ORNL-CF-59-1-26.txt

X822

 

OAK RIDGE NATIONAL LABORATORY
Operated by

UNION CARBIDE NUCL EAR COMPANY

Division of Union Carbide Corporation

uee

Post Office Box X
QOak Ridge, Tennessee

DATE: January 13, 1959

SUBJECT: A Preliminary Study of a Graphite Moderated.
Molten Salt Power Reactor

TO: Liszted Distribution

FROM: H.
L.
D.
E.
M.
Lo
Jo
G.
Jo

Ga
G.

B“

w‘
E.
A,
W.
D,

MacPherson
Alexander
Grimes
Kinyon
Lackey
Mann
Miller
Whitman

Zasler

Abstract

g ¥

»

:
f ;i

~&

. . ) 4 .
STER COPY

MAL

-

External
Distribution Limited

 

 

ORNL

CENTRAL FILES NUMBER
59-|-=2¢

 

COPY NO. -{é

A preliminary desjgn and cost study has been made on a one region

unclad graphite moderated molten salt power reactor.

Included are conceptual

plant laysuts, basic information on the major fuel circuit components; and a
discussion of the nuclear characteristics of the core.

For a plant electricsl output of 315,000 kw and & plant factor of
80 percent, the energy cost was approximately T.k mills/kwh.

NOTICE

This document contains information of a preliminary nature
and was prepared primarily for internal use at the Oak Ridge
National Laboratory. It is subject to revision or correction
and therefore does not represent a final report. The information
is not to be abstracted, reprinted or otherwise given public
dissemination without the approval of the ORNL patent branch,
l.egal and Information Control Department.

 
1. General Features of the Reactor

A power reactor of the molten selt type using a graphite moderator achieves
a high breeding ratio with a low fuel reprocessing rate. The graphite can be in
contact with the salt without causing embrittlement of the nickel alloy container.

The salt selected consists of a mixture of LiF, BeFE, and UFh (70, 1o,
20 mol %), melting at 9%32°F. The uranium is 1.30% enriched. The core is 12.25
feet diameter by 12.25 high, with 3.6" dismeter holes on 8" centers. 16% of

the core volume is fuel.

The choice of the power level of this design study was arbitrary, as the
core is cagpable of operation at 1500 Mw(t) without exceeding safe power densi-
ties. An electrical generator of 333 Mw(e) was chosen, with 315 Mw(e) as the
station output, which requires 760 Mw(t).

The heat transfer system includes a fluoride salt to transfer heat from
the fuel to either primary or reheat steam. The salt selected has 65 mol %
LiF and 35% BeF,, which is completely compatible with the fuel. The Loeffler
steam system, at 2000 PSI, 1000°F, with 1000°F reheat avoids the problems
associated with a high température fluid supplying heat to boiling water.

The fuel flow from the core is divided among four circuits, so that
there are four primary heat exchangers to teske care of the core heat genera-
tion. Two superheaters, one reheater, three steam generators are required
for each circuit. This arrangement is based on the practical or economic

size of the respective components.

While it would have been possible to design this graphite moderated
molten salt reactor plant identical to the homogeneous plant described in
ORNL 2634, Molten Salt Reactor Program Status Report, an effort was made
to include new designs evolved since then for a number of festures and
components. These include the meintenance concept; heat exchanger design,
fuel transfer and drain tank system, gas preheating, barren salt inter-
mediate coolant and the Loeffler steam system.

In most of these, the actual design chosen for a plant will not greatly
affect the overall economy and operation. It is highly probable that the
-3-
Table 1

REACTOR PLANT CHARACTERISTICS

Fuel 1.30% U°>°F, (initially)
Fuel carrier 70 mole % LiF, 10 mole % BeF,,
20 mole % UF),
Neutron energy near thermal
Moderator carbon
Reflector iron
Primary coolant fuel solution circulating at 35,470 gpm
Pocwer
Electric (net) 315 Mw
Heat 760 Mw
Regeneration ratio
Clean (initial) 0.79
Estimated costs
Total 279,250,000
Capital g252/kw
Electric 7.4 mills/kwhr
Refueling cycle at full power semicontinuous
Shielding concrete room wall, 9 't thick
Control temperature and fuel concentration
Plant efficiency 41.5%
Fuel conditions, pump discharge 12250F at ~105 psia
Steam o o
Temperature 1000°F with 1000 F reheat
Pressure 2000 psia
Second loop fluid 65 mole % IiF, 35 mole % BeF,
Structursl materials
Fuel cirecuit INOR-8
Secondary loop INOR-8
Steam generator 2.5% Cr, 1% Mo steel
Steam superheater and reheater INOR-8B
Active-core dimensions
Fuel equivalent dia 14 ¢
Reflector and thermal shield 12-in. iron
Temperature coefficient (&k/k)/oF negative
Specific power 1770 kw/kg
Power density 117 kw/liter
Fuel inventory 35
Initial (clean) 700 kg of U2
Critical mass clean 178 kg of 0255

Burnup unlimited
el

features of the actual plant built would consist of a mixture of those de-
scribed in this report; in the previdfia'reporta and evolved as _a result of

future design and development work.

A plan view of the reactor plant layout is presented in Figure 1, and
an elevation view is shown ingFigure éo The reg&to:@and the primary‘heat
exchangers are contained in a large rectangular reactor cell, which .is
sealed to provide double containment- for any ieakage of fission gases.

The rectangular configuration of the plant permits the grouping of similar
equipment with a minimum of floor space and piping. The superheaters and
reheaters are thus located in one bay, under a crane. Adjacent to it are
the turbogeneratcr, steam pumps, and feed water;heaterpand pumps. The
plant includes; in addition to the rgactor and heat exchanger systems and
electrizal generation systems, the control room and fill-and-drain tanks

for the liquid systems.

