ORNL-2796.txt

ORNL.-2796

Contract No. W-7405-eng-26

REACTOR PROJECTS DIVISION

EXPERIMENTAL MOLTEN-SALT-FUELED 30-Mw POWER REACTOR

. C. Alexander

L

B. W

M. E. Lackey
H. G

. MacPherson

DATE ISSUED

MAR 8

J. W. Miller

F. C. VonderlLage
G. D. Whitman

J. Zasler

1960

 

OAK RIDGE NATIONAL LABORATORY
Oaok Ridge, Tennessee

UNION CARBIDE CORPORATION

U.S. ATOMIC ENERGY COMMISSION

 
 

CONTENTS

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General Description of Reactor and Plant Layout........coriiiiieeeeecee e ]
Molten-Salt System AuUxiliQries .....ccoiieiieercecre e et e e e b e e s et sb e et er b s enenenas 10

Enriching and Sampling Sy stem. ...t e e st e e e sebee e ceaeerste e ebe e e sebebre s 10

Fill and Drain Sy stemu. i it ettt sttt et et e e te b s e s asee st araseaseesaessneetesessenseeseesnseseans 13

Preheating and Afterheat Removal Equipment ..ottt 14

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Molten-Salt Pumps and Heat Transfer EqQuUipment ..ottt s 16
Steam Generating Equipment and Turbine-Electric System ........ccociiiniii 20
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EXPERIMENTAL MOLTEN-SALT-FUELED 30-Mw POWER REACTOR

L. G. Alexander
B. W. Kinyon

M. E. Lackey
H. G. MacPherson

J. W. Miller
F. C. Vonderlage

G. D. Whitman
J. Zasler

ABSTRACT

A preliminary design study has been made of an experimental molten-salt-fueled power

reactor. The reactor considered is a single-region homogeneous burner coupled with a Loeffler

steam-generating cycle.

Conceptual plant layouts, basic information on the major fuel circuit

components, a process flowsheet, and the nuclear characteristics of the core are presented.

The design plant electrical output is 10,000 kw, and the total construction cost is estimated

to be approximately $16,000,000.

 

INTRODUCTION AND CONCLUSIONS

The molten-salt-fueled reactor system described
in this report represents a preliminary design and
is to be considered as a reference design for
further experimental molten-salt-fueled reactor
studies. Designs have been developed for the
major components which are sufficiently detailed
to permit an initial evaluation of costs and con-
struction problems. Information on nuclear per-
formance was obtained to give a basis for study
of the major problems involved in operating a
molten-salt-fueled power reactor.

The molten-salt concept has been considered
for a variety of reactor types. These may be
classified as homogeneous or graphite-moderated
systems, in the main, with a variety of modifi-
cations depending on particular objectives.
Single- and two-region homogeneous reactors have
been considered in burner and converter cycles,
and, most recently, unclad-graphite-moderated
reactors have been emphasized for breeding
cycles. The major aim of this study was to design
a nuclear power plant which, with a minimum of
developmental effort, could be built in the near
future and would provide considerable information
applicable to larger or more complex molten-salt
power plants. With this objective in mind, a
single-region homogeneous burner employing a
semidirect-maintenance concept, operating on
a Loeffler steam cycle, and producing 10 Mw of
electricity was chosen as a reference design.

The system embodies a simple core from which
heat is transferred through a coolant salt to a
steam system to produce a useful amount of power

by a proved cycle. The semidirect-maintenance
philosophy circumvents the more complicated oper-
ations involved in complete remote-maintenance
schemes, but does not prevent the replacement of
the more vulnerable components of the fluid-fuel
system.

This reactor plant could be used to demonstrate
the nuclear performance of a homogeneous core,
the reliability of components, the stability of the
molten-salt fuels, the handling of gross quantities
of molten salts, the chemical processing of
fissioned fuel, the containment and processing
of fission gases which would be stripped from
the fluid fuel, the application of a particular high-
pressure, high-temperature steam cycle to the
molten-salt system, and the practicality of certain
maintenance techniques. In addition to the
primarily technical data, cost information would
be obtained on the construction and operation of
a molten-salt-fueled power reactor. |t has been
estimated that this experimental reactor would
cost approximately $16,000,000 and that about
two and one-half years would be required for
construction. The general characteristics of the
system are listed in Table 1.

GENERAL DESCRIPTION OF REACTOR
AND PLANT LAYOUT

The molten-salt-fueled reactor system adopted
for this study incorporates a complete electric
generating station but does not include on-site
fuel reprocessing facilities or a major-equipment
remote-maintenance repair shop. The successful
operation of an experimental reactor would not

 
Table 1. Reactor Plant Characteristics

F uel

Fuel carrier
Neutron energy
Moderator

Primary coolant

Power
Electric

Heat

Estimated cost
Refueling cycle
Shielding
Control

Exit fuel temperature

Steam
Temperature

Pressure
Intermediate coolant

Structural materials
Fuel system
Coolant system

Steam superheater

Core diameter
Temperature coefficient of reactivity, (Ok/k)/°F
Specific power

Power density (core)

Fuel inventory
Initial (clean)

After second year

Critical mass (clean)

depend on continuous fuel reprocessing, and,
indeed, long-term exposure of the fuel carrier
would be required to achieve fission product
concentrations comparable to those expected in
prototype molten-salt-fueled power systems. Small
quantities of fuel could be withdrawn from the

>90% UZ3F,

63 mole % LiF, 37 mole % BeF2
Intermediate

Lil"'--Bel"'2

Fuel solution circulating at 1480 gpm

10 Mw
30 Mw

$16,000,000

Semicontinuous

Concrete

Temperature and fuel concentration

1235°F

1000°F
1450 psia

65 mole % LiF, 35 mole % BeF2

INOR-8
INOR-8
INOR-8

6 ft

~6.75 x 10~>
417 kw/kg
9.4 kw/liter

65.8 kg of U235

107 kg of U233

40.6 kg of U239

drain system and transferred to shielded flasks,
which could then be transported to the ORNL
facilities for experiments in chemical reproc-
essing.

One of the primary objectives of the reactor
experiment would be to determine the metallurgical
and structural reliability of components. Major
remote repair facilities were not included in the
design, but on-site decontamination and de-
structive disassembly facilities were provided so
that manageable specimens of the fuel components
could be obtained and transported to ORNL hot
shops for inspection. |t is intended that part of
the hot storage cell be used for such work.

The fuel, LiF-BeF,-UF,, is circulated in a
single-unit primary container by a centrifugal
pump, and a single coolant-salt system containing
LiF-BeF, is used as the heat transfer coupling
between the fuel and the steam superheater. A
salt coolant has several advantages over a liquid
metal coolant in this application, such as com-
patibility with the fuel salt, lower induced radio-
activity coupled with a much faster decay rate, !
nonflammability, and elimination of cold-trap cir-
cuits. The major disadvantage of the coolant salt
is a relatively high liquidus temperature (865°F)
which increases the preheating load and increases
the danger of freezeup at off-design conditions.

A Loeffler steam cycle is used because of its
unique adaptability to the molten-salt-fueled re-
actor. The relative merits of this system have
been described in connection with earlier molten-
salt-fueled reactor studies. ?

The primary fuel container is made up of a
single structure which forms the core, heat ex-
changer, expansion tank, and fuel pump wvolute.
This assembly, along with the gas heating and
cooling jacket, is shown in Fig. 1. The primary
fuel container weldment is shown in Fig. 2.

