EIR-332.txt

IR -Bericht Nr 332

EIR-Bericht Nr. 332

Eidg. Institut fur Reaktorforschung Wirenlingen
Schweiz

Fast Reactors Using Molten Chloride Salts as Fuel

Final Report (1972-1977)

M. Taube

Sily

Warenlingen, Januar 1978

FAST ~ REACTORS  USING
MOLTEN ~ CHLORIDE  SALTS AS  FUEL

FINAL REPORT (1972 - 1977)

| , [ -

e are

"
-~

SWISS FEDERAL INSTITUTE FOR REACTCOR RESEARCH

CH-5303 Wlrenlingen

Januar 19/8
Co-authors and contributors to this work

Physics, Neutronics
J. Ligou

Cross sections and codes
E.H. Ottewitte
J. Stepanek

Thermohydraulics
K.H. Bucher
M. Dawudi

Chemistry and experimental
Dr. E. Janovici
Ur. M. Furrer

Other assistance

Programming - 5. Padyiath
Code operations -~ B. Mitterer
Text preparation - R. Strattan

Support and encouragement

Prof. H. CGrénicher EIR, Wirenlingen

Or. P. Tempus EIR, Wdrenlingen

Or. J. Peter EIR, Wirenlingen

Cr. H. Schumacher EIR, Wirenlingen

Support and encouragement

7. Fougeras ‘Fontenay aux Roses

J« Smith Winfrith U.K.

C. Long Harwell U.K.

L.E. McNeese Oak Ridge National Laboratory U

France

C

e Do

A
Ssummary

This report deals with a rather exotic "paper reactor” in which
the fuel is in the form of molten chlorides.

(a) Fast breeder reactor with a mixed fuel cycle of thorium/
uranium-233 and uranium 2538/plutonium in  which all of
the plutonium can be burned <4 44fu and in which a dena-
tured mixture of uranium-233 and uranium-238 is used to
supply further reactors. The breeding ratio is relatlively
high, 1.583 and the specific power is 0.75 GW(th)/m® of
core.

(b) Fast breeder reactor with two and three zones (internal
fertile zone, intermediate fuel zone, external fertile
zone) with an extremely high breeding ratio of 1.75 and
a specific power of 1.1 GW(th)/m’ of cocre.

(c) Extremely high flux reactor for the transmutation of the
fission products: strontium-90 and caesium-13/. The effi-
ciency of transmutation is approximately 15 times greater
than the spontaneous beta decay. This high flux burner
reactor is intended as part of a complex breeder/burner
system.

(d) Internally cooled fast breeder in which the cooling agent
is the molten fertile material, the same as 1n the blanket
zone. This reactor has a moderate breeding ratio of 1.38,
a specific power of 0.22 GW(th)/m® of core and very good
inherent safety properties.

A11 of these reactors have the fuel in the form of molten chlo-
rides: PuCl3 as fissile, UCl4 as fertile (if needed) and NaCl

as dilutent. The fertile material can be 2*®UCl1, as fertile and
NaCl as dilutent. In mixed fuel cycles the **’UCly is also a fis-
sile component with 232Th614 as the fertile constituent.

In some special cases a hypothetical mclten fluoride has been

checked using PuFs3 as the fissile, UF3 as the fertile and NaF

and Zrl, as the dilutents components. In this case one obtains
a lower but still respectable breeding ratio of 1.5,

In all cases a directly coupled continuously operating roproc-
essing plant is proposed. Some of the technological problems of
reprocessing are discussed. Furthermore the report touches on some
of the difficulties associated with corrosion arising from the
use of these molten media coupled with the irradiation effects
such as structural damage from fast neutrons.
The thermohydraulic studies show that even under the extreme op-
erating conditions of very high neutron fluxes and high specific
power, cooling is possible, in most cases by out-of core cooling
but also in one or two cases cooling internally in the core.

Some molten salt reactor specific safety problems are discussed.
The influence of fast neutrons on the chlorine, forming sulphur
by the (n,p) reaction has been experimentally investigated and

the results are reported briefly.

With this report the work of several years at the Swiss Federal
Institute for Reactor Research is brought to a conclusion.
CONTENTS
p3ge
Forward 8
1. Molten Salt Reactors. General Description 9
1.1 Methods of classification 9
1.2 Method of cocoling. External: Internal 9
7.3 Intensity of Neutron Flux 10
T4 Number of core zones 12
1.5 Type of Fissile Nuclide. Plutaonium: Uranium 12
1.6 Neutron esnergy: Thermal and Fast 12
1.7 Purpose of the reactors 15
1.8 Fuel Components. For Molten Salt, Fluoride
and Chloride 16
1. Short resume of the classification 16
1.170 Method of Neutronic Calculation 17
2.1 Breeder Reactor with Plutonium Burning 4n sLtu 19
2.7.7 Aim of this Concept 19
2.17.2 Reactor in the Build-up Phase 2
2.17.3 The Transient Reactor 29
2.17.4 The steady state reactor 29
2.17.5 Comparison of the three phases 35
2.2 Impact of some parameters on the mixed zone
two zone fast breeder 43
2.3 Material balance of the steady state reactor 49
2.4 Conclusions 49
3.1 The Three Zone Reactor 51
3.7.7 Introduction 51
3.17.2 The three zone breeder with thorium/
uranium-233 57
3.2 A Three zone breeder reactor with a mixed
tfuel cycle U-238/Pu-239 plus Th-232/U-233 52
3.3 The three zone reactor - uranium-plutonium
fuel cycle 57
3.4 The three zone breeder reactor: Very high
breeding gain 63
I>

B A

The Two zone fast breeder. Fuel of
uranium plutonium fluorides

3.5.7 Introduction
3.5.2 Arbitrary assumptions and uncertainties

High Flux Reactor with Fluoride Fuel

High flux burner reactor for transmuation

A

o~ O

Need for fission product transmutation

—

1. Introduction
1.7 Why some opinions concerning trans-
mutations in a fission reactor are
rather pessimistic
4,1.3 Which fission products are suitable
candidates for ftransmutation and
in what quantities?
4.17.4 In what way could a burner reactor be
coupled to a system of breeders:
4,17.5 1Is the rate of transmutation
sufficient?
In what reactors is transmutation
possible?
4.1.7 What are the limitations of a solid
fuelled reactor?
4.1.8 The liquid-fuelled fast reactor with
central thermal zone

4,
4.

N

=
RN
)

The Neutron-physical aspects of the High
Flux Reactor (according to Ligou, 1972)

1 Introduction
2 Neutronic calculations
.3 Moderation requirements
4 Influence of other parameters

B
MM

Thermohydraulic considerations
Some results

Comments on hazard coefficients
Secondary processes

Conclusions

page

/70

/0
/0

83
An internally cooled breeder with uranium-
plutonium fuel

g1 oy U U

LN -

Design features adn objectives
The reference design

Neutron physics

Satfety problems, comments

Chemical and Related Problems

6.

6.

Experimental Work (according to Ianovici, 1976)
7

/.

1

7

B w

1

2

Physical and chemical criteria for slat
components
Corrosion of structural material

6.2.17 General criteria
6.2.2 Molybdenum as structural material
6.2.3 The irradiation of molybdenum and

iron in a fast high flux reactor

Fission product behaviour in the fuel
Some comments on reprocessing
In core continuous gas purging

The proposal

6.5.1
6.5.2 Delayed neutron emitters

Chemical behaviour of radiosulphur obtained
by *°Cl{n,p)?°S during in-pile irradiation
Temperature dependence of sulphur species

(accaording to Furrer, 13/77)

P3ge
137
137
147

147
154

177
wl

page

Thermohydraulics 185

.17 Introduction 185
8.2 High Flux reactor with the core as a spherical

shell 135

5.3 Power Reactor with spherical core 169

8.4 The external heat exchanger 191

8.5 The internally cooled reactor 194

References 199

9.1 List of EIR publications used in this report
(chronological order) 199
9.7 Former publications concerning molten
chlorides fast breeders and the fluoride

thermal breeder 200
3.3 Publications concerning transmutation 204
J.4 Publications concerning the thorium fuel

cycle 206
9.5 References to the experimental work

(chemistry) 208

5.6 References to the physics calculations 208
FORWARD

The history of the development of fission reactor concepts
using molten salt as fuel media is as old and as complex as
the history of the development of nuclear power 1tself. The
ups and downs have followed those of the parent technology
but the swings have been if anything more violent. In 1976
for example molten salt technology all but died out but then
in 1977 a new attempt at revival was bepun this time associa-
ted with the new interest in proliferation-proof systems.

The author of the present paper has a profound belief that
the concept of molten slat rsactors coupled with continucus
reprocessing and the associated waste management will become
an important feature of nuclear strategy perhaps in 10 or

20 years time.

In this report the efforts in this field over the last =six
years are summarised.
1. MOLTEN SALT REACTORS. GENERAL DESCRIPTION

1.1 Methods of classification

There are many ways of classifying a reactor type. One such pos-
sibility 1s shown here.

a) Method of cooling

b) Flux intensity related also to specific power density

c) Number of zones in the reactor

d) Kind of fissile nuclides and fuel cycles

e) Neutron energy

f) Purpose of the reactor

g) Dilutent for the molten salt
It 1s clear that such an arbitrary classification is not neces-

sarily internally compatible and not all reactor types fall easily
into the scheme chosen.

1.2 Method of cooling. External: Internal

Molten fuel reactors differ from the point of view of the cooling
system. The following are three types of molten fuel reactors:

to the external heat exchanger. In this type of reactor, only fuel
and fertile material are present in the core (no cooclant). The
large amount of molten fuel ouside the core does not of course
contribute to the critical mass.

This type of reactor has been discussed for example by Nelson,
(Argonne 1967) and Lane (USA 1870) especially as a high flux ma-
terials testing fast reactor.

In externally cooled fast reactors the loss of a portion of the
delayed neutrons could adversely affect reactor contrcl. Also the
biological shielding outside the core is very expensive. In this
paper most of the reactors discussed are externally cooled.
Internally, direct cooled reactors: here the cooling agent is
pumped directly into the core where, after mixing, the fuel in
the lower part of the core is separated and pumped out of the
core to the heat exchanger. The direct contact of molten fuel
with molten coolant has several particular advantages! very
good heat transfer, no coolant tubes (or cladding), possibility

of transporting fisslon products.

The disadvantages are unfortunately, also numercus: problems
of mixing and separating the fuel and ccolant, corrosion, etc.
This type of reactor has also beern studies, e.g. cooled by
molten lead (Long, Harwell and Killingback, Winfrith 19679,
cooled by boiling mercury (Taube, Warsaw 196E8) and ccooled by
boiling aluminium chloride (Taube, Warsaw 1566). This type of
reactor must bhe considered as an "extremely exotic type”, and
only some references are given here.

Internally indirectly cooled reactor: here the cooling agent
flows through tubes in the core. Heat is transferred from fuel
to cooclant across the tubes. No direct contact between molten
fuel and liquid or gaseous coolant 1s permitted. These types
have alsc been studied, 1In most cases using sodium as a coolant,
(Nelson, Argonna 19687]) or molten chlorides of uranium (Taube,

1370}, See Fie. 11.

1.3 Intensity of neutron Flux

The molten salt reactors discussed here can be used for two more

or less guite different purposes.

- power production and fissile breeding, which is self evident

- neutron producticon for nuclear transmutation of the long
radionuclides produced in power reactors.

In this report both types have been considered

- power breeding ractors with a mean power level of approx
3 GWith) and steam production with over critical parameters.

- burner reactors with a very high neutron flux particularly in
the internal zone for neutron moderaticn when the thermal flux
reaches 3 x 10'% n cm™ %571,
Fig. 1.1 TYPES OF REACTOR COOLING SYSTEMS
TYPE SCHEME CHAPTER
INTERNAL CHAPTER 8§
INDIRECT
COOLING
\. J
____>
i fanuun
EXTERNAL
INDIRECT CHAPTER
COOLING 2, 3, 4
9 J L
>
INTERNAL : ( I
f
DIRECT ) 1
B Ta Ve W W
COOLING o
OO (o)
(BOILING) o O O
o o
© Qo ©O oo here not
O discussed
oo O
. e J U —J

1.4 Number of core zones

The division of the reactor into several zones must be consid-
ered from the point of view of neutronics, thermohydraulics
and safety.

The organisation of multiple zones is easier in the case of
molten fuel reactors than for solid fuel reactors. In this re-
port two types are discussed

- with two zones

- wlth thrse zones including outer and inner fertile zone,
(soe Fig. 1.7).

1.5 Type of Fissile Nuclide. Plutonium: Uranium

The fast reactors show excellent neutron properties, not only
tfor the fuel cycle:

Uranium -238/Plutonium bhut also
Thorlum -232/Uranium-233

Also a mixed fuel cycle of both types has some spectial advan-
tages. Fig. 1.3 shows the nuclear properties of the fissile nu-
clides.

1.6 MNeutron energy: Thermal and Fast

The reactors discussed here are all fast reactors. Thermal reac-
tors however have also been extensively and intensively investi-
gated during the 1360's and 18/0's in Oak Ridee National Labora-
tory USA. (Rosenthal at all, 19727
_"]3_

Fig. 1.2 TYPES OF REACTORS

PROPERTIES
NUMBER ZONE GEOMETRY
POSITIVE NEGATIVE
SIMLICITY OF RELATIV
ONE TECHNOLOGY LOW BREEDING
RATIO, BECAUSE
OF TOO SOFT
NEUTRON.
FERTILE
BLANKET
TWO THE ORTIMAL
DESIGN
GODD USE OF THE GEOMETRY
NEUTRONS IS VERY COMPLEX
THE NEUTRON FLUX
THREE BDISTRIBUTION IS

DISTURB

Thorium Uranium Cyecl
/\ Stable in Nat

() beta stable
(Q beta unstable

thermal
a

is ri
\(n,y) 15

O

e

ure ’neutron capture

—* heta decay

’[ fission

elativ big

Tt Fa S

T T T

BB a1 i z
{ross section Cross Section of Pu-239

(simplified)

108

varns (13
1000 4
100 4
107 “fission oM, )
Gcapbure\
aln,y) \\
\
1 \
\
\
\
\
\
AY
N\
\\
0.1 1 ~
~
AN
N\
\
\
0.01 T T T T 1
1072 1 10° 10" 10°
Neutron energy, eV

Uranium-Plutonium Cycle

- . 2
= cross section for neutron reaction (cm”)

g =
A $ = neutron flux (neutron cm'Zs—l)
X = decay constant (s-l)
all data for ¢ 2 10%%n en 57t 1.7x1077
241 | fast e e
op :j107%?
240 4
A
)
[}
t 1o
ofast is small af =12 *
i
1
! [}
x oz o st
€337 O 2.5 min O .7 days \’
o8 o 10740
238 A
U Np Fu
T T T
92 93 RE Z
Value of n for fissile nuclides
{simplified)
n
34
Ve
Pu-"f2 /////
—-—n  minimun
2
1
Thermal Fast
Reactors Reactors
0 2 T E T 4 LI 1 q
- :
10 1 10 10 10” 107

Neutron energy, eV
1.7 Purpose of the reactors

15

The principle purposes of the large ractors proposed can be

classified as follows.

Table 1.7

(Table 1.1)

Reactor type

Primary Aim

Secondary Aim

Comments
to be found
in chapter

Power Electrical Production of ch.5
energy . fissile nuclides
T > 600 C 2R > 1

dreeder Production of Production of ch.3
fissile nuclide electrical
B.R. Vv optimum ENergy

High flux Neutron Flux Production of ch.4

d (n cm %5 1)
> 1pt® for
transmutation

electrical
energy

High Temperature

T > 850°C
for chemical
reactions

Production of
clectrical
energy

not discussed
here

Non-proliferating

Maximum
security.

No plutonium
output

Production of
electrical
energy

ch.?

Propulsion

Heat for steam
turbine

not discuseed
here

Space Heating

Heat with

100 < T < 200%C

not discussed
here

1.8 Fuel Components. For Molten sal+, fluoride and chloride

In the thermal molten salt reactor the best fuel compound 1is
unduobtedly the flucrids.

For fast reactors the use of chlorine as the compound seems to

be preferable but the use fo fluorine (as zirconium and sodium
fluoride) as dilutent is not excluded.

1.9 Short resumé of the classification

Table 1.2 brings together all these characteristics 1n an at-
tempt at classification.

Table 1.2
This work:
yea no
Method of External X
Cooling Internal Direct X
Internal Indirect X
Flux Intensity High X
Low X
Number of One X
ZONRS Two X
Three X
fFissile Plutonium
Uranium
Mixed
Lnergy of Thermal
Neutrons Intermediate
Fast X
Aim Power X
Hreeder X
High Flux burner X
Dilutents Fluoride
Chloride

17.70 Method of Neutronic Calculatiaon

Almost all results given here have been obtained using the
following calculational method

- the reactor code: ANISN

- number of zones: 5, 6 or 7

- 40 - 100 spatial positions

- order of quadrature S5, checked by Sg

- neutron groups: Z2 or 23 groups including the thermal neutrons
(see Tabic 1.3)

- anisotropy by first order Legendre expansions
- library ENDF/BI, BII and BIV processed by code GGC-3 and GGC-4

- the management of additional sub-routines have been realised
by RSYST.
Table

1.3 Relative Fluxes in Each Group

18 -

Upper Mean Au Core
o Centre
boundary value (Lethargy) boundary
AL_S
1 15 MeV 12.2 MeV 0.4 - 107"
(.0002)
L0010
2 10 8.18 0.4 0o
(.0025) note
L0045
3 6.7 5.4 0.4 0020
(.0118) oz
0112
4 4.5 3,687 0.4 L .
(.0286) o7z
5 5.0 2.46 0.4 222 L .0146
U ‘é]
nan7y
; 35
g 2.0 1.65 0.4 < naot) .0285
L0264
1.35 1.23 0.7 019
/ (.0536) 3
. L0281 o
a 1.11 1.00 0.7 Coean) 0215
.nazeg
q n.9 78 0. o 0371
§ e e . (.06837) |
10 .87 0,55 n.4 L2 0787
: : ‘ (.1373) |
.45 0,37 0 He 007z
11 .45 . 37 L4 ( 4ona) 977
_ 1162
2 0.2 .25 . 1130
d 0.25 0.4 (.1047) 113
1044
3 20 N.165 0.4 ne4
q - (.0749) 1
10486
/ 0,135 0.10 .45 A1
14 13 108 0.4 Cane 11118
_ i L1061 i
15 86.5 keV 50.5 kel 0.75 C esa) 1200
1043
6 4 5, i 6
16 10, 8 5.0 1.00 * ha04) 1262
L0589
15.0 2,0 .25 : n748
17 5 o 1.2 14 n7
0047
' . 2.9 a. 008
18 4.3 2,94 0.75 Lo 1051
| .N148 _
2.03 1,20 0,75 _ 0708
‘ (.0025) e
.0048
20 0.96 N.67 0.75 .0079
(.0004) o7
.0008
21 0.45 0.24 1.25 0017
l C10- 001
5.75 < 10-" |
22 0,13 kel 0.4 eV < 10
§ < (107%)
Total 0.4 eV - 15 MeV = 17.40 1.0000 1.0000

2.1 BREEDER REACTOR WITH PLUTONIUM BURNING IN SITU

NI
-
[N

Alm of this Concept

The aim here is to demonstrate the possibility of using a molten
chlorides fast breeder reactor with external cooling as a device
for consuming all plutonium produced, 4n ALtu. At first the re-
actor is fuelled with denatured uranium -233/uranium -238 and
this 1s changed stepwise to a feed of thorium and depleted or
natural uranium only.

Such a reactor will have the following phases in its fuel cycle

(Table 2.1) See Fie. 2.1
Table 2.7
Fuel input Fissile Fuel output
Phase PUF”Fd
Fertile fissile o setu Fissile Fertile
start. Build-up| U-238: 70% {U-233: 30% U-233 none none
nhase. Th
(Fig. 2.1 A)
Transient Phase| U-238: 70% [U-233: 30% Pu-239 none none
(Fig. 2.1 B) Th U-233
Steady State U-238 none Pu-2393 mixed
Th + other -233: 30% =738
Pu-isotopes

20 -

Flg. 2.1 Two-zones Reactors with uranium-233/plutonium-239

A) BEGINING
(NOPU-417)

from the reactor

B) TRANSIENT
(NOPU-501)

BLANKET

C) STEADY STATE

(UTMOST
NOPU-302)

\

V

J
U238
o>
U233 for a new
reactor
j

2.17.2 Reactor in the Builild-up phase

At the start of the cycle the reactor core is fuelled by ura-
nium-233 denatured with uranium 238 {(see Fig. 2.1). [(Table 2.7]

Table 2.3 gives information concerning

- the method of calculation

- densities of elements in each of the 5 zones
(core, wall, blanket, wall, reflector)

Table 2.4 shows the neutron balance in the core and blanket

Table 2.5 gives the breeding ratio calculated by a microscopic
metnod of the form:

- and the macroscopic method by

_ production rate of fissile nuclide
macr rate of destruction of fissile nuclidse

BR
- and the maximum neutron flux which gives informatlicon on the
flux spectrum in the core.
Table 2.6 shows the geometry of this reactor e.g. see Figp.

- radius: 0.955 m

- volume: 3.65 m?

and

- specific power: 0.75 CW(therm)/m’ of core

- total power: 2.8 GWltherm)
and inventories of fissile and fertile materials.

Table 2.7 gives some informatlon concerning

- the material flux in this type of reactor. (for more ser sectiaon

2.3).
OBJECT Thorium-uranivm Breeder with Plutoniom

burnine in situy

REACTOR TYPE : Power, Hreeder
CEOMETRY : INTERNAL ZONE : Fuel
WAL L
Ser
INTERMEDIATE ZONE .
Fle
WALL : P

EXTERNAL ZOUNE : Fertile zone
WALL, REFLECTOR

POWER (GW thermal)

a3
B

FOWER DENSITY (GW therm/m3 core) @ 0,75

NEUTRCON FLUX, MZAN (n/cmzs] s 1.6 x10

FISSILE NUCLIDE . Pu-22¢/Pu-241 in core, fuol
U233 in fertile

FERTILE NUCLIDE - U238 in core, ThZ327 in fertiie

DTILUTENT : Chloride

COOLING SYSTEM Cuter

BREEDING RATIO 1.58

PARAMETER STUDIED : Make up reactor with U233/0238
transient reactor with U233 + FHy23@

steady state reactor with PouZ3¢

METHOD OF NEUTRONIC : See table 1.10
CALCULATION
— 23 -

Fig. 2.2 Three zones reactor °.8 GW th (meter)

TITLE 3
DATE ¢

NA
PA

UTMOST-417,NU2U.

SRSV FIFFERRERL PR TR R

SPHZRIC GcOM.
C3/s709/77

INPUT OJATA (RSYST FORMALIS

M)

RN ERELEYE RRRERRRER RN R EEFEERD

ORJ_R OF SCATTIRING

QUL DRA TURE ORUER

NO. OF ZONES

NOe OF INTERVALS

NO. OF GROUPS

CI5_NVALUE MODIFICR -2
PR-CISION UESIRZD L
NORYs FACTOR 2
MoW PARM. MOD. SRCH. 1

g e SO &

DENSITIES(ATIMS®L L -24/0M3

FEFRERERELERFEFRRERRRFEXERDE

CORE
= 7.04E=-03 CL = 1,30c2=-02
= D.obE-U4 PU2C39 = 7,31c-1i7
= 7T.31t=-47 FP239 = 1.6lip=-4d7
Ui/ (PUTOT + U3) = Be37
FIRST HWALL
= B.b2E~ud Fr = ¢J.88E-d¢
LA
= SeklE=-ud oL = 1.u8-0¢
= 1,3092=0C5 u2a33 = 1,03E8-15
2=NDJ WALL
= 1,37E~u?d F = 7e3lce=-ud
REFLEETOR
= .06 Fo = Babdp=-ud

»

ue23s
PUC 44

THZ 32

4ol 5c=ud
7317

'D.:-}i\"_‘t43
TITLE ¢
DATE ¢

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SPAZRIC GoOM,
Usrbari?

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- 29 -

The data given indicates that in the given geometry the nominal
power and all related values: temperature, temperature gradient,
velocity of circulating fuel, heat exchange etc. are as for the
steady state reactor with only plutonium fuel in the core (see
section Z2.1.4).

2.17.3 The Transient Reactor

Very shortly after the start up of the reactor, a significant
amount of plutonium has been produced in the core.

The total amount of plutonium chloride, after having been sepa-
rated from the fission products but mixed with the uranium chlo-
ride and sodium chloride is circulated back into the core. The
amount of fresh U-233 required is correspondingly smaller.

The reactor now burns two fissile nuclides, the uranium-233 and
the reprocessed plutonium. The data given below refers to the
case where approximately half of the fissile uranium-233 is re-
placed by plutonium-239 and plutonium-241.

Based on rather simplified assumptions concerning the isotopic
composition of the plutonium it can be shown that the same core
cesign is capable of burning the mixture of fissile materials
maintaining approximately the same power level.

Tables 2.8, 2.9, 2.10, 2.11% and 2.12 give the equivalent values
previously shown for the build-up phase.

These confirm the sulitablility of the core design for both the
build-up phase and the transient phase.

2.1.4 The steady state reactor

This reactor phase 1s in fact the main object of the work covered
in this report, the steady state fast breeder reactor having the
following featurses.
TITLE
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TITLE
DATE

UTHOST=5(1,

lable Z.11

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# £ & # & % & & # B X & B B % € £ & # ¥ X & B & B ¥ £ B & #H ¥ B 4 X R & ¥ X % &

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T IS B R RS ERLE SRR REE R EENRE SRR EEESEE AL SRR R E R R R E SRS R R RN ERR FU

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WHI RE

PR = FLOW FRO2M [WE ORE CF SYNTH, I8 K6 / YEAR
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- all the freshly bred plutonium can be burned in the same re-
actor (the history of the transplutonium elements is neg-
lected here).

- to achieve this 4n s4fu burning of plutonium the reprocess-
ing of the irradiated fuel 1is limited to the separation of
all or just the most neutron absorbing fission products and
directly coupled to the reactor

- for the next generation of reactors the fuel is produced in
the form of denatured uranium, that is an isotopic mixture
of

10.5% uranium -233 (produced in this breeder)
89.5% uranium -238 (from the mine cor depleted uranium stock-
piles)

- for over 80% of the lifetime of this plant, that is 30 years,
the reactor burns its own plutonium and produces the uranium
-233 for the next breeder generation. Tables 2.13 to 2.17
glve the corresponding data for this reactor phase as before.

2.17.5 Comparison of the three phases

In spite of this rather over simplification it seems that the
praoposed reactor design 1s suited for burning:

uranium -233 in th first phase

uranium -233 and plutonium in the second phase
plutonium in the steady state phase.

(see Fig. 2.3)

As a first approximation the timetable of such a reactor will
be as follows

- doubling time approximately 4 years (including the out of core
inventoryl.

- transient pericd (going from U-233 of plutonium] approximately
2 years.

