ORNL-3996.txt

 

   

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_';-IUC-SO Reactor Technology

 
 

 

    
   
    
     

 

" ‘BREEDER REACTORS

--‘-.:Pa'ul R -Kcst'en
E. S. Beths
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ORNL-3996

Contract No. W-7405-eng-26

DESIGN STUDIES OF lOOO-Mw(e) MOLTEN-SALT BREEDER REACTORS
Paul R. Kasten

E. S. Bettis
Roy C. Robertson

Molten-Salt Reactor Program

R. B. Briggs, Director

W

RELEASED FOR ANNOUNCEMENT

| IN NUCLEAR SCIENCE ABSTRACTS

e b gt A

 

 

AUGUST 1966

OAK RIDGE NATIONAL LABORATORY
Oak Ridge, Tennessee
operated by
UNION CARBIDE CORPORATION
for the
U.S. ATOMIC ENERGY COMMISSION

 

 
 

 

 

«t

-

iii
SUMMARY

Design and evaluation studies have been made of thermal-energy
molten-salt breeder reactors (MSBR) in order to assess their economic and
nuclear performancehand to identify important design and development prob-
lems. The reference reactor design presented here is related to molten-
salt reactors in general.

The reference design is a two-region two-fluid system, with fuel salt
separated erm the_blanket salt by graphite tubes. The fuel salt consists
of uranium fluoride dissolved in a carrier salt of 1lithium and beryllium
fluorides, and the blanket salt contains thorium fluoride dissolved in a
similar carrier salt. The energy generated in the reactor fluid is trans-
ferred to a secondary coolant-salt cifcuit, which couples the reactor to a
supercritical steam cycle. On-site fuel-recycle processing is employed,
with fluoride-volatility and vacuum-distillation operations used for the
fuel fluid, and direct-protactinium-removal processing applied to the
blanket stream. The resulting power cost for the reference plant, termed
MSBR(Pa), is less than 2.7 mills/kwhr(e); the specific fissile-material
inventory is only 0.7 kg/Mw(e), the fuel doubling time is about 13 years,
and the fuel-cycle cost is 0.35 mill/kwhr(e). The associated power dou-
bling time based on continuous investment of bred fuel is less than 9

years.

Reference MSBR Plant Design

Flowsheet

Figure 1 gives the flowsheet of the lOOO-Mw(e) MSBR power plant.
Fuel flows through the reactor at a rate of about 44 OOO gpm (velocity of
about 15 fps); it enters the core at 1000°F and leaves at 1300°F. The
prlmary fuel C1rcuit hag four loops, and each loop has a pump and a pri-
mary heat exchanger. Each of these pumps has a capaC1ty of 'about 11,000

- gpm. The four blanket-ealt pumps and heat exchangers, although smaller,

are similar to corresponding components in thegfuei system. The blanket
salt enters the reactor vessel at 1150°F and leaves-at 1250°F. The
blanket-salt pumps have a capacity of about 2000 gpm.

 
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551.7° ' |
BLANKET SALT HEAT COOLANT SALT COOLANT SALT! ! GEN.
EXCH. AND PUMPS ' ; ;.
PUMPS PUMPS ' TURBINE [ TuriNg [] 3077 Mwe
FUEL SALT HEAT i ., | _ Gross
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BOOSTER MIXING TEE
PUMPS
BLANKET SALT FUEL SALT COOLANT SALT PERFORMANCE
DRAIN TANKS DRAIN TANKS DRAIN TANKS NET OUTPUT 1000 Mwe
‘ LEGEND - GROSS GENERATION  1,034.9 Mwe
BF BOOSTER PUMPS 9.2 Mwe
FUEL == STATION AUXILIARIES 257 Mwe
BLANKET === wwe REACTOR HEAT INPUT 2225 Mwi
COOLANT —--— NET HEAT RATE 7,601 Blu/kwh
STEAM  ———=—— NET "EFFICIENCY 449 %
o 10% 1b /hr
Pmmmmmn P8I0
Mo Bt /1b.

e .. Fraeze Valve

Fig. 1. Reference MSBR Flow Diagram

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Four 14,000-gpm pumps circulate the coolant, which consists of a mix-
ture of sodium fluoride and sodium fluoroborate. The coolant enters the
shell side of the primary heat exchanger at 850°F and leaves at 1112°F.
After leaving the primary heat exchanger, the coolant salt is further
heated to 1125°F on the shell side of the blanket heat exchangers. The
coolant then circulates through the shell side of 16 once-through super-
heaters (four superheaters per pump). In addition, four 2000-gpm pumps
circulate a portion of the coolant through eight reheaters.

The steam system flowsheet is essentially that of the new Bull Run
plant of the Tennessee Valley Authority system, with modifications to in-
crease the rating to 1000 Mw(e) and to preheat the working fluid to 700°F
prior to entering the heat exchanger-superheater unit. A supercritical
power-conversion system is used that is appropriate for molten-salt appli-
cation and takes advantage of the high-strength structural alloy employed.
Use of a supercritical fluid system results in an overall plant thermal

efficiency of about 45%.

Reactor Design
Figure 2 shows the plan and elevation views of the MSBR cell arrange-

ment. The reactor cell is surrounded by four shielded cells containing
the superheater and reheater units; these cells can be individually iso-
lated for maintenance. The fuel processing plant, located adjacent to
the reactor, is divided into high-level and low-level activity areas.
The elevation view in Fig. 2 indicates the position of equipment in the
various cells.

Figure 3 gives an elevation view of the reactor cell and shows the

- location of the reactor, pumps, and fuel and blanket heat exchangers.

The Hastelloy N reactor vessel has a side-wall thickness of about 1.25
in. and a head thickness of about 2.25 in.; it is designed to operate at
1200°F and up to 150 psi. The plenum chambers at the bottom of the ves-
sel commmnicate with the external heat exchangers by concentric inlet-
outlet piping. The inner pipe has slip joints to accommodate thermal
expansion. Bypass flow through these slip joints is about 1% of the

total flow. As indicated in Fig. 3, the heat exchangers are suspended

 
 

vi

from the top of the cell and are located below the reactor.

Each fuel

pump has a free fluid surface and a storage volume that permit rapid

drainage of fuel fluld from the core upon loss of flow. In addition,

the fuel salt can be drained to the dump tanks when the reactor is shut

down for an.extended time.

The entire reactor cell is kept at high tem-

perature, while cold "fingers" and thermal insulation surround structural

support members and all special equipment that must be kept at relatively

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FUEL HEAT EXCH. Z REHEATERS

HEAT EXCH.

Fig. 2. Reactor and Coolant-Salt Cells — Plan and Elevation.

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BLANKET PUMP
MOTCR

FUEL PUMP
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COOLING =
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viii

low temperatures. The control rod drives are located above the core,
and the control rods are inserted into the central region of the core.
The reactor veséel, gbout 14 £t in diameter and about 19 ft high,
contains a 12.5-ft-high 10-ft-diam core assembly composed of reentry-
type graphite fuel elements. The graphite tubes are attached to the two
plenum chambers at the bottom of the reactor with graphite-to-metal
transition sleeves. Fuel from the entrance plenum flows up fuel passages
in the outer region of thé fuel tube and down through a single central

passage to the exit plenum. The fuel flows from the exit plenum to the

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heat exchangers and then to the_pumprand back to the reactor. An 18-in.-
thick molten-salt blanket plus a 3~in.-thick graphite reflector surround .
the core. The blanket salt also permeates the interstices of the core
lattice, and thus fertile material flows through the core without mixing
with the fissile fuel salt. |
The MSBR requires structural integrity of the graphite fuel element.

In order to reduce the effect of radiation damage, the fuel tubes have

been made small to reduce the fast flux gradient across the graphite wall.
Also, the tubes are anchored only at one end to permit axial movement.
The core volume has been made large in order to reduce the flux level in
the core. In addition, the reactor is designed to permit replacement of
the entire graphite core by remote means if required. _
Figure 4 shows a cross section of a fuel element. Fuel fluid flows
upward through the small passages and downward through the large central .
passage. The outside diameter of a fuel tube is 3.5 in., and there are
534 of these tubes spaced on & 4.8-in. triangular pitch. The tube as- -
semblies are surrounded by hexagonal blocks of moderator graphite with
blanket salt filling the interstices. The nominal core composition is
75% graphite, 18% fuel salt, and 7% blanket salt by volume.
In determining the design parameters of the MSBR, two different
methods were considered for removal of bred fuel from the reactor. The
designation MSBR(Pa) represents a plant in which protactinium is removed
directly from the blanket stream, whereas the designation MSBR corre-
sponds to remcval of uranium per se from the blanket. With the exception
of the blanket-processing step, the MSBR(Pa) and the MSBR plants have (;J
essentially the same design. Development of an MSBR(Pa) plant is the '

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FUEL PASSAGE (UP

 

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BLANKET PASSAGE

FUEL PASSAGE (DOWN)

 

 

 

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t - BRAZED JOINT
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= .\ : r. Ny b=
= 2 \&E ~\. ——FUEL INLET
: i s PLENUM
= -
1 3 FUEL OUTLET

e re PLENUM

Fig. 4. MSBR Graphite Fuel Element.

 
 

 

present goal of the molten-salt reactor program. A summary of the
parameter values determined for the MSBR(Pa) and MSBR designs is given
in Table 1.

Fuel Processing

 

The primary objectives of fuel processing are to purify and recycle
fissile and carrier components and to minimize fissile inventory while
holding 1osses to a low value. The fluoride volatility—vacuum distilla-
tion process fulfills these objectives through simple operations. The
process for direct protactinium removal from the blanket also appears to
be a simple one. o |

The core fuel for both the MSBR and the MSBR(Pa) is processed by
fluoride volatility and vacuum distillation operations. For the MSER,
blanket processing is accomplished by fluoride volatility alone, and the
processing cycle time is short enough to maintain a very low concentraé
tion of fissile material. The effluent UFg is absorbed by fuel salt and
reduced to UF, by treatment with hydrogen to reconstitute a fuel-salt

mixture of the desired composition. For the MSBR(Pa), the blanket stream

is treated with molten bismuth containing dissolved thorium; the thorium

displaces the protactinium from solution (as well as uranium). The metal-

lie protactinium and uranium are deposited on a metal filter and hydro-
fluorinated or fluorinated for recyéle of bred fuel.

Molten-salt reactors are inherently suited to the design of process-
ing facilities integral with the reactor plant; these facilities require
only a small amount of cell space adjacent to the reactor cell. Because
all services and equipment available to the reactor are available to the
processing plant and shipping and storage charges are eliminated, inte-
gral processing facilities permit significant savings in capital and
dperating costs. Also, the processing plant inventory of fissile mate-
rial is very low.

The principal steps in core and blanket stream processing of the
MSBR(Pz.) and the MSBR are shown in Fig. 5. A small side stream of each

fluid is continuously withdrawn from the fuel and blanket loops and circu-

lated through the processing system. After processing, the decontaminated

fluids are returned to the reactor system. Fuel inventories retained in

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xi

Teble 1. Reactor Design Values

 

 

MSBR(Pa.) MSBR
Power, Mw
Thermal 2225
Electrical 1000
Thermal efficiency, fraction 0.449
Plant load factor 0.80
Reactor wvessel
Outside diameter, ft 14
Overall height, ft ~19
Wall thickness, in. 1.5
Head thickness, in. 2.25
Core
Height of active core, ft 12.5
Diameter, ft 10
Number of §raphite fuel passage tubes 534
Volume, ft 282
Volume fractions :
Fuel salt 0.169 0.169
Blanket salt 0.073 0.074
Graphite moderator 0.758 0.757
Atom ratios
Thorium to uranium 42 40
Carbon to uranium 5800 5440
Neutron flux, core average, neutrons/cmzosec
Thermal 7.2 x 1014 6.7 x 1014
Fast 12.1 x 10*#4 12.1 x 104
Fast, over 100 kev 3.1 x 10*4 3,1 x 10t
Power den31ty, core average, kw/liter
Gross 80
In fuel salt 473
Blanket
Radial thickness, ft 1.5
Axial thickness, ft 2.0
Volume, £t3 1120
Volume fraction, blanket salt 1.0
Reflector thickness, in. 3
Fuel salt
Inlet temperature, °F 1000
Outlet temperature, °F 1300
Flow rate, ft’/sec (total) 95.7
£pm 42, 950
Nominal volume holdup, £t3
Core 166
Blanket 26
Plena 147
Heat exchangers and plping 345
Processing plant .33
717

- Total

 
 

 

 

xii

Table 1 (continued)

 

 

MSBR(Pa) MSBR
Fuel salt (continued)
Nominal salt composition, mole %
LiF 63.6
BeFa 36.2
UF; (fissile) 0.22
Blanket salt
Inlet temperature, °F 1150
Outlet temperature, °F 1250
Flow rate, ft3/sec (total) 17.3
g£pm 7764
Volume holdup, £%3
Core 72
Blanket - 1121
Heat exchanger and piping 100
Processing : 24
Storage for protactinium decay 2066
Total 1317 3383
Salt composition, mole %
LiF 71.0
T11F4 27.0
UF; (fissile) 0.0005
System fissile inventory, kg 681 769
System fertile inventory, kg 101,000 260, 000
Processing data
Fuel stream
Cycle time, days 42 47
Rate, ft3/day 16.3 14.5
Processing cost, $/ft> 190 203
Blanket stream
Equivalent cycle time, days
Uranium-removal process 55 23
Protactinium-removal process 0.55
Equivalent rate, ft3 per day -
Uranium-removal process 23.5 144
Protactinium-removal process 2350
Equivalent processing cost (based on 65 7.3
uranium removal), $/ft>
Fuel yield, %/yr 7.95 4.86
Net breeding ratio- 1.071 1.049
Fissile losses in processing, atoms per 0.0051 0.0057
fissile absorption S
Specific inventory, kg of fissile material 0.681 0.769
per megawatt of electricity produced
Specific power, Mw(th)/kg of fissile material 3.26 2.89
Fraction of fissions in fuel stream 0.99% 0.987
Fraction of fissions in thermal-neutron group 0.815 0.806
Net neutron production per fissile 2.227 2.221

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Fuel- and Fertile-Stream Processing for the MSBR and MSBR(Pa).

 

TTIX

 

 
 

 

xiv

the processing plant are estimated to be about 5% of the reactor system

for core processing and less than 1% for blanket.processing.

Heat'Exdhangg_and Steam Systems

The structural material is Hastelloy N for all components contacted
by molten salt in the fuel, blanket, and coolant systems, including the
reactor vessel, pumps, heat exchangers, piping, and storage tanks. The-
primary heat exchangers are of the tube-and-shell type, with fuel salt
on the tube side. Each sheil contains two concentric tube bundles at-
tached to fixed tube sheets. Fuel flows through the two bundles in series;
it flows downward in the inner section of tubes, enters a plenum at the
bottom of the ekchanger, and then flows upward to the pump through the
outer section of tubes. The coolant salt enters at the top of the ex-
changer and flows on the baffled shell side down the outer annular re-
gion; if then flows upward in the inner annular section before exiting
through a pipe centrally placed in the exchanger. '

Since a large temperature difference exists in the two tube sections,
the design permits differential tube expansion. Changes in tube lengths
due to thermal conditions are accommodated by the use of a sine-wave type
of construction, which permits each tube to adjust to thermal changes.

The blanket heat exchangers increase the temperature of the coolant
leaving the fuel heat exchangers. The design of these units is similar
to that used in the fuel heat exchangers.

The superheater is a U~tube U-shell heat exchanger'that has disk and
doughnut baffles with varying spacing; it is a long, slender exchanger.
The baffle spacing is established by the shell-sidé pressure drop and by
the temperature gradient across the tube wall; it is greatest in the
central portion of the exchanger where the temperature difference between
the fluids is high. The supercritical fluid enters the tube side of the
superheater at 700°F and 3800 psi and leaves at 1000°F and 3600 psi.

The reheaters transfer energy from the coolant salt to the working
fluid before its use in the intermediate pressure turbine. A shell-and-
tube exchanger is used that produces steam at 1000°F and 540 psi.

Since the freezing temperature of the secondary coolant salt is about

700°F, a high working fluid inlet temperature is required. Preheaters,

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XV

along with prime fluid, are used in raising the temperature of the work-
ing fluid entering the superheaters. Prime fluid goes through a pre-
heater exchanger and leaves at a pressure of 3550 psi and about 870°F.
It is then injected into the feedwater in a mixing tee to produce fluid
at 700°F and 3500 psi. The pressure is increased to about 3800 psi by

a pressurizer (feedwater pump) before the fluid enters the superheater.

Capital Cost Estimates

Reactor Power Plant

 

Preliminary estimates of the capital cost of a 1000-Mw(e) molten-
salt breeder reactor power station indicate a direct construction cost
of about $80.7 million. After applying the indirect cost factors asso-
ciated with reactor construction, an estimated total plant cost of $114.4
million is obtained for private-financing conditions and $110.7 million
for public financing. A summary of plant costs is given in Table 2.
The relatively low capital cost estimate obtained is due to the small
physical size of the reactors and associated equipment, the high thermal
efficiency, and the simple control requirements.

The operating and maintenance costs of the reactor power plant were
estimated by standard procedures and were modified to reflect present-day

salaries. These costs amount to 0.34 mill/kwhr(e).

Fuel-Recycle Plant

The capital costs associated with fuel-recycle equipment were ob-
tained by itemizing and costing the major process equipment required and
estimating the costs of site, buildings, instrumentation, waste disposal,
and building services associated with fuel recycle.

Table 3 summarizes direct construction costs, indirect costs, and
total costs associated with an intégrated processing facility having
approximately the capacity required for a 1000-Mw(e) MSBR plant. The
total construction cost was estimated to be about $5.3 million; in ob-
taining this figure, the indirect charges amounted to about 100% of the

direct construction cost. The high value used for the indirect charges

 
 

 

XVl

 

 

Table 2. Preliminary Cost-Estimate Summary® for a 1000-Mw(e) Molten-Salt
Breeder Reactor Power Station [MSBR(Pa} or MSBR]
Federal
Power Costs
Commission (in thousands of dollars)
Account
20 Land and Land Rights ] 360
21 Structures and Improvements !
211 Ground improvements 866 .
212 Building and structures
.1 Reactor building® o 4,181
.2 Turbine building, auxiliary building, and feedwater 2,832
heater space )
.3 Offices, shops, and laboratories 1,160 2
.4 Waste disposal building 150
.5 Btack 76
.6 Warehouse 40
.7 Miscellaneous 30
Subtotal Account 212 8,469
Total Account 21 9,335
22 Reactor Plant Equipment
’ 221 Reactor equipment
.1 Reactor vessel and internals 1,610
.2 Control rods 250
«3 Bhielding and containment 2,113
+4 Heating-cooling systems and vepor-suppression system 1,200
.5 Moderator and reflector 1,089
.6 Reactor plant crane 265
Subtotal Account 221 6,527
222 Heat transfer systems
.1 Reactor coolant system 6,732
+2 Intermediate cooling system 1,947
.3 Steam generator and reheaters 9,853
+4 Coolant supply and treatment 300
Subtotal Account 222 18,832
223 Nuclear fuel handling and storage (drain tanks) 1,700 =
224 Nuclear fuel processing and fabrication (included in (c)
fuel-cycle costs )
225 Redioactive waste treatment and disposal (off-gas 450 <
system)
226 Instrumentation and controls 4,500
227 TFeedwater supply and treatment 4,051
228 Steam, condensate, and feedwater piping 4,069
229 Other reactor plant equipment (remote maintenance) 5,000d
Total Account 22 45,129

 

 

 

 

 

aEstimates are based on 1966 costs for an established molten-salt nuclear power plant industry.
bContainment cost is included in Account 221.3.
®see Table 3 for these costs.,

dThe allowance for remote maintenance may be too high, and some of the included replacement
equipment allowances could be classified as operating expenses rather than first capital costs. o

"
 

 

 

 

 

wh

w)

Xvii

Table 2 (continued)

 

 

 

 

 

Federal
Power Costs
Commission (in thousands of dollars)
Account
23 Turbine-Generstor Units
231 Turbine-generator units 19,174
232 Circulating-water system 1,243
233 Condensers and auxiliaries 1,690
234 Central lube-oil systen 80
235 Turbine plant instrumentation 25
236 Turbine plant piping 220
237 Axuiliary equipment for generator 66
238 Other turbine plant eguipment 25
Total Account 23 22,523
24 Accessory Electrical
241 Switchgear, main and station service 500
242 Switchboards 128
243 Station service transformers 169
244 Auxiliary generator 50
245 Distributed items 2,000
Total Account 24 2,897
25 Miscellaneous 800
Total Direct Construction Cost® 80,684
Private Financing
Total indirect cost 33,728
Total plant cost 114,412
Public Financing
Total indirect cost 30,011
Total plant cost 110,695

 

®Does not include Account 20, Land Costs.
However, land costs were included when computing indirect costs.

Land is treated as a nondepreciating capital item.

 
 

 

xviii

Table 3. Summary of Processing-Plant Capital Costs
for a 1000-Mw(e) MSBR

 

Installed process equipment $ 853,760
Structures and improvements 556,770
Waste stofage 387,970
Process piping 155,800
Process instrumentation 272,100
Electrical auxiliaries | 84,300
Sampling connections 20, 000
Service and utility piping | 128,060
Insulation 50,510
Radiation monitoring 100, 000
Total direct cost $2,609, 270
Construction overhead 782,780
(30% of direct costs) — e
Subtotal construction cost $3,392,050
Engineering and inspection 848,010
(25% of subtotal construction cost) —_—
Subtotal plant cost | $4, 240,060
Contingency (25% of subtotal 1,060,020
plant cost)

Total capital cost $5,300, 080

 

should more than compensate for the higher rates of equipment replacement
in the fuel-processing plant as compared with the power plant as a whole.

The operating and maintenance costs for the fuel-recycle facility
include labor, labor overhead, chemicals, utilities, and maintenance mate-
rials. The total annual operating and maintenance costs for a processing
facility having a throughput of 15 ft3 of fuel salt per day plus 105 ft>
of fertile salt per day is estimated to be about $721,000. A breakdown
of these charges is given in Table 4.

These capital and operating costs were used as base points for ob-
taining the costs for processing plants having different capacities. For

each fluid stream the capital and operating costs were estimated separately

3]

[

("
 

 

 

wh

)

.

 

Xix

Table 4. Summary of Annual Operating
and Maintenance Costs for Fuel
Recycle in a 1000-Mw(e) MSBR

 

Direct labor $222,000
Iabor overhead 177,600
Chemicals 14,640
Waste containers 28,270
Utilities 80, 300
Maintenance materials

Site 2,500

Services and utilities 35,880

Process equipment 160, 040
Total annual charges $721,230

 

as a function of plant throughput based on the volume of salt processed.
The results of these estimates, given in Fig. 6, were used in calculating
the nuclear and economic performance of the fuel cycle as a function of
fuel-processing rate.

For the MSBR(Pa) plant, the processing methods and costs were the
same as those for the MSBR, except for blanket-stream processing. The
cost of direct protactinium removal from the blanket stream was estimated

to be
c(Pa) = 1.65R0°45 , (1)

where C(Pa) is the capital cost of protactinium-removal equipment, in
millions of dollars; and R is the blanket-stream processing rate for prot-
actinium removal, in thousands of cubic feet of blanket salt per day.
Thus, the cost of fuel recycle in the MSBR(Pa) was estimated to be equiva-
lent to the costs given by Eg. (1) and Fig. 6 based on uranium being re-
moved from the blanket stream by the fluoride volatility process and the

rate of uranium removal being influenced by the rate of protactinium re-

moval.

 
 

 

 

) ORNL-DWG 66-T455
BLANKET STREAM PROCESSING RATE (ft*/day)

2 5 10° 2 ’ 5 10

AL COST OF CORE PROCESSING

TING COST OF CORE PROCESSING

N

CORE STREAM PROCESSING COST (:S7£3)
BLANKET STREAM PROCESSING COST (S7ft%)

COST OF
OPERATING COST OF ——, BLANKET PROCESSING
BLANKET PROCESSI CINO

 

2 5 10 2 5 100
CORE STREAM PROCESSING RATE (ft¥day)

Fig. 6. MSBR Fuel-Recycle Costs As a Function of Processing Rates.
Fluoride volatility plus vacuum distillation processing for core; fluo-
ride volatility processing for blanket; 0.8 plant factor; 12%/yr capital
charges for investor-owned processing plant.

Fuel-Cycle Performance

 

The objective of the nuclear desigh calculations was primarily to
find the conditions that gave the lowest fuel-cycle cost and, then, with-
out appreciably increasing this cost, the conditions that gawve highest
fuel yield.

Analysis Procedures and Basic Assumptions

The nuclear calculations were performed with a multigroup, diffusion,
equilibrium reactor program, which calculated the nuclear performance,
the equilibrium concentrations of the various nuclides, including the

fission products, and the fuel-cycle cost for a given set of conditions.
 

'y

 

 

xxi

The 12-group neutron cross sections were obtained from neutron spectrum
calculations, with the core heterogeneity taken into consideration in the
thermal-neutron-spectrum computations. The nuclear designs were optimized
by parameter studies, with most emphasis on minimum fuel-cycle cost and
with lesser weight given to maximizing the annual fuel yield. Typical
parameters varied were the reactor dimensions, blanket thickness, frac-
tions of fuel and fertile salts in the core, and the fuel- and fertile-
stream processing rates.

The basic economic assumptions employed in obtaining the fuel-cycle
costs are given in Table 5. The processing costs afe based on those given
in the previous section and are included in the fuel-cycle costs. A fis-
sile material loss of 0.1% per pass through the fuel-recycle plant was
applied.

The effective behavior used in the fuel-cycle-performance calcula-
tions for the various fission products was that given in Table 6. A gas-
stripping system is provided to remove fission-product gases from the
fuel salt. In the calculations reported here, a *3°Xe poison fraction of

0.005 was applied.

Table 5. Economic Ground Rules Used in
Obtaining Fuel-Cycle Costs

 

Reactor power, Mw(e) 1000
Thermal efficiency, % 45
Load factor 0.80

Cost assumptions B
Value of 33U and 233Pa, $/g 14
Value of 235U, $/g 12
Value of thorium, $/kg 12
Value of carrier salt, $/kg _ 26

Capital charge, %/yr
Private financing

Depreciating capital 12
Nondepreciating capital 10
Public financing |
Depreciating capital 7
Nondepreciating capital 5

Processing cost: given by curves
in Fig. 6, plus cost given by
Eq. (1), where applicable

 

 
 

 

 

xxii

Table 6. Behavior of Fission Products
in MSBR Systems

 

Behavior | Fission Products

 

Elements firesent as gases; assumed to be Kr, Xe
removed by gas stripping (a poison -
fraction of 0.005 was applied)

Elements that form stable metallic colloids; Ru, Rh, Pi, Ag, In
removed by fuel processing |

Elements that form either stable fluorides Se, Br, Nb, Mo, Tec, 3
or stable metallic colloids; removed by Te, I
fuel processing _
Elements that form stable fluorides less . Sr, Y, Ba, Ia, Ce,
volatile than LiF; separated by vacuum Pr, Nd, Pm, Sm,
distillation , Eu, G4, Tb '
Elements that are not separated from the Rb, C4, Sn, Cs, Zr

carrier salt; removed only by salt discard

 

The control of corrosion products in molten-salt fuels does not
appear to be a significant problem, and the effect of corrosion products
was neglected in the nuclear calculations. The corrosion rate of Hastel-
loy N in molten salts is very low; in addition, the fuel-processing
operations can control cbrrosion-product buildup in the fuel.

The important parameters describing the MSBR and MSBR(Pa) designs
are given in Table 1. Many of the parameters were fixed by the ground :
rules for the evaluation or by engineering-design factors that include
the thermal efficiency, plant factor, capital charge rate, maximum fuel
velocity, size of fuel tubes, processing costs, fissile-loss rate, and
the out-of-core fuel inventory. The parameters optimized in the fuel-
cycle calculations were the reactor dimensions, power density, core compo-
sition (including the carbon-to-uranium and thorium-to-uranium ratios),

and processing rates.

Nuclear Performance and Fuel-Cyclé Cost

The general results of the nuclear calculations are given in Table 1;

the neutron-balance results are given in Table 7. The basic reactor QEJ

!

L3
 

1]

¥

 

xxiii

Table 7. Neutron Balances for the MSBR(Pa) and the MSBR Design Conditions

 

 

 

 

 

 

 

 

 

MSBR(Pa) MSBR
Neutrons per Fissile Absorption Neutrons per Fissile Absorption
Material
Total %:igr:?d Neutrons Total g:sgrb?d Neutrons
Absorbed Fi u.lng Produced Absorbed ? u?lng Produced
ission Fission
232my, 0.9970 0.0025 0.0058 0.9710 0.0025 0.0059
233p, 0.0003 0.0079
233y 0.9247 0.8213 2.0541 0.9119 0.8090 2.0233
234y 0.0819 0.0003 0.0008 0.0936 0.0004 0.0010
2357y 0.0753 0.0607 0.1474 0.0881 0.0708 0.1721
236y 0.008% 0.0001 0.0001 0.0115 0.0001 0.0001
2375p 0.0009 0.0014
238y 0.0005 0.0009
Carrier salt 0.0647 0.0186 0.0623 0.0185
(except 6Li)
614 0.0025 0.0030
Graphite 0.0323 0.0300
135%e 0.0050 0.0050
149gn 0.0068 0.00692
1539m 0.0017 0.0018
Other fission 0.0185 0.019
products
Delayed neutrons 0.0049 0.0050
lost®
LeakageP 0.0012 0.0012
Total 2.2268 0.8849 2.2268 2.2209 0_. gaz28 2.2209

 

aDelayed neutrons emitted outside core.

b

Leakage, including neutrons absorbed in reflector.

design has the advantage of zero neutron losses to structural materials

in the core other than the moderator.

Except for the loss of delayed

neutrons in the external fuel circuit, there is almost zero neutron leak-

age from the reactor because of the thick blanket.

The neutron losses

to fission products are low because of the low cycle times associated

with fission-product removal.
The components of the fuel-cycle cost for the MSBR(Pa) and the MSBR

are sumarized in Table 8.

The main components are the fissile inventory

 
 

XXiv

Table 8. Fuel-Cycle Cost for MSBR(Pa) and MSBR Plants®

 

MSBR(Pa) Cost (mill/kwhr)

MSER Cost [mill/kwhr(e)]

 

 

el T oatoray om Rl feriile g gnd
Fissile inventory? 0.1125 0.0208 0.1333 0.1180 0.0324 0.1504
Fertile inventory 0.0000 0.0179 0.0179 0.0459 0.0459
Salt inventory 0.0147 0.0226 0.0373 0.0146 0.0580 0.0726
Total inventory 0.188 0.269
Fertile replacement 0.0000 0.0041 0.0041 0.0185 0.0185
Salt replacement 0.0636 0.0035 0.0671 0.0565 0.0217 0.0782
Total replacement 0.071 0.097
Processing 0.1295 0.0637 0.1932 0.1223 0.0440 0.1663
Total processing 0.193 0.166
Production credit (0.105) (0.073)
Net fuel-cycle cost 0.35 0.46

 

®Based on investor-owned power plant and 0.80 plant factor.
Prncluding 23%Pa, 233y, ana 22%U.

and processing costs. The inventory costs are rather rigid for a given

reactor design, since they are largely determined by the external fuel

volume. The processing costs are a function of the processing-cycle

times, one of the chief parameters optimized in this study. As shown

by the results in Tables 1 and 8, the ability to remove protactinium

directly from the blanket stream has a marked effect on the fuel yield
and lowers the fuel-cycle cost by about 0.1 mill/kwhr(e). This is due

primarily to the decrease in neutron absorptions by protactinium when

this nuclide is removed from the core and blanket regions.

In obtaining the reactor design conditions, the optimization pro-

cedure considered both fuel yield and fuel-cycle cost as criteria of

performance. The corresponding fuel-cycle performance is shown in Fig. 7,

which gives the minimum fuel-cycle cost as a function of fuel-yield rate

based on privately financed plants and a plant factor of 0.8. The de-
sign conditions for the MSBR(Pa) and MSBR concepts correspond to the

designated points in Fig. 7.

</

e

o

"
 

*

»h

 

XXV

ORNL-DWG 66-T456

 

 

 

 

 

 

 

 

 

 

 

 

 

 

0.6
l%l
2 05
x
x
2 ®MSBR DESIGN POINT
& /
-
2 04 _ /
o
W ///
o MSBR(Pa) DESIGN POINT
o
d /
S 03 e
w

0.2

2 3 4 5 6 7 8 9

FUEL YIELD (% /yr)

Fig. 7. Variation of Fuel-Cycle Cost with Fuel Yield in MSBR and
MSBR(Pa) Concepts.

Power-Production Cost and Fuel-Utilization
Characteristics

The power-production costs are based on the capital costs given
above, operation and maintenance charges, and fuel-cycle costs. Table 9
sumearizes the power-production cost and the fuel-utilization charac-
teristics of the MSBR(Pa) and MSBR plants. The results illustrate that
both concepts produce power at low costs and that the fuel-utilization
characteristics for the MSBR(Pa) plant are excellent and those for the
MSER are good. Measuring these characteristics in terms of the product
of the specific fissile inventory and the square of the doubling time,
the MSBR(Pa) concept is comparable to a fast breeder reactor with a
specific inventory of 3 kg of fissile material per megawatt of electriéity
produced and a doubling time of 6 years, while the MSBR plant is compa-
rable to the same fast breeder with a doubling time of 10.5 years.

 
 

 

 

Table 9. Power-Production Cost and Fuel-Utilization Characteristics
of the MSBR(Pa) and the MSBR Plants®

 

 

 

MSBR(Pa) MSER
Specific fissile inventory, 0.68 0.77
kg/M(e)
Specific fertile inventory, 105 268
kg/Ma(e)
Breeding ratio _ 1.07 1.05
Fuel-yield rate, %/yr 7.95 4,86
Fuel doubling time,P years 12.6 20.6
Power doubling time,© years 8.7 14.3
Private Public Private Public
Financing Financing Financing Financing
Capital charges, mills/kwhr(e) 1.95 1.10 1.95 1.10
Operating and meintenance cost, 0.34 0.34 0.34 0.34
mill/kwhr(e)
Fuel-cycle cost,d mill/kwhr(e) 0.35 0.20 0.46 0.29
Power-production cost, mills/kwhr(e) 2.64 1.64 2.75 1.73

 

&Based on 1000-Mw(e) plant and a 0.8 load factor. Private financing con-
siders a capital charge rate of 12%[yr for depreciating capital and of 10%/yr for
nondepreciating capital; public financing considers a capital charge rate of
7%/yr for depreciating capital and 5%/yr for nondepreciating capital.

bInverse of the fuel-yield rate.

cCa.pability based on continuous investment of the net bred fuel in new re-
actors; equal to the reactor fuel doubling time multiplied by 0.693.

dCosts of on-site integrated processing plant included in this wvalue.

Studies of Alternative Molten-Salt Reactor Designs

Modular-Type Plant

An important factor in maintaining low power-production costs is
the ability of the power plant to maintain a high plant-availability
factor. A modular-type MSBR plant, termed MMSBR, was therefore investi-
gated to determine the practicality of a four-module plant. Stoppage
of a fuel pump in such a system would shut down only one~quarter of the

station capacity, leaving 75% available for power production.
L)

 

«¥

 

 

 

xxvii

The MMSBR design includes four separate and identical reactors,
along with their separate salt circuits. The designs of the heat ex-
changers, the coolant-salt circuits, and the steam-power cycle remain
essentially as for the MSBR. Each reactor module generates thermal power
equivalent to that required for a net production of 250 Mw(e). The flow
diagram given in Fig. 1 is applicable to the MMSBR. The new features of
the MMEBR design are indicated in Fig. 8, which illustrates the four
distinct reactor vessels and cells, along with their adjacent steam-
generating cells.