The fuel is recirculated through the system by
means of a sump-type centrifugal pump. The fluid
enters the upper heat exchanger header from the
pump discharge and flows down through the tubes.
After passing through the heat exchanger, the fluid
enters the 6-ft-dia core via the central pipe and
flows to the bottom of the sphere. The flow then
reverses, washes the reactor vessel wall, and
returns to the expansion tank region through the
annulus surrounding the heat exchanger assembly.
The flow is directed into the pump by means of
a horizontal baffle located on the level of the

 

D, J. McGoff et al., Activity of Primary Coolant in
Molten-Salt Reactor, ESP-X-400, Oak Ridge, Tenn,,
MIT Practice School, UCNC-ORGDP (1958).

24, G.MacPherson et al., Molten-Salt Reactor Program
Status Report, ORNL-2634 (Nov. 12, 1958),

pump inlet. The entire upper region of the fuel
container assembly serves as a fuel expansion
tank.

Some flow is directed up out of the heat ex-
changer inlet header for fission-gas removal and
is recirculated in the expansion tank through the
pump. The fission gases are purged from the fuel
and held temporarily in tanks located in the
reactor cell. A dip line is inserted to the bottom
of the core for fuel filling and draining.

The heat exchanger and pump are removed or
inserted from above. These major components
can be changed without breaking liquid seals,

‘since their closures are made in a gas volume

above the fuel level.

The fuel container is surrounded by a jacket of
stainless steel, which serves as a container for
forced-circulation gas preheating and cooling.
The cooling unit was added so that afterheat
could be removed, in an emergency, without
draining the system. The package comprising
the gas blower, heater, and cooler is arranged to
be removed vertically from above, as in the case
of the heat exchanger and fuel pump.

The entire assembly is housed in a two-layer,
gas-buffered steel enclosure, which is surrounded
on the sides by 8 ft of concrete shielding. The
heat exchanger and pump project upward through
S5 ft of concrete so that semidirect maintenance
operations may be performed in a shielded area.
A multilayer organic-cooled boron steel shield
surrounds the reactor to reduce the radiation level
in the reactor cell.

The coolant salt, which forms the intermediate
coupling between the fuel and steam systems, is
directed into the heat exchanger through a central
pipe. This fluid flows on the outside of the tubes
and out of the heat exchanger through the annular
pipe.

An elevation drawing of the plant is shown in
Fig. 3, and two plan views at elevations 818.0
and 846 ft are shown in Figs. 4 and 5, respec-
tively. A perspective view of the building is
shown in Fig. 6.

The area immediately above the reactor cell is
the base for all fuel system maintenance oper-
ations. Contaminated parts may be removed from
this maintenance bay to a hot storage cell, which
is included in the building. The maintenance bay
is shielded and is sealed off during reactor
operation.

 
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The intermediate-coolant pump and the steam
superheater are housed in a separate shielded
area which is accessible from above. Direct
maintenance is planned for this portion of the
system. As previously mentioned, the induced
radioactivity in the coolant salt would decay
rapidly, and direct maintenance could be performed
after a few minutes’ delay to allow the 1l-sec
fluorine activity to decay.

The Loeffler system evaporators and steam
pump are located in the turbine-generator bay.
The bridge crane, in the turbine-generator bay, is
located to service the Loeffler components, feed-
water heaters, deaerator, and turbine-generator
system.

The control room is located between the reactor
area and the turbine bay. Office and laboratory
space is located to one side of the plant and
extends over four floors. An extended work area
surrounds the reactor cell on three sides at the

834.5-ft elevation.

A site was chosen along the Clinch River at
Gallaher Bend in the Oak Ridge area. The site
plot is shown in Fig. 7. The plant is located on
sloping terrain, which would minimize the ex-
cavation requirements, since the general arrange-
ment placed the reactor cell well above the
condenser to prevent the possibility of flooding.
This layout minimizes cooling water pumping
requirements, but increases the building costs.
This relationship has not been optimized, and
the arrangement presented was selected as a
reasonable first approximation.

MOL TEN-SALT SYSTEM AUXILIARIES
Enriching and Sampling System

Fissionable material is to be added to the
circulating fuel on a semicontinuous basis to
sustain criticality, within design temperature
limits, and fuel samples are to be withdrawn
throughout the course of operation for chemical
assay. A relatively small volume of fissionable
material will be added at any one time, and a
comparable, or smaller, volume will be withdrawn
as a sample.

A single mechanism is provided to accomplish
both these operations. The fuel will be added
as solid U235F4 at a free liquid surface in the
fuel system expansion tank or in a separate vessel
in which a free surface is presented. Sampling
will be accomplished by reversing this operation.

10

That is, a small portion of the circulating fuel
will be removed by ‘‘thief sampling,’’ allowed to
solidify, and then placed in a shielded container
for transfer out of the area.

It is estimated that 70 kg of would have to
be added to the system during the first two years
of power operation (30 Mw at 100% plant factor)
to take care of burnup and fission product over-
ride.  This quantity is equivalent to a daily
addition of 96 g of U?33. Considering the density
of UF, to be 6.70 g/cm®, it may be seen that
approximately 14.3 cm? of UF, would have to be
added daily.

The frequency of additions would be contingent
on the allowable mean temperature deviation in
the fuel system. The quantity of fissionable
material, AM, consumed per day was calculated,
as mentioned above, to be 96 g. Without fuel
additions, this would result in some mean temper-
ature decrease of the fuel, AT . These quantities
can be related as follows:

U235

AM
AT = ——
m  aBM

/

where a is the temperature coefficient of reactivity
per °F, B is the mass reactivity coefficient, and
M is the critical inventory in grams. For a temper-
ature coefficient of —6.75 x 10~>, an assumed
mass reactivity coefficient of 9, and a critical
inventory of 90 kg, a mean temperature drop of
less than 2°F per day would be experienced. The
system could be operated at full power for several
days before the decrease in fuel temperature would
seriously affect the thermodynamics of the power
cycle.

Several experiments have been run in which
slugs of UF, weighing approximately 40 g have
been inserted into 1200°F LiF-BeF, salt to
study solution rates and homogeneity of the UF,
in the melt.® The UF, slugs were held in per-
forated copper tubes, and the tubes were dipped
into the salt mixture. The results of these tests
indicate that solid UF, dissolves quite rapidly.
The chemical assays of the resultant mixture gave
excellent material balances.

The fuel sampling and enriching device is
shown schematically in Fig. 8. The entire

 

Zzgm Quar. Prog. Rep. Apr. 30, 1959, ORNL-2723,
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UNCLASSIFIED
ORNL-LR-DWG 37811A

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UNCLASSIFIED
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Fig. 8. Fuel Sampling and Enriching Mechanism.

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mechanism is enclosed in a vacuum-tight struc-
ture. All the material transfers in and out of the
- reactor system are made at ‘‘air lock’’ sections
which can be purged and isolated from the process
and cell side of the equipment. This equipment
would be maintained by the semidirect type of
techniques employed on the other major com-
ponents of the fuel system.

The mechanism consists of a sample carrier
elevator, enricher elevator, horizontal conveyer,
and reactor sample enricher elevator. The ele-
vator sections transport the capsules between
their respective containers and the horizontal
conveyer section which interconnects the three
vertical sections. Actuation of these elevator
elements is accomplished by motor-driven gear
packages.

Isolation of the various sections of the mecha-
nism is accomplished by remotely operated ball
and gate gas valves. Gas connections made
between pairs of valves allow contaminated gas
to be purged from the connections before dis-
assembly and inert gas to be charged into the
system after a connection has been made. All
operation must be remotely controlled and will
be interlocked to prevent improper manipulation
of the equipment. Preliminary layout drawings
have been made of this mechanism, and details
for the gas valve operators and capsule handling
have been developed.

