ORNL-2348.txt

   
   
   

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ORNL-2348
Metallurgy and Ceramics
TID-4500 (13th ed., Rev.)

February 15, 1958

aa, )33

 

COMPONENTS OF THE FUSED-SALT AND SODIUM CIRCUITS
OF THE

AIRCRAFT REACTOR EXPERIMENT

 

CENTRAL RESEARCH LIBRARY
DOCUMENT COLLECTION

LIBRARY LOAN COPY
DO NOT TRANSFER TO ANOTHER PERSON

If you wish someone else to see this
document, send in name with document
and the library will orrange o loan.

 

OAK RIDGE NATIONAL LABORATORY
operated by

UNION CARBIDE CORPORATION
for the

U.S. ATOMIC ENERGY COMMISSION

 

 

 
 
Contract No. W-7405-eng-26

REACTOR PROJECTS DIVISION

COMPONENTS OF THE FUSED-SALT AND SODIUM CIRCUITS

OF THE

AIRCRAFT REACTOR EXPERIMENT

H. W. Savage
G. D. Whitman
W. G. Cobb

W. B. McDonald

DATE ISSUED

 

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

ORNL-2348

MARTiNHMARIETTA EiiR||G|V SYSTEMS |LIBi|A|RIES

3 4456 03BLLY?7? 7

 

 
CONTENTS

AD SIFACT ettt et ettt et et e ettt e et et e et et ater et e n e et ense st ersans 1
IETOAUCTION Lottt et b et e e e e et e ee et e et ee et e e e s e e ereeteoteneetesaearense e seresaass 1
Engineering Development for the ARE ... et 1
Forced-Circulation Test LooOps ..ottt e et e et ae e eneoes 3
Mixing of Fuel and SOdiUm ....o..ooiiiiiicc et ettt e 3
BeO-Sodium Compatibility .....cccoooiiiiiiiiie e et 8
REAEIOE COTE oottt ettt e ettt oot ettt et et eee e e e ee e e e e e s e et e et s e et e e e et raneeens 8
Fill and Drain EQUIPMENT ..ot ettt e e ee et et e et ee e st eee e e eeenaans 8
Fill and Drain TanKs ....ooooio e e ettt et er e et e e et et et eeeeeae e, 8
Fuel ENFiChmMent oottt e te et ee et s e ettt e e eeeee e 8
Piping and Leak Detection ..ottt sttt ettt e 11
P i e ettt e e s ere e et e er et b et eaereesae e crereer e 11
ek D et O ION ittt ettt ettt e, 12
Preheaters and InsUlation ..ottt 13
Heat DUMP Sy Stems ..o ittt ettt ettt e bs et e e bt sttt es st eeeeees e 13
Fuel Heat DUMP Sy Stem ..o e e e ettt ee e e e eae s eae e 13
Sodium Heat DUmp System ..ottt ettt 13
Heat ExXchanger Test .ot e et e e eeeaeane e 15
High-Temperature Isolation Valves ...t 19
Sodium Oxide Cold Traps oo ettt e e e e ete et et e et et e e 19
INErt GOS8 Sy STEMS ..ttt ittt ettt et sttt e b b re et e e et b er b ssest e e te ke eatt s e et s raabessenbere et ereea b s antesseas oo 20
2T TGP S s b s 20
P UMD S et e ettt bbb st er e et e s et et st eee e 21
Packed Seals ... e e et sttt et e 21
Rotary Gas-to-Lubricant Seals ... e 23
Centrifugal Sump Pumps oo e ettt ae e 23
PUMP D SigN .. i et bbbt st ne 25
PUMP AUXTTIGRIES oottt h et b e e e et e et e st e e bt e e eb e ettt aeas e enn e s e nbeeaabeantesaensane e 34
P UMD DIIVES oottt e et e e et e e eat b ete e e ae s es st et e eee e st e e e ans eeesae ek s aan snsen e e enne e entneeneeenns 34
LU Cation Sy SEEmS L it ettt ettt e sttt et aen e e et n et ean 34
High-Temperature INstrument@tion ...........cooooiiiiiiiioiiieiie ettt ettt et e er e et ae s e 34
T emperature MeasUremMeEnt ..ot et reer e s ettt et ee e e e e e e et e e e ettt et e anate et e e 34
PressUre MeasUrEemENnT ... . ettt e et ettt et r et a et e reraes 34
it

 
F oW MO SUICMENT ooeeeereeee oot e e e oo e eeeee e e e e e s e ee e ee e e e e ee e mae e e ae e ae e e e m e aeee et ettt A bbe bbb sb bbb s raaesneebsrnrnrenn s 34

L eVel MEASUIPEMENT oottt e e e e e e e e e et e et e e e e esbbe aa e e eabbe b e e e s ernb e an s renae e e an s 36
Component Installation ...ttt bbb 36
SUMMIGIY et ciete ittt cteet et et e ekt et e e ese et e et e e ese £ eteems e b e et et et e ebeshee et rones e e ead e R e s s saen Ao es e e R e a e e re e a e e s e e s eeae e e ab et sebe s 36
ACKNOWIEAGMENT 1ottt ettt bR a e e et 38
COMPONENTS OF THE FUSED-SALT AND SODIUM CIRCUITS OF THE AIRCRAFT
REACTOR EXPERIMENT

H. W. Savage
G. D. Whitman

W. G. Cobb
W. B. McDonald

ABSTRACT

The Aircraft Reactor Experiment (ARE) successfully demonstrated the feasibility of generating

heat by fission in a fused-fluoride circulating fuel. Most of the heat was removed from the reactor

by the fused fluoride at 1580°F.

Sodium at 1350°F was used to cool the BeO moderator. With

minor exceptions all the components proved to be adequate.,

The development of components and fabrication techniques for this reactor consumed a four-

year period, during which time the technology for handling high-temperature fluids was extended

to equipment operable above 1500°F. The methods used for determining compatibility of materials

under static and dynamic conditions, standards for materials, ond techniques for welding, fabri-

cation, and assembly and the design criteria for pumps, seals, valves, heat exchangers, cold

traps, expansion tanks, instrumentation, preheating devices, insulation, etc., are described,

 

INTRODUCTION

A high-temperature Aircraft Reactor Experiment
(ARE) generated heat by fission of U233 in a
fused salt composed of UZ235F , ZrF ., and NaF
for a period of 221 hr, ending on November 12,
1954. As shown in Fig. 1, the maximum equilibrium
temperature of the saltwas 1580°F, with @ maximum
temperature gradient of 380°F. About 25% of the
heat was transferred to sodium, which was circu-
lated at a maximum equilibrium temperature of
1350°F with @ maximum temperature gradient of
120°F. The heat was then transferred from both
salt and sodium to helium in separate closed
circuits and was finally transferred to water. The
maximum heat power generated was 2500 kw, and
the total amount of energy was 96,000 kwhr. The
entire operation was performed in remotely
controlled equipment in three concrete enclosed
pifs.]

Prior to the nuclear power operation the systems
and components were checked out? by operating
the fused-salt circuit for a period of 388 hr at
temperatures above 1200°F and the sodium circuit
for a period of 561 hr at temperatures above 600°F.

 

]E. S« Bettis et al., **The Aircraft Reactor Experi-
ment ~ Design and Construction,’
Eng. 2, 804-825 (1957).

2E, S, Bettis et als, **The Aircraft Reactor Experiment —
Operation,’t Nuclear Sci. and FEng. 2, 841-853 (1957).

Nuclear Sci. and

The ARE than four years of
research and development and is believed to be
the first reactor to generate nuclear power above
1500°F.

It is presumed that the reader is familiar with
the basic concepts of the design, the physics, the
chemistry, and the metallurgy which led to this

required more

particular reactor system, since these topics have
been covered in other reports,1+3=6 as has the
operation? of the In the
following discussion emphasis is placed on the
principal components of the reactor experiment
and on the solution of some of the important de-

reactor experiment.

velopment problems involved.

ENGINEERING DEVELOPMENT FOR THE ARE

Figure 2 shows the principal phases of engi-
neering development. Since the original concept
of the ARE was that of a solid-fuel-pin, beryllium

oxide—moderated, sodium-cooled reactor, the first

 

3A. M. Weinberg et al., ‘*Molten Fluorides as Power
Reactor Fuels,’” Molten Fluoride Reactors, ORNL
CF+57-6-69.

W, K. Ergen et als, **The Aircraft Reactor Experi-
ment — Physics,’" Nuclear Sci. and Eng. 2, 826-840
(1957).

SW. R. Grimes et als, *"Chemical Aspects of Molten
Fluoride Reactors'' (to be published).

bW, D. Manly et als, ORNL-2349 (Sept. 17, 1957)

(classified).

 

 
SCDIUM
CIRCULATED AT 42D gpm

 
 
  

HELIUM
ELOWER

ABSORBER RCD

HEAT £EXCHANGER

 

 

 

 

 

Fig. 1.

UNCLASSIFIED
ORKNL-LR-DWG 282t6

 

 

 

 

1950 19951 1952 1953 1954

. e , R -
BASIC 1500°F SCDIUM HANDLING h ‘
BASIC 1500°F MOLTEN FLUCRIDE

SALT HANDLING e
CORROSION STUDIES h
MECHANICAL PUMPS C e e
ARE COMPONENT DEVELOPMENT | | ..

