EIR-259.txt

EiR-Bericht Nr. 259

EIR-Bericht Nr. 259

Eidg. Institut flr Reaktorforschung Wirentingen
Schweiz

A High-Flux Fast Molten Salt Reactor for the
Transmutation of Caesium-137 and Strontium-80

M. Taube, E.H. Ottewitte, J. Ligou

l__l

Wiuarenlingen, September 1975

[ I

EIR-Bericht Nr. 259

A High-Flux Fast Molten Salt Reactor for the

Transmutation of Caesiumn-13%7 and Strontium-90

M, Taube, E,H. Ottewitte, J. Ligou

September 1975
sSummary
1. Introduction
2., rormulation of Reactor Requlirements

2.1 Minimal System Doubling Time Requirements
2.2 Heed for Central Flux-Trap
2.5 Determination of Flux-Trap Size from Destruc-

tlon-Production Balance Reguilrements

Weutronic Consideration

5.1 Burner Reactor Calculations

5.2 IPoderator Requirements

4.4 Motilvations for Molten Salt Fuel

5.4 Outer-Reflector Zone Consideration
Tnermohydraulic Considerations

Murther Parameters of the Burner Reactors

Effect of Keplacing Chlorine with Fluorine

Remarks about Transmutation and Hazard

Coefficients
Conclusicns
Acknowledgment

neferences

page
Summary

A high flux molten salt (plutonium chlorides) fast reactor

(7 GWth) with internal thermal zone for transmutation of Sr-=90

and Cs-137 is here discussed. These fission products have been
produced by breeder reactors with total power of 23 GWth and by the
the 7 GWth fast burner reactor.

For the case when the power breeder reactors achieve a breeding
gain G >0.2 the doubling time for whole system, including the

high-flux burner reactor, equals ~30 years,

The transmutation of 3r-90 and Cs-137 in a total flux of
3,8-1016n cm_gs_l, and thermal flux ~2,O'1016n cm_gs"l can
achleve the steady-state which corresponds to an effective
half-life of 1,8 year for Sr-90 and of 8,9 years for Cs-137.
In terms of hazard coefficient the transmutation system gives

an improvement of 14 times.

The impact of numerous parameters is discussed, e.g., chemical
form of both nuclides, nature of moderator in thermal zone,
moderator, layer, fuel composition, radius of thermal zone,
ratio of burner/breeder reactors etc, The effects of replacing
chlorides with fluorides as fuel for fast core is also dis=-

cussed.
1. Introduction

The alm of this work was to study the concepfi of transmuting
Sr-90 and Cs-137 in a very high-flux reactor. These two
isotopes are of particular interest because of their high
yield in fission, their longevity and their biological hazard.

The rather high yields (Y: mol %) from fission are

U-233 U-235 Pu-239
Sr-90 ~6,2 ~5,1 2,2
Cs=137 6,6 ~6,0 ~6,7

Assuming each of the above fissile fuels to be equally prevalent
in the future, the mean yield is 4.1% for Sr-90 and 6,4% for
Cs=137.

The half-lives (28 years for Sr-90 and 30 years for Cs-137)
indicate the longevity of the isotopes. To "significantly" reduce
their activity (by a factor of 1000) through natural decay would
take 300 years,

The blological hazard 1s reflected in the maximum permissible
concentrations (IAEA, 1973):

Nuclide in air3 on water
(uCi/em”) (uCi/em?)

Sr-90 1.207° 1.107°

Cs=-137 6-10"8 4-10'Ll

This 1indilcates that Sr-90 is roughly 50 times more hazardous than

Cs=13%7 and warrants prime attention.
These facts have understandably prompted numerous transmutation
studies (Schneider 1974; see also Taube, 1975). The most opti-
mistic has been ftransmutation in a controlled thermonuclear
reactor CTR (Wolkenhauer, 1973). However, even unrealistic flux
levels in CTR could only reduce the transmutation half-1life
from 30 years down to 5-15 years. The time required for
"significant" reduction is five to ten times longer than indi-
cated above, That would be longer than the usual life of a power
plant. Furthermore, Lidsky (1975) has pointed out that for any
reasonable ratio of fusion burners toc fission reactors, the
burners soon contain a much higher radicactive burden than the
fission reactors themselves. This is certainly undesirable, and
probabvly intolerable. A motivation for alternate solutions

remains.