Fill and Drain System

A fuel fill and drain system having a storage
capacity of approximately 400 3 is provided.
This system is made up of four vertical cylinders
2.5 ft in diameter and 20 ft high. The drain vessel
size was selected to give adequate heat transfer
surface for afterheat removal in a subcritical
geometry. These vessels are manifolded in two
pairs so that a spare fuel system inventory of
salt can be made available in the plant. This
excess salt can be used for system flushing
during startup operations or system cleanup during
any phase of the operation.

Fill and drain lines extend to the bottom of each
vessel and are coupled to the fill and drain line
in the reactor vessel. Fuel is transferred between
the reactor and drain tanks by a pressure-siphon
principle. Each pair of drain vessels is mani-
folded through two mechanical valves in series.
A gas-connected surge chamber is located between

each pair of valves. Fluid is transferred from the
drain vessels to the reactor or vice versa by
applying differential gas pressure to establish
flow. During this operation the mechanical valves
are open and the gas vent to the surge chamber
is closed. When this gas vent is opened and
pressures in the reactor and drain gas systems
are equalized, flow is stopped and further fluid
transfer is impossible. The mechanical valves
may then be closed. They would not normally
be exposed to liquid. The surge chamber also
provides a source of buffer gas between the
mechanical valves when the fluid is in the drain
system and the reactor vessel must be opened
for maintenance.

The preheating and afterheat removal schemes
are discussed in the next section of this report,
and the fuel temperature rise from afterheat
production for various values of heat loss are
plotted against time after shutdown in Fig. 9.
The data presented are based on one year of
operation at full power; no credit was taken for
fission-gas removal.

It was assumed that the fuel temperature rise
would be limited to 1500°F and that a heat sink
of 300 kw would be adequate to maintain the
temperature below this maximum, even with an
excursion of 100°F resulting from fuel flow
stoppage. The effects of a flow stoppage incident
are discussed in the section on ‘‘Reactor Haz-
ards.”’

A preliminary check of the criticality problem
in the drain vessels showed the system to be
safe. One hundred kilograms of U3 was con-
sidered to be contained in a single vessel and

UNCLASSIFIED
ORNL-LR-DWG 43506

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

— 500 "O% ~
L Q °
> L7 < \,OV
w % w2
A 400 X W
@ < \O(y
& 300 -~ 055 T——
: ABE
o -
'%J 200 7 ~ i ‘
L
= 400 v 300 kw HEAT Loss
0
0 i 2 3 4 5 6 7 8 9 10
TIME AFTER SHUTDOWN (hr)
Fige 9. Fuel Temperature Rise from Afterheat Pro-

duction. Running time, one year; volume of salt, 175 ft3;
heat capacity of salt, 64 Btu-ft_so(oF)-l.

13

 
was presumed to be settling out of the carrier
salt as an oxide compound, so that all the
fissionable material would be contained in the
lower part of a vessel and the top surface would
be reflected by the carrier salt. When the u23°
was contained in the lower half of the vessel, a
10-ft-high by 2.5-ft-dia volume, the multiplication
constant was 0.596. When the U%3° was con-
sidered to be contained in the bottom as a cube
2.5 ft on a side, the multiplication constant in-
creased to 0.827. This value was considered
uncomfortably high; however, the drain-vessel
diameter could be reduced at the bottom to offset
this rise in multiplication.

Fuel carrier and spent fuel salt would be loaded
into and out of the plant via the drain vessels at
a fuel transfer area adjacent to the drain vessels.
Remote liquid connections, probably flanges,
would be provided in the lines used to transfer
radioactive material out of the plant. |t was
presumed that freeze-flange junctions could be
used in this application. These transfers would
be made infrequently, and the transfer equipment
would not be used for long periods.

Estimates have been made of the shielding
required for carrier flasks containing 2 f13 of
fuel.  The radiation level for various shield
thicknesses is shown in Fig. 10. This calculation
was made on the basis of one year of operation
exposure at full power with 2.32 days of cooldown
and no self-shielding in the fuel.

Preheating and Afterheat Removal Equipment

The liquidus temperatures of the fuel and
coolant salts are above 800°F; therefore, means
are provided for preheating all the process equip-
ment containing these materials.

Those components which contain the fuel salt
are preheated, in the main, by forced gas circu-
lation. This scheme eases the remote-maintenance
problem in that a single, replaceable gas heating
and blower package may be used to heat a large
portion of the fuel system.

The main preheating system for the reactor is
shown in Fig. 1. This equipment includes a
blower, a heater, and a cooler section. The cooler
section was added to take care of emergency
situations when it would be desirable to remove
fission product decay heat without transferring
the fuel to the drain tanks.

UNCLASSIFIED
3 ORNL-LR-DWG 43507
10 —

 

 

 

 

 

DOSE RATE (rads/hr
S,

 

 

 

 

 

 

 

 

 

 

6 8 10 12 14 16 18
SHIELD THICKNESS (in.)

Fige 10. Dose Rate from Carrier Flask. Fuel cyl-
inder, 13,5 in. ID and 24 in. high.

The heater-cooler package, shown schematically
in Fig. 11, is connected to a thermally insulated
stainless steel jacket which surrounds the entire
reactor assembly. A centrifugal blower capable
of delivering 3000 cfm with a pressure differential
of 5 in. of water at a gas temperature of 1200°F
recirculates gas through the heater-cooler package
and then around the reactor assembly.

One hundred eighty kilowatts of heater capacity
is designed into the heater sections in the form
of high-temperature (1500°F) alloy sheath tubular
elements. A water-cooled fin-and-tube heat ex-
changer capable of handling 250 kw is used as
a heat dump. Hydraulically operated baffles are
manipulated to direct the gas through either the
heater or cooler unit. A preliminary estimate
indicates that a cold reactor system can be
preheated to 1000°F in less than two days by
using forced air circulation.

Resistance-type heater elements are provided in
the upper section of the fuel heat exchanger.

<y

{l
 

Gl

 

 

METAL BELLOWS

GAS FLOW

DAMPERS '

REMOVABLE
PACKAGE

BUFFER
PASSAGE

  
   
   
   
 

UNCLASSIFIED
ORNL-LR-DWG 39469

UICK CONNECTS

    
  
  

BLOWER -

HIGH TEMPERATURE
DUCT HEATERS

ATER COOLER TUBES

REMOVABLE PACKAGE
HEAT DUMP

Fig. 11. Reactor Preheating and Cooling Package.
Similar preheating units are included in the fuel
pump region and in the removable portion of the
pump. These heaters may be replaced from above
the top shield.

The fuel drain lines and drain vessels are
preheated by circulating gas systems. The drain
vessels have heater-cooler packages similar to
the reactor package.

The coolant-salt system is preheated with re-
sistance heaters attached to the lines and in-
dividual components. These elements will be
maintained directly, and no particular service
problem is envisioned.

Off-Gas System

As previously mentioned, a portion (approxi-
mately 10%) of the fuel flow is bypassed through
a gas stripping system. Directly above the fuel
heat exchanger inlet header there are nozzles
located in a plate to jet the fuel into the gas
space.

Helium is admitted into the expansion tank
from a fuel pump purge at a rate of 1 scth. This
gas and the gaseous fission products stripped
from the recirculating fuel are taken out of the
fuel surge tank to primary holdup tanks, which
are located in the main reactor cell. The effluent
from these tanks is then circulated out of the
system into two stages of cooled charcoal beds
which hold up the krypton and xenon. The purge
gas, then essentially free from activity, is purged
back into the reactor expansion tank. Parallel
installations of charcoal beds and gas recircu-
lating pump are installed in the system to ensure
continuous operation.