The reactor core consists of 210 graphite fuel cells operating in
parallel within the reactor tank. The core region is cylindrical, with
a diameter of about 6.3 ft and a height of about 7.9 ft. Each reactor
vessel is about 12 ft in diameter and about 14 £t high.

The nuclear and fuel-cycle performance of the MMSER was also studied
for protactinium removal from the blanket stream; this case is termed
MMSBR(Pa). The results indicate that the nuclear and fuel-cycle per-
formance of a modular-type plant compares favorably with that of a single-
reactor plant; the modular plant tends to have slightly higher breeding
ratio, fissile inventory, and fuel-cycle cost; the power-production cost

is virtually the same as for the MSBR plant.

Additional Design Concepts

Other molten-salt reactor designs were studied briefly. In general

'the technology required for these alternative designs is relatively un-

developed, although_there are experimental data that support the feasi-
bility of each concept. An exception is the molten-salt converter reactor
(designated MSCR), whose application essentially requires only scaleup of
MSRE and associated fuel-proéessing technology. Howéver, the MSCR is not
a breeder, although it approaches breakeven breeder operatidn. The addi-
tional concepts are termed MSBR(Pa-Fb) , SSCB(Pa), MOSEL(Pa-Fb), and MSCR.
The MSBR(Pa-Pb) designation refers to the MSBR(Pa) modified by use of
direct-contact cooling of the molten-salt fuel with leten lead. Iead is
immiscible with molten salt and can be used as a heat exchange medium

within the reactor vessel to significantly lower the fissile inventory

 
 

* (4ISWA) $UBTd YESW odAT-TeTnpoW °8 °JTd

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—1

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"

XXix

external to the reactor. The lead also serves as a heat transport medium
between the reactor and:the steam generators.

The SSCB(Pa) designation refers to a Single-Stream-Core Breeder with
direct protactinium removal from the fuel stream. This is essentially a
single~region reactor having fissile and fertile material in the fuel
stream, with protactinium removal from this stream; in addition, the
core region is enclosed within a thin metal membrane and is surrounded
by a blanket of thorium-containing salt. Nearly all the breeding takes
place in the large core, and the blanket "catches" only the relatively
small fraction of neutrons that "leak" from the core (this concept is
also referred to as the one-and-one-half region reactor).

The MOSEL(Pa-Pb) designation refers to a MOlten-Salt Epithermal
breeder having an intermediate-to-fast energy spectrum, with direct prot-
actinium removal from the fuel stream and direct-contact cooling of the
fuel region by molten lead. No graphite is present in the core of this
reactor.

The MSCR refers to a Molten-Salt Converter Bgactor that has the
fertile and fissile material in a single stream. No blanket region is
employed, although a graphite reflector surrounds the large core.

The fuel-cycle performance characteristics for these reactors are

summarized in Table 10; in all cases the methods, analysis procedures,

Table 10. Suwmary of Fuel-Cycle Performance for
Reactor Designs Studied

 

 

Specific
. Fuel — poeeding ~ Fuel-tyele  picsite
eactor Yield Ratio .007t Inventory
(%/yr) (mil1/loenr) o Age(e)]
MSBR(Pa) 7.95 1.07 0.35 0.68
MSBR 4.86 1.05 0.46 0.77
MMSBR(Pz.) 7.31 1.07 0.38 0.76
MSBRsPa-Pb) 17.3 1.08 0.25 0.34
SSCB(Pa.) 6.63 1.06 0.37 0.68
MOSEL(Pa-Pb) 10.3 1.14 0.13 0.99
MSCR 0.96 0.57 1.63

 

 
 

and economic conditions employed were analogous to those used in obtain-
ing the reference MSBR design data. In general, fuel recycling was based
on fluoride volatility and §acuumrdistillation Proceésing; direct prot-
actinium removal from the reactor system was also considered in specified
é cases. |
% IThe results indicate the potential performance of fluoride-salt
systems utilizing a direct-contact coolant such as molten lead and the
versatility of molten salts as reactor fuels. They also illustrate that
single-region reactors based on MSRE technology have good performance
characteristics. Since the capital, operating, and maintenance costs of
the MSCR should be comparable with those of thé MSBR, the power-production
cost of an investor-owned MSCR plant should be about 2.9 mills/kwhr(e)

based on a lcad factor of 0.8. However, the lower power costs of the

 

MSBR(Pa.) and MSBR plants and their superior nuclear and fuel-conservation

characteristics make development of the breeder reactors preferable.

 
 

 

 

 

L)

XXX
CONTENTS
S[JWARY llllllllll 4 % & g 9% 2 & 3 8 e # & 8 0 0@ 4 & " % 0 0 % 8 0 98 & BB BB S O BB . *
l 2 INTRODUCTION IIIIIII “« 5 o » 8 8 * 8 0 ® 5 0 P " 8 * % 9 " & 9 & 0 B 8 8 0 * 3 & & 8 & 8 2
1.1 General Purpose of Study ..voeveevevee sesassssases sesane
1.2 Power Cost and Nuclear Performance GoOalsS .eeeeescsss .

103 Scope Ofstudy LI I I T D IR N B B B B B B B O RN R RN Y N I B TN B B N N B I I R R A R R ]

1.4 Study Organization and Participating Personnel ........

1.5 Acknowledgements ..eeveieeees seenses cecsesnnsanos creeae
2. MOLTEN-SALT REACTOR TECHNOLOGY +evevesncascencss vevessncesaas
2.1 Chemical Develbpment ............. cesecrtacenseus ciesae
2.2 Structural Material Development ......... cesessses ceans
2.3 Fuel Processing Development seeeeeesessrsessosnse ceensas
2.4 Component Development ..... ettt eeeneenete e e
3. INITIAL DESIGN OF A 1000-Mw(e) MSBR POWER STATION ....eevee.

3.1 General Design Criteria, Cost Bases, and Ground
Rl-]—]—es LR IR I R N BN BN B A S I I I D I I I Y B RN RN I BN BN RN I AN R DN Y Y BT BT RN TN IR R BN NN B B B RN B N )

3.2 General Plant Layoul .iceeevstrvecsercossscosscasrssonas
3.3 TFlowsheets and General Description eeeececesecsecesanes
3.4 Reactor SystTem .veeieeeresescrreceerscsorrascesscssacsaes
3.4.1 Description .cveveveresens toacecssessessensnnsus
3.4.2 Reactor Materials ..eivieecoceesnonnsas cesesenee
3.4.3 MSBR Load-Following CharacteristicCs .seeveeeeeses
3.5 Nuciear Fuel-Cycle PErTOTMANCE s eeseosnneocecnnnnesane

3.5.1 Design and Cost of Processing Plant for
- FuelReCyCle & 8 & 2 80 8 0 6 3 8 00 PN NSNS S AE S esEe

3.5.2 Nuclear Design Method I T
3.5.3 Nuclear Performance and Fuel-Cycle Cost ........
3.5.4 Critique of Nuclear Performance Calculations ...
3.6 Off-Gas SySTem cvvesnrsessoseransssssenssosscsssacsensnss
3.7 Heat Exchangers ........,.....................;........
3.7.1 Fuel-Salt Heat EXChangers t.eeecvesecccases veeas

3.7.2 Blanket-Salt Heat Exchangers ..... cheerecenaas oo
3 -7 o3 BOiler-SuPerheaterS @ & F 0 P 0 4P e e N “ s 0 000 0 .

Page

 

iji

0 o0 NN P

i~
o W

16
19
26
34
34
41
42
43

43
54
56
61
65
68
68
73
77

 
 

 

 

 

3.8
3.9
3.10

3.11
3.12

xxxii

3.7.4 Steam Reheaters ceeceeses
3 07 05 Reheat-steam Preheaters (R IEEN]
Salt Circulating Pumps «.eeccescevcans

Steam-Power System ......

® 8 & 5 & 5 8 0 0 0 D OB PP QRN

Other Design Considerations cevececesecssoscacososne

3.10.1 Piping and Pipe Stresses .veeseses

¢ 8 &9 0 8 0

3'10.2 MaintenanCe 8 80 ¢ 0 2 8 ¢4 8 5 0 800 2 S 0GP SR SR N A

3.10.3 Containment ....

Plant Construction Costs

Power-Production Cost ..

LR N BE BE BN BN BN

4 & & & & 508 0 288NN R

4, ALTERNATIVE CONDITIONS FOR MSBR DESIGN .....

4ol
4.2
4.3
4oy

"5 0 & o s 00

4 86 8 85 8 0 5 0 0 0 "0 BB D

* % &0

. e o 00

* P

Protactinium Removal from Blanket Stream eeeeeevesscesss

Alternative Feedwater Temperature Cycle ...cevevees
Modular-Type Plant ......

Additional Design Changes

5 080823

L I N BN

.00

5. ALTERNATIVE MOLTEN-SALT REACTOR DESIGNS ....

5.1
5.2

5.3
5.4

6. EVALUATION

References

MSBR (Pa-Pb) Concept
SSCB(Pa) Reactor Concept seseeses

« & o & 0 8

* v b F e

* &P s et ae

L2 BN N B B

4 6 & 0 % 0 8 & 0 8 s a0 s 0 0

5.2.1 SSCB(Pa) Reactor Concept with Intermediate

Coolant

8 8 ¢ 0 8 @

* o8 e S e s

5.2.2 SSCB(Pa) with Direct-Contact Cooling .......

MOSEL(Pa-Pb) Reactor Concept seeveeosss

MSCR Concept ...

o 8 b 89 0 o

6 0 0 0 & &P B B s E IS

80
83
87
90
95
95
95
o7
97
105
110
110
118
123
129
135
136
140

140
141
143
145
147
148
 

 

 

 

1. INTRODUCTICN

1.1 General Purpose of Study

 

An important objective of the AEC commercial nuclear power program
is to develop reactors that produce low-cost power and at the same time
conserve nuclear-fuel resources. Since the most important factor in com-
mercial application of reactors is power production cost, fuel utilization
aspects should be consistent with generation of low-cost power over a
given period of time. However, in evaluating economic factors, future
conditions must also be properly weighed and taken into consideration.

The general purpose of the studies discussed here was to determine
the incentive for molten-salt reactor development within the context of
low power cost and good fuel utilization. An associated objective was
to define important problems that need to be overcome prior to commercial
application of molten-salt reactors. -

1.2 Power Cost and Nuclear Performance Goals

The desirability of developing a given type of power reactor depends
on its performance relative to that of alternative concepts. This per-

formance is measured in terms of the power-production cost and the fuel-

utilization characteristics., Based on the accounting practices of in-

vestor-owned utilities, present-day light-water reactor plants generating
1000-Mw(e) appear capable of pfoducing power for about 4.0 mills/kwhr(e).
At the same time, substantiél AEC support is being given to the high-
temperature gas-cooled (HTGR) and the heavy-wafler-moderated organic-
cooled (HWOCR) reactor concepts, which appear capable of proaucing power
for about 3.5 mills/kwhr(e) in privately owned 1000-Mw(e) plants. For a
new type of reactor to_merit sérious attention, it should be judged capable
of producing even lower cost power in iOOO-MW(é) ififiestor-owned plants;
therefore, a goal of this study was to estimate the power-cost performance
of molten-salt breeder reactors to determine their competitive position.
As more nuclear power plants are built, the efficient use of our

nuclear fuels becomes increasingly important. New reactors must have the

 
 

 

potential of producing low-cost powér from more expensive fuel resources
orrpreferably of conserving fuel so the use of expensive resources is un-
neéessary. There is general agreement that breeder reactors are required
to attain this objective. Important factors related to conservation of
fissile fuel resources are the fuel doubiihg times of the breeder reac-
tors, the associated specific inventories of fuel, and the total nuclear-
electric generating capacity at the time when breeder plants begin-to
-COmpete cdmmercially and to be installed in large numbers. Also, the
mined fissile fuel needs are decreased if breederétype reactors can be
operated economiéally when initially fueled with 225U (initial operation
as fuel converters to produce plutonium or 233U). Such ability influences
the time at which reactor plants having good fuel utilization character-
istics can be introduced on a large scale. However, in order for 235y
to serve as the-initial fuel, the associdted Speéific inventory require-
ments and conversion ratio must be consistent with economic operation.

It is desirable that breeder reactors have both low fuel doubling
times and low specific inventories, since mined fissile fuel needs depend
on both factors. In general, it appears prudent.that the needs of the

nation for mined fissile material be below the quantity associated with

‘low-cost uranium reserves. Use of breeder reactors having specific in-

ventories of 1 kg fissile/MW(e) and fuel doubling times of 20 yearé ap-
pears to make this possible. Also, the capacity of existing gaseous dif-
fusion plants appears sufficient to provide the enriched-uranium require-
ments of the nation if such breeder reactors can be developed and-built

in large numbers by about 1985. Thus, a major objective of this study
was to determine whether a molten-salt reactor can achieve the performance
discussed above. Specifically, this goal is the simultaneous achieve-
ment of powér production costs of about 3 millé/kwhr(e) in a lOOO-Mw(e)
investor-owned station, a specific inventory of 1 kg fissile/Mw(e) or

less, and a fuel doubling time of 20 years or less.

1.3 BScope of Study

The molten-salt reactors being developed are fueled with solutions

of uranium and thorium fluorides dissolved in lithium and beryllium
 

actors.

fluorides. They operate at high temperature and relatively low pressure.
Fuels and materials are commercially. available for operating such systems
at temperatures at least as high as 1400°F, with pressures determined
primarily by fluid flow requirements. Since the salts do not undergo
violent chemical reactions with air or water, equipment and containment
design problems are minimized. Since the molten-salt fuels are compat-
ible with unclad graphite, a breeder core having low parasitic-neutron-
capture cross sections is practical. The combination of the high spe-
cific heat of the molten-salt fuels, their large operating temperature
range, and their radiation stability permits the attainment of wvery high
fuel specific powers. Also, fuel processing and reconstitution involve
inherently simple processes that allow inexpensive fuel recycle at high
processing rates in compact on-site integrated processing plants. In
this study these features were incorporated into a 1000-Mw(e) power plant
design, and the nuclear and economic characteristics of the plant were
evaluated as functions of design and operating conditions.

Only the Th-223U fuel cycle with fluoride salt fuels is considered
because the fuel-recycle processes employed apply uniquely to it (in
general, the chemical, physical, and nuclear characteristics of the
Th-2337y cycle favor its use over the uranium-plutonium cycle in thermal
molten-salt systems). Uranium can be recovered readily without affecting
the chemical form of the fercile material by fluorinating the molten
fluoride mixture. Also, the ThF, dissolved in the carrier salts does not
undergo oxidation-reduction reactions as does UF,; this reduces mass
transfer effects in systems constructed of Hastelloy N that circulate
salts with high fertile material concentrations. In addition, the nu-
clear properties of 33U that determine the fuel-utilization character-

istics are superior to those of 23°U or plutonium fuels in thermal re-

The initial reference moiten—salt breedef reaétor (MSBR) éonsidered
here is a two-region fluid-fuel concept with fiséilé material in.the core
stream andrfertile material in the blanket stream. The fuel and blanket
galts are in direct contact with the graphite moderator, and graphite

tubes are used to separate core and blanket streams. The fertile stream

 
 

 

 

 

not only surrounds the core to form a blanket region but also circulates
through the core region in spaces between the fuel tubes. Energy generated
in the reactor fluid is transferred to a secondary coolant-salt circuit, |
which couples the reactor to a supercritical steam plant. Fuel proceésing
is accomplished.in an on-site plant that utilizes fluoride-volatility

and vacuum-distillation processing. Although most of the design effort
centered on this system, it is not to be inferred that this concept is
necessarily the best or involves the best :processes. It’was chosen as

a logical starting point that would permit definitionrof a specific sys-
tem, help in determining design problems of molten-sait:reactors in gen-
eral, and provide a standard of performance against which the incentive

for design, development, and operating improvements could be measured.

In order to indicate the depth of experience presently available with
molten-salt reactors, Chapter 2 presents a summary of the technological
development and status. Following a description of the initial reactor
study (Chapt. 3), Chapter 4 presents alternate design conditions for the
reference design. Chapter 5 briefly presents alternate reactor designs
and their performance characteristics. Finally, Chapter 6 evaluates the

overall results of these design studies.

1.4 Study Organization and Participating Personnel

 

The areas investigated in the studies and the personnel involved are

given in Table 1.1,

1.5 Acknowledgements

 

The encouragement and support of Mr. R. Beecher Briggs, Director of
the Molten-Salt Reactor Program, during the course of these studies was
very much appreciated. In addition, we are grateful for his critical re-

view of this report and his helpful comments.
 

 

n

Table 1.1, Organization of Molten-Salt Reactor Design Studies

 

Report coordination
Engineering design studies

Mechanical design

Nuclear design and fuel-cycle cost
Processing plant design

Heat exchanger design

Pump design

Capital cost evaluation

Molten~salt chemical studies

Analytical chemical studies

Structural material studies

P. R. Kasten, W. B. McDonald,?
R. C. Robertson

E. S. Bettis

E. S. Bettis, J. H. Westsik,
R. C. Robertson, D. Scott,
W. Terry

H. F. Bauman, H. T, Kerr
W. L. Carter, C. D. Scott

C. E. Bettis, T. W. Pickel,
R. J. Braatz, G. A. Cristy,
A, E., Spaller

A, G. Grindell, L. V. Wilson

R. C. Olson, R. C. Robertson,
M. L. Myers

W. R. Grimes, H. F. McDuffie,

C. J. Barton, F. F. Blankenship,
3. Cantor, S. S. Kirslis,

J. H. Shaffer, R. E. Thoma

J. C. White

A. Taboada,P W. H. Cook,
C. R. Kennedy, G. M. Tolson

 

*Now with Pacific Northwest Laboratories, Hanford, Washington.

bNow with Division of Reactor Development and Technology, AEC,

Washington.

 
 

6

" 2. MOLTEN-SALT REACTOR TECHNOLOGY

-

TThe‘initialrtechndlogical develqpmenfi for molten-salt réactofs“fias
done in the early 1950's in the Aircraft Nuclear Propulsion (ANP) Program
at the Oak Ridge National Laboratory. This program involved extensive
fluoride-salt chemistry and materials compatibility studies, ¢omponent
development, material and fabrication development, and develqpment of
reactor maintenance methods,‘ In 1954 the Aircraft Reactor Experiment
(ARE), a 2.5-Mw(th) molten-salt reactor was built and operated success-
fully at outlet salt temperatures up to 1650°F. The ARE was fueled with
UF, dissolved in a mixture of zirconium and sodium fluorides, moderated
with beryllium oxide, and constructed of Inconel. |

‘The present molten-salt reactor program, initiated in 1957, has
drawn upon the information from the AfiP program and has also initiated
new investigations. By 1960.enough fnvorable'experimental results had
been obtained to support authorization of a 10-Mw(th) molten-salt reactor
experiment (MSRE). Power operation of the MSRE was initiated in early
1966. The system provides facilities for testing fuel salt, graphite,
and Hastelloy N (the container material) under appropriate reactor oper-
ating conditions. The basic reactor performence to date has been out-
standing and has demonstrated that the desirable features of the molten-
salt concept can be embodied in a practical reactor that can be constructed,
operated, and maintained with safety and reliability.

As indicated above, the successful operation of the MSRE is based
upon a broad technological development program. In order to give a better
understanding of present knowledge useful in the design of molten-salt
breeder reactors, a summary of selected work is given below that covers
chemical development, structural material development and corrosion
studies, fuel-processing development, and component development. Addi-
tional information is presented in other reports in this series that

amplifly the present discussion and give épecific resul‘os_.l"'6

2.1 Chemical Developm.entl

 

The chemical and physical characteristics of a large number of molten-

fluoride-salt compositions were studied extensively, with measurements
 

 

N

involvifig melting temperature, vapor pressure, heat capacity, enthalpy,
heat of fusion, thermal conductivity, and surface tension. These studies
showed that melts containing fissile and/or fertile material are available
which possess adequately low liquidus temperature, excellent phase sta-
bility, and good physical properties. Also, these salt mixtures appear
compatible with Hastelloy N and with graphite under irradiation as well
as nonirradiation conditions. The primary fluids proposed for the molten-
salt breeder reactor (MSBR) are a ternary mixture of “LiF, BeF,, and UF,
for the fuel salt, and a mixture of 7LiF, BeF,;, and ThF, for the blanket
salt. The choice of these compounds is based on their nuclear, chemical,
and physical properties, as discussed in Ref. 1. Briefly, fluoride car-
rier salts were chosen because of their chemical stability, their ability
to produce fuel solutions with relatively low melting temperature, low
neutron-capture cross section, low vapor pressure, and good heat transfer
properties. The fluoride fuel salts are also thermodynamically stable
with respect to the structural metal, Hastelloy N. Graphite was chosen

as a moderator because of good moderating ability, compatibility with
molten-salt fuels, low neutron-absorption cross section, and good struc-
tural properties.

There have been extensive investigations of the stability and com-
patibility of MSBR fuels and materials under irradiation conditions.
Capsule tests have been carried out with fission-power densities of 80 to
8000 kw/liter at temperatures from 1500 to 1600°F and for irradiation
times of 300 to 800 hr. Chemical, physical, and metallurgical tests have
indicated that no significant changes take place in the fuel or in the
structural material that can be attribufedato irradiation conditions.
Also fuel irradiation tests have been performed in graphite capsules con-
taining structural material, with initial fuel-power densities in the
range 200 to 1000 kw/liter and exposures of the order of 1000 hr. The
results indicate excellent radiation stability and compafiibility between
Hastélloer, graphite, and molten flubride fuels. ©Subsequent detailéd
tests at lower power densities substantiated these findings.

The very low solubility of the fission?product gases in molten-salt
fuel suggests that they can be readily removed from reactor systems; this

has been demonstrated in the ARE and MSRE operations. In addition,

 
 

 

experimental studies have shown that iodine, the precursor of xenon, can

" be removed directly from the fuel fluid by stripping with hydrogen fluo-

-~ ride gas.

- Although the physical chemistry of the fission products is not known
completely, thermodynamic considerations lead to the conclusion that the
fission process per se is oxidizing to Hastelloy N. The results of many
in-pile tests of metals and graphite in fuel salts suggest, however, that
fission does not lead to corrosion of the container material. Even if
the overall fission process is oxidizing, no real corrosion problem.need.
exist in an MSBR, since preferential oxidation of uranium would take place
if "burned" uranium were partially replaced with UF3 (rather than UFy).

Fuel and blanket salts of high purity are required to obtain the
very low corrosion rates observed in MSRE operation. The methods used
in purifying commercially available fluoride salts for the MSRE are di-
rectly applicable to the large-scale production operations required to
supply the salts for MSBR systems.

Continuous monitoring of the salt composition is highly ‘desirable
“and advantageous in operating a fluid-fuel reactor, although not essen-
tial. Current methods give accurate measurements of the composition and -
purity ol the reactor salts on a routine basis, but not as rapidly as
desirable for an MSBR. Thus, investigations are being performed to de- -
velop appropriate instrumentation and new analysis techniques. Results
indicate that new composition-analysis methods can be developed for "on-

line" reactor use.

2.2 Structural Material -Develgpm_ent2

The structural material for containing the molten flubride salts
must have desirable structural properties, be easily fabricated, and be
metallurgically stable over a wide temperature range. A mostrimportant
requirement is that of adequate resistance to corrosion at elevated tem-
peratures under reactor conditions. Since molten fluoride salts‘are ex-
cellent fluxing agents, surface films cannot be relied upon-as protective
membranes. Therefore; the structural material must be basically inert

to corrosion processes under conditions of thermodynamic equilibrium.
 

 

 

it

Extensive corrosion studies were conducted in which various structural
materials were exposed to the salt in both thermal-convection and forced-
circulation loops with hot-leg temperatures of about 1500°F. These studies
led to the development of INOR-8,* a nickel-base alloy containing about
16% molybdenum, 7% chromium, and 5% iron. This alloy has good to excel-
lent mechanical and thermal characteristics that are superior to those

2>7 Tt has good resistance to oxi-

of many austenitic stainless steels.
dation by air, and it retains favorable mechanical properties at tempera-
tures up to about 1500°F. Results of long-term corrosion experiments
(exposures of up to 20,000 hr) have demonstrated its basic inertness to
molten fluoride salts at temperatures up to about 1500°F. Corrosion rates
appear to be controlled primarily by impurity levels in the molten salts
and by the temperature-dependent mass transfer associated with the reac-
tion

2UF, + Cr = 2UF3 + CrFp .

Based on experimental data from test loops, the corrosion rate of Hastel-
loy N in MSBR fuel systems will be less than 0.5 mil/yr with a core outlet
temperature of 1300°F, and probably will not exceed that with a 1500°F
outlet temperature under equilibrium conditions. Even less corrosion
should occur in the blanket-salt and secondary-coolant-salt systems,
where the UF, concentration will be extremely low and zero, respectively.
These test loop results have been substantiated by data obtained in the
MSRE, where no significant corrosion of the Hastelloy N has taken place
in 2500 hr of exposure at 1200°F (on the average, chromium was removed
from a 1ayer 0.006 mil in thickness over loop surfaces, with v1rtually
zero corrosion after the initial months of operation).

Extensive tests of the mechanlcal and physical properties of Hastel-
loy N as a function of temperature up to about 1800°F indicate charac-
teristics suitable for MSBR use. The creep and stress~rupture properties

are equivalent to and in most cases superior to those of Inconel. Iong-

time ageing studies have shown that the material does not embrittle with

 

*This alloy is commercially available as Hastelloy N or INCO-806;
throughout this report, the designation Hastelloy N is employed.

 
 

 

 

10

time. Further, the mechanical properties of Hastelloy N are virtually
unaffected by long-time exposure to the molten fluoride salts.

The structural material must retain its good mechanical properties
when exposed to reactor radiation. Irradiation studies have shown that
the (n,o) reaction in structural materials tends to decrease ductility.
This reaction and its effects on Hastelloy N have been studied in detail,
and it appears that the deleterious effects can be minimized by maintain-
ing a low *°B content, adjusting the concentration of minor constituents
in the alloy, and improving heat-treatment practices. Development work
in these areas appears dapable of producing an improved Hastelloy N whose
ductility will not decrease below aéceptable values during long-term ex-
posures to MSBR fluxes.

The melting and casting of Hastelloy N can be carried out with the
conventional practices for nickel and its alloys. Conventional methods
of hot and cold forming have been used to produce it on a commercial basis
in a variety of shapes, such as plate, sheet, rod, wire, and as-welded
and seamless tubing. Cold working operations can be performed, such as
rolling, swageing, tube reducing, and drawing. Cold forming has been
successfully used for fabricating Hastelloy vessel heads. The material
is readily weldable by the inert-gas-shielded tungsten-arc process.

In addition to Hastelloy N, the other prime structural material
used in the MSBR is graphite. This material does not react chemically
with the molten fluoride mixtures under consideration, and since it is
not wetted by molten-salt mixtures, there is little salt permeation of
the graphite. 1In general, the graphite needs to have low permeability
to salt and gases, to have adequate structural properties when.exposed
to high radiation fluxes, and to be fabricated into tubes and other mod-
erator shapes. These properties were obtained, at least partialiy, in
the MSRE graphite, which was produced by extruding petroleum coke bonded
with coal-tar pitch and applying multiimpregnations and heat treatments.
The resulting product has a high specific gravity (1.86), low permeation
(0.2% bulk volume penetration by molten salt — surface penetrations .
only — when a 150-psi pressure was applied to the salt), and high strength
(ability to withstand 1500-psi tensile strain and 3000-psi flexural strain

was shown by all bars fabricated). This material represents a successful

o
 

11

first step in developing a graphite acceptable for MSBR use. Graphite
tubing having l/2—in.—thick walls has also been successfully fabricated:
the product had no visible cracks.

The graphite in regions of high flux in an MSBR will be irradiated
to doses above 10°? neutrons/cm® in five years and will be exposed to
radiation flux gradiefits. The magnitude of the graphite differential
shrinkage that will occur under these conditions will depend on the
graphite creep coefficient, flux gradient, and geometry of the particular
structural component. Isotropic graphite has demonstrated the ability to
withstand high radiation exposures. Also, the ability of the graphite to
absorb the creep strain regardless of the stress intensity has been shown
experimentally. Thus it appears that graphite satisfactory for MSBR use
can be developed.

Techniques are required for attaching graphite to metal with reliable
joints. Graphite has been brazed successfully to metals, with brazing
alloys that were found resistant fo corrosion by molten salts. Alloys
of gold, nickel, and molybdenum and other alloys under development
appear to be satisfactory brazing materials. Brazes made with these
materials can be used for Jjoining graphite to graphite or graphite to
molybdenum (molybdenum has a thérmal expansion coefficient near that of
graphite). Metal-to-graphite joints have maintained their integrity in
molten-salt enviromments at 1300°F and at pressures of 150 psi for periods
of 500 hr. In addition, mechanical joints may be useful in MSBR cores,
gince zero leakage between the core and blanket fluids is not required.

Finally, compatibility of molten salts, Hastelloy N, and graphite
appears excellent. Tests have shown no carburization of Hastelloy N

under MSBR conditions.

2.3 TFuel-Processing Development?

Experience in processing molten-fluoride-salt fuels at Oak Ridge
National ILaboratory dates from 1954 and began with fluoride volatility
processing studies. The initial laboratory and development work formed
the basis for successful operation of a pilot plant. The associated

process is designated the Fluoride Volatility Process after the principal

 
 

 

 

 

 

 

12

operation of volatilizing uranium as the hexafluoride. Although also
applicable to the treatment of solid fuel elements, fluoride volatility
processing is uniquely suited to molten-salt fuels because the fuel salt
can be treated directly with fluorine. Elemental fluorine reacts with
the UF, in the molten salt (at about 930 to 1020°F) to produce volatile
UFg. The reaction is rapid and essentially quantitative for uranium;

it easily reduces the uranium content of the molten salt to a few parts
per million. The UFg product can be treated in absorber beds to give
decontamination factors of 10° ahd more. Recycle uranium is easily con-
verted to UF, dissolved in carrier salt by absorbing the UF6 in molten
salt containing some UF,; and hydrogenating in the liquid phase. This
treatment also reduces any corrosion product contaminants to metal that
can then be filtered from the fuel solution prior to returning fuel fluid
- to the reactor system.

The fluoride volatility process can be used for both the core stream
and the blanket stream. When applied to the core stream it is used to
separate the uranium from the carrier éalt before that stream is pro-
cessed (by another method) for fission-product removal. Essentially all
the uranium must be recovered, and this leads to relatively severe fluori-
nation conditions. Requirements for processing the blanket stream are
less stringént. Uranium that is not removed during the fluorination is
merely returned to the reactor blanket and is removed during subsequent
passes through the processing plant. Discard of 3% annually or process-
ing by other methods keeps the fission products at a very low level in
the blanket salt.

The ease of removal of xenon gas from molten-salt fuels has been
demonstrated in both the ARE and the MSRE. It thus appears practical to
obtain very low xenon poisoning by sparging the salt with an inert gas

1351, the precursor of 13°Xe,

such as helium or nitrogen. In addition,
can be stripped from fuel salts by sparging with HF and hydrogen. Such
processing would virtually eliminate xenon poisoning in MSBR systems.
The discovery that vacuum distillation permits the economic separa-
tion of carrier salts from fission products has been a vital factor in
improving the economic and nuclear characteristics of MSBR systems.

Laboratory experiments have demonstrated that carrier salt can be readily
 

 

 

13

separated from rare-earth fluorides at distillation pressures of 2 mm Hg,
with separation factors of 50 to 100 and 95% recovery of carrier salt.
These process characteristiés appear adequate for MSBR application.
Fluoride volatility processing appears well suited for keeping the
uranium inventory and the fission rate in the blanket low and thereby
maintaining low neutron leakage from the blanket. An even better process
would be one for recovering protactinium directly from the blanket fluid.
Recent work toward providing such a process has been encouraging; at
least two possible methods are being considered. One involves removal
of protactinium from the process stream by precipitation as the oxide
through reaction with Zr0O;. After the protactinium decays, the product
U0» can be recovered by reaction with ZrF, to give UF,; in solution.
Even more encouraging results have been obtained by treating fluoride
gsalts containing PaF,; with thorium dissolved in molten bismuth. The
thorium metal reduced the protactinium to the metal which subsequently
deposited on a stainless-steel-wool filter. These results indicate that
inexpensive methods can be developed for removing protactinium directly

from the blanket stream of an MSBR.

2.4 Component Development4? ?

Nearly all molten-salt component development work has been for ex-
perimental molten-salt reactors (the ARE, the planned Aircraft Reactor
Test, and the MSRE). The components required for these systems were de-
veloped at ORNL, including pumps, seals, valves, heat exchangers, fuel
sampler-enricher units, freeze flanges, remote-maintenance tools, heaters,
and instrumentation forrmeasuring pressure, fluid flow, liquid level,
and temperature under molten-salt reactor conditions. A major effort
has been devoted to developing pumps that have long-term reliability at
temperatures of about 1300°F. These pumps are vertical-shaft sump-type
centrifugal pumps with a free surfaée in the pump bowl; all‘parts wetted
by molten salt are constructed of Hastelloy N. Various pump models with
capacities up to 1500 gpm have been manufactured and tested, and present
models have circulated molten salt continuously for more than 25,000 hr

at temperatures above 1200°F without maintenance. Stopping and starting

 
 

 

14

of pumps does not appear to produce any corrosive attack; thermal and
pressure stresses associated with thermal cycling and reactor operations
do not appear excessive. For MSBR application, it appears feasible to
use a vertical sump-type pump similar to present models, with the upper
end of the pump shaft supported by oil-lubricated radial and thrust
bearings and the lower end supported by a molten~salt-lubricated journal
bearing. The present experience with molten-salt-lubricated bearings
consists of 3900 hr of operation in development of the bearing and
operation for 13,500 hr of a pump containing a salt-lubricated bearing
at temperatures of 1000 to 1400°F. The results obtained indicate that
the development of salt-lubricated bearings is feasible; testing of these
bearings is continuing. | .
Molten-salt heat exchangers have been designed and constructed and
successfully demonstrated in the ARE and the MSRE. Numerous heat ex-

changer designs have been tested, and the results show that the required

performance capability and mechanical integrity can be obtained with
straightforward design and fabrication methods. The use of Hastelloy N
as the construction material introduced no major difficulties. Experi-
ments and experience with the MSRE have shown that conventional heat-
transfer-coefficient correlations with minor modification are applicable
to molten-salt heat exchanger design; also the physical properties of
molten fluorides make them good to excellent heat transfer media. Since
the molten salts are good fluxing agents and keep all surfaces clean,
scale formation does not occur on heat transfer surfaces.

An important feature of molten-salt reactors is the ease of adding

or removing fuel fluid from the reactor system. This permits ready com-

pensation for fuel burnup, and the fluid removed can be easily transported

to processing areas. The successful operation of the MSRE sampler-
enricher system indicates that adjustments in fuel concentrations can
be accomplished readily and reliably with relatively small and simple
equipment. |

The high melting point of MSBR fluoride salts provides a means of
sealing a system, without the need for mechanical valves, through use of
"freeze" valves in which a frozen plug of salt prevents leakage from the

system. Although slow acting, the performance of freeze valves in the

&
 

 

15

MSRE has been excellent. It appears that such valves will be useful in
MSBR subsystems. Freeze flanges have also been developed because of
their proven reliability in containing fluid salts under all anticipated
thermal-cycling conditions. Such flanges appear appropriate for Joining
components and piping in MSBR subsystems.