2o Fuel Circuits

" The primary reactor cell which encloses the fuel circuit is a .con-
crete structure 22 ft wide, 22 ft long, and 32 ft high. The walls are
made of 9 ft thick concrete to provide the biological shield. Double
steel line%sflform 8 buffer zone to ensfire that no fission gas that may
leak into the cell can escape to the atmosphere and that no air can enter
‘the cell. An ifieft atmosphere is maintained in the cell at all ‘times.

The fiumps, heat exchangers, and instrumentstion are so arranged that
the equipment may be removed through plugs at the top of the cell leaving

the fuel-containment shell behind in the reactor cell.

In the reactor cell are located the reactor, four fuel pumps,; and
four heat exchangers. _The fuel system, gas heating, and cell cooling
equipment as well as the fission gas hold-up tanks are in connected side

passages.

The reactor core consists of a graphite moderator, 12.25 ft in dia—‘
meter and 12.25 £t high. Vertical holes 3.6 inches in diasmeter on an
UNCLASSIFIED
ORNL-LR-DWG, 35086
COOQLANT SALT rRUMR

COOLANT SALT CELL
SUPERMEATERS

EQUIP. PEMOVAL HATCH PREHEATER

EVAPORLATORS

LOOFR MHEAT

 

 

 

 

- ————————— - —_— —

   
 
 
  
 

 

 

< ;;?’%EA p EPENEAT LINE
NEICHE
CELL —— = TURBOGENELATOR
CooL/ING
ACCESS
LOC K

& PLUM Pz HGH PRESSURE LINE
VEL

SALT QOOLING LINES

O O

 

AYEL T
LXQMANGER
(4)

 

FUEL DRAIN TRMNKS

o O |

 

FI@GURE | - PLAN VIEW - 760 MW (t) GRAPHITE MODECRATED
MOLTEN SALT POWERL REARCTOR PLANT

tmxgm
et 00 v

 

 
 

 

UNCLASSIFIED
ORNL-LR-DWG. 35087

   
   
    
 
   
    
    
   

 

 

COOLANT PUMF

STERM PUMPR

— B — — mr————— i e mmmnm—

 

 

 
  
  

 

 

- — - e e — e — - - | S——— —

-
\

N rbmm— - r——

  

SUPERMHERTER

-

 

 

FOEL PUOMPR —

HERT EXCHANGCER

REACTOR

EIGURE 2 - ELEVATION VIEW - 760 MW (t) CRAPHITE MODERATELD
MOLTEN SRLT PLOWER BEMRCTORL PLONT™ |

be—r20 "
eight inch square pitch form the fuel passages. The core is mounted in a 1-1/2"
thick INOR-8 cantainer. Fuszl azrters at the bottom afid passes through the fuel
passages and a two inch annulus between the core and shell which coonls the shell.
At the top of the reactor is the fission gas holdup dome described elsewhere.

This is shown in Figures 3 and L.

From the reactor; the fuel gees to centrifugel pumps of which there are
four in parailel. The lower bearing is sali-lubricated, submerged in the fuel
above the impeller. Above the fuel surface is a shielding section, to protect
the upper bearing lubricant and motor. The beering includes a radial bearing,
a thrust bearing, and a face sesl. The motor rotor is on the shaft above the
bearing. The foter is canned, so the field windings may be replaced without
breaking the reactor seal. KCmoling is profiided for the shielding section and

the shaft. The entire pump may be removed and replaced as a unit.

The coclant salt pump is of a similar design, with modifications permitted
by the lower radiaticn level.

The primary exchangers are of the bayonet bundle type, to permit semi-
direct replacement. The incoming coolant passes through the center of the
exchanger to bottom, then upward on the shell side through the exchanger,
leaving inlan annulus surrounding the incoming coclant. Helical tubes are

between flat tube sheets.

3. Off-Gas System

‘ An efficient process for the zontinuous removél of fission-product gases
is provided that serves several purposes. The safety in the event of a fuel
- 8pill is considerably enhanced if the radiocactive gas concentration in the
fuel is reduced by stripping the gas as it is formed. Further, the nuclear
stability of the reactor igger changes of power level is improved by keeping

the high cross secticn Xe continuously at a low level. Finally, many of
the fission-product poisons are, in their decay chains, either noble gases
for a period cf time or end their decay chains as stable noble gases, and

therefore the buildup of poisons is considerably‘reduced by gas removal.
 

 

 

UNCLASSIFIED
ORNL-LR-DWG. 35088

 

 

 

 

 

 

 

 

 

   
   

 

 

 

 

 

 

 

 

 

 

 

 

C;/?,Apd TE
C oRE

 

 

 

 

 

 

 

 

G 3
REQCTOR & PUMPS ~ FLEVEBT IOV

 

 

 

 

 

 

 

 

 

 

 

 

 

£AL WELD
) Db b ‘E'U’
i
5Hin.p
i | —~O-RING
. | 1
u?‘ . ?
LR " ’4‘ ' AA
i
Exvansidry  Tank
e SO N\
i
1 T~
b ’

 

Cc)er&;y:gQ

/SALT

 

 

 

/’Gfi ¥ Hffir/N[, S‘H:&&'
——f UL FhssAacE
C7pr)

 

 

{} ] Foee in
UNCL.ASSIFIED
ORNL-LR-DWG. 350839

 

 

 

F/6G 4
REACTOR & FUNTEPS ~ FLAN
The solubilities of noble gases in some molten ssiits are given in Table 2,
and it is indicated that sclupliiities of similar corders of magnitude are likely
to be found in the LiFmBng

cbeys Henry's law, so that the equilibrium soiubility is proporticnal to the par-

gsalt of this study. It was found that the solubility

tial pressure of the gas in contact with the salt. In principle; the method of
fission=gas removal consists of providing a quiet free-surface from which the

gases can be libersated.

