000°C lithium than do the
austenitic grades.

Additional work is in progress to investigate the corrosion behavior of

the extra-low-carbon grade alloys.

209
Metallic Elements in Sodium. Tests were initiated on several elements
and alloys to determine their corrosion resistance to sodium at high tem-
peratures. The sodium used was of commercial purity (99.95%) from Merck and
Company.  The first tests were made with the specimens in Armco-iron capsules,
examination of which showed that Armco would not contain sodium as it had
apparently leaked through the capsule bottom, sometime during the test period.
Metallographic examination showed that these leaks possibly occurred as the
result of reaction of sodium with the manganese sulfide stringers. Because of
the uncertainty of the length of time the specimens had been exposed to the

molten sodium, these tests were discarded.

A trial test was made using nickel for the capsule and it was found that
this material would contain sodium barring any imperfect welds. Since this
time, the tests in sodium have been in evacuated nickel capsules. Preliminary
inspection, on the basis of weight change data only, indicates that nickel has
good resistance and that cobalt, molybdenum, tantalum, Alloy N-155, inconel,

and inconel X have fair resistance to sodium at 1000°C for 400 hr.-

The stainless steels are also being studied extensively in the form of
both flat stock and tubing exposed tosodium at high temperatures. The metallo-
graphic work on the stainless steel alloys has not been completed although
data have been compiled on several special tests that have been run. Two
samples of 304 stainless steel (18-8) flat stock, with an extra low carbon
content (0.006%) were tested for 40 hr in sodium at 1000°C. Both had a high
weight loss.” A heavy precipitate was seen at the grain boundaries throughout
the specimens and a new phase (possibly sigma) appeared in crystallographic
planes as needles. The specimens seemed to be fairly heavily corroded although
the attack did not appear selective. A dark, nonadherent, unidentified film

formed on the specimens. The corrosion depth was 0.006 to 0.008 cm.

Two ferritic type 446 stainless steel samples with low carbon content
(0.006%), flat stock, were also tested in nickel capsules for the same period
and temperature.’ These samples also showed a precipitate both in the grain
boundaries and within the grains, but in these it appeared very angular in
shape; it also could be sigma." Some irregular corrosion which was also seen
was confirmed by the small weight loss.: The surface exposed to the sodium was
rough and had a blue-gray compound adhering to it. The corrosion depth for
type 446 was 0.002 to 0.005 cm.-

210
Two tests were performed on stainless steel tubing filled with sodium.
These were heated for 400 hr at 1000°C (1832°F). The sample of type 310
showed an extensive intergranular attack with large voids in the grain bound-
aries at the exposed surface.  The matrix appeared to have complex carbides
precipitated both in grain boundaries and along crystallographic planes.: The
sample of type 420 showed essentially no attack and the structure was that of
a typical sample of 420 stainless steel.

In general the tests of stainless steels in sodium indicate a mild attack

1n comparison to similar tests conducted in lithium.-

Compatibility tests were run for possible mass transfer between 316
stainless steel capsules and beryllium, niobium, molybdenum, and L-605 alloy
specimens in liquid sodiumat 1000°C for 40 hr. The results are incomplete, but
essentially no attack was shown by the materials tested except extruded
beryllium, which was uniformly attacked and had a general weight loss. Niobium
and molybdenum both had films which were very thin and continuous but which
have not yet been identified. Alloy L-605 was decarburized slightly but
showed no other visible attack. It did show a slight weight gain.

Metals in Uranium-Aluminum Alloy. Some exploratory corrosion tests were
run on a few selected pure metals in the hope of finding potential materials
to contain molten uranium-aluminum alloy (2 atom % uranium by weight).- The

test temperature was 1000°C (1832°F) and the time at this temperature was 4 hr.-

Tests were performed by immersing samples in the uranium-aluminum bath
contained in a BeO crucible under a blanket of inert gas. Metals tested in-
cluded titanium, iron, zirconium, columbium, and molybdenum. Results indicate
that the above materials are not resistant and cannot be used to contain this

uranium-bearing coolant at 1000°C.-

Macrophotographs of the specimens after the test are shown in Fig. 8.1.-

DYNAMIC CORROSION TESTING

Anton Erasiman

The current dynamic corrosion testing has been limited to experimentation
on thermal convection loops. The work is being carried out by the ANP Experi-

mental Engineering Group with metallurgical control and examinationto be done by

211
2le

e

   
 

 

BERYLLIUM

O O

NOT CLASSIFIED
PHOTO NO. Y-2625

 

 

 

  

 

 

 

 

 

 

 

 

 

 

TITANIUM ZIRCONIUM NIOBIUM (COLUMBIUM)

 

 

 

 

  

 

 

 

 

L

 

 

MOLYBDENUM TANTALUM WOLFRAM (TUNGSTEN)

FIGURE 8.1 CORROSION SPECIMENS IN MOLTEN U-AL
| ALLOY FOUR HOURS AT 1000°C

 

 
the ANP Materials Group in the X-10 Metallurgy Division.  No results are yet
available although 19 loops are being, or have been, operated with sodium at
Y-12.- So far, no loops have been examined metallographically. More loops are

on order."

Thermal Convection Loops. The loops are filled with filtered sodium and
operated under a helium atmosphere at temperatures up to 1500°F for 1,000 hr
or until failure occurs. At present, temperature differentials between 60 and

100°C are being obtained around the loops.:

Immediately after the loops are filled a sample of sodium is taken for
determination of oxygen. After the loop has been operated, a sample of sodium
is taken from the top cup for determination of oxygen and samples are taken
from the top and bottom cups for spectroscopic analyses for carbon and metallic
corrosion products. From both the hot and cold zones of the loop metallo-
graphic specimens will be taken from the pipe and welds. These samples will
be examined for corrosion and any evidence of a local build-up of corrosion

products or film formation.-

FUEL-ELEMENT FABRICATION

George Adamson

The equipment for the powder metallurgy laboratory is arriving and is
being placed in operation. At the time of this writing it was thought that
the second room for the laboratory would be available about December 15. The
only major piece of equipment now missing 1s the molybdenum-wound furnace,

which 1s not expected before February.

Some work has started on cold pressing and extrusion of fuel-element com-
ponents. A series of pellets containing various concentrations of U0, and
iron have been pressed and sintered. Photomicrographs show that continuous
iron networks were obtained with 50 and 70% iron but not with 30%.- In all
cases the iron and UO, tend to agglomerate instead of forming a continuous
thin network. The physical properties of these pellets have not yet been
determined. Efforts are being made to obtain various sizes of UQO, particles

since all that are available are very fine, being of the order of 10 microns.

213
Using a 75-ton hydraulic press and makeshift dies, some extrusion work

has been attempted. With carbonyl iron powder and a cellulose acetate binder,
tubes of 1/8 in. diameter and 0.020-in. walls and containing 30% UO, were
made. " The tolerances were very poor but were all that could be expected with
the dies available. A regular extrusion die for a %-in.-diameter tube with
a 0.020-in. wall has been received from the shop and is being tested with iron

powder.-

Static Corrosion Tests. The high-temperature compatibility of the
various materials now being considered for use 1in reactors 1is being examined.
This information must be obtained before fuel-element design can advance
beyond the preliminary stages. A "quick and dirty" investigation to guide in
selection of materials 1is being made at ORNL, while a more complete long-term
program is being launched by NEPA. For this work materials in forms immedi-
ately available were assembled in various combinations inside stainless steel
tubes.- These tubes were evacuated and sealed, swaged at approximately 1000°C
to a 40% reduction in area, and then held at 1100°C for 100 hr to allow for
diffusion or for reactions to taée place.” They were then cut and examined

metallographically.

The results obtained with these combinations,

table,

summarized in the following
are discussed below. " In this discussion the penetration depths are
measured from present surfaces since there is no means of determining the
position of the original faces.' The only stainless steel that has been tested

is type 316.

 

 

 

 

 

 

 

 

 

 

STAINLESS
STEEL 316 Mo Cb
Uuo, Moderate No - Moderate
reaction reaction reaction
BeO No No Moderate
reaction reaction reaction
Mo ' Moderate Not
reaction finished
Cb Moderate _ Not
reaction finished
Be Strong Not Not
reaction finished finished

 

214

 
(1) Stainless Steel--U0, (Fig. 8.2a). There is a reaction at the inter-
face which involves both a general attack and preferential leaching. The
preferential attack is shown by the stringers extending about 0.003 in. into

the material. The dark spots in the stainless steel are carbides.

(2) Stainless Steel-—BeO. There appears to be no reaction at this in-
terface. The surface of the stainless steel did contain a considerable number
of smooth indentations, but these are thought to be a result of folding during

swaging. This conclusion is being confirmed with additional samples.

(3) Stainless Steel—Molybdenum (Fig. 8.2b). The molybdenum showed a
recrystallized structure and in many samples had numerous small tears on the
inner periphery adjoining the stainless steel. These are fabrication rather
than corrosion defects. A third phase, which extended into both metals, was
found in the intersection. Layer depths were of the order of 0.001 in. Some
selective leaching is evident in the narrow stringers extending deeper into
the steel. The number and depth of these stringers varied tremendously from

sample to sample.

(4) Stainless Steel—Beryllium (Fig. 8.3a). The sample of this compact
ruptured after 52 hr at 1100°C. On examination it was evident that an extensive
reaction had taken place and that,quite likely, a molten or at least a plastic
phase of some type had been present. A series of samples at 800, 900, 1000,
and 1100°C was then run. The only one of this series that has been completed
is the one at 800°C, which is shown in Fig. 8.3a.  Even at this temperature
considerable reaction had taken place. There is a continuous light diffusion
zone of about 0.002 in. around the stainless steel, followed by another zone
of indeterminate depth (estimated at 0.0005 in.) which dissolves during
etching.  The gap between the two phases also contains many broken particles

of a reaction product.-

(5) Stainless Steel—Columbium (Fig. 8.3b). The swaging resulted in
considerably more deformation of the columbium rods than of the stainless
steel matrix tube. A continuous reaction layer of about 0.001 in. is found in
the interface. The surface of the stainless steel appears to have been etched

quite evenly.:

(6) Molybdenum-UO,. Although it was enclosed in a stainless steel tube,

the inner surface of the molybdenum was cracked considerably. There did not

215
 

 

REGAET
d DRAWING NO.Y-2652

 

 

 

 

 

| @
\
|
|
FIGURE a. STAINLESS STEEL & UO, HELD
! 100 HOURS AT 1100°C (mag 250 X)
C NG No
DRAWING NO Y-2519
|
\ FIGURE b. STAINLESS 8 MOLYBDENUM HELD
o 100 HOURS AT 1100°C (mag250X)

| FIGURE 8.2

216

 

 
O

 

 

DRAWING NO. Y-2695

     
  
  

 

- 1 Py o0
e TR - ~
. :
r--'a""' !fl L A ]
L s - l' st L

=
3 k o o 3

I L b ¥
agm P T Ll .f‘:k\

e T T T{-
1 L e -

e N A i

e . o

e ":..‘:‘ - f

 

w AT

    

el . = L + 3 s . P 5 TER . Y

et N u e - - i U Y ek

N . e . i R - N e T

' : l._:._ ‘ -fiu_'.. -t 7
=, ok T 1

FIGURE a. STAINLESS STEEL & BERYLLIUM
HELD |00 HOURS AT 800°C(mag.250X)

- SEeGRET W
DRAWING NQ Y-2522

  
     

FIGURE b. STAINLESS STEEL & COLUMBIUM
HELD 100 HOURS AT 1100°C (mag.250X)

FIGURE 8.3

217
b =

seem to be any reaction products or evidences of leaching present. This test

will be repeated with a molybdenum rod to avoid breaking-up of the surface.

(7) Molybdenum-BeO. The surface of this sample, which was in the same
tube as the one discussed above, was also broken but not so badly. There was

no evidence of any reaction having taken place.

(8) Columbium-UO, (Fig. 8.4)., A reaction is evident on the columbium
surface. The reaction products present are of the order of 0.001 in. but
there is evidence that an additional layer has been removed in polishing.

This is being investigated further.

(9) Columbium-BeO. The surface of this interface also shows a thin
layer of reaction products. The appearance is identical with that in Fig. 8.4
but the thickness is 0.0005 in.

WELDING LABORATORY

P. Patriarca

Welding research to date has been confined to preliminary experiments,
using the welders and existing facilities of the X-10 Research Shops pending
the completion of a welding laboratory in the Metallurgy Division. A satis-
factory technique for the welding of molybdenum has not yet been developed.
Preliminary experiments on welding niobium are more promising. In addition,
technical assistance is being supplied to the Dynamic Corrosion Testing group

in the welded fabrication of convection loops.

