ORNL-1556.txt

3 wish somenns

 
   

 

ORNL-1556

This document consists of 14) poges.
Copy ?3 of 249 copies, Series A.

Contract No. We7405-eng-26

AIRCRAFT NUCLEAR PROPULSION PROJECT
QUARTERLY PROGRESS REPORT

for Period Ending June 10, 1953

R. C. Briant, Director
A, J. Miller, Assistant Director

Edited by:
W, B. Cottrell

DATE ISSUED

O AK RIDGE NATIONAL LABORATORY
Operated by
CARBIDE AND CARBON CHEMICALS COMP ANY
A Division of Union Carbide and Coarbon Corporation
Post Qffice Box P
Oak Ridge, Tennessee

MARTIN MARIET TA ENERGY §YS8TEMS LIBRARIES

TR

3 4456 0349535 3
 
 

       

    
 

 

   
   

  
  

ORNL.-1556

Progress

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93-94,

. Reactor Expe;:gmentuu
Engineering Le;bmry

   

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vi

 

Reports previously issued in this series are as follows:

ORNL-528
ORNL-629
ORNL.-768
ORNL -858
ORNL-919
ANP-60
ANP-65
ORNL-1154
ORNL-1170
ORNL.-1227
ORNL-1294
ORNL.-1375
ORNL-1439
ORNL-1515

Period Ending November 30, 1949
Period Ending February 28, 1950
Period Ending May 31, 1950
Period Ending August 31, 1950
Period Ending December 10, 1950
Period Ending March 10, 1951
Period Ending June 10, 1951
Period Ending September 10, 1951
Period Ending December 10, 1951
Period Ending March 10, 1952
Period Ending June 10, 1952
Period Ending September 10, 1952
Period Ending December 10, 1952
Period Ending March 10, 1953

 
 

FOREWORD ...........

CONTENTS

--------------------------------------------

PART |, REACTOR THEORY AND DESIGN

INTRODUCTION AND SUMMARY .. i i it i i i e e et ittt e e nans

1. CIRCULATING-FUEL AIRCRAFT REACTOR EXPERIMENT . ... .. it in.

Fluid Circuit .........
Reactor ., .,.........
Instrumentation . .. .....
Dff-Gas System .. ......
Reactor Control . .. .. ...
Electrical System .., ....
Fuel Loading Facility ...

Fire-Fighting Preparations

--------------------------------------------

--------------------------------------------

-------------------------------------------

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--------------------------------------------

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--------------------------------------------

............................................

2, EXPERIMENTAL REACTOR ENGINEERING .. .. ittt i i e i e ee e

Pumps for High-Temperature Liquids © . .o i i it ittt ittt i i e i ie et te s nnenn
Cenfrifugal pump with combination packed~frozen flucride seal ... . ... .. ... ... ...

ARE-size sump pump ...

--------------------------------------------

Frozen-sodium-sealed pump for ARE moderater-coolant circuit . . . .o v vt it it v s
ARE packed-frozen sealed fluoride pump . . . i it it it i e e e e e s
Allis-Chalmers conned-rofor pump . v v v v it vt s it ot vttt o s ot ssnnaes e te s
Frozen-lead-sealed fluoride pump . . o o ittt i et e e e e
Laboratory-size sump-type pump with gas seal ... i it ittt ittt i e e
Rotating Shaft ond Yalve Stem Seal Development . .o 0 i i ittt ii it n e

Graphite-packed seals ..

--------------------------------------------

Grooved-shaft packed-frozen seal fests . .. L i il i it i e e
Bronze wool and MoS, packed-frozen seal tests .. ... ... ..., et et

Frozen-lead seal tests . .
Voringseal .........
BeF2 seal tests , ... ..

Packing penetration tests

Instrumentation .., ......

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B ) BRI B el oend s bomd med e i) wwd cemd el mwemd el pmed mad  met
WKIMNMRO VOGNS NN e =d OO

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B

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Lo

vii
 

4. REACTOR PHYSICS ..o i i ittt ittt i it e et e saenannnaaneesas

Kinetics of a Circulating-Fuel Reactor ., . . . ittt ittt ittt eanerans
Behavior at long time after adisturbance .. . .. it ittt i e e e e e
Oscillations .. . it ittt i i ittt e ttettetensaaraeaaeaeaeee e
The first overswing following adisturbance ... ... ... . i it nn.
Statics of the Circulating-Fuel ARE Critical Experiment .. .. . . 00 it ittt it oo

5. REFLECTOR-MODERATED REACTOR CRITICAL EXPERIMENTS . ...............

Power Distribution . . .. ... ittt ittt i it it it ettt i ettt
Leakage Flux . oo i it it it it s i e s vttt sttt as s eeonennnseesasnnsenss
Control Rod Measurements .. ... i ittt ie ittt ennertonoseasenseesononas
Danger Coefficient Medasurements . . ... ..ot ittt et s e esaennoneansansensonssas
Correlation with Theory . ... . . ittt ittt it toteotnetnssntosesasssnnnsassss

PART il. MATERIALS RESEARCH

INTRODUCTION AND SUMMARY . .. .. i i ittt i i ittt

6, CHEMISTRY OF HIGH-TEMPERATURE LIQUIDS .. .. .. ... it ieennens

Fluoride Mixtures Containing UF , . ... .. o i i i i i
e A e U T
RbF-ZrF -UF o o e e e e
NaF-ZrF -BeF -UF , o i i s
Fluoride Mixtures Containing UF, ... .. ittt
Chloride Mixtures Containing UCL, .. ... .t it ittt e

KCl-Ucl,
Licl-ucl,
TICl-uCl,

------------------------------------------------------

-------------------------------------------------------

------------------------------------------------------

KC-LiC-UC e e i i e
Coolant Development . .. .. i it ittt it ittt et ittt eieonnsseetonessonanss

NGF-ZrF rBeF ) 4 vt e v et et s e et

LiF-ZrF4

-----------------------------------------------------

Differential Thermal Analysis . . v v i it it it ittt it it et e e e s s nnoasnanaenens
Production and Purification of Fluoride Mixtures .. . .. i i i i it ittt ittt ee e o e nnn
Removal of HF from fuel batches .. ... .. . . i i it it e it it et
Reduction of metal fluorides with Hy, ... .. o o i oo,
Pilot-scale fuel purification . . .. ... . .0 it iiie it iteetneeroeesnesnns
Fluoride production facilities . ... i ittt ettt it et nennonsnnensensas
Preparation of Complex Fluorides . .. ... ... i i ittt it
X-ray Diffraction Studies . ..o i ittt ittt ittt it et e e e e e e e

viii

NaF-UF,
NaF-ZrF

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7. CORROSION RESEARCH

Fluoride Corrosion of Inconel in Static and Seesow Tests ... ........ s
Effect of exposure time .. ....... e e e e C e e
Residual hydrogen Huoride . ... ........ e e e e e
Fluoride pretreatment . .. ... e e e e e s et e e
Structural metal fluoride additiens . ... ........ ... c e c e
Structural metal oxide additions . , . ... ... . . . oL, e ‘e
Chromium metal addition . ... ... ... i i i e e
Zirconium hydride addition .................. e r e ey
Carbon addition ... .. . ittt ittt n et s s ater et r e
Crevice CoMmOSION & .t it et i it s sa e snseessononronnnsnennennn
Inconel container pretreatment . .. ... ....... e e

Static Corrosion of Stellite by Fluorides e e e

Fluoride Corrosion of Inconel in Rotating Test .. ... .. .0 eon.. e o

Fluoride Corrosion in Inconel Thermal Convection Loops . ... ... .. ... ...

Effect of exposure time .. .........
Temperature dependence . .. ... ... .
Chromium and nickel fluoride additives .
Fuel purity . ........

Loopoxide films ., ... .,..........
Nonuranium bearing Huorides ... ....

Fluoride Corrosion of Nickel Thermal Convection Loops

------ - s » e 8 e
» u & s & @ * . . 2 v a w
< * 2 = ®F N 8 W A A B *T B & O n
----- 4 5 &8 2 v & w B &

. . “ 8 4 s » . a LI

4 + 8 B @ & & & & ¥ ¥ B s & 3 4 s

Fluoride Corrosion of Type 316 Stainless Steel Thermal Convection Loops . . . .

Liguid Metal Corrosion .. ..........

Mass transfer in liquidlead ... ... ....... e e e e .
BeO in sodiumand NaK . . . . ... .. .o oo, C e e e e e e
Coated beryllium in sodium . .. .. ot i i e e o
Carboloy in lead and sodium .. ... ... .. .. f e e s e

8. METALLURGY AND CERAMICS .....

Welding and Brazing Research ., .. ...
Welding of heat exchanger tube bundles

Brazing of air radiators .. ... . ... ...

Dilution and diffusion of brazing alloys .
Brazing Carboloy to stainless steel . ..

--------

B4 A sy s

High-temperature brazing alloy evaluation tests , ..
Creep and Stress-Rupture Tests of Structural Metals

Fabrication Research ... ...........
Extrusion of high-purity Inconel tubing .
Air oxidation of columbivm ., . ... ..
Control rods for the G-E reactor . ....

Hot-Pressed Pump Seals . . .
Fabrication of small compacts with Mo$

B4 a oW s 4o

« 8 8 P 2 e »

2 % & » 4 » v a

a2 %« 3 * & N s

Metallographic examination of compacts with MeS, .

--------

 

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Fabrication of face seal rings . . . .. .. it i i i it e

Surface sulfurization of molybdenum , .
High-Conductivity Metals for Radiators .
Chromium-clad QFHC copper ......

Chromium and nickel electroplate on copper
Inconel claddingon copper .. vttt ittt e
Type 310 stainless steel cladding on copper ... ...... ..

Type 310 stainless steel cladding on sil
Copperalloys . .............. .

--------------

oooooooooooooo

ver ------------------------ * 5 5 & ® s a s

Solid Fuel Formed in Spheres .o .ttt i it it ittt s et enietannonnonnnnss
Heating under vacuum . ..o i ittt e ittt i et et aann Cee s
Three-phase carbon GrC . v it i i it ittt et tn e st e onsenaonsans

Molten alloy passed through a small orifice ,....... e e Cere s e

Settling of particles through o molten salt

Inflammability of Sodium Alloys . .. . .. i ittt ittt ittt e s

ot amiCS o v v it i it v e v et o s o s s s e s oseesseesonsnessensons .

Development of cermets , ... ......

Effect of heating rate on the beryllium oxide blocks for the ARE

9. HEAT TRANSFER AND PHYSICAL PROPERTIES

------------

ooooooooooooooooooooooooooo

Thermal Conductivity . . i ittt i iiientte et ensennnnsesenans et e .

Density and Viscosity . ...........
Heat Capacities of Liquids ... ......

Surface Tensions of Fluorides . . . . o ittt ittt ittt ittt et i s ennnss e
Vapor Pressures of Fluorides .. . . ..o ittt it it ittt st taesronnnnnn
ON LooOPS . vttt ittt ittt e s e e e

Velocity Distributions in Thermal Convect

Forced-Convection Heat Transfer with NaF-KF-LiF

High-Temperature Reactor Coolant Studies

--------------

Turbulent Convection in Annull . . 0 Lttt it i it it it e sttt oe o enononoernsnens

Circulating-Fuel Heat Transfer , .. ...
Bifluid Heat Transfer Experiments ., ..

10, RADIATION DAMAGE .. . it it ittt ettt e e i e eaaaens

Irradiation of Fused Materials .. .....

fnpife Circulating Loops + . oo i i ittt i it it it it e et e st cessas et onaas Cene
Creep Under Jrradigtion . 0 i i ittt ittt ittt it is s s e e ennannonsnns v

Radiation Effects on Thermal Conductivity

11. ANALYTICAL STUDIES OF REACTOR
Chemical Analyses of Fluorides and Their

Determination of oxygen in metallic oxides with bromine trifluoride

Volumetric determination of zirconium
Determination of uranium trifluoride and
Colorimetric determination of zirconium

---------------------------------

MATERIALS ...... ... vt coas

Contaminants . . v v v v vt o v o enas e ee e aees

--------------

-----------------------------------

metallic zirconium ... .....

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““Tiron' as a reagent for the determination of uroniom . .. ..........
Petrographic Examination of Fluorides .. ... .. 0oy
Summary of Service Chemical Analyses ... .....................

12. FLUORIDE FUEL REPROCESSING ... ci v v it it neenntanonsns
Dissolution of ARE Fuel .. . . it ii i i it ittt et s et neraesanns

Solvent Extraction . v v v v it ot v o vt oo et o ranonsnsasanss o
Corrosion in Dissolution of ARE Fuel ... .. e e h e
Plant Processing .+ . v v v v v ot v vt ot oot a asansnenesansoscnssos

INTRODUCTION AND SUMMARY .. ... ...t iiinnienenns

13. BULK SHIELDING FACILITY .. .. ittt et ii i e i e e

Neutron Spectra for the Divided Shield . ... ....... .. ... ot
Energy per Fission and Power of the Bulk Shielding Reactor ... .......
Gamma-Ray Air-Scattering Caleulations . ... .. oo oo v e
Differential results . ... .. ... ... ... e e e e
Integral results ., . . . i e e it e e e i e

M., LIDTANK FACILITY ... it ittt i i s et n s e asas .

Reflector-Moderated Reactor Shield Tests . ... ........ . .00
Shield configurations . . . ... i v it ittt it n i e s e
Gamma-ray attenuation data . . o . v oo it b i i i e s e

Neutron attenuation data . .. ... v v v s ek e r s s n e e ma e e n e e as e

Removal Cross Sections . . .. 0o i i i n it ittt i it i
Facility Modification .. . it vt e e i ettt i as v annosaons

15. TOWER SHIELDING FACILITY . ...van i iionn,

PART IV. APPENDIX
16. LIST OF REPORTS ISSUED DURING THE QUARTER . ...........

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xi
ANP PROJECT QUARTERLY PROGRESS REPORTY

FOREWORD

This quarterly progress report of the Aircraft Nuclear Propulsion Project at ORNL records the technical
progress of the research on the circulating-fuel reactor and all other ANP research at the Laboratory
under its Contract W-7405-eng-26. The report is divided into three major parts: |. Reactor Theory and
Design, [l, Materials Research, and lll. Shielding Research. Each part has o separate introduction
and summary. _

The ANP Project is comprised of about three hundred technical and scientific personnel engaged in
many phases of research directed toward the nuclear propulsion of aircroft, A considerable portion of
this research is performed in support of the work of other organizations porticipating in the national ANP
effort. However, the bulk of the ANP research at ORNL is directed toward the development of a circu-
lating-fuel type of reactor.

The nucleus of the effort on circulating-fuel reactors is now centered vpon the Aircraft Reactor Experi-
ment -~ a 3-megawatt, high-temperature prototype of a circulating-fuel reactor for the propulsion of air-
craft. This reactor experiment is now being assembled; its current status is summarized in section 1,
However, much supporting research on materials ond problems pecuhar to the ARE will be found in other
sections of Parts | and |l of this report,

The ANP research, in addition to that for the Aircraft Reactor Experiment, falls into three generdl
categories: (1) studies of aircraft-size circulating-fuel reactors, (2) material problems associated with
advanced reactor designs, and {3) studies of shields for nuclear aircraft. These three phases of research
are covered in Parts |, 11, and 1], respectively, of this report.
 
INTRODUCTION AND SUMMARY

The Aircraft Reactor Experiment (sec. 1) is now
being rapidly assembled; most of the component
parts have been received ond installed in the ARE
Building. There were no significant design modifi-
cations during the quarter, but a supplementary off-
gas system to handle contaminated pit atmosphere
is contemplated. Installation of the fuel-circuit
piping is nearly complete, and the detailed testing
and cleaning procedures for both the fuel and the
sodium circuits are being established. Integral
parts of the system, including the water system,
the main off-gas system, and the rod cooling sys-
tem, have been completed. The fill and flush tanks
are installed, wired, and insulated. The reactor
core has been completely assembled, welded, and
is ready to be shipped to the ARE Building. All
instrument indicators have been installed, and the
interconnecting lines to the control room are es-
sentially complete, Routine and emergency pro-
cedures for handling sodium and the fluorides have
been established.

The development of a satisfactory ARE pump
seal has required major research effort (sec. 2).
Conventional centrifugal pumps have the required
pumping characteristics, but the aftainment of a
reliable seal which will retain the high-temperature
fluorides with leskage rates of less than 35 cm?
per day continues to be an outstanding development
problem for the ARE. This seal problem is being
investigated in several pump loops, as well as in
numerous seal testers, with the result that several
seal-types now either appear to be potentially
capable of fulfilling the leakage requirement or
have actually dome so. The frozen-lead seals for
fluorides have operated with vertical shafts with
negligible leakage for extended periods of time;
gas-sealed pumps have also been satisfactorily
operated. At present, however, the ARE plumbing
arrangement requires a horizontal shaft pump for
which the following packed-frozen seals appear
promising: {1) alternate layers of copper (or bronze)
wool and a grophite-MoS, mixture, (2) triangular

rings of Graphitar bearing on sintered Cu + MoS,,
and (3) foyers of NaBeF, packing. I has been
demonstrated that oxygen has a deleterious effect
on all seals, so they are now blanketed with
helium. The frozen-sodium seal pumps for the
ARE reflector-coolant (sedium) circuit have been
fabricated.

Recent studies of an aircraft-size reflector-mod-
erated- reactor system have included analysis of
the control of a 200-megawatt power plant and some
parametric studies of fuel volume in the heat ex-
changer and in the reactor for vorious core diame-
ters and power levels (sec. 3), From the 200-
megawatt power plant control study, it appears
that the reactor temperature will not tend to rise
ot a rate greater than 100°F/sec, even if cooling
of the engine radiators is sfopped instantaneously.
Insuch a case, the negative temperature coefficient
will adequately control the reactor,

The kinetics of circulating-fuel reactors and the
interpretation of the ARE critical experiment com-
prise the recent reactor physics studies (sec. 4).
Introduction into the core of 1% A%, if not compen-
sated by control rods, would cause a 100°C rise
in the reactor equilibrium temperature. On the
other hand, power oscillations damp out quickly
and no serious overswing in power or in tempera-
ture is expected. The ARE critical experiment
dato are being extrapolated to provide additional
information on the statics of the ARE.

The critical assembly of the reflector-moderated
reactor has provided measurements of power distri-
bution, leakage flux, and control rod effective-
ness, as well as danger coefficient measurements
of materials in both the fuel and reflector regions
{sec. 5). The end leakage of fast neutrons is six
times the side leakage when the end fuel ducts
are shielded with boral sheets, The boral reduces
the ke” by 2% and correspondingly modifies the
power production in its vicinity. Extrapolation of
the uranium required in the experiment to an air-
craft reactor with Inconel structure indicates that
the critical mass of the latter is about 30 pounds.
ANF PRQJECT QUARTERLY PROGRESS REPORT

1. CIRCULATING-FUEL AIRCRAFT REACTOR EXPERIMENT
E. S. Bettis, ANP Division

Duwring this quarter, effort on the ARE Project
has been concentrated on the installation of equip-
ment in Building 7503, No significant design
modifications or alterations were made, and the
work has progressed steadily with no serious inter-
ferences. The major emphasis was on the installa-
tion of the fuel-circvit piping, which is now neorly
completed.

The checking out of auxiliary systems was
started. The water-cooling system has been com-
pletely checked and put in stand-by condition for
actual operation, All water lines, water heat ex-
changers, valves, tanks, and lines were flushed,
the flows were checked, and the system was filled
with a rust inhibitor to keep it in readiness for use,
A check has been started on the off-gos system,
but this has not yet been completed.

All major components, except drive motors for
the pump, have been received. All volves have
been received, except two of the 2-in. sodium
valves,

A large amount of effort went into preparing the
testing and preliminary operating procedures. The
first draft of these procedures has been completed
and is being carefully studied for revision, As was
expected, some bosic changes in the original con-
cept must be made, ond during the next few weeks
all possible effort will be made to complete these
procedures.

Much research that bears directly on the experi-
ment is contained in other sections of this report.
Of particular significance are the following: pump
seals {sec. 2), kinetics of circulating-fue! reactors
(sec, 4), composition and production of the fluoride
fuel (sec. 6), corrosion of the fluoride-Inconel
system (sec.7), high-temperature strength of Incone!
in fluorides (sec. 8), physical properties of the fuel
(sec. 9), effect of radiation on creep, thermal con-
ductivity, and comosion of Inconel (sec. 10), and
reprocessing of the fuel (sec. 12).

FLUID CIRCUUT
G. A, Cristy

Engineering and Maintenance Division

The fuel piping system is almost completed and
initial testing is expected to begin the middle of

June. This work has proceeded satisfactorily and
according fo schedule, Approximately 5% of the
welds have foiled o pass inspection and have had
to be reworked. This represents less trouble than
waos anticipated, since the standard for an accept-
able weld is high.

The hot fuel dump tank is completely insulated
and ready for use. The fill and flush tanks have
been installed, and installation of the heaters and
the insulation is almost completed. Figure 1.1
shows the status of the tank pit as of the middle
of the quarter,

Installation of the piping for the sodium circuit
will not be started until the fuel piping is com-
pleted. The sodium-to-helium heat exchangers are
now in place and the sodium pipe lines can be
installed as soon as the fuel piping is completed,
The final installation of the sodium piping must
await installation of the reactor; according to the
present schedule, the reactor will be installed
during the latter port of June.

Pumps for the sodium circuit have been completed
and should be ready for installation about the middle
of June.
seal,

These pumps employ a frozen-sodium -
which has been thoroughly tested by the
Experimental Engineering Group. No decision has
yet been made as to the type of seal to be used
for the fluoride pump (cf., sec. 2),

Strain gages have been installed on the fue!
piping system and the calculated prestress has
been verified by measurements made with these
gages. Additional checks will be made as the
system is warmed up. A complete report on the
stress analysis of the system is being compiled.

REACTOR

The reactor, including the beryllium oxide block,
the fuel tubes, and the pressure shell, is completed,
except for the final closure welding (Fig. 1.2).
The reactor core will be transferred from the Y-12
shops to Building 7503 about the latter part of
June, but it will not be installed in the system
until the prechecking of the fuel circuit has been
completed. Prior to installation in the system, the
reactor will be helium checked at a temperature of
1200°F to maoke sure that the fuel circuit is tight
with respect to the moderator volume,
Wi @ o
HOT FUEL

PERIOD ENDING JUNE 10, 1953

- j
PHOTO 1412107

"DUMP TANK

 

Fig. 1.1. ARE Tank Pit.

INSTRUMENTATION
R. G. Affel, ANP Division

All instrumentation has now been completed and
most of the pneumatic lines have been checked.
The leak detection system has been further simpli-
fied, but no other alterations are contemplated.
Liquid-level and flow instruments have been com-
pleted and tested insofar as is possible without
actual operation. Interconnection of the control
room with instrument panels, relay racks, amplifier
boards, and control actuators is 98% complete.
A very complete labeling system, with a permanent
wall-mounted check print for each control point,
has been effected.
checked as rapidly as an integral part is completed,
and by the middle of June, the control room will
have been completed and checked.

The control wiring is being

The annunciator points have been checked and
tested insofar as the preliminary status of the
sending points allows. The same condition exists
with the connections to the graphic panel.

No significant problem remains, so far as the
instrumentation of the ARE is concerned.

OFF-GAS SYSTEM

The main off-gas system as originally designed
has been installed and partly checked. A con-
siderable amount of thought and calculation has
been given to a secondary system for handling pit
gas in the event of pit atmosphere contamination.
The evaluation of this system has not been com-
pleted, but it is fairly certain that it will contain
a charcoal adsorber maintained at liquid-nitrogen
This adsorber will probably consist
of two tanks located outside the building and

temperature.
ANP PROJECT QUARTERLY PROGRESS REPORT

PRESSURE SHELL S

FISSION
CHAMBER
THIMBLE

REFLECTOR &

COOLANT TUE!E‘% y ;
b BeO MODERATOR BLOCKS

B A
MODERATOR CAN

FUEL TUBE INLETS PHOTO {1208

JFISSION CHAMBER
THIMBLE |

FISSION
CHAMBER THIMBLE .

BRREFLECTOR CODLANT TUBES

 

Fig. 1.2. ARE Reactor Core. (a) BeO moderator blocks partly stacked. (b) BeO moderator blocks in

place.
arranged to permit alternate adsorption and re-
generation cycles.

This auxiliary off-gas system is designed pri-
marily to handle contaminated pit atmosphere, but
it can also be used for system-gas disposal in the
event of trouble with the primary off-gas system.

REACTOR CONTROL

The active fission chamber has passed several
hundred hours at 800°F with no increase in back-
ground or deterioration of the octive cylinder. This
chamber therefore appears to be satisfactory for
ARE use., The control console and its associated
servo, safety rod actuator mechanisms, efc. are
completely installed and almost entirely checked.

The control rod ‘‘igloo’’ assembly is nearly
complete. It was necessary to rewind the safety
rod drive motors in order to make them reversible
without limitation of the number of reversals per~
mitted in any unit of time. All safety rod drive,
control rod servo, fission chamber, and source
drive mechanisms have been assembled.

A dummy control console for checking the ac-
tuator assembly has been built, The control actu.
ator assembly is 75% complete.

ELECTRICAL SYSTEM

The electrical work in the building is on schedule.
All  major electrical circuits have been wired;
however, the fluid circuit pumps have been changed
from 5 to 15 hp, and therefore new circuits will be
needed for these motors. |Installation of annulus
heaters is proceeding wherever completion of the
lines permit., The motor generator sets have been
moved to a shed outside the building.

FUEL LOADING FACILITY
G. J. Nessle, Materials Chemistry Division

Preliminary work has been started to provide
adequate facilities for loading the processed fluo-
rides into the ARE. A platform is to be built
inside the storoge tank pit for installation of the
loading furnace to be used in transferring the
processed fluorides into the storage tanks.

PERIOD ENDING JUNE 10, 1953

FIRE-FIGHTING PREPARATIONS
D. R. Ward, ANP Division

In view of the use of hazardous materials and the
high-temperature operations connected with the
ARE, some fire-fighting preparatims have neces-
sarily been made. The materials and techniques
to be employed in case of fire are based upon the
considerable experience acquired during the past
two years by the Experimental Reactor Engineering
Group.

The test pits will be open while the fuel system
is being cleoned and leak tested so that the usuadl
direct fire-fighting methods can be employed if
NaK is used and a fire occurs. Graphite powder
will be the preferred extinguishing material, with
Ansul extinguishers filled with Metal-X on hand for
additional safety. Several buckets of graphite will
be strategically located in each of the three pits.
Standard safety items, such as proper gas masks,
fire-fighting suits, etc., will be on hand.

During the reactor operation period, the test pits
will be closed and access for fighting fires will
be impossible. During this time, any fires must
be combated remotely. Reservoirs of graphite
powder will be located over the most vulnerable
parts of the system and will be fitted with dumpers
which may be operated from the control room, It
is expected that the (approximately) 80% helium
plus 20% qir atmosphere in the pits during reactor
operation will help to retard fires, but combustion
tests have not been conducted in such atmospheres.

Metal catch-trays containing graphite powder
will be placed in wulnerable locations, such as
beneath pump seals and heat exchangers, to catch
leakage both during leck testing and during reactor
operation.

All temporary flammable material, such as wooden
stairways, will be removed from the pits before
high-temperature testing is started. Exposed pit
wiring will be wrapped with asbestos tape ond
coated with flame-retardant paint. A flexible
exhaust hose will be in readiness to reach any
fires to withdraw noxious fumes from their points
of generation.
ANP PROJECT QUARTERLY PROGRESS REPORT

2. EXPERIMENTAL REACTOR ENGINEERING
H. W. Savage, ANP Division

During this quarter, the developmental effort of
the Experimental Reactor Engineering Group has
been devoted mainly to determining final specifi-
cations for the pumps and the pump seals to be
used in the fuel and moderator-coolant circuits in
the ARE. Numerous seal testers and several pump
loops are being used concurrently to obtain the
necessary information.

Fabrication of three 100-gpm frozen-sodium-sealed
horizontal-shaft centrifugal pumps for use in the
moderator-coolant circuit of the ARE has been

completed. The operational parameters of these
pumps are being determined in detail. A 100-gpm
double-cell a-c electromagnetic pump has been

delivered to the ARE building for use in the pro-
posed NaK precleaning operation. A frozen-sodium
seal tested on a 21/2-in. (ARE-size) shaft has given
in excess of 1,000 hr of continuous, reliable, and
predictable performance.  Successful stop-start
procedures have been established. The seal is
formed by freezing sodium in the seal gland. Also,
since the seal is the coldest part of the system,
sodium oxide from the circulating sodium precipi-
tates and accumulates in the seal.

An extensive program is under way to determine
within the next few months whether a horizontal
shaft seal can be proved sufficiently reliable for
operation with fluoride fuel in the ARE. The three
types of packings tested to date that show promise
of sufficiently low leakage and reliable, low-power
consumption are (1) alternate layers of copper or
bronze wool and a gmphii‘e-MoS2 mixture, (2) an
arrangement of rings of triangular cross section in
which clternate rings are Graphitar and sintered
copper impregnated with MoSz, and (3) a series of
annular cells packed with a material of high BeF,
content, such as NaBeF,. Materials that are high
in Bef,_ content are glasses, and the temperatures
at which they soften vary over a wide range. To
provide o steep thermal gradient over the seal
length, one end of the seal is heated and the other
end is liquid cooled.

Two alternate types of pump systems for fluo-
rides are being constructed for tests that are to
begin in June. One system will have a frozen-lead-
sealed pump with o moiten lead—molten fuel inter-
face above the seal. In a test run, this type of seal

10

operated successfully for 500 hours. The other
system will have o gas-sealed sump pump, the
sump of which would replace the ARE expansion
tank,

An Allis-Chalmers 5-gpm canned-rotor pump is
being set up for test, and a small 2-gpm 100-psi-
head air-driven pump for inpile operation is being
modified for seal testing.

The use of thin-walled bellows (0.005 in.) has
been proved entirely practicable by tests operated
in excess of 2500 hr, with many cycles, provided
the temperature of the contacting fluoride is held
below 1100°F,

It has been demonstrated that an external oxygen
atmosphere is deleterious to the successful opera-
tion of frozen-sodium, frozen-lead, and frozen-fluoride
seals. Consequently, all such seals will be ex-
ternally blanketed with inert gas. Parameters of
operation of inert-gas-blanketed seals are calculable
on the basis of coolant temperature, liquid film
thickness, seal length, and the pressure differen-
tial across the seal.

PUMPS FOR HIGH-TEMPERATURE LIQUIDS
W. B. McDonald A. G. Grindell

W. G. Cobb G. D. Whitman
W. R. Huntley A. L. Southern
J. M. Trummel P. W. Taylor
ANP Division
Centrifugal Pump with Combination Packed-

Frozen Fluoride Seal. Three tests were conducted
in which the 50-gpm centrifugal pump(!? was used
to circulate the fluoride NaF-ZrF,-UF, (50-46-4
mole %). The operating conditions for these tests
were: pump speed, 900 to 1400 rpm; pump suction
pressure, 4 to 6 psi; pump discharge pressure,
30 to 60 psi; fluid flow, 10 to 30 gpm; fluid tem-
perature, 1200°F. The shaft of the pump was con-
structed of 2'/2-in.-dia stainless steel coated with
Stellite No., 6; the fuel leakage rate at the post
seal was 40 to 150 g per 24 hr; and the seal L/D
ratio was ¥.

In the first of these tests, the gland was packed
with Dixon’s ‘‘Microfyne’’ flake graphite powder
retained by close-fitting APC graphite rings.

 

{Dy, B. McDonald et al., ANP Quar. Prog. Rep. Mar.
10, 1953, ORNL-1515, p, 18.
During this 340-hr test, the seal leckage was low
and performance was generally satisfactory., The
gradual loss of graphite from the seal caused the
termination of the test,

For the second test, the seal was packed with
braided copper sheathing filled with MoS, powder,
The power required to rotate the shaft was quite
low, but operation was characterized by consider-
able leakage of MoS, powder from the seal. After
260 hr of operation, sufficient MoS, powder had
been lost to cause fuel leakage through the seal.
Post-run examination of the seal showed the copper
sheathing to be generally worn at the shaft surface.
Stop-start tests were difficult to make during this
run, and they always involved the danger of a
massive leak, because the temperatures in the seal
could not be measured accurately or readily con-
trolled.

For the third test, the packing gland was filled
with strands of copper rope impregnated with Mo3,.
Smooth operation was experienced in the early
stages of this run; however, power fluctuations
became greater, ond more power was required to
drive the shaft as the test progressed. The test
was terminated after 385 hr because of o gross
sea! Jeak that resulted from frictional overheating
of the seal.

ARE-Size Sump Pump. All parts for the high-
temperature model of the ARE-size sump pump(z)
have been received, and the pump is being as-
sembled in a test loop. A sectional view of this
pump is shown in Fig. 2.1. The sump tank of this
pump is designed to serve as one of the ARE surge
tarnks, and low-temperature tests show that priming,
vortexing, and fluid degassing will not be problems.
The gas seal for this pump consists of a Graphitar
ring rotating against a stationary hardened-tool-
steel surface, The gas-seal wear faces will be
maintained at a low temperature during operation
by circulating coolant in three different regions
at or near the seal. Hydraulic tests of a mockup
of this pump with water, water and glycerine, and
tetrabromoethane indicated that the design provides
for stable pump operation and adequate degassing.
Tests planned for this pump during the month of
June will determine the operating characteristics
at ARE tempercture and flow conditions.

 

(Dipid., p. 20.

PERIQOD ENDING JUNE 10, 1953

Frozen-Sodium-Sealed Pump for ARE Moderator-
Coolant Circuit. The ARE mode! FP pump(®) with
a frozen-sodium seal has pumped sodium for more
than 1000 hours. Operation has been at near the
ARE moderator-coolant pump design conditions of
100 gpm at 20-psi developed head with sodium
temperatures of 1100 to 1250°F., The pump suction
pressure has been varied from 10 tc 50 psi to get
data on seal leakage. Six pump shutdowns have
been made to correct loop difficulties or to make
loop changes., Shutdowns and startups have been
made without difficulty, although considerable
time (1 to 2 hr) and care are required to effect g
safe startup, For a safe startup, it is necessary
to warm the frozen seal.

The most critical element of this pump, as in
most other high-temperature liquid pumps, is the
shaft seal. The sodium freezing gland for this
seal is 513/16 in, long on a 21/2-in.-dia shaft with
a 0.030-in, radial cleorance in which the sodium
is frozen. Earlier difficulties with seal seizure
ond roughness have been eliminated by mointaining
a helium gos blanket between the cold end of the
seal ond the atmosphere, Seal operation has been
sufficiently reliable and reproducible to permit
taking quantitative data on seal power and leakage
while varying speed, pressure, and seal coolant
temperature.

An analysis of seal data indicated that viscous
liquid film theory(*) can be used to at least partly
describe seal characteristics, in particular, the
scul power requirements and the seal leckage rate.
In the viscous film model, the equilibrium thickness
of the liquid sodium film at the melting point is
determined by the rate at which the heat energy of
viscous friction is removed from the liquid film.
In turn, the rate of addition of heat energy is the
mechanical power supplied te the sec!, and this
must be equal to heot removed from the seal by the
coolant. The rate of leakage by viscous flow is
related to the film thickness and to the pressure
differential.

The viscous film theory can be used to obtain
significant dimensionless numbers which may be
used to describe seal characteristics. Two such
numbers are the power number, Nps and the leak

 

(pid., p. 21.

)H. Rouse, Elementary Mechanics of Fluids, Chap. 6,
Wiley, New York, 1946,

11
ANP PROJECT QUARTERLY PROGRESS REPORT

    
 
 
       
 

UNCLASSIFIED
DWG. ESK 160234

 

-—-GAS SEAL

 
  

LABYRINTH

SEAL

EVEL
INDICATION

PUMP SUMP AND
SYSTEM SURGE

 

DISCHARGE

SUCTION

N

Fig. 2.1. ARE-Size Sump Type of Pump (Centrifugal Pump Model DAB Assembly).

12
number, NQ, which are defined in the following:

N P
P AT UrL
and
ouU3 (AT) L
e T T .. !
AP +7 208
where
P = power,

AT = temperature difference between the melting
point of sodium and the seal coolant tem-

perature,

U = over-all coefficient of heat transfer from
liquid film to coolant per unit area of film,

r = shaft radius, '

L. = seal length,

O = volume leakage rate,

AP = pressure difference across seal,
i = viscosity of sodium at its melting point,

n = speed of shaft,
According to viscous film theory, the numbers
N, and N, should be constants. The experimental
data obtained by varying AP and AT are shown in

PERIOD ENDING JUNE 10, 1953

Figs. 2.2 and 2.3, The slopes of the lines agree
fairly well with theory. The scatter may be due to
error in instrumentation ond failure to observe the
presence of pertinent factors, such as shaft surface
roughness and contamination of seal sodium, which
were included in the analysis. Shaft surface rough-
ness may explain the fact that the absclute value
of the leakage number (N ) determined experi-
mentally is several times larger than the theoretical
value,

The presence of oxygen around the frozen-sodium
seal may account for scme of the discrepancy
When
the cold end of the seal was open to air, operation
deteriorated over a period of three or four days and
usually ended in sudden shaft seizure., That this
was due to oxide accumulation in the seal oppears
obvious, since the condition was alleviated by
addition of the helium gas blanket. However, the
seal also accumulates oxygen from within the
system, since it contains the coldest sodium of the
system. Seal roughness and relatively high leakage
rates have been observed following periods of no
flow through the bypass filter in the loop. The

between theoretical and experimental data.

 

 

1O e [T peeemeees T

|

 

 

SEAL POWER (hp)

 

 

 

 

 

LINE DRAWN FOR N, = ——
Z ....................................... |

 
  
 

120

DWG. 19895
2Yp-in.- dia SHAFT, 5'%g-in.-LONG COOLING GLAND.
' WATER COOLANT WITH ABOUT S-gpm FLOW
PUMP FLUID AT ABOUT t2Q0°F AT MOST POINTS |
—" SPEED AT MOST POINTS, 1700 rpm ST

|
C
= CONSTANT
o (ATYUrL

 

 

 

 

 

 

180 200 220

140 160

MEAN GCOOLANT TEMPERATURE (°F}

Fig. 2.2. Seal Power vs, Seal Coolont Tempercture for Frozen Sodium Secled Pumps,

13
ANP PROJECT QUARTERLY PROGRESS REPORT

Ay

DWG. 19896

2Ys-in.-dic SHAFT, 1700rpm, 5'%g-in-LONG GOOLING GLAND.
FP PUMP WITH FROZEN SODIUM SEAL
LINES DRAWN FOR LEAK NUMBER

3 3
N, = QUIATYL CONSTANT

(AP r P a®

WATER COOLANT WITH FLOW OF ABOUT 5gpm

PUMP FLUID AT ABOUT 4200°F

1000
SEAL PRESSURE DIFFERENCE, AR
50 psi
5 25 psi
10 psi
= 5 psi
o
9 2
~
M
&
©
~ 100
L
o
<1
&
= 5
-
I
<

 

10 20 50 100 200

SODIUM MELTING POINT MINUS MEAN COOLANT TEMPERATURE (°F)

Fig. 2.3. Seal Leckage vs. Seal Cooling Temperature Difference for Frozen Sodium Sealed Pump.

14
deterioration of operation resulting from this condi-
tion is best alleviated by mointaining clean sodium
in the system aond, if necessary, by infrequent
clean-outs of the sodium in the seal.

ARE Pocked-Frozen Sealed Fluoride Pump. An
ARE model pump®® circulating the fluoride NoF-
ZrF ,-UF, (50-46-4 mole %) with a packed-frozen
seal was operated for o period of 49 hours, Shaft
seizure occurred, and the Carboloy sleeve under the
seal was found to be cracked. The seal packing
consisted of strands of copper wire rope which were
socked in on oil suspension of MeS, for a period of
24 hr before installation, During the test, the pump
operated at 1250 rpm with a flow of 45 gpm and
produced o head of 55 psi. Power input to the pump
drive motor was 5.5 kw, with an estimated 3-hp
input to the pump shaft and 0.75 hp absorbed in the
seal. Fluoride chip leakage from the frozen end of
the seal was checked for each 24-hy period of
operation, ond measurements of 133 and 132 g were
obtained,

The seal configuration is shown in Fig. 2.4,
Heat was supplied to the seal by electrical tubular
heaters on the outside of the shaft housing and a
cartridge heater inside the shaft, as shown. Cool-
ing of the seal area was achieved by introducing
compressed air between the anchor flange and the
packing gland Hange and allowing the air to pass
out through multiple holes around the nose of the
gland slightly oblique to the shaft axis. This
cooling was supplemented by two small centrifugal
blowers directed against the rear of the packing
gland and seal.

The pump and the test loop of ll/é-in.-lPS pipe
were constructed of 316 stainless steel. Flow was
measured with a water-calibrated venturi, Pres-
sures were measured with Moore Nullmatic trans-
mitters, which have air balancing the loop pressure
across a bellows, The transmitters contained
triple-ply Inconel bellows and were operoted in a
downward position so that the transmitters were
flooded and free surfaces were eliminated,

When pump operation begon, the frozen seal ap-
parently established itself chead of the packing
area, but as the test progressed, the frozen zone
moved rearward until, after approximately 12 hr,
the seal existed in the annulus between the shaft
and the packing gland, However, after an 18-kw
power surge, a small amount of heat was ploced on
the packing and the shaft housing next to the im-
peller, and no other power surges occurred until

PERIOD ENDING JUNE 10, 1953

the shaft seized. At the time the shaft seizure
occurred, all heat had again been off the seal for
30 minutes.

