EIR-276.txt

EIR-Bericht Nr. 276

EIR-Bericht Nr. 276

Eidg. Institut fur Reaktorforschung Wiirenlingen
Schweiz

Breeding in Molten Salt Reactors

Lectures at the University of Liege/Belgium
15th May, 1975

M. Taube

| I

,_l

Wirenlingen, April 1975

EIR-Bericht Nr., 276

BREEDING IN MOLTEN SALT REACTORS

Lectures at the University of Liége / Belgium

15th May, 1975

M. Taube

April 1975
Remerciements

Je remercie M., le Prof., G. DUYCKAERTS de 1l'Université de
Liége, de m'avoir permis de faire cette conférence dans

le cadre de la licence speciale en Science Nucléaire.

M. Taube
WHAT IS BREEDING?

WHY IS BREEDING POSSIBLE?

WHAT IS BREEDING GOOD FOR?

Breeding is a process in which two mechanisms are occurring simultane-
ously. 1) Fissile nuclides are 'burning' and producing energy and
neutrons. 2) Some of the neutrons are transforming the so called
fertile nuclides into fissile nuclides at a rate greater than the

rate at which the fissile nuclides are being consumed.

Breeding is possible because of three factors: 1. Some fissile nuc-
lides have a rather large nett production of neutrons so that more
than half of them can be used for breeding. 2. The fertile
nuclides are converted to fissile nuclides only by the simple, and
energetically not expensive act of neutron capture followed by two
spontaneous processes which occur at a much faster rate than the rate
of neutron capture. 3. Fertile materials are present in the earths
crust at relatively high concentrations. (This also means that these
fertile nuclides must be rather stable, beta stable of course, like
all heavy nuclides they are alpha unstable. The a-decay rate must be
matched to the age of the solar system).

Breeding makes it possible to use not only the uranium-235 (the only
‘naturally occurring fissile nuclide) but also the uranium-238 which
occurs in amounts 140 times greater than U-235. These factors permit
the use of ores with low uranium and thorium concentrations even

down to granites (>50 ppM U+Th).

Practically only some reactor types are suitable for effective bree-

ding:

- all reactors with a fast neutron spectrum (e.g. sodium cooled fast

reactor, gas cooled fast reactor, molten salt fast reactor)

- one reactor type with a thermal flux spectrum (molten fluoride

thermal reactor).
FISSILE AND FERTILE NUCLIDES

Definitions

Fissionable Nuclides

Fissile nuclides

Fertile nuclides

Characteristics Neutrons of any energy Only neutrons of high energy
can produce fission (>1 MeV) can induce fission;
Neutrons with lower energy
are captured and after gamma
emission (and in some cases
then by beta-decay) a trans-
formation into a fissile
nuclide occurs.
Examples U-233% U-235 Th-23%2 U-238
Pu-239 Pu-201 U-2310 Pu-240

Binding eénergy of
neutrons and
barrier to fis-

Binding energy of the
captured neutron is
greater than the fis-

Binding energy of the cap-
tured neutron is lower than
the fission barrier

sion sion barrier (more %@ < Q
exactly: fission n a
activation energy Qg) only fast neutrons with
kinetic energy E, can cause
Qn > Qg the fission
Qn + Ec > Qa
Fertile Nuclide [Fissile Nuclide
°q re s ) TP §
\/ t 0.8 MeV MeV
k 74
29 ‘ bar =4 Upatilug Transformation
\ after neutron
capture
U-236

N

Qn = Neutron Binding Energy

> Qac

Qbar

Fission Energy Release

Activation Energy for
Fission

= Energy barrier

198 o

and Fission activation energy Q

n
-
—

Neutron binding energy Q

54

200 N/\r*
Deformation

233 234 235 236 237 238 239 240 241 242

atomic mass, A

THERE IS ONLY ONE FISSILE NUCLIDE

EXISTING IN THE EARTH'S CRUST

Atomic

Mass, A 5
Fissionability Parameter % = 36
2704
@
2604 Fm@
©
Es
non-fissionable c (] o
[
250 _ o ®
qg Bk
@
A -
Aing o
@ Pu-241
240 @FPu-239
230 -
:
Rn fissionable
220
At
104 51—209%’
le-208 0=-209
200
v T 4 T T T T T T | - | VLT v T v 1
g2 84 86 88 90 92 24 96 98 100
Atomic number, Z
Half-1life
years
Es
Md
T R N Feo B M e (RN P S S WA S SRR RS (LS
82 34 86 88 90 92 94 96 98 100

Atomic number, Z

for the following reasons

1. The fissile nuclides must have
2 2
%

Z
I >%6 (=

I fissionability para-

meter)

2. The fissile nuclides must be beta
stable, because the half 1life of
beta-decay in this region of A
1s smaller th%n 10 years.

3. Because for Z >36 gives in the

realistic case Zz 92; all
fissile nuclides are alpha un-
stable.

i, To exist in the earth's crust the
half 1life of alpha decay of the
nuclide must be of the same
order of magnitude as the age
of the solar system 2109 years.

5. Only one nuclide fulfills all
these criteria: - U=-235

Relative 10

abundance t 1/2 = 10x10" "y: Th-232
1 S

A

t 1/2 = 0.7x109y: U-235

Supernova

t 1/2 = 8.3x107y

Pu-244

10

-2k

|
4 3 2 1 Present

time, Giga years

J
=
1
l
ONLY TWO NUCLIDES EXISTING IN THE EARTH'S CRUST CAN PLAY THE ROLE OF

FERTILE NUCLIDES

The fertile nuclides must fulfill the following obvious criteria

1. They must be abundant on the earth. At least a mean value equi-

valent to 1 ppM throughout the earth's crust (or more exactly in

the outer layer of the earth's crust.

2. Thus they must be beta-stable

3, Thus the Z value must lie between 90 and 94 and be even

4, These are the isotopes of thorium and uranium with t 1/2 >lO9 years

The only two 'natural' isotopes are

Th-232 and U-238

These nuclides can be transformed into fissile nuclides by simple

neutron capture and spontaneous beta-decay.