Welding of Molybdenum. All welds were made manually and found to be
brittle regardless of the welding process or welding conditions used. A
massive copper jig was constructed which permitted a steep thermal gradient
during welding in order to minimize recrystallization, this being one oi the
factors believed to be responsible for the brittleness of molybdenum welds.
Ample inert-gas shielding was also provided by the jig by supplementing inert
gas flow from the welding torch with a separate inert gas cover of the work
within the copper welding jig. Both direct-current straight polarity and
alternating current were used with the tungsten-arc process under commercial

helium and argon as a shielding gas. Direct-current straight polarity was

218
 

 

612

 

FIGURE 8.4

 

DRAWING NO. Y-2653

COLUMBIUM & UO, HELD 100 HOURS AT 1000°C
(mag 250 X)

 

 
found to provide better control of the weld bead. Atomic-hydrogen welding was
also attempted. All welds were edge welds using Fansteel molybdenum sheet
0.020 in. thick. Becent research 1ndicates that present grades of molybdenum
must be improved by further purification or by minor alloy additions to
render them weldable. Future experiments are contemplated using Climax arc-
cast molybdenum and available grades of molybdenum alloyed withminor additions.
It is hoped that near-ideal welding conditions can be achieved by conducting

future experiments within the confines of an inert-gas-purged dry box.

Welding of Niobium. Preliminary experiments indicate that welding of
niobium will not present so imposing a problem as does joining of molybdenum.
Welds have been made which demonstrate some ductility. Edge welds made using
inert-arc alternating current and inert-arc direct-current straight polarity,
using argon and helium, respectively, as inert-gas shields, were relatively

strong and ductile as indicated by manual cyclic bend tests.:

Fabrication of Thermal Convection Loop. Technical assistance 1s being
supplied to the Liquid Metals Dynamic Corrosion Group in the form of supervi-
sion of welded fabrication of all future loops. Written welding specifications
and schedules will be set forth on each thermal convection loop and will in-
clude specific instructions on edge preparation of component parts, welding
sequence, and welding conditions to be used. It is believed that this quality

control will minimize premature convection loop failures.

CREEP-RUPTURE LABORATORY

R. B. Oliver

Eight Baldwin lever-arm machines and two Baldwin screw type stress-rupture
machines have been received and erected. Six additional lever-arm machines
have been designed and are in the process of construction. Sixteen Leeds and
Northrup Speedomax-DAT recorder-controller units have been received and in-
stalled, and one 56-point precision indicator and one 12-point recorder have
been received and installed.  Delivery of the power distribution equipment 1is
expected by the middle of December. A 20-kw gasoline driven emergency power
plant has been ordered and should be installed by February. - The chambers for

testing in vacuum or in inert atmosphere have been designed, and production of

220
these chambers and their accompanying furnaces will be started in the near

future.

Since most of the metal going into the reactor and i1ts auxiliaries will
be in the form of pipe and tubing, attempts are being made to devise a creep-
rupture test for such shapes. Tests will be performed over the range of
temperatures expected 1n service and in the several environments chosen for
the conventional creep- rupture tests. It is hoped that a method can be devised

to measure the tubing creep, both radially and longitudinally.

The overall ANP Creep-Bupture program will test pipes and tubing in the
temperature range from about 600 to 1000°C in air, in liquid metals, and in
inert atmospheres. These data will not only indicate possible design stresses
but will also give information on stress corrosion in liquid metals and
oxidizing atmospheres. The initial effort will be with the austenitic stain-
less steels with subsequent efforts being indicated by the interim ANP materi-

als studies.-

221
9. RADIATION DAMAGE

222
9. RADIATION DAMAGE

D. S. Billington, Institute for Solid Studies, Metallurgy Division

Radiation damage experiments are underway at a number of installations.
The Y-12 cyclotron is being adapted for such work, and targets are now being
constructed. Both stationary and rotating targets are being developed, and
initial measurements of creep and resistivity under bombardment will be made.
The lithium-iron corrosion experiments in the Perkeley cyclotron will substitute
Globe iron for Armco. The Purdue University cyclotron group is preparing to

run creep experiments.

Initial experiments on various metals have indicated no significant
changes in shape, electrical resistivity, permeability, or hardness after an
exposure to a neutron flux of 1.2 X 10!2 and while at 400 to 500°F and 1000
psi. In-pile creep tests will be performed on a cantilever type apparatus
because of its reliability. Operational techniques with this apparatus are

now being perfected.

Y-12 CYCLOTRON EXPERIMENTS

R. S. Livingston, Electromagnetic Research Laboratory

The construction of targets and the adaptation of the 86-in. cyclotron at
Y-12 for radiation damage work is well underway. Modifications consist, in
the main, of a vacuum lock and handling equipment for the targets. Considera-
tion has also been given tochanging the resonance frequency and magnetic field
so that low-energy protons of 10 to 15 Mev can be obtained at a larger radius.-
Initial plans, however, call for using the cyclotron as it is presently

operating.

The basic design for the targets consists essentially of two concentric
tubes through which a cooling (or heating) fluid is passed (Fig. 9.1).  With

a large flow of liquid through the system, small variation in beam intensity

223
vee

WATER HEADERS

010 WALL —

-

s

W7

     
  

SCRIBE MARKS FOR
MEASUREMENT OF CREEP

   

      

 

THERMOGCOUPLE JUNCTION

FLOW OF COOLANT

 

 

 

 

 

 

 

 

(((( ))‘7—— BEARING

 

 

 

 

 

I

FIGURE 9.1 ROTATING TARGET

 

11201 ON 9NIiMvida
is not expected to affect the target temperature seriously. When running at
high temperatures and using liquid metals to heat the target, the power input
of the beam is small compared to the power used in the fluid. Rotation of the
target is achieved by turning the entire assembly through a vacuum seal. The

rotation ensures uniform exposure of the target material to the beam.

Controlled stresses must be applied tc the target so that creep of the
material under bombardment can be measured. The inner "squirt" tube provides
not only a path for coolant flow, but also transmits a force which keeps the
target area under tension. ©Creep which may occur under these conditions 1is
determined by marking the target in two places around the circumference.
FPictures will be taken of these targets and shown on a calibrated screen.

Thermocouples will probably be used for temperature measurements.:

In cases 1n which water-cooled targets are used, the assembly will look
essentially the same as that in Fig. 9.1.- The water-cooled type of target is
under construction; the liquid-metal target 1s in the design stage. Figure 9.2
illustrates a stationary target to be used for exploration work in the 1initial

stages.

IN-P?ILE CREEP

Pata obtained from the creep of a cantilever beam of type 316 stainless
steel at 1050°F under a stress of 40(C" psi show at least a transient increase
in creep rate upon subjecting a creeping specimen to neutron bombardment in

the ORNL reactor.

The tensile creep apparatus described earlier‘!’ proved erratic, mainly
because of pulley friction difficulties, and, furthermore, 1ts removal from

the pile was hazardous.  The cantilever creep test is not commonly used, but

(1) Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending May 31, 1950,
ORNL-768, p. 96 (Aug. 14, 1950,

225
922

     

SAMPLE
INSERT

WATER PASSAGES

 

 

| SAMPLE

\THIN Al

WINDOW

 

 

 

 

 

 

 

 

 

 

. D ‘
"isib 4 | WATER HEADER & SUPPORT

=
I{‘ \ — -J

sz

o AN
= z
= Z
S Zi
= ~
= 2
| I
{* il
i

FIGURES.2 STATIONARY TARGET

2120l ON 9NIMVviya

g
since some correlation between tensile and cantilever creep is possible,{?/ it
may be used for the determination of tensile creep. From the standpoint of
ease of construction and operation the cantilever has much to recommend it for
in-pile operation. A very smali weight i1s all that 1s required for stressing
to any reasonable stress level so that only electrical leads need be run to
the apparatus in the pile. The strains are magnified mechanically by the long
lever arm carrying the weight; furthermore, the effect of temperature on the
apparatus (other than the effect on intrinsic creep behavior; is much smaller

than for a conventional tensile creep test

Figure 9.3 shows a partially assembled and unwired apparatus of the type
used i1n these experiments. Supported on a long aluminum base plate are the
test specimen with the furnace windings in place, the weight on the end of the
beam, and the microformer for the measurement of displacements of the end of
the beam.  In the foreground are an unmounted specimen, two lavite furnace
supports, and the concentric nickel-~foil radiation shields which surround the
specimen upon assembly. The specimen, i%s support on the base plate, and the
loading beam are integral. The gauge length (approximately ¥ in.) 1is the
reduced portion extending between the loading beam and the support. The fur-

nace windings are nichrome wound around lavite spacers and through ceramic

tubes.  The furnace 1s composed of three separate windings to facilitate re-
moval of temperature gradients across the gauge length  Thermocouples are
welded directly to sides of the tezt bar at three points. Figure 9.4 shows

the apparatus itself and an outer furnace tube {(controlled temperature) to
keep the microformer at constant temperature. Control of this temperature
also aids in the control of the test bar temperature by its separate controller.
This outer furnace is wrapped with aluminum foil and inserted in the rectan-

gular can for insertion in one of the stringer holes in the ORNL reactor.

The apparatus was charged during an extended pile shutdown. Difficulties
in removing the temperature gradient across the bar and burnout of one of the
furnace sections prevented operation at the anticipated temperature (1500°F)
and caused large temperature gradients across one end of the bar. A deflec-
tion (strain) time curve plotted from the data showed an increase in strain
rate immediately (within 30 sec} upon starting up the pile (to 4400 kw, fast

im
L0

flux approximately 4 ¥ 10"},  The deflection rate for the half hour after the

(2) Harris, Go Te, and Chiid. He Co; *“C-eep Tesving by the Cantlieve -bending Method |, " J. Jron Steel
Inst. 165, 139 (1950, -

227
   

DRAWING NO. Y-2732

CANTILEVER BEAM TYPE IN-PILE CREEP APPARATUS

   
  

FURNACE WINDINGS

BASE PIATE

  

8227

RADIATION SHIEID

IAVITE FURNACE PARTS

GAUGE

   
  

LENGTH ( 3// inch )

UNMOUNTED SPECIMEN

FIGURE 9.3
6¢¢

     

CANTILEVER BEAM TYPE IN-PILE CREEP APPARATUS
(WITH FURNACE)

FIGURE 9.4

N
DRAWING . Y-2731

INNER FURNACE (1500 F )
pile was started was approximately five times the rate in the half hour
preceding (pile off). The temperature gradient across the bar and the present
lack of correlation between tensile and cantilever creep preclude any quanti-

tative conclusions.

Although the deflection rate fell off with time, subsequent transient
increases 1in deflection rate were again observed upon applying the flux after

each pile shutdown during the following week.

Inasmuch as the stress was applied to the specimen all during the heating
cycle from pile ambient temperature to operating temperature, and because the
amount of time at constant temperature before the pile was operating was
short, the state of the metal (whether in so-called "first" or "second" stage

of creep) cannot be described with any accuracy.

There is a possibility that the changes in the deflection rate when the
pile is started are being caused by gamma or neutron heating of the specimen.
However, the behavior of the controllers and the chart record of the tempera-
ture indicate no appreciable heating here. The controller on the outer furnace
(although unrecorded) and the controller on the inner furnace showed no ab-
normal variations at the time of pile start-up. The chart on the controller
for the 1nside furnace was unaffected by pile start-up. Separate temperature
measurements at three points on the bar showed variations of less than 2°F
during the period covered by the measurements described above. The Duration
Adjusting type control unit used is very sensitive to temperature variations,
and rates of variation, i1f any existed, would in all probability show up on
the chart as changes in the on-off cycle, provided their magnitude was greater

than 2°F.

The data obtained are only qualitative to date, and such variations 1in
rate might not be nearly so marked in tensile creep tests. Some theoretical
calculations by F. R. N. Nabarro¢®’ predict that the greatest effect of
neutron bombardment on creep rate will occur when there i1s a stress gradient
in the specimen.- Of course, the stress gradients in the cantilever test are
greater than would be expected in pure tension of a polycrystalline metal.
However, most engineering structures have many points of complex stress distri-

bution where gradients exist, so the observations reported above are not

{3) Nabarro, F. R, N., Report on a Conference on Strength of Solids Held at the H. H. Wills Physical
Laboratory, University of Bristol,on 7-9 July 1947, p. 85, Physical Society, 1948.:

230
without value.

At present a number of bench test rigs are being built to obtain compara-
tive data under conditions free from radiation. The apparatus is being im-
proved to make possible loading of the beam at any time while in the pile and
to permit extended operation at 1500°F and higher. Also being investigated
are various forms of strain measuring devices, since microformers would be
difficult to cool under the conditions of high radiation heating that might be

encountered i1n the MTR.

CORROSION EXPERIMENT OF NORTH AMERICAN AVIATION, INC.*

This group is continuing the lithium—=Armco iron accelerator corrosion
experiments proposed in the last quarterly report.(*) Plans have been made
calling for one or two completed irradiations by January, 1951.- It has been
decided to substitute Globe iron for Armco because of inherent fabrication

difficulties when using thin-walled Armco.

The analyses of Armco and Globe iron are as follows:

ELEMENT GLOBE IRON ARMCO IRON
C 0.03 0.015
Mn 0.24 0.035
P 0.015 0.005
S 0.025 0.025
Si 0.01 0.003

The NAA group is also extending their work to include accelerator cor-
rosion experiments with stainless steel and sodium. It 1is thought that this
work can proceed rapidly as there should not be the fabrication difficulties

encountered with the Armco iron.