MOLTEN-SALT PUMPS AND HEAT TRANSFER
EQUIPMENT

The flow diagram for the experimental reactor
is shown in Fig. 12. Fuel is circulated in the
reactor system by means of a sump-type cen-
trifugal pump. This pump has a salt-lubricated
lower bearing which is submerged in the fuel.
The upper bearing assembly is oil-lubricated and
includes a radial bearing, a thrust bearing, and a
face seal. The electric motor rotor is on the shaft
above the upper bearing. The rotor is canned,
so that the field windings may be replaced without
breaking a reactor seal. A shield section is
installed between the fuel surface and the upper
bearing assembly. Cooling must be provided for

16

the shield and shaft. The entire rotating pump
assembly may be removed as a unit, while the
volute section is a part of the fuel surge tank
and reactor assembly.

The coolant pump is of a similar design,
modified for operation at a lower radiation level.
Since less shielding is required, the over-all
height of the coolant pump is less than the height
of the fuel pump. Heat is exchanged from the
fuel to the coolant salt in a bayonet-type heat
exchanger. The fuel salt is contained in the
tubes, and the flow is countercurrent. The heat
exchanger is shown in Fig. 13, and the design
data are presented in Table 2.

The tubes are coiled in helices and are located
in nine concentric rows around the coolant-salt
inlet pipe. The fuel salt enters the upper header
from the fuel pump and flows down through the
tubes and out of the heat exchanger. The tubes
were coiled to reduce the length of the heat
exchanger and to increase the flexibility of the
tube bundle.

Fuel bypass leakage around the heat exchanger
is reduced by a close-fitting section between the
heat exchanger outer shell and the reactor weld-
ment. Leakage from the top of the heat exchanger

Table 2. Design Data for Heat Exchanger

Fuel salt
In 60 psia, 1235°F
Out 25 psia, 1100°F
Flow 3.30 cfs
Coolant salt
In 90 psia, 1000°F
Out 70 psia, 1160°F
Flow 2.17 cfs
Configuration Multiple coiled helices in multiple

concentric rings in annular shell

Flow Countercurrent, with fuel salt in
tubes
Tubes
Size %y-in. OD, 0.049-in. wall
Number 450
Active length 12.6 ft
Pitch 0.875 in. center line to center line,
rings

0.750 in. center line to center line,

tubes in rings (normal to tubes)

&
Ll

 

UNCLASSIFIED
ORNL—LR—DWG 39470A

 
    

 

 

 

  

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

   

 

 

 

 

 

 

 

 

 

 

 

 

  

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

1800 Ib/hr
}
70 psia 247 cfs N 326,700 Ib/hr fi - 96,3001b 1000 F TURBINE -GENERATOR
1 1160° F i R ! 1010° F 1 | 1450 psi -
ATTEMPERATOR T —
SRIMARY 232,200 Ib/hr
HEAT ‘gD
EXCHANGER fi -
\
REACTOR SUPERHEATER — -— '
(COOLANT-STEAM) v
. —
1000°F J A
1100 °F 90 psia - -
3.30 cfs COOLANT 328,500 Ib/hr
PUMP 600°F 1474 psia CONDENSER
STEAM 960 Ib/hr 1
PUMP \ | MAKEUP ~——
13,500 Ib/hr fiil
»
592° F
345°F 253° F 169°F 92°F
EVAPORATOR & ! 1 1 ! 1
97,2601b 450°F '
/o - DEAERATOR
: ' CONDENSATE
PUMP
960 1b/hr / FEED-WATER
BLOWDOWN CLOSED FEED-WATER ’ PUMP '
HEATERS — =
1
CLOSED FEED-WATER

 

 

 

HEATER

Fig. 12, Schematic Diagram of Heat Transfer System,
UNCLASSIRED
ORNL-LR-DWG 39164

 

]
;

 

A-A

VIEW

1)

Fige 13. Fuel-to-Coolant-Salt Heat Exchanger Bundle.

18
back into the surge tank is restricted by a
Belleville-type spring seal between the upper
section of the heat exchanger and the fixed inner
housing around the exchanger shell. The volume
immediately above the fuel inlet header is part
of the fuel off-gas system. Some fuel is bypassed
up into the surge volume and then flows back
into the free surface of the expansion tank, where
it is recirculated through the pump.

The coolant salt enters the heat exchanger
through the inner tube and flows down the center
of the heat exchanger. This fluid then flows out
above the lower tube sheet and up through the
heat exchanger on the outside of the tubes. Five
vertical baffles are located on the shell side of
the exchanger, so that the coolant salt tends to
be directed across the tubes. The coolant then
flows across the bottom of the upper tube sheet
and out of the heat exchanger in the annular
section located on the outside of the inlet pipe.
Two thermocouple wells are included in the heat
exchanger assembly to measure the fuel temper-
ature into and out of the unit. These thermo-
couples can be inserted into or withdrawn from
the wells from the top of the reactor shield. The
entire assembly is bolted down against a lower
buffer seal mounted in the reactor container, and

COOLANT SALT IN

a seal weld is made at the very top of the heat
exchanger.

The steam superheater is a U-tube, U-shell
design with steam in the tubes. The superheater
is shown in Fig. 14, and the design data are
presented in Table 3. The shell side of the

Table 3. Design Data for Superheater

Coolant salt

65 psia, 1160°F

Out 37 psia, 1000°F
Flow 2.17 cfs

Steam
In 1473.5 psia, 600°F
Out 1450 psia, 1010°F
Flow 326,700 Ib/hr

U-tube, U-shell

Configuration

Flow Countercurrent, steam in tubes
Tubes

Size 5/8-in. OD, 0.065-in., wall

Number 311

Active length 23.0 ft

Pitch 0.875 in. triangular

Total length 7150 ft

Surface area 926 12

UNCLASSIFIED
ORNL—LR—DWG 39169A

 

SUPERHEATED
STEAM OUT

 

 

 

 

  
     
  
 
  
      
   
 
   

|

*z

 
 

 
 

 
     

 

SATURATED
STEAM IN

:::f'

|
fi‘f.

" 3

i
a‘fff:if |
i
|

Fom——— )

‘l
4‘

1
|

——————————— e ey
e e ey e
e e e e e

 

0

I

|

  
  
 

I ———

 

  

 

COOLANT SALT OUT

 

Ll 2 & Z

Fig. 14. Coolant=Salt=to-Steam Superheater (Schematic).

19

 
superheater is baffled for cross flow, and the
salt is separated from the high-pressure steam
by a single wall.

An emergency dump system is tied into the
coolant-salt plumbing. This system is required
to protect the coolant-salt circuit, in particular
the fuel-to-salt heat exchanger, from a pressure
failure if a gross rupture occurs in the steam
superheater. A drain vessel is connected to the
coolant system through a rupture valve, which
is designed to fail at an off-design high pressure
that the coolant system could safely contain for
short periods of time.

STEAM GENERATING EQUIPMENT
AND TURBINE-ELECTRIC SYSTEM

The two major components of the Loeffler
system are the evaporator drums and the steam
pump. Saturated steam leaving the evaporators
is pumped through the superheater, and a portion
of the superheated steam is recirculated back
into the evaporators in a regenerative cycle to
produce steam.

Two evaporator drums 4 ft in diameter and 20 ft
long are used for steam generators. These drums
are half filled with water, and steam nozzles
project down into the water for direct-contact
heat transfer. These vessels have been sized
to give a maximum liberation rate of approximately
10 cfm per square foot of water surface.

A bypass line around the superheater is included
to attemper the steam entering the tubrine. The
steam temperature rises with a decrease in load,
and de-superheating is required to keep the steam
temperature within tolerable limits.

