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ORNL
P 1784
TEEFEREE
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MICROCOPY RESOLUTION TEST CHART
NATIONAL BUREAU OF STANDARDS - 1963
ORNI f.1984
Cont-651101-41

NUCLEAR AND FUEL-CYCLE PERFORMANCE OF
MOLTEN-SAIT BRIWEDER REACTORS*
RELEASED POR ANNOUNCEMENT
IN NUCLEAR SCIENCE ABSTRACOS
Howard F. Bouman and Paul R. Kasteta
Oak Ridge National Laboratory
Oak Ridge, Tennessee
DEC 21 1965
Thermal- and intermediate-energy molten-salt reactors appear capable
of fuel doubling times in the range of 10 to 20 years, specific inventories of
about 1 kg libsile/Mwle), and fuel-cycle costs under 0.4 mi 11/kwhr. Thermal
systems (termed MSBR's) are graphite-moderated and fueled with circulating
molten salts consisting of the fluorides of uranium, thorium, lithium-7,
and beryllium. Intermediate systems? (termed MOSEL reactors) are similar
but without graphite moderator. The results presented here are based on
1000-Mw(e) plants having 45% efficiency and 80% plant factor. The calcu-
lations were performed with OPTIMERC, an optimization code combined with the
MERC multigroup, equilibrium reactor code.”
Fluoride Salt Processing
IO
Molten fluoride salts are promising reactor fuels due to their ability to
dissolve thorium and uranium fluorides, their very low corrosion rates in nickel
alloy systems, their low vapor pressures at high temperatures, and the ease with
which uranium can be recovered from the salt. The latter feature stems from the
high volatility of UFG (boiling temperature less than 100°C), and the low
volatility of UF, (boiling temperature over 1400°C). The fluorination of UF,-
bearing salts (the fluoride volatility process) 18 a well-developed process that
can be integrated with a molten-salt reactor to achieve continuous, rapid, and
inexpensive fuel reprocessing.
vitae
Research sponsored by the U. S. Atomic Energy Commission under contract
with the Union Carbide Corporation.
For presentation at the American Nuclear Society Meeting, November 15-18,
1965, Washington, D. C.
Presidente de munca sex

The reactors considered here consist of a core surrounded by a thorium-
bearing fertile salt blanket. Typical compositions for fuel and fertile
salts are shown in Table 1. Tre fertile salt is processed by fluoride
volatility to remove the 'bred uranium; the non-volatile-thorium remains in
Table 1. Typical MSBR Fuel and Blanket salt
Compositions and Liquidus Temperatures
Fuel Salt
Blanket
Salt
MSBR I MSBR II I
Liquidus temperature, ec
450
480
565
Composition, mol %
Li'F
BeF,
UFC(11usile)
.510
73.0
565
71.0
68.75
31.0
0.25
0.0
71.0
20.0
0.25
8.75
71.0
0.0
5.0
24.0
13.3
2.7
0.0
0.0
29.0
InF4
13.0.
the salt, which is simply returned to the blanket. The fuel salt 18 also
a sa
processed by fluoride volatility. Ninety-five percent of the stripped salt
from this process can be recovered by vacuum distillation, which separates
the rare-earth fission products; the remaining 5% is discarded to control the
buildup of other fission products. The fuel salt 18 reconstituted by adding
part of the uranium recovered from the core and blanket, and 18 returned to the
i
sent
reactor.
The excess uranium is available for sale.
retir eminded mestihämtades per
Integrated processing is one of the important bases for the high neutron
economy and low fuel-cycle costs of these reactors; others are the absence of
structural materials (other than graphite) in the core; the nearly immediate
maintensitation and its
release of xenon and other fission gases from the fuel stream; and the low
neutron leakage associated with a thick blanket. Neutron losses to Pa255 are
controlled by either holdup of the fertile stream outside the core for Pa decay,
and out their creation
mentre side
"
.
or by considering that Pa can be removed from the fertile stream by processing
on a very short cycle time.