The Carboloy sleeve was found to have a crack
approximately 3/&4 in. wide that originated in the
corner of a driving notch at the cool end of the
sleeve. Subsequent heating and cooling tests on
the sleeve-shaft assembly alone indicated thot a
Carboloy sleeve cannot withstand any apprecicble
axial temperature gradient, A stainless steel
sleeve with Colmonoy hard surfacing will be tested
in the next pump assembly, since the Carboloy
sleeve does not appear satisfactory for this appli-
cation,

Allis-Chalmers Conned-Rotor Pump. A 5-gpm
conned-rotor pump built by Allis-Chalmers is being
assembled in a test loop and will be tested with
NoK af temperctures approaching 1500°F., This
pump has a hydrostatic bearing, pressurized by o
smoll pump impeller at the rear of the pump, which
bypasses some of the system fluid through the
bearing. The motor windings are made of small-
diometer copper tubing with fiberglas insulation.
The windings are cooled by circulating a coolant
through the tubes. The pump is completely in-
strumented so that temperatures at all critical points
can be determined during operation,

Frozen-Lead-Secled Fluoride Pump. Sea! tests
have demonstroted that a frozen-lead seal has
smoother operating characteristics than a frozen-
fluoride seal and that molten lead and molten
flucrides do not intermix when in intimote contact
with each other. Consequently, o centrifuga!l pump
of ARE size has been redesigned as a frozen-lead-
sealed pump for fluorides. In this redesign, the
pump is mounted in o vertical position with the
seal on the hottom side of the pump. A chamber
equipped with immersion heaters located immedi-
ately below the pump impeller housing contains
both molten lead ond molten fluorides at 1000°F,
Because of the difference in the densities of the
two mixtures, the fluorides float on the lead, and
a sharp fluoride—lead interface is maintained. A
frozen-lead seal is accomplished in the bottom of
this chamber in a liquid-cooled seal annulus that
is approximately 1 in. long ond has approximately
0.030-in, radial clearance around the 21’2-in.-diu
shaft, This redesign permits the modification of
the frozen-fluoride-sealed pump to a lead-sealed
pump that utilizes oll the parts of the original
pump, except the seal.

15
91

UNCLASSIFIED
DWG CSK-15993A

 

 

2¥%s
3 i 3 ——
NOTE: ALL DIMENSIONS ARE IN INCHES far==1e }//4 oD TUBE
45°
{ !
1 : ~COOLING DUCT
{
%4DIA SCREW 3 EACH SIDE Y32 THICK SHEET ?fi

. 7 TUBULLAR HEATERS
J-11 STUD AND NUTS—__ ¢ ==

 

 

78°TYP&CAL§

boss

 

3ip IPS STAINLESS STEEL PIF’E/7 ‘ / Y NE

[ P—% 3 -2'/32

 

 

 

 

 

SEAL GLAND

 

 

 

 

 

 

 

 

 

 

 

   
   

 

 

 

 

 

 

 

o
| T
| L
L o403
2 Scoows (I | S - §§E ‘:_fl_{"e CARTRIDGE HEATER
Ry ""’""‘\3&;%1'1"/8 o !
' REEEY % 5
’ ! ! N bbb —_ |
! : }\\‘\\\X N M
1 ; N
N i T T |
o Tt N SESESES TINESESES
| (it .f.{‘,“.‘"‘.'-‘w e
by o
I|II'||"|’ e
i , ;
L AN S THERMOCOUPLE HOLE

LEAK CATCHER ——=!

 

 

 

e
‘L% °

e e e e

 

 

 

-

TWIST LOCK JO:NT\E !

4 a4

TUBE DOWN TO
CATCH POT

 

 

Fig. 2.4. Seal Gland for ARE Pump with Packed Frozen 3eal.

 

130d 3 SSF¥908d AT4ILIVND LIIF0¥d INY
L.aboratory-Size Sump-Type Pump with Gas Seal.
It was reported previously“) that the laboratory-
size sump-type pump operated 965 hr with a
stuffing-box seal ond 2300 hr, after redesign, with
a rotary face seal of Graphitar Ne. 30 running on
hardened tool steel, An additional 1000 hr of
operation has been accumulated on another pump of
the some design; however, Morganite MYIF, a
silver-impregnated graphite with self-lubricating
properties, was substituted for Grophitar No, 30
in the seal. Speeds of up to 3600 rpm were used
with a flow rate of 5 gpm that produced a 55-psi
head to pump the fluoride NaF-ZrF ,-UF, (50-46-4
mole %). Suction pressure varied from 10 to 30
psi, and the fluid femperature ranged from 1200 to
1500°F, There was regular dropwise oil lubrica-
tion at a rate of 6 to 8 drops per 25 hours ot first,
but later, lubrication was used only when power
fluctuations occurred in the drive-motor input,

Gas leakage during the test was negligible up to
24 hr prior to termination of the test, when a marked
increase was noted. The graphite seal piece was
found to be pitted and scored. The stationary tool
steel foce had a work depression. Operation of the
pump will be continued with continuous oil-mist
lubrication on the seal exterior.

ROTATING SHAFT AND VALVE STEM
SEAL DEVELOPMENT

W. B. McDonald W. R. Huntley

W. C. Tunnell L.. A, Mann

R. N. Mason D. R, Ward

P. G. Smith J. M. Cisar
ANP Division

Graphite~-Packed Seals. A seal with layers of
natural-flake grophite and MDS2 retained by Graphitar
rings and machine turnings of an Ag-Mo$, compact
was operated to seal against the fluoride NaF.
ZrF -UF, (50-46-4 mole %) at 1300°F and 10 psi
for 53 hr before termination of the test. Thermeo-
couples were attached within the rotating shaft,
Cperation was smooth; the power requirements
were low; and consequently, the temperatures were
low. It was possible to make 2-min stops without
it being necessary to add heat, The indicated seal
temperature was above the melting point of the
fuel (515°C). However, there was no leakage when
the shaft stopped and froze, which indicated that
some fluoride was leaking into the sea! area und
freezing, The test was terminated when continuad
heating caused gross ligquid leckage.

PERIOD ENDING JUNE 10, 1953

In an attempt to toke advantage of the apparent
nonwetting characteristic of grarhite with fluoride,
several dry runs were made with graphite as the
packing material. In this series of tests, ths
shaft rotated in o right-hand direction and o left-
hand Y thread was machined on the shaft. This
arrangement tended to pack the graphite ot the
flucride end of the seal. An air cylinder wos con.
nacted to the gland so that gland pressure could be
controlled, and with this st up, it was determined
that the packing pressure must be less than 30 psi
to avoid excessive friction between the graphite
and the rotating shaft,

Another seal comprised of bronze wool as o
retainer and artificial graphite plus 5% MoS, as
the packing materia! sealed helium ot 10-psi
pressure.  This seal also satisfactorily sealed
the fluoride NaF-ZrF UF, (50-46-4 mole %).
There was some indication frem stop-start tests
that an interface or meniscus wos established at
the hot end of the seal. However, in the start-stop
tests, performance was erratic. In some instances,
heat was required for starting ofter stepping, and
ot other times the shaft started freely, After more
than 500 hr, operation was continuing ot 1200°F
and 20-psig pressure, but there was the slight
leakage of granulated material that is typical of
frozen seals,

Grooved-Shaft Packed-Frozen Seal Tests. Various
packing matericls hove been tested in a sealing
gland in which both the shaft and housing sides
of the gland contain meshed annulor grooves to
minimize loss of scal moterial. Three tests were
made of seals with different packings but with
similar geometry and eoperating conditions. The
tests were operated with the fluoride NaF-ZrF -UF
(50-46-4 mole %) at temperctures ronging from
1050 to 1350°F. The shaft diameter was ]3/16 i,
and the seol glond length was 1% inches. la all
these tests it was noticed that the first leakage
of fuel from the seals was preceded by leckoge of
the packing material, which indicates that there is
need for improved retention of the packing material,
Start-stop tests werz made saotisfactorily only
after the addition of sufficient heat to compensate
for frictional heat loss when the shaft was stopped,
These thiee tests are summarized in Table 2.1,
after the first test showed the
arooved section of the shaft to be badly worn to
the extent that there was o single wide groove
covering about half the widih of the original grooved

Examination

17
ANP PROJECT QUARTERLY PROGRESS REPORT

 

 

 

TABLE 2.1. GROOVED-SHAFT PACKED-FROZEN SEAL TESTS
GROOVE FLUID FLUID DURATION
SHAF T DIMENSIONS PACKING PRESSURE | TEMPERATURE | OF TEST | RCASONFOR
MATERIAL . . o TERMINATION
(in.) (psi) (°F) (hr)
Type 316 stain- | % by ko by % | Dixon's No. 2 grophite 7.5 1050 to 1225 92 Bearing failure
less stee! plus Inconel braid
impregnated with MoS,
Stellite-coated 3/16 by 1/a by l’a Dixon's No, 2 graphite 10 1100 to 1200 53 Scheduled
type 316 plus Inconel braid
stainless steel impregnated with MoS,
:}‘16 by ]/8 by 1/8 Dixon’s Na, 2 graphite 10 1200 1o 1350 148 Loss of seal
plus superflake material during
graphite start-stop test

 

 

 

 

 

 

 

section and approximately as deep as the original
grooves, Neither of the other seals showed as
severe damage, although some wear occurred at
the retainer end of the packing used in the second
test,

Bronze Wool and Mo3, Packed-Frozen Seal Tests,
Several packed-frozen seals in which the packing
gland contains bronze wool and a lubricant, such
as MoS, or MoS, and graphite mixture, have been
tested with the molten fluoride NaF-ZrF -UF,
(50-46-4 mole %). A leakage rate on the order of
1 g/hr has been obtained with a ]3’]6-in.-diq shaft
and 10 psi pressure across the seal. The leakage

rate is not only a function of the pressure across
the seal but also of the location of the actual
point of seal within the gland, The attempt to
adapt this seal geometry to a larger shaft (2]/2 in.
in diameter) has not been entirely successful.
Applications of successful small-diameter-shaft
packings to larger shafts have usually been ac-
companied by perturbations attributable to the
increase in surface speed, higher friction, and
increased area for heat conduction from the hot
liquid being sealed. A summary of design and
operating conditions for the four seals tested is
given in Table 2.2,

TABLE 2.2, SUMMARY OF BRONZE WOOL AND M052 PACKED-FROZEN SEAL TESTS

Shaft material: Stellite-coated type 316 stainless steel

 

 

 

SHAFT SEAL FLUID FLUID DURATION | bEasON FOR
DIAMETER | DIMENSIONS PACKING PRESSURE | TEMPERATURE | OF TEST | 1 ERMINATION
(in.) (ind) (psi) CF) (hr)
1:}16 Y by 5 Y=in, layers of bronze 10 1200 to 1275 78 Seal leak
wool plus M052
z-in. layers of bronze 10 1200 to 1275 620 Still operating
wool plus M052 plus
graphite
2} % by 5 | Vi, layers of bronze 10 1200 to 1275 Seal leak
wool plus Mc>S2 plus
graphite
]{l-in. layers of bronze 10 Still operating
wool plus graphite
plus 5% M052

 

 

 

 

 

 

 

18
Although the first secl remained free ofter stop-
start tests of up to 10-min duration, the seal lecked
excessively when raised to a temperature above
the melting point of the fluoride, and the test was
terminated. The power requirement for this seal
was approximately ¥ kw, with slight surges.

Start-stop tests of up to 101/2 hr have been made
in the second test, but heat was added to maoke up
for friction heat loss. Although the seal is believed
to have moved up inte the glond region, the test
is operating smoothly with a shaft rpm of 1500,
The initial leakage rate of 1 g/day increased to
1.5 g/day, and it was necessary to raise the seal
temperature about 100°F above its operating tem-
perature to enable the motor to restart the shoft
seal without assistance.

In the first test with the 2)-in.-dia shaft, the
seal lecked liquid which appeared to be an MoS,
mixture, Power fluctuations were greater than
hitherto observed. During a 3-hr stop, the shaft
remained free. Although the seal usedin the second
test had to be repacked after excessive initial
graphite leakage, operation is continuing smoothly
with a low leakage rate and o somewhat variable,
but usually smooth, power requirement,

Frozen-Lead Seal Test. A series of tests has
been conducted to determine the feasibility of
sealing a high-temperature fluoride pump with a
frozen-lead seal. These tests show this type of
scal to be feasible, since it has been determined
that molten lecd and molten fluorides do not inter-
mix when in intimate contact with each other.
Rather, the fluorides Hoat on top of the lead, and
there is, apparently, a sharp fluoride-lead interface.

The test equipment consists of a 13/1 -in.~dia
type 316 stainless steel shaft powered by a 2-hp
motor. The shoft was operated in avertical position
at a speed of 1500 to 2500 rpm and the seal was
below the shaft, A helium gas blanket was pro-
vided for the cold end of the seal, and above the
seal was a pot contuining lead at 1000°F, A series
of five test runs was conducted for a total operating
time of 550 hours.
by extremely smooth operation and low power input
during the first 30 10 100 hr of operation. Continued
operation resulted in shaft seizure in the seal
region, and the test was terminated. The conditions
which appeared most likely to be contributing to
seal foilure were oxidation of the lead in the frozen
seal because of the air at the cold end of the seal,
insufficient shaft clearance at the hot end of the
lead sealing annulus, and impurities in the lead.

These runs were characterized

PERIOD ENDING JUNE 10, 1953

In the new seal test rig design, which eliminates
the difficulties mentioned, the frozen seal region
was shorteried to approximately }’2 in, in length
and liquid coeling was substituted for gas cooling.
High-purity lead was obtained and pretreated by
bubbling hydrogen through it for approximately 100
hours. A frozen-lead seal has operated in the new
rig for upproximately 500 hours. The power input
to the shaft has been low, there have heen no power
tluctuations, ond the lead leckage rate has been
negligible. The shaft speed during this test was
varied from 2000 to 4200 rpm, and performance over
the entire speed range was excellent,

V-Ring Seal.
rings machined from Graphitar No. 14 and a hot-

A seal consisting of alternate

pressed compact of copper and 14% MoS, is being
tested. The rings are assembled as shown in Fig.
2.5, A 13/16-in.-diu hardened shaft rotates against
the Graphitar. A gas cylinder is used as the means
of applying force to the gland to compress the
rings, The seal was designed to operate at ap-
proximately 1100°F.

A dry run of this seal indicated that helium leak-
age through the seal was almost zero with the seal
area at a temperature of above 1000°F. Operation
during the dry run consisted of rotating the shaft
at approximately 1650 rpm with the seal area at a
temperature of above 1050°F, Power dissipation
in the seal is approximately 50 watts, With these
conditions, an ottempt was made to force helium
through the seal with pressures of up to 30 psig.
Flow of helium to the seol was read on a gas
rotnmeter. The unit was operated for one day with
indicaoted flow and then disassembled for
inspection. When reassembled, the unit was oper-
ated for three days with an indicated flow of about
0.4 cm3/sec, so preparations were made fo intro-
duce fluorides. Calculations indicate that for o
gas leakage of about 2 em 3/sec of helium the cor-
responding fluoride ledkage will be about 0.5
cm3/day,

BeF, Seal Tests. A series of tests has been
conducted with a stuffing-box type of seal pocked
with BeF , to determine whether the high viscosity

ere

of BeF, at its softening temperature would give a
satisfactory szal against the fluoride mixture NaF-
ZrF4-UF4 (50-46-4 mole %) at high temperatures.
A summary of the seal design and operating data
is given in Table 2.3,

In the first test, a !@-in.-déa shaft was sealed by
a series of surrounding annular compartments con-
taining BeF , powder. During operation of the shaft

19
ANP PROJECT QUARTERLY PROGRESS REPORT

 
    

LS LY

N

A ‘I"‘ ‘fi;’?&z

        

 

s

£

et R i N

ey

i

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AR\ P a)
PR ac N e g

A
Z

 

 

 

 

 

ey »'. . »‘l',
e e P

 

 

 

 

 

 

 

 

L-._._H
0.002 +0.0005 (COLD) 2

UNGCLASSIFIED
DWG. ASK 16050A

  
 

—0.0000

SHAFT
SURFACE 1

 

 

CONTAINER WALL

Fig. 2.5. Y-Ring Seal

at 440 rpm, the power needed to overcome friction
was too lowto read on a 10-kw full-scale wattmeter.
Leakage through the seal wos zero. A packing of
l/4 in. of graphite-impregnated asbestos was used
at the cold end to contain the dry powder. The
shaft was stopped and started many times for
periods of up to 1 hr, and only motor power was
used for restarting. After fermination of the test,
the seal was cooled and sawed lengthwise. The
fuel had penetrated only into the second compart-
ment, that is, less than ?’/4 inch. None of the pack-
ing material beyond the second compartment had
melted. Chemical analysis showed 0.96 mole %
BeF, in the first compartment and 50.2 mole %
BeF , in the second compartment,

In the second test, a 21/2-in.-dia shaft was sealed
by o packing glond which was packed with copper-
braid sleeves filled with BeF, powder. One turn
of graphite-impregnated asbestos was used on the
air end of the seal to contain the dry powder. This
seal showed none of the characteristics of a BeF,
seal, but operation and failure were similar to the
experience with frozen-fluoride seals.

20

In the third test, a vertical ]3/]6-in.-dic1 shaft was
packed with BeF, in annular compartments, as in
the first test. Failure after ”’2 hr of operation at
880 rpm appeared to have been the result of the
air end of the seal becoming too hot (870°F), The
seal ran well until it failed.

For the fourth test, a horizontal 2]/2-in.-diu shaft
was assembled with annular compartments filled
with BeF,. Improper design caused thermal dis-
tortion and mechanical binding. The shaft was
started 15 to 25 times, but each time it was stopped
by metal to metal contact in the seal.

In the fifth test, a 23/8-in‘-diu shaft, which simu-
lated a horizontal ARE pump shaft, extended
through a packing gland into a body of the fluoride
fuel NaF-ZrF -UF, (50-46-4 mole %). Provision
was made for selectively heating the end of the
packing gland adjacent to the hot fluorides and for
cooling the gland at its outer end. A metal sleeve
around the shaft, sealed at the inner end of the
gland, provided a narrow annulus ]/1 in. wide and
obout 2 in. long for the fluorides to pass through
and be cooled somewhat before they entered the
PERIOD ENDING JUNE 10, 1953

TABLE 2.3. TESTS OF Ber-PACKED SEALS ON ROTATING SHAFTS
OPERATED IN NnF—ZrF4-U Fy {50-46-4 mole %)

 

 

 

 

SEAL PACKING CHAMBER
PACKING - T PACKING MET SHAFT H N
TEST MATERIAL Over-all Llflr:zgoridz Ne. of ANNULUS gfl‘fmf? spégn PRE[s"slfigE Dgrf '}qur REASON FOR
No. COMPOSITION Length | oo " | Compart- |  THICKNESS (in) ( . L TERMINATION
o . 9 ; in, rpm) {psig) {hr)
{mole %) (in) (in.} ments {in.}
1| 100% BeF, 4 2 5 0.25 Y 440 30 50 Scheduled
2 | 100% BeF, 1.25 0.B5 5 In coppar 2k 550 10 1 Senl leak
braid sleeves
3 | 100% BeF, 3.1 2 6 0.25 % 880 30 1% | Seal l=ak
4 | 100% BeF, 4 2 7 0.375 2} 650 5 Off ond on | Mechonical
worpage
5 | 75% BeF, plus 25% 5 2 1 0.4375 2% 380 ta 1460 1110 3
flucride mixture
6 Various percentoges 3.1 2z 7 0.25 Hfllfi 1350 to 3800 5t % 104 Failed during
of BeF , plus ZrF stop-start tests
3
7 100% NaBeF 2.1 2 7 0.25 J 1%, 1350 5 J

 

 

 

 

 

 

principal packing zone, Strands of copper rope
were then placed in the seal annulus to a depth
of ]/2 in. to serve as the principal deterrent to entry
of the fluorides into the seal proper, which was
formed by a 2-in. copper sleeve. The seal cavity
was filled with 75 mole % BeF, ond 25 mole %
fluoride fuel granules, and as a result there was
25 to 30% void space in the cavity. The seal was
heaoted and instrumented so that the temperature
gradient along the seal could be controlled. Pene-
tration of the fuel proceeded until the resulting
mixture in the seal cavity was too cold or too
viscous to penetrate further. Shaft speed wos
maintained at 1400 rpm for approximately 400 hr,
and the test ended at 448 hr because of the failure
of the external seal heaters and plugging of the
coolant line. Power consumption in the seal be-
cavse of friction {primarily at the ocuter end of the
seal) was of the order of 400 watts. Many per-
versities were encountered, as well as long periods
(days) of stable operation with low friction in the
seal. During the stable periods, fine powder leaked
from the seal at a rate of 10 to 15 cm® per day.
Shaft wear was light except under the outer copper,
which was cold.  Sufficient powdered material
accumulated in this location to act as a grinding
compound for the shaft material. Nearly all the
major perversities of this seal could be attributed
to the accumulation of fluoride powders under the
external copper, augmented by oxidation of the
fluorides at the cold end.

 

 

 

 

 

 

The sixth test was run in the same equipment as
was the third test, but the following alterations
were made. (1) Two turns of ]/A-in.-dia copper
cooling coil was wrapped (not soldered) arcund the
top of the seal housing to carry cooling water,
(2) The fluoride packing was divided into seven
compartments by wusing type 316 stainless steel
spacers and washers, and copper-rope packing was
placed at the hot and cold ends of the seal. {3) The
fluoride packing consisted of 8- to 20-mesh grains
of BeF, plus ZrF, in the following weight per-
centages of ZrF, in the layers from the hot to the
cold end, respectively: 0, 15, 25, 40, 55, 70, 100.
The unit was operated 104 hr at o sea! temperature
of 1200 to 1250°F on the hot end and approximately
650°F on the cold end. The speed was varied from
880 to 3800 rpm. The helium blanket pressure was
5 to 10 psig. The main purpose of this test was
to see whether ZrF, would prevent leakage (be-
cause of its high melting point)., Although the
leakage was not great, it waos fairly continuous,
and the stop-start test was unsuccessful for a 10-
min stop.

Exomination of the shaft after the fourth test
revealed that a high percentage of NaBeF, was
possibly the most effective sealant. Therefore it
was decided to operate a seal with NdeF3 packing
in the same equipment as was used in tests 3 and
6.

This seventh seal was similar to that used in
the sixth test, except that all seven compartments

21
ANP PROJECT QUARTERLY PROGRESS REPORT

were packed with NaBeF,. Gronules of 8 to 20
mesh (Tyler Standard) were used to pack the seal,
and the copper cooling coil was soft soldered to
the seal housing. The shaft speed was increased
from 1500 rpm, that is, until the shaft peripheral
speed was at ARE design point (15,3 fps), and was
maintained there for the balance of the 160-hr test.
Several successful stop-start tests were made with
stop periods of up to 10 minutes. On most restarts,
some assistance to the 2-hp motor was required;
however, several restarts were made with no as-
sistance. There was no detectable leakage of
sealant, The friction drag was estimated at about
150 to 200 wotts. Temperatures along the seal
remained constant. The temperature range was
from approximately 1250°F at the hot end to ap-
proximately 200°F at the cold end. The seal was
cooled and separated from the unit and sectioned
axially. Inspection revealed that the fuel had
penetrated four of the seven compartments and that,
in the remaining three compartments, the NaBeF3
had fused or sintered and changed color from almost
white to a gray-brown. The only shaft scoring
found was at points of contact with the stainless
stee!l washers used as separators between com-
partments.

Packing Penetration Tests, Packing penetration
tests have been continued in which graphite or
graphite mixed with another material is used as
the primary sealant, The use of graphite has been
emphasized in these tests because it appears to
be one of the few materials capable of preventing
fluoride leakage.

A test in which Baker Chemical Company pow-
dered graphite was used to seal against the fluo-
ride NuF-KF-LiF-UF4 (10.9-43.5-44.5-1.1 mole %)
was terminated after 240 hr with no leakage, as
reported previously.’3) Since other tests of similar
seals used 1o seal against the fluoride NaF-ZrF -
UF, (50-46-4 mole %) had leaked, it was thought
that the sealing might be dependent either on the
presence of LiF or on the absence of ZrF,. How-
ever, during this quarter, another test was operated
for 690 hr with the same type of graphite and
zirconium-bearing fluoride mixture, and there was
no leakage. This test was terminated at the end
of the 690-hr period, because excessive oxidation
occurs if the graphite tests are allowed to run for
extended periods of time., The oxidation begins

 

(5w, B. McDonald et al., ANP Quar. Prog. Rep. Mar.
10, 1953, ORNL-1515, p. 28.

22

at the opening in the bottom of the container and
then extends upward toward the fuel region. Upon
postrun examination, a large cavity was found in
the graphite around the opening. If the test had
continued, there would have been leckage due to
the oxidation of the graphite rather than to pene-
tration,

All the graphites tested this quarter, except the
Baker Chemical Company graphite, have been of
the artificial type, which should have a very low
amorphous carbon content, Tests of these groph-
ites have had one thing in common: leakage began
at a very slow rate and could only be detected by
the small amount of corroded metal that fell from
the furnace in which the packing containers were
heated. The tests were continued and terminated
only when fuel was detected in the receptable
beneath the furnace. Because of the slow leakage
and continuation of the tests, the outer surfaces
of the containers were very badly corroded; they
were covered with o heavy scale that could be
removed easily. Results of these tests are summa-
rized in Table 2.4,

One combination of packing material being tested
is a mixture of 80% artificial graphite powder (ob-
tained from the Y-12 Carbon Shop) and 20% pow-
dered BeFZ. The mixture was compressed ond
then heated to 1150°F and further compressed.
Heating and compressing were done three times,
and the packing lost approximately 40% of its
original volume, The test of this packing material
is being run at the usual pressure, 30 psi, but the
test temperature is 1150°F instead of the usual
1500°F, The test has been running for 600 hr with

no feakage.

INSTRUMENTATION

P. W, Taylor D. R. Ward
ANP Division

Pressure Measurement. A large pressure trans-
mitter, pattemed after the Moore Nullmatic instru-
ment, () that utilizes a triple-ply Inconel bellows
as the pressure-sensing element has been in inter-
mittent service on a fluoride pump loop. The
transmitter operates upside down to measure pump
suction pressure (about 5 psi) and it is completely
filled with 1100°F molten fluorides., The tempera-
ture of the fuel in the loop is approximately
1200°F, The total operating time to dote is 360

 

(6)y, B, McDonald et al., ANP Qugr. Prog. Rep. Dec.
10, 1952, ORNL-1439, p. 30.
PERIOD ENDING JUNE 10, 1953

TABLE 2.4, SUMMARY OF PACKING PENE TRATION TESTS PERFORMED DURING THIS QUARTER

Pressure: 30 psi

Temperature: 1500°F, except as noted

 

 

 

TEST DURATION
PACKING MATERIAL OF TEST REMARKS
NO,
(hr)
n Stainless steel wool impregnaoted with MoS,, 0 L.eaked immediately
12 Boker Chemicol Company powdered graphite 6%0 Did not leak
13 MoS, in copper sheath ]/2
14 Graphite from Y-12 Carbon Shop 262 Scale indicated slight leakage
15 National Carbon Company graphite 2301 191 Scale detected at 72 hr
16 80% graphite from Y-12 Carbon Shop and 20% Be F2 Operating temperature 1150°F; had not
leaked at 600 hr; test continuing
17 50% graphite from Y-12 Carbon Shep and 50% M052 40 Scale detected at 30 hr

 

 

 

 

hours., The fuel froze in the transmitter during
twe pump shutdown periods, and it was found that
subsequent remelts and startups did not damage
the transmitter.

A small nickel transmitter operating completely
filled with fluoride fuel on a static test (10 to
30 psi, one cycle every hour) at 1100°F failed
after 2500 hr of continuous operation. Postoper-
ative inspection revealed two small holes in one
of the inner bends of the 0,005 in. (wall thickness)
bellows. Throughout the 2500-hr period, operation
was excellent, The combined range and zero shifi
over a 1500-hr period wos less thon 1]/2% of full
scale (30 psi).

Two other type 316 stainless steel transmitters
that are being operated on a similar static test
are still functioning smoothly, One of these trons-
mitters has logged 3400 hr, to date, in fuel ot
1100°F; the other has operated 3500 hr in lead at
700°F,

The initial test of the single-diophragm force-
balanced pressure transmitter is being run at
1050°F with no fuel. This instrument, constructed

entirely of Inconel with a 0,014-in, Incone! X
diaphragm, is being operated at a static pressure
of 50 psi for 700 hr, with o cycle every 4 hr to
20 psi. Upon heating to 1050°F from room tempera-
ture, there was an approximate 2% increase in
output ot 50 psi. After 1200 hr of operation at
1050°F, the output has increased 3%. Response
to pressure chonge is excellent, with ne visible
delay. This fast response is obtained at the ex-
pense of a very large, balancing, «ir flow. The
instrument uses approximately 24 cth of air,

NaK Leak Detection Methods, Since NaK may
be used os the final leak-checking material for the
ARE fluid circuit, a religble NaK leak detector for
small leaks is desirable. Some preliminary testing
has been done to determine the effect of NoK vapor
on copper wool, With a pure helium atmesphere
at temperatures from 250 to 650°F, the copper wool
gave a substantial basic indication in phenol-red
indicator solution (in H,O) after an exposure time
of 1 hour. The addition of about 5% oxygen to the
helium atmosphere decreased the strength of the
indication but did not eliminate it entirely.

23
ANP PROJECT QUARTERLY PROGRESS REPORT

3. REFLECTOR-MODERATED CIRCULATING-FUEL REACTORS
A. P. Fraas, ANP Division

An extensive description of a proposed reflector-

moderated circulating-fuel reactor suitable for

powering an aircraft was presented in the previous |

Subsequent studies of the full-scale
power plant system have included an examination
of reactor control, and a revision of calculations
on the basis of new data from the critical assembly
of this reactor configuration. The reflector-moder-

report, (1

 

(DA, P. Fraas ot al., ANP Quar. Prog. Rep. Mar. 10,
1953, ORNL-1515, p. 41-84,

ated reactor critical experiment is discussed in
sec. 5, and shielding configurations applicable to
this reactor are described in sec, 14,

REACTOR CONTROL

The control problems of the full-scale power plant
are being examined. The fluid flow rates, transit
times, and thermal capacities for the various ele-
ments of the full-scale 200-megawatt sea-level
aircraft power plant are presented in Table 3.1.
These values should be considered as those for a

TABLE 3.1. FULL-SCALE 200-MEGAWATT SEA-LEVEL AIRCRAFT POWER PLANT SYSTEM DATA

MaK in each radiator core circuit
NaK in radiator header drums
NaK in radiator tubes
NaK in 3.3-in.-ID 20-ft-long outlet line
NaK in 3.3-in.-ID 20-ft-long inlet line

MNaK in intermediate heat exchanger

NaK flowing in enfire circuit

NaK in header tank

Total NaK in sach system
Total NoK for sach engine

Total NeK in complete power plant

81b
8
50
56
28

i

150
30

 

180
720
2880 Ib

NaK flow rote through each radiator circuit (for a 400°F Az and a 12,500-kw heat

rejection rate)

MNaK transit time through system
NaK transit time through intermediate heat exchanger

NaX transit time through radiator tubes

Fuel in core circuit (flowing)
Fuel in header tank

Fuel in core

Fuel flow rate

Fue!l cireuit transit time

Thermal capacities of power plant
Radiator cores (1b)
Lines and pumps

MNaK (flowing)

Intermediate heat exchanger
Total for NaK systems
Fuel

Total for power plant

24

2.5 #t3/sec or 120 Ib/sec

1.25 sec
0.23 sec
0.067 sec

6 ft3

1 £e3

1.2 #3
12.0 #13/sec
0.5 sec

600 Btu/°F
250
600
250

 

1700 Btu/°F

300 Bru/°F

Ao r———

2000 Btu/°F
typical case; considerably less favorable values
might be required with some designs (for example,
NaK line lengths would be considerably greater for
designs with engine nacelles in the wings), while
on the other hand, more favorable values might be

obtained for some of the other factors through

design changes, with small penclties in turbojet
engine weight or performance.

It is important to note that the NaK transit time
for the intermediate heat exchanger is 0.23 sec,
which shows that even if a step change in NaK
temperature could be effected in the secondary
circuit, such a temperature change would be damped
out over a period of more than 0.23 sec in the
intermediate heat exchanger. It is interesting to
note, too, that the combined thermal capacity of
the fluoride and NaK circuits is 2000 Btu/sec;

hence, even if the reacter were operating at 200

PERIOD ENDING JUNE 10, 1953

megawatts and heat removal in the engine radiators
was stopped instantaneocusly (something that would
be impossible to effect), the temperature in the
reactor could not rise at a rate greater than 100°F
per second. Thus the negafive temperature co-
efficient of the reactor should be able to cope with
a condition even so impossibly severe as this with
a wide margin of safety,

GEMNERAL DESIGN PARAMETERS

Several parameters of importance in both control
and shielding calculations have been compiled in
Tables 3.2 and 3.3 for ¢ variety of reactor power
outputs and core diameters. A constant power
density of approximately 10,000 kw/ft3 in the
intermediate heat exchanger was assumed for all
cases.

TABLE 3.2, FUEL VOLUME IN CORE AND FAST-NEUTRON ESCAPE FOR VARIOUS CORE DIAMETERS

Constant Heat Exchanger Power Density: 10 I‘a.'iegcmrflfls/f’r3

 

Core digmeter, in, 14.3
Escape factor for neutrons* 0.40
Escape factor for 3-Mev gammas 0.47
Fuel volume in core, > 0.625

 

 

18
0
0
1

22.7 28.5 36 45.3
.35 0.32 0.29 0.26 0.23
.43 0.39 0.35 0.30 0.27
.25 2.5 5 10 20

 

 

 

 

 

*Fraction of neutrans escaping from reactor cors as fast neutrons.

TABLE 3.3,

FRACTION OF FUEL IN HEAT EXCHANGER FOR VARIOUS

REACTOR POWERS AND CORE DIAMETERS

Constant Heat Exchanger Power Density: 10 Megawafls/fi‘3

 

 

 

 

REACTOR FUEL IN HEAT FRACTION OF FLOWING FUEL IN HEAT EXCHANGER
POWER EXCHANGER Core Diameter (in.)
(megawatts) (1) 14.3 18 22.7 28.5 36 45.3
50 1.25 0.67 0.50 0.33
100 2.5 0.80 0.67 0.50 0.33
200 5 0.89 0.80 0.67 0.50 0.33
400 10 0.99 0.80 0.67 0.5 0.33

 

 

 

 

 

 

 

 

25
ANP PROJECT QUARTERLY PROGRESS REPORT

4. REACTOR PHYSICS
W. K. Ergen, ANP Division

A review of the status of the work on kinetics
of the circulating-fuel reactor brings out three main
points: (1) permanent introduction into the reactor
of more than 1% of the total uranium investment
in excess of the amount of uranium required to
make the reactor critical, has to be avoided or
quickly compensated for by control rods or the
reactor will be damaged by the resulting permanent
rise in the equilibrium temperature, (2) oscillations
in reactor power will damp out quickly if the
reactor can be regarded as rigid and the flow
pattern remains constant, and {3) following a sudden
introduction of excess reactivity, no serious over-
swings in power and in temperature are expected.
The possibility of an excessive pressure surge has
not yet been ruled out,

A number of results were obtained by further
analysis of the critical experiment data for the
ARE. The reactivity values of both the regulating
and the control rods, as well as the axial and the
radial cadmium fractions of the reactor, have been
determined., The critical mass is, however, still
uncertain. Since the earlier estimate(!! of 28 Ib
was made, the coolant composition was changed to
eliminate the highly absorbing potassium. This
would allow a reduction in critical mass, but other
revisions in the calculations, such as revision of
the data on the absorption of Inconel, increase the
critical mass,

KINETICS OF A CIRCULATING-FUEL REACTOR
W. K. Ergen, ANP Division

If a disturbance is introduced into a reactor, the
kinetic behavior of the reactor introduces three
problems: (1) the behavior at long times after the
disturbonce, (2) the question of whether oscil-
lations initiated by the disturbance build up or are
damped, and (3) the extent of the first overswing
of reactor power, temperature, and pressure.

Behavior ot Long Times After a Disturbance, [t
is known that an excess reactivity Ak introduced
into a reactor with a negative temperafure coef-
ficient —a, sets up a new equilibrium temperature
which exceeds the original equilibrium temperature

by AT = Ak/a. Thus if a = 1074/°C and Ak = 1%,

“)W. B. Cottrell (ed.), Reactor Program of the Air-
craft Nuclear Propulsion Project, ORNL-1234, p. 45
(June 2, 1952).

 

26

AT = 100°C, which is barely tolerable. However,
if o were smaller or Ak greater, the reactor would
be in danger. A Ak of 1% con be caused, for
instance, by introducing into the reactor 3 to 4%
more uranium than is required for criticality, that
is, less than 1% of the total uranium investment,
The HRE, with its much larger temperature coef-
ficient, can tolerate much more excess fissionable
material in the core; but, on the other hand, the
large concentration differences occurring in the
HRE fuel ore mainly due to the reprocessing of
the fuel circulation, and such reprocessing will
not be attempted during the operation of the ARE.
Though the possibility of bringing excess fission-
able material into the reactor is more or less a
specific difficulty of the circulating-fuel reactor,
other aircroft reactor proposals have similar prob-
lems, such as the possibility of the deformation
of voids or the compression of supercritical low-
density water,

Oscillations, It has been shown that the circu-
lation of the fuel causes damping of power oscil-
lations of the reactor, This damping can now be
demonstrated, even in cases in which the power
distribution varies along the fuel path and perpen-
dicularly to it and in cases in which different fuel
particles have different fransit times through the
reactor. Also, this domping is added to the damping
afforded by whatever delayed neutrons are left, and
there is no destructive interference between the
damping by delayed neutrons and the damping by
circulation, However, the present theory is as yet
incomplete, because it dees not take info account
possible mechanical deformation of the reactor or
variation in the hydrodynamic flow of the fuel,

Professor Cornelius Weygandt has obtained, on
the ditferential analyzer of the University of Penn-
sylvania, a few numerical solutions to the kinetic
equation of a circulating-fuel reactor, and some of
them are shown in Fig. 4.1, The solutions refer
to the constant-power-distribution constant-tfransit-
time case without delayed neutrons, [t was as-
sumed that at times ¢ < 0, the transit time is equal
to the period p of the oscillation, At t = 0, the
transit time changes, while all other parameters
of the equation remain constant, Figures 4,1a and
4,16 refer to a small oscillation; for the undamped
UNCLASSIFIED
DWG 19087

 

099

 

 

(a) 8=09p
P/P,
tCQ e *—-J!‘ e 120
|
104
e — —_—— t
0se”” |
b B=11p
PP
‘U
e {30

PO e

   
  

6.3

0.0t (€} 9=09p

P/Pq

10 mm—m——— A {30

85

e}
oo | @ 8=11p

P/P,

1< == {30

6.5\-“

ollel]

 

(8) 8215

Fig. 4,1, Differential Analyzer Solutions to the
Kinetic Equation of a Circulating-Fuel Reactor,

PERIOD ENDING JUNE 10, 71953

condition at £ <0, the maximum and minimum power
deviate by only 1% from Py, At ¢ = 0, the transit
time in Fig. 4.10 abruptly gets 10% smaller than
the period of the undamped oscillation, ond in
Fig. 4.1b, 10% greater, The oscillation is practi-
cally sinusoidal, and the damping is rather weak,
though present, Figures 4.1c, 4,1d, and 4.1 use
a different scale and refer to a more violent oscil-
lation, In the undamped condition, the moximum
of the oscillation is 6.5 times P and the minimum
is only 0.01 Py; that is, the reactor almost shuts
itself off, At ¢ = 0, the transit time jumps to 0,9,
1.1, aond 1,5 times the period of the undamped
oscillation in Figs. 4.1c, 4.1d, and 4.le, respec-
tively, Figures 4.l1c and 4.1d show that the
damping is small initially because of the small
deviation of the transit time from the period, As
the amplifude decreases, the period decreases. In
Fig. 4,1c, this means that the period more closely
approaches the fransit time, and hence the damping
decreases. In Fig. 4.1d, the opposite is the case.
In Fig, 4.1e, the large deviation of the transit time
from the pericd results in strong damping during
the first cycle, Then the period happens to have
decreased to one half the transit time, and the
oscillation continues essentially undamped. It
should be noted, however, that undaomped oscil-
lations of this kind are unstable, because a small
disturbance which decreases the amplitude and
hence the period would ‘‘detune’’ the system and
cause further domping. More important, the possis
bility of undamped oscillations is largely a conse-
quence of the simplifying assumptions used in the
computation, in porticular, the assumptions of
constant transit time and of constant power and
flux distribution, Under more general conditions,
the undamped oscillations are much less likely
to occur, and with proper design, their occurrence
can be prevented,

The First Overswing Following a Disturbance,
Sufficient work has been done to show that under
conditions envisaged for the aircroft reactor, even
a sudden application of considerable excess reac-
tivity will only result in a moderote temperature
overswing. However, since this temperature in-
crease occurs in a short time, it is not casy to
show, and it has not yet been shown, that the fuel
expansion occurs without accompanying large pres-
sure surges,

27
ANP PROJECT QUARTERLY PROGRESS REPORT

STATICS OF THE CIRCULATING-FUEL ARE
CRITICAL EXPERIMENT

C. B. Mills, ANP Division

The analysis of the data of the ARE Critical
Experiment is not yet complete, However, several
interesting results have been obtained.

1. The multiplication constant of the reactor
with a simple loading pattern of about 61 tubes in
the 70 fuel-coolant tube holes(? in the core was
recalculoted, and aoll the correction factors now
known were used. The result was 1,003, as com-
pared with the experimental value of 1,000, As
reported previously,(s) the assembly also went
critical with a higher U235 concentration and about
42 fuel tubes. Recalculation of this arrangement
gave o multiplication constant of 1,009, These
slight errors are consistent with a slightly in-
correct value for neuvtron chamneling from the un-
filled fuel-tube holes., No large errors have as yet
been found,

2. The danger coefficient of Inconel, the struc-
tural material to be used in the hot ARE, was
inconsistent with the assumed dbsorption cross
section of the material. An error was found in the
basic data which increased the values of the
thermal macroscopic absorption cross section from
0.310 em™! to 0.356 cm™!, The reactivity coef-
ficient for Inconel thus changed from 0,170 to
0.231, and the predicted critical mass of the ARE
was increased by about 4 Ib, Previously,!V a
value of 28 |Ib was predicted as the uranium re-
quired in the ARE core. However, since that
prediction was made, the fuel and reflector coolant
were changed in composition and the highly ab-
sorbing potassium was eliminated. The Inconel
correction and the elimination of the potassium
almost compensate for each other, but because of
the possibility of other corrections, the critical
mass is still somewhat in doubt,

3. The computed and the experimental cadmium
fractions, both radial and axial, for the ARE (CA-8)
are given in Fig, 4.2, The discrepancy at the core
boundary is largely due to the assumption of core
homogeneity, with the boundary specified by the
mean first neutron collision distribution.