Thorium Uranium Cycle
/\ Stable in Nature

(D veta stable
(Q bpeta unstable

—» beta decay

’f fission

*neutron capture

e X
od = 2xi;;jL/
o = 10710
e ?
thermal . .
g is relativ big _
(n,y) o0 = l1o 10
Ox =5x10" %5~ Oz .9x107 7571 ®
2334
22 min 2.7 days
‘ i o = 2x107°
232 ] A
Th Pa U
1 1 )
90 91 92

A

2‘%1T

240 -

239 7

238 7

Uranium-Plutonium Cycle

o = cross section for neutron reaction (cm2)
@ = neutron flux (neutron cm—2s_1)
A = decay constant (s-l)

all data for ¢ E 1015n cm-zs_l L 7x10

fast O
/c:m: 10710
e {

ofast fls small o =410

TO PERMIT BREEDING THE NUMBER OF NEUTRONS EMITTED PER FISSION
SHOULD BE LARGE

The scheme of neutron induced fission of Pu-239

240 fi i 1 1
giPu+n > 9uPu 1581ons}22184 + 2?Nb59 t v ong (prompt neutron)
delayed ’/// - V8~ (beta-minus decay)
neutron lOOM
emission 42 O58
Y stable
136I
53783
lB%XB- (beta-minus decay)
54482
stable

The value of prompt neutrons fission = v

. The value of v for uranium isotopes 2.5

The value of v for plutonium isotopes 2.9

. The value of v increases when the energy of the captured neutron
is ~1 MeV

W

Table All data approcimate

v Value of Neutrons per Fissioned Nucleus: v

Nuclide | Thermal o .o .oactor

reactor
Y n n
U-235 2.431 2,09 2.57 2.50
U-233% 2.48| 2.25| 2.51 2.51
Pu-239 2.87| 2.08| 2.88 2.40
U-238 - - 2.66 very
small
Th=23%2 - - 2.36 very
small

Pu-240 2.87
Pu-241 2.97
Pu-242 2.18

3.00 8
L1801 3.1 2.06
Y 1.26 habig

oo

Pu-240 [lu~238

T L] 2 L 1
10 1 10 10 10 10

Neutron Energy, eV
ONLY SOME OF THE CAPTURED NEUTRONS CAUSE FISSION THE OTHERS BRING
ABOUT A GAMMA EMISSION. THIS REDUCES THE BREEDING POTENTIAL

fission fragment

83 fissions—=fission fragment

~\\*239 neutrons (83%x2.88=239)

100 8% atoms
100 atoms prompt neutrons (from
neutrons —= of 17 atoms Pu-239=2,88 per fissioned
fast Pu-239 of Pu-240 atom)
gamma

The ratio of non-fissioned atoms to fissioned atoms 1s called «o

a = %%%f%% (in this example o = %% = 0.,20)

We define: OYabsorption = o(n, ) + o(n,f)
The real number of emitted neutrons per neutron absorbed equals:

in this example -
n = 2.88 ('§217) - 2.40
. om,f) .. L
B O(H:f) + O(H,Y) l+a - 1 -
L? = 2.88 T3 0.2 ° 2.u?d

The value of n is strongly energy dependent
The value of n differs for different fissile nuclides

Value of n for fissile nuclides
: y (simplified)
Cross section Cross Section of Pu-239
BADHE (10—24cm2 (simplified) n
1000 1

~

100 o

e ___.___..__-———-_,/—' ——/——n minimum
o~ '-__________,—’

10

ocapture‘\
o(n,y)

Thermal Fast
Reactors Reactors

T T ] 1

0.01 T T T T ! 2 2 4 6 8
1072 1 102 10" 108 108 10 1 10 10 10 10

Neutron energy, eV
Neutron energy, eV
THE NEUTRON BALANCE FOR FISSION AND BREEDING IS AS FOLLOWS:

Molten Salt (thermal) fluoride

breeder for

~100 neutrons

Molten salt (fast) chloride
breeder for ~100 neutrons

Nuclide Absorption Fission
232y, by, 7 0.03
233p, 0.02
233y 41,4 92.3
23L‘U Y 0,01
235y 3.4 6.8
236y 0.3
237Np 0.3

6L1 0.1

14 0.7

IBe 0.3 0.9

g 0.8

Graphite 2.3
Fission 0.7
products :
Leakage 1.0
(nE 2.2317)
Breeding ratio (total) 1.0708

Nuclide Absorption Fission
2350 (n,v) 22.51
(n,f) 2.99 8:.2%
239py (n,y) 5.58
(n,f) 28.98 85.55
2LmPu (n,vy) 2.214 v
(n,f) 1.54 4.72 | §
Na 0.26 @
Cl nat 3,16
Fe 1.3%0
Mo 2.04
Fission
products A
23%(n,y) 23.15 N
(n,f) UeB5 1.50 O
Na 0.08 =
Cl nat 2.22 3
Leakage 2.90 £
Breeding ratio core 0.716

blanket 0.670
total 1.386

Balance for Thermal Fission U-233

100

neutrons for
fission

100

neutrons
for fission

Fission of 91
atoms of U-233

225 neutrons

6 atoms of U=~233

breeding gain

Y £

Balance for Fast Fission of Pu-239

(Fission Products not shown)

100 neutrons

u

O

Fission of 83
atoms
v = 2,98

239 neutrons

100
neutrons
for fission

Pu-240
7 atoms

DEFINITION OF BREEDING RATIO AND BREEDING GAIN

BR = average rate of production of fissile nuclides
average rate of loss of fissile nuclides
Rmaximum = Breeding potential = n-1
G = Breeding gain = BR-1 = n-2; for breeder G>O
1) for typical fast reactors; value of breeding gain, G
oxide fuel 0.25 (at present for 300 MWe LMFBR;
G = 0,12 £ 0.03)
nitride fuel 0.3%0
carbide fuel 0.4 + 0.47 (gas cooled FBR)
molten chloride U955
2) for thermal breeder
molten fluoride 0.06
all other fuels
solid and ligquid <0.00
Typical value
BR = [Mechanism for Pu fuelled
fast reactor
(+v |No. of neu-
trons per (+2.96
fission
-1 one neutron
for fission -1
claim £
-0 loss due to .
absorbtion 104
in Pu-239 0.24 Metallic fuel
-A |losses in 131 Rt
Structural - Molten chlorid
material, etc. (0.1:0.25) e
-L |leakage of 13] Fruoe
neutrons -(0.04+0.06) ride
-T losses due to
absorbtion in |-(0.01+0.015)
P
+F ., |rate of fission
of U-238 +(0.19+0.22).
.(v=1)neutrons from
U-238 fission |.(2.70-1))
( 1 rate of ab- ( 1 ) 0.8
l+a/ |sorbtion in Pu-239 U-233 U-235
Pu-239 in (1 + 0.24)
(n,y) reaction
BRfast= 1.25 + 1.40