This work examines the basic reguirements to obtain a system cf
reactors which provide adequate breeding and self-destruction of
figsion products. These requirements lead towards the inclusion
in the system of a molten salt burner reactor with a central
flux-trap (Taube 1974).
2. Formulation of Reactor Requirements

2.1 Minimal System Doubling Time Requirements

1t was assumed at the outsef that a very-high flux reactor for
fisslon product transmutation would be special, only one being
bullt for every L (breeder) power reactors. Such a system of
L+l reactors should provide a doubling time T <30 years;
this 1s estimated to satisfy the needs of future world power

development.

Doubling time i1s defined by

1 GWatt thermal operating day 1 year

1.1 kg fuel fissioned 365 days

1 kg fissiles destroyed (absorbtion) InZ doubling time
X X
G kg fissiles net galn per e-folding time e-folding time

(I+F) kg fuel fissioned 1 kg fissile fission 1

kg fissile fissioned 1+a kg fissile destroyed (abs)

1 calendar day SI kg fissile s in systen

C operating day MWatt thermal

)5S L . I+ 1
(I+1) Isys 1ne 1.73 (I+F)S Sys

(1+u)CuSyS x 1.1 x %65 (l+a)CGSyS
N

where fuel = fertiles and fissiles

fissiles = Pu-239, U-235 and U-233

SIi = speclfic fissiles inventory in the reactor sub-
system 1 = I./P.,
i" 71
Ii = fissiles inventory in reactor subsystem i (inclu-

ding fuel cycle).

Pi = thermal power level 1n reactor type 1
I = fertile to fissile fission rate =~.2
Q = capture to fission cross section ratio ~.2
C = fraction of time at full power ~.8
i = I
oI bu/Ibr
bu = burner reactor subsystem
br = breeder rcactor subsystem
T) -
“p Pbu/Pbr
shen 1+G. P+ .7 LG, + B_G
a - br “br bu ~bu _ br P bu
sys : - L+
s Pb bu b BP
T + +
o _  8ys  _ ler Ibu ~ (L BI) a7
- s5ys P ) LP + P ) (L + B.) br
SyS br ou P

It

wetting (conservatively) Gbu -1, and dropping "br" subscripts

one nas

Tnen, assumlng the previous nominal values for a, C, and F, the

doubling time reqgulrement 15

2.16 (LB,
T F < 30

Solving for X = L/BP, the ratio of thermal power in the L bree-

ders to that consumed in the burner, the requirement becomes

B /B, + 13.89/81

X >
(13,89/8I) ¢ - 1

In Table 1 this relation is studied over reasonable ranges of
values for G, SI and BI/BP' As one mlght expect, the possible
variations in G produce the largest changes in the requirements
on X, Furthermore, Table 1 suggests a minimum requirement that

X 23, with nearterm practical requirements approaching X = 10.

Table 1 Minimum "X" Requirements to Accomplish a Burner/Breeder

System with 30 Years Doubling Time

G = Breeding Gain = BR-1
MwWwatt thermal Advanced Current
speclific fissile Breeder Breeder
inventory Art Art

kg fisslile in system

ST BI/BP G = .4 G = .2
1 3.09 7. U3
0.8 2 3,26 7.83
3 3. 42 8.24
1 5,27 &.37
1.0 2 5,49 8,94
3 3,71 9.50
1 3,47 9,57
1.2 g 3,74 10.3%

2.2 Need for Central Flux Trap

To accomplish a significant fission product transmutation rate
will reqguire high neutron absorption rates. For both Sr-90 and
Us=1357, the thermal absorption cross section (in barn) is one

or two orders of magnitude greater than in the fast neutron

region:
EP o(.0253% eV),b o(fast reactor),b c(.,025% eV)/ofast
Sr=50 0.8 0.0076 ~100
Cs=-137 O.11 0.0137 ' ~ 8

Consequently, the transmutation might be best accomplished in a
central thermal flux trap surrounded by a fast fuel region (for
high flux levels), If o¢(total spectrum, flux trap) =

1/2 « o(E = .0253% eV) the estimated total flux étot just to

match (K=2) the natural decay rates is

ffor Cs=1357:

hgd3T 16 —2 -1 . 137

1.%*107 " n ecm s ("R = 7.35-10_103_1

o
11

.