A recirculating superheated steam connection
is located ahead of the steam pump to ensure
that no moisture enters the superheater. As
designed, slightly superheated steam leaves the
pump at all times. The heat added to the system
by the steam pump was not considered in de-
termining the mass flows shown in Fig. 12,

Steam at 1450 psia and 1000°F is supplied to
a 3600-rpm condensing turbine. There are four
stages of feed-water heating, and the turbine
heat rate is 9220 Btu/kwhr. With a net efficiency
of 33.3%, the net electrical power output of the
plant is 10 Mw.

20

REMOTE MAINTENANCE

The major requirement for the maintenance of
the reactor is the ability to remove and replace
limited-life equipment in the radioactive portions
of the complex, that is, equipment such as pumps,
heat exchangers, preheating packages, instru-
ments, and valves. The experimental molten-salt
reactor has been designed to achieve such main-
tenance by semidirect means. Replaceable equip-
ment is located so that it penetrates the top

shield and may be severed from the system by

direct contact outside the concrete primary reactor
cell and then removed remotely into a secondary
shielded gastight volume.

The procedure for removing a major component
such as a heat exchanger would be as follows:
The fuel would be transferred to the drain vessels
and the drain line isolated and gas-buffered with
the mechanical valves. The coolant salt would
also be drained and the reactor cooled down with
the circulating gas system. Personnel would then
enter the area above the reactor, uncouple the
electrical and instrument connections from the
heat exchanger, attach the crane to a heat ex-
changer lifting lug, sever the coolant lines, unbolt
the main hold-down flange, and cut the seal weld.
The personnel would then leave the area, and the
heat exchanger would be lifted out of the reactor
vessel by remote crane manipulation. The ex-
changer would be withdrawn into a plastic bag
or metal coffin to limit the spread of contami-
nation. The unit would then be lowered into a
storage pit interconnected to the hot storage cell.
A new element could then be lifted from a storage
rack and lowered into the reactor and put into
service by reversing the operations. The main-
tenance area above the reactor is equipped with
remote manipulators and special tools to assist
in the operations. Viewing windows are provided
in the walls around the maintenance areq, so that
complete visual coverage is possible.

No equipment was designed for replacement or
repair of the main fuel container, since this item
was assumed to last the design life of the experi-
ment. |f a failure should occur in the reactor
vessel, it would be possible to make weld repairs
from the inside through the heat exchanger
opening. |f complete removal were desired, the
structure could be sectioned remotely from the
inside and removed in manageable sections.

w
 

NUCLEAR PERFORMANCE

The details of the core design (Fig. 1) were
modified slightly for the purpose of evaluating
the nuclear performance. The fuel is considered
to be contained in a quasi-spherical INOR-8
vessel having a diameter of 72 in. and a wall
]/2 in. thick. The care is surrounded by a neutron
and gamma-ray ‘‘thermal’’ shield consisting of
the following annular regions: a gas annulus
3 in. thick through which dry air or nitrogen may
be circulated to preheat the shell during startup
or to cool it during operation; a 6-in. layer of
aluminum silicate insulation (6 Ib/ft3); a 2-in.
layer of ordinary steel; a 4-in. layer of organic
coolant (diphenyl or related compounds); and four
successive 2-in. layers of boron steel (1% B)
separated by four 4-in. layers of organic coolant,
Draining the organic coolant from the first layer
(adjacent to the ordinary steel) will result in a
reactivity decrease (1.89%) expected to be suffi-
cient for emergency shutdown.

The fuel carrier consists of a mixture of LiF
(63 mole %) and BeF, (37 mole %) having a
liquidus temperature of 860°F and an estimated
density of 1.9 ¢/cm3. The critical concentration
of U2%? (as UF,) is estimated to be 3.25 x 107
atoms/cm® (0.1 mole %), giving a critical mass
of 40.6 kg and a critical inventory of 65.8 kg for
a total fuel volume of 171 ft3. For purposes of
the nuclear calculation, the effect of the fuel in
the inlet-outlet duct was ignored; thus the critical
concentration estimated is probably somewhat
high. The nonsphericity of the thermal shield
was also ignored. The Oracle codes Cornpone
and Sorghum were used for criticality and lifetime
calculations, respectively. |

The relative distribution of the fission source
density is shown in Fig. 15. At a power level
of 30 Mw(th), the fission rate adjacent to the
reactor vessel is approximately 7.5 x 1010
fissionsecm~3.sec™!, and the power density is
approximately 2.4 w/cm>. A graph of the spectral
distribution of fission density is presented in
Fig. 16. It may be seen that 55% of the fissions
are epithermal.

Operation at power will require an increase in
the fuel loading to override the effects of the
accumulation of fission products and nonfis-
sionable isotopes of uranium. Inventories and
neutron balances for the initial state and after
operation for two and five years at 30 Mw(th) are
given in Table 4. The sum of the inventory

4/5 T x fissions x 10° per cm® per fission in core

UNCLASSIFIED
ORNL-LR-DWG 43508

 

 

 

.

 

S

T~~~

 

 

 

 

 

 

 

 

 

 

0 1 2 3 4 5 6 7 (x109)
RADIUS CUBED (cm3)

Fig. 15. Relative Distribution in 30-Mw

Experimental Molten-Salt-Fueled Reactor,

Fission

UNCLASSIFIED
ORNL-LR-DWG 43509

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

20
c>5 18
% ' 2
5 " N £
5 o \E
LLSL.L ° y% )
5 N \\w\‘
£ \\\ “\i

) crlniintnmy

4

0 2 4 6 8 10 12 1
LETHARGY

o
®

Fig. 16. Spectral Distribution of Fissions in 30-Mw
Experimental Molten-Salt-Fueled Reactor,

21

 
¢c

Table 4. Nuclear Performance of Spherical Molten-Salt-Fueled Reactor

Core diameter: 6 ft
Power: 30 Mw(th)
Plant factor: 1.0

Fuel volume: 171 §#°
Fuel geometry: Spherical

Fuel processing rate: 0 times per year

A. lInventories and Neutron Balances

 

 

After Cumulative Power
Generation of 60 Mw-years

 

After Cumulative Power
Generation of 150 Mw-years

 

 

 

 

 

Initial State During Two Years During Five Years
Inventory Absorption of Operation of Operation
(kg) Ratio Inventory Absorption Inventory Absorption
(kg) Ratio (kg) Ratio
Fissionable isotopes
u23s 65.8 0.513 107 0.522 226 0.554
Py239 0.50 0.009 1.6 0.011
Fertile isotopes 4.9 0.0 9.0 0.018 19.1 0.029
Fuel carrier
Be’ 912 0.009 912 0.008 912 0.007
Li’ 1200 1200 1200
19 0.099 0.069 0.045
F 7124 7124 7124
Fission Products 22.7 0.035 56.7 0.053
Parasitic isotopes (U236, etc.) 6.5 0.012 18.5 0.024
Miscellaneous: core vessel and leakage 0.368 0.327 0.277
B. Performance Data
After Cumulative Power After Cumulative Power
Generation of 60 Mw-years Generation of 150 Mw-years
Initial State . . .
During Two Years During Five Years
of Operation of Operation
Neutron yield, n 1.95 1.88 1.77
Total fuel inventory, kg 65.8 107 228
Cumulative net burnup, kg 0 28.7 73.3
Net fuel requirement of U235, kg 65.8 136 301

 
U235 amounts to 70 kg

increase and burnup of
during two years of operation.

In the initial state, the neutron leakage from
the spherical thermal shield amounts to about
108 neutronsecm™2.sec™! at a power level of
30 Mw(th). The spectral distribution of these

neutrons is shown graphically in Fig. 17.