Instrument development carried out for the MSRE also appears useful
for MSBR systems. Liquid-level measuring devices have operated success-
fully, as have instruments for fluid flow, differential pressure, and
temperature measurements. Development work has also been performed on
control-rod drive units capable of operating reliably for long periods
while located in a strong gamma field.

Since the inception of molten-salt reactors, there has been signifi-

®  Remotely

cant engineering development work on maintenance operations.
operated tools and procedures for remote maintenance have been devised,
and the required operations have been studied in a maintenance facility.
The results of these studies, along with other experience, were used in
developing the MSRE maintenance tools and procedures. Also, equipment
for remotely cutting pipes and brazing them back together was developed
for replacement of MSRE components, and the results obtained with this
equipment indicate that a remotely operated cutter and welder for MSBR

maintenance operations is feasible. Experience to date with maintenance

. of radiocactive molten-salt systems 1s encouraging.

 
 

 

 

16

3. INITIAL DESIGN OF A 1000-Mw(e) MSBR POWER STATION

The MSBR design discussed here is for a 1000-Mw(e) power station
that appears technically sound, maintainable, and attractive from the
power cost, reliability, and fuel utilization standpoints. This refer-
ence design is not necessarily the best design for a molten-salt reactor,
but it represents a logical starting point based on the information
available at the time of this study. The report is intended to illus-
trate the general merits of molten-salt reactors for power applications,
delineate design problems and possible solutions to them, and indicate
areas where research and development programs could improve MSBR per-
formance.

A complete power station is considered, including all major equip-
ment and a fuel-processing facility that is integral with the reactor
plant. Very little optimization work was done, and layouts and designs
were detailed only to the extent necessary to establish feasibility and
to permit preliminary estimates of construction and operating costs.

The design is based only on those materials and techniques that appear
feasible based on present-day technology. In addition, several alter-
native molten-salt reactor designs were examined briefly (see Chapt. 5)
in order to show the influence of design concept and technology require-

ment on the performance characteristics of molten-salt systems.

3.1 General Design Criteria, Cost Bases, and Ground Rules

The following design criteria, costs bases, and ground rules were
used in making the study:

1. The power station will have a net electrical output of 1000

Mw(e) and will be used solely for the production of power.

2. The reactor will be a two-region two-fluid graphite-moderated
and -reflected thermal breeder with graphite separating the fissile and.
fertile materials. The reactor will be designed to achieve low power
cost, high specific power, and low fuel doubling time.
3. Equilibrium fueling conditions will apply, with mixtures of
BeF, and "IiF used as carrier salts for 233U and ThF,. O
 

 

17

4., Because of the present uncertainties concerning long-term ex-
posure of graphite in a high neutron flux, the MSBR core size will be
relatively large in order to reduce the graphite irradiation rate. The
fuel cell dimensions will be small to reduce flux gradients in the
graphite. The fuel velocity in the core will be limited to 15 fps.

The graphite tubes will be attached to a fixed Structure at one end only
to give freedom of movement for shrinkage and thermal expansion. Pro-
visions will be made for removal and replacement of the core by remote-
maintenance procedures.

5. A control rod will be incorporated in the design, primarily as
a convenience feature.

6. The reactor core will be arranged so that the fluid will drain
by gravity to make the reactor subcritical in event of loss of electric
power or other scram-initiating disturbance.

7. The reactor vessel, pumps, heat exchangers, and drain tanks for
the fuel- and blanket-salt systems will be housed in a heavily shielded
structure. This structure, and the more lightly shielded structure
housing all portions of the system containing the coolant salt, such as
the boiler-superheaters and reheaters, will be housed in a shielded con-
tainment vessel that meets acceptable leak-rate standards for this ser-
vice. This containment vessel will incorporate a pressure-suppression
system. The reactor containment vessel, but not the turbine room, will
be located in a confinement-type building with controlled air-cleaning
and venting systems.

8. Heat will be transported from the primary heat exchangers to
the steam-power system by a circulating secondary coolant that must be
compatible with the fuel- and blanket-salt systems in case of accidental
mixing. This coolant must have suitably low vapor pressure and liquidus
temperature.

9. The salt pumps will be limited in size to about 15,000 gpm;
that is, they will be about an order of magnitude larger than the fuel-
salt pump used in the MSRE. 8

10. The reactor system will incorporate an off-gas system for con-

tinuous removal, retention, and disposal of the fission-product gases.

 
 

 

 

18

11. The fuel and blanket salts will be continuously processed in
a processing facility that is an integral part of the reactor plant.
In the initial design, the fluoride-volatility—vacuum~distillation pro-
cesses will be used for the fuel salt, and the fluoride-volatility pro-
cess will be used for the blanket salt. A system will be provided for
cleanup of the coolant salt.

12. An afterheat removal system will be included in the design.

13. The core outlet temperature of the fuel salt will be 1300°F.
The temperature of the coolant salt entering the primary heat exchangers
will be above the liquidus temperatures of the fuel and blanket salts.
The feedwater entering the boiler will be above the liguidus temperature
of the coolant salt. The temperature of the steam entering the reheaters

will not be more than 50°F below the liquidus temperature of the coolant

- salt.

14. The cells in which the fuel and blanket salts will circulate
will be maintained above the liquidus temperature of both salts (ébout
1040°F). The cells in which only coolant salt is circulated will be
operated above the liquidus temperature of the coolant (about 700°F).

The cell temperatures will be maintained by radiant heating surfaces.
Thermal insulation and water cooling will be applied as required to pro-
tect concrete, equipment supports, instrumentation, and other items.

15. The boiler will operate with supercritical-pressure steam in
a once-through counterflow arrangement.

16. The steam~power cycle will operate with 3500-psia 1000°F steam
to the turbine throttle, with single reheat to 1000°F.

17. A1l salt-containing portions of the system will be constructéd
of Hastelloy N. The allowable design stress will be 3500 psi at 1300°F,
6000 psi at 1200°F, etc., in accordance with the MSRE design literature®
and Ref. 2.

18. All portions of the system will conform to the applicable por-
tions of the ASME Codes. Specifically, points of suspected high stresses
will be examined for pfacticality of the proposed concepts.

19. All major equipment for the plant will be included in the study

up to, but not including, the station high-voltage output transformer
 

 

 

 

19

and the switchyard. Iand and site development costs will be the same as
those used in the advanced-converter reactor studies.??*0
20. Both capital and power production costs will, where applicable,

11 In-

be estimated and presented in accordance with the AEC cost guide.
direct and operating costs will be estimated on the same hases as those
used in the advanced-converter reactor studies.? The plant life will be
30 years. Power costs will be estimated on the basis of both private
financing (12% fixed charges) and public financing (7% fixed charges),
with private financing as the base case. A plant factor of 80% will be
assumed for both cases. In estimating all costs, it will be assumed
that equipment and materials are obtained from a large and established
molten~-salt reactor industry.

21. The reactor-plant financing rate will apply to the fuel-pro-
cessing and -fabrication plant, which will be a part of the power plant.
To account for a higher equipment replacement rate, the indirect costs
for the fuel-recycle plant will be 100% of the direct costs.

22. Inventory charges on fissile, fertile, and carrier-salt inven-
tories will be computed with a reference value of 10% per year for the
base case and with 5% per year to represent public ownership.

23. The value of core and blanket fluids will be based on the
following: 223U and ?33Pa at $14/g, 22°U at $12/g, Th at $12/kg, and
carrier salt at $26/kg.

24. Losses of materials through fuel recycle will be based on
uranium losses of 0.1% per pass, thorium and blanket-carrier-salt dis-

card on a 30~year cycle time, and core-carrier-salt losses plus discard

of 6.5% per fuel-cycle pass.

3.2 General Plant ILayout

 

The MSBR site is that described in the AEC handbook for estimating
costs!? and also used in the advanced-converter reactor studies.? In
brief, the site is a 1200-acre plot of grass-covered level terrain ad-
jacent to a river having adequate flow for cooling-water requirements.
The ground elevation is 20 ft above the high-water mark and is 40 ft

above the low-water level. A limestone foundation exists about 8 ft

 
 

20

below grade. The location is also satisfactory with respect to distance
from population centers, meteorological conditions, frequency and in-
tensity of earthquakes, and other external conditions.

As shown in Fig. 3.1, the plant area proper is a 20-acre fenced-in
area above the high-water contour on the bank of the stream. The usual
cooling-watér intake and discharge structures are provided, along with
fuel-oil storage for a startup boiler, a water-purification plant, water-
storage tanks, and a deep well. This plant area also includes radiocactive
waste-gas storage, treatment, and disposal systems. Space is provided
for the output transformers and switchyard. A railroad spur serves for
the transportation of heavy equipment, and parking lots are provided.

A large single building houses the reactor and turbine plants,
offices, shops, and all other supporting facilities. This building, as
shown in Figs. 3.2 through 3.5, is 250 £t wide and 528 ft long; it rises
98 ft above and 48 ft below grade level. The construction is of the
typical steel-frame type, with steel roof trusses, precast concrete roof
slabs, concrete floors with steel gratings as required, and insulated
aluminum or steel panel walls. The wall joints are caulked or otherwise
sealed on the reactor end of the building.

The reactor complex occupies less volume than the steam-generating
equipment in a conventional plant, and the turbine floor dimensions are
the same as those used in the Bull Run Steam Plant of the Tennessee
Valley Authority (TVA), but there are slightly larger allowances for the
shops, offices, control rooms, and other facilities of the reactor plant.

The reactor end of the building is 168 ft long and consists of a
high-bay portion above a reinforced-concrete reactor containment struc-
ture. A single crane is pictured as serving both the turbine room and
the reactor plant, but separate cranes would probably be required, and
the cost estimate allows for two units. The reactor plant building is
sealed sufficiently for it to serve as a confinement volume in the un-
likely event of a radioactivity incident, and it is provided with posi-
tive ventilation, air filtration and dilution equipment, and an off-gas
stack. | |

The arrangement of the reactor plant cells is shown in Fig. 3.6.

The thicknesses of concrete required for shielding against reactor
 

 

ORNL DWG 66-7133
RIVER FLOW

  
    
 

DISCHARGE

   
   
  
    
     
   
 
 
 
 
 
  
 
 
   

\

ELEV. 0' NORMAL RIVER LEVEL

  

INTAKE
20"

60

80’

  
   
  

FUEL OIL STORAGE
PUMP

WASTE STORAGE POND STATION

O

 

WASTE GAS CHARCOAL ADSORBER

TER PURIFICATION
WASTE GAS TREATMENT & STACK

TER STORAGE
TE STORAGE
TRANSFORMERS
PARKING

REACTOR PARKING

Fig. 3.1.

MSBR Plant Site Layout.

e

 
 

 

 

22

ORNL DWG 66-7138

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

528"
D ' c
ROLL-UP —_| j _Tll 1
DOOR
HOT SHOP ELECTRICAL&EQUIPMENT 1
8 35
STORAGE BUILDING VENTILATION SYSTEM
i__ ——————— o] TURLB‘I)NE g
5 , | L e d
| |
‘ | REACTOR | 130'
i
| | weate | oo
‘ ]' TURBINE
OFFICES, FEEDWATER SYSTEM 30
TREATMENT ROOM,
4 LUNCH ROOMS,
& ETC
i SHOPS '
3 FLOORS HIGH ' \ . 55
68 x 85' 336 x 55 x 30 HIGH
od cd
Fig. 3.2. MSBER Building — Plan AA.

   

Fig. 3.3. MSBR Building — Elevation BB.

ORNL DWG 66-7137

 

:52' ABOVE NORMAL

RIVER LEVEL

3

LYl
'
 

 

 

23

ORNL DWG 66-7140

r —250 —
‘ 55' T3O'T'H - -130' 35';1

| 198’

 
  
     
    

2-i50 TON
OVERHEAD
CRANES

CRANE BAY

TUNNEL

Fig. 3.4. MSBR Building - Elevation CC.

ORNL DWG 66-7130

CRANE BAY

Fig. 3.5. MSBR Building — Elevation DD.

radiations were estimated on the basis of previous reactor design experi-
ence. A minimum of 8 ft of high-density concrete separates the reactor
vessel from an occupied area. A minimm of 4 ft of concrete is used
around equipment containing the coolant salt, which is at a relatively
low level of activity durifig reactor operation and this level decreases
a short time after power shutdown.

As illustrated in Fig. 3.6, the reactor vessel is housed in a cir-
cular cell of reinforced concrete. This cell is about 36 ft in diameter
and 42 ft high. The four fuel- and blanket-salt primary heat exchangers

and their respective circulating pumps are placed around the reactor.

 
 

24

ORNL DWG 66-7111

REHEAT STEAM

    
 

   
  
  

  
       
    
  

H.P. STEAM “WASTE GA mh FEEDWATER I
FEEDWATER CELL nn H.P. STEAM
, o LP. STEAM - 28
COOLANT SALT - FUEL M
PUMPS HEAT EXCH. i

 

 

DECONTAMINATION:

16 SUPERHEATERS
AND STORAGE

BLANKET
HEAT EXCH.

 

 

 

 

“.CONTROL AREA—~

 

 

ORNL DWG 66~7110

CONTROL ROD DRIVE

  
  
 

    
 
    
     

COOLANT SALT

FUEL CIRCULATING—
PUMPS—\\\L PUMP

e

BLANKET CIRCULATING
PUMP

SUPERHEATERS

GROUND LEVEL

FUEL HEAT EXCH.

 
 

  

KET

, REHEATERS
HEAT EXCH.

3.6. Reactor and Coolant-Salt Cells — Plan and Elevation.

 
 

 

 

25

The wall separating the reactor cell from the adjoining cells is 4 ft
thick, and the removable bolted down roof plugs total 8 ft in thickness.
The pump drive shafts pass through stepped openings in the special con-
crete roof plugs to the drive motors, which are located in sealed tanks
pressurized above the reactor cell pressure. The special roof plugs are
removable to permit withdrawal of the pump impeller assemblies for mainte-
nance or replacement. The control-rod drive mechanism passes through the
top shielding in a similar manner. The coolant-salt pipes passing through
the cell wall have bellows seals at the penetrations.

The cells are lined with 0.25- to 0.5-in.-thick steel plate haVing
welded joints, which, together with the seal pan that forms a part of the
roof structure, provide a cell leak rate that meets the requirement of
less than one volume percent per 24 hr. The reactor cell is heated to
about 1050°F by radiant heating surfaces located at the bottom. The heat
is supplied either electrically or from gas-fired equipment. The liner
plate and the concrete structure are protected from the high temperature
by 6 in. or more of thermal insulation and cooled by either a circulating-
gas or water-coil cooling system. The reactor and heat exchanger support
structures are also cooled as required.

The four circuits that circulate cooling salt are housed in indi-
vidual compartments, or cells, having 4-ft-thick reinforced concrete walls
and bolted down removable roof plugs. Each compartment contains four
boiler~superheaters, two reheaters, one coolant-salt pump that serves the
boiler-superheaters, and one coolant-salt pump that supplies the reheaters.
A1l pipes that pass into these cells from the turbine plant have sealed
penetrations and valves outside the walls. The pump drive shafts extend
through the roof rlugs, and the cells are sealed and heafied in the same
manner as the reactor cell. The temperature is only maintained above
750°F, however.

The design pressure for the reactor cell and the four adjoining com-
partments is aséumed to be about 45 psig. Pressure-suppression systems
are provided, with the reactor cell system.beifig separate from the sys-
tems for the other compariments. These systems consist of water-storage
tanks through which vapor released into a cell would pass and be condensed

to maintain the cell pressure below the design value.

 
 

26

As indicated, the reactor plant structures have not been optimized
nor have they been studied in any detail. Likewise, the cell heating
and cooling systems, the pressure-suppression systems, and the building
ventilation, filtration, and air-disposal systems have received no de-
| tailed study. However, the allowances made in the cost estimates for
| these items should not require adjustments large enough to affect the
overall conclusions drawn from this study.

The turbine plant is standard in the utilities industry and needs
little description. Space has been allowed for offices, control rooms,

shops, storage, change and locker rooms, and other facilities.

3.3 Flowsheets and General Description

The general flow arrangements and operating conditions of the MSBR

 

power station at rated output are summarized in the flowsheet presénted
in Fig. 3.7. The 2225-Mw(th) reactor consists of a vessel about 15 ft in
diameter and 19 £t high that contains a 10-ft-diam core made up of 534
graphite fuel tubes, which are fastened to two plenum chanmbers at the
bottom of the reactor vessel. As shown in Fig. 3.8, fuel salt is pumped
into one plenum, flows upward through eight 0.53-in.-diam passages in
each graphite tube to the top of the reactor core, and turns downward to
flow through the central 1.5-in.-diam passage to the other plenum at the
bottom of the vessel. The graphite tube construction is indicated in
Fig. 3.9. A matrix of hexagonal graphite blocks surrounds the fuel tubes
§ and serves as moderator. A 1.5-ft-thick annular'space filled with the
fertile, or blanket, salt surrounds the core. Outside the blanket volume

 

is a 3-in. thickness of graphite that acts as a reflector. A 1.5-in.-
wide space separates the graphite reflector from the wall of the re-
actor vessel; the vessel wall is 1.5 in. thick and is-constructed of
Hastelloy N. .

The fuel salt is pumped into the reactor plenum at lOOO°F and about
144 psig at a rate of about 95.7 cfs (43,000 gpm).' It flows upward
through the fuel tubes and then downward through the central passage, as
described above, at an average velocity of about 15 fps. During its
passage through the core, the fuel salt is heated to about 1300°F by
 

 

 

   

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

   
  
  

 

CRNL DWG 66-7175.

 

 

 

 

 

 

  

 

 

 

 

 

REACTOR VESSEL 5.134 # -
5500~ 1000°F - 10.067# E _ /7.I52#4_ i|5l8.5h-540p-I000'
p-looo® f ) ”' - '
5183, . i 11424 h-3615-1000° i
BLANKET SALY FUEL SALT | — 3 o - |
F | o1 [ GEN
PUMPS PUMPS . 3600p-1000°F | bogy 1 | HP |
' . | | IISOlt'Isec 201 Y i'm' | g i Luraing [1'P TURBINE S 52-2"20”““
173 ft¥sec | 95.7 #Ysec. I , BOILER | P 600p- : .
. (150 J; _1000°F | | REHEATERS SUPERHEATERS 1 2sean | 4 B
T, 1250°F T300°F j850°F | | j8sor - r r
A BLANKET SALT .—’ COOLANT SALT COOLANT SALY B GEN. |
4| HEAT EXCH. iF PUNPS PUMPS 507.7 Mwe
I Yy 250 FUEL SALT.; K B ,wfj-,s;c ; _ . Gross
- . HEAT EXCH. L)} Tiae BT | | P |
| ‘ ¥ o T PREHEATERS CONDENSER 8 FEEDWATER
I 1 ] ) . : !
i I ' | b _J 38009-7100'F gy | DOTEH SYSTEMS
. i i ! . - 3500p-866"F ) i .
JL ' WZ5°F "'"‘1’""—" T sk 4 L 1 T [P0
[% | u) i:l ~ - 570.-650.'F§ -—Q WISp-695F 1
—~ - - ' 766.4h
B00STER MIXING TEE
| | . . PumPps
BLANKET SALT FUEL SALT - { COOLANT SALT \ | FERFORMANCE
DRAIN TANKS / ~ DRAIN_TANKS /- _DRAIN_TANKS | NET OUTPUT LO0O' Mue
LEGEND P GROSS GENERATION 1,034.9 Mwe
FUEL , - BF BOOSTER PUMPS 9.2 Mwe
BLANKET e m et _ STATION AUXILIARIES  25.7 Mwe
COOLANT ———— REACTOR HEAT INPUT 2225 Mwt
STEAM ————— NET HEAT RATE 7,601 Btu/kwh
ho NET EFFICIENCY 449 %
2
ok 10° 1b/he
[ I psio
Moo Btw /1D,

—-______Freeze Volve

 

 

  
  
    

 

 

Fig. 3.7. MSER General Flowsheet.

      

    

 

 

 
  

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

L

 
 

 

 

     

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

COOLANT—
SALT TO

SUPER
HEATERS

COOLANT SALT
RETURN

 

 

 

 

 

 

 

e
. . . :
S PRIMARY
HEAT EXCH.

 

CONTROL. f;
ROD
DRIVE

£

 

 

 

 

 

 

i A
N

“ ‘:!.. -
36 FT. DIA,
REACTOR CELL

 

 

  
     
 
 
 
 
 

 

   
  
   

HEAT EXCH.
SUPPORTS

 

Fig. 3.8. Reactor

+ ¢j'-*}. i '::: G w;,fisi@&f?««’ R

_ PRIMARY |
, HEAT EXCH.
|

!

 

 

t
v
1

'i

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

    

o

 

  
  

 

M —REAC

 

      

 
   

TR
PEDESTAL _

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

28
ORNL DWG 66=-6968
FUEL PUMP BLANKET PUMP
MOTOR CONTROL ROD MOTOR
DRIVE
;‘ ;.‘.’. 4 "" . ._“\-, .‘:“ ' "‘._-' ‘.. ? '. . Vl. ~a: . N : . .". . . \.' u '.'“" ‘ :" .‘ s 0 " /\ y "‘.." . i' ' ‘ -fi
consTANT | e R R G BN S
SUPPORT S - ’ —f
HANGERS | o | ;
] e O
;‘,‘ ,‘; Sl
L
ORNL DWG 66-6967 L
» e ? — —— b te=mo LD W g i e "
U ‘ ’ * , . , * 5 ‘ ; * <\ '
o - ; ‘ . Sy | » . B T |~ BLANKET
S 4 lff . ! : HEAT EXCH
o, ; 1 ' ¢ 4":.2‘, :
o \ I N | '
BLANKET R N Y FUEL DUMP Ho
HEAT g e | } i I
EXCH. , : ¥ ; ; 7/ STEAM Cogtfnng N 'i :
‘ . ‘ N / . GENERATOR w -
§ ’ e CELL COILS FOR ~
T ’ D TR @ AFTER HEAT ™0 F7. D
— BLANKE TS ~ : REMOVAL. — i | CORE
‘ REGION J \ ‘ . 0
. —CORE A @ 420 14+ T—REACTOR
= REGION (1 . VESSEL

4.1
[

- T~—FUEL SALT
| DISTRIBUTION
PLENUMS

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

REACTOR CELL

Cell — Ej;evation and Plan.
@

 

 

 

- L
& ;"‘ P Vi 4
. fos
o . AR
. p a
;Ej‘* ; »
Y 4 <
K. 3 ' “
g §
wal. %
o | .
* Q‘ - .
ey
e »
]
Y 24
+ A v s
s ¥ s
. T e LTIt |
T e T T e o
4 1 - T AL .
= P A IO E M G ST
VRTINS T e .
Ta " Ta At e a
. . & «
. . . « .
e - . * b . . T e s .
S RS L e N + - . . 1
ki g —tecnp
29

ORNL DWG 66-7139

   
  
    
 

MODERATOR GRAPHITE) —BLANKET PASSAGE

FUEL PASSAGE {UP) FUEL PASSAGE (DOWN)

 

 

 

 

 

 

3% OD. FUEL TUBE :
1 :
MODERATOR HOLD
=48 PIT
8 PTCH DOWN NUT GRAPHITE)
{
N |

 

REACTOR

 

 

 

 

 

 

 

 

 

 

 

—SPACER
AL TO GRAPHITE
SLIP-JOINT

METAL TO GRAPHITE
BRAZED JOINT

BRAZED JOINT

~——FUEL INLET
PLENUM

~———FUEL OUTLET
PLENUM

Fig. 3.9. Graphite Core Tubes.

 
 

 

 

30

nuclear fission. About 95% of the heat generated, or 2114 Mw(th), is
removed by the fuel salt.
Concentric pipes connect the core plenum chambers to the heat ex-

changers. The 1300°F fuel salt leaves the lower plenum of the reactor

vessel and flows downward through the 18-in.-diam inner pipes to the top

of the heat exchangers, where the pressure is about 96 psig.

The fuel salt is circulated in four loops that operate in parallel.
Each loop contains a vertical shell-and-tube heat exchanger about 5.5 ft
in diameter and 18 ft high, with a fuel-circulating pump mounted on the
tdp. Each pump impeller operates in a bowl that is integral with the ‘
shell of the associated exchanger. Above each bowl and connected to it
by open passages is a salt storage volume of about 45vft3, which is suf-
ficient to store about one-fourth of the fuel salt needed to fill the
reactor core. The general afrangement of the four blanket heat ex-
changers and the four fuel heat exchangers around the reactor is shown
in Fig. 3.8. |

In the heat exchanger, the fuel salt flows downward at about 11.3
fps through the outer row of 0.375-in.-diam tubes into the lower head,
where the salt conditions are about 1170°F and 51 psig. It then flows
upward at about 13 fps through the 0.375-in.-diam innermost tubes to the
bottom of the pump bowl, where the conditions are approximately 1000°F
and 5 psig. The pump discharges through the annular flow passage between
the 18- and 24-in. concentric pipes, and the salt returns to the reactor
plenum to repeat the cycle. Each pump is rated at 11,000 gpm at a 150-ft
head and requires a 1250-hp motor.

The blanket salt is pumped into the reactor vessel at about 1150°F
at a flow rate of about 17.3 cfs (7700 gpm). The blanket salt flows
downward through the space between the graphite reflector and the re-
actor vessel to cool the wall and the top head of the vessel, and then
flows upward through the blanket volume and the interstices of the core
lattice (the blanket salt occupies about 7% of the core volume). About
111 Mw(th) is deposited in the blanket salt as it passes through the re-
actor, and it leaves the reactor at about 1250°F through the inner pipe

of the 8- and 12-in.-diam concentric pipes.
/
 

 

 

31

The blanket salt is cooled in four circulating loops in a manner
similar to that used for the core salt. The blanket salt flows downward
through 0.375-in.-diam tubes in a 3-ft-diam, 9-ft-high vertical shell-
and-tube heat exchanger at about 10.5 fps. After passing through the
lower head, the fluid flows through another section of 0.375-in.-diam
tubing at about the same velocity and enters the pump bowls at about
1150°F. (These pumps do not have the large storage volume above the
bowls.) ZEach of the four blanket-salt pumps has a capacity of about
2200 gpm at a 150-ft head and uses a 500-hp motor. The salt flows to
the reactor through the annular region between the 8- and 12-in.-diam
concentric pipes connecting the heat exchanger and blanket volumes and
repeats the above cycle.

The volumes above the four fuel pump bowls have a combined capacity
sufficient to hold all the fuel in the reactor core. Since the reactor
is at a higher elevation than the fuel pumps, stoppage of the pumps will
cause the salt to drain from the core by gravity. It is estimated that
the reactor would become subcritical in 1 to 1.5 sec. Loss of one pump
would also cause the core to become subcritical because of salt drainage.
Thus all blanket and fuel-salt pumps need to be operative for the re-
actor to generate power. {An alternate modular design, discussed in
Chapter 4, permits partial power generation even though a fuel pump
fails.)

Afterheat generated in the salt stored in the volumes above the
pump bowls is removed by coils through which a coolant is circulated.
Salt remaining in the heat exchangers, piping, and reactor plenum cham-
bers is circulated through the exchangers by a gas 1lift to permit after-
heat removal from these volumes. The gas 1lift is provided by helium,
which is normally introduced continuously at the bottom of the heat ex-
changer to purge fission~product gases from the fuel salt. The fission-
product afterheat is transferred to the coolant salt, which will circu-
late through the primary exchangers by thermal convection and in turn
transfer energy to the steam cycle.

The fuel-, blanket-, and coolant-salt systems are provided with
"ever-safe"” tanks for storage of the salts when the systems are drained

for maintenance or other purposes. The drain valves for these lines

 
 

 

32

have not been specified, but they could possibly be freeze-type12 or

..mechanical valves developed for salt service.

As indicated above, fission-product gases such as xenon and krypton
are sparged from the fuel-salf circulating system by introduction of
helium in the bottom head of the heat exchangers. The off-gas system
and flowsheet for the handling of these radioactive gases are described
in Section 3.6. ,

A helium system provides cover gas for fhe pump bowls, drain tanks,
fuel-handling and -processing systems, and other equipment. This system
is briefly described in Section 3.6.

For processing pfirposes, small side streams of fuel salt (about
14.5 £t3/day) and blanket salt (about 144 ft3/day) are taken from the
main circulating loops and sent to the fuel-processing plant located in
cells adjacent to the reactor proper. The fuel-recycle system and its
flowsheet are described in Section 3.5.

An intermediate coolant salt is utilized to transfer energy from
the primary circuit to the steam cycle. The coolant salt is pumped
through the shell sides of the four fuel-salt heat exchangers and then
through the four blanket~salt exchangers by a total of eight pumps.

Four of these, each rated at 14,000-gpm capacity at a 150-ft head (1250-
hp motor), pump the coolant salt through the 16 boiler-superheaters.

The other four, individually rated at 2200 gpm at a 150-ft head (200-hp
motor), pump the coolant salt through the eight steam reheaters.

The coolant salt enters the shell side of each of the fuel-salt ex-
changers at about 850°F and at a rate of 37.5 cfs. The salt is thus
above the 842°F liguidus temperature of the fuel salt. The coolant salt
flows across the tube bundle, as directed by the baffles, to the exit at
the bottom. It then enters the top of the shell side of the blanket-salt
heat exchangers at about 1111°F, which is above the 1040°F liquidus tem-
perature of the blanket salt. It leaves the bottom of the shell at
about 1250°F, | |

About 87% of the coolant-salt flow, or about 32.5 cfs for each of
the large coolant-salt pumps, supplies a total of 1931 Mw(th) of heat
to the boiler-superheaters. The remainder of the flow, or about 5 cfs

for each of the small coolant-salt pumps, supplies about 293 Mw(th) of
 

 

 

 

33

heat to the steam reheaters. The coolant salt exits from the heat ex-
change equipment at 850°F.

The coolant salt enters the 16 vertical U-shell-and-tube heat ex-
changers, which serve as the boiler-superheaters, at the top of one leg
at a temperature of about 1125?F. Tt passes downward through the 18-in.-
diam baffled shell and upward through the other leg of the shell to
emerge at 850°F. The high-purity boiler feedwater, at about 700°F (the
estimated liguidus temperature of the coolant salt) and 3800 psia, is
introduced at the tube sheet at the top of one leg, flows through the
1/2-in.-diam tubes, and exits at the top of the other leg as steam in a
once-through arrangement. The steam leaves the units at 1000°F, 3600
psia, and a total rate of about 10,067,000 1b/hr.

As shown in the steam system portion of Fig. 3.7, about 7,152,000
lb/hr of the steam enters the throttle of the high-pressure turbine at
about 1000°F and 3500 psia. About 5,134,000 1b/hr leaves this turbine
at 552°F and 600 psia and flows to the eight U-tube vertical shell-and-
tube heat exchangers, which preheat the "cold" steam before it enters
the reheaters. It flows through the 20-in.-diam shells and is heated
to about 650°F by about 2,915,000 1b/hr of the 1000°F throttle steam,
which flows through 0.375-in.-diam tubes. The supercritical steam leaves
the tubes at 866°F and 3500 psia and is mixed with the 552°F 3500-psia
feedwater from the No. 1 feedwater heater in the regenerative steam cycle
to give a fluid temperature of about 695°F. The water is then boosted
in pressure to 3800 psia and raised in temperature about 5°F by two par-
allel 20,000-gpm 6200-hp motor-driven pumps. This produces the 700°F
feedwater for the boiler—superheatefs, as mentioned above.

The 650°F reheat steam from the preheaters flows through the tubes
of the eight vertical straight-tube 28-in.-diam shell-and-tube heat ex-
changers, which serve as the steam reheaters. The tubes in these units
are 0.75 in. in diameter. The heat source for the reheaters is the
1125°F coolant salt mentioned above, which raises the ten@erature-of the
steam to 1000°F. The steam returns to the double-flow intermediate-
pressure turbine at about 540'psia; this turbine is on the same shaft as
the high-pressure turbine. These two 3600-rpm prime movers drive a gen-

erator on the same shaft to give a gross electrical output of 527.2 Mw.

 
 

 

 

 

34

The steam leaves the intermediate~pressure turbine at about 172 psia and
706°F and crosses to the 1800-rpm four-flow low-pressure'turbine, where
it expands to 1.5 in. Hg abs and produces 507.7 Mw gross electrical power.
The regenerative feedwater heating system empioys eight stages of
feedwater heating, including the deaerator, and_two-tufbine-driven boiler
feed pumps. Condensing water, boiler makeup, and condensate-polishing
systems are also included.
The gross electrical generatlon of the plant is 1034.9 Mw; the net
statlon output is 1000 Mw(e). The overall net thermal efficiency is
4de 9%,

3.4 Reactor System

3.4.1 Description

Top and sectional views of the reactor wvessel and core are shown in
Figs. 3.10 and 3.11. Pertinent data on the reactor system are summarized
in Table 3.1.

The reactor vessel is about 14 ft in diameter and has an overall
height of about 19 ft. It is constructed of Hastelloy N; it is designed
for 1200°F and 150 psi; and it has walls 1.5 in. thick. The torospheri-
cal heads are 2.25 in. thick. The bottom head is an integral part of
the vessel, but the top head is arranged for grinding away the weld so
that the head can be removed. The vessel is supported on reinforcing
rings in the bottom head that rest on a structural steel stand mounted
on a reinforced-concrete pedestal in the center of the reactor cell.

As shown in Figs. 3.10 and 3.11, the fuel salt enters and leaves
the reactor through four concentric pipes (diameters of 18 and 24 in.)
in an arrangement that tends to minimize the stresses due to temperature
differences. These pipes communicate with plenum.chambers in the bottom
head of the reactor vessel. The fuel salt flows through the annular pas-
sage between the two pipes and enters the outer plenum chamber. It then
flows upward through the fuel-salt passages to the top of the reactor
and downward to the inner plenum chamber, where it leaves through the

18-in.-diam pipe.

3 ]
 

Ac—_1 BLANKET RETURN ORNL DWG 66-7128

CONGENTRIC FUEL LINES

 

      
   

    
 

    

 

 

 

 

r..,"' CORE CELL
REACTOR HOLD DOWN CLAWPS _ V@QO . O‘ e JV. \
(02610060 (000,00 \\
@@QOO@. .. %)! Q. o “\ REFLECTOR (GRAPHITE)
CONCENTRIC BLANKET LINES
’ . ’ P
\ ",

 

Fig. 3.10. Reactor — Plan BB.

 
 

 

 

~I7' 0"

T TT T T TTe
3

 

 

 

 

 

 

TO FUEL HEAT EXCHANGER

    

 

 

36

 

 

 

 

 

 

 

 

 

 

 

ORNL DWG 66-7129

TO BLANKET
HEAT EXCHANGER

3. oll ’

 

 

 

 

  

S
10 FT. DIA. CORE

 

 

 

 

 

 

 

 

 

Fig. 3.11.

 

Reactor — Elevation AA.

 

 

“ . - A k1
) A .
j o |
|- | ———Core cELL
- — !‘-_—
| | .
: b /—REFLECTOR (GRAPHITE)
A ! 1
: g
! —— Bl REACTOR VESSEL
' A
z
- FUEL SALT
Al DISTRIBUTION PLENUMS
1 K
| Al _~——BLANKET
o’ 4
~</' /|
U 114
M ! * A
H ~ F 1K
s 7N i
i \
\
T N
REACTOR PEDESTAL -
e TN e e
 

 

 

 

37

Table 3.1. Reactor System Data

 

Gross thermal power, Mw

Core
Blanket

Total
Reactor vessel

Outside diameter, f%
Overall height, ft
Wall thickness, in.
Head thickness, in.