Reactor Concepts
There is no less a variety of possible concepte for molten-salt reactors
than for soll.d-fueled reactors. . Molten-salt reactors may have a single-
stream core, with both fissile and fertile materials in a single salt, or
two-stream cores, with f18bile and fertile salts separate ; they may be cooled
internally (11ke solid-fueled reactors), externally (by circulating the fuel),
or by direct-contact of the fuel with an immiscible coolant; they may be
graphite moderated, salt moderated, or unmoderated, etc. The purpose of the
calculations presented here is to evaluate the nuclear and economic per-
formance of some of these combinations.
Taree reactor concepts were considered, all externally cooled. In the
first, fissile and fertile materials were naintained in separate streanas
within the graphite-moderated core (MSBR-1); in the second, also graphite
moderated, the core fissile and fertile materials were combined in one stream
(MSBR-2). The third, a variation of the MOSEL concept, had a homogeneous,
wamoderated, fluid core containing both fissile and fertile materials. Two
versions of the MOSH were evaluated, one direct-contact cooled with molten
lead," the other with a conventional heat exchanger. These four conceptual
designs are shown in Figs. 1 through 4.
In MSBR-I, (Fig. 1), the fuel salt is circulated through the reactor
within graphite tubes. The core moderator is supported from the tubes, and
18 immersed in a pool of fertile salt so that no structural barrier separates
the core and blanket regions, Fertile salt lills the spaces between moderator
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11
*
blocks, and circulates through passages provided in the moderator.
The MSBR-II (Fg. 2) consists of a cylindrical core tank, of Hastelloy N or
niobium, filled with graphite moderator blocks. A fuel salt containing both
thorium and uranium Pllls the spaces between the moderator blocks and flows:
upward through passages in the moderator. The core 18 surrounded radially
by a blanket filled with fertile salt. The metal wall separating the core and
blanket regione 18 only 1/8 in. thick, to reduce neutron absorptions, and 18
supported from the wall of this blanket vessel.
The MOSHL consists of a sphericai core vessel, containing thorium-bearing
fuel salt, but no moderator. 'The core is surrounded by a fertile salt blanket.
In the externally cooled version (Fig. 3) the fuel salt 18 pumped tarough a sumber
of parallel shell-and-tube heat exchangers. In the direct-contact cooled"
version (Fig. 4) the fuel is contacted with a Jet of molten lead, in each of several
parallel loops. The jet Aerves botih to cool and circulate the fuel.
Calculation Method
The calculations were performed with OPTIMERC, a combination of an
optimization code with the MERC multigroup, diffusion, equilibrium reactor
code. MERC calculates the nuclear performance, the equilibrium concentrations
of the various nuclides, including fission products, and the fuel-cycle cost
for a given set of conditions. OPTIMERC permits up to twenty reactor parameters
to be varied, within limits, in order to determine an optimum, by the method
of steepest ascent. The reactors were optimized essentially for minimum fuel-
cycle cost, although, in the objective function, minor weight was given to
maximizing the annual fuel yield. Typical parameters varied were the reactor
dimensions, blanket thickness, fractions of fuel and fertile salts in the core,
and fuel and fertile stream processing rates. The economic assumptions followed
..::!.
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AEC standards wherever possible.
The basic assumptions for these
evaluations are given in Table 2. The cross sections were calcužated from
modified GAM-THERMOS 11braries."
Table 2. Basic Assumptions Employed
in the MSBR Evaluations
1000
Reactor power, Mwe
Thermal efficiency, %
Load factor
Cost assumptions
Value of u233 and Pa255, $/8
Value of 1235, $/8
Value of thorium, $/kg
Value of carrier salt, $/kg
Capital charge, annual rate, %
Plant
Non-depreciating capital, including
Pissile inventory
Processing cost, $/ft3 salt
Fuel (at 10 P43/day)
Blanket (at 10 ft3/day)
Processing cost scale factor (exponent)
Processing losses, fraction of liesile
per pass through processing
Fission gas removal, % per core pass
0.325
0.001
100
Results
A characteristic of externally cooled, fluid-fueled reactors is the
large fuel volume required for the heat-removal system, often on the order
of five times the fuel volume in the core itself. The influence of external
volume on MSBR-I performance is shown ir. Table 3, where case 1 (424 it)
represents a nominal external inventory, and case 2 (316 ft)) represents that
associated with an advanced system. For comparison, the external fuel volume
.