MO
it

)

7.88+10 1% "1y

HH
H
:
N
I
ro
=
o
O

@)
3
—~
w
I
2.3 Determination of Flux-Trap Size from Destruction-Production

Balance Reguirement

Another important requirement will be that the FP transmutation
rate 1n the burner flux-trap (FT) exceed the production rate in

the power and burner fuels:

flux - Mtransm. Velux L N
br bu
trap trap
° _ dNi _ . JFP e1~10 .
where Ni = 3T Pi Y 3,110 (fiss/Ws)
where Ni = the number of F.P. atoms in subsystem 1.
FP _ _ ,FP. FP
transm, (00 + kB>FP =g
using Pbu = BP Pbr and X = L/BP - get
FP
Y ©P
NFP ____bu_ (X + 1) - 3.1_1010
. FP, FP
fine K AB
trap

(Any additional transmutation beyond this would help to remove FP
inveatory from reactors outside this system. However, for trans-
mutation to be a significant improvement over tne 1000-fold
decrease 1in 300 years by natural beta decay, it should then be
accomplished 1n a much shorter time. A reasonable goal 1s 30 years

(K=10), the lifetime of a power plant.)
Assuming a central thermal flux trap in spherical geometry, the

racgius 1s tnen

| | T1/3
R (cm) I 3.1'1010(X+1)109 P {(gigawatts) . YFP ;
A1 ux | Hm 5,023 - 1025 atoms KFPQFPA FP,
- mole B
trap

an

P . . - :
where pi = density of FP nuclei (moles cm 3) 1n the flux trap

reglion,
10

3. Neutronic Consideration

3.1 Burner Reactor Calculations

To analyze the burner reactor, calculations were made by ANISN-
code in the Su transport-theory approximation and checked in 88
with 23 energy groups and approx. 100 spatial positions (checked
by 160 spatial positions). Pl—approximation cross sections were
produced mostly with the GGC-3% code which utilizes ENDF/B-1 and
-2 and GAM data. Sets for Cs=-137, Sr-90 and F were made in

less exact fashion from AAEA and Hansen-Roach data, Fig., 1

shows the group structure,.

A reference burner reactor model is shown in Figure 2, The flux
trap 1s surrounded by a BeO spectrum-converter, a critical fuel

thickness, and an outer wall {(see Table 2).

Figure 3 shows the calculated flux distributions. The total flux
in the fuel is similar to that in the flux trap. The calculated
fluxes lead to the conclusion (used in section) that o(total

spectrum, flux trap) = 1/2 « o(E = ,025% eV),
10

12
J

11 -

11

Pig., 1 Neutron spectrum in the core
- -0 -
Total mean flux H.OB'lOlO 27
Specific power 10.1 kwem

Total power 7 GWth

arbitrary
units

Hedytrom energy group:

21 19 17 15

N

15 12 11 10 9 3 ! 6 5

Lethargy

Flg. 2

12

High flux burner reactor
Reflector
Target
J ob g 100 " 2
-
] Total fluxX ———jéi
T Thermal flux B \ /
o (2% groups) \ ///
o ‘/
_ Va7
,/" 1//
//”” *//
] Fast I'Lux "
O. group .~ 4
7 (1,35 —/?.,Kfew )/
4 ‘/' /
./
1 ..~ /
EEE

13

Table 2 High-flux burner reactor with chloride fuel
Total power 7 GWtn
Zone Components Neutron flux |Specific power
o 24, -3 16 -2 -1 -3
Radius (cm) atom 10~ /cm 107" n em “s (kW cm )
Volume (cmB) total transmutation
thermal rate (s“l)
I
0 + 78.5 cm Ce=-137 00,0116 w9
Target . Sr-90  0.0016 5,83 Cs=i50 ~ 1,7°10
Vol. 2+10 cm 8] 0.0145 2,05 -8
0 2 o1kLn Sr-90 1,110
11
78.5 - 88 cm Be 0.060 %l%%
Moderator 0 0,060 ’
with thin 4,365
graphite layer 0,201
IIT
58.0 - 94,56 em| Pu-23% 0.0014
tuel Pu-240 0.0004 -5
* 0
Pu-241 0.0002 4,05 L0, 1 kiem
) 0,0156
Vol.6.9-107cm” | & v.lel
e Cl 0.0180
IV
94,6 - 97.6 cm| Fe 0.08 4,00
Wall 5,10"5
V
¢7.6 - 200 cm Fe 0.083 5,39
Fellector 1,8‘10_5
o, 41077
10-12