10

FOR THE CHEMIST, THE POSSIBILITY OF INTERNAL OR EXTERNAL

BREEDING IS OF IMPORTANCE

Primary Step .. Secondary Step
Cycle Fertile Intermediate | Fissile Intermediate Fissile
Thorium Th~232 Pa-23%3% U=-233 U=-234 U=-235
90 91 92 92 92
Uranium U-238 "Np=-238 Pu-239 Pu-240 Pu-241
g2 93 94 QY olu
Fissile and Fertile Materials
‘Micro-mixing (U,Pu)O2
Macro-mixing UO2 particles + PuO2 particles
In fused salt: PuCl3 in fuel, UCl3 in coolant/fertile material
External
Small critical mass, Using the fertile medium
spectrum hard, breeding as coolant in a 'Chlorophil'
ratio high cooling prob- type of reactor.
Fissile TFertile lems, low Doppler effect,

loss of reactivity due
to burn-up

<100 MWD/kg possible
higher enrichment

Internal
No blanket

spectrum softer, 1
BR, high cross se

299,99,
tion of structural
materials, good \’”"”
Doppler effect, \\’”
longer burn-up to \.A’,‘
100 MWD/kg lower
enrichment

Note: for ~3000 MW(th) radius of core is ~100 cm

thickness of blanket is ~100 cm

Liquid fertile
material as
coolant

11

DEFINITION OF DOUBLING TIME

Doubling time To is the period of time (years) in which a breeder
produces enough fresh fissile material to fuel a new breeder react
with the same power level (Teff includes the inventory of fissile
materials out of core e.g. cooling, being transported and reprocessed).
The compound doubling time Tigip'takes into account that in a breeder
system, the new breeder can be fuelled with fissile material coming

from the total system and not only from one reactor.

_ Specific inventory (gram fissile/MWth)
To - (L + o) *° G (days)

G = BR—l; breeding gain

r = fuel inventory out of core

L = load factor (hours per 8760 hours in year)
I

= fraction of fission in fertile nuclide

Teff - To (1+F)
comp L(l-r)

1n?

Doubling

Time T
(years)

- 12
Doubling-Time and T 40w
Breeding Ratio 30

Sodium cooled, burnup 10 MWD/kg fuel

1 worid Energy Production

Oxide fuel, specific power
0.7 MW(th)/kg

20

/ Nuclear Pdwer

15 = L

10

G e ] s e e e ] e

|
I
I
I
) I
’2 years _J
|
J

1 20 yea \\\:::;:
; 8 ‘yea 4
BR
0 T i il 7 \ T

i 1 1 1 L] 1 Ll
1 1.1 1.2 1.3 1.4 1.5 1970 1980 1990 2000 2010 2020 2030

12

DOUBLING TIME IS COUPLED WITH FUTURE ENERGY DEVELOPMENT

The rate of doubling (doubling time) of the total energy consump-

tion is ~18 years and may increase in the future,

Doubling time of electrical energy consumption is ~9 years.
Doubling time of nuclear power capacity is less than 9 years.
Doubling time of breeders should be of the same magnitude.

In the future, civilization should reach a steady state when the
doubling time for the breeder will satisfy the demand with

Ts> ~30 years.

Combined Breeder/Reprocessing Plant

Uranium+Plutonium

Uranium=-238 Fuel
.3 kg/da Preparation
? e Y 1 kg/day
Fresh Fuel
2.3 kg/day
(= \
0.33 GW(el) =
Breeder Reactor Reprocessing
1 GW(th) Plant
Energy
0.33 GW(el) Pu for sale
—_— 0.3 kg/day
0.67 GW(th) E==——"" P
- e et
Irradiated
Fuel L
2.3 kg/day ission Products

kg/day
BREEDERS ARE NECCESSARY BECAUSE THE WORLD RESOURCES OF 'GOOD'
URANIUM ORES ARE RATHER LIMITED

if one con-
but the very

A tremendous increase in available uranium ores occurs
siders not only the classical ores (>1000 ppM uranium)
abundant granites with 80 ppM uranium and thorium.

Even if the price of uranium should increase a hundred fold the price

of the raw fuel per kWhr will be no higher than it is now for light

water reactors (since the uranium contains only ~0.4% of burnable

U-235).

1 kWhr (e) £ 3 cents US, of this fuel % 0.8, from this plutonium

0.5 ecents.
1l g Pu 2 1 MWDth = 24000 kWhr(th) = 8000 kWhr(e) % 40$/g Pu
for 1 g Pu £ 1.5 g natU

e

(114

present price 1 kg U = 30 $; 1 g 3¢
assume future extreme price 1 kg U = 2000 $ lg

2 3

1l g Pu will require 3 $ worth of Unat
Each gram of Pu will be not 40 $ but 42 $ per g.

The price of electrical energy increases to 3.25 ¢/kWh(e)

World Resources of Nuclear Fuel

UppM
2000 | 5 iD
16 ppM in Ocean natBLi ~30 ppM
16 in crust
1000 10 2.2xlO2 kg i 6 -
ut Li is only
6x10 MJ or 2.4 ppM
/ 6x1016 kg 6Li
00 4 i
> / Granitp 20 nat g U 10.0x102% Mg
e 4 ppM in
& crust
2 16 5%
$/kg 8x10™" kg
: 5%
7x102u MJ ’ug%

—_—

Us’g 200 o | 50

Consumption
23Th 12 ppM in crust

s NSZi_gan.

100 = o — o —  Granitp 100
A (y 8x10”° popu-U + Th 16
= lation at 21 x 10 kg
15 kW per capita 24
50 o & Cumulative { L 500 17 x 107" MJ
=Y World needs
to 2000 |
| without breeder |
I | | Complex Ores
20 o | | - 500
l : l D, Li = 2.75 MeV/atom = 2.75 x 10%" J/kg
‘y & High Grade Ores 1000 Th, U = 0.83 MeV/atom = 8.65 x 1013 J/kg
‘16 Earth's crust down to ~20 km % 2,4 x 19%° kg

10 T = = T-
100 109 12080 1011 1ol2 1913 1ot 1015 1o g
Ocean = 0.14x 10 kg
14

FUTURE ENERGY NEEDS COULD BE VERY GREAT BUT WOULD STILL BE MET FROM
NUCLEAR SOURCES IN THE FIRST INSTANCE
TODAY:  3.8x109 people x 2 kW/cap. = 7.6 TW = 2.4x10°°J/year

FUTURE: 8x109 people x 15 kW/cap. = 120 TW_ = 38x1020J/year
1 gram (U,Th) = 1 MW/day = 8.64x1010J

38x1020J/year 4,4x1010g/year = 44000 tons Uranium and/or
Thorium per year

But granite contains 80 ppM U+Th so the annual need for granites
as '"fissile' ores will be 550 million tons.
At the present time coal production alone is 4 times greater!

