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ORNL-2751.txt
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LB S gfl‘ ;
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’i“* ORNL-2751
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TID-4500 (14th ed.
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NUCLEAR CHARACTERISTICS OF SPHERlCAL,
HOMOGENEOUS, TWO-REGION,
Lo AN
- Tl e
MOL TEN-FLUORIDE-SALT REACTORS |
. Alexander
L.G
D. A. Carrison
H. G
J. T
£ otur
. MacPherson
. Roberts
T
OAK RIDGE NATIONAL LABOQM%Y
operated by ;
UNION CARBIDE CORPORATION |
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R U.5. ATOMIC ENERGY COMMISSION'
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_document, send in name with
arid the librory will orrange
-~
Printed in USA. Price Z?_q_,_.enfs. Available from the
Office of Technical Services
Department of Commerce
Washington 25, D.C.
LEGAL NOTICE
This repors was prepared as an account of Government sponsored work. Neither the United States,
nor the Cammission, nor any person acting on behalf of the Commission:
A, Makes ony warranty or representation, expressed or implied, with respect to the accuraey,
completeness, or usefulness of the information contairned in this report, or that the use of
any informotion, apparatus, method, or process disclosed in this report may not infringe
privately owned rights; or
B. Assumes any liabilities with respect to the use of, or for damages resulting from the use of
eny information, apparatus, method, or process disclesed in this report.
As used in the above, ‘'person acting on behalf of the Commission’" includes any employee or
contractor of the Commission, or employee of such contractor, 1o the extent thot such employee
or contracter of the Commission, or employee of such contractor prepares, disseminctes, or
provides access to, any information pursuant to his empleyment or contract with the Commission,
or his employment with such contractor,
ORNL.-2751
Reactors—Power
TID-4500 (14th ed.)
Contract No. W-7405-eng-26
REACTOR PROJECTS DIVISION
NUCLEAR CHARACTERISTICS OF SPHERICAL, HOMOGENEOUS,
TWO-REGION, MOLTEN-FLUORIDE-SALT REACTORS
L. G. Alexander H. G. MacPherson
D. A, Carrison J. T. Roberts
DATE ISSUED
SEP 161909
OAK RIDGE NATIONAL LABORATORY
Oak Ridge, Tennessee
operated by
UNICN CARBIDE CORPORATION
for the
e AL
3 445k 03L1387 3
CONTENTS
ABSTRACT tuturrerretiiteeereserisriresereesiessersessstessessessrerseessesessesssessasesssesesesseeersseseerenenes 1
PRIOR WORK tvvevttrttvrrersvrietereereressereereeessessresssssssesessssnsssssessssssssssnsssssssssnsnsssssessens 2
METHOD OF CALCULATION tttttetteeeeeviieeiriesiussstrtrnrerereeesesseesesesssssensssseesessessaessaarons 4
HOMOGENEOUS REACTORS FUELED WITH U?%3% (it vninniriiinr e e e e seeessennans 4
It Al ST S Lttt ittt it et it er et ittt aesaenerneeaeneansnaesrnsraarnatnransenreranerns 4
Neutron Balances and Other Reactor Variables cuvirviiiissiieeieiiisieersiostsniensonnseenmacansssncans 7
HOMOGENEQUS REACTORS FUELED WITH U233 1ottt 9
NUCLEAR PERFORMANCE OF A REFERENCE DESIGN REACTOR ceveevviiennnnnnvererrnenes. 14
COMPARATIVE PERFORMANCE +vvvvvttuentttetiniiersersseseessesssesssssssnsssassnsesssnsssssannssesssnnn 18
ACKNOWLEDGMENTS 1vvvereeiiiisierssnunereetestrerresessessssssesossssseesssssesessesssessesnsnnsessessess 18
NUCLEAR CHARACTERISTICS OF SPHERICAL, HOMOGENEOUS,
TWO-REGION, MOL TEN-FLUORIDE-SALT REACTORS
H. G. MacPherson J. T. Roberts
L. G. Alexander D. A, Carrison
ABSTRACT
The use of a molten-salt fuel makes possible the production of high-pressure, superheated
steam with a nuclear reactor aperating at low pressure. The corrosion resistance of the INOR-8
series of nickel-molybdenum alloys appears to be sufficient to guarantee reactor component life-
times of 10 to 20 years, Proposed continuous fuel-processing methods show promise of reducing
fuel-processing costs to negligible levels. With U233 45 the fuel and Th23245 the fertile material
in both core and blanket, initial regeneration ratios up to 1.08 can be obtained at critical masses
less than 600 kg.