CREEP EXPERIMENTS OF PURDUE UNIVERSITY

The Purdue University group is setting up creep experiments to be run on
the Purdue cyclotron. The first materials to be studied will be stainless

steel types 316 and 347 and molybdenum.
'(4) Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending August 31, 1950,
ORNL-858, p.: 86 (Dec.:4, 1950).

* See also Section 12 of this report, ‘‘Vapor-cycle Reactors,’ where other corrosion experimentation
of North American Aviation is discussed. :

231
Much of the preliminary work involving the metallography of molybdenum
and the construction of furnaces has been completed.” In addition an a-c Hall-

coefficient apparatus has been designed.

PROPERTIES OF METALS

The Radiation Damage Group has completed some measurements of samples of
stainless steel types 304, 309, 316, and 347, nickel "A," and hastelloy C done
in conjunction with the ANL-—Naval Reactor Program. The samples were exposed
at ORNL for a nvt of 1.2 X 10'° (flux 1.2 %X 10'2) at 400 to 500°F and a
pressure of 1000 psi. - Samples were above 400°F for four months. Before and
after measurements of the electrical resistivity, dimensional stability,

magnetic permeability, and hardness showed no significant changes.

OTHER ACTIVITIES

The Radiation Damage Group is actively studying the problem of conducting
experiments in the MTR. The gamma-ray flux is so high that considerable
heating will apparently take place in all uncooled samples.: The problem of
conducting suitable experiments under this condition is not a simple one.-
Plans have been made for activation of the Thermal Conductivity Group by

Jan. 1, 1951.

The samples being exposed at Hanford in the ANL-—-Naval Reactor test loop

will be discharged in January, according to the latest estimate.-

232
10. CHEMISTRY OF LIQUID FUELS

233
10. CHEMISTRY OF LIQUID FUELS

W. R. Grimes, Materials Chemistry Division

A previous report{!) has discussed some of the general characteristics of
liguid-tfuel svstems and has descrihecd the rreliminary phases of the research
program designed to produce satisfactory liquids. During the past quarter,
interest in such liquid fuels has increased materially. Two types of liquid
fuels, suspensions of uranium compounds in sodium hydroxide and solutions of
UF, in alkali and alkaline earth fluorides, have been under intensive study.
It is possible at this time to view with optimism the prospect of such a fuel
in the reasonably near future; it is true, however, that it is not possible at

this time to state the composition of a liquid which will prove suitable.

It has been demonstrated that all uranium materials tested to date react
rapidly with sodium hvdroxide to form a finely divided material which 1s almost
certainly sodium uranate. Late in the period a satisfactory method for evalu-
ation of the suspensions in terms of settling rate has been established.
Studies designed to improve the stability of these suspensions are currently

underway.

The equilibrium diagram of the system LiF-UF, has been established by
thermal analysis techniques, and study of the ternary systems NaF-BeF,-UF,,
NaF-KF-UF,, and NaF-LiF-UF, is well underway. Of these the NaF-BeF,-UF,
system appears the most promising in so far as low melting point and reason-
able uranium content are concerned. It must be emphasized, however, that a
great deal of effort 1s still required before these systems are established

sufficiently to be recommended for use.

Concurrently, the corrosion of metal immersed in NaF-UF, and LiF-UF,
eutectics has been examined. There is considerable correlation in the
corrosive properties of these eutectics, molybdenum, inconel, and hastalloy C
(in that order) being the least corroded in the lithium eutectic, and hastelloy
C, inconel, and molybdenum the least corroded in the sodium. All other of the
many metals under study were far inferior. A test was performed to determine

~which, and to what extent, fission products react with the proposed eutectic.

(1) ‘‘Chemistry of Liquid Fuel Systems,” The Aircraft Nuclear Propulsion Project Quarterly Progress
Report for Period Ending August 31, 1950, ORNL-858, p. 104 (Dec. 4, 1950).

234
SUSPENSIONS OF URANIUM COMPOUNDS IN SODIUM HYDROXIDE

J. D. BRedman D. E. Nicholson
L. G. Overholser

Uranium compounds are not appreciably soluble in sodium hydroxide.(?)
It has, however, been established that uranium compounds react with sodium
hydroxide to produce finely divided materials. The prime objective of this
study is to establish those conditions which will produce a suspension of
maximum stability in the temperature range 600 to 900°C.  The work to date has
been concerned primarily with attempts todetermine the stability of these

suspensions and the identity of the compound or compounds present.:

Preliminary Observation of the Suspensions. Suspensions containing 4 to
5% uranium were prepared from various uranium compounds by adding the uranium
compound to sodium hydroxide contained in a silver crucible at 700°C. The
suspensions were held at -700°C for 30 min, poured into platinum dishes, and
allowed to solidify.  The button was removed and portions of the material were
examined microscopically.: The following compounds were used: ammonium uranyl
pentafluoride, uranyl nitrate, uranyl acetate, uranyl formate, uranyl tartrate,
uranyl pyrophosphate, sodium peruranate, peruranic acid, ammonium diuranate,
potassium diuranate, sodium diuranate, and uranium trioxide (commercial product
and specially prepared samples from ammonium diuranate and peruranic acid).:
Suspensions were also prepared from uranium metal by heating with sodium
hydroxide at 500°C in an atmosphere of nitrogen.: Uranous chloride and uranous

fluoride were similarly treated at 500°C.-

An evaluation of these different suspensions, based on very rough settling
rate observations and approximate particle size determinations by observation
with the microscope, indicated that the suspensions prepared from uranium
trioxide and those from uranium metal were probably superior to any of the
others; the average particle size in these preparations was approximately 2
microns.  Actually, none of the suspensions examined had an average particle
size much greater than 10 microns and the majority of, them appeared to have
about equal particle size.- The specially prepared uranium trioxides yielded
suspensions comparable to those obtained from the commercial uranium trioxide,
indicating that the original particle size is not the most important factor

involved in preparing these suspensions.-

(2) ORNL-858, op. cit., p. 107.

235
By the same technique it was demonstrated that use of carbonate-free
NaOH, mixtures of N32CO3 and NaOH, and KOH did not yield suspensions which
were superior to those described. Mixtures of Na,0, and NaOH yielded suspen-

sions which had definitely larger particles.

A few experiments were performed in which the suspensions were held at
700°C for 24 hr or longer to ascertain whether such treatment promotes growth
of the particles. Evidence was found suggesting that such growth does occur
on prolonged heating, but the microscopic examination of these materials was

complicated by the silver oxide present.

Settling Rate Measurements. Although the microscopic examinations
proved useful in the preliminary work, the evaluation of the suspensions by
measuring the settling rate appeared to be of more practical value than the
approximate particle size determination based upon microscopic examination.
Consequently, studies were made using various techniques for measuring the

settling rate of the suspensions.:

The first method that afforded any degree of success made use of a stain-
less steel nipple (1 in. I.D. and 3 to 4 in. in length) provided with a cap
having a poured silver disk in the bottom to afford a seal against sodium
hydroxide. The nipple, with the cap in place, was loaded with 20gof sodium
hydroxide and heated to 700°C before addition of approximately 1 g of the
uranium compound. The suspension was heated at 700°C with frequent stirring
for 30 min, removed from the furnace, and air cooled (about 10 min was re-
quired for the sample to freeze). The cap was removed from the nipple, and the
sodium hydroxide plug was driven out. Samples were cut off the plug by use of
a hot copper blade. Analyses of sections of the plug for uranium yielded dats

of which those shown in Table 10.1 are representative.

These results confirm the earlier observation that the uranium trioxide
suspensions are the most stable of any studied and also show rather conclu-

sively that growth of the particles occurs upon aging for 24 hr at 700°C.

Although this method was capable of measuring relative settling rates, it
was not so flexible as desired and might be subject to serious error arising
from convection currents. Further studies by this method were, therefore,

abandoned upon delivery of a silver reactor. The apparatus being used at

236
TABLE 10.1

Settling Rates of Uranium Suspension Measured in

Stainless Steel Nipples

 

 

 

COMPOUND ADDED TO NaOH PERCENT OF QCOMPOUND FOUND
TOP BOTTOM
U0, 4.1 4.1
U0, 4.2 4.5
U0, 1.3 6.2
Uo, 2.8* 6.0
U0 1.7 6.8
U304 3.9 7.1
U,04 2.9 6.7
U30g 3.2 6. 1
Na,U,0, 3.0 5.3
Na,U,0, 2.0 6.5
(U0,),P,0, 2.0 6.7
(10,),P,0, 3.4 4.5
(U0, ), P,0, 3.5 4.0
(U0,) ,P,0, 3.3 4.1
(U0,),P,0, 1.6 4.1

 

 

 

 

 

* Digested for 24 hr at 700°C (no agitation).

present for determining relative settling rates is shown in Fig. 10.1.  1In
using this apparatus the temperature is raised to 700°C, 180 g of sodium
hydroxide is added, and, after all water is removed, the uranium compound .is
carefully added and the suspension is agitated by stirring for the desired
length of time.: After this time has elapsed, the stirrer is cut off and
samples are removed after various intervals of time by inserting a silver rod
through a hole in the cover. The silver rods are collared to sample the upper

one-fourth of the molten sodium hydroxide. After the silver rods are cooled

237
—SILVER STIRRER

THIN WALL SILVER VESSEL
/WITH COVER
L

 

 

 

 

 

 

\\\\ /

 

 

//////

 

 

 

 

 

 

llllllllll
and weighed, the uranium is dissolved and the solution is analyzed for uranium.
By this method the system may be digested with agitation for any desired

period and may be sampled with a minimum of error due to convection currents
in the liquid (Table 10.2).

TABLE 10.2

Settling Rates of Uranium Suspensions

Measured in the Silver Reactor

 

 

 

. PERCENT OF COMPOUND FOUND
SEr{I‘;?:G TIME DIGESTIOIEIthE, 700°C |
U0, Na,U,0, (U0,) ,P,0,
1 % 2.4 0.2 0.3
2 % 1.7 0.1
3 % 1.6 0.09 0.2
4 A 0.5 0.04 0.2
5 Y 0.4 0.09 0.2
10 A 0.2 0.04
1 % 3.1 0.08 0.4
2 4 2.2 0.06 0.3
3 1% 1.3 0.04 0.2
4 4 0.4 0.04 0.4
5 VA 0.3 0.07 0.4
10 /7 0.2 0.03
0 4 3.6 2.6
1 4 2.3 0.2
2 4 1.2
3 4 1.0
4 4 0.5
5 4 0.4

 

 

 

 

 

 

 

The results agree with those obtained by use of the steel nipples inas-
much as they show that the uranium trioxide suspensions settle less rapidly

than those prepared from the diuranate or pyrophosphate. However, it will be

239
noted that the settling rate for the uranium trioxide suspensions measured in
the silver reactor is greater than that observed by the nipple method.  For
the best suspension studied in the silver reactor, the upper one-fourth of the
molten sodium hydroxide is virtually free of uranium after 3 to 4 min as con-
trasted to the relatively uniform distribution found in the nipples after
10 min. This suggests that convection currents played an important role 1in

prevention of the settling in the earlier studies.

It is significant to note, however, that only very slight agitation 1%

required to keep the particles well suspended.

Identity of the Uranium Compound. For simplicity, repeated reference has
been made in the preceding discussion to uranium trioxide suspensions although
the uranium is certainly not present as the trioxide. A pregram for identifi-

cation of the uranium compound 1is still incomplete.

The X-ray diffraction patterns from caustic buttons prepared at 700°C
with uranium trioxide or sodium diuranate are identical, but they differ con-
siderably from the patterns for uranium trioxide or sodium diuranate. Caustic
buttons (prepared with uranium trioxide), when extracted with anhydrous ethyl
alcohol, yield a residue of sodium diuranate; the alcohol extraction alters
the X-ray diffraction pattern, indicating that the uranium compound originally
present in the caustic cannot be isolated by this technique. Samples of sodium
diuranate have also been prepared by precipitation of uranyl nitrate from
methyl or ethyl alcohol by sodium hydroxide. Precipitation from ether failed

to give the diuranate.

It is believed that sodium monouranate is present in the caustic buttons,
but the X-ray diffraction pattern for the monouranate is not available for
comparison, Vacuum distillation of the excess sodium hydroxide from the
suspensions is being tried, using the apparatus shown in Fig. 10.2.- A suspen-
sion prepared from uranium trioxide was heated at 800 to 850°C [vapor pressure
of NaOH(l), 4.0 mm Hg at 800°C] for approximately 50 hr. The sodium hydroxide
distilled off leaving an orange residue in the silver crucible, but this resi-
due was contaminated by particles from the stainless steel pipe fittings and
by sodium hydroxide which had fallen back into the crucible. - Another run is
planned using a cover over the silver crucible to prevent this contamination.
Results of X-ray analysis on the first sample are not available, and identifi-

cation of this orange residue has not as yet been accomplished.