\
_ - o i o

ARE SHAKEDOWN AND OPERATION | ol

— | e e

EXPLORATQORY MAJGR DEVELOPMENT - IMPROVEMENTS

OR SHAKEDOWN: PEQIOD; 7

Fig. 2. ARE Development Phases.

year was devoted to studies of the corrosion,
dynamic, and engineering problems of handling
sodium at 1500°F. In 1951 the concept was
changed to that of fuel elements containing
stagnant molten salt, and shortly thereafter to
that of a reactor containing no fuel elements but
circulating a fused fuel salt.! The sodium circuits
were retained to cool the beryllium oxide moderator

   
 

 

UNCLASSIFIED
ORNL-LR-DWG 14562AR

FLUQORIDE
CIRCULATED AT 46 gpm

 
    

HE

BLOW

s

HEAT EXCHANGLA
m ‘
T : +

5
M
pral

S

 

 

  

 

Schematic Diagram of the ARE.

and the reactor pressure vessel. The second year
was devoted primarily to determining compatible
structural materials and fused fluoride compositions
and to investigating the engineering and fabrication
problems involved in handling fused fivorides at
temperatures between the melting temperature,
~950°F, and the proposedreactor operating temper-

ature of 1500°F.

Development of pumps, heat exchangers, valves,
pressure-sensing instruments, cold traps, and
other components began in late 1951 and continued
to the summer of 1954, culminating with the
defivery of pumps and other components to the
experimental reactor facility, Many of the tech-
niques developed were extrapolations from data
already available from the extensive experience in
the temperature range of 800to 1000°F with sodium
and sodium-potassium alloy at Argonne National
Laboratory, Knolls Atomic Power Laboratory,
and Mine Safety Appliances Co. Had not this
experience been available, the development period
would certainly have been much longer.

Design of the reactor, development of pumps,
valves, heat exchangers, and other components,and
containment of sodium and molten sait at 1500°F
presented new and perhaps fascinating problems.
Equally challenging were the problems concerning
fabrication,  construction, preheating, instru-
mentation, and insulation of reliable leak-tight

high-temperature circuits made of Inconel. Much
of the technology of these developments has been
d.7=9 We wish also to acknowledge the
important contributions to the technology by
hundreds of engineers and scientists, and regret
that it is impossible to give them individual

recognition.

reporte

Although some of the avenues investigated in
developing components were not fruitful, many of
the solutions used in the ARE have become
accepted practice. Among these are the following:
1. use of fire-resistant insulation,

2. use of all-welded construction,

3. standardization of specifications for quality
and inspection of reactor and heat transfer
circuit materials,

4. standardization of welding procedures,

5. standardization of inspection specifications
for assembled components as to fabrication and
leak-tightness,

6. development of high-temperature instrumentation
to include pressure, temperature, flow, and
liquid level measurement,

7. development of the high-temperature sump-type
centrifugal pump — now in routine use in the
laboratory.

Completely adequate designs were not available
for valve seats, bearings in sodium or salt, pump
operable against the liquid, or pumps
in circuits with more than one free

seals
operating
surface. Mechanical valves were used in the high-
temperature reactor circuits, but because of their
dubious qualities provisions were made for freeze
valves and frangible diaphragms. Overhung shafts
were used in the pumps to avoid bearings and
seals in the liquid, and the pump tank (or sump)
was enlarged sufficiently to become the expansion
tank of the system, thereby preventing multiple
free liquid surfaces.

During pump development a number of pumps for
sodium were operated successfully with frozen
sodium shaft seals. On the other hand, frozen or
packed seals for fluorides invariably resulted in
excessive wear and failure.

Fabrication of reliable leak-tight high-temper-
ature circuitry required the use of all-welded

 

7C. B. Jackson (Editor-in-chief),
Handbook, Sodium-NaK Supplement, 3d ed,,
Washington, 1955.

8w. B. Cottrell and L. Ae. Mann, Sodium Plumbing,
A Review of the Unclassified Research and Technology
Involying Sodium at the Oak Ridge National Laboratory,
ORNL-1688 (Aug. 14, 1953).

IA. M. Weinberg, **The Nature of Reactor Technology
and Reactor Development,!’ Molten Fluoride Reactors,

CRNL CF-57-6-69.

Ligquid Meials
GPOC,

stress-free structures, use of seamless tubing and
pipe, and adherence to carefully defined welding
and inspection techniques. All material is 100%
dye-checked, is vltrasonically and radiographically
inspected for defects before use, and is rejected
when there are any observable defects. Materials,
labeled to
minimize errors in selection; analyses are per-
formed to avoid mislabeling; and all critical welds
radiographed before ac-

including weld rod, are carefully

are dye-checked and
ceptance. With these controls, leaks are rare,
unless the part involved has been severely over-
stressed.

FORCED-CIRCULATION TEST LOOPS

Much of the technology and component develop-
ment was accomplished in forced-circulation loops.
These loops, which were also used to determine
corrosion rates,® were of two types. Figure 3
shows a typical sodium test loop which uses an
electromagnetic pump. Figure 4 shows a salt test
loop which employs a down-flow centrifugal sump
pump capable of 1600°F operation. This pump has
been improved since its development in 1951 and is
still used routinely in the laboratory. It provided
some basic ideas for reactor pumps subsequently
developed. It was partly on the basis of corrosion
data obtained from such loops that Inconel and a
composition®:® were chosen for the
reactor. These loops established that use of a
single alloy in a high-temperature system would
minimize corrosion and mass transfer.® Inconel
was chosen for both the sodium and the fuel
circuits, although type 316 stainless steel is more
resistant to attack by sodium, in order to avoid

fuel salt

duplex-material construction for reactor fuel tubes
(see Fig. 5). (Mass transfer in sodium-Inconel
systems is considerably higher than in sodium-—
stainless steel systems, but was not expected to
be excessive in the periods of operation anticipated
for the reactor experiment.)

Screening tests were conducted with hundreds
of natural-convection loops, Fig. 6, to eliminate

structural materials and fused
6

fess desirable
fluoride compositions.

Mixing of Fuel and Sodium

Other determined the effects of

sudden
could have occurred if one of the reactor fuel tubes

experiments
mixing of fuel salt and sodium, which

had cracked or ruptured during operation. The
reaction was known to be exothermic. Insoluble

reaction products were frequently found in sufficient
 

Fig. 3. Sodium Corrosion Test Loop.

! UNCL ASSIFIED
PHOTO 22434
SAFETY
ALLES
YLST 8¢ gy
1% "f-!5 4?5:

 
 

Fig. 4. Fused=Fluoride Corrosion Test Loop Employing a Down-Flow Centrifugal Sump Pump (See Fig. 22 for Cross Section of Similar Pump).

 
UNCLASSIFIED
DWG. 16336
REGULATING ROD

ASSEMBLY

SAFETY ROD
GUIDE SLEEVE

 
 
 
 
   
 
 
 

——TUBE EXTENSION
THERMAL SHIELD CAP
_—— THERMAL SHIELD TOP

SAFETY ROD ASSEMBLY FUEL INLET MANIFOLD

 
  
 

THERMOCOUPLE
LAYOUT

 
 

TOP HEADER

      

CORE ASSEMBLY TOP TUBE SHEET

    

HEATERS

  
 
 

REFLECTOR COOLANT
TUBES

8eC MODERATOR
AND REFLECTOR

    
  

FUEL TUBES

THERMAL SHIELD
ASSEMBLY

   
 

PRESSURE SHELL

TUBE
SHEET

STUD

SUPPORT ASSEMBLY
BOTTOM HEADER

FUEL OUTLET
MANIFOLD

e e e 8O Y2 i, REF.

|

THERMAL SHIELD

THERMAL SHIELD CAP BOTTOM

e 235%g in. REF. —

 

  

\SUPPORT ASSEMBLY

02 4 6 i &
ELIUM MANIFOLD s;ghaéuaggzfisde

SCALE IN INCHES

Fig. 5. The Reactor {(Elevation Section).

 
UNCLASSIFIED |
PHOTO 22432

 

 

 

 

 

 

 

 

Fig. 6. Natural-Convection Corrosion Test Loops.
quantity to stop circulation. While the pressure
rise observed was small, local temperature

transients of 200 to 300°F were observed.

BeO-Sodium Compatibility

Specimens of BeO were suspended in sodium
and examined for attack. Little or no beryllium
was found in the sodium, but porosity of the BeO
was clearly evident, since the specimens exuded
sodium for long periods after they were removed.
Visually, the BeO showed no damage of conse-
quence at any temperature of interest in the reactor.

REACTOR CORE!

In the original concept the reactor core provided
sixty=six ]]/4-in.'rubes in parallel vertically through
holes in hexagonal beryllium oxide bricks and
seventy-nine ]/2-in. tubes in the outer reflector
section of the core, When the shift was made to a
circulating fused salt fuel, the higher viscosity and
reduced over-all flow rate required made it essential
to reduce the number of parallel paths through the
reactor in order to maintain Reynolds numbers in
the turbulent range.! Consequently, the 66 tubes
arranged in six parallel routes, each
comprising a serpentine (see Fig. 5) of 11 tubes
in series connected by U-bends at the top and
bottom. This arrangement introduced the question
of whether the core could be filled without first
being evacuated and also made complete draining

The outer sodium tubes

were

of the core impossible.
were left unchanged.