Granite: Energy cost for 15 ppM U and 60 ppM Th.
Granite 1s the main constituent of the earth's crust (up to
20 km deep)

lO6 g
68.6 g/mol

Free egthalpy of formation of granite = 210 kcal/mol '
8.8x10°J/mol
?eghfiglogical free energy (electrolysis?) = 1.8 MJelec/mol +

’ therm/mol 10
Technological free energy for 1 ton granite = 5.2x107J.
Amount of wuranium and thorium = 75 ppM = 75 g =

T5MWD total =
A 6.48x1012%

Amount of electricity at 40% efficiency = 2.6x1014g

= 1.46 x 104 mol

1000 kg granite =

World Population
X 109

1205000 1000 %/oo Solar Energy
9 N on Earth's Surface

8 s f$>

120 4 1 %/00 {promille)

prodfction

present 160 y 200 y
14 T

Power per capita
kW
USA 50

>
si°&

> 4 10 J

I
T I 1] fi
present 100 y 200 years present 90 100 208 years
15

ONE OF THE BIGGEST CONSTRAINTS TO FISSION ENERGY IS THE PROBLEM
OF FISSION PRODUCT MANAGEMENT

Each fission releases ~200 MeV; 1 Joule = B.lxlolo fissions.

1 watt = B.lxlO:LO fissions per second equivalent to 6.2X1010 fission

product atoms per second.

In steady state very roughly 1 Watt of power ~1 Curie of fission pro-

ducts (1 Curie = 3.7x1010 disintegrations per sec).

All fission products are beta unstable (neutron rich nucleil)

Some fission products are long lived, comparable with a human life

span or even the life span of an element of social organisation.

@ beta stable Watt
Obeta unstable T

J \beta decay -~
83 -

Sm-l.Sl

| \‘ Cs=-135
-8

1 daly lqu 109'd kyez'n' 10 yr 100 y : 1000 y
57 10" 0% 10° 107 10% 10° 1020 1ot
a

time, sec
16

THE PROBLEM OF FISSION PRODUCT MANAGEMENT TS VERY DIFFICULT AND
REQUIRES A SOPHISTICATED SOLUTION

The possibilities of radioactive waste management are:

1. Without the use of nuclear transmutation; that is the fission
product nuclides are not changed, but merely removed and isola-
ted; as for example a rocket to the sun, in discused salt mines

or in the polar ice-cap.

2. By using nuclear transmutation in which the nuclear properties are
so fundamentally changed that the transmuted products are short
living nuclides which decay after a short retention time to a
stable nuclide.

In this second class of waste management techniques a number of exotic

methods have been discussed

- gamma laser excitation (does not exist)
- underground neutron irradiation due to fission or thermonuclear

explosion (e.g. more than 3000 explosions of 100 kT per year in
USA 2000 only)

- bombarding by protons (e.g. a 10 GeV accelerator with 1 Amp)

- neutron irradiation in a thermo-nuclear fission reactor (does not
exist)

- neutron irradiation in a fission reactor (only pessimistic opinions).
1%

Radioactive
Waste
/’ \
high active low active
waste ‘waste
any management management dissipation

(forbidden!!

no transmutation transmutation
direct Antrolled without with with
controlled storage changing changing changing
storage on Z and A A Z
the surface (b7<2)
coulomb gamma (n,y) (n,2n) (p,y) (p,spallation)
exitation laser
Earth cosmic
space
(rockets)
)
litosphere hydrosphere solar deep continously periodical
(underground) impact cosmic
///////\\ escape
salt rocks water ice neutrons neutrons fission thermonuclear
mines (deep sheet (primary) (secondary) underground underground
see (polar) explosion explosion
floor
stable geologic fusion fission accelerator
conditions nonstable reactor reactor
conditions (CTR)
fast thermal thermal fast

neutron neutron

with without
thermal thermal
column column
THERE IS A METHOD FOR

PRODUCTS

18

'INCINERATION'

OF THE MOST DANGEROUS FISSION

Shown here are the most dangerous nuclides: per 100

Caesium

Strontium

Iodine

Technetium

Krypton

(stable)*
(2x106 years)
(30 years)
(stable)*
(28 years)

(stable)
(1.7x107 y)

(2.1x10° y)

(stable)
(stable)
(10.7 years)
(stable)

133
135
137

88
90
127
129
99

83
84
85
86

6.91
7.54
6.69

1.44
2.18

0.38
Lsly

5.81

0.35
0.56
0.672
0.882

21.14

302

1.55

| R S S

2. 47

fissioned atoms.

\ 33,18

*problem of iso-
topic separation

The incineration is possible due to irradiation in a very high neutron

flux, which gives, for example

Cs=-137

53 -

(:)Sr-9l

gcapfure
very| small

(n,y)

®Sr-90

9.7 hr

52 =

o

50 =
() beta unstable
® the most

dangerous
stable nuclide

(n,y) Cs=-13

'Decay' and Transmutation of
Strontium=~90

64 hr

. Zr-90

} (n,y) reaction:

4 B-decay

8 e 9in

Ba-138 (stab

"Decay' and Transmutat

(n;p)

le)

ion of Caesium=-137

(:) Cs-138
acapt 32 min
very| gmall
82 ‘ Cs-137 Ba~138
A
0y
(n,2n) Cn )
\J
81 4 (:) Cs-136 Ba-137
A
N
2.9 d
80 - . @s-us Ba-136
Qbeta unstable , 106
thhe most dangerous N y
@stable nuclide
75 Q -5
T T T
54 55 56 Z
Xe Cs Ba
19

THE NEUTRON TRANSMUTATION OF F.P., NEEDS A VERY HIGH THERMAL NEUTRON

The efficiency of the transmutation is limited by the criteria that the
probability of the bombardement of an atom by a neutron is ~10 times
greater than the probability of spontaneous beta decay.

A(neutron capture) 2 10-A (beta decay)

We know that for a neutron flux @ (n cm 2 ) and the capture cross
section o, (cm?/particle) the probability of reaction equals.

A reaction = @+0; therefore @#:-0 >Adecay

For caesium-137 we have

Adecay = T.7 X 1 P i
ccapture (thermal) = 0.06 x 10_2Ll cm2
occapture (fast) * 0.01 x 10°2% cn?

For the postulated neutron reaction probability
Rt 0 - =15
?> l%%%él > @fast; @fast L. 7x10 — = 8X1016; Pthermal Y. A50 i
P 0.01x10 10.6x10
= 1,2x1016
AL Possibilities of transmutation
Thermal
Particles
Fast Source of
Tower _@Q Particles Protons neutrons
rating (W/cm ) 200 Accelerator 30 MeV -
Fissile _2h_ ~30 MeV protons
concentration (N) 10% for
30x10 (secondary)
(10—24 700 Accelerator 1 GeV High flux
9fission =2 - >1 GeV protons for of neutrons
cm ) 2
Flux 0 _ §;2§l915 (p,spall) _
nem2g + , B = 2.6x10 primary
Nuclear - very high
explosion flux in very
short time
C.T.R. = High flux,
(controlled continously
Fluence, 0.t thermonuclear
t = (days) 700 reactor
-5 £ 50 Fission = Very high
.t = (ncm ) é-ZEipgz reactor flux
2. 2%l continuously

20

IT MAY BE POSSIBLE TO COUPLE A BREEDING SYSTEM WITH A HIGH FLUX
TRANSMUTATION FISSION REACTOR

We postulate a system with breeding potential and with transmutation
potential.