Core

Height of active core, ft
Diameter, ft
Number of graphite fuel passage tubes
Volume, ft>
Volume fractions
Fuel salt
Blanket salt
Graphite moderator

Blanket

Radial thickness, ft

Axial thickness, ft

Volume, ft3

Volume fraction, blanket salt

Reflector thickness, in.
Fuel salt

Inlet temperature, °F
Outlet temperature, °F
Flow rate, ft3/sec (total)
1b/hr
gpm
Volume holdup, ft3
Core
Blanket
Plena
Heat exchangers and piping
Processing plant

Total
Salt composition, mole %
LiF
BeF;
UF, (fissile)

Blanket salt

Inlet temperature, °F
Outlet temperature, °F
Flow rate, ft?/sec (total)
1b/hr
gpm
Volume holdup, ft3
Core
Blanket ,
Heat exchanger and piping
Processing
Storage for Pa decay

Total

2114
111

2225

14
~19

2.25

12.5

534
282

0.169
0.0735
0.7575

2.0
1120

1000
1300

95.7
43,720,000
42,950

1lé6
26
147
345
33

717

63.6
36.2
0.22

1150
1250
17.3
17,260,000
7764

72
1121
100
24
2066

3383

 
 

 

 

 

38

Table 3.1 (continued)

 

Blanket salt (continued)

Salt composition, mole %
LiF

UF, (fissile)
System fissile inventory, kg
System fertile inventory, 1000 kg
Processing data

Fuel stream
Cycle time, days
Rate, ft3/day
Processing cost, $/ft?
Blanket stream
Cycle time, days
Rate, ft?/day
Processing cost, $/ft3
Fuel yield, % per annum
Net breeding ratio

Fissile losses in processing, atoms per fissile
absorption

Specific inventory, kg of fissile material per megawatt
of electricity produced

Specific power, Mw(th)/kg of fissile material
Core atom ratios

0

Fraction of fissions in fuel stream

Fraction of fissions in thermal-neutron group
Mean 7 of 233y

Mean 7 of ?3°U

Net neutron production per fissile absorption (Me)
Power density, core average, kw/liter

Gross
In fuel salt

Neutron flux, core average, neutrons/cmz-sec

Thermal
Fast
Fast, over 100 kev

Core thermal flux factor, ratio of peak to mean

Radial
Axial
Qverall plant data

Net electrical output, Mw

Gross electrical generation, Mw

Boiler feedwater pressure-booster pump power, Mw(e)
Station auxiliary load, Mw(e)

Net heat rate, Btu/kwhr

Net efficiency, %

Assumed plant factor

71.0
2.0
27.0
0.0005

769
260

47
14.5
203

23
144
7.33

4.86
1.0491
0.0057

0.769

2.89

41-7
5800

0.987
0.806
2.221
1.958
2.221

80
473

6.7 X 1014
12.1 X 1014
3.1 X 1014

 

o
 

 

 

39

The active portion of the reactor core is 10 ft in diameter and
about 12.5 ft high. It contains 534 graphite tube assemblies through
which the fuel salt flows and around which the blanket salt circulates.
Each tube assembly, as shown in Fig. 3.9, consists of a 3.5-in.-0D graph-
ite tube with eight 0.53-in.-ID holes regularly spaced on a 2.62-in.-
diam circle. The fuel salt flows upward through these eight tubes at
about 15 fps. The salt reverses direction at the top of the fuel as-
sembly, flows downward at about 15 fps through the 1.5-in.-ID central
passage, and enters the inner plenum at the bottom of the reactor. The
3.5-in.-0D graphite tubes are slipped into hexagonally shaped passages
inside hexagonal graphite tubes that are approximately 5 in. across the
outer flats. Blanket salt circulates in the passages between the circu-
lar and hexagonal graphite tubes. Thin portions of each outside face of
the hexagonally shaped graphite are cut away, as indicated in Fig. 3.9,
to form passages for circulating blanket-salt. The fuel tubes are con-
tinuous along their lengths, whereas the hexagonal tubes are made up of
stacked graphite pieces. The upward and downward fuel flow passages com-
municate at the top of the fuel tube, where a threaded-graphite plug
tightly closes the top end of the tube, as shown in Fig. 3.9. This plug
is provided with a threaded-graphite stud, washer, and nut assembly for
holding the hexagonal pieces in place.

Stubs of 4-in.-0D Hastelloy N tubes that vary in length from about
6 to 15 in. are welded to the upper diaphragm in the lower head of the
reactor vessel. This diaphragm is about 0.75 in. thick. Metal transi-
tion pieces with an outside diameter of 4 in. and a length of about 8 in.
are brazed to each of the stubs; previous to this, the 3.5-in.-0D graph-
ite tubes for the fuel salt are brazed under carefully controlled shop
conditions to shoulders on the inside of the metal transition pieces.

The hexagonally shaped graphite tubes rest on top 4-in.-diam by about

4=in.-long metal spacers, which in turn rest on top the metal transition

- pleces.

Other Hastelloy N stubs, 2 in. OD and varying in length from 8 to
30 in., are welded to the 0.25-in.-thick top of the inner plenum chamber
at the bottom of the reactor vessel. These stubs neck down to about

1.62 in. OD at the top and are a sliding fit into the bottom of the inner

 
 

 

 

 

40

passage of the graphite fuel tube (the tubes are machined at the bottom
end to permit this fit). Any salt leakage through this joint constitutes
only a small bypass of the core. |

The blanket salt leaves and enters the reactor vessel through con-
centric 8- and 12-in. pipes located near the top of the reactor vessel
(see Fig. 3.8). The inner pipes of these concentric connections, like
those in the fuel-salt system, are provided with slip joints near the
heat exchanger nozzles to allow for the relative movement between pipes
due to temperature differences. Small leakage through the joints is
inconsequential. '

The blanket on the sides and top of the core averages 1.5 ft in
thickness. Outside this blanket, and 1.5 in. from the vessel wall and
top head, is a 3-in. thickness of graphite which serves as a reflector
for neutron economy and also helps to protect the wvessel from irradiation
damage. The annular space between the reflector graphite and the wall
is a flow passage for the incoming blanket salt; the stream enters at a
temperature of 1150°F and serves to cool the vessel wall and the top
hesd. |

The basic design of the reactor has the advantage of low neutron
losses to structural materials other than the graphite. Except for some
unavoidable loss of delayed neutrons in the external fuel-salt circuit,
there is almost no neutron leakage through the thick blanket. Neutron
losses to fission products are minimized by the continuous treatment of
a side stream of the fuel salt in a processing plant that is part of the
MSBR power station. The nuclear performance is discussed in more detail
in Section 3.5.

The reactor system described above provides for support of the
graphite when the vessel is empty of salts, prevents the graphite from
floating during normal operation, and allows for thermal expansion and
growth or shrinkage of the graphite. The core can be removed as an
assembly after the holddown clamps are unbolted and removed and the seal
weld is cut (see Fig. 3.11). The upper plenum diaphragm, which carries
the load of the graphite in the reactor core, can then be removed for
replacement should this prove necessary. Tools must be developed for

seal-weld cutting, joint preparation, and rejoining. The drawings do

m

i
 

 

41

not indicate a means of guiding a new core assembly into position, but
this could be readily provided.

Replacement of a graphite tube with the core in place may also be
practical. This could be accomplished by first cutting off and removing
the top head of the reactor vessel to expose the tops of the fuel pas-
sage tubes. Removal of the graphite nut at the top of the defective or
suspect tube would permit withdrawal of the graphite hexagonal section
surrounding the tube. The Hastelloy N spacer at the bottom could then
be lifted out to make it possible to lower an induction coil heater and
break the metal-to-metal brazed joint between the metal stub and transi-
tion pieces. A replacement tube could be installed with a reverse pro-

cedure.

3.4.2 Reactor Materials

 

Fuel and Blanket Salts. The chemical compositions of the fuel and
blanket salts and the pertinent physical properties employed in the de-
sign are shown in Table 3.2. The phase diagrams of these salts and a
general discussion of the chemistry, physical properties, and behavior
of fluoride salts are given in Ref. 1. The feagibility of the use of
these salts in reactors is well established on the basis of many experi-

mental studies® and MSRE experience.

Table 3.2. Physical Properties of MSBR Fuel, Blanket, and Coolant Salts®

 

 

Fuel Salt  Blanket Salt  coolant
, Salt

Reference temperature, °F 1150 1200 988
Salt components ‘ , - LiF-BeFy=UF, LiF-ThF,-BeF; NaF-NaBF,
Nominal salt composition, mole % 68.3-31.2-0.5 71.0-27.0-2.0 61.1-38.9
Molecular weight, approximate 34 o 103 68
Iiquidus temperature, °F 842 1040 700
Density, 1b/ft> 127 277 125
Viscosity, 1b/hr-ft 27 38 12
Thermal conductivity, Btu/hr-ft® (°F/ft) 4 1.5 1.3
Heat capacity, Btu/lb.°F 0.55 . 0.22 0.41

 

%The values listed are those used in the MSBR heat-power cycle studies to es-
tablish heat transfer coefficients, flow rates, etc. Many of the properties are not
known with certainty, and a few, such as the viscosity of the cooclant salt, are
little better than rough estimates. In addition, the values used for thermal con-
ductivity appear at present to be slightly high (Ref. 1).

 
 

42

Coolant Salt. The coolant tentatively selected for the MSBR is a
sodium fluoroborate salt that appears to be compatible with the materials
in the system and with the fuel and blanket salts; it has a liguidus tem-
perature of about 700°F and appears to have heat transfer and fluid flow
Properties that make it generally suitable for MSBR application. Several
of the physical properties shown in Table 3.2 need to be verified but
are believed to bé sufficiently accurate for the purposes of this study.

Graphite. The MSBR core graphite would be an improved grade of
that used in the MSRE (properties of the MSRE core graphite are given
specifically in Ref. 2). The MSBR graphite tubes should have no signifi-
cant cracks and should be able to withstand high radiation exposures ex-
ceeding 10°2 neutrons/cm? (neutron energies above 300 kev).

Hastelloy N. The salt-containing portions of the MSBR are fabri-
cated of Hastelloy N, since it has excellent compatibility with molten
fluorides at high temperatures and under severe radiation conditions.

The chemical composition, mechanical and physical properties, and corro-
sion resistance of this material are discussed in Ref. 2. The mechanical-
property values given in Ref. 2 were used in conjunction with ASME Code
requirements in specifying equipment. Although Hastelloy N has exhibited
radiation embrittlement when irradiated to MSBR exposures, major improve-
ments in the radiation stability of the material can be obtained by minor

changes in composition and by modifying the heat treatment.?

3.4.3 MSBR Load-Following Characteristics

The negative temperature coefficient of reactivity makes the MSBR
independent of the need for control rods for load following.' The pre-
liminary nature of this report did not permit a study of reactor safety,
but on the basis of MSRE studies 15 and experience,®
converter reactor‘safety studies,l7 it appears that the fuel, blanket,

and molten-salt

and coolant-salt temperatures will be quickly self-adjusting with no
oscillations or reactivity perturbations of consequence following changes
in turbine-generator load. In recognition of the need to control the _
throttle-steam superheat temperature at 1000°F and the reheat steam tem-
perature at 1000°F independently of eaéh other and of turbine load,
separate variable-speed coolant-salt pumps were specified for the boiler~

superheaters and the reheaters.
 

 

 

 

 

43

3.5 DNuclear Fuel-Cycle Performance

 

It is desirable that the rate of fissile fuel yield be maximized
consistent with low fuel-cycle costs. Since two nuclear designs can
have about the same fuel-cycle cost but significantly different fuel
doubling times, MSBR nuclear design optimization studies were performed
to find conditions corresponding to both low fuel-cycle costs and high
fuel~yield rate. '

An important feature of the MSBR concept is the fuel-recycle plant,
which is an integral part of the reactor plant. Fuel-recycle costs play
an important role in determining the rate at which fuel can be economi-
cally processed and thus significantly influence the breeding ratio and
fissile inventory of the MSBR. In order to properly consider this in-
fluence, a detailed design and cost study was made of the fuel-recycle
plantl® and is summarized below. The costs obtained, including those
for capital and operation of the fuel-processing plant, have been kept
separate from the costs of the main reactor plant.¥* This was done in
order to show a fuel cost that can be more readily compared with fuel
costs of other reactor plants utilizing off-site fuel-recycle facilities,

where fabrication and processing charges include such facility costs.

3.5.1 Design and Cost Study of Processing Plant for Fuel Recycle

The MSBR core fuel consists of fissile UF, dissolved in an inert
carrier salt containing 7LiF4 and BeFp. The blanket salt contains the
fertile material, ThF4,.which is also dissolved in a carrier salt con-
taining "LiF,; and BeF,. The flowsheet for the MSBR processing plant for
recycling the fuel is shown in Fig. 3.12. ' '

The fuel stream is processed by the well-established fluoride vola-

tility process to separaté the uranium from the carrier salt and fission

- products. The valuable carrier salt is separated from the fare—earth

fission products by the vacuum-distillation process; about 6.5% of the

 

*¥An exception to this is the capital cost of the building for the
fuel-recycle plant. This has been included with the reactor plant,
since the fuel-recycle system is housed within the reactor building.

 
 

 

 

 

ORNL-DWG 65- 6194 Al

 

UFg RECYCLE TO REACTOR

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 
 
   
 
   
   

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

WYl | 7 soaaens/ s
e N 000005
°C > /| A
V¢ A A | 7
R
sl .
g e ¥ | PRODUCTION /
UFg + Nor MR/ " waste /] (L LS
VOLATILE FP i
MAKE UP | aF MR/ TP
LIF/BeE, /ThE, ok + vz lr
VOLATILE FP MAKE UP
_ LiF/BeF,
) | fonTinudus Vioots Vegiiscos A A 7 AN
e ke Rl AN ik Mo el
/| MAKE B , VOLATILITY /] _[FP DECAY |/ voLATILITY |/} DisTiLLATION | - LiF/BeF, /|7 ) REDUCTION || FILTRATION
7/// 7// //// // //// c NN A~s80%C /// ~1000°c [ /|”~500°c /|, /|, 550-600°¢C // //
/ ~550°C / /”1.5?/ UFq'+F2=UF6 7 {mm Hg //////// /%/
000 N 0 00 0
. & A A
LIF/BeF, / ThE, /FP F— F— LiF + RARE —H;

    

 

EARTH FP REDUCED METALS

 

 

 

 

 

7777 77 7 Cr, Fe. N
/DISCARD FOR/ WASTE /

- FP REMOVAL [ STORAGE

L L L L L2272 /

 

 

 

 

FERTILE STREAM RECYCLE LiF/BeF,/UF4 RECYCLE

 

 

 

Fig. 3.12. Fuel- and Blanket-Salt Processing for the MSER.

 
 

 

45

carrier salt is either discarded or unrecovered in the distillation pro-
cess in order to control fission-product buildup and reduce recovery
costs. | ':.HV -

The fuel salt is reconstituted by absorbing UFg in uranium-containing
carrier salt”and.feducing it to UF, by bubbling hydrogen through the melt.
Excess urénium.from fihE‘reactor is sold as an equilibrium mixture of the
fuel isotopes. _ - |

The blanket‘saitisproéessed by the fluoride volatility process
alone. Any uranium nbt removed during blanket processing returns to the
blanket and isvremovedfby‘subsequent processing.

Smell side streams of about 14.5 £t3/day of fuel salt and 144 t3/day
of blanket salt are contifiupusij withdrdwn from the reactor circulating
systems and routed to fihe\fiquéssing plant located within the same build~
ing. The inventoriesuretained in'the'processing plant are estimated to
be about 10% of the reactor system fuel-salt inventory. The correspond-
ing value for the blanket system is about 1%.

An_imporfiant factor affecting both the breeding gain and the fuel
cost is the loss of fisSile‘material in processing. There is considerable
engineering experiénce_in fluéride volatility prbcessing that indicates
an MSBR fissile material loss of 0.1% per pass or less through the pro-
cessing plant. Therefore le.l%'loSS*per pass has been assumed in this
study. | o |

Based on the fuele}ecycle proceSsing schemes indicated above, capital
cost studiesl® were made of an MSBR integrated processing plant. The
plant throughput was assfimed to be 15vft3/day of fuel salt and 105 ft3/day
of blanket salt, with'each stream being treated separately. These through-
put rates correspond roughly to the needs of a 1000-Mw(e) station.

In performing the proéessing plant cost study,l8 the equipment flow-

sheet given in Fig. 3.13 was developed, the required equipment was de-
signed, and cost estimates were made for the process equipment and asso-
ciated structures. The basic processes considered involve fluorination,
purification of UFg, vacuum distillation, reduction of UFe and reconsti-
tution of the fuel, off-gas processing, waste storage, flow control of
the salt streams, removal of decay heat, provisions for sampling of the

salt and off-gas streams, and provisions for shielding, maintenance, and

 

 
 

 

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46

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47

repair of equipment. Based on these considerations and associated opera-
tions, a total direct capital cost for the plant was obtained along with.
a. direct operating cost. From these results, and consideration of in-
direct costs, the total fuel-recycle processing costs were obtained.

The major novel pieces of processing equipment are the fluorinator,
fuel reduction equipment, and distillation unit. The designs considered
are shown in Figs. 3.14, 3.15, and 3.16. Figure 3.14 illustrates the
fluorinator, which utilizes a frozen wall of salt and performs continuous
fluorination of a flowing stream of uranium-containing molten salt. The
NaK coolant flowing through the jacket, as shown, freezes a layer of salt
on the inner surface of the column to protect the structural material
(alloy 79-4) from corrosive attack by the molten-salt—fluorine mixture.
Figure 3.15 illustrates the equipment for reducing UFg to UF,. Barren
salt and UFg enter the bottom of the column, which contains circulating
LiF-BeF5-UF,. The UFg dissolves in the salt, aided by the presence of
UF,, and moves up the column, where it is reduced by hydrogen. Reconsti-
tuted fuel is taken off the top of the column and sent to the reactor
core. Figure 3.16 illustrates the design of the vacuum-distillation unit.
Barren fuel-carrier salt flows continuously into the still, which is held
at about 1000°C and 1 mm Hg. LiF-BeF, distillate is removed at the same
rate that salt enters, and thus the volume is kept constant. Most of
the fission products accumulate in the still bottoms and are drained to
waste storage when the heat-generation rate reaches a prescribed limit.

The fuel~-recycle processing plant is located in two cells adjacent
to the reactor shield; one contains the high-radiation-level operations
and the other contains the lower radiation-level operations. Each cell
is designed for top access through a removable biological shield having
s thickness equivalent to 6 ft of high-density concrete. Both cells are
served by a crane used in common with the reactor'plant. Process equip-
ment is located in the cell for remote removal and replacement from above.
No access into the cells with be required; however, it is possible with
proper decontamination to allow limited access into the lbwer radiation
level cell. A general plan of the processing plant and a partial view of
the reactor system are shown in Fig. 3.17. The highly radioactive opera-

tions involved in fuel-stream processing are carried out in the smaller

 
 

 

48

ORNL DWG €5-3037

 

 

 

 

 

 

 

 

Fig. 3.14. Continuous Fluorinator for Salt Processing.
 

ORNL OWG. 65-3036

 

 

 

 

 

 

Fig. 3.15. Fuel-Reduction Column for Salt Processing.

 

 
 

 

 

 

 

 

 

 

 

LiF-BeF2
PRODUCT

i

 

Z, NaK OUTLET
-«

817 TUBES

fe———————34" |D—————=-

50

ORNL DWG 65-1802R2

AR COOLANT

;—\
/\

 

770N
]lil’é b o

T

B I l l AIR OUTLET

 

1/2" x 16 GAUGE

1/2" INOR-§

TUBE SHEET

FUEL SALT
INLET

 

Fig. 3.16.

 

N

 

 

 

 

INOR-8
1/2" WALL

 

 

amal) —+ NaK COOLANT

 

 

 

LiF WASTE DRAIN

30" ID——e-

 

Vacuum Still for Carrier-Salt Recovery.
 

 

 

51

ORNL-DWG 66-7459

   
      

 

 

v .

C

't

- :

v 4

v ’

, ¥

PROCESSING < "

CELLS -

AR-NaK - R
HEAT 4 ;
EXCMANGER ¥ 3
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e
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d

  

      
   
  
  
  

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SR N ,. JReacTor
MAKEUP § CELL
AREAN

B g WL VA g g Wk W R PLAW LA L  y

 

 

1
i s
= r .
o ® @ Fflf
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4

 

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ol
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YEOK

i
r
Y

 

 

 

.j_nig% —

Fig. 3.17. Arrangement of Salt-Processing Equipment.

cell (upper left). The other cell houses equipment for the fertile-stream
and the "cooler'" fuel-stream operations.

The highly_radioactive cell contains only fuel-stream processing
equipment: the fluorinator, still, waste receiver, NaF and MgF2 sorbers,
and associated vessels. The other cell houses the blanket-processing

equipment and fuel- and fertile-stream makeup vessels.

 
 

 

 

 

52

A detailed cost estimate for the fuel-recycle plant was made and is
reported in Ref. 18. The results for the total capital costs are sum-
marized in Table 3.3. The operating and maintenance costs for this plant

were also estimated and are shown in Table 3.4. The direct operating

Table 3.3. Summary of Cost Estimate for a Typical
Fuel-Recycle Processing Plant for a
1000-Mw(e) MSBR Stationa

 

Installed process equipment : | | $ 853,760
' Structure and improvements | 556,770
Interim waste storage 387,970
Process piping 155,800
Process instrumentation 272,100
Electrical auxiliaries 84,300
Sampling connections 20,000
Utilities (15% of installed process equipment) 128,060
Insulation (6% of installed process equipment) 51,220
Radiation monitoring 100,000
Total direct plant cost $2,609,980
Construction overhead (30% of total direct 782,990
plant cost) -
Subtotal construction cost $3,392,970
Engineering and inspection (25% of total con- 848,240
struction cost)
Subtotal plant cost | $4,241,210
Contingency (25% of subtotal plant cost) 1,060,300
Total construction cost $5,301,510
Inventory® cost of NaK coolant (at $100/ft3) 40,000
Total capital cost $5,341,510

 

%Based on throughput of 15 ft3/day of fuel salt and 105
£t3 /day of blanket salt.

bInventory of fuel and blanket salts is considered as
part of the reactor inventory.

o
 

 

 

 

53

Table 3.4. Summary of Annual Operating and Maintenance
Costs of Fuel-Recycle Processing Plant for
1000-Mw(e) MSBR Station®

 

Direct labor | $222, 000
Labor overhead 177,600
Chemicals 14,640
Waste containers 28,270
Utilities 80, 300
Maintenance materials

Site , 2,500

Services and utilities 34,880

Process equipment : 160, 040
Total annual charges $721,230

 

®Based on throughput of 15 ft3/day of fuel salt
and 105 £t3/day of blanket salt.

cost includes the cost of immediate supervisory, operating, maintenance,
laboratory, health physics, clerical, and janitorial personnel; also in-
cluded are costs of chemicals, waste containers, utilities, and mainte-

nance materials.

These capital and operating costs were used as base points for ob-
taining the costs for salt-processing plants having different through-
puts. Specifically, the capital and operating costs were estimated
separately for each fluid stream as a function of plant throughput,
based on the volume of salt processed.l9 The results of these estimates
are given in Fig. 3.18, and were used in calculating the nuclear and
economic performance of the MSBR fuel cycle.
| It may be noted that in Table 3.3 the indirect charges (overhead,
‘ehgineering; and éontingencies) amount to a total of about 100% applied
against the direct construction cost of the processing plant. This
compares with a similar value of about 41% used in the cost estimate of
the MSBR reactor and’tufbine;generator plant (see Sect. 3.11). The high

value used here should more than compensate for the higher rates of

 
 

54

ORNL-DWG 66-7455
BLANKET STREAM PROCESSING RATE (ft%day)

2 5 10° 2 5 10

AL COST OF CORE PRCCESSING

TING COST OF CORE PROCESSING

N

CORE STREAM PROCESSING COST (:S7f13)

BLANKET PROCESS!

'BLANKET STREAM PROCESSING COST (S/t3)

COST OF
OPERATING COST OF \BLANKET PROCESSING

 

2 5 10 2 5 100
CORE STREAM PROCESSING RATE (ff:'f’doy)

Fig. 3.18. MSBR Fuel-Recycle Costs As a Function of Processing
Rates. Fluoride volatility plus vacuum distillation processing for
core; fluoride volatility processing for blanket; 0.8 plant factor;
12%/yr capital charges for investor-owned processing plant.

equipment replacement in the fuel-processing plant as compared with the

power plant as a whole.

3.5.2 Nuclear Design Method

 

Values of the MSBR nuclear design parameters, which were largely
fixed by the design criteria in conjunction with nuclgar—economic calcu-
lations, are listed in Table 3.1. The criteria helped to establish the
design of the salt-circulating loops external to the reactor (the volumes
associated With these loops constitute the largest portion of the total
volume of salt holdup). Additional parameters which were optimized by

the fuel-cycle-performance calculations were the reactor dimensions, the

»

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o
 

 

 

55

power density, the core composition, including the carbon-to-uranium and

thorium-to-uranium ratios, and particularly the fuel-recycle rates through

- the processing plant. Table 3.1 also lists the parameter values obtained

through nuclear design optimization.

The fuel-cycle calculations were performed with OPTIMERC, a combina-
tion of an optimization code and a multigroup, diffusion, equilibrium
reactor code. Details of the program are summarized in Ref. 20. 1In
brief, the program initially calculates the nuclear performance, the
equilibrium concentrations of the various nuclides (including the fission
products), and the fuel-cycle costs for a given set of conditions; fol-
lowing this, performance optimization is done by permitting up to 20 re-
actor parameters to be varied, within limits, in order to determine the
most desirable values based on the method of steepest ascent. Typical
input parameters were the reactor dimensions, blanket thickness, frac-
tions of fuel and fertile salts in the core, and fuel- and blanket-stream
processing rates. These parameters were varied in a logical fashion,
with final values based on designs optimized primarily for minimum fuel
cost, with lesser emphasis given to maximizing the annual fuel yield.

In addition to fuel-cycle cost per se, OPTIMERC includes several
equations for approximating certain capital and operating costs that vary
with nuclear design values, such as the size of the reactor vessel and
the cost of graphite. These costs were automatically added to the fuel-
cycle cost in the optimization routine so that the optimization search
would take into account all known economic factors. However, costs other
than the fuel-cycle cost are reported undér capital investment (Sect.
3.11).

Standard neutron-cross-section libraries were used in obtaining the
broad-group cross sections for the MSBR physics calculations (1.2 groups
were employed, with one effective thermal group). The cross sections
were_evaluatéd and modified where necessary to be consistent with present
information (see also Sect. 3.5.4). In obtaining the nuclear constants
for nonthermal neutron groups and for a particular region, a transport-
type multigroup calculation was performed (B-1 approximation to the
Boltzmann equation for a single region); the three specific regions con-

sidered were the homogenized core, the blanket, and the reflector regions.

 
 

 

 

56

The effective thermal-neutron reaction rate was based on transport calcu-
lations, which generated the thermal-neutron spectrums in the various
reactor regions. In the core, the thermal-spectrum calculation considered

the core lattice cell to consist of concentric cylindrical regions; the

- resulting neutron reaction rates were used to determine the effective

thermal-group cross sections for the various nuclides.

The broad-group cross sections were employed in a one-dimensional
multigroup diffusion program modified so as to approximate a two-dimen-
sional calculation. The concentrations of the various nuclides were
based on equilibrium neutron-reaction rates, which were consistent with
criticality considerations, the fuel-processing rate, the assumed be-
havior of fission products and higher isotopes, and the sale of uranium
having an isotopic composition equal to the average in the reactor plant.

These reactor-physics calculations were incorporated in the fuel~
cycle-performance optimizations carried out by the OPTIMERC program, in

which various parameters were allowed to vary within specified limits.

3.5.3 DNuclear Performance and Fuel-Cycle Cost

 

The nuclear performance of the MSBR is significantly influenced by
the physical behavior of the fission products. In particular, the be-
havior of 13%Xe and other fission gases is important. A gas-stripping
system is provided to remove these gases from the fuel salt. However,
part of the xenon could diffuse into the moderator graphite. In the
calculations reported here, a *>°Xe poison fraction of 0.005 was assumed.

The disposition of the various fission products in the reactor and
processing system, based on their estimated physical, chemical, and
thermodynamical properties, was assumed to be as shown in Table 3.5.

Another factor to consider is the behavior of corrosion products.
However, the control of corrosion products in the MSBR does not appear
to be a significant problem, so the effect of corrosion products was ne-
glected in the nuclear calculations. Not only is the corrosion rate very
low, but the fuél-processing methods considered here can remove corrosion
products from the molten salts (by reduction with hydrogen followed by
filtration).

te
 

 

i s i

 

 

57

Table 3.5. Disposition of Fission Products in
Reactor and Processing Systems

 

Group Assumed Fission-Product Behavior Fission Products

 

1 Elements present as gases; assumed to be Kr, Xe
removed by gas stripping, with a small
fraction absorbed by graphite

2 Elements that plate out on metal sur- Ru, Rh, Pd, Ag, In
faces; assumed to be removed in-
stantaneously

3 Elements that form volatile fluorides; Se, Br, Nb, Mo, Te,
assumed to be removed in the fluoride Te, I
volatility process

4 Elements that form stable fluorides less ©Sr, Y, Ba, Ia, Ce,
volatile than LiF; assumed to be Pr, Nd, Pm, Sm,
separated by vacuum distillation Eu, Gd, Tb

5 Elements that are not separated from the Rb, Cd, Sn, Cs, Zr

carrier salt; assumed to be removed
only by salt discard

 

The calculation of fuel-cycle cost involves economic factors as
well as those given above. The economic ground rules used here are given
in Table 3.6. The values of the fissile isotopes were taken from the
current AEC price schedule. The capital charges of lZ%/yr for depreciat-
ing items and 10% for nondepreciating materials correspond to those for
a privately owned plant; the corresponding values used for publicly
owned plants were 7 and 5%/yr, respectively.

The processing costs are based on the specific fuel-recycle plant
design and cost study given above and are included in the fuel-cycle
costs. The results, given in Fig. 3.18, were used to estimate the pro-
cessing cost as a function of fuel-processing rate. FProcessing losses
corresponded to a fissile material loss of 0.1% per pass through fuel-
recycle processing.

The results of the fuel-cycle calculations for the MSBR design are
sumarized in Table 3.7 and the neutron balance is given in Table 3.8.
The reactor has the advantage of no neutron losses to structural mate-

rials in the core other than the moderator. Except for some unavoidable

 
 

 

 

58

Table 3.6. Basic Economic Assumptions Used
in Nuclear Design Studies

 

Reactor power, Mw(e) | | 1000
Thermal efficiency, % 45
Load factor . 0.80
Cost assumptions
Value of 233U and 233Pa, $/g 14
Value of 23°U, $/g 12
Value of thorium, $/kg 12
Value of carrier salt,'$/kg 26
Capital charge, %/yr
Plant 12
Nondepreciating capital, including 10

fissile inventory
Processing cost, $/ft> salt

Fuel (at 10-ft3/da processing rate) 252
Blanket (at 100-ft?/day processing rate) 9.3
Processing-cost scale factor See Fig. 3.18

 

Table 3.7. MSBR Fuel-Cycle Performance

 

Fuel yield, % per year | 4,86
Breeding ratio 1.0491
Fissile losses in processing, atoms per fis- 0.0057
sile absorption

Neutron production per fissile absorption (Tme) 2.221
Specific inventory, kg of fissile material per 0.76%9
megawatt of electricity produced
Specific power, Mw(th)/kg of fissile material 2.89
Power density, core average, kw/liter

Gross 80

In fuel salt 473
Fraction of fissions in fuel stream 0.987
Fraction of fissions in thermal-neutron group 0.806
Mean 1 of 433y 2.221

Mean 1 of 233U : 1.958

 

“)

tm
 

 

 

 

»

59

Table 3.8. MSBR Neutron Balance

 

Neutrons per Absorption
in Fissile Fuel

 

 

Material
Total Absorbed Giving Neutrons
Absorbed Fission Produced

232M 0.9710 0.0025 0.0059
233pg 0.0079
233 0.9119 0.8090 2.0233
234y 0.0936 0.0004 0.0010
235 0.0881 0.0708 0.1721
<36y 0.0115 0.0001 0.0001
23TNp 0.0014
238y 0.0009
Carrier salt (except 6Li) 0.0623 0.0185
611 0.0030
Graphite 0.0300
135%e 0.0050
149gm 0.0069
151gpm 0.0018
Other fission products 0.0196
Delayed neutrons lost® 0.0050
Leakage® 0.0012

Total 2.2209 0.8828 2.2209

 

 

 

 

aDela,yed neutrons emitted outside the core.

bLeakage, including neutrons absorbed in the reflector.

loss of delayéd'neutrons in the external fuel circuit, there is almost
zero neutron leakage from the reactor because of the thick blanket. The
neutron losses to fission products are minimized by the availability of
rapid and inexpehéive integrated processing. |

The fuel-cycle cost for the MSBR is given in Table 3.9. The main
items are the fissile inventory and processing costs. The inventory
costs are rather rigid for a given reactor design, since they are largely

determined by the fuel volume external to the reactor core region. The

processing costs are, of course, a function of the processing-cycle times,

one of the chief parameters optimized in this study.

 
 

 

 

60

Teble 3.9. Fuel-Cycle Cost for MSBR*

 

Costs [mill/kwhr(e)]

 

 

Fuel Fertile  Sub- Grand
Stream  Stream total Total
Fissile inventory® 0.1180 0.0324  0.1504
Fertile inventory 0.0459 0.0459
Salt inventory 0.0146  0.0580 0.0726
Total inventory 0.269
Fertile replacement 0.0185 0.0185
Salt replacement 0.0565 0.0217 0.0782
Total replacement 0.097
Processing 0.1223  0.0440 0.1663
Total processing 0.166
Production credit (0.073)
Net fuel-cycle cost 0.46

 

%Based on investor-owned power plant.
bIncluding 233Pa, 233U} and #3°U,

The fuel costs in Table 3.9 are based on use of private financing.
Fuel-cycle and power-production costs based on public financing are also
of interest. With public ownership, the fixed annual charge on depre-
ciating capital is taken asr7% and on nondepreciating items as 5%. This
difference in the financial conditions results in slightly different
optimization points for the fuel cycle that affect the volume fractions
of fuel, the thorium-to-uranium and carbon-to-uranium ratios, etc. Re-
optimizing such parameters has only minor effects on the nuclear per-
formance. However, the difference between the 12 and 7% annual fixed
charges on the cost of the fuel processing plant and the lower charges
on nondepreciating items (5% versus 10%) results in lowering the esti-
mated fuel cost from 0.46 mill/kwhr to about 0.29 mill/kwhr. Table 3.10
summarizes the fuel-cycle costs for investor-owned and for publicly owned

plants.
 

 

 

®y

»

61

Table 3.10. MSBR Fuel-Cycle Costs for Investor-Owned
and Publicly Owned Plants

Plant factor: 0.8

 

Cost [mill/kwhr(e)]

 

Investor Public
Ownership®  OwnershipP

 

Fissile-, fertile-, and carrier- 0.269 0.135
salt inventory
Replacement cost of fertile and 0.097 0.097

carrier salts

Core- and blanket-processing costs

Operation and maintenance 0.075 0.075

Capital costs 0.091 0.053
Bred fuel credit (0.073) (0.073)
Net fuel-cycle cost 0.46 0.29

 

®Based on 12%/yr capital charges for processing plant
and inventory charges of 10%/yr.

bBased on 7%Vyr capital charges for processing plant
and inventory charges of 5%/yr.