A full-size core, Fig. 7, was mocked up
with glass and metal tubing, and its hydraulic
characteristics were studied with water-glycerin
solutions for viscosity effects and with tetra-
bromoethane (manometer fluid) for density and
viscosity effects. It was found that complete
filling of the core could not be accomplished by
pressurization of the fluid from the fill tanks,
because gas became trapped in the multiple
vertical rises of each parallel circuit through the
reactor core; nor was it possible to expel this gas
and establish full flow within the maximum head
provided by the fuel pump. With a partial vacuum
above 400 mm Hg, filling was certain and flow
could be established in each of the six parallel
paths without difficulty. Full blowdown draining
was impossible because liquid expulsion ceased
as soon as one circuit was opened to the gas flow;
however, the liquid could be forced or chased out
with another liquid. As the result of these experi-
ments, the reactor was filled under vacuum, and

spent fuel was displaced with barren salt, followed
by several flushings.

FILL AND DRAIN EQUIPMENT
Fill and Drain Tanks

Tanks were attached to both the sodium and the
salt circuits to receive the initial charges of
sodium and barren salt. With the use of helium
pressurization the materials were transferred from
the tanks into the reactor Once the
circuits filled, the interconnection was
closed by a mechanical valve and the liquid could
not be drained back. (As explained elsewhere,?
the sodium valves did not seal tightly, but this
situation was tolerated.)

circuits,
were

There were three fill tanks for sodium and two
for barren salt, all located in a tank pit as shown
in Fig., 8. In addition, a hot-fuel dump tank con-
taining 89 vertical through tubes for convection
cooling with helium in the pit was reserved for
the spent fuel. Each tank was equipped with
external electrical heaters and thermocouples and
was of welded construction. Valved connections
to a supply of spectroscopically pure helium
(<10 ppm 02, —60°F dew point) permitted the
fluids to be blanketed at any desired pressure and
prevented exposure of them to air or water vapor,

Fuel Enrichment2:4,5

The approximate quantity of U235 needed in the
fused fluoride mixture for the reactor to reach
criticality was known from an earlier, low-temper-
ature critical experiment. Consequently, the first
phase of nuclear operation of the ARE was a high-
temperature critical experiment.24  The initial
charge of fused fluorides to the reactor was a
highly purified® 50-50 mole % mixture of NaF
and ZrF,. To this a mixture of 33% mole %
U235F4, 66%_13 mole % NaF was added?:4¢5 in
known increments wuntil criticality, and subse-
quently the desired excess reactivity, was
reached. At the completion of this operation the
reactor fuel was approximately 6.18 mole % U235F4,
40.73 mole % ZrF ,, and 53.09 mole % NaF, which
had a melting temperature of 1000°F.

The highly enriched fuel was stored in several
containers under helium, and was batched down to
smaller containers in quantities ranging from 2 to
33 |Ib. The smaller containers were connected
successively to the fuel circuit as shown in Fig. 9,
and 23 transfers were made in order to fully enrich

the fuel. A Z-in.-dio tube, heated by a current

 
"UNCL ASSIFIED
PHOTO 21918

 

Fig. 7. Full-Scale Reactor Core Mockup.

 

 
 

0l

 

      
  
  
  

//SOD\UM FILL AND
: DUMP TANK

SODIUM FILL AND
DUMP TANK NG.5

 

HOT FUEL DUMP TANK

. ' o — i
SODIUM FiLL AND DUMP TANK NC 4 - \

TANK NO. 3 (NOT USED)—

Fig. 8. ARE Fill and Drain Tanks.

 

UNCLASSIFIED
ORNL-LR-DWG 6222R
UNCLASSIFIED

TO HELIUM SUPPLY ORNL-LR-0OWG 6395

   

| VENT
TO HELIUM SUPPLY

TO VACUUM
PUMP
VENT
"N ELECTRICAL CONNECTIONS
1 FOR RESISTANCE HEATING
1

E E/S#’«TCH Can
. 5> |
E 1 g TRANSFER POT
MM

L—O\./'EJ\I

= TRANSFER LINE

   
 
  
    

 

 

ELECTRIC HEATERS

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

g{TO HELIUM SUPPLY

 

 

FLOOR LEVEL AUXILIARY

|

|

;

|
g

L —><—— VENT
5
= VENT

 

FUEL PUMP

SAMPLING LINE SUMP
DISCHARGE

| - SUCTION

|

\

|

. Fig. 9. Equipment for Addition of Fuel Concentrate to Fuel System.

passed through it, led to the sump of the fuel UNCLASSIFIED
pump, and injections were made with the pump in ORNL-LR-DWG 28217
operation and the fused salt circulating. The
high melting temperature of the enriched salt
(1185°F), unanticipated cold spots in the transfer
tube, and leaks in tube fittings used in the transfer 100 deg
line made enriched-fue! transfer more tedious than
had been expected. The exact amount transferred

 

 

 

was determined after completion of the operation
by comparing differences in weights of the con- W /////@1 fi&\ j&
tainers emptied. A-J L

g in 1/46 in.

PIPING AND LEAK DETECTION 1 .
/16—|n.MAX.

Piping

 

All the reactor high-temperature piping was

Y
seamless sched-40 inconel pipe with full-penetration V /%// /m\\k N

welds, Fig. 10, at joints. The pipe was installed
between anchor points, prestressed sufficiently

- at room temperature to be approximately stress-free Vg —in. MAX.
at operating temperature. All piping connections
to the pumps, the heat exchangers, the reactor, Fig. 10. Full-Penetration Weld.

11

 

 
and other components were also made with full-
penetration welds.

Development studies of flanged joints, with or
without oval-ring or other gaskets, revealed that
they were not satistactory for critical high-temper-
ature service. An oval-ring-gasketed flanged joint,
Fig. 11, was usually successful for a time if the
flanges and bolts were exposed to convection
cooling by air {thus introducing a cooled region in
the piping), but ultimately the bolt tension would
relax sufficiently that if it were not discovered and
tightened a leak would ensue. In addition each
flanged joint required two welds, whereas only
one was needed to join the pipes.

Leak Detection

All high-temperature piping was jacketed by a
stainless steel annulus system which, in principle,
could be monitored for indications of a leak in the
primary piping. Before these jackets were installed,

 

 

the piping was subjected to gas pressurization and
a soap bubble check.

The monitoring principle designed into the
annulus was based on the detection of fluid vapors
in  the circulated in the jackets
surrounding the piping.

helium gas

Copper wool was placed in strategic spots
throughout the annulus and was withdrawn periodi-
cally and placed in a phenolphthalein solution,
which would turn a characteristic red if an alkali
metal was present, In addition, samples of the
annulus gas were passed through a flame photometer

for more sensitive detection of sodium vapor.

A high-temperature leak check was run on the
fuel system by pressurizing krypton gas into the
fuel piping before fluid was added to the system
and by spectroscopically checking the annulus
helium for traces of krypton.

UNCLASSIFIED
DWG 115014

GAS

e

 

   

Zm‘ - LIQUID LEVEL |

- .,

 

 

 

 

 

 
 
 
 

FLANGES =<

 

  

 

 

 
  
   
  

 

OVAL-RING GASKET — T

 

 

 

 

 

 

 

 

 

 

 

 

/‘/SUCTION BAFFLE PAN L
N
. LT ]
//, 2
/,,/ THIMBLE //—-’/Z/:{/?////?{;/_ ““““““ %
S /
\ N g _;_:_‘//f’/‘{’/// - —— 174
X ‘\ W % ; 7
WY

 

 

 

 

 

 

 

 

DIAPHRAGM PLATE

 

DISCHARGE

Fig. 11. Typical High-Temperature Flange Joint.

 
No positive indication of a leak was detected by
any of these methods, and the systems were
deemed to be leak-free.

Helium in the annulus served to even out temper-
ature differences during the preheating period, and
the annulus wall would have helped to contain
leaking material had a leak occurred.

Preheaters and Insulation!

Piping was preheated electrically. All the

straight runs of high-temperature piping were
surrounded by refractory clamshell heaters con-
taining totally enclosed Nichrome V wire; the
heaters were held in position by external straps.
Groups of nine nonadjacent heaters, each rated at
1 kw, connected in closed delta, constituted a
circuit regulated by a Variac. There were several
hundred circuits, and preheat temperatures were
monitored by Chromel-Alumel thermocouples welded
to the pipe at strategic locations between heaters.
Bends, tees, and odd geometry parts were either
fitted with appropriately shaped tubular heaters or

were wrapped with flexible heaters.,

Vessels such as the dump tanks were fitted
with flat strip-type heaters, and the pump tanks
used flat, ceramic heaters containing embedded
Nichrome V wire. The reactor vessel was also
equipped with strip-type heaters to bring its
temperature to 600°F so that the sodium could be
circulated through it to bring the temperature of
the core to 1200°F before the fused-salt circuit
was filled.

Development work with smaller systems had
indicated that a temperature of at least 100°F
melting temperature at the coldest

needed to avoid the

above the
observable
danger of freezing during filling.
both systems

location was
As soon as
circulation became nearly

isothermal.

began,

The entire high-temperature structure was covered

with preformed insulation made of calcined
diatomaceous silica, bonded with asbestos fibers.
This is one of several fire-resistant insulations
commercially available which will not react
significantly with sodium or fused fluorides. A
good discussion of thermal insulating materials is

given in the Liquid Metals Handbook.”

HEAT DUMP SYSTEMS!
Fuel Heat Dump System

The heat generated in the reactor was removed
by circulation of the fuel through two heat ex-
changers in parallel in the main fuel-process
As shown in Fig. 12, the two heat ex-
changers, each designed for 600 kw, were located
in a closed duct in which helium was circulated by
a centrifugal blower to transfer heat from the two

fuel-to-helium exchangers to two helium-to-water

stream.

exchangers. The water was dumped at a low-
temperature level, and no attempt was made to
extract useful power from the system.