The compound doubling time Ts must be at least 30 years,

. _ 1000 . M. (1+F) i B}
sincey; Ts(year) = Z65  (BR-1) (1+a) L 1n2 = 30 years

M, (1+F)
(30.(1+a) - L

from this;minBR =1+ 2.75 « 1n2

for typical values;

minBR = 1.077
we know that: maxBR - vol-a- (A+L+T) +F(v=-1): when T 1s very small
(1+a) (T=transmutation rate)
For the desired BR . = 1.077 we obtain:
min
T = wv-1-a-BR(l+a)=-A-L+F(v-1)

For typical values: T = 0.364

That is, for a breeding system with a breeding potential of BR = 1.4 and
for a compound doubling time of 30 years, approx. 0.36 of FP nuclides
could be transformed in the fast system.

Organisation of transmutation and
breeder system

Breeding and Transmutation

(history of only F.P. and
neutrons)

Our system is:
- breeding system with compound
doubling time 30 years.

for/fission ~ breeding potential BRpot

Losses ete. = transformation rate 0.36
~ 1 burner reactor for transmutation
- breeder power reactors

= 1.4
100

Fissions

00 , We obtain:
Producis 140 ' X ° BRmaX = (X+l ) BRmin T
. but we know BRmaX = BRmin + TEIET
B o from this, for typical values
_ l+a _
reedng with X - BRmin T - 3 . 7

dRubliyg time
framsmutation M 30 years Result: for about 3.7 power units of

breeder reactors and 1 power unit of
the breeder transmutation reactor
0.36 of FP nuclides of all reactors
can be 'incinerated'.

mutation
of ‘selected

Stal
Long

21

WHAT IS THE CONNECTION BETWEEN BREEDING AND THE FUEL CHARACTERISTICS
(PHYSICAL STATE: SOLID/LIQUID, CHEMICAL STATE: METALLIC, CERAMIC, SALT)?

Breeder reactors are feasable using both thermal and fast neutrons. But
for thermal neutrons, breeding is possible only in a very limited region
the thorium-232/uranium=-233 cycle in a liquid fuel reactor.

Breeder Reactors

Neutrons Thermal Fast
Fertile/fissile Th-233 U238/Pu239 Th232/U239 U238/Pu239
Cycle
Potential Bree-
ding gain ~0.1 <0 ~0.3 ~0.45
G=BR~-1-losses
Impact of inter- Pa-233 breeding Pa-233 Np-239
mediate mole- th_ not possib- _fast_ fast e
— cab—large 1e Oab =small Oab negligi

ble be-
cause
tl/2 small
Consequence Continuous _ no limitations, periodic
Pa extraction rechanging possible
Results liquid fuel no limitations, solid fuel

(aqueous or
molten)

possible

Maximum losses
in core struc-
tural material
and coolant

ad i P

~0.2 ~0.3

Consequences Moderator D.O _ Metals possible, graphite
or graphite forbidden
Consequences No metallic metallic tubes in the core

tubes in the
core

possible

Impact of coo-
ling system

out of core

no limitations, in core coo-
ling possible

Fuel system

liquid fuel

solid and liquid fuel, cooling
in/or out of core. Solid fuel:
ceramis, metals; liquid fuel:
chlorides only.

Example only
for liquid
fuel reactors

only, cooled
out of core
molten |aqueous
alt solu-
LiF tion
BeF2 suspen-
iy 560 ho
Ty 5! PR
UFH- in D2O
graphite

molten salt molten salt

NaCl NaCl
232Th01u 238U013
233UCl3 259Pu013

22

THE LIQUID FUEL REACTOR ALSO HAS OTHER ADVANTAGES IN RELATION TO
THE CLASSICAL SOLID FUEL REACTOR

Solid Fuel Liquid Fuel
Pressure pressure of volatile fission no FP pressure: Very low
products. pressure of fuel components
Cooling cooling only in the core cooling in the core or
Problems out of core
Burn=-up periodic recharging continuously recharging e.g.
typical case PWR: specific MSBR: 22 KW/litre
power 40 KW/kg fuel,
burn-up 30000 KWd/kg fuel,
dwell time 750 days ~dwell time 10 days
Reprocessing | periodic continous
Heat Transport in the Core
solid fuel liquid fuel
"’,——~—“-‘-\\\\~ [’/’,,——-~———-§\\\\
fsggeraturef‘ A ¥ g
30004
20001 é
Coplant
lanty Temp.
1000 J \

H
4 '
fd

41
4

23

Fuel
thermal
fast ambient pressure overpressure
liquid solid
metal now not
possible
i \Ha) projected LMFBR
0]
0
e MSBR _
= I= salt (fluoride)
3 % MSFBR
& |- Chlorophyl _
% - (chloride) .
=
- LWR
E water HWR
P not not
~
g possible possible
T
o] - English
§ co French
. )
o | 2 GCR
2 o0 No No
[63]
g - HTGR
He HHT
- GCFBR
GCR = Gas Cooled (thermal) Reactor
GCFBR = Gas Cooled Fast Breeder Reactor
HHT = High Temp. Helium (thermal) Reactor
HWR = Heavy Water (thermal) Reactor
LMFBR = Liquid Metal Fast Breeder Reactor
LWR z Light water (thermal) Reactor
HTGR = High Temp. Steam Turbine Reactor
MSBFBR = Molten Salt Fast Breeder Reactor
M3BR = Molten Salt (thermal) Breeder Reactor

24

MOLTEN SALT FUEL APPEARS TO BE THE BEST LIQUID FUEL

Only limited possibilities exist for choosing a fuel to operate in

the liquid state

in a reactor.

The limiting criterium seems to be the radiolysis in the extremely
high field of neutrons, fission fragments and gamma rays.

In addition the chemical stability of the dissolved species of

thorium, uranium and/or plutonium in the presence of the products of
radiolysis is a severe problem. The ionic liquids, that 1s the molten
appropriate media for fission fuel

ionic salts seem to be the most
(see table of electronegativity

overleaf).