240
v

DISTILLED MATERIAL

 

 

N Nt
TO VACUUM __
SYSTEM

e e—————

 

[

 

FURNAGE AT 800°C a"

 

\\\\\\§

 

Nt

 

[0

 

SILVER CRUCIBLE WITH CHARGE

VACUUM DISTILLATION OF SODIUM HYDROXIDE
FIGURE 10.2

¥1201 'ON ONIMVYAQ
LO¥-MELTING FLUORIDE SYSTEMS

R. E. Moore J. P. Elakely
G. J. Nessle C. J. Barton

Studies of molten fluoride systems which might serve as fuels have been
continued in the directions indicated in the last quarterly report.(3) During
the past quarter the equilibrium diagram for the UF,-LiF system has been
established, and considerable progress has been made on the ternary systems
UF, -NaF-BeF,, UF,-NaF-KF, and UF,-NaF-LiF. The equilibrium diagrams for these
systems are, however, still far from complete, and it is not as yet possible
to state the precise composition of a fuel which would be completely satis-
factory from a phase stability point of view.  Although the UF,-NaF-BeF,
system has shown most promise in preliminary studies, 1t 1is not yét certain
that this system is best for the purpose. The general characteristics of the

systems studied are discussed briefly under individual headings below."

Experimental Methods. The conventional techniques of thermal analysis
have been used in virtually all the experiments to date. The procedures and
materials are essentially as described earlier¢?’ although the actual apparatus

used has undergone considerable modification sincé that report was issued.-

The melts are still contained in high-density graphite crucibles, but the
. graphite stirrer—thermocouple well combinations have been replaced by motor-
driven stirrers of monel. These stirrers are drilled toserve as a thermocouple
sleeve from which the couple is insulated by a ceramic tube. Lubrication of
the ceramic tube with graphite permits the stirrer to turn freely without
twisting of the thermocouple wire. The stirrer is rotated at about 150 rpm by
means of a belt drive which slips without injury to the equipment when the
melt solidifies. A graphite sleeve bearing supports the stirrer shaft to

facilitate smooth operation of the stirrer.

Cooling curves have usually been recorded by automatic pen-type recorders
of the Brown or Leeds and Northrup variety. Manual recording of the curves

"~ using type K potentiometers is used to verify points of special importance.-

(3) ORNL-858, op. cit., p. 104,
(4) Ibid., ps. 110.°
242
Efforts have been made to separate by filtration the phases present under
equilibrium conditions. While this has been accomplished at temperatures
below 420°C in the NaF-BeF,-UF, system, apparatus difficulties have so far
precluded general application of the method.

At low temperatures the melts are essentially without effect on glass
equipment, provided the samples are well dried and outgassed. Systems which
contain sufficient amounts of liquid near or below 400°C can, therefore, be
separated easily by use of the apparatus shown in Fig. 10.3. The materials,
which have previously been melted under an inert atmosphere, are placed 1in the
apparatus on the sintered glass disk, and the entire assembly i1s carefully
evacuated. The sample i1s then heated to the desired temperature while the
pumping 1s continued. After 1 to 2 hr at the desired temperature helium is
admitted to the apparatus to accomplish the filtration. The liquid phase that

accumulates in the lower tube 1is removed for analysis.

Apparatus of similar design but with sintered stainless steel filters
and of all-metal construction should permit general application of this direct

procedure.-

Properties of the Fluoride Systems. Lithium Fluoride—Uranium Fluoride.
This system has been investigated in considerable detail by thermal analysis,
and some information has been obtained from X-ray diffraction photographs.:
While some further study would be necessary to establish with certainty the
identity of the compound with incongruent melting point, the main features of
the diagram (Fig. 10.4) are well established. Only one eutectic exists in
- this system; this material, containing 26 mole % UF, and melting at 490°C,
would be quite interesting as a possible fuel mixture 1f separated lithium
1sotopes were available.: The compound of incongruent melting point is prob-
ably LiF-3UF, although LiF-2UF, is a possibility. The compound does undergo a
phase transition at about 600°C.

- NaF-BeF ,-UF, System. The binary NaF-BeF, system has been reported as
showing a eutectic at 45 mole % BeF2 which melts at about 360°C. Such a
mixture prepared in this laboratory has been shown to melt at 342°C. The
purity of the only beryllium fluoride available here is unknown; it is, there-
fore, quite possible that impurities are responsible for this discrepancy.

The extremely low melting point available has encouraged studies of the

243
DRAWING NO.10218

TO VACUUM SYSTEM

MM

s

g

 

TO HELIUM
SUPPLY =) =

 

 

 

 

THERMOGOUPLE § fl«-—GLASS TAPERED JOINT

 

 

 

 

— FURNACE

POWERED
— CHARGE

. SINTERED
GLASS DISC

 

 

 

 

 

 

 

 

FILTRATION OF LOW-MELTING SYSTEMS
FIGURE 10.3

244
Sbe

TEMPERATURE °C

1100

 

1000

900

800

700~

600

5001

PHASE

DIAGRAM OF UF-LiF BINARY SYSTEM

 

 

 

 

400

 

 

 

 

MOLE PERCENT UF,

FIGURE 0.4

80

90

100

 
solubility of UF, in this material. It must, however, be recognized that the
vapor pressure of BeF, at high temperatures is not known and may be relatively
high (perhaps as high as a few atmospheres at 1000°C). The well-recognized
toxicity of beryllium compounds, especially in the form of air-borne dusts, 1is

a factor which necessarily makes the study a time-consuming one.

Study of this ternary system has been attempted by thermal analysis and
by the filtration techniques described above. It has been demonstrated that
uranium fluoride dissolves in mixtures of NaF and BeF, even at temperatures
below 340°C.- Unfortunately, however, the uranium content of this liquid phase
is less than 5% by weight.  Since the volume available for an unmoderated fuel
in reactors of current design is limited to about 2 cu ft, such dilute solu-
tions of uranium are of little interest. It is likely that, using reasonable
estimates for the density of such liquids, liquids of about 50% UF, by weight

are required.’

Thermal analyses of such mixtures, in the range 10 to 20 mole % UF,, are
far from complete.. Such studies as have been completed, however, indicate
that mixtures containing 15 to 18 mole % UF, and 35 to 45 mole % BeF, are
completely liquid below 550°C. It is extremely likely that the optimum
composition has not yet been reached and that the goal of critical amounts of

uranium in 2 cu ft with melting point below 500°C may be realized.

NaF-LiF-UF, System. The relatively low temperature (490°C) obtained in
the LiF-UF, binary system and the known binary eutectics of NaF-KF (640°C) and
of NaF-UF, (600°C) have encouraged the study of this ternary system. A large
number of compositions within this system have been checked by thermal analy-
sis methods. Not all regions of this diagram have been explored as yet. It
would appear that the lowest melting point in the system 1s at a ternary
eutectic at 26 mole % UF,, 48 mole % LiF, and 26 mole % NaF. Material of this
composition shows only the eutectic halt at 450°C., It appears that in regions
considerably removed from this composition,i.e., 25 mole % UF,, 45 mole % NaF,
and 40 mole % LiF, a new ternary eutectic of higher melting point is observed.
A large number of compositions must be investigated before this could be

definitely stated.

The uranium concentration of the 450°C ternary eutectic is sufficiently
high and the melting point sufficiently low to make it of considerable inter-

est, provided the separated lithium isotopes are available.

246
NaF-KF-UF, System. Studies of this ternary system have been initiated
quite recently, and, to date, the few cooling curves available are insufficient
to characterize the system.- It seems likely, however, that the lowest melting
points available are higher than those indicated for the two ternary systems
discussed previously. 1In spite of this fact, there is evidence of a ternary
eutectic at 505°C, and samples containing up to 25 mole % UF, have remained
liquid to 550°C, ‘Determination of the eutectic composition is the object of

present efforts.:

CORROSION TEST OF METALS

P. J. Hagelston, Isotope Research and Production

The first corrosion test was made to determine the effects of NaF-UF,
eutectic on various metals and to select the least susceptible for the next
test. The samples were cut from sheet materials approximately 1% by 2 in.,
weighed, and placed in a reactor containing the molten eutectic at 700°C for

approximately 160 hr. The following samples were tested:

Durimet Inconel
International Ni Special Stainless steel 316
Illium G Nickel

Hastelloy A Tungsten

Hastelloy B Molybdenum
Hastelloy C Monel

Hastelloy D Stainless steel 304
Illium R Platinum

Tantalum Stainless steel 347
Inconel X Zirconium

It was found that tantalum and zirconium disintegrated. Some of the
samples lost weight from erosion while others gained weight from penetration

by the salts.

The test with LiF (74 mole %) — UF, (26 mole %) eutectic was made using

samples from the following materials:

Hastelloy A Illium R

Hastelloy B Stainless steel 316
‘Hastelloy C Stainless steel 321
Nickel Stainless steel 347
Inconel Ticonium
Molybdenum

247
The samples were placed in a graphite box filled with molten eutectic
which was maintained at 700°C.- The graphite box was surrounded by a tightly
lidded mild-steel "shoe box" through which argon was circulated. The samples
were removed and weighed after 24, 68, 134, and 204 hr run time. It was
necessary to replenish the eutectic each time the run was interrupted, owing
to the creeping of the mixture over the sides of the container. During the
final run (134- to 204-hr period), the mild-steel shoe box oxidized and the

samples were exposed to the atmosphere. The test was. therefore terminated.

It was found that some materials showed an initial weight gain which
gradually diminished with time. Scale formations were not removed. In the
order of initial weight gain, starting with the least weight gain, these

materials were:

Molybdenum

Inconel

Il1lium R and hastelloy C
Ticonium

Hastelloy A

Nickel

Hastelloy B

Stainless steels 316, 321, and 347 all showed a loss in weight which in-
creased with time. Listed in the order of weight loss, least weight loss

first, they are as follows:

Stainless steel 316
Stainless steel 321
Stainless steel 347

The attack on 347 stainless steel represented the extreme case among those

tested. Figures 10.5, 10.6, and 10.7 picture this attack.

In order to compare the corrosion effects of LiF-UF, and NaF-UF, on
identical samples, two tests were started on November 30 using the following

materials in interior containers made of hastelloy C instead of graphite:

Stainless steel 304 (eextra low carbon) Stainless steel 410
Stainless steel 316 (extra low carbon) Stainless steel 430
Stainless steel 304 Stainless steel 446
Stainless steel 316 Allegheny 590
Stainless steel 321 Molybdenum
Stainless steel 347 Inconel

Stainless steel 321 Hastelloy C

Stainless steel 405

248
6ve

 

347 STAINLESS STEEL PRIOR TO TEST

FIGURE 10.5

347 STAINLESS STEEL (UNEXPOSED)
(mag. 100X)

 

2¥201'ON 9NIMY YA

 

 
062

 

347 STAILESS STEEL, SHOWING ATTACK (DARK OUTER EDGES) AFTER 24
HOURS EXPOSURE TO LiF-UFg EUTECTIC

FIGURE 10.6 347 STAINLESS STEEL (24 HR. EXPOSURE)
(mag.100X)

€201 ON ONIMVYA
162

oy

 

347 STAINLESS

O

STEEL , SHOWING ATTACK AFTER 135 HOURS EXPOSURE TO
L1 F-UF, EUTECTIC

LIGHT SECTION IS UNATTACKED METAL

FIGURE 10.7 347 STAINLESS STEEL (135 HR.EXPOSURE)

(mag. 100 X)

.
#
i

¥¥201 ON ONIMVYHQ
Both of these runs were terminated because of oxidization of the mild-
steel shoe boxes when the argon lines were plugged by condensation of the
vaporized eutectics.' The test with LiF-UF, eutectic failed after 113 hr, and
the one using NaF-UF, eutectic failed &fter 71.25 hr. The samples were re-
moved and cleaned by first dipping into a molten mixture of NaCl and Na,CO,
and then into cold water. After washing in water, the samples were free from

scales and a fairly accurate weight loss was obtained. Results are summarized

in Table 10.3.-

TABLE 10.3

Comparison of Material and Eutectic

 

 

 

WEIGHT LOSS (mg/dm?/day)
MATERIAL

LiF'-UF4 NaFlUF;
Mo 2.1 64. 4
Inc 68. 5 —44.7*
HaC 198.5 19. 67
'A1-590 607.0 514.0
304 SS (ELC) | 1698.0 876.0
316 SS(ELC) | 1732.0 1138.0
347 SS 1850.0 856.0
321 SS 2475.0 1482.0
304 SS 3270. 0 2050.0
316 SS 3490.0 2070. 0
410 SS 5000.0 3360. 0
430 SS 6510.0 3670. 0
405 SS 7175.0 4830.0
446 SS 7280.0 5470.0

 

 

 

 

 

* Weight gained.