- steady state period approx 30 years including shutdown periods
for the exchange of the gpre vessel etc. See Table 2.18
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BN g RS R K WS W e Wk K ok ok W W o kW b e g o R ok o R R K

«COMMEMT

*FASE, N*SIGMA FISS,FERT / N*SIGMA F1SS,r ISSIONARLE,
*AHPAR, N«SIGMA PAPAS / lNeCAPT (FFRT + r15S),

TOTAL BREEDING

HeyTRONICS MIcP

S 2 N TR R R RN

SXUR0E

RPENLRATIO MICD

BALA3S
BALAJ®

MACR 8BRTOTY

*| EAG, LEAKAGE /
¢PRODUG,
sFAST RONUS,
‘AHSOR.PARS,
*LFAKAGE, LEAG#ZINIM /
*RRTOT, M«SIGMA THRAMSMUTATION (TH =+

*CAPT,
ERAKE R IR R E N R AN TR R g g R R B RN Aoy BN RN R kRSN & k%

IOTAL AFSQR.(FROM
(FASH =
ABPAR / (1 « ALF),
(1 + ALF),

(113 « PU9 « Pyt),

e 4+ PLD)

VAaLUE

T.187
3,020
2e454
2,515
2947
2,993
200
L 455
1173
» COH
1,501
AT
REY.
144
1,951
5,371
5,666
630
137
5,402
-, 495
1.581

MaX, FLUX IS 1 2,84C+15 FOR GPOUP 14 N INTERVAL

Tar| g NEUTROMICS).,
(INIM e 1 « ALFM) ,/ (1 + AlLr),
(MNIE = 1)) / (1 +« ALF),

/ NeSTGHMA,

1

<

n
»
L]
-
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%
L]
|
»
¥
*
L
w
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HTHNgT=302,N0P;,

L ERFEREEE N D R R QS gy

TITLE
DATE

SPHERIC GEOM,
R N B

*m ®wa

T T T N

» &
. cnrrzy SXHBOL UNTX MALUE *
* .
* GEOMETRY *
| »
* CORp RADIUS Bl M 2954 "
* WALL 41 THICKNESS THICK {2y M 0410 .
. PLANKET THICANFSS THIcK(Z)y M e 550 *
» FALL 2 THICKNESS THICK{4)y M e 040 *
* HETGHT 77 SYLINDER  HEIGHT M w000 *
» HETGHT/ZD 0y wATIg  YDRAT - 000 *
* VOL UMD OF (Owe VOLCOR Mw g 3 T.633 *
. VOIL UME OF RLanNF[T VoL el Moyl 14.299 *
b L
* POMIR ®
" »

POWER 1M 10w AR IS Gw 2.722 »

POWrR In HLANKET POTNK Gw 063 *
» TOTAL Powee FOWRTO G 2.765 %
* FOWER JENS, QURD rorco OW /My e 752 *
“ POWED OBRENS, =LANKET TODEH CW/Muncz 2006 *
* POWER RATING @ pyy "OLRAT MW /K0 2.570 *
L »
* THYENTORY *
% *
" TOTAL PUZ3¢ INVENT, "STQOT KG 736,542 *
" TOTAL PU INVENT, cyTe T Kg 1063,452 *
e (223 IN CORE var KG 1.035 *
* U232 IN RLANRKRET v Keg 47 .3C7 »
. TOTAL U238 INvEINT, VRTOT Kn 6763.900 -
» TOTAL TH INVenT, THTAT K 23551 ,7E9 »
* FPATIO U/ (Pu=ti33y TATUS - 6362 "
L3 *
. THEN GaTa *
] *
* PR . AT IO fwildy T URMR - 50963 *
® LU TN T Mk T Ivk YE ARG 1.984 *
. PEAN FLUX COKRE fLMEAN NEHT/S«CH2 1,460E+16 *
"« *
L EEREEEEELEEEEREEREEEEEESENEREEEEEREE RS FWCECENEE RS FEEEE R R R
IS RN FFTEEAEEEENEER RS S EEE RN REEESER AR N FFESEASSENEEEEERE RS B BV

* TARLFE *
* "R RN R N N *
* MATERIAL RALANCE OF THE BnEFNnFR SYSTEM *
* INCLUDING NEW REACTUR *
PRI YR E AR R R R R R E SRR TR SRR E RS R TS SRR SRR RE SRR R S
* " THOR . pA - uesad * ur3e - PUZ® *
A L I A  E F E F E N T e A R R S Y R LA RS2SR X R TR I M)
*0OKS3 1.26E+04 AeBICT2 6,7 E+nn Ae04F+03 6,/ OF+(0>*
* Sl 2. IBE+N4 R,680+"1 4,730+ "
*FHL Q.55E+00 P417E+70 1,540+04 "
*FTR 6.83E+12 Ay7Al+r2 6e6OFE+02 .
»S0 1,03E+0n e 76E+03 7e37E+0n
¥FEP De78E+02  7.34E+00w
*FCR 6eP4E+2 Ged47E+02 *
*F LR 2e30E+ D B e68F=r 2 4,73C=n" 6e76E+00 7¢37E=04"
aSii 2.30E+0)0 5 6B =t ] 1.03F =07 te76E+00 7e37E=04 =
*FwA feF4E= 31 6,500 ™04 Hel6E=04 7:37E=0>2*
*E AW {.19E+;4 6,56 +n> 2.c90+02 *
#S MW DG 30E+i1 4 1,010+ et 3E+NT "

g L R s A A T T T R S IR A R T IE E Y

WHFRE

ORrRS
SHL
FRL
FTR
SC-
FF#
FOw
F ik
Skt
FuA
F M
SN W

" 8 R 3 0N

1 n

STRADY
FLOW
FLOW TO

FLOW FROM
STEADY STATE IN THI
FLOW FISS.

STFADY STATE IN
FLOW FISS.+ PARASITIC
FLOW TO THE CORE OF [FLANKET IN g , YEAR
FLOWTO REPWUCESS AMD PETURE IN
STATE IN THI
TO THE
THE

I THE

I QRFE

THI

NASTD 1IN

MEW TN KG
NEW AFTER DT Inr

I

OR SYNTH, IM
GLANKET IN «n,
+ PARAS.IN BLANKECT
FLOW FOR TRaMSMUTATION

o

/ YEAR

KG 7/ YEAR

I FG 7/ YpaPp

CorRE IM Kg

N CNRRE

FC=PROCESS1Ne
K G
/ YEAP

/ YLAR

It

TN OKG

/ YEAR

/ YEAP
IN KE
Fig. 2.3 Fissile atom concentretion in the core

R = constant
P * constant
(:E:) (:E:>
} nn*%&?&1w'
Yy \
NOPU
(V17]
B 5.73 -
___—_—__—_—_ 5.68
. s
-
H3
2
1
T T T T T T T T T
0 0.5 1.0
Pu23§ Pu239 430 kg
Pu241 U233 499 kg U733_800 kg
I« 843 kg T+ 839 kg L - 800 kg

Fig. 2.4 Thorium and uranium-236 breeding
{NOPU-21)

P
R
g

L : ///// Bybedi in -23

%

1 /] /
U'I i < r

U-238/Pu-2%9

ratio in core
Table 2.18 Core Diameter = 0.95 m

Power = 2.8 GWl{th)
Build-up Transient Steady State
A B C
Figsile - U233 U233, Pu233 | o 5ag 4 o4y
+ 247
Core inventory ke 800 409, 430 840
Fertile - U238 U238 U238
Core inventory kg 6864 6978 6763
Fiasi]
2227 (enrich- % 10.4 1.7 11.0
Fertile
ment )
Hreedi i
reeding ratio AR 1.34 1.45 1,58
total tot
Mean reutron Flux [1nte DELErons 1.26 1.33 1,06
cm2s
GW(th _
Power density (Lherm) 0.75 0.745 0.75
m core

It should be emphasised that the era when fusion reactors are
raeplaced by other energy sources and hence the plutonium burners
are shut-down, has not been dealt with.

bven here it is conceivable that a reactor design could be procoe-
cd which only consumes and thus be used in this shutdown phase.
This however has not been calculated.
2.2 Impact of some parameters on the mixed zone two zone fast
breeder

In this section some intermediate results are presented concern-
ing the influence of selected parameters in this design, and giv-
ing the breeding capability of a two zone fast breeder reactor.
(Table 2.19)

The influence of mutual displacement.

T(Pu -239) + T(U-233) = 9 x 10°" (10°*atom. cm~ )

by five steps from A to £ 1s given in Table 2.20 and Fig. 2.5

and Z2.6.

The influence of the thickness of the external blanket, which
contains thorium, on the total breeding ratio and the volume of
the core for a given case of the following concentrations of
fissile nuclides

PU-239 3.5 x 10°% (10%%atcom. cm™?)
U-233 10.0 x 10°% (10%“atom. om-?)

is given in Table 2.20.

For the same case the partial BR's and BR total are shown in

i)

Fig. 2.7 and Fig. 2.8
In addition the followling 4 reactor designs have been calculated

having the following arrangement

Core Pu-239 1.1 x 10% —> 1.8 x 10" (x10°*atom. cm™ )
U-233 3.4 x 107 %— 5.5 x 107" (x10%*atom. cm ?)
1-233/Pu-239 = 3.1

Blanket thickness 120 cm.

The data are given in Tab. 2.217. Other results are found in
Fig. 2.9, 2.170 and Z2.117.
Tahle 2.189
OBJECT: Plutonium burning «4n s4tu: some parameters.
REACTOR TYPE : Power, Breeder,
GEOMETRY : INTERNAL ZONE : Spherical core,
WALL : v 4 om
INTERMEDTIATE ZONE -
WALL : v 4 cm
EXTERNAL ZONE : Blanket: 100 cm
WALL, REFLECTOR : Iron,....
POWER (CW thermal) 3 GW
POWER DENSITY (GW therm/m®CORE)
NEUTRON FLUX, MEAN (n/cm?s) .2
FISSTLE NUCLTIDE - Core - PuZz39d
: Out blanket - U-233
FERTILE NUCLIDE Core - UZ38 0Out, Blanket ThZ32
DILUTENT : Chlorides

COOLING SYSTEM : External to core
BREEDING RATIO :

PARAMETER STUDIED : Ratio UZ2368/PUZ3Y in core
: Ratio PUZ38/0U233
FF Concentration
Wall (Molybdenum) thickness
Hlanket thickness

METHOD OF NEUTRONIC
CALCULATION : see 1.10 (page 17)

Two zones
Nvobh o em
Table 2.20 Mixture of Pu-238 and U-233 1in the core

Pu/U3 Ratio Py gnly PU g U3 | o 5 U3 Pu S U3 3 Enly
Fission/sec 2.74x10%0 2.6x1029% | 2.51x10%°% | 2.41x10%°%}  2.32x207°0
Pu-238, core, 3 9,107" 8.10:2 4.10:: 2.10:: 0.0 3y

U-233"  x10%%at/cm®)| 0.0 3.10 5.10 7.10 0.10
AR total 1.76% 1.64 1.56 1.49 1.42
Partial BR U/Pu 0.353 0.312 0.289 D.271 n.253
Partial BR Th/U-233 1.408 1.33 1,27 1.21 1.16
Flux tot. (ncm ?s ') | 2.05x10'¢® 1.99x101§ 1.94x10'% ] 1.86x101%) 1.76x101°
Radius, (cm) 67.28 56.22 65.72 65.5 £65.8
Volume, (cm) 1.28 1.22 1.19 1.18 1.0
Spec. power (GW m~3) | 2.34 .48 2.52 2.54 2,57
Ratio of partial 3.99 4,76 4,39 4.48 4.48
breeding ratios
“ission ratios %igggfi 0 0.89 2.28 5,40 =

Fig. 2.5 Thorium and uraniue-238 breeding
(NOPU~-22)

BR 1

n

.,
w
W

NV

27/SNNN\
1%, NNNINNY,
NN

- SN \
N Ny N E
0.0 P T
Pu-239 9 8 7 6 2 1 0x10"*x10*%/cm’
U=-233 0 1 2 3 L] 5 7 8 9x10"*x10%*/cm?®
Approxunatelly
Fig. 2. Radius of core versus radius of blanket
(NOPU-23)
Py = 3.5x10°° 2 ,
T 20.0“0..) x10?atom/em’®
13
p
1.3
4 Breeding
1.2 ratio
4
1.1 4
BH tot
1.0
¢-9 A Radius of
core, cm
0.6 - 4 50
e 8
-
< "1 Radiusjof
core LT3
v.7 -
LL)
< N2
U.6 40
T
9 20 40 00 8o

Radius of blanket, cm

Fig. 2.6 PRatio of breeding ratios (NOPU-22)

Ratio

Pu-239:

)

]
|
|
/

G

Fig. 2.8 Partial breeding ratio

(NOPU-23)
Pu = 3,5x10"°
uzs =

Ratio of Breeding m]tio I !
T h-233/0}238 {
Ratio qf
Burni
ratio | ¢
e Ul-233/Pu-2
-1 Ratip of
’ Breeding
4 ratio
/ U-233/Pu-239
e !
" B [+
/ -
* T T L 1 T T
9 8 7T 6 5 L} 3 2 1

10 OXIO"} xlO"/om/cm’
o/

-
.
E BR Uranjum
4 =
T
0 20 40 60 80

Radius of blanket, cm

Table 2.21

47 -

Influence of Pu/U33 ratio

FarameLers A B C D
U-238 1.0 0.85 0.85 0.85
PU-739 - 3 1.8x10"" 1.5x10"" 1.3x10-" 1.4x10°"
0%%at |
pu-24g X107 atom/cm 2.5%10- % 2.1x10- 5 1.8x10-5 1.56x10-5
U-233 5.5x10°" 4,7x107" 4.0x40"" 3.4x40°"
U3/Pug ratio 3.08 3.13 3.08 3.09
BR total 1.5 1.51 1.52 1.528
Partial BR thorium 1.12 1.04 0.85 0.852(see fig.2.9
Partial BR uranium 0.39 0.47 .564 0.6/77
o breeding ratio 2.37 2.21 1.68 1.26(see fig.2.10)
Flux tot (nem-%s-1) 1.3x10%°8 0.92x1018 N.62x10t6 0.463x101°
Radius of core, cm 78,06 4941.4 10/7.7 122.0
Volume of core, m3 Z2.03 3.7 5,23 7.6
Spec. power (GW m™3) 1.48 0.938 0.574 0.39
Thickness of blanket m 1.0 1.0 1,0 1.0
Volume of blanket m3 80.6 93, 4 109.7 124.0
PuU-39 in core, atoms 3.65x10%° 4,8x10%°8 £.80x10%¢ 8.37x10°%°
mol 509 800 1130 1400
kg 146 191 271 333
J.33 in core, atoms 1.12x10%7 1.5x10%7 2.09x1027 3.50x102%7
mol 1860 2510 3490 431G
kg 434 584 512 1000
Th in blanket, atoms 1.01x10°°2 1.21x10%° 1.49x10°° 1.806x10%°
' mol 169000 202000 248000 203000
g 39200 45800 57600 58000
J-238 in core, atoms q,54%x10%7 1.5x10%8 2.46x10n%8 3.568%x10°%8
mol 15300 25100 41000 596000
kg 3780 5970 a750 14200
Th-32
: T inventory ratio 10.6 8.1 B, 0 4.4
Jurning rate
Py-239 in 107 atom/s 1.44 1.39 1,39 1.35
U-233 in 10'7atom/s 5.08 5.1 6.08 6.09
U3
Ratio-ga burning 1.30

Fig. 2.9

Partial dreeding ratio

(NOPU-24)

-238
Concentration ratio Y-233 5 3.1

. BR total

Pu-2139

1.5
1.0 7
4
BR
4
0.5 =1
4
4
p
0.0 v ’ ’ v
4 & i 7
: 20 -3
concentravion of fissile, 10 atom cm
Fig. 2.17 Ratio of burning rate snd ratio of
breeding rate
5"
U-~2
Ratic of burning rate Pu-239
4 od
37 A
Hatio off r;me-ding“M
- PR g
L | Fo=23g
. Fu-2 3¢
PRt
- !
o |
i !
3
|
i
1 !C B A
i
0 T T T
4 5 T
20 -3
concentration of rissile, 10 atom cm

—4
Fig. 2.11 Impact of aome selected parameters
(NIOPU-26 reference)

2 em wall
S
Mo + Pe

1.u5 =

referengs

1.60 <

smaller

1.50 — RS
blanket

80 cm

20 times more ¥.P.

2.3 Material balance of the state reactor

The most important feature of this reactor is the (n s4fu
consumption of plutonium. The studies show this to be feasable.
Fig. 2.12 gives the resulting material balance for the steady
state reactor.

2.4 Conclusions

The molten chlorides fast breeder is ideally suited to the (-
s4tu burning of plutonium matching the current requirement for
a "proliferation safe” concept.
_50_

Fig. 2.12

S S

ORES

L L L

12600

4240 kg

N *
N \
23600 |» N\ ' ‘
Th 9; 388 BLANKR ? o @ o
N L6760
560
Pa) 9.17 & tfpisgf : 27
N 734
U3)15.1 N\ 757 |- —
RPN N I P B :: |
i ) t: chE AN
: N\
N

WASTES

Th: 23600 kg
U3: 1310 kg
Ug: 6530 kg
AFTER
DOUBLING
PERICD
3.7 THE THREE ZONE REACTOR

3.1.1 Introduction

The reactor discussed now is rather unconventional because of
its three zones (see Fig. 1.2])

- internal blanket zone
- fuel zone

- external blanket zone

This concept has been compared with the more conventional type.
Holding most of the parameters the same the breeding gain comes
out about equal. However, one large difference is that although
the total power 1s the same the specific power changes by more
than one magnitude being higher in the conventicnal central fuel
zone reactor. Also the mean neutron flux increases from

1.7 x 10'%n em™%s”™! for the non-conventicnal central blanket zone

to 2 x 10'"’n em™?s~! for the conventional central fuel zone.

Since the specific power and intensity of neutron flux 1s clearly
a major pronlem form the point of view of the englineering design
of the reactor (cooling, radiation damage of structural materiasl

and fuel), both systems have been studied, that 1s without a cen-
tral blanket region and with a radius up to 110 cm. The results
are given below for fuel without uranium and with uranium

fuel for both cases: no internal blanket zone and with an inter-
nal blanket zone.
3.1.2 The three zone breeder with thorium/uranium-233

The first case is the fast breeder molten salt reactor with
uranium-233 as fissile and thorium-232 as fertile material
with a three region layout.

- internal fertile zone 1, 50, 90, 110 ©m
- wall 2 cm
- spherical shell core 19 cm
- wall 3 cm
- external fertile zone 110 cm
- reflector 40 cm

Table 3.1 gives the main details of the reactor.

L)

.2 A Three zone breeder reactor with a mixed fuel cycle
U-238/Pu-239 plus Th-232/U-233

The next step is a study of a mixed fuel cycle in the three zange
fast breeder reactor. This problem has received much attention
and the published papers given in list D should be referred to.

A three zone fast breeder reactor with the geometry shown in
table 3.1 has been calculated.

The range of variations covered include

Pu-239 or U-233 in the core
U--38 Th-232 in the blanket

For a given geometry and concentration of ertile and fissile nu-
clides the influence of the fission products, when the concentra-
tlons are increased by a factor 10 results in a reduction in the
breeding ratio by 5%. An increase by a factor 10 in proactinium
decreases the breeding ratio by only 2% (Fig. 3.4).

For the three zone reactor calculations have also been made far
mixed fuel cycles. For the fissile materials Pu-239 and U-233
and for fertile materials with U-238 in the core only and Th-232
in the blanket only.
Table 3.1
OBJECT: Three zones thorium cycle
REACTOR TYPE : Power, Breeder
GCREUOMETRY « INTERNAL ZONE :  Spherical, fertile material
(™) WALL : Metallic
INTERMEDIATE ZONE “hell; fissile material, active core
WALL : Metallic
EXTERNAL ZOMNE : “hell, fertile material
WALL, REFLECTOR : Metallic

6
A6 - 9.hY

10

POWER [(GW thermal) 2.
POWER DENSITY (GW therm/micore) : 0.
NEUTRON FLUX, MEAN (n/cm?s) v

FISSILE NUCLIDE : Internal Blanket
: Core - Pu + UZ33
Outer Blanket U233
FERTILE NUCLIDE : Internal blanket Core U238
OQuter Blanket - Th 237

OILUTENT : Chlorides

CODLING SYSTEM : External

BRECDOING RATIO < 1.08 - 1.14

PARAMETER STUDIED Thickness of blanket

Thorium concentration in balnket material
FP and PA concentration

METHOD UF NEUTRONIC @  ANISN, 54
23 Groups
CALCULATION : 060 spatial positions
P1 approxim.
GGC-3 code

ENDF/B-1 and B-2 Dats
Table 3.2 (THOC-300)

Three zone reactor. Volumes and breeding ratio

Geometry: internal blanket, radius 110 cm

wall, width 3 cm
core, width 27 Cm
wall 3 cm
external blarket, width 100 cm
wall reflector 140 cm
/ariable Results
Internal | Internal | U-233 concen. Yolume Specific Oreeding
Case fertile wall in core of core DowWer ratio,
(cm) (cm) (107" /em?) (m?) (GW m~3) total
B 1.0 1.0 0.0012 0.783 3.36 1.08
C 50 2.0 0.0012 1.66 1.57 1.10
D 890.0 3.0 0.0017 3.73 0.70 1.13
E 110 3.0 0.0017 5.67 0.46 1.11
A 3.0
roference 110 0.0C18 3.13 0.80 1.16
f: 110 3 00,0003 9,57 N.28 1.4
see Fig. 3.7

(THOCL-300)

Three zone reactor: ataomic composition (atomic concentration x 102%)

Internal External

fortile zone Wall Core Wall Dlanket Reflector
o 4.5x107° Fe 7402 | Th 2.5x107° the same Fe /x107°
L 1x10-" Mo 1x10-7 Pa 1x10°" as inter- Mo A1x10-7
=33 1x107" =23 1x1n-? nal
c1 2.2x1077 -2 axant® nlanket (remark:
Ha 4.0x107 3 wrong
variable variable Fe 7x1n~7° reflector
radius 1-233 con- Mo 7x10° 2 with Mo!}
(see Fig. 3.1) centration

(see Table

abovel.

.1 Impact of the radius of thre internal

55

olanket (ThCL-300) on the breeding ratio

1,10 -
¥ | BR * breeding
' ratio
o il
] y
1,05 "/c i
C
1,0 T T T 1 T T
[+ 20 &0 100{ 120
internal blanket rgdius, cm
&
calculated cases:
A, B, C, D, E
Tpecific 9
rower
(cwem™ )
1 c
5 —
4
2 -
E (more U-33)
4
J
1A l A (reference)
1 ! (less U-33)
[¢]
Pip., .3 Uranium concentrations for given geometry
(THOCL-301) for A, B, C, D, E, F, G, H, T,
1.164 Th mol ratio
p
0.6
4 o 1
0.5
1.15 D r
1 A g pe
BR 0.4
<
-
1.10
Power
rating
1.5 J
-
- 0.60
1.3 >
specific specific -
power,
P e ] | 0.50
= kg U-2
| MW tot
- - owe
1.0 JNFning L ob
e 4
-
0.8 4 I
0.7 — ——r—r 0.3
0.17 0.21 0.25%

U-233 mol ratio

of the
Rate of breedins

Pig. 3.2 Concentratign of thorium and uranium-233
(in 102:7“3)
{THOCL-301)
0.0048 ] 4
H
mol ratio
T™ = 0.6 1
- 4
0.0040] P—
E
T™h d
ol ratio \ ’
a ™ * 0.5 b
0.0035 4 1
- mol ratio
Th = 0.4
- \{
6.0032 T T T T vr%
0.0010 0.0015 0.0019 J.0021
Uranium in core
Pig. 3.4 Three zones, mixed fuel cycle
Impact of fission products and protactinium
{THOCL-305)
Rate of breeding
4
1.55 4 J ~
ratio
4
1.50 +
st Paom
1.45 ~
= 0.18
1.%0
1.35 o
4
1.30

10 times reference
less PP FP = 1
Pa = 1 Pa = 1

Table 3.3

- 56 -

(THOCL-302)

Influence of FP and Pa concentration

gecmetry

internal wall core wall external
blanket balnket
110 cm 3 cm 19 cm 3 cm 100 cm

reflector

40 cm

Fission product Protactinium concentration Breeding ratio
content corresponds to dwelling time,
Case days.
1024 -3
* e 2 5 18
A 3 x 10°°, 1 x 107° 1.11¢
B dwelling time = 3 x 1077 1.124
C 3 days 9 x 10°° 1,123
5 x 1077,
0 dwelling time = 1 x 107° 1.1238
10 days

fon)

S

also Fig. 3.4

Remark Tabhle 3.4

cance led

|
[
~J4
I

ihe influence of parameter variations 1s given in Fig. 3.5.
The following results can be noted:

the breeding ratio increases (under the given conditions) when
- the concentration of U-233 in the core decreases and U-238

increases

- when the internal wall thickness 1s halved (note this is very
sensitive due to the presence of 15 atom% of molybdenum)

- when the outer fertile radius increases

- when the inner fertile zone radius 1ncreases.

3.3 The three zone reactor - uranium-plutonium fuel cycle

The reference reactor i1is described in table 3.5 and 3.5. The
thermal flux in all three zanes, the external breeding zone,
the fuel and the internal breeding zone is aonly 10°°% of the
total flux and in the external hlanket reaches 10°% of the to-
tal flux. The total flux has a relatively flat distribution and
even in the fuel region the max. to mean ratio 1s only about
.13 (Fig. 3.6 and Fig. 3.7).

The neutron flux 1s rather hard and the mean neutron energy
(calculated as the mean of the no. of fissions) is around

370 keV {see Fig. 3.3). In a typical LMFBR and in a gas cooled
fast breeder this value is 120 keV and 176 keV respectively.

A pood illustration of the influence of the most iImportant
papameters aon the breeding ratio is given in table 3.7. The
differences between these calculations and the computur cut-
put is approx 8%.
Fig. 3.5 Three zones; mixed fuel cycle
Impact of selected parameters on the breeding ratio
{NOPU-1)
wall 1.5 cm
h K ———
1.7 o

greater ouf.

D gblanket
125 om
BR ] greater inner
covmm—
p P4 vlanket 150 cm
) T™/U in
more U-238 in core, o _
1.6 inner blanket
no U-233
more U-238
1 C erm——

half of
i u-233
reference
1A

out blanket 100 cm
wall 3 cm
1.5 M inn. blanket 110 cm

Fig. 3.6 Three zones: uranium-plutonium fuel cycle
Neutron Flux in reference reactor

fotal neutron llux L~y
10t d
1647
Fast ne
(~370 KeV)
101"_
(Neutraqs)
cm< sef
1015.'
1012
Internal fertile
zone
wall
11
10%% ] el
pond
1010_ WLl
External fertile
zone Reflectdr
1Y ‘
b ‘nermal neutrons
o flux
IUU -
10 T ég.r Tt
, 20 40 60 B0 11)" 14C 160 180 200 260

Radius {(cm)
- 59 -

Table 3.5

OBJECT : Uranium-plutonium cycle - Maximum breeding ratio in
three zones

REACTOR TYPE : Optimised breeder
CEOMETRY :  INTERNAL ZONE :  Spherical fertilile zone
(M) WALL i Iron, Maolybdenum
INTERMEDIATE ZONE : Core, fFuel
WALL : Iron, Molybdenum
EXTERNAL ZONE : Fertile zone
WALL, REFLECTOR ¢ Iron
POWER (GW thermal] 1 b
POWER DENSITY (GW therm/m® CORE) : 1.1
NEUTRON FLUX, MEAN {n/cm®s) couinte - oty
FISSILE NUCLIDE Internal Blanket: small amounts of FPuZ3Y,
Core Pu 238, PuZ40, PuZd1, O.7: /.2: 0.7:
External Blanket: small amounts of PuZ3%,
FERTILE NUCLIDE : Internal Blanket UZ38
DILUTENT : Chlorides, Sodium
COCLING SYSTEM :  QOuter
HREEDING RATIO . 1.864 - 1,044
PARAMETER STUDIED : Flutonium-uranium ratio

With and without uranium
Refloctor, Fe, Pb

MEHTOD OF NEUTRONIC : see chapter 1
CALCULATION :
Table 3.6

Three

/ones reactor:

(200/C)

uranium-plutonium fuel cycle.

Radius Width /one Compoailion Flux Specific
cm of atoms/10%“cm’ thermal; power
Z0ne total, GW/ma,
cm Sreeding Temperature:
roa-io
0 I 1-238 6.4x10°°
et Pu-239  B6.0x10-° 1.05%x1016 lnlpt—7naoc
- ) h*S el
0.0 fortile F.P. 2.0x10 3.7x10° .
" one Ma 3.4x10°° 1o SB00°C
: C1 2.27x10°3 BR 0.490 OULLE
110.0
- IT Fo 7x10° 7 1.10x100° 2e0°c
o Wall Mo 1xq0-2 Ix107 R
113,10
I11 Pu-233 1.3x107° g
ny-240 4.2x107" 1. 1GW/m
17.9 Fuel cy-241  2.9x107" 1,02x10L® _ =750°C
o 3 —— i inlet
zone =238 4.7x10°° B,.Ex10
A=E 2.0x10° T . _=1050°0
N 3.4x107 3 RR 0,22 outlet
] 2.6x10°7
130. 9
TV Fo 7.0x1077 8.24x10%°
.0 - r
3 Wall Mo 1.0x10° 7 2.4x10° SR
133.9
v the same as 3.6x101" T, 7ot
Uxternal central fertile 1.0x107 HhLet
1000 fertile “one, 1 _ ZSOUOC
~one AR: 1.040 cutlet
233.9
VI e 8.0x10°7° 5.2x10%7?
40 TrITETT
Hx10
273.0

Fig. 3.7 ‘hree zones: uranjum-plutenium fuel cycle
Total flux in the ruel zone
Internal
fertile \j Fuel zone \*
zone i\ 1
1,0 . \\ '\
1,1 =
Neutron flux
1oke
.2 -
nem s \
1,0 = ::::
- \s
[T \
~
Fig. 3.8 Three zones: uranium-plutonium fuel
Spectrum in fuel zone
lo(é_ T
1041
1029 |
-
4
1019 N
Median
<
<
10i5 L It I i Y " 4
T T L) . T T 1
102 103 10 10° 10 107

Neutron Energy (eV)
Table 3.7 Three Zane reactor: uranium-plutonium fuel

Simplified calculation of the breeding ratio and neutron balance.

Component of
Parameter . .
breeding ratio
Median energy Kol 370
Pu-239 of{barn) 1.83
(from computer output) 0. {barn) G.180
V 2.95
o 0.0984
Nreooding potential (n-1) 1.6857 1.685
Fission ratio
S 0.37

fertile/fissile

(v-1)
Fast bonus 0.539 0.538

1+
Total positive 2.275
Losses (sbsorption in

0.160
P, C1, Na, Mo, Fe) ‘
Leackage (arbitrarily]) 0.10
Loss+o

Total losses = 0.32 0.320

1+0
Calculated BR {(micro) 1.830
Computed BR (macro) 1.752

- 63 -

3.4 The three zone breeder reactor: Very high bresding gain

Cne of the most important factors in achieving a very high breed-
ing ratio i1s the hardness of the neutron flux. This is strongly
influenced by the fuel compositicon.