Liquid state Range of 1li- Stability Reactor type
quid state
(°C) Thermal Fast
Aqueous solution| 0<T<300 with Radiolytic decom-|Possible but
pressure position. Instabi-negative ex-
lity of U and Pu |perience no
salts'ln aqueous |wlth homoge- possible
solution neous
thorium due to
reactor
neutron
O?ganlc solu- -50<T<309 . Rad}o}ytlc decom-|Possible (?) moderation
tion or suspen- | decomposition | position. Very but no
sion unstable compound|experience
of Pu and U
Metallic solu- -50<T<2000 Very stable, Possible. Possible.
tion or suspen- some limitations |Some expe- LAMPRE,
sion in the solubility|riments in Reactor
of Umet and Pumet Brookhaven Experi-
Nat.Lab. USA|ments
(with Pu
alloys?
Los Alamos
USA
Molten salt 400<T<1000 Stable to irra- Possible Possible
solution diation but using fluo- |using
corrosive! rides. MSBR |chlorides.
Experiment No ex-
(Oak Ridge, |perimental
USA), experience
8 MW(th) EIR
Switzerland
Ceramics 1500<T
practically - ~ -

impossible

25

BOTH THERMAL AND FAST FLUX MOLTEN SALT BREEDER REACTORS ARE POSSIBLE

Thermal Fast
Breeding ratio max. 1.06 Le3 = 145 £2)
Possibility of »
transmutation 1o yes, 0.35
Fuel cycle Th-Upzz only because, in U-Pu239 and/or Th-U233

this case only, is n
thermal >1.2

Chemistry of molten
salt

Fluorides only. positive
experience exists

Chlorides, or fluorides
(?) no experience with
chlorides

Structural materials

Only graphite-direct
contact, no metals (no
parasitic capture)

metals in core
permitted.

Cooling

Out of core cooling

Either in core or out of
core cooling

Present day develop-
ment

USA - Oak Ridge; France
Fontenay aux Roses;
India - Trombay?

UK., -Harwell, Winfrith
(2); Switzerland:
Wirenlingen EIR

Moderation of
neutrons

Fluoride in large
amounts acts as a weak
moderator

The moderation factor for
chlorine is 3 times lower
and the concentration is
also smaller

Electronegativity

Super
Actinide

Components of 0l -of
molten salt reactors

[] thermal

Zl fast

Energy Decrement

6 8 10 12 1l 16 18 20 22 o4 26 28
A

26

MOLTEN FLUORIDE VERSUS MOLTEN CHLORIDE FUEL - I

Criteria for molten salt as fuel:
Melting point: lower than 500-600°C
Boiling point higher than 1500°¢ (partial pressure)

Density: as high as possible, p= 3 gem -
| . thermal . _fast
Capture cross section low Jm—— <0.,16; 9 s mpts small

Cross section for nuclear reactions low; e.g. for 3501 (n,p)BSS
Elastic scattering: high for thermal, low for fast (see previous page)
Radiolytic stability high and fast recombination (previous page)
Chemically stable (large free enthalpy of formation), no precipitates
Simple technology, low corrosivity

Low viscosity (low pumping requirements) ~1 centipoise

Low price, good avallability, non poisonous

1500 o 3 X only as a small component

of a solution

ThF . . ;
.ll (111°%) limit for single substance

PuFu

1000 =

Melting
Temperature
(°c)
limit for complex mixture
500 - - -

UCl6 non stable
5.35°

density of UClx (gcmflj) PuF6

UFg

T I i | 1
I5E v v VI

Oxidation State
MOLTEN FLUORIDE VERSUS MOLTEN CHLORIDE FUEL - IT

27

- MELTING POINTS

Fluoride

Chloride

Reactor Type

Thermal (fast possible)

only fast because

Gcapt Cl is large
Components - fertile ThFu only UC‘l3 or UClu(unstable!)
| fissile UFM or UF3 PuCl3 only
dilutent| LiF Possible but also LiCl forbidden since
Bel, plays the role of BeCl, they act as
2 2
moderator moderators
NaF forbidden since .
T is too high NaCl possible
melt
Moderator Li.F, BeFo not suffi- none

cient. Graphite 1s nec-
cessary: does not react
with molten fluoride

Metallic tubes in the
core

not permitted due to
neutron balance

permitted, e.g. a
molybdenum/iron
duplex

Melting Point of -

Fluoride

1100 +

1000 4

700 T

- Selected Salts

11004

700

500 7

Chloride

0 0:.25 0.5

mol fraction

0.75 1.0

mole fraction

28

MOLTEN FLUORIDE VERSUS MOLTEN CHLORIDES FUEL - IIT CHEMICAL STABILITY

The free energy of formation for fluorides 1s greater than that for

chlorides.
The formation of oxides of uranium is retarded in fluorides.
The ratio AG (HF/HZO) 1s not the same as AG (HCl/HzO).

Carbon tetra-fluoride is a product of graphite and fluorides.

Free Enthalpy of Formation Free Enthalpy of Formation

1007 1oofl

G

-2007 -200 ‘ ;.
.——‘:::::—

(KJ/mol EL_ (KJ/mol C —

-300 300

NaCl
-400
" -1400 4 —
/o/
/u/o
-500 47, <5004 4o, "
——’b.
-600 T T T T T T — -600 6‘0 50 850 JO 1550 1100
500 600 700 800 900 1000 1100 1200 S ° ! ’

Temperature (K)
Temperature, K

1
1200
29

MOLTEN FLUORIDE VERSUS MOLTEN CHLORIDES FUEL - IV

STABILITY AGAINST FISSION AND CORROSION PRODUCTS

For molten fluoride the crucial mechanism

UFu(d) + 1/2 Cr(Ss) UFB(d): K = 1.38 x 10
i P 1000K . . .
Numerical Values given for AG in 2L1F—BeF2 solution
Fuel Structural Fission Fuel Structural
R Components Materials Products Components Materials
100 +100 ~
o ™ Moc1 4-10C13
2
CFU -
-100 -100 — -FeCl,
' CrCl. .4
AgL000K TeFg AGLO00K 2
RuF
o¥ i Rqu _UClu
-200 oF , MoFy L pop -200 -
3 Th014-~UCl
v (KJ /molC}) o 3
(KJ/molf") L i PuCl3 —PaCl
— Cerq-Fer -300 | FAmClB
BeF2 CrF_ NaCl
— 3
rUFll
-~ UF
3
-400 PuFB- ThFu Zer N -400
Aij_
CsF -
PaFBA Rbi? L Sml,
[~ PrF3
-500 + LiF ga{ - CeI*.j =500 =
Skl -
-600 -600

Fission
Products

- SeCl

= ZrCl2

SrCle—_ RbC1
BaClz“_ CsCl

50

MOLTEN SALT FLUORIDE BREEDER (THERMAL) POWER REACTOR

1 GW(el), MSBR (Oak Ridge concept)

7

Fuel: 'LiF (71.7 mol%), BeF, (16%), ThF, (12%), *33yr, (0.3%)

2

Tfuel = 839-978 X Tmelt = 172 &

Breeding ratio = 1.06 Doubling time = 22 years

1.-1

Density (~908 K) = 3330 kg/mB. Heat Capacity = 1.36 J g K

Viscosity (~908 XK) = 0.01 N s m °

Structural material: graphite (in core)
Hastelloy N

Problem of delayed neutron precursors.