3.5.4 Critique of Nuclear Performance Calculations

 

The performance characteristics given above show that the MSBR has

a high specific power [about 1.2 Mw(e)/kg fissile] and a relatively low

 breeding gain (about 0.05 net fuel bred per unit of fuel burned). Un-

certainty in the specific power is due to uncertainties in the fuel in-
ventory requirements external to the reactor core (related to the fuel
heat exchanger design and flow-distribution systems), as well as to in-
accuracies in the critical-mass calculations. _it is estimated that about
a 10% uncertainty exists in the fuel volume réquirements.external to the

core of the MSBR because of uncertainties in heat transfer, fluid trans-

- port, and flow distribution requirements. Relative to critical mass,

experience with the MSRE indicates that the calculational methods and

applicable neutron cross sections employed are reliable (the calculated

 
 

62

MSRE critical mass was within 1% of the experimental value). Also, the
methods and cross sections employed are similar to those used by other
groups who have had good success in calculating the reactivity of criti-
cal assemblies. As a result, the uncertainty in the critical concentra-
tion is estimated to be less than 5%. Thus the uncertainty in the
specific power appears to be less than 15%. In addition, because of the
use of fluid fuel, compensating changes in fissile and fertile material
concentrations can be made if the calculated quantities are in error.
Finally, since the specific power is high, a small change in its value
cannot change the fuel-cycle cost appreciably. Thus uncertainties in
speéific power do not appear to significantly affect MSBR performance.

With a 10w‘breeding gain, however, uncertainties in nuclear con-
stants, fuel-processing losses, and/or physical properties of the fis-
sion prodficts can have a significant influence on the fuel doubling time
through their influence on the net breeding ratio. A detailed appraisal
of the MSBR huclear-performance uncertainties due to the above factors
is given in Ref. 21, and the results are summarized below.

The importent nuclides in the MSBR are C, Li, Be, F, U, Th, Pa,
and fission products. Changes in the neutron-absorption cross-section
values of these nuclides can influence the breeding ratio, with some
nuclides having more importance than others. The cross-section values
are not known in an absolute sense, but they can be inferred from the
precision of the various measurements available. On‘this basis, a range
of values was assigned to each nuclide that represents a "best judgment"
of the values within which the true value will fall.

The neutron balance given in Table 3.8 shows the relative absorp-
tions in the various materials based on the studies performed. Of the
nuclides indicated, only two or three have cross-section uncertainties
that could individually affect the breeding ratio by as much as 0.0l.

By far the most important nuclide is 233U, and its most important charac-
teristic is the value of eta averaged over the reactor neutron spectrum.
The 2200-m/sec value and the variation of eta with energy are not known
accurately enough to establish eta in MSBR spectrums to much better than
about 1% (the 2200-m/sec value used for n?3 was 2.292). The associated

uncertainty in breeding ratio is about #0.02 to 0.03, of which the major

8

i
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i

 

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£

 

 

 

 

 

63

fraction is due to the uncertainty in the effective thermal value (the
uncertainties associated with the 2200-m/sec value and the variation with
energy in different energy regions are independent of each other).

One of the most abundant materials in the MSBR is fluorine; although
its neutron-absorption cross section is low, its high concentration makes
it an important material relative to neutron absorptions. For fluorine,
the high-energy absorption cross sections are estimated to be uncertain
by about #30%. Also, the high-energy neutron reactions in beryllium are
uncertain by about *10 to 15%. Uncertainty in the gross cross section
for fission products (other than xenon and samarium, whose cross sections
are so high that fission yield is the important quantity) is estimated to
be about *30% for resonance-energy neutrons, and about *10% for thermal
neutrons. Uncertainties in other nuclide cross sections are estimated to
be about +10% or less.

Based on these uncertainties in cross-section values, the uncertainty
in breeding ratio is about #0.02 to 0.03 due to *33U, *0.004 due to *3°U,
+0.002 due to protactinium, #0.006 due to fluorine, +0.002 due to 14,
+0.002 due to beryllium, and *0.004 for gross fission products. Breaking
down these summed uncertainties into their independent uncertainties and
taking the square root of the sum of the squares of the independent un-
certainties gives a mean uncertainty of about #0.024 in breeding ratio.

This result illustrates that the uncertainty in breeding ratio can
have a significant effect on the MSBR fuel-yield rate; changing the
breeding ratio by *#0.024 would change the fuel-yield rate by about *50%.

‘In addition, the above analysis illustrates the relative importance of

the thermal value of 1°? in the MSER.

The cross sections actually used in the MSBR studies did not always
correspond to values présently considered to be the most probable. For
example, the high-energy neutron-absorption cross sections used for fluo-
rine are higher than present estimates; also, the graphite absorption
cross section (a 2200-m/sec value of 4 millibarns was used) did not allow
for burnout of trace impurities. Incorporating such changes would im-
prove the breeding ratio by about 0.005. In addition, the assumed be-
havior of fission products did not always correspond to present estimates

of their behavior in MSBR systems. In particular, it appears most probable

 
 

 

 

 

64

that molybdenum, technetium, and other members of group 3 in Table 3.5
will form intermetallic compounds with other fission products rather than
remain in solution as fluorides. Under such circumstances the elements
will most likely circulate as colloidal~-like metal suspensions (this is
indicated by MSRE experience with iron and chromium). In this event the
group 3 elements would be removed in fuel-recycle processing, so the
effect of the assumed behavior in the MSBR studies was correct.

There is a slight possibility that the group 3 fission products will
form metal carbides and remain indefinitely in the MSBR core. Such action
by a few percent of the group 3 nuclides could lead to a decrease in
breeding ratio of about 0.02 or more.

As shown in Table 3.5, it was assumed that the group 2 fission prod-
ucts would plate out on metal surfaces; at present it appears most likely
that these noble metals will remain in colloidal suspension and be removed
during fuel-recycle processing. The change in breeding ratio dque to the
above change leads to a decrease in breeding ratio of only 0.0CL.

It is important that xenon be removed from the MSBR core in order
to maintain breeder operation. Experience in the MSRE indicates that the
gas removal assumed in Table 3.5 is realistic.

Relative to group 5 fission products, it appears that at least cad-
mivum and tin will behave like the group 2 fission products and therefore
be removed in the fuel-recycle processing. The MSBR calculations assumed
that these fission products would be removed through salt discard alone.
Changing the behavior of fhis group to that indicated above would increase
the breeding ratio by no more than 0.003.

Although not discussed previously, it was assumed that 237Np would
be removed‘during fuel reprocessing. It appears that this removal can be
accomplished by proper operation of the absorber beds. If not removed,
the accumulation of 237Np in the fuel stream would decrease the breeding
ratio by about 0.01.

The fuel-processing losses were assumed to be 0.1% per pass through
the fluoride volatility process, and this loss is consistent with experi-
mental results. Doubling the losses would decrease the breeding ratio
by about 0.006.
 

 

 

»

"

)\

N

65

The nuclear calculations were made with the assumption that all
nuclides in the reactor were at their equilibrium concentrations. When
starting with 235U as the initial fuel, there will be a period of opera-
tion during which nonequilibrium conditions apply. To check the adequacy
of the assumption used, the operating time required to approach equilib-
rium concentrations with 43°U startup was examined. It was found that
233 and 235U were within 95% of their equilibrium concentrations in less
than two years, 234U’was within 95% after eight years, while *2°U was
within 80% after ten years. Since 236U buildup is detrimental, startup
with 435U fueling will lower the breeding ratio. However, the net effect
of 3%y startup is equivalent to increasing the MSBR specific fissile in-
ventory by 10 to 15% and considering the equilibrium breeding ratio to
apply. This is due to the higher critical mass with 23°U fueling and its
decrease with time as the bred fuel is recycled; this keeps the effective
fuel "production" rate close to that associated with equilibrium condi-
tions, after the first year of MSBR operation.

In summary, although there are sufficiently large uncertainties in
neutron-cross-section values and in the behavior of fission products to
significantly influence nuclear performance, the net nuclear performance
presented in Section 3.5.3 appears consistent with present information
based on equilibrium fueling of the reactor. Initial fueling with 235U,
rather than having equilibrium fueling conditions, will tend to be equiva-
lent to a slightly higher specific fissile inventory and a fuel production

corresponding to equilibrium conditions.

3.6 0Off-Gas System

Xenon and krypton are stripped from the fuel salt in the reactor cir-
culating system by sparging with an inert gas, such as helium. Since a
xenon-removal cycle time of about 1 min is required to maintain the xenon
poisoning at a satisfactorily low level, an in-line sparging system is
provided. The sparging gas is introduced at the bottom of the primary
heat exchangers to provide some circulation of the salt in event of pump
failure. This gas and the fission-product gases are withdrawn in a full-

flow gas separator in the pipe between the heat exchanger and the reactor.

 
 

 

 

66

The flowsheet for the off-gas system is shown in Fig. 3.19. As
mentioned above, xenon, krypton, and other fission-product gases are
sparged from the fuel-salt circulating system; these gases are removed
from the loop in a full-flow centrifugal separator located in the dis-
charge of each heat exchanger, with each loop unit discharging about 50
gpm of salt and about 4 scfm of gas.* A jet pump is used to aspirate
the fuel-salt-gas stream from the separator; the pump discharges into
the salt storage volume above the pump bowl and circulates the helium
carrier gas. After passing through the salt storage volume, the carrier
gas enters a 1000-ft> decay tank, which is cooled by evaporation of water
(similar cooling is used in the MSRE drain tanks®?). The gases then pass
through water-cooled charcoal beds, where xenon is retained for 48 hr,
and reenter the fuel system at the bottom of the primary heat exchanger.
In addition to removing the l35Xe, this system of circulation effectively
transfers a large fraction of thé other gaseous fission products to areas
where the decay heat can be removed more readily.

About 0.1 scfm of the gas stream leaving the charcoal beds (or 0.4
scfm total for the four fuel-salt circulating loops) passes through other
charcoal beds and then through a molecular sieve (operated at liquid
nitrogen temperature) to remove 99% or more of the 85Ky and other gaseous
products. The effluent helium can be recycled into the system or passed
through filters, diluted, and discharged into an off-gas stack. The
molecular sieves can be regenerated and the radioactive gases driven off
can be sent to storage tanks.

A helium system also provides cover gas for blanket-salt pump bowls,
drain tanks, fuel-handling and -processing systems, etc. The cover gas
discharged from these systems passes through charcoal adsorbers and ab-
solute filters prior to dilution with air and disposal through the off-

gas stack.

 

*¥The full-flow gas separators have been studied only in laboratory-
size equipment but are considered to be within the range of present tech-
nology. The MSBR loop installation requires 15 small separators arranged
in the annulus between the 18- and 24-in. concentric pipes. These small
separators would be capable of removing essentially all bubbles larger
than 0.0l in. in diameter.

"
 

 

Cw. OUTLET

    

FUEL PUMP & ¢

CW INLET

DECAY TAN®

48 hri HOLDUP XENON WATER
COOLED CHARCOAL ADSORBERS

Fig. 3.19.

Off-Gas System Flowsheet.

ORNL DWG 66-7135

STORAGE FOR
RANE GASES

L9

 

 
 

 

 

68

3.7 Heat Exchangers

The system heat exchangers consist of thé primary heat exchangers
used to transfer heat from the fuel and blanket salts to the coolant salt
and the boiler-superheaters and reheaters that transfer heat from the
coolant salt to the supercritical fluid in the.steamrpower system. Also
included, although more closely associated with the steam system than the

salt systems, are the reheat steam preheaters.

3.7.1 Fuel-Salt Heat Exchangers

Four shell-and-tube two-pass vertical heat exchangers transfer heat
from the fuel salt in the tubes to the coolant salt circulated through
the shell. The conceptual design is shown in Fig. 3.20, and the perti-
nent data are listed in Table 3.11.

Each exchanger has a capacity of about 528 Mw(th) and is about 5.5
ft in diameter and 18.5 ft high, including the bowl of the circulating

pump, which is an integral part of the heat exchanger shell. Shell, tube,‘

and tube sheets are fabricated of Hastelloy N.

The reactor fuel salt enters the heat exchanger from the 18-in.-
diam inner passage of the concentric pipes connecting the reactor and
exchanger. In the heat exchanger, the fuel flows downward through the
annular, outer rows of tubes at a velocity of 11.3 fps. In each unit
there are 4167 of these 0.375-in.-0D tubes on a 0.75-in. pitch. Upon
reaching the bottom head the salt reverses direction and moves upward at
about 13 fps through a center bank of 0.375-in.-diam tubes. There are
3624 of these tubes on a 0.625-in. pitch. Thus each fuel-salt primary
heat exchanger has 7791 tubes and about 9665 ft? of effective surface
area.

The coolant salt enters the heat exchanger at the top and flows
dowmward, countercurrent to the flow of fuel salt. It initially flows
through the center section of the exchanger, and on reaching the bottom
of the shell it turns upward to flow through the tubes in the annular
section and leave the exchanger through an annular collecting ring at
the top.

i
 

 

 

 

 

 

 

 

 

 

 

      
    
    

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

3 69
| - ORNL DWG 65-12379
i
_ g
/
FUEL SALT DUMP TANK ; . FUEL SALT PUMP
| T . : : s >
FUEL SALT FROM NY/ % , /// AN
REACTOR — ALl
(¥
1%5‘,‘
COOLANT SALT
L. : FROM STEAM
- HEAT EXCHANGER
| -
FUEL SALT TO ——/‘
REACTOR
COOLANT SALT
) TO BLANKET
[ - HEAT EXCHANGER
DOUGHNUT murruaZ T T DISK BAFFLE
| —————————— LONGITUDINAL BAFFLE
LONGITUDINAL R 7—7%55
BAFFLE i
- :::?‘“TlE RODS & SPACERS
CORE TUBE | ANNULAR
SHEET —\ : 1 /"TUBE SHEET
EXPANSION
BELLOWS SHROUD
COOLANT SALT DRAIN FUEL DRAIN
|- Fig. 3.20. Fuel-Salt Heat Exchanger.
|
; ]
i
{ »
|

 
 

70

 

Table 3.11. Fuel-Salt Heat Exchanger Design Data

 

Type Shell-and~tube two-pass
vertical exchanger with .
disk and doughnut baffles

 

 

 

Number required 4
Rate of heat transfer, each,
Mwr 528
Btu/hr 1.80 x 10°
Shell-side conditions
Cold fluid Coolant salt
Entrance temperature, °F 850
Exit temperature, °F 1111
Entrance pressure, psi 80
Exit pressure, psi 29
Pressure drop across exchanger, psi 51
Mass flow rate, 1b/hr 1.68 x 107
Tube~side conditions
Hot fluid Fuel salt
Entrance temperature, °F 1300
Exit temperature, °F 1000
Entrance pressure, psi 9%
Exit pressure, psi 10 (pump suction)
Pressure drop across exchanger, psi 86
Mass flow rate, 1b/hr 1.08 x 107
Mass velocity, l'b/hr-ft2
Center section 5.95 x 10°
Annular section 5.18 x 10°
Velocity, fps
Center section 13.0
Annular section 11.3
Tube material Hastelloy N
Tube 0D, in. 0.375
Tube thickness, in. 0.035
Tube length, tube sheet to tube
sheet, ft
Center section 13.7
Annular section 11.7

Shell material
Shell thickness, in.

Hastelloy N
0.5

o
 

 

i

71

Table 3.11 (continued)

 

Shell ID, in.

Center section
Annular section

Tube sheet material
Tube sheet thickness, in.

Top annular section
Bottom annular section
Top and bottom center section

Number of tubes

Center section
Annular section

Pitch of tubes, in.

Center section
Annular section

Total heat transfer area per
exchanger, ft2

Center section
Annular section
Total

Basis for area calculation
Type of baffle
Number of baffles

40.2
66.5

Hastelloy N

=W
O~ 0
oo

3624
4167

0.625
0.750

4875
4790
9665

Tube outside diameter

Disk and doughnut

Center section 5
Annular section 2
Baffle spacing, in.
Center section 27.4
Annular section 21
Disk 0D, in.
Center section 30.6
~ Annular section 55.8
Doughhut 1D,
* Center section - 25.0
Annular section _ 51.0
Overall heat transfer coefficient, U, - 1110

Btu/hr-ft2

 
 

 

 

72

Table 3.11 (continued)

 

Maximum stress intensity,® psi

Tube _
Calculated Py = 413; (Pp + Q) = 12,000
Allowable Pnp = Sy = 4600; (Py + Q) =
38y = 13,800
Shell '
Calculated Py = 6160; (Pp + Q) = 21,600
Allowsble Py = = 12,000; (Pp + Q) =
35Sy = 36,000
Maximum tube sheet stress, psi
Calculated \ | 10,750
Allowable 10,750

 

®The symbols are those of Section 3 of the ASME Boiler and Pressure
Vessel Code, with

Py = primary membrane stress intensity,
Q = secondary stress intensity,
Sy = allowable stress intensity.

The general configuration and arrangement of the exchanger'were
largely dictated by the design requirement that the fuel-salt circulating
system have a minimum fuel inventory consistent with practical design
considerations. Associated factors were permissible stress values and
the ability to remove afterheat and drain the core. The heat exchanger
calculations were concerned primarily with determining the lengths and
number of tubes, the tube pitch, the number of baffles, the baffle spac-
ing, etc., which would best suit the specified conditions. A computer
program was developed for this optimization work. The program and the
details of the calculations for all the MSBR heat exchangers are reported
elsewhere.?3 |

To distribute coolant-salt flow on the shell side of the exchanger,
disk and doughnut baffles are used in the center section. In the annular
region there are two baffles, one extending inward from the exterior shell
and one extending outward from the barrier that surrounds the core sec-

tion. These baffles improve the shell-side heat transfer coefficient;
 

 

 

h

73

however, no baffles are used at the top of the annular section, because
the hottest fuel fluid enters here and an improved heat transfer coeffi-
cient would result in an excessive temperature drop across the tube wall.
Also, a baffle is located near each tube sheet to partially insulate it
and thereby reduce the temperature drop across the sheet. Fuel or cool-
ant salt can be drained from the bottom of the primary exchangers through
the concentric drain lines indicated din Fig. 3.20.

The stresses that tend to be deveiqped in the heat exchanger due to
the temperature differences between the shell and the upflow and downflow
tubes are relieved in the design concept by a bellows expansion joint at
the lower tube sheet. The stresses in the present design are given in
Table 3.11.%

3.7.2 Blanket-Salt Heat Exchangers

The four shell-and-tube vertical heat exchangers used to transfer
heat from the blanket salt to the coolant salt are very similar to the
fuel-salt exchangers, but they only have a capacity of 27.8 Mw(th) each.
They are illustrated in Fig. 3.21. Pertinent design data are given in
Table 3.12.

The coolant salt passes through the fuel-salt heat exchangers and
then through the blanket exchangers, in series, entering the latter at
about 1111°F and leaving at 1125°F. Since the flow rate is relatively
high and the temperature change is small, the exchangers are designed for
a single shell-side pass of the coolant salt. The blanket salt in the
0.375-in.-0D tubes makes two passes, however, moving downward at about
10.5 fps in the outer annular section and upward through the inner bank

to the pump suction.

 

- *Other exchanger designs were also studied that utilized bent tubes
rather than the bellows to absorb the differential expansion; these ex-
changers had the pump discharging fuel from the reactor into the heat
exchanger so that the point of highest pressure in the system was the
exchanger rather than the reactor. The results of these studies are
presented in Section 4.4. 1In general, it is believed that the present
exchanger design can be improved to minimize engineering development
problems but that the estimated capital costs of heat exchangers based
on the present design are representative of developed heat exchanger
costs.

 
 

 

 

 

74

ORNL DWG 65-123 80

     
     

 

_ BLANKET SALY
FROM REACTOR

 

BLANKET SALT
TO REACTOR

 

 

 

 

 

DISK BAFFLE
ODOUGHNUT BAFFLE

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 
  

il
I

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

   

 

 

 

 

TO BLANKET DRAIN TANKS

Fig. 3.21. Blanket-Salt Heat Exchanger.

BLANKET SALT PUMP

COOLANT SALT
TO STEAM _HEAT

EXCHANGER

COOLANT SALT

FROM PRIMARY
EXCHANGER

<
 

 

 

 

 

»

75

Table 3.12. Blanket-Salt Heat Exchanger Data

 

Number required

~Rate of heat transfer per unit, Mw

Btu/hr
Shell-side conditions |

Cold fluid

Entrance temperature, °F

Exit temperature, °F

Entrance pressure, psi

Exit pressure, psi

Pressure drop across exchanger, psi
Mass flow rate, lb/hr

Tube-side conditions

Hot fluid

Entrance temperature, °F

Exit temperature, °F

Entrance pressure, psi

Exit pressure, psi

Pressure drop across exchanger, psi
Mass flow rate, lb/hr

Mass velocity, lb/hr-ft?

Velocity, fps

Tube material

Tube 0D, in.

Tube thickness, in.

Tube length, tube sheet to tube sheet, £t
Shell material

Shell thickness, in.

Shell ID, in. |

Tube sheet material

Tube sheet thickness, in.
Number of tubes

Pitch of tubes, in.

Total heat transfer area, ft°

Basis for area claculation

Shell-and-tube one-shell-
pass two-tube-pass exchan-
ger, with disk and doughnut
baffles

4
27.8
9.47 X 107

Coolant salt
1111

1125

27

9

18

1.68 X 107

Blanket salt
1250

1150

100

10

90

4.3 X 108
10.5 X 10°
10.5

Hastelloy N
0.375

0.035

8.25
Hastelloy N
0.25

36.5
Hastelloy N
1.0

1641 (~820 per pass)

0.81
1330

Outside diameter

 
 

 

76

Table 3.12 (continued)

 

Type of baffle : ’ Disk and doughnut
Number of baffles 3
Baffle spacing, in. . 24.8
Disk 0D, in. 26.5
Doughnut ID, in. 23
Overall heat transfer coefficient, U, 1020
Btu/hr«ft2
Maximum stress intensity,® psi
Calculated Py = 410; (P, + Q) = 7840
Allowable Pp = Syp = 6500; (Py + Q) =
- 35Sy = 19,500
Shell
Calculated ~ Pn = 1660; (Pp + Q) = 11,140
Allowable Pp = Sm = 12,000; (Pp + Q) =
38m = 36,000
Maximum tube sheet stress, psi
Calculated 2220
Allowable 5900 at 1200°F

 

®The symbols are those of Section 3 of the ASME Boiler and Pressure
Vessel Code, with
Pn = primary membrane stress intensity,
Q@ = secondary stress intensity,
Sm = allowable stress intensity.

Straight tubes with two tube sheets are used rather than U-tubes in
order to permit drainage of the blanket salt. Disk and doughnut baffles
are used to improve the shell-side heat transfer coefficient and to pro-
vide the necessary tube support. Baffles on the shell side of the tube
sheets reduce the temperature difference across the sheets to keep thermal
stresses within tolerable limits. Calculations show that the relatively
low pressures and small temperature differences produce stresses that are

well within the allowable range.?23

im
 

 

 

 

#

77

3.7.3 Boiler-Superheaters

Sixteen vertical U-tube U-shell heat exchangers-are used to transfer
the heat from the 1125°F coolant salt to the 700°F feedwater to generate
steam at 1000°F and about 3600 psia. Four of these exchangers are in
each of the coolant-salt circulating circuits and are supplied by a
variable-speed coolant-salt pump (adjustment of the pump speed permits
control of the outlet steam temperature). Each exchanger has a capacity
of about 121 Mw(th) and has a U-shaped cylindrical shell about 18 in. in
diameter; each vertical leg stands about 34 ft high, including the spheri-
cal head. The tubes and shell are fabricated of Hastelloy N. The unit
is shown in Fig. 3.22. Pertinent design data are given in Table 3.13.

Because of marked changes in the physical properties of water as the
temperature increases above the critical point (at supercritical pres-
sures), heat transfer calculations for this particular exchanger were
made on the basis of a detailed spatial analysis with a computer program.23
The calculations established the optimum number of tubes, tube length,
nunber of baffles, and baffle spacing, in terms of specified design cri-
teria. The results indicated that the optimum design was an exchanger
with a long, slim shell and relatively wide baffle spacing. The spacing
was greatest in the central portion of the exchanger where the temperature
difference between the bulk fluids is high.

The 3600-psi fluid pressure on the inside of the tubes dictates that
the heads and tube sheets be carefully designed. The relatively small
diameter of 18 in. selected for the shell and the spherical heads on the
ends of the exchangers allows the stresses to be kept within permissible
limits. A baffle on the shell side of each tube sheet provides a stag-
nant salt layer that helps-to reduce gtresses in the sheet due to tempera-
ture gradients. L e |

The coolant salt can be cofipletely drained from the shell. The
water can be partially removed from the tubes by gas pressurization, or
by flushing, but completeldrainability was not considered a mandatory

design requirement.

 
 

 

 

78 |

ORNL DWG 65-12383

  
 

SUPER CRITICAL

SUPER CRITICAL
FLUD OUTLET

_ FLUI T
—qFLUID WLE

COOLANT SALT _

COOLANT SALT
INLET

OUTLET

 

 

 

| TIE ROD & SPACER

 

 

 

 

 

 

 

 

 

 

AFFLES

 

 

 

 

 

 

Fig. 3.22. Boiler-Superheater.

i
 

M o b e el 88 b A1 L 50w L AL e koL s

 

79

Table 3.13. Boiler-Superheater Design Data

 

Type

Number required

Rate of heat transfer, Mw
Btu/hr

Shell-side conditions

Hot fluid

Entrance temperature, °F

Exit temperature, °F

Entrance pressure, psi

Exit pressure, psi

Pressure drop across exchanger, psi
Mass flow rate, lb/hr

Tube~side conditions

Cold fluid

Entrance temperature, °F

Exit temperature, °F

Entrance pressure, psi

Exit pressure, psi

Pressure drop across exchanger, psi
Mass flow rate, lb/hr

Mass velocity, 1b/hr.ft?

Tube material

Tube OD, in.

Tube thickness, in.

Tube length, tube sheet to tube sheet, ft
Shell material |
Shell thickness, in.

Shell ID, in.

Tube sheet material

Tube sheet thickness,.in.
Number of-tubes

Pitchrof tubes, in.

Total heat transfer area, £t2
Basis foraarea calculation
Type of baffle

Number of baffles

' U-tube U-shell exchanger

with crossflow baffles
16

121
4,13 X 108

Coolant salt
1125

850

150

92

58

3.66 X 106

Supercritical fluid
700

1000

3770

3600

166

6.33 X 10°

2.78 X 106

Hastelloy N
0.50
0.077
63.8
Hastelloy N
0.375
18.2

4Hastelloy N

475
349
0.875

2915

Outside surface
Crossflow
9

 
 

 

80

Table 3.13 (continued)

 

Baffle spacing - Variable
Overall heat transfer coefficient, U, 1030
Btu/hr - £t2 |
Maximum stress intensity,® psi
Tube
Calculated P, = 6750; (P, + Q) = 40,700
’ ’ 48,000
Shell :
Calculated Py = 3780; (P, + Q) = 8540
Allowable : Pp =-10,500; ?Bm + Q)allow =
31,500
Maximum tube sheet stress, psi
Calculated . <16,600
Allowable - - 16,600

 

®The symbols are those of Section 3 of the ASME Boiler ahd Pressure
Vessel Code, with

Py = primary membrane stress intensity,
Q = secondary stress intensity,
Sm = allowable stress intensity.

3.7.4 Steam Reheaters

 

Eight shell-and-tube heat exchangers transfer heat from the coolant
salt to the high-pressure-turbine exhaust steam (~57O psia) and raise its
temperature to 1000°F. The steam enters the exchanger at about 650°F,
having been heated from the 552°F exhaust temperature in a preheater
described below. There are two reheaters to each coolant-salt circu-
lating loop, each pair being supplied by a variable-speed coolant-salt
pump in an arrangement that permits control of the outlet steam tempera-
ture. The general arrangement of the reheaters is shown in Fig. 3.23,
and design data are givén in Table 3.14. |

Each of the eight units has a capacity of about 36 Mw(th); the cool-
ant salt enters a unit at 1125°F and leaves at 850°F. Straight vertical
shells about 28 in. in diameter and 24 ft long are used.¥ Both shell and

 

*The straight shell occupies less cell volume than a U-tube U-shell
design and requires slightly less coolant-salt inventory.

ra
 

 

 

 

81

ORNL DWG 66-7118

STEAM OUTLET

   
   

TUBE SHEET

TIE ROD & SPACER

 

TUBULAR
SHELL

gl!&g' TUBES

DISC BAFFLE

 

 

DOUGHNUT
BAFFLE

 

 

 

SALT OUTLET

 

 

 

 

 

 

 

 

 

 

 

 

 

 

INSULATION BAFF

 

ORAIN LINE

STEAM INLET

Fig. 3.23. ©Steam Reheater.

 
 

 

 

82

Table 3.14.

Steam Reheater Design Data

 

Type

Number required

Rate of heat transfer per unit, Mw- :
Btu/hr
Shell-side conditions

Hot fluid A

Entrance temperature, °F

Exit temperature, °F

Entrance pressure, psi

Exit pressure, psi

Pressure drop across exchanger, psi
Mass flow rate, 1lb/hr 1

Mass velocity, lb/hr- P12

Tube-side conditions

Cold fluid

Entrance temperature, °F

Exit temperature, °F

Entrance pressure, psi

Exit pressure, psi

Pressure drop across exchanger, psi
Mass flow rate, lb/hr

Mass velocity, 1b/hr- 12

Velocity, fps

Tube material

Tube OD, in.

Tube thickness, in.

Tube length, tube sheet to tube sheet, ft
Shell material

Shell thickness, in.

Shell ID, in.

Tube sheet material

Tube sheet thickness, in.
Number of tubes

Pitch of tubes, in.

Total heat transfer area, ft?

Basis for area calculation

Straight tube and shell ex-
changer with disk and dough-

nut baffles

8

36.2
1.24 X 108

- Coolant salt

1125

850

106

90

16

1.1 X 106
1.44 X 10°

Steam
650

1000

570

557

13

6.3 X 10°
4.0 X 10°
147

Hastelloy N
0.75

0.035

22.9
Hastelloy N
0.5

28
Hastelloy N
4.75

628

1.0

2830
Outside of tubes
 

 

83

Table 3.14 (continued)

 

Type of baffle Disk and doughnut

Number of baffles 14 and 15
Baffle spacing, in. 8.75
Disk OD, in. 23.2
Doughnut ID, in. 16.0
Overall heat transfer coefficient, U, 275
Btu/hr -ft?
Maximum stress intensity,® psi
Tube
Calculated P, = 5240; (P, + Q) = 15,100
Allowable Pp = 14,500; (Py + Q) =
| 43,500
Shell
Calculated Py = 4350; (Pp + Q) = 14,800
Allowable P, = 10,600; (P, + Q) =
31,800
Maximum tube sheet stress, psi
Calculated 9,600
Allowable : 9,600

 

e symbols are those of Section 3 of the ASME.Boiler and Pressure
Vessel Code, with
Pp = primary membrane stress intensity,
Q = secondary stress intensity,
Sm = allowable stress intensity.

tubes are fabricated of Hastelloy N. Disk- and:dbughnut—type baffles
support the tubes at closé intervals to prevent.excessive vibration.
Baffles on the shell sides of the tube sheets provide a stagnant layer
of coolant salt to reduce thermal stresses in the sheet. A special drain
pipe at the bottom provides for drainage of the coolant salt.

Analyses of the stresses'ihdicated that the values were within per-

missible limits.
3.7.5 Reheat~Steam Preheatérs

Throttle steam at 3500 psia and 1000°F is used to heat the high-

pressure turbine exhaust from about 552 to 650°F before it enters the

 
 

84

reheaters. (Heat transfer studies indicate that no freezing of the cool-
ant salt takes place in the reheaters if steam enters at 650 rather than
700°F, due to the low value of the steam-side heat transfer coefficient.)
Use of this preheater permitted the adoption of the TVA Bull Run Sfieam
Plant operating conditions without significant changes affecting costs
or performance. This factor was given priority over designing for maxi-
mum thermodynamic efficiency and minimum cost, * since the difference
would be swall and have little effect on the findings of this study.
The design concept for the reheat-steam preheaters is shown in Fig.

3.24, and the design data are listed in Table 3.15. ZEight preheaters

 

 

*A thermodynamically more efficient arrangement would be to exhaust

the steam from the high-pressure turbine at 650°F rather than 552°F,
i which would also have the advantage of eliminating the preheating equip-
! ment (estimated cost, $275,000).

. ORNL DWG 65-12382

   

SUPER CRITICAL

SUPER CRITICAL
FLUID INLET

FLUID OUTLET

   
     
   

~/:

TUBE SHEET

BAFFLE
” STEAM QUTLET

| BY+PASS RING

STEAM INLET = ’

 

J———TIE ROD & BPACER

? TUBES

/nv-mss RING

 

 

 

Fig. 3.24. Reheat-Steam Preheater.

 

e
 

 

 

 

 

85

Table 3.15.

Reheat~Steam Preheater Design Data

 

Type

Number required

Rate of heat transfer, Mw
Btu/hr

Shell~side conditions

Cold fluid

Entrance temperature, °F

Exit temperature, °F

Entrance pressure, psi

Exit pressure, psi

Pressure drop across exchanger, psi
Mass flow rate, lb/hr

Mass velocity, 1b/hr.ft?

Tube-side conditions

Hot fluid

Entrance temperature, °F

Exit temperature, °F

Entrance pressure, psi

Exit pressure, psi

Pressure drop across exchanger, psi
Mass flow rate, 1b/hr

Mass velocity, 1b/hr-ft?

Velocity, fps

Tube material

Tube 0D, in.

Tube thickness, in.

Tube length, tube sheet to tube sheet, ft
Shell material

Shell thickness, in.

Shell ID, in.

Tube sheet material

Tube sheet thickness, in.
Number of tubes

Pitch of tubes, in.

Total heat transfer area, ft°

Basis for area calculation

One-tube-pass one-shell-pass
U-tube U-shell exchanger
with no baffles

8

12.3
4,21 X 107

Steam
552
650
595.4
520.0

Supercritical water
1000

869

3600

3544

56

3.68 X 107

1.87 X 108

93.5

Croloy
0.375
0.065
13.2
Croloy
7/16
20.2
Croloy
6.5
603
0.75
781

Tube outside diameter

 
 

 

 

 

86

Table 3.15 (continued)

 

Overall heat transfer coefficient, U, 162
Btu/hr-f£t?
Maximum stress intensity,2 psi
Tube .
Calculated Pp = 10,500; (Pp + Q) = 15,900
Allowable Pp = Sp = 10,500 at 940°F;
TPm + Q) = 38y = 31,500
Shell
Calculated P, = 14,400; (P, + Q) = 33,100
Allowable P, = Sp = 15,000 at 650°F;
ey + Q) = 38, = 45,000
Maximum tube sheet stress, psi
Calculated 7800
Allowable 7800 at 1000°F

 

SThe symbols are those of Section 3 of the ASME Boiler and Pressure
Vessel Code, with

Py = primary membrane stress intensity,
Q = secondary stress intensity,
Sm = allowable stress intensity.

are used. This number was determined almost entirely by the selection
of reasonable dimensions for the units. The preheaters are part of the
steam power system and no biological shielding is required. They are
located in the portion of the plant assigned to the feedwater heaters.