In this case the fuel is postulated to be a mixture of

a PuCly - b NalCll - ¢ UCL where
a = 0.1 - 0.2, b = 0.7 - 0.8, c = 0.1 - 0.2

Untortunately not all data are available for this system. (see
Fig. 06.9)

The rough calculations on changing the concentration of PuCl, in
the melt of NaCl (Fig. 3.9) shows a rather sharp decrease of
breeding gain (BG) for decreasing plutonium concentration, espe-
cially when the plutonium molar ratio to sodium is lower than
025,

In spite of these uncertainties of the PuCla-NaCl-UCly system,
the influence of the U-238 in the fuel has also been calculated.
For a constant PuCly concentration with a simplified assumpticn
for the NaCl concentration the results are given in Fig. 3.10
and 3.11.

Increasing the ratio of uranium to plutonium in the fuel from

0 to 3 causes the total breeding gain to increase from 0.55 o
.95, This 1s rather clear and thus the reference reactor con-
cept includes uranium In the fuel in a ratio of 2 : 1 to pluto-
nium.

Such a high breeding gain is a special feature of this type of
reactor for producing large quantities of fissile material.

Fig. 3.12 and 3.13 give the results of calculation when the ra-
dius of the central fertile zone is varied. Table 3.8 shows the
simplified calculation of central and external fertile zone
brecding ratins.
- 4 -

fig. 3. Impact of Plutonium Concentration Fig. 3.10 Three zones: uranium-plutonium fuel cycle
Impact of U-232 concentration in the fuel
Pu-Concentration: 0.0021x102* atom/cm’

ouD g
Pucla 1 -
NacCl
700 -
Tempegature
ompeyatu 0.9
(c)
600 - Breeding
gain
500 0.8 4
400 NS N S MO SEN SR
1.0 0.7 0.5 0.3 0.2 0.1 0.0
Mol ratio of plutonium 0.7
0.6
0.5 =
Specific 0.6
breedfng power
gain (kW/cm”)
). o
o 1 U/Pu ratio 2 3
L 1 i L
0.8 . 4 —i L— T—
0.7 ) 1 P4 3 L 5 6
*-¢ 24
Atom U in fuel / (0.001:10 /cm,)
0.2 0.5
0.7 0.5 0.3
mol ratio of plutonium
Pig. .11 Three zones: uranium-plutonium fuel cycle Fig. 3.12 Three zones: ursnium—n%utonium rue} gele
" Impact of Uranium concentration in the fuel Impact of central fertile zone radius
(202/201) versus central fertile zone (No U in fuel)
1.0 o Uranium 0.y
Concentration x10*!* at/em?
o Breeging
‘vith u "5_3 103 gainjtotal
0.5 o
Core
0.9 vl
1 p 4.2 10 Core 28.0
0.7 ] 133
. Breeding Gain
ore °
L 16
0.6 =15
0.8 o
3,15 16° 1
Breeding gain -13
-2
¢ -3
2,1 10 <11 Spec.,
4 power
0.7 Kw/cm™3
) 0.0
‘Mm ¢
0.6
Core 200
5 Core
Ve
o 110 cm T T T M M T v T M Y T

9 10 20 30 40 50 60 70 89 30 100 110
Central fertile zone, ratios (cm)

Central fertile zone, radius {cm)
- bh -

From these results the following conclusions can be drawn

- 1increasing the raius of the internal fertile zone up to 110 cm
increases tLhe breeding gain for a given type of fuel. The ef-
fect of wall and fertile material changes are insignificant

- at the same time the specific power decreases dramatically

- increasing the U/Pu ratic from 2 to 3.6 does not influence
the total breeding gain (see Fig. 3.13)

- the FP concentrations play a rather significant role (Fig. 3.74)

Table 3.8
T i
Internal fertile External fertile OLa?
Fuel zone breecing
zone zone ,
ratio
Case Pu-239 Uf, qU Pu—239jc Pu—241F Pu-2349 U{_ ]
(Number) iss cap f fiss cap
OXV OxV Ox\V OxV OxXV Oxw
three
zones | U.14 10.308 0.8 3.05 0.351 0.30 0.47 1.83 1.70
(200)
two
ZOnes - 3.63 0,367 .04 0.46 2.50 1.53
(180)

The influence of the 40 cm reflector if changed from iraon to lead
is not very great as shown in table 3.0

Fission product concenfiration however plays a very important role.
For a given reactor design and given fuel and fertile compositions,
increasing the concentratlion of fission products (simulated here
with Cs-133 only) from 2x107° to 2x10°% (in 10%%/cm?®) decreases the
breeding gain from 0.85 to 0.38 when the specitic power docreases
less than a factor two.
Table 3.9

Central fuel (Core 180)
(wall 2.5 cm Pu =

Three zone reactor:

— 6 6 —

uranium-plutonium fuel

2.1x10°3% x atom/10%%cm?)

Case A B C
Uranium 238 in fuel no VeSS no
4,2x107°
x10%%atam/cm’

Reflector 40 cm: material Fe Fe Pb
Volume fuel x 10°cm’ 2.95 Z2.40 2.97
Spec. power in fuel kW/cm® 18.4 23.0 18.3
Hreeding ratio total 1.54 1.394 1.66
Tot T Z20Ne 2
y 2l Tl of fuel zone 1.18x1017 1.25x1917 1.187x107
n/cm’ s

Three zoves: ararium-plutonium fuel
Breediry ratio versus radius of
internal fertile zone

Fig. *.1°

Breeding

T
50 100

Radius of internal rertile zone, (cm)

Fig. 3.14 Three zones: uranium-plutonium fuel gele

Impact of Fission Products Concentration
in fuel

(very simplirfied, from different calculations)

J,7 -
breefing
gain
B.G. B
U,b
S.T7.
(kw/em?)
=3
0.5
-2
S.P.
0.b
dwelling time of fuel in ¢ore (days) i
v.3
t-2 ~10 "20
T .
107> 127

Concentration of F,P, atoms -« 10° /cm
- 6/ -

In the steady state reactor a concentration of 2x10°° atoms
F.P. x 10%%/cm?® for a fuel having 2.1x1073 atoms Puxi02%/cm?
is reached for a specific power of 2 GW/m® after a time t of

_(2x107°) x 102" ) .
b STy (5onio) w2 C 1-61 x 107 s

that 1s after 1.87 days. The higher fission product concentra-
tion - that is 2x10-* corresponds to 18.7 days of mean dwell
time f or the fuel in the reactor,

The influence of chlorine-3/ separation may now be looked at.
The influence of each absorber on the breeding ratio is given

by

- A+ D+ L +
T+ o
B = decrement of breeding ratio
A = absorbtion rate in a given absorber

absorbtiaon rate in rest of absorbers

J
1

L = leakage

o = Oc/ojc

It can be postulated that for a strong absaorber in a hard [(fast)
spectrum that

A - 0.15
0 + L - UJ.15
o = 0.15

The relative influence on the rather high breeding ratio of
1.6 results in a case where the profit of the separaticn fac-
tor will be for example 0.2 then

and In relation to the breeding gain

AG —4—7 = 0.20
- 606/63 -

which results in an increase in doubling time of

T2 )

NI

= .83

It can be seen that introducing three zones does not result in

any significant increase in the breeding gain

(table 3.10}).

Therefore the two zone reactor must be preferred for its simpler

layout.
Table 3.10 Fuel in central zone (2 zone reactor) versus
fuel in middle zone (3 zone reactor)
iwo zones Three zones
Core, Case Nymber {conventional) ‘nonconventional

(1230 (2nn)

eometry Central Zone Fuel 100 cm Slanket 110 cm
Middle Zone -—- Fuel V18 om
Outer Zone Blanket 100 cm

Thermal power, GW 6 B
PLU/FP ratio 2.1x1073/2x10°°% | 2.1x10-3/2x10° ¢
Spec. power, kW/cm® 17 .7 1.41
Power in fuel, % 90.9% /B.2%
Flux total left 2.04x10"7 1.2x101°
‘ , boundary ————
in fuel right 1.15x1017 1.08x1016
Flux in left 8.99x10'°® 9,7x10'°
outer } boundary
blanket right 2. 16x101° 1.5x101"
Breeding gain 0.63 0.70
Median energy (group) g 10

3.5 The two zaone fast breeder. Fuel of uranium plutonium
flucrides

3.5.1 Introduction

The aim of this section is to give a rough idea of a fast
breeder power reactor having the fuel in form of plutonium
trifluoride in the molten state instead of molten chloride.
(Table 3.11)

The earier suggestions for a reactor of this type came from
A.M. Weinberg.

The first attempts at carrying out calculations on a reactor
of this type were not successful because a fuel was chosen
having a high concentration of light metals, lithium and be-
ryllium.

A very rough attempt by J. Ligou and the author (1372) shows
the possibility for a fast breeder reactor with molten pluto-
nium fluoride where the light metals were eliminated and the
melting point increased.

F. Faugeras (Fautenay aux Roses) claimed that the fluoride of
U-233 and Th-237 can be used for non-thermal reactors.

Some preliminary results for a three zones reactor are eciven
in a short form in Table 32.12. The neutron flux remains rathe-
hard (Fig. 2.15].

3.5.2 Arbitrary assumptions and uncertainties

The fuel composition has been arbitrarily chosen since the
appropriate data is lacking in the literature. In most cases
the following fuel composition has bean used

1 PuF, / 1.2 NaF (T i © 727°0)
+ 2.4 NaF /1 Ir F, (T 1 7 510°C)
another alternative would be
PuF, / 2 Naf
CaF. / 3 NaF (T = 820°C)

% melt
Table 3.11

OBJECT Fast Breeder, Molten Fluoride,

REACTOR TYPE FPower
GEOMETRY INTERNAL ZONFE
(M) WALL

INTERMEDIATE ZONE

WALL

EXTERNAL ZONE

WALL, REFLECTROR
POWER (GW thermal) : 6
POWER DENSITY (GW therm/m’CORE) : 3.5
NEUTRON FLUX, MEAN (n/cm?s) : o 1nte

FISSILE NUCLIOE Flutonium.

FERTILE NUCLIDE U-2368

DILUTENT NaF, 7rF,
COOLING SYSTENM Cbornal
HREEDING RATIO 1.38 (up to 1.0

PARAMETER STUDIED
Wall: beryllium,
Fission product

METHOD ON NEUTRONIC
CALCULATION

FLUDRTOE+

Fission product
only as Cs-133

Three zones

Fuel caomposition PuF

wall structural material:
Iron,

recalculated
from Hansen-Roach

Fertile zone

Iron, Graphite
Fuel, fluoride
Iron, Mo, Craphite
Fertile zone

Iron

a7 Naf, 4

with higher specific power)

graphite

thickness
concentration

ANISN
Regions O
Meshes 110
Order of

quadrature 84

Anlstropy P1

23 neutron groups
incl. thermal ENDF
B/111
-~ /7

Table 3.12 Design of fast breeder molten fluoride reactor
(Three zone: 6 GCWithermal]})
Power rating, total = 1 kg Pu/MwWwth, Doubling time = 5.5 years.
Radius Zone Components total Broedine
cm molecules Flux ————— L
per cmix102Y thermal ratio
0 I | Internal | UF, 5x107° 1.6x10'° 0.42
Blanket NaF 5x10°° 5.8x1012 )
liquid Pufg Bx10°° T = 800C
state
80.0
I Wall Fe 7x107 2 1.86x1018 - -
Mo 1x19°2 2.2x1017?
(graphite is
also possible)
81.C
11 Fuel Pu-239 F3 1.47x107% | 1.57x10"° “ertile
liguid Pu-240 T3 4,2x107" 1.00x1016 material
state PU-241 F1 2.1x107" . B N R
NaF 7.5x1073 "inlet O”
Irf, 5.1x1077 T_. T 10s07C) 0,056
F.P. (Cs133)| 0.2x10-° ou
98,2
TV Wall Fe 7x1072 1.37x101 8 -
Mo 1x10-2 5.2x10172
(graphite is
also possible)
100. 2
v | External | U, 5x107° 8.8x10"° n.sa9
blanket NaF 5x10-3 2.5x107T
liquid | Pufjg 6x107° T = 800°C
state Mean
200.0
S 7.5x10%! -
VB2 —
. 5. 3% 106
240.0
Total TR = 1.38

Fig. 3.15 Neutron spectrum in molten fluoride fast breeder
(roughly)
arpitrary
units

14

10°*

107"

————g» Median enargy. 60 kaV

Lethargy
The results obtained, in spite of the uncertainties are en-
couraging. The simplified breeding ratio calculation gives a
BRyy+ of 1.517 and the computed value is BRy,t = 1.465 (table
3.13).

Table 3.14a shows that the influence of fission products 1is
very significant (see also Fig. 3.16). However the change of
structural material of the wall form beryllium to 1ron has
ocnly a small effect on the breeding ratlio and specific power
(table 3.14b and Fig. 3.17).

In the calculations for three zones studies here, 1ncreasing
the radius of the internal fertile zone from b0 cm to 80 cm,
that is a volume increase of 2.3/ has little effect on the
total breeding ratio in spite of changes in the regional
hreeding ratios (table 3.15a and Fig. 3.18]).

Altering the small amounts of plutonium 1n the fertile material
as 3 result of reprocessing efficlences also has only a small
effect on the breeding ratic and specific power (tablie 3.75hb
and Fig. 3.19).

Becouse of the good experience of American and French groups
using graphite as a structural material for molten fluordie
thermal breeders, calculations have been made using grapnite
for gseparating walls for fast breeders. Graphite 2 cm thick
as the wall material was chosen.

The results are rather encouraging. The breeding ratio using
graphite is still very hiph, even slightly higher and the
specific power in the fuel is lower (table 3.10a and Fig. 5.20].

Changing from the complicated design of three zones with fuel
in the intermediate shell, to the "classical” two zone design
having fuel in the central region results in a dramatic 1in-
crease in specific power to a prohibitive 26 kW.cm™? (table
3.1060, Fig. 3.27, tabie 3,177,
Table 3.13 GSimplified calculation of the breeding ratio

Median energy {from computer calculation)

Pu-239

J-238 (arbitrary)

Fast fissicon of ertile component
(from computer output)

Fast bonus
Total positive
Losses by absorption (from computer)
Leakage (assumption)

Total losses

Total negative

fOreeding ratio = total positive - total
negative

Breeding ratio (from computer output)

kel
o  {barn)
a
C
v
o
n
n - 1
\) L}
S
(V-1) x §
1+
Y positive
Lab
leack
L
tot
Ltotfig
1+

BR (macro)

— ———— — —— ——

Tahle 3.14a Three zones reactor: fluoride fuelled fast breeder
reactor

Influence of Fission Products.

BR = Breeding ratio, total

Wall .. o
Fission Specific
Core fuel/external
Numb blanket products (Cs-133) B.R. power
e ,? | atoms/cm® x 1n?" oW/m?
2 ¢
163 Be 0.0001 1.437 4,33
164 He 0.0002 1.427 4,24
166 Be 0.0040 1.14 2.32
167 e 0.004n .17 2.0

Table 3.14bh  Three zonex reactor: flucoride fuelled fast breeder
reactor

Influence of the beryllium-moderatnr

F.P. = 0.0002 atom x 10%%/cm?

Core Material ] AR, Sp@cific3power
Number 7 cm 1.5 cm GW/m

164 e 1.427 4.74

162 Be 1.464 4.19

165 e 1,38 3.45

161 Ce 1.37 3.47

(see also Fig. 3.17)

Fig. 3.16 Impact of fission products

1.5 _
with Be
1.4 -
with B
<«
BR 1.3 -
1.2 - =
wlth Fe
1.1 |
193 144 165
; 166
1.0
[ | T |
104 1073 1072

Fission products

Fig. 3.17 Impact of the beryllium and iron moderator on BR
162
1'45_1 [
Be
BR tot 164
1.40 —
162
164 Be
1.35 —

5
4
spec.
power
kw/cm?
3
2
1
0
and spec.
5
spec.
power
kW/cm3
4

power
Three zone reactor:

fluoride uranium-plutonium fuel

Table 3.1ba
Influence of the radius of the internal fertile zone
& —
Radius of Breeding ratio
of Core fertile fertile fuel fertile total
zone, &m zone zone
165 50 J0.42 g, 056 0.884 1.364
1649 50 0.34 0.053 0,985 1.379
Ratio Ratio of
of cores volume
165
237 1.21 1.05 0.B98 (0.4984
169
(see Fig. 3.18)
Toble 2,150 Thpee —one Teactor: fluoride uranium-plutonium fuel
ITnfluence of Pu-739 in the fertile material
Pu-239 1r Spec.
Number , fi_ H F.FP. B.R. Pec
f fertile concentration tot pOWET
[ ) .
o cor material ou/m’
166 0. 007 2 x 10" 1.49 3.79
1657 .01 72 x 10 1.37 3.477
prrme m— e b e e e e ] — e e e . e e e e . e e i [ mn e emem — —
Ratio
ot corec
15&- 8.1 10 1. 049 1.09
e . , . ¢ . 0¢
(see Fig. 3.19)
Fig. 3.18 Three zones fluoride uranium-plutonium fuel
Impact of radius of internal blanket

1.20—
relat1v1.1o_
BR
BR fuel
1.00
0.9 —
R =160 cm R = 80 cm
T T 1
0 1 2 2.37 3

internal fertile zone volume, relativ

Fig. 3.19 Three zones:fluoride uranium-plutonium fuel
Impact of Pu-239 in fertile material

1.50

BR sSpec
BR power

1.40— — 3.70

Spec
power

1.30—
— 3.40

1072 1072

Relativ Pu concentration; U-238 = 1.0
Table 3.188 Tirrpe zonegreactor: fluoride uranium-plutonium fuel

Influence of graphite as structural materisl

st wall between Znd wall between
Number internal fertile |[fuel and external B.R. Spec.
of core zone and fuel fertlle zone Total power;
2 cm 2 cm GW/m?
164 Fe, Mo Be met 1.427 4.24
171 C graphite C graphite 1.45 3.65
-.———————-—-—r-—--———-—w--—-»—-——-—- —_——— e ——— ——e — — — —— — o —— —— — =
Ratio
of cores
/ hite hite '
i graphite graphite 1.00 0860
164 Fe Be

(see Fig. 3.20)

Table 3.16b Two zones reactor: fluorlide uranium-plutonium funol

Influence of the geometry of the reactor

Structure o
Number B.R obEes
of core Internal | Intermediate External total powEs
zone zone zane GW/m
160 Fuel Blanket Cooling 1.424 26.7
two zone Zone zane
reactor
161 Blanket Fuel Blanket 1.37 3.47
three zZone cooled
reactor out of
core
Ratio
of cores
160
i 1.04 /.69
101

(see Fig. 3.21)
Fig. 3.20 Three zones-fluoride uranium-plutonium fuel

Graphite 1nstead of beryllium as a structural material

1.50 — 5
BR tot spec
power
k'/cm3
1.40 A - 4
1.30 — 3
Fe-Mo, (1st wall) graphite (1st wall)
beryllium (2nd wall) graphite (2nd wall)
Fig. 3.21 Two zones: fluoride uranium-plutonium fuel
Impact of geometry
— 30
1.50
BR tot 5 spec
power
kw/cm3
1.40
1.30
— 10
0

Two zones Three zones
t
o
.

]

Table 3.17 Three zones reactor: fluoride uranium-plutonium fuel

Influence of the plutonium concentration

Pu total
Number ;; ;1;1 Atoms B.R. B.0G. Spec. Power
of Core Mol o cmox 107 tot. gain GW/m’
165 14 0.0027 1.38 .38 3.45
1772 7 0.0011 1.05 05 4 .05
————— —_ — ————— = — e — — — —— — A — e e —— —— ———- — o]
Hatio
of cores
/<
5 .
ToE 0 0.5 0.76 0.13 1.1/

i, “..2 Burner reactor with molten fluoride fuel

Fuel: Pqu-B.Bu NaF-2,41 Z”Fu
3

Zpecific power ~20 kWem
fotal power 11 Gw (thermal)

] 100
98,8 109 112,6

Radius, cm
3.6 High Flux Reactor with Fluoride Fuel

Can a fluoride fuelled burner, as opposed to a chloride fuelled
reactor be considered as a reactor for transmuting fission pro-
ducts? (See chapter 4, for the high flux transmutation reactor).

In such a reactor a fuel made up of PuF3/5.4 NaF/2.4 ZrF, has
been assumed. The calculations have been carried out for a
larger bruner of 11 GW(th) and the fission products are assumed
to be generated in a total system of 55 GW(th).

The results are not encouraging in spite of the fact that the
neutron lfux in the target region was 1.05 times higher for

the fluoride fuelled reacotr as opposed to the chloride fuelled
reactor (Table 3.18, Fig. 3.22).

The effective half life of both fission product nuclides was
(in years)

In fluoride In chloride
fuelled fuelled
reactor reactor

(reference)
Cs-137 8.57 5.93
Sr-90 1.73 1.83

These "benefits” must be balanced against a specific power which
is twice as high as the reference case. In the fluoride cors this
is 19.9 GW.m™* and for the chloride reactor 10.1 GW.m"-?. Such s
high specific power is not realisable.

In addition since graphite might be used in place of beryllium
oxide as moderator (possible for a neutronic viewpoint) a signifi-
cant improvement in corrosion problems is obtained. This has been
proved by the excellent experience of Oak Ridge National Labora-
tory with one proviso - at ORNL the fuel was LiF - BeF, - ThFg -
UFg.
- 8/1 -

Table 3.18 High-flux burner reactor with fluoride fuel

Total power 11 CW(th)
Hurning fission products form a total system of

55 GW(th)
s Neutron flux Specific power
ol Dni ] Components 1018 cm 2571 (GW m=?)
Callué [C?] (atom 107 *cm™?) total Transmutation
oLume A thermal rate (5™ 1)
I
T~ 88.8 cm Cs-137 0.0116
Target zone Sr-90  0.0016 4.01 C5-137 : 1.8x107°
Vol: 4.1 m? Oxygen 0.0145 2.21 Sr-90  : 1.2x10°°8
Deuter 0.0145
IT
A8 - 109 cm Be .060 5,08
“oderator, wall Oxygen 0,060 1.83
with thin 5.25
graphite layer 0,235
111
7.4 - 112.5 cm Pu-2338 0.0017
susl ozone Pu-240 0.00042
Vol: 0.55 m’ PuU-241 0.00021 5.02 s
—_— 9.9 GW
Na 0.0075 0.0457 1 "
r 0.0051
z 0.0340
IRV,
"Mz.6 - 118.6 cm He 0.060 4.87
wall Oxygen 0,060 0.034
3.9
0,041
\
3.6 - 218.0 cm e .06 0.0023
Feflector zone 2.8x10°°

4. A HIGH FLUX BURNER REACTOR FOR TRANSMUTATION

4.1 Need for fission product transmutation

4,11 Introduction

The problems associated with the management of highly radio-
active fission product waste has been intensively and exten-
sively discussed (Fig. 4.1).

Here only the transmutation of fission products (F.P.) is
dealt with. The recycling of the actinides is not treated.
Transmutation occurs by using neutron irradiation in a fis-
sion reactor.

A snort outline of this chapter can be presented in ths form
of the following questions

- why, contrary to many assertions, 1s neutran transmutation
in a fusion reactor not feasable? - this in spite of the
fact that the fusion machine has often been proposed far
this purpose

- why are recent opinions concerning transmutation in fis-
s5ion reactors rather pessimictic?

- could transmutation 1n a fission reactor be possible taking
into account the neutron balance in a breeding system?

- which fission products are candidates for irradiation in a
fission reactor?

- 1s the rate of tramnsmutation sufficiently hieh in a fissicn
reactor?

- in what t [BAS of reactor 1is the transmutation hysicall
Y Y
D(]SSith?

- what are the limiling parameters for transmutatbion in a
solid fuelled filssion reactor?

- 15 a very high flux fission reactor possible if the fuel is
in the liguid state instead of the solid state?
Fie, 4.1

POSSIRBRILITIES FOR
(see also table 4

1
1)

'‘RANSMUTATION OF

Long-tived
fission products

Other waste
management
methods

Transmutation

Changing neither

<4f—””’/”’/”’7

30sr AND

Anor Z

Changing A but
not Z

Poton bombardment

Changing 2

Neutron bombardment

neutrons from
accelerator

N:'ustr_on Neutron capture

emission ()

{(n,2n) 7
Periodic

irradiation Contmupus

{explosion) irradiation

Secondary

Primary neutrons
from reactor

Neutrons from
fusion

Neutrons from
fission

— —

Proton bombardment,

<‘\\\\\\\\~\“‘-;

Thermal reactor

Fast reactor
with thermal
central Z011€

Fast reactor

137CS
Table 4.1 Transmutaticn possibilities for differsn. devices

Machine Clux/ ~eactions, and remarks of authors
Cnergy of original reports.
Accelerator of Protans Reaction p,xn!] not promising. Ruled out
medium and high 100 MeV on basis of energy balace criteria.
energy protons Protons Spalation (p,xn) and (n,2n) (n,¥y)
1.70 GeY with secondary neutron flux
Cs-137 as Not feasible within limits of current
target and/or technology. The capital cost is
thermalised prohibitive.
flux of (see table 4.2)
neutrons

Fusion

{thermonuclear)
reactor in all
cases with wall

Fast flux of

14 MeV neutrons
from (0D-T) & =
5x101 %0 em2g7 !

Neutran reactions (n,2n) and (n,y).
Fast Flux of 5 x 10'°n cm™ 257!

Thermalised
flux 1in
ceryllium trap

Practically only (n,Yy)
Thermal flux 6.7 x 10'°n cm 257!

Attractive transmutation rate has naot been
demonstrated but possible to transmutate
all Cs-13/ and Sr-90 creatad by fission
reactors

Nuclear

Fissile
explosive or
thermonuclear
explosive

Technically not feasible. No. of explosions

per year very high. Appr. 3800 per year each of
100 kton. (For USA in year 2000 Cs-137 and
Sr-40) Probably not ecceptable to public!

/Y
Fig. 4.2 FISSION PRODBUCTS

@®veta stable Wate
QObeta unstable

\beta decay

10 b

1 daly 1\3 d 1ui) d l‘ year ]lJ y lC‘\ ¥ li‘;\l ¥

1 T T T T T 1,
i 10" 10° 10" 107 108 1? 10t il

time, sec

Fig. 4.2 FISSION PRODUCTS
CANCIDATS FOR

Half-1ife, years TRANSFUTAT 10N

>>10'* | 107

stable | short
\ nuclided living
F.P. F.P.

\ /
\ b ey /

/

— — — —— Thansmutation,
shortning the
half-11fa

very long
living F.P.

quasi stable
F.P.

Table 4.2 Passibility for transmutation of F.P. particulariy Cs-137 and S5r-80 in
2 fission reactor according to BNWL - 1500
Reactor Referance Flux Remarks
Steinbtersg The authors use a wrong value:
Thermal Gwer reactor Wotzak Kr-85 with large 0 = 15 barns
= cact
P Manowitz, 1964 instead of ¢ = 1.7 barns.
Isotopic separation of Kr-isotopes.
3 x 10'? thermal Only I-129 can be transmuted.
10'® in the trap An equal or greater no. of F.P.
L smaller 1n the would be formed in the fission
Steinberg, 1364 , )
presence of the process per transmutation event.
F.P. target.
This reactor does not meet the
high flux (trap) Claiborne, 1972 2 x 10'° thermal criteria of overall waste balance
and of total transmutation rate.
Meutron excess 0.15 - 0.3 at the
expense of being no longer a
. . viable breeder of fissile material.
liquid metal i _ 15 .
Fast Claibtorne, 1972 1 x 10 fast Alzo this flux does not allow the

fast hreeder

attainment of a sufficiently high
transmutation rate and is, therefore,
not a feasible concept.

Fast with
thermalil

liquid fuel
fast reactor

with thermal

column.

this paper

- how could such a high flux reactor with circulating liquid
fuel and a thermal column cperate as a "burner” for some
F.P. (Cs-137, 5r-90 etc.) transmutations? (Fig. 4.3)

- 1s such a system feasable?

Comment

In BNWL-1800 it was noted that the calculation (in a moderating
blanket of the CRT) represents a more realistic blanket config-
uration with a neutron wall loading of 10 MW/m? (This is still
a very optimistic value. M.T.]).

In this case the following data have been obtained for a therma-
lised neutran flux from a CTR with a 10 MW/m? wall loading.(Tab.l.3)

Table 4,3
For 80% ¢ thermal b0 beg effective
fraction (n.om 2s-1) (n,v) (n,2n) tVZ effective
291 ke
CoT g 51 w1010 | o - 0.117 | o = 0.10° | A = 22.2x107 157!
(n,y) (n+Zn)
barn barn
7.91 x 10710 7.0 x 10°1° 3.9 years

The conclusions of this study are that useful quantities of
Cs-13/ could be transmuted under the projected CTR blanket
loading conditions. The reduction in Cs5-137 "toxicity” 1s still
expected to be at most a factor 3 down. In addition a study of
the build-up of fission product nucledl in order to establish
the roquirements of periodix chemical processing and associated
costs has not been carried out.
Fig. 4.3 Accelarator for transmutation
BALANCE FOR 1 FISSIONED ATOM

< '0@.«. XA
SRS

HEAT WASTE

0.008 PROTONSat GeV
x 40 = 0,32 NEUTRONS for
TRANSMUAT 10N

MUTATION

Fig. 4.4 Neutron balance

Camment

H.W. Lefevre (appendix to BNWL-1300) makes an interesting
comment on the study of the transmutation of Cs-137 and Sr-30
in CTR: "Everyone knows that a CTR will be "clean”. Don't
spoil that illusicn. I think that I would worry some about

a CIR loaded with 50 kg of Cs-137".