 

(2)D. Scott and C, B, Mills, ANP Quar. Prog. Rep,
Sept, 10, 1952, ORNL-1375, p. 43.

B3)p, seott, C. B. Mills, J. F. Ellis, D, V. P. Williams,
ANP Quar. Prog, Rep, Dec. 10, 1952, ORNL-1439, p. 54.

28

4, The assumed cos? @ for control rod sensi-
tivity and the cos? @ 40 for control rod value are
compared to the experimental values in Fig. 4.3,

DWG. 12897

REACTOR RADIUS (in)
25 0 25

 

 

 

 

 

 

 

 

 

 

 

1 :
FUEL-TUBE SPACING, 3}, in, &3
TUBE DIAMETER 1Y, in.
EXPERIMENTAL
NS h RADIAL \'
\ | / } _
z > . £ £
O

= | =
= | 2
& 5 THEORETICAL —_ Z
w o0—% . - o0 4

[ae] @
z | ° 2 | g
= o b

a AXIAL o
g & - e 3
y_..fl - CORE - &3 '&J

o Q

hd I

7 INCONEL” 7~ 7 7 7 7]
1 25

 

 

 

 

1 0
CADMIUM FRACTION

—

Fig. 4.2, Rodicl and Axial Cadmium Fraction in
the ARE Critical Experiment,

A
DWG. 19898

o

   

A
&
=
o
S
&

°© o o
H OO ®©

9
\?Cu
¥

o
n

o

el END OF REACTOR

ARBITRARY SCALE

-0.4

-0.6 o EXPERIMENTAL
COMPUTED
-0.8

-1.0
5C 40 30 20

10 0O
RADIUS, LENGTH FROM REACTOR CENTER (cm)

10 20 30 40 5C 60 70

Fig. 4.3. Control Rod Values for the ARE Crit-

ical Experiment,
PERIOD ENDING JUNE 10, 1953

 

   
 

 

 

 

 

 

 

 

 

 

 

00 ow'e'.||9“g|33 5. Computed ond experimental values for the
cadmium fraction vs, computed values for per cent
thermal fission are shown in Fig. 4.4,
~ 6. The three safety rods are worth 13% in k_ /e
= This is twice the value required to reduce tf;e
9 ) o .
3 operating temperature 1000°F, The preliminary
B 50— N — guess for the value of the rods had been 15%,(4)
L
‘_XJ e ,,-_._.,,,,,,,4___L\_*
= COMPUTED | CADMIUM FRACTION
£ THERMAL 1 ca CL’LATEDt EXPERIMENTA (4)
* FISSIONS (%) “A-CY XPERIMENTAL W, B. Coitrell (ed.}), Reactor Program of the Aifr-
T sar | o329 | craft Nuclear Propulsion Project, ORNL-1234, p. 49
58 0 0345 | 036 (June 2, 1952).
59.6 0353 | 0.355T0 0.39
90.5 0.662 | 066
0 0.50 100

CADMIUM FRACTION FOR THE BARE ARE,C.A-8

Fig. 4,4, Per Cent Therma! Fissions as a Fune-
tion of the Cadmium Fraction for the Bare ARE,

29
ANP PROJECT QUARTERLY PROGRESS REPORT

5. REFLECTOR-MODERATED REACTOR CRITICAL EXPERIMENTS

A, D. Callihan
D. V. P. Williams

R. C. Keen
J. J. Lynn

Physics Division

D. Scott

C. B. Mills

ANP Division

The second critical assembly of the reflector-
moderated circulating-fuel reactor was described
previously. (1) The fuel, a mixture of Zr0,, NaF,
and C (66-24-10 wt %) with sufficient enriched UF ,
added to make the U235 density 0.2 g/cm3, was
placed around a central ‘‘island’’ of beryllium and
was, in turn, surrounded by a beryllium reflector.
The critical mass of this system was 7.7 kg of
U235_

POWER DISTRIBUTION

The power distribution and the method used in
obtaining it in a direction parallel to the reactor
axis were given in a previous report.?) The power

 

BERYLLIUM REFLECTCR

FuU

BERYLLIUM (SLAND

RELATIVE FISSION FRAGMENT ACTVITY
o

 

-5 G 5

distribution has been remeasured in the same
location, with the fission rate reduced in the ends
of the fuel coolant channels by shielding the chan-
nels on four sides with boral sheets 0.25 by 2.87
by 9 inches. (Boral is a mixture of aluminum
powder and boron carbide sandwiched between
aluminum sheets; the boron thickness is 0.25
g/ecm?.) Figure 5.1 gives the two power distri-
butions normalized at a point 5.25 in. from the
mid-plane,

 

“)D. V. P. Williams et al,, ANP Quar. Prog. Rep.
Mar. 10, 7953, ORNL-1515, p. 53.

pid., p. 58.

DWG. 19900

 

0.25-in.- THICK BORAL

i
0.002-in.- THICK URANIUM DISK PLUS
ALUMINUM CATCHER SOIL

 

DISTANCE FROM REACTOR MID-PLANE (in)}

Fig. 5.1. Power Distribution Parallel to Axis of the Reactor.

30
LEAKAGE FLUX

The fast neutron leakage through the end of the
reactor, both with and without the boral around the
fuel coolant channels, was measured by a fission
chamber placed outside the end reflector and
perpendicular to the axis of the reactor. The
chamber was lined with almost pure U238 gnd
covered with boral.
neutron leakage through the side of the reactor,
the chamber was placed in the mid-plane outside

For measurements of the fast-

the reflector and parallel to the axis of the reactor.
The fast-neutron leakage flux through the end was
30 times the corresponding flux through the sides
without the boral around the fuel-coolant channels,
and only about six times the flux through the sides
with the boral shield in place. Of course, this
reduction in fast-neutron leckage couses some loss
in reactivity; the introduction of the boral on one

end reduces the keff by 2%.

CONTROL ROD MEASUREMENTS

The loss in reactivity as a function of the po-
sition of g %’lé-in. stainless steel tube filled with
boron when inserted along the axis into the island
of the reactor is shown in Fig. 5.2. The boron
density was about 0,14 g per inch of the 0.1375
in.-1D tube, or about 50% of the theoretical density.
A solid stainless steel rod of the same size in-
serted along the axis to the mid-plane gave o loss
in reactivity of 5 cents,

DANGER COEFFICIENT MEASUREMENTS

A large number of danger coefficient measure-
ments were made with the test material placed in
both the fuel and reflector regions. The results

DG 9t
1

REACTIVITY LOSS (cents;

 

DISTANGE FROM REACTOR MID-PiLANE (in)

Fig. 5,2. Change in Reactivity Produced by
Boron-Filled Rod us a Function of Its Position
Along Reactor Axis,

PERIOD ENDING JUNE 10, 1953

are reported in Table 5,1. The last two columns
list the change in reactivity which occurred when
an empty test space was filled with the sample.
The samples of Li7, KF, NoF, Zr, Cr, ond RbF
were in aluminum containers 5/1‘5 by 2]/2 by 9 in.,
which had ]/32—in.-thick walls. These containers
were assumed to have no significant effect on the
reactivity. The sample of sodium was contained in
The reactivity value
of the can when empty was compared with that
obtained with the sample in place to obtain the
reactivity value for sodium. The other materials
required no containers. The long dimensions of
the samples were parallel to the reactor axis, and
the samples were symmetrically locoted about the
mid-plane. When in the reflector, the samples were
placed about 12 in. from the axis. In this location
the total neutron flux was greatest, (3)

The reactivity change incurred by the substitution
of a sample of D,0 for beryllium was measured at
three locations in the assembly, The sample was
99.2% D,0 and weighed 2.65 kg. It wos placed at
the assembly mid-plane in the island adjacent to
the fuel, in the reflector adjacent to the fuel, and
in the reflector 3 in. from the fuel layer, In each
of the first two locations the gain in reactivity
resulting from the substitutions was 12 cents; in
the latter position the reactivity was decreased
24 cents,

In the present design of a three-region reflector-
moderated reactor, the fuel-coolant flows in an
Inconel shell.) To evaluate the loss in reactivity
caused by the insertion of such a poison around
the fuel region, Y-in.-thick stainless stee! sheets
9 in. long were placed around one-quarter of the
fuel annclus. The section covered was symmetri-
cal cbout the mid-plane. The addition of this
stainless steelresulted in o 1.4% loss in reactivity,
which was compensated for by the oddition of
468 g of U235 distributed in the fuel region,

a thin stainless stee!l con.

CORRELATION WITH THEORY

Theoretical extrapolation of the various measure-
ments, as well as of the danger coefficient meas-
urements presented in Table 5,1, indicates that
the distribution of 1 vol % of Incone! throughout
the moderator and the addition of a !g-in.-thick fuel-
coolant container would increase the critical mass

Lid., p. 60, Fig. 4.22.
ipid., p. 62, Fig. 4.25.

31
ANP PROJECT QUARTERLY PROGRESS REPORT

TABLE 5.1. DANGER COEFFICIENT MEASUREMENTS

 

 

 

 

S AMPLE SAMPLE WEIGHT SAMPLE REACTIVITY CHANGE (cents/g)
(9) SIZE* In Fuel In Reflector
Li” 152.4 A 0.0000 ~0.0415
KF 198.0 A ~0.0162 ~0.0578
NaF 97.2 A —0.0058 --0.0354
Zr 287.7 A +0.0150 —0.0498
Cr 249.4 B ~0.0237 ~0.0680
RbE 156.3 B +0.0118 +0.0026
Inconel 1956.4 A ~0.0276 ~0.0448
Ni 2061.0 A ~0.0274 —0.0437
P 2544.8 C ~0.0008 ~0.0007
Pb 4592.7 D ~0.0012
Stainless Steel 1990.4 C ~0.0243 ~0.0385
Stainless Steel 3195.0 D ~0.0233
Stainless Steel 995.2 E ~0.0509
Be 753.6 D +0.0090
Na 366.9 D ~0.0190
Bi 3931.8 D +0.0006
Graphite 376.2 A +0.0108 +0.0043
Graphite (Density, 1.72 g/em®) 700.1 D +0.0064
Graphite (Density, 2.0 g/cm>3) 825.6 D +0.0065
BeO 375.5 F +0.0055

 

 

 

 

 

A = 5/‘]6I:;y5by9in.

B =
C =

32

%16 bY 2}2 by 9 in.
1/4by 5374 by 9 in.

1 by 27/8 by 85/8 in,
]/B by 5% by 2 in,
7/8 by 27/8‘by 5in.
to about 40 |b, compared with 17 Ib for the critical
experiment not containing the Inconel. A critical
mass of only 30 Ib results if the pressure shell is
only ys in. thick and the Inconel structure in the
reflector decreases rapidly from 1% at the core-
reflector interface to 0.04% ot the outside of the
reflector, with the average being 0.2%. This distri-
bution is probably o good approximation to the
distribution in on actual reactor.

The cadmium fraction for one point at the core-
reflector interface of the reactor was previously
reported(3) to be significantly different from the
computed value. The discrepancy has now been
resolved by shifting the radic!l coordinate in the
computation to properly take into account the void
which existed ot the boundary. Figure 5.3 shows
the revised computed curve; some difference still
exists between the shapes of the experimental and
the theoretical curves. These differences may be
due to (1} the deviation of the experimental, flat
geometry from the spherical geometry assumed in
the computation, (2) boundary effects not quite
accurately described by simple diffusion theory, or
(3) the radial orientation of the foils in the experi-
ment. The radial orientation makes locating the
effective centers of the foils difficult in the rapidly
varying flux.

PERIOD ENDING JUNE 10, 1953

 

   

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

A—
DWG. 19902
0.7 fw-— FUEL COOLANT~— i - Be REFLECTOR -]
3 '
FIT AT 19, in.
0.6
=4
o
"—
)
a
[ong
u.
=
2
=
0
<
[
QA [ 4 ® EXPERIMENTAL
o CALCULATED
0 — e )
3 2 1 0 1 2 3
INCHES -
Fig. 5.3. Computed vs, Experimental Indium

Activation and Coadmium Fractions at the Fuel.
Coolant-tg-Reflector Interface of the Reflector-
Moderated Reactor.

33
 
INTRODUCTION AND SUMMARY

The research on high-temperature liquids has
been primarily concerned with the detailed study
of the complicated NaF-ZrF -UF, system, the
production of the fluorides required by the ARE,
and, to a lesser extent, with other halide systems
containing UF, or UCI, (sec. 6). The pseudo-
eutectic that has been shown to exist at or neor
the composition 65-15-20 mole % is now designated
as the ARE fuel concentrate. Although its melting
point (515°C) is comparable to that of the previ-
ously considered 50-25-25 mole % mixture, the
new mixture is the more desirable, since it does
not segregate on cooling and, in addition, has o
lower vapor pressure. Use of this new fuel-
concentrate in conjunction with the NquF5 carrier
results in a fuel composition of 53-43-4 mole % of
NaF-ZrF ~UF . Construction of the facilities for
production of these fluorides in the quantities
required by the ARE is complete., These facilities
provide for the necessary purification by hydrogen-
ation-hydrofluorination to effect the removal of
such corrosive contaminants as NiF,, FeF, and

CrFa.

The recent corrosion studies have been almost
entirely devoted to the general problem of the
corrosion of Incenel by fluorides (sec. 7). Other
studies include secondary systems associated with
the ARE, such as fluorides on Stellites ond be-
ryllium oxide in sodium, as well as the longer
range problem of mass transfer in lead, The cor-
rosion of Inconel by the fluoride fuel NaF-ZrF .-
UF4 (50-46-4 mole %) haos been reduced to around
5 mils in 500 hr, and the corrosion mechanism is
now fairly well established. The corrosion rate
decreases with time ond is primarily a function of
the fluoride contaminants NiF and FeF,, which
react with the chromium in the mefal. At present,
the attendant formation of UF,; limits the extent
to which the NiF ond FeF, contaminants may be
removed from the fuel. With regard to beryllium
oxide in sodium, it has been fairly well established
that the corrosion mechanism.is in reality only the
mechanical erosion of the beryllium oxide surface,
In convection loop tests of various metals in lead,
only molybdenum and columbium did not show any
mass transfer, while type 446 stainless steel
showed a slight amount, Other metals, including
Inconel, Armco iron, ond types 304 and 347 stain-
less steel, showed extensive mass transfer,

Welding, brazing, creep-rupture tests, develop-
ment of cermets, and the fabrication of various
pieces of equipment constitute the metallurgical
research program (sec. 8). Manual inert-arc welds
have been successtully applied to tube-to-header
heat exchanger welds which did not lend them-
selves to the semiautomatic cone-arc technique.
The flowability of the Nicrobraz brazing alloy
used to assemble the liquid-to-air radiator section
is adversely affected by the presence of nitrogen,
The use of high-conductivity oxidation-resistant
fins for these radiators is obviously desirable,
Coating copper with chromium, nickel, Inconel,
and stainless steel has been undertaken. Severdl
copper alloys fin material.
Creep and stresse-rupture tests in air, argon, hy-
drogen, and fluorides indicate that Inconel ond
nickel are more sensitive to environmental changes
than are the austenitic stainless steels, The
creep rate of Inconel in tluorides is comparoble
to that in argon, Of the several methods for the
fabrication of spherical solid fuel elements that
were investigated, forcing the molten alloy through
a small orifice produced the most uniform and
least oxidized particles. MNumerous compacts with
14 vol % MoS5, have been hot pressed for use as
pump seals,

show promise as

While the physical properties of molten fluorides
and hydroxides are being measured at temperatures
of up to 1000°C, the heat transfer characteristics
of these liquids are being studied in vorious
systems (sec. 9). Among the physical property
measurements pertinent to the ARE are (1) the
viscosity of the recently designoted ARE fuel
NaF-ZrF ,-UF, (53-43-4 mole %), which decreases
from 13.5 cp at 620°C to 8.8 ¢p of 757°C, (2) the
density of this fluoride, which is 3.5 g/cm’® at
653°C, and (3) its vapor pressure, which ronges
from 4.5 mm Hg at 790°C to 39 mm Hg at 958°C.
The experimental velocity profiles which have
been determined in a convection system differ
significantly from the parabolic characteristic of
isothermal laminar flow, Forced-convection heat
transfer for the NaF-KF-LiF eutectic—inconel
system in 0.175.in.-ID tubes is one half that for
a comparable fluoride-nickel system, probably be-
cause of a corrosion layer found on the Inconel
surface. A mathematical analysis has been made
of the effectiveness of a reactor coolant with

37
ANP PROJECT QUARTERLY PROGRESS REPORT

regard to duct dimensions ond spacing, amount
of heat to be removed, coolont temperature rise,
and coolant physical properties,

The radiation damage studies are primarily con-
cerned with the evaluation of the results of the
fluoride-Inconel somples irradiated in the LITR
and MTR, as well as the effect of radiation on
creep and thermal conductivity, An inpile fluoride
loop is being constructed (sec. 10). The irradiated
fluoride-containing Inconel capsules show an inter-
gronular attack of up to 3 mils, which does not
occur in control samples, Chemical analyses of
these fluoride mixiures indicated uneven distri-
bution of uranium. Cantilever-type creep measure-
ments made on Inconel in a helium atmosphere in
the LITR indicated no serious change in creep
properties as a result of irradiation; also, there
was no significant change in the thermal conduc-
tivity of an Inconel specimen irradiated in the
MTR.

The analytical studies of reactor materials in-
clude the development of chemical, petrographic,
and x-ray anclyses of fuel compositions and/or
identification of fuel corrosion products (sec. 11).
A procedure, which uses BrF as a reagent, has
been developed for the determination of oxygen in
fluorides. The determination of zirconium in fuels
may be made volumetrically by the use of p-

38

bromomandelic acid as a reagent, or colorimetri-
cally with reference to zirconium alizarin sulfonate,
““Tiron'" is shown to be a suitable reagent for the
determination of uranium, The concentrations of
UF3 and Zr° in Nc:l"'-Zrl:4--UF:4 (50-46-4 mole %)
have been determined by the evolution of hydrogen,
upon treatment of the mixture with hydrochloric and
hydroflueric acids, Optical data, from petrographic
examination, are reported for many new fluoride
compounds.

Studies of reprocessing of fluoride fuels indicate
that reprocessing may be accomplished readily in
existing facilities (sec. 12). Although the NaF-
ZrF4-UF4 (50-46-4 mole %) fuel was used in these
studies, the information obtained is generally
applicable to other compositions within this
system. After an aqueous solution of the fluoride
that is suitable for solvent extraction is obtained
by reaction with dilute aluminum nitrafe—nitric
acid solution under reflux conditions, the uranium
can be recovered and decontaminated by solvent
extraction in batch countercurrent runs with losses
of less than 0.01%. Processing of the fuel in this
manner is possible in the present ORNL Metdl
Recovery Plant without major equipment additions,
olthough the corrosion of existing equipment will
be severe and some precautions will have to be
taken to avoid criticality.
PERIOD ENDING JUNE 10, 1953

6. CHEMISTRY OF HIGH-TEMPERATURE LIQUIDS

W. R. Grimes, Materials Chemistry Division

An effort is being made to obtain a better under-
standing of the solid-phase relationships in the
very complicated NaF-ZirF -UF, system because
of its importance to the aircraft reactor program.
Some additional thermal data were obtained, but
the greater number of samples prepared was for
solid-phase studies. A pseudoeutectic has been
shown to exist at or near the composition 65-15-20
mole %, and it has been fentatively designafed as
the ARE fuel concentrate rather than the previously
considered 50-25-25 mole % mixture. Data on a few
compositions in the RbF-Z¢F -UF, system indicate
that this system may possible be of interest as «
fuel material, Because of the current interest in
BeF, pump seals, the change in the melting point
of the 50 mole % NaF—46 mole % ZeF -4 mole %
UF, mixture when various amounts of BeF, are
odded was restudied.

Studies of the phose diagroms of mixtures of
metal and uranium halides, other than those con-
taining UF,, have continved. Several binary
fluoride systems containing UF, have been ex-
amined, but no satisfactory phase diagrams have
yet been obtained. Lack of sufficiently pure UCH,
has hampered the thermal analysis of chloride
system; however, low-melting-point compounds are
indicated in several mixtures.

Modifications of the hydrogenation-hydrofluori-
nation process previously described for the fuel
preparation have been studied on laboratory and
pilot scales. There has been considerable study
of the kinetics of removal of HF from the melts by
stripping with helium and hydrogen, and efforts
have been made to evaluate the kinetics of re-
duction of NiF,, FeF, FeF, CiF; and UF, by
hydrogen in these melts. [t appears that stripping
of the HF by use of hydrogen rather than helium
is to be preferred, since more complete removal of
the structural metal fluorides will be effected at
the same time,

Construction of the equipment for production of
3000 Ib of ARE fuel solvent (NaZrF,} is compleate,
and preliminary testing of the equipment has been
satisfactory. Production of this material is sched-
uled for June and early July.

X-ray diffraction studies have been made on two
fluoride mixtures to establish the compounds and
phases found in the respective melts. The mixtures

studied were NaF-UF, (75-25 mole %) and Nak-
ZrF , (50-50 mole %). Several ternary phase regions
were also examined, and the x-ray data, together
with petregraphic observations for compounds found
in these regicns, are presented.

FUEL MIXTURES CONTAINING UF4

L. M. Bratcher J. Truitt
C. J, Barton
Materials Chemisiry Division

NoF.ZrF -UF . The discovery that UF  sepo-
rates from the mixture NaF-ZrF4»UF4 (50.25-25
mole %) when cooled without efficient stirring made
this mixture undesirable as a fuel concentrate for
the addition of enriched uranium to the ARE, The
need for a nonsegregating composition with a high
uranium concentration was also demonstrated in
radiation damage studies.

X-ray diffraction data for compesitions on the
NaUF ,-Na,ZrF, line indicated that these com-
pounds may form the pseudobinary system for
which thermal data are shown in Fig. &.1. The
data seem to indicate o minimum composition at
about 45 mole % NGUFS (22.5 mole % UF4). How-
ever, petrographic examination of the 40 and 45
mele % compositions showed that both compositions
had the appearance of o eutectic (aggregates of
very fine-grained crystals), and the results of
settling tests (cf., “"Analytical Studies of Reactor
Materials,”” sec. 11) seem to indicate that the
lower concentration of NaUF, is closer to the
sutectic.  This composition (NaF-ZrF -UF ,, 85-
15.20 mole %} is, af present, tentatively designated
as the enriched-uranium composition for ARE
startup.

It seemed undesirable to change the composition
of the uranium-free fused salt (NQZrFS) that will
be circulated in the reactor before the addition of
uranium. Mixing the new uranium-rich composition
with NaZrF o in the proper proportions results in a
fuel with the composition NaF-ZrF -UF , (53-43-4
mole %). This mixture shows a melting point of
220°C, which is approximately the same as that of
the 50-46-4 mole % fuel composition previously
considered. Vapor pressure data for the new fuel
composition are reported in sec. 9, ‘"Heat Transfer
and Physical Properties.’’

39
ANP PROJECT QUARTERLY PROGRESS REPORT

B50 prrereeeeens

 

   

800 |

750 |- -

 

700

TEMPERATURE (°C)

 

650

 

~_DWG. 19903

 

 

 

 

 

 

 

600

 

 

 

 

 

 

 

 

 

 

 

 

 

NaUFg {moaole %)

Fig. 6.1. The System Na,ZrF_-NaUF .

RbF-ZrF4-UF4. The low melting point of RbZrF,,
recently redetermined to be 405 * 5°C, makes this
material seem gttractive as a possible base for a
fuel preporation. Mixtures containing various
proportions of RbZrF . and RbUF . (melting point,
710 + 10°C) were prepared, and data on thermal
effects were obtained from cooling curves (Table
6.1).

Thermal data for the 92 mole % RbZrF ~8 mole
% RbUFS mixture were checked on samples pre-
pared in sealed capsules. This material showed
a halt only at 402°C, which is the lowest melting
point yet recorded for o nonberyllium fluoride
mixture containing as much as 4 mole % UF ;. For
mixtures with higher uranium concentrations, the
rather sharp rise in the melting point probably
indicates the formation of a higher melting complex,
such as Rb3UF7 (melting point, 990°C). Further
work on this system will be undertaken as the
supply of RbF permits

NaF-ZrF ,-BeF,-UF,. Data on mixtures resulting
from the addition of BeF, to the binary compo-
sition with NaF.-ZrF UF . (50-46-4 mole %) were

40

TABLE 6.1. THERMAL EFFECTS FROM
COOLING CURVES FOR MIXTURES OF
RbZrFs AND RLUF,

 

 

 

COMPOSITION THERMAL EFFECTS
(mole % RbUF ) (°C)
4 400, 385 (halt)
8 400 (hal1)
12 560, 400 (halt)
16 505, 400 (halt)
20 575, 400 (halt)

 

 

given in a previous progress report;(? however,
o more detailed study of such mixtures was carried
out during this quarter, The thermal effects noted
are given in Table 6,2, At the higher BeF, con-
centrations, the thermal effects were not well
marked on the cooling curves. The dato indicate,

 

My, s, Coleman and W. C. Whitley, ANP Quar. Prog.
Rep. Sept, 10, 1952, ORNL-1375, p. 78.
however, that liquid probably exists in mixtures
with 60 or 70 mole % BeF, at as low as 430°C,
The uranium concentration in these mixtures is
very low, and the data are of interest only in

TABLE 6.2. THERMAL EFFECTS FOR MIXTURES OF
NaF-ZrF4-UF4 {50-46-4 moie %) AND BeF2

 

 

CONCENTRATION | THERMAL EFFECTS FROM
OF BeF, COOLING CURVES
{mole %) (°C)

1 510 (halt), 490

2 520 (hal?)

5 493, 445

10 476 (halt), 435, 365
20 465 (halt)
50 602, 543, 444 {halt)
60 505, 430
70 620, 430
80 880(?), 415

 

 

PERIOD ENDING JUNE 10, 1953

connection with the possible use of BeF, in pump
seals,

FLUORIDE MIXTURES CONTAINING UF3
V.S, Coleman C. J. Barfon

Materials Chemistry Division
T. N. McVay, Consultant, Metallurgy Division

Thermal analyses of binary mixtures of NeF-
UF, and KF.UF,, supplemented by examination of
the solid phases, have not, to date, provided satis-
factory information for the preparation of phase
diagrams of these systems, A number of compo-
sitions in other systems were prepared to aid in
the identification of the solid phases occurring in
reduced zirconium-base fuels. The complex com-
pound 27rF -UF,, mentioned in a previous re-
port,® has been definitely identified. Table 6.3
shows the thermal data obtained from cooling
curves and the optical data obtained from petro-
graphic examinations for severa!l preparations in

the UF -UF ;-ZrF , system,

(2w, C. Whitlay, V. S. Coleman, and C. J. Barton, ANP
Quor. Prog. Rep, Dec. 10, 7952, ORNL -1439, p. 109.

TABLE 6.3. THERMAL AND OPTICAL DATA FOR UF:;«Z:-FJ1 AND UF -ZrF, MIXTURES

 

OPTICAL DATA

 

MOLECULAR COMPOSITION | THERMAL EFFECTS (°C)
UF 4-27:F 715, 631

UFS-ZrF4

UF UF,-2Z¢F 730

UF -UF -4 ZF, 742

UF -3UF,-8ZrF, 712

UF -UF  Z/F, 735

AUF -UF .. Z:F, 847

3UF -8UF 5 ZeF , 784, 773, 600

15% UF ,~35% UF ;-50% ZrF 725, 590

 

Orange-red compound; RI* = 1,556

Yellow and brown crystals; Rl = 1.616; free UF,

Predominantly brownish, some olive drab

Predominontly olive drab; biaxial +, alpha = 1.556,
gamffld = ].568, 2V = 80 deg

Mixture of olive-drab and red-orange striated crystals

Brown isotropic phase; elpha = 1.560, gamma = 1.568;
same UF3

Predominantly UF4 surrounding brown-orange

crystals of Rl = 1.56 %0 1.57

Homogeneous clive-drab phase different from the 1:1:4
preparation; uniaxial +, plecchroic, algha =1.57Q,
gamma = 1.578

Predominantly olive drab, trace of UF 4 and brown-

orange crystals

 

 

*Refractive index.

41
ANP PROJECT QUARTERLY PROGRESS REPORT

CHLORIDE MIXTURES CONTAINING UCl4
R. J. Sheil C. J. Barton

Materials Chemistry Division

A considerable part of the work with uranium
tetrachloride during the past quarter was devoted
to efforts to obtain pure material. Neither sublis
mation of product from hexachloropropylene chloris
notion of UO; nor from chlorination of UH, pro-
duced material of the desired purity., Although the
accepted volue for the melting point of UCIH, is
590°C, the highest melting point obtained to date
from these preparations was 567°C. Plans have
been made to procure pure UCl, and pure UCI,
prepared by vapor-phase chlorination of UO; with
ccl, (and reduction of the UCld with Hz) from the
Y-12 plant,

Although one significant experiment in the KCi-
UCl, system is reported below, little study of the
NaCl-UCl, aond KCI-UCl, systems waos attempted
during the quarter, This work will be resumed for
final exploratory checking when sufficiently pure
UCI, is available. Studies of the LiCI-UCI,,
TICI-UCH,, and LiCI-KCI-UCI, systems were in-
itiated during this period. The best material avail-
able was used.

KCI-UCi,. Thermal data obtained in this labo-
ratory$3) indicated a eutectic between K,UCI, and
KUCI which melted at 320 + 10°C, while the
previously published(‘” diagram for this system
shows a melting point for this composition of about
600"C. To ascertain whether a solid phase sepao-
rates from the melt without giving o detectable
thermal effect, a mixture containing 57 mole %
KCl and 43 mole % UCI,, the approximate eutectic
composition, was heated to 355°C, and a portion
of the liquid was drawn by suction through a fritted-
glass filter. |t was not possible to filter the liquid
at lower temperatures, probably because of its high
viscosity. The comparison of observed and calcu-
lated volues for composition of the liquid are
shown below:

Found (%) Calculated (%)
Potassium 10.9 10.8
Uranium 48.8 48.1

 

(B)R. J. Sheil and C. J. Barton, ANP Quar. Prog. Rep.
Mar. 10, 1953, ORNL-1515, p. 108.

(4)C. A, Kraus, Phase Diagrams of Some Complex Salts
of Uranium with Halides of the Alkali aond Alkaline
Earth Metals, M-251 (July 1, 1943).

42

The agreement between the calculated and the
analytical results shows that no significant amount
of material separated from the melt at 355°C and
indicates that the previously accepted value for
the melting point of this composition is erroneous.

LiCi-uct,. Thermal data for compositions in
the Lill-UCl, system containing 25, 40, and 75
mole % UCl, were reported in a previous study, (4}
On the basis of the few compositions studied, it
was reported that there were no compounds and
that the eutectic temperature was 410 to 430°C,
The thermal data obtained in this laboratory with
the use of impure UCI, prepared by hexachloro-
propylene chlorination of UO, are shown in Fig.
6.2. The data are considered tentative and subject
to revision when purer UC|4 becomes available.
The values obtained thus far ogree reasonably well
with those reported in the previous study, but they
seem to indicate that two eutectics melting at
415°C (approximately 29 mole % UCI,) and 405°C
(approximately 48 mole % UCI )} exist in the sys-
tem. The compound Li,UCl, appears to melt con-
gruently ot 430 + 10°C; however, the flatness of
the liquidus curve suggests that this compound is
rather unstable, The existence of the compound
is not yet supported by petrographic examination,
since compositions in this region were too poorly
crystallized to permit microscopic characterization
of the solid phases.

TICI-UCI,. On the basis of data obtained to
date, it appeors that there is a eutectic in the
TICI-UCI, system at approximately 16,5 mole %
UCI, which melts ot 362 £ 10°C and a congruently
melting compound TIZUClé which melts at about
480 + 10°C. The data are subject to the reser-
vations stated above regarding purity of materials.

KCI-LiCl-UCl,, The five compositions studied
to date in the KCI-LiCl-UCl, system were prepared
by adding increasing amounts of UCI, (HCP
product, as received) to the KCI-Lill eutectic
(41 wmole % KCI; melting point, 352°C), The ad-
dition of 2 mole % UCI, to the binary eutectic
raised the melting point about 16 deg, and subse-
quent additions of up to 20 mole % raised the
melting point further. For UCl, concentrations
of up to 10 mole %, the cooling curves indicated
the probable existence of a eutectic whose compo-
sition is not yet known which melts at 345°C and
contains a very low concentration of UCIH,, At the
higher UCl, concentrations (30 and 35 mole %
PERIOD ENDING JUNE 10, 1953

DWG. 19304

 

700

 

 

 

 

 

 

 

 

 

 

 

 

e :

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

600 —
5QQ oot
5 -
Lud
o
;{3 o
<( -
E 3OO b e e el
Lt
a
=
L)
o
300 1 . et
>
200 e 1 8 -~
o
2
o
1001 e e
Licl t0 20 30 40

 

50 60 70 80 90 ucl,

UCI4 (mole %)

Fig. 6.2. The System LiCl-UC|4 (Tentative),

UCl,), the cooling curves showed halts at approxi-
mately 275°C. The halts are probably due to a
high-uranium eutectic in the vicinity of the KCl-
UCI, binary eutectic (43 mole % UCl,; melting
point, 320 % 10°C). Work on this system is con-
Yinuing.

COOLANT DEVELOPMENT

L. M. Bratcher V.S, Coleman
C. J. Barton
Materials Chemistry Division

Most of the work with coolants during the past
quarter was devoted to checking some of the
results of earlier work and to preparing possible
compounds for petrogrophic ond x-ray diffraction
examination. An effort was also made to extend
the data on the binary ZrF, systems to higher
ZrF, concentrations through the use of sealed
capsules. It seems likely that a part of the earlier
data on systems containing KF, RbF, and CsF
with ZrF, was erronecus because of the lack of

sufficient care in keeping water out of contact
with these hygroscopic materials, Work on a
number of systems is still in progress and will
be reported at a later date.

NoF-ZrF,-BeF,. Although some data on the
NaF-ZrF ,-BeF, system were reported earlier,
renewed interest in the system as a result of the
recent use of BeF, in pump seals made further
work with compositions in this ternary system
desirable, The lowest melting point found was
345°C for a mixture containing 15 mole % ZrF,
42.5 wole % NaF, and 42.5 mole % BeF,. Halts
on the cooling curves were observed with this
composition and other compositions in its vicinity
at temperatures ranging from 290 to 320°C. It is
not clear from the avaiiable data whether this
thermal effect is due to a solid transition or fo a
evtectic of unknown composition. On the NaZrF .-
BeF., line, the lowest melting point observed in
the range of 4 to 45 mole % BeF, was 487°C at
12.5 mole % BeF,.

43
ANP PROJECT QUARTERLY PROGRESS REPORT

Lif-ZrF,. Preliminary data on the LiF-ZrF,
system were reporfed in a previous progress ree
por‘r.(s) Most of the early data have been subse-
gquently proved to be incorrect, prohably because
of the presence of oxide or oxyfluoride in the
mixfures. 1he minimum melting point in the system
was reported more recently.(®)  Publication of an
equilibrium diagram for the system has been de-
layed by an apparent conflict between the thermal
data and the results of petrogrophic and x-ray
diffraction identifications of the solid phases in
the 75 mole % LiF—25 mole % ZrF, mixture. Since
it appears that a rather detailed examination of
this region will be necessary to clear up the dis-
crepancy, the thermal data are presented in Fig.
6.3 as a tentative equilibrium diagram,

 

(3)y. P. Blakely, L. M. Bratcher, R. C. Traber, Jr.,
ond C. J. Barten, ANP Quar, Prog. Rep. Mar. 10, 1952,
ORNL-1227, p. 104.

(6). M. Bratcher and C. 1. Barton, ANFP Quar. Prog.
Rep. Dec. 10, 1952, ORNL-1439, p. 112.

 

The liquidus line shown in the diagrom indicates
LijZrF, to be a congruently melting compound,
but the solid material of this composition was
found to contain Li,ZrF,, LiF, and a crystalline
phase of lower refractive index than Li,ZrF .
This latter phase has not been prepared in pure
form, but it predominates in the mixtures con-
taining 15 and 20 mole % ZrF,. The thermal data
also show some indication of a compound con-
taining more than 1 mole of ZrF, per mole of LiF.
An unknown phase, in addition to Li,ZrF, and
ZrF,, was reported to be present in the x-ray
diffraction pattern of the 67 mole % ZrF , mixture,

DIFFERENTIAL THERMAL AMNALYSIS
R. A. Bolomey, Materials Chemistry Division

Experience during the quarter with the differ-
ential thermal analysis apparatus has shown that
both the precision and the accuracy of the data
obtained with this apparatus are increased by

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

1000 B DWG. 19905
900 e e— . e e ——— — - — _— - fg.-—i,--%b
/
/
¢ | //
800 |-—- 1--— _— e
N P
o
— 7
O
2.
E:J 700 ——— O e & —1 e
=
=
< O
o
o ™ ;
= / o
w 600 f----———— S N | B ] f“——"f—J‘I— —
1 \po by <
. 5 . L)
o™ . !L- Cromeer s (ymrroonl) rarmeums el smmen
300 00 o o0 o o o L rTTT T T
o
400 e e
300 . v —
LiF 40 50 &0 70 80 90 Zrf,
ZrE  (mole %}

4

Fig. 6.3, The System LiF-ZrF, (Tentative).
referring the differential temperature to the sample
temperature rather than to the Al /O, reference
material, Further improvement might be effected
by using reference materials which have the same
thermal properties as those of the sample but
which do not exhibit transition temperatures in
the temperature range of interest,

PRODUCTION AND PURIFICATION OF
FLUORIDE MIXTURES

F. F. Blankenship G. J. Nessle
Materials Chemistry Division

Removal of HF from Fuel Batches (C. M. Blood,
F. P. Boedy, R, E. Thoma, Jr., Materials Chem-
istry Division). Poor corrosion performance of the
fluoride mixtures could be caused by HF that re-
mains in the finished fuel, Although there is no
reason to believe that HF is appreciably soluble
in NaZrF, at 800°C, further experimentation to
check this point ond fo evaluate the rate of re-
moval of HF by stripping techniques were per-
formed during the quarter. Generally speaking,
vacuum techniques for removing the HF must be
avoided because of the risk of introducing air and
water vapor through undetected small leaks.

The HF content of gas which has been bubbled
through the molten salts is best measured by the
method of White and Manning{?) i which the con-
ductivity of a dilute HBO, solution is measured
as g function of the HF it cbsorbs from the strip-
ping gas. In this way, the concentration of HF
in the effluent gas can be followed almost con-
tinuously,

it has been shown that HF can be removed from
the molten salts ot 850°C much more rapidly than
from the unheated valves, gages, traps, and con-
nections of the purification assembly, Conse-
quently, numerous blanks have been run ic ascer-
tain the rate of removal of HF from empty as-
semblies under various conditions; these runs
served only to indicate that the lost fraces of
adsorbed or trapped HF are almost impossible to
remove, Diminishing returns were reached at a
level of about 107> mole of HF per liter of strip
gas when routine procedures were applied to an
empty apparatus with the usual lines aitached,
and 107% mole of HF per liter of strip gas was
regarded as close to the limit of removal.

 

(7)J. C. White and D. L. Manning, ANP Quar. Prog.

Rep. Mar. 10, 1953, ORNL-1515, p. 173.

PERIOD ENDING JUNE 10, 1953

A mathematical analysis (L. Alexander, private
communication) based on models resembling the
system showed that a first-order process can be
expected. |t o first-order process is actually
followed, the amount of HF remaining in the melt
at any time during the stripping operation can be
calculated and then combined with the equilibrium
vapor concentration to provide the Henry's law
constant for solubility of HF in the melt.

When helium is used as the stripping gas, the
concentration of HF in the effluent gas decreases
with time in a manner agpproximating that expected
for a first-order process. When 2 kg of NaZrF
is stripped at 800°C with a 5-cm bubble path, the
half life appears to be 10 liters or less, Larger
values seem to be due to siow removal of residual
and adsorbed HF from cooler sections of the
apparatus; with the most rigorous efforts to avoid
extraneous contributions, half life values between
5 and 10 liters have been obtained with NaZrF ¢
as the melt. If 10 liters is used as the half life
volume and 0.6 as the fraction of equilibrium con-
centration maintained in the effluent gas, an upper
limit estimate of 0.25 mole of HF per liter of strip
gas per mole of HF per kg of fuel is obtained for
the Henry’s law constant. When the stripping rote
is 1 mmole of HF per liter of strip gas, the equis
librium concentration is 4 mmoles of HF per kg
of fuel; this amount of HF would produce about
80 ppm of CrF, by the reaction

2C¢° + 6HF‘::A"2CI‘F3 + 3H2 .

The conclusion has been reached that melts
siripped to a concentration of 1074 mole of HF
per liter of strip gas connot contain HF in amounts
sufficient to cause appreciable corrosion; pre-
cautions must, of course, be taken to avoid recon-
tamination of the melt from tramp HF from other
portions of the system,

When hydrogen is used to remove HF from the
molten salts, the stripping operation is compli-
cated by the production of HF by reduction of
various fluorides. The differences observed in
the use of hydrogen ond helium for stripping o
uranium-containing fue! in Fig. 6.4,
which shows that helium stripping occurs at a
rate corresponding to a half-life volume of about
20 liters but that the change to hydrogen results
in the production of HF ot a higher rate, The
steeper portion of the hydrogen curve dppears fo
be related to the reduction of structural metci

are given

45
ANP PROJECT QUARTERLY FROGRESS REPORT

 

  

0_3 DWs. 19908
3 e e e e e ws. 19208
{ T T

j i .
: [ 40-liter HALF LIFE | | |
5 . ] u ! L . . ! ‘
| | J . |
o UG + Y H,m URg+HF ]
( _CHANGE FROM H, \
2 |- T TO HELIUM ‘ -

 

|

|

N |
5/ \ ‘{‘--*HZ TREATMEPJTM-—E =

|

MOLES OF HF per liter CF STRIP GAS

T } o , ! o Lo

P 3kg OF 50 50 NoUFy-NeZrFy AT 700°C [
5-in. BUBBLE PATH i

2o e I . ; T ] e s \

{ f | | | |

[ TR I ‘7,, L } J

50 100 150 200 250 300
STRIP GAS (liters)

 

 

Fig. 6.4, Stripping of Hydrogen Fluoride from a
Molten Fluoride with Helium and Hydrogen,

ions; a leveling takes place at about 1.8 x 1074
mole of HF per liter of gas, which corresponds to
the rate of reduction of UF, to UF ..