4 uynits
1.8 MWel

Primary
Salt Pump

1m3 /s NaBFy-NaF

Secondary
Salt Pump

727 K

Purified
Salt

“Graphite
Moderator

Reactor
2250 MWth

Heat '
e 5 menanser] [
TLiF-BeF_~ThF, -UF
2 4 4 Steam Generator
Fuel Salt

e f——

Chemica
Processing
Plant

Turbo-
Generator

Steam
31

A FAST BREEDER REACTOR WITH MOLTEN CHLORIDES COULD HAVE IN CORE COOLING

Total power: 2050 MW(th) Core volume 7.75 m3 (2m x 2.36 m)

Fuel: PuCl3 (15 mol%) NaCl (85 mol%). Tmelt = 685005 0 2,34 gem

Coolant, fertile: ZB8UCl3 (65 mol%) NaCl (35 mol%) Tm = 71000;
3

p= 4.00 gém-

5

elt

Specific power: 220 W/cm3 core; 705 W/g Pu
Velocity: Fuel 2 m s—l; coolant 9 m s+
Number of tubes in the core: 20,000 (1.26/1.20 cm dia.)
Reprocessing: continous, 3 gs-l; dwell time: 10 days
Breeding ratio: internal 0.716; outer 0.67

total 1.386  (corrected)
Mean fast flux across core: 7 X 1015 n cm-2s—l

fast flux in the centre: 1.2 X 1016n cm_2>s-l

Temperature reactivity coefficent (8k(%/6 T(OC))

fuel = 3.8 x 10°°; coolant +1.29 x 10 °

Structural material: Fe (1.45 kg/l): Mo(0.465 kg/l)

;753 750° 790°¢ k:i) -
_::::F:ji=10m % 1L
o a4
f?fi: /
/;?f;/ /
CEEEE:::: 93;g°c fuel ?EEEEEEE 1y m
) e
I //
¥ //

” %‘ //////

32

THE VERY COMPLEX PROBLEM OF CONTINOUS REPROCESSING OF THE FUEL IN A
MOLTEN FLUORIDE REACTOR HAS BEEN SOLVED

The operation of a molten salt thermal reactor as a high performance
breeder is made possible by the continuous reprocessing of the fuel
salt in a plant located at the reactor site. The most important opera-
tion consists in removing fission products (rare-earths) and isolating
Pa-233 from the high flux region during its decay to U-233 in order

to reduce neutron absorption.

It 1is also neccessary for:

- the removal of excess U-233

- addition of fresh amounts of Th-23%2

- the proper redox potential to be maintained
- the oxides to be removed

- corrosion products to be maintained at tolerable levels

In order to achieve the breeding ratio of 1.06 - 1.07 the following

fission products must be removed as shown.

Dwelling Time of F.P. in Molten Fuel

Rave Earths

v

Noble Metals

_.
Se, Nb, Mo, Te, Ru, Rh, Ag, Sb, Te, Pd

Semi-Noble Metals
Zr, Cd, In, Sn

Noble_Gases
Kr, Xe

Volatile Fluorides
Br, I

Y

Non-Volatile Fluorides

v

Rb, Cs, Sr, Ba

Protactinium

v

Actinides

Wp, Pu

1 day 10d 100 d 1 year 10 y
T T T T l! T ll = T - L
1 1 102 100 10" 10 10° 107 108 109

time, sec
33

Reprocessing in MSBR

Power
Reactor
~2250 MW(t

UF
)

6

Collection

SR—
705%¢
Processing A
time
10 days
D.02 g/sec A
B
| |
[rradiated
Huel Efficiency
of RBemoval
~95%
\‘-Fluorination
U-233

or

Bi in fuel <2 ppM Extr ion
< 1
(3-6
stagks
j
Salt + R.E )
i
- A
¥
Extpadtion #
| 2
L/
Reductive '—‘ *
Extractién A
k ExtrT
T=640°( 5
Molybdqnum
Separa -|as g
ti¢n ot | Structyral e,
R.E and .
Pa Mater1q1
Extrl
uy

Bi/Li (0.5 m) |

EL!

THE EXISTING REPROCESSING SCHEME IS TOO CONSERVATIVE.
IT DOES NOT TAKE INTO ACCOUNT THE PROBLEMS OF RADIOACTIVE WASTE MANAGEMENT

PRESENT STATE POSSIBLE FUTURE FORM
Reprocessing periodilc quasi=-continuous
Removal of F.P. in a mixture as individium
Transmutation of hagzar- no possible in a special
dous F.P. central station
Actinides (Np, Am, Cm..) removal to radio- recycled in a special
active waste power reactor?
Reprocessing today Reprocessing in the Future
Fresh
Fuel
Fuel
Pr:epara- ‘uel 2% Pu + ~100% Np, Am, Cm
tion Preparatio
98% Pu
+ U
| Power Sink A Irra-
Reactor diated Reprocessiyig Fuel
Electrical |_Reactor 2 = .
Energy 9 Electrical SpIatessIng
/ Energy
U losses ;
RE_D QSSegS ,; Irradiated Fue 2% Pu lossef
/ Sink
Np, Am, Cm / Stable or Np, Am, Cm lossed
100 % ¢ short lives
“ 0.001 %
¢ V
-
Storage Transmutatian ;
F ‘:)_ Reactor ;
/ “
L
L
long lived F.P. v
Stable Nuclide

55

THE LONG TERM HAZARDOUS TRANSPLUTONIUM ELEMENTS CAN BE FULLY
BURNED-UP IN A FISSION REACTOR

Parallel to the burning of U and Pu also synthesis of Np, Am, Cm,
Bk and Cf is going.

At present time the actinides, other than U and Pu, are removed
to the waste.

The problem of the hazard from these actinides in long term

must be solved.

The recycling of actinides bach to the power reactors gives the so-

lution of problem.

The methods of the recycling must be developed.

Relative
Toxicity

.

Hazard from Nuclides

actinides

242

all other F.P.

240
239 4 Cs + Sr

238

36

CHEMISTRY OF FLUORIDE FUEL

Fluorides are the only salts with acceptable Oobs? requisite stability
(AG) and melting point.