Table 3.
Influence of External Fuel Inventory
on MSBR-1 Performanca
Cabe 1
Case 2
424
316
Heat exchange system fuel salt
volume, 1t3
Specific inventory, kg fissile/
MW (electrical)
Breeding ratio
Fuel yield, % per annum
Fuel-cycle cost, mills/kwhr
0.68
1.048
5.4
0.34
0.60
1.058
7.2
0.34
in a recent detailed design for such a system was 345 ft". The low-inventory
case has a higher annual fuel yield, not only because of the smaller fissile
'Inventory, but also because the optimum conditions shift to gi.ve a higher
breeding ratio.
The external fuel volume was even more important in the MOSHL reactors,
because the fissile concentration in the fuel stream in these intermediate-
spectrum reactors 18 necessarily higher than in the thermal reactors. Table
4 illustrates this by giving the reactor performance for the direct-contact,
molten-lead-cooled system, MOSHL(1), (50 ft? external fuel volume), and for
the externally cooled MOSH(2), (316 it' external fuel volume). The ruel
yield is about four times greater in the lower volume system - a strong
incentive for developing high performance heat-exchange systems.
In general, the breeding gain of fast reactors is higher for the u_30.
Pi259 cycle than for the Th232-0259 cycle. The solubility of plutonium in
fluoride fuel salt is limited to about 1 mol %, whereas the nominal fissile
concentration for a MOSHL reactor is about 5 mol h; therefore, these reactors
normally employ the ucycle. However, they may be operated on a mixed
cycle, up to the solubility limit for plutonium. We performed a series of
Tahle 4. Influence of External Fue.l. Inventory
on MOSEL Reactor Performance
Case 1
Case 2
50
316
Heat exchange system fuel salt
volume, ft3
Speciric inventory, kg fiskile/
Mw (electrical)
Breeding ratio
Fuel yield, % per annum
Fuel-cycle cost, mills/kwar
3.0
1.1
1.156
10.3
1.094
2.4
0.58
0.14
calculations, substituting u 90 for part of the Th232 in the MOSEL, with
the results shown in Table 5. The fertile material can be as much as 17%
0290 without exceeding a 1% plutonium concentration, and the breeding gain
18 increased by nearly two points. However, we made no allowance for the
additional cost and complexity of processing a plutonium-bearing fuel. In
practice, the straight Th-U233 cycle, with its still very attractive 9.3%
yield and 0.17 'mill fuel-cycle cost, may be preferred.
more remory. Si Nuri
Table 5. Influence of Core uegº Concentration on MOSEL Fuel
Cycle Performance for High-Performance Core
1,238 Core
Fertile
Fraction
Plutonium
Concentration
(mole %)
Breeding
Ratio
Fuel Yield
(% per annum)
Cost
(mills/kwar)
0.10
0.20
0.30
0.59
1.19
1.79
1.140
1.149
1.159
1.168
9.9
10.5
0.17
0.15
0.14
0.12
inand
In the MSBR-I, in which the fuel is segregated within fuel tubes, the
self-shielding of the fissile isotopes has an important effect on the fissile
concentration and inventory, and, indirectly, upon the fissile-to-Perille ratio
"Naitwaerd
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and the fuel yield, as shown in Table 6. Thus, from a wuclear standpoint,
...
the fuel tube diameter should be as small as possible; the minimum practical
size depends upon engineering considerations. In the MSBR-II (one-stream
core), the fuel passages may be cut into the moderatur blocks almost at will,
and are readily arranged to minimize self-shielding.