CORROSION TEST OF CERAMICS AND FISSION PRODUCTS

R. 0. Hutchinson

A test was performed to determine which, and to what extent, of the better

than 1% yield of fission products reacted with the liquid fuel.: When the

252
working fluid is decided upon for the ARE, experiments may be conducted to
determine what effect the fission products present in the system will have on
the various materials of construction under consideration.  Some ceramics were

tested to determine which would be least attacked by the liquid fuel at 700°C.

The better than 1% yield of fission products desired as samples were
rubidium, strontium, yttrium, zirconium, columbium, molybdenum, technetium,
ruthenium, rhodium, palladium, cesium, barium, lanthanum, cerium, praseodymium,
neodymium and samarium. Of these metals, only six were found in a slug form.-
The others were either not available at all or were very expensive. Zirconium,
neodymium, palladium, hafnium, barium, and strontium were used. The last two,
because of their known activity, were first tested by dropping into a boat
containing molten UF,-LiF eutectic. A spontaneous reaction occurred with both

metals.

The remaining four metals were immersed in a boat of molten eutectic and
heated up to 1300°F. They were then suspended on molybdenum wire and lowered
into the salts.” Beryllium metal, BeO, Zircon, lava, and graphite were treated
in the same manner.  The beryllium metal began immediately to change the salts
to a red-looking slag. The slag formed a coating over the zirconium and
hafnium, but the palladium and neodymium were gone. The zirconium and hafnium

gained about 1 g owing to the slag formation.

The second run was the same as the first except for the ommission of
beryllium metal.- The four metals were removed at the end of 24 hr.- The
neodymium and palladium were completely dissolved, and the hafnium metal was
still in original shape but completely carbonized when the sample was crushed.

The zirconium was also completely dissolved.

Information regarding the weights of samples after 24 hr of immersion is
given in Table 10.4.

TABLE 10.4

corrosion Tests of Ceramics and Fission Products

 

 

 

 

 

 

 

 

WEIGHT (g)
SAMPLE OBSERVATION
IN [ our
FIRST RUN
Hf 25. 4592 26.7331 Slag formed over
Nd 7.8377 90% or better dissolved
Pd 2. 4264 90% or better dissolved
Zr 13.8474 12.9677 Slag formed, easily removed
SECOND RUN
Hf 21. 1092 Carbonized 100%
Nd 7.7303 90% dissolved
Pd 1. 1102 90% dissolved
Zr 7.7169 Completely dissolved

 

 

 

 

9K
The molybdenum wire used in the suspension of samples was noted to be
much more ductile after subjection to extreme heat and immersion in the bath.

" After repeated bending, it still remained soft.-

The following ceramics were also tested in LiF-UF, eutectic at 700°C:

Graphite Zr0,
CaF, Al,0,4
Lava A MgO
Zircon

After 22 hr CaF,, Zr0,, Al,0,, and MgO showed signs of complete attack and

were eliminated from further test.

A new test was started using graphite, lava, Zircon, and BeO. At the end
of 90 hr the zircon was completely dissolved and the lava badly eroded. The
graphite and BeO showed no signs of attack.:

After 190 hr the test was terminated; it was found that graphite had lost
1.15% of its original weight, BeO had gained 5.52% owing to formation of slag,

and the lava was about 75% dissolved.:

254
11. ANALYTICAL CHEMISTRY

255
11. ANALYTICAL CHEMISTRY*

C. D. Susano, Analytical Chemistry Division

The Analytical Chemistry Division (Y-12 Section) has assisted in the ANP
program by performing service analytical work and by carrying out development,
as required, of methods for use, principally, in analyzing materials of
construction, liquid-metal coolants, and materials for use in fuel elements.
Summaries of the development and analytical activities of this Division which

are concerned with the ANP project are here presented.

DETERMINATION OF OXYGEN IN SODIUM

J. C. White and W. J. BRoss

Although the method of Pepkowitz and Judd(!) has been adopted for the
routine determination of oxygen in sodium, it has not proved entirely satis-
factory with regard to precision and it does not permit a spectrographic
analysis to be made on the same sample. A search is being made for a reagent
other than mercury which will take sodium into solution but will not react
with or dissolve sodium monoxide. Of 25 organic liquids tested for this
purpose, all except two (ethyl acetoacetic ether and acetonyl acetone) have
been eliminated on the grounds of unreactivity with sodium, lack of selectivity
of action, high viscosity, or production of insoluble products with sodium.
Ethyl acetoacetic ether and acetonyl acetone, together with other as yet un-

tried reagents, will be further investigated.

An additional assembly of the Pepkowitz and Judd type for the determina-
tion of oxygen in sodium has been set up and tested. The equipment consists

of a purification train and two reactors.

Preparation of Standard Samples (J. C. White and C,LL:Boyd).:Difficulty
was encountered in preparing suitable samples to check the accuracy attained
in the determination of sodium monoxide in sodium by the method of Pepkowitz
and Judd. Samples were prepared by drawing molten sodium into an evacuated
glass vial containing a weighed amount of mercuric oxide. The reaction be-
tween sodium and mercuric oxide to produce sodium monoxide was so vigorous

that it shattered the glass vial.

* This section, except for minor editorial changes for conformity with other sections, is the same as
Report Y-B31-221, -

(1) Pepkowitz, L, P.; and Judd, W. C., * Determination of Sodium Monoxide in Sodium,’’ Anal. Chem. 22,
1283 (1950). -

256
Samples were next prepared by adding sodium and mercuric oxide 1n separate
glass vials to the reactor of the apparatus  After the addition of mercury,
the vials were broken and the reaction was allowed to proceed inside the re-
actor. Test results for samples formed in this manner indicate that the
reduction of mercuric oxide 1is complete and that recovery of the added oxygen

is possible. Representative results were as follows:

OXYGEN ADDED (%, OXYGEN FOUND <%,

0.13 0. 14
0.057 0.065
0.17 0 14

Work is continuing on the improvement of the accuracy and precision

attainable by the Pepkowitz and Judd method.

Time Study of the Pepkowitz and Judd Method (J. C. White) A detailed
study was made of the Pepkowitz and Judd method with regard to the actual time
spent in the various operations when the method was performed on a routine
basis. It was possible, on the basis of the study, to make several changes
designed to speed up the operation. It was conciuded that one man could be
expected to do eight determinactions per day and two men sixteen per day with
apparatus consisting of two purification trains and four reactors. A report(?)
was written in which the resuits of the time study, along with a detailed

procedure, were given.

MICROANALYSIS OF SILICON CARBIDE

J. C. White and W. J. BRoss

Some fourteen sampies of irradiated silicon carbide have been submitted
for the determination of free silicon and silicon dioxide. Since the average
weight of these samples is of the order of 50 mg, a microanalytical method

must be used.:

A method described in the literature, based on the removal of Si10, by
treatment with HF and H,SO, and the subsequent removal of silicon by treatment

with HF, H,S0,, and HNO,, was shown to be unsuitable. Silicon is not 1inert to

(2) White, J. C.; Suggested Procedure for the Determination of Oxygen in Sodium by the Pepkowitz and
Judd Method, Memorandum to D, ‘R. ‘Rowen, Analytical Chemistry Divizion, Oak Ridge National Labora-
tory, Y-12 Site (Nove 17, 1950}, -

257
HF and H,SO,, as was supposed by the authors of the method, but undergoes

continuous loss 1n weight

A method has been developed based on the removal of silicon as SiCl, by
reaction with dry chlorine at 350°C and the subsequent removal of Si0O, by
treatment with HF and H,S0,. Repeated tests have established that silicon can
be gquantitatively removed from known synthetic mixtures of the metal and the

carbide by this procedure.

Test results representative of those obtained in the analysis of ap-

proximately 5-mg portions of the NEPA samples are listed in Table 11.1.°

TABLE 11.1

Analyses of Silicon Carbide Samples for

Silicon and Silicon pioxide

 

 

SAMPLE SILICON (%) | SILICON DIOXIDE (%)
Bureau of 1. 74 1.63
Standards 1.72 1.90

1.71 1.35
1 0.95 3.23
1.09 3.12
2 0.1 2. 38
0.1 2.56
3 0.59 2.61
0.57 2. 45
4 0.12 1.63
0.17 1.93
6 0.41 0.69
0.22 0.66

 

 

 

 

 

258
FLAME PHOTOMETRIC ANALYSIS OF LITHIUM

O. Menis

An i1nvestigation was made of the flame photometric method for the de-
termination of alkali and alkaline earth metals in a lithium matrix. The
principal purpose of the study was to evaluate the previous work done in the
Analytical Chemistry Laboratory with a view toward bringing this project to a

conclusion.-

In brief, the general conclusions and suggestions for further work are as
follows: for dilute solutions containing the ions in the range of 1 to 10 ppm,
the method as now developed should serve without major modification for the
determination of sodium and potassium, but for the determination of calcium,
and especially for magnesium, higher temperatures of flame excitation are
imperative. The minor constituents should be concentrated by chemical means.
A pure lithium standard should be prepared by one of the several methods to be
found in the literature. Attention should be given to the following: anion
effects, the use of more volatile solvents, choice of wavelength to minimize

interference, and proper laboratory conditions to assure instrument stability.

STABILITY OF A SILICONE OIL IN CONTACT WITH SODIUM

J. C. White and W. J. Ross

At the request of the ANP Experimental Engineering Group, an investiga-
tion was made of the stability of Dow Corning silicone fluid DC-550 in the
presence of sodium at elevated temperatures. This material 1i1s a clear,slightly
yellowish liquid of exceptional heat stability, high resistance to oxidation ,
low volatility, and high flash point (600°F, minimum). Because of its favor-
able properties, 1t 1s being considered as a pressure transmitting agent
in equipment in which the fluid must be in contact with sodium metal at tem-
peratures of 300 to 350°F. 1In the event of certain types of failure of the
apparatus, however, the oil could come in contact with sodium at considerably

higher temperatures.

In order to simulate conditions approaching those which might be en-

countered in the case of an apparatus breakdown, tests were carried out 1in

259
which samples of 01l and sodium were heated together at temperatures up to
1050°F,. 1In these experiments the sample was heated under an atmosphere of
helium in a test tube which was placed in a lead bath. Temperatures were
measured by means of a thermocouple which was inserted in the o1l. Two
tests were made in which the o1l was heated with sodium for 30 min at 1050°F,
No violent reaction took place but the oil darkened 1n color. Upon cooling,
the material solidified into a hard, oily, crystal-like mass. A qualitative

test showed the presence of sodium

In a test inwhich the o1l was heated without sodium at 1000°F for 15 min,
the only effect was an i1ncrease i1n viscosity and refractive index, apparently

due to the loss of some of the lower boiling constituents of the o1il.

When the 01l was heated for 6 hr with sodium at 400 to 450°F there was no
visible change, but the basic reaction shown by a water extract of the oil was

evidence that some reaction between oil and sodium may have taken place.

The results of these experiments indicate that silicone o1l DC-550 1is
stable in the presence of sodium at 400 to 450°F for several hours and prob-

ably stable at 600 to 700°F for short periods.

SERVICE ANALYSES

J. W. Bobinson and L. J. Brady

During this period the analytical work carried out in furtherance of the
ANP program consisted chiefly of the determination of minor impurities 1in
liquid metals (coolants) and of the analysis of materials of construction and

of uranium compounds and mixtures.

Sodium Analysis (J. M. Peele and L. H. Jenkins). Corrosion-test samples
of sodium after exposure to various materials of construction at elevated
temperatures were analyzed spectrographically for minor components using the

porous cup method, and for oxygen content by the method of Pepkowitz and Judd.

Solutions of test samples and of standard samples were prepared in the
chemical laboratories for spectrographic analysis; these solutions were then
examined spectrographically in the Physics and Spectrographic Department of

the Isotope Research and Production Division.

260
Analysis of Boron Carbide <(J R. Lund}. Acceptable methods have been
established for the determination of boron oxide, total boron, carbon, and
moisture 1in boron carbide  The boron oxide is determined by titration of a
hot-water leach of the sample. In determining total boron complete dissolution
of the sample is required. It was found that the product of a sodium carbonate

fusion would dissolve completely only 1f the material was ground to pass a

No. 120 sieve prior to fusion. The resuiting solution was titrated for total
boron in the usual manner. Carbon was determined by combustion at 1400°C,
using copper as an accelerator. Moisture was calculated from loss in weight
at 110°C,

Analysis of Ferrous and Nonferrous Alloys (C. K. Talbott). - In general,
standard methods were used in the analysis of stainless steels and nickel
alloys. A number of special alloys such as Co-Ni, Zr-Ni, and Mn-Cu, required

less common methods of analysis.

Analysis of Uranium Compounds (E C. Lynn and A F, Roemer, Jr.} Sodium
and potassium uranate, oxides of uranium, and fusion products containing small
amounts of uranium were analyzed for uranium, alkali metal, hydroxide, and
carbonates. Potentiometric, colorimetric, or fluorimetric methods—depending
on the concentration-—-were used for the determination of uranium. The
alkali metals were determined by means of the flame photometer, carbonates by
gas evolution, and hydroxides by titration with acid after precipitation of

the carbonate as barium carbonate.