In the larger scale apparatus for treating 50-lb
batches of material, the flow rate has been in-
creased from 1 to 5 liters/min, and the bubble
path is 10 em. Stripping rates from the 50-lb
apparatus are about the same as those for the
smaller rigs, Experimental data for four runs in
the 50-1b apparatus, in which hydrogen was used
for the stripping operation, are shown in Fig. 6.5.
It appears from these data that HF concentrations
in the gas phase can be brought to moderately low
values by using hydrogen, even in large apparatus,
without excessive stripping times,

Reduction of Metal Fluorides with H, (C, M
RBlood, F. P. Boody, R. E. Thoma, Jr., Materials
Chemisiry Division). Nicke! is one of the metals
that is resistant to HF ot high temperatures be-
cause of a protective film of NiF, that forms on
the metal. However, when nicke! is used to con-
tain a melt capable of dissolving this film, the

46

3"
x

5\

- _!u
|

—
|

. _!
|

]
|

~
|

—

i

 

 

2 _k b l - |1 S - _ L [
(MELTS PREVIOUSLY TREATED WITH
HF FOR TWO HOURS. NICKEL CONTAINER)

\ REACTION MONITORED BY MEASURING

 

 

 

o A RATE OF HF EVOLUTION }
L A .\ 815°C 10-in. BUBBLE PATH
- ! i [ i
o | I
= TN i \
g | \e | | |
¥ o - e \g o [ | | -
{ o\ * |
Ué I T N Li, . - _
- . | |
l_j ,,,,, 17 A RO Afl J‘,io e l —— | ‘
g | ~N !
2 NG e
| AN |
R R S N
: \ AN A
I f Ak \lr ’
5 e s b i 1 — F— = {\ — !
‘ : ‘ ’ t
axyo L - j 7,\,,,_,7,fi,._% | A
200 400 800 800 1000 1200
H2 {liters)
Fig. 6.5. Stripping of Hydrogen from Hydrogen

Fluoride in 4- to 50-Ib Batches of NuF-ZrF4-UF4
(50-46-4 mole %).

metal is slowly attacked by MF to the extent that
0.1% or more of Ni' " can be dissolved by the melt
during a 2-hr HF treatment at 800°C, Unless this
Ni** is subsequently removed, the me!t is quite
corrosive to Inconel because of the nearly gquanti-
tative replacement of Ni*" by chromium from the
metal ,

Accordingly, several attempts have been made
to study the rate of reduction by hydrogen of
structural metal fluorides in NaZrF . In general,
the NaZri, contained in graphite was siripped
with hydrogen until the HF concentration in the
exit gas concentration reached 107 mole/liter,
After the melt had cooled, the predetermined
quantity of metal fluoride was added. The sample
was reheated to B0O0°C while being stirred with
helium; hydrogen was then readmitted and analysis
of the exit gas was begun,

Data from some preliminary runs with NiF, are
shown in Fig. 6.6, The trial run with T wt % Ni*"
indicated that a clear-cut first-order reaction was
involved., However, the data shown by 0.1% ad-
ditions were poorly reproducible and did not agree
well with those obtained with the higher concen-

tration. In most cases, the integrated total yield
N b |
. { e e g

tm{i\ { NiF, + H af\fu + 2HF \ I

romrmea e et R T L _ ,,,,J..___,,,, - dmme s ermmesaan .. |

     
 

 

10’24% { .
| i 10wt % Nitt
f , 35-liter HALF LIFE _
4
1 5 40 -liter f 3-kg BATCH AT 800°C ’
HALF LIFE

_ 5-in. BUBBLE PATH
- o
e 2’;; T it

A .
f \#
% .
i Y J .
\

 

MGLES GIF HF per fiter OF HYDROGEN

 

 

 

4 &0 -liter a;\ j
T HALFLIFE e 1 S
\
f \\ . }
\L a
l RN ‘ .
1G5 _‘A"fiAw‘_LMN~
50 100 150 200 250 300
H2 (iiters)

Fig. 6.6. Rate of Reduction of NiF in NaZrF by
Hydrogen.

of HF was about 25% low; this is well beyond
the limits of accountable experimental errors, It
seems likely that some reduction by the graphite
from the container is responsible for these dlff;-
culties.

Preliminary data for other structural metal fluo-
rides reveal that Fe*" is reduced about four times
less rapidly than Ni** and Cr*** about 10 times
less rapidly if equal quantities of these ions are
present during the usual fuel-production procedures,
in practice, no significant quantity of chromium is
present in the starting materials, About 500 to
700 ppm of iron is present and, as described above,
significant concentrations of mckel are added from
apparatus,

Data from analyses for iron, chromium, and nickel
in the finished fuel are plotfed as a function of
hydrogen treatment in Fig. 6.7. It is apparent that
substantial benefits are obtained by hydrogen
treatment for reasonable mfervals of time,

[t has been established that the reduced metuls
tend to accumulate as spongy masses on the
hydrogen delivery line, on the vessel walls, and
at the surface of the melts. Although the accumu-
lation is valuable in prolonging filter life, it may
prove to be a serious problem in equipment which
cannot be opened for cleaning after preparation of

the molten mixture. To date, however, there is no

PERIOD ENDING JUNE 10, 1953

    
     
 

e BWE, 19908
T

 

 

[¥p]
- £
£ TG
a =
a <
= &
o) oy
W Q2

L
= o |
> - 2
a uwi

=
2 3

 

 

 

 

 

 

 

 

HYDROGEN STRIPPING (hr)

Fig. 6.7. Hydrogen Reduction of Structural Metal
Fluorides in NaF-ZrF -UF, (50-46-4 mole %) at
1500°F.

evidence that continued re-use of the equipment
leads to less pure fuel mixtures.

Attempts have been made to prepare very pure
fuel by using graphite-lined nickel equipment. A
batch of NaZrF. so handled and given 20 hr of
hydrogen treatment showed 30 ppm of iron and less
than 20 ppm each of chromium and nickel.

A more thorough study of the effect of UF,
concentration on the rate of reduction of UF4 by
hydrogen is planned. The preliminary indication
from trials with compositions containing 20 ond
25 mole % UF, was that the rate of reduction is
approximately the same gs with 4 mole % UF or,
in other words, that the reaction is of zero order
with respect to UF ; concentration,

Pilot-Scale Fuel Purification (G. J. Nessle, J. P.
Blakely, J. E. Eorgon, Materials Chemistry Di-
vision). During this quarter a fotal of 318.9 kg of
mixed fluorides was processed. This quantity of
fluorides comprised 12 small batches of approxi-
mately 2 kg eoch and 11 large batches of approxi-
mately 27 kg each.

47
ANP PROJECT QUARTERLY PROGRESS REPORT

The gas treatment of the fluorides during prepa-
ration has been altered to effect a more thorough
removal of retained HF in the melt and to yield a
final product lower in metal impurities. The treat-
ment consists of the following steps:

1. melt fluoride charge under HF atmosphere,

treat with hydrogen for 1 hr,

treat with HF for 11/2 hr,

flush with hydrogen for 3 to 4 hr,

transfer molten fluorides to receiver with helium,

flush with helium during cooling and solidifi-

cation.

This oltered procedure substitutes a hydrogen
flush for the helium flush; hydrogen will strip out
HF as efficiently as helium aond, in addition, will
reduce more of the metal impurities. With the
uranium-bearing mixes, the longer hydrogen flush
may produce a small amount of UF,. Careful con-
trol of the amount produced will be necessary to
avoid precipitation of the UF,.

An aodditional change in the treatment procedure
is the use of a continuous flow of gas of approxi-
mately 5 liters/min during the hydrogen-flushing
period.,  Previously, intermittent flow had been
obtained by pressurizing to 10 Ib and then bleeding
off to atmospheric pressure in a 2-min cycle.

The data concerning the preparation of niixed
fluorides in the pilot-scale facility during this
quarter are given in Table 6.4.

Fluoride Production Facilities (G. J. Nessle,
Materials Chemistry Division). The installation of
process equipment to be used for production of the
nonuranium fluoride base for the ARE fuel has been

*

O hWwho

completed, and preliminary tests are under way.
The first ““dry run’’ in one of two process cubicles
proceeded with no difficulty, ond data were ob-
tained concerning heating time and gas flow; no
leaks were found either before or after the test,
During a second dry run in this cubicle, a short
circuit developed in the receiver furnace and
necessitated a complete shutdown for two weeks
for installation of new elements in the furnace,

A dry run for the other process cubicle is sched-
uled in the near future. It is hoped that some
valuable information concerning the behavior of HF
in the process equipment can be obtained simul-
taneously with the trial runs.

The installation of the process equipment to be
operated by the Y-12 Production Division for the
production of the enriched-fluoride mix for the ARE
is nearly complete. Arrangements have been made
to train the personnel directly concerned with this
operation. This short training program will probably
be started during May.

PREPARATION OF COMPLEX FLUORIDES

B. J. Sturm L. G, Overhol ser
Material s Chemistry Division

Methods of preparation and partial listings of
properties of complex fluorides of the types re-
sulting from interaction of alkali and structural
metal fluorides have been reported previously, (8
During this quarter, additional batches of the

 

@) ANP Quar. Prog. Rep. Dec. 10, 1952, ORNL-1439,
p. 115,

TABLE 6.4, PILOT-SCALE FLLUORIDE PREPARATIONS

 

 

 

 

COMPOSITION
MIXTURE per cent ppm pH
F U Ze Ni Cr Fe
NaF-ZrF ;-UF , (46-50-4 mole %)% 42.1 8.52 | 38.8 125 20 375 2.03
NaF-ZrF -UF , (50-46-4 mole %){®) 42.3 8.92 38.0 135 30 345 2.32
NaF-ZrF , (50-50 mole %)) 43.8 (d) 43.2 37 20 245 2.38
NaF-ZrF ;-UF ; (50-25-25 mole %){¢) 33.6 | 41.2 16.3 20 20 40 2.92
NaF -UF , (50-50 mole %){) 27.3 | 67.4 20 120 190 2.85

 

 

 

 

 

 

 

 

(a)Average of 4.

Average of 16, (e)

<) Average of 3.

43

(d)Nof reported.
Average of 1.
complex fluorides, as well as batches of the
simple structural metal fluorides, were prepared.
These materials were identified or characterized
by the use of chemical analyses, x-ray examination,
and crystallographic data. '

The foliowing list gives the materials prepared
during this quarter. Similar batches of some of
these materials were prepared previously by this

group, but some other batches represent new
syntheses made during this period:

NiF, KNiF,

Na.FeF CsNiF

BT N tentatively identified

LiCrF RbNiF,

CrFy NiCl,

NaCrF, FeCl,

Fusions were made that consisted of NaZrF
with each of the following: CF, NoCrF, and
NoCrF,. In each case, the chromium compound
lost its identity; x-ray analysis indicated a possi-
ble solid solution of Na CrF, and NaZrF ofter
fusion.

Fusions of mixtures of KF and MnFB, corre-
sponding to KMnF ,, K, MnF , KBM”F(H and K MnF_,
all yielded compounds with the same x-ray pattern,
The product approximated K,MnF, which may be
a solid solution of KMnF ; and KMnF .

X-RAY DIFFRACTION STUDIES
P. A. Agron, Chemistry Division

In order to resclve some of the complexities in
the NaF-ZrF -UF, system, examinations have been
made of several incongruently melting binary com-
pounds and a number of ternary regions. '

NaF-UF, (75-25 mole %). The binary region in
the vicinity of the NaF-UF (75-25 mole %) compo-

PERIOD ENDING JUNE 10, 1953

Cooling curves, aleng
with x-ray examination of the solid phases, defi-
nitely established that Na,UF is an incongruently
melting compound. The tetragonal N03UF7{9) ap-~
pears to be stabilized by
zirconium ion,

sition was investigated.

small additiens of
Another phase, which may be an
allotropic form of Na,UF,, will be investigated
further. '

NoF.ZrF, (50-50 mole %). Further study of the
NaF.Z:F, (50.50 mole %) region indicates strongly
that NaZrF . has an incongruent melting point and
Preparations of this
binary salt generally yield the normal hexagonal
compound(m) on cooling the melt with adequate
stirring. On examination of large-batch prepardtions
of NaZrFs, the appearance of an unknown phase is
marked in many instances. Two new phases in
the NqF-ZrF4 (50-50 mole %) composition region
have been observed in quenching experiments aft
600°C and on fusing small quantities of NiF, and
CrF, with normal NquFg. Further work will
definitely establish the nature of these two new
crystal structures. The phase induced by rapid
observed earlier ond erroneously
aftributed to o reduced form of zirconium fluoride.
This unidentified phase was found to be unstable
when exposed to atmospheric conditions. The lack
of maintenance of equilibrium

the attendont complexities,

cooling was

conditions on
possing through the incongruent freezing point is
doubtless responsible for the muitiplicity of phases
found in the solid. *

 

W’W. H. 2aahariasen, The Crystal Structure of NQ3UF7,
AECD-1798 (Mar, 3, 1948). :

(10)p  pgrom, ANP Quar. Prog. Rep. Sept. 10, 1952,
ORNL-1375, p. 86.

49
ANP PROJECT QUARTERLY PROGRESS REPORT

7. CORROSION RESEARCH

W. D. Manly, Metallurgy Division
W. R. Crimes, Materials Chemistry Division
H. W. Savage, ANP Division

During this quarter, the static and seesaw test
facilities have been used almost entirely for the
study of the corrosion of Inconel by a proposed
ARE fuel mixture, NaF-ZrF -UF . Among the
corrosion parameters examined in 100-hr tests at
1500°F, were exposure time, crevice corrosion,
container cleanliness, and the effect of various
additions, including potential corrosion inhibitors,
Neither
crevices nor container cleanliness seems to have
a significant effect on corrosion. The corrosion
rate decreases with time and is shown to be «
function of the fluoride contaminants, particulorly
Nif and FeF,. Hard-facing alloys of the Stellite
type thot were deposited on Inconel and static
tested in the fluorides at 1500 and at 1300°F were
rather severely aottacked,

Several corrosion tests have been completed with
fluorides in Inconel on the NACA type of rotating
apparatus. |his type of apparatus is advantageous
becouse velocities can be obtained that are higher
than those available in thermal convection loops or
in the seesaw tests; however, no large temperature
gradient can be attained and the flow of liquid is
not continuous. In tests conducted so far ot 1500°F
for 100 hr, there has been no indication of plugging,
and the corrosion is similar to that obtained in the
static tests, even though a flow velocity of 10 fps
is employed.

as well as fuel-processing contaminants.

Fluoride corrosion studies in dynamic systems
provided by thermal convection loops have been
continued, The loops are normally operated for
500 hr with a hot-leg temperature of 1500°F, Most
of the work during this quorter was with Inconel
loops circulating the fluoride NaF-ZrF -UF
(50-46-4 mole %), which was the proposed ARE
fuel. It has been shown that the depth of attack
increases quite rapidly during the first 250 hr of
operation but that thereafter the attack is essen-
tially constant, These effects are also found
with the fluoride NaF-ZrF , ~ the ARE fuel carrier -
but the attack is not so deep. The depth of attack
is not temperature sensitive; only a slight increase
in depth occurred when the temperature was in-
creased from 1250 to 1600°F, The type of attack,

50

The depth of attack was
reduced when the small amount of weld scale found
on the loop was removed by precleaning with
NaF-ZrF , (30-50 mole %). Some purification of the

fuel is desirable,

however, does change,

but the present production-
purification facilities seem to be adequate. It was
shown that the depth of attack is a function of the
fuel and not of the diffusion of chromium from the
solid-solution [nconel, since two batches produce
almost twice the penetration of ¢ single batch.
Addition of NiF, to the fuel increased the depth of
attack.,  Chromium metal added to the fluoride
NaF-KF-LiF-UF , (10.9-43.4-44.5-1.1 mole %) had

no effect on the attack.

In a nickel loop which circulated the fluoride
Nc1i7:-Z:'F4-UI::4 (50-46-4 mole %) there was practi-
cally no attack; however, a small amount of metallic
mass transfer was found. Two type 316 stainless
steel loops were operated for 500 hr with this
same fluoride composition; these loops showed
considerable mass transfer but no plug formation.

The study of the mass transfer of various metals
in high-purity lead is being conducted in quartz
thermal convection loops. So far, the following
container materials have been examined: Inconel,
molybdenum, columbium, Armco iron, and types 304,
347, and 446 stainless steel. In these tests, molyb-
denum and columbium did not show mass transfer,
while type 446 stainless steel showed a slight
amount of mass transfer, The remaining metals
showed extensive mass transfer and premature
plugging of the thermal convection loops.

The coirosion resistance of treated beryllium
oxide specimens has been studied in both static
and dynamic corrosion tests,  The corrosion
mechanism is, apparently, only the mechanical
erosion of the beryllium oxide surface. Beryllium
samples plated with chromium and nickel have been
tested in sodium contained in Inconel tubes. The
plating did not adhere to the specimens. Several
Carboloys have been tested in sodium and lead at
1500°F and, with the exception of Coarboloy 608,
these alloys appear to have good corrosion re-
sistance.
FLUORIDE CORROSION OF INCONEL N
STATIC AND SEESAW TESTS

H. J. Buttram C. R, Croft
R. E. Meadows

Material s Chemistry Division

D. C. Vreeland L. R. Trotter
E. E. Hoffman J. E. Pope
Metallurgy Division

The test conditions of the static tests in which
the Inconel specimen is sealed with the fluoride
mixture in an Inconel capsule are 100 hr at 816°C,
unless otherwise stated in the following sections.
The conditions of the seesaw tests in which the
copsule containing the specimen and the fluoride
is rocked in a furnace are 4 cpm with the hot end
of the capsule at BO0°C and the cold end ot 650°C.

Effect of Exposure Time. Although a large
number of studies have been reported in which
fluoride mixtures in capsules of Inconel have been
exposed in the tilting furnace for intervals of up to
500 hr, very little information has been available
on the course of the attack during the first few
hours of such exposures. To supplement previous
data, copsules containing the luoride NaF-ZrF -UF
{50-46-4 mole %) with and without zirconium hydride
odditions hoave been exomined ofter exposures of
1, 5, and 10 hours. '

Metallographic examination of the capsules with-
out zirconium hydride after these short exposures
indicated only very slight attack, Chemicaol analy-
sis of the fuel after tests (Table 7.1) revealed that

PERIOD ENDING JUNE 10, 1953

the chromium content increased in direct proportion
to the exposure time during these short exposures
but that the iron and the nickel contents of the
material dropped to nearly constant values after
5 hours.

In the tests made with zirconium hydride added
to the fuel, somewhat less corrosion was found
upon metallographic examination. With zirconium
hydride added to the fuel, the iron and nickel com~
pounds ore apparently reduced guite rapidly, and
they remain at low concentration levels for the
durotion of the tests., The linear increase in
chromium content with exposure time is prevented.
In previous experiments“? with fuel to which
zirconium hydride had been added, the chromium
content increased by about 100 ppm in the interval
between 100 and 500 hr of exposure.

Effect of Hydrogen Fluoride. The value of
previous work(?2) on the effect of HF on the cor-
rosion of ARE fuel was somewhat in doubt because
of the possibility that the presence of UF, in the
melt might have, according to the reaction of
2UF3 + QIHF:W_T“2UF4 + H,, masked the dele-
terious effect of small additions. Additional studies
have been performed with a mixture which should
have contained fess UF, than was present in the
previous mixtures. To produce hydrogen fluoride

concentrations of 0.016, 0.032, 0.097, and 0.29%,

 

“)D.C, Vreeland et al., ANP Quor. Prog. Rep. Mar, 10,
1953, ORNL-1515, p, 118,

2pid., p. 120.

TABLE 7.1. EFFECT OF EXPOSURE TIME ON CORROSION OF INCONEL BY
NoF.ZrF ~UF , {50-46-4 mole %) IN A TILTING FURNACE

 

 

 

 

EXP OSURE zem, METAL CONTENT AFTER CHANGE IN METAL CONTENT
TIME ADDITION TEST* (ppm) DURING TEST {(mea/kg)

(hr) (wt %) Fe Cr Ni Fe Cr Y

1** None 215 45 65 ~5.1 +1.5 +0.4

Fx* None 145 120 35 76 +6.0 0,6

10** Mone 155 190 40 ~7:3 +9.8 ~0.5

1= 0.5 a5 30 <25 ~2.9 +0.8 -0.,9

5 0.5 55 25 <20 -10.8 +0.4 -1.0

10** 0.5 75 <25 <20 -~10,2 +0.3 ~1.0

 

 

 

 

 

 

 

 

*Metal content of fuel before test:

**Average of two tests,

Fe, 380 ppm; Cr, <20 ppm; Ni, 50 ppm.

51
ANP PROJECT QUARTERLY PROGRESS REFPORT

NaHF ., was added to the mixture. The tests were
conducted in the tilting furnace at 800°C, The data
shown in Table 7.2 reveal that, as before, the two
smallest additions were not harmful; however, the
0.097 and the 0.29% HF additions caused very
marked increases in the chromium concentration,
Thus the previous results were confirmed. There
is no obvious explanation for this behavior. Similar
studies will be conducted with uranium-free fuel
solvent to eliminate uranium as a possible source
of difficulty. It should be pointed out that the
lowest concentration of HF used in these tests is
probably higher than the amount that will be left
in the fuel during its preparation. Accordingly,
it seems likely that residual HF will not cause
enhanced corrosion by carefully hydrogenated fuel,

Flvoride Pretreatment. Since the corrosion
characteristics of the fuel might be strongly de-
pendent on the manufacturing process, the effect

of some of the process variables has been evaluated
by the standord tilting-furnace technique.

The standard processing for fuel mixtures in-
cludes high-temperature treatment with anhydrous
HF. Since HF couses increased corrosion, at
least when used in large amounts, parallel tests
have been conducted on otherwise comparable
materials from which the HF was stripped by
purging with helium for various intervals of time.
The data obtained by chemical analysis of three
batches of fuel after 100-hr seesaw tests are shown
in Table 7.3. Since the initial concentration of
soluble structural metal compounds was similar in
the three batches, the decrease in chromium con-
tent with increasing stripping time is probably due
to the effective removal of HF. Metallographic
observation of the test capsules showed moderate
corrosion to a depth of 1 mil for the shortest strip-
ping time; corrosion by the other two batches was

TABLE 7.2, EFFECTS OF chHl"'2 ADDITIONS ON THE CORROSIOM OF INCONEL BY
I~lc=F-ZrF4-UF’4 (50-46-4 mole %) DURING A 100-hr EXPOSURE

 

 

 

 

 

METAL CONTENT CHANGE IN METAL
NaHF, ADDITION FINAL HF AFTER TEST* CONTENT DURING
CONCENTRATION (ppm) TEST (meq/kg)
(wt %)
{wt %) (meq of HF/kg) Fe Cr Ni Fe Cr Ni
None 0 0 70 830 60 | ~8.6 47 | +0.5
0.05 8 0.016 45 755 | <20 | -9.5 43 | -0.8
0.1 16 0.032 110 850 | <20 ~7.3 49 | -0.8
0.3 48 0.097 80 1705 | <20 | ~8.4 97 | ~0.8
0.9 144 0.29 75 | 3065 85 | -B.4 | 176 | +1.2

 

 

 

 

 

 

 

 

 

*Metal content before test:

Fe, 310 ppm; Cr, <20 ppm; Ni, 45 ppm.

All data are average of two tests,

TABLE 7.3, EFFECT OF HELIUM STRIPPING ON CORROSION OF INCONEL BY NaF-ZrF4-UF4 (50-46-4 mole %)

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

STRIPPING METAL CONTENT METAL CONTENT CHANGE IN METAL CONTENT
TIME BEFORE TEST (ppm) AFTEFE-I_EST* {(ppm) DURING TEST (meq/kg)
(hr) Fe Cr Ni Fe Cr Ni Fe Cr Ni
1 410 <20 30 g0 930 20 ~11.4 +57.8 ~0.3
3 680 <20 110 135 665 25 ~19.6 +41.0 ~3.0
5 470 <20 35 75 580 20 ~15.4 +35,5 -0.5
*Average of iwo tests.

52
considerably less severe, and there were no ap-
preciable differences between the results for the
3- and 5-hr stripping times.

Hydrogen should be more advantageous than
helium for stripping the HF, since reduction of
FeF,, FeF,, and NiF, to the metallic state and
reduction of some UF, to UF,; would be simul-
taneously dccomplished. In Table 7.4, the resulis
of tests with untreated fuel, fuel treated at 750°C
for ]]/2 hr with hydrogen, and fuel treated at 750°C
for 1‘/2 hr with hydrogen and then filtered are shown
for comparison. Results of metallographic exami-
indicate that in 100 hr at 800°C in the
seesaw apparatus, the untreated material caused
heavy attack, fo a depth of 4 mils, while the fuel
which was treated with hydrogen and filtered
caused slight attack, to o depth of 1 mil. The
material which was freated with hydrogen but not
filtered caused moderate attock, to o depth of 3
mil s,

Structural Metal Fluoride Additions. In additional
studies, FeF,, CrF,, and NiF, were added to
molten fluorides in an attempt to evaluate the
importance of these materials in the selective
leaching of chromium from Inconel. Previous
experience had shown that FeF., increased the
corrosion of Inconel by fluoride mixtures, and in
the recent experiments in which FeF, was used,
there was severe attack, as was expected. How-
ever, the recent experiments indicated that in-
creasing the FeF, concentration from G.1 to 0.9
wt % had very little effect on the quantity of

nations

PERIOD ENDING JUNE 10, 1953

chromium dissolved or on the severity of attack
during the 100-hr exposure. Additional experiments
will be required to confirm and explain this be-
havior,

The addition of large amounts of CrF, (0.5 to
3.0 wt %) to the ARE fuel mixture prior to exposure
caused moderate to heavy subsurface void formation
to a depth of 6 mils in the hot porfion of the cap-
sule. It is possible that the reaction

CrF3 + Cr® = ZCer

is responsible for the very high corrosion observed.

Structural Metal Oxide Additions. A series of
static tests of Inconel in the fluorideNaF-ZrF ,-UF
(46-50-4 mole %) with additions of Fe,O,, Cr,04,
and NiO have been completed. The tests were run
for 100 hr at 816°C. In these tests, additions of
Fe 0, and Cr,0, did not affect the depth of attack,
but NiQO increased the attack, The metallographic
results of these tests are listed in Table 7.5.

In all tests in this series, a bluish-gray phase
was detected on the surface of the Inconel in what
have in previous tests been considered voids. It
was found on repolishing some of the specimens
that by eliminating the last two polishing steps on
the long-nap cloths, a large amount of this phase
could be retained. The material in place in the
voids is shown in Fig. 7.1, enlarged 2000 times.
In order to confirm the presence of material in the
voids, a Bergsman microhardness determination was
run on this bluish-gray phase. It was found to
have a DPH value of 892 (inconel matrix 135).

TABLE 7.4, EFFECT OF HYDROGEN STRIPPING ON CORROSION OF INCONEL
BY I‘\h:tl""-Zer;-lJF4 (50.46-4 mole %)

 

 

 

 

 

L CONTENT ME DN CHANGE IN
N SLroRE TEST AFTER TESTY ME TAL CONTENT
PROCESSING e f ) DURING TEST
VARIATION ppm ppm (mea/kg)
Fe Cr Ni Fe Cr Ni Fe Cr
Reference
(untreated) 2100 35 200 75 4000 30 w74 229
Hydrogen stripped 690 70 190 265 2035 -4 118
Hydrogen stripped
and filtered A70 20 | 35 205 610 ~-10 35.2

 

 

 

 

 

 

 

 

*Average of two tests,

53
ANP PROJECT QUARTERLY FPROGRESS REPORT

TABLE 7.5. EFFECT OF OXIDE ADDITIONS ON STATIC CORROSION OF INCONEL IN
NuF-ZrF4~UF4 AFTER 100 hr AT 816°C

 

 

 

ADDITIVE METALLOGRAPHIC NOTES
None Specimen attacked 3 to 6 mils, average 4 mils; tube attocked less than 0.5 mil
1% Fe203 Specimen and tube attacked 3 te 4 mils
1% NiO Specimen attacked 5 to 7 mils; tube attacked 7 to 8 mils
1% Cr203 Specimen attacked 3 to 4 mils; tube attacked 2 to 3 mils

 

 

  
  

 

 

UNCLASSIFIED
PHOTO Y-8708

Fig. 7.1. Static Corrosion of Inconel by NaF-ZrF ,-UF, (46-50-4 mole %) After 100 hr ot 816°C. Note

material in voids., 2000X

Microspark spectrographic examination of these
specimens indicated that the material in the voids
contained Fe, Ni, Cr, Zr, and probably U, although
only a faint trace of uranium was detected.
Chromium Metal Addition. Molten ARE mixture
was pretreated in a tilting furnace by exposure to
chromium metal in Inconel capsules. Most of the
excess chromium was then removed by sedimenta-
tion, and the pretreated mixture was transferred to
another capsule for testing, In o series of experi-
ments, this pretreated mixture caused only a slight
roughening of the surface at the hot end of the
capsule and a slight deposit on the cold wall, The
addition of NiF2 to samples similarly pretreated
increased the corrosion considerably; in these

54

tests the chromium content of the melt increased
linearly with the NiF, addition.

Zirconium Hydride Addition. Zirconium hydride
reacts with UF , in fluoride melts to form uranium
trifluoride. A certain amount of the trifluoride can
be retained in solution, but precipitation of this
material occurs if extensive reduction takes place.
An effort has been made to check previous data on
the tolerable limit of ZrH, addition by allowing
the insoluble UF, to separate by sedimentation
after exposure.  The additional data obtained
(Table 7.6) indicate that the uranium concentration
of the melt is decreased considerably with addi-
tions of 0.7% or more of ZtH,. These data, in
general, agree with data from previous experiments
performed with other techniques. The chromium
content of the melt is sharply decreased by the
addition of as little as 0.1% of ZrH,; further
increases in the ZrH, concentration have only o
slight effect. [t appears that although up to 0.7%
of ZrH, may be folerable in this system, much
lower concentrations, of the order of 0.2% may
show the beneficial action desired without the
risk of precipitation of uranium from the melt,

Carbon Addition. Static tests were run for 100 hr
at 816°C with o mixture of 25% powdered graphite
and 75% NaF-ZrF -UF , (50-46-4 mole %) in type
316 stainless and Inconel tubes to determine

carburization of these moterials would
No evidence of carburization was detected
in these tests. After testing, the grain boundories
of the type 316 stainless steel had large amounts
of carbides present, but the deposits were uniform
throughout the specimen and no hardness change
could be detected from the inner to the outer surface
of the tube wall, |

whether
oCCur.

PERIOD ENDING JUNE 10, 1953

Crevice Corrosion. Crevice corrosion tests have
been run in the seesaw furnace with the tapered,
or V-shaped, crevices described previous]y.“)
Corrosion in these crevices seemed to be somewhat
erratic, with some surfaces being apparently un-
attacked and others having the usual subsurface
voids. No acceleration of corrosion was noted in
these crevices, and the depth and the extent of
afttack in the crevices was not beyond what might
be usually expected for the materials tested, Table
7.7 summarizes the results obtained in these tests.
Similar results for crevice corrosion were obtained
in the rotating tests reported in a following section.

Inconel Container Pretreatment. The descaling
properties of the fluoride NaZrF _ have been checked
on oxidized inconel at 1000 and at 1300°F. As
can be seen in Fig. 7.2, the oxidized Inconel was
cleaned up after 4 hr at 1300°F, but it was not
cleaned up after 4 hr at T000°F,

It was thought that perhaps o passivation treat-
ment, similar to that given fo stainless steels,

TABLE 7.6, EFFECT OF VARYING QUANTITIES OF ZrH, ON CORROSION OF
~ INCONEL BY NaF-ZrF -UF ; (50-46-4 mole %)

 

 

 

 

ZiH, ADDED URANI;M CONTENT AFTER METAL CONTENT AFTER TEST*
(wt %) rH, ADDITION (ppm)

(wt %) Fe Cr Ni

None 2. 11 95 560 ‘ 35

0.1 .01 70 195 B8O

0.3 8.94 125 150 50

0.5 9.17 220 175 90

0.7 8.65 405 145 155

0.9 6.76 350 130 125

 

 

 

 

 

*Metal content before test: Fe, 310 ppm; Cr, 20 ppm; Ni, 45 ppm, All results average of two fests.

TABLE 7.7. RESULTS OF SEESAW TESTS OF INCONEL WITH TAPERED CREVICES IN
NaF-ZeF UF , (50-46-4 mole %) FOR 100 hr WITH A HOT-ZONE TEMP ERATURE
OF 800°C AND A COLD-ZONE TEMPERATURE OF 600°C

 

 

POSITION OF CREVICE

METALLOGRAPHIC NOTES

 

Hot zone

Hot rone

Cold zone

 

2 mils of subsurface voidz in upper poart of crevice, increased to

3 mils in lewer part
Subsurface voids te 2.5 mils

Apparently no attack

 

35
ANP PROJECT QUARTERLY PROGRESS REPORT

!nnhm!nu?;.;;;
AS POLISHED

gfi.&. LR imETen gy g -
¥ .
4

-

{9.
f

 

  

UNCLASSIFIED
PHOTO Y-9192

’u!rmimslnnl:zzzizm

AS OXIDIZED

 

 

Fig. 7.2. Temperature-Dependence Tests on the Descaling of Oxidized Incone! by NaZr Fs.

would be beneficiol in increasing the corrosion
resistance of Inconel to the fluoride mixtures.
Therefore, Inconel tubes were hydrogen fired and
the specimens were electropolished, as usual,
and both the tubes and the specimens were treated
in a 25% solution of nitric acid at approximately
120°F for 30 minutes. The tubes and specimens
were then used for static corrosion testing with
fluoride NoF-ZrF -UF, (50-46-4 mole %) for 100 hr
at 816C. The pretreatmant did not appear to be
especially beneficial. Attack inthese tests seemed
more erratic than is usual. In two of the tests,
subsurface voids to a depth of 5 mils were noted;

56

in the other test, subsurface voids were noted 1o a
depth of 3 mils.

STATIC CORROSION OF STELLITE
BY FLUORIDES

D. C. Vreeland E. E. Hoffman,

L. R, Trotter Je E. Pope
Metallurgy Division
In conjunction with the hard-facing problems of
the ARE, Stellite No. 6 (27.5% Cr, 4% W, 2 to 3%
Fe, 1% C, balance Co) was deposited on pieces of
Inconel, and these composite specimens were
static tested in the fluoride NaF-ZrF4-UF4(46-50-4
mole %) in Inconel tubes at both 816 and 538°C,
The structure of Stellite No. 6 has been described
as ‘‘dendrites of a cobalt-rich solid solution sur-
rounded by a mixiure of carbides.'”” Apparently it
is the carbides which are attacked by the molten
fluorides., There appeared to be no acceleration of
corrosion on either the Stellite or the Inconel as
a result of the presence and contact of these
dissimilar metals. The extent of attack appeared
to be the same in the tests at 538°C as in the
tests at B16°C. In all cases, the Inconel showed
about 1 mil of subsurface voids, while the Stellite
showed an average attack of 3 to 4 mils and a
maximum attack of 9 mils on the carbide phase in
the Stellite. Figure 7.3 shows the type of attack
which occurred in these tests..

FLUORIDE CORROSION OF iNCONEL
IN ROTATING TEST

D. C. Vreeland L. R. Trotter
E. E. Hoffman J. E. Pope
Metallurgy Division

Several tests have been completed with the fluo-
rides in Inconel on the NACA type of rotating

' UNCLASSIFIED
{ PHOTO Y-8951

A .
1 3 o -
- ( - . - .‘
- Ve .
- e *
! -
’ - R v l
<
L
%

 

wn
LsJ
X
O
=z
‘ K . 0.005
"'-:-.
,0.: "‘. ” T ? —
v ’ # -(x' . X
C e bes v oE s
.Y oo ) ' s *Y . I?)
.o o
..’ fl r -
‘ - 0010
RN
)"’
.

PERIOD ENDING JUNE 10, 1953

apparatus.t3)  This type of apparatus is advan-
tageous because velocities can be obtained that
are higher than those available in thermal convec-
tion loops or in the seesaw tests, In the tests
conducted to date, there has been no indication of
plugging, and the corrosion results have been
similar to those obtained in static tests, even
though a fluid velocity of 10 fps is being employed,
The depth of penetration has not exceed=d 2 mils
in any of the tests in which leaks hav> not oc-
curred.

Atypical 100-hr test with NaF-ZrF 4-UF , (46-50-4
mole %) in which a temperature of 816°C was
maintained throughout the apparatus resulted in
uniform attack in the form of subsurface voids to
a depth of 2 mils. Additions of sodium have re-
sulted in the development of surface layers on the
inconel, as was also noted in static tests. The
fluoride, with 2% sodium added, was tested for
94 hr with a hot-zone temperature of 816°C and a
cold-zone temperature of 783°C., Although the
attack was in the form of subsurfoce woids, a
surface layer was visible in many places.. The
total depth of voids and ldyer did not exceed 0.5

 

Ghbid., p. 121,

 

Fig. 7.3. Static Corrosmn of Stellite No. & After 100 hr at 816°C in NoF-ZrF 4-UF ; (46-50-4 mole %).

250X

57
ANFP PROJECT QUARTERLY PROGRESS REPORT

mil, nor did the attack vary in extent or nature
from the hot zone to the cold zone.

Some of the crevices near welded joints tested
in the whirligig opparatus in fuel mixture with the
2% sodium addition were examined metallograph-
ically for evidence of crevice corrosion, and no
accelerated corrosion could be found. The greatest
depth of attack noted was 0.5 mil in the form of
subsurface voids. On the other side of the crevice
with the greatest depth of attack, there was attack
to 0.5 mil that was definitely intergranular. As is
usually noted in these crevices, the attack was
somewhat erratic, with some sections unattacked.

FLUORIDE CORROSION IN INCONEL
THERMAL CONVECTION LOOPS

G. M. Adamson, Metallurgy Division

The use of thermal convection loops for determin-
ing dynamic corrosion by liquids has been pre-
viously described.’#) Unless otherwise specified,
for the tests described in the following sections,
the temperature of the hot leg of the loop was
maintained at 1500°F and the temperature of the
vninsuloted cold leg was approximately 1300°F,
With the fluoride salts, this temperature difference
results in a fluid velocity of aboyt 6 to 8 fpm, The
usual testing period is 500 hours.

Effect of Exposure Time. A series of loops has
been placed in operation to determine the effect of
time on the depth of maximum penetration. The
loops were filled from a single batch of fuel at as

 

(4)D. C. Vreeland, R, B, Day, E. E. Hoffman, and
L. D. Dyer, ANP Quor, Prog. Rep. Mar. 10, 1952, ORNL -
1227, p. 119,

near the same time as possible and were then
allowed to circulate for times varying from 10 to
3000 hours. The tests of short duration have been
terminoted, and the loops have been examined.
The data obtained are given in Table 7.8, along
with the data obtained from loops run previously
for periods of 100, 250, 500, and 1000 hours, The
data show that the depth of attack increases quite
rapidly for about 250 hours., After 250 hr, the
depth of attack remains almost constant, but there
is some increase in the intensity of attack. When
data are available from the other loops filled from
the same fuel batch, a time curve will be prepared.

Since chromium must diffuse to the surface of the
hot leg before it can be removed, the diffusion rate
could be the limiting factor in the corrosion rate.
Two Inconel loops were operated for 500 hr to test
this hypothesis, The first loop operated for two
250-ht periods, with g new chorge of fuel for the
second period. The second loop operated for 500 hr
with a single charge from the same fuel. If dif-
fusion was the limiting factor, the attack should
have been the some. The loop which had two
charges showed heavy subsurface void formation of
8 to 17 mils, with the voids primarily in the grain
boundaries. The loop with a single charge showed
moderate to heavy attack of from 3 to 8 mils, with
general attack of 5 mils, The hot-leg sections
from both of these loops ore shown in Fig. 7.4.
The doubling in depth of attack for two charges
waos also confirmed by the chemical analyses, The
chromium content of the fluorides increased more
during the second 250-hr period in Loop 281 than
it did in the entire 500-hr period in Loop 28Q.

These lcops show that the decrease in attack rate

 

 

 

TABLE 7.8, EFFECT OF EXPOSURE TIME ON THE CORROSION OF INCONEL
BY NaF-ZrF4-U Fy (50-464 maole %)
EXPOSURE TIME MAXIMUM DEPTH
(hr) OF ATTACK REMARKS
(mils)
10 1 Widely scattered grain boundary attack
50 3 Scattered groin boundary attock
100 4 l.ight, typical subsurface voids
250 9 Moderate attack
500 9 Moderate attack
1000 11 Moderate to heavy attack

 

 

 

58
PERIOD ENDING JUNE 10, 1953

 
     
  

 
 

UNCLASSIFIED
PHOTO T.3037
"> T

  

’

“
*

| 04 % e :
o - | !
“. ‘ _\\ ’ .
S . -
- 3 —8
o @ . Q.
Son o ® o -
. W ) . Y . ‘ O < . 9
. ’ " - e
- - - . - g
\ oy —?_w . > - o FES . -, ;a * ‘O
ST ~ :
T ' Iy | - MILS
. .- > - e Y £
: e . 7 - ¢
4 - ;'_’ ’,/ .
* . L £ . “’ -
. e i / ! & -
. oo B f : ’
P Lt ey .
LA - L %
’ - : / =, t 3
v ¥ - e . . .
-~ i * i * ‘ .‘ P ¥
bt &

  
  

UNCLASSIFIED
tPHOTO T-3042

Fig. 7.4. Corrosion of Inconel by NaF-ZrF -UF (50-46-4 mole %) After 500 hr at 816°C. (a) 500 hr
with same fuel charge. (b) 250 hr on each of two fuel charges. 250X

59
ANP PROJECT QUARTERLY PROGRESS REPORT

with time is caused by changes in the fuel rather
than by the limiting action of diffusion, The in-
crease in attack is probably coused by an increase
in impurities with a double charge.