14

5.2 Moderator Reguirements

To accomplish a thermal neutron flux trap one must naturally
employ neutron moderating materials in and about it. As is well
known, light materials can scatter neutrons past the neutron-

absorbing intermediate-energy resonance region. iH 1s the most

efficient nuclide in this respect but also exhibits appreciable
2D 9
172 5
1s a blt heavy though frequently already present in a molecular

thermal absorption. Be and 120 are usual alternatives. 80
combination. Other light nuclides have unacceptable nuclear or

physical limitations,

Considering chemical and physical properties, the logical mate-
rials to be used inside the flux trap are hydroxide and/or
deuteroxide compounds of the FP, Figure 4 shows that just a small
proportion of H molar fraction has a large deleterious effect on
the Cs=-137 transmutation rate., This is due to the H absorption

cross section, Therefore, CsOD and SP(OD)2 are preferred.

As Sr-90 and Cs-137 also have their fair share of resonances it

1s advantageous to thermalize the flux before reaching the flux

trap region containing these targets. Therefore a spectrum con-
verter between flux trap and fast fuel is needed. Bearing in mind
the high temperatures to be obtained in this reactor and possible
chemical reactions with molten salt, HEO and D20 are unacceptable.
This leaves Be, BeO and graphite or some variant therefore for
consideration., Be {(and D) compounds, of course, have also to
their advantage a relatively low (n,2n) threshold (1.67 MeV),.
Location next to a fast region can therefore produce considerable
extra slow neutrons 1n the flux trap - which is a main objective
of the burner reactor. Replacement of Be by C or Mo wall material
should therefore lower the FP transmutation rate, and it does
(Table 3)., Figure 5 indicates an optimum thickness of about 5 cm Be.

For the sake of safety and higher melting temperature, BeO 1is

preferred over Re,
Fig., 4 Hydrogen versus deuterium as moderator

Strontium-90 and
Caesium-137

activity

120 %7 7

Caesium-13

110 4
100 % = Strontium-90
‘_ *
—9
90 5 -
80 %

0 0,1 0,2 0,%

Hydrogen mol ratio
b
(@A

Table 3 ffect on Replacing Be Converter upon the

Relative FP Transmutatiocn Rates

!
Be | Be !
! nmolten-salt
C : Be
flux | ffast driver
trap Be I'e /Mo
. . A . A
canse materials transm, (Cs=-137) transm., (Sr-90)
1 Be, Be 1.0 1.0
2 C, Be 0.72 0.68
3 Be, Fe/lMo 0.84 0.82

(Remark: transmutation rate in arbitrary units)
17

1
H.
072

> Thickness of Dberyllium moderator zone

Caeslum—-13%7 and
Stront tim=90

activity

1504

Caesium-137

110 =

Strontium-90 L

9 1 2 3 I 5 6 7

hickness, cm
18

5.3 Motivations for Molten Salt Fuel

In a fast plutonium reactor the average fission cross seetion is
estimated to be 2.5 barns. The Pu atom density is ,002 atoms
cm_lb—l. The total flux is of the same magnitude as in the flux
trap. The specific power for the Cs-137 burnup is then (using ¢

from section 2.2)

15

N(Pu)o_, ©/3,1+10 = 2.1 (K=1) kW Cm_5

f

The rise of the thermal flux in the fast fuel (Fig. 6) leads
Lo an order of magnitude increase in the fission density at
the Inner fuel boundary. For a solid fuel core the resulting
peak power posltion would present extreme demands upon coolant

veloclty and flow distribution.

To minimize thilis, the addition of boron in and about the fast
fuel was considered. Reactor calculations were made for 5 to
100 micron-thick intermediate walls as well as for distributed
boron inside the fast region. As seen from Figure 7 the FP
transmutation rate is always reduced; although the spectrum in
the fuel region 1s harder, the loss of thermal neutrons to the

fiux trap has a greater effect.

From the above one sees that the fuel melting point and the
thermohydraulic reguirements effect a loss in FP transmutation
rate for solid fuels. One solution might be to use liquid fuel,
cooled out-of-core, Turbulent flow in the core will alleriate
the thermal peakling problem. Also, the melting point limitation

15 removed,
}‘_l
09

107 o

Fission
density

Cission

D i
cnorsec

19

O Fission density in the fuel zone

seryllium

/

/

2z

F¥ssiq
Gensxty

o te

/.