SUMMARY OF SERVICE ANALYSES

A brief summary of the analytical work for the ANP program during the

period is tabulated below.

NO. OF SAMPLES

NEPA Y-12  TOTAL

Backlog of samples 9-16 50 50 1 51
Samples received 9-16-50 to 11-30-50 inclusive 157 71 228
Total 207 72 279
Samples reported 9- 16-50 to 11 30- 50 inclusive 128 8 116
Backlog as of 12 1-50 79 24 103

NO. OF DETERMINATIONS

Samples from NEPA 581
Samples from Y- 12 128
Total 709

261
Part II

LONGER RANGE ACTIVITIES

262
12. VAPOR-CYCLE REACTORS

263
12. VAPOR-CYCLE REACTORS¢!’

North American Aviation, Inc.

C. Starr, A. S. Thompson, and Others

Two types of mercury-vapor compressor jets have been studied. A prelimi-
nary design study of aircraft performance indicates that neither type appears
suitable for flight at 45,000 ft and Mach 1.5. Even if a very optimistic value
of L/D is choseh, indications are that the better of the two 1is still only

marginal.-

Some work has been done on a gas-cooled reactor for a turbojet cycle,
but no refined weight estimates have yet been made. Two characteristics
appear at this stage: (1) The reactor must operate at a very high tempera-
ture, and (2) the pumping power required 1s so great that the turbine size
required 1s several times that required to drive the propulsion duct air

compressor.

The corrosiveness of sodium vapor i1s being measured on potential structural
materials of the sodium-vapor cycle reactor. In addition a program has been
initiated to study the effects of radiation oncorrosion in systems of possible

interest to the ANP,

MERCURY-VAPOR COMPRESSOR JET

Two types of mercury-vapor compressor-jet power plants are being stud-
ied. (2> The first employs a mercury booster heater downstream of the mercury
condenser, and the second employs, instead, a liquid-sodium booster. The
mercury turbine cycle which drives the air compressor is the same in each
case. Mercury is heated in the intermediate heat exchanger and passes through
the turbine to the condenser in the duct air stream.  The mercury booster
heater draws some of the mercury from the intermediate heat exchanger and
condenses 1t; the sodium booster heater draws sodium directly from the re-
actor and cools it in the duct air stream.: The temperature of the mercury is
lower than that of the sodium because of the temperature drop in the inter-
mediate heat exchanger. The reactor wall temperature is 2260°F. The duct air

temperature for each case is given in Table 12.1.-

(1) See North American Aviation, Inc., ANP Quarterly Progress Report July 1, 1950 to Sept. 30, 1950,
ANP-56 (Nov. 24, 1950). !

(2) See Schwartz, H., Dean, A,, and Malone, J. L., Aircraft Nuclear Propulsion Program Progress Report
to July 1, 1950, NAA-SR-86, p. 23 (Aug. 7, 1950).

264
TABLE 12.1

Results of Cycle and Weight Analyses

 

 

Hg BOOSTER Na BOOSTER

Mach No. - 1.5 1.5
Altitude (ft) 45, 000 45,000
Maximum mercury temperature (°R) 1835 1835
Maximum sodium temperature (°R) 2192 2192
Maximum propulsive duct air temperature (°R) 1601 1849
Specific thrust 37.3 42.2
Reactor power (kw/lb of thrust) 7. 44 8.18
Specific power plant weight exclusive of re-

actor (El)b/lb of thrust 3.8 2. 64
Specific power plant weight including re-

actor (1b/1b of thrust 5.2 4.04

 

 

 

 

 

Weight studies were carried out assuming a power plant having a thrust of
100,000 1b at Mach 1.5 and 45,000 ft. The properties of the power conversion
equipment are most sensitive to cycle selection and attention was therefore
confined there. For aircraft configuration studies, a figure of 70 tons (for
the weight of reactor core and shield) was chosen as a representative value.

The results of the cycle and weight analyses are shown in Table 12.1.-

Aerodynamic studies and preliminary aircraft design studies were begun
as soon as preliminary values for specific thrust, weight, and equipment size
were available. The first design study presumed a swept wing, but this proved
impractical because of the poor wing-weight distribution and poor aerodynamic
characteristics at inlet to the outer engines. Consideration was given to a
split wing with no sweep containing a total of 20 engines within the wings.
The reactor and intermediate heat exchanger were contained in the fuselage.:
Some work has also been done on the ducted body configuration but no conclusive

results have been obtained as yet.:

In considering the practical application of a.nuclear power plant to
aircraft, the tentative maximum gross weight of 450,000 lb was assumed for

preliminary investigations.  Planes in excess of this weight could probably be

265
built,but it was the general opinion that the cost of the necessary facilities
for construction would be prohibitive. Of the total weight of power-conversion
equipment (264,000 1lb) approximately 164,000 lb or about 70% is due to the dry
weight of heat exchangers, compressors, turbines, pumps, etc.  The balance of
100,000 1b represents the weight of the charge of working fluids, viz.,
mercury and sodium. Effort is being made to reduce this charge to a realistic
minimum but any possible further reduction due to ingenuity in design would

not be expected to be significant.-

It is the present opinion of the aircraft designers involved that this
plane with an L/D of from 4.5 to 5 cannot be built to fly at Mach 1.5 and
45,000 ft unless the power-plant weight including reactor does not exceed
3.4 1b per pound of thrust. Accordingly, it appears that the mercury-vapor
compressor-jet cycle even with the sodium booster (estimated specific weight =
4.04) 1is not suitable for this application. Even i1f, on the other hand, an
L/D of 6 or greater could be realized, and also if a gross weight in excess of
450,000 lb were permissible, then the mercury-vapor cycle would still be
marginal. Under these conditions the compressor jet, as compared with a
turbojet, would probably not warrant the additional complexity. A more
detailed report relative to the mercury-vapor compressor jet is now in prepa-

ration.

GASEOUS POWER CYCLE

Preliminary analysis of this turbojet cycle, utilizing a helium-cooled
reactor which transfers heat into the propulsion duct through a heat ex-
changer, indicates that the cycle inherently requires a high temperature.
This is because of the gas-to-gas heat exchanger which necessitates high tem-
perature differences to yield useful heat-exchanger weights.: Work on this
cycle 1s now concentrated on an effort to arrive at a preliminary design of a
heat exchanger which would offer hope for successful application to an air-
craft power plant. The analysis is still in such a preliminary stage that

refined weight estimates are yet to be made.

Concentrated effort on this compressor-jet cycle 1s not contemplated
until further investigation of the turbojet cycle has been made.: It may be
that the power to drive the propulsion duct air compressor can be obtained

from the turbine driving the gas-circulating compressor by increasing the

266
turbine size slightly. This is true because of the very large amount of

pumping power required.

SODIUM- VAPOR COMPRESSOR JET

Some very rough weight estimates indicate low values of power-plant
weight (sp.-wt. = 1.1 to 1.3, exclusive of reactor weight).- This fact,
combined with the relatively high specific thrust (57 to 60) at Mach 1.5 and
45,000 ft, indicates promise for the cycle. The maximum temperatures involved
here (3500 to 4000°R) are, of course, much higher than those in the mercury-
vapor cycle discussed above (2260°R). Analytical work on this cycle, inclfiding
aircraft configuration studies, will continue after completion of mercury-

vapor and gaseous cycle investigations.:

CORROSION EXPERIMENTATION

Current activity 1in this experimental program is concentrated on the
corrosion in systems of probable interest to ANP. Initial results are reported
on (1) the corrosive effect of sodium vapor on various potential structural
materials in absence of irradiation, and (2) the corrosion in systems (fluids

and containers) subjected to irradiation.-

Sodium-vapor Corrosion; No Irradiation. The test consists in suspending
a weighed sample of the test material from the top of a stainless steel
"test tube,'" evacuating and outgassing the tube, introducing a small quantity
of purified sodium, and holding at a temperature of approximately 1650°F for
a period of 16 to 24 hr. This bathes the specimen with saturated sodium
vapor at a pressure slightly in excess of 1 atm absolute pressure. At the end
of the test period, the test tube is opened and the sample is removed, washed,
weighed, and subjected to metallographic examination.  The techniques involved
have now been worked out and one test each has been completed on the following

materials:

1. Graphite (AWG).- No attack indicated by weighing or by metallo-
graphic examination.:

2. Stainless steel (type 316). No attack indicated by weighing or
by metallographic examination.:

3.- Tantalum. Attacked at rate of 0.007 in./year as indicated by
weight loss. A rerun of this material is planned for the near
future since there is evidence that the sodium was contaminated
with oxygen.

267
These tests will be extended to other materials. Upon completion of the
relatively low-temperature tests, those materials still in contention will be
tested in sodium vapor at higher temperatures. Equipment for these latter

tests 1s now being designed.:

Corrosion Tests with Irradiation. Fxperiments were undertaken to study
the effects of radiation on corrosion, in systems of possible interest to ANP,
bv exposure to charged particles from the 60-in. cyclotron at Ferkeley. A
comparison of cyclotron and reactor irradiation indicates{3J) that accelerated
tests are possible on the cyclotron, particularly in regard to corrosion
phenomena where i1onization energy density 1s important. Where lattice dis-
placements arising from collisions with the cyclotron beam particles are
important, the relative advantage of cyclotron over reactor irradiation 1is

reduced. The systems under study include:

1. Be2C—air, at elevated temperatures.
2. Pure iron-lithium, at 900°C.

3. Type 316 stainless steel—sodium, at 900°C.-

Two ‘samples of beryllium carbide have been irradiated in the Berkeley
cyclotron. Both samples definitely contained no free beryllium prior to
irradiation.  In order to test the applicability of the analytical method
to interstitial beryllium, the first irradiations were made in a helium atmos-
phere at room temperature. The approximate limit of detectability corresponds
to an irradiation of 30 wpa-hr of 40-Mev alpha particles in which each alpha
particle decomposes one Pe,C molecule. Tests made on the 30 wpa-hr irradiation
sample indicated beryllium present at just about the detectable limit.: Im-
provements in the analytical method, which are in the process of development,
will permit more accurate determination. X-ray diffraction studies showed no
lattice expansion (as would accompany the presence of interstitial atoms) and

indicate that no precipitated phase is present.:

Capsules of Armco iron have been prepared for the iron-lithium experiments
with the assistance of Dr. W. Parkinson of OBNL.  These are to be provided
with thin windows of the same metal, filled with pure lithium, and then
welded shut. The assembled capsules are to be exposed to 40-Mev alphaparticles
while being maintained at 900°C in a pure helium atmosphere. It has been

possible tomake vacuum-tight capsules in only a small percentage of the total

(3) Martin, A, B., and Mills, M. M., The Application of Particle Accelerators to the Study of Radiation
Damage, NAA-SR.56 (July 24, 1950),

268
tried because of the difficulty of welding the window to the capsule. Samples
of Armco iron were irradiated at 800 and at 850°C in a helium atmosphere,
using 40-Mev alpha particles. The duration of the runs was about 2 pa-hr. No
change in the metal microstructure was observed in metallographic examination

after the irradiation.:

Windows and capsules of type 316 stainless steel were fabricated for use
in sodium test assemblies in order to determine the feasibility of making the

window capsule closure by welding.-

INVESTIGATION OF MECHANICAL PROPERTIES AT HIGH TEMPERATURES

Test machines now in the prrocess of design are intended to determine the
following properties of various possible high-temperature materials at a

maximum of 2000°C (3630°F).
1. Short-time tensile strength,-
2. Creep strength.
3. Fatigue strength.
In the tensile machine, stress-versus-strain curves will be determined, and,
if desired, tests determining the time to rupture at a fixed stress can be
run. After elimination tests at 800°C, it 1s contemplated that tests will be

repeated on the remaining materials at a temperature of approximately 2000°C

(3630°F) .-

269
13. CIRCULATING-FUEL REACTORS

270
13. CIRCULATING-FUEL REACTORS*

The H. K. Ferguson Company, Inc.

The design of the major components of the homogeneous NaOH-uranium powered
aircraft has progressed to the extent that a fairly definite picture of the
aircraft may be presented. The 3-in.-diameter reactor 1s provided with a
single 7-in.-diameter control rod from which extremely rapid action is not
required. Xenon is expected to collect in the vapor space provided for the
thermal expansion of the fuel and the control problem is thus considerably
simplified.. NaK 1s proposed as the coolant and a divided shield is mandatory."
The reactor and auxiliary chemical engines (six XJ-53) will fit within the
B-52 airframe with no major modifications.: Weight and thrust calculations

indicate that the nuclear aircraft will exceed the design specification of
Mach 0.8 at 35,000 ft.

Reactor. The reactor is spherical, 3 ft in diameter (excluding re-
flector), and provided with a single 7-in.-diameter control rod. Flow through
the reactor is maintained by a single centrifugal pump which delivers the

fuel near theperiphery with high tangential velocity topromote flow stability.:

The reflector consists of 8.5 in. of stainless steel in three concentric
shells, which serve as the thermal shield, and the 2-in. pressure vessel.
Heat is removed from the thermal shield by partial diversion of fuel flow
between the shells. Each shell is made up of pieces which are individually
pinned to the vessel wall to permit clearance for thermal expansion. Laminated

construction of each piece reduces thermal stress.