Fach preheater has a capacity of about 12.3 Mw(th). Each vertical
leg of the U-shell is about 21 in. in diameter and the overall height is
about 15 ft, including the spherical heads. The tubes, tube sheets, and
heads contain the 3500-psia throttle steam and are designed for this_high
pressure and temperature. Selection of the U-shell rather than a divided
cylindrical shell permits use of small head diameters and reduces the
required tube-sheet and head thicknesses. Stress analyses indicate that
the stresses are within the allowable limits. Both tubes and shell are
fabricated of Croloy.

The flow in the preheaters is countercurrent and no baffles are
needed in the shell. The U-tube construction accommodates the thermal

expansion that occurs. The relatively high steam film resistance to heat

—
 

 

o

87

transfer on the shell side reduces the temperature gradient across the

tube wall to permissible levels.

3.8 BSalt-Circulating Pumps

Fach of the four separate salt-circulating circuits contains a fuel-
salt circulating pump, a blanket-salt pump, a coolant-salt pump for the
boiler superheaters, and a coolant-salt pump for the reheaters. The de-
sign data for these pumps are listed in Table 3.16. The pump designs
utilize the technology developed over the past 15 years, with present

pump capacities being extrapolated a factor of 10 for MSBR use.

Table 3.16. Salt Pump Dimensions and Performance Requirements

 

Reheat Superheat

 

Si:iim g;z:g;# Coolant Coolant
System System

Number of pumps 4 4 4 4
Temperature, 1,000 1,150 1,125 1,125
Flow, gpm (each) 11,000 2,200 2,200 14, 000
Head, ft 140 100 110 150
Speed, rpm 1,170 1,750 1,750 1,170
Impeller diameter, in. 24 13 13 24
Pump tank diameter, in. 36 60
Suction diameter, in. 18 8 8 18
Discharge diameter, in. 6 14
Nominal motor power, hp 1,250 500 200 1,250
Motor length, in. 92 72 37 92
Motor diameter, in. 64 40 29 64

 

A1l pumps are of centrifugal type, with a vertical shaft supported
at its lower end in a hydrodynamic journal bearing lubricated by molten
salt. The fuel- and blanket-salt pump bowls have diffuser vanes and are
an integral part of the pfimary heat exchanger vessels. The equipment
arrangements are illustrated in Figs. 3.20 and 3.21. Figure 3.25 shows
the fuel- and blanket-salt pumps and Fig. 3.26 shows details of the

‘coolant-salt pump.

 
 

 

 

88

ORNL-DWG 66-69T71

ELECTRIC
MOTOR

 

 

 

 

 

 

 

N

 

 

 

 

 

 

 

  

DIMENSIONS ARE LISTED

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

N TABLE 3.15
$— {
%g” 7
IR
v GAS puaee-—-,..f ~ / |
/)
1
TEN SALT .
JOURNAL BEARING |/
| /
]
DIFFUSER TR= = %EE'E{;&L&GQ&E%ORT 2
A
/1
/1
/
; /
Y
— 9
SUCTION g

 

 

Fig. 3.25. Tuel- and Blanket-Salt Pump.

A continuous purge of inert gas flows through each pump during opera-
tion. Thus purge gas enters the labyrinth annulus near the upper end of
the pump shaft. The labyrinth seals the motor cavity from the gas space
in the pump tank. The purge gas flow splits into two paths; one portion
flows upward in the annulus to keep lubricating vapors from entering the
pump tank, and the other flows downward to prevent the migration of radio-
active gases into the mofior cavity.

The pumps éonstitute a part‘of the primary containment of the reactor

fluid. As such, they would be constructed in accordance with the applicable

i
 

 

89

ORNL-DWG 66-6972

     

ELECTRIC
MOTCR

PURGE

       
    
 
 

PUMP TANK

SALT
GAS QUT JOURNAL BEARING

SELF ALIGNING

GAS VOLUME BEARING SUPPORT

DISCHARGE

DIMENSIONS ARE LISTED
IN TABLE 3.15

SUCTION

Fig. 3.26. Coolant-Salt Pump.

portions of the ASME codes, with proper allowances made for the thermal
strain fatigue that would accompany reactor startups, power cycles, and
radiation heating. The long shafts on the fuel- and blanket-salt pumps
permit the drive motors to be shielded from the reactor radiation and
temperature. The electric drive motors are located outside the biological
shielding, with the hermetiélcans around these motors serving as part of
the reactor containment vessel. A squirrel-cage induction motor is used,
with the ball bearings lubricated with radiation-resistant grease capable

of withstanding 3 X 10° rad. The electrical insulation also uses special

 
 

 

 

 

90

materials having a radiation tolerance up to 10° rad. Motor heat is re-
moved by circulating a coolant through coils inside the hermetically

sealed motor wvessel.

3.9 Steam-Power System

 

The thermal power of the MSBR is 2225 Mw(th), which provides a full-
load net electrical output of 1000 Mw(e) plus about 35 Mw(e) of power
for auxiliary equipment. Throttle steam conditions are 3500 psia and
1000°F/1000°F; these conditions are representative of modern steam-power -
plant practice and correspond to those employed in the recently completed
TVA Bull Run Steam-Electric Plant.

The steam-power system flowsheet is shown in Fig. 3.27, and the de-
sign and performance data are summarized in Table 3.17. Energy balances
were made in determining the thermodynamic performance of the system based
on a 700°F inlet feedwater tem;pera.ture.24 The flow rates and steam prop-
erties at various points in the system are shown on Fig. 3.27. The net
efficiency of the plant is about 45%.

The MSBR system has a conventional 1035-Mw(e) gross output cross-
compounded 3600/1800—rpm four-flow turbine-generator unit with 3500-psia
1000°F steam to the throttle and reheat to 1000°F after the high-pressure
turbine exhaust. The exhaust pressure at rated conditions is 1l.5-in. Hg
abs. Eight stages of feedwater heating are used, with extraction steam
taken from the high- and low-pressure turbines and also from three points
on the turbines used to drive the boiler feedwater pumps.

The feedwater leaves heater 4 at about 357°F and 200 psia; it is
raised to 3800 psia and 366°F by two turbine-driven centrifugal pumps.

The pumps have six stages, run at 5000 rpm, and deliver 8100 gpm against
a head of 9380 ft. The drive turbines, which have eight stages, are sup-
plied with throttle steam at 1069 psia and 700°F, and they exhaust at
about 77 psia. There is one drive turbine per pump, and no standby pump-
ing capacity is provided.

The MSBR steam-power system differs from the TVA Bull Run plant in
having higher feedwater and reheat-steam temperatures. The temperature of -

O

the feedwater entering the boiler-superheaters was governed by the estimated

ik
 

 

-

"

"

ORNL DWG 66-7125

91

 

 

 

 

 

 

 

 

o . . e et s S

 

 

 

 

 

 

 

 

 

 

   
       

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

NT
3.2 e
1054.9 Wwe
29.4 N
2EES Mt
sTe%
TAN Dk

1000 hwe
i

QROSS EFF OF CYGLE
NET EFF. OF CYCLE

REACTOR HEAT WRUT

NET HEAT RATE

OF PREDNUNL-BO0ETER FUMPS
OUTAUT
8 PUNPE

SUMMARY OF PERFORMANCE AT RATED LOAD
NET OUTRUT
BLEC. POWER FOR AUX.
Lo

et ——— du e m e, — - —————— - —————

e

 

 

 

 

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. ;o
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1 “ |
J-E.lb_lllti. -.ll-l
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P i
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%l _w R
oI 1 5
—M —w HII_ llllllll ID-
| T
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b L —sem
P — e — =
- e
(I
“ b - ———————n
| .
| mmw
| | w [
Ly
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&
l 2 g ww
£ 1

 

 

- SB00P- 368,17 MIBH

 

 

 

 

 

 

 

 

 

Steam System Flowsheet for 700°F Feedwater.

Fig. 3.27.

 
 

 

 

92

Table 3.17. MSBR Steam-Power System Design and
Performance Data with 700°F Feedwater

 

General performance _
Reactor heat input, Mw 2225

Net electrical output, Mw 1000
Gross electrical generation, Mw 1034.9
Station auxiliary load, Mw 25.7
Boiler-feedwater pressure-booster pump load, Mw 9.2
Boiler-feedwater pump steam-turbine power output, — 29.3
Mw (mechanical)
Flow to turbine throttle, lb/hr 7.15 X 106
Flow from superheater. 1b/hr 10.1 X 10°
Gross efficiency, % (1034.9 + 29.3)/2225 47.8
Gross heat rate, Btu/kwhr 7136
Net efficiency, % 4,9
Net heat rate, Btu/kwhr - 7601
Boiler-superheaters
Number of units 16
Total duty, Mw(th) 1932
Total steam capacity, 1b/hr 10.1 X 106
Temperature of inlet feedwater, °F 700
Enthalpy of inlet feedwater, Btu/lb 769
Pressure of inlet feedwater, psia 3770
Temperature of outlet steam, °F 1000
Pressure of outlet steam, psia ~3600
Enthalpy of outlet steam, Btu/lb 1424
Temperature of inlet coolant salt, °F 1125
Temperature of outlet coolant salt, °F 850
Average specific heat of coolant salt Btu/1b*°F 0.41
Total coolant-salt flow, lb/hr 58.5 X 106
cfs 130
gpm 58,300
Coolant-salt pressure drop, inlet to outlet, psi ~60
Steam reheaters
Number of units | 8
Total duty, Mw(th) 294
Total steam capacity, 1b/hr 5.13 X 108
Temperature of inlet steam, °F 650
Pressure of inlet steam, psia ~570
Enthalpy of inlet steam, Btu/lb 1324
Temperature of outlet steam, °F 1000
Pressure of outlet steam, psia 557
Enthalpy of outlet steam, Btu/lb 1518
Temperature of inlet coolant salt, °F | 1125
Temperature of outlet coolant salt, °F 850

Average specific heat of coolant salt, Btu/lb °F 0.41

h
 

 

 

 

93

Table 3.17 (continued)

 

Steam reheaters (continued)

Total coolant salt flow, lb/hr
cfs
gpm
Coolant-salt pressure drop, inlet to outlet, psi

Reheat-steam preheaters

Number of units
Total duty, Mw(th)
Total heated steam capacity, lb/hr
Temperature of heated steam, °F
Inlet
Outlet
Pressure of heated steam, psia
Inlet
Outlet
Enthalpy of heated steam, Btu/lb
Inlet
Outlet
Total heating steam, lb/hr
Temperature of heating steam, °F
Inlet
Outlet
Pressure of heating steam, psia
Inlet
Outlet

Boiler-feedwater pumps

Number of units
Centrifugal pump
Number of stages '
Feedwater flow rate, total, 1b/hr
Required capacity, gpm
Head, approximate, ft
Speed, rpm
Water inlet temperature, °F
Water inlet enthalpy, Btu/1b
Water inlet specific volume, ft3/1b
Steam-turbine drive
Power required at rated flow, Mw (each)
Power, nominal hp (each) |
Throttle steam conditions, psia/°F
Throttle flow, 1b/hr (each)
Exhaust pressure, approximate, psia
Number of stages
Number of extraction points

8.88 x 10°
19.7
8860
~117

8
100
5,13 X 10°

552
650

595
590

1257
1324
2.92 X 108

1000
869

3600
3544

2

6
7.15 X 10°
8,060
9,380
5,000

358

330
~0,0181

14.7

~ 20,000

1070/700
414,000
77

8

3

 
 

 

 

94

Table 3.17 (continued)

 

Boiler-feedwater pressure-booster pumps

Number of units . 2
Centrifugal pump
Feedwater flow rate, total, 1lb/hr 10.1 X 10°
Required capacity, gpm (each) 9,500
Head, approximate, ft 1413
Water inlet temperature, °F 695°F
Water inlet pressure, psia ~3,500
Water inlet specific volume, ft2/1b ~0.0302
Water outlet temperature, °F ~700
Electric-motor drive ,
Power required at rated flow, Mw(e) (each) 4.6
Power, nominal hp (each) 6,150

 

liquidus temperature of the coolant salt; it was decided that the coolant
salt should not be permitted to freeze, so the feedwater must enter the
boilers at 700°F or higher. In the reheaters, however, the heat-transfer
resistance of the steam film is high, so the steam can enter at 650°F.
The feedwater leaves the conventional eight stages of regenerative
feedwater heating at about 551°F, the same as in the TVA Bull Run steam-
power cycle. The steam leaves the high-pressure turbine at about 552°F
and is heated to 650°F in a shell-and-tube type exchanger (described
in Sect. 3.7.5), with supercritical fluid at 3515 psia and 1000°F. The
high-pressure heating steam leaves the heat exchanger near 866°F and 3500
psia and is directly mixed in a "mixing tee" with the 550°F feedwater to
raise its temperature to about 695°F. The mixture is then boosted to
boiler-superheater pressure by motor-driven pumps, and the pumping effort
raises the pumped water temperature to the requisite 700°F. The density
of the supercritical fluid pumped by the booster pumps is about 34 1b/ft3,
and very little compressive work [~9.2 Mw(e)] is involved in raising the
fluid pressure. The pumps employed are similar tbrthose used for forced-
convection flow in supercritical-pressure steam generators. Each of the

two booster pumps has a rating of about 20,000 gpm and 6200 hp.

5
 

 

95

3.10 Other Design Considerations

 

3.10.1 Piping and Pipe Stresses

The stresses in the salt and steam piping were studied briefly to
determine whether the reactor and turbine plant layouts contained grossly
impractical arrangements. The calculations were made with the MEC-21/7094

25 In these estimates, it was assumed that the centers of the re-

code.
actor and the turbines were fixed and the rest of the system was allowed
to move in accordance with the thermal-expansion forces. Stresses were
examined at about 150 points with particular emphasis on the locations

of suspected high stresses.

The sizes of the main piping in the steam-power system are shown in
Table 3.18. The assumed vélocities, materials, and other conditions are
also given. The maximum calculated stress in piping fabricated of Hastel-
loy N was found to be about 10,000 psi, which is about a factor of 3 less
than that allowable based on ASME Code requirements. The Croloy steam
piping has a maximum calculated stress of 2200 psi, which is well within
the allowable value.

One case was calculated in which the coolant-salt pumps were re-
strained in the direction transverse to the main coolant-salt pipe rum.
The maximum stress in the Hastelloy N piping in this case was about
22,000 psi; the maximum stress in the Croloy steam piping was essentially
the same as before. BSince the wvertical deflections at the pump location
are apparently small, it appears that use of vertical and transverse re-

straints will not cause thermal-expansion effects to overstress the

piping.
3.10.2 Maintenance

The MSBR equipment was designed and arranged so that inspection,
maintenance, and replacement of all major equipment would be practical.
Most of the maintenance would be done by use df remotely operated tools
through openings in roof plugs. The feasibility of such methods has been
demonstrated in the MSRE, and information is available relative to the

special tools required.

 
 

 

 

 

Table 3.18. MSER Steam-Power Piping and Operating Conditions

 

 

Steam Line Cold Rehesat Cold Reheat Hot Reheat Feedwater Line Heating Steam Heating Steam
S8izes and Conditions Leaving Boiler- Line to Line to Line Leaving to Boiler- Line to Line from
Superheater Preheater Reheater Reheater Superheater Preheater Preheater
Number of pipes 8 2 8 8 8 2 2
Nominal pipe OD, in. 14 35 18 16 12 12 12
Wall thickness, in. 3 0.69 0.5 0.5 1.3 2 2
Pipe material A335, Gr P-22 A155, Gr KC-70 Al55, Gr KC-70 A335, Gr P-22 Al06, Gr C A335, Gr P-22 A335, Gr P-22
Operating temperature, °F 1000 552 650 1000 700 1000 866
Allowsble stress at operating 7,800 15,750 15,750 7,800 16,600 7,800 7,800
temperature, psi
Flow rate, 1b/hr 10.1 x 10° 5.1 X 106 5.1 X 10% 5.1 X 108 10.1 X 10® 2.9 x 10° 2.9 X 108
Pressure, psia 3600 600 570 540 3800 3600 3500
Specific volume, ft3/1b 0.20 0.88 1.07 1.57 0.029 0.20 0.16
Total volume flow, cfm 33.4 X 10° 78.9 X 10° 90.0 x 10° 132 x 10° 4.9 X 10% 9.6 X 10° 7.9 %X 10%
Caleulated velocity, fpm 11.9 X 10° 6.0 X 10° 5.7 x 10° 13.5 x 10° 1,12 x 10°  11.5 x 10° 9.5 x 10°
Assumed velocity, fpm 10 to 12 X 10° 5.8 to 7.4 X 10° 5.8 to 7.4 X 10® 15.4 X 10° 15.4 X 10° 1.1 X 10° 1.1 x 102
Total flow area, in.? 481 177 1756 1235 657 138 114

 

26

 
 

 

 

o«

97

3.10.3 Containment

The primary circulating systems containing the fuel and blanket salts
are constructed of HaStélloy N and designed for about 150 psi and 1200
to 1300°F. These systems — consisting of the reactor, heat exchangers,
pumps, and connecting salt piping — are all housed in the reactor cell.
This cell volume is contained by a reinforced concrete structure lined
with steel plate; beneath the roof plugs are seal pans with welded joints.
This containment assures a cell leak rate below 1% of the total cell
volume per 24 hr. The reactor cell design pressure is about 45 psig.

The cells adjoining the reactor cell contain the boiler-superheaters,
reheaters, and coolant-salt circulating pumps. These cells, which are
also designed for a pressure of about 45 psig, are of reinforced concrete
and are sealed in the same manner as the reactor cell. Pressure-suppres-
sion systems are provided for both the coolant and reactor cells. These
systems are separate and independent and contain underground water tanks
for condensing steam.

The amount of water present in the reactor cell proper will be small,
probably consisting mainly of the water circulated through the shielding
and equipment-support cooling coils. The coolant-salt cells, one for
each of the separate coolant circulating circuits, are not intercon-
nected, and one cell could not credibly receive more than one-fourth of
the total coolant salt. However, it is conceivable that all the approxi-
mately 1,000,000 1b of steam and water inventory in the steam-power sys-
tem might flow into a single coolant cell. The pressure-suppression
system is designed to limit the cell pressure to 45 psi in such an acci-
dent. ,

The reactor and coolant-salt cells and the fuel-processing cell are
located in a building with a controlled ventilation system. The usual

adsorption and filtration equipment are provided.

3.11 Plant Construction Costs

The methods and assumptions used in estimating the MSBR cost conform

to those used in the advanced-converter reactor studies;® particular

 
 

 

 

28

reference was made to the Sodium Graphite Reactor, since its circulating
systems were similar to those of the MSBR.

The construction cost estimates of the reference MSBR design are
listed in Table 3.19 in conformance with the AEC Cost Guide.'! The di-
rect construction costs totalled sbout $80.7 million, and to this must
be added the indirect costs for engineering, contingencies, etc. These
indirect costs, listed in Table 3.20, correspond to about 41% of the
direct construction costs and givé total MSBR plant construction costs
of about $114 million. The‘exisfence of an‘established molten-salt re-
actor industry was assumed in estimating costs covering materials, fab-
rication, inspection, tranéportation, installation, and testing.

Additional information concerning the cost estimates for the vari-
ous accounts are given below. In obtaining the turbine plant costs, the
estimates used were influenced by the actual costs experienced in the
TVA Bull Run Steam Plant (completed about March 1966).

Land and Tand Rights (Acet. 20). An investment of $360,000 was as-
sumed for land. This is the same cost that has been allowed in other re-
actor studies. As is customary, the land was treated as a nondepreciating
capital cost and was subject to a lower fixed charge rate, as indicated
in Table 3.19.

Structures and Improvements (Acct. 21). The preliminary character
of the MSBR study did not warrant extensive optimization of the plant
layout. The turbine-room floor dimensions of the TVA Bull Run plant
were incorporated in the MSBR drawings.

The reactor plant portion of the building was considered in two
parts. The portion partially below grade and containing the more mas-
sive structures was estimated at $1.30 per cubic foot of building volume.
The upper high-bay portion was costed at $0.8O/ft3. These costs do not
include the containment, shielding, and overhead cranes, all of which
are included in separate accounts.

The total estimated direct construction cost of $9.3 million for
buildings and structures appears to be typical of 1000-Mw(e) nuclear
power stations.

Reactor Equipment (Acct. 221). The MSBR reactor vessel is about
14 ft in inside diameter and 19 ft high with torospherical heads; the
 

 

 

99

Table 3.19. Cost Estimate for MSBR Power Station

 

Account
No.

Item

Amount,
(in thousands of dollars)

 

20
21

22

aIncluded in indirect costs and in total plant cost.

Land and Land Rights®

Structures and Improvements

211

Ground improvements

ol
2

3
o4
5
.6
7

Reactor buildingb

Turbine building, auxiliary building, and feedwater
heater space

Offices, shops, and laboratories

Waste disposal building

Stack

Warehouse

Miscellaneous

Total Account 212

Total Account 21

Reactor Plant Equipment

221 Reactor equipment

222

223
225

226
227

228

d
2
.3
o
5
.6

Reactor vessel and internals

Control rods

Shielding and containment

Heating-cooling systems and vapor-suppression system
Moderator and reflector

Reactor plant crane

Total Account 221

Heat transfer systems

1

2
.3

)

Reactor coolant system

.11 Fuel-salt system

.12 Blanket-salt system

Intermediate coolant system

Power system )

.31 Steam generators (boiler-superheaters)
.32 Reheaters

Coolant supply and treatment

Total Account 222

Nuclear fuel handling and storage (drain tanks)

Radioactive waste treatment and disposal (off-gas
system)

Instrumentation and controls

Feedwater supply and treatment

o1
.2
'3
‘4
l5

Makeup supply and feedwater purification
Feedwater heaters

Feedwater pumps and drives

Reheat-steam preheaters

Pressure-booster pumps

Total Account 227

Steam, condensate, and feedwater piping

capital expense in estimating fixed charges.
‘iiJ Phoes not include containment cost; see account 221.3.

"

866
4,181
2,832

1,160
150
76

30

 

8,469

 

9,335

1,610
250
2,113
1,200
1,089
265

 

6,527

5,054
1,947

6,530
3,323
300

18,832

1,700

450

4,500

470
1,299
1,600
275
407

4,051

Land is classified as a nondepreciating

 
 

 

100

Table 3.19 (continued)

 

Account It Amount
No. : : o : . . (in thousands of dollars)

 

229 Other reactor plant equipment (remote maintenance)

.1 Cranes and hoists 500
.2 Special tools : 1,500
.3 Decontamination facilities 1,000
.4 Replacement equipment ‘ 2,000

 

Total Aceount 229 5,000
Total Account 22 ' - 45,129
23 Turbine-Generator Units

231 Turbine-generator units 19,174
232 Circulating-water system ' : 1,243
233 Condensers and auxiliaries 1,690
234 Central lube-oil system : 80
235 Turbine plant instrumentation ' 25
236 Turbine plant piping _ 220
237 Auxiliary equipment for generator 66
238 Other turbine plant equipment ‘ : 25

 

Total Account 23 22,523
24 Accessory Electrical Equipment

241 Switchgear, main and station service . 550
242 Switchboards 128
243 sStation service transformers 169
244  Auxiliary generator 50
245 Distributed items 2,000

 

Total Account 24 2,897
25 Miscellaneous 800

 

Total Direct Construction Cost 80,684
Privately Owned Plant
Total indirect costs (see Table 3.20) 33,728
Total plant cost® 114,412
Less nondepreciating capital
Land 360
Coolant-salt inventory 354
Total Depreciating Capital 113,698
Publicly Owned Plant
Total indirect costs (see Table 3.19) | 30,011
Total plant cost® 110,695
Less nondepreciating capital
Land 360
Coolant-salt inventory 354

Total Depreciating Capital 109,981

 

cInclu.des land and coolant-salt inventory costs.

h
 

101

Table 3.20. Distribution of Indirect Costs®

 

 

 

 

Percentage of Ingigzct Acc;gtiited
Ttem Acc;gziited (in thousands (in thousands
of dollars) of dollars)
Direct construction cost 80,684
General and administrative 6 4,841 85,525
Miscellaneous construction 1 855 86,380
Architect-engineer fees 5 4,319 90,699
Nuclear-engineering fees 2 1,814 92,513
Startup costsP 0.7 646 93,159
Land 360 93,519
Coolant~salt inventory 354 93, 873
Contingency 10 9,387 | 103, 260
Interest —-private financing 10.8 13,152 114,412
Total indirect costs 33,728
(private financing)
Interest — public financing 7.2 7,435 110,695
Total indirect costs 30,011

(public financing)

 

a'Indirec‘l; costs follow those used in the advanced converter reactor
studies.?

bStartup costs are based on 35% of first year's nonfuel operating
and maintenance costs.

vessel walls are about 1.5 in. thick and the heads about 2.25 in. Based
on fabrication experience with similar vessels and materials, $8.00/1p
was used to cover the installed cost of the vessel, supports, etc.

There are 534 Hastelloy N tubes 1.5 in. in diameter and 18 in.
long, and a like number of tubes 3 in. in diameter and 18 in. long.
These were estimated to have an installed cost of $6.00/1b. The cost
of brazing the graphite tubes to these Hastelloy N tubes was estimated
at roughly $100/braze, or about $107,000. An additional $393,000 was

 
 

 

 

102

allowed for special inspections, assembly of the graphite, etc., to bring
the total cost of the vessel to about $1.6 million.

The control rods for the MSBR do not employ expensive drive or scram
mechanisms and were not studied in detail for this preliminary report.
An allowance of $250,000 was made for the rods.

Réihfbrced concrete for'shielding and containment was estimated to
cost $80/yd3, in place. The 0.25- to 0.5-in. steel liner plate was esti-
mated to cost $l.50/1b, in place. The thermal insulation was considered

to have an installed cost of $6.OO/ft2. An allowance of $100,000 was made

for water cooling of the structures.

The MSBR reactor cell requires heating, cooling, and a vapor-suppres-

sion‘system to limit the pressure in an emergency condition, but con-
ceptual studies of these systems were not undertaken. An allowance of
$1.2 million was made for these items.

The graphite used as the MSBR moderator must be of high density and
high quality and be closely inspected. About 76% of the core volume is
graphite. It is assumed to cost $10/1b based on information from a manu-
facturer and the apparent feasibility of extruding the required shapes.
The reflector graphite was assumed to be 6 in. thick. The cost of this
graphite was estimated at $5/1b.

Heat Transfer Systems (Acct. 222). The costs of the shell-and-tube
heat exchangers were determined by breaking down each component into
weights of shells, tubes, etc., and using typical costs for materials
and fabrication to arrive at a total estimated cost per square foot of
surface. These values checked well with costs of similar reactor plant
heat transfer equipment for both actual equipment and for estimates used
in other studies.

The cost of Hastelloy N piping carrying molten salts was estimated
at $10/1b. | |

The costs of salt-circulating pumps were estimated by extrapolating
experience with existing molten-salt pumps and using costs for liquid-
metal pumps.2® The costs were increased 5% to include supports and by
10% to include installation and testing.

The quantity of coolant salt required for one filling of the circu-

lating system is about 2833_ft3. At an estimated cost of $1.00/1b, the

h
 

 

 

103

coolant-salt inventory cost is about $354,000. This is & nondepreciating
type of capital expense and was treated in the same manner as the land
cost.

Nuclear Fuel Handling and Storage (Acct. 223). No conceptual design
work was done on the MSBR fuel- and blanket-salt drain tanks. An allow~
ance of $1.7 million was made for the eight tanks.

Feedwater Supply and Treatment (Acct. 227). The estimated costs for

 

the makeup supply, feedwater purification, and feedwater pumps and drives
are largely based on values used in other 1000-Mw(e) reactor plant studies
and on the TVA Bull Run Steam Plant data.

A value of $44/1b was used for the eight Croloy reheat-steam pre-
heaters. The two high-pressure low-head 20,000-gpm pressure-booster pumps
in the feedwater supply to the boiler-superheater were estimated at
$4/gpm capacity. The motor costs were based on unit costs of $lO/hp.

The variable-speed drives were also based on unit costs of $10/hp.

Steam, Condensate, and Feedwater Piping (Acct. 228). The cost of
condensate and feedwater piping for the 915-Mw(e) TVA Bull Run Plant is
reported to be $3.62 million. On this basis, the piping for the 1000-
Mw(e) MSBR was estimated to be $4.07 million.

Other Reactor Plant Equipment (Acct. 229). Maintenance of the MSBR
will probably require remotely controlled cranes and hoists and the use
of special tooling and remote-brazing and -welding equipment. Decontami-
nation and hot storage facilities are needed. No conceptual designs were
made for this equipment. The costs listed correspond to judgments, with
estimates tending to be high due to lack of design data and of mainte-
nance experience. |

The MSBR maintenance procedures will involve replacement and subse-
quent repair rather than in-place repair of items such as salt pumps and
primary heat exchangers. This will entail an inventory of replacement
equipment, an expense that could be interpreted as part of the initial
capital investment rather than as an operating expense. An allowance of
$2 million was made for this replacement equipment.

Turbine-Generator Units (Acct. 231). The turbine-generator founda-
tions were estimated at $370,000, a more or less standard allowance for

1000-Mw(e) station studies.? Erection costs were taken to be $700,000,

 
 

 

 

104

again a standard value. The cost of a cross-compounded four-flow turbine-
generator unit with 43-in. last-stage blades was based on General Electric
Company pricing data, with a 78% discount factor applied to the book wvalue.
The excitation equipment was assumed to be of the brushless type, with
no provisions for standby excitation.

Circulating-Water System (Acect. 232). The estimated cost of about

$1.24 million for the circulating-water equipment was taken from the SGR
9,10

 

cost estimate. The TVA Bull Run cost data were not applicable because
the circulating-water installations include provisions for future plant
expansion.

Condensers and Auxiliaries (Acét. 233). The total cost of the four-
section horizontal single-pass 320,000-ft? units for the 915-Mw(e) TVA
Bull Run Plant was $1.3 million. This was extrapolated to $1.5 million
for the MSBR. The $l90,000 allowance for the MSBR condensate pump was
also extrapolated from the TVA data.

Central Lube-0il System (Acct. 234). An allowance of $80,000 was
made for this account on the basis of TVA cost information.

Turbine Plant Instrumentation (Acct. 235). This account covers tur-
bine plant control boards and instruments not included with the steam
piping (Acct. 228) and instrumentation (Acct. 226). An allowance of
$25,000 was made on the basis of the SGR estimate.®’*® (The TVA Bull Run
data were not available in a form such that this account could be ex-
tracted conveniently.)

Turbine Plant Piping (Acct. 236). The TVA Bull Run Plant reported
cost of $160,000 was extrapolated to the 1000-Mw(e) plant size; in addi-
tion, $12,000 was added for the preheater, booster-pump, etc., to make a
total of $220,000 for this account. '

Auxiliary Equipment for Generator (Acct. 237). Although the esti-
mated cost of the turbine-generator unit is presumed to include the
auxiliary equipment, the preliminary nature of the estimate led to in-
clusion of $66,000 for miscellaneous equipment end uncertainties.

Other Turbine Plant Equipment (Acct. 238). This miscellaneous ac-
count is of little significance in the total cost, and other reactor
plant studies have not always included it. On the basis of TVA experi-
ence, however, $25,000 has been included in the MSBR estimate.
 

o

105

Accessory Electrical Equipment (Accts. 241—245). This account
covers the cost of hundreds of electrical items, such as motor starters,
etec., scattered throughout the plant, and amounts to a significant por-
tion of the total plant investment. The estimate of about $2.35 million
for the total of accounts 242 through 245 is the same as that used in
the SGR study.g’lo Account 241, which covers both main and station ser-
vice switchgear, was reduced below the SGR estimate because the MSBR has
smaller pump motors, utilizes turbine-driven boiler-feedwater pumps, and
does not require large motor-driven pumps for emergency cooling of the
type needed in the SGR. However, the total of about $2.9 million for
Account 24 is only slightly less than the total of $3.0 million used for
the SGR.

Indirect Costs. The indirect costs, which amount to about 41% of
the total direct construction cost, have a very important bearing on the
total capital cost and the final production expense. The indirect costs
for the MSBR follow those used in the advanced-converter study9 and are
listed in Table 3.20. The percentages used appear to be more representa-
tive of present practice than those suggested in the AEC cost evaluation
handbook.t! Each percentage expense is applied to the accumulated cost
total preceding the particular item.

The land and coolant-salt inventory costs are included in the in-
direct costs so that the contingency and interest costs reflect these
expenses. However, the land and coolant-salt costs are deducted from

the total plant cost to obtain the depreciating capital outlay.

3.12 Power-Production Cost

Power costs are made up of capital charges, operating and mainte-
nance costs, and fuel-cycle costs. In computing capital charges, an
important quantity is the fixed charge rate. For an investor-owned MSBR
plaht, a fixed charge rate of 12%/yr was applied to depreciating capital,
while 10%/yr was'applied to nondepreciating capital. These fixed charge
rates are the same as those used in the advanced-converter reactor
studies;? the distribution of the charge rate for depreciating capital

is given in Table 3.21.

 
 

 

 

106

Table 3.21. Fixed Charge Rate Used for
Investor-Owned Power Plants

 

Tten : (%/yr)

 

Return on money invested® 6
Thirty-year depreciationP 1
Interim replacements® | 0.
Federal income taxesd 1l
Other taxes® 2
Tnsurance other than liability® 0

 

Total 12.0

 

&Return was based on one-third equity capital
financing, with a return of 9% after taxes, and
two-thirds debt capital drawing 4.5% interest.

bThe sinking-fund method was used in deter-
mining the depreciation allowance (plant life of
30 years assumed).

°In accordance with FPC practice, a 0.35%
allowance was made for replacement of equipment
having an anticipated life shorter than 30 years.

dFedera.l income taxes were based on "sum-of-
the-year digits" method of computing tax defer-
rals. The sinking-fund method was used to nor-
malize this to a constant return per year of 1.8%.

®Ihe FPC recommended value of 2.4% was used
for "other taxes."

A conventional allowance of 0.20% was made
for property damage insurance. Third-party
liability insurance is listed as an operating
cost. :

For publicly owned plants, the fixed charge rate employed was 7%/ yr
for depreciating capital; the distribution of this chargé rate is given
in Table 3.22. For nondepreciating capital the charge rate was 5%/yr .

The operation and maintenance charges are given in Table 3.23 and
are consistent with those used for the advanced-converter studies;®
however, the staff payroll costs were increased by 35%, since preliminary
infbrmatidn regarding the proposed revision to Section 530 of the AEC &;J

Cost Guide!! indicates that such an increase is required to be consistent

1k
 

 

 

 

9)

107

Table 3.22. Fixed Charge Rate Used for
Publicly Owned Power Plants

 

 

Ttem Rate

(%/yr)

Return on money invested 4.00
Thirty-year depreciation 1.75
- Interim replacement 0.35
Local taxes plus insurance 0.90
Total 7.00

 

Table 3.23. Operation and Maintenance Costs for
a 1000-Mw(e) MSBR®

 

Ttem Annual Cost

 

Operating

Total payroll, 70 employees with 20% $ 900,000
for fringe benefits and 20% for
general and administrative expense

Private insurance 260, 000

Federal insurance, at $30/Mw(th) 67,000
Maintenance

Repair and maintenance materials 1,065,000

Makeup coolant saltb (2% replacement 7,000

per year)

Contract services 72,000

Total operating cost $2,371,000
Unit cost, mill/kwhr(e), 0.8 load 0.34
factor

 

a . . .
The operating and maintenance costs associated
with the fuel-recycle processing plant are included in
the fuel-cycle costs.

bMakeup carrier salt for the réactor salt cir-
cuits is included under fuel-cycle costs.