Wall

130

104,06

106,6

108

111,14
20

Fig., ¢ Impact of natural boron

0.5 10t \
Actavity \\\ o
. Thermal
of Thermal \t\ f1ux
Cs-137 |[flux in N
lozb/secfue}gzgfie \\‘\
dnem™<es
Cs=137
8 activity
0,7 =10
0,0_107
0,5 —10°
A
=)

Boron thickness, mm
21

The disadvantages arise from the increased fuel inventory in
the burner, e.g. a factor BI=2 + 0.5 higher. To keep the
system SI low will then require an specific fissile inven-
tory SI in the burner core considerably lower to accomodate

the factor (ratio of inventory in both subsystem) B_-higher

I
lnventcry. Other problems include reduced doubling time,

increased capital costs, and decreased Be the effective

rf?
fraction of delayed neutrons in the core.

Proceedling with this concept, the molten salt burner fuel is
assumed to be a mixture of PuCl3 and NaCl., Critically will de-
pend upon the Pu concentration and the thickness of the fuel
region. Figure 8 shows the effect of PuCl3 molar fraction upon

specifiic power, core exit temperature, and Cs-137 transmuta-

tion rate.

5.4 Outer-Reflector Zone Consideration

The large size of the thermal flux trap results in the fuel re-
zion approaching slab geometry with attendant high neutron

leakage., To better econcomize on neutrons several possibilities

arise

(1) use of an optimum reflector such as Fe, Ni, Cu or Be to
minimize the critical mass

(2) use of the outer neutron leakage for breeding

(3) use of the outer neutron leakage for additional FP trans-

mutation.

To begin with a solid Fe reflector was assumed (Fig. 9).
800 =

Fig., &8

PuCl,
5

22

Impact of plutonium concentration

in the fuel

Ligquid phase

Eutecti

1alCl

100 =

—

50 =

Activity
of Cs=-137
(relativT
unitse)

60 7]

ho =

0

Specific
power
(relativ)

SpeciZlc power

Caesium activity

Plutonium molar ratio
20

im

Activity
of Cs-137

(%)

1

U0

L
—

80

%

Fig, 9 Impact of reflector

Reference
Reactor

Uranium
reflector

120
Volume
of fuel

(/é)

100

o)
=

Beryllium 5 cm
as additional
reflector

24

', Thermonydraulic Considerations

We now examine the thermohydraulic implications in more detail.

For a typlcal burner reactor consider (see Fig, 10).
total power, P
power density in core, Pc = 11 kW cm_>
Pu density, pPu = ,002 atomns cm_lb_l 0.9 g em””
fuel density, pf = 2.35 g em”
volumetric specific heat, Cp = 1.95 Joules cm—BK
temperature rise across core, At = 250 + 500 K

length of channel 1in the core, Lch =

80 + 120 cm

The cruclal parameter here 1s core power density. The given value

is nigh but still near the present state-of-the-art (Table 4).

Table 4

Feinberg, research reactor

Melekeg CHM-2
FPETH

Lane (cnhlorides)
AFIR  mean

max
Phicenix 250

Chilorotrans (here)

kW/cm5

in the core

Coclant only

in the core

25

Fig, 10 Cooling of the burner reactor
Total power of reactor 11 GW

Heat excnanger 2,75 GW 4 units
& s

S S S S S S volume of fuel

in tubes 0.825 m

5

24/' Z,
1 1] .
1N g :b
< 17
q % #
L/
% 1
4
Partial g ; ; Specific
power % power
5 %% ] |1 Kwem™?
o 5l 9 ; Total volume
rar % =5
e ; ; # 2.75 m
VO lur % 1 ¢
‘ %
0. ; [
% L/
% “ ;)E
S % L
S LA
TR /

1.08 m
26

Using the above one gets

power rating, .07 g Pu / MWth

(2.2 + l.l)'lO7 emos L

H

volumetric flow rate of fuel

residence time in core 0.045 = 0.091 s

. . -1
velocity of the fuel in the core = 26,7 + 3.8 m s

The deduced fuel velocity is similar to that postulated by Lane
(1969) for the HFIR reactor: 21 m s_l. (The HFIR is also a high

flux irradiation facility).