initially about 1300 kg. With U
in excess of 0.6 con be obtained with critical masses of less than 300 kg. The corresponding
The corresponding inventory for a 600-Mw(th) central station power reactor is
235 45 the fuel, U233 is produced, and initial regeneration ratios
critical inventories for 600-Mw(th) central station power reactors are 600 kg or less, depending on
the thorium loading. It is concluded that homogeneous, molten-salt-fueled reactors are competi-
tive in regard to nuclear performance with present solid-fuel reactors, and they may be economi-
cally superior because of lower fuel and fuel-processing costs,
Molten fluoride salts provide the basis of a new
family of liquid-fueled power reactors, The range
of solubility of uranium and thorium compounds in
the salts makes the system flexible and allows the
consideration of a variety of reactors. Svitable
salt mixtures have melting points in the range
850 to 950°F and are sufficiently compatible with
known alloys to assure long-lived components, if
the temperature is kept below 1300°F. As may be
seen, molten-salt-fueled reactor systems tend to
operate natyrally in o temperature region suitable
for modern steam plants; they have the further
advantage that they achieve these temperatures
without pressurization of the molten salt,
The molten-salt system, for purposes other than
electric power generation, is not new. Intensive
research and development over the past nine years
under ORNL sponsorship has provided reasonable
answers to a majority of the obvious difficulties.
One of the most important advances has been the
development of methods for handling salt melts at
high temperatures and maintaining them at tem-
peratures above the liquidus temperatures. In-
formation on the chemical and physical properties
of a wide variety of molten salts has been ob-
tained, and methods were developed for their
manufacture, purification, and handling that are in
use on a production scale. It has been found that
the simple ionic salts are stable under radiation
and that they suffer no deterioration other than the
buildup of fission products.
The molten-salt system has the usual benefits
attributed to fluid-fueled systems., The principal
advantages claimed over solid-fuel systems are:
(1) the lack of radiation damage that can limit fuel
burnup; (2) the avoidance of the expense of fabri-
cating new fuel elements; (3) the continuous re-
moval of fission products; (4) a high negative
temperature coefficient of reactivity; and (5) the
avoidance of the need for excess reactivity, since
makeup fuel can be added as required. The last
two factors make possible a reactor without control
rods that automatically adjusts its power in re-
sponse to changes of the electrical load. The
lack of excess reactivity con lead to a reactor
that is safe from nuclear power excursions.
In comparison with aqueous systems, the molten-
salt system has three outstanding advantages: it
allows high-temperature operation at a low pres-
sure; explosive radiclytic gases are not formed;
and thorium compounds are soluble in the salts.
Compensating disadvantages are the high melting
points of the salts and poorer neutron economy;
the importance of these disadvantages cannot be
assessed properly without further experience.
Probably the most outstanding characteristic of
the molten-salt systems is their chemical flexi-
bility,
solutions which are available for reactor use. In
this respect, the molten-salt systems are practi-
cally unique, and this is the essential advantage
which they enjoy over the uranium-bismuth sys-
tems. Thus the molten-salt systems are not to be
thought of in terms of a single reactor — rather,
they are the basis for a new class of reactors.
Included in this class are all the embodiments
which comprise the whole of solid-fuel-element
technology: Y233 burners, thorium-uranium thermal
converters or breeders, and thorium-uranium fast
converters or breeders, Of possible short-term
that is, the wide variety of molten-sait
interest is the U%35 burner. Because of the in-
herently high temperatures and because there are
no fuel elements, the fuel cost in the salt system
is of the order of 1 mill /kwhr in graphite-moderated,
molten-salt-fueled reactors.