4.1.2 Why some opinions concerning transmutations in a fis-
sion reactor are rather pessimistic

A recent and mecst intensive study of the use of a fission re-
actor for the transmutation of fission products has been pub-
lished by Claiborne (1972). He writes:

"The problem fission products cannot be eliminated by any
system of fission power reactors ogperating in either a stag-
nant or expanding nuclear power economy since the production
rate exceeds the elimination rate by burnout and decay. Unly
abt eguilibrium will the production and removal rates be equal
a condition that 1s never attained in power reactors. Equilib
rium can be obtained, however, for a system that includes the
stockplle of fission products as part of the system inventory
since the stockpile will grow until its decay rate equals the
net production rate of the system. For the projected nuclear
power economy, however, this will reguire a very large stock-
pile with its associated potential for release of large quanti-
ties of hazardous radio-isotopes to the environment. It is this
stockpile that must be greatly reduced or eliminated from the
biosphere. A method suggested by Steinberg et al. i1s ftransmuta-
tion in "burner reactors”, which are designed to maximize neu-
tron absorption in separated fission products charged to a
reactor. If sufficient numbers of these burners are used, the
fission products inventory of nuclear power system can then
reach equilibrium and be maintained at an irreducible minimum,
which 1s the quantity contained in the reactors, the chemical
processing plants, the transportation system, and in some in-
dustrial plants.

-
- 43 -

If the assumption 1s made that burner reactors are a desirable
adjunct to a nuclear economy, what are the design requirements
and limitations: it is obvious that they must maximize (with
due regard to econcmics) the ratio of burnout of a particular
fission products to its producticn rate in fission reactors,
and the neutron flux must be high enough to cause a significant
decrease in its effective half-1ife. 0Of the fission types, the
breeder reactor has the most efficient neutron economy and in
principle would make the most efficient burner if all or part
ot the fertile material can be replaced by a Sr-Cs mixture
without causing chemical processing problems or too large a
perturbation in the flux spectrum because of the different
characteristics of these fission products. The cost accounting
in such a system would set the value of neutrons ahsorhed in
the fission product feed at an accounting cost equal to the
value of the fuel bred from those neutrons.

The maximum possible burnout of fission products would occcur
when the excess neutrons per fission that would be absorbed

in a fertile material are absorbed instead in the fission pro-
duct feed. The largest possible burnout ratic would then be the
hreeding ratio (or conversion ratio for nonbreeders) divided
by the fission product yield. The estimated breeding ratio for
the Molten Sall Breeder Reactor [(MSBR), a thermal breeder, is
1.05 and for the Liquid Metal Fuelled Fast Breeder Reactor
(LMFBR), 1.38. The yield of '*7Cs + %%Sr is 0.12 atom/fission,
but a number of other isotopes of these elements are produced
which would also absorb neutrons. However, if the fission pro-
duct waste 1s aged two years before separation of the cesium
and strontium, the mixture will essentially be composed of
about 80% '*7Cs + ?%3Spr and 20% '*°Cs (which will capture neu-
trons to form '?°®Cs that decays with a 13-day halflife conse-
guently the maximum burnout ratio for '?7Cs + 2%y Wwill be de-
creased by 20%. This leads to a maximum possible burnout ratio
ot about /7 for the MSBR and about 9 for the LMFBR. Unfortunately,
however, the neutron fluxes in these designs are well below
10'°%n em™?s” !, Any modificetions of these desligns to create
high neutron fluxes will increase the nectron leakage and de-
crease the burnout ratios significantly”. (Claiborne 1072
Comment

it is not clear why Claiborne claimed that after 2 years ageing
and separation of strontium and caesium the isotope composition
will be

From Crouch (1873) the fission products of U-235 have the fol-
lowing composition (2 years ageing) (in at % per fissioned nu-
cleocus). {see Table 4.4)

Sr-88 (stahle) 3.63
Sr-90 (28 years) 4.39
Cs-1233 (stable) B.57
Cs-134 (2 years) 3.5 (7.08 0.5 from independent yield)
Cs-135 6.26
Cs-137 5.99
Subtotal 30.34

The realistic dala are unfortunately more than twice those citod
by Claiborne.

The same negative opinions concerning the use of Flssion Reactors
for F.P. - transmutation are given by the following authors:

- A.S. Kubo (BNWL - 1300):
"Fission products are not conductive to nuclear transforma-
tion as a general solution to long term waste management”.

- BNWL - 1800, itself:
"In summary it is improbable that transmutation of fission
products in flission reactors could meet any of the technical
feasibility requirements for the production of stable daugh-
ters”.

- Claiborne (1972):
"Developing sprcial burner reactors with the required neutron

flux of the order of 10'7n cm ™ ?s™! is beyond the limits of
current tecnhnology”.
4.1.3 Which fission products are suitable candidates for
transmutation and in what quantities?

A simplified breakdown of neutrons and fission products produced
by fission of 1 fissile plutonium atom is given in Fig. 4.4

From this it must be clear that only a very limited amount of
fission products can be irradiated by neutrons of the whole
system to retain a good breeding gain and doubling time - in
other words a s elf sustaining and expanding breeding system.
For further consideration it is postulated that the maximum num-
ber of transmutable nuclides equals T = 0.3.

the proposed system for the transmutation includes two types
of reactor:

- power reactors in the form of fast breeder reactors with s
total power of three to four times that of:

- a high flux burner reactor.

The crucial F.P. nuclidas are characterised in table 4.4 to-
pether with others. The data available now makes it possible
to estimate the number of candidates for transmutation in our
breeder/burner system, using the tollowing criteria

~ thr total amount of all nuclides to be ftransmuted carnot be
greater than the estimated value of T = 0.30, that is 30 atoms
of F.FP. nuclides for each 100 fissioned nuclides.

- the order of priority taken from this table is given as
Cs > 5r> 1T > Te

- in Lhe first instance no isotopic separation process is
postulated.

Table 4.4 shows the F.P. nuclides selected for transmutation.
Table 4.4 The

priorities for the tr-asmutation of fissioned products

Soloctod Yield for fission of Atom/170 atom Pu-235 Assuming isotopic separation

SSLEELE 109 atoms of Pu-23°9 Subtotal stoms/100 atoms Pu-230
Cs-133 (stable! .97 6. 31 0.14
Cs-135 .54 140 14,450 7.54 } 14.37
Cs-13/ .69 21.140 £.09
Sr-390 2.716 23.32 2.18

205 Z2.208
Sr-88 (stablel) {1.44 x O 23.349 0.029 J
(2% isotopic
separation
efficiency)
I-129 N7 24.5193 1.17
.55 1.18

1-127 (stable) .38 24.899 0.01 J
Tc-389 5.81 .81 30.709 5.81 5.81
Kr-83 (stable) .36

- > .56
Kr-84 (stable) 5 474
Kr-85 672 0.67 } 9.7
Kr-86 (stable) 882 33.183 0.04 )
Total 33.183 24.28

9b

4.17.4 In what way could a burner reactor be coupled to a
system of breeders?

The aim of the calculations used here is to show

- Lhat given a system containing some breeding power reactors
with a breeding gain of G

- the fission products from all of these reactors can be tans-
muted in the high flux burner reactor, which includes of
course the fission product transmutation for the burrer re-
actor itself.

The calculation of the ratio of breeder power to burner power
1s as shown here

Transmutation rate (atoms s7!), T
Yield of fission products (atoms/fissiaon), VY
Effectiveness of transmutation device £
T = e 1 (Y(?%Sr) + Y('37Ce) + Y{other F.P.))
Y(3%3p) = 0.0471; Y('?*7Cs) = 0.064

Preeding gain for the total system without transmutation
GY = 0.375 (arbitrary)

Freeding gain for the total system with transmutation G
Ratlo of fission to total capture: o = 0.24 (arbitrary)

C = G2 - T/(1+a)

Taking a numerical example with the same values arbitrarily
chosen {including also another fission product with ayield of 0.17)

T = 0.57Y (0.041 + 0.064 + 0.1) = 0,40
5T = 0.375 - =10 = 0,07
e 1+0.24 T

which 1s sufficient for a power system having a slowly increas-
ing capacity with a doubling time of 100 yr. This corresponds
to Lhe near steady state case.
To determine X, the number of power reactors

X-Go-1 = (x+1) &
T
G +1 _  0.05+1 )
. “EB_GT" 3.375-0.05 ~ °43

The correspondence ratio of the power of the breeder reactors
and the burner reactor in this case equals

This means for example for 8 power reactors each of 3 GW(th)
(i.e. a system total of 30 GW(th) can deal with the transmuta-
tion of the fission products chosen here. The electrical output
assuming an aefficlency of 40% is (3x8 + 1x7)x0.40 = 12 GW(e)
(Fig. 4.5).

4.17.5 Is the rate of transmutation sufficient?

It 1s clear that the rate of radioactive nuclide removal 1in a
field of particles 1s given by

- inZ
AQFF ) Adpca ' >\trarmmut(:fi:ion ot
ey - Y2 (efF)
where
A, = constant of radicactive decay (s~ ')
decay
trans - trans
o = cross section (em?) for a given reaction
® = flux of the reacting particles (particles cm “s” ')

this value of Agoep will be used later for the calculations of the
neutron flux required to permit the transmutation rate to match.
Fig. 4.5 Scheme of the proposed breeding power system

with "self-cleaning". (For the sake of
simplicity, only the routes of the fission
products *%°3Sr and '?7Cs are shown).

8 Fower Reactors, each GW (thermal) total ~24 GW (thermal)

) @@

Fission Products .

Partition of
Fission
Products
BURNER (inctuding
isotope
REACTOR ' separation}
Vi GW (therm) T
Other Fission
Products
Fast
Core ¢ goSr
Fuet 1370

Reprocessing
of lrradiated
Nuclides

- —

l Stable uclides

g g‘Zr, 138,
- 100 -

Let us assume that the energy production is based on a set of n
burners and nX breeders. At a time t, (see Fig. 4.5) when it

is decided to stop the use of fission energy production in favour
of other sources the total amount of a selected fission product
present 1is

(1) N(E ) = (X+1) ne—t
eff
with K = Y«P/E
Y = yield of the selected F.P.
P = power per burner (or breeder) (Watt)
E = energy per fission (Joule)

This amount of F.P. is located only in the burners, therefore
sach bhurner can receive

Kk

ke??

(X+1)

although then any production should only represent

K

AEFF

in the steady state.
- 1071 -

At time t, the nX breeders are shut down and only n burners are
in cperation. Later on {(time t,_4) the nuclide removal is such
that a re-arrangement is possible and one burner can be stopped,
ist F.P. contents will be loaded in the remaining burners etc.
At the beginning of each time step tp the p burners which are
still working contain the maximum possible amount of F.P.

(X+1) =
eff
) N(tn) ] N(tn_qJ _ N({t ) ) N(tp 1] _ N(tzl _ N[th )
“ n n-1 D p-1 Z 1 )
S Xe1) e 5
eff

where N(t) represents the total amount of the selected F.P.

One could imagine other schemes: for example the rearrangement
could be made only when 2 burners are to be shutdown. From the
reactivity point of view this solution is worse than the one
preposed. Coming back tec the original solution, one has still

to sclve at each time step (tp, tp-1) the burn-up eqguation.

dh -, \ N where the right hand side i1s the
dt eff F.P. production.

L

then the solution is

Using (2) the time needed to go from p burners to (p-1) can be
deduced

102 -

with a summation one obtains the time t1 after which one burner
only is in operation.

p=n
. - 1
(63 A ps (Ey-E ) Z :1”1_X+1
p=2 peX

A more direct evaluation can be obtained iF¥ n is so large that
the number of operating burners changes continously with time
(p = n(t)) then by a single elimination of p between (2) and
(3) one gets

4N .
P v | o
(4 M) = N(E Jer 2t sty )
n. eff X+1 ’ n
NCL )
N X1 o xel
(5 A _pplt t ) = = In s = 1n (n )

The two approaches give similar results except at the end when
few burners are in operation.,

For tLimes longer than ty only one burner 1s operated and the
amount of F.P. would decrease from

We shall postualte that 1t has no sense to operate this last
burner when the amount of F.P. is only 1.2 times longer than
the asymptotic value which requires a new time interval

(eq. 4 p = 1).
- 103 -

The total time t, - t,, will be the sum (6) + (7) which corresponds
to the reduction factor

n{xX+1)
1.7

Further reductions can only be obtained by natural decay (t>tOJ.

Numerical application: with X = 4,n = 100 which means the esconomy
was based before t, on 400 breeders, the initial F.P. amount is
reduced 415 times when the last burner is shutdown. Then the re-
quired time is defined by ggerltg-tp) = 8.93 (8.76 with the approx
expression). If this time is to be less than say 60 years 2 re-
actor generations) then Agpp = 4.7x107°s~' (ty, eff = 4.7 years).

Now the problem of the intensity of the neutron flux desired for
transmutation arises. Since the most hazardous F.P. nuclides are
those which apart from their high metabolic activity and high re-
tention in living organismus alsoc have a half life of the same
order as a human life span of B0-70 years we arrive at the fol-
lowing list of hazardous isotopes which are the mest important
for transmutation.

Kr-85 10 years, 20.9 x 1071071

Lo

A
dec
_ - - -10_-1
Sr-90 tVZ 28.2 years, Adec /.78 x 10 s
- = = -10 ~-1
Cs-137 tV2 30 years, Adec 7.32 x 10 5
desired 'half life’' = 4.7 years, A : = 4.7x10"%s 7!
desired
35 we know xdesired i Adecay ' xtransmutation

The most important problem arises from the fact that the two
nuclides Sr-90 and Cs-137 have very small cross sections for
neutron absorption in both the thermal and fast regions.
104

Cross section, olbarn = 1072%cm?) Ratio

thermal fast therm./fast
Sr-90 0.8 barns 0.0076 barns " 400
Cs-137 0.11 barns 0.0137 barns g 8

therefore to achieve Adesired

fluxes should be:
fast flux for transmutation of

Cs-137: ¢

fas

®Fast

desired—Adecay

4,7 x 107%s~! the neccessary

4x10°°

t o]

4.0 x

(Cs-

137 fast) v0.01x10” 2

107 (n em™ 257 1)

thermal flux for transmutation of

Cs-137: ¢

th

Ax107°

’g

3.2 x 10'% (n cm %57 1)

0.11x10

24

thermal flux for transmutation of

Sr-90: o

“th

‘see also Fig.

The
- a
ing
des
achieved)

gquestion t
fast flux

4.,0x10"

9

5.0 x 10'° (n cm 25~ 1)

0.8x10°

4.7).

hen arises,
of 4x10'7 of

i

in what device are such fluxes possible
a thermal flux 6x10'%, It is interest-

to point aout that during the period of 60 years which provi-
the reduction factor of 415
the natural decay of Cs-137 would have reduced it only

1

(if the Agf¥f 4.7%x10°%s~! can be

by a factor 4 which demonstrates the efficiency of the burner.

Also the burning which occures during the first period

reduces the am
A
Fop._8ff
deca

ount of

times b.

y

(t tn)

7 times for Cs-137
Pig. 4.7

1016 4

Effective half life and neutron flux for

transmutation

wmcevemes  THIS PAPER
——cwmee PARISH  (1977)
o= CLAIBORNE (197..>

MSFR
Mclten Salt
Past Reac-or

Neutron J
flux
(n/em's)

\ %
Intensive
Neutron
\ ontroll generator)
Ther-nl

AMATTITRFTETTTETTTEEECTEERRKRRTT

(Light Water
Reactor)

Effective half life (years)
- 106 -

4.17.6 In what reactors are the transmutations possible?

From the point of view of this paper the most important process
is the transmutation of some of these nuclides by neutrons in a
fission reactor. The criteria given in chapter 2 limit the choice
of system. That 1is

- the number of F.P. nuclei cannot be too iarge in relation to
the number of fissioned atocms in the burner reactor (reactor
for transmutation) because the latter process also produces
new fission products.

- the fission reactor should be self-sustaining - that is a
breeding system.

- Lthe specific power of the reactor is preoportional to the neu-
tron flux. Hiegh neutron flux means high specific power which
is controlled by the effectiveness of the core cooling.

- the specific power F and the neutraon flux ¢ are coupled by
the fission cross section and the concentration of fissile
nuclide (Ng)

For thermal neutrons ogf 1s approx. /00 barns and for fast neutrons
only 1.8 barns, that 1s 400 times smaller.

For the given total power and the same specific power the product
Ne+d for the thermal reactor must be approx. 400 times smaller
than for a fast reactor. Since the critical concentration of fis-
sile nuclides 1n a thermal reactcor can only be 10 times smaller
than fer a fast reactor then for a given specific power the neu-
tron flux 1n a fast reactor can be about 40 times higher than that
of a thermal reactor.

The cross section for thermal neutrons for the nuclides considered
here 1s from 3 to 10 times larger than in a fast flux and this
must be taken into account.

All these factors bring us to the following solution of the prob-
lems under discussion.

- the highest specific power and hence the highest neutron flux
1s pogseible 1if the cooling process is carried out by the fuel
itself and not by a separate cooling agent only.

This directs our interest towards a reactor with molten fuel
in spite of the exotic nature of this solution.

- the high flux reactor must be a fast reactor (small for fast
fission)
- 107 -

- because Otp > Ofggt the thermalisation of the high flux in a
internal thermal zone is postulated, then it is possible that

zone core
therm. fast

- the first approximation is made for an i1sotopically pure radio-
nuclide e.g. Cs-137 without Cs-133 (stable) and Cs-135 and also
Sr-90 without Sr-88 (stable).

The discussion then results in:

- transmutation of Cs~-137 (and some other nuclides) in a ther-
malised central trap of high flux neutrons:

®therm = 5x10'® n cm™?%s7!

- production of a high flux of fast neutrons 5x10'® n ecm™?s” ! and
the high specific power of 15 kW em™® is achieved by means of

liquid fuel circulating through an external cooler.
- transmutation of other selectred fission products in an external
thermalised zone with a thermal flux of 5x10'° or

Tx10' n oecmm%st,

- coupling of one burner - high flux fast burner reactor with a
system of 'normal' power breeder reactors.

A.1.7 What are the limitations of a solid fuelled reactor?

Can the desired specific power of 15 kWem™® be achieved in a solid

fuel reactor? These are the self-evident limits in this case.

- the rate of burning of fissile nuclides is limited due to
depletion of fissile or an increase of F.P. nuclides

- the heat transfer limitation of fuel/clad to coolant

- the temperature and temperature gradients 1in the fuel and
cladding (melting, mechanical properties)
- 103 -

- the boiling of the coolant
- the limitation of coolant velocity, pumping power, stability.
Now we discuss these limitations in more detail

- the dwell time in a solid fuelled reactor in core for the
fissile nuclides must not be too short.

concentration of fissile nuclides*maximum burn-up
dwell fission rate

We could write:

where N = concentration of fissile nuclide

N = — and f = 3.1 x 101°% fissions/joule

= maximal burn-up
= power (watts)

b
=
R = fission rate [(fission s~ 1)
R

from this:

: . Pofib b

dwell o*QP.T UF'Q

b = 0.03 (corresponds to 30.000 MwWd/t)
LRED  5pg w 10-2% om?
fiss

d = 5 x 10'® n cm 25 !

tdwell = 850 s = 14.3 minutes

hut also for b

i

.10 we achieve t = 4/7.6 minutes
dwell
- 103 -

For a fast reeactor {(some arbitrary values)

b = (1,10
$?St = 1.8 x 10°2?% cm?
fiss

o] = 5 x 10'® n gm™ % !

tdwell

Conclusion:
- the dwell time in a thermal reactor is prohibitively short,

~ in a fast reactor it is more reasonable but still very short,
especially in the case of a solid fuel reactor

- the limitation of specific power by heat transfer is as follows:
Specific power, Pspec' in a "good” 3 GWip power reactor and wilth
the appraopriate flux taken from literature is:

0.05 kW/cm?; o = 5 % 10' n cm%g !
spec th

fast = = kW/ecm?; @
spec fast

thermal P

= 5 x 10'° n em™ 257!

in a high flux reactor: (see also Fig. 6 1.8)

thermal:P

i
]
o=

kW/cm? 3 x 10 nocm %g7!

a=y
I

spec ) th
P = 1.5 kW/cm?®; @ = 3 x 10'® n ocm %g”!
spec th
fast: = = 1.0 kW/cm?®; & = 1.,5x10'® n cm™?s7!
spec fast

With the same geometry the very high flux reactor desired here
would have the following flux for Cs-137 transmutation:

I
~J
™3
Sl
P
=
.y
T
3

[

5.0 x 10'%, the specific power P

-b
G
3
HH
i

th

l
nNJ
-
x
=
-
0
3
w

= 4, 107, P f
for ¢ < 0 x 10 the specific power PFast
For a solid fuel we postulate the following "unit-cell”

Dimension Volume Cross- Surface-
section area
area
Cell: 0.9%x0.9x1.0 cm 0.81 om’ 0.81 cm? 3.60 om?
Fuel: = 0,60 cm 0.783 cm? 0.283 cm? 1.6885 cm?
Cladding: diam = 0.63 cm
wall: s = 00,03 cm
diam = 0.60 cm N.568x10 " %cm?®  0.568x107%cm?  1.904 cm?
Coolant: G.521 cm? 0.521 cm?

"desired”
per unit fuel element surface area:

In this specified cell of a
would achieve a heat-flux,
(for both bypes of reactors,

high-flux-reactor,

thermal and fast)

L2 21 kw/em’-0.81 om’

-2
fs 1. 885 cm? 4 kioem

we
Using now a simplified model for the first guess of the tempera-
ture gradient we can say: the amount of heat generated in the
fuel must be the same as that leaving the surface of the cladding
material.

A

5
Tclad H?S By

wWwhere s = wall thickness and A heat conductivity {(Weem™1ek™ 1) an
optimistic value for stainless steel is A = 0.4 Weom 'ek ™!
- 0.03 _ o
ATclad = 9000 T T 6/57C

It is evident that this result is not realistic.

The solution of this problem may be the thermalisation of neutrons
in a high flux fast core and the irradiation of Cs-137 in a thermal
central zone.

In such a thermal central zone we postulate (and this must be based
later on corocalculations)

®th = e CI)Fast

ta reach &+y = 6.0 x 10'® we require dfaet = 5.0 x 10'% n om 25710,
For this fast flux the specific power can be assumed, if weo take
into account the effective increase of the fission cross section
because of the influence of the thermal trep. The simplified cal-

culation results in a specific power of 10 kW cm™?.

The corresponding heat-flux is therefore reduced to

g PF.V ell
H = PeEC  CP 4.3 kW cm”?
fs AFS

and the temperature gradient to

_ 4 L.o-2y _0.03 (em) _ 0
ATGlad = 4300 (Weem™7) 0.4 (Weem™ K- g

M~J
L

This

value

is still rather high.

A rough estimate of the thermal

stress in the wall can be taken as

coefficient of linear

& . expansion (K™1)
5 = % . Bth C AT - F, - - TESU%Uim?E)QIBStiGity
and the corresponding values for stainless steel (13-9 0OL)
g = % < 151075 (KT e 323 (K) 2.5+10% (Kpecm™?) = 5500 Kpecom
It is also evident that this result is not realistic!

The resulting thermal stress in the cladding wall coupled with the

high flux is prohibitive.

Hlere, The most important problem 1s the local overheating by the
thermal flux present at the interface of the thermal column and
the solid fuel core. If, in the fast region near to the interface
approx. 20% of the enrgy comes from the thermal neutrons, the fol-
lowing ratio of fluxes must be achieved:

T T %tn east T Peast?! U7

0
fast 1.8

“eh TER fast T U7 Ton Tt eagy

i - G0 7 "% . ¢

TR Pfast
That the thermal flux must be 2000 times lower than the fast, would

seem very difficult to realise.
1.7.8 The liquid-fuelled fast reactor with central thermal zone

A much better solution is using a liquid-fuel. The transfer of
the heat generated is done, by pumping and cooling the liguid
fuel ocut of core.

For a unit cell of 7 cm3, which in this case consists of fuel
only, we can write the following hesat-balance.

) = . [ ] —— . = *“‘ - 1 . -3 . -1
%SDQC (p+c) W i (pec) heat capacity (Jecm K™)

W = velocity of fuel (em/s)

T/1 = Lemperature increase per unit

cell (Kecm™ 1)

= = apecifi ) Wepemm™ 3
Spec pecific power (Weecm™ )

It we allow In the core a temperature increase of AT/l W 2 dececm
<0

This velocity appears to be within the practical limits proposed
for 2 fuel-velocity of 40 m.s™! for a reactor with 10 kW.cm™?
specific power. This point however has to be seriously investi-
gated, as erosion due to high velocities is a problem.

For a /7000 MWith) core with a specific power of 21 kWecm™? the
fuel volume is about 0.330 m?3.

The target volume, that is the volume of irradiated (transmuted)
fission products e.g. Cs-137, is postulated as 1.3 m®. The dia-
meter of a sphericael core is therefore 145 cm. The temperature

increase of the unit cell of fuel, in one pass through Lthe core,

rguals approximately

T = 3 (dacecm™ ') « 146 (cm) = 438°C
fuerl e :

and for a fusl inlet temperature of 550°C we reach an outlet tem-
perature of approx. 988",
The

for socme years. Lane
fas reactors:

"As an alternative

to the possibilitly
test facility. The
ability to achileve
time eliminate the
actor.

just mentionsd 1is

idea of a liquid fuel high flux reactor has

A fast spectrum molten
hiph fissile concentration
crder to get mean neutron energiles
switching from an NaF - UF, salt
based)

been discussed

(18971) for example writes about high flux

some consideratilion has been given at Oak Ridge
of using a molten salt reactoras a fast flux
primary virtue of this approach includes the
very high power densities and at the same

down time associated with refuelling the re-
salt reactor however reguires a
(i.e. 300 to 500 ¢ 27°U/1litre) in

in the range of 10 to 50 keV.
(on which the energy range

to a chloride-salt reactor would permit

a higher mean energy for the same fuel concentration but would
require the development of a new technology assocliated with the

use of chlorides. Since the fast flux level 1is largely debtermin-
od by the power density a flux of the order of 10'°% or more cor-
responds to a peak power density in the fuel salt In a range of
5 oto 10 MW/litre and to a power level of about 1000 MWith)., This

means that there will be only 100 to 200 litres in Lhe core:
however the external volume would be about 10’000 litres”.

4.7 The Neutron-physical Aspects of the Hipgh Flux Reactor

4,720 Introduction

The idea of destroying the beta active long lived radionuclides

is based on the following:

B A Fi?oo= primary lsalon
{ > { ‘ T
Soanbanooea denay ST re ot
! ! | e T ol o
‘:, l - '[ i J Tfl l\ ™ !
l NG 1s unft L
L,y ) Ao b trradiavion A - abomic maso
| /= atomic number
v -
LA+ L) 6™ [A+1}F Fo= stable nuolide
_ - . -
ST spontaneous decay (Z-1)
= 1o very short
‘|/J 2 -
Fig. 4.10 gives some of the given transformations which may
occur under a high flux. (Remark: Fip. 4.3 and 4.9 omited)

In this system the following simple assumptions are made

- the amount of fission products come from both the fast power
reactor breeders and the thermal burner.

- the fuels and materials are continually reprocessed

- the irradiated fission products are continously (or periodi-
cally) separated in order to eliminate the daughter stable
nuclides (e.g. Zr-90 and Zr-91) from the decay and burning
of Sr-30

- the amounts of transmutated nuclides in the steady state [(85)
irradiation are calculated by the obvious relationships for
the i-th nuclide.