The control rod has two separate functions for which mechanisms within
the shield will be designed.: Shim control will be provided for 7% in 8k/k
which covers the temperature range 1200 to 1500°F (3.5% &k/k) plus various
other contingencies (8m, burn-up, etc.).- Lower average slurry temperatures in
the reactor are not permissible. Although a homogeneous reactor has con-
siderable stability by reason of the temperature coefficient of reactivity,
fine control will still be desirable.” This will be obtained by motion of the
rod between shim stops (which are set for withdrawal at less than 0.3% §&k).

Initial calculations on reactor kinetics indicate that neither of these

* This section, except for minor editorial changes for conformity with the rest of the report, is from
a letter of Oct. 27, 1950 to Dr, A. M, Weinberg from The H, K. Ferguson Company, Inc. on the current
status of their design of a liquid.fuel reactor,’

271
mechanisms will be called on for extremely rapid action. Details of the
kinetics depend on pump characteristics, heat—exchanger properties, etc., and

are being worked up to determine required characteristics of control response.

Although data on xenon solubility in NaOH are lacking, consideration of
analogous systems indicates 1t to be too low at reactor operating condi-
tions for xenon to be present in solution. Since there will be a vapor space
in the circuit for thermal expansion of the fuel, it 1s expected that the
xenon will collect there and will not affect reactivity. The control problem
is then considerably simplified; this seems an important advantage for the

circulating-fuel type of reactor.:

The current estimate of fuel volumes in the primary circuit is as follows:

Core 13.5 cu ft
Reflector cooling channels 3.5 cu ft
External circuit 14.1 cu ft

Thus the ratio of external volume to core volume is 1.05. These figures are

based on fairly detailed designs for reactor, pump, and exchanger,

Reactor Coolant. NaK is the preferred coolant for the intermediate heat
exchanger. The relatively mild corrosion properties of NaK may eliminate the
need for a bimetal heat exchanger to resist both coolant and fuel. The low
melting point makes it easy to handle in start-up and shutdown and permits
control of the power cycle by by-passing the engine radiators.  Although the
radiation from this coolant will make a crew shield necessary, the use of

divided shielding is desirable for other reasons as well.

An alternate design using lithium as coolant with a unit shield will be
prepared for comparison.: Some crew shielding may be necessary even in this

case.

Shield. The use of NaK as coolant makes the divided shield the only
possible choice. Since the divided shield will be considerably lighter than a

unit shield, it is a natural choice in any case.

The principal drawback of a divided shield, difficult servicing by the
ground crew, is largely obviated by the nature of the fuel. It is compara-

tively simple to drain and flush both the fuel and the NaK from the power unit

272
with all necessary connections being made by remote control. With this ex-

ception the unit may be serviced as easily as that with a unit shield.

A preliminary study of airplane geometry, weight, and balance shows an
advantage in having part of the weight forward.: With natural position of the
reactor and engines for minimum modification of the contours, up to 25,000 1lb
can be accommodated at the crew compartment without adverse effect on the
airplane balance. Thus the divided shield is consistent with preserving the

aerodynamic features of the XB-52 plane.

The reactor shield is designed to fit within the 9- by 12-ft oval of the
E-52 fuselage. The weight of everything inside the shield boundary (reactor
and contents, coolants, heat exchanger, shield, and structure) is about
75,000 Lb. At present the uncertainty is restricted to last-minute changes to
fit in the fuselage. The shield at the crew compartment consistent with this
shield weighs about 25,000 lb. The calculations are for an air-safe shield
only. Full advantage is taken of shadow shielding.

Engine. The plane will be powered by six engines of the XJ-53 type,
modified to contain the necessary radiator surface as well as the chemical
fuel burners.. The cruising conditions (35,000 ft at Mach 0.8) and sea-level
climb will be met by nuclear power only, the chemical burners being provided
only for take-off and emergency go-around after a missed approach. The
chemical fuel burners will have enough capacity to run the plane entirely on
chemical fuel, since there is little to be saved by installing partial capacity
burners.” Aside from addition of the radiator section and lengthening of the
drive shaft between compressor and turbine, there will be no essential modifi-

ation of the XJ-53 engine.

Calculations of thrust, confirmed by similar calculations by G.E., show
that the present engine design is satisfactory even with the lower turbine
inlet temperature and extra pressure drop between compressor and turbine.

Estimated thrust for six engines is as follows:

Altitude Sea level 35,000 ft
Speed 270 knots Mach 0.8
Turbine inlet temperature 1200 °F 1300°F

Pressure drop, proportion of
inlet pressure

In radiator 12% 13%

In ducts 3% 3%
Thrust 50,400 1b 24,800 1b
Heat to radiators 293, 500 kw 145,000 kw

273
The speed of 270 knots at sea level was picked for best climbing perform-
ance (200 ft/min) on nuclear power alone for a 360,000-1b plane. The 24,800-1b
thrust available at 35,000 ft is actually more than 1s required for level
flight at Mach 0.8 (approximately 19,200 1b) so that the plane could actually
fly somewhat higher or faster.-

A radiator of the shell and tube type using internally finned tubes on a
triangular lattice 1is being considered.” While this arrangement may not be
quite so efficient as the NEPA radiator on a surface-per-unit-volume and
pressure-drop basis, it has the advantage of ruggedness and easy construction
in almost any geometry. Thus, this type of radiator may be made in the shape
of a U which can be easily slipped over the drive shaft or can be made in two

D sections to use the total cross-section area available.-

The actual design of the radiator cannot be completed without further
details on the XJ-53 compressor, which are expected shortly, but tentative
figures show a radiator with frontal area of 7.8 ft? and length of 21 in.
This size will just fit inside the present XJ-53 shell with no bulge.' Inclu-
sion of both radiator and burning section will lengthen the drive shaft about
36 in. and will probably require an additional bearing between compressor and

turbine.

Plane Configuration. The best estimates of thrust available per engine
and of lift-to-drag ratio show that six engines of the XJ-53 type will be
sufficient. An arrangement of six engines immediately behind the reactor in
the fuselage has a number of advantages in shielding, piping, and servicing,
and may actually result in an improved lift-to-drag ratio even when the in-
creased angle of attack demanded by the additional structural weight in the
wing is taken into account. There may be no increase in airplane weight with
this arrangement because the fuselage installation of the engines allows a
saving of approximately 25,000 lb in piping, insulation, supports, and coolant.
Thus the improved lift-to-drag ratio gives a proportional reduction in re-

quired thrust.

A further advantage in this compact arrangement is that development and
testing of the plane could proceed more quickly.: The plane aerodynamics could
be checked with a dummy engine group powered by chemical fuel, while the

nuclear power plant could be completely designed and tested on the ground.

274
The final step of installing the nuclear unit in the plane would be relatively

uncomplicated."

Ground Facilities. The airport for a nuclear plane must be equipped for
certain routine tasks.. The plane must be warmed up, filled with fuel and
coolant, and checked before flight. Upon returning from a mission it may be
completely overhauled or it may need only a few minor repairs, such as changing
tires, before taking off on the next flight.- Whenever nuclear fuel 1is in the
reactor, its temperature must be in the range 1200 to 1500°F because of control

limitations and danger of freezing.

Normal start-up is as follows: The NaK system is first filled and
started circulating through an auxiliary heater. The reactor is then heated
up by circulating first hot helium and finally hot sodium hydroxide through
the reactor—heat-exchanger system. When the system reaches a temperature
within the range 1200 to 1500°F, the sodium hydroxide is drained and replaced

with fuel preheated to the correct temperature.:

When the plane comes in for servicing, it is first located at a dump
point at which the fuel and NaK systems are drained and flushed.  The plane 1is
then towed to the regular loading buildings for servicing. If the plane is to
spend only a short time on the ground, the reactor may be filled with sodium
hydroxide and kept hot during the turn-around. The dumped fuel may be replaced
"as is" in the reactor or may be chemically processed to remove fission
products. The actual cycle will depend on cost and holdup in the chemical

cycle and degree of fuel burn-up per flight.

275
14. SUPERCRITICAL WATER REACTOR

276
14. SUPERCRITICAL WATER REACTOR

For the past several months various members of the Aircraft Reactor Branch
of the Reactor Development Division in the Washington office of the AEC have
been developing a design for an aircraft reactor cooled and moderated by
supercritical water, along the lines mentioned in the Lexington Report. (1)
This work has now been summarized(2?) and the Oak Ridge National Laboratory has
been asked by the Washington office of the AEC to comment upon the usefulness
of such a system for an aircraft reactor. In order to develop the details of
the design a little further than they were carried in the Aircraft Reactor
Branch Report, WASH-24, ORNL asked its contractor, Nuclear Development
Associates, Inc., to prepare an independent design study on this subject. NDA
has now submitted an informal interim report{3) stating that it should be
possible to construct a reactor substantially the same as the one proposed in
WASH-24. It is also felt that it would be no harder to shield this reactor
than others considered for aircraft use. However, it was stressed that the
difficulties of operating a nuclear aircraft on a supercritical water cycle
lie much more in the engine and aircraft phases of the problem than in the

reactor.

Both NDA and ORNL are continuing the study of the supercritical water

system.

(1) Lexington Project, Nuclear-powered Flight. A Report to the Atomic Energy Commission, Lex P-1, Sec.:
III, pe 57 (Sept. 30, 1948),:

(2) Aircraft Reactor Branch of USAEC, Application of a Water Cooled and Moderated Reactor to Aircraft
Propulsion, WASH-24 (Aug. 18, 1950).°

(3) Gruber, A.; Preliminary Comments onSupercritical Water Cycle, NDA, ANP-55 (Nov.;6, 1950).:

277
15. SUPERSONIC TUG-TOW SYSTEM

278
15. SUPERSONIC TUG-TOW SYSTEM

During the closing week of the quarter, in response to suggestions by
Dr. E. P. Wigner and by Nuclear Development Associates, Inc. (see Appendix A),
the Oak Ridge National Laboratory began the study of tug-tow aircraft reactor
systems with particular interest in supersonic tug-tow. If, as suggested by
NDA, it is possible to build a shield for the tug plane which weighs 1in the
neighborhood of 15,000 1lb, then the tug-tow system might permit the use of
very small aircraft.: Although the operational difficulties of tug-tow might
be fairly great, the present international situation does not permit over-
looking any systems which could possibly result in a supersonic aircraft of
global range within a relatively short time. The investigation of a Supersonic

Tug-tow Aircraft Beactor, with its associated engine and planes, is being

called Project STAR.:

279
16. LITHIUM-ISOTOPE SEPARATION

280
16. LITHIUM-ISOTOPE SEPARATION

W. M. Leaders, Materials Chemistry Division

During the last quarter all the research aimed at the separation of
lithium isotopes has been concentrated on the electroexchange process, which
is the process based on the countercurrent chemical exchange reaction between
lithium amalgam and aqueous lithium hydroxide. This process was described in
detail in the last quarterly report.{!) Subsequent development of the 48-
compartment electroexchanger has eliminated the transport problem through
constructed openings and has greatly reduced back-diffusion. With these
modifications, samples of 94.5% Li’ (when operated to enrich Li’) and 90.3%
Li7 (when operated to enrich Li®) have been obtained. Other experimental
electroexchange apparatus has been designed and built, and is now being
operated. The production problems of lithium separation by electroexchange
have been studied, and production appears to be feasible; the major difficulty

1s in reaching sufficiently high reflux ratios in the amalgam-forming section.

Operation of the 48-Compartment Electroexchanger. The 48-compartment
apparatus was equipped so that eight of the chambers could be utilized as an
internal refluxer for the preparation of amalgam. (This modification further
reduced the transport difficulty mentioned above although itdid not completely
eliminate the deposition of lithium hydroxide.) Operating the equipment with
only 40 exchange compartments both as an enricher for Li’ and as an enricher

for Li® yielded the following limits of separation:
1. Up to 94.5% Li’ when operated as a Li’ enricher.
2. Down to 90.3% Li’ when operated as a Li® enricher (or Li” stripper).
Based on an a factor of 1.05, these separations amount to seven stages of
enrichment and six stages of stripping. It is considered that the results

have definitely established that the fundamental principle of the method is

sound for enrichment of either of the isotopes of lithium.:

Considerable difficulty was encountered during the early phases of these

experiments in effecting the transport of the amalgam through small pipes and

(1) Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending August 31, 1950,
ORNL-858, pe.: 122 (Dec, -4, 1950).:

281
other constricted openings. This problem has been diagnosed as being caused
by the deposition of solid lithium hydroxide on the surfaces of the pipe and
other containing walls which are in contact with the amalgam. It appears that
some surface wetting by the aqueous solution 1s occurring in all parts of the
equipment (constructed of lucite and plexiglas) and that reaction with the
lithium in the amalgam takes place to such an extent as to exceed the solu-
bility limits of lithium hydroxide in the limited supply of water available.
No particular difficulty has been encountered in portions of the equipment
being agitated. Larger sized pipes have proved satisfactory for the current
experimental program, however, and it is realized that the problem might recur
in these larger pipes 1f the duration of the experiments should be greater

than a few hours.