Temperature Dependence, In the previous re-
port, (5} results for a single loop operated at a
hot-leg temperature of 1650°%F were compared with
those for a loop operated at a hot-leg tempera-
ture of 1500°F, An additional series of three loops
has now been operated to confirm those results,
The loops in this series were all filled from a
single fuel pot on the some day. The results of
this temperature study are presented in Table 7,9
and illustrated in Fig. 7.5. These results show
that the change in depth of penetration of the attack
is not very temperature sensitive. The maximum
penetration at 1300°F is slightly less than at
1500°F, but it is within the usual spread in re-
sults, and the results for all the loops are definitely
within the experimental variation.

The nature and distribution of the ottack does
change with temperature. At the lower temperature,
the voids are small and evenly distributed. As the
temperature increases, the voids concentrate in
the grain boundaries and become quite large. The
growth and concentration in the woids are caused
by the increasing mobility of the fluid as the
temperature increases. In contrast with the pre-
vious temperature tests, the recent tests showed
no systematic distribution of impurities in the
fluorides at any temperature.

(5)G, M. Adamson, ANP Quar. Prog. Rep. Mar. 10, 1953,
ORNL-1515, p. 123.

Chromium ond Nickel Fluoride Additives., One
more attempt was made to reduce the depth of
attack by pretreating the fluoride NoF.KF-LiF-UF,
(10.9-43.5-44.5-1.1 mole %) with chromium meta!
powder. The powder was added to the fluoride
while in the charging pot. The mixture was held
at 1300 to 1400°F, agitated with helivm for several
hours, and then blown through a stainless steel
pressure snubber before the sample was taken,
The
chromium content of the fluorides as charged to
the loop was 3500 ppm. After the 500-hr test, the
chromium content was 2400 ppm. The attack in this
loop was to a depth of 10 to 14 mils (which is a
standard depth of attack), and a corrosion layer
was found in the cold leg. This was the fifth
loop operated with the fluoride NaF-KF-LiF-UF
to which chromium metal was added by some
Of the five loops, four definitely did not
show any improvement in corrosion resistance, and
the improvement of the fifth is doubtful, An attempt
will now be made to add chromium metal to NaF-
Zr'FA-UF4 (50-46-4 mole %) to determine whether

an improvement can be obtained in this system.

The snubber served os a very coarse filter,

means.

Three loops were operated with material from the
same batch of NaF-ZrF4-UF4 (50-46-4 mole %), but
NiF, was added to two of them. The additions
were made in the fill pot, which was agitated for
1 hr before the loop was filled. The attack in-
creased with increasing amounts of NiF,. The
control loop showed moderate aitack of up to 8
mils. With the addition of 0.05% NiF,, the attack

was reported as heavy, with a moximum depth of

TABLE 7.9. EFFECT OF TEMPERATURE OM CORROSION OF INCONEL BY I«h::F-ZrFA-UF-"4 (50-46-4 mole %)

 

 

 

HOT-LEG TOEMP ERATURE HOT-LEG ATTACK

°F)

1650* Light to moderate subsurfuce voids of 4 to 11 mils; voids larger
than normal and concentrated in grain boundaries

1500* Light to moderate subsurface voids of 3 1o 10 mils; general
attack fo 4 mils

1300 Moderate to heavy attack of 4 to 9 mils; voids very small and
well distributed

1500 Moderate to heavy attack of 7 to 13 mils

1650 Moderate attack of 3 to 12 mils; voids very large and concentrated
in groin boundories

 

 

(5)

*Results reported previously.

60
PERIOD ENDING JUNE 10, 195

 
    
 

TUNCLASSIFIED
PHOTO T-3088 1§

- P T . . y
: k L e ’ ? A el ey 9 3 .
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s % . 7 —2
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i & . - MY ' . s
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» . .. -'~.$ - ARSI '_‘r‘ s » 's,, _‘_‘V 5 I
o e 2 cet ¥ ~_‘*a ‘14_}: N . v A ‘.‘ - . v . \!’-; ® ’ e e
ol )\. . . ‘o‘ . -7 T ' - - i s . . ‘ o .
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Tl e BT oA N ! SN ., _—
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eV PEN 2 e e i S SN "
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N . W -8 ". e . o [ s, Y e e 7 - . . .
‘ “" - ' \A~ - : ;' ;:. =’ \" e ""- i ‘ ‘. . 4 - l“ ) .’ . » - .1 . ¥ ; - .
Lo T . s b . T e - . ’ . L s . —7
™ . \ - "" o sy‘ » > 4 - : ","\ ’é 7 . oy
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o2 (- w . o~ . S - e g
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v M ' - . _ __..9.
% ! % -
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UNCLASSIF [ED
: PHOTO T-3071

  
  

 
        

-

Crecepr @t ‘ | LW TN —_
x\ fui\. e T A N ; S L. ’r ‘ g . - !‘ _ v 3 » .

. . 1 g e o - S e .
- ..* ¢ A, e e e aN - T §:
-~ 4 * T N VyoenT £ s €

Y L8 ,: . ) . fr
'J}‘. L A R

.. o 0 0.
’”,o . . P v re #a. —3
T « 0 : . o ‘

 

   
     
 

UNCLASSIFIED
PHOTO Tf3083
» 9§ :

» \. ¢ . e ‘ # ‘ , 4 .
- . . ‘ ,f . =
% ’g‘. - ‘ .'q. ’ - \ ‘w 5
» ’ /‘ ¥ ' ’ : * —
. ¢ ¢ 7

- . 5 —8
o . ;
* % . +
) )
~4 ' : A" . .7 .
‘ L e ey, "o ® MILS

Fig. 7.5, Corrosion of Inconel by NaF-ZrF -UF, (50-46-4 mole %) After 500 hr As a Function of
Temperature, (a) 1300°F. (b) 1500°F. (c) 1650°F. 250X

61
ANP PROJECT QUARTERLY PROGRESS REPORT

10 mils. With a 0.2% addition, a rough surface and
heavy attack of up to 11 mils was found. The
nickel analyses of three samples taken as the
loops were filled were reported as 120, 280, and
500 ppm, respectively. It has not yet been de-
termined whether insufficient time was allowed for
all the nicke! fluoride to go into solution. In the
samples with the NiF, addition, the iron content
increased from 620 to 1100 ppm, while the chromium,
which is the usual reaction product, only increased
from 230 to 620 ppm.

Fuel Purity. Several loops were operated with
high-purity botches of NaF-ZrF ,-UF,
(50-46-4 mole %) prepared in graphite-lined vessels.
One loop contained o batch of high-purity fuel
prepared from standard materials, while two other
loops were filled with fuel prepared from hafnium-
free zirconium oxide, which was low in iron im-
purity. However, the actual chemical analyses
showed the iron and the nickel contents to be about
the same in all three charges, that is, a total of
about 300 ppm in each. In addition, one of the
loops with fuel prepared from hafnium-free zir-
conium oxide was cleoned by circulating NaZrFS.
In the specially cleaned loop, the attack was less
than normal, with @ maximum penetration of 5 mils;
in the other two loops, the attack was normal, with
a maximum penetration of 9 mils, However, the
corrosion attack of the high-purity fuel batch was
not measurably different from that of the standard
fuel in the comparable test.

In another test, two batches of fuel that had
received no purification other than a vacuum drying
during melting were run in two other loops., Heavy
attack, with a maximum penetration of 15 mils, was
found in both loops, This is the deepest pene-
tration found in any loop in which the fluoride
Nc1F-ZrF4-UF:4 (50-46-4 mole %) has been circu-
lated.

It is apparent from these tests that although
purification of the fuel is desirable, the present
standard production procedures appear to be ade-
quate.  Little, if any, additional reduction in
attack can be obtoined with additional fuel purifi-
cation processes,

special

However, the contaminants
(FeF, and NiF,) that remain in the fuel may be
further reduced by additions such as ZrH,,.

Loop Oxide Films. To determine the effect of
oxide scale on the maximum depth of attack,
several loops that had been subjected to various
surface treatments were operated for 500 hr at

62

816°C with fluoride NaF-ZrF ,-UF, (50-46-4 mole
%). The various treatments included (1) circu-
lating NaZrF_ for 2 hr ot 1400°F, (2) degreasing,
and (3) oxidizing before filling. The maximum
attack in the three loops was 6, 7, and 8 mils,
respectively. The attack on the NaZrF .-cleaned
loop and the attack on the oxidized loop are shown
in Fig. 7.6.

Although the treatment with NaZrF . removed the
oxide film from the loop, there was a brown film on
all sections of the loop which varied in thickness
from thin in parts of the cold leg to granular in the
hot horizontal section. A diffraction test showed
the film to be over 90% Na,ZrF,, with some zir-
conium oxide present. The effect of this compound
on the fuel is not known. No measurable attack
took place during the cleaning.
small

Removing the
amounts of oxide film by treating with
NaZrF, results in what oppears to be o slight
reduction in depth of attack; however, a large
oxide content did not cause a large increase in
depth of attack.

Nonuranium Bearing Fluorides, During the test-
ing and filling of the ARE system, flucride mixtures
which do not contain uranium will be used. Cor-
rosion work on such mixtures was started in the
previous quarter and was continued. The corrosion
results contirm those previously reported in that
nonuranium bearing fluorides are found to be less
corrosive than the fuel mixture. With the fluoride
NaZrF, alight attack with a maximum penetration
of 5 mils was found. In the cold leg, the maximum
attack was 2 mils. A maximum attack of 5 mils
was al so found in a second loop, which was similar
but was allowed to circulate for 1000 hours. A
very thin corrosion layer was found in the cold leg
of each loop. Such a layer is not usually discerni-
ble with NaF-ZrF,.-UF, (50-46-4 mole %). The
nature of the attack is the same with or without
the uranium in the mixture,

FLUORIDE CORROSION OF NICKEL
THERMAL CONVECTION LOOPS

G. M. Adamson, Metallurgy Division
The fluoride NaF-ZrF4-UF4 (50-46-4 mole %)

was circulated in a nickel thermal convection loop
which was cleaned with dry hydrogen. After 500 hr
at 816°C, o few metallic crystals were found in
the trap and in the fluorides in the lower part of
the cold leg. An extremely thin corrosion layer
PERIOD ENDING JUNE 10, 1953

| UNCLASSIFIED
| PHOTO T-3148

+

 

. ' L S e Y MILS
. . . - . - L - - t

(£)

Fig. 7.6. Effect of Oxide Scale on the Corrosion of Inconel by NoF-ZrF -UF , (50-46-4 mole %) Atter
500 hr at 816°C, (a) Deoxidized with NaZrFSV (b) Oxidized. 250X

63
ANP PROJECT QUARTERLY PROGRESS REPORT

was found on the cold-leg surfaces by metallo-
graphic exomination. The hot-leg surfaces were
quite smooth, with no intergranular or subsurface
voids. The wall of the hot leg was 0.002 in.
thinner than that of the cold leg, but since the
thickness was still within commercial tolerances,
no definite conclusion may be drawn from this one
measurement. Sections of both the hot leg and the
cold leg of this loop are shown in Fig. 7.7. A
small amount of mass transfer took place, but it
was less than that found with type 316 stainless
steel and many times smaller thon that found when
the fluoride NaF-KF-LiF-UF, (10.9-43,5-44.5-1.1

mole %) was circulated in nickel.

FLUORIDE CORROSION OF TYPE 316
STAINLESS STEEL THERMAL
CONVECTION LOOPS

G, M. Adamson, Metallurgy Division

Two type 316 stainless steel thermal con-
vection loops operated for 500 hr at 1500°F with
NaF-ZrF ,-UF , (50-46-4 mole %)} without plugging.
These were the first loops of stainless steel to
circulate uranium-bearing fluorides at 1500°F with-
out plugging. Typical hot- and cold-leg sections
from these loops are shown in Fig, 7.8. The hot
legs in both loops were very rough, with both the
intergranular and the typical Inconel types of
attack. In loop 133 the depth of attack was 5 to
11 mils, while in loop 134 it was 4 to 7 mils. Some
grains had been removed from the surface, but the
average thickness had not been reduced. In both
loops, a heavy multilayer deposit was found in the
cold leg. The top layer of the deposit consisted
of fine, dendritic, metallic crystals, These crystals
were also found in the fuel. Diffraction and
spectrographic studies show these crystals to he
mainly iron, with some chromium in solution. The
zirconium content was higher than would be found
if fuel particles were mixed with the crystals.
Both the nicke!l and the molybdenum contents were
very low. Under the first layer was another layer
which appeared to be nonmetallic, and in some
areas there was another metallic layer beneath the
nonmetallic layer. The attack found in these loops
appears fo be much more comparable to that found
with Inconel than was the attack in the loops that
plugged with the fluoride NaoF-KF-LiF-UF, (10.9-
43,5-44.5-1.1 mole %).

64

LIQUID METAL CORROSION

. A, Knox E. E., Ketchen
L.. G. Overholser

Materials Chemistry Division

J. V. Cathcart D. C. Vreeland
G. P. Smith E. E. Hoffmon
W. H. Bridges L.. R. Trotter
J. E. Pope
Metallurgy Division

L. A. Mann D. R. Ward
ANP Division

Mass Transfer in Liguid Lead., Studies of the
mass transfer of various solid metals in small
thermal convection loops containing liquid lead
have continued. These tests have been performed
in thermal convection loops made of quartz tubing.
Samples of the test metals are fastened in the hot
and cold legs of the loops. Details of the con-
struction of the loops and of the results obtained
for Inconel, columbium, molybdenum, and type 304
stainless  steel were reported previously, (67}
Armco iron and types 347 and 446 stainless steels
were tested during this quarter.,

The loop containing type 347 stainless steel was
operated with hot- and cold-leg temperatures of
800 and 500°C, respectively, and failed after 140
ht because of the formation of a plug in the cold
leg. Figure 7.9 is a cross section of the hot leg
of this loop; the intergranular type of corrosion
which occurred, in addition to the mass transfer,
can be seen. A similar, but less marked, corrosive
attack occurred in the cold leg.

The loop containing Armco iron failed because
of plug formation after 250 hr of operation. The
hot- and cold-leg temperatures were 800 and 545°C,
respectively, As may bhe seen from Fig. 7.10,
which shows a cross section of the hot leg of the
loop, the iron specimen suffered almost no inter-
granular attack. A decrease in wall thickness of
0.003 in. was measured for the hot-leg specimen,
There was no change in the wall thickness of the
cold-leg specimen,

A marked increase in resistance to mass transfer
was noted in the loop containing type 446 stain-
less steel. This test was terminated on schedule
after 619 hr of operation with the hot leg at 810°C
and the cold leg at 520°C. Although insufficient

 

)G, p, Smith et al., ANP Quar, Prog. Rep. Mar. 10,
1953, DRNL-1515, p, 128,

7)Met. Prog. Rep. April 10, 1953, ORNL-1551 (in press)e
PERIOD ENDING JUNE 10, 1953

 
 
  

UNCLASSIFIED
PHOTO T-3149

»

 
    

\‘s. 7
™, ; ‘
\\ , : -9
Y, : ' -c;: \
N 4 - T R [+
% A e
5, o o ., s
\% e - ‘°‘1\ / \."\ - M | L“S
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» i £ //‘ 5 &
& /,‘ r“‘d
e &
W
. -//

 
 
  

UNCLASSIFIED
' PHOTO T-3150

 
    

—9

—10
MILS

 

(£)
Fig. 7.7. Corresion of Nickel Thermal Convection Leop by NaF.ZrF .UF ( (50-46-4 mole %) After
500 hr at 1500°F. (a) Hot leg. (b) Cold leg. 250X

65
ANP PROJECT QUARTERLY PROGRESS REPORT

UNCLASSIFIED
PHOTO T-2709

v T
e ., M_'j;,,“ K. .

¢

 

Fig. 7.8. Corrosion of Type 316 Stainless Steel Thermal Convection Loop by NaF-ZrF4-UF4 (50-46-4
mole %) After 500 hr at 1500°F. (a) Hot leg. (b} Cold leg. 250X

66
PERIOD ENDING JUNE 10, 1953

 

UNCLASSIFIED
PHOTO Y-9255

INCHES

0.005 -

 

 

 

 

(500 Xx)

Fig. 7.9. Corrosion of Type 347 Stoinless Steel by Lead in Quartz Thermal Convection Loop After
140 hr ot 800°C. 500X : _ : :

 

 

 

 

 

C . 3
s H}i - [ -
! e - s w
£ , - 1 T -
5 e U .
- %, - ,“" ‘:(:.-Bs ‘ >
’:\ » i% “a . A
% - - _ % e
™, & t bt - 1&} ” :
\'\,,\w‘*lk' « -v_‘_“ -
- : T( i
» . 3,
£ ? e .
- N ey - 10.005
i E
»» ) ~m L 3
/ e Y flwwwrwn”a - |
& ";“ o g : oy # -
; - ' - s » -
— v § v R,
k ’ . B / . : %
e / % v - 1 ' o

Fig. 7.10. Corrosion of Armco Iron by Lead in Quartz Thermal Convection Loop After 250 hr ot 800°C.,
500X '

67
ANP PROJECT QUARTERLY PROGRESS REPORT

mass transfer occurred fo form a plug large enough
to stop the circulation of lead in the loop, a small
quantity of mass transferred material collected at
the bottom of the cold leg. It is to be presumed
that had the loop been dllowed to operate for a
sufficiently long time, this deposit would have
grown to sufficient size to stop the circulation of
lead in the loop.

The results obtained thus far indicate that alloys
with a high nickel content — for example, |nconel
and type 347 stainless steel ~ show little resist-
ance to mass transfer or corrosion in lead. On the
basis of the results with type 446 stainless steel,
it appears that stainless steels containing no
nickel suffer relatively little mass transfer as
compared with the nickel-rich alloys, Columbium
and molybdenum showed no mass fransfer in lead.

BeQ in Sodium and NaK. The compatibility of
beryllium oxide with sodium or NaK is of such
concern to the ARE that the problem is being
attacked simultaneously in seesaw, rotating, and
thermal convection loop tests, with both treated
and untreated beryllium oxide specimens. The
treated specimens were sprayed with slurries of
various  fluorides, including CaF,, CaF -AIF ,,
and CaF ;-MgF,, and then dried. In general, the
cootings seemed to have some beneficial effect
in reducing corrosion,

Each of the tests was operated with a hot-leg
temperature of about 1500°F for 100 hr in the
seesaw and rotating tests and for 212 hr in the
convection loop tests. The results in terms of
weight change of the 1{‘- by ‘4- by ]/Q-in. beryllium
oxide specimens vary from 1% in the seesaw
tests up to 10% loss in the convection loop tests
and to 71 to 100% loss in spinner tests., The
sodium velocity in the rotating test was 400 fpm
and, in thermal convection loop tests, it was only
of the order of 6 to 10 fpm. The data obtained are
summarized in Toble 7.10. It now appears that the
corrosion mechanism is, in reality, a mechanical
erosion of the hot-pressed BeO surface. Therefore
the ARE has been designed so that the sodium in
contact with the BeO will be virtually stagnant,

In an attempt to determine where the missing
beryllium oxide ends up, two convection loops
were sectioned for analysis. Each loop had ope-
rated for 212 hr ot 1500°F; one loop contained
sodium and the other NaK, The distribution of the
residual beryllium oxide in the liquid metal drained
from the loop, in the wash solution, and on the
wall is given in Table 7, 11. These data indicate
that the beryllium oxide is mainly suspended in the
liquid metal rather than being firmly attached to
the walls, as was previously reported.

Coated Beryllium in Sodium. A series of beryl-
lium specimens, some of which were plated with

TABLE 7.10. EROSION OF BERYLLIUM OXIDE BY SODIUM AT 1500°F

 

 

 

 

TEMPERATURE (°C) WEIGHT LENGTH
TYPE OF TESTY SPECIMEN CHANGE OF TEST
Hot L.eg Ceold Leg (%) (hr)
Seesaw A|F3~C0F2 treated BeO 825 760 -1.1 100
AIF3'CGF2 treated BeO 820 675 +0.20 100
Special grade of BeO 800 650 +0.89 100
Special grade of BeO 805 720 +0.32 100
Rotating A|F3-—C0F2 treated BeO 816 816 ~100 100
AIF3-COF2 treated BeO 816 816 -71 100
Special grade of BeD 816 816 —~94 100
Special grade of BeO 816 816 -90 100
Convection loop Untreated BeQC 816 600 -16.8 212
A|F3-C0F2 treated BeO 816 600 ~9.8 212
A|F3-C0F2 treated BeQ 816 625 ~1.8 212
C<:|F2 treated BeD 816 650 ~10.5 212
CaF, treated BeO 816 650 -7.3 212
MgFZ-Cch treated BeD 816 600 0.0 212

 

 

 

 

 

 

68
PERIOD ENDING JUNE 10, 1953

TABLE 7.11. DISTRIBUTION OF BERYLLIUM OXIDE RECOVERED FROM CONYECTION LOOPS

 

 

 

 

 

 

RECOVERED BeO (%)
LiQUID METAL CIRCULATED
On Wall In Washings in Ligquid Metal
NaK 1.2 36.9 61.9
MNa 2.4 77.2 20.4
chromium and others with nickel at Gerity-Michigan UNCLASSIFIED
Company and at Y-12, were tested in sodium in PHOTO Y-8612

Inconel tubes for 100 hr in the seesaw furnace.
Tubes were crimped so that the specimens were
restricted to the hot zone. All specimens showed
attack, and in only one nickel-plated specimen
was the plating at all adherent. Since both chro-
mium and nickel have good resistance fo sodium,

 

‘E! o
e

 

   

   

     
          

it is assumed that the poor showing of these

i T A W AR

specimens is due fo porous plating. The appear- '.(q .HQT ZE _ 10°C
ance of some of these specimens after test is COLD 7ONE — 640°C
shown in Fig., 7.11,

Carboloy in Lead and Sodium, Static tests: of
several Carboloys have been run with sodium and
lead in Inconel tubes for 100 hr of 816°C. Except
for one test with Carboloy 608 in lead, the Carbo- (b) HOT ZONE - 830°C
loys (907, 779, 44A, aond 55A) appeared to have COLD ZONE —-T770°C
good resistance both to sodium and to lead. While
Carboloy 608 was satisfactory in sodium, when in
lead, a light color phase appeared to be preferen-
tially attacked to a depth of 26 to 28 mils. The )
brittleness of these alloys, however, appears to (c) HOT ZONE—830°C

 

 
   

make them prone to cracking and spalling at the COLD ZONE—725°C

edges, as can be seen in Fig. 7.12, It is not known

whether this spalling takes place during testing or Fig. 7.11. Corrosion of Chromium-Plated Beryi.
during metallographic preparation. fium by Sodium After 100 hr in Seesaw Test,

69
ANP PROJECT QUARTERLY PROGRESS REPORT

  

 

 
     

 

 

R
" .
UNCLASSIFI%? i - kS
PHOTO Y-8872 | » -
G TR gy g P . P
.. ' 4 e T TR ., N
% , . s . ) . el
- -7 - - S ¥ T
™ : * . . . * * Q
- . 7 - * ‘ . > v' * * — 2
4 4 oy . s - e ~ 4 =
’ ‘. r ’ * ‘ ‘ . - . * - ’ ‘.‘ -
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; " - . . ( . . ’ ¥ . . é
t 4 , - . v . % - Q005
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‘ . . A P . . , . Lo . 4 -
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v 0o N I & - ¥, folee TN e x
1 1N ‘ . ’ L , \' » ) o
’ ~ . » ¥ 1. ” . - -
w7 : . . N e - ’ ' 0
SN - % 1 . $ - ~ * .,! TN
.D o « 1 . poL , L a o . ., Nt
L - “ . ' . . “fi. s + . ‘.l f !\ . 4 - , . .
’ , : - ‘ f‘l {\ e { o ' r : .‘ ’ .- ;c . =~ . ’ - 2 v ¥ &* " - -*
< . AN Ul T ) . - e~ " 0010
; * » ” + 4 - :'.. 2 : . " 3
. . " - 2 .
t - 4 L - 4 . . o . .. ' - -
r - ot P - - 13 “a (‘ L. ? ¢ - y . f’ LHoe
-”“ . * - " - 4 - » ~ - . - * . - F
» * . ” My . : - »
* . € - “ ". .
~ ' s - il . , - e . i :‘ ‘. 4 ’ r . B .
-, 5 a 1 - “ > »
e ' -4 ! woN ¢ ‘n‘ 't ’ . - .

Fig. 7.12. Carboloy 608 After Testing in Sodium for 100 hr ot 816°C. Note cracking and spalling at
upper edge. 250X

70
PERIOD ENDING JUNE 10, 1953

8. METALLURGY AND CERAMICS

W. D. Manly

J. M, Warde

Metallurgy Division

The fabrication of an intricate liquid-to-liquid
heat exchonger was effected by manual inert-arc
welding of tube-to-header welds, The individual
tube bundies were reinforced ond spaced through
the use of grid-type spacers fabricated by cone-arc
wire-to-header welding techniques.

In a study of the effect of the various brozing vor-
iables on the flowability of the Nicrobraz brazing
alley, it was found that contamination of the
brazing otmosphere with nitrogen has a serious
effect on the flowability of the alloy. Dilution
and diffusion studies on three high-temperature
brazing alloys ~ Nicrobraz, 60% Mn~40% Ni, and
82% Au-18% Ni —~ have been completed. Brozed
tube-to~fin joints were prepared to study the effect
of the quantity of brazing alloy used, the bruzmg
temperature, and the brazing time,

A technique has been developed to braze ce-
mented coarbide rings to shafts for use os pump
seals. Standard brazing techniques could not: be
used since there is approximately u factor of four
differerce between the thermal coetficient of ex-
pension of Carboloy ond that of stainless stesl.
Additional information on the elevated temperature
strength of Nicrobrazed Inconel and type 316 staine
less steel joints hos been obtained.

The study of the effect of different environments
such as air, argon, hydrogen, and the fused fluoride
salts on the creep and siress-rupture properties
of Inconel has continued, and o wide variation of
properties haos been found. This study is being
extended to include austenitic stainless steels,
ferritic stainless steel, ond cobalt-base alloys.
Preliminary results indicote that Inconei and nicke!
are much more sensitive to environmental changes
than are the austenitic stainless steels, such as
types 316 and 304. A design curve for Inconel
tested in the fused fluorides at 1500°F has been
completed over the stress range of 2500 to 8000
psi.

Special high-purity alloys of the Inconel type are
prepared in o vacuum melting furnace ond then
extruded to form a draw blank. The drow blank is
then reduced to produce 1/2-in.-c!ia 0.035-in,-wall
tubing for corrosion tests. The activation energy
for the oxidation of columbium in air was found to

be 13,400 cal/mole for the temperature range of
600 to 900°C and 4350 cal/mole for temperatures
above 900°C. The oxidation resistance of chro-
mium on OFHC copper, chromium and nickel elec-
troplate on copper, and Incone! and type 310 stain-
fess steel clad on copper has been studied for the
production of high-conductivity fins for an ANP
type of radiator, Special oxidotion-resistant alloys
of copper have been prepored, and their oxidation
characteristics have been determined.

Work on the fabricaotion of control rods for the
GE-ANP program has continued. Aftempts to braze
hot-pressed segments of B, C-Fe mixture fo stain-
less steel tubes were not successful; therefore an
alternate method of fabrication is being tried which
consists of tilling a tube with a mixture of A_I-BAC
or Cu-B,C powders and reducing to final size by
swaging and hot rolling to consolidate the mixture,
as well as to form a bond between the matrix and
the tube wall, Powder metallurgy compacts of
various metals with additions of MoS, were pre-
pared for use as pump seals, Additional work on
the preparation of spherical particles of uronium-
bearing alloys has been performed with the use of

the following methods:

1. heating in a vacuum to gbove the melting point,
2. dropping particles through o three-phase three-
electrode arc, which provides the heat source,

3. forcing the molten alloys through a small ori-

fice,

4. letting precut pqrtlcles of the alloy settle
through o molten salt both with o temperoture
gradient that extends beyond the liquidus and
solidus of the alloy.,

Research has been initiated to determine to what
extent the burning of jets of liquid sodivm cen be
reduced by alloying the sodium with some of the
heavy metals. In the initial phase of this project,
a large number of alloys will be surveyed quali-
tatively, Preliminary experiments indicate that
alleying with lead, mercury, or zinc does not ap-
preciobly change the inflammability of sodium at
800°C. A collection of this type of qualitative
data will serve to indicate whether a detailed
study of the numerous variables of the inflamma-

 bility process will be worth-while.

71
ANP PROJECT QUARTERLY PROGRESS REPORT

A cermet has been developed in which 10% USi,
is included in the central portion of a silicon car-
bide fue! element. Other ceramics research has
included @ study of the heating rate of beryllium
oxide. |t appears that the safe heating rate is less

than 200°C per hour,

WELDING AND BRAZING RESEARCH

P. Patriarca Y. G. Lane
G. M. Slaughter C. E. Shubert
Metallurgy Division

Welding of Heat Exchanger Tube Bundles, Pre-
liminary cone-arc welding experiments for deter-
mining the feasibility of the consiruction of the
six-tube-bundle liquid-to-liquid heat ex-
changer were described in a previous reporh“)
Further efforts to apply the process to this fabri-
cation merely verified the evidence that the pro-
duction of relichle, consistently leak-tight cone-
arc tube-to-header welds was complicated by the
curvature of the dished headers of the heat ex-
changer. Rather than to delay the production of
the heat exchanger, several experimental headers
were fabricated by manual inert-arc welding., It
was found that a qualified operator can fabricate
consistently leak-tight tube-to-header welds by the
use of manual welding techniques. Consequently,
the manual welding of the six tube bundles come
prised of 35 tube-to-header welds on each end of
each bundle was undertaken, pending further de-
velopment of the cone-arc welding process.

One tube bundle consists of 35 Incone! tubes,
0.148 in. in outside diameter with 0.025-in.-thick
walls, welded on each end into a 0.125-in.-thick
dished (4-in. radius) heoder. FEach bundle was
helium leak checked and Dy-Cheked prior to weld-
ing of the nozzle assembly. The nozzle was at-
tached by manual inert arc welding. Six of the
420 tube-to-header welds were found by the helivm
leak detector inspection procedure to contain
leaks, which were subsequently successfully re-
paired,

Since the tube bundles contained a free span of
approximately 40 in., the design required that grid-
type spacers and clamps be used at 8-in. intervals
to provide reinforcement of the bundle and to
ensure free flow of liguid around the tubes. A
cone~arc wire-to-header plug-welding technique was

lncone!

”)P. Patriarca, G. M. Slaughter, ¥. G. Lane, and
C. E, Shubert, ANP Quar. Prog. Rep. Mar. 10, 1953,
ORMNL-1515, p. 140,

72

developed and applied to the fabrication. Svitable
jigs were used to provide the dimensional toler-
ance required with respect to curvature and spac-
ing, Figure 8.1 shows three completed tube bundles
welded into a half section which will constitute
the cylindrical vessel of the six-tube-bundle liquid-
to-liquid heat exchanger when welded to its mate,

: UNCL ASSIFIED
s 2. PHOTO Y.90460

 

 

 

 

Fig. 8.1. Three Tube Bundles Welded into Half
Section of Cylindrical Six-Tube-Bundle Liquid-te-
liquid Heat Exchanger. Note spacers,
Brozing of Air Radictors, In the fabrication of
previous heat exchanger test assemblies by furnoce
Nicrobrazing,(” experience indicated that furnace
temperatures of 2300°F were necessary to obtain
suitable flowability of the brazing alloy. At braz-
ing temperatures below 2300°F, a granular braze
was often obtained, and as o result, the joints
leaked. PBrazing ot 2300°F wos found to cause
severe dilution of the parent material and inter-
granular dilution and/or diffusion, which resulted
in severe embriftlement of the parent assembly,
The design of the heat exchanger test units usu-
ally required a multiple brazing operation. Conse-
quently, leaky assemblies that were practically
impossible to repair were sncountered on the sub-
sequent rebrazing operations, possibly because of
excessive dilution and/or cracking of previously
embrittled joints as a result of the thermal stresses
set up during the rebrazing operation.
to remedy the deleterious effects of the high braz-
ing temperature, a series of tests was run to de-
termine the cause of poor flow at normal brazing
temperatures.

Brazing cycles were accurately determined by

In an effort

multiple temperature measurements made with chro-
mel-alumel thermocouples encased in impervious
ceramic tubes. The dew point of the hydrogen
atmosphere was measured ond verified as being
less than —70°F, The effect of rate of rise of
temperature on diffusion of boron and the effect
of boron on flowability were also experimentally
discounted as major factors responsible for poor
flow when the clloy is used in the form of a thick
slurry and the contact area for boron diffusion is
small. A determingtion of the effect of nitrogen as
an impurity in the hydrogen atmosphere, however,
produced significant results, There is indication
that elimination of the hydrogen purge being used
in the brazing operation is necessary. Controlled
experiments performed with the use of a small
muffle furnace revealed that the presence of ni-
trogen impurity in the hydrogen resulted in a defi-
nite impedance to the flow of Nicrobraz by very
rapidly forming a poorly reducible nitride ond/or
a new brazing alloy with a higher melting poini.
These experiments led to the conclusion that in
future can-brazing operations a helium purge should
be substituted for the hydrogen purge. To verity
this thinking, a smail 15-fin/in. heat exchanger
was rebrazed after two previously unsuccessful
operations at 2200 and at 2300°F. The joints were
prepared with a fresh slurry of alley and brazed of

PERIOD ENDING JUNE 10, 1953

2150°F for 40 minutes.
tained, and helium ledk testing did not detect the
presence of any leaky joints.

Future work will further verify whether complex
heat exchangers can be fabricated in a single
brazing opsrotion at a reasonable temperature
range (between 2050 and Z2150°F) in which the
excessive dilution, embrittlement, and development
of leaks experienced with the multiple brazing
operations performed ot 2300°F can be avoided.

Excellent flow was ob-

Dilvtion and Diffusion of Brazing Alloys, It was

indicated in a previous reporf(” that studies were
being made of dilution and diffusion of three high-
temperature brazing alloys — Nicrobroz, 60% Mn-
40% Ni, and 82% Au~18% Ni. Brazed fube-to-fin
joints were prepared for a study of the metallo-
graphic effects of (1) the quantity of brazing alloy
used, (2} the brazing temperature for o given time,
and (3) the brazing time for a given temperature,
The joints were prepared from 0.010-in.-thick type
302 stainless stee! fin material and 0.150-in.-0OD
0.020-in,-wall type 304 stainless steel tubing.

Smoll, medium, and large amounts of brazing
alloy were investigated, while the time ot tempera-
ture was varied from 10 min to 18 hr for an over-
night brazing cycle. The brazing temperature was
varied from 20 to 120°C above the melting point,

The results of metollographic examination of the
test specimens are presented in Table 8.1 as
average percentages of the tubes wall thickness
actually dissolved during the brazing.

The specimens brozed with Nicrebraz alioy olso
exhibited o diffusion zone which sometimes ex-
tended completely through the wall thickness, The
maximum penetration, which woas primarily intere
granular diffusion of clioy constituents, was re-
corded, but the scatter of the data was such gs fo
No dif
fusion zones wider than 0.5 mil were found in
specimens brazed with the gold-nickel ond nickel-
manganese olloys,

make a definite correlation impossible,

As a result of this investigation, it became op-
parent that base-metal dilution increases rapidly
with the use of increased amounts of Nicrobraz
alloy. Severe alloying of the tube plus complete
collapse of the fin is shown in Fig. 8,20, which is
a photomicrograph of a joint brazed at 1220°C for
30 min with a large quantity of alloy,
tube-wall and fin dilution can be seen in Fig, 8.25,
which is a photomicrograph of a joint brazed under
identical conditions but with less alloy. Dilution

Much less
ANP PROJECT QUARTERLY PROGRESS REPORT

 

 

 

 

 

 

 

TABLE 8.1, RELATIVE AMOUNT OF DILUTION OF VARIOUS ALLOYS INTO 0.020-in.-WALL
0.187-in.-0D TYPE 304 STAINLESS STEEL TUBING
TYPE AND QUANTITY EXTENT OF DILUTION OF STAINLESS STEEL
OF BRAZING ALLOY TUBING (%)
At 1220°C T At 1180°C At 1125°C
Nicrobraz
1 ring of 31-mil wire* 8 3 2
2 rings of 31-mil wire* 28 15 10
3 rings of 31-mil wire* 26 26 5
At 1130°C At 1080°C At 1030°C
60% Mn—40% Ni
1 ring of 21-mil wire 3 1 0
1 ring of 32-mil wire 11 4 0
1 ring of 45-mil wire 15 7 2
At 1130°C At 1030°C At 980°C
82% Au~-18% Ni
1 ring of 21-mil wire 8 9 4
2 rings of 21-mil wire 13 10 6
1 ring of 37-mil wire 19 19 12

 

 

 

 

*Mixture of Nicrobraz alloy powder in acryloid cement binder.

also increased rapidly with increasing brazing
temperature, but it was virtually independent of
time at temperature.

Intergranular diffusion during Nicrobrazing was
greatly increased with the use of high temperatures
and/or the use of large quantities of brazing dlloy.
Time seemed to affect the depth of the diffusion
zone somewhat, but no definite relationship could
be determined from this investigation,

Dilution of the stainless steel by the manganese-
nicke! and gold-nickel alloys increased substan-
tially with both temperature and gquantity of alloy,
but time ot temperature did not seem to be an
important variable. In general, dilution was greater
with the gold-nickel alloy than with the manganese-
nickel alloy.

Brazing Carboloy to Stainless Steel., It has be-
come necessary to braze cemented carbide rings to
stainless steel for use as rotating mechanical pump
seals for operation at 1500°F in both fluoride and
air environments, Standard brozing techniques re-
sult in severe cracking of the brittle Carboloy upon
cooling becouse of differences in coefficients of

74

thermal expansion. Carboloy, for example, has a
representafive coefficient of thermal expansion of
5 x 107¢ in./in.-°C, while stainless steel has a
representative value of 20 x 107° in./in..°C.
Nicrobraz, which is highly resistant to high-
temperature oxidation and moderately resistant to
tluoride aottack, was used in preliminary investi-
gations. However, severe distortion of the stain-
less steel with subsequent fracture of the Carboloy
occurs when standard techniques are used (Fig.
8.3). A small rod of the cemented carbide fractured
completely along its length. A nickel shim, which
was placed between the two parts to be joined to
reduce the effects of contraction, was so em-
brittled by the Nicrobraz that yielding was pre-
vented and the Carboloy fractured. Thus it became
evident that a ductile brazing alloy with the re-
quired corrosion resistance was needed. An 82%
Au-18% Ni alloy was selected, and actua!l as-
semblies were brazed by wusing 0.020-in. rings of
brazing alloy, together with various thicknesses
of nicke! shim material. It was found that a nickel
thickness of ]/8 in. was needed to consistently
PERIOD ENDING JUNE 190, 15953

E UNCLASSIFIED !
E PHOTO Y-B981

§ UNCLASSIFIED B
| proTO Y-8982 §

 

Fig. 8.2. Stainless Steel Tube-to-Fin Joint Nicrobrazed at 1220°C for 30 min, (o) With a large quantity
of brazing alloy, severe tube wall dilution results. (b) With a small quantity of brazing alloy, little tube
wall dilution is evident. Eiched‘ with aqua regia. 50X

75
ANF PROJECT QUARTERLY PROGRESS REPORT

UNCLASSIEED
PHOTO Y-8780

INCHES
i

\tl_’__LLE i l..l..l..l_l._l...i‘. ll’ i f_!_i
0 w 2

 

Fig. 8.3, Carboloy Rod Nicrobrazed to Type
316 Stainless Steel,

produce cracksfree pump seals. Cooling rates of
approximately 1000°F/hr were also employed to
decrease cracking tendencies. One successfully
brazed seal consists of a Carboloy ring brazed to
a nicke! shim which is in turn brazed to the stoin-
less steel body.

It is expected that similar techniques may be
applied advantageously in joining other materials
with widely different coefficients of
expansion,

High-Temperature Brazing Alloy Evaluatien
Tests. The results of several room- and elevated-
femperature butt-braze tensile tests on Nicrobrazed
Inconel and type 316 stainless stee! joints indi-
coted that the low-temperature properties of stain-
less steel joints were definitely superior to those
of Inconel joints. However, the difference de-
creased rapidly above 1200°F, as determined from
a plot of tensile strength vs. testing temperature.

A summary of the existing data is given in Table
8.2. Each value listed in the table is an average
of five tests on 0.252-in.-dia test bars. The tests
were shori-time tensile tests on buit-brazed joints
that were machined under nearly ideal conditions.
It should be remembered that the strength properties
of actual joints may vary quite markedly from those
shown for these more ideal cases.

thermal

CREEP AND STRESS-RUPTURE TESTS OF
STRUCTURAL METALS

R. B, Oliver C. W. Weaver
D. A, Douglas J. W. Woods
Metallurgy Division

The effects of various environments, such as air,
argon, flucrides, and hydrogen, on the creep and
stress-rupture properties of Inconel were investi.
goted, and wide variations in properties were
found. This study is being extended to include
the following additional materials: types 316, 304,
and 405 stainless steel, nickel, and Hastelloy

“C", Table 8.3 shows the results of the tests
made to date, A stress was picked for each ma-
terial which would cause rupture in an argon

atmosphere in the time range of 700 to 1000 hours.
The data seem to indicate that there is little
environmental effect on either type 316 or type
304 stainless steel under the present testing con-
ditions, An A nickel specimen which is still in
test in an air atmosphere indicates that nickel may
hove considerably more creep strength in air than
in any of the other environments presently being
used. If this is true, it is possible that the
strengthening agent or mechanism is the same for
nickel as Incaonel, which exhibits a similar effect.
Fourther tests are in progress so that a more com-
plete evaluation of the environmental effects can
be made,

As a part of this program, fine-grained Inconel
was tested at 3500 psi in alternate cycles of argon
and hydrogen. The first 200 hr of the test was in
argon, ond a constant creep rate was reached
before the first hydrogen cycle was started. Each
hydrogen cycle resulted in on accelerating creep
rate. Each argon cycle tended to maintain es-
sentially the same creep rate as that observed at

TABLE 8.2, TENSILE TESTS OF BUTT-BRAZED JOINTS

 

TENSILE STRENGTH (psi)

 

TEMPERATURE (°F)

 

Stainless Steel Joint Inconel Joint
Room 68,500 35,000
1000 60,500 33,000
1500 26,000 24,500
1700 16,000 13,500

 

 

 

76
PERIOD ENDING JUNE 10, 1953

TABLE 8.3. RESULTS OF CREEP AND STRES5S-RUPTURE TESTS OF STRUCTURAL METALS

 

 

 

MATERIAL STRESS RUPTURE TIME (hr) CREEP RATE (%/hr)
{psi) in Air In Argon | In Hydregen |In Fluorides In Air In Argon | In Hydrogen | In Fluorides

Fine-grained

Inconel 3500 Over 2500 730 275 750 0.002 0.007 0.020 0.015
Type 316 stainless

steel 6300 855 830 770 865 0.009 0.009 0.010 6.020
Type 304 stainless

steel 3500 1410 1100 1400 ¢.008 0.005 0.002
Nickel 1500 80O* 1004 935 0.0002 0.003 0.001

 

 

 

 

 

 

 

 

 

 

*Still in test.