The fuel-to-helium heat exchangers were of
all-welded, two-pass, fin-and-tube design fabricated
from Inconel pipe and tubing and are shown in
Fig. 13.

The manifolds were fabricated from 21/2-in.
sched-160 pipe, and the tubes were fabricated from
1-in. No. 12 BWG tubing bent to produce 43 straight
lengths in five fluid circuits. Fuel from an inlet
header passed through two of the circvits in
parallel to an intermediate header and thence
through the remaining three circuits in parallel to
the outlet header.

The 2-in.-OD by 0.024-in.-thick fins
fabricated from type 304 stainless steel and were
swaged into continuous spiral grooves on the tube

were

surface.

Since the fuel had a melting point in excess of
950°F, provision had to be made for preheating the
heat exchangers during filling and low-power
circulation of the fluid fuel. Retractable heat
barrier doors, containing heaters, were located in
the ducts at each face of the heat exchangers, and
heaters were placed around the exterior of the
heat exchanger. When power was to be extracted
from the system, the barrier doors were raised and
helium was circulated through the duct.

The helium-to-water heat exchanger, Fig. 14,
was of conventional design, and the tubes were
fabricated from ASME SA-179 seamless steel

tubing covered with copper fins.

Sodium Heat Dump System

The sodium circuit, which removed heat from
the BeO moderator-reflector and the core pressure
shell, employed a similar scheme, in which heat
was transferred from the sodium to helium and then
from the helium to water.

13

 

 
vl

Fig- ]20

 

 

Fuel Heat Dump System. (a) Fuel-to-helium exchanger, (4) helium-to-water exchanger,

UNCL ASSIFIED
PHOTO 10902

 

 
 

UNCLASSIFIED
PHOTO 22531

Fig. 13, Fuel-to-Helium Heat Exchanger.

The sodium-to-helium heat exchangers, Fig. 15,
each designed for 325 kw, were of all-welded
construction and of the same materials as those
used in the fuel-to-helium heat exchangers. Fifteen
parallel fluid passes having 75 straight lengths
in the helium duct formed the active section of
this exchanger. Two separate ducts, each con-
taining a centrifugal blower, a sodium-to-helium
heat exchanger, and a helium-to-water heat ex-

changer, were used. The sodium heat exchangers

operated in parallel and were connected to the
suction side of the sodium circulating pumps.
Preheating of the sodium heat exchangers was
achieved by retractable heat barrier doors and duct
wall heaters.

Heat Exchanger Test

A spare exchanger and a pump were set up, and
are shown in Fig. 16 (the snow trap included in the
system is not shown); the characteristics of the

15

 

 
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DWG. A-3-2-21AR

 

 

   

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

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=, DESIGN TEMPERATURE | 1500°F | 340°F
< e
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HALF VIEW OF STAT HEAD WITH COVER REMOVED

Fig. 14. Helium-to-Water Heot Exchanger.

16

 
- a7 -
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al -

SUPPORT PLATES |
ES |

12%¢

  
  

“VENT TUBE

 

-~ 205 S e
SECTION A-A
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—— ‘—,I,fff 25V . o
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ELEVATION

Fig. 15.

 

INCHES

2 0 4 8 2 16

NOTE: ALL DIMENSIONS IN INCHES

 

 

UNCLASSIFIED
DWG. A-3-3-Z21AR

 

 

 

  

il.l,,,,, s

_._161/2 — .

 

 

 

 

 

 

 

SECTION C~-C

Sodium-to-Helium Heat Exchanger.

 

 
18

 

PUMP

 

 

 

UNCLASSIFIED
ORNL-LR-DWG ic88

="

r
{

 

PUMP TANK

 

 

 

  

tY,~in. SCHEDULE 40 PIPE

 

 

 

 

— -

 

 

 

 

 

HEAT EXCHANGER

L D)

 

 

 

 

 

 

 

 

VENTURI

HAIRPIN TUBES

 

 

 

 

FILL AND DRAIN
TANK

 

 

Fig. 16. Fuel-to-Helium Heat Exchanger Test Loop.

 
system were determined from isothermal tests and
from periodic coolings by removal of insulation from
the heat exchanger. Facilities were not available
for applying sufficient heat load for heat transfer
experiments; consequently, corrosion and mass-
transfer characteristics requiring a thermal gradient
could not be determined. A non-enriched-fuel
fluoride was used in this test. After the heat
exchanger had been operated at 1400°F for 2000 hr,
it was removed for examination; testing of the
pump continued for an additional 1750 hr and was
then terminated for other use of the space and the

pump.

HIGH-TEMPERATURE ISOLATION VALVES

High-temperature fuel and sodium valves were
required for isolation of the dump tanks and the
spare sodium pumps. The fuel circuits included
frangible diaphragms to isolate the spare fuel
pump. Freeze valves were provided for use in the
event that the mechanical fuel drain valves failed.
During operation only the sodium mechanical
valves leaked, the freeze valves did not have to
be used, and the fuel drain valves were kept
closed except for the initial filling and final
draining operations.

The valve, Fig. 17, was fabricated from Inconel
and was designed with a double bellows seal so
that the pressure on the inner or primary fluid seal
bellows could be balanced with inert gas. The
body was fabricated from a 4 by 2 in. reducing tee,
and the plug and seat were hard-surfaced with
Stellite No. 6. This material was found to cause
fusion bonding of seat and plug when the valve
was closed and held at temperatures above 1200°F.

Interchangeable spring-loaded pneumatic actu-
ators were available for holding the valves in
either normally open or normally closed positions.
The spring force was used to hold the valve stem
in its normal position.

The spare fuel pump was isolated from the fuel
system by frangible valves to ensure that enriched
fuel would not leak into the spare system. These
isolation
plug and seat geometry of the high-temperature
valves and clamping a 0.013-in. nickel diaphragm
in the valve body to produce a positive seal,
Actuation of the valve stem would rupture the

devices were made by modifying the

diaphragm and connect the spare pump to the fuel
In the event of the spare pump having to
there was sufficient inventory

system.
be put in service,
in the main pump to prime the spare pump and to

UNCLASSIFIED
ORNL-LR-DWG 29252

    
   
 
    

  

Sy \\\\% \\\\\\\\\\\
T O Tz

   
   

 

 

 

 

 

w\\\\\\\\\§\\\'\\\ J \\““\ TRt

   

 

 

\L | fl

Ponns -s

2 & =
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1 za E
i @:g ' s _.- NORMALLY CLOSED
‘-s RS ..,.. ACTUATOR
= IIIIIIIIA\\:IIIIIIIII

] \-..__

 

\0

-

 

Y

?rg\'\ L

N

F L rA]

 

TLLT S TINANY

 

 

i
‘ ! NZZZ272

 

LA

‘} UPPER BELLOWS SEAL

A S S S SN \

|

 

SRR IR

BELLOWS BACKUP
GAS CONNECTION -~/

7//]

VALVE BODY

   

LOWER BELLOWS
SEAL

 

INCHES

 

 

 

Fig. 17. High-Temperature Valve.

freeze the residual fuel in the lines to the main
pump and isolate it from the fuel system.

SODIUM OXIDE COLD TRAPS

The sodium circuit included two bypass cold
traps for the removal of oxide. Design, based on
KAPL experience with traps in a sodium circuit
operated at a maximum of 1000°F, was completed
before sufficient information on the more stringent

 

 
requirements for service at 1500°F was available.
The circuit to one cold trap leaked at a thermo-
couple weld, and as a result both traps were
removed from the system. The absence of cold
traps thereafter probably caused the sodium to be
of poor quality; and even though corrosion and
mass transfer might have been thereby accelerated,
it was believed that the short period of power
operation would prevent the consequences from
being serious.

INERT GAS SYSTEMS

The free surfaces of high-temperature sodium and
fused fluorides were blanketed with high-purity
helium. In the case of the fuel pump, where gaseous
fission products also escaped, an off-gas line led
through a snow trap adjacent to the pump to a
charcoal bed and exhausted into a high stack.
Helium was chosen as the blanket gas because it
is chemically inert and is readily available in a
very pure, very dry state. Its lightness is a minor
disadvantage.

Experiments indicated that to hold the sodium
and fluoride quality to the desired levels, helium
with an oxygen content of less than 10 ppm and a
dew point below —60°F was essential. Cylinder
helium, generally, did not meet these specifications;

! " .
0.010-in. SHIM STOCK = l

therefore every cylinder was examined, and those
in which the helium did not meet specifications
were rejected. By this practice and with careful
control of cylinders used, rejections were reduced
to about 10%. A distribution system in which the
helium is piped directly from tank car or tank
truck has been found to be superior to the use of
individual cylinders and to avoid completely the
problem of quality control.