>

A,7.7 Neutronic calculations

A reference burner reactor concept is shown in Flg. 4.771. The
flux trap 1s surrounded by a BeO beryllium oxide spectrum con-
vertor, a critical fuel thickness and an outer wall (see Table
4.5 and Table 4.8). Fig. 4.17 shows the calculated flux distri-
hbution. The tontal flux in the fuel is similar to that in the
flux trap [(Fig. 4.13)

I
™J
8]

Moderation requirements

To form a thermal neutron flux trap one must naturally use neu-
tron moderating materials. As is well known, light materials can
scatbter neutrons past the neutren-absorbing intermediate-ennrgy
resonance region. jH is the most efficient nuclide in this re-
spect but als exhibits appreciable thermal absorption. Deuterium.
10, beryllium %Be and arbon 1§C are usual alternatives. Oxygen WfiD
1s rather heavy though frequently already present in a molecular
combination. Other light nuclides have unacceptable nuclear or
physical limitations.
oL 4010

140

139 -

138

137

136 -

135

Cs-135/137

Atonilc

Transmutat ion

Number

nf sele

92—

91 +

103

102

101 —

100

99 -

Te-99

-
57

cted Nucolides-Tission products

Sr~90
Zr-93%

B~ B~
jfiosad
3 B

=

28y 6Uh
ST Y ir Nb Mo

T
38 39 40 41 42

Loneg lived fissinn product
belng the obiect of Lrancmutation

@ ciable Nuclide

() Heta unstabla, intermediabe
nucl 1oe

(:)ljjng Lived nuclide

R-
—> Snontaneous beta decav
nalf-1life -

t+ neutron capture (n,y)
- 117 -

Table 4.5

OBJECT: HIGH FLUX BURNER REACTOR WITH THERMAL ZONE

REACTOR TYPE

CEUOMETRY:  INTERNAL Z0ONE : THERMAL, HIGH FLUX
WALL : IRON, GRAPHIT
INTERMEDIATE ZONE: CORE, FUCL
WALL : IRON
EXTERNAL ZONE -

WALL, REFLECTOR : IRON

POWER (C% thermal) .7
POWER DENSITY (GW thearm/m? core):
MEUTRON FLUX, MEAN (n/cm?s) v

FISSILE NUCLIDE:  Pu-235
DILUTENT : NaCl
COOLING SYSTEM ¢ out of core
BREESING RATIOC 0 -

PARAMETER STUDIED:  TModerators in thermal zone
Wall, thickness, beryllium, graphite
Volume, specific power
cooling parameters

METHOD OF NEUTRONIC: ANISN S4

CALCULATION : 23 Groups
Pq APP.
GCC3 CODFE
END
- 118 -

CEOMETRY OF THE HICGH FLUX REACTOR

Neutron flux, ncm ¥ s’

Reflector

Fe
A
Centra! Region /.-/
ya /"/".'Sr(IODl)z,ICSJOD /
0 radius, cm 78,5 875 94.6 97.6 200

Diam =
120 cm

______________________

o — —— ——

Thermal

Fast 1.6 MeV
Table 4.2 Gacmotrey and

CTet 41 Dowern =

7 GWlth): fission

leutronics of the iligh-Flux Burner Resactor

oraoducts from the total system of 30 GWlth)).

Region Radius Componcnts Meutron Flux
(cm] (atams 1077 "em™?) (n cm %5 1)
total
Inner Duter =
thermal
Cetral for trarmsmutation 0.0 78.5 13705 0.0116 3.83 16
90
0ol : L
Volume 2.1 m° =T . 0U15 203
0 0.0145
3 0.0145
(Cesium and
strontium deuteroxide)
Wall, moderator /3.5 88.0 Be G.060 4.48 « 1016
0 (. 08d e
Volume 0.82 m’ ). U8 1.7
{Beryllia: 8 cm,
raphite: 0.5 cm)
Core (fuel) 86.0 34.5 2390y, 0.0014 4.03 1018
Jelume .57 3 240p 0.0004 0.0158
‘ oo “Hlpy 7.0002
Na 0.012
Cl J.018

(Plutonium, sodium
chlorides: PulCl_.+f NaCl)

3
i i 4.00 x 10'°®
Wall 34.0 )7 .5 Fe 0.08 v T
Reflaector 37 .6 200 fri 2.08 Boundary flux

2.4 x 1013
10°

SR
Fig.y Heutron spectrum in the core
Total mean flux “.O}'lolb n
Specific power 12.1 kvlcm-)
Total power 7 dwth
arbitrary
units
W
1ot
e
1 [
11l -
Jdedtrop erergy grpup
Q1 19 17 5 3 1 o sl |t 1
T T T T T T T T l
i 1. 11 10 1 3 7 t S 4 3 g 1 J

Neu%rlo;\d

Wik

nem

Letnargy

i.13 High flux burner reactor

Cross-section of
the reactcr

Reflector

Target zone
(Sr(0D Yzt Cs{0OD))

T2t v ud T et 7 od TTUIhy T 1o 200

Radius, cm

—_——
- Thermal flux =

(2% th groups) \

y 4

: \

-

h' Fast 1'lux
5 th group
(1,39 - 5

Wall

Modegaton Reflector

Target zone

SESSUNAGNNNNANNY

200

N
3
"
m
2
2
9
=,
-

Radius {(cm)
Considering chemical and physical properties, the logical mate-
rials to be used inside the lux trap are hydroxide and/or deuter-
oxide compounds of the FP. fFig. 4.14A shows that just a small
proportion of H molar fraction has a large deleterious effect

on the Cs-13/7 transmutation rate. This is due to the H absorp-
tion cross section. Therefore, Cs00 and Sr(0D), are preferred.

As Sr-80 and Cs-137 alsoc have their fair share of resonances it
1s advantageous to thermalise the flux before reaching the flux
trap region contalning these targets. Therefore a spectrum con-
verter between flux trap and fast fuel is needed. Begaring in
mind the high temperatures to be obtained in this reactor and
possible chemical reactions with molten salt, H50 and 0,0 are
unacceptable. This leaves Be, Bel and graphite or some variant
therefore for consideration. Be {and B) compounds, of course,
have also to thelr advantage a relatively low (n,2n) threshold
(1.67 MeV). Location nect to a fast region can therefore pro-
duce considerable very slow neutrons in the flux trap - which
1s a main objectlive of the burner reactor. Replacement of Be

by € or Mo wall material should therefore lower the FP trans-
nutation rate, and it does. (Table 4.7 Fig 4.14B) indicates an
optimum thickness of about 5 cm He. For the sake of safety and
higher melting temperature, Be0 is preferred over Be.

Table 4,7 Effect on Replacing Be Converter upon the
Relative FFP Transmutation Rates

Case materials A (Cs-137) . (Sr-00)
i transm. transm.

L Be, Be 1.0 1.0

2 C, Be 0.77 0.64

3 Be, Fe/Mo 0.34 J.82

(Nnte. transmutation rate in arbitrary units)

S.204 Influence of other parameters

The influence of plutonium concentration (Fig. 4.14C) is an im-
portant parameter of thls reactor. Increasing the plutanium con-
centration sipgnificantly improves the transmutation rate but also
increases the power density above a technically feasable level
(Fig. 4.140). Sne also Fig. 4.15,
Transmutation Rate, Arbitrary Units

Transmutation Rate

122

L.14  Impact of some selected parameter v
(b) thickness of moderator (Rel), (
(d) power density, P (kW cm
A
1.2 —
- “sr
' =
-
-~
-
-~
s
\\
v37CS -
0.8 —
Reference
0.6 A
nE T T
0 0.1 02 03
H/D Ratio
(a)
C
1.2 o —
>
”
/ 1370
7/
1.0 ~ //
/ K
cm
0.8 -
Reference
0.6 -1 A
T T T T )
0 0.1 0.2 03 0.4

Molar Ratio of Pu

(c)

05

Transmutation Rate, Arbitrary Units

1.2 —

{

ariations:

(a) hydrogen/deuterium,

plutonium concentration, and

0.6 = Reference
T
3 5 7 9
Thickness, cm
(b)
D
10 -
5 -
2 -
1 =
A Reference
L 1 | ] 1 R
0 0.1 0.2 0.3 0.4 0.5

Molar Ratio of Pu
{d)
- 123 -

Increasing the reactor power (Fig. 4.16A) from 5 to 11 GW(th]
improves the transmutation rate. However a power unit above

7 GW(th) seems to be beyond the technological limit even for
the distant future. The reference case has been taken as

7 CWlth).

The use of beryllium [(Fig. 4.168B and 4.16C) instead of iron 1in
the reflector of the core improves the transmutation rate sig-
nificantly. At the same time the power density increases pro-
hibitively (Fipg. 4.160). The raference case contalins 1ron 1n
the reflector.

Molten fuel offers the only way of handling the very high power
densities of 10 GWm-°®. In addition the very steep gradient of
fission rate makes a molten fuel core essential since the local
fission density can be one order of magnitude greater than the
mean density. In a solid fuel core the high heat removal rates
would not be achlevable.

The use of boron to absorb the thermal neutrons results in a de-
finite decrease in the rate of caesium transmutation (Fig. 4.1/,

The large size of the thermal flux trap results In the fuel re-
oion approaching slab geometry with attendant high neutron leak-
agr. To better epconomise on neutrons several possibilities may

e tried

- use of an optimised reflector such as te, Ni, Cu or He to
minimise the critical mass

- use of the outer neutron leakage for breeding
- use of the neutron leakage for additional FP ftransmutation

To begin with a solid Fe reflector was assumed (Fig. 4.18]).

1.2 Thermohydraulic consideraltions

We now examine the thermohydrauvlic implications [(for more detail
see ch, 8).

The crucial parameter here is the core power density. The given
value is high but still near the present state of the art
(Table 4.8).

For comparison, power densitiss for some high flux reactors.
(Table 8)
Transmutation Rate, Arbitrary Units

1.2

0.8 -

0.6 -

Effect of variation in some selected parameters: (a) power, (b) power

124 -

(b) reflector, (c) power density, P(GW m-3), and (d) fission

density in fuel, F (fission em=?)

Transmutation Rate, Arbitrary Units

20 -

15 -~

¥
9

Power, GW(th)

(a)

(c)

|
1
(
1
1
'
f
1
|
}
I
t
t
1
L}
1
1
i
i
1
1
:
]
)
¥
]
[}
,
]
H
I

Beryllium

. . . . . 3
Fission density (fission/cm )

1.2

0.8 —

0.6 -

1015 -

1015 e

10|4 -

BeO

(b)

Fuel

B Vo Y

Beryllium

Wall

88

Radius, cm

(d)

94.6
Impact of plutonium concentration

125 -

Fig.4.17 Impact of natural boron

0,9 41010 \
Actavity \
of Thermal \E‘.err{al
Cs~137 |[flux in ~Jux
1020/Se rue}zzo e ‘\‘
-—ncm 3
=137
3 activity
0,7 410
0,60’
o
u,s 10
M ™
0 10 10 10 510
Boron thickness, mm

Impact of reflector

800 in the fuel /
Tuu -
Temp.
°c
¢0o0 ~ Liquid phase
500 =
PuCl ¢ a1
3
4u0
Caesium activity
100 =
Specific
-4 power
(relativ)
ol -
Activity b
ol vs-1%
{(relativ
units)
£
Wy =
;-
;
T T T T T T T
1, to N LY 0.t g4 D04 U 2.2 MY 0
Filutonium molar ratio
Fig. 4.18
Uranium
reflector
120 71 120
Volume
Activity of fuel
of Cs~-131
- k)
£ (
100 —100
SRIE BN
8o+ 80 /
il /
70 70 /
L
OO -1 60
4

Reference
Reactor

(iron

/

y{iv'ny

Beryllium 5 em
as additional
reflector

lTable 4.8 Thermohydraulics of the High-Flux Burner Reactor

FParameter Unit Value
Total power CWlth) 7.0
Core volume m’ 0.69
Fower density kW em™? 10.5
Fuel density g om”? 2.35
Heat capacity, mass J g'lK“l .83
HHeat capacity, volume J oom ikl 1.95
Diameter of tube inlet cm 120
Inlet velocity m s ! 15
Volumetric velocity mig ! 17
Hrat capacity E 7,233
Temperature increase 0. .
(outlet-inlet) - A

Temperature of fuel 8

. € 7010

inlet

Temperature of fuel OC 590

out lrt o

Mean velocity of fuel - -

. m 5 ‘

in core

Cooling --- Out of core in

heat exchanger
cooled by sodium

fhis seems tn indicate that a total specific power rating of
about 1T kgPu//™W(lth) may bhe achievable.
- 127 -

Table 9 Power density in high-+lux reactors

Power density GW(th)/m®
in core volume in coolant volume

Feinberg, research reactor 3- 8-10
Merlekes CM-2 [(Soviet Union) 25 5
FETE (USA) 1.0 2

Lane (Molten chlorides) 5-10 5-10
HFEIR (USA) mean 2 a
maximum 438 8¢5

Phenix 250 (France) .48 1.0
this paper 108 10«9

Thio crucial problem will be the efficiency of the external heai
pxchanger. In the following example some typical heat cxchancor
characteristics are taken to demonstrate the

culated for 11 GWIlEh).

~pecific heat exchanger power
(consarvative data)

Total volume of heat exchanger for
11 GW(ith)

Volumetric fuel ratio

Fuel volume in heat exchanger

Fuel In the core heat exchanger piping

Total fuel out of core
Fuol in core
Total fuel 1n system

Mean specific power of fuel in the
whole system 11 GW(lEh)
H.3m?3

Plutonium content of fuel
Power rating of whole system

The postulated power rating for the
whole system

For this case calculated the power
rating 1in the breeder power reactors

possibilities (oAl

[

oy}

O3
- 3
L Om

07 OW/m3

L3 gPus/em’

L3685 kePu/MWlth)

kgPu/Mw(th)

.15 kgPu/Mu(th)
- 128 -

4.4 Some results

Parametric studies were made as variations around a reference
system which assumed P = 11 GW(th) (X = 2+9, K = 4-2) and Rg7 =
78¢5 em. The flux trap 1s surrounded by 5 cm BeD converter, a
critical fuel thickness of 6+6 cm and an outer wall. Fig. 4.19
shows the calculated flux distributions for such a burner re-
achtor. Note that the total flux in the fuel 1s similar to that
in the flux trap. The calculated fluxes lead to the conclusion
that

(total spectrum, flux trap) = V2«(E=0.0253)

Tho result indicates the relative effect of X on the ratio R of
the FFP transmutation rate to FP production rate for the reactor
system. It can be seen that X should be kept as low as possible.
Absclute results will depend on the Cs and Sr densities in the

flux trap. A value of R = 1 was achieved for both FP nuclides
at X = 4«65, The F.P. atom ratio there was (Cs-137) / (Sp-90) =

- i)
//.,’;5-

Another Important problem is the relatively high flux in the outer
zaone or leakage from the core.

this flux can be used for two purposes:

1) for transmutation of other fissiocn products which have
rather a high abscrption cross section e.g¢.

Te-88  o°' = 22 harns tJi = 2.7 x 10° years
[-129 oth = 728 barns tJ; = 1.7 x 107 years
In both cases a fiux of 10'" - 10'°n cm-'s™!

permits a rather effective transmutation rate of

1 years

T-120 X = 7 x 10 %s ! ¢ = 2 years

To still further improve these reactions the use of a beryllium
mocerator in the ferm of a 5 cm wall in the ocuter region of the
core has been calculated. This gave an improvement in the trans-
mutation rate of Cs-137 in the inner target region but also a
very slgnificant increase of specific power due to the scattering
of the neutron flux in the fuel region.,

IThe possibilities for transmutation of these two long lived fis-
sion products will be descussed further elsewhere.
4.1% Fission density in the fuel zone

101

Fission
density

fission

s
emtsec

H
1

14

107

124

Fission density in the fuel zone

Beryllium

s S

Wall

190

104,6

106,6
- 130 -

The neutron flux outside the core may alsc be used for breeding
in a uvranium blanket. The breeding ratio may be higher than 1
but the decrease of the transmutation ratio is critical and
seems to be too low for a burner reactor.

Nevertheless 1t could be useful to check these possibilities in
more detail using some of the available neutrons for breeding
in an external blanket region.

Table 4,9 summarises the data for the reference case. It can be
shown that the transmutation rates obtained are egual:

90+, b = 2 = . - 0 —85_1
ol "?"F'F tf' + >\8 ] @"'XB q.flj X /]O
/2 i
te?? = 1.85 vyr
‘7
et 710
P37Cs: = U.2460 107871
Cs AS{¥ L ooox 10 e
1772
A -
tQFF .05 yr
I, ]/7
LY/ T = 3.3
B eff

The results obtained are significant but rather pessimistic. To
estimate the "profit” of the transmutation process, the hazard
index (H) must also be taken into account. From Table 4.70 1t can
be seen that the total reduction in hazard from both fission pro-
ducts mguals 13.5, which is a better indication.

L5 Comments on hazard coefficients

It is perhaps valuable to estimate the usefulness of using the
concapt of hazard coefficients 1n fission product management.
Table 4.10 gives socmo values for Sr-90 and Cs-137. From this it
can be seen thalt in a steady state transmutations requiring the
amount of hazardous substances 1s reduced by a factor ~ 15 1In
relation to the steady beta-decay.
Table 4.9 The Transmutatbion Process faor Soleocted Fission Products

rner reactor 7 o« A0%W) s fission rate in

il

)

)
>
S
|

W
I

notal power P

Lel

Dat= faor
Froperty Symbal Unit
902, 13704
Yield of fission prcducta Y --- 3.041 0.064
Production rate RD = F .y atom 57! 3.81 x 10172 5.05 x 10173
tom/10% " em?®
Concentration C a{o?gw trm 0.0016 0.0116
@] & DEE'
Volume of targe: ) cm’ 2.03 x 10°
Number of atoms N =V s C atom 3.25 x 1027 2.35 x 10°%°
Pecay constant AR 57! 7.86 x 10710 7.33 x 10710
Dzcay rate R = NAR atom s ! 2.55 x 10%°8 1.72 x 10t°
Mean cross section 0. 074 om? .29 3.045
Total Flux, moan D n cm_z 5 ! 3.83 x 10t®
Transmutation rate Rtr = Neg oD atom =t 3.61 x 10%° 4.05 x 10%°
C
Total destruction rate Rt = R+Rtr atom 5 ! 3.80 x 101° 5.77 x 10t?
Vz i
Destruction canstant tpf{ = IHZ/AdfF year 1.85 8.495
ici ( -rancsmubati 1 |7
Efficiency of T?ér mu 3.10H - . t/Z . /2 o 19 3.9
Inventory reduction ratio tr B Teff
Steady-state equilibrium RO R 3.8 x 10b? 5.8 x 10t°
B
YTar the mean value of She yield, see bt S

b,
Tacle 4,17 =azard 1ndex for Che Improved Reforenoc Case”
Ratio
- - . - 37 90
Carameter Symbol nit ERere L7y 5
1
370
Gaximum permissicle
concentratian
water 3.7 o« 107! 14,8 49
air H 3.7 x 10770 2.2 x 107 ° 55
umeann SO
Yield of Ffission products
from “¥3(0 Y stomn/fiesion 1.0572 0.935
from 23°U Y atom/Fissian (51 1.1599
from 23°u Y stom/fiasion .24 J.10669
Mean value for fission
rate aof
) 2 ) o
L2307 %py 1011 Y 7.041 N.754
Hazard HeY - 2.5 0,064 Z2.114
Efficiency of hazard
reduction by transmutation HeY /B4 . - - 137 .13 7.156
(Table ITI)
Mean effective hazard 2.114 i3 s
reduction 0.156 :
*The hazara ooefficient (H) is defirn-d as the amnuat oFf air and/or water neei-d to dilute o2
amount of a ;pivaen nuc_ide present levels propoosaerd £or the maximum permissible concentration.

B

AN’
Fig. 4.21 Inventory of Strontium-90 in the power
reactors and in the burner reactor

Power of total system %% uWth
Yield of Sr-90: 4,1 3%

Number']
of Sr-40
atoms
Inventory in the steady satate
due to teta-decay only
1079 o
(eg. in sal L1 — = ar—
o
/' 8.m,~10‘q atoms
A ”
28
10 - inventory power reactors, mean
4 b il atoms
Inventeory in burner reactor
0. 461077 atoms
107 o
1026
Charging,
discharging
Production rate of in power
5r-90 reactors
2 ~7v1019 atoms/sec
107
Y T T T
9,01 0,1 1 1 1990 1000
Time, years
z aard ol T products: Ure-doand Cs=1137
1
Heacltor
nacarid
(B
Total
hacard
Larie
tranamat at fon 1
-
time, years
4
1 - —
-4
1 T T T T ¥ T

i, years
- 134 -

The amcunt of strontium-30, the most hazardous nuclide, in the
high flux burner i1s about the same as that found in the power
reactors after 3 years of operation.

However the most impressive result comes from considering the
end of the fisslon power area. Where by compared to storage
(natural decay with a combined half life of 39 years) without
transmutation, the transmutation case shows that the amount of
nuclides remalning will be reduced by a factor of 1000. (Fig.
4.22).

In a high flux transmutation the reduction by a factor 1000
would be achleved with 26 years in the lifetime of one reactor
generation,

4.7 Secondary processes

It must be remembered that the relatively high neutron flux re-
sults 1n the irradiation not only of the non-desirable radio-
active nuclides but also the stable fission products, including
the stable components of the fuel and structural material. A
very short review of these partivular processes is given here.

Natural chlorine contains two stable isotopes (underlined)

15 36 R~ 36
/8L Guy) OCL STTET s 64T
volatile
37 38 B _ 0
e N VA I F LV -t Y tata

However not only (n,y]) reactions are important here. Much more
important is the followling raction

35, 8-

35,
Cl (n,p) 18~ mg

T

—
N

‘1
yat
(see ch. / for experiment results)

The presence of sulphur also influences the chemistry of the
molten fuel (see ch.G6).
Sodium having only one stable isotope is also transformed

23, 24 B 24 25 26

fhis chailn gives rise to increasing amounts of stable magnesium
and over larger periods also stable aluminium.

2.9 Conclusions

The system proposed for the transmutation of °°Sr and '?7Cs,
fulfills the following criteria:

1. The energy halance 1s positive.

J. The hazardous waste balance 1s stronegly negative. That is,
the amount of hazardous material destroyed oreatly exceeds
the amount of freshly produced, e.g. tritium, '°B, and the
activation products of the structural material.

3. The rate of destruction (transmutation) is aporoximately at
ieast one order of magnitude greater than that due to spon-
taneous beta decay.

4. The periocd in which a thousand-fold reduction in the hazard
can be achlieved is the same as the lifetime of one reactor,
Chat 1s, 20U to 40 yr.

The neutron halance of the system is positive. That is, i1t
permits breeding to occur alone with the transmutation.

B. The weakest feature is shown in the relationship between the
nrobability of catastrophic release to the environment for
the transmutation operation, Pyring, to the probability of
a simllar event in the case of storage, where it is
desired that

Pstorg’
7

- 1306 -

where A, 1s the decay constant.

B

Further optimisations of the system are possible.

The proposed system, a molten-salt fast reactor, while rather
exotic from a technological point of view, is not as far re-
moved from the present state of technology as some other trans-
mutation proposals (e.g. high-flux high-energy accelerators,
controlled thermonuclear reactors) may be.
- 137 -

5. An internally cooled breeder with uranium-plutonium fuel

5.1 Design features and objectives

In this chapter a further variant of the molten salt breeder is
described.

The core 1s cooled by a cooling medium circulating in tubes. The
molten fuasl is 1Intensively miced. It remains in the core and no
fuel other than that being drawn off for reprocessing is present
outside the core.

The unique feature of this concept 1s the use of & molten fertile
material as the cooling medium for the core.

Such a dual function for the fertile coolant results in some
unusual properties for this reactor type. The study here is
concernaed with a molten chloride breeder reactor. The most im-
portant features are: (see table 5.1)

- thermal power: 2.05 GW - 1.94 GW in core + 0,17 GW 1in the
nlanket

- electrical power: 0.85 GW in the optimum case | = 0.4)

Derf

- molten fuel consisting of {in mol %)

15% Pulls (of which Pu-239 + Pu-241

80% and Pu-240 = 20%)
85% NaCl {(no 2?80CI1,in fuel)

1

and fission products in the form of chlorides or in an elementary
state.

- molten fertile material (in mol %)

nhe 7PfUCTS

A0% NalCl

and newly bred Pull,y and fission products

- conlant flowlng in tubes: made up of the fertile material above.
No other coolant in core.

- blanket material: also made up of the fertile material above

- the core is internally cooled. There is no circulating fuel
outside the core,.
- 138 -

Table 5.1

OBJECT . INTERNAL COGCLED FAST BREFEDER POWER REACTOR

REACTOR TYPE

CECOMETRY : INTERNAL Z0NE CURE, Fuel

(1) WALL : Iron, ‘lolybdenum
MTERMED SIS i . )
PHTERMEDTATE  Z00E cooling tube with fertile
WALL : =
EXTERNAL Z0ONE Fertile zone
WALL, REFLECTRO Tron

POWER (GW thermal) DR

POWER DENSITY (GW therm/m? core) : 0.07

NEUTRON FLUX, MPAN (n/cm?s) 7 x 171°

FIaSILE Pu-239/241

NUCLTDE -

DILUTENT NaCl, UCL.

COOLING SYSTEM @ Internal, in tubes by fertile material
DRECDING RATIO : 1.38
PARAMETER STUDIED NDesign

Tube materials

Pu/lU ratio

Temperature

METHOD GF NEUTRONIC @ ANISH 54

CALCILATION ;23 Sroups
P,oATD.
G

CODE

material
- 139 -

- the fuel and coolant flow concurrently (Fig. 5.1)

- the reprocessing plant is in close proximity tec the reactor
("under same roof")

- the fuel in the core, and the coolant is pumped with a velo-
city of 2 and 9 m.s™! respectively.

- structural material: possibly molybdenum along with small
amounts of other metals e2.g. Ni, Fe.

The advantages of the proposed reactor are as follows -

- no separate coolant, no "foreign” cooling agent (e.g. sodium,
helium etc.) in the core which results in a more satisfactory
system with improved neutron balance.

- the fuel inventory is very small due to lack of a separate
cooling system and the small out of core inventory due to
the directly coupled reprocessing plant.

- the fuel contains only plutonium and no uranium which simpli-
fies the processing technology and removes the danger of
uranium trichloride oxidation which alsc improves the cor-
rosion properties of this medium

- the high velocities of both fuel and coolant, significantly
reduces the temperature gradients in the equilibrium state
and reduces the mass transport mechanisms. These are very
sensitive to ftemperature gradients and play a large role in
corrosion.

However the diadvantages are numerous:

- the first and most improtant is of course corrosion. The mol-
ten chloride medium especially in neutron and gamma fields,
at high temperatures and velocities with chlorine being vir-
tually freed in tLhe fission process from plutonium chloride
presents a very serious problem which must, and probably could
be solved.

- the most likely structural material seems to be molybdenum
alloy which among other thing gives rise to parasitic abscrp-
tion of neutrons.

- the fuel is circulated by a pump which must be located in or
close to the core which increases the corrosion problems.

- the high fuel and collant velocities result in high pumping
costs and could cause severe erosion.
1o

i

hemat

b} [:

(=

rcuit

i

HE = HEAT EXCHANGER

cuT 1

CiR

CRCUIT [T

- 141 -

Table 5.7 Molten chloride fast breeder reactor (MCFBR)

"CHLOROPHIL"
Electrical power, approx MW(elect) 800
Thermal power, total/in core MW (thermal) 285@/1940
Core volume m* 8.75
Specific power MW om” 3 220
Core geometry m height2.02/2.36 dia.
Fuel: liquid PuClgq NaCl mol% 16/84
Liguidus/bociling point °c 685/1500 (appr. .
Fuel mean temperature OC 584
Fuel volume fraction in the core 0.386
Paower form-factors radial/axial 0.60/0.78
-ast flux, mean across caore ncm %s”! 7 x 10b°
“uel density at 984°C kg m 3 2344
Heat capacity, 984°C kJ Kg~ 1 k7! 0,95
Viscosity, 984°C ¢ em le! 0.0217
Thermal conductivity, at 750°C Woem b oK 0.007
Fuel sallt in core Kg 7900
Total plutonium in core/in system kg 2900/3150
Plutonium in salt weight = 36.4
Mean plutonium specific power MWth) kg_l 0.67
::i?rzlzgzziim specific power in MA(ER) ke 1 067
Coolant liguid U-238 C13/NaC1 mol % 65/35
Liguidus/boiling pint °c 710/1700
Coolant temperature inlet/outlet °c 750/783
Coolant volume fraction in the core 0.555
Coolant density kg m” 3 4010
Coolant salt in core/in blanket kg 19,500/165,000
Total volume (85 cm) m? 47 .85
fuel (shell side, pumped), velocity m s ? 2
Coolant velocity ms } 3
Number of coolant tubes 23,000
Tubes inner/outer diameter cm 1.20/1.26
Tubes pitch cm 1.38
Breeding ratio, internal/total - 0.716/1.3886
Coubling time, load factor 1.0/0.8 yr. years 8.5/10.5

- 147 -

Mean flux across core 7.0 x 10'°n cm™ %5

in centre 1.2 x 10'%n em 2?2571
Temperature coefficient of reactivity §k(%)/6T(°C)
- fuel - 3.8 x 10°°2
- coolant + 1.29 x 10 2

Reactivity wall fo vessel 0.72% {10 mm thickness).

SRR

5.3 Neutron Physics toceordine bo J0 Licou, 107

The 7 eroup transport calculation gives 125 cm {(8-18 m®) for tho
critical radius of the core with a blanket thickness of 295 eom
(36,42 m?). The detailed neutron halance is given in table 5.73.
The relative fluxes for each eroup are to be found in tahle 1.3
for core centre and core boundary. The corresponding one group
cross sections are given in table 5.4, In Fig. 5.2 the neutron
spoetra (flux per lethargy unit) are compared to that of the fast
critical facility ZPR - 3 - 48, This last spectrum is slightly
harder but the spectrum of the molten chlorides fast breeder com-
parcs facourably with that of a power LMFBR. From table 5.3 one
can drduce the following parameters

Ko = 1.384

Dreeding ratio Brcore - 0,716
Rblanket V. b70
HRtotal 1.386

For the eiven core (125 cm radius) the blankel Lhickness was variocod
botween /75 cm oand 115 cm. Fig. 5.4 shows the variation of brording
ratios obltained from the new transport calculations. The reactivity
and the core brecding ratlo remain practically constant in this
rankge making the adjustment of the core valume UNNECEeSsSsary.