The specific power in the coolant is about % times higher for
the burner than for the indicated breeder concept. Furthermore,
1t must be increased by BI to account for the increased fuel
inventory in the burner reactor cycle. The crucial problem will
be the efficiency of the external heat exchanger. In the follo-
wing we use some typical heat exchanger characteristics to

examine the possibilities.,

Specific power for heat exchanger ~1 kW/cm

(rather conservative data)

Total volume of heat exchangers for

11 GW(th) ~11 n°

Ratlo of fuel, volumetric 0,3

Fuel volume 1n heat exchanger 3,3 m5

Fuel 1n the pipes core-heat exchanger 1,0 m5

Total fuel out of core b,3 m>

Fuel in core 1,0 m3

l'otal fuel 1in whole system 5,5 m5

flean specific power of the fuel in

whole system E£%—%¥— = 2,07 KW/cm3
’ 3

Plutonium amount, in the fuel 0,8 gPu/cm
27

Plutonium inventory in whole system hoao kg
Power rating in whole system 0,385 kgPu/MWth

The postulated power rating for

whole system 1 kgPu/MWth

In this calculated case the power

rating in the breeder power reactors 1,15 kgPu/MWth

The above indicates tnat the achievement of a total specific

power rating of about 1 kgPu/MWth may be feasible.
28

5. Further Parameters of the Burner Reactors

Parametric studies were made as variations about a reference

system which assumed P = 11 GWatts, (X=2.9, K=4.2) and RFT = 78.5 cm,
The flux trap is surrounded by 5 cm EeQ converter, a critical fuel
thickness of 6.6 cm, and an outer wall, Figure 11 shows the calcu-
lated flux distributions for such a burner reactor, Note that the
t.otal flux in the fuel is similar to that ir the flux trap. The
calculated fluxes lead to the conclusion that o (total spectrum,

flux trap) = 1/2 +* o(E=0.0253).

Figure 12 shows the dependence of Cs-137 transmutation rate upon
the burner power level, The higher the power, the greater the
rate. However, power must, of course, be subject to thermohy-

draulic restraints, such as pump and heat exchanger capabillities.

Figure 13 indicates the relative effect of X upon the ratio R
of FP transmutation rate to FP production rate for the reactor
system, One observes the need to keep X low here. Absolute results
will depend upon the Cs and Sr densities 1n the flux trap. An
actual case of R = 1 for both FP nuclides was achieved at X = 4,5,

The FP atom ratio there was (Cs-137)/(Sr-90) = 7.25,

Another important problem is the relative high flux 1n the outer

zone, the leakage from the core.

This flux can be used for two purposes:

1) for transmutation of other fission products which have
rather high absorption cross sectlon e.g.
Tc-99 oth = 22 barn; tg 1/2

1-129 Oth = 28 barn; tp 1/2

2,1°10° a

1l

1,7-107 a
29

Radius, cm

Fig, 11 Neutron flux in burner
il ux 1. Power 11 GWth
"%, _, 2. Power 7.5 Gwth \
174em “s §
Tnerral Ll § ‘
— — S ey — N N
- %
= §/
Thermal flux \\\
NN
16 ;
\1
=
4/ \
15_] V/////
e "
—
4 /’a"
Fast neutpd
R rlux ~ id
7/ (6. group)
14 7 Reflector N
N :
-// ii Wall 1
Efl%‘/ A
Beryll u(
i Oxiae s;/
L e \/
| \//
Target zone
_ \/1/
§ */
4 s/
/u/
M
81 86 103 for 11 GWth
73 76 92 for 7.5 GWth
30

Fig., 12 Impact of the power of burner for

breeder/burner ratio ~3
Caesium=-137
activity in
relativ units
124 -

120 -

~3
CO el
Ne)
}_l
@)

11 12 13

Power, GWth

31

Fig, 15 Caesium-137 and Strontium-90 activity

in the target zone

Activity
- Caesium
o
110
Jesired level
709 Strowgium=90
-
T
! | ' 1 | T 1
1 2 3 b 5 6

atlo of breeder/burner power
3

In both cases a flux of 1014-1015

n cm_2s L (see fig.ll) can

result 1n a rather effective transmutation rate of

Tc=-99 o = 7-10_95—1 t 1/2

1%
i

a

I-129 o = 7-10'95'l £ 1/2

12
Mo
o

For improvement of this possibility use of a beryllium moderator

in form of 5 cm wall in the outer region of core have been calcu-
lated. This gave an improvement of transformation rate of Cs-137

in the innter target region but alsc a very significant increase

of specific power, due to the softening of the neutron flux in

the fuel region has been obtalined (Fig. 11).