UNCLASSIFIED
6 ORNL-LR-DWG 43510

 

24

 

22

 

20

 

 

 

 

 

////V//%Z} Z llrZ4

Ll ALL 0l L Ll d

 

 

 

 

L/
//I/ LA LN )

 

 

77

 

PER CENT OF NEUTRONS PER UNIT LETHARGY
]

 

0000

N &\\\
AT T e

0 2 4 6 8 10 12 14 6
LETHARGY

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

N WLLL LA A A A

@

0

Fig. 17. Spectral Distribution of Neutrons Leaking
from Thermal Shield of 30-Mw Experimental Molten-
Salt-Fueled Reactor,

The gamma-ray heating in the core vessel was
estimated by means of the Oracle code GHIMSR
to be 2.5 w/cm®, and the strength of the gamma-
ray current at the surface of the core vessel was
estimated to be 5 w/cm?. The estimated spectral
distribution is shown graphically in Fig. 18.
The attenuation of this gamma-ray current in the
thermal shield is substantial. It was estimated
that the photon current leaking from the thermal
shield would not exceed 0.001 w/ecm?.  The
estimated spectral distribution is given in Fig.
19.

A breakdown of the gamma-ray escape-current
energy in terms of gamma-ray sources is given in

UNCLASSIFIED
ORNL-LR-DWG 4354t

FRACTION OF GAMMA ENERGY PER Mev

 

O 1 2 3 4 5 6 7 8 9 10 1" {2
GAMMA -RAY ENERGY (Mev)

Fig. 18. Spectral Distribution of Gamma Energy
Leaking Through Core Vessel in 30-Mw Molten-Salt-

Fueled Reactor.

Table 5. |t may be seen that the major source
of gamma rays is inelastic scattering of neutrons
by fluorine. These gamma rays are, however,
relatively low in energy. The fission and fission-
product-decay gamma rays will probably provide
the major biological shielding problem.

REACTOR HAZARDS

Exhaustive studies of the effects of nuclear
transients induced by off-design operation or
malfunctioning and breakdown of equipment of
molten-salt-fueled reactors of this general design
have not yet been made but must, of course, be
completed before a design is approved for con-
struction. Exploratory studies were made of a
generally similar reactor of 600-Mw(th) cc:pc:cit‘y,4
and it was tentatively concluded that the reactor

 

4H. G. MacPherson et al., Molten-Salt Reactor Program
Status Report, ORNL-2634, p 42-59 (Nov. 12, 1958),

23

 
is inherently stable and that a minimum of safety
control equipment would be needed. The failure
which led to the most serious transient was that
resulting from a sudden stoppage of fuel flow,
and therefore such a failure has been investigated
in relation to the design proposed here.

UNCLASSIFIED
ORNL-LR-DWG 43542

FRACTION OF GAMMA ENERGY PER Mev

 

O 4+ 2 3 4 5 6 7 8 9 10 1 12 13 44
GAMMA-RAY ENERGY (Mev)

Fig. 19. Spectral Distribution of Gamma Energy
Leaking from Thermal Shield of 30-Mw Experimental
Molten-Salt-Fueled Reactor.

The flow stoppage incident is defined as the
one in which fuel circulation instantaneously
ceases at a time prior to which the reactor was
operating at design power. Practically, because
of inertial and thermal convection effects, this
precise incident is impossible to achieve, but,
in the limit, it approximates a pump failure. The
peak temperatures computed for such an incident
are upper bounds of those which would occur as
a result of a pump failure.

During power operation of a circulating-fuel
reactor some delayed-neutron precursors decay
from the fuel while it circulates through the
regions external to the core. Thus the fuel
leaving the core is richer in delayed-neutron
precursors than is the entering fuel. The steady-
state precursor concentration in the core of the
reactor when the fuel is circulating is accordingly
less than when the fuel is not circulating. In
order to compensate for the partial loss of delayed
neutrons arising from the precursor deficiency,
the reactor is just critical when k& exceeds 1 by
the appropriate amount.

When circulation is suddenly stopped, the
precursor concentration gradually rises to the
steady-state value appropriate to a stationary-fuel
reactor (i.e., # = 1), and the corresponding in-
crease in delayed-neutron production tends to
make the reactor supercritical. Simultaneously,
cessation of heat removal by circulation of the
fuel tends to cause the core temperature to rise,
and the reactor, because of its negative temper-
ature coefficient, tends toward subcritical. The
result is a complex nuclear power transient and
a corresponding temperature transient which, in
the present reactor, after a moderate overshoot

Table 5. Components of the Gamma-Ray Current from the Core Vessel of a 30-Mw

Experimental Molten-Salt-Fueled Reactor

 

Fraction of Energy

Mean Energy of Photons

 

Gamma-Ray Source of Gamma-Ray Current (Mev)
Fission 0.245 1.1
Fission product decay 0.191 0.9
Inelastic scattering of neutrons 0.487 0.2

in fluorine
Captures in beryllium 0.007 5.6
Captures in fluorine 0.031 5.8
Captures in uranium 0.037 0.3

 

24

>
finally settles down to a higher critical temper-
ature.

The above-described transient was studied for
the present reactor with the aid of the ORNL
Analog Computer Facility, using methods de-
scribed by C. S. Walker.® The pertinent pa-
rameters that were used to describe the reactor
and the computed effect on the critical temperature
are given below:

Fuel u233
Spherical core volume 113.2 13
External fuel volume 57.5 £+
Fuel circulation rate 3.18 cfs

Volumetric heat capacity of fuel 64 Btu'f'#—a-(oF)*]

Design power (initial condition) 30 Mw
Nuclear temperature coefficient ~6.75 X 1073 per °F
of reactivity
Prompt-neutron lifetime 8.8 x 107 sec
Rise in critical temperature 19.5°F
(computed)

As expected, for the case in which no heat
escapes from the core, the temperature rises
asymptotically to a maximum, where it remains
indefinitely, maintaining the reactor in a sub-
critical condition. The temperature rises about
35°F in the first 10 sec, and the total rise in
temperature is less than 110°F; 90% of the
maximum is reached 90 sec after the start of the

transient. Nuclear afterheat was not taken into
account in the above calculations, and, of course,
in any real reactor the temperature would not be
asymptotic but would continue to rise because
of afterheat.

To limit the temperature during a prolonged
stoppage of fuel circulation, afterheat must be
removed by some alternative means. In general,
if the removal rate is a fraction of the initial
reactor power and exceeds the afterheat production
rate, instantaneous stoppage of circulation will
result in an initial temperature rise followed by
a fall. After perhaps oscillating a few times,
the temperature will settle on the new critical
value appropriate to the noncirculating core. This

 

3. s. Walker, Simulation of the ORSORT Buttermilk
Reactor,) Loss of Fuel Flow, ORNL CF-58-7-64 (July

31, 1958

stagnant-mean core-critical temperature is 19.8°F
above the circulating-mean core-critical temper-
ature.

To gain insight into the maximum temperatures
which might be attained and the times to reach
them, the behavior of the reactor was investigated
under a condition of auxiliary cooling wherein the
cooling rate less the afterheat rate was held
constant and the fuel circulation was instan-
taneously stopped. The peak temperatures and
times to reach them as a function of the assumed
power-removal rate (in excess of that required for
afterheat), which is constant throughout the
transient, are indicated in Fig. 20. As expected,
the peak temperatures and times to reach them
decrease with increasing power-removal rates.

UNCLASSIFIED
ORNL—-LR—DWG 43513

120 \ ‘ l 120

® TIME TO REACH
\ \,/PEAK TEMPERATURE (sec)
100 ‘ 100

 

 

 

 

 

 

 

3 |

Q

= \ PEAK MEAN CORE _

@ ° - A °F &

5 50 3 TEMPERATURE RISECF) | g ¢

< n

& \ x
x

= 60 N, 60 2
<

= a

> '\ =

S 40 a0 &

O

F_.