To guard further against possible contamination
of reactor fluids, the helium gas lines were equipped
with scrubbers containing sodium-potassium alloy
at room temperature. While these devices were
very effective in removing oxygen and water, they
introduced the danger of transported NaK vapor
and liquid and subsequent fouling of downstream
gas lines. Such fouling was not encountered in

the reactor experiment, however.
ZrF4 TRAPS

The fuel system cold trap, more commonly called
a ‘‘snow’ trap, is in Fig. 18; it was
in order to eliminate the moderately
problem associated with ZrF

shown
designed
serious vapors
escaping from the free surface of the fused fluoride
in the fuel pump sump. Since ZrF4 has no liquid
sufficiently cold

phase, it will condense on

UNCLASSIFIED
ORNL-LR-DWG 28223

: —1,-RESISTANCE HEATING
: LUG

  

 

 

 

 

 

 

 

BAFFLES - }

1Y

 

 

 

 

i ;Q‘CJ;

 

  

| '
7

50 O

 

 
 

134 -in. RADIUS

 

 

 

 

 

 

 

 

 

 

SUPEREX INSULATION-----

 

 

 

 

 

N L :
“BAFFLES
e

 

 

 

 

'- sl O
o T T CALROD

 

 

Fig. 18. ZrF4 Snow Trap,

20

 

 
surfaces as a solid, which does not drain back
into the circulating system. Instead, it builds up
on the cold surface to the extent that it may
ultimately cause fouling. Fortunately, the material
is flocculent and powdery and, although underneath
layers slowly form well-developed crystals, does
not foul moving parts readily.

Pump development work pointed up the necessity
of maintaining the line between the pump tank
and the snow trap at o temperature of approxi-
mately 1300°F to prevent condensation of ZrF
ahead of the trap. The temperature was mainfcineé
by direct resistance heating of the interconnecting
tube and by additional heat applied to the inlet
end of the trap. The remainder of the trap was
allowed to lose heat to its surroundings. The
baffled path was perhaps 95% effective, but some
condensate was found beyond it. At the low power
level of this reactor operation, such carry-through
was not significant, but significantly higher power
would have required higher gas flow rates and a
better snow trap to remove beta heat in the off-
gases and to prevent fouling of downstream piping
and fittings.

PUMPS

Each reactor circuit required a pump and a spare.
The flow and head requirements for the reactor
pumps are given below:

Fuel Sodium
Impeller diameter, in. 8.125 8.125
Flow, gpm 46 152
Head, ft 28 78
Speed, rpm 1080 2000
Suction pressure, psig 0.3 36
Discharge pressure, psig 41 63
Specific speed 602 940
Specific gravity of liquid 3.38 0.79

Electromagnetic pumps were considered initially
for the sodium circuit, and a 150-gpm experimental
pump (Fig. 19) similar to the Mine Safety Appliances
Co. double-stage, single magnetic circuit design
was built and tested. Since electromagnetic pumps
could not be used for fused salt because of its
low electrical conductivity and since a single
centrifugal pump development program with minor
modifications would solve both flow problems,

UNCLASSIFIED
DWG 115044

 

MAGNETIC COIL

Fig. 19. Schematic of Two-Stage Electromagnetic
Pump.

development of the electromagnetic pump was not
carried forward.

Space limitations in the heat exchanger pit of
the reactor test facility appeared to indicate that
horizontal-shaft pumps would be preferable. Such
pumps required cantilevered shafts and impellers
and a means of sealing the liquid in the pump. In
addition, since the fuel pump was to handle radio-
active liquid containing fission products, the seal
had to be reliable at all times and to leak only in
an absolutely controlled fashion.

A canned-motor type of pump, which would
appear to have been the most desirable, was not
available at the time for the required operating
temperatures.

Packed Seals

The hydraulic problems of impeller and volute
design were solved by straightforward adaptations
of conventional practice, but the problem of sealing
liquids around horizontal shafts was more compli-
cated. Packings, labyrinths, high-viscosity ma-
terials, and frozen pumped fluid seals were con-
sidered, and many versions were tried. All
resulted in shaft seizure or periodic galling and
leakage to some degree. Frozen sodium seals

21
(Fig. 20)

promise, but each required nearly constant atten-

and frozen lead seals showed real

tion and was subject to periodic ejection of
frozen material. In the frozen lead seals, there

was a liquid lead-~to=liquid fluoride interface

ahead of the frozen lead seal, and this combination
liquids were

was considered because the two

nearly inert to each other. Since at least one
reactor type!® now built uses frozen-sodium-sealed
pumps, it is apparent that this seal is considered
practical.

Various packings were tried for fluorides, and
braids and

the most promising were metallic

 

107e SRE, a sodium-graphite power reactor buiit by
Atomics International  Division, North American
Aviation, Los Angeles, California.

SEAL COOLANT —-
\

4

LEAKAGE
CONTAINER|

FROZEN Na —!

— MOLTEN Na 215°

graphite powders impregnoted with metal powders
such as nickel, These failed as soon as the

fluoride penetrated sufficiently to freeze and

the shaft,
“high-viscosity seal,”” the packing consisted of

abrade In another type, called the
namely, glasses having
It was believed that
the fuel fluoride would mix physically with the

fluorides containing BeF .,

no sharp melting temperature.

salt packing and establish a liquid-to-glass viscous
interface. The seal was extremely temperature-
sensitive and required very hard shaft surfacing;
a water-cooled packing gland had to be provided to
These
seals failed because penetration of fuel salt was
progressive until the packing

largely dissolved. Also, the shafts (Fig. 21) were

remove the heat generated by friction.

material was

UNCLASSIFIED
ORNL-LR-DWG 28394

F MAX. ~HEATER

 

.

Fig. 20. Frozen Sodium Rotary Shaft Seal,

22
UNCLAASIFIED
PHOTO-21340

-

 

Fig. 21. Shoft from BeF, Seal.

abraded badly, the abrasion progressing toward
the outer end of the gland.

Rotary Gas-to-Lubricant Seals

Prior to the work on packed seals, a 7-gpm down-
flow centrifugal pump for laboratory use, Fig. 22,
had been developed and was found to be very
satisfactory at temperatures up to 1600°F, Hence,
concurrently with the packed-seal investigations,
larger versions of the gas seal were being tried.

In  this packings of graphite-
impregnated asbestos and Teflon were tried, but

development,

leakage rates of gases were too high for the

to be used with radicactive liquids.
seals offered the most promise, but

packings
Face-type
long-term wear-resistant materials were required,
and there were important questions concerning the
kind and amount of lubrication to use,
Silver-impregnated graphite running against case-
hardened cold-rolled steel gave the best results,
It was found that if the surfaces of both members
were optically flat and lubricated and if the mass
and inertia of the movable parts were sufficient
for relative positioning of the parts to be insensi-
tive to vibration, oil leakage could be reduced
consistently to less than 5 cc/day and frequently
to less than 2 cc/day, and gas leakage to less

than 15 cc/day.

Centrifugal Sump Pumps

Use of a rotary gas seal required a vertical-
shaft sump-type pump in which the high-temperature
liquid could be gas-blanketed and the liquid level
carefully controlled. A centrifugal-type pump with
an overhung shaft was selected to avoid close-
running surfaces and bearings in the pumped
liquid.

The next problem involved a solution to degassing
requirements. In the reactor fuel system there were
two degassing problems; in the sodium system
there was one. Both systems had to be purged of
trapped helium gas at startup, and during power
operation it was desirable to remove fission gases,
particularly xenon, from the circulating fuel,

The sump-type pump operated essentially in a
pot of the fluid to be pumped. Early designs
recommended geometric displacement of the tank
and pump inlets; this led to ingassing due to
turbulence in the tank. In the pump configuration
finally evolved, liquid in the pump tank returned
to the pump inlet only in a laminar fashion and
without vortexing, and the tank and pump inlets
were in axial alignment at the bottom of the tank.
Sufficient clearance was allowed between the
pump shaft and its housing for 5 to 10% of the
pumped liquid to continually bypass the main fluid
circuit by flowing up the shaft and back through
the pump tank to the pump inlet, Some continuous
bypass in that flow direction is a requirement of
a sump pump to prevent ingassing, and the resi-
dence time for liquid bypassing the main circuit
through the sump tank must be sufficient to allow
gases to escape. When the bypassing liquid is
spread out thinly on a surface of large area, escape
of entrained gas is aided. On the other hand, xenon

23

 

 
   
      
  
     
  

P SHAFT

 

LEY

EARINGS

SEAL SLEEVE

SEAL RING

DRAIN

SPLASH GUARD

UNCLASSIFIED
OwWG. 17877

SPARK PLUG PROBE

SEAL ASSEMBLY

GRAPHITE

SOFT-IRON OvaL
RING GASKET

=1 CONTROL GAS

. ———MAXIMUM LEVEL

     
  

PUNY.
£
WLEL

~—— IMPELLER

0030 in. IMPELLER CLEARANCE

 
 

PUMP
DISCHARGE

Fig. 22. Gas-Seal Centrifugal Pump.

24

. —_ MINIMUM LEVEL

2 3

U

INCHES
from fission-product decay was removed principally
by scrubbing the liquid fuel with helium.”

Thus the processes were somewhat competitive,
and the liquids were never free of gases in solu-
tion, Undissolved gases, on the other hand, were
very rapidly minimized during pump startup to an
equilibrium value too low for detection. As was
pointed out in other repor'rs,]'2 fission gases were
certainly liberated to the heat exchanger pit due to

leak in the off-gas line from the fuel pump. The
degassing means were proved to be effective, since
the gases could have come from only the free
liquid surface in the fuel pump. Postrun esti-
mates2'? indicated that at least 97% of the generated

  
 
  
  
  
 
 
   
  
   
  
 
 
 
  
 
  
   
    
  
   
 
   
   
  

DRIVE SHEAVE -

BEARING CLAMP RING -

BEARING HOUSING -

TOP SHAFT SEAL ASSEMBLY
SEAL FACE RING

 

INTERNAL SHAFT COOLING
LUBE OIL GUIDE TUBE ---

SEAL FACE RING---..
SLINGER - .~
LOWER SEAL ASSEMBLY - -- . __
SPACER AND HEAT DAM - .. .
OIL DRAIN HEADER RING L

PUMP BODY ASSEMBLY.