The test was dis-
Metallographic exami-
revealed profuse internal intergranular
cracking, Other tests of fine-grained Inconel ot
3500 psi in argon only were discontinued after
200 and 450 hours. Microscopic examination did
not reveal any intergranular cracking. It has been
shown that, in both argon and hydrogen, failures
will result from initiotion and propagation of inter-
granular cracking. Hydrogen apparently tends to
premote and accelerate the rate of nucleation and
the propagafion of the cracking.

Testing of Inconel in the fused fluoride NaF-
ZrF ~UF , (46-50-4 mole %) is continuing, and the
remaining tests are to be performed in the low
stress range. A preliminary design curve for fine-
grained material is shown in Fig. 8.4, The dotted
portfions of the curves that are below 3500 psi
were obtained by extrapolation. Data for a design
curve for coarse- and fineegrained inconel will be
completed to a stress of 2500 psi in opproximately
three months. Tests at 1500 psi are scheduled,
and it is estimated that about 12 months will be
required for rupture of the specimens.

the end of the hydrogen cycle.
continued after 550 hours.
nation

FABRICATION RESEARCH

E. S Bomar’ R. W. Johnson
J. H. Coobs H. lnouye
Metallurgy Division

A. Levy, Pratt and Whitney Aircraft Division

Extrusion of High-Purity Inconel Tubing. It was
indicated in the previous repori(z)thof the principal
problem involved in extrusion of high-purity cast

 

(2, S. Bomar, Jr., et al., ANP Quar. Prog. Rep. Mar.
10, 71953, ORNL-1515, p. 152.

lnconel was that of lubrication. During this period,
several lengths of Inconel tubing was extruded by
using glass wool on the billet ard rock woo! in the
container. T1he results indicate that malleabilizing
agents are unnecessary in these vaccum melts,
The data are tabulated in Table 8.4.

Analyses of the several heats show that the total
impurities are about 0.4%, of which 0.3% is silicon
and manganese from the ferrochromium used. The
extruded tubes are being drawn to 0.50 in. in out-
side diameter and 0.035 in., in wall thickness for
corrosion testing.

Air Oxidation of Columbium. The oxidation rate
of columbium is linear between 400 and 1200°C.
The logarithm of the rate constant plotted vs. 1/T
is a straight line with a break at 900°C. This
temperature corresponds with o change in the modi-
fication of the Cb,0, from the low, or “'T", form
to the high, or ““H’", form. The energy of activation
between 600 and 900°C was found to be 13,400
cal/mole, Above 900°C, a value of 4350 cal/mole
was obtained. The only oxide found in these tests
was Cb,0,. Embrittiement of the metal occurs ot
approximately 800°C,

Control Rod for the G-E Reactor. The prototype
contral rod, 30 in. long and Y% in. in diameter, was
assembled and delivered for testing. This rod
comprised several segments of hot-pressed 56%
B ,C~44% Fe absorber canned in o stainless steel
tube, However, brazing the hot-pressed mixture to
the tube did not result in a satisfactery bond, so
a decision was made to abandon this method of rod
fabrication.  An alternate method of fabrication
that is being investigated consists of filling o
tube with a mixture of Al-B,C or Cu-B4C powders
and reducing the tube to 1/2 in. in diameter by hot

77
ANP PROJECT QUARTERLY PROGRESS REPORT

CREEP RATE (% /hr) m

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

01 1.0 10
N T
\RUPTURE /" CREEP RATE
< o T
N P
"\\* ***** - .
o
NID N
rofl
) NATINONS N
< \\\ N, \
® I & . X
4000 ; \\ A1 ™ \\ \\\ N
= & e \ \ \ \
2 d N NN
e \ ® @ Q%
@ A N SN ™R
4@ 3000 r b LR RN \-.‘\
& A N NONTNY
@ ] \ \ ™ \:\
v ~d ™ Y
.\\ \\\ ‘\\ \
2000 L Jod \\\ \\;2&._ g
~L \_\\
\\ NN
o R
NN
\\\‘\
1000 .
1 10 {00 100QC 10,000
TIME {hr)

Fig. 8.4, Creep ond Stress-Rupture of Incone! Sheet Tested in o Fluoride Fuel at 1500°F, Heat
NX8004; cold rolled and annealed at 1500°F. Grain size: approximately 90 grains per square millimeter.
Times were measured for 0.5, 1, 2, 5, and 10% extension and rupture vs. stress. Extension measured by
diol gage.

TABLE 8,4, EXTRUSION DATA FOR VACUUM-MELTED INCONEL BILLETS*

 

 

 

BILLET EXTRUSION EXTRUSION UNIT
INGOT NGO, TEMPERATURE®** PRODUCT RATIO EXTRUSION DIE
(°F) PRESSURE (tsi)
7,9,10 2150 Tube 22:1 58 to 72 25 deg cone
8 2200 Tube 23:1 60 30 deg cone
8,2,10 2200 Rod 14:1 45 45 deg \

 

 

 

 

 

 

*Billets 3 in. in cutside diameter and 4 in. leng; front radius ]{4 in,
**Heated in ND-15 salt bath.

78
swaging or hot rolling to consclidate the mixture,
as well as to form a bond between the matrix and
the tube wall. When about 50  vol % of fine B ,C
powder is used, the core matericl should have
nearly the same nuclear cross section as the hot-
pressed B ,C-Fe composition.

Mixtures containing 47% Al-53% B,C and 75%
Al-25% B ,C were prepared by mixing ~100-mesh
atomized alummum powder, grade ‘‘C'" electrolytic
copper powder, and metallurgicdl-grade B C powder
for 1 hr in an oblique blender. The B,C was
ground for 16 hr ond contained about 60% -~325
mesh particles. The tap densities of these mix-
tures were 63% for the B C-Al mixture and 50%
for the B C-Cu mixture. The B ,C-Al mixture was
used w:'rh % -in.~dia tubing and wcxs reduced a total
of 40%. However, since the B ,C-Cu mixture re-
quired more reduction for complete consolidation,
it was loaded in a ¥-in.-dia tube and reduced o
total of 58%. The finished rods were }/2 ine in
diameter.

The tubes used for both mixtures were AlS! type
304 stainiess steel with an original wall thickness
of 0.049 inch. They were bright annealed in dry
hydroger, loaded by tapping for 3 to 5 min, sealed,
and evacuated through a porous sintered-metal
plug. The 84C~AE rods were reduced at 600°C in
two steps, while the B C-Cu rods were reduced at
1000°C in four steps.

A total of seven experlmentai rods has been
prepared, including one 20-in.-long rod of each
material, for reactivity tests in the LITR. In
addition, short sections about 6 in. in length of
each type of rod have bzen prepared for thermal
conductivity ‘and strength tests. In each cose the
core material was consolidated to 95% or more of
theoretical density and showed good distribution

PERIOD ENDING JUNE 10, 1953

of B,C. The B,C-Cu core material showed good
bondlng to the stainless steel tube wall in the as
swaged condition, but the B C-Al mixture did not.
However, after heating to 700°C for 15 min, the
B,C-Al exhibited excellent bonding.

The stainless steel tube wall was increased in
the swaging process from 0.049 in. to a final thick-
ness that varied from 0.053 to 0.057 in.; thus
the final wall thickness was somewhat less than
the required 0.065 inch., However, if the original
wall thickness were 0.058 in., an increase of the
same proportion during the swaging would give «
final thickness very close to that required.

HOT-PRESSED PUMP SEALS

E. S. Bomaor - R. W. Johnson
J. H. Coobs H. Inouye
Metaliurgy Division
A, Levy

Pratt and Whitney Aircraft Division

Fabrication of Small Compacts with Mo5,. Addi-
tional small experimental compacts with 14 vol %
MoS., were prepared by hot pressing. The data from
these experiments are given in Table 8.5. The
first four experiments showed that the 95% Ag~5%
Cu alloy composition could not be prepared by
using coarse silver powder. In all cases in which
the pressing temperature exceeded the eutectic
temperature of 780°C, o portion of the ailoy was
lost by extrusion of the eutectic composition past
the rams. A single compact pressed at 770°C with
no loss of material was heated at 820°C in helium
for a total of 36 hr in an attempt to produce a
homogeneous alloy matrix. This treatment was
not successful. Thus, it was necessary to prepare

TABLE 8,5. EXPERIMENTAL COMPACTS OF FACE SEAL COMPOSITIONS

 

 

 

MATERIAL HOT~PRESSINC; DENSITY REMARKS
TEMPERATURE (VC) (% of theoretical)
(Coarse 95% Ag—5% Cu)-MoSz. 790 94.5 8% loss
280 97.3 T0% leoss
a50 95.9 4% lass
770 94.6 No lass
Type 304 s;tuiniess steef-MoSz 1120 84.0 Highly magnetic
1220 94.0 Highly magnetic
Molybdenum ‘ 1530 92.0

 

 

 

 

79
ANP PROJECT QUARTERLY PROGRESS REPORT

this composition by using the ultrafine precipitated
silver powder.

Metallographic Exaemination of Compacts with
MoSz. The compacts of stainless steel and MoS,
are being studied metallographically to determine
the extent of reaction, if any, between the MoS,
and stainless steel. The sulfide phase seems to be
interspersed with a metallic phase, which becomes
more prominent as the pressing temperature is
increased.  X-ray diffraction analysis revealed
only ferrite; the identity of the sulfide phase could
not be established and austenite could not be de-
tected. |t thus seems that molybdenum is dissolved
or nickel is removed from solution, or both, with
the consequent formation of ferrite. Thus the com-
position may be unsatisfactory for the pump seal
tests.

Metallogrophic  examination showed that the
Cu-MoS, compact and the Ag-MoS, and (95% Ag-
5% Cu)-MoS, compacts prepared with precipitated
silver powder had excellent structure. The flake-
like particles of MoS, were oriented perpendicular
to the direction of pressing and were distributed at
random in the continuous, homogeneous, highly
dense matrix.

The coarse silver compact, on the other hand,
had a rather poor structure in which the Mo3, was
distributed exclusively at the particle boundaries,
This compact showed rather poor interparticle
bonding and had poor physical properties.

Fabrication of Face Seal Rings. On the basis of
results obtained with the small compacts, face
seal rings 37/8 in. OD by 23/8 in, iD by 7/8 in. long
with the compositions shown in Table 8.6 were
fabricated by hot pressing at 2500 psi. The Cu-MoS,
composition seemed quite satisfactory except that
the density was not so high as desired. A slightly
higher pressing temperature might have increased

the density, but during the second hot pressing
operation the die broke. Since the compact ob-
tained could be used, the operation was not repeated,

The Ag-MoS, compacts were prepared with coarse
silver powder, and it was necessary to approach
quite close to the melting temperature of silver to
obtain the low porosity desired. Difficulty in
accurately controlling the temperature of the large
die assembly resulted in a substantial loss of
material from one of the compacts by extrusion past
the roms ond infiltration of the graphite die body.

The (95% Ag—5% Cu)-MoS, compact was prepared
with the fine precipitated silver powder which
gave excellent results in small compacts. However,
the density of the first compact was rather low,
and in an attempt to obtain a density of 95%, the
maximum temperature was exceeded ond again a
subsiantial amount of the material was lost by
extrusion ond infiltration. All compacts were
judged to be satisfactory for testing and were
delivered.

An additional request was received for a number
of smaller rings of the Cu-MoS, composition to be
used in a packing gland to seal the shaft of a
fluoride pump. A total of 20 rings, ]]}/16 in. OD by
]3/16 in. ID by ¥ in. thick, is required for the first
test, Six cylinders, 13/4 in. QD by ”/8 in. 1D by
2 in, long, from which sufficient rings could be
machined, were prepared from the 92% Cu--8% MoS$,
composition, |t was found that densities as high
as 97.5% of theoretical could be attained by hot
pressing the rings at 950 to 960°C and cooling
under pressure.

Surface Sulfurization of Molybdenum. Efforts to
form MoS, on the surface of molybdenum have been
continued, The pyrex reaction vessel used initially
was replaced with a vessel of material that permits

TABLE 8.6. FACE SEAL TEST RINGS

 

 

P A - NSI
COMPOSITION TEM EOR TURE DE TY REMARKS
(wi %) ") (% of theoretical)
Cu—-8% M052 910 89.5
200 ~00 Die broken
Ag-9% M052 260 92.5 15% weight loss
955 95.0 3% loss
(95% Ag~5% Cu)-9% MaS, 835 85.5 No loss
880 86.5 20% weight loss

 

 

 

 

80
the use of a higher temperature. Also, H,S was
used in one experiment in place of the sulfur vapor.,

For the experiment in which H,5 was used, the
molybdenum was introduced in the form of an 18-in.-
long by 0.057-in.-dia wire. Thewire was suspended
in a vertical tube furmoce that had a temperature
gradient of from room temperature to 1020°C, The
H,S was passed through the tube at the rate of
1 liter/min during an exposure of ]!/2 hr at tempera-
ture., The exposure was terminated because of
the escape of H,S§ when the gas outlet became
plugged with condensed sulfur,

Data on the free energy of formation of H_§(3)
indicate that at 1000°C the equilibrium pressure of
sulfur gas is about 0,08 atomic weight. An appreci-
abl e portion of the decomposed H,S must have been
swept out of the reaction chamber before recombi-
nation could take place. The molybdenum wire
reacted with the gaseous atmosphere at tempera-
tures of from 535 to 1020°C. X-ray diffraction
patterns showed the reaction layer to be MoS,.
However, the difficulty described here plus the
psychological and physiological hazards associated
with the escape of H,S make the use of sulfur
vapor more attractive, _

Two furnaces were arranged to permit the heating
of a molybdenum sample contained in a tube to one
temperature and the heating of a sulfur reservoir at
the other end to a lower temperature. Type 446

 

3y, . T. Ellingham, J. of the Soc. of Chem. Ind.
63-65, 125-133 (May 1944).

PERIOD ENDING JUNE 10, 1953

stainless steel fubing was used for the container.
Data from the four exposures made by using this
arrangement are summarized in Table 8.7. These
tests indicate that a MoS, layer can be formed by
exposure to sulfur vopor and that a degree of con-
trol may be exercised by varying the molybdenum
temperature and the sulfur vapor pressure. Further
checks will be made to determine whether the
temperature at which the MoS, is formed has an
appreciable effect on the nature of the compound
formed.

HIGH-CONDUCTIVITY METALS FOR RADIATORS

E. S. Bomar R. W. Johnson
J. H. Coobs H. Inouye
Metallurgy Division

A. Levy
Pratt ond Whitney Aircraft Division

Finned tubes could be advantageously used for
the air radiators that will be required by most
nuclear aircraft reactors to transfer the reactor
heat to a turbojet air stream. The design of the
fins requires that they be 0.010 in. or less in
thickness, that they be oxidation resistant at
1500°F, and that they possess high thermal con-
ductivity. The object of the initial investigation
has been to determine the most promising method of
fulfilling these specifications.

The mu’ior effort, to date, has been directed
toward the protection of copper by cladding it with

TABLE 8.7. REACTION OF MOLYBDENUM WITH SULFUR VAPOR

 

 

SUL.FUR SULFUR VA.POR MOLYBDENUM TIME AT
TEMPERATURE PRESSURE TEMPERATURE | TEMPERATURE REMARKS

(°C) (mm Hg) (°C)* {min)

445 760 800 60 No Mc:oS2 formed

490 1441 900 120 Complete conversion of 4-g molybdenum
sample; layer of Fel formed on inner
surface of tube

490 1441 900 B Powdery MoS, % _ in. thick on molybde-
num compact .

460 948 900 10 Porous molybdenum sample, ]/Ié'in.‘
layer of MoSz; dense molybdenum
sample, slight reaction

 

 

 

 

 

*Data from Lange's Handbook of Chemistry, Sixth Edition, p. 1421,

81
ANP PROJECT QUARTERLY PROGRESS REPORT

the specified maoximum permissible thickness of
0.002 in. of oxidation-resistant materials. Some
attention has been given fo the use of oxidation-
resistant copper alloys as the fin material or as a
cladding on copper. Type 310 stainless steel clad
silver was also investigated. The results, to date,
are of a preliminary nature.

Chromium-Clad OFHC Copper. Both vapor-de-
posited and electroplated chremium claddings on
oxygen-free high-conductivity (OFHC) copper were
tested at 1500°F for 100 hours. The results indi-
cate that this combination offers protection to the
copper if the plate does not peel off or crack during
temperature cycling. Diffusion does not appear to
be o problem. Samples of ‘‘ductile’’ chromium on
copperhave been requested from the Ductile Chrome
Process Company.

Chromium and Nickel Electroplate on Copper. A
composite that was intended to be 0.0005 in. of
chromium plus 0.0015 in. of nickel on copper was
prepared; however, measurements indicated that
the composite tested actually had 0,002 in, of
nickel and less than 0.0001 in, of chromium. When
tested, as plated, at 1500°F, the surface showed a
few areas of failure after 100 hours. A 500-hr test
showed no failure. After a pretreatment of 24 hr
at 1000°C in hydrogen, a test for 100 hr showed no
surface failure, but a test for 500 hr showed com-
plete failure because of separation of the plate
from the core. Metallegrophic examination of the
specimens tested for 100 hr showed that the copper
core had suffered exireme diffusion and that
numerous voids had appeared near the original
interface.  The amount of unaltered copper was
approximately 25% of the original core,

Incone! Cladding on Copper. In 100-hr tests of
Inconel-clad copper, diffusion occurs that affects
a zone that is from 0 to 4 mils thick. In 500-hr
tests, the characteristic color of the copper nearly
disappears, and the cladding is penetrated by a
copper-rich phase thai extends to the surface in
some locations., The 100-hr tests showed that a
core of commercial copper develops substantially
more voids near the interface than does OFHC
copper. In all instances, surface failure (that is,
oxidation effects) is limited to the development of

copper oxide nodules, This type of failure is

82

attributed to pinholes developed in the cladding and
rolling of the composite. Suitable rolling tech-
niques reduce the number of pinholes developed to
the extent that this type of failure becomes in-
consequential.

The wide variations in thickness of the diffusion
zones in the several different composites tested
suggest that the method of fabrication is responsi-
ble for the diffusion. It is thought, at this time,
that assembling the composites with both the copper
and the cladding as clean as possible promotes
diffusion and that consequently a ‘‘not so clean”
interface is desirable.

Type 310 Stainless Steel Cladding on Copper.
Tests of type 310 stainless steel cladding on copper
have been completed, These tests were conducted
in air for 100- and 500-hr periods. The results
indicate that the diffusion which occurs is pre-
dominantly of a different type than that found in
specimens of Inconel cladding on copper. In most
cases the interface becomes highly irregular, to
the extent that islonds of o copper-rich phase
appear in the cladding., In only one instance did a
test piece show alteration of the copper core. The
number and the size of the copper-rich islands
that appear in the cladding are greater when com-
mercial copper is used than when OFHC copper is
used. All surface failures were pinholes such as
found with Inconel cladding.

Type 310 Stainless Steel Cladding on Silver,
Tests of type 310 stainless steel cladding on
silver were made at 1500°F for 100- and 500-hr
periods., The material tested showed a cladding
thickness of 5 mils. There was no failure in the
cladding, and diffusion at the end of 500 hr was
negligible.

Copper Alloys. Copper alloys were tested for
100-hr at 1500°F with the results given in Table
8.8. The thermal conductivity at 20°C of the
copper-aluminum alloys is about 0.20 cgs units,
with a positive temperature coefficient. Extrapo-
lated data indicate a value of about 0.4 cgs units
at 1500°F {for Inconel and type 310 stainless steel
the value is about 0,06 cgs units). These copper-
aluminum alloys may prove to be adequate as they
are, but the present plans are to clad copper with
both the 6% and the 8% aluminum-bronze alloy.
PERIOD ENDING JUNE 10, 1953

TABLE 8.8, TESTS OF COPPER ALLOYS IN AIR FOR 100-kw PERIDDS

 

 

ALLOY LOSS OF COPPER CEMARKS
{%)

99% Cu—1% Be 7.0 Uniform block oxide that scaled freely
98% Cu~2% Al 28.5 Thick uniform black oxide
96.5% Cu~3.5% Al 1.0 Numerous areas of localized oxidation
95% Cu—5% Al | 0.16 Few areas of lacalized oxidation
94% Cu~6% Al | No change Polished sucface bacame dull
92% Cuwm8% Al No change Polished surface remained bright
90% Cu~10% Al Weight increased Visible coating of Al,O, found
Copper 42.4 Thick black oxide

 

 

 

SOLID FUEL FORMED IN SPHERES

E. S. Bomar R. W. Johnson
J. H. Coobs H. Incuye
Metatlurgy Division
A. Levy

Pratt and Whitney Aircraft Division

The initial work on the preparation of spherical
particles of solid fuel was reported pre\fiousl),f.(“‘3
For the previous work, alioys containing 5% uranium
in copper, nickel or molybdenum were subjected to
four different methods of processing. Only one
method, spraying from a metallizing gun, gove an
appreciable yield of particles that approached
spherical geometry. :

Additional work has been bused on modifications
of earlier methods or the use of additional methods,
Some success has been achieved. Briefly, the
work has included:

1. heating to above the alloy melting point in a
vacuum, '

2. substitution of a three-phase three-electrode arc
for o single-phase double-electrode arc as a
heat source for particles dropped through the arc,

3. forcing the molten alloy through a small orifice,

4. letting precut particles of the alloy settle
through a molten salt bath having a temperature
gradient extending beyond the liquidus and
solidus of the alloy.

Each of these methods of preparing spherical
particles is discussed below, and additicnal de-
tails, including photomacrographs of the particles

 

(4YE_S. Bomar, J. H. Coobs, and H. Inouye, ANP Quar.
Prog. Rep. Dec. 10, 1952, ORNL-1439, p. 155.

prepared by these and the previous methods, are

presented in other reports.(sfl'_“ There are no plans

at present for pursuing further the problem of fabri-
cafing spherical particles., However, two odditional
methods which might prove of interest are:

1. passing of a high velocity stream of inert gas
through an electric arc formed between elec-
trodes of the desired alloy, ()

2. forming spherical geometry from small right
cylinders by random impacts in g hammer mill,

The second method was suggested in ¢ private

communication from Doyle Rauch, Research Divi-

sion, NYO-AEC. This method and the more con-

ventional ball-bearing manufocturing techniques are
used o produce the spheres used in ballpoint pens.

Heoting Under Yacuum. Particles clipped from a
0.010-in. wire of 5% U-95% Ni alloy were degassed
under a vaccum of 1077 wmm Hg ot 900°C  while
resting on an alumina plate. They were withdrawn
from the hot zone of the furnace ond examined
through a glass port, still under vacuum. The
particles appeared bright and clean. The furnace
temperature wos then raised to 1450°C, and the
charge was gradually moved into the hot zone.

The vacuum did not exceed 5 x 107¢ mm Hg during

any of the heating cycles. After opproximately

5 min at temperature, the charge was wifhdrfiwn,

alfowed to' conl, and removed from the vacuum

 

Slg, 5. Bomar and H. Incuye, Fubrication of Spherical
quflcfes, ORMNL CF-52-11-7 {Mov, 1, 1952).

}A Levy, Fabrication of Spherical Forticles, ORNL
CF.53.3-183 (Murah 19, 19531,

(Y)H. H. Hausner and H, Mansfield, Avomization Method
of Making Urdgnivm Powder, NY0-1133 {Aug. 7, 1950).
ANP PROJECT QUARTERLY PROGRESS REPORT

furnace for examination. Melting had occurred, but
the formation of a surface film dgain prevented
spheroidization.

Three-Phase Carbon Arc, Three carbon arcs
were symmetrically arranged about a common axis
in a housing purged with tank argon. Passage of
cut particles of 5% U-95% Cu and 5% U-95% Ni
alloys through the zone of the arc produced a very
low yield of melted particles ot currents of up to
35 amperes. At 40 to 45 amp, the major portion of
the particles melted. Many particles were flattened
when they struck the bottam of the heusing because
there was insufficient free fall. As in the eorlier
two-electrode set up, alignment through the arc
was very critical. Examination of the melted parti-
cles revealed surface oxidation,
depletion of uranium,

porosity, and
Any or all of these effects
may have been due to inadequate control of the
atmosphere,

Molten Alloy Passed Through a Small Orifice,
Spherical particles were produced, with moderate
success, by forcing molten uranium-copper alloy
through o small orifice in the base of a refractory
container. A positive pressure of argon was re-
quired to force the alloy through 0.020., 0.025., or
0.031-in. holes in alumina crucibles. For a run, a
crucible 10 to 15 in, long and ]/2 in. in diameter
was passed through a rubber stopper into a quartz
tube containing argon. A small graphite sleeve
placed over the lower 3 in. of the crucible served
as a heat source when expased to the field of an
induction coil. Upon heating to approximately
1200°C, particles ranging in size from 0.010 to
0.020 in. were ejected from the crucible, The
orifice size appeared to have no effect on the
resulting particle size. Particle surfaces showed
little oxidation. Macroexamination revealed a cast
dendritic structure of primary copper plus UCu ,-Cu
eutectic. Some of the surface defects were found
to be due to shrinkage upon solidification,

Settling of Particles Through a Molten Salt.
BaCl, was charged into a graphite crucible, which
was, in turn, positioned in a quartz tube that was
being continuously purged with argen. The upper
half of the crucible was heated by induction to
produce a temperature gradient in the molten salt
that ronged from 962 (mp of BaCl,)) to 1200°C.
Particles of 5% U-95% Cu alloy clipped from 0.010-
in. wire were partly spheroidized under these con-
ditions, When the maximum temperature of the salt
was increased to 1300°C, most of the particles

were spheroidized, Slight surface oxidation of the
particles occurred.

INFLAMMABILITY OF SODIUM ALLOYS

G. P. Smith L. L. Hall
Metallurgy Division

Studies have been initiated to determine to what
extent the inflammability of jets of liquid sodium
can be reduced by alloying the sodium with some
of the heavy metals. In the initial phose of this
project, a large number of alloys will be surveyed
qualitatively. The qualitative data will serve to
indicate the worthwhileness of a detailed study of
the such as porticle-size
distribution, rate of evaporation, ond gas-phase

reaction kinetics, which probably influence the

numerous variables,

inflammability of jets of liquid alloys.

The following method of measurement is being
used. A 15-cm® sample of alloy is sealed in a
metal capsule under 1 atm of argon. The capsule
is made so that one end may be broken off to form
a small hole. After the capsule is heated to a
temperature of 700 to 900°C, the end is broken
off, and the argon pressure developed by heating
at constant volume ejects the molten alloy through
the small hole to form a jet. Visual observations
are made of the extent of combustion.

One unalloyed sodium jet was burned at 750°C
and three jets were burned at 800°C. The dif-
ference in temperature did not make any significant
difference in the flammability. In all cases, the
jet burned with a brilliant yellow flash. Copious
quantities of white smoke were produced. Motion
pictures were taken of a speed of about 700
frames/sec of one of the burning jets as it left the
capsule. The only source of illumination for these
pictures was the light produced by the burning
sodium, It was observed that the jet consisted of
many, small, brightly glowing droplets accompanied
by a less luminous cloud. This cloud may have
consisted of very fine particles or may have been
sodium vapor lighted by resonant radiation. When
the cluster of droplets forming the jet had moved
a few centimeters away from the capsule, there
was a sudden increase in light intensity.

Three jets of unalloyed lead at 800°C did not
take fire, and most of the lead was recovered in an
unoxidized condition. At 950°C, a jet of unalloyed
lead oxidized considerably but did not take fire.

Alloys of sodium containing 10 and 50 wt % lead
appeared to burn more intensely at 800°C than did
pure sedium. Additions of 5 and 10 wt% mercury
did not seem to change the burning of sodium jets
to an appreciable extent at 750 and 800°C. A
single test was made at 800°C with sodium con-
taining 10 wt % zine. This alloy burmed at about
the same intensity as did pure sodium.

The apparatus being used for these measurements
is of a very simple design. A small furnace for
heating the capsules and the mechanism for break-
ing off the ends of the capsules are mounted on a
transite board which rests on top of a 55-gal steel
drum.  The drum, which serves to confine the
burning alloy, has a long narrow safety glass
window down one side for observation. This ap-
paratus is enclosed in a transite hood of temporary
construction, After sodium or one of its alloys has
been bumed, the dense cloud of white smoke
(probably a mixture of sodium oxide, hydroxide, and
carbonate) which is produced is removed through
the exhaust system of the building.

A new apparatus is being constructed in which
the combustion of sodium dlloy jets can be studied
in air at low pressures with controlled water-vopor
concentration. The pumpose of this apparatus is to
simulate more nearly atmospheric conditions at
high elevation.

CERAMICS
L. M. Doney J. R. Johnson
J. A, Griffin A. ) Taylor

Metallurgy Division
Development of Cermets. Further developments
of the silicon carbide—silicon fuel elements in-
clude the following:

1., The inclusion of 10% USi, in the center
portion of a cross-shaped element has been suc-
cessfully achieved. A layer of material opproxi-
mately l/16 in. thick that does not contain fuel
completely surrounds the fuel-bearing portion.

2. lsostatic techniques for pressing the porous
carbon buse material have been initiated.

PERIOD ENDING JUNE 10, 1953

3. A new means of fabrication has been developed
that utilizes a dipping technique and induction
heating for melting the silicon,

4, Preliminary autoradiographic tests have been
developed to show the location of the fuel after
impregnation with silicon has faken place.

Effect of Heating Raote on the Beryllium Oxide
Blocks for the ARE. Experiments were carried
out to determine the effect of hieating rate on the
full-size hot-pressed beryllium oxide blocks in the
as-received condition. All heating was carried out
in a large, Harper, glo-bar furnace; the samples
were shielded from the direct radiation of the
heating elements by silicon carbide plates 3/4 in,
thick. Several of the blocks were checked for
incipient cracks before the experiment was started,
but no cracks were found in any of the blocks
checked.

In the first experiment, two blocks were placed
in the furnace and heated at 100°C/hr to 1500°C;
the furnace was then shut off with no holding time
at the peak temperature. The cooling rate was
slightly over 100°C/hr down to 1300°C, but it was
less thon 100°C/hr from 1300°C down to room
temperature. Both blocks were intact at the end
of the experiment and had no visible cracks.

Two unfired blocks and one from the first test
were then placed in the furnace, heated to 1500°C
at 200°C/hr, and cooled as before. Both new
pieces were cracked through ot the center, per-
pendicular to axis; the lower half of one was also
split parallel to axis. However, the piece which
had been heated previously was unchanged, In a
third test, the second heot treatment was repeated,
but this time, no pieces were broken.

It cannot be claimed that enough samples were
used to allow definite conclusions to be made; the
results should be considered only as indicative of
a irend., However, there is some evidence that
these pieces do have internal strain and that their
thermal endurance might be improved by an initial
heat treatment at a very slow heating rate. On the
basis of these few tests it would appear that a
safe heating rate for the as-received blocks is

below 200°C /hr.
ANP PROJECT QUARTERLY PROGRESS REPORT

9. HEAT TRANSFER AND PHYSICAL PROPERTIES
H. F. Poppendiek

Reactor Experimental Engineering Division

Some preliminary density and viscosity measure-
ments were made on the new ARE fue! NaF-ZrF .
UF, (53-43-4 mole %). The viscosity was found
to vary from about 13,5 cp at 620°C to about 8.8 cp
at 757°C. The density was found to be about 3.35
g/em® at 653°C. The enthalpies and heat capaci-
ties of lithium hydroxide and sodium hydroxide
eutectic and of fuel mixture NaF-KF-LiF-UF,
(10.9-43,5-44.5-1.1 mole %) have been obtained.
The heat capacity of LiOH-NaOH eutectic was
determined to be 0,60 + 0.02 cal/g-°C over the
temperature range 260 to 850°C; the heat capacity
of the NaF-KF-LiF-UF, was found to be 0.44
+0.03 cal/g-°C over the temperature range 500 to
1000°C.  The surface tension of NaF-ZcF -UF,
(50-46-4 mole %) was determined to vary from 160
dynes/cm at 530°C to 115 dynes/cm at 740°C.

Experimental velocity distributions have been
determined in a pyrex convection harp system; the
profiles differed significantly from the parabolic
profile characteristic of isothermal laminar flow,
At Reynolds' moduli as low as 85, large scale
viscous eddies were observed that were considered
to be due to thermal turbulence. Forced convection
heat transfer data for NoF-KF-LiF flowing in a
small Inconel tube were found to lie 50% below the
data previously obtained in q nickel system. Some
evidence of a thin, green corrosion layer was found
at the surface of the Inconel. Such a layer or a
nonwetting phenomena could account for additional
thermal resistance in the thermal circuit,
reactor coolant is a
function of coolant duct dimensions and spacing,
the amount of heat to be removed, the total coolant
temperature rise, and the physical properties of the
coolant. A mathematical onalysis of the effective-
ness of reactor coolants has been made with the
aid of IBM machines; this analysis can be used to
optimize coolant systems in solid-fuel
systems.  An apparatus which
transfer

The effectiveness of a

reactor
simulgtes heat
in a circulating-fuel piping system by
passing an electric current through a flowing
electrolyte has been successfully operated. An
experiment has been analyzed in which heat was
transferred in an annular system from a molten
fluoride to NaK, The data are compared to Hausen’s
and to Colburn’s equations,

86

THERMAL CONDUCTIVITY

W. D. Powers S. J. Claiborne
R. M. Burmnett

Reactor Experimental Engineering Division

The longitudinal thermal conductivity apparatus
for measurements on solid and liquid metals has
been completed, and preliminary runs have been
made. The solid samples used are in the form of
rods, the liquid samples ore placed in cylindrical,
thin-walled capsules. The heat used is passed
through a heat meter before it flows through the
sample being studied. The conductivity of the
sample is determined by measuring the temperature
gradient within the sample and the heat flow
through the heat meter. To eliminate radial heat
losses, guard heating is provided. All heaters are
individually controlled by variacs.

The uninsulated and uninstrumented conductivity
device is shown in Fig. 9.1. The two top calrods
guard-heat the sample heater, and the other eight
calrods are aftached to the eight annular guard
disks. A copper disk, which is cooled with water
(the heat sink), is located at the bottom of the
apparatus. When in operation, the entire device is
surrounded by a metal shell which is filled with
fine Sil-o-cel powder. Platinum~platinum-rhodium
thermocouples are welded to the sample at the
l[evel of each guard heater disk; similar thermo-
couples are attached to the guard heaters. After a
sample heater has been turned on and o tempera-
ture gradient has been established in the sample,
the guard heaters are turned on and adjusted so
that no temperature difference exists between the
sample and the guard heater disk at each level.
At present, the radial temperature differences can
be held to within less than 1°C when a total axial
temperature difference of 300°C exists.

Preliminary thermal conductivity measurements
have been made on type 316 stainless steel. A
value of 10 Btu/hr-ft2.(°F/ft) at 750°F, which
compares favorably with the literature value, was
obtained, Before the thermal conductivity of
liquid lithium is determined, the apparatus will be
checked with molten sodium.
PERIOD ENDING JUNE 10, 1953

- .. UNCLASSIFIED
: ' PHOTO 202"

    
     
    

SPECIMEN
HEATERS

 

SPECIMEN

GUARD
HEATERS

 

HEAT SINK

Fig. 9.1. Longitudinal Thermal Conductivity Device.

87
ANP PROJECT QUARTERLY PROGRESS REPORT

DENSITY AND VISCOSITY
S. |. Cohen T. N. Jones

Reactor Experimental Engineering Division

A new drybox designed to contain instruments
for measuring densities, viscosities, and surface
tensions by each of two methods has been com-
pleted. Included in the system is an elaborate gas
train for removing the undesirable impurities from
the cylinder gas used to provide the inert atmos-
phere. This drybox will be described in detail in
a future report,

A study of the density and the viscosity of the
fluoride fuel Nc:F-ZrF“-UF4 (53-43-4 mole %) is
being made and preliminary data have been ob-
tained. The viscosity was found to vary from about
13.5 ¢p ot 620°C to about 8.8 cp at 757°C. The
density was found to be about 3.35 g/cm? at 653°C.
These preliminary viscosity and density measure-
ments are in good agreement with the corresponding
values for the fluoride fuel NaF-ZrF, -UF, (50-
46-4 mole %), whose composition is very similar,

Several glass mockups of possible capillary
viscometer designs have been fabricated and
tested. However, no final design for the device
has been decided upon to date. The gas-bubble
densitometer has been completed and tested. M
will be used in conjunction with the plummet
method to determine densities.

HEAT CAPACITIES OF LIQUIDS
W. D. Powers G. C. Blalock

Reactor Experimental Engineering Division

Investigations of the enthalpy and heat capacity
of hydroxides and fluoride salt mixtures are being
continued. The following measurements have been

made:{1+2) for NaOH-LiOH eutectic (73-27 mole %)
at 260 to 850°C,

HT(quuid) Hooc(solid) 44 + 0.60T ,
c =

» 0.60 = 0.02 ,
and for NaF-KF-LiF-UF , (10.9-43,5-44.5-1.1 mole
%) at 500 to 1000°C,

Ho(liquid) ~ Hgo (solid) = 21 + 0.44T ,
C =

» 0.44 + 0.03 ,
where H is the enthalpy in cal/g, T is the tempera-
ture in °C, and <, is the heat capacity in cal/g-°C.

i

 

“)W. D. Powers and G, C. Blalock, Heat Capacity
of the Eutectic of Lithium Hydroxide and Seodium Hy-
droxide, ORNL CF-53-5-103 (May 18, 1953}.

(2w, D. Powers and G. C. Blalock, Heat Capacity of
Composition No. 14, ORNL CF-53-5-113 (May 18, 1953).

88

SURFACE TENSIONS OF FLUORIDES
S. I. Cohen T. N. Jones

Reactor Experimental Engineering Division

The surface tension of the fluoride fuel NaF-
ZrF ,-UF, (50-46-4 mole %) was measured with a
commercial tensiometer, which was
modified for the experiment. A spring with a low
spring constant was fabricated and used to give
greater sensitivity to the instrument. The liquid-
surface temperatures were obtained from tempera-

somewhat

ture profile measurements made with a traversing
thermocouple probe. The surface tension was

found to vary from 160 dynes/cm at 530°C to 115

DWG. |89f9A

dynes/cm at 740°C (Fig. 9.2).

160

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

140
E 120
°
c
-
ol
S 100
10
o |
500 600 700 800
TEMPERATURE (°C)
Fig. 9.2. Surface Tension of i*lc|F-ZrF4-UF4

(50-46-4 mol %).

VAPOR PRESSURES OF FLUORIDES

R. E. Moore R. E. Traber
Materials Chemistry Division
Additional vapor pressure determinations of

fluoride mixtures and components have been made
by the method described by Rodebush and Dixon. (3)
Data for the vapor pressure of NaF-ZrF -UF,
(53-43-4 mole %), the composition presently con-
sidered as the fuel for the ARE, range from 4.5 mm
Hg at 790°C to 39 mm Hg at 958°C. The data are
best represented by the equation

-7160
=+ 7.37 ,

T(°K)
from which the heat of vaporization of 33 kcal/mole
is obtained.

|°910 P(mm Hg)

 

(S)W. H. Rodebush and A, L. Dixon, Phys. Rev. 26,
851 (1925).
Work has begun on the mixture NaF-ZrF -UF
(65-15-20 mole %), which is the new composition
proposed for the addition of enriched uranium to
bring the ARE to criticality., Preliminary results
indicate that the vapor pressure of this composition
at 900°C is approximately 8 mm Hg.

Vapor pressure data for the mixture NaF-KF-
ZeF -UF, (5-51-42-2 mole %) were given in @
previous repor’r.(‘” This composition was the first
of the ZrF ,-bearing mixtures to be prepared, and it
was the only mixture prepared before the hydro-
fluorination treatment became the practice. Be-
cause it had not been subjected to such treatment
and because the zirconium tetrafluoride in use at
the time was not so pure as that used in mixtures
prepared more recently, vapor pressure measure-
ments were repeated on a new sample.

The recent data, which range from 8 mm Hg at
900°C to 64 mm Hg at 1123°C, fall considerably
below the values obtained previously, A small
amount of a more volatile impurity (possibly zir-
conium tetrachloride) in the zirconium tetrafluoride
used to prepare the first samples may account for
the difference in pressures. The equation

~6789

o 4 6,743,
T(°K)

which was obtained from data on the new mixture
over the range 900 to 1123°C, is believed to be
more nearly correct than the equation obtained
previously., The heat of vaporization derived from
the equation is 32 kcal/mole.

It is expected that mixtures containing beryllium
fluoride will be subjected to investigation in the
future. During the quarter, data were obtained for
beryllium fluoride at higher temperatures than those
reported previously.(4) The two sets of data are
in satisfactory agreement, although the slope of
the plot of log P vs. the reciprocal of the absolute
temperature is slightly greater when values ob-
tained af the higher temperatures are included.
The equation 4

log10 'P(mm Hg)

~9236
~ T(°K)
was derived from data given before, as well as from

recent dafa, over the temperature range 891 to

1073°C. The heat of vaporization (42.5 kcal/mole)

long(mm Hg) + 9.136

 

(4R, E. Maore, ANP Quar. Prog. Rep. June 10, 1952,
ORNL.-1294, p. 150-151.