LJ

=

F_.

 
     

 

 

 

O 10 15
CONSTANT POWER REMOVAL (Mw)
(IN EXCESS OF THAT REQUIRED FOR REMOVAL OF AFTERHEAT)

Fig. 20 Effect of Instantaneous Reduction of Heat

Flow from Reactor on Temperatures.

The initial rate of rise, which is important for
thermal shock considerations, is very modest,
as shown by the curve giving the rise in the first
10-sec period. Oscillations about the new critical
temperature are well damped; the damping in-
creases and the times between peaks decrease
with increasing power-removal rates. Typically,
for 3 Mw heat removal, the first maximum occurs
in 82 sec following initiation of the transient,
with a 55°F peak (above the new critical temper-
ature); the second maximum occurs 9 min after

25

 
initiation of the transient, with a 5°F peak. It
is clear that, if need be, ample time is available
to adjust heat-removal rates to near the afterheat
production in order to maintain the reactor above
the critical temperature and, hence, subcritical
without oscillations, so long as afterheat is
sufficient to maintain control.

COSTS

An estimate of construction costs has been
prepared and is presented in Table 6. It was
concluded that the plant could be built for about
$16,000,000. Primary emphasis was placed on
obtaining estimates of the cost of the INOR-8

Table 6. Capital Cost Summary

FPC Account No.
310 Land and land rights

311 Structures and improvements
Site improvement
Site facilities
Station buildings
Reactor building

Turbine-generator building

Total structures and improvements

No cost

$ 10,000
160,000

600,000
800,000

$1,570,000

312 Reactor and steam-generating equipment

Primary system
Fuel container and gas shroud
Pumps and pump drive

Primary heat exchanger

Subtotal

Primary system auxiliaries
Fuel drain and storage
Enriching and sampling system
Purge system
Off-gas and effluent system
Inert-gas system

Other auxiliary systems

Subtotal

Intermediate system
Pump and pump drive
Superheater
Drain system
Piping and valves

Emergency dump system

Other intermediate system equipment

Subtotal

Reactor cell: shielding and containment

275,000
250,000
150,000

$ 675,000

218,000
100,000
50,000
150,000
50,000
50,000

$ 618,000

150,000
194,000
50,000
100,000
75,000
50,000

$ 619,000

1,150,000

Heating, cooliing, and ventilating systems

Reactor primary heating and cooling system

Intermediate heating system

Cell cooling and ventilating system

Stack

Subtotal

26

50,000
50,000
100,000
15,000

$ 215,000
312

314

315
316

Table 6 (continued)

Reactor system instrumentation and controls

Steam generator and feed-water system
Loeffler evaporators
Loeffler steam pump
Feed-water heaters
Boiler feed-water pumps and piping

Other equipment

Subtotal

Total reactor and steam-generating equipment

Turbine-generator equipment
Turbine-generator and accessories
Condenser and circulating-water system
River intake, weirs, and shore-line structures

Other equipment
Total turbine-generator equipment
Accessory electrical equipment

Miscellaneous power plant equipment
Reactor maintenance equipment
Cranes
Manipulators
Viewing equipment
Cutting and welding

Miscellaneous tools
Subtotal
Spare parts
Pumps (molten salt)
Heat exchangers

Steam pump

Miscellaneous parts

Subtotal

Original inventory of molten salt
Total miscellaneous power plant equipment
Transmission line
Total direct construction
Contingency
General expense
Engineering and design

Interest during construction

Total cost

750,000

200,000
50,000
50,000
50,000
50,000

400,000

800,000
175,000
200,000

125,000

75,000
150,000
50,000
75,000
100,000

450,000

300,000
150,000
20,000

100,000

570,000

825,000

4,427,000

1,300,000
200,000

1,845,000

50,000

3,000,000
1,400,000
1,500,000

900,000

$ 9,392,000

6,800,000
$16,192,000

 
components of the molten-salt systems. These
items made up the more unique portions of the
plant and, in an over-all analysis, were subject
to the greatest estimating errors.

Most of the design time was concentrated on
the major components of the salt systems to
facilitate cost estimating. Engineering layouts
were prepared of the fuel container, heat ex-
changer, and superheaters. These drawings were
reviewed by the Y-12 shop personnel for fab-
ricability and first estimates of manufacturing
time. This group has had considerable experience
in nickel-base alloy fabrication, and they have
built reactor-grade components. In addition,
several outside fabricators were asked to review
the drawings from the standpoint of design and
fabricability.

Many constructive comments were obtained from
these reviews, but no serious objection with
regard to concept or design was raised which
would invalidate the system or seriously change
the manufacturing estimates. Some development
work would be required in the tube fabrication of
the fuel-to-coolant-salt bayonet heat exchanger,
and test work would be required to check out some
of the welding and brazing on the major INOR-8
components.

The finished INOR-8 components in the plant
were estimated to weigh approximately 50,000 Ib.
Most of the material required to fabricate these
components would be in the form of plate and
tubing.  Facilities for supplying INOR-8 stock
are available, and a sizable inventory is currently
on hand. The raw-material prices used were
obtained from vendors’ estimates on sizes and
quantities needed for the experimental reactor
and from actual costs of material received for
the molten-salt-fueled reactor development pro-
gram.

The reactor buildings and site were not suffi-
ciently specified to get complete cost information.
A site was chosen near an area that had been
studied for a similar reactor installation, and
estimates for site improvement and facilities were
available. No cost was applied to land acqui-
sition, since the reactor was assumed to be built
in the Oak Ridge area.

The gross volume of the reactor building,
offices, hot cells, laboratories, and control room
was estimated at 200,000 3. The turbine-
generator building was calculated to have a

volume of 400,000 f+3.

28

The cost of most of the auxiliary systems was
estimated without detail. These systems, or
subsystems, were estimated as gross packages
on the basis of general experience. One exception
was the reactor heater-cooler unit, which was
developed to the point of engineering layout and
specification of the major components.

‘The more conventional portions of the steam
generator plant were determined from manufac-
turers’ data.

Included in the first plant cost was the molten-
salt inventory. This quantity included a spare
fuel volume and 50% overage for the coolant-salt
volume, resulting in a total of 550 ft3. This
entire quantity was assumed to contain Li’ at
the 99.99% assay level, which contributes one-
third of the total cost of $1500 per cubic foot.

An over-all contingency factor of approximately
30% of the first cost was used, and the other
indirect costs were set at values considered
applicable to this type of construction.

The operating costs for the system were studied.
After the completion of shakedown and planned
experiments, it was concluded that the plant
could be operated for $635,000 a year. The

breakdown of this estimate is as follows:

Wages (including supervision) $250,000
Supplies 10,000
Maintenance 75,000
Fuel burnup and inventory charge 270,000
Fuel preparation 30,000

$635,000

ALTERNATE DESIGNS

A suggested modification of the fuel container
assembly and reactor is shown in Fig. 21. The
fuel pump is located on the vertical axis of the
core and is a removable part of the heat exchanger
assembly.  The maintenance concept for this
system is the same as that previously described,
with the exception that the fuel pump must be
removed when major heat exchanger maintenance
is required. The pump can be removed or replaced
independently.