TN )
THERMOCOUPLE GLAND - ;j::*,;*-’ il i}@‘
- (,—-—-r-‘,w -

 

FLEXIBLE CONNECTOR

 

INLET
DISCHARGE

 
 
   
 

fission gases were liberated continually by this
technique.

As shown in Fig. 23, each pump tank was equipped
with sloped troughs to conduct the bypassing
liquid to the walls of the tank, where it joined
smoothly with the main body of liquid in the tank.
Despite bypass flows as high as 8 gpm, the liquid
in the tanks appeared to be quiescent, which
simplified level indication problems in the tanks.

Pump Design

Except for differences in pump tank design
because of the tanks being used also as the

UNCLASSIFIED
ORNL-LR-DWG 6218R

-UPPER SEAL OIL RETAINER
UPPER SEAL LEAKAGE DRAIN

- PRESSURE SLEEVE
SPACER SLEEVE

~-0IL SHELD
Y[””———r OIL INLET
- BEARING {UPPER)

B.H. BREATHER
.. HEATED GAS VENT AND
BEARING SPACER -~ LIQUID INJECTION NOZZLE

t——e” .- RESISTANCE
.~ BEARING (LOWER) = HEATER CONNECTION
SEAL RING ’

- DIL OVERFLOW : . LAVA SPACER
- g f

 

v - RESISTANCE
g LEE =" HEATER CONNECTION
Ol DRAIN i e
T 48 .. COOLANT
He CONNECTION = ~REPLACEABLE
-4 ! "H‘"h SEAL WELD
A g PUMP TANK  — ey
i “COVER FLANGE '

   

CLAMP

 

 

 

 

PUMP  TANK

   
  
 

LEVEL INDICATOR FLOAT

IMPELLER 1%
{30 VARIABLE INDUCTANCE LEVEL INDICATOR

HOUSING i}
N
ANTISWIRL 1 ,
BAFFLE o
e

Fig. 23. ARE Fuel Pump.

25

 
expansion tanks of the liquid circuits and except
for the difference in volute size because of the
different pumping rates required for the sodium
and the fused fluorides, the pumps and their
auxiliaries were of similar design and many parts
were interchangeable, The basic design features
of this pump have been described previously;”
however, this discussion merits repetition of the
more important aspects,

Operation of a high-temperature sump-type cen-
trifugal pump was assured by separating high-
temperature components such as the pump tank,
impeller, impeller housing, and shaft extremity
from the bearings, seals, lubricant, and drive motor
by appropriate thermal barriers so that the latter
parts could be operated nearly isothermally at
temperatures below 200°F,

The structural connections also provided for
continuous alignment of the rotary and stationary
parts and for maintenance of running clearances
to avoid binding over a temperature range of
1500°F. The parts were carefully stress-relieved,
and preheating was applied as symmetrically as
possible to minimize asymmetries in temperature
distribution which would have distorted the align-
ment of the parts.

In the ARE, shielding of the lubricant and pump
elastomeric seals was judged to be unnecessary
since the total dose would not exceed 108 r, at
which dose the viscosity of petraleum lubricants
would have increased approximately 10% due to
decomposition and buna-N elastomers would not
have taken a permanent set. (Neoprene is less
resistant to radiation damage than buna N.)

Bearings, Shaft, and Seals. -- The bearings,
shaft, and pump seals were supported in a type
316 stainless steel housing flanged at the bottom
for bolting to the pump body assembly, which in
turn was bolted to the top of the pump tank with
an oval-ring gasket for precision alignment. The
flange and housing enclosed a hollow torus (see
Fig. 23), through which water was circulated as
a primary thermal barrier,

The bearings (Fig. 23) were conventional ground
ball bearings. A pair of angular contact ball
bearings mounted face to face at the upper end of
the shaft supported it and accepted the thrust and
provided for some thermal distortion in the bearing
housing, The bottom bearing was a single radial
deep-grooved ball bearing, providing axial align-

 

TH, w. Savage and W. G. Cobb, ‘‘High-Temperature
Centrifugal Pumps,’” Chem Eng. Progr. 50, 445 (1954).

26

ment of the shaft without axial restraint., These
bearings were spray-lubricated by a petroleum-
base mineral oil of approximately 39 SSU viscosity
at 170°F, which was diverted from the main stream
of lubricant.

The shaft, which carried external sheaves for
the belt drive, was Inconel. |t was hollow to the
elevation of the lower seal and had an upper
opening to receive the main lubricant stream to
cool it and a lower opening just above the primary
rotary pump seal and just below the lower bearing
to deliver the lubricant to the seal cavity, The
seal cavity was run partially filled with lubricant,
which was returned by gravity to the lubricant
circuit sump. A breather line from the lubricant
sump opening into the bearing housing at a level
below the upper bearing supplied the gas entrained
by the oil turbulence. This maintained the pres-
sure on the lower seal at oil sump pressure.

The lubricant flowed at 3'/2 gem and removed
about 4 kw of heat (when pumping 1500°F liquid),
which was dumped to water in an external lubri-
cant cooler. Operating temperature of the bearing
housing and associated parts averaged about 170°F,

The lower or primary pump seal, Fig. 24, con-
sisted of a case-hardened cold-rolled steel,
optically flat (<3 helium light interference bands)
runner fixed on the shaft, and a stationary, tool-
steel, equally flat runner brazed to a formed
bellows sealed with a copper gasket and bolted to
the bearing house near the bottom of the seal
cavity,

11

/lé-in.-fhick annular floating ring of silver-

These surfaces were separated by an

impregnated graphite with 3/3 -in. wearing lands on
each side, also optically flat. Exposed areas of
the seal parts were equalized for pressure balance.
Pressure across the seal was held at 1 *+ 0.5 psi
positive on the oil side to minimize oil and gas
transport across the oil films between the optically
flat surfaces. Laboratory tests showed that oil
leakage of less than 5 cc/day (frequently less
than 2 cc/day) and gas leakage of less than
15 cc/day were usual with carefully made seals
of this type. Oil leaked to a catch basin in the
bearing housing, which was continually gravity-
drained.

A secondary seal (see Fig. 23), consisting of a
wearing ring of bearing bronze mounted on a brass
bellows and of a case-hardened steel runner fixed
to the shaft, was provided at the upper end of the
shaft just above the upper bearings. Since oil
leakage from this seal was accessible for disposal,
SEAL FACE RING

SLINGER

LOWER SEAL ASSEMBLY

SPACER AND HEAT DAM

PUMP BODY
ASSEMBLY

UNCLASSIFIED
ORNL—-LR-DWG 28293

 
  
  
  
 
 
 
 
  
  
    

BEARING (LOWER)
SEAL RING
OIL OVERFLOW

OIL DRAIN

HEAT BAFFLE
1 | |

Fig. 24. Pump Primary Rotary Gas Seal.

tightness specifications were less strict, 20 cc/day
feakage being considered acceptable.

Other pump seals which were not subject to
relative motion were metal O-rings and buna-N
rubber. Buna N was used instead of natural rubber
for increased chemical resistance to lubricant.

Each assembly of these components was suk-
jected to a 100-hr mechanical shakedown test at
170°F,

was taken apart, repaired, and reassembled and

If any fault was observed, the assembly

the test repeated until no fault was observakle.
Each

high-temperature sodium pumping test, which in-

assembly was then subjected to a 100-hr

cluded two complete temperature cycles from 600
to 1400°F.
of the sequence of tests until each assembly
passed both tests. (It had been observed that with
this degree of mechanical operability assured, the

Any fault resulted in full repetition

pump mechanical assembly would nearly always
operate for more than 2000 hr without serious
mechanical difficulty.)

Impeller and Impeller Housing. — Figure 23

shows the impeller and impelier housing for the

fuel pump, and Fig. 25 shows the corresponding
parts for the sodium pump. VYolutes and impellers
were designed according to conventional practice.

After many futile attempts to obtain satisfactory
defect-free castings, the volutes and impellers were
all made as weldments, heat-treated, and machined.
Since this method of fabrication of impellers of
this small size, 8 in. in diameter, is unusual,
the fabrication steps are illustrated in Figs. 26
through 29.

Running
its housing were minimized empirically, labyrinths

being provided around the inlet hub to reduce

clearances between the impeller and

internal pump leakages to acceptable values.
Clearance in the shaft labyrinth annulus was
established at about 0.060 in. radially, which

permitted priming of the pump and establishment
of the degassing bypass flow, without which the
pump would ingas. Other clearances around the
impeller permitted a % -in. axial dislocation of
the impeller and housing; the axial dislocation
resulted from a thermal expansion differential

occurring during preheating.

27

 

 
UNCLASSIFIED
ORNL-LR-DWG 29253

  

 

L -~ VARIABLE INDUCTANCE

LEVEL INDICATOR

 

 

/CO\/ER FLANGE

           

-~ FLANGE CLAMP

  

 

e
i

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

,_ N c o
‘LHL;M z 2 =
1] ,.,.__._.__:__.fiH <38
== 5 E
ok
m

  

DISCHARGE ELBOW

DISCHARGE

INLET

Fig. 25. ARE Sodium Pump.