PERIGD ENDING JUNE 10, 1953

and the boiling point (1207°C) were derived from
this equation.

VELOCITY DISTRIBUTIONS IN THERMAL
CONVECTION LOOPS

D. C. Hamilton F. E. Lynch
L. D. Palmer

Reactor Experimental Engineering Division

The objective of the investigation of velocity
distributions in thermal convection loops is to
determine the accuracy with which mean cirevlation
velocities in thermal convection harps can be pre-
dicted by the numerical solution of the heat transfer
equation obtained by using experimental wall
temperature data, The velocity distribution has
been measured in the cold leg of a 17-mm-1D pyrex
hamp similar to the one described previously,(5?
Velocities were determined from observations of
suspended particles illuminated by a collimated
light beam which passed through the center of the
tube. The internal wall temperature was measured
at various positions in both the hot and the cold
legs. These temperature data are being used to
obtain a numerical solution to the heat conduction
equation, When the numerical solutions are comr
pleted, the resulting predicted velocities will be
compared with the observed velocities to evaluate
the method.

Observed velocity distributions are presented in
Figs. 9.3 and 9.4 for two values of the Reynolds’
modulus. At even greater vaolues of Reynolds’
modulus, negative velocities were observed in the
center of the tube. The large deviation of the
velocity distribution from the fully developed
isothermal laminar distribution is apparent. Large-
scale viscous eddies were observed to be super-
imposed on the mean flow. The swirling or eddying
increased in intensity and decreased in size as the
temperafure difference (or Reynolds’ modulus) was
increased. This eddying appeared to be sufficiently
intense to contribute to the transport of momentum
and heat. Since no data are available on ‘‘thermal
turbulence’’ the accurate prediction of the hegt
and momentum transfer in such o system is ex-
tremely difficult,

 

Blp, c. Mamilton, F. E, Lynch, and L. D. Palmer,
ANP Quar. Prog. Rep. Dec. 10, 1952, ORNL-1439,
p. 182,

89
ANP PROJECT QUARTERLY PROGRESS REPORYT

UNCLASSIFIED

 

  

DWG.1929114
2z ] 1 1 T
20 —h_-fi':;-.. DEVELOPED ISOTHERMAL LAMINAR

DISTRIBUTION, g— 22 (1-R2)
m

18— _\“\ o
16 e e LY ‘

 

 

 

 

 

at

 

 

 

 

 

 

SY)y

Fig. 9.3. Experimental Velocity Distribution in
the Cold Leg of a Glass Convectian losp with «
Reynolds Modulus (2r U /1) of 85.

UNCLASSIFIED
DWG. 19912

!

 

2z l l

 

 

 

 

DEVELOPED ISOTHERMAL LAMINAR

2.0 r———~—— U 2 —
\ DISTRIBUTION, /[~ =2(1-R%)
| 1

S L

\\
1.6 T e ‘-K-O' """"""
(4t vd

A <\

12 1 P

 

)=

 

 

 

 

 

 

 

 

 

 

 

 

 

06 foo —
04 |
—
02 f—m -
o L |
o 0.2 o 0.6 0.8 10
R=F
o
Fig. 9.4, Experimental Velocity Distribution in

the Cold Leg of a Glass Convection Loop with a
Reynolds Modulus (2r U/ _/v) of 105,

90

FORCED-CONVECTION HEAT TRANSFER
WITH NaoF-KF.LiF

H. W. Hoffman J. Lones

Reactor Experimental Engineering Division

Additional heat tronsfer data have been obtained
for the NaF-KF-LiF eutectic(11.5-42.0-46.5 mole %)
flowing through a heated Inconel tube, The ex-
perimental results in terms of Colburn’s ; function
vs. Reynolds’ modulys are presented in Fig. 9.5,
The previously reported data, which were obtained
in a nickel tube, are also presented for comparison.
The data obtained in the Inconel tube fall approxi-
mately 50% below the curve obtained from the
equation which correlates heat transfer in ordinary
fluids, in sodium hydroxide,(®? and in NaF-KF-LiF
in nickel tubes. One would expect NaF-KF-LiF
to behave as the ordinary fluids do, as for as heat
transfer is concemed, ond yet there is a difference
between the data obtained in Inconel tubes and that
obtained in nickel tubes. In order fo resclve this
disparity, temperoture measurements and physical
property data were reviewed. Checks on the thermo-
couple calibrations showed no significant devia-
tions from the original calibrations. Checks on the
mixed mean fluid temperature measurement tech-
nique also indicated that no errors appeared to be
present, The heat capacity of the NaF-KF-LiF
mixture was never in question, because good heat
balances were obtained during the heat transfer
experiments,  The viscosity of the mixture had
been checked previously and would have had to be

 

(6)H. W. Hoffmen, Turbulent Forced Convection Heaf
Transfer in Circular Tubes Containing Molten Sodium

Hydroxide, ORNL-1370 (Oct. 3, 1952).

UNCLASSIFIED
DWG. 19913

   

 

 

0850 FTTT T T (7T ""“'J"T"L"“"'"'"“”T“*‘"”"“’

0_ 9 e o 4 n ramna o - e e ——— .........‘.A....<|'.....

0008 | ] H 1L L —-
Q007 - - + NoF-KF-LiF IN NICKEL TUBE |
0006 f———- ~4- a NaF-KF-LiF IN {NCONEL TUBE

— 0005 b—

2

 

7 MODULUS (St-Pr
Q Q
o Q
S o
I o
|
/
L
o

T

|

 

 

 

 

 

0001
1000 5000 10,000

REYNOLDS MODULUS

 

20,000

Fig. 9.5. The j-Modulus vs. Reynolds Modulus
for NaF-KF-LiF Systems,
an unreasonably low value to account for the dis-
crepancy. A several-fold change in thermal con-
ductivity would account for the difference in heat
transfer, but the checks on conductivity, which are
currently being made, indicate that the original
values are correct. :
The explanation of the difference in heat transfer
between the nickel and the Inconel system is ap-
parenily that on interface resistance, such as a
layer of corrosion products, exists between the
tube wall and the NaF-KF-LiF in the Inconel
system. For example, it would only take a l-mil
filmhaving athermal conductivity of 0.5 Btu/hr-ft-°F
to account for the reduced j modulus observed; if
the conductivity of such a film were lower, it could
be much thinner than 1 mil. Examination of the in-
side surface of one of the Inconel tubes used in
the experiments showed the presence of a thin,
green film, which has recently been identified as
K3CrF6, with some Li,CrF,. These corrosion
products are typical of those found when KF- and
LiF-bearing fluoride mixtures are contained in
Inconel. It is significant, however, that no such
film appears when the NaF-ZrF4-UF4 system (o
which the ARE fuel belongs) is contained in
Inconel, The thickness and, possibly, the thermal
conductivity of the green film will be examined.

HIGH-TEMPERATURE REACTOR
COOLANT STUDIES

M. F. Poppendiek J. I Lang

Reactor Experimental Engineering Division

The effectiveness of a reactor coolant is a func-
tion of coolant duct dimensions and spacing, the
amount of heat to be removed, the total coolant tem-
perature rise, and the physical properties of the
coolant. An analysis of the effectiveness ofreactor
coolants from a heat transfer standpoint has been
completed. Four simultaneous heat transfer equo-
tions were reduced fo a single equation which con-
tains several dimensionless moduli. This equation
was evaluated, with the aid of {BM machines, over a
wide range of the various parameters involved, and
the results have been plotted in chart form. The
pressure-drop and pumping-power equations can
readily be evaluated next, In particular, the pump-
ing power per heat extracted (a dimensionless
pumping-power modulus) can be obtained.

Recently the pumping-power moduli of several

reactor coolants were calculated for heat removel
from & hypothetical solid-fuel-element aircraft

PERIOD ENDING JUNE 10, 1953

reactor with the following characteristics: total
heat flux, 200 megawatts; length of the cylindrical
core, 2.75 ft; cylindrical diameter, 3 ft; coolant
volume, 30%; total coolant temperature rise, 200°F,
The coolants lithium, bismuth, sedium, and NaF-
KF-LiF were considered, and the pumping-power
modulus for each coolant was evaluated over
g range of coolant tube diameters, The results
indicated that therewas a minimum in each pumping-
power modulus vs, tube-diameter curve and that of
the four coclants lithium was the best and bismuth
was the worst, (NaF-KF-LiF and sodium were, in
general, about the same for the particular case
considered.)

It is felt that the charts described above can be
used quickly to make effectiveness optimizations
of coolant systems for solid-fuel-element reactors
from a heat transfer — momentum transfer standpoint,

TURBULENT CONVECTION IN ANNULI
J. O. Bradfute J. 1. Lang

Reactor Experimental Engineering Division

The design of the previously described annulus
flow system for measuring velocity profiles has
been completed, and fabrication of the machined
part is under way. The structural steel bracket
which will rigidly support the annulus has been
installed. A direct-drive blower has been obtained
and modified to a belt-driven device so that the
rotor speed can be increased until an adequate
head is developed. '

A small flow system has been constructed so
that the photographic techniques can be con-
veniently developed and refined while the equip-
ment is being fubricated. A few preliminary photo-
graphs have been made which revealed the need
for increasing the light intensity in the beam.
Several photographs of dust particles that were so
small as to be invisible in diffuse light have been
taken by using the light scattered from a Tyndall
beam. The resolution or grain size difficulties
anticipated have not materialized; the particles
appear in the photographs as well-defined streaks.
The grid, which is simply d 1/]6-in.-thick sheet of
lucite with scratches ruled on one side to form a
cartesian coordinate system, has been photographed
in light emitted from the scratches by illuminating
one edge.  This procedure exposes the negative
only on the image of the grid lines and leaves the
remainder of the film unexposed and, hence, trans-
parent after development.

91
ANP PROJECT QUARTERLY PROGRESS REPORT

CIRCULATING-FUEL HEAT TRANSFER
H. F. Poppendiek G. Winn
N. D. Greene

Reactor Experimental Engineering Division

The apparatus which simulates heat tronsfer in a
circulating-fuel pipe system by passing an electric
current through o flowing electrolyte{”} has been
successfully operated. Preliminary wall and mixed
mean tluid temperature measurements were obtained
over the Reynolds’ modulus range of about 5,000 to
10,000 and at a power level of about 0.1 kw/em3,
The limited wall-to-fluid temperature differences
measured were in the range predicted from the
previously developed theory. Some gassing was
observed at the stainless steel electrodes at high
current flows. Experiments have indicated that if
the electrode surfoces which are in contact with
the electrolyte are made of platinum, the gassing
can practically be eliminated.

A brief study of strong electrolytes(®! for use in
volume heat source experiments was made to find
an electrolyte having a low electrical resistivity
and, aiso, flat resistivity-temperature character-
istics. The study indicated that phosphoric acid
satisfies these requirements, [t was also found
that this electrolyte would be satisfoctory from
the corrosion standpoint.  When the platinum-
surfaced electrodes ond a centrifugal pump with a
somewhat higher head have been installed, the
volume heat source experiments are to be resumed,

BIFLUID HEAT TRANSFER EXPERIMENTS
D. F. Salmon, ANP Division

Heat transfer data from the previously de-
scribed!?) bifluid heat transfer system containing
NoK and the fluoride mixture NaF-ZrF -UF,
(50-46-4 mole %) have been analyzed. Fifteen runs
were made which yielded about 80 data points, The
only data points that were considered valid were
those obtained before a gradual reduction of the
inside diameter of the tube occurred; the reduction
was evidenced by increased pressure-drop meas-
urements, Subsequent examination showed that
the reduction of the tube diameter was the result

(7)H. F. Pappendiek and G, Winn, ANP Quar. Prog.
Rep. Mar. 10, 1953, ORNL-1515, p. 162,

(B)N. D. Greene, A Preliminary Study of the Electrical
Conductivity of Strong Electrolytes for Possible Appli-
cation in Volume Heat Source Experiments, ORNL
CF-53-5-149 (May 19, 1953).

(9)9. D, Whitman and D. F. Salmon, ANP Quar. Prog.
Rep. Mar. 10, 1953, ORNL-1515, p. 30,

92

of build-up of mass-transferred material in the heat
exchange element of this trimetallic system:
nickel heat exchange tube, type 316 stainless
steel pump, and Incone! piping.

The heat transfer occurred in a concentric tube
section; the center tube contained the fluoride and
the NaK was in the annulus. The center tube was
made of nickel, 0.269 in, ID with o 0.030-in. wall.
The L/D ratio of the tube was 40, while the 1./D
ratio for the annulus was 22, The fluoride Reynolds’
numbers ranged from 5,000 to 20,000, and the NaK
Reynolds' numbers ranged from 20,000 to 100,000,

Inlet and outlet temperatures of the two liquid
streams were measured with chromel-alumel probes
welded into the piping at the center line of the
streams. Inlet fluoride temperatures varied from
1200 to 1500°F. Fluoride radial temperature dif-
ferences varied from 50 to 230°F,

The over-all heat transfer coefficient, based on
the outside tube wall surface, was calculated from
the equation

q
AO
U, =

0
Dt;

where Dt is the log-mean temperature difference.
The fluoride heat transfer coefficient was obtained
from the over-all heat transfer coefficient in con-
junction with a Wilson Plot.{1?) The fluoride data
are compared to the Colburn’s(11) equation,

Nu = 0,023 Re0:8 p,!/3

in Fig. 9.6, the viscosity of the fluoride was

 

10y, H, McAdams, Heot Transmission, p. 272, 2d
ed., McGraw-Hill, New York, 1942.

Uibid., p. 168, Eq. 4D,

UNCLASSIFIED
DWG. 19915

bomerd

 

100

   

 

 

 

10,000
REYNOLDS MODULUS

Fig. 9.6. Comparison of Fluoride Heat Transfer
Data with Colburn’s Equation,
evaluated of the so-called ‘'film temperature,”’

1
T --2-..(1.?- zw) .
In Fig. 9.7, the fluoride data are compared with
the Hausen's equation,{1?) which
account the L/D effect:

takes into

PERIOD ENDING JUNE 10, 1953

The bifluid system is now operating with a con-
centric tube exchanger constructed of Inconel to
operate ot fluoride velocities of 8 fps and oxial
temperature differences of 350°F. The current
tests are to supply information on dynamic corrosion
and mass transfer. | '

 

2/3 0,14
Nu = 0.116 (Re2/3 - 125) prl/3 [1 +<—D-> } (i) .
i

In this equation, p is evaluated at the bulk tempera-
ture and p_ is evaluated at the wall temperature,
The fluoride center line temperature rather than
the bulk temperature was used in Fig. 9.6. Theo-
retical investigations have shown thot for the
Reynolds’ and Prandtl’s numbers involved the ratio
t - f
Bl v x oo .

t, -1
c u

The heat balances were in error by a maximum
of 28%, with an overage deviation of 11%. The
fluoride data fell within approximately 11% of
Colburn’s equation and about 27% below Housen's
equation,

The bifluid loop was sectioned at the conclusion
of the test ond examination of the nickel heat
exchange tube showed that a metallic fayer had
built up on the inside to a thickness of approxi-
mately 30 mils at the exit end. This large deposit
probably was built up after the represented heat
transfer data were obtained and during the 300 hr
of operation with a fluoride velocity of 8 fps.
Spectrographic and x-ray diffraction analyses both
identified this 30-mil layer as iron. The iron was
undoubtedly transferred from the type 316 stain-
less steel pump to the nickel heat exchange tube
by the fluoride. This mass transfer, which re-
sulted from the presence of more than one metal
in the same fluoride system, would mask such
corrosion as existed., [t is significant that this
mass transfer phenomenon has not been observed
in any monometallic system; the ARE fuel sysfem
is monometallic. Dynamic tests of the fluoride
corrosion of Inconel in all-lnconel systems are
reported in sec. 7, ‘‘Corrosion Research."’ '

Mo

 

Uz)E. Eckert, Introduction to the Tronsfer of Heat and
Mass, p. 115, McGraw-Hill, New York, 1950,

UNCLASSIFIED
DWG. 19314

100

 

 

 

 

 

o
Q

n
<

 

1 -
W { Py 73 (O

 

 

 

 

 

 

 

 

0 LD
1000 4000 10,000
REYNOLDS MODULUS

 

 

30,000

Fig. 9.7. Comparison of Flueride Heat Transfer
Data with Hausen’s Equation,

93
ANP PROJECT QUARTERLY PROGRESS REPORT

10, RADIATION DAMAGE

D. S. Billington, Solid State Division
A. J. Miller, ANP Division

Additional radiction damage studies of reactor
fuels and structural materials were carried out in
the LITR and the MTR. Efforts are being made to
determine and explain the apparent uneven uranium
distribution in the fuel irradiated in Inconel cap-
sules. Since the state of the fuel in the capsules
with regard to turbulence and temperature gradients
is quite different from that found under service
conditions, the design and construction of an in-
pile loop for circulating fluoride fuel are being
carried out as rapidly as possible, The cantilever
type of creep measurements made on Inconel in
a helium atmosphere indicated no serious change
in creep properties upon irradiation, and a speci-
men of Inconel irradiated in the MTR showed no
change in thermal conductivity to within the ac-
curacy of the measurements,

IRRADIATION OF FUSED MATERIALS

G. W. Keilholtz P. R. Kiein
J. G, Moargan M. T. Robinson
H. E. Robertson A. Richt
C. C. Webster W. R. Willis
M, J. Feldman

Solid State Division
K. J. Kelly, Pratt and Whitney Aircraft Division

Exaominations have been made of the Inconel
capsules containing small amounts of the ARE
type of fuel that were irradiated in the LITR and
in the MTR for various periods of up to several
hundreds of hours. In the LITR, 230 watts were
dissipated in each cubic centimeter of fuel, while
in the MTR, approximately 3400 watts were dissi-
pated by increasing the uranium fluoride content
of the fuel, In general, the samples irradiated for
long periods of time showed an intergranular type
of corrosion which does not occur in out-of-pile
control samples; the penetration occurred occa-
sionally to a depth of 3 mils.

Numerous chemical analyses indicated, as in
the case reported previously for two capsules
irradiated in the MTR, (") that the uranium content
of the fuel which could be readily melted from
irradiated capsules was lower than would be
expected after allowance for burnup. On the other
hand, several samples of fuel bored from various

(1)G. W. Keilholtz et al., ANP Quar. Prog. Rep. Mar.
10, 1953, ORNL.-1515, p. 166.

94

sections of the capsules were analyzed ond found
to be deficient in uranium, while in other such
samples, an excess of uranium was indicated.
Capsules irradiated in the LITR and in the MTR
were sectioned and examined visually at low
magnification, No evidence could be detected of
changes in the fuel or of segregation in this ex-
amination; thersfore odequate facilities for petro-
graphic examination are now being constructed in
one of the hot cells, An investigation of the
validity of the methods usad for the chemical
analyses of the irradiated materials is continuing.
it is possible that thermal gradients in the fue!
contained in the capsules are causing concen-
tration gradients in the fuel, ond an inpile pump
loop would be a much better method of simulating
service conditions than the static copsule.

A few, simple, preliminary experiments were
carried out to see whether it would be possible to
readily obtain information concerning phencmena
caused by temperature gradients or overheating of
the capsules. Qut-of-pile control samples were
heated to 1800 and 2280°F for 140 hours, Very
little intergranular corrosion occurred; subsurface
voids to a depth of about 1 mil were observed; and
the uranium concentration of the central portion of
Other control capsules
were heated in a manner that provided a tempera-
ture of 2240°F at the top and 1475°F at the bottom,
approximately 3 in. away. No significant change
occurred in the uranium concentration between the
hot end and the cold end.

the fuel was unaffected.

INPILE CIRCULATING LOOPS

0. Sisman R. M. Carroll
W, W. Parkinson A. S. Olson
W. E. Brundage C. Ellis

C. D. Baumann F. M. Blacksher
Solid State Division

Developmental work continued on a pump for an
inpile fused-fluoride-fuel loop to be operated in the
LITR and in the MTR., Design and fabrication of
other portions of the loop are in progress.

Beryllium oxide in contact with sodium at 1500°F
was exposed in the LITR to determine the effect
of radiation on its stability. Specimens from both
the high- and low-density regions of ARE moderator
blocks were sealed in capsules containing about
2 em? of sodium to give a surfoce-to-volume ratio
of 3to 1. Eight such copsules were enclosed in a
steel can that provided a helium atmosphere and
had heaters attached to maintain the temperature
at about 1500°F. The irradiation was carried out
for three weeks, thaot is, until an exposure of
greater than 10'8 fast neutrons/em? was obtained.
Inspection of the beryllium oxide and sodium is
under way. The preparation of samples for out-of-
pile control runs is almost completed.

CREEP UNDER IRRADIATION

J. C. Wilson J. C. Zukas
W. W. Davis
Solid State Division

The inpile cantilever creep apparatus used previ-
ously has been modified to permit fests in an inert
atmosphere. Figure 10,1 shows that previous re-
sults obtained in the Metallurgy Laborafory(z) are
confirmed by the bench tests; that is, creep of

), Potriarca and G. M. Slaughter, ANP Quar. Prog.
Rep. Sept. 10, 1952, ORNL.-1375, p. 135.

 

PERIOD ENDING JUNE 10, 1953

Inconel appears to proceed more ropidly in an inert
atmosphere than in air. The creep curve in the
only inpile inert-atmosphere test to date lies above
that resulting from a test in air but below the
corresponding inert-otmosphere bench test. Ap-
parently, neutron irradiction slightly reduces creep
in an inert atmosphere. The results are not yet
conclusive, inasmuch as the purity of the inert
atmosphere is not yet high enough to prevent
oxidation and some difficulties in temperoture
control have been encountered because of the
parasitic emf’s generated in the seals through
which the thermocouple wires enter the sealed
experimental can. The main effort at present is
directed toward eliminating these difficulties. In
addition, an inpile test (in air) is running at the
conditions of stress and temperature expected in
the ARE pressure shell.

Two constant-strength cantilever specimens of
0.065-in.-thick Inconel sheet were tested ofter
irradiation in the LITR at 100°F to an integraied

S8D-B-756

 

 

o_ 30 e A m e e Am e e mm Mk e e e f e S e e s e ki

ANNEALED 2hr AT 1700°F
PRIOR TO TESTS

o

 

 

CANTILEVER BEAM EXTENSION (%)

010

 

 

 

200

     
  

“HOLE MB-3 LITR, IN AIR, FAST FLUX, > 10'2

 

 

__ HOLE HB-3 LITR, IN HELIUM,
FAST FLUX, >10'%

"fl.\‘ -

-
-
-
"""
-
-

 

 

::..""-’—-’/
T

TP BENGH TEST IN AR e

HOLE 20 ORNL GRAPHITE REACTOR, IN AIR,

FAST FLUX, 4 x10'C

 

 

 

500 800

TIME (hr}

Fig. 10.1. In-Reactor Cantilever Creep Tests on Inconel at 1500°F and 3000 psi in Air and Helium,

95
ANP PROJECT QUARTERLY PROGRESS REPORT

fastneutron flux (greater than 1 Mev) of 2.5 x 1017,
The postirradiation test was carried out ot 1500°F
and at a stress of 3000 psi. In Fig. 10.2 it can be
seen that the irradiated specimens showed less
creep in the earlier stages than did the unirradi-
After several hundred hours, the
creep rates of both the irradiated specimens and

ated controls.

the one control sample were the same within the
margin of experimental error. This result indicates
that at least part of the effect of irradiation under
these conditions is to reduce the creep of Inconel,
because the radiation-induced hardening effect
apparently persists over tens of hours at 1500°F,
In similar tests (at 1500°F) on type 347 stainless
steel, the creep strength did not appear to be
influenced by prior irradiation. The difference in
the behavior of stainless steel and of Inconel was
expected from the results of hardness data reported
previously.3) Several additional samples are now
being irradiated.

A satisfactory extensometer design has been
evolved for the MTR tensile-creep apparatus, and

I, W. Davis, J. C. Wilson, and J. C. Zukas, ANP
Quar. Prog. Rep. Mar. 10, 1953, ORNL-1515, p. 167.

 

0.4 [ _[

 

| BENCH TEST

03— ————

0.2

 

CANTILEVER BEAM EXTENSION (%)

0 50

100

    
 
 

~= HOLE HB-3 LITR, FAST FLUX,=10

ANNEALED 2hr AT {TOO®F

—_

means for mounting and connecting the apparatus
to the specimen have been worked out. The be-
havior of the thermocouples under the unusual
furnace conditions is now understood sufficiently
well to assure reliable results. A heat transfer
and fluid-flow mockup of the experimental appa-
ratus is being made to test the ability of the unit
to dissipate the 15 kw of induced gamma heating.

RADIATION EFFECTS ON
THERMAL CONDUCTIVITY

A. Foner Cohen .. C. Templeton
Solid State Division

A small Inconel sample was irradiated in the
MTR at approximately 600°F for 400 hr in a flux
greater than 1013 fast neutrons/cm2.sec.  After
removal from the reactor, the thermal conductivity
of the sample was measured at 300°F and com-
pared with the thermal conductivities of unirradi-
ated specimens. The irradiation coused no change
in the thermal conductivity within the accuracy of
the measurements, which was 10%.

S5D-B-757
DWG. 194694

/__
Tt

12

 

 

 

 

150 200 250 - 300

TIME {hr}

Fig. 10.2. Post-lrradiation Creep of Inconel in Air at 1500°F and 3000 psi.

96
PERIOD ENDING JUNE 10, 1953

11, ANALYTICAL STUDIES OF REACTOR MATERIALS

C. D. Susano, Analytical Chemistry Division
J. M. Warde, Metallurgy Division

Studies were continued on the use of bromine
trifluoride as a reagent for the determination of
oxygen in reactor fuels., A procedure was stand-
ardized in which the oxygen liberated by the
fluorination of uranium trioxide is measured, A
calibration curve was established by plotting the
increase in terminal pressure as ¢ function of the
concentration of oxygen added in the form of
uranium frioxide, Efforts to fluorinate zirconium
oxide quantitatively were continued.

The reaction between zirconium tetra p-bromo-
mandelate and sodium methylate was shown to be
stoichiometric, and therefore p-bromomandelic acid
was established as a reagent for the volumetric
determination of zirconium,

The concentrations of uranium trifluoride and
metallic zirconium in NcF-ZrF4-UF4 {50-46-4 mole
%) were determined by the evolution of hydrogen
upon treatment with hydrochloric and hydrofluoric
acids, Preliminary studies were conducted on the
hydrolysis of uranium frifluoride as a function of
temperature and time, After 48 hr, approximately
22% of the vuranium originally present as uranium
trifluoride was in solution,

";Ditferential colorimetry’’ was applied to the
colorimetric determination of zirconium as zir-
conium alizarin sulfonate, The optimum concen-
fration of the reference solution of the alizarin
complex, which represents the maximum precision
theoretically possible, was found to be 1.0 mg of
zirconium per 10 ml of solution. The method will
be applied to the determination of zirconium in
reactor fuels when a sufficient order of precision
i s obtained,

““Tiron,” disodium-1,2-dihydroxybenzenedisulfo-

nate, was shown to be a suitable reagent for the
determination of uranium. The composition and
the stability of the complex were determined, in
addition to the sensitivity of the reagent for the
determination of uranium,

Petrographic examinations of over 700 samples
of fluoride mixtures were completed, Optical
data are reported for NaUF,, Na,UF ., NaZrF,,
Nuzsz6, Na,ZrF_, Li3CrF6, K3CrF6, N03CrF6,
FeF,, K2NaFeF6, NaMF,, LEZZrFé, and Li4ZrF8,

The activities of the service laboratory during
this quarter included the analysis of 872 samples
involving 9335 determinations,

CHEMICAL ANALYSES OF FLUORIDES AND
THEIR CONTAMINANTS

J. C, White, Analytical Chemistry Division

Determination of Oxygen in Metallic Oxides with
Bromine Trifluoride (J. E. Lee, Jr.,, Analytical
Chemistry Division). The use of bromine tri-
fluoride as a reagent for the determination of the
oxygen which probably exists in reactor fuels
as oxides of uranium ond zirconium is being
studied.{"  In order to conduct the fluorination
at the temperatures and pressures which have been
indicated by Emeleus(? as necessary for quanti-
tative reaction with zirconium oxide and to resist
the high corrosivity of the reagent, an apparatus
was fabricated from nickel ond stainless steel.
The fluorination step is accompanied by quanti-
tative liberation of oxygen from the metallic oxide,
and the oxygen is collected ond measured in the
glass portion of the apparatus, Attempts to apply
simple gas law relationships for calculating the
oxygen by measuring the increase in pressure of
the system were unsuccessful because of the ex-
treme temperature gradients encountered in the
system, An empirical procedure was established
whereby conditions of the fluorination were stand-
ardized, and the increase in pressure was plotted
as a function of oxygen added in the form of
uranium trioxide. The relationship over a pressure
range of 0 to 1100 p and 0 to 5 mg of oxygen is
shown in the calibration curve (Fig. 11.1). Pre-
vious experiments showed that a maximum of 85%
of the oxygen was released from zirconium oxide.
When uranium trioxide was mixed with zirconium
oxide and the mixture was fluorinated for 30 min
at 300°C, the results, in some cases, were high
by as much as 50%. Subsequent investigation
revealed that the stopcocks were not adequate for
continbous use under high vacuum, The faulty

 

(D), E. Lee, ANP Quar. Prog. Rep. Mar. 10, 1953,
ORNL-1515, p. 173.

(2)H. J. Emeleus and A. A, Woolf, J. Chem.. Soc.,
p. 164-168 (1950).

97
ANP PROJECT QUARTERLY PROGRESS REPORT

UNCLASSIFIED
OWG. 19946

T 1 ( N

s !

 

S

 

OXYGEN {mg)

 

 

N

 

 

 

 

 

  

 

Q 400 BOO 1200
TERMINAL PRESSURE (u)

Fig. 11.1. Calibration Curve for Determination
of Oxygen by Reaction with BrF .,

stopcacks have been replaced by stopcocks with
longer barrels which are better suited for high-
vacuum work, Further work to produce quantitative
liberation of oxygen from zirconium oxide and to
apply the reaction to reactor fuels is under way,

Volumetric Determination of Zirconium (T. R.
Phillips, Analytical Chemistry Division). The
advantages of halomandelic acids as reagents for
the determination of zirconium were listed in «
previous report.t®) The application of a volumetric
method in which p-chloromandelic acid is used
was also described previously. A method of syn-
thesis of p-bromomandelic acid was obtained
(private communication from R, B. Hahn to J. C,
White, April 3, 1953), and suitable quantities of
the acid were prepared fo investigate its applica-
bility to the volumetric technique. In generdl,
the bromo derivative is equally as good as the
chloro acid. The reaction

Zr(p-BrC H ,CHOHCOO),
+ 4Na(OCH,) —> 4Na(p-BrC H ,CHOHCOO)
+ Zr(OCH,),

proceeds stoichiometricolly, Complete precipi-
tation of zirconium takes place within 30 min when
a 40-fold excess of reagent is used, The standard
deviation of the method is approximately 1%. The
compilation of the data on zirconium in the ARE
fuel mixture is presently under way in preparation
for the writing of a final report on this project,
Determination of Uranium Trifluoride and Me-
tallic Zirconium (W. J. Ross, Analytical Chemistry
Division), The uranium trifluoride and metallic

 

(e, K. Talbott and J. M. McCown, ANP Quar. Prog.
Rep. Mar. 10, 1953, ORNL-1515, p. 172.

98

zirconium contents of the fluoride NaF-ZrF -UF
(50-46-4 mole %) to which 1 wt % of zirconium
hydride was added were determined by the two-step
oxidation method described in the previous re-
port,(4) Zirconium was calculated from the volume
of hydrogen liberated by treatment with 0.2 M
hydrofluoric acid, and the trivalent uranium was
calculated from the volume of hydrogen liberated
by treatment with 9.6 M hydrochloric acid.

In connection with this problem, preliminary
studies were conducted on the hydrolysis of ura-
nium trifluoride as a function of temperature and
time, Uranium trifluoride (124 mg) was placed in
Erlenmeyer flasks containing 100 ml of water and
stirred at 25 + 1°C for various periods of time,
An dliquot of the supemnatant liquid was with-
drawn, and the concentration of the uranium was
determined. The results are shown in Table 11,1,
The dota are inadequate ot present for any con-
clusions to be drawn, After 48 hr, however, ap-
proximately 22% of the uranium originally present
in the form of the trifluoride was in solution, The
oxidation state of the uranium in solution has not
been definitely established, Tests to determine
the extent of hydrolysis at 100°C are also planned,

TABLE 11.1. CONCENTRATION OF URANIUM IN
WATER AFTER CONTACT WITH URANIUM
TRIFLUORIDE AT 25°C

 

 

 

CONTACT TIME URANIUM
(hr) (mmol es)

1 0.016

3 0.046

5 0.056

24 0.061

48 0.096

 

 

Colorimetric Determination of Zirconium (D. L.
Manning, Analytical Chemistry Division). A method
for the colorimetric determination of microamounts
of zirconium as zirconium alizarin sulfonate was
reported by Green,!5) The method is of insuf-
ficient accuracy to permit application to the de-
termination of zirconium in reactor fuels, However,

 

(4)W. J. Ross, ANP Quar, Prog. Rep. Mar. 10, 1953,
ORNL-1515, p. 172.

Glp, g, Green, Anal. Chem. 20, 370 (1948).
‘*difterential colorimetry,”” in which the absorb-
ancy of reference solutions of known concen-
trations rather than the absorbancy of distilled
water is used, has been shown to yield an order
of precision comparable to that obtained by gravi-
metric techniques. By using Hiskey’s method,(®
the optimum concentration of the reference solution
of zirconium dlizarin sulfonate was found to be
1.0 mg of zirconium per 10 ml of solution; this
reference solution yields the maximum precision.,
The relative accuracy of the method, as calculated
from the formula derived by Bastian, ") was calcu-

{oted to be 14:

 

Relative accuracy = s¢ x 2.7 ,
0.434
where
§ = slope of the differential Beet’'s law
curve,
¢ = concentration of the reference stond-
ard,
0.434 = mathematical optimum absorbancy,
2.7 = instrument factor,

The differential colorimetry technigque has been
applied to the determination of zirconium in fuel
mixtures, and a standord deviation of opproxi-
mately 2% was found, This standard deviation is
arger than thot of either the gravimetric or volu-
metric methods, |t is believed that greater pre-
cision can be obtained by using extremely careful
techniques. The method is of great potential value
because of its extreme simplicity and ropidity,

“Tiron'' As a Reagent for the Determination of
Uranium (D. L. Manning, Analytical Chemistry Di-
vision), A new colorimetric reagent for uranium
has been found, Disodium-1,2 -dihydroxybenzene-
disulfonate (tiron) forms o reddish-brown complex
in basic solution with uranyl ion, The complex
is unimolecular, it forms immediately, and it is
stable for several hours. The sensitivity of the
reagent for uronium is equal to that of ammonium
thiocyanate and ascorbic acid, Tests on anionic
that sulfate, chloride, and
Phosphate in concen-

interference reveol
nitrate do not interfere.
trations greater than 20 times that of vranium does
interfere.  lron, titanium, and vanadium are the
major cationic interferences. [t is planned to test
the feasibility of applying differential colorimetry

 

(8)c, F. Hiskey, Anal. Chem. 21, 1440 (1949).

()R, Bastian, R. Weberling, and F. Palilla, Anal.

Chem. 22, 164 (1950).

PERIOCD ENDING JUNE 10, 1953

to the determination of uranium with this reagent
in reactor fuels in which the interference by iron
is not serious.

PETROGRAPHIC EXAMINATION OF FLUORIDES
G. D. White, Metallurgy Division

T. N, McVay, Consultant
Metallurgy Division

Petrographic examinations of about 700 samples
of fluoride mixtures were carried out. The optical
data collected for various new fluoride compounds
are given below,

NaUF .
Color: green
Interference figure: uniaxial negative
Birefringence: crystals show first-order gray and
yellow between crossed nicols
Refractive indices: ¢ = 1.510
E = 1,500
Na,UF
Color: green
Interference figure: uniaxial negative
Birefringence: interference colors are first-order
gray
Refractive indices: O
E = 1,490
NalrF
Color; colorless
Interference figure: uniaxial negative; some
crystals produce a biaxial
negative figure with a small
optic angle
Birefringence: interference colors are first-order
white to blue
Refractive indices: 0 = 1,508
E = 1.500
Forms solid solution with NaUF
Na,ZrF
Color: colorless
Interference figure: biaxial positive with 2V =75

deg
Birefringence: interference colors up to first-
order blue
Refractive indices: « = 1,419
y = 1.430

Na Z:F
Color: colorless
Interference figure: unioxial negative
Birefringence: interference colors to first-order
white

99
ANP PROJECT QUARTERLY PROGRESS REPORT

Refractive indices: 0 = 1.386
£ = 1,381
LiCrFy
Color: pale green
Interference figure: biaxial negative with 2V = 40
deg
Birefringence: interference colors through the
second order
Refractive indices: a = 1.444
y = 1,464
K3ch6
Color: pale green
Interference figure: isotropic (no figure)
Refractive index: 1,422
Na,CrF
Color: pale green
Interference figure: isotropic
Refractive index: 1,411
FefF
Cofor: Colorless
Interference figure: uniaxial positive
Birefringence: interference colors through second

order
Refractive indices: 0 = 1,524
E = 1,540

K NaFef

Color: colorless

Interference figure: isotropic

Refractive index: 1,414
NaHF

Color: colorless

Interference figure: uniaxial positive
Birefringence: interference colors through the

third order
Refractive indices: 0 = 1.261
E = 1,328

LJZZrF6
Color: colorless
Interference figure: biaxial positive with 2V =20

deg
Birefringence: interference colors through the
first order
Refractive indices: o« = 1.462
y = 1,482

Li4ZrF8
Color: colarless
Interference figure: biaxial negative with 2v = 30
deg
Birefringence: interference colors through the
first order

100

Refractive indices: a = 1.445
y 1,465

SUMMARY OF SERVICE CHEMICAL ANALYSES

J. C, White A. F. Roemer, Jr.
Analytical Chemistry Division

The work of the Analytical Chemistry Laboratory
during the quarter has continued to consist chiefly
of the analysis of fluoride fuel mixtures, alkali
metal fluorides, ond NaK.

Zirconium is being determined gravimetrically by
precipitation with mandelic acid and ignitien to
the oxide. The ignited oxide has been much purer
with respect to contamination by iron and silica
than the oxide obtained from ignition of the phenyl-
arsonic acid salt. Precision ond accuracy are
somewhat improved also, Uranium in these mix-
tures is being determined by reduction with the
Jones reductor and oxidation with standard ceric
sulfate solution with the use of ferroin as the
indicator.  This method is used because of its
simplicity and rapidity.

A number of Incone!l specimens, which had been
used as containers in corrosion tests of fluoride
fuels, was analyzed for concentration of fluorides.
The pyrohydrolysis method was used with satis-
factory results,

During the quarter the laboratory reported 872
samples involving a total of 9335 determinations.
These analyses were mode for various groups in
the ANP project, as indicated in Table 11,2,

TABLE 11.2. SUMMARY OF SERVICE ANALYSES

 

 

NUMBER OF NUMBER OF
REQUESTOR SAMPLES |DETERMINATIONS
REPORTED REPORTED

 

Reactor Chemistry
Group 267 2274

Corrosion Group 309 3762

Experimental Engi-
neering Group 284 3240

Heat Transfer and
Physical Prop-
erties Dept. 12 59

Total 872 9335

 

 

 
PERIOD ENDING JUNE 10, 1953

12. FLUORIDE FUEL REPROCESSING

D. E, Ferguson

G. . Cathers

O. K. Tallent
Chemical Technology Division

Development of a method for chemical processing
of ARE fuel was initiated to ensure recovery and
U235 prior to its allocation
and use. The processing method now proposed
was adapted from that in use by the Y-12 Chemical
Production Division for the processing of uranium
fluorides., It was desirable that the method be
adaptable to existing facilities at ORNL without

much additional equipment and cost, since the fuel

decontamination of the

moy be further modified in future developments of
the ANP program.

It was considered in this study that the fuel to
be processed was NoF-ZrF -UF, (50-46-4 mole
%), which has a melting point of 504°C and o
density of 3.0 g/cm?. (Although the ARE fuel is
now the 53-43-4 mole % mixture, the considerable
research on the 50-46-4 mole % mixture is appli-
cable, in general.) The amount of U233 to be
processed is 70 kg per batch, that is, 16 to 18 3
of fused salt containing equal parts by weight of
NaF-ZrF -UF , and flush material consisting of
ZrF , and NaF (50-50 mole %).

DISSOLUTION OF ARE FUEL

The initial work on the preparation of an aqueous
solution of ARE fuel was carried out on material
containing 4 mole % UF ,, that is, undiluted ARE
fuel. The fused salt was used in the form of very
coarse lumps to simulate probable conditions in
ARE fuel processing.

Exploratory solution experiments showed ‘that
the fuel mixture could be dissolved in an alumiaum
nitrate~nitric acid solution at reflux temperature.
In effect, this leads to oxidation of guadrivalent
uranium to uranyl ion, complexing of the fluoride
ion with aluminum, and conversion of insoluble
NaF and ZrF, to soluble compounds. Although in
this preliminary work, solutions containing some
insoluble precipitate were obtained, the method
promised to be of value with further development.

Additional tests with different proportions of
aluminum nitrate nonahydrate (ANN), concentroted
nitric acid, and water were run to see whether
complete solution was possible and to observe
dissolution rates. The results showed that fast

and complete solution was possible with various
material proportions.  The optimum proportions
seemed to be an ANN-to-fuel ratio of 4 to .5 with
20 to 30 ml of acid and 40 to 50 ml of water for a
5-g batch of fuel. Less acid and less ANN led to
a decrease in rote of dissolution and some residual
precipitate. Best results were also obtained when
an optimum amount of water was present to dilute
the ANN-nitric acid solution.

The actual composition of the fuel used in the
above tests was 4.25 47.9, and 47.9 mole %, re-
spectively, of UF ,, NaF, ond ZrF,. An ANN-to-
fuel ratio of 5 therefore corresponded to a tlueride-
to-aluminum mole ratio of 1.7, while the lower limit
ratio of 3 corresponded to a mole ratio of 2.8. The
value of 1.7 appears preferable, because it ensures
more complete complexing of fluoride ion and
thereby less corrosion to processing equipment.
It also aveids approaching too closely to the mole
ratio of 3 represented by the insoluble compound
AIF,.  The use of more aluminum would lead to
faster dissolution and less corrosion, but it would
also lead to a lower uranium content in the feed
for solvent extraction processing.