 
 

e o b v e e ©

 

108

with present-day salaries. The operation and maintenance costs associated
with the fuel-processing plant could also be included here but, instead,
are included under fuel-cycle cost so that the latter can be more di-
'rectly compared with the fuel-cycle costs of reactor plants employing
off-plant fuel fabrication and processing. |

Combining the capitai costs, operation and maintenance costs, and
the fuel-cycle costs gave the power-production costs summarized in Table
3.24 for investor-owned and publicly owned utilities. As shown, the
power-production cost would be about 2.75 mills/kwhr(e) in an investor-
ovned plant and about 1.73 mills/kwhr(e) in a publicly owned plant.

In a utility system complex, the incremental cost between zero-power
and full-power operation influences the load factor of an individual
plant. This incremental cost for the MSBR is shown in Table 3.25, along

with other costs that are independent of power level. As shown, the

Table 3.24. Power-Production Cost in 1000-Mw(e) MSBR

Load factor: 0.8

 

Capital Cost Annual Cost
(in thousands (27te) (in thousands [migzz?kSEi?e)J
of dollars) yr of dollars)

 

Private~-Ownership Financing

Fixed charges

 

Depreciating capital 113,700 12 13,644 1.947
Nondepreciating capital 714 10 71 0.010
(1and plus coolant-salt
inventory)
Operation and maintenance costs 2,371 0.338
Fuel-cycle cost 0.459
Total estimated production cost 2.75

Public Financing
Fixed charges

 

Depreciating capital 110,000 7 7,700 1.099
Nondepreciating capital 714 5 36 0.005
(1and plus coolant-salt -
inventory)
Operation and maintenance costs ' 2,371 0.338
Fuel-cycle cost 0.287

Total estimated production cost 1.73

 

120
 

 

9

109

Tabie 3.25. MSBR Power Cost Breakdown into Fixed
and Incremental Costs

 

 

 

 

Financing
Ttem
Private®  PublicP
Annual fixed charges, $/kwyr 16.2 9.04
Fixed operating costs,C mill/kwhr(e) 0.39 0.39
Total fixed power cost,® mills/kwhr(e) 2.70 1.68
Tncremental power cost,® mill/kwhr(e) 0.05 0.05
Total power-production cost, mills/kwhr(e) 2.75 1.73

 

a12%/yr fixed charges on reactor plant, including process-
ing plant; lO%/yr inventory charges for nondepreciating items.

b7%/yr fixed charges on reactor plus processing plant;
5%/yr inventory charges for nondepreciating items. Not opti-
mized for changed conditions.

®Includes 0.06 mill/kwhr(e) for fixed operating cost of
the processing plant.

dBased on 0.8 load factor.

eIncrem_ental cost in going from zero- to full-power opera-
tion (0.8 load factor); includes incremental fuel-cycle cost
and incremental operating costs.

incremental cost between operation at zero power and at full power is
only 0.05 mill/kwhr(e) and would provide a high incentive for operating
with a high plant factor. Since the reactor has "on-line" refueling,
there is no basic reason why the plant has to be shut down except for
maintenance; operation with a 0.9 load factor would decrease MSBR power
costs to 2.49 mills/kwhr(e) for investor-owned plants and to 1.59
mills/kwhr(e) for publicly owned plants,

 
 

 

 

110

4. ALTERNATIVE CONDITIONS FOR MSBR DESIGH.

As a part of this study, various alternative conditions were con-
-sidered for the initial MSBR design in order to improve the plant and to
measure the incentive for achieving such conditions. One of the more
impoftant conditions is the ability to economically remove protactinium
directly from the blanket stream of the reactor. Another desirable con-
dition is that of introducing feedwater into the boiler~-superheaters at
580°F rather than 700°F. The ability to maintain a high plaht factor at
all times is also of importance. These items, as well as others, are

discussed below.

4,1 Protactinum Removal from Blanket Stream

Even.though fluoride volatility processing appears to be a satisfac-
tory process for removal of uranium, the ability to remove 233Pa directly
and economically from the blanket region of an MSBR would significantly
improve the performance of the reactor. One possible process involves
oxide precipitation of protactinium. Several laboratory experiments27’28
have demonstrated that protactinium can be readily precipitated from a
molten fluoride mixture by addition of thorium oxide and that the precipi
tate can be returned to solution by treatment with HF. Experimental re-
sults also indicate that treatment of protactinium-containing salt with
Zr0, leads to oxide precipitation of the protactinium and that after beta
decay of the protactinium, the resulting UOp; will react with ZrF, to give
UF,.

More recent experimental results have indicated another method for

removing protactinium directly from the blanket fluid. This involves

treating the molten blanket salt with a stream of bismuth containing dis-

solved thorium metal. The thorium reduces the protactinium (and also any
uranium) to metal, which can then be accumulated on a stainless-steel-
wool filter. The deposited metal can be hydrofluorinated and/or fluori-
nated to return the protactinium (and any uranium) to the fuel-recycle

- process as the fluoride. Thus there is experimental evidence that simple

processes are available for direct removal of protactinium from the blanket
 

 

 

 

111

stream of an MSBR. Practicable application'of such processes would de-
crease absorptions of neutrons by protactinium to a negligibly low level
and also remove economic restrictions as to the permissible average neu-
tron flux in the circulating-blanket volume (related to thorium inventory
needs ).

The mechanical design of the MSBR with protactinium removal would
be essentially the same as that given previously, and the primary change
would be in the nuclear design and fuel-cycle performance. The resulting
reactor plant is termed the MSBR(Pa) and refers to the initial MSBR design
modified for protactinium removal from the blanket stream.

Either the oxide-precipitation process or the liquid-metal extraction
process appears feasible as a method of removing'protactinium from the
blanket stream. It was estimated thafi for either process the blanket-
processing costs would be equivalent to those associated with uranium re-
covery by fluoride volatility processing plus an additional capital in-
vestment for equipment. This additional investment varies with the
blanket-processing rate associated with protactinium recovery and is esti-
mated to be about $1.65 million at a blanket-salt processing rate of 1000
ft3 per day; for other processing rates the capital investment is esti-
mated to vary in accordance with the throughput rate raised to the 0.45
power.,

The same design methods used for the MSBR were employed in obtaining
the MSBR(Pa) design conditions, except that the blanket-processing method
and costs were altered in accordance with the above discussion. The re-
sulting MSBR(Pa) design conditions are given in Table 4.1,

The results of the MSBR(Pa) nuclear performance calculations are
summarized in Table 4.2, while Table 4.3 gives the neutron balance for
the associated design conditions. These results can be compared with
those in Tables 3.7 and 3.8 to give the relative nuclear performance of
the MSBR(Pa) versus the MSBR. The essential differences are the decreased
neutron absorptions by protactinium and the lower thorium inventory for
the MSBR(Pa) design conditions.

In obtaining the reactor design conditions, the optimization pro-

cedure considered both fuel yield and fuel-cycle cost as criteria of

 
 

 

112

Table 4.1. MSBR(Pa) Design Conditions

 

Power, Mw

Thermal
Electrical

Thermal efficiency
Plant load factor
Dimensions, ft

Core
Height
Diameter
Blanket thickness
Radial
Axial
Reflector thickness

Volume fractions

Core
Fuel salt
Fertile salt
Moderator
Blanket
Fertile salt

Salt volumes, ft>

Fuel .
Core
Blanket
Plena
Heat exchanger and piping
Processing

Total
Fertile
Core
Blanket
Heat exchanger and piping
Processing

Total
Salt compositions, mole %
Fuel
LiF
BeF»
UF, (fissile)

2225
1000

0.45
0.80

166
26
147
345
33

717
72

1121

100
24

1317

63.6
36.2
0.22
 

 

113

Table 4.1 (continued)

 

Salt compositions, mole % (continued)

Fertile
LiF 71.0
UF, (fissile) 0.0005
Core atom ratios
Thorium to uranium , 41.7
Carbon to uranium 5800
Fissile inventory, kg 681
Fertile inventory, 1000 kg 101
Processing
Fuel stream Fertile stream
Equivalent cycle time, days
Uranium removal process 42 55
Protactinium removal None 0.55
process _
Equivalent rate, ft3 per day
Uranium removal process 16.3 23.5
Protactinium removal None 2350
process
Unit processing cost, $/ft’ 190 652

 

aEquivalent unit processing cost based on recovery of ura-
nium by the flouride volatility process and protactinium concen-
tration in accordance with protactinium removal rate, which gives
the same processing cost as that associated with direct protac-
tinium removal from fertile stream.

performance. Although most.emphasis was given to obtaining a low fuel-
cycle cost, a fractional weight was given to maximum fuel yield, so the
design conditions do not correspond to minimum fuel-cycle costs. This
is illustrated in Fig. 4.1, which shows the minimum cost as a function
of fuel yield. The design conditions for thé MSBR(Pa) and also the MSER
correspond to the designated points in Fig. 4.l.

The MSBR(Pa) fuel-cycle costs are listed in Table 4.4. Comparison
with results in Table 3.9 shows that direct protactinium removal from

the blanket stream reduces fuel-cycle costs by about 0.1 mill/kwhr(e)

 
 

 

114

Table 4.2. Nuclear Performance for MSBR(Pa)
Design Conditions

 

Fuel yield, % per annum 7.95
Breeding ratio 1.0713
Fissile losses in processing, atoms per fis- 0.0051
sile absorption
Neutron production per fissile absorption (ne) 2.227
Specific inventory, kg/Mw(e) 0.681
Specific power, Mw(th)/kg 3.26
Power density, core average, kw/liter ,
Gross 80
In fuel salt 473
Neutron flux, core average, neutrons/cmz-sec
Thermal 7.2 X 10*%
Fast 12.1 X 104
Fast, over 100 kev 3.1 X 104
Thermal flux factor in core, peak-to-mean
ratio _
Radial 222
Axial - 1.37
Fraction of fissions in fuel stream 0.996
Fraction of fissions in thermal-neutron group 0.815
Mean n of 233y 2.221
Mean 1 of 23°U 1.958

 

and the thorium inventory requiréments by nearly a factor of 3. Table
4.5 summarizes fuel-cycle costs for privately and publicly financed
MSBR(Pa) plants, while Table 4.6 gives estimated power-production costs.
Table 4.7 gives MSBR(Pa) fixed and incremental power costs similar to
those given in Table 3.24 for the MSBR. As shown, it is more economical
to operate the plant at full power than to let the plant idle at zero
power; operation at 0.9 load factor rather than 0.8 would lead to power-
production costs of 2.35 and 1.46 mills/kwhr(e) for private and public

financing, respectively.
 

 

 

 

 

@

115

Table 4.3. Neutron Balance for MSBR(Pa) Design Conditions

 

Neutrons per Fissile Absorption

 

 

 

 

 

 

 

 

 

Material
Absorbed Total Absorbed by Fission Produced
232 0.9970 0.0025 0.0058
233pg, 0.0003
233y 0.9247 0.8213 2.0541
234y 0.0819 0.0003 0.0008
235y 0.0753 0.0607 0.1474
2367 0.0084 0.0001 0.0001
237y 0.0009
238y 0.0005
Carrier salt (except 6Li) 0.0647 0.0186
611 0.0025
Graphite 0.0323
135%e 0.0050
149gm 0.0068
151lsm 0.0017
Other fission products 0.0185
Delayed neutrons lost® 0.0049
Leakagéb 0.0012
Total 2.2268 0.8849 2.2268
a‘Delayed neutrons emitted outside core.
bLeakage, including neutrons absorbed in reflector.
Table 4.4. TFuel-Cycle Cost for MSBR(Pa) Design Conditions
Cost (mill/kwhr)
Fuel Stream Fertile Stream  Total Grand Total
Fissile inventory® 0.1125 0.0208 0.1333
Fertile inventory 0.0000 0.0179 0.0179
Salt inventory 0.0147 0.0226 0.0373
Total inventory | 0.188
Fertile replacement 0.0000 0.0041 0.0041
Salt replacement 0.0636 0.0035 0.0671
Total replacement _ 0.071
Processing 0.1295 0.0637 0.1932
Total processing ' 0.193
Production credit 0.105
Net fuel-cycle cost 0.35

 

aIncluding 233pg, 233y, and 235y.

 
 

 

116

: ORNL-DWG 66-7456
0.6

/

MSBR DESIGN POINT

 

 

\

 

 

FUEL CYCLE COST [mitts/kwhe}]

 

 

 

 

 

 

 

 

 

04 /
MSBR(Pa) DES'DW/
0.3 g
02
2 3 4 5 6 7 8 9

FUEL YIELD (%/yr)

Fig. 4.1. Variation of Fuel-Cycle Cost with Fuel Yield in MSBR and
MSBR(Pa) Concepts.

Table 4.5. MSBR(Pa) Fuel-Cycle Costs for Investor-Owned
and Publicly Owned Plants

Load factor: 0.8

 

Cost [mill/kwhr(e)]

 

Item

 

Privated PublicP
Ownership Ownership
Fissile-, fertile-, and carrier-salt inventory 0.188 0.094
Replacement cost of fertile and carrier salts 0.071 , 0.071
Core- and blanket-processing costs
Operation and maintenance 0.069 0.069
Capital costs 0.124 0.073
Bred fuel credit (0.105) (0.105)
Net fuel-cycle cost 0.35 0.20

 

®Based on lZ%/yr capital charges for processing plant and inventory
charges of 10%/yr. ’

bBased on ‘7%/y"r capital charges for processing plant and inventory
charges of 5%/yr.

h
 

 

 

117

Table 4.6. Power-Production Costs of 1000-Mw(e) MSBR(Pa)

Load factor: 0.8

 

Power Cost

 

 

 

 

[mills/kwhr(e)]
Item
Private Public
Financing Financing
Fixed charges
Depreciating capital 1.947 1.099
Nondepreciating capital (land plus 0.010 0.005
coolant-salt inventory) |
Operation and maintenance costs 0.338 0.338
Fuel-cycle cost 0.348 0.202
Power-production cost 2 .64 1.64

 

Table 4.7. MSBR Power Cost Breakdown into Fixed
and Incremental Costs

 

 

 

Private® PublicP
Item . . . .
Financing Financing
Annual fixed charges, $/kwyr 15.9 8.90
Fixed operating costs,® mill/kwhr(e) 0.38 0.38
Total fixed power cost, d mllls/kwhr(e) - 2.65 1.65
Incremental power cost,® mlll/kwhr(e) —0.01 —0.01
Total power-production cost,‘mllls/kwhr(e) ' 2.64 - 1.64

 

12%Vyr fixed charges on reactor plant, 1nclud1ng Processing
"‘plant lO%Vyr inventory charge for nondepreciating items.

7%Vyr fixed charges on reactor plus processing plant; 5%/yr
inventory charges for. nondeprec1at1ng items. Not optimized for
changed condltlons.

“This 1ncludes 0. 055 mlll/kwhr e) for fixed operatlng cost of
the processing plant.

dBased on 0.8 load factor.

®Incremental cost in going from zero to full-power operation
(0.8 load factor); this includes incremental fuel-cycle cost and
incremental operating costs.

 
 

 

 

118

For comparison, a summary of the power cost and fuel-utilization
characteristics of the MSBR(Pa) and the MSBR is given in Table 4.8.

- Table 4.8. Comparison of Power-Production Cost and Fuel-Utilization
Characteristics of the MSBR(Pa) and the MSBR

 

 

MSBR(Pa ) MSBR

Specific fissile inventory, kg/Mw(e) 0.68 0.77

Specific fertile inventory, kg/Mw(e) 105 268

Breeding ratio 1.07 - 1.05

Fuel-yield rate, %/yr 7.95 : 4 .86

Fuel doubling time,2 years 12.6 | 20.6
Private Public Private Public

Financing TFinancing Financing Financing

 

 

Capital charges, milils/kwhr(e) 1.95 1.10 1.95 1.10
Operating and maintenance cost, 0.34 0.34 0.34 0.34
mill /kwhr(e)

Fuel-cycle cost,b mill/kwhr (e) 0.35 0.20 0.46 0.29
Power-production cost, mills/kwhr(e) 2.64 1.64 2.75 1.73

 

% Inverse of the fuel-yield rate.

bCosts of on-site integrated processing plant are included in this value.

 

4.2 Alternative Feedwater Temperature Cycle

The 700°F feedwater temperature and the 650°F temperature of the
"cold" steam to the reheater in the initial design were dictated by the
700°F liquidus temperature of the coolant salt. It would be an obvious
advantége if it were not necessary to divert almost 30% of the throttle
steam for heating of the feedwater and reheat steam, since this diversion
leads to a loss of available energy. An even more significanthsaving
could be achieved if the 9.2 Mw(e) of power required to drive the feed-
water pressure-booster pumps could be eliminated; also, removal of the
- reheat~-steam preheaters and the booster pumps would reduce capital in-

vestment requirements. Thus, savings can be achieved by lowering the

(1]
 

 

 

 

119

temperature of the steam-cycle fluid entering the boilers and reheaters.
To determine the incentive for developing a coolant salt having a low
liquidus temperature, the MSBR steam-power cycle was studied with condi-
tions of 580°F feedwater temperature and 550°F reheat steam. In order
to differentiate and compare cases, use of 700°F feedwater and 650°F re-
heat steam is designated case A, while case B represents the alternative
conditions.

The cycle arrangement for the case B conditions is shown in Fig. 4.2.
In this cycle the 552°F steam from the high-pressure turbine exhaust is
introduced into the reheaters without preheating. The feedwater is heated
from 550 to 580°F by the addition of one more stage .of feedwater heat-
ing; steam extracted from the high-pressure turbine is used. The con-
densate from this new heater is cascaded back through the feedwater heaters
to the deaerating heater in the usual manner. The heat balances and the
analysis of the steam cycle with case B conditions were performed in the
same manner as for case A conditions.?% Table 4.9 compares the design
data for the two cases.

Thé elimination of the feedwater pressure-~booster pumps required in
case A saves about 9.2 Mw(e) of auxiliary power, which, together with the
improvement in the cycle thermal efficiency due to the additional stage
of feedwater regeneration, makes about 9.7 Mw(e) additional power avail-
able from the casé B cycle. The overall net thermal efficiency is thus
improved from the 44.9% dfitained from case A to 45.4% in case B.

The cost estimates for fihe MSBR steam station are given in detail
in Section 3.11 for caseJAQ‘ To qémplete the discussion of case A versus
case B conditions, the cost eSfifiates forfthe affected items of equipment
were compared; the reéfiltégare summarized in T&bfié 4.10. As shown, the
case B arrangement requirés about $465‘OOO less capitél_expenditure,,pri-
marlly due to removal of the pressure-booster pumps . ¥

The net effect of changing from case A to case B condltlons, assuming

that inexpensive coolant salt is available for both cases, is to increase

 

*¥In this cost study it was assuimed that the 580°F llquldus-temperature
coolant salt has the same cost (about $1.00/1b) as the MSBR coolant salt.

 
 

 

120

 

    
       

 

 

 

 

 

 

 

 

 

 

 

    

 

 

 

 

 

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——— COOLANT SALT

 

rlllllll-lé)\P ‘ @ SovDer T

 

 

 

Alternate Steam System Flowsheet for 580°F Feedwater.

Fig. 4.2.
 

121

‘Efi) Table 4.9. MSBR Steam-Power Steam Design and Performance Data for Case A and Case B Conditions

 

Case A — MSBR Steam Cgse B — MSBR Alternative

 

 

 

 

 

Cycle with 700°F Steam Cycle with
Feedwater 580°F Feedwater
General performance
Reactor heat input, Mw 2225 2225
Net electrical output, Mw 1000 1009.7
Gross electrical generation, Mw 1034.9 1035.4
Station auxiliary load, Mw(e) 25.7 25.7
Boiler-feedwater pressure-booster pump load, Mw(e) 9.2 None
: Boiler-feedwater pump steam-turbine power output, Mw 29.3 30.6
j Flow to turbine throttle, 1b/hr 7.152 X 106 7.460 X 10°
% Flow from superheater, 1b/hr 10.068 x 10°¢ 7.460 X 10°
Gross efficiency 47.83 ' 47.91
Gross heat rate, Btu/kwhr 7136 7124
Net efficiency, % 449 45.4
Net heat rate, Btu/kwhr 7601 7518
Boiler-superheaters
Number of units 16 16
Total duty, Mw{th) 1931.5 1837.0
Total steam capacity, 1b/hr 10.068 X 10° 7.460 X 10°
Temperature of inlet feedwater, °F 700 580
Enthalpy of inlet feedwater, Btu/lb 769 .2 583.6
Pressure of inlet feedwater, psia ~3800 ~3800
Temperature of exit steam, °F 1003 1003
Pressure of exit steam, psia ~3600 ~3600
Enthalpy of exit steam, Btu/lb 1424.0 1424.0
Temperature of inlet coolant salt, °F 1125 1125
Temperature of exit coolant salt, °F 850 850
Average specific heat of coolant salt, Btu/lb-°F 0.41 0.41
Total coolant-salt flow
1b/hr 58,468 X 106 55,608 X 106
cfs 129,93 123,57
gpm 58,316 55,463
Steam reheaters
Number of units 8 8
Total duty, Mw(th) 293.5 388.0
Total steam capacity, lb/hr 5.134 X 106 5.056 X 106
Temperature of inlet steam, °F 650 551.7
Pressure of inlet steam, psia ~570 ~600
Enthalpy of inlet steam, Btu/lb 1323.5 1256.7
Temperature of exit steam, °F 1000 1000
Pressure of exit steam, psia ~540 ~540
 Enthalpy of exit steam, Btu/lb 1518.5 1518.5
Temperature of inlet coolant salt, °F 1125 1125
Temperature of exit coolant salt, °F 850 850
Average specific heat of coolant salt, Btu/lb*°F .41 C.41
Total coolant-salt flow
: 1b/hr 8.884 X 108 11.744 X 108
{ efs 19.742 26.008
: gpm : : 8861 11,714
; Coolent salt pressure drop, inlet to outlet, psi ~60 ~60
: Reheat-steam preheater
Number of units 8 None
Total duty, Mw(th) 100.45
Total heated steam capacity, lb/hr : 5.134 X 106
Inlet temperature of heated steam, °F 551.7
Exit temperature of heated steam, °F 650
Inlet pressure of heated steam, psia ~580
: Exit pressure of heated steam, psia ~570
Inlet enthalpy of heated steam, Btu/lb 1256.7
Exit enthalpy of heated steam, Btu/lb’ 1 1323.,5
Total heating steam, 1b/hr - 12.915 X 10°
Inlet temperature of heating steam, °F 1000
; , Exit temperature of heating steam, °F 866
; \iiJ Inlet pressure of heating steam, psia 3515

Exit pressure of heating steam, psia

 

!
!
|
!

 
 

 

 

122

Teble 4.9 (continued)

 

Case A — MSER Steam Case B — MSER Alternative
Cycle with 700°F

Steam Cycle with

 

- Feedwater 580°F Feedwater
Boiler-feedwater pumps
Number of units 2 -2
Centrifugal pumps
Number of stages 6 6 .
Feedwater flow rate, 1b/hr total 7152 X 10% 7460 X 10°
Required capacity, gpm - 8060 8408
 Head, ft ~9380 ~9380
Speed, rpm oo 5000 5000
Water inlet temperature, °F 357.5 357.5
Water inlet enthalpy, Btu/lb 329.5 329.5
Water inlet specific volume, £t°/1b ~0.01808 ~0.01808
Steam~turbine drive .
Power required at rated flow, Mw (each) 14.66 15.30
Power, nominal hp (each) 20,000 20,000
Throttle steam conditions, psia/°F 1070/700 1070/700
Throttle flow, lb/hr (each) 413,610 431,400
Exhaust pressure, psia ~7 ~77
Number of stages 8 8
Number of extraction points 3 3
Boiler-feedwater pressure-booster pump
‘Number of units 2 None
Centrifugal pump
Feedwater flow rate, 1b/hr total 10.067 X 108
Required capacity, gpm (each) 9500
Head, ft ~1413
Water,inlet temperature, °F 695
Water inlet pressure, psia ~3500
Water inlet specific volume, £t3/1b ~0.03020
Water outlet temperature, °F ~700
Electric-motor drive
Power required st rated flow, Mw (each) 4.587
Power, nominal hp (each) 6150

 

 

the thermal efficiency from 44.9 to 45.4% and to reduce construction costs
by about $465,000. The lower construction cost reduces power costs by
about 0.008 mill/kwhr(e), while the increased efficiency lowers power

cost by about 0.026 mill/kwhr(e) (private financing), to give a total
saving of about 0.034 mill/kwhr(e) [0.021 mill/kwhr(e) for public financ-
ing]. This saving in a 1000-Mw(e) plant (0.8 load factor) corresponds to
about $238,000 per year. The present worth (6% discount factor) of this
saving over a 25-year period is about $1.5 million. For several MSBR
power plants, the saving would be proportionally greater. Thus, there

is an economic incentive for developing a coolant salt with a low liquidus
temperature so long as its inventory cost ddes not outweigh the potential
saving. If the inventory cost of the coolant salt for case B were about

$2.4 million more than that for case A, the potential saving would be

W
 

 

 

123

Table 4.10. Cost Comparison of 700°F and 580°F
Feedwater Cycles for MSBR2

 

 

Number  oose A — 700°F Case B — 580°F
of Feedwater Feedwater
Units
Feedwater pressure-booster pumps 2 $ 400,000 None
Reheat-steam preheaters 8 180,000 None
Special mixing tee 5,000 None
Feedwater heater No. OP None $ 150,000
Charge for extra extraction noz- None 45,000
zle on turbine for heater No. O
Boiler-superheaters 16 6,000,000C 5,900,000
Reheaters 8 2,720,000€ 2,880,000
_ $9,305,000 $8,975,000
Cost differential
Direct construction cost $ 330,000
Total construction costf $ 465,000

 

%Table shows only those costs different in the two cycle arrange-
ments and is not a complete listing of the turbine plant costs.

bThe high-pressure feedwater heater added in case B was designated
"No. 0" in order not to disturb the heater numbers used in case A.

“Estimated on basis of $130/ft2.
dpstimated on basis of $140/ft2.
®Estimated on basis of $125/ft2.

fIndirect costs were assumed to be 41% of the direct costs.

cancelled by the increased coolant-salt inventory cost (for a privately

owned plant).

 

4e3 Modular-Type Plant

An important factor in low poWer costs is the ability of the power

plant to maintain a high plant-availability factor. Thus design features

 
 

 

 

 

124

that can improve this factor are desirable if these features do not them-
selves introduce compensating disadvantages.

A feature of the MSBR plant design is the use of four heat exchanger
circuits in conjunction with one reactor vessel in such a manner that if
one pump in the fuel circuit stops, the reactor is effectively shut down.
If, on the other hand, it were practicable to have four separate reactor
circuits, with each connected to one of the four heat exchanger circuits,
stoppage of a fuel pump would shut down only one~quarter of the station
capacity, leaving 75% available for power production. In ordér to deter-
mine the practicality of using a number of reactors in a single 1000-Mw(e)
station, the design features of a modular-type MSBR plant, termed MMSER,
were investigated. |

The MMSBR design concept considers four separate and identical fe-
actors, along with their separate salt circuits. The only connections
of the four reactors are through the fuel-recyéle plant. The designs of
the heat exchangers, the coolant-salt circuits, and the steam-power cycle
remgin essentially as for the MSBR. Each reactor module generates ther-
mal power equivalent to that required for producing 250 Mw(e) net.

The flow diagram given previously for the MSBR (Fig. 3.7) also is
essentiglly valid for the MMSBR. Salt flow rates and capacities of the
various components remain as in the MSBR design.

Figures 4.3 and 4.4 give plan and elevation views of the four dis-
tinct reactor cells, along with their adjacent steam-generating cells.
Any reactor module can be shut down and serviced while the other three
remain operating.

The reactor core consists of 210 graphite fuel cells operating in
parallel within the reactor tank. The design of the graphite tubes sepa-
rating the fuel and blanket salts is similar to that used in the MSBR.
The reactor core region is cylindrical-with a diameter of about 6.3 ft
and a height of about 7.9 ft. The reactor vessel is agpproximately 12 ft
in diameter and about 14 ft high. Except for the use of four reactor
vessels instead of one, all design features of the MMSBR are similar to
those of the MSBR. The design conditions associated with one reactor

module are summarized in Table 4.11. (;J
   

 

   
 
  
   
    

STE TERS

GENERATORS

COOLANT
PUMPS

PRIMARY
EXCHANGER AND

AN

FUEL PUMP BLANKET
HEAT
THERMAL EXCHANGER
SHIELD AND PUMP

o

INSULA REACTOR

Fig. 4.3. Modular Plant Reactor Cell — Plan AA.

 
Y

 

 

ORNL DWG 66-7115

CONTROL ROD DRIVE

FUEL AND BLANKET
PUMP DRIVE MOTORS

   
 

T SALT PUMPS

 

 
 
 
 
 
 
  
 
 

STEAM GENERATORS

REACTOR
REHEATERS
BLANKET HEAT
EXCHANGER
=
60-0" -3 Eé
=3 STEAM
PIPING
PRIMARY HEAT =
EXCHANGER
=3

 

 

 

Fig. 4.4. Modular Plant Reactor Cell — Elevation BB.

 
 

 

 

 

127

Table 4.11. MMSBR Design Conditions for One Module

 

Power generation

Thermal
Electrical

Thermal efficiency
Plant factor
Dimensions, ft

Core
Height
Diameter
Blanket thickness
Radial
Axial
Reflector thickness

Reactor volumes, ft°

Core
Blanket

Salt V‘olumes‘,rft3

Fuel
Core
Blanket
Plena
Piping
Heat exchanger and pump
Processing

Total -
Fertile
Core
Blanket , .
Heat exchanger and piping
Processing

- Total
Salt compositions, mole %

Fuel '
o TIiF

BeF»

UF, (fissile)
Fertile S

TLiF

BeF>

ThF,

Average power density in core fuel salt, kw/liter

556
250

45
0.80

245
1.000

41.5

22

25

82
7.5

 

185
12

1000
25
24

1061

63.6
36.2
0.22

71

473

 

 
 

 

128

The nuclear and fuel-cycle performance of a four-module plant gen-
erating 1000 Mw(e) was studied both for protactinium removal from the
blanket stream and for the case of no direct protactinium removal. The
same methods and bases as those for the MSBR studies were employed. Analo-
gous to previous terminology, these cases are termed MMSBR(Pa) and MMSER.
The results obtained are summarized in Table 4.12. Comparison with the
results obtained for the MSBR(Pa) and the MSBR indicates that the nuclear
and fuel-cycle performance of a modular-type plant compares favorably with
that of a single-reactor-type plant; the modular plant tends to have

slightly higher breeding ratio, fissile inventory, and fuel-cycle cost.

Table 4.12. Nominal Nuclear and Fuel-Cycle Performance
o of 1000-Mw(e) Modular Plants

Investor-owned plant: 0.8 load factor

 

MMSBR (Pa.) MMSBR

 

Fuel yield, % per year 7.3 5.0
Breeding ratio 1.073 1.053
Specific fissile inventory, kg/Mw(e) 0.76 0.80
Specific fertile inventory, kg/Mw(e) 125 310
Fuel-cycle cost, mill/kwhr(e) 0.38 0.48
Doubling time, yr® 13.7 20

 

% Inverse of fractional fuel yield per year.

Capital cost estimates were also made for the modular plant. The
primary difference between the MMSBR and MSBR-type plants is the use of
four reactor vessels and cells in the modular plant rather than the one
in the MSBR. However, the reactor vessels in the modular plant are
smaller, and their combined cost is not much more than that of the single
1arge vessel. Also, the modular plant permits better placement of cells
and a reduction in building volume. The resultant capital cost estimate
for the modular plant was essentially the same as that obtained for the
single-reactor plant. Using a cost estimate of $112/kw(e) for a privately

owned plant, along with the MSBR estimate for operation and maintenance

.costs, and the fuel-cycle costs from Table 4.12 gives the power-generation
 

 

 

 

 

129

costs summarized in Table 4.13. These costs are virtually the same as
those for the MSBR-type plants (see Table 4.8) and thus indicate the
desirability of a modular-type plant if the plant availability factor is

improved by its use.

Table 4.13. Power-Production Costs for Modular-Type
Molten-Salt Breeder Reactors

Investor-owned plant: 0.8 plant factor

 

Cost [mills/kwhr(e)]

 

MMSBR(Pa,) MMSBR

 

Fixed charges 1.93 1.93
Operation and maintenance costs 0.34 0.34
Fuel-cycle costs? 0.38 0.48

Total power-production costs 2.65 275

 

aCapital charges of processing plant are included
in fuel-cycle costs.

4.4 Additional Design Changes

 

In reactor design studies it often occurs that certain features of
the detailed design undergo changes as more understanding is obtained of
the overall problems and as new wajs are discovered to solve a given de-
sign problem. Such changes'have*taken place during the MSBR design
studies; of these, the most important are those associated with the pri-
mary heat exchanger designs and the pressures that exist in the various
circulating-salt systems. The revised design conditions are discussed
below.

An objectional feature of the MSBR heat exchanger design shown in
Fig. 3.20 is the use of expansion bellows at the bottom of the exchanger.
These bellows permit tubes in the central portion of the exchanger to

change in length relative to those in the annular region due to thermal

 
 

 

 

130

conditions. Since such bellows may be impractical to use under reactor
operating conditions, a new design was developed that eliminated them.

Figure 4.5 shows the revised heat exchanger design. The expansion
bellows were eliminated, and changes in the tube lengths due to thermal
conditions are accommodated by the use of sine-wave type of construction,
which permits each tube to adjust to thermal changes. In addition, the
coolant salt now enters the heat exchanger through an annular volute at
the top and passes downward through a baffled outer annular region. The
coolant then passes upward through a baffled inner annular region and
exits through a central pipe. |

In Fig. 4.5, the flow of fuel salt through the pump is reversed from
that shown in Fig. 3.20 in order to reduce the pressure in the graphite
fuel tubes. TFuel salt enters the heat exchanger ifi the inner annular
region, passes downward through the tubes; and then flows upward through
the tubes in the outer annular region before entering the reactor.

The blanket-salt heat exchanger was also revised to give g design
similar to that of Fig. 4.5. The general features of these exchangers
and their placement in the reactor cell are shown in Fig. 4.6 (for com-
parison with the initial MSBR design see Fig. 3.8). The blanket-salt
pump was also altered so that blanket salt leaving the reactor now enters
the suction side of the pump.

From the viewpoint of reactor safety, it is important that the blanket
salt be at a higher pressure than the fuel salt.?® Under such circum-
stances, rupture of a fuel tube would result in leakage of fertile salt
into the fuel and a reduction in reactivity. In order to achieve this
condition with a minimum operating pressure in the reactor vessel, the
fluid flow was reversed from that in the initial MSBR design, with fluid
leaving the reactor entering the suction side of the pumps. The result-
ing flow diagram is shown in Fig. 4.7 (for comparison with initial design
see Fig. 3.7). |

In addition, it is desirable that any leakage between the reactor
fluid and coolant-salt systems be from the coolant system into the fuel
or blanket system. In order to achieve these conditions, the MSBR op-

erating pressures were revised to those shown in Table 4.14.

o
 

131

ORNL DWG 66-7136

FUEL LEVEL (DUMP)

FUEL SALT DUMP TANK

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

o.g" FUEL LEVEL (OPERATING)
! ALT PUMP
; FUEL SALT TO FUEL SALT PUM
" & FROM REACTOR
D
| COOLANT SALT FROM
| STEAM HEAT EXCHANGERS
|
I —COOLANT PaSS
il SEPARATING BAFFLE
I4l_4u
|
|
1§
; i i FLOW ARRANGEMENT
| ; i 1R FUEL SALT-IN TUBE SIDE
| it b | (LR COOLANT SALT-IN SHELL SIDE
1l 10" 1[I}
| '
6'0'bo -
. (it
“’1 K =: i \I! ‘ 1‘1 .
U | |
~I ]I 7 A
i AT, AHIA A
N S
| | e
TO FUEL i ¥
DRAIN TANKS ™ COOLANT SALT TO

BLANKET HEAT EXCHANGER

 

Fig. 4.5. Revised Fuel Heat Exchanger for MSBER.