In the system chosen; approximately 50% of the fuel flow is allowed to
flow intc the reactor expansion tank. The tank provides a large fuel-tc-gas
interface; whicfi promotes the establishment of equilibrium fission gas concen-
trations in the fuel. The expansion tank provides a iliquid surface area of
epproximately 52 ft2 for removal of the entrained fission gases. The gas re-
moval is effected by the balance between the difference in the density of the
fuel and the gases and the drag of the oprosing fuel velozity. The surface
velocity downward 1n the expansion tank is epproximately C.75 ft/sec, which
should screen cut all bubbles larger than 0.06 in. in radius. The probability
that bubbles ¢f this size will enter the remctor is reduced bty the depth «f the
expansion tank being sufficient to aliow time for smail hubb1$s toy conlesce

and be removed.

Table 2

SOLUBILITIES AT 600°C AND HEATS OF SOLUTION
FOR NOBLE GASES IN MOLTEN FLUORIDE MIXTURES

 

 

In NaeF-ZrF In LiF-NaF-KF In LiF-BeFpn

(53-47 mole %) (L6.5-11.5-42 mole %) (6Ua’6 mole %)
Gas k* k¥ k#*

x 1070 x 107° % 167°
Helium 21.6 + 1 11.3 * 0.7 11.55 + 0.07
Neon 11.3 + 0.3 holb + 0.2 L7 +0.02
Argon 5.1 + 0.15 0.90 + 0.05 0.98 *+ 0.02
Xenon 1.94 + 0.2 - .28 festimated}

 

Henry's law constant in moles of gas per cubic centimeter of sclvent per
atmosphere.
-} -

>

The liquid volume of the fuel expansion tank is approximately 50 ft” and

the gas volume is appraximaéély 2k0 ftfi. With none of the fission gases purged,
approximstely 3300 kw of beta heating from thekdeéay of the fission gases and
their daughters is deposited in the fuelfand on metal surfaces of the fuel ex-
pansion tank. This 3300 kw ¢f heat is partly removed by the byfiass fuel circuit
‘and. the valiancs is traneferred thréugh the expansion tank walls to the secondary

| loop coclant.

The fission product gases will cause the gas pressure in the reactor to
rise approximately 5 psi per m@fith. Tfiifl pressure is relieved by bleeding the
tank cnce a month at a controlled rate of approximately 5 psi per day to a
hold tank. ({See Figure 5.) The gases in the hold tank are held until they
have decayed sufficiently tc be disposed of either through a steck or a noble

- gas recovery system.

A small emcunt of fission gases will collect -above the free surface of
each pump. These gases are continususly pfifged”with’helium. The purge gases
from the pfimps are delayed in a hold voliume for approximately 5 hours to allow
e large ffaétiafi of the shorter lived fission products tc decay before enter-
ing the cooled carbon beds. The carbon beds provide a holdup time of approxi-
mately 6 days for krypton and mnch‘langer for the xenon. The purge gasss from
the carbon beds, essentially freeAfrom activity, are compressed and returned

to the reactor o repeat the cycle.

L, Molten Salt Transfer Equipment

The fuel transfer systems are shown schematically in Figure 6. Fluid is
firansferred between the reacter.and drain tanks thrcugh a presgureasiphbn Bys-
- tem. Two, mechanical valvea; in series, are placed in this line with a siphon-
breaking connection bestween them. ¥Fluid is transferred.frem one system to a-
nocther by isolating the siphon«bresker and applying differential gas pressure
to establish flow‘and finally complete the %transfer. When the gas equalizing
valves in the siphon-~bresker circuit are opened the fluid will drain cut of

the transfer line and the valves are then closed.
UNCLASSIFIED
ORNL-LR-DWG. 35090

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

MOTOR
— MM e ; - ]
FueEL DEGASSER  puel i l _ !
4 EXPANSION TAIK PUNPS vew | |
[ . ~ | BY PAss '
IC%F};’ .4 = i
17 :
| 39,5¢F5 ) 1225°F ]0.553 SCFM |
I YR \L r |3'35 FT? l‘l £i~ jo FT3 ‘w-l A__»_N__G—N_“l/ *
' ) ‘ \ ! !
* L‘g' %m .X____w_u_‘ GAS PUMPS L-*-—‘J
CORE k * * , ,
1 l ;
o 3T Laegre :
: PUEL BUED °F o= |
"‘?‘? .4%006..AUT g
; e —{BS‘QFT.’ |— — FUEL LINE
éX * —-—— GAS LINE ALUE
—p4—CLOSED U
] L. o To STACK OR —p<— OPE N UALVE
INTERMITTENT FLOwW % X NOBWE GAS
O.0LTBRLFM @ 1250°F izg{) ;:'1-,3’...1 RECOUERY

FIG.S - SCHEMATIC FLOW PIAGRAM FoR REMIVAL OF FISSION PRODUCT GASES,
UNCLASSIFIED
ORNL-LR-DWG. 35091

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

INERT GAS INERT (3AS SUPPLY
-t PP T - -—-—--—“l—-;M”T
- VENT i | . q UE
SUPPLY % SIPHON | BREAK X X
PoINT | | |
SAMP LING - | I ,J . '
A ENRICH\WG ¥ l —(—)——M—T_N c C ( <
POINT | ISOLA TION | | | SALT
- VALLOES | | l CHARA ING
I—L ! . i
[ | ' | I
| ; | .
| I l | |
HEAT REACTOR : : : :
EXCHANGER | ' | | | t
(| oF W) | 0 | | | '
— DRAIN  TANKS
iy
LEQ GE LD
FUELL LIME
—-— GAS LINE

—b<t— REMOTE L
OPER ATE D VALVE

Fle. 6 SCHEMATIC DIAGRAM OF MAIN FUEL SALT
TRANSFER SYSTEM,

-E'[..
=1h-

Normally the isolation valves are dry and when the gas equalizing systems
at the siphon-bresker point are open fluld cannot rise to the isolation valves

even though they are open or leasking.