The present technology suggests that homo-
geneous converters using a base salt composed of
BeF, and either Li’F or NaF and using UF, for
fuel and ThF | for a fertile material are more svit-
able for early reactors than are graphite-moderated
The chief virtues
of the homogeneous converter reactor are that it is
or plutonium-fueled systems.
based on well-explored principles and that the use
of a simple fuel cycle should lead to low fuel cycle
costs,
With further development, the same base salt
(using Li’F) can be combined with a graphite
moderator in a heterogeneous arrangement to pro-
vide a self-sustaining Th-U23? system with a
breeding ratio of about 1. The chief advantage of
the molten-salt system over other liquid systems
in pursuing this objective is, as has been men-
tioned, that it is the only system in which a solu-
ble thorium compound can be used, and thus the
problem of slurry handling is avoided.
PRIOR WORK
The applicability of molten salts to nuclear
reactors has been ably discyssed by Grimes and
1.2 The most promising systems are those
the fluorides and chlorides of the
These
appear to possess the most desirable combination
others.
comprising
alkali metals, zirconium, and beryllium.
of low neutron absorption, high sclubility of ura-
inertness. In
general, the chlorides have lower melting points,
but they appear to be less stable and more cor-
rosive than the fluorides.
The fluoride systems appear to be preferable for
use in thermal and epithermal reactors. Many mix-
tures have been investigated, mainly at ORNL and
at Mound Laboratory. The physical properties of
these mixtures, in so far as they are known, have
been tabulated, and the results of extensive phase
studies have been reported.’®
Lithium-7 has an attractively low capture cross
section, 0.0189 barn at 0.0759 ev, but Li®, which
comprises 7.5% of the natural mixture, has a cap-
ture cross section of 542 barns at the same energy.
The cross sections at 0.0759 ev and 1150°F for
several lithium compositions are compared below
with the cross sections of sodium, potassium,
rubidium, and cesium.
nium and thorium, and chemical
Cross Section
Element (barns)
Lithium
0.1% Li® 0.561
0.01% Li 0.0731
0.001% Li® 0.0243
0.0001% Li® 0.0194
Sodium 0.290
Potassium 1.130
Rubidium 0.401
Cesium 29
The capture cross sections of the lighter ele-
ments at higher energies presumably stand in
approximately the same relation as at thermal. |t
may be seen that purified Li’7 has an attractively
low cross section in comparison with the cross
sections of other alkali metals and that sodium is
the next best alkali metal.
The sodium-zirconium fluoride system has been
extensively studied at ORNL.®> A eutectic con-
taining about 42 mole % ZrF, melts at 910°F.
'W. R. Grimes, ORNL CF-52-4-197, p 320 ff (April
1952) (classified).
2. R. Grimes, D. R. Cuneo, and F. F. Blankenship,
in Reactor Handbook, ed. by J. F. Hogerton and R. C.
Grass, vol 2, sec 6, p 799, AECD-3646 (May 1955).
3J. A. Lane, H. G. MacPherson, and F. Maslan (eds.),
Fluid Fuel Reactors, p 569, Addison-Wesley, Reading,
Mass., 1958.
Small additions of UF, lower the melting point
appreciably. A fuel of this type was successfully
used in the Aircraft Reactor Experiment (ARE).?
Inconel, a nickel-rich alloy, is reasonably re-
sistant to corrosion by this fuel system at 1500°F.
Although long-term data are lacking, there is
reason to expect the corrosion rate at 1200°F to
be sufficiently low that Inconel equipment would
last several years.
However, with regard to its use in a central-
station power reactor, the sodium-zirconium fluo-
ride system has several serious disadvantages.
The sodium capture cross section is less favorable
than the Li’ cross section. In addition, there is
the so-called “*snow’’ problem; that is, ZrF , tends
to evaporate from the fuel and crystallize on sur-
faces exposed to the vapor. In comparison with
the lithium-beryllium system discussed below, the
sodium-zirconium system has inferior heat transfer
and cooling effectiveness. Finally, the expecta-
tion at Oak Ridge is that the INOR-8 alloys will
prove to be as resistant to the beryllium salts as
to the zirconium salts and that there is therefore
no compelling reason for selecting the sodium-
zirconium system.
The capture cross section of beryllium appears
to be satisfactorily low at all energies. A phase
diagram for the system LiF-BeF, has recently
been published.® A mixture containing 31 mole %
BeF, reportedly liquefies at approximately 980°F.