 

i
!
|
i
i
i
5
i

 
 

 

 

   

   

 
 
   

132
ORNL DWG 66-7109
FUEL PUMP . BLANKET PUMP
MOTOR SgyTEOL ROD MOTOR
IVE -

  
  

CONSTANT — &
SUPPORT |~P=T"
HANGERS

    
   
    
 
 
 
 
  
 
 
 

FUEL DUMP | o+ 4 ;
TANK WITH mw [ oy W FTTwhRTTTT - Lt
COOLING
COILS FOR
AFTER HEAT
REMOVAL

 

, FT. DIA.
. ¢ CORE

~ TOR
"+ VESSEL

   
 
 

, SALT
' %L DIST.
v - +. PLENUMS

 
     
    
  
  
     
    

)
. * . v .‘.' “ 'l ’
PRIMARY — SRR
HEAT | * LA
EXCH. * «' s
‘4 "‘ _ " .
' - ; . ‘. 'l:: L
Y LS.
¥ . B ..n
v - . “ &
< PR v CESE P TE) 5 '_::‘ ,.:"‘ » ¢ @ 3 . * & ah LR : . ,~.."' -' - . v
I o. *. - REACTOR CELL HEATERS " e o '.¢ ° . e . e ae " -, -
]. . . ‘A.'. 0.. S . h. [P 704.[\‘ & ! ’ o : * L s A v " .. \En L] n “: ot e

 

Fig. 4.6. MSBR Cell Elevation Showing Primary Heat Exchangers and
Their Placement.
 

Juizser

REACTOR . VESSEL

 

 

 

 

 

 

 

 

 

 

   

 

 

 

 

 

HO5°F

 

 

 

 

 

 

  
    

BLANKET SALT
DRAIN TANKS

LEGEND

BLANKET e o s
COOLANT —--—

 

STEAM -————
H,0 -
o _10% b/

" PecemmnPsia
MevreoemnnBlU./ 1D,

N . Fraeze Valve

    

FUEL SALT
DRAIN TANKS

Fig. 4.7.

 

Revised MSBR Flow Diagram.

 

 

     
  

  
      
   

  

 

 

 

 

 

 
   

 

 

 

 

 

 

 

 

 

 

 

 

NET EFFICIENCY 449 %

- .
ORNL DWG 66-7022
5.134 # _
f -
| 10.067 # 7152 # 1 1518.50-540p-1000°
| 550p - 1000°F F —— T - :
i i il424h-35|5r|000' i
> 1 ] i
. ' GEN.
3600p- 1000°F | , ; HP
201t Yue ) | surgiNg [P TURBINE [ 5272 Mwe
| | | Gross
BOILER | - . oo S -
SUPERHEATERS | 2seis| 3 |
850°F \ J\ )L 3 {" {'"
551.7° ! !
COOLANT SALT GEN.
PUMPS | TURBINE [ TuRBINE [] 5077 Mwe
vl Gross
130 f1 Ysde :
Q"uzs—? | L |
wor | REHEAT STEAM ; bl ;
| ! PREHEATERS CONDENSER 8 FEEDWATER
i 3800 p_foo o 1307.8h SYSTEMS
] | |m692h 3500p-866°F 5000 550.9°
= mash 1| ZOE|  geanl
57°"'°5°'Fj3\ usp-eesF 1 |
766.4h
BOSSTER MIXING TEE
PUMPS
COOLANT SALT PERFORMANCE
DRAIN TANKS NET OUTPUT 1,000 Mwe
GROSS GENERATION 1034.9 Wwe
BF BOOSTER PUMPS 9.2 Mwe
STATION AUXILIARIES 25.7 Mwe
REACTOR HEAT INPUT 2225 Mwt
NET HEAT RATE 7,601 Biu/kwh

eeT

 

 

 
 

 

 

e e i a2 s g £ o e o e AR e 0 Bt T B T e

134

Table 4.14. Pressures in Various Parts
of Revised MSBR Salt Circuits

Flow diagram given in Fig. 4.7

 

 

Nominal
Location Pressure
| (psig)
Fuel-salt system
Core entrance 50
Core exit 25
Pump suction 10
Pump outlet , 150
Heat exchanger outlet 60
Blanket-salt system
Blanket entrance | 66
Blanket exit 65
Pump suction 64
Pump outlet 155
Heat exchanger outlet 67
Coolant-salt system
Pump suction before boiler-superheaters 130
Pump outlet before boiler-superheaters 280
Inlet to fuel heat exchangers 220
Outlet from fuel heat exchangers 160
Outlet-inlet to blanket heat exchangers 142
Pump suction before reheaters 130
Pump outlet before reheaters 240

Reheater outlet 220

 

As given in Table 4.14, the minimum pressure difference between the
core and blanket regions is about 15 psi plus the static head differential
or a minimum total difference of about 30 psi. If it is desirable to
increase this pressure differential, the blanket-salt pump could be
changed so that it discharges into the reactor blanket'region, giving a
minimum differential pressure between the core and blanket fluids of about
120 psi. Whether this change is necessary or-whethér it would increase
the reactor vessel design pressure is dependent upon the safety criteria
that need to be satisfied. A design pressure of 150 psia was used in

determining the thickness of the MSBR reactor vessel.
 

 

 

135

5. ALTERNATIVE MOLTEN-SALT REACTOR DESIGNS

A number of possible molten-salt reactor designs were considered,
and some of these are discussed below. Generally, the alternative de-
signs were studied only in concept and not in detail, so the results are
more qualitative than those given previously. Also, the technology re-
quired for these alternative designs is relatively undeveloped, although
there are experimental data which support the feasibility of each con-
cept. An exception is the molten-salt converter reactor (designated
MSCR), which was studied in detail by Alexander et al.20 and whose appli-
cation essentially requires only scaleup of MSRE and associated fuel-
processing technology. However, the MSCR is not a breeder, although it
approaches a bréak—éven breeder system. It is included to place molten-
salt breeders and converters in perspective relative to nuclear perfor-
mance, fuel-cycle cost, and power-production cost.

The terminology employed for each design concept will be discussed
first, along with a summary of the associated design conditions and fuel-
cycle performance. Additional information for each concept is given in
individual sections below. 1In all cases, a 1000-Mw(e) power plant is
considered.

The designations MSBR(Pa) and MSBR have the same meanings as before
and represent the reference breeder reactor design with and without di-
rect protactinium removal from the blanket stream, respectively. The
MMSBR(Pa) designation also has the same meaning as before and represents
the modular version of the MSBR(Pa). These concepts were presented above
and are included here for-completeneés.-

The MSBR(Pa-Pb) designation refers to the MSBR(Pa) modified by use
of direct-contact cooling of the molten-salt fuel by molten lead. Lead
is immiscible with molten salt and can be used as a heat exchange medium
within the reactor vessel to significantly lower the fissile inventory
external to the reactor. The lead alsc serves as-a heat transport medium
between the reactor and the steam generators.

The SSCB(Pa) designation refers to a Single-Stream-Core Breeder with
direct protactinium removal from the fuel stream. This is essentially a

single-region reactor having fissile and fertile material in the fuel

 
s et e 1o A s e e e

e A L R s

 

136

stream, with protactinium removal from this stream; in addition, the
core region is enclosed within a thin metal membrane and is surrounded
by a blanket of thoriumrcontaining salt. Nearly all the breeding takes
place in the large core, and the blanket "catches" only the relatively
small fraction of neutrons that "leak" from the core (this concept is
also referred to as the one-and-one-half region reactor).

| The MOSEL(Pa-Pb) designation refers to a MOlten-Salt Epithermal
breeder haVing an intermediate-to-fast-energy spectrum, with direct pro-
tactinium removal from the fuel stream and direct-contact cooling of the

fuel region by molten lead. No graphite is present in the core of this

" reactor.

The MSCR refers to a Molten-Salt Converter Reactor which has the
fertile and fissile material in a single stream. No blanket region is
employed, although a graphite reflector surrounds the large core.

The design conditions and fuel-cycle performance for the above-
mentioned reactor concepts are summarized in Table 5.1; in all cases the
methods, analysis procedures, and economic conditions employed were
analogous to those used in obtaining the reference MSBR design conditions.
In general, fuel recycling was based on fluoride volatility and vacuum
distillation processing; direct protactinium removal from thé reactor

system was also considered in specified cases.

5.1 MSBR(Pa-Pb) Concept

The MSBR(Pa-Pb) concept is essentially identical to the MSBR(Pa)
concept, except that heat is removed from the fuel salt by direct contact

with circulating molten lead. The lead is pumped in a circuit external

to the reactor and transports the reactor energy to the steam-generating

equipment; the circulating-lead circuit takes the place of the coolant-
salt circuit used in the MSBR design.

A conceptual arrangement for this reactor is shown in Fig. 5.1.
The lead is discharged through many jet pumps located under the reactor
core; the aspirating action of the jet pumps causes circulation of fuel
salt through the fuel tubes of the reactor. To effect this action, each

inner fuel tube terminates below the core in a venturi head; lead, flowing
 

 

" ¥ Ry ¥

Table 5.1. Summary of Design Conditions and Fuel-Cycle Performance
for Reactor Designs Studied

 

Design Conditions

Reactor Designation®

 

 

MSBR(Pa.) MSBR MMSBR(Pa) MSBR(Pa-Fb) SSCB(Pa) MOSEL(Pa-Fb) MSCR
Dimensions, ft
Core
Height 12.5 12.5 7.9P 12.5 16.0 3.0¢ 20.8
Diameter 10.0 10.0 6.3b 10.0 9.8 6.5¢ 16.6
Blanket thickness :
Radial 1.5 1.5 2.0 1.5 1.2 3.0
Axjal 2.0 2.0 2.0 2.0 0.0
Volume fractions, core
Fuel 0.169 0.169 0.17 0.169 0.193 0.5 0.105
Fertile 0.073 0.074 0.05 0.076 0.0 0.0 0.0
Moderator 0.758 0.757 0.78 0.755 0.807 0.0 0.895
Salt volumes, ft3
Fuel
Core 166 166 166 166 230 63.5 476
External 551 547 574 110 600 0.7 654
Total 717 713 740 276 830 64.2 1130
Fertile, total 1317 3383 1570 1324 983 758 0.0

 

%See text for explanation of reactor designatiocns.

The core dimensions for this case refer to one module of a four-module station.

®For this case, the core had annular geometry; the fuel annulus inside diameter was
3 ft, and the outside diameter was 6.5 ft.

LeT

 
 

 

Table 5.1 (continued)

 

Reactor Designation®

 

Design Conditions

 

MSBR(P=a) MSER MMSBR(Pa) MSBER(Pa-Pb) SSCB(Pa) MOSEL(Pa-Pb) MSCR
Fuel-salt composition, mole %
LiF 63.6 63.6 63.6 63.6 71.0 71.0 70.0
ThF4 0-0 O-O ODO 0-0 8068 24-0 16-55
UF, (fissile) 0.22 0.23 0.21 0.23 0.23 5.0 0.45
Core atom ratios
Th/U 41.7 39.7 28.4 41.5 37.7 4.76 36.7
c/u 5800 5440 5980 5520 6280 0.0 6525
Power density, core average,
Xw/liter
Gross 80 80 80 80 66 618 17
In fuel salt 473 473 473 473 341 1236 165
Neutron fluxé core average,
neutrons/cm® «sec
Thermal 7.2 x 10 6.7 x 104 7.3 x 10*% 6.8 x 10** 6.1 x 10 0.0 x 10'* 1.9 x 10*4
Fast 12.1 x 104 12,1 x 10*%  11.7 x 10*% 12.1 x 10*% 10.0 x 10** 72.2 x 1014 2.7 x 104
Fast, over 100 kev 3.1 x 10 3.1 x 10*% 3.0 x 10*% 3.1 x 10** 2.6 x 10*% 23.3 x 10** 0.7 x 10*4
Weutron production per fissile 2.227 2.221 2.229 2.226 2.226 2.280 2,201

absorption (ne)

Nuclear and fuel-cycle performance

Fuel yield, % per year 7.95 4.86 7.31 17.3 6.63 10.3

Breeding ratio 1.07 1.05 1.07 1.08 1.06 1.14 0.96
Fuel-cycle cost, mill/kwhr 0.35 0.46 0.38 0.25 0.374 0.13 0.57
Specific fissile inventory, 0.68 0.77 0.76 0.34 0.684 0.99 1.63

kg/Mw(e)

 

dUse of direct-contact lead cooling would lower the fuel-cycle cost to about 0.32
mill/kwhr(e) and the specific fissile inventory to about 0.41 kg/Mw(e).

8eT

 
 

 

 

 

-

139

CONTROL ROD ORNI~-DWG 66-6677

  
 
  
 

REFLECT

10-0"DIA. X 10™-0"

LEAD LEVEL
(OPERA

  
 
 

FUEL
LEVEL

 
 
 
 
 
  
 
 
 

FUEL DUMP
TANK WITH
AFTER HEAT

BLANKET
TO HEAT
EXCHANGER

GAS

  
 
 
 
 

FUEL
LEVEL
(DUMP)

LEVEL
DUMP)

   

LEAD IN

 

| —— - e e S e g -

Fig. 5.1. Two-Region Circulating-Lead Reactor — Elevation.

upward to this point, discharges horizontally out of the venturi tube

and in the process draws fuel salt into the venturi to cause intimate
mixing’of the salt and lead. This mixing generates large areas for heat
transfer between the salt and lead and resulis in efficient heat exchange
between the two media. After passing through the venturi, the lead and
salt separate by gravity due to density difference, with the lead flowing

dovnward to the lead outlet lines.

 
 

140

The separated fuel salt floats on the lead and forms a 4-in.-deep
layer. The core fuel tubes are submerged in this salt layer, and open-
ings into their annular regions provide flow passages through which fuel
flows into the core wvolume.

There are no mechanical pumps in the reactor cell. The only heat
exchange within the reactor cell is that provided by the direct-contact
lead-and~fuel jet pumps. The only iiquid lines leaving the reactor cell
are the lead lines and the fuel-processing line, which communicates with
the fuel layer at the bottom of the reactor. The blanket probably would
be cooled with lead also; however, since the blanket volume is not criti-
cal, the blanket salt could be cooled by pumping it through a tube-and-
shell exchanger as in the MSBR.

Use of lead cooling requires niobium cladding of metal parts of the
system. However, this requirement does not appear to introduce a signifi-
cant economic penalty. At the same time the primary heat exchangers are
eliminated, with their attendant costs and dperating requirements.

The significant advantage produced by direct-contact cooling is the
reduction in fissile~-fuel holdup external to the core proper. As shown
in Table 5.1, the MSBR(Pa-Pb) concept has a very high fuel-yield rate of
about 17%VYT, corresponding to a fuel doubling time of 5.8 years.

5.2 8SCB{Pa) Reactor Concept

5.2.1 SSCB(Pa) Reactor Concept with Intermediate Coolant

In the single~stream-core breeder reactor, or one-and-one-half re-
gion reactor, the fuel salt contains fertile as well as fissile material.
Within the core proper there is no separation of fluids, so graphite tubes
of the type needed in the MSBR are not required. A thin metallic mem-
brane of Hastelloy N, niocbium, or similar structural material about 0.12
in. thick surrounds the core and separates the core region from the
blanket region. |

The reactor core is cylindrical and is about 14 ft high and about
10 £t in diameter. The core structure is an assembly of graphite blocks
with passages for flow of fuel salt. An annular, cylindrical graphite

barrier divides the core into two regions so that the fluid makes two
 

 

 

 

141

passes through the core. Leakage between the regions is permissible,
and therefore the barrier can be constructed by simply interlocking the
graphite sections.

The core structure is built on a tube-sheet-like support plate,
which also serves as a flow distributor for the incoming fuel salt and
a collector for the discharge stream. Below this plate are the plenum
chambers for fuel distribution. These plenums consist of a central cir-
cular region and an annular region, which are separated by a curtain-
like barrier. The center plenum directs the fuel to the central region
of the reactor, while the annular plenum receives fuel salt as it leaves
the annular region of the reactor core. Some bypass of fuel salt be-
tween the reactor inlet and outlet plenum chambers is permissible.

The energy generated in the fuel salt is transferred to an inter-
mediate coolant as in the MSBR concept. The steam-power cycle is also
the same as for the MSER.

The blanket region contains ThF,; in a carrier salt. Neutrons dif-
fusing from the core regioh are absorbed by thorium in the blanket to
produce about 5% of the bred 233y, Cooling of the blanket stream is
done in a manner similar to that used in the MSBR concept.

Direct protactinium removal from the fuel stream is an important
feature of this concept. The ability to do this practically in the pres-
ence of relatively hlgh uranlum.concentratlons has not been demonstrated
conclusively; however, the ox1de—pre01p1tatlon process shows promise of
being applicable to protactinium removal from molten salts containing

both thorium_and uranium.

5 2. 2 SSCBLP&) with D:Lrect Conta.ct Cooling

The performance of the SSCB(Pa) can be 1mproved if molten lead is
found to be practlcal as a direct~contact coolant for molten salts con-
rtalnlng thorium and uranlum This concept which is termed SSCB(Pa-Pb),
is shown in Flg. 5.2, which also 1llustrates features of the SSCB(Pa)
conqept As in the MSBR(Pa Pb) concept, the lead coolant not only ab-
sqrbe fihermal_energy from the fuel salt, but also supplies the motive

power for circulating the fuel salt through the core.

 
 

 

 

142

ORNL-DWG 66-6680
CONTROL ROD DRIVE

  

 

 

 

 

 

 

 

 

;:’/T—FUEL LEVEL

| ——LEAD LEVEL
_—LEAD RETURN

 

 

BLANKET RETURN

BLANKET TO !
HEAT EXCHANGER |

 

 

- —ILEAD TO PUMP AND
¢ |HEAT EXCHANGER

 

 

 

 

 

 

Fig. 5.2. One-and-One~Half Region Circulating-Lead Reactor — Ele-
vation.

As illustrated in Fig. 5.2, the reactor core is mounted above a
pool of lead. TFuel salt, which is floating on the lead, flows through
the suction pipes into the inlet plenum below the central region of the
core and then—through the core in a two-pass arrangement.

From the outlet plenum the fuel is channeled radially out and down-
ward to peripheral lead-activated ejectors. These ejectors discharge
the mixture of lead and fuel salt into the lead pool. Dfiring this con-
tact the cooler lead extracts heat from the fuel salt. In the pool, the

~less dense fuel salt rises to the top and is returned to the core. The

heated lead is piped away from the pool to a pump and is passed through
 

 

e

143

the steam superheaters and reheaters. Cool lead is returned to the
ejectors.

The blanket salt may be cooled in a similar fashion, as indicated
in Fig. 5.2, or the blanket salt may be passed through a shell-and-tube
heat exchanger cooled by lead returning to the fuel loop.

Direct-contact lead cooling reduces the external fuel inventory by
permitting efficient heat exchange in a system requiring short runs of
fuel piping. Although niobium is needed as a structural and/or cladding
material in systems containing lead, fewer heat exchangers may be re-
quired. As indicated in footnote d of Table 5.1, the SSCB(Pa-Pb) concept
gave a fuel-cycle cost of 0;32.mill/kwhr(e) and a specific fissile in-
ventory of 0.41 kg/Mw(e). |

5.3 MOSEL(Pa-Fb) Reactor Concept

The MOSEL reactor concept has no moderator (in the sense that no
material is introduced for moderating purposes) and operates in the
intermediate-to-fast energy range (mean fission energy of 10 to 20 kev).
The core contains only molten-salt fuel and the lead introduced for cool-
ing, while the blanket contains ThF, in a carrier salt. Niobium is used
as the structural'or'cladding material where there is the possibility of
contact with lead. ’

Figure 5.3 1llustrates the reactor concept; the core is toroidal in
shape, having a cross section about 3 £t wide by 4 ft high. The internal
diemeter of the torus is 4 ft. The core is in a tank of blanket salt,
and except for the lead outiet pool ét the bottom; is nearly surrounded
by blanket salt. ’ e

Lead is pumped in through a perforated header at the tqp of the
toroidal core. The lead falls through the fuel salt and extracts energy
from the core. In the process, the falling lead causes circulation of
fuel salt w1th1n the core region in a rotatlonal pattern, ‘with salt flow-
ing upward on each side of the central region. The central reglon con-
tains about 50 vol % lead, and the lead separates from the salt by
gravity, with the fuel salt floating on the Tead. "Although a protac-

tinium removal scheme was assumed in the nuclear design calculations,

 
 

144

ORNL-DWG 66-6673

T0 FutL
REPROCESSING 8

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

   

oo —mmmm= OFF GAS SYSTEM
LEAD COULANT™._ i
HEADER N TO BLANKET
“ REPROCESSING
.
~
COOLING LEAD
RETURN ™.
\\
ANNULAR CORE _
REGION = ™~
.
\\ el __ BLANKET
1 BLANKET REGlONS"—-\\ ‘ ~o ' Ty o / MAKE-UP
: \ ! ' ' ' ' =
| , v —
& v ' f
' : T v v oy ¢
i Y J\\' ' ' v
1 v v t \ I ' ' '
T 1 1 ' 1 ¥
t 1] v l ! ! Y
¥ .o v .
~I7,0" A ' EnS . AV
( v e , - v - FUEL TRAP
] . Al
L 1 ¥
— - Y
COOLING LEAD L —1 _
TOPUMP ] .
8 HEAT - o
B HEAT . SR FUEL MAKE-UP
-~ N o\ J
; AN
L . FUEL SALT
LEAD POUL - ' DRAIN LINE
! u
N—————— 3~ -

 

 

; LEAD AND BLANKET _ _
* SALT FILL 8 DRAIN

Fig. 5.3. MOSEL(Pa-Pb) Lead-Cooled Reactor — Elevation.

the reactor performance given in Table 5.1 would change only slightly
if fuel recycling was accomplished with only fluoride volatility and
vacuum distillation processing.

The design shown in Fig. 5.3 is conceptual in nature, and the actual

requirements for separation of the salt and lead phases may involve more

 
 

 

 

 

 

145

than simply separation by gravity forces alone. However, mechanical
methods of separation are permissible, and preliminary work indicates
that they are feasible. Although preliminary, the results obtained for
the MOSEL(Pb) concept indicate the potential performance of an inter-
mediate-to-fast energy molten-salt reactor and the versatility of molten
salts as reactor fuels.

These studies also illustrate that MOSEL-type reactors need efficient
methods for removing energy from the reactor core without requiring a
large fuel inventory external to the core, since the fissile concentra-
tion in the carrier salt is high (about 20 times higher than in a thermal
reactor). Direct-contact cooling with lead appears to lower the external
inventory requirements to a level sufficient for attaining low fuel dou-

bling times and low fuel-cycle costs.

5.4 MSCR Concept

The molten-salt converter reactor is a single-region single-fluid
reactor moderated by graphite, with the fertile material physically mixed
with the fissile fuel salt. The graphite is an arrangement of vertical
bars, with fuel passages permitting single-pass flow through the core.
The reactor concept is described in detail in the report by Alexander
et al.?0 The essential differences between the MSCR concept referred to
here and that described by Alexander et al. concern the steam-power cycle
and the processing scheme. In the previous report, a Loeffler boiler
was used in conjunction with a subcritical steam cycle, while here a
supercritical steam-power system and once-through boiler-superheaters
are considered that are identical to those given for the MSBR. These
changes substantially increase the thermal efficiency and lower the unit
capital cost of the previous MSCR plant. Also, the previous system did
not use vacuum distillation processing, since the discovery of its appli-
cation came at a later date. Incorporation of the vacuum distillation
process for carrier-salt recovery, as considered here, leads to substan-
tial improvements in fuel-cycle performance. The fuel~-cycle cost of the
MSCR concept is given in Table 5.1. The capital costs were not studied

specifically but should be comparable with those for the MSBR, that is,

 
 

 

 

146

about $ll4/kw(e). Assuming the operating and maintenance costs to be
0.34 mill/kwhr(e), as for the MSBER, gives power-production costs under
2.9 mills/kwhr(e) based on an investor-owned plant and a 0.8 load factor.
 

&Y

 

 

 

"

 

 

147

6. EVALUATION

Of the reactor designs and concepts considered in this study, the
MSBR(Pa) plant appears to have euperior power-production cost and nuclear
characteristics, as well as technology requirements that demand only a
reasonable amount of developmental effort. The estimated power-production
cost of 2.64 mills/kwhr(e) for investor-owned MSBR(Pa) plants with a load
factor of 0.8 indicates that their development can lead to large economic
savings. Also, the low specific inventory requirements (less than 1 kg
of fissile material per megawatt of electricity produced) and the low
fuel doubling time of about 12.6 years, which corresponds to a capability
for doubling the installed power capacity every 8.7 years; leads to ex-
cellent fuel-conservation characteristics.

The results obtained for the MSBR design indicate that this plant
also has good performance characteristics, although not so good as those
for the MSBR(Pa). At the same time, the MSBR plant appears'less demand-
ing of its fuel-recycle technology.

Molten-salt reactors appear well-suited for modular-type plant con-
struction. ©Such construction causes no significant penalty to either
the power-production cost or the nuclear performance, and it may permit
MSBR's to have very high plant-availability factors.

Use of direct-contact cooling of molten salts with lead signifi-
cantly improves the potential performance of molten-salt reactors and
indicates the versatility of molten salts as reactor fuels. However,
in order to attain the technology status required for such concepts, a
51gn1f1cant development program appears necessary. |

The molten-salt reactor concept that requires the least amount of
development effort is the MSCR, but it is not a breeder system. The
equilibrium.breeding ratio and thé power-production cost of the MSCR
plant were estimated to be about 0.96 and 2.9 mllls/kwhr(e), resPec—
tively, in an investor-owned plant with a load factor of 0.8. Although

this represents_excellent performance as an advanced converter, the de-

velopment of MSBR(Pa) or MSBR plants appears preferable because of the

lower power-production costs and superior nuclear and fuel-conservation

characteristics associated with the breeder reactors.

 
 

 

10.

11.

12.

13.

14.

148

References

W. R. Grimes, Chemical Research and Development for Molten-Salt

Breeder Reactors, ORNL~IM series report to be issued.

G. M. Adamson, Materials Development for Molten-Salt Breeder Re-
actors, ORNL-TM series report to be issued.

W. L. Carter, Fuel and Blanket Processing Development for Molten-
Salt Breeder Reactors, ORNL-TIM series report to be issued.

D. Scott, Components and Systems Development for Molten-Salt
Breeder Reactors, ORNL-TM series report to be issued.

J. R. Tallackson, Instrumentation and Controls Development for
Molten-Salt Breeder Reactors, ORNL~TM series report to be issued.

R. Blumberg, Maintenance Development for Molten-Salt Breeder Re-
actors, ORNL-TM series report to be issued.

R. W. Swindeman, The Mechanical Properties of INOR-8, USAEC Report
ORNL~-2780, Oak Ridge National Laboratory, January 1961. See also
Ref. 2.

R. C. Robertson, MSRE Design and Operations Report, Part I.
Description of Reactor Design, USAEC Report ORNL-TM-728, pp.
133-160, Oak Ridge National Laboratory, January 1965.

M. W. Rosenthal et al., A Comparative Evaluation of Advanced Con-
verters, USAEC Report ORNL-3686, Oak Ridge National Laboratory,
January 1965.

R. C. Olson and R. E. Hoskins, Capital Cost Breakdown for Compara-
tive Evaluation of Advanced Converters, unpublished internal docu-
ment, August 1964.

Guide to Nuclear Power Cost Evaluation, USAEC Report TID-7025,
September 1962.

Op. cit., Ref. 8, pp. 190-201.

B. E. Prince and J. R. Engel, Temperature and Reactivity Coefficient
Averaging in the MSRE, USAEC Report ORNL-TM-379, Oak Ridge National
Iaboratory, October 1962.

C. W. Nestor, Z@RCH — An IBM 7090 Program for the Analysis of Simu-
lated MSRE Power Transients with a Simplified Space-Dependent
Kinetics Model, USAEC Report ORNL-TM-345, Oak Ridge National Labo-
ratory, Sept. 18, 1962.
 

 

 

15.

16.

17.

18.

19.

20.

21.

22,

23.

24 .

25.

26.

27 .

28,

29.

149

T. W. Kerlin and S. J. Ball, Stability Analysis of the Molten Salt
Reactor Experiment, USAEC Report ORNL-TM-1070, Oak Ridge National
Laboratory, December 1965.

Oak Ridge National Laboratory, Molten-Salt Reactor Program Semi-
annual Progress Report for Period Ending February 1966, USAEC Report
ORNL~3936.

Y. G. Grunzweig et al., Molten-Salt Converter Reactor Hazards
Studies, ORSORT Reactor Hazard Evaluation Course Report, September
1963.

C. D. Scott and W. L. Carter, Preliminary Design Study of a Con-
tinuous Fluorination-Vacuum Distillation System for Regenerating
Fuel and Fertile Streams in a Molten Salt Breeder Reactor, USAEC
Report ORNI~-3791, Oak Ridge National Laboratory, January 1966.

W. L. Carter, personal communication, 1965.

H. F. Bauman, OPTIMERC, A Reactor Design Optimization Program,
ORNL~-TM series report to be issued.

A, M. Perry, Physics Program for Molten-Salt Breeder Reactors,
ORNL-TM series report to be issued.

Op. cit., Ref. 8, pp. 220-238.

T. W. Pickel et al., Design Study of a Heat Exchanger System for
One MSBR Concept, USAEC Report ORNL-TM-1545, Oak Ridge National
Laboratory, September 1966.

R. C. Robertson, MSBR Steam System Performance Calculation, un-
published internal document, July 5, 1966.

J. H. Griffin, Piping Flexibility Analysis Program for the IBM 7094,
USAEC Report IA-2929, Los Alamos Scientific Leboratory, July 1964.

Tnternal correspondence, A. G. Grindell to E. S. Bettis, Nov. 26,
1965.

J. H. Shaffer et al., The Recovery of Protactinium and Uranium from
Molten Fluoride Systems by Precipitation as Oxides, Nucl. Sci. Eng.,
18(2): 177-181 (1964).

W. R. Grimes et al., Reactor Chemistry Division Annuel Progress
Report for Period Ending January 31, 1965, USAEC Report ORNL-3789,
pp. 137 £f, Oak Ridge National Laboratory.

P. R. Kasten, Safety Program for Molten-Salt Breeder Reactor,
ORNL-TM series report to be issued.

 
 

150

30. L. G. Alexander et al., Molten Salt Converter Reactor Design Study
and Power Cost Estimates for a 1000 Mw(e) Station, USAEC Report
ORNL-TM-1060, Oak Ridge National Laboratory, September 1965.

 

 
 

 

s

eI O0vwui WMo

43 .
bite
45.
46-145.
146.
147.
148.
149.
150.
151.
152.
153.
154.
155.
156.
157.
158.
159.
160.
161,
lé2.
163.
164.
165.
166.
167.

151

Internal Distribution

 

R. E. Adams

L. G. Alexander
E. D. Arnold

C. F. Baes

S. J. Ball

W. P. Barthold

C. J. Barton

H. F. Bauman

S. E. Beall, Jr.
M. Bender

L. Bennett

C. BE. Bettis

E. S. Bettis

H. Beutler

J. E. Bigelow

A. M. Billings

F. F. Blankenship

C. M. Blood

E. G. Bohlmann
R. J. Braatz
E. J. Breeding
R. B. Briggs
0. W. Burke

S. Cantor

R. S. Carlsmith
W. L. Carter

C. E. Center (K-25)
R. L. Clark

W. G. Cobb

C. W. Collins

E. L. Compere

W. H. Cook

G. A. Cristy

¥, L. Culler

J. E. Cunningham
S. J. Ditto

E. P. Epler

W. S. Ernst, Jr.
A, P. Fraas

R. E. Gehlbach
E. H. Gift

W. R. Grimes

R. C. Gwaltney
o. H. Hanauer

168.
169.
170.
171.
172.
173-197.
198.
199.
200.
201.
202.
203.
204.
205.
206.
207.
208.
209.
210,
211.
212.
213,
214,
215,
216.
217,
218.
219.
220.
221,
222,
223,

224.

225,
226,
227 .
228,
229,
230.
231,
232,
233,
234.
235,

ORNL-3996
UC-80 ~ Reactor Technology
TID-4500
P, G. Herndon
H. W. Hoffman
G. H. Jenks
W. H. Jordan
Lincoln Jung
P. R. Kasten
C. R. Kennedy
T. W Kerlin
H. T. Kerr
S. 5. Kirslis
J. A. Lane
C. E. Larson (K-25)
R. B. Lindauer
A. P. Litman
D. B. Lloyd
A. L. Lotts
M. I. Lundin
H. G. MacPherson
R. E. MacPherson
C. D. Martin, dJr.
F. C. Maienschein
W. R. Martin
H. E. McCoy
H. F. McDuffie
D. L. McElroy
T. L. McLean
J. G. Merkle
H. J. Metz
A, S. Meyer
S. E. Moore
W. R. Mixon
J. F. Mardock
M. L. Myers
L. C. Oakes
R. C. Olson
A. M. Perry
T. W. Pickel
J. W. Poston
J. W. Prados
A, 5. Quist
J. L. Redford
R. C. Robertson
D. P. Roux
R. Salmon

 
 

 

152

236. Ann W. Savolainen 255. D. R. Vondy

237. C. D. Scott 256. D. D. Walker

238. D. Scott 257. T. N. Washburn

239, J. H. Shaffer 258. G. M. Watson

240, M. J. Skinner 259. J. R. Weir

241. G. M. Slaughter A 260. R. C. Weir

242. A. E. Spaller 26l. A. M. Weinberg

243, I. Spiewak 262, W. J. Werner

244. R. S. Stone 263. J. H. Westsik

245. J. C. Suddath 264. M. E. Whatley

246. J. R. Tallackson 265, J. C. White

247. W. Terry - 266. H. D. Wills

248. R. E. Thoma 267. L. V. Wilson

249. D. E. Tidwell 268. M. L. Winton

250. G. M. Tolson 269—271. Central Research ILibrary
251. D. B. Trauger 272—273. Y-12 Document Reference
252. R. W. Tucker Section

253. J. W. Ullmann 274—~514. Laboratory Records Department
254. W. E. Unger 515, Laboratory Records, RC

External Distribution

516. R. A. Charpie, UCC, New York
517. C. B. Deering, AEC, ORO
518. F. C. Di Luzio, Office of Saline Water, Department of Interior
519. M. C. Edlund, University of Michigan
520. E. A. Eschbach, Pacific Northwest Laboratory
521. H. Falkenberry, Tennessee Valley Authority, Chattanooga
522. A. A. Giambusso, AEC, Washington
523. R. E. Hoskins, Tennessee Valley Authority, Chattanooga
52%. Lenton Long, Pacific Northwest Laboratory
525. W. B. McDonald, Pacific Northwest Laboratory
526. M. W. Rosenthal, AEC, Washington
527. S. Sapirie, AEC, ORO
528. M. Shaw, AEC, Washington
529. W. L. Smalley, AEC, ORO
530. J. A. Swartout, UCC, New York
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