The volume between the two isolation valves is designed to be an inert
gas buffered region to effectively isolate the molten fuel in the drain vessels

when the reactor system must be opened for maintenance operations.

The fuel added to the reactor will have a high concentration of U255Fh,
with respect to the process fuel, so that additions to overcome burnup and

fission product buildup will require only small volume transfers.

Solid U235Fh or highly concentrated U235Fu in an alkeli metal fluoride
mixture in the solid or liquid state could be added to the fuel system. Solid
additions would be added through a system of "air locks' over a free surface
of fuel, while the liquid asdditions would be mede from a heated vessel fron
which fluid is displaced by a piston to meter the quantity transferred.

Samples of the fuel would be withdrswn from the fuel system by the "thief
sampler’ principle which in essence is the reverse of a solid fuel addition

system.

5. Heating Equipment

 

The melting points of the process salts are well above room temperature.
It is therefore necessary to provide a means of heating all pipes and equip-

ment containing these fluids.

For the mssf part high temperature gas circulastion will be ermployed to
heat the major fuel components and conventional electric heater-insulation
instellations will be used to heat the remsinder of the systems. Gas blower
and heater packages are installed so that they may be removed and replaced
as a unit thus easing the meintenance problem as compared to direct electric

heater instellations.

6. Auxiliary Cooling

Cooling is provided in the reactor cell to remove the heat lost through
=15-

the pipe insulation‘and the heat generated in the structural steel pipe and
equipment supports by gamma-ray absgorption. The heat is removed by‘means of
forced gas circulation through radistor-type space coolers. A cooling medium,
such as Dowtherm, in a closed lcop removes heat from the space coolers and
dumps it_ta a water heat exchanger. Gas blower and cooler packages are in-

stalled so that they may be removed and replgced as a unit.

T Remote Maintenance

The major requirement for maintenance of the reactor is the ability to
remove and replace the pumps and heat exchangers. These are designed so that
they may be removed through plugs in the top shield by a combination of direct

and remote maintenance methods.

The removable parts of this equipment are suspended from removable plugs
in the top shield, as shown in Figure 3. The primary reactor salt container
seal is made with a buffered metal O ring seal backed up by a seal weld at
the top of the shield. : -

The room at the top of the reactor is sealed and shielded to safely con=-
tain the radiocactive equipment that is removed, and is provided with a crane,

boom mounted manipulators and viewing windows.

When it has been ascertained that & piece of equipment should be replaced,
the reactor will be shut-down and drained and the faulty equipment will be re-

moved according to -the fdllowing procedflre.

The plug clemp bolts will be removed, the seal weld cut, all electrical
and instrument connections will be broken, the crane will be attached and in
the case of a heat exchanger, the secondary coolant lines will be cut by
direct means. After this has been accomplished and all personnel has left
the room, the equipment will be withdrawn into a plastic bag or metal con-
tainer and dropped through the transfer hatch into a storage coffin. The
spare equipment will then be dropped into place, and the room purged of any
fission gas that may have escaped during the removal operation. It will
now be possible to enter and make the seal weld and all other connections
directly. a
wlB-

Boom mounted manipulators; that can covér the entire area of the room

are provided to assist in the remcte operations and for emergencies.

8. Fuel Fill-and-Drain System

A fuel fill-gnd-drain system has been provided to serve as a molten
salt storage facility before the plant is started and as a drain system
when the fuel process circulits have to be emptied after the reactor has

been in operation.

The draining operation has not been considered as an emergency procedure
which must be sccomplished in a rela?ively short time to prevent a catastro-
phe. In the unlikely event. that all heat removal capasbility is lost in the
process system, the fuel temperature would not rise to extreme levels, 1600°F
sf greéter; in less than one hour. There could be an incentive to remove the
fuel from the process systém as fast as possible to prevent contamination in
the event of a leak. Considering any reasonable drain time there is sub-
stantial after-hest prcductién and the drain system must have a heat re-

moval system.

The éféin vessel and heat removal system is shown in Figure 7. The
fused salt is contained in afcylindrical tank into which a number of bayonet
tubes are inserted. -These tubes are“'welded to the upper tube sheet and
serve &as the'prim&ry after-heat removal radiating surface. In addition they
contribute substantial nucleer poison to the geometry.

The water boiler, which is Qperated at low pressure;.is of the Lewis
type and is inserted into the vessel from the top. The radiant heat ex-
changer heat transfer system results in double contingency prctection against
leakage. of either system.

Electric heaters are installed around the cutside of the vessel for pre-
" heating. During the pre-heat cycle the boiler would be dry end the boiler
tubes would be at temperatures above the melting point of the fuel. If
after-heat is to be removed from the fuel, the heaters would be turned off
and water admitted to the system through the inner boiler tube. This water
-17 -

UNCLASSIFIED
ORNL-LR-DWG. 35092

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

WATE R
REMOUABLE
BOILER
STEAM
S ou T
N7 |11 # L, WERT GAS
- e : l CONNECTION
|
FUueL Fian 1] |
AND DRAIN | ‘
FUEL ™

: N
- FUEL TANK

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Fica7- DRAIN TANK FOR FUEL SALT
-18~

would be introduced at a .slow rate so that it would be flashed to steam. The
steam would then cocol the outer boiier tube and the boiler would be grafiuaily

flooded with water for a maximum heat removal rate.