Oxidation state UF3 due to stability and Iow Tmelt
PuF3 the most stable but in both cases
ThFu the only one Tmelt 1ls very high
; .
Components 1) 'LiF and BeF2 for lowering Tmelt
2) for moderating
3) 7Li because: gLi(n,a) g Oip = 953 barn

) LiF/BeF2 for low viscosity

Tritium production: in spite of 7Li the products are (for ~1 GW(el))
253

7

U(n,f) ternary:30 Ci/day

Li(n,on) 3‘I‘ : 1200 *

®Li(n,a) °T: 1200 "

Graphite - as moderator and structural material (p = 1.9 gcm-s)

b UFM + C CFH + MUFB: e.g. pressure CF) = 10—8 bar.
The problem of xenon-135: This F.P. is adsorbed in the goen pore volume
in the graphite; accesible void volume ~10%

graphit replacement in the core ~each 2 years

Disproportionation

K9OOK

YUF,(d)  3UFy(d) + U: - 6.3 x 10 ¢

Instability in the presence of oxygen-containing materials:

UFM + 02 UO2 T oeeen

solubility ofUO2 in molten salt ~40 ppM

also precipitation of Pa20 PuO

52 2
27

FLUORIDES

F.P. Balance: UFu > 2 atom F.P. + 4UF -

i,
> MelF + MeIIFZ + Me™"TFy + Me® +~ + F,

The Breeding Ratio is very sensitive to the removal rate of RE and
Pa-233,

Structural material for reprocessing: TZM: molybdenum alloy

for fuel system Hastelloy N (modified) (wt%)

Ni 75 8 0.1 C 0.06

Mo 12 W 0.1 P 0.015

Cr 7 Al 0.1 S 0.015

Fe 4 Ti 1.0 B 0.001 (10 ppM)
Mn 0.2 Co 0.2

Hf 1.0 Ta 2.0

Cost of salt 57$/pound, in this U=-233 $13/g; total cost 23x10 $

Breeding Ratio v Reprocessing

days Pa-cycle e [J=2 33

1.05

0.95 4

0.90 «

T T T T
100 200 . 300 400 500
Recycle Time days
38

ANALYTICAL PROBLEMS FOR MOLTEN FLUORIDE FUELS ARE OF THE HIGHEST
IMPORTANCE

In order to fully exploit the unique features of the molten salt reactor
concept it will be neccessary to carry out all analyses automatically

on line with transducers located directly in the salt streams in the
reactor and reprocessing plant. The most significant items to be

measured are

- Redox conditions of the fuel: the U3+/ULl+ ratio which influences the
rate of corrosion and the distribution of certain fission products

and tritium in the reactor. This is due to the disproportionation

WUF5 + 3UF, + g® (only in molten state)

(by spectrophotometric methods also voltammetry)

2 ..+3

- corrosion product concentration (Cr+ Ti -, voltammetry)

- oxide levels (by hydrofluorination: MeO + HF » H

0 +‘MeFX)
50 ppM precision |

2

- bismuth (polarographic method 50 ppG)

- hydrogen and tritium (diffusion to Pd-electrode some ppM)

- protactinium

- certain fission products (remote gamma spectrometry and for noble

metals mass spectrometry)

Note: Most of these techniques are carried out in hot laboratories.
The electrochemical technique appears to be the prime candidate for
practical on-line fuel analysis because of the simplicity of its trans-

ducers.

The analytical and general chemical problems in the molten chlorides
are also of interest because an important step in the continuous
reprocessing technique is the liquid bismuth/molten chloride reduction

extraction metal transfer process.
59

Chlorine 35 reacts with neutrons: 35Cl(n,p); 35Cl(n,a)32P

Separation of 3501/3701 is relatively easy because

1 3Te1 in the naturally occurring Cl is 24.5%

2 the ratio 2%%25 = 0,057

3 Chlorine is volatile

WITH MOLTEN CHLORIDE FUEL IT SEEMS THAT A CONTINUOUS GAS EXTRACTION
SYSTEM IS POSSIBLE

Iodine isotopes as precursors of delayed neutrons

Isotopic separation e.g. Caesium

Chlorine in Neutron Flux Z 51 52 53 54 55 5p
L L 38

—+Sb = Te—t I—Xe —»Cs —=Ba A=137
from — 4
fission Aj j I
21 T
[:] stable 100
0captur 3¢ min Independant
(:) beta-unstable very sfall Yield 10
1
0] jom [ O g

5 Cs-137 Ba-137
Half L
a te 103 30 year stable

Seconds 2
10

10

19 =

1 Y T T T T

=

Extraction
Rate

18 A s-l 14

17 -

4o

THE BALANCE OF CHLORINE DURING THE FISSION PROCESS IS "NEGATIVE"

PuCl, (n,f)

100 PuCly (n,f)

o+ 4+ + =+

100 PuCl, (n,f)

LITERATURE

E!

+ E" + 3C1
0.008 Se + 0.003 Br + 0.942 Kr + 1.05 RbCl + 5.49 Sr‘Cl2
3.03 YCl3 + 215 Zr’Cl3 + 0.29 NbClS(?) + 18.16 MoCl2 +

.01 Tc + 31.45 Ru + 1.73 Rh + 12.66 Pd + 1.88 AgCl + 0.66
0.66 CdCl, + 0,06 InCl + 0.325 SnCl2 + 0,67 SbCl, +

2 3
7.65 TeCl2 + 6,18 I + 21.23%3 Xe + 13%3.35 CsC1l + 9.50 Ba012
5.78 LaCl3 + 13%.98 CeCl3 + 4,28 PrCl3 + 11.87 NdCl3 +
4y PmCl3 + 3,74 SmCi_, + 0.60 EvCl. + 0.03 CdCl
3 2 3
200 ECll,.5

For molten salt reactors only, estimated to the recent publication

Harder B.R.,
Long G.,
Stanaway W.P.

McNeese L.E.
Rosenthal M.

3

W.
.Rosenthal M.W.,
Haubenreich P.N.,
Briggs R.B.

Taube M.

Taube M.,
Ligou J.

Taube M.,
Ligou J.,
Bucher K.H.

Compatibility and reprocessing in use of molten
UCly alcalichlorides mixtures as reactor fuel in
"Sympos. Reprocessing Nuclear Fuels"

Ed. P. Chiotti, USAEC, Con-690801 (1969)

MSBR- a review of its status and future Nuclear
News. Sept. 1974

The development status of molten-salt breeder
reactor.
Oak Ridge, ORNL-4812, 1972

Steady-state burning of fission products in a fast
thermal molten salt breeding power reactor.
Ann.Nucl.Sci. Engin. 1, 283 (1974)

Molten plutonium chlorides fast breeder cooled by
molten uranium chlorides.
Ann.Nucl.Seci. Engin. 1, 277 (1974)

The transmutation of fission products (Cs-137,
Sr-90) in a liquid fuelled fast fission reactor
with thermal column.

EIR-Report Nr. 270, Feb. 1975