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Table 6.
Influence of Fuel Tube Diame'cer
on MSBR-1 Performance
.---:.ansininen Pan
Fuel tube diameter, in.
Breeding ratio
Fuel yield, % per annum
Fuel-cycle cost, mills/kwhr
4.0
1.048
5.4
0.34
-
ibuYiTMMIT
nie
In the two-stream core, neutron losses to protactinium may be controlled
by holding up a large volume of fertile material outside the core, to allow
the decay of Pa3 to 95 outside the high-flux region. This is not practical
when the fissile and fertile materials are in the same stream, because of the
prohibitive Inventory charge for holding-up a large amount of fissile material.
In the one-stream case (MSBR-II), therefore, a process was proposed for
separating protactinium from fuel salt and the processing cost estimated. This
case is compared with the MSBR-I in Table 7, and indicates that, 11 Pa losses
can be controlled in this manner, the greater homogeneity of the one-stream core
can give a significant improvement in the breeding ratio and fuel yield. ..
Table 7. Reactor Performance of MSBR-1 and MSBR-2
MSBR-1
MSBR-2
Streams in core
Breeding ratio
Fuel yield, % per annum
Fuel-cycle cost, mills/
kwhr
Two-stream
1.048
5.4
One-stream
1.065
7.2
:
0.34
0.33
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The individual compcnents of the fuel-cycle cost for the four reactors
are given in Table 8. In general, f1881le Inventory and processing are the
major components of the cost. The processing cost is high for MSBR-II,
because of the rapid processing for Pa removal. The inventory cost is highest,
in the MOSEL reactors, but this is offset, at least in MOSEL(1), by a high
production credit for sale of bred fuel.
Table 8. Fual-Cycle Cost
Cost, mills/lwhr
MSDK
MSBR
II
MOSEXL
(1)
MOSEL
(2)
Fissile inventory
Fertile inventory
Salt inventory
Fertile replacement
Salt replacement
Processing
0.133
0.035
0.054
0.015
0.057
0.118
0.133
0.018
0.030
0.005
0:066
0.166
0.207
0.013
0.015
0.002
0.005
0.119
0.565
0.011
0.016
0.002
0.008
0.115
Total costs
Production credit
Net fuel-cycle cost
0.412
0.070
0.342
0.418
0.085
0.333
0.361
0.224
0.137
0.717
0.137
0.580
· Discussion
The breeder reactor criteria of short doubling time and low fuel-cycle cost
can be met by all three of the molten-salt concepts. Each requires the
solution of somewhat different development problems, however, MSBR-I requires
the ability to separate the fertile and fissile salts by graphite tubes
under high irradiation conditions. In MSBR-II, the core graphite require-
ments are less stringent (minor flaws become unimportant), but inexpensive
processes for Pa removal and thorium recovery must be developed. The MOSEL
2
I
CITY
concept requires no graphite, but tha core tank material must be resistant
termome
to fast neutron damage, and a high-performance heat-removal system with
low fuel inventory 18 essential.
The three reactor concepts above are not the only attractive possibilities.
The fast molten-salt reactor, employing chloride rather than fluoride salts,
has a higher breeding potential than the corresponding fluoride system. It
was omitted from this study, not because it lacks merit, but because fluoride
salt processing 18 already developed, and the development of comparable chloride
salt processing and handling technology would require a major effort.
The performance of MSBR-I can be aproved by processing the fertile
stream rapidly to remove protactinium. Laboratory scale experiments indicate
that this may be done more easily than the corresponding process for the fuel
stream, as is proposed for MSBR-II; bot possibilities are under study.
The performance of either of the thermal reactors, as well as the MOSEL
reactors, can be improved by the development of high performance heat-removal
systems, either direct-lead-cooled or conventional. We plan to evaluate these
and perhaps other possibilities, and hope to present the results in a future
paper.
Acknowledgements
The cooperation of J. L. Lucius and W. L. Kephart in preparing the
OPTIMERC code, and of E. S.Bettis in developing the reactor conceptual designs,
is gratefully acknowledged.