Substantial concentrations of ThF, in the core
fluid may be obtained by blending this mixture with
the compound 3LiF-ThF,. A temperature diagram
for the ternary system has been published.® The
liquidus temperature along the join appears to lie
below 930°F for mixtures containing up to 10
mole % ThF,. Small additions of UF, to any of
these mixtures should lower the liquidus tempera-
ture somewhat.
The ARE was operated with a molten-fluoride-
salt fuel in November 1954, The reactor had a
moderator consisting of beryllium oxide blocks.
The fuel, which was a mixture of sodium fluoride,
zirconium fluoride, and uranium fluoride, flowed
through the moderator in Inconel tubes and was
pumped through an external heat exchanger by
bid., p 673.
31bid., p 573.
Stbid., p 579.
means of a centrifugal pump. The reactor op-
erated at a peak power of 2.5 Mw. It was dis-
mantled after carrying out a scheduled experi-
mental program.
In 1953 a group of QRSORT students, under the
leadership of Jarvis,” investigated the applica-
bility of molten salts to package reactors., More
recently, another ORSORT group led by Davies
prepared a valuable study of the feasibility of
molten-salt U232 burners for central-station power
production.® Fast reactors based on the U?38.py
cycle were studied by Addoms ez al. of MIT and by
an ORSORT group ied by Bulmer.® Both groups
concluded that it would be preferable to use molten
chlorides, rather than the fluorides, because of the
relatively high moderating power of the fluorine
nucleus, although it was recognized that the chlo-
rides are probably inferior with respect to corro-
sion and radiation stability. Bulmer et al. also
pointed out that it would be necessary to use puri-
fied CI®7 on account of the (n,p) reaction ex-
hibited by CI33, Because of the disadvantages of
the chloride systems and, further, because the
technology of handling and utilizing neptunium-
and plutonium-bearing salts is largely unknown, it
was decided to postpone consideration of chloride
salt reactors.
In 1953 an ORSORT group led by Wehmeyer!'®
analyzed many of the problems presently under
study. The proposals set forth in that report have
influenced the present program. A study by David-
son and Robb of KAPL!! has also been helpful.
Both studies considered the possibility of using
thorium in a U233 conversion-breeding cycle at
thermal or near thermal energies.
A recent conceptual design study'? of a 240-Mw
(electrical) central-station molten-salt-fueled re-
actor was used as a basis for examining the eco-
nomics and feasibility of a reactor using molten-
salt fuel. An attempt was made to keep the
7T. Jarvis et al., ORNL CF-53-10-26 (August 1953)
(classified).
8R. W. Davies et al, ORNL CF-56-8-208 (August
1956) (classified).
9J. Bulmer et al., Fused Salt Fast Breeder, QORNL
CF-56-8-204 (August 1956).
IOD. B. Wehmeyer et al., Study of a Fused Salt
Breeder Reactor for Power Production, ORNL CF-53-
10-25 (September 1953).
”J. K. Davidson and W. L. Robb, A Molten-Salr
Thorium Converter for Power Production, KAPL-M-JKD-
10 (Oct. 15, 1956).
]2ane, MacPherson, and Maslan, op. cit., p 681.
technology and the processing scheme as simple
as possible,
METHOD OF CALCULATION
Reactor calculations were performed by means of
the UNIVAC program Ocusol, '3 a modification of
the Eyewash program.'® Ocusol is a 31-group,
multiregion, spherically symmetric, age-diffusion
code. The group-averaged cross sections for the
various elements of interest that were used were
based on the latest available data.!® Where data
were |acking, reasonable interpolations based on
resonance theory were used. The estimated cross
sections were made to agree with measured reso-
nance integrals where available. Saturations and
Doppler broadening of the resonances in thorium
as a function of concentration were estimated.
The molten salts may be used as homogeneous
moderators or simply as fuel carriers in hetero-
geneous reactors, Although graphite-moderated
heterogeneous reactors have certain potential ad-
vantages, their technical feasibility depends upon
the compatibility of fuel, graphite, and metal,
which has not as yet been established. For this
the homogeneous reactors, although in-
ferior in nuclear performance, have been given
prior attention.