The sump-type centrifugal pump has an annular
diffuser. The fluid fuel is directed out of the
diffuser section into the tubes of the heat ex-
changer.  After passing through the heat ex-
changer, the flow is directed into the annular
COOLANT OUT

 

l COOLANT IN

UNCLASSIFIED
ORNL—LR—DWG 43514
i

 

PUMP MOTOR

 

 

 

OFF-GAS
LIN

 

|
FILL AND
DRAIN LINE

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

M _GAS-DUCT OUTLET
X XSO KNS
'y =
R/ EXPANSION
/l SPACE
d
PUMP ——
M EXCHANGER
1
I REACTOR
/

 

 

 

 

 

 

 

 

 

 

 

 

 

 

W)&X XXX XAy

GAS-DUCT INLET

 

 

Fig. 21. ConcentricsPump Molten-Salt-Fueled Reactor Layout.

29

 
volume surrounding the reactor and then dis-
charged up into the core at the bottom opening.
The fuel leaves the core at the top and goes
vertically up into the pump section. The fuel
expansion tank surrounds the pump barrel and is
cooled by the intermediate salt circuit.

The coolant salt enters the heat exchanger shell
at the bottom and flows upward around the tubes.
After it leaves the heat exchanger it passes
around the pump region and out of the assembly.
This arrangement has several important advan-
tages. The primary fuel container geometry is
simplified, and the support problem is more
straightforward. The fuel inventory is reduced,
and the expansion tank region is completely
surrounded by coolant to ensure more positive
temperature control of this region.

The major disadvantage of the concept is the
more complicated heat exchanger upper structure.
This region, as presently conceived, includes
the stationary parts of the pump assembly, which
would be discarded in the event of a heat ex-
changer failure. Even so, this concept does
render more of the fuel system components readily
replaceable and would simplify the operations
required to remove the entire fuel container from
the reactor cell. This fuel system is sufficiently
attractive to warrant further design study.

An alternate steam-generating concept has also
been considered for molten-salt-fueled reactor
systems., There appears to be little question as
to the operability of the Loeffler cycle, but the
system does result in a cost penalty because of
the massive evaporation drums required in larger
power plants. The proposed steam generator,
shown in Fig. 22, would be used in place of the
superheater, steam pump, and evaporator drums.

The molten-salt coolant is circulated on the
shell side of the bayonet tubes. Feed water is
recirculated through the bayonet-tube assemblies
and superheated and/or saturated steam is with-
drawn from the steam generator.

The feed water is pumped up into the tubes
through the inner annulus and spills back down
through the central tube to the sump. The water
boils as it is pumped up this annulus by virtue
of the heat transferred from the molten salt through
the steam as it flows down the outer annulus.
The system is regenerative in that the steam
formed is used as a heat transfer fluid coupling
the molten salt and the boiling water. This

30

system, therefore, allows low-temperature water,
600°F or less, to be put into the same vessel
with the salt having a liquidus temperature of
865°F.

The steam output is regulated by reducing the
flow through the steam annulus so that less heat
is transferred to the boiling water and the load
is reduced. Conversely, when more flow is
channeled through the steam annulus, the load
is increased. A heat baffle 51 in. long was added
at the outlet end of the steam annulus to insulate
the steam from the water so that 1000°F superheat
could be obtained.

Saturated steam would be removed from the
region above the sump, and this flow would be
blended with the superheated steam to regulate
the steam temperature. At lower flows, the outlet
steam is at a higher temperature and a larger
fraction of saturated steam is used for tempering.

The design data for a 30-Mw steam generator
are presented in Table 7. This unit is considered
to be more easily fabricated than the U-tube,
U-shell steam superheater, and experiments are
planned to study the control and heat transfer
performance of this geometry.

Table 7. Design Data for Once-Through

Steam Generator

Coolant salt

In 65 psia, 1160°F
Out 37 psia, 1000°F
Flow 2.17 cfs
Feed-water inlet conditions 96,400 Ib/hr at 475°F,
1600 psia

96,400 Ib/hr at 592°F
192,800 Ib/hr at 535.5°F
96,400 Ib/hr at 1000°F

Recirculated feed water
Total feed water
Steam outlet conditions

Configuration Concentric bayonet tubes

in cylindrical shell

Flow Steam in tubes
Tubes
Size 1.25-in. OD, 0.109-in.-
wall
Number 91
Pitch 19/]6 in., triangular
Length 27 £+ 9 in.

)
UNCLASSIFIED

ORNL~-LR—~DWG 39168A

 

   

N

THERMAL
BAFFLE

 

 

 

 

 

 

 

 

 

 

 

 

  
  

 
 
      

 

 

\

\\

§ WATER

\

§ = SUPERHEATED

N\ | STEAM D
\
N\
\ L WATER AND STEAM

SECTION THROUGH TUBE ASSEMBLY

 

 

 

 

 

 

SUPERHEATED STEAM OUT —=—o

 

 

 

 

 

 

 

 

 

 

 

 

 

 

=

il 2e
il

 

¢ SATURATED WATER OUT

Fige 22. Once-Through Steam Generator,

 

31

 
It was previously mentioned that unclad graphite
was being investigated for use in molten-salt
breeder reactors. Graphite test specimens could
be inserted in the core of this reactor without
undue complication, or the entire core could be

modified and a graphite moderator incorporated.
Neither of these approaches would alter the basic
fuel container concept, and the latter would result
in a system with a much smaller liquid inventory
and probably a smaller core vessel.

LIST OF DRAWINGS

A list of ORNL drawings that were prepared for this design study is presented below:

Title

Plant Section Looking West
Plant Section Looking North
Plan - Elevation 838

Plan - Elevation 803
Reactor and Cell

Fuel Container Weldment

Fuel-to-Coolant-Salt Heat Exchanger

Steam Superheater

Bayonet Tube Boiler-Superheater
Reactor Heater-Cooler Assembly
Concentric Reactor Assembly
Enricher Sampler Assembly
Enricher Sampler Assembly
Preliminary Site

Perspective, S. W, Corner
Perspective, S. E. Corner

Perspective, N. E. Corner

32

ORNL Drawing No.

F-2-02-054-7942
F-2-02-054-7943
F-2-02-054-7938
F-2-02-054-7941
F-2-02-054-7939
F-2-02-054-7940
F-2-02-054-7972
F-2-02-054-7897
F-2-02-054-7928
F-2-02-054-9057
F-2-02-054-9028
F-2-02-054-7657
F-2-02-054-7658
F-2-02-054-7944
D-2-02-054-7945
D-2-02-054-7946
D-2-02-054-7947

]
is

PPENO N AN~

. Affel

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. Bettis

. Billington

. Blankenship
. Blizard

. Boch

. Borkowski

. Boudreau

. Boyd

. Bredig
Breeding

. Briggs

. Browning

. Campbell

. Carr

. |. Cathers

. E. Center (K-25)
. Charpie

. Coobs

. Culler

. DeVan

. Douglas

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. Ergen

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. Ferguson

. Fraas

. Franco-Ferreira
. Frye, Jr,

. Gall

. Gresky
Gregg

. Grimes

. Guth

. S. Harrill

. R. Hill

. W. Hoffman

. Hollaender

. S. Householder
. H. Jordan

. W. Keilholtz

. P. Keim

. 1. Kelley

. Kertesz

TOMWP>PMTACT TV OO

MEOQEPPIZTOME-PE-MPUSEMD-T-F00EDEAIMZOEOPMIMOME™D
AT 4TI >omMmXXxXo>ITrI>»

. B. W. Kinyon

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@EFPOTMOAMPE-OMP>IMIE->IZ-V0OPVEOZF-AAMAL-IETrMERLIAN-F

ORNL-2796
UC-81 Reactors -~ Power
TID-4500 (15th ed.)

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oOOmzT=TNwmI

33

 
34

95. G. C. Williams 102-121. Laboratory Records Department

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97. J. Zasler 123-124. Central Research Library
98-101. ORNL — Y-12 Technical Library,

Document Reference Section

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