Above 100 cm improvement of the breeding ratio by Increasing the
blanket thickness gives a poor return. For example to increase

the breeding ratio from 1.40 to 1.45 requires a thickness increase
of 70 em or a blanket volume increase of 32%.
143

Table 5.3

Neutronics of internal cooled fast breeder

Core atomic

densities

(atoms x 102%)

PU-239 6.5797 x 10 * atoms cm ?
Pu-240 1.5699 x 107"
U-238 3.5629 x 10 3
C1 1.8495 x 1072
Na 6.3017 x 1073
Mo 7.386 x 10"
Fe 5.078 x 10738
Blanket (coolant) densities (atoms x 1024)
U-238 5.4023 x 103 atoms cm ?
C1l 2.2718 x 1077
Na 3.457 x 10 3
Neutron Balance
Region Nuclide Atoms Abs tion Leak Producti
o . i (em® x 1021 orp eakage roduction
2138 (n,y) 22.57
. 56”79 5. .
. 3 25.30 (n,f} 2.99 6.23
239 [H,Y] 5.58 R
FPu 0.66796 34,56 (h.f) 28.98 85.55
240 (n,y) 2.24
Pu 0.1663889 3.78 (h.f) 1.54 4,772
Na 6.3017 0.26 -
(in fuel 1.10)
C C1 19.485 3.16 -
ore ! (in cool. 2.08)
Fe 5.0/78 1.30 -
Mo g.73886 Z2.04 -
F.P. 0.0687 0.50 -
Total core 71.10 27,40 38.50
238 {n,y) 23.15
U 6.427 23.70 (n.f) 0.55 1.50
Blanket Na 3.457 0.08 -
Cl 2272 2.272 -
Total blanket 26,00 2.9 1.50

Fig. 5.2

Chlorine absorpt. cross-asctions ENDF_B/III (Jan. 1972)

Absorption cross-section of chlorine

IT‘YT‘]

Disggs

: >
2 ]
= -
= E] \
/./
: iF
o s #
L o
(BN -
U e Lugdi 1o i luuLl L
S gevs (mb) 2 = °
8 2
va Fig. 5.4 ¢
welh Breeding Ratio C and blanket thickness t
( Core volume = 8.18 m’) o 148
i
Bisnket Yeiume ¥ wo— LR}
»- ’
Ref. Case 1.3
kB o
t {cm)
. i 1 A 1 L 130
n / " [ "e 1
L K2 b
Fig. 5.5 J
Sk Reactivity change with Core volume 1
b (Blanket Thicknesss95cm) 4
4 -
* 1% P~ - b 1%
L Ret.Case i
. | b
—=)c
. I 1 (") b .
(Y} ) 12
~1% | EERLS

144

1187

e’

Fig. 5.3 Neutron Spectra and Chlorine Cross Sections

(qu) %% 0 o =4
LA S SR BN LA 2R B S § T T T
L N - q
E 45 1.
- e = = K]
= i =
el -1
=

] e
I~ -

g

R

e 82 +
1) 2 -
= +« @& Cc L]
| €858 3

a o O <
[ 887 =
o e
S
F &8~ 3 § T
e &z °
PPRT & 2
I €068 ©

R +

T IAARASLE A

T

T

T T T

0.1 Mev

AINN AQHWHIZT ¥3d XM14

Radiatl distributions ol the l.'. g™
fies (Sph. Geom.)
transport calcuiations

Asymptotic mede In The core

i i P ——

s e 190 r (cm)
Fis. 5.7

Redist distributions of the 12'*11us P

and specitic pawer (Cyl. geom.)

On the other hand the reactivity depends on the core radius.
Fig. 5.5 shows the variaticn of reactivity with increasing
core size. Such acurve 1s very useful when 1t 1s required to
translate the cost in reactivity of a supplementary parasitic
absorbticn into an increase in core volume (or plutonium in-
ventory). The transport calculations used in Fig. 5.5 could
not be used to obtain the carresponding variations in breeding
ratio, rather they have been calculated from the information
in table 5.3. (assuming small core volume variation < 10% and
no important changes in spectra).

This gives

1) BRtotal - 1.386 - 2.45 8k (breeding in core unchanged)

The effects of modifying the core composition can be evaluated
by the same method. This is arranged to avold altering the
preperties of the coolant and fuel by varying the coolant tube
diameter. Ihe reactivity is very sensitive to this parameter.
If ¢ and n are respectively the relative increases in fuel

Table 5.4 0Une group cross sections based on the reference
22 group transport calculation

(barn = 107 2%zm?)
Type aof Core Soectrum Hlanket Spectrum
Nuclide o Or v 9 Og v
238y 0.249 3.30 1072 | 2.746 || 0.326 7.71 107% | 2.715
239 0.329 1.709 2.951 || 0.536 1.841 2.917
240py, 0.527 0.364 3.052 || 0.790 0.174 2.998
Na 1.62 103 - - 2.044 1073 - -
- .38 1077 i _ || 8.83 1073 i i
(8.64 107 3)* (9.11 10 %)+
Fe 1.01 1077 ~ - 1.43 1072 - -
Mo 0.109 ~ - 0.166 - -
Fp 0.225 - - 0.362 - -

* These valuos were computed on the bases of ENDE/A. IIT data.
- 146 -

and coolant volume for a constant pitch are as

- 1.44 ¢

1 + §.23 x 107 %¢ + 0.9027 n
17 + D.2/56¢ + 0.3944 n

3
1l

1e

and k

which gives

Sk - - 0.924 ¢
n = 1.56 &k

The corresponding relationships for breeding ratios (BR) are

12

0.716 - 1./72 6K

core
BRblanket - 0.670 - 1.05 dK
1 - 1.336 - 2. SK
&Rtotal 1.336 VAV

It can be seen that the penalty on the total breeding ratio for
the same 8K is anly slightly greater while the penalty on the

increase of plutonium inventory is five times lIess. Therefore a
reduction in the diameter of the coolant tubes is preferable to
an increase 1in coolant diameter, provided of course that an in-
crease in coolant velocity is admissible. This last assumption

is implicit in these calculations since the coolant density was

kept constant.

In the whole system the chlorine absorption represents 5.38%.
We have seen that the GGC-3 values are different from the more
up to date ones (ENDF/B-T1I1). On the basis of these new cross
sections given in Fig. 5.3 (curve 6] and assuming that the re-
ference spectrum is unchanged, a computation of the one group
corss secltions pives 8.40 mh instead of 6.353 mb in the care,
and 9.11 mb instead of 8.43 mh in the blanket. In this last re-
cion the spectrum is softer and the increase of cross sections
in the eneregy range [(E A 0.8 MeV) is almost compensated for by
the decrease at bthe lower energies (10 keV - 0.0 MeV).

The total absorbtion by chlorinn for the whole system is B.5/%
ingstead of 5.38% giving a loss of reactivity of 1.2%. This loss
could be replaced by a 1.8% increase in Pu inventory 1f a very
small decrease (0.65%) of the conlant tube diasmeter is accepted.
Otherwise by changing only the core radius a greater Increase

of Pu inventory (10%) is required (fFig. 5.5)
Fven with this latest data the problem of parasitic capture 1In
the chlorine 1s not dramatic and there 1s no reason to believe
that there is no need to enrich the chlorine *’Cl. This is con-
sistent with the conclusions of Nelson. Fig. 5.3 eclearly shows
the importance of the energy distribution (spectrum corresponds
to the chlorine cross section minimum (65% of neutrons are 1in

an energy range where opp <€ 5 mh). This fact was not perhaps re-
copnised 15 years ago when fine spectrum calculations were not
possible. This could explain the pessimistic conclusions of
several eminent physiclsts.

For the molbydenum alloy chesen (20% Mo) the reactivity penalty
(2%) is guite acceptable. However the cost could rapidly become
nrohibitive if the volume of the structural material and/or the
molybdenum content should increase for design reasons. In the
context of more detailed design studies this polnt may become
more important than the definition of the proper chlorine cross-
sections. It does however seem likely That the molybdenum cross
sections used in GGC-3 were overestimated.

In the simplified calculations, no core vessel was allowed for
at the core/blanket boundary but all the required information

is available - fluxes, one group cross sections etc. (Table 5470,
Using the same alloy for the vessel (20% Mol the one group ma-
croscopic absorbtion corss section is 2.56 x 10 Jem”™! giving a
loss of reactivity:

where e is vessel

o, — P
Sk (%) U727 e thickness in cm.

For 19 mm thickness a value of 1.37% is obtained for loss of
reactivity which would have to be compensated for by an increase
of core volume of about 10% i.e. about 9 m?® instead of §.18 m’.

A hetter solution would be an increase in the plutonium inventory
of about 2.1% in the reference core (loss in breeding ratio

1.336 >~ 1.350).

Transport calculations (GGC-3 + SHADUK]) have been made for dif-
forent coolant fuel densitites, different temperatures [(in this
rase with the density constant to determine only the fDoppler
offect). The reactivity changes with respect ot the reference
core are given in table H.bh.
- 148 -

Tnble 5.5 Reactivity changes

Type of modification Parameter Reactivity changes
~1.Y (Keo)
51 - 5% -3.0%
Fuel density 5. 3 { 1.1 (leakage)
Coolant densit 59 .77 200 ke
- 3% . e
poient geEnsitby -1.23 (leakage)
Fuel temperature +300°C 0.01%
Coolant temperature +300°C -0.14%
Ve high
Full loss of coolant - #17 ne f 0 VEDY TLER ON

Koo [ e GJ

The partlial changes 1in ke or leakage are only approximate but the
total reactivity changes are evaluated directly and are therefore
more precise.

The Doppler effect in the fuel is quite neglicgible due Lo com-
pensation between the capture and fission praocesses.

The effect of full loss of coolant is large and positive but
considerably lower than might be expected from crude calculations
(kee Changes in the reference spectrum).

From table 5.5 one can deduce the feed back effect which is very
important for kinetic studies

o - ) _nlbkl | S IV N R
k p fuel o tu=l pocoolant SR o

The vold coefficient of the fuel (1st therm) is strongly negative.

1% voild
CLENEEPEEE
0

gives a 0.6% loss in reactivity. If boiling occurs in the fuel it
will be rapidly arrested by a decrease in reactor power.

Lo oand
I+ one considers that all density modifications come from thermal
expansions (liquid phase only) one can define general temperature
coefficlents.

The thermal expansion coefficients are

CS _ _ - -3 -1
(557 fuel - - 0.853 x 10°° K
(90 ) - - 0.89 x 1073 K71}

Replacing these values in above given equation leads to the fol-
lowing expression

°) T - "2 (3 "
(%) 3.6 x 1077 (8T) . 4o+ 1.29 x 1072 (8T o

in the second term the part played by the Doppler coefficient
(4.8 x 107%) is quite negligible. For + 100°C in the fuel the
loss of reactivity is - 3.8% which is very important from the
safety point of view. Compared to any kind of power reactor
(even the BWR) the advantage of this kind of reactor is qguite
evident.

For the Nelson (198687) value of the thermal expansion -3.107°
instead of -6.3 x 107" one gets - 1.8% which is very close to
the Nelson result -1.5%.

If we postulate an accident condition and assume that the same
increase of ccolant temperature immedistely follows the fuel
temperature rise, the overall change in reactivity is defined
by:

This important isothermal and pessimistic coefficient 1s still
nepacive. Nevertheless during a detailed study of this reactor
concept 1t would be necessary to check the values of the ther-
mal expansion coeffiliclent for fuel and coolant more carefully.
The relative value of the coolant term which is positive might
prove to be too high 1f the differences between fuel and coolant
became too marked. This problem did not arise with the present
data.
- 150 -

In a spherical assembly the fluxes in the core are given to a
good approximation

- sinBr

AL fE)

dlr,E)

where the space function is called the "fundamental mode” (soclu-

tion of

V2y+B2y = 0

in spherical geometry). The critical buckling B? is obtained from
homogeneous calculations based only on the cross sectlon data of
the core,f(E) is the asymptotic spectrum which is space indepen-
dent far from the core bhoundary. For the same 22 energy groups
one gets: B? = 4.08 x 107" cm™?. Using this value a good fit of
the "exact fluxes” have been obtalned from the complete trans-
nort calculations (Fig. 5.6). The asympototic fluxes cancel for

= 155.5 cm

whero R, is the extrapolated radius. By definition the blanket
saving 1s given by

where Re 1s the core critical radius. The blanket saving depends
mainly on the nuclear properties of core and blanket and on
blanket thickness. However for thlicknesses greater than 60 cm
this last effect is very weak. Finally the shape and size of the
core have almast no influence on this saving. This parameter,

for this reason so important in reactor physics, will be used In
the next section for the one dimensional cylindrical calculatlions.

The axial blanket thickness 1s taken to be equal to the radial
thickness (95 cm), and the core height as H_ = 200 cm. The criti-
cal radius of this cylindrical core has to be determined. The

two dimensional transport calculations are too expansive (and un-
safe) and only one dimensional calculations have been made, whilch
is sufficiently accurate. Axial transport calculations are not re-
quired since the balnket saving is known from the spherical geo-
metry calculations. One can therefore assume the following flux
shape.
- 151 -

®(r,Z,E) - (c0s RZY(r,E) for any
r value (including the radial blanket) and

H H
_c
Zz

IA
™~
A

_c
-2

-8? is the axial buckling computed from the extrapolated
height: H8 = HC + 28 = 261 cm which gives

g = fil = 1.204 x 10 2em™ !
[

(B2 = 1.450 x 107 %cm™2)

The computation of core radius and spatial distribution have been
made with the SHADOK code (cylindrical version) by introducing
axial leakage defined by B?. Befare that a first approximation

is obtained by introducing the radial buckling a?, that is to

say assuming for the core only, the shype Y¥(r,E) = J (ar}f(E)
where J_ is the usual Bessel function. One obtained a? = B? - g?
where the critical total buckling is 4.08 x 107% cm™? which

gives a’ = 2.63 x 107" ecm™? and a = 1.625 x 1072 cm~ 2. Then the
extrapolated radius of the cylindrical reactor is

R = = ~ = 148.0 cm.

Finally with the previcus blanket saving we get a core radius
proper of R = 117.5 cm. The direct transport calculations with
SHADOK -code gives Ry = 118 cmi This cleariy indicates the value
of the blanket saving concept. Nevertheless these transport cal-
culations are still necessary because they give the radial dis-
tribution of fluxes and specific power over the whole system and
more detalled informations. Fig. 5.6 shows some of the radial
distributions of flux and specific power. The energy production
in the blanket is quite small (1.7% of the core power) because
no fissile materials are present (and only fast fissions oceur
in ??%U). In praectice it would be higher (say 5%) since the re-
processing process would not be able to remove all the fissile
nuclides produced even with continuous fuel (coolant) reprocessing.

The radial form factor for specific power distribution is, for
this core

mean pPower

o = ; 0.60
r maximum pPOower

Note; it would be possible to improve this coefficient by choice
of different lattices particularly the most reactive at the
peripheral region.

The axial distributicon of specific power is given with a good
approximation. If the axial mean value is unity then this dis-
tribution is:

BH
P(Z) = —_EJEEE—;OS B7
o
28in—
] juad l = T[ = P 1 2 Fl
wilth R v T o8 1.7 x 10 Cm
3; o
and H_, = 200 cm giving
P(Z) = 1.284 cosn Z—é%”—;[_mo < 7 s 100
and a-, = 0.78 Ffor the axial form factor.

This axiael distribution 1s very close Lo that used by Nelsan.

The critical volume is higher for cylindrical geometry 8.75 m’
compared to 8.18 m’: this increase was expected.

Usually the number of energy groups required for a good defini-
tion of neutron spectrum is at least 17 for fast critical assembly
studies and 22 can be considered desirable. Therefore a 27 group
cross section set has been prepared, the code GGC-3 which allows
99 pgroup calculations for a rather simple geometry has been used
for this condensation. The cross sections were produced separa-
tely for core and blanket and the scattering anisotropy was
limited to Py which 1s sufficient for this reactor type. Most

of the GCC-3 library data were evaluated by G G A before 1967

but some are more recent.

- Iron - evaluatoed from ENDE/BT data (Feb. 1363)
- Molybdenum - evaluated from isotopes of ENDF/BI date (July 1363)

- Pluteonium 239 - evaluated from KFK -750 Resonance Nuclide
{(Feb. 1969)
New data concerning chlorine absorbtion cross sections are
available at EIR (Fig. 5.7) they are obtained from ENDF/B-III
(Jan. 1972). Unfortunately this information came too late to
be used for the transport calculatiocons. Fig. 5.3 (curves 5 and
6) shows that the GGC-3 values were underestimated above 0.6 MeV
and overestimated between 10 keV and 0.6 MeV. The effect of this
on the reactivity is not great. Taking also the molybdenum cross
section from ENDF/B-III one can see that the GGC-3 values are
too high. (experiments made with molybdenum control rods in fast
critical assemblies could not be reproduced with ENDF/B-T1 which

makes a new evaluation of data necessaryi.

Fission product data are from Bodarenko
cross sections for resonent nuclides are obviously shilelded.
theory is included in the GGC-3 code,
shown 1in table

Nordheim

(1961)

some special data as

(1664).

The absorbtion

The

it requires

Tanle 5.6
Resonant Atomic density - . R R
T (K] C a om 2 ar
Nuclide R. Np (x102%) R R i " "
238, \
65.47 107 10472 0.58 925 215 7.10 20.
(Coolant) ’ x ] - c b3
2395,
1,725 x ’iD—?’ 1257 0.44 83 25.3 23.0 1/8.3
(fuel)
4.32 x 107% 1257 0.40 83 105.0  [111.7 731.7
(fuel)

- 154 -

5.4 Gafety Problems, Comments

The molten chlorides reactor seems to be a relatively safe system
due to the following rasaons

- an extremely high negative temperatur coefficient of reactivity,
since during a temperature rise part of the liquid fuel 1is
pushed out of the core into a non-critical geometry buffer tank.
The dumping cof fuel in case of an incident is also possible in
an extremely short time.

- 1n a more serious incident when the fuel temperature increases
to 1500-170009C (depending on external pressure) the fuel hegins
to boil. The vapour bubbles give rise to a new and unigue, very
high negative "fuel vold effect”

- the leakage of fuel to the coolant 1s probably not a serious
nroblem because the coolant is continuously reprocesced.

- the leak of coolant to the fuel for the same reason cannct
cause large problems (provided the leak remains small).

A rahter adverse property of such a molten fuel reactor is the
need to Initially heat the sclidified fuel in a non critical
perometry with external power. (e.g. from the electrical grid).
This problem has been fully overcome in the case of the molten
fluoride fthermal reactor (Oak Ridge National lLaboratory).
5.

CHEMICAL AND RELATED PROBLEMS

6.

Physical and chemical criteria for salt components

The limiting criteria in the search for fuel, fertile material
and coolants for internally cocled systems are as follows

1.

2.

10.

1.

small elastic scattering for fast neutrons (Fig. 6.1)
small inelastic scattering.
low neutron capture cross-sections for fast neutrons

thermodynamic and kinetic stability of plutonium and
uranium compounds (Fig. G.2, t£.3, 6,47,

melting point below 700°C in the pure state or in the
dissolved state (Fig. B.5, 6.6, 6.7, 6.3, 6.9, 6.10).

boiling point above 1500-1600°C for both pure and
dissolved states {(low vapour pressure) [(Flg. G.117.

stability against atmospheric constituents, oxygen, water
carbon dioxide. (Fig. B5.727.

good heat transfer properties and specific heat capacity
(low viscosity, high conductivity ete.) (Fig. 6.13)

good corrosion properties if possible (Fig. B.14)
adequate technological or laboratory experience.

relatively cheap.

These wide ranging criteria are fulfilled best by the following
compounds

PuClq, UClR, NalCl (Table 6.1, 5.2)
Flectronepativity Fnergy Decrement

I
Aot i Li— T T
' ST A AR AEET VNN 20T on 2F
i 1A piues rar sl "o solutivn
5.1 Frey B.4
Lol Fael
. trerial Tompaonent s
v 4 4] .
* ¢l
N Lo tely
i R B
i -~ 1 1 Fell,
B J
N r 4 A'l ¥ &2
I 1 i
a4 L uel
- B - — onn o h
. - ool
Ul
veJl ]
PR i) -
S tacl, Ladl [~
] s - Anc \
L [N e~ sadi Ao ool
I -+ 1
- all L
- 4 [ S
) — -
B
— -+
- A | + RINEIE
- —tan

. . -
Fie. 6.6 delting Point of - - Selected Salts }‘lg . 0.5
v s Chloride
Fluoride 11004

T T T mole fraction

mol fraction

Fig. f.7 Furl,-Alkali Metal Chlorides Fig. 6.8 Phase Diagram for PuCl, /NaCl and ¥C1,/NaCl

342%5°C

Coolant

TEMPERATURE (°C)

700_
Fuel
o
w
o 4
2
~ 600
<
w
o B
=
w =4
—
500+ UCls solid
b i
| |
] | {
| NacCt , solid ! |
b
400 T N T T T T T T LOO Yy Y T Lo 'B]ST T T fi[
0 1 20 30 40 50 60 70 60 90 100 5 101520 40 60 80 100
PuCly in SALT (MoL%) NaCl Mol%  UCls

- 158 -

4o

Fig. 6.9 The System PuCl -WaCl-iCl, Fig. 6.10 Phase diagramme with thorium

00 1Y)

50Q,
NaCl

NaCl

PuCl ucl

Thel, PuCl}
Fig. "3l Vapeur pressure - metai chioreies “ig. .t CHLORIDES-OXIDES EQUILIBRIUM
DIAGRAM AT 1000K
Cs ta
- 400+ °
CHLORIDES MUCH MORE
STABLE THAN OXIUES
Crile  NiCL MnGly ZrCls L1750 ¢ CHLORIDES Id
815 (987 190 J 1207 1490 C mg' a EQUIL:IBRIUM
Y N P2 Na ) WITH 041,
y .
= ~300+ Pr
o
2 '
104 2
.
N ir ®ite
— ) ug® Pu
- S -2004 T OLIDLS MUCEH AUEE
104 -~ 5 ® Al SVABLE THAJ Cho o RITEG
w
4 x
2 o
a cr ®
6 w o X ® Mn
10 4 x Maxmmum L2 Fog !
tuel temperature -100 . ® .y
in this reactor
-
104 ¢
[ ]
Mo
T T T T Y Y T
16" -100 -200 - 300
v v v y v T v v
200 400 600 800 1000 1200 1400 1600 aat%%  ox1pes (k3.1/2 mo17l0)

r
Temperature  [°C) o
- 159 -

Pig. 6.13 Salts properties at 650°C v. Chemical composition Data
derived from Nelson for UCIB/PuCIB/MgCLZ/NaCI

0014 4
) 0-0086{[wemaeg’]
0-00A
1-031
083 [%Jd'f’]
Ce
0634
' o
l_ B
Lo
454 VISCOSITY COOLANT ‘g o
' FUEL & €
3 4 a "
T] 404 Eo;—an s’] w8
< Q
35 i 58
From Nelson (interpolated)MgCy °3
{ o
~
35 3
[g.cm'fl ! OLANT O w
P 30 DENSITY I FUEL
254 ]
T T T T
0 20 30 40 50 [ua %)
Fig. 6.14 Free enthalpy of formation chlorides
01 octe

300 sbo 1000 500 2boo
TEMPERATURE [K]
Table 6.1 Properties of

- 160 -

fuel components

PuCl UC1
u 3 3 NaCl
Molecular weight 348.3 347 58.4
Postulated molar ratio-fuel 0.15 - 0.20
- blanket material - 0.65 0.20
Density solid state (kgem™3) 5.7 5.57 2.14
Melting point (°0) 767 835 800
Boiling poént at atmospheric 1730 1790 1065
pressure ( C)
Melting enthalpy (kJemol~™!) 64.0 64 28
Enthalpy of vapourisation
- 240 300 188
(kJemol™ 1)
Temp coeff. of density (K™1) 0.0010 0.0010 0.0005
Specific heat (Jemol~! K71 140 140 77
hermal conductivity
(Weem™ ! deg™ 1)
Viscosity (geem™! 71 Fuel: 0.025 Coolant: 0.045 0.0143
:ET?]COQ$€. of viscosity 0.0005 0.0005
N _ £
ree enthalpy of formation 70 575 990

a4t 1000 K (kJemol™ 1)

Table 6.2

- 161 -

Other chlorides of plutonium

and uranium

Melting point ("c)

Boiling point (70)

Froe enthalpy of
formation at 1000

(kJ/mol)

Plutonium Uranium
PuCl e LUC1 ucl
4 4 g 5
All efforts to produce
pure solid PUCI4 have
been unsuccessful; 550 (287) 178
only in gaseagus
atate with free 747 (417 (372
chlorine, or in molten
salt solution or in
aqueaus solution as
complexes
T =180 = =760 Ax-18 x-165 Ex-130
= -/B8 = -B70 = ~/8(

- 167 -

5.2 Corrosion of structural material

H.7 Ceneral criteria

It 1s clear that one of the most problematic areas in molten salt
reactor technology i1s the area of corrosion. Some criteria can be
formulated as follows

- Lhe free energy of chlorides formation for structural materials
must be relatively low, significantly lower than those of pluto-
nium and dJranium chlorides but still lower than those of the
main fission products

- the partial pressure of the chlorides formed from the structural
materials must be rather low which corresponds to & relatively
high boiling point for these chleorides (Fig. 6.16)

- the neutron capture cross secticn for (n,y), (n,p) and n,2n)
must be low (see later]

The structural materials are in principle different for the two

typrs of core discussed.

- internally cooled, using tubes plus the effect of the cooling
agent

- externally cooled by pumping the liquid fuel cut of the core

These two variants call for different structural materials and
different requirements

hoat conductivity .
: mechanical

Cocling method of the structural A
L, behaviour
material

internally by tubes veTry  good very pood

(thin wall tubes)

externally not important not so Important

f formation (KJ mol™h)

Free energy

Fig, f.16 chlorides = b ling p(vh_‘t v
Free enthualpy of Pormation
ch’ .PuCh
J SALT- FUEL
COMPONENTS
15004 M’Clz N.aCl
ZrCla
[
)
[\
LY
3
? Wely
:5; PtCle
- 5004 AuCl
0 100 200 300
Free Enthaipy of Formation Afix [KTMOR:‘]
Fig. 6.13
0—
50
-1m_
150
H,0
HZCIZ
~-200
1/2 00,
- 250 ————
300 500 1000 1500

Temperature K

163

FREE ENTHALPY OF FORMATION

Periodic Table

[K.T- mol“Cl]
o
bl

80 se0 T~ T idos 1500

Motybdenum —Metal "
Melting Point-2610°C (P=1022 g.cm
Boiling Point-5560°C

o4 —— - - =

04

(4lpropcnimnion

1
~3
o

r

1000°C

TEMPERATURE °K

Fig. 6.20 Pissjon Products in Molten Chlorides Media

36 Kr | 54 Xe fast
o Gas 1
35 Br 53 1 FPE Extraction °4°%
very slow
34 Se 52 Te
33 As 51_sb FPS Volatile chlorides
32 Ge 50 Sn
31 Oa 49 In FPS Low volatile
30 Zn 48 cd chlorides
47 Ag
46 pd Non
45 Rh FPE volatile
44 Ry metals
43 Tc
42 Mo Low
41 Nb FPS volatile
chlorides
40 zZr
64 ad
63 Eu
62 Sm
61 Pm Non

volatile
chlorides

60 Na | FPA

37 Rb |55 cs FPA Low volatile

chlorides
- 164 -

From these and other criteria the following choices may be made
- for the internally cooled core tubes: molybdenum

- for the wall of the gspherical core, in the case of the
externally cooled reactor: graphite and berrylia.

6.2.2 Molybdenum as structural material

The main corrosion processez result from the followling mechanisms
(m = metallic phase, s = salt phase, Me = metallic component of
irradiated fuel or coolant)

2
X

> frd
%4

Mfa) +
(m)

For the behaviour of fresh fuel PuCl., in NaCl tho most likely
reaction is (T = 1250 K]

AG T =+ 450 kd/mol C1.
The equilibrium constant of this reaction is small and nguals
10717 50 that this reaction is completely unimportant.

In the bionkelt zone the most dangerous reaction is connected with
uranium trichlorides (chlorine from the fission aof PUC]H].

the nontrol of the UC1,/UC1 ., ratio in the fertile ccolant might
he frasible due to the continuous reprocessing of this material
togother with the cantrol of zirconium from the fission products
oxidation state.
An additional problem comes from the fact that molybdenum has
different oxidation states +2, +3, +4, +5 and all of them have
the corresponding chlorides. (see Fig. 6.17)

Futher difficulties arise from the problem of the reactions
between metal chlorides and oxygen and water.

These reactions (for oxidation state +7) could be written in
simplified form

The metal oxides are mostly insoluble in molten chlorides which
results 1n a serious disturbance of the fuel system. From this
point of view the metallic elements could be divided into three
classes (see Fig. 06.12).

- those which are stable with H.,0 and 0,, that is the chlorides
are more stable than the oxides (e.g.LNa, Cs, Bal) and partially

Ca.