A preliminary study indicated that, if the in-
tegrity of the core vessel could be guaranteed,
reason,
the nuclear economy of two-region reactors would
probably be superior to that of bare and reflected
one-region reactors, The two-region reactors were,
accordingly, studied in detail. Although entrance
and exit conditions dictate other than a spherical
shape, it was necessary, for the calculations, to
use a model comprising the following concentric,
(1) the core; (2) an INOR-8
reactor vessel, ‘/3 in, thick; (3) a blanket, approxi-
mately 2 ft thick; and (4) an INOR-8 reactor vessel,
2/3 in. thick. The diameter of the core and the con-
spherical regions:
centration of thorium in the core were selected as
independent variables, The primary dependent
variables were the critical concentration of the
3L, G. Alexander et al., Operating Instructions for
the Univac Program Ocusol-A, a Modification of the
Eyewash Program, ORNL CF-57-6-4 (June 5, 1957).
MJ. H. Alexander and N. D. Given, A Machkine Multi-
group Calculation. The Eyewash Program for Univac,
ORNL-1925 (Sept. 15, 1955).
ISJ. T. Roberts and L. G. Alexander, Cross Sections
for Ocusol-A Program, ORNL CF-57-6-5 (June 11, 1957).
fuel (U233, U233, or Pu?%%) and the distribution
of the neutron absorptions among the various
From these, the
inventory, regeneration
ratio, burnup rate, etc., could be readily calculated.
atomic species in the reactor.
critical mass, critical
HOMOGENEOUS REACTORS FUELED WITH U235
U233 would be a superior fuel
While the isotope
in molten-fluoride-salt reactors, it is unfortunately
not available in quantity, Any realistic appraisal
of the immediate capabilities of these reactors
must be based on the use of U233,
The study of homogeneous reactors was divided
into two phases: (1) the mapping of the nuclear
characteristics of the initial (i.e., ‘‘clean’’) states
as a function of core diameter and thorium con-
centration and (2) the analysis of the subsequent
performance of selected initial states with various
processing schemes and rates. The detailed re-
sults of the first phase are given here. Briefly, it
was found that regeneration ratios of up to 0.65
could be obtained with moderate investment in
U235 (less than 100 kg).
Initial States
A complete parametric study was made of molten-
fluoride-salt reactors having diometers in the range
of 4 to 10 ft and thorium concentrations in the fuel
ranging from 0 to 1 mole % ThF ,. [n these re-
actors the basic fuel salt (fuel salt No. 1) was a
mixture of 31 mole % BeF., and 69 mole % LiF,
which has a density of about 2.0 g/cm® at 1150°F,
The core vessel was composed of INOR-8. The
blanket fluid (blanket salt No. 1) was a mixture of
25 mole % ThF , and 75 mole % LiF, which has a
density of about 4.3 g/cm® at 1150°F. In order to
shorten the calculations in this series, the re-
actor vessel was neglected, since the resuitant
error would be small. These reactors contained
no fission products or nonfissionable isotopes of
uranium other than U238,
A summary of the results is presented in Table 1,
in which the neutron balance is presented in terms
of neutrons absorbed in a given element per neu-
tron absorbed in U233 [both by fission and the (,y)
reaction]. The sum of the absorptions is therefore
equal to 7, that is, the number of neutrons pro-
duced by fission per neutron absorbed in fuel.