The possipility of transformation of these two longliving fission

products will be further discussed elsewhere,

The neutron flux outside of the core can be used for breeding in
an uranium blanket., The breeding ratio can be really very good,
higher than 1, but the decrease of the transmutation ratio is
critical, and seems to be too low for a reasonable burner reac-

tor,

In spite of this, in a further study the possibilities must be
checked of using part of a neutron for the breeding process in

an external uranium blanket region,
53

6. Effect of Replacing Chlorire with Fluorine

Is a fluoride-fuelled burrner, Instead of one with chloride-fuel,

possibley (Table 5, Fig. 14, 1%)

For fuel 1n tnils reactor we nave used an arpbitrarily chosen
mixture of Pur, «%,54«ar-2,4 ZPFM. The calculaticons have been
done for a pigger burner reactor of 11 GWth and, the whole

T

system totaling -h UWon,

The neutron flux 1n the target region was in the case of fluoride
fuel appr. 1.05 times higher than in the reference case with

chloride fuel,

The elfective nall-1life of both the fission product nuclides

was (in years):

In fluoride- In chioride~fuelled
ffuelled reactor reactor (reference)
Cs=137 &.57 6.973
Sr=90 175 1.83

But all these benefits must be paid by a twice higher specific
5

power: 1n [luoride-fuelled core of 19.9 kWem 7, and 10,1 chmB in

criloride-contalining core.

A1ls0o because the possibility to use graphite, instead of part of
ceryllium oxlde as moderator, which seems to be possible from the
point of view of neutronics, a rather important improvement of the
coerroslicn problems can be achieved., We must remember that the com-
pativility of molten fluoride fuel and graphite had been proved by
the excellent experience of Oak Ridge Nat. Laboratory (with one

difference: in ORNL experiments the fuel was LiF-Bel —ThFu-UFM).

2
34

Table 5 High-flux burner reactor with fluoride fuel
Total power 11 GWth
Power of total system 55 GWth
Zone Components Neutron flux Specific power
) 24 =3 16 -2 -1 -3
Radius (cm) atom 10° cm 10 cmo s (KWem 7)
Volume (cmB) total transmutation
thermal rate (s_l)
il
0 - 98.8 Cs-137 0.0116 _ 2 n=9
Target 6 z|Sr=90  0.0016 4,01 Cs=137 1,8 10—8
Vol.~4,1+10 7 em” |0 0.0145 —— Sr=90 1,2+10
D 0.0145 2,21
1T
38.8 - 109 em Be 0,060 5,08
Moderator 0 0.060 1,83
with thin 5. 25
3 —_—
graphite layer 0,239
111
109 - 112.6 em |Pu-239 0.0017
Fuel 5 3 Pu-240 0.00042 E -3
Vol, 5.5+.107cm” |Pu-241 0,00021 TO5T 15,9 kWem
Na 0.0075 2V
4r 0.0051
B 0.0340
IR
112,6-118.6 cm |Be 0.060 4,87
Wall 0 0.060 0,034
5,9
0,001
Vv
118.6-218.0 cm |Fe 0.08 0,0023
RFeflector 2,8-10“9

55

Fig. 14 Burner reactor with molten fluoride fuel
Fuel: PuFB—B.SM NaF-2.41 ZrFu
Specific power ~20 chm_3
Total power 11 GW
Breeder/burner ratio: 4

Reryllium Reflector (Fe)

1de

| | LI
96,8 109 112,6

Radlus, cm
36

Fig. 15 Burner reactor with molten fluyoride fuel

Transmutation Specific
rate of nucleil power
- — 20
: kWem
110 = =15
100 _] Desired level =10
80 = 0

without with
beryllium EeQ :one
in outer in outer
core core

region region
7

57

Remarks about transmutation and hazard coefficients

It seems to be worthwnile

to make some estimation of the

advantage given by this type of fission-product management.

To operate for sale with only one figure for both nuclides:

Sr-90 and Cs-137 the following mean values have Dbeen arbitrary

postulated here:

Ratio of
Sr=90 Ca=137 hazard
Sr=90
Cs=137
Maximum permissible (uCi/CmB)
water 1-107° he1o™" 40
air 1-1077 61070 60
mean value 50
vield from fission 7
U-233% 6,2 6,6
U-23%5 5,1 5,99
Pu-23%9 2,18 6,69
mean value for power
production
1:1:1 for U-23%3:U=-235:Pu-239 4,1% 6,4%
relative hazard coefficlent 50 1
nazard coeff, X yield 2,050 0,064
total hazard coefficlent 2,114
in transmutation the
effective half-1life 2 years 6 years
hazard coeff, in transmutation 0,144 0,013