The rate of heat removal could be controlled by varying the water rate
in the beiler or by segregating the boiler tubes ints sections; thus varying

the effective area of the heat transfer system.

A leak in one of the bayonet tubes in the drain tank could be contailned -
within the vessel shell by;maintaining the gas pressure above the fuel below

a vaiue that would elevate the ligquid to the tube sheet level.

The entire boiler system may be removed or inserted from cverhead_withaut
disturbing the fuel system. Arleaky fuel tube could be ?lugge& at the tube
" sheet without removal of thé veésel. Heaters may be installed or removed
from overhead and only failure of the vessel wall would necessitate com-

plete removal of the unit.

Four tanks, 7 ft in diameter by 1C ft high, would be required to handle
the fuel inventory. The boiler system would have to be:designed té remove &
maximum of 8 megawatts of heat. Approximately 80 pipes L4 inches in dismeter
would have to berinstalied in each. vessel to achieve this capacity.

9. Heat Transfer Systems

Heat is transferred from the fuel to the steam by a cirgulating molten
salt. The selected salt, a mixture of 65% LiF-35% BeFe,'is completely com-
- patible with the fuel and does not cause activation in the secondary cell.
Thus, after & few minutes delay for the 11 second- fluorine activity to de-

cay, the secondary cell can be entered for direct maintenance.

Four systems in parallel remove heat from the reactor fuel; each is
independent up to the point where the superheated steam flow paths Jjoin

shead of the turbine. Figure 8 is a flow diagram cf the system.

Bach coolant salt circuit has one pump. The flow ¢of the hot sslt from
the pump is divided between 2 superhesters and 1 reheater; joining sgain a-
head of the primery exchanger. No valves are required in the sgalt circuit,
UNCLASSIFIED
ORNL-LR-DWG. 35093

 

 

 

7.Gs g

 

 

 

 

 

 

 

 

\ L __Juprg/»
yoop /000" TURBINES
I —
f-.._d
Rewsnren ‘
f.—.—.
. . = Y
[ _2,05%0007" 425p 625 |
- V—HF Tursine
l
’;f_._...
7, 210,000°
2000 p , 1000°F 2,80, 900 =

 

 

Ve

“"‘“":5 j22s " ;135‘61
7? CFS 67&?;

Cooving e O

 
  

C onoENSER

S2A¥5 c £y

 

 

 

 

 

> Y

 

 

 

 

 

X Ly

UPER WEATER
(2 #a. cinceir)

C ONDENSATE
VM

 

L&

}‘\/
C3o, 00

 

 

 

 

 

 

Fuee SALT

 

 

 

 

 

|

 

| ;\j Clozs® F45°
( £
L

Feow HRartées Lwown

 

ARE FOR COMPLETE SVYITEAL

Fre. &8

Feen warer
- AEstens

 

   

 

 

Steam Fume D — I series
Boreer ].._.w
—
|
| DEAERATOR

 

{2, 789 ooD‘{M’“‘& \

M95p Y75 .

FEFOwATER
M TERS

Feepwrrer
AN

Feow Pracram

uétm
=20 -

as control can be achieved by vaeriations of the steam flow.

The primary exchangers are o5f the bayonet bundle typefi to permit semi-
direct replacement. The superheaters are of U-tube in U-shell, while the
reheaters are of straight tube constructicn. The stgam generators are
horizontal drums, 4 foot diameter-by 24 feet long, with 5w1/2 inch walls.
These are half-filled with water, into which the siteam nozzles project for
direct contact heat tramsfer. Recircuiation of steam provides the heat for

genération of stean.

Heat exchangef data is summarized in Table 3.

10. Turbine and Electric System

Steam at 2000 PSIA and 1000°F, with reheat to 1000°F, iz supplied to the
333-Mw-rated turbine, which has a single shaft, with 4 exhaust ends. The tur-
bine heat rate is estimated to be 7670 BTU/kwh, for a cycle efficiency of '
4i,7%. The generator and station heat rates are, respectively, 7785 and 8225
BTU/kwh. The supply to the bus-ba; is 3195 Mw. These estimates are bhased c¢n
Tennessee Valley Authority*heat balances for a similar turbine(l}5 with ad-

(2)

Justments for the modified steam conditicn and different plant reguire-

ments of the molten salt reactor system.

1l. Nuclear Calculstions

A number of age theory thermal resctor calceulations were made 0 sur-
vey the nuclear characteristics of graphitewmsderafied mélten galt reaatfira(i).
In sll cases the salt was of the compcsition LiTFwBeFEwUFk {70+10-20 mole %},
and was located in cylindrical channels spaced on an 8 inch center-to-center
square array. The volume percent Qfmfuel in the reactor core was varied and
calculafiions vere made for k = ;.OS and 1.10.‘ The calculations yielded the
percentage of‘U~235 required in the initial invest@ry of ufaniums the dimen-

- sions of the reactor required“forkkeffective to be equal ta one; and the
initial conversion ratioc. Table 4 gives the results of the cslcoulaticons.
Fuel and Sodium to
Sodium Exchangers

 

Number required
Fluid

Fluid location
Type of exchanger

Tbggeratures

Hot end, °r

Cold end, °F
Change o
AT, hot end;, F
AT, cold end; OF
AT, log mean, °p

Tube Data

Material

Qutside dia, in.
Wall thickness, in.
Length, ft

Number

Pitch; in.