28

 
UNCLASSIFIED
Y—10924

 

Fig. 26. Rough-Machined Parts of Inconel Impeller,

The upper portion of the impeller housing was
attached to the pump body assembly (see Fig. 23)
and was sealed to the lower portion of the impeller
housing by a radial sealing metal O-ring carried
by the upper impeller housing.

Pump Tanks. — The pump tanks were of two
types but performed essentially the same functions.
Each tonk had a central inlet with antiswirl vanes
at the bottom and a vertical discharge outlet at
one side.

Each tank cover plate (see Figs. 23 and 23) was
the support member for the pump parts, and carried
thermal shields on the under side, the lower
portion of the impeller housing, and liquid level
sensing devices. Thus if the cover plate became
distorted, all critical pump parts moved together
so as to maintain alignment. A movable dumbbell-
shaped section of pipe having ball and socket
type of surfaces at one end and ball and cylinder
at the other end connected the volute discharge to
the tank outlet.

the pump tank with a replaceable weld and was

The cover plate was sealed to

held by heavy C-type bolted clamps. The tank had

four welded-on lugs around its periphery for

supporting the entire assembly. (Each pump tank
was an anchor point for the prestressed piping
attached to it.)

The service requirements of the pump tanks
differed.

system to eliminate problems of multiple dynamic

Each was the expansion tank of its
free surfaces. Each was designed to accept the
total fluid expansion
operation and was sized so that the liquid level
would always be above the pump inlet and below
the thermal shields carried by the cover plate.

expected during reactor

Table 1 gives a comparison of volumetric data for
Each pump tank was provided
with several spark-plug-type probes and a float-
These are described

the pump tanks.

type liquid level indicator,
later.

Each system was filled to the normal operating
level (which was higher than the minimum prime
level) and then degaossed. The liquid temperature
was then reduced to the minimum expected and the

29
UNCLASSIFIED
Y-12437

 

Fig. 27. Impeller After Heliarc Welding of Vanes to Drive-Shaft Hub. Impeller is tack-welded to o carbon steel

strong-back.

level trimmed to the minimum acceptable pumping
level, 1 in. above the pump inlet bell. Thereafter,
if liquid temperature increased, the level could
not exceed the permissible maximum unless reactor
design operating temperatures were exceeded.
(Little danger could ensue if these temperatures
were exceeded, insofar as the pumps were con-
cerned, unless the transient should be sufficient
for the pump tank to be flooded and thus endanger

30

the pump seals and off-gas lines.) If the pump
should lose prime after initial trimming of liquid,
it could be reprimed by increasing the system
temperature to expand the fluid, or by adding liquid
temporarily.  Reactor operation did not require
either technique.

A greater volumetric expansion of fuel than of
sodium was expected; consequently the fuel pump
expansion tank was 32 in. in diameter and the

I e
UNCLASSIFIED
Y-12487

 

Fig. 28. Impeller with Fluid-Entrance Hub Plug-Welded into Position.

sodium pump tank was 24 in. in diameter. Another
criterion was that the parasitic volume in the fuel
pump tank (below the minimum pumping level) be
held to a minimum,

Pump Performance. — The performance ot each
pump model was first determined with water and
then checked on sodium and, for fuel pumps, on
fuel salt. No significant disagreement between

performance on water and reactor fluids was

The performance characteristic for the
sodium pump is shown in Fig. 30 and in Fig. 31
for the fuel pump.

found.

The pumps installed in the reactor circuits were
also equipped with crystal sound detectors at
each bearing location, with the amplifier located
in the control room. The detector at the lower
bearing of the fuel pump indicated excessive noise

shortly before scheduled reactor enrichment, The

31

 
source of the noise could not be isolated specifi-
cally but was analyzed to include a dominant fre-
quency nearly the same as the calculated pre-
cession frequency of one ball in the lower bearing
race. Since the noise did not increase, reactor
operation was continued without pump difficulty.

Postrun examination revealed that the noise had

not been coused by bearing wear but rather by o
slightly loose, vibrating, dumbbell-shaped pump
discharge piece.

Pumps built according to these specifications
are now in routine laboratory use and have proved
to be very reliable,

 

 

UNCLASSIFIED
¥-10794

Fig. 29. Completed Impeller.

Table 1. Compatison of Yolumetric Data for Pump Tanks

 

 

Fuel Sodium Fuel Sodium
Ins ide diameter (in.) 32 24 Distance (in.) between
3 Minimum operating level and 2,'.":‘ 2}/4
Parasitic valume (in.”) 208 1060 minimum prime level
Minimum prime level and normal 121’32 12]%2
Tomlausuble expansion velume 3175 2170 operating level
{in.”)
Normal cperating level and 21]"32 12;’32
maximum operating level
To+u3| usable exponsion volume 1.84 1.26
(Ft9)
Volume (in.a:l between
Moximum permissible pressure 22 70 Minimum operating level and 320 800
at 1400°F (psi) minimum prime level
Minimum prime level and normal 1083 570
Inlet pipe size (in.) 3 3 operating level
1 1 Normal operating level and 1772 800
Discharge pipe size (in.) ]’IE 2;’2 max imum operating level

 

 

 

32

 
120

8O

UNGLASSIFIED

ORNL-{R-0WG 348

 

2100 rpm

 

 

  

 

 

[ BOO rpm T s

 

 
 
   
 
 

 

 

 

 

 

 

 

 

 

  

 
 
 
   

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

CF_O'Q()\(}%
£ ARE DESIGN POINT
a 60
<l
u 1500 rpm
40 -
1200 rpm
900 rpm j;
20 TN e - _- S
600rpm ;
|
0 i
0 40 80 120 160 200 240 280 320
FLOW (gpm)
Fig. 30. Performance of ARE Sedium Coolant Pump.
UNCLASSIFIED
ORNL—LR—DWG 4524R
60 i ‘ 60
i |
| |
! |
| \
50 |- e et 400 rpm T 50
| . \
40 — - 40
‘ \EFFIC!ENCY AT 1200 rpm
14200 rpm 59
= . T
= 1 O
=
< 30 [ - W
bt o
T / W
b Ly
Ll
|
20 |-
800 rpm
600 rpm
10 ———
0 1
0 10 20 30 40 50 60 70 80
FLOW (gpm}
Fig. 31. Performance of ARE Fuel Pump.

 

 
PUMP AUXILIARIES

The principal pump auxiliaries are the drive and
the lubrication system.

Pump Drives

All pumps were driven by vertically mounted
15-hp d-c motors to obtain variable speeds. For
flexibility each motor drove its pump through
V-belts. The motors were an air-cooled type with
class B windings to withstand ambient tempera-
tures of 125°F., Continuity of pump motor opera-
tion was assured by storage batteries in parallel
with a normally operating motor-generator set,

Because of possible failure of the fuel pump
V-belts due to radiation damage, several belts
were irradiated to doses of 10° r and tested on
pumps in the laboratory without failure. As a
possible substitute, chain drives were ftried ex-
perimentally but were abandoned when the driving
and driven sprockets were badly damaged due to
the vertical positioning of the shafts,

Lubrication Systems

Each pump was provided with a hermetically
sealed forced-circulation lubrication system con-
sisting of a gravity return from the lower pump
seal cavity, an oil reservoir, two parallel-connected
canned-rotor pumps, a filter, a water-cooled heat
exchanger, a thermostatically controlled valved
bypass line around the heat exchanger, and a line
leading back to the pump. These systems operated
without difficulty,

HIGH-TEMPERATURE INSTRUMENTATION

High-temperature instrumentation was required
in the ARE for record, alarm, and control purposes.
Sensing elements inserted in or attached to high-
temperature system components provided signals
for the measurement of temperature, pressure,
flow, and liquid level. Low-temperature and most
inert-gas instrumentation was essentially con-
ventional and is not discussed further. Controls
and nuclear instrumentation are discussed else-

where, 142

Temperature Measurement

The ARE included about 800 (0.032-in. wire)

Chromel-Alumel thermocouples. Each thermocouple

34

was crimped in an Inconel sleeve and Heliarc-
welded to the piping or other parts. The signals
from these thermocouples were indicated on 27
strip-chart recorders and multipoint temperature
indicators, A few thermocouples gave erroneous
signals due to inadequate thermal or electrical
insulation, and by the end of the experiment about
5% of the thermocouple circuits, randomly located,
were open (i.e., failed).

Previous experimentation indicated that a pair
of thermocouple leads could be welded to the
appropriate part separately, be twisted together
and welded as a whole, or inserted in the flowing
stream in a well. In welded arrangements it was
essential to weld the leads in such o way that
neither alloy of the couple penetrated the parent
metal of the part to which it was attached.

The above methods provided response in the
millisecond range which was adequate for the
ARE and most high-temperature experimentation,

Pressure Measurement

Three high-temperature pressure transmitters
were used for reactor liguids, two for fluoride as
it entered the reactor and one for sodium as it
left the reactor.! Each was a null-balance bellows-
sealed transmitter,'? Fig. 32, which translated
liquid pressure to a balancing gas pressure for
Experiments indicated that the
transmitters had an accuracy of 2%, and there

measurement.

were only minor difficulties experienced with zero
shifting, No difficulties were encountered in the
reactor experiment,

In development work it had also been found
possible to measure pressures from high-temperature
gas regions directly with Bourdon-type instru-
ments, and these were used extensively.