Further, the sum of the absorptions in U238 and
thorium in the fuel and in thorium in the blanket
Table 1. Initial-State Nuclear Characteristics of Two-Region, Homogeneous, Moiten-Fluoride-5alt Reactors Fueled with U
Fuel salt No, 1: 31 mole % BeF2 + 69 mole % LiF + UF’4 + ThF4
Blanket salt No. 1: 25 mole % ThF4 + 75 mole % LiF
600 Mw (heat)
External fuel volume: 339 ft°
Total power:
235
Case number
Core diameter, ft
ThF4 in fuel salt, mole %
U235
in fuel salt, mole %
U 235 3tom den sity*
Critical mass, kg of y23s
Neutron absorption ratios**
U233 (fissions)
U235 (n,y)
Be, Li, and F in fuel saolt
Core vessel
L.i and F in blanket sait
Leakage
U238
Th in fuel salt
in fuel salt
Th in blanket salt
Neutron yield, i
Median fission energy, ev
Thermal fissions, %
Regeneration ratio
1 2 3 4 5 6 7 8 9 10 11
4 5 5 5 5 5 6 6 6 6 6
0 0 0.25 0.5 0.75 1 0 0.25 0.5- 0.75 1
0.952 0.318 10.561 0.721 0.845 0.938 0.107 0.229 0.408 - 0.552 0.662
33.8 11.3 20.1 25.6 30.0 33.3 3.80 8.13 14.5 19.6 23.5
124 81.0 144 183 215 239 47.0 101 179 243 291
0.7023 0.7185 0.7004 0.6996 0.7015 0.7041 0.7771 0.7343 0.7082 0.7000 0.7004
0.2977 0.2815 0.2996 0.3004 0.2985 0.2959 0.2229 0.2657 0.2918 0.3000 0.2996
0.0551 0.0871 0.0657 0.0604 0.0581 0.0568 0.1981 0.1082 0.0770 0.0669 0.0631
0.0560 0.0848 0.0577 0.0485 0.0436 0.0402 0.1353 0.0795 0.0542 0.0435 0.0388
0.0128 0.0138 0.0108 0.0098 0.0093 0.00%90 0.0164 0.0116 0.0091 0.0081 0.0074
0.0229 0.0156 0.0147 0.0143 0.0141 0.0140 0.0137 0.0129 0.0122 0.0119 0.0116
0.0430 0.0426 0.0463 0.0451 0.0431 0.0412 0.0245 0.0375 0.0477 0.0467 0.0452
0.0832 0.1289 0.1614 0.1873 0.1321 0.1841 0.2142 0.2438
0.5448 0.5309 0.4516 0.4211 0.4031 0.3905 0.5312 0.4318 0.3683 0.3378 0.3202
1.73 1.77 1.73 1.73 1.73 1.74 1.92 1.82 1.75 1.73 1.73
270 15.7 105 158 270 425 0.18 5.6 38 100 120
0.052 6.2 0.87 0,22 0.87 0.040 35 13 3 0.56 0.48
0.59 0.57 0.58 0.60 0.61 0.62 0.56 0.61 0.60 0.60 0.61
* 1019 cxfoms/cms.
**Neutrons absorbed per neutron absorbed in u23
5
Table 1 (continued)
Case number
Core diameter, ft
ThF4 in fuel salt, mole %
U2:35 in fuel salt, mole %
U235
atom density*
Critical mass, kg of U235
Neutron absorption ratios**
U235 (fission)
U235 (n,y)
Be, Li, and F in fuel salt
Core vessel
Li and F in blonket salt
Leakage
238
u in fuel salt
Th in fuel salt
Th in blanket salt
Neutron yield, 7
Median fission energy, ev
Thermal fissions, %
Regeneration ratio
12 13 14 15 16 17 18 19 20 21 22
7 8 8 8 8 8 10 10 10 10 10
0.25 0 0.25 0.5 0.75 1 0 0.25 0.5 0.75 1
0.114 0.047 0.078 0.132 0.226 0.349 0.033 0.052 0.081 0.127 0.205
4.05 1.66 2.77 4.67 8.03 12.4 1.175 1.86 2.88 4.50 7.28
79.6 48.7 81.3 137 236 364 67.3 107 165 258 417
0.7748 0.8007 0.7930 0.7671 0.7362 0.7146 0.8229 0.7428 0.7902 0.7693 0.7428
0.2252 0.1993 0.2070 0.2329 0.2638 0.2854 0.1771 0.2572 0.2098 0.2307 0.2572
0.1880 0.4130 0.2616 0.1682 0.1107 0.0846 0.5713 0.3726 0.2486 0.1735 0.1206
0.0951 0.149 0.1032 0.0722 0.0500 0.0373 0.129 0.0915 0.0669 0.0497 0.0363
0.0123 0.0143 0.0112 0.0089 0.0071 0.0057 0.0114 0.0089 0.0073 0.0060 0.0049
0.0068 0.0084 0.0082 0.0080 0.0077 0.0074 0.0061 0.0060 0.0059 0.0057 0.0055
0.0254 0.0143 0.0196 0.0272 0.0368 0.0428 0.0120 0.0153 0.0209 0.0266 0.0343
0.1761 0.2045 0.3048 0.3397 0.3515 0.2409 0.3691 0.4324 0.4506
0.4098 0.4073 0.3503 0.3056 0.2664 0.2356 0.3031 0.2617 0.2332 0.2063 0.1825
191 2.00 1.96 1.89 1.82 1.76 2.03 2.00 1.95 1.90 1.83
0.16 Thermal 0.10 0.17 5.3 27 Thermal Thermal 0.100 0.156 1.36
33 59 45 29 13 5 66 56 43 30 16
0.61 0.42 0.57 0.64 0.64 0.63 0.32 0.52 0.62 0.67 0.67
* 1019 cfloms/cms.