Bundle dia, in.

Exchanger dimensions
Heat transfer capacity, Mw
Heat transfer ares, fté
Average heat flux,

1000 Btu/hr-fte
Thermal stress¥*, psi
Flow rate, ft3/sec,

1000 1b/hr
Fluid velocity, ft/sec
Max Reynolds modulus/1000
Pressure drop, psl

* ([ OF X Max AT wall
1 - v 2

 

L]

Table 3 - DATA FOR HFAT EXCHANGERS

 

Primary System Superheater Reheater
L 8 b
fuel salt coolant salt coolant salt steam coolant salt steam
tubes shell shell tubes shell tubes
bayonet bundle U=-tube in U-shell straight
counterflow counterflow
1225 1125 1125 1000 1125 1000
1075 %5 9%65 650 9%65 620
150 160 160 350 160 380
100 125 125
110 315 335
105 207 21%
INOR-8 INOR-8 TNOR-8
0.500 0.750 0.750
0.049 0.083% 0.065
21.8 23 17.5
3173 925 750 ds
0.638 () 1.00 (a) 1.00 (A) =
33 28
50.75 in. dia x 17.5 £t long
190 81.2 27.6
5830 3315 2100
111 84 45
2000 6100 4600
19.8 16.8 7.18 2.46
90l 512
8.80
10.0 L 265 270
L7.6 16.5 23 13 23 12.5
P

 

 

Table 4
Dimension of
Vol fraction % Enrichment Unreflected Initial
of fuel ‘of cylindrical Conversion

in core uranium reactor ratio

Case F Kk e D (ft) H (ft) ICR
1 0.05 . 1.05 1.30 26.3 24,3 0.55
2 0.05 1.10 1.45 17.9 16.4 0.49
3 0.075 1.05 1.25 2k.o 22.2 0.63
L 0.075 1.10 1.39 16.6 15.3 0.58
5 0410 1.05 1.28 22.4 20.8 0.71
6 0.10 1.10 1.6 15.6 14.3 0.65
T 0.15 1.05 1.53 20.5 18.9 0.80
8 0.15 1.10 1.80 14.3 13.1 0.73
‘Q 0.20 1.05 2. 24 19.4 18.0 0.86
10 0.20 1.10 2.88 13.5 12.4 0.79
11 0.25 1.05 L.36 18.8 17.4 0.90
12 0.25 1.10 7.05 13.1 12.0 0.82

 

Case 8 of this table is quite similar to the reactor that forms the de-
sign basis of this study. The nuclear performance of the actual reactor
| design chosen for this study was calculated by the multigroup code Cornpone
on the ORACLE. The neutron balance obtained under initial conditions is
given in Table 5 below. Also given are the inventories of materials, based
‘on & total fuel balance inside the reactor and in the external circuit of
900 cubic feet. It should be noted that the enrichment of U-235 predicted
by the machine calculation is 1.3 vs the 1.8% of case 8, Table 4, and the
conversion ratio is 0.79 instead of 0.75. The lower enrichment requirement

results from the higher value of eta and some reflection.

The long term performance of the reactor was calculated for a case,
assuming an initisl inventory of U-235 of 700 kg and an initial breeding
ratio of 0.73.
~23.

Table 5

NUCLEAR CHARACTERISTICS

 

 

 

Element Inventory gkgz Neutron Absorption
U-235 717 1.000
U-238 55,000 0.790
Ii 5,760 0.053
Be + C - 0.031
F 37,900 0.026
Teakage - 0.166
eta 2.07
Conversion ratio 0.79
Table 6

CROSS SECTIONS USED FOR REACTOR LIFETIME CAILCULATIONS

Effective Cross Section Neutron Yield Capture to Fission Ratio

 

 

Tlement barns 1 o
U-235 605 2.029 0.23%
U-23%6 25

U-236 2.5

Pu-239 1903 1.84 0.58
Pu-240 3481

Pu-241 1702 2.23 0.36

Pu-242 491

 
~2l

The calculation was m&fié by an adaptation of the method described by
Spinrad, Carter, and Eggler(h). The buildup of U-236, Pu-239, Pu-2L0, Pu-241,
Pu-242, and fissién*products wvas calculated as a function of the integrated
flux-time variable. The reactor was kept critical by additions of U-235.

The cross sections of the heavy isotcopes were taken as a consistent
set from the 1958 Geneva Conference Paper P/lclé(S) and are given in Table 6.
For this long burnup reactor, the fission product crecss sections were taken
as & Tunction of time based on the celculations of Blomeke and Todd(é)i
Xenon=-135 was cqpsifiered tq be removed continuously, but all other fission
- products retainéfi. In the early phases of reactor operation this amounted
to an initial 1.3% poison plus sbout 48 barns per fission. The rate of
fission product poisoning then tapered off té about 18 barns per fission,
a value that:was assumed for the refiainder‘of the reasctor lifetime. This

(7).

latter value is consistent with that proposed by Weinberg

Figure 9 shows the accumuelative addition of U«235 required to keep
the reactor critical. In the first few months the velue is negative, that
_is, U-235 should be removed from the reactor. In actusl practice the extra
reactivity would be ccantrolled by adding high cross-secticn burneble poisons,
or possibly by contrcl rods. After the first few years the sddition of U-235
is about linear, and at the end of 32.5 years some 4100 kg have been added.