- those which are not stable with H, 0 and O, and the resulting
nroduct is a mixture of chloride, oxychlofide and oxide (e2.¢.
Pu, U but also Zr, T1, Al, Fe, Cr, Mn, Mg - this is the most
numerous group of metals).

- those in which chlorides are converted to the most stable oxide
in the presence of H,0 or O, (e.g. Mo, W) metals of this class
seem to bhe less numerous than thaose in the other two classes.

This property causes the rapid elimination of traces of water or
oxygen in the molten chloride salts of Pu and U. It is also well
xnown that traces of H,0 and 0, have a very big influence of the
corrosion rate. Molybdenum belongs to the last class, the oxide
is much more stable than the chloride {(Fig., B.14)

This means that the traces of exygen or aven water will result in
the production of molybdenum oxide. This effect requires consider-
able further study.
— ’] 6 [“) ~

5.2.3 The irradiation of molybdenum and iron in a fast high
flux reactor

The high neutron flux irradiation causes physical and chemical
changes 1n structural materials.

Molybdenum is a mixture of stable 1sotopes. The most
important by-product of neutron irradiation is the Tc-99 beta-

emlitter with ty, = 2.1 X 10° years and belongs to the decay chain
shown here.

Mo -498 R B
- , i - 94 — -94
(;%Q] (n Y] 10 'LU = (_]Dh TC _t_’!r? _ > ,I % ,lflsv
(n,y)
g .
~ - 400 ; -100
T P00 y I Ry T

For approx 13090 kg molybdenum in the core in the form of coocline
tulbies or about 13,300 moles. the Mo-898 gives 2300 mol.

The irra-
diation rate N (mols/s! equals

r\'f}’,_Lfi‘A"‘
RIS e 40 23 W =D 7 L1 R
N = (2LANTEY) s (Bx1073) o (I0x10727) it e = 1 2k 0 Tatam/s

After an irradiation of /00 hrs a steady state concentration of
Mo-98 is reached

ByeYe i b7
rteagy T TEomeT 7 3x107'atons - 0.005 mol
state

The radioactivity of the Te-384

after three years irradiation of 1000 kg of molybdenum in the fast
reactor core:
- 167 -

Activity Ig—izarJ = 1.2x10" "atoms/s x (3x3.7x107s/year)
107172
T T 3 Curle/tonne of Mo

The diffusion rate of hydrogen from the molten fuel to the
coolant and blanket {(here also UCl3 - NaCl) must also be
mentioned.

Une can assume that this melt containing hydrogen is saturated
so that the porosity of the wall {(molybdenum) will play a minor
role. The most important factor is the variation in the mechani-
cal properties of the molybdenum caused by the uptake of hydro-
gen.

The problem of molybdenum corrosion in chlorine containing media
is particularly complicated by the numerous molybdenum chlorides:

MoCl.,, MOCIB, MDCld, MDCIE (Fig. B.17)

A0

Fission product behaviour in the fuel

The fission of Pull, causes the formation of two fission products

' oand BT and three atoms of chlorine
PuCl, S O T

For the fissloning of 100 atoms of Pu the following balance has
been sugegested

Li;i; 0,008 Se + 0.003 Br + 0.842 Kr
3 gas gas g8

+ 1.05 RbC1 + 5.43 5rCl, + 3.03 YE13 + 21.5 ZrClq

100 PulCl

+

0.29 NBC1_.(?) + 18.16 MoCl, + 0.28 MOClS + 4,01 Tcm
Fa

5 et

+ 31.45 Ru + 1./73 Rh + 12.066 Pd + 1.88 AgCl
m m m

et et et
+ 0,66 CdC17 + 0,06 InCl + 0.325 SnCl, + 0.687 5bC1,

+ /.0b TeCl, + 6078 1 v 21023 Xe + 13.34 CsCl
z £as oas

+ d050 Ball, + 5.74 LaClg + 13,98 Cefilq + 4,28 PrCl3

+ 11.87 NdCl? + 1.44 Pm[filq + 3.74 SmCl% + LB EuCl

+ 0.03 CdClB

the average balance of fission can be represented in the fol-
lowing manner

Wl ogg een

100 PUC13 15
- 1RA -

for

Table 6.3 The behaviour of fisslon products in the molten
chlorides fuel. (Yields given represent products
for 100 Pu atoms fissioned).

Principal fuel Fission Remarks
and structural products
material
04 -Mo metal Pd(12.66), Tcl(4.01), Rh(1.73),
(struct.) 2™ Rul(3.744), 1(5.17), Te(7.865)
- Xel21.2), Kr(0.%4), Br
Q

— &

— 0}

- ol e no chlorin=e

. c é the synthesis of

= il possible metal

£ . chlorides
L] D—_
]'_:)} -
- 14 =facl -1 -MeCl (18.18)
7 X "
e (corrosion -MoCl., (0.28) 0
3

c product) c

N o

T in -AeCl (1.;8) 3

| ; —Sb613 (0.57) o
| -CdC1. (0.66)

: | 501G (0.32) | B
5 | -InC1" (0.08) "
12 !
£ Couct. (1) ad
(, 200 g Eh
- i
"+ -UC T, -ZrCl, (2.15) | 5°
e - - no0
Q - L &
- —Lr612 I
CL C m
— .
- -Pull, -YCl, (3.02) | %o
- - -PrCl, (4.28) o
s -EaCl, (1,068 =
N —C@Cl% (13,94 .
5 -LaCly (5.78) 0
& 3

L {

E
A0+ -NaCl -RLHCT (1.05) 5
-CsC1I (13,75 I
-omCl, (3.73)
-S5rC17 (5.48) -
-Ball, (8.50) )

From the earlier published data (Chasanov, 1965; Harder et al.,
19685 Taube, 1861) it appears that the problem of the chemical

state (oOxidation state) in this chloride medium for the fission
product element constituent requires further clarification.

From a simple consideration it seems that the freeing of chlorine
from the fissioned plutonium is controlled by the fission preduct
elements with standard free enthalpy of formation up to ~ 20 KJ/mol
of chlorine, that is up to molybdenum chloride. The more ’noble’
metals such as palladium, technetium, ruthenium, rhodium and prob-
ably tellurium and of course noble gases: xonon, krypton plus
probably iodine and bromine, remain in their elementary state
because of lack of chlorine. Molybdenum as a fission product with
a yleld of 18% from 200% all fission products may remain in part
in metallic form. Since molybdenum alsc plays the role of struc-
tural material the corrosion problems of the metallic molybdenum
or its alloys are strongly linked with the fission product be-
haviour in this medium.

The possible reaction of UCly and PuCly with Mool

resulting in further chlorination of the actinigdes-trichlorides
to tetrachlorides seems, for PuCly very unlikely (AG'?00K -

- 450 kKI/mol C1) but this is not so for UCL,.

A rather serious problem arises out of the possible reaction of
oxygen and oxygen containing compounds (e.e. water) with FulCl
and UCls which results in a precipitation of oxides or oxychlo-
rides. The con®incuus reprocessing may permit some control over
the permissible level of oxygen in the entirs system as well as
the contiinucus gas bubbling system with appropriate chemical re-
ducing agent.

Corrosion of the structural material, being molybdenum is also
strongly influenced by the oxygen containing substances, A pro-
tective layer of molybdenum however, may be used on same steel
materials using electrodeposition or plasma spraying technigues.

Note that all these considerations have been based on standard
free enthalpy: but even a chanre in the thermodynamic activity
trom = 1 to = 0.001 which means a change in free enthalpy of
174kJ mnl~ ! thus appears Insignificant as far as these rough
calculations po.

In the fertile material reactions also occur and the most impor-
tant are

fission process: Jd€l, ——= Fiss.products + 3C1
. 1250K
oxldaticn process: UCl3 + V2C12 — UC14; AG “7 =25k emo; !
: : o -
disproportionation UCl3 + 3UC13 3UCl4 + Umet
- 1/n -

t.4 Scme comments on reprocessing

Breeder reactors as is known form part of a breeder system which
includes not only the power reactor hut also the reprocessing
plant.

The advantages cf molten salt bresder reactors become particularly
apparent when the reprocessing plant is under the same roof as

the power reactor and when chemical separation processes take
place in the high temperature molten salt media in a continuous
cycle. From Fig. 6.2° it can be seen that all fission products
might be classified into 3 classes.

FPA = fissiaon products of alkali and alkali earth but also
rare earth elements which have free enthalpy of chlo-
ride formation greater than those of PUCIB.

FPS = fission products of seminoble metals with free ent-
halpy of formation smaller than those of PuCla.

FPE = fission products existing in elementary form because

of the low free energy of chloride formation oar
negative balance of chlorine.

The separatlion of plutonium or Uranium form the irradiated fuel
by means of pyrochemical technigues cculd be carried out for
oxample 1n the following way.

Molten salt, primary phase Pu, 7 [(part of I remains)
Metallic phase (part of FP remains)
MJolten sall, secondary phase containing only Pu.

This 1s the so called metal transport process. The proposed
schematic of separation processes utilizing metal transport is
given in Fic., 65.21 and b.22.
Fig. 6.21 Fuel reprocessing flow Scheme

FROM REACTOR

PRIMARY SALT PHASE
irradisted Fuel
UCly  PuCly
FPA+FPS+FPE

in NaCt

PRIMARY SALT PHASE
irradiated Fuel

PA Chlori
\ METALLIC PHAS SALT METAL in NaCy o g Ol \
Mg Metat in EXTRACTION
Notten Metal RED!;COL‘-ON
Tex 8007 ME TALLIC PHASE.

Mg in molten metat

Ymet Plimet
0FP§n + FPE

ME TALLIC PHASE
ME TAL SALT
49 m moltenmetal | EXTRACTION SECONDARY SALT /
* OXIDATION PHASE
3 [Pha FPE et T = 800°C MgCly 1n NaCl
=
3
=
O
' z' PuCIgNoNcglCl:
n Na
z|2 FRESH FUEL
o iBrELE
MA
g‘ SALT PHASE PRE PARATION MOLTEN FUEL
ucly in T = 700°C
NaCl FERTILE MATERIAL

T0_REACTOR

FPA — Alkali and Alkati earth fission products e.g- Cs Ba, Sr
FPS — Semi and noble metals and metal chiorides

FPE — Noble metals in metallic states and noble gases

Fig. #,20 Tuel Heprocessing Mate T obala o

IRRADIATED FUEL
FROM CORE

379 Salt 7' K
210g Py s CONTINUOUS 00229 Pu-s
00459 FP- 5! FUEL REPROCESSING

(50% EFFICIENCY FOR

ALl FP)

370 St Pu INVENTORY ~ §00kg
70 Sait-s-
22gPu. st RENTENTION TIME
0022gFP- 3 DAYS

-t

0-0225g FP-s

FRESH FUEL TO dituted in

v
THE CORE Tesgsan B
o
24
FROM REACTOR
o
£
21-6gSalt-s” =
1359 U-s~! v
00860 Pu-s” CONTINUOUS :
0-005gFP- ' FERTILE MATERIAL

REPROCESSING

50% EFFICIENCY Py recovery 0.033gPu s’

INYENTORY ™ 11600kg

Udep!

FERTILE MATERIA RENTENTION TIME

10 THE Fission Products c
REACTOR ~10 DAYS to waste kS
2189 Snlt.s': 2
0033gPy- s Udepleted 0-0025g.¢~

0:0025g FP- ' input In10'8g Satt. s~

13-533g U+ s~ +10-8g fresh salt

U-depleted

00355 g. s~ Pu gain
gs 0-0115g-5~
- A7 -

6.5 In-core continuocus gas purecing
g ging

6.5.1 The proposal

In this type of reactor an in-core continuous gas purging of
the molten fuel which can significantly improve the safety in
an in-core accident, is possible.

A mixture of hydrogen-helium gas is continuously bubbled through
the liguid fuel in the core. The mean dwell time of the gas-
bubbles needs toc be controlled and the mean transport time of
the molten components to these bubbles must also be controlled
(e.g. 1if speed-up is desired-intensive mixing, if delay-local
addition of a further g¢as stream).

The aim of the gas stri ing 15 as follows:
g £

1) to remove the volatile fission products which in the case
of an asccldent control the environmental hazard. (1-131,
xe-133, Kr-85 and precursors of Cs-137) and a2t the same
time for the thermal reactor, removal of the I1-135, precursor
of Xe-135, improves the neutron halance.

2] to control the production of delayed neutrons since most of
the precursors and nuclides of this group are very volatile,
p.g. + Hr-JI-isotopes.

3) removal of oxygen and sulphur, continuously (see Chapter 7)
4) «n s4tu control of corrosion problems on structural materials

For the sake of a first approximation a gas flux of 30 cm® per
sec. {(normal state) of Hy/He 1s arbitrarily assumed. At 20 bar
pressure and with a dwelling time in core of 20 seconds, the gas
bubbles will only ocoupy a fraction of the core equal to 10°°

of its volume and have little influence of the criticality, (but
the collapsing of bubbles results in a positive criticality
coofficient ).

The system proposed for continuous romoval of the volatbtile fission
product from the core ilself has a retention time of some hundreds
of seconds only. Each reprocessing mechanism which operates out

of cere is limited by the amount of molten fuel being pumped from
the core to the reprocessing plant. This amount, due to the high
capital cost of the fuel and high operation costs cannot be prea-
ter than that which gives a fuel in-core dwell time of about one
week, Even with a 1 day dwell time, that is, if after one day the
fuel goes through the reprocessing plant, no acceptable solution
to the I1-131 problem is obtained since the activity of this nuclide
is only diminished by one arder of magnitude. Nnly

8 direct in-core removal gives the dwell time in core as low as
some hundreds of seconds.
- 173 -

6.5.2 Delayed neutron emitters

The principal question arise out of the fact that some of the
short lived iodine and bromine {(perhaps also arsenic, tellurium)
lsotopes are the precursors of the delayed neutrons.

Table 6.4 Precursars of delayed neutrons for Pu-239 fast fission

Probable
Croup Half life Fraction Nuclide
t V2 (seconds) %
1 52.75 3.8 Hr-8/
Z 22.79 28,0 I-137, Br-36
3 5.19 21.6 1-138, Br-29
al 2.09 32.8 T-135, Br-90
5 0.5493 10,3
5 0.216 3.5

As other possible nuclides the feollowing can be considered +
As-8h; Kr-492, -93; Rb-92, -94; Sr-87, -98; Te-136, -15/;
Cs-142, -143. The remcval of these delayed-neutron precursors
from the core reduces the value of B, which is lower for Pu-2390
than U-235. Thus we have a problem of reaching a compromise
between an as rapid as possible removal of the hazardous I1-131,
and as long a dwell time In the core for the delayed neutron
precursors: 1-140, I1-13%, 1-138, 1-137 and the appropriate
bromine isotopes.

In this case the mean dwell time of iodine in the steady state
reactor is about 100 seconds. Tt can be seen that the activity

of lodine for a 2.5 GW(t) reactor is of the order of only 10 kilo
curies (for seconds) activity of approx. 10% (or 10% for 1000
second extraction rate].
The gas-extractiocn also influences the other volatile nuclides.
From a very rough estimation for these molten salts (with a
small excess of free hydrogen) the following fission products
and thelr associated precursors of iodine and bromine can be
volatile at 1000°C.

In elementary form: Xe, Kr, Te (7)

In simple volatile hydrides: 82rH, IH

In simple volatile chlorides: SnElz, SbClB, NbCle, CdC17.
This amount of finally volatile components including I, Br, Xe,
Kr amounts to approximately half the total fission products
{i.e. 100 micromoles per second). In addition there is the cor-
responding amount of tritium (fromternary fission). This amount
of all fission products corresponds to a gas volume ratio of
about 2 cm®/s or 10 times smaller than the postulated amount of
hydrogen flow at 30 cm’/s.

The extraction removes all short lived fission products which
are valatlile under these conditions. Thus not only is the re-
moval of the lodine isotopes and the consequent reduction in
nroduction of xenon (e.g. for the atom number: A = 135, 130,
137, 138, 139) achieved but it slows down the in-core production
of Cs-13/, Cs-138, Cs-1349, and then also barium-139

The higher components of the liquid fuel: PuCl, and NaCl. The
Fuel consists of: i

- 1o mols Pull,; boiling point 2040 K

- 85 mol% NaCl ; boiling point 1728 K

One can as a first approximation say that it would have the
following compositicn in the vapour phase: 5 mol% PuCly - 95 mol%
NaCl. )

The aorder of magnitude of vapour for pure components at a tem-
nerature of about 1250 K is

NaCl v 5 x 1077 bar; Putl, ~ 107"  bar

For the PuCl3-NaCl system one assumes here a lowering of the
vapour pressure (thermodynamic activity coefficient approx. 0.1).
At the postulated volumetric flow rate of 230 cm’ Ho normal per
second, the vapourized amount of plutonium is given by:

30 cm?

. -1 _ -8
TSE00 ot Tmn] X 10 bar x 10 10 mol Pu/s

- 175/170 -

This amount of plutonium is of the order of 107% relative to
the amount of plutonium fissioned in the same time (approx 107"
mol Pu/s). However, it still has to be recovered, which unfort-
unately makes the reprocessing more complicated.

Last but not least is the in-core gas extrection of two other
alements

- oxygen in the form of H2D : oxygen from impurities (i.e. PuBC1)

- sulphur in the form of H.S: sulphur from the nuclear reaction:
SCL (n,p) P8 :

(s Chapter 7).
I
-
~J
~~J

|

/. EXPERIMENTAL WORK

/.17 Chemical behaviour of radiosulphur ocbtained by
*°Cl(n,p)?®°s during in-pile irradiation
(accordineg to Janovici, 19/5)

The rather large concentration of sulphur formed by *°Cl(n,p)®°S
reacticn 1n the molten chlorides system proposed for the fast
reactor makes it necessary to obtain the full information on the
chemical behaviour of the radicsulphur. The most recent studies
on the chemical states of radiocsulphur obtained by n-irradiation
of alkall chlorides have shown the complexity of this problem.

To obtain new data aon the behavicur of radiosulphur we have in-
vestigated the influence of the time and temperature of irradia-
tion and of post-irradiation hecating on the chemical distribution
of the sulphur.

EXPERIMENTAL

— e e e = = —

Sodium chloride ("Merck” reagent) was heated for B0hr at 2007
In an oven in vacuo. The dried samples of 100 mg sealed in
cvacuated (107 "torr) quartz tubes were irradiated near the core
of the "Gaphir” swimming poaol reactor for different periods at
a neutron flux of 5 x 10'% and 4.3 x 10'% n cm™? s”!'. Reactor
irradiations were carried out at an estimated temperature of
1509C and -190°C. After irradiation the samples were kept for
5 days to allow the decay of ““Na.

The method of *°S-species separation. The crushing of the ir-
radiated ampoule was made in a special device from which the

alr was removed by purging with a nitrogen stream containing
10ppM of oxygen. After crushing a gentle stream of nitrogen was
allowed to flow for about 10 min. The gases evolved were collected
in coolerd traps. The irradiated slat was dissolved in 2 M KCN
snlutioncontaining carriers of 877, CNS™, S05°%-, Squh. During
dissolution oxygen was not completely excluded although nitrogen
gas was passed continucusly throueh the system. The radiosulphur
found 1n gaseous form was determined as barium sulphate. For

the *°S-species separation the chemical method described recently
by Kasral and Maddock was used. The radioactive samples were
counted under a thin window Gelger counter. All measurements

were made in duplicate with and without Al-absorber.

Fost-irradiation heating. The sealed irradiated ampules were
heated in an electric ocven at 770°C for 2 hr or at 830°C for
gbout 5 min. and then cooled and crushed in a closed system under
a stream of nitrogen.
Results and discussions

Effect of length of irradiation time. As can be seen from

Fig. 7.1 S2- remains the preponderent fraction independent of
the irradiation time. Formation of S°- is indicated by charge
conservation during the *°Cl(n,p)3°S reaction. Alternatively it
can be supposed that reduction of sulphur takes place by capture
of electrons due to the discharge of F-centers. The presence of
5% in this oxidation state in the lattice is no longer contested.
The precursors of higher forms may be S* as a result of an elec-
tron loss from 5°. However, the interaction of chlorine entities
formed by irradiation with radiosulphur tao form species as S5C1,
sCl>7, SCl, may be an important mechanism 1in forming the per-
cursors of sulphate and sulphite.

During longer irradiations some of the sulphide 1s converted into
higher oxidised forms. This may be a consequence of radiation-
produced defects with oxidising character (e.g. V-centres or deri-
vatives). It 1is possible that the concentration of defects re-
sponsible for reduction of the sulphur decreases by annihllation
when new traps are formed. The oxidation of radicsulphur with
increase of radiation damage concentration may alsc be due to the
reaction of recoil sulphur with chlorine atoms. The presence of
OH™ in the crystal must not be rneglected. It has been suggested
that radiolysis of OH™ can be responsible for accelerating the
oxidising process.

Effroct of post-irradiation heating. The effect of post-irradiation
heating (including melting) can be seen in Table 7.1.

Comparisons between heated and unheated samples are made for
irradiations of 2, 12 and 724 hr. For 2 hr irradiation, results

on samples heated at a temperature below the melting point of

NaCl are also presented. As is seen, on heating, a part of the
radiosulphur is found in a volatile form. The veclatile radio-
sulphur appears at the expense of S° and higher oxidation forms.
The results show that with temperatures above the boiling point

of sulphur and above melting point of NaCl the S° and &% and/or
5¢Cly receive sufficient kinetic enerpgy to migrate to the surface
or evoen to escape from the crystal and be collected as volatile
radliosulphur. However, there are some differences in the 3og-
chemical distribution on heating below and above the meliing point
of NaCl (experiments Z2-3). 1t seems that for relatively short
periods of irradiation (2 hr) only the sulphate and sulphite
precursors account for the volatile radiosulphur fraction. For

a longer time of irradiation, on melting the S° value decreases

to about 2% and this corresponds f£o an increase in the volatile
radiosulphur (experiments 5, 7). However, a small and practically
constant yield of S° is found in the melt after longer ilrradiation
Igble /]

Irraad Fost-irrad. 2 - 0 2 JENU .
o} [0 M * o
Expt. e b ot e 55 S SUy ; 302 S volgtlle
1 Z hrs no /3.1 £ 0O, 9.8 * .3 16.9 £ 0.8 0.01
Of_‘
2 " K 75.4 5.3 + 0. 3.6+ 2.3 15,4 + 1.1
Z hrs.
nec
" 59 ) /702 £ 2,1 11.0 = 1), 5.6 £ 2.5 5.0 £ 1.4
5 min.
12 hres. no 57.5 = (1.7 12.1 £ 1.1 Z0.4 + 0.6 .01
0
n
" BSU,C 7.15 1.9 18.49 7.6
5 min.
24 nrs nag o4.4d4 + 2.5 1.3 £ 0.5 23.7 + 2.0 0.0
§ 330°C )
7 ) 8.2 + 3.4 2.3 £ 0.5 Z1.4 + 2. 7. + 0./
5 min.
Expt -5 = 4,3 q0t? -m~ 2 57!
Expt. B-7 = & 104 noem™? 57!
Culpnite fraction 1o less than 5% in nur
fractlion

GXOeTLIMmeNt s

i

Always

lower than

sulphate

~

s
times. No significant changes are observed for the sulphide and
higher oxidation forms. Comparison of these results and those
presented in Fig. 1 shows that in the post-irradiation melted
samples the radiation damage does not have the same effect as

in the unmelted samples. Supplementary information can be cb-
tained by studying the effect of high temperature irradiation

on the distribution of the radicsulphur. It is possible that

in the molten state the active oxidising agents have a different
identity from those present below melting. The presence of oxygen
and probably sodlum oxides during melting may have a determinant
role in deciding the state of the radiosulphur.

Effect of irradiation temperature.

A compariscn of results obtained by irradiation at 423 K and

/7 K (Table 7.2) shows that the higher oxidation fraction is
lower (3%) at 77 K. As is seen the increased S° after low tem-
perature irradlation occurs at the expense of the sulphate +
sulphite and sulphlde fractions. The defects with oxidising and
reducing character formed by low tLemperature irradiation such
as F and V-centres (or derivatives] become important factors in
determining the radiosulphur behaviour.

/.2 Temperature dependence of sulphur species (according to
Furrer, 19/77)

Sipgnificant amounts of the order of magnitude of thousands of
opm °°S would be present as steady-state concentration in a
proposed fast breeder reactor fuelled with molten Pu/U-chlorides
diluted in NaCl. To obtain information about the chemical be-
haviour, mainly the distribution of oxidation-states, the in-
fluence of irradistion temperatures (-130 and 159C) and the
ffects of a post-irradiation heat treatment, solid NaCl was
investigated and the results published. The subject of the
present note are Investlpations at higher irradiation tempera-
tures, especially with samples molten during irradiation.
Fig.

/0

Effect of length of irradiation time on the
35 -species distribution

100

1
- 2. .
] S0, +5042
] 5°
: 52‘
%, 50
4
=
o T T T T T
| 2 3 4 5678910 2 3 456789102 2

Irradiagtion time, hr

Irradiation-temperature dependence of the oxidation-
state distribution of *°S-species

sulphur species (%)

T T T T T
100 300 500 700
irradiation temperature (°C)
Table /.7
Irrad. COHQitions Trrad. Sf_ G0 c0?” + sp?”
flux, time temp. (K] % % 4 o 3
5.107 20 em 257! 4723 73.1 + 0.4 9.8 + 0.8 16.9 + 0.8
2 hrs.
" 77 Shiod £ 0.8 31.5x 1.1 3.0 = .1

Caperimental

Cquimolar mixtures of NaCl and KCI ("Suprapur”, Merck) were studied
instead of pure NaCl {m.p.0658 vs B00YC) because the irradiation
device permitted temperatures of up to 750°C only. The finely
crushed slat-mixture was dried in vacuo at 250°C for 24 hr and
subsequently treated with dried FHCl-pas at 300°C for 24 hr in

order to remove traces of water and hydroxildes. 100 mg samples

were weighed into quartz ampoules in a pglove-box with a purified
nitropen atmosphere (07, H,0 T0pom).

In order to study the influence of oxygen from the surface of the
Si0--ampoules, parts of the samples were weighed into small cruci-
hles made of pold-foil and cleosed by folding the foll. The ampoules
wore ovacuated to a pressure of less than 137 mm Hg (24 hr) and
during evacuation heated to 2Z509C for about 3 hr to remove ad-
hering traces of HCl, sealed and irradiated for 2 hr at 500, ©00

or 7000C in the swimming-pool reactor SAFPHIR near the core at a2
total flux of about 4 x 10'% n cm™? s°'. The neutron-spectrum is
not well characterised but known to be rather hard.

The post-irradiation treatment followed closely the work of (8)
as described In detail elsewhere (4.5). The ampoules were crushed
at room remperature in a nitrogen atmosphere, volatile reaction
products carried by a nitrogen g¢as-stream into cold-traps cooled
with liquid nitrogen and the samples subsequently dissolved in

an oxypen-froe carrier-solutlon containing cyanide and sulphlde,
thiocyanate, sulphite and sulphate as carriers. The fractions were
seporated. independently oxidized to sulphate, precipitated as
HQS:fl and thn activities measured with a Lhin-window GM-countoer.
Some experiments were made at 15000 irradiation-temperature in
order to be able to compare the CaCl/KCl-system with pure NaCl.
New irradiations of pure NaCl were carried out in order to study
the influence of the HCL pre-irradiation treatment.
- 183 -

The data for at least 2 independent experiments at 150 and 500°C
or 4-5 experiments at /00°C for each set of parameters (pre-
irradiation treatment, gold-foil packing, irradiation-temperature)
are shown in the following table. Sulphite, sulphate and volatile
fractions are tabulated together as SX*, While the sulphite-
fraction was always less than 2% of the sulphate-content, the
volatile part showed large fluctuations, especially with samples
packed into gold-foil, (0.20% of the sulphate content) caused

by the variations in sample surface and tightness of the package,
rven 1f mechanically destroyed before dissolving. The volatile
species still remaining In the slat at dissolution, presumably

as S5y¢Cly,, are immediately hydrolized to sulphate and only to a
small extent to sulphite in the basic cyanide-carrier-solution.
The following figure shows the temperature-dependence of the
oxldatilon state distribution.

The experiments at 1509 on Nall treated with HCl-gas confirmed
the existing published results (4.5), which were obtained with
a slat dried in vacuo only.