**Neutrons absorbed per neutron absorbed in U23
5
.
salt give directly the regeneration ratio. The
losses to other elements are penalties imposed on
the regeneration ratio by these poisens.
A graph of critical mass plotted as a function of
core diameter, with thorium concentration as a
parameter, is presented in Fig. 1. The masses
vary from about 40 kg of U?3°% in a 7-ft-dia core
having no thorium in the fuel to about 450 kg in
the 10-ft-dia core having 1 mole % ThF, in the
UNCLASSIFIED
ORNL—LR—DWG 39521
SOO ¥ T T T l
CORE AND BLANKET
SALTS NOt mole % Tth
! IN FUEL. SALT ®
° -
g ¢
"
ND //
s
@ 300 ra .
; / O.T?___________-u
< '/"-__'"—'—-—Ii—-—-"'—""
= ¢
&J 200 28 |
i, 9" e
2 2 N 0.50 o
E ” ® -..____"—/T—/‘
© I‘r"\ \
100 " '." 0.25 .
! \ ®
| — e /
' NO ThF, IN FUEL SALT
0 | i
CORE DIAMETER ()
Fig. 1. |Initial Critical Masses of U235 iy Two-
Region, Homogeneous, Molten-Fluoride-Salt Reactors.,
tuel.
plotted in Fig. 2, range from 0.5 for the minimum
mass reactor to 0,63 for the largest mass reactor.
It does not seem likely that further increases in
diameter or thorium concentration would increase
the regeneration above 0.65.
The effects of changes in the compositions of
the fuel and blanket salts were studied in a series
of calculations for salts having more favorable
The corresponding regeneration ratios,
melting points and viscosities. The BeF, content
was raised to 37 mole % in the fuel salt (fuel salt
No. 2), and the blanket composition (blanket salt
No. 2) was fixed at 13 mole % ThF4, 16 mole %
Ber, and 71 mole % LiF. Blanket salt No. 2 is a
somewhat better reflector than No. 1, and fuel salt
No. 2 is a somewhat better moderator than No. 1.
As a result, at a given core diameter and thorium
UNCLASSIFIED
ORNL-LR—-DWG 39522
1.0 T T T T
CORE AND BLANKET SALTS NO.1
0.8 mole % ThF, IN FUEL SALT
o 1 AND 0.75
= | e
= AV e
z 06 =2 + 0.5
o
e 0.25
<
o
>
w >
* NO ThF, IN FUEL SALT/
0.2 '
0
q 5 6 7 8 S 10
CORE DIAMETER (ft)
Fig. 2. [Initial Fuel Regeneration in Two-Region,
Homogeneous, Molten-Fluoride-Salt Reactors Fueled
with U235,
concentration in the fuel salt, both the critical
concentration and the regeneration ratio were
somewhat lower for the No. 2 salts.
Reservations concerning the feasibility of con-
structing and guaranteeing the integrity of core
vessels in large sizes (10 ft and over), together
with preliminary consideration of inventory charges
for large systems, led to the conclusion that a
feasible reactor would probably have a core di-
ameter lying in the range between 6 and 8 ft. Ac-
cordingly, a parametric study of the No. 2 fuel and
blanket salts in reactors with core diameters in
the 6- to 8-ft range was made. |n this study the
presence of an outer reactor vessel consisting
of 2/3 in. of INOR-8 was taken into account. The
results are presented in Table 2. In general, the
nuclear performance is somewhat better with the
No. 2 salts than with the No. 1 salts,