Figure 10 shows the inventories of fissionable isctopes as a funztion
of time. After seven years of operation, the inventory of U-235 starts to
rise above its initial value and is approximately twice its initial value
after 32.5 years of operation. It is evident thgtg on the basis of these
nuclear calculatiéns, the fuel could be retained in the reactor without re-
processing for the life of the reactor. The calculations are open te the
criticism that epithermal absorptions by fission products are neglected and
they might be import&nt as they build up to high concentrations. The effect
will probably be small during the first ten years of operation, however.
UNCLASSIFIED
ORNL-LR-pwG. 35094

35 MWe Fased Sa/t feoc/sr

*  Shabtly Lrriched
9000 Grz,o/ e Modors 7o

S
Q
&

!

2000}

/000

Cumuletve U5F Addition, Ky

 

& ) } i y 1 {
O 5 10 /5" 20 25 20 35

e J
0}9@/’07,(5/7_9 @Zis YEqrs

/:/-yz/re S - Cumulalive V5 Adhitor  vs Opei’cj?‘mf/ oy

-42-
UNCLASSIFIED
ORNL -LR-DWG. 35095

    

/500
315 MWe Fused Salf FegTlor

9htly Enriched
Graphile Modera?ed

  
  

3
2

500

 

 

/:/’55//«9 /50%@;)@ /fiV@/yfiJ/‘)/j /fg"

 

/5" TO & 20 35

0/0 erqting 7’7?’)6’) JPars

F ;941;/9 /0"'/&755//@ /Sof‘w/oc’-’ //71/&/775/// VS, Opefafifiy Time

—98-—
-27-

12. Fuel Cycle Costs

 

The fuel <ycle cost has besn: calouiated by Co E. Guthrie on the basis of
the nuclear caiculations and is given in Figure 1i. In each case the cost is
based on a uniform rate that is the sum of current charges and the amount that

would be placed in a sinkifig fund to téke care of future expenditures. The
| calculations include'use charges of 4% on the inventory, cost of -burnup of
U-235, and the cost of replacement of the base salt including its Li-7 con-
tent. The upper curve does not assume recovery of the U-235 and Pu inventor-
ies atu%he end of the operation, while the lower curve assumes recovery of

the fissionable isotopes.

This recovery would taske place at some central processing plant where
the operations are on large enough scaie to assure uranium proéessing costs
of no more than $100/kg of uranium, including transportation of the frozen
fuel salt to the processing plant. Such costay of processing seem entirely
(S)u With fissionable isotope recovery, the fuel cycle costs
would be about 1.25 mills/kwh at a cycle time of 5 years, 1.1 mills/kwh at
a c¢cycle time of 10 years or 1.0 mills/kwh for a 20 year cycle time. At the
end of this cycle tim&‘ihe sinking fund accumulated plus the value of the

reasonable

accumulated fissionable isotopes would pay for the purchase of a new batch
of base salt and for the cost of recovery of fissicnable isctopes from the
0ld salt.

The practical life of the fuel without any prosessing other than re-
moval of fission gases is prcbably limited by the sclubility of rare earth
Tisgion products, since when they start to precipitate they will carry
plutonium down with them. The sclubility of rare esrth fluorides in this
fuel salt has not been determined, but if the sulubility is as low as in
the LiF-BeF, base salt, it could limit operation toc as little as three
years of operation, which would indicate a fuel cycle cost of about 1.5
mills /kwh. ~

From analogy with the Zth base salts, the rare earth flucrides may be
considerably more soluble in the 20% UF), salt than in the LiF-BeF, base galt;
this will be determined. If the rare earth solubility is low, it could pro-
~-28.
UNCLASSIFIED

ORNL-LR-DWG. 35096

 

 

F/D 2hrow, away cyc/e
L w recovery ot /00,//“:"{0

foef G/c:/e (;;52‘,‘) M///g/éw/;

 

0 ¢ ! t i { [ !
o 5 /0 /5 20 25 30 35

Cyele Tme, Years

@wr@ /= Fuel Cyc/e (os? vs. Cyc/e Trme
315 MWe Frsee) Sal? Feoc7sr
=20~

vide a means of keeping the rare earth fission products, which provide a high
percentage of the fission product poisoning, from accumulating. A device
similar to a sampler :zould te suspended in the salt maintaining a surface at
a lower temperature than at any point in the salt circuit. This would
accumalate a rare earth fission product precipitate (ccntainifig some PuF5)
which could be withdrawn through locks in the same manner as the regular
samples, and provide a concentrated product for shipment to a Pu recovery
plant.

From the above discussion we feel that it is reasonable to assume a
fuel cycle cost in the range of from 1.0 to 1.5 mills/kwh. Considering
that this plant will not be built until several more years of resesrch and
develépment have elapsed, a figure of 1.1 mills/kwh seems reasonable.

13. Capital Costs

The capital cost summary is presented in Table 7. It shculd be noted
that a 40% contingency factor has been applied toc the resctor portion of the
system. A contingency value of this order is warranted because of the num-
ber of uncertainties in the large components. A 7.5% contingency factor was

applied to the remainder of the direct costs.

The general expense item charged to the plant was set at 16-1/2% of
the direct costs and the design charges were set at 5% of the ssme value.
The total capital cost of the plant leads to a value of $252 per kilowatt

of generating capacity.

Table 8 presents a more detailed cost breskdown of the reactor porticn
of the plant. The major items have been listed and in some cases an indi-

cation of the basis for the evaluation has been presented.

14. Power Costs

Power coste have been divided into three categories. These are:

fixed costs; operating and maintenance costs; and fuel cycle costs.
10.
1l.

13.

1k,
15.

16.

18.