The results of the investigations at higher temperatures and

with molten samples show a monotone decrease of the S™?-species
and a corresponding increase of the SX*-species with increasing
irradiation-temperature. The content of 5% is influenced neither
by the pre-irradiation treatment not by the irradiation-temperature
and 1Is always about Z20%. Melting of the samples during irradiation
does not i1nfluence the distribution of axidation states. The
studies at 1500C show for NaeCl/KCl-mixtures a shift of about 20%
in 5% towards higher oxidation-states, mainly $°%, compared to
pure NaCl. No influence of the pre-irradiation sample treatment
could be shown at 1500C, but it is of importance for work with
molten samples. Untreated samples without gold-foil protection
showed S5~°-levels of less than 2%. HCl-treatment increased this
value Lo about 13%, an additional pold-foil protection to 26%.
Uxygen of the guartz surface in contact with the melt is clearly
significant at /00YC. The assumption that oxygen from the guartz
surtface should be of importance for reactions cver the cas-phasne
i1s not plausible, without mentioning that oxygen-levels in the
evacuated ampoules (10°°, Hg) must be much greater but have no
significant influence, as low temperature ecxperiments show.
vol

r X+
Irrad. 5-? g? S
5 - -treatm, 5 - i —~ 5t ,
salt-type HCl-treatm cld-packing femp. (9C) state (%) (%) (%)
NaCl [ No 15C /3 10 17 (4)
NaCl Yes Mo 150 /2t 7 13 £ 2 15 £ 2
NaC1l/KC1] Yo No 150 solid 55 + 7 25 £ 2 g x Z
NaCl/KC1 Yi2g Yes 500 30 £ Z 19 2 2 51 £ 3
NaCl/KC1 Y253 Ho €00 17 = 3 272 3 61 £ 4
Nall/KC1 No Me 70 2 1 22 * 3 /6t 5
NaC1l/KC1 g Mo 70 Jolten 13 £ 3 27 3 G5 *+ 4
NaCl/KC1 Yes Yes Vasld Zb 4 10 = 3 55 + 4
Table 7.3 O«<idaticn state distribution of radiosulphr procduced by 3561[n,p)35Cl in NaCl,
'

N e
i

snlld mixtures and MNaCl/¥Cl-melts
8. THERMOHYDRAULICS

5.1 Introduction

This chapter gives the results of studies on the thermal proper-
ties of the following units.

- High flux liguid fuel core with external heat exchanger. This
core has the geometrical form of a spherical shell (Fig. 8.1A)

- Power-breeder-core with molten fuel and external heat exchanger.
The core is spherical. (Fig. 8.108).

- Sodium/molten salt heat exchanger.

All the calculations were made in simplified form and were
intended only to roughly define the scope of the prcocblems.

4.2 High Flux reactor with the core as a spherical shell

This reactor was designed to meet the following reguirements

a) In the shell type core to achieve the highest possible flux
in having a fast spectrum requiring a relatively high power
density. Very good mixing in the fiuid, i.e. a hiph turbulence
is a vital requirement. For this the flow rate of the fuel has
to be kept high.

b} In the central regions a thermal neutron flux is aobtainded
(Beryllium coated Deuterium hydroxide as moderator!. The
thermal neutron flux extends partly into the fast core which
causes a very high and locallised power density. This 1s the
reason for requiring a high fuel turbulence.

) In the termal zone having a neutron flux of approx 4x10'%n cm?s !

it is possible to transmute the most 'dangerous' fisslon products
Strontium-90 and Caesium-13/. Table 8.1 shows the most important
parameters.
Z

=\

-
hugelcore
A
—
G
<

- 13/ -

Tanle 8.1 ©Shell form High Flux reactor

Property Unit Value
5 §)
Total power Ewtot
Core Geometry see Fig. B8.Z
3 6
Core volume cm 3.7 x 10
Mean power density kwecm 3 2

Specific heat of the

J em k7! 0.83 (1-0.5 x 103 (T-850))
fuel (T in OC) e S 0

o
0
3
!
L)
M

Density at 850°C 12

The Lhermal properties of this core were calculated using the
following paramefers

- Flow rate at input 10-22 m.s !
- Uiameter of the inlet pipe 0.5 - 0.6 m

the inlet temperature was kept constant at 7509C

The most important variable was the cutlet temperature

which was varied between Y93509C and 96500
For various reasons a reference concept was used as a basis for
the calculations, see Fig. 8.7. A description of the reference
design is also given in Fig. 8.3,
The relevant characteristics are given in the following fipures:

Temperature curve Fig. 3.4 A

Flow rate curve Fig. 6.4 B

Mean circumference and Fia. 8.4
shell thickness & M
Number of Pu atoms 1in & Fig. 6.5 A
layer

the sudden changes at the start and end are dues to the uncertain-
tles in the geometry at the pipe/shell interface.

Density Fig. 8.5 B
?ESSTTindigkeit T = Austrittstemperatur des Lrennstoffes (9
ms 7
1 T = 120
10 9180 1140 1100 1230

D

1360
1542

300

&

Geferenzfali sconalerc

—tp-
Radius des

sintrittronres

Felative .rza.l ier ru-.tome ir ..r.:ontaler .:
Lo des cores
Je
arcitrale
~ln.elit
|
o :
I}
) 1
[ v
4! !
s i l ‘ ,
R4 \
iy v v
I\
20 I
)
s £ 120 149 i il
hihe des Uores
(goem

srennsteffdichte

n

~

20 LG 100

h8ne des Joures

heferenzfall Schalencore

Brennstof{temperatur

40 €0 80 100 12C 143 162 180 2.2 fz) 260

Brennstoffgeschwindigkeit

20

50 60 80 100 129 140 160 180 202 220 240

Mittlerer Umfang und Schalendicke

2J

40 &0 80 100 120 140 160 130

ii8he des Cores

Referenzfall Xugelcore

230 823 4T

-
]

R P,

- 189 -

8.3 Power Reactor with spherical core

This reacltor was designed to meet the following requirements

- a spherical core with the minimum of structural meterial in
order to optimise the neutron balance and to achieve a hard
spectrum.

- a maxlmum bLreeding ratio in the core and blanket

- the liquid fuel cooled irn an external heat exchanger.

Table 9.7 shows the most important characteristics.

Table 8.7 Power breedine reactor

Charactoarictic Ut Value

lotal power L (th) 3

Core geometry
spherical radius

Core volume cm’ 2.3 x 10°
Mean power density ks om?® 1.3
sprcific ho:s -3 -
opeciflc heat Jeem IKT! N.83x(1 -0.5x10%(T-850)
(T in "C)
0 -
Density at 850 C geem 2,90

(e thermal parameters of this core have been caloulated varying
the following parameters

- inlet velocity

- dilameter of inlet tube

. SRS BN
The i1nlet temperature was kept constant at 7507

Th@jmain parameter varied was the outlet temperature in the range
13570 to 365°C (see Fig. 8.7).

The given reference case (see Fig. 8.7) is shown in Fig. 8§,
S

parameter changes under consideration are shown in fipure
4, C and Fig. 85.9.

o .
8.9 A,
rig. 8.7 T = Austrittstemperatur des brennstoffes (OC) Fig. 3}

Heferenzfail Kugelcore

Qeschwindigkeit
(ms™ 1) 4 T » 1100 1050 1000
180
10 °c)
760 Brennatofftemperatur
11 950 350
4035
940 5
Referenzfall 859
12 — 7
[
20
Y 750
13 20 60 100 140
(muvl)
13
% 9300 Brennstoffgeschwindigkeit
11
I
15
7
5
16
3
880 10¢ 140 -
17
(em)
86
18 .
Su Radius
950 e
19 -
5%
78
v &
35 36 37 38 39 LC 41 42 (em) e
3 : Tu
Ragdius des Lintrittsronrs a0 5o 100 o L
hdhe des Cores (cm)
3 - Vo : totales Volure: .Jes orennstulfs i ie.. <o re;

= pirtrittstemperdatur ies datr.um:

Heferenzfall Kugelcore

Jescow. Jes

(g.em )
3.0 Brennstoffdichte
2.0
2.8
2.6
20 63 130 143 (em)

r8he des Cores

1e 11 12 1t 1u 1 it LT

vescowindiprest Latri.c .ing

crennstelfvolarern (m

Latriumtemperatur
8.4 The external heat exchanger

The most important parameter for the external heat exchanger was
selected as the volume of fuel contained in it. The decisive
factor is the minimisation of the fuel volume existing outside
the core, which influences:

- the effective specific power
(MW{th] per kg guel in the whole system)

- the effective doubling time
- safety considerations under accident conditions.
- minimisation of the loss of delayrd noutrons,

For bhis reason a simple molten salt/sodium heat exchangar was
~haooe
L,J:WO\_)&' IW-

The most Important properties of the two media are shown in
Lable 4.3,

Tha following parameters were arbitrarily chesen
- VYolocity of fuel 10-15 mes !
- Velocity of sodium 10-16 mes

- Inlet temperature of the fuel 35070

- Furl outlet temperature 750°¢C
- Sodium outlet temperature BOOOC
salt on the tube side dia 0.0 cm

- vodium on bthe shell side 0.0/1.2 to 2.2 cm pitch

Important variabhles

- Inlet sodium temperature

- Length of tubes

- Volume of fuel in the tubes

All these data are given in Fig. 9.10 for the case of a pitch
of 1.2 cm,
Table 8.2 Heal “xcnhanger:

3 GW thermal

Property Unit Fuel Coclant
Composition Mol 17 PuCl_ = 2 UCl3 » 3.65 MalCl 50dium metal
Density gom 3.526 x (1-0.007 (T-850)) 7.784  (1-0.001 (T-350))
Heat Capacity Jg tkt 1.0 (1.0.0005 (T-850)) 1.256  (1-0.0005(T-350))
Thermel conductivity Wem™ ! oKk 0.015 (1.0.091  (T-850)) 0.663 (1+0.001 (T-350))
Viscosity ocm gt .05 (1-0.6501 (T-855)) 0.0018 (1-0.001 (T-350))

U n. .4 0.8 0.8
Heat transfer co-fficiont vom 2K .23 x BE < pr ot 5.3 + 0.019 RE ‘PR
(Rr/RmyHe3
Inlet temperature °C 750 variable (see Fig., 10)
. o . o
Jutlet temperature C 5657 23818
Tnlet velocity men 11 = 1A 10+ 16

cbl
Austrittstemp. Natrium

Salezvelunen
x

°c) e
L 7,
L}
\
\ L 2,4
\
<
690 \
b 2,0
50C "
1,3
1,0
40y o Temperat
,_1,“
Referenzfail
valovolumen
1,2
3Ly -
v L L A 1
1,0 1,2 1,4 1, 1,0
—GEHCnW. Salz 1does t - 17 B
—_— Gescuw. ia 14 -x:l —_———- i .:l
.13 Referenzfall
RO
ENIOE,
trennstof:
guo
7o,
BUL o
datriunm
PRIV \\
P~
L6

1., 20C 359

e Jes wirmeaustauscrers

4oL cem

TEMPERATURE °C

1000°4

800"

7504

Fig. B.12

Keferenzfall

998°C

0453
00218
0484
14330

asa'cE
83T

—_ total

|
COO/LANT
9 st"
|
I
i

|

781°C

2.188
0-0937
0957 |

0,3 em
3,0
] Lrenn-
o storr
atrium
e
—t
1,6 cm
a
L
Heat Transter 2
coefficient ENC"‘ dtg]
Yiscosity

Heat Capacity [J'gni'deg"
Reynolds Number
For other pitches between 1.2 and 2.2 and for two selected fuel
and sodium velocities values can be seen in fig. B8.11. The re-
ference case was arbitrarily chosen (see Fig. 8.11). The geometry
of the heat exchanger for this reference case is shown in Fig.

8.172. Further parameters are summarised in table 8.4.

lable 3.4 Heat Exchanger

Characteristic UInit Fuel Metallic sodium
Tnlet velocity mes™! 14 14
Viscosity peom ! gl 0.0524 0.0N183
Heat capacity Jeg k! 1.024 1.275
Number of tubes 5787
Volume cm?® 1.32x10°

The temperature curves for both media are given in Fip. 3.13.

8.5 The internally cooled reactor

This type of reactor, has fuel circulating in the core only, with
a relatively low temperature gradient of approx. 35°C and a nigh
heat capacity, and a high velocity coolant 9m s7! with apain a low
temperature gradient of approx. 43°C and high heat capacity. This,
coupled with the very high nagative temperature coefficient of
reactivity results in an unusually hieh necative thermal and
'reactivity’' stability.

Jocrease of the fuel circulation and/or coolant velocity (in the
U-tubes) results in a definite and "automatic” decrease of reaclor
power without recourse to engineered methods. This points to such
a reactor being a surprisingly stable and self regulating device.

The aschievement of the required fuel velocity in the core seems
te require a forced circulation system since the rough estimate
using natural convection gives a heat transfer coefficient which
not acceptable.
Such a forced circulation system (core only) can be one of the
following types - pump installed directly in the core, pump
cutside the core, an external pump with injector, a gas 1ift
pump using inert gas. Consideration of the factors involved
Uusing criteria such as reduction of the out of core inventory,
elimination of additional hesat exchangers, minimisation of the
fuel leakape, minimisation of the ausiliary power, coptlimisation
of the fuel flow regulation - all point ot an in-core pump soclu-
tion. of cource this ¢ives rise to considerable technical prob-
blems (cooling of the rotor, corrosion and erosion, mailntenance,
noutron activation etal.

The caleculations for this type of reactor have been based on the
following more or less arbitrarily selected parameters:

- fuel on the shell side, with tube pitch fTo diameter ratio
cqual to 1.70 to 1.18.

1
ot
|

- fuel velocllby: (1.

- core dia: 2 and .7 m

-]
ot
)

- coclant in tube with tube internal diamebter equal to: 1.0

- velocity of conlant: 1 to 1/ m s !

- coolant inlet emperature 750Y and 3009C

The neutronic calculation and thermo-hydraulics were made for
T om of the core helght.

The detaliled representation for the temperature distribution 1in
a typical power reactor with a core output of 1.0938 GWith) are
given in Fig. 6.14 (for a position 43 cm above the bottom of the

core where the neutron flux 1s normalised to 1.

, 0

The bulk temperature of the fuel 13 here U893 0, Che tomperature
af the tube walls 857 - 389 C and the bulk temperature of Lhe

L

.
coolant 810,

For the total output of the core 1.936 GWith), the power distri-
bution 1s as shown in Fig. 5.175.

0f course a flatter power ditribution could be obtained by ad-
justing tube deameters and pitch across the core. (Note that In
this calculation the radial neutron flux distribution has been
taken as unperturbed).
Fig. 8.15

Power distribution inthe core

Mean total
value
200 1-— ———n 1936-27 MW(t)
17%
150+4+—- - —
125 _
p-3
S
S0 .
T
T
“
& 75
[¥)
S0
25
0 v — ——
0 1 3 4 5 6 7 8 8 10 1 12
Power produced per cm height of cell MW(t)
2008 W
. P
200 1 1w ey 7 e
~ Te08°C
- xmyc !
~ |
P |
|7 ‘
| 1584 W) ,
Tt=1007C 1
150 A
CORE
POWER
OUTPUT

500

1

M

2

T . \ T +

3 4 5 6 7 8 9 © 1 12
coolant velocity [ms"]

Fig. 8.16 Temperature of fuel and coolant relerence core

Perox1 0"heri's™
Coolant Fuel 2ms™’
260 e A
7
] '
175 f¢/
150 4 - + ,,
\
T A\
S5 -
100 4 p
S \
s \
© \
78 of e
\
\
\
)
S0 *
|
!
I
WY \ 4
\ /
\ /
\ ’
el o
750 60 70 80 90 800

2

900 910 920 940 950 960 920 963 990 1000
TEMPERATURE [c}

Fig. 8.18 Core power output versus fuel velocity

4
000 - N A
96.9’ A
] i
P

AN

1 ims

0 — T —
o 20

FUEL VELOCITY ms™

- 157 -

A very encouraging indication of the good temperature distribu-
tion with very small temperature gradients 1s shown 1in Filg.
8.16 which 1indicates the axial bulk temperature distribution

in the fuel and in the coolant 1n the core.

The Fuel bulk temperature changes From 380°C at the bottom to
965°C at V4 core height and is ggg" Coat 3/4 of core hPtht The
coolant temperature lies between /50 C inlet and 7537C outlet.
Both these small temperature gradients in the fuel and in the
coolant (fertiel material) may prove beneficial in reducing
corrosion processes due to the minimizing of mass transport
phenomena.

The stable behaviour of this type of reactor results from many
parameters. Two of them are the veleocity of the coaolant and 1ts
bulk temperature. The mean power output of the core is strongly
dependant on the velocities of both fuel and cooclant., Fig. 8.17.
For a fuel vplocity of 2 mes” !, when the coolant velocity falls
from 17 m* s ' to 1 mes~ ! the coolant ocutlet temperature 1mcrraseq
from 784°C to 993°C for constant inlet temperature of 7507C. This
change of coolant velocity and its bulk temperature results 1in

thr decrease of the mena core output from Z2.088 GW(th) to 0.548
CW(th) - that is approximately a factor 3! It is clear since the
lower coolant velocity results in a higher coolant cutlet tempe-
roture and lower power ocuftput we have definite negative temperature
coefflicient (power ocutput) varying with the given coolant velocity.

If the fuel velocity falls from 2 mes ' to 0.8 mes™ ' we again get
an important decreass of power outpulbt (see Fig. 3.18). The de-
crease in both fuel and coolant velocity results in a sharp
decrase of reactor power. This means that such a reactor can be
considered as a surprisingly stable and self regulating device.

In the case of a sudden fall in coolant and/or fuel velocjities

the power output decreases to a safe level without interventilon.
The achievement of the required fuel velocity in the core seems
to require a forced circulation system since Lhe rough estimate
ucing natural convection gives a heat transfer coefficient which

15 too low.

Such a force circulation system (core oniy) can be aone of the
following types

- pump instailed directly in core
~ pump outside the core
- an external pump with injector

- a gas-1ift pump using inert gas (argon)
Intensive consideration of the facltors involved, using criteria
such as - reduction of the out of core inventory, elimination of
additional heat exchangers, minimization of the fuel leakage,
minimization of the auxilliary power, optimisation of the fuel
flow regulation, points to an in-core pump solution. Of course
this gives rise to consliderable technical problems (cooling of
the rotor, corrosion and eraosion, maintenance, neutron activation
eto. ).

The postulted fuel velocity makes it possible to make some cal-
culation of the heat transfer problems and also gives a feel for
the kinetics of the reactor under discussion.

It must be stressed that these kietics studies have no strong
physical sense and use an 1tterative approach but 1t 1s clear
that they give some useful information about the general reactor
stability. (see Fig. £8.19).

Fig. B.19 Fuel temperature during several passes thprouth oo
under different inital conditions

Fuel velocity 2ms‘:
1200- Coolant ISms”

L,
1
5
£
¥
“900 -
Cold fuel case
BUO‘I §=02x10 nem 'y
l TM-'ISO'C
»we+—T—r-vrr-rr-rorT-T T T Ty
1 2 k| 5
Botton Top Botton Fuel No. of passes

of core
4. REFERENCES

9.7 List of EIR-publications used in this report (chronoleogical

order)

M. Taube Molten chlorides fast breeder reactors

J. Ligou (problems, possibilities)

EIR-Report 215 (13972)

J. Ligou [flolten chlicorides fast breeder reactor.
Reactor physics calculations
FEIR-Report 228 (1872)

M. Taube wWwarme-Tauscher fUr schnelle Reaktoren mit

Ko Dawudi geschmolzenen Zalzen
TM-HL-2/70 (139/73)

M. Taube A molten salt fast/thermal reactor system
with no waste
EIR-Report 249 (1374)

Taube Thorium-Uranium fast/thermal breeding system
with molten salt fuel
FIR-Report, 753 (1374)

e Taube The possibility of continous in-core gaseous
rxtraction of volatile fission products in o
malbten fucl roactor
CIR-Roport 07 (1070)

M. Taube Modlten plutonium chlorides fast breeder cooled

Joo Lipou by molbon uranium chlorides

Ann. Nucl. Gei. Engin. 1, 277 (1874)
- 200 -

M. Taube Steady-state burning of fission products in
a fast thermal molten slat breeding power
reactor

Ann. Nucl. Sci. Engin. 1, 283 (1974)

. Janovici Chemical state of sulphur obtained during
M, Taube in pile irradiation
EIR-Report 267 (1974)

M. Taube A high-flux fast molten salt reactor for the
J. Ligou transmutation of Cs-1537, r- 0

. Ottewitte EIR-Report 259 (1975)

M. Taube The transmutation of fission products (Cs-127,
J. Ligou Sr-90) in a salt fuelled fast fission reactor
K.H. Bucher with thermal zane

EIR-Report 270 (15/75)

M. Furrer Arbeitsvorschrift zur Auftrennung des nach
°C1(n,p)*°S in NaCl-Proben produzierten
Schwefels
TM-HL-247 (1875)

M. Taube Fast breeder power reactor with molten
plutonium trif
TH-HL-258 (1575)

M. Taube Hreeding in molten sall reactors, [oolurog
on the University of Liege
FIR-Report 276 (19/75)

L. Janovici Chemical behaviour of radiosulphur obtained
M. Taube by *°Cl(n,p)?°S during in-pile-irradiation
J. Nucl. Inorg. Chem. 3/, 2581 (1875)
M.

Taube

Furrer

Furrer

Taube

Transmutaticn of 5r-80 and Cs-137 in a high-

flux fast reactor with thermalized central
region
Nucl. Sci. Engineer. 81, 212 (13976)

Chemischer Zustand von *°Clin,p)3%S in
Alkalichloridschmelzen
EIR~-TM-HL-275 (18786}

Chemical behaviour of radiosulphur obtained
during in-core irradiation
J. Inorg. Nucl. Chem. 712771, 30, %7

Is the transmutation of strontium-90 and
carsium-12/7 in a high flux fission reocactor
frasible?

in "Lon--Life Radionucolide Transformabion”
Froce . 70, Nuclear Fnergy Aponcy
Ispra, 10 March-1977
M.

M.
M.
SN

A

Former publications concerning molten chlorides fast

- 2072 -

breeders

and the fluoride thermal breeder

Taube

Taube

Kowa lew
Miaolcarski
Potura]

Taube
Miclecarski
Lowalow
Prtura]

Taube

Taube
Kowalew
Potural]
Micleoarski

Tauhe

Taube
Micloarshi
Fobtura]
Kowalow

Fused salt fast breeder
Cf-56-5-204, Oak Ridge 19586

Fused plutonium and uranium chlorides as
nuclear fuel for fast breeder reactors, Symp.
React, Eperiment. International Atomic
Energy Agency, Symp. (SM-21/18]) Vienna 1961

The concept of salt-boiling fast breeder
reactor (AICl, as cooling agent)
Nukleonika 9-10, 639 {1885)

of fast breeder reactor with
and bolling mercury
9-10, 641 (1968R/)

The concept
fused salt
Nukleonika

in siit-boiling ractors
2317 (1967)

Reactivity
Nukleonika 3,

der Salzsiedereaktoren
184 (1867)

Konzeption
Kernenergle b,

Fast breeder boiling reaclbors with fused salts
(in Russian)

Atomnaya tnergiya 1, 10 (19607)

New boiling salt fast
Nucl. Engin. Design 5,

hreeder reactor concepts
100 £1967)
1. Taube Fast breeder reactor with direct cooling by
boiling agents (in Russian)
IBJ-Warszawa, 042/C
Conf. Fast Breeder, USSR, Dubna, December, 1967

L.6. Alexander Molten “ali Reactors
Froceed. Breeding Large Fast Reactors
ANL-6792 (13963)

P.A. Nelson Fuel properties and nuclear performance of
D.K. Butler Tast reactors fuelled with molten chlorides
M.G. Chasanov Nuclear Applications, Vol. 3, 540 (1467)

D. Nemeghetti

2.R. Harder Compatibility and reprocessing in use of molten
G. Long UCly alkaelichlorides mixtures as reactor fuel
W.F. Stanaway in "Sympos. Reprocessing Nuclear Fuels”

Ed. P. Chiotti, USAEC, Con. £30801 (196¢)

J.A. Lane Test-reactor perspectives
React. Fuel. Proces. T=ch. 12, 1, 1 (1363)

M.W. Rosenthal Molten sait reactors: Series of papers
et al. Nucl. Applic. Techn. &, 2, 105-219 (13/0)
9.3 Publications concerning transmutation

M. Steinberg Neutron burning of long-lived fission
C. Wotzka products for waste disposal
B. Manowitz BNL-8558 (1964)
M.GC. Chasanov Fission-products effects in molten chloride
fast reactor fuels
Nucl. Sci. Eng. 23, 189 (13865)
M.V. Gregory A nuclear transformation system for disposal
M. Steinberg of long-lived-fission product waste
Rrookhaven Nat. Lab. BNL-11915, 1867
W.H., Walker Fission Products Data, Part 1
ALE.C.L. - 3037 (1983)
E.A.C. Crouch Calculated independent yields
AERE-R-B056 (1963
S.M. Feinberg High flux statlonary experimental reactors
and their perspectives (in Russian)
Atom. Energ. 29, 3, 162 (149/70)
R.D. Cheverton H.F.I.R. Core Nuclear Design.
T.M. Sims ORNL-4621 (1971)
E. Clayton Thermal caplture cross-section
AAEC-TM-613  (1972)
H.C. Claiborne Neutron-induced transmutation of high-level

radioactive waste

ORNL-TM-3964 (1872)
W.C. Wolkenhauer The controlled thermonuclear reactor as a
fission product burner
BNWL-4232 (1873)

J.0. Blomske Managing radiocoactive wastes
Nichols Physics Today 8, 36 (1873)
W.C. McClain

-]
U

F.A.C. Crouch Fission products chain yields from
gxperimaents
AERE-R-7394 (1973)

A.S5. Kubo Disposal of nuclear waste

D.J. Rose Science 182, 4118, 1205 [(1973]

W.C. Wolkenhauer Transmutation of high-level radicactive
H.K. Leonhard waste with a CTR

2.F. LGore BNWL-1772, Pac. North West Lab., (1873)
.J. Schneider Advanced waste management studies, High

i2vel radioactive Waste Jdisposal alternatives
USAFEC, BNWL-1900, Richland 19/4

T.0. HBeynon The nuclear physics of fast reactors
Rep. Prog. Phys., 37, 251 (18/4

R.F. Gare Transmutation in quantity of Cs-137 in 2
D.HR. Leonard controlled thermonuclear reactor
Nucl. Sci- Engin. 93, 249 (1974

W.F. Vogelsang Transmutations, radilioactlivity in D-T

R.C. Lottt TOKAMAK fusion reactor

C.L. Kulcinski Nucl. Technology 22, 379 (19/4)

T.Y. Sung

H. Wild Radioaktive Inventare und deren zeitlicher

Verlauf nach Abschalten des Reaktors
KFfK-1797, Kernforschungszentrum Karlsruhe, 1874
- 206 -

9.4 Fublications concerning the thorium fuel cycle

A.I. Lejpunskii Feasability of using thorium in fast power
0.0. Kazachkovski reactors

S.B. Shikhov Atomnaya Etnergiyva, 18, 4, 342 (1965)

V.M. Murogov

C.A. Renni Selfsustaining mixed fast and thermal
reactor system
0.P. Report 5817, AEE Winfrith, 1267

ASH-1097 The use of thorium in nuclear paower reactors
USAEC 19869

- - Thorium fuel cycle
Bibl. Series No. 39
Int. Atom Energy Agency, Vienna 18710

F. Fortescue A reactor strategy FBR's and HIGR's
Nuclear News, April 1872

G. Grazianni Plutonium as a make up 1in the thorium
C. Rinaldini integral block HTR buel element
H. Bairiot

Fur 50Z0e Joint Nuclear Research Centre
. Tranwaert

Tspra 19/3
R.t. Hrogli Thorium uvtilization in an FOR/HTIOR power
KWR. Schulz system

34 LC 3730 1973 7/ 74
D. Broda

P.J. Wood
J.M. Driscoll

2.R. Seghal
C. Lin
J. Naser

W.HB., LOwensteln

F.R. Merz

.. Kasten
J

Homan

- 207 -

Eine Welt aus Flutonium?
Naturwiss. Rundschau, 28, 7, 233 [197h)

Can we live with plutonium?
Some articles: Now Scientist, 29 May 1375

The ecconomics of thorium blankets for FBER
Proceed ANS 1975 10na

Thorium-based fuels in FBER
Proceed ANS, Jure 1974

Thorium fuel cycle
Inter. Conf. ducl. Power, Fuel Cycle
TAEA, Salzburg, May 1977, CN - 36/98

tvaluation of Pu, U and Th use in power

Feactor Tuel cyele

Inter. Conf. Nucl. Fower, Fuel Cycle
TAEA, Galzburg, May 14977, CN-36/407
- 208 -

9.5 References to the chemical experiments

M. Kasrai J. Chem. Soc. A, 1105 (1970)

A.G. Maddock

C.N, Turcanu Radicchem. Radioanal. | %%, E, 287 (1970)
J.L. Baptists J. Inorg. Nucl. Chem. 36, 1683 (1974)
No5S.5. Magues

G.6 References to the physics calculation

AT, Weinberg The Physical Theory of Neutran Chain

E.FP. Wigner VI p.143, Univ. Chic. Press (1353)

L.W. Nordheim A Program of research and ralculations of
resocnance ~hsorption
- CA - 007, 01hn)

I'.J. Bodarenko Group constants for nuclear reactor

ot al. Calculations (13G4)

J. Adir et al. User's and Frogrammer’'s Manual for the GGC
Multigroup Cross-Section Code - Part 1,

GA 7157 (18967)
AR WY

=
i jn
-

SRS

A5

.1. Bell

Glasstone Nuclear Reacltor Theory. Van Nostrand

Reinhold Comp. (1970) 244 - 247

A. Ducat Fvaluation of Lhe parfait blanket concept
.J. Driscoll for fast breeder reactor
.E. Tudreas M.I T. Cambridege, MITNE-157 (1974)
4. Hardie A Comprehensive expression for the doubling
We Little time of fast breeder reactors

.. Omberg Nuclear Technology, 26, 115 (1975)