38

i 16 Inventory of Strontium-90 in the power
reactors and in the burner reactor
Power of total system 55 GWth
Yield of Sr-90: 4,1 %
Number |
of Sr-40
atoms
Inventory in the steady state
o due to beta-decay only
1043 -
(eg, In salt Ll s -
. 28
’////' 8.96+10 atoms
4
28
10 - nventory power reactors, mean
" 6.6-1027 atoms
Inventory in burner reactor
6.&6'1027 atoms
o
10‘7 -
ol
102¢
Charging,
discharging
Production rate of in power
Sr-=90 reactors
19
. ~7+1077 atoms/sec
1092 =
v | | | |
0,01 0,1 1 10 100 1000

Time, years
59

Thus 1n the steady state of transmutation the amount of hazardous
substances 1s reduced by factor ~15 in relation to the steady-

state of beta-decay.

The amount of strontium, the most hazardous nuclide, is in a
high-flux reactor burner approximately the same as in the power

reactors after 3 years of fuel burning (fig. 16),

But the most impressive result comes from the considerations
about the "end of the fission power era". The storage (without
transmutation) Caesium-137 and Strontium-90 decayed within
half-1ife of 39 years, core after ~300 years the amount of

these nuclide will be reduced by factor 1000 (fig. 17).

In a high-flux transmutation the reduction with factor 1000 could
be achlieved after approximately 26 years, (fig. 17) i.e. in the
lifetime of one reactor generation (for more exact consideration

see (Taube, Ligou, Bucher, 1975).
Fig, 17

Hazard of the both fission products:

4o

Sr-90 and Cs-137

1
Spa
~‘~_ be
-~ -
~§~-~
Reactor S -
hazard T
1071
Relativ
i Total hazard
hazard
during
\ transmutation
LY
Ts-1
Hurin
_ L ransmudatio
10 ° .
7] 1
100
\ time, years
-
\ N
107° = \
Sr=9
durin *
T transm‘fation
-y \
10 | } | I
0 10 20 30 ho 50 60

Time, years
41

8., Conclusions

These consliderations show that a high flux burner reactor for the
transmutation of selected most hazardous fission products as Sr-90

and Cs-137 may be achieved for reasonable assumptions:

1) the whole system 1s a steady state fransmutation system

2) 1in the steady=-state the amount of transmutated nuclides is
more than one order of magnitude lower than in a spontaneously

beta-decay steady-state.

%) the ratio of conventional breeder power reactors to the burner
is to about 4, making it possible to organize the fission

power industry.

)Y the whole system 1s still breeding, with a relatively long
doubling time of 30 years. The relative high specific power-
rating of 1 kgPu/MWth, inclusive of the amount of plutonium
out of core (cooling transport, reprocessing) is rather
opftimistic but not impossible, especially if a more sophisti-
cated system will be used e.g. GCFBR or MSFBR (Molten Salt
Fast Breeder Reactor) with a quasi-continuously pyrochemical

reprocessing plant
42

9. Acknowledgement

The authors wish to express their best thanks to Mr. Stepanek
for helping in some data processing, to Mr, K.H. Bucher for
discussing some thermo-hydraulic problems and Mr., S. Padiyath for

realizing the computer calculations,
b3

10. References

- International Atomic Energy Agency

Safe Handling of Radionuclides,
Vienna, 1973

Lane J.,A, Test-reactor perspectives

React. Fuel. Proces. Tech. 12, 1, 1 (1969)

Lidsky L.M, Fission~Fusion Systems:
Hybrid, Symbiotic and Augean
Nuclear Fusion, 15, 151 (1975)

Schneider K.J,, Advanced Waste Management Studies, High Level
Radiocactive Waste Disposal alternatives

Platt A.M,
USAEC, BNWL-1900, Richland 1974
Taube M., The Transmutation of Fission Products
Ligou J. , (Cs-1%7, Sr-90) in a Liquid Fuelled Fast
Bucher K.H. Fission Reactor with Thermal Column
EIR-Report 270, Wirenlingen 1975
Taube M. Steady~-State Burning of Fission Products in a

Fast/Thermal Molten Salt Breeding Power Reactor
Ann. Nucl. Sci. Engin. 7, 283 (1974)

Wolkenhauer W.C. The Controlled Thermonuclear Reactor as a
Fission Product Burner

BNWL-4232  (1973)