Table 7
CAPITAL COSTS

Tand and land rights
Structures and improvements
A. Reactor systems

B. Steam system (less major Ioeffler
components )

Turbine~generator plant
Accessory electrical equipment

Miscellaneoug equipment

7.5% contingency on 10, 11, 13B, 14,
15, and 16

40% contingency on 13A

Contingency Subtotal
General expense
Design costs

Total Cost

4 500,000
7,250,000

2% 790,000

4,000,000
14,000,000
4,600,000

125002000
g 55,440,000

2 b 370 ;OOQ

9,520,000
# 11,890,000
9,150,000

2,770,000

# 79,250,000
IT.

I1I.

-31-

Table 8

REACTCR SYSTEM CAPITAL COST SUMMARY

Fuel System

A.

Core {12.5 ft x 12.5 ft right cylinder)
Reactor vessel at $10/1b # 515,000
Graphite at §4/1b 685,000

Inspection, assembly, etc. 200,000

Four fuel pumps including drives and
shielding {~ 9,000 gpm each)

Four fuel-to-salt heat exchangers at SSO/fte
having 5830 ft2 each

Piping
Main fill-and-drain system
Off-gas system
Enriching system
Preheating and insulation
Subtotal

Coolant System

A.
B.

Four pumps

Fight coolant-to-steam Superheaters at
#80/£t2 having 3315 ft= each

Four coglant-to-steam reheaters at
480/ft° having 2100 £t° each

Piping
Fill and drain
Preheating and insulation
Subtotal

Ipeffler Comporenhts

A.

B.

Four steam pumps and drives at $50/cfm
6,000 efm each

Twelve evaporator drums

1082 £t2 (total liberating surface in a b £t dia)

drums at $5000/linear ft 1082

5 X 5000
Subtotal

é 1,400,000

2,000,000

1,170,000
139,000
750,000
200,000
250,000

150,000
6,050,000

1,200,000
2,120,000

670,000
50C ,000
500,000

200,000
5,190,000

1,200,000
1,350,000

2?55030(}0
~32-

Table 8 - continued

IV.

Ve

VI,

Vil.

VIITI.

IX.

Xo

Concrete Shielding {m~ 20,000 yd3 )
Main Containment Vessel
Instrumentation

Remote Maintenance Equipment

Auxiliary Systems

Spare Parts

Fuel pump rotary element 200,000
Salt pump rotary element 120,000
Fuel heat exchanger 300,000
Salt-steam superheater 265,000
Salt-steam reheater 165,000

Original Fluld Inventories
Fuel salt - 900 £t required at SlSOO/ft3

Coolant salt - 1900 ft5 required at $1500/f‘t5 2,850,000

1,350,000

Reactor Systems Tctal Cost

2,000,000
500,000
750,000

1,000,000

500,000

1,050,000

4,200,000

g 23,790,000
«35

The fixed costs are the charges resulting from the capital investment in
the plant. This smount was set at 14% per annum of the investment, which in-

zluded taxes, insurance, asnd financing charges.

The annual operating snd maintenance expenditure was assumed to consist
of the following:

Annual Charge
Labor end Supervision g 900,000
Reactor System Spare Parts | 1,000,000
Remote-Handling Equipment 400,000
Conventional Supplies 400,000
TOTAL Anniiel Expenditure g2,700,000

The fuel cycle costs are discussed elsevhere and the three contributing
categories add up as follows: ‘
Annusl Charge millsg/kwh

Fixed Cost $11,100,000 540k
Operating and Maintenance 2,700,000 1l.22
Fuel Charges 2,400,000 1.10
TOTAL Annusl Charge 816,200,000

TOTAL Power Cost 7. 36%

* Based on an 80% plant factor.
=3l

References
The data used were for Unite Nos. % and 4, the. Gellatin Steam Plant,
Gallatin, Tennessee.

Reese, H. R. and Carlson, J. R., "The Performsnce of Modern Turbines",
Mech. Engr., March 1952, p. 205

MacPherson, H. G.;, ORNL-CF-58-10-60, "Survey of Low Enrichment Molten-
Salt Reactor", October 17, 1958. '

Progress in Nuclear Energy, Vol. VIII, pp. 289-320, McGraw-Hill (1957)
Pigford, T. H., Benedict, M., Shanstrom, R. T., Loomis, C. C., and

Van, Ommesloghe, B., "Fuel Cycles in Single-Region Thermal Power Reactors',
Conf. 15/P/1016, June 1958.

Blomeke, J. 0., and Todd, Mary F., "U-235 Pission Product Production gs

a Function of the Thermsl Neutron Flux Irradistion Time and Decay Time",
ORNL-2127 (1958)

Progress in Nuclear Energy, Vol. VIII, p. 282, McGraw-Hill (1957).

See ORNL-CF-59-1-13, Guthrie, C. BE., "Fuel Cycle Costs in a Grephite
Moderated Slightly Enriched Fused Salt Reactor.
WO 0] O
9 3

10,
11.
12.
135=-37-
38.
29

L1,
Lo,
43,
Ly,
L5,
L6,
L7,
L8,
L9,
50.
51.
52-53.
54-55.
56,
57

58,

=35 -

Distribution
L. G. Alexander
M. Bender
EO Sa Bettis
A. I.. Boch

W. F. Boudreau

E. J. Breeding

R. A. Charpie

W. Re. Grimes

W. H. Jordan

B. W. Kinyon

M. E. Iackey

Je. A. Iane

H. G. MacPherson

E. R. Mann

L. A. Mann

W. B, McDonald

A. J. Miller

J. Wo Miller

H. W. Savage

A. W. Ssvolainen

Jo A. Swartout

A. Taboada

F. C. Vonderlage

A. M. Weinberg

G. D. Whitman

C. E. Winters

J. Zasler

Central Research Iibrary
Iaboratory Records
Iaboratory Records, ORNL-RC
ORNL ~ Y-12 Technical ILibrary,
Document Reference Section
M. J. Skinner