References
1. R. B. Briggs et al., Molten-Salt Reactor Program Semiann. Prog. Rep.
for Period Ending July 31, 1964, USAEC Report ORNL-3708, Oak Ridge
National Laboratory, November 1964.
2. Paul R. Kasten, The MOSEL Reactor Concept, Third Geneva Conference,
Paper 538, September 1964.
3. T. W. Kerlin, Jr., c. W. Craven, Jr., L. G. Alexander, and J. L. Lucius,
The MERC-1 Equilibrium Reactor Code, USAEC Report ORNL-TM-847, Oak
Ridge National Laboratory, April 22, 1964.
4. Proposed by E. S. Bettis, Oak Ridge National Laboratory.
5. GAM-THERMOS cross sections modified by E. H. Gift, Oak Ridge National.
Laboratory.
6. L. G. Alexander, "Molten-Salt Fast Reactors," Proceedings of the
Conference on Breeding, Economics and Safety in Large Fast Power Reactors
October 7-10, 1963, ANL-6792, p. 553-7, December 1963.
A
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ANTIKVITET
totale
FUEL SALT
OUT, 1300°F
"
FERTILE SALT
keliones con men mere om di
AL
OUT, 1200°F
w
TUBE,
HITE-
-BLANKET
(UNMODERATED)
I
FUEL
GRAPI

CORE
(GRAPHITE)
MODERATOR
BLOCK,
GRAPHITE
Illi
FUEL SALT
FERTII
PASSA
CROSS-SECTION, TYPICAL CORE
'TYPICAL
FERTILE SALT
PASSAGE
FUEL SALT
IN, 1000°F
FUEL SALT
IN, 1100°F
(Th232,
(U233,
Fig. 1. MSBR-I Conceptual Design for a Thermal, Two-Stream Core Reactor.
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Tara
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2
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CORE SALT
OUT, 1300°F
BLANKET SALT
OUT, 1200°F; :
-BLANKET
(UNMODERATED)
-
.-
--
-
E==============
di CORE
!! (GRAPHITE)
CORE
PASS
PARTITION,
· INOR-8 OR NIOBIUM.
TYPICAL CORE
SALT PASSAGE-
GRAPHITE-
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CROSS-SECTION, TYPICAL
MODERATOR BLOCK

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4
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...
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BLANKET SALT
- CORE SALT IN, 1100°F.
IN, 1000°F. : (Th<33);
(U233 Th)
Fig. 2. MSBR-II Conceptual Design for a Thermal, One-Stream Core Reactor.
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ORNL-DWG 65-11766
BLANKET SALT
OUT, 1200°F
FUEL SALT,
1400°F
PRIMARY HEAT
EXCHANGER
(BLANKET
(Th SALT)
COOLANT SALT.
OUT, 1200°F
CORE TANK,
INOR-8
CORE
(FUEL SALT)
BLANKET SALT
IN, 1100°F 7
COOLANT SALT ..
IN, 900°F
TYPICAL COOLANT LOOP
FUEL SALT
1100°F
Fig. 3. MOSHL(1) Conceptual Design for an Intermediate Energy, Externally cooled Reactor.
...
incore
1
.com

...
ORNL-DWG 11768
:
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...
-
:--
•
...
..
..
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BLANKET SALT
OUT, 1200°F
Pb IN,
900°F
CORE TANK,
INOR-8-
BLANKET:
(Th SALT)
: CORE
(FUEL SALT)
-TYPICAL DIRECT
CONTACT COOLANT
LOOP, NIOBIUM
LINED
-
.-
.
-..-
-
w
BLANKET SALT
IN, 1100°F
.
CORE SALLE
MOLTEN Pb
...Bo Pb OUT, 1200°F
Fig. 4. MOSEL (2) Conceptual Design for an Intermediate Energy, Direct-
Contact Cooled Reactor.
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DATE FILMED
0 / 21 /66
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