Flow Measurement

Flow in the fluoride circuit was measured by a
high-temperature rotameter, and flow in the sodium
circuit was measured by two electromagnetic flow-
meters of standard design. The high-temperature
rotameter, Fig. 33, had a float which could move
vertically in response to fluid flow rate within a
tapered barrel inserted in the piping. The float
was attached to one end of a vertical Inconel rod,
and the other end carried a tapered iron core
chrome-plated to resist corrosion. The rod and

 

125 Moore Products Cos development.
BELLOWS
PRESSURE CHAMBER
/ GAS CHAMBER

Z // /////// TLLL

  

 

UNGLASSIFIED
DWG. 17879

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

  

Fig. 32. Moore Nullmatic Pressure Transmitter.

 

UNCLASSIFIED
DWG-DSK-15781B

 

 

2 1/2 in

s
NEEDLE VALVE
TO PRESSURE GAGE—
F_i,,, - -
- 133, m REF T
— n "“
- 3 g
! ul W
! a @ |
i o [0l
i O O |
2 3] ‘
T ‘?fl -

SR AR \\fi" ‘

I - N

/U—\VA COIL FORM
S (FIRED AFTER MACHINING)

 

Trrrzozaa TrrEn

Z o um x

 

 

s | - Moo ias
A ~«\-“4<«wfl‘e“‘\“\<"\ i

Y,
= = Ygin
TAPERED BARREL”

ROTAMET ER FLOAT

 

Fig. 33. Rotameter-Type Flowmeter

core were contained in a vertical pipe filled with
static fluorides and were held at a constant tem-
perature near 1300°F, An inductance coil, operable
at 1300°F, was placed around the pipe at the
section containing the tapered portion of the iron
instrumentation measured the

core, and external

variation in inductance of the coil circuit as the
iron core and rod changed position due to motion
of the attached float.

The principal difficulties with this device were
variations in temperature at the inductance coil

 

e aroToor -

 

1 \

L4 TAPERED CORE

/ (%= =D x15-in - LONG CHROME -
PLATED MILD STEEL BAR)

INDUCTANGE COIL

with Varioble Inductance Level Indieator.

which affected its accuracy, and oxidation of the
coil wire due to the high temperature, which changed
its resistance and limited its life. Refinements
have shown that the coil life'3 could have been
extended by the use of different materials of con-
struction and that accuracy could have been im-
proved by alteration of rotameter system circuit

design,

 

134, L. Southern, Closed-loop Level Indicator for
Corrosive Ligquids Operating at High Temperatures,
ORNL-2093 (May 16, 1956).

35
Level Measurement

 

Dynamic liquid surfaces blanketed by helium
existed in the fill and drain tanks and in the ex-
pansion (pump) tanks of the reactor circuits.

Spark-plug-mounted electrical probes were used
in all tanks, and floats were used in the pump
tanks to provide information on liquid level and
in some applications to actuate alarms,

A typical spark-plug-mounted electrical probe is
shown in Fig. 22. Four were provided in each fill
and drain tank to indicate maximum and minimum
levels and two intermediate levels. Four probes
were also used in each pump tank: one each at

the minimum operating level, at the maximum
operating level, both of which actuated alarms, at
and at the normal

the minimum priming level,

operating level. (The significance of each of
these levels is discussed in ‘“‘Pump Tanks.”’)
Each pump tank also included a float within an
enclosing liquid stilling well (see Figs. 23 and
25), and the location of the liquid level was de-
termined in the same fashion as with the high-
temperature rotameter described in "‘Flow Measure-
The float provided
continuous liquid level indication,
type
Some shorting of probes

ment’' and shown in Fig. 33.

During each presented
difficulty,
perienced,
by the introduction of an increased current in the

This burn-out feature

operation some
was  ex-

This was eliminated in the fuel system

circuit to burn out the short,
did not work with sodium because of its low elec-
trical resistance.
float
the sensing coils caused shifts in the zero and

The principal objection to the

system was that temperature variations in

range settings.

COMPONENT INSTALLATION

The facility for the reactor experiment provided
three concrete-walled pits, one for the reactor,
one for the fill and drain tanks, and the third, and
largest, for the bulk of the high-temperature plumb-
ing, pumps, heat transfer equipment, and essential
process auxiliaries. The pits were very crowded,
and accessibility, needed for installation or for
corrective measures during installation and shake-
down, was very limited.,

The amount of U%3° available necessitated re-
duction of the fuel volume external to the core,
and to gccomplish this one-half the heat-dump equip-
ment planned for the fuel circuit was eliminated.

36

Figures 34 and 35 show the arrangements of the
sodium and fluoride circuits used, Unfortunately,
the fuel

further crowding of the remaining fluoride circuitry,

volumetric reduction also resulted in

The pits were closed at the top with concrete
shield blocks during nuclear operation, so that
during this phase of the experiment all the equip-
ment in the pits was completely inaccessible and
could be operated only remotely and as indicated
by the instrumentation provided,

SUMMARY

In summary it is noted that

1. the reactor circuits were at temperatures above
1200°F for 609 hr, 221 of which comprised the
nuclear experiment;

2. most of the reactor circuit components performed
their assigned functions adequately;

3. the reliability of all-welded high-temperature
liquid circuits was proved;

4, leaks did not occur in the liquid circuits during
reactor nuclear operation;

5. the only failures of major significance were
loss of leak-tightness in the fuel off-gas system
and loss during prenuclear operations of cold
traps in the sodium circuits;

6. most of the individual high-temperature com-
ponents of this system could probably have
operated successfully many hours beyond the
termination of the experiment;

7. the effect on continuity of nuclear operability
imposed by the lack of drainability of the
reactor core could not be assessed since no
situation occurred which demanded draining;

8. the

cells

leakage of fission-product gases to the
complicated gas and radiation control
problems but did not prevent successful opera-
tion of the reactor nor orderly termination of
the experiment,

The ARE and related engineering development
tests have shown that it is feasible to provide
the necessary equipment for the reliable removal
of heat power generated by nuclear fission in a
forced-circulation fused fluoride fuel and to transfer
this power by means of molten alkali metals and
the fused fluorides at temperatures as high as
1580°F.
with decreasing temperature and, with materials
used in the experiment, appears to be of the order
of 2000 hr when the equipment is subjected to
1500°F liquids.

Durability of the equipment increases
  
  
  
    

UNCLASSIFIED
ORNL-LR-DWG €219

 
   
    
   
    
  
   
  

309 i

2 VALVE
‘5:;.} s 3
VALVE U-21 U= ?3

 

=l \-.:a'\q 0y

 
   

PRESSURE
TRANSMITTER -

  

VALVE U=
VALVE U-31 -2 "
;/". - -

 

VAL_VF U300

 

VALVE U-26 " -

e - o

S
7
©VALVE U-27 . l -
e

/// ! . / KEY:

somum FILL AND DUMP TANK NO. 6 — - _
/ l’ /’ £ 9 CROSS BARS ON LINES INDICATE WELDS.

// - T ALL SODIUM PIPE LINES NUMBERED IN
A ,/ 300 SERIES.

Fig. 34. lsometric Drawing of ARE Sodium System,

LE

 

 

 
 
 
  

 

M
;‘/ ‘ .
T ,/ ", FRANGIBLE <
- Yy DISK VALVE U-2 ——=y
PRESSURE Y ' e

TRANSMITTERS

5 HEAT
* EXCHANGER
\ NO.2

  

VALVE U-{
A

YALVE B-55—
.

VALVE B-56-
N

    
  
 
 

§- . DUMP TANK 7 g

   
    

RESERVE TANK NO. 4

UNCLASSIFIED
QRNL-LR-DWG 6217

_~FRANGIBLE .
DISK VALVE U-3

 
 
  
 

 
 
 

 

    

~-HOT FUEL

     

2 7 '35 CROSS BARS ON LINES INDICATE WELDS,
ALL FUEL PIPE LINES NUMBERED (N 100 SERIES,

Fig. 35. Isometric Drawing of ARE Fuel System.

ACKNOWLEDGMENT

The authors wish to express their appreciation to
their many associates and consultants, without
whose invaluable assistance and many contribu-
tions the work described could not have been
accomplished. The report embraces the engineering
which was accomplished in several sections of the
ARE but which, instances, was not
necessarily the direct responsibility of one of the

in many

38

authors. A project such as this crosses a wide
of professional disciplines and craft
activities, and it is with regret that we admit the

impossibility of naming each.

variety

The assistance of A. L. Southern, ORNL
Instrumentation and Controls Division, in preparing
the section ‘‘High-Temperature Instrumentation®

is particularly acknowledged.,

 
J:-J:-.n-J:-h.h-AwwwwwwwwwwmmwmmmwMMM—-—-—-—-—-—-—-—-—-—-
EE RN E S 0ENECHAONEODINGORON OO NG RO =0

0@ N LA WN

MAAONME-OEZTAICAATPENNPOPCOOMEOMEQNOIMMEZIAIMEIED P

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Lane

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Bresee

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Morris

Ring, Jr.

A. Cristy

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G. Bohlmann

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DempmmOommA4RXAO0O0M

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EH-P>O0-mMmE=x=nITOMMMO O

CMEM-OMIBIPPPEOCZTACOCTE=QR

ORNL.-2348

Metallurgy and Ceramics

TID-4500 (13th ed., Rev.)

February 15, 1958

E. Beall

. M. Reyling
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S. Bettis

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WAOAPHELS<QOQUIMER0LD

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39
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40

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