During the course of the experimental work on the 48-compartment apparatus,
considerable time was spent in studying the problem of back-diffusion. It is
a well-recognized fact that in any apparatus operating on the countercurrent
principle, any back-diffusion or countercurrent flow greatly reduces the efficiency
of the apparatus. In the 48-compartment apparatus all the cells are 1n one
line. This arrangement 1s most readily assembled, and this was the deciding
factor in the design of this piece of equipment. As originally built, the
openings between the cells were all on the center line of the apparatus, The
excessive amount of back-diffusion encountered with this particular design of
cell dividers necessitated a major revision. Smaller slots (1/16 in.) were
cut on alternate sides of the cell dividers to provide the most tortuous path
for the liquids. The apparatus was operated in this condition during the most
successful runs, the results of which are quoted above. The major difficulty
with these small openings was the limit placed on the throughput of the
equipment because of the resistance offered by these constrictions to the flow
of the liquids. This particular problem should become less troublesome in

larger scale equipment.

72 Compartment Electroexchanger. A second device was delivered from the
shop during the quarter. This piece of equipment, although more difficult to
construct, embodies several features which should reduce if not eliminate some
of the difficulties inherent in a straight-line apparatus.  This exchanger
contains 72 individual compartments, 1 in. in diameter and 2 1in. deep. These

are arranged in two rows of 36 each and the interconnections between chambers

282
are arranged to give a zigzag flow through the equipment. The agitators are
synchronized by a central worm shaft. Although the connecting canals are % 1in.
wide, the back-diffusion encountered in this device 1s extremely small when
compared to the best results obtained in the 48-chamber in-line model. The
apparatus has only recently been reduced to trouble-free operation and no re-
sults are as yet available. It 1s currently being used to determine more
accurately the a factor of the method under various rates of production. This
may be one of the more reliable methods of definitely establishing the exact

value of a for this process.:

Alternative Type Electroexchanger. When the decision was made to curtail,
at least temporarily, all work on molecular distillation, the group which had
been conducting this research turned their attention to the amalgam system.:
They designed a somewhat different type of apparatus for countercurrent con-
tacting of the amalgam and aqueous solution. A schematic drawing of the
current design 1is shown in Fig. 16.1. The apparatus has only recently been
operated for sufficient time to reach isotopic equilibrium. The result from
the first successful run showed an increase in Li’ from 92.5 to 93.6%. One
of the features of this type of equipment is the fact that back-diffusion 1is
absolutely eliminated. The only problem from a flow standpoint is the possible
short-circuiting of liquid through a chamber before equilibrium has been
achieved. As the work progresses it should be possible to evaluate this method

more accurately in comparison with the horizontal types of equipment.

Analyses of Production Problems. After it became apparent that the
general principle of the electroexchange process was sound, the scale-up
problems of production were of major interest. In order to get a more accurate
indication of what these problems might be, a large size in-line model ex-
changer was built. A special baffling device was incorporated in this apparatus
in an attempt to reduce back-diffusion. The effectiveness of this baffling
device has not been completely established, but preliminary tests indicate
that it is quite effective. The apparatus consists of 18 cells operating in
groups of 6, one group being used as a Li’ refluxer for continuous preparation
of amalgam, one group as an exchange section, and the third group as a strip-
ping section to remove lithium from amalgam. All agitators are individually

powered. "

283
" DRAWING NO, 10245

ROTATING DRIVE SHAFT
(COMMON ANODE)

  
     
    

STAINLESS
STEEL

AGITATOR PLEXIGLAS

AGITATOR

    
  
  
   
  
 
  
   
  

'

——
=

   
    
 

= —
L
b
—

Li1OH
FLOW

LiOH

- == CATHODE
e CONTACT PLATE

=T ' CATHODE
AMALGAM | /5= ‘
FLow TERMINALS

LiOH
FLOW

  
  
  

{
|
\

   
 

R

Iy

      
 
 
 
 

AL
W

 

 

. N :; A AMALGAM

AMALGAM
FLOW

FIGURE 16,1 VERTICAL EXCHANGE APPARATUS
(DIAGRAMMATIC)

284

 
Although the work to date has been essentially of a preliminary nature
with the purpose of pointing up the major problems with the apparatus, it has
indicated that scale-up of the process is feasible. The one major difficulty
encountered is the inability of the present apparatus to reach sufficiently
high reflux ratios in the amalgam-forming section (Li” reflux). Research work

1s now directed toward finding a solution to this problem.

Considerable research has been done in an effort to find a suitable anode
material from the standpoint of cost, electrical properties, and product
purity. Platinum 1s the only material found which 1s suitable for use 1in the
high-voltage electrolysis section of the Li’ refluxer. Fecause of the obvious
cost disadvantages of platinum it would be most desirable to find a substitute.
In the exchange section of the apparatus, where the potential applied 1s

quite low, 1t has been found that type 302 stainless steel 1s a suitable

electrode material. An investigation into the possible use of graphite 1is now
underway. To date no sample of graphite has been found which possesses the
necessary mechanical properties. It appears from early observations that the

gas (0,) released at the anode is causing the surface to disintegrate as a
very fine powder which remains in suspension in the aqueous phase. Research

is continuing on this problem.

Recently a closer examination has been made of the factors i1nfluencing
the product cost in a large-scale plant, and among the several fundamental

aspects shown to be of prime importance in future work were the following:

1. The quantities of mercury required will be very large, and every
effort must be made to increase plant capacity and reduce mercury
holdup. Some study should be made of the probable losses of
mercury per cycle since even small losses will add materially =o
the product cost.

2. Power costs are likely to be fairly high, and every effort
should be made to obtain maximum current efficiency. Since most
of the power is used in the refluxing sections, this points up
the necessity of continuing the effort to design an efficient
refluxing mechanism.

o
n
APPENDIXES

286
APPENDIX A

REPORT OF NUCLEAR DEVELOPMENT ASSOCIATES, INC.

Rather than separate the brief discussions of the diverse topics discussed
in the "Quarterly Report on ANP Activities"(1) from the Nuclear Development
Associates, Inc., they are presented here together, but with identifying

paragraph headings.

Xenon Cross-section Computations. Computations on the xenon cross
section averaged over a Maxwell distribution, giving the cross section and its
temperature derivative as a function of temperature, have been completed and

reported (Goertzel, Oppenheim).

Critical Mass of Bare Beryllium-moderated Piles. Calculations on the
critical mass of bare Be-moderated piles over a wide range of fuel concentra-
tions, including a perturbation method for finding the effect of added absorbers
on fuel requirement, have been carried out and a report is being written
(Preiser, Shapiro). A similar study for H-moderated reactors has been started.
Some additional considerations pertaining to aircraft reactor control are
also intended. Beryllium absorption at thermal (n-y¥) and high (n-a) energies,
and production (n-2n) at high energies, is being reviewed with Goldstein of

the NDA data group and Hughes of Brookhaven.

Supercritical Water Reactor. A preliminary review of the Williams
supercritical water proposal has been undertaken. The reactor and shield were
looked at briefly, and liaison has been established with United Aircraft which

is studying the propulsion system and airplane aspects of the proposal.

Tug-Tow System. The shield weights needed with the Lexington Tug-Tow
arrangement appear to be considerably less than the values indicated by the
Lexington project. For example, if 40 cm of water (Lex P-1, p. IV B-16) 1is
placed around a 60-cm-radius spherical reactor, and in addition a lead disk of
1 meter radius with a thickness (about 100 g/cm?) sufficient to provide a 7y
attenuation of 60 (Lex P-1, p. IV B-14) 1is provided, the water and lead weight
will be a little over 3 metric tons each, or about 7 tons total.: Thus
relatively midget planes, with midget power outputs and substantially smaller

fuel reguirements, might be used for tug-tow flight.

(1) Letter to C, ‘B, Ellis from NDA, Nov. 30, 1950.

287
APPENDIX B

LIST OF REPORTS ISSUED

 

 

 

 

 

 

 

88 ¢

 

 

Cylindrical Geometry

 

 

i

 

REPORT NO. TITLE OF REPORT AUTHOR(s) DATE ISSUED?
Design of the ARE
Y-F&-3 t eactor Maintenance and [lisassembly . V.. Schroeder 106-24-5¢
Y-Fg&-4 tvaluation of Coaxial (C linder Insulation System S. V. tanson 10-30-50
Y-F8-5 Lecture Notes of Heat [xchanger System of GE Project SIk S. V. Manson 10-13-50
Reactor Physics
% ANP-5201 { Calculational Procedure for Control Data N. M. Smith 9-13-50
é Y-F10-20 | Besults of a Test of the Linear Approximating Multigroup D. K. Holmes 11-6-50
§ Calculations
E Y-F10-21 IBM Multigroup Numerical Procedures D. K. Holmes 11-28-50
i O. A. Schulze
gY-F10-22 FResults of Some Bare Calculations of Critical Mass and J. W. Webster 12-1-50
: Reactivity Effects L. T. acauley
gY-FlO-BO Perturbation Equations for the Kinetic Response of a Licuid- N. M. Smith to Le
fuel Reactor T. Rubin 1ssued
, M. J. Nielsen
: R. Coveyou
' Y-F10-17 ! Xe Effect in an Epi-thermal Reactor J. W. Webster 10-10-50
L Y-F10-15 Calculation of Average Lifetimes of Neutrons Using the Results D. K. Holmes 10-2-50
of Multigroup Calculations
Y-F10-18 Bare Pile Adjoint Solution M. J. Nielsen 10-27-50
Y-F10-19 ‘Self-shielding in Uranium J. W. Webster 11-8-50
Y-F5-22 Numerical Integration of the Multigroup Reactor Equations 1in M. C. Edlund 9-27-50

i
686

 

 

 

 

 

 

REPORT NO. TITLE OF REPORT AUTHOR (s ) DATE ISSUED E
Critical Experiments
CF 50-10-1151 Fabrication of Fuel Discs for ANP Critical Experiments A. D. Callihan 10-23-50
CF 50-11-120| Storage and Handling of SF Material for ANP Critical A. D. Callihan 11-30-50
- Experiments
Nuclear ieasurements
CF 50-12-25 Estimate of the Effects of Higher Levels on the Hesonance G. E. Arfken 12-6-50
Integral of Xel?S | T. A. Welton
CF 50-12-45 High Level Sources of 92-hr Xenon z G. W. Parker 12-15-50
Shielding Research g
ANP-53* Keport of the ANP Shielding Board . NEPA, CHNL,and | 10-16-50
OKNL-710 Theoretical and Practical Aspects of Shielding ? A. S. Kitzes 9-29-5(Q
Y-F15-4 Divided Shields for More Conservative Ground Specifications ; T. B. Mitchell 11-20-50
Y-F20-1 Unmanned Nuclear Powered Aircraft and RW E V. K. Ergen 10-10-50
Y-F20-3 Measurements of Fission Fragment Decay Gammas in the Swimming é W. K. Ergen 11-28-50
FPool !
i
Liquid-metal and Heat-transfer Research {
i ,
CF 50-9-24 A Survey Concerning the Use of a Liquid Sodium—Austenitic i R. J. Radebaugh | 9-7-50
Stainless Steel Heat Transfer Unit Under Conditions Antici- !

 

pated in Nuclear Aircraft QOperations

* This report incorporates many earlier reports on various aspects of shielding.

A o~ S

 

 
068

 

REPORT NO.

TITLE OF REPORT

AUTHOR( s)

 

 

 

Y-F8-6

Y-F8-7
Y-F19-2

CF 50-9-83
CF 50-9-127
Y-F15-3

HKF-109

ANP-55

ANF-54

 

Theoretical Equations for Estimating Thermal Conductivity of
Liquids

Physical Data for Substances in the Fused State

Design Charts for Sodium to Lithium and Lead to Lithium

Intermediate Exchangers

Metallurgy
ANF Materials Projects
ANP Materials Program

Transient Temperature and Thermal Stress in an Infinitely
Long Solid Cylindrical Kod with Surface Temperature Chang-
ing at a Constant Rate

Circulating Fuel Reactors

Homogeneous Reactor for Subsonic Aircraft

Supercritical Water Reactor

Preliminary Comments on Supercritical Water Cycle

Miscellaneous

Interim Report of the ANP Control Board

 

S. V. Manson

V. Manson

W. C. Cooley

E. C. Miller
E. C. Miller
J. G. Duffy
H. K. Ferguson

Company

. R. Gruber(NDA)

NEPA, ORNL, and

C&CCD

DATE ISSUED

11-13-50

 

——
- —k

11-20-50
9-1-50

9-19-50
9-26-50
10-27-50

12-15-50

11-6-50

11-1-50

&