On the basis of 1.7 as the fluoride-to-aluminum
mele ratio, the following weight proportion was
selected for feed preparation in solvent extraction
work:

Weight
Proportion
ARE fuel (210 4 mole % UF4) 1
ANN 5
Ceoncentrated nitric acid 7
Water 11

This weight proportion leads to uranium concen-
trations of 2.2 and 4.3 mg/mi, respectively, for
fuel plus flush material (2 mole % UFA) and for
fuel alone (4 mole % UF,). Other concentrations
used are, approximately, 4 M HNO, 0.2 M Zr,
0.2 MNa, 1.2 M F, and 0.67 M Al

Loter work showed that if the femperature was
raised too rapidly in heating the inhomogeneous
mixture of all sofution components, the ARE fuel
began to dissolve in the lower layer that was rich

101
ANP PROJECT QUARTERLY PROGRESS REFPORT

in aluminum nitrate before thorough thermal mixing
had occurred. This led to formation of a yellowish
precipitate which was not redissolvable. [t is
preferable, therefore, that a homogeneous solution
of ANN, nitric acid, and water be obtained before

addition of the ARE fuel.

SOLVENT EXTRACTION

Feed solution made from 2 mole % UF& material
(fuel plus flush) was used in batch countercurrent
runs to test the feasibility of a TBP solvent
exiraction process. 1he low uranium concentration
of 2.2 mg/ml made operation of such a process at
the usual 60 to 70% uranium saturation of the TBP
impossible. Too low a concentration of TBP in
Amsco to obtain an opprecioble saturation would
result in low exiraction factors and o large uranium
loss in the hot wasie, A compromise was there-
fore tried in three batch countercurrent runs with
3, 5, and 7% TBP in Amsco 123-15 with cold feed
salution. A nonocid scrub of 0.67 M ANN was
vsad to maintain high salting strength in the scrub
secticn and to secure an acid-free product suitable
for concentration and isolation by ion exchange.

The distribution coefficients in both the extrac-
tion and the scrub sections showed that 5 and 7%
TBP was very satisfactory and that the uranium
losses in the waste were less than 0.01%, The 3%
TBP run, however, gave a uranium distribution co-
efficient of less than 2 in the scrub section, and
the vranium loss was 0.2%. Zirconium carryover
with the uranium varied over the range of 0.02 to
0.06 mg/ml for the three cases; thus, zirconium
carryover does not appear to be dependent on TBP
concentration. Fluoride and acid decontaminations
were likewise satisfactory in the 5% TBP run.

Two runs with feed spiked with trace fission
products have shown that adequate first cycle
radicactive decontomination is obtainable with

both 5 and 7% TBP in Amsco. Gross beta decon-
tamination was better than 104 in both cases; thus,
the TBP concentration is not critical. The wranium
saturation of the TBP, 10 and 7%, respectively,
was far below that for which TBP concentration
is critical,

A rough schematic flowsheet can be drawn on
the assumption thot decontamination will be suf-
ficient in one solvent extraction cycle for coupling
A second
cycle will probably be necessory only for material
of very high burnup.

to an ion exchange column (Fig. 12.1).

CORROSION IN DISSOLUTION OF ARE FUEL

Corrosion tests have been carried out on types
309 and 347 stainless steel with a solution of ARE
fuel (2.2 mg of uranium per m!) operated with reflux
at aboutr 110°C.
desirable, since it was expected that a 4 M nitric
acid—1.2 M fluoride ion solution would be very
corrosive, despite the high concentration of
aluminum, and since most equipment available for
scaling up the process would be made of either
type 309 or type 347 stainless steel.

The tests were made by suspending square plates
of the materials at the liquid-vapor interface, in
the liquid, and in the vopor of the test solution.
Glass supparis were used to avoid metallic junc-
tion potentials. The total time of the tests was
33.5 hr, extended over a five-day period, with
weighings being made at the end of each day’s run.
The data are summarized in Table 12,1. The re-
sults clearly indicate a very large corrosion rate,
Type 309 stainless steel was somewhat better
than type 347, presumably because of its higher
The highest cor-
rosion rates were observed duringthe initial period,
but they decrease to a third or less during later

Information on corrosion was

nickel and chromium contents.

periods. The lower rates will probably prevail in

TABLFE 12.1, CORROSIOM TESTS ON TYPES 309 AND 347 STAINLESS STEEL IN
ARE FUEL DISSOLVER AT 110°C

 

 

CORROSION RATE (mpy)

 

 

 

 

 

 

Type 309 Stainless Steel Type 347 Stainless Stes!
At Interface In Liquid in Vapar At Interface in LLiquid In Vapor
First 3.5-hr period 153 131 117 238 166 129
Last 7.5-hr period 31 24 19 57 62 38
Over-all 33.5-hr period 53 44 35 g1 67 59

 

 

 

 

 

102

 
PERIOD ENDING JUNE 10, 1953

DWG {97084

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

SCRUB
9.67 M ANN WATER
(AsNBng) { VO*L‘UME 1.25 VO*LUMES
ko) /// - \\w__._._. SOLVENT wasTE
WATER
(128 kg) / \
[ 1 \ //% \\\\ SECOND SOLVENT EXTRACTION CYCLE
FEED PREPARATION - voLumes ///gé \\\ié\ OR ION EXCHANGE COLUMN
arerue==l 2z m0aag i N\ T
6 kg) OF U PER LITER ?’U¢ &8\%
/S/ \xi“z"’\ PRODUCT
ézé \\@% B.8 TO17.2 g
///i/ \\\E\ OF U PER LITER
% NEX
EXTRACTANT / §
o Top 5 VOLUMES /A hk\\ -
IN AMSCO *
HOT WASTE

Fig. 12.1. Flow Sheet for ARE Fuel Processing. Basis: 1kg of U235,

equipment which has already suffered from nitric
acid corrasion. The corrosion was highest at the
liquid-vapor interface and lowest in the vapor.

PLANT PROCESSING

It appears, on the basis of present knowledge,
that batches of ARE fuel plus flush material con-
taining 70 kg of U?35 can be decontaminated and
recovered by using two solvent extraction cycles
coupled with an ion exchange column, The use of
Metal Recovery Plant equipment for this process

is possible without major edditions provided the

following precautions are observed to avoid criti-
cality.

1. Proceeding with processing of ARE material
on a crash basis will require batch operation and
a high order of supervision to ensure safety with
respect to criticality. Frequent sampling will be
necessary and carewill have to be taken to prevent
any heel buildup in the dissolution step. The
process should be scheduled on the basis of 1-kg
batches of U235,

2. One-kilogram batches of U235, that is, about
1 gal of ARE fuel or 2 gal of fuel plus flush, con

be transported in cylindrical aluminum cans of
noncritical geometry with adequate shielding.

3. Dissolution of the material can be carried out
in a 500-gal tank in the Metal Recovery Plant.
The charge would be 60 gal for ARE fuel or 120
gal for fuel diluted with an equal weight of flush
material. Criticality is to be avoided partly by
slab geometry,

4. Two batches of feed may be stored in the
2500-gal tank of the Metal Recovery Plant that has
a concave bottom. Noncriticality is favored by
the ring geometry.

5. The 1A pulse column available for this work
unfortunately extends into a sump well in which
neutron reflection must be considered. A safe
concentration limit, however, would be 12.5 mg of
U235 per ml, or a concentration 3- to 6-fold above
that of the feed being considered.

Several coincident errors, such as the use of
water in place of 0.67 M ANN solution for the
scrub and reflux and buildup of uranium at the
feed plate with subsequent leakage from the column
into the sump well, would be necessary to approach
crificality.

103
 
INTRODUCTION AND SUMMARY

E. P. Blizard

J. L. Meem, Associate

Physics Division

The measurements of neutron and gamma spectra
in the divided-shield mockup in the Bulk Shielding
Facility constitute an important advance in shield-
ing research (sec. 13). These spectra give the
energy and angular distribution of neutrons from a
fission source after attenuation in water and are
directly comparable to some calculations being
carried out under subcontract by Nuclear Develop-
ment Associates. [t is by the extension of this
approach that a real understanding of shield design
will be achieved.

The Lid Tank Facility is being used for the engi-
neering type of studies of unit shields for which it
is especidlly applicable (sec. 14). Many combina-
tions of materials and of thicknesses of materials
are being tried in an effort fo obtain the lightest
possible unit shield for the reflector-moderated
liquid-fuel reactor. The majoruncerfainty remaining
is the possibility that the delayed neutrons and
gamma rays from the circuldting fuel in the heat
exchanger cannot be measured in the Lid Tank
Facility, Efforts have been initiated, however, to

calculate these components in order to evaluate
their possible importance. If their importance
appears fo be overriding, which might well be the
case, some special experiments will have to be
devised., With the use of this facility, on effective
removal cross section of 0.99 b was obtained for
oxygen. ' :

Construction of the Tower Shielding Facility,
which was delayed by strike, has now been resumed
and should be completed by the end of 1953 (sec,
15). The report to the Reactor Sofeguard Committee
revealed no untoward hazards,

A Shielding Session is scheduled for the month
of June in Oak Ridge. The pumose of the session
will be to stondardize methods of shield design,
and members of a number of interested organiza-
tions will porticipate. In order to ensure that the
problems studied will be generic to the basic
question of shield weight, actual shield designs
will be studied for the several cycles which are of
most interest. A report will be issued, probably
late in July.

107
ANP PROJECT QUARTERLY PROGRESS REPORT

13. BULK SHIELDING FACILITY

J. L. Meem

R. G. Cochran
M. P. Haydon

K. M. Henry

H. E. Hungerford

E. B. Johnson

J. K. Leslie

T. A. Love

F. C. Mgienschein
G. M. McCammon

Physics Division

During this quarter, the major effort at the Bulk
Shielding Facility has been devoted to measure-
ments on the top-plug mockup for the SIR shield.
This work has been very useful in that it offorded
an opportunity to study the perturbations to a
shield introduced by engineering features. A
description of these experiments will be given in
the next Physics Division semionnual progress
report. Some results of the neutron spectral meas-
urements, as well as results of air-scattering cal-
culations on gomma rays for the divided aircraft
shield, are presented here; the final results on the
energy-per-fission experiment are also given,

NEUTRON SPECTRA FOR THE DIVIDED SHIELD

The fast-neutron spectrometer(” at the Bulk
Shielding Facility has been used to measure the
reactor spectra as attenuated by the beryllium
oxide reflector and various amounts of water,{?)

The experimental setup used is shown in Fig.
13.1.  An odluminum collimator tube 6 ft long
(0D, 1 in,; 1D, 7/8 in.) was used for the measure-
ments close to the reactor (0 and 12.5 em). The
0-cm distance was the position in which the
collimator was pushed through a hole in o beryl-
lium oxide reflector can so that the end of the
collimator tube touched a fuel element. For the
27,5~ ond the 37.5-¢m distances, the collimator
length was reduced to 4 ft to increase the neutron
intensity, Previous experiments at this facility
have indicated that as long as the ratio of length
to diameter is large, the collimator does not distort
the neutron spectrum,

The principal purpose of the lead spectrometer
housing (shown in Fig. 13.1} was to ottenuate the

 

(”R. G. Cochran and K. M, Henry, A Proton Recoil
Tgpe) FastNeutron Spectrometer, ORNL-1479 (Apr, 2,
1953).

(Q)R. G. Cochran and K, M. Henry, Neutron Spectra of
the BSF Reactor, ORNL CF«53.5-105 (in press).

108

infense gamma rays encountered in making measure-
ments close to the reactor. In addition, it served
as a waterprocf container for the spectrometer
components. lo ensure water exclusion, the box
was pressurized with 2 to 3 psi of dry air.

A logarithmic plot of the data obtgined out to
37.5 cm from the reactor is given in Fig. 13.2
The data indicate that the spectrum hardens ap-
preciably as the neutrons penetrate largerdistances
of water,

In addition to the proton-recoil spectrometer data,
threshold detector measurements(®! have been made
in a fuel element of the BSF reactor. Figures 13.3
and 13.4 show the spectrum obtained by this
method.

ENERGY PER FISSION AND POWER OF
THE BULK SHIELDING REACTOR

The energy per fission in the BSR has been
measured as 193 Mev.*) In addition, the graphite
sigma pile, which is the stondard for thermal-
neutron flux measurements ot ORNL, hos been
recalibrated. ) Since all shielding dato at the
BSF have been reported as so much radiation per
watt of reactor power, a correction should be
applied to all previous data. The fast-neutron and
gamma data should be multiplied by o factor of
0.8306 and the thermal-neutron data should be
multiplied by a factor of 0.9250. A list of reports

to which these corrections apply has been pub-
lished.(®)

 

(B)J. B. Trice, Neufron Threshold Measurements in the
BSR, ORNL CF-53-5-13%9 (in press).

(4)J. .. Meem, L. B. Holland, ond G. M, McCammon,
Determination of the Power of the Bulk Shielding Re-
actor. Part {{l. Measurement of the Energy Released per
Fission, DRNILL-1537 {to be issued).

(S)E. D, Klema, R, H. Ritchie, and G. M, McCammon,
Recalibration of the X-10 Standord Graphite Pile, ORNL-
1398 (Oct, 17, 1952).

(6)J. l.. Meem, E. B, Johnson, and H. E, Hungerford,
Energy per Fission and Power of the BSR, ORNL CF-
53.5-21 (May 12, 1953).
=7

 

 

 

 

 

 

 

 

HOLE IN
REFLECTOR CAN

Fig. 13.1,

merged in water,

GAMMA-RAY AIR-SCATTERING CALCULATIONS

Measurements have been reported”) on the spec-
tra of gamma radiation emerging from a mockup of
the reactor portion of the aircraft divided shield.(®
Calculations have been made for the gamma radie-
tion arriving at the outside of the aircraft crew
shield, and further calculations are plonned on the
penetration of the gamma radiation through the

 

(7)F. C. Mdienschein, Gamma-Ray Spectral Measure-
ments with the Divided Shield Mockup, ORNL CF-52-3-1
(Mar. 3, 1952); Part |l, ORNL CF-52-7-71 (July 8, 1952);
Part 111, ORNL CF-52-8-38 {Aug. 8, 1952).

(S)Reporf of the ANP Shielding Board for the Aircraft
Nuclear Propulsion Program, ANP-53 (Oct. 16, 1950).

\ FUEL ELEMENTS
. //
P '/hBERYL_LlUM OXIDE

REFLECTOR CANS

-COLLIMATOR

Experimental Arrangement for Hewtron Spectral Measurements,

PERIOD ENDING JUNE 10, 1953

OWG. 19887

      
 

 

- LEAD SPECTROMETER
HOUSING

 

Apparatus operates sub-

crew shield, The data collected recently by the
Bureau of Standards(®) will be used, in part, for
these calculations,

The gamma-ray spectral measurements were
recorded?) as a function of energy and as a
function of the angles, as shown in Fig, 13,5,
Data obtained through the lead shadow shields
shown in Fig. 13.5 and through various thicknesses
of water were available, The data included the
dependence of I" (the emerging gamma-ray flux in

 

(?)F.. S. Kirn, R, J. Kennedy, and H. 0. Wyckoff,
Oblique Atftenuation of Gamma Rays from Cobalt-60 ond
Cesium 137 and Polyethylene, Concrete and [ead, NBS-
2125 (Dec, 23, 1952).

109
ANP PROJECT QUARTERLY PROGRESS REPORT

DWG. 193692A

CCRRECTED TO I-WATT
REACTOR POWER

O COLLIMATOR AGAINST
FUEL ELEMENT

AScm H20+?.50m BeO
¢ 20cm Hy0+7.5¢m BeO

V 30cm H;0+7.5¢cm BeO

¢ (neutrons /cm? /sec/Nev/Q /watt }

 

6o 1 2 3 4 5 & 7 8 9 10
NEUTRON ENERGY (Mev)

Fig. 13.2. BSF Reactor Neutron Spectra,

photons/cm?/sec/Mev/watt/steradian) on the ele-
vation angle 9 and also on the azimuth angle ¢ by
virtue of the existing symmetry along the aircraft
axis,

Gamma-ray dose measurements indicated that
data token behind the lead shadow shield would

 

%(Eifi) mf;faffefqé

be applicable for « less than 40 deg, whereas the
open-water data would be applicable for a greater
than 70 deg, where a is the polar angle indicating
the point of departure of the photons from the re-
actor shield. The intermediate region at the edge

110

5SD-A-709
D'wG. 184004

T

 

 

 

- ]

 

e nfemiem
- S !7,7

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

P
- - - " A|27 A!ZB”"
BULK SHIELDING FACILITY A (7.

REACTOR POWER: 100 kw

12

V0" e e N

 

 

 

 

 

 

! ] o L S
| R

| - —t R g e

 

 

 

 

 

}
T <
!

 

 

 

 

 

 

 

$(£), DIFFERENTIAL FLUX (neutrons/cm?/sec/Mev}

 

 

 

 

 

 

 

 

 

 

 

 

 

11 i |

 

10 | !| R —
0.0001 0.001 0.01 0.05
ENERGY {Mev)}

Fig. 13,3, Neutron Energy Spectrum for the

Epithermal Energy Region (0.5 kev < E < 10 kev)
by Threshold Detector Measurements,

of the shadow shield had to be investigated in
further detail. Measurements were made at a = 50
deg for all values of 6 and at a = 60 deg as a
function of both @ and ¢, The measurements then
afforded an approximate knowledge of 1'(E,q,0,0)
over the entire reactor shield, where E is the
energy (Mev) of the emerging gamma ray. The
gamma radiation arriving at the crew shield was
then obtained by integrating over the energy and
the surface of the reactor shield and by applying
the differential scattering cross section and inte-
grating over all space. Single scattering and no
air attenuation were assumed in the calculations,
which should be adequate with the reactor-crew
separation distances being considered.

1(E,a,0,¢) KR™2 dE dA 4V

WI'I ere

¢
~— = gommaeray flux arriving at the crew shield
P (photons/em?/sec/watt),

E’ = degraded gamma-ray energy {(Mev),
receiver angle,
= energy of the emerging gamma ray (Mev),

T
|

a = polar angle at the source,
r = distance from the point of emergence from
the reactor shield to the point of scatter
(em),
¢ = elevation angle,
¢ = azimuth angle,
K = Klein-Nishina differential scattering cross
section,“o)
R = distance from the point of scatter to the
center of the crew shield (em),
dA = surface element at the reactor shield,
dV = volume element of scattering space.

These calculations were simplified considerably
because, for the most part, the gomma radiation
emerged radially in the shield regions away from
the edge of the shadow disks., Calculations for a
typical element of volume were carried out by
using the measured dependence on @ ond ¢ and by
assuming that all the radiation was emitted radial ly,
Since the results agreed fo within 30%, the radial
assumption was used for all regions away from the
shadow-disk edges. This component was called
the radial beam, while the contribution from the
edge of the shaodow shield was called the skew
beam.

The direct (unscattered) beam calculation was
given simply by

¢ , 2
— (EB) = IT'R™ dE du .
E EF "a

The actual integrations were performed numeri-
cally, in steps, as shown in Table 13.1. The

 

(]D)C. M. DPavisson and R. P. Evans, Revs. Mod.
Phys., 24, 79 (1952).

TABLE 13,1,

PERIOD ENDING JUNE 10, 1953

computations were made on a UNIVAC after they
had been coded by the Mathematics Panel.
Differential Results, The specirc of the scattered
gamma-ray flux arriving at the crew compartment
are shown in Fig. 13.6 for the front region of the

SED-A=-T10
DWG. 1B40IA
7x1c"

n

7

7
Aia {r, ,c:v)Mg2

A

FLUX {neutrons /om?- sec)

BULK SHIELDING FACILITY
5 REACTOR POWER: 100 kw

 

0 2 4 6 8 10 12
THRESHOLD ENERGY (Mev)

Fig. 13.4. Neutron Flux Above Threshold Energy
as a Function of Threshold Energy. :

STEPS CHOSEN FOR NUMERICAL INTEGRATION

 

 

 

 

CALCULATION
VARIABLE
Radial Beam Skew Beam Direct Beam
7 6 steps of 1.2 to 2.5 Mev 4 steps of 1.2 to 2.5 Mev 6 steps of 1.2 to 2.5 Mev
a 18 steps of 10 deg 2 steps of 10 deg 9 steps of 10 deg
fi 18 steps of 10 deg -
r 40 steps of 5 to 500 +

 

 

 

 

111
ANP PROJECT QUARTERLY PROGRESS REPORT

N 90°
)
90° .
‘ \\“ POINT OF SCATTERING
y
LEAD

DWG. 19374

SHADOW SHIELDS pa

REACTOR—~ e

-

'REACTOR SHIELD

  
  

 

 

 

 

 

 

 

CREW GHIELD

 

MEASURE: T'(E,a, 8, ¢) CALCULATE: T"(E’, B)

Fig. 13.5. Definition of Angles for Gamma-Ray Air-Scottering Calculations,

 

 

 

 

 

 

 

 

 

 

 

 

 

DWG. 19868
0.012 r ‘
—_ S AREA UNDER CURVE = &/p
2 soo Lo SOURCE ANGLE (@)=0 TO 65 deg )
> (SHADOW-SHIEL.D REGION)
o
=
~
D
>~ 0008 | RECEIVER ANGLE, B (deg) CODE
£ 0 TO 180 —A— T
= 0 TO 30 B
2 30 TO 80 — G
o 60 TO 90 meme Do
o 0008 — 90 TO 120 e =
5 120 TO 150 e
& 150TO 180 e Grenne-
o
E 0.004 |- AT ]
E s e ——
o B
m
> - ~D--1 A
2 e -
g 0002 f— |c oA -
Q ~TB
~
o _B__ - "“"%”"‘“
------- ¢ F--p--- ] . | L
0 ¢ oG ] A I
0 1 2 3 4

E/, POST-SCATTERING ENERGY (Mev)

Fig. 13.6. S3pectra at the Crew Compartment of the Scottered Gamma Rays from the Front of the Re-
actor Shield,

112
 

 

 

 

 

 

 

DWG. 19869
48 | ]
g SOURCE ANGLE (@) =65 TO 125 deg
£ 40— r—-a (SIDES OF REACTOR SHIELD)
=
o
5
o
S RECEIVER ANGLE , B{deg) CODE
< O TO 180 —A—
£ 0 70 30 ——B-—
3 0 10 80 -
o 070 90 wmnDee
£ 24— 90 TO 120 .
2
L
8
&
E 15— AREA UNDER CURVE =Dsp
[ey)
o = C——]
= - =B
a - D---
o 8
.
<)
b ene]
0 Ay I I
0. 0.5 1.0 1.5 20

£7, POST-SCATTERING ENERGY (Mev)

Fig. 13.7. Spectra at the Crew Compariment of

the Scottered Gamma Rays from the Side of the
Reactor Shield,

shield.

from this region.

All the high-energy flux arrives

The side of the reactor shield
contributes much more low-energy radiation as
shown in Fig. 13.7 and, finally, Fig. 13.8 shows
the small contribution of low-energy radiation from
the rear of the shield.

A graphical representation of the contribution
from various volume elements in space is shown in
Figs. 13,9 and 13.10. Figure 13,9 gives the flux
contribution in photons/cm?/sec/watt, while Fig.
13.10 is shaded to demonstrate the same informa-
Thus it may be
seen that regions far from the aircraft contribute
a substantial portion of the gamma radiation arriving
ot the outside of the crew compariment.

reactor

tion for larger volume elements.

However,
since this radiation has to be scattered through a
large angle, it is of low energy and it loses some
of its importance after penetrating the crew shield.
All gamma rays arriving at the crew shield with an

PERIOD ENDING JUNE 10, 1953

 

 

 

 

 

. DWG. 19B70
32
3 l & |
@
< SOURCE ANGLE (@)=125 TO 180 deg
B (REAR OF REACTOR SHIELD)
=
5 el -
3 RECEIVER ANGLE, S{deg) CODE
e 0 TO 180 —p
< 0 TO 30 —— -
Q
2 30 TO 60 -
8 18—
o
o
£
S
S AREA UNDER CURVE = $/p
£
E
Q
&
o
2, L L
a 0 0.5 1.0 1.5 2.0
% £/, POST-SCATTERING ENERGY (Mev)

Fig. 13.B. Spectra at the Crew Compartment of
the Scattered Gamma Rays from the Rear of the
Reactor Shield,

energy greater than 0.5 Mev, for example, must have
a+ 3 < 90 degrees. This condition implies
scattering from a point closer thon B.6 meters from
the aircraft axis.

The direct-beam radiation orriving at the crew
shield is shown in Fig. 13.11. Since this contri-
bution is ploited as a function of the angle g,
rather than as that of the solid angle, the dose will
tend to increase with a. Counteracting this, how-
ever, is the fact that @ increases with increasing
a and that the radial nature of the emerging radia-
tion causes the contribution to rapidly decrease.
The marked increase at the edge of the shadow
shield indicates that the shadow shield should be
made larger and that it should be tapered in thick.
ness toward the edges.

Integral Results, The integral results are shown
in Table 13.2, for which the gamma flux was
converted into dose units.!17?  The results are
compared with the calculations of ANP-53,¢8) and
they agree to within a factor of about 3. This is

 

(”)E. P. Blizard, Introduction to Shield Design,
(]:);?Ng_ CF-51-10-70, Part | Revised, p. 41 {Jan. 30,
52}.

113
DWG, 194834

 

Noooose

     

     

 

       

"
<
o
o
o
o
e o
w
g -—
o T
S
S :
: S60000" el
o 00 QMOOOO,O —
~
© 3 0.000048
g g ———
o o el
o o \\\u\\\n\n 0 OOOO.mU
S ~ ;
e 15
L~ g
&
S
S e
W o
Qo &
o o
o
Q
o
<

Q.20002Q

N,

   

Q.000024
0.000053

0.000070

0.000063

          

0.000058
0.000074

0.0000%90

   

    

 

0.000023
0.00007%

0.00C08

0.00015

0.00022

0.00032

 

    

            

0.0003049

         
   

0.085%4

                  

\ 0.000065
\ \0.000048
| \0.000038
0.000034

i

0.000027

                        

  

//

0.000019
0.000024’

/

0.000018
0,000022

0.000048'
0,000048
00000208

 

 

 

 

VOo

{10

 

Jgo

-O)
5
s

/qé‘o

20°

&

“y
O
3

155

165°

 

 

SEE INSERT

¢80°

2

VALUES ARE IN photons/em™/sec/ watt

0.000035

0.0000179

./
REACTOR

 

10°

 

0.0859y

180°

 

 

CREW COMPARTMENT

Fig. 13.9. Gamma-Ray Scottering for an Aircraft Divided Shield,

115
DWG. 19470A

e

REACTOR CREW COMPARTMENT
0.0005 TC 00002 0.0002 TO 0.0004

 

VALUES ARE IN pho‘rons/cmd/sec/wa’r*r

Fig. 13.10. Gammo-Ray Scatfering for an Aircroft Divided Shield,

TABLE 13.2, TOTAL GAMMA DOSE AT THE OUTSIDE OF THE CREW SHIELD

 

 

 

 

SCATTERED BEAM
N -6 DIRECT BEAM
METHOD (+/hy/watt x 1077
{r/he/ watt x 10'6)
Radia! Skew

This report 2.3 . 0.034 29
ANP-53 3.0*
Experiment 50

 

 

 

 

*Corrected for a leckage ratio of 5.

117
ANP PROJECT QUARTERLY PROGRESS REPORT

S

DWG. 194774

0.8 TO 2 Mev

% {gammos/cmzfsec/wafl/io deg}
o

2 TC 3.5 Mev

1072

3.5 TO 5 Mev

 

-5 5 15 25 35 45 55 65 75 85
a{deg)

Fig. 13.11. Direct-Beam Gamma Rays Arriving
at the Crew Shield as a Function of a.

118

surprisingly good agreement, since for the calcula-
tions in ANP-53 it was assumed that all the gamma
rays had an energy of 2.5 Mev and that they emerged
radially.

A comparison was also made with the air-scatter-
ing experiment{12) which was carried out at the
Bulk Shielding Facility. After corrections had
been made for the difference in reactor-crew sepa-
ration distance, the calculations just described
indicated o dose about five times less than that
measured in the air-scattering experiment. This
disagreement is not to be interpreted as fundo-
mental, since the experimantal arrangement was
but a poor replica of the shield which was calcu-
lated.

 

{IZ)J. L. Meem and H. E. Hungerford, Air Scattering
Experiments at the Bulk Shielding Facility {Preliminary
[ssue), ORNL CF-52.7-37 (July 8, 1952); H, E. Hunger-

ford, private communication,
PERIOD ENDING JUNE 10, 1953

14. LID TANK FACILITY

J. D. Flynn
G. T. Chapman

J. N. Miller
F. N. Watsen

Physics Division

M. E. LaVerne

F. H. Abernathy

ANP Division

Experiments on a unit shield for the circulating-
fuel reflector-moderated reactor, as described in
the previous quarterly report, are in progress.
Fifty-four configurations have been tested to dafe.
Effective removal cross sections have been ob-
tained for oxygen, nickel, ond Inconel.

REFLECTOR-MODERATED REACTOR
SHIELD TESTS

A preliminary series of shielding tests has
been run in the Lid Tank Facility on the basic
type of shield for the reflector-moderated re-
actor.V  Three basic refl ector—intermediate-heat-
exchanger—pressure-shell configurations have been
tested, Various thicknesses and spacings of lead
layers were tested with each configuration, A
typical assembly was then used to determine the
effects of such factors as boron concentration,
replacement of iron by Inconel, water by trans-
former oil, lead by tungsten carbide, etc,

The only berylliium available for this first series
of tests consisted of one tank of beryllium pellets
approximately 1 ft thick with an average beryllium
density of 1.23 g/em® and two beryllium slabs,
21 by 42 by 3.6 in., with a density of 1.84 g/em®,
By placing the two beryllium slabs together to form
a 42-in. square and placing this square in series
with the tank of beryllium pellets, an equivalent
thickness of 11.3 in. of solid beryllium was ob-
tained. The heat exchanger was simulated by
using thin, slab-shaped steel tanks that were 3.5
cm thick and filled with NaF. The sheet-steel
tank walls of the tanks served to simulate the
matrix of tubing that would be present in a full-
scale heat exchanger. Provision was made for the
insertion of capsules of NaF in each tank to permit
measurements of sodium activation at wvarious
points in the heat exchanger. A typical shield
configuration is shown in Fig. 14.1.

 

“)A. P. Fraas and J. B. Trice, ANP Quar. Prog. Rep.
Mar, 10, 1953, ORNL-1515, p. 74.

The capsules consisted of 0,250-in.-0OD aluminum
tubing attached to the lower end of 0.250-in.-dia
aluminum rod. Even though the heat exchanger
tanks were loaded with NaF in as dry a condition
as possible and were closed off, it was feared that
the moisture content might change during the
course of the test and have an effect on the amount
of neutron moderation in the simulated heat ex-
changer. Therefore the central position of each
heat exchanger tank was filled with cans of sodium
rather than with NaF powder. The cans of sodium
were borrowed from the Critical Experiment Fa-

cility.

 

 

     
   

 

 

 

 

 

 

 

 

 

 

 

 

DWG.19871
/VOID
T // - N, S e —— T
- - "l e % 1 -
] ) — e
‘-;‘_\ E ™ - -/, _Cfl;"’ e
W8 hE! ol el T
) / ™ ~ -0 . /, __B\. v
N ~ N, AN B —
Lt '/Cr‘& <t - g // o C\i b et
;(j:) = A0 ST . —
.7 o T o - o
2% . JVoQL LT Hzo
O U7 - N W7 de v
w E—J S dom f///// A |
N L e e
WA g SR
oy e / A WA ]
N ol T L
) @ e W le | -—
'S\\ S 1M ; S 52
AL . N ey ——
= // o // Y . //// _..5() —_—
= L s - 7S o 9
7S 1 : //
A oA T T
T el N i TP 8 _
e AT e
/f"f =20 —| ...
———————————— oy i
""""" o e
Fig. 14,1, Typical Reflector-Moderated Reactor

Shield Configuration in Lid Tonk.

119
ANP PROJECT QUARTERLY PROGRESS REPORT

The regular Lid Tank slabs of lead, iron, and
boron carbide were used wherever these materials
were needed. Dry tanks selected from those used
in previous Lid Tank experiments were employed
to hold the various configurations tested.

Shield Configurations, The first set of experi-
ments was carried out with a simulated heut ex-
changer region that was only 2.75 in. thick. To
determine the shielding requirements for a reactor
surrounded by a heat exchanger of this thickness,
some 14 configurations were tested. The first
configuration mocked up a shield arrangement in
which no heavy shielding material would be placed
at the reactor, that is, an arrangement suitable for
use with a divided shield. The other configurations
tested employed various thicknesses of lead,
various arrangements of the lead relative to the
pressure shell, ond various concentrations of boron
in the water region immediately outside the pres-
sure shell,

Twenty-two other configurations in this first
series of tests employed o heat exchanger region
6.9 in. thick. Four of the configurations were used
for investigating the effect of various thicknesses
of a lead region placed immediately outside the
pressure shell. A few tentative tests indicated
that there was little or no over-all advantage to be
gained from moving the lead away from the pressure
shell, since its increased effectiveness would be
largely offset by its greater radius, and the greater
structural complexity would be undesirable.

Examination of the data from the first 18 configu-
rations indicated that secondary gamma rays were
responsible for most of the gamma dose measured
well out in the water region of the shield. [t
seemed likely that the iron tank walls between the
beryllium pellets and the beryllium slab were a
major source of secondary gamma rays that would
not exist in a full-scale reactor-heat exchanger
assembly. Also, it was decided that a thicker
beryllium region would be desirable and could be
simulated by introducing o layer of water between
the tonk of beryllium pellets and the beryllium
slab. A configuration was assembled with this
extra water region and with 0.250-in,-thick boral
sheets placed against the iron tank walls to inhibit
the production of copture gamma rays in the iron,
Also, a 0.125-in.-thick boral sheet was placed
against the outer face of the lead slabs to prevent
the thermal neutrons from the borated water from
entering the lead region. Five configurations
involved the use of various thicknesses of lead

120

with this arrangement. Three more configurations
were tested to investigate the effectiveness of the
boral sheets and the water layer which had been
inserted between the tank of beryllium pellets and
the beryllium slab.

Five additionol configurations that were similar
but contained no lead were tested to compare the
relative effectivenesses of iron, nickel, copper,
There was practically no difference
between these four materials when allowance was

and lnconel.

made for density variations,

Inasmuch as it is possible that the problem of
containing lead at high temperatures may be solved
at some future date, a test was made with a lead
slab between the source plate and the beryllium
region to simulate o reactor design that might be
used if lead were employed as a moderator coolant.

A substantial saving in shield weight might be
effected through the use of a material more dense
than lead for the gamma shield. Since some tung-
sten carbide was available from supplies that had
been acquired by NEPA, a series of tests was run
with tungsten corbide instead of with lead. Various
arrangements of boral sheet ond lead layers were
employed in an effort to investigate the production
of secondary gamma rays in the tungsten,

Gamma-Ray Attentuation Data, The test results
for the various configurations tested in the Lid
Tank Facility have been plotted directly, without
correction for air gaps. Figure 14,2 shows curves

3 DsG. 19872

10 Ty T L T T T T T T
" 14.3-in-THICK BERYLLIUM REFLECTOR — |

e AT75-in-THICK IRON PRESSURE SHELL
5t NO LEAD LAYER

NO BORON IN THE wATER |

 

 

 

 

 

 

 

   

 

 

 

 

 

 

 

 

GamMma DOSE (mr/hr)

 

 

 

 

 

 

 

10 ‘ !

 

100 10 120 130 140 150 160 170
DISTANCE FROM SOURCE PLATE (em)
Fig. 14.2. Gammo Attenuation with No Lead

Layer for Two Heat Exchanger Thicknesses,
for operation with only water in the Lid Tank
Facility, together with curves for both a thin and
a thick heat exchanger with no lead layers. Note
that despite the relatively large void present in the
assemblies, the mockups for the reflector~heat
exchanger~pressure shell configurations give lower
gamma-ray doses af any given distance from the
source plate than would be obtained with water
only, The effects on gamma dose of various
thicknesses of lead placed immediately outside
the pressure shell are shown in Fig. 14.3 for the
2.75-in.-thick heat exchanger. Figure 14.4 shows
test results for essentially the same configuration
as that of Fig. 14.3, except that the thickness of
the heat exchanger region was increased from
2.75 to 6.9 inches. Note that the thicker heat
exchanger gove some increase in the gamma-ray
dose when used with thick layers of lead (because
of an increase in secondary gamma-ray production),
while with no lead layer, it gave a reduction in
gamma-ray dose by absorbing a substantial amount
of primary gomma radiation,

A careful examination of Figs. 14.3 and 14.4
indicated that the secondary gamma-rays consti-

 

   

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

DWG, 19873
2 T T .
1.3 - in.- THICK BERYLLIUM REFLECTOR
2.75-in-THICK HEAT EXCHANGER
Vi ToTTI 75 -in- THICK IRON PRESSURE SHELL -7
""" T 2.8% BORO!}! IN THE WATER ooy
E
S
E
it
o
Q
Q
<
=
=
<
o
-2 I~ L~
f0 [=7.5-in OF Ph, 13-cm VOID AN O
[“4~in. OF H,0 AHEAD OF OUTER IS5 1= i S
4l TN . .
5x 103 {3-in. OF Pbo——=- 5 |
00 N0 120 130 140 150 160 {7C

DISTANCE FROM SCGURCE PLATE (cm)

Fig. 14.3. Gamma Aftenuation with Yarious Lead
Loyers and a 2,75-in.-Thick Heat Exchanger,

PERIOD ENDING JUNE 10, 1953

tuted the major portion of the gomma-ray dose well
out in the shield when relatively thick layers were
employed. When the moderator region was thickened
and boral sheets were inserted to inhibit neutron
capture in the iron tank walls, the data plotted in
Fig. 14.5 were obtained. Examination of Fig. 14.5
discloses that the oddition of lead just outside
the pressure shell, up to a thickness of 6 in.,
gives a configuration in which the lead is as
effective as could be hoped for. The addition of
.5 in. of lead to give an over-all lead thickness
of 7.5 in. attenuated the gamma flux by a factor of
only 2.4. Thus, the apporent attenuation length
in the last layer of lead was approximately 5.3 cm,
which indicated that the secondary gamma-ray
production in the lead is important in this region,

Neutron Attenuation Data. Figure 14.6 shows a
neutron attenuation curve for a typical configuration
superimposed on o reference curve obtained with
only water in the tank. Note that in spite of the
void space in the assembly, the beryllium was so
etfective in slowing down fast neutrons that the
neutron dose 140 cm from the source plate is well
below that for water alone, and the steeper slope
of the curve indicates a shorter attenuation length

DWG. 188 E4

eregeeeeeeenae —

  
   
  
  

  

 

 

 

 

 

 

GAMMA DOSE (mr/hr)

 

 

 

 

N

", 3-in- THICK BERYLLIUM R
REFLECTOR
6.9 -in-THICK HEAT EXCHANGER
2| 1.75-in.-THICK IRON PRESSURE SHELL
107172 8% BORON IN THE WATER "

 

sy 1073 L
00 110

 

 

 
 

 

1200 13 140 150 160 170
DISTANCE FROM SOURCE PLATE (cm)

Fig. 14.4. Gomma AHenuation with Varicus {.ead
Layers and a 6.9-in.-Thick Heat Exchanger,

121
ANP PROJECT QUARTERLY PROGRESS REPORT

 

 

 

 

 

 

 

 

 

 

 
  
   

sl 1 |
14-in - THICK BERYLLIUM ™~
(EQUIVALENT) REFLECTOR J
6.9-in.-THICK HEAT EXCHANGER |-
1.75-in.- THICK IRON
PRESSURE SHELL
1/8-in. BORAL SHEET
1% BORCN IN THE WATER
=1
LT'F‘?R"ST RUN - — .|
* RERUN.

[.. . =t
) o

      
 

ny

 

 

 

 

 

 

GAMMA DOSE (mr/hr)
S

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

100 110 120 130 140 150 160 170
DISTANGE FROM SOURCE PLATE (cm)

Fig. 14,5, Gamma Attenuation for Various Lead
Layers with Thick Reflector and Heat Exchanger.

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

THERMAL FLUX (neuTrons/crn2 -58c)

 

 

 

[ 11 3-in-THICK BERYLLIUM 1UM (EQUIVALENT) REFLECTOR |
6.9-in-THICK HEAT EXCHANGER |

1.75-in.-THICK IRON PRESSURE SHELL
29% BORON IN THE WATER

19 L J 1 |

100 "o 120 130 140 150 160 17C

 

n

 

 

n

 

DISTANCE FROM SOURCE PLATE [cm)

Fig. 14.6. Thermal-Neutran
Typical Shield Configuration.

Attenuvation for a

122

for those neutrons that do get through the reflector—
heat exchanger—pressure shell assembly. '

Changing the thickness of either the heat ex-
changer or lead regions had little effect on the
neutron attenuation, The neutron-flux curves for
all the configurations covered by Figs. 14.3, 14.4,
and 14,5 fell within 20% of the values given in
Fig. 14,6 for the configuration with the thin heat
exchanger, thin reflector, and 6-in.-thick lead layer.

REMOVAL CROSS SECTIONS

Several measurements have been made to deter-
mine a true cross section for oxygen.
Previously reported values, for example, those
reported by Welton and Blizard,u) have been based
on an average removal effect for oxygen distributed
throughout a hydrogenous shield and are therefore

not comparable to other removal cross sections

removal

because the material should be placed close to the
source and followed by a large thickness of water.

For the recent measurements, a slab of carbon
was inserted next to the source,and an effective
removal cross section for carbon was determined.
Then, a slab of oil (CHz) was measured fo obtain
a difference, by comparison with water datq, in
the effective removal cross section for oxygen and
carbon. From the difference, the value for oxygen
is obtained by addition. The oil measurement was
verified by another measurement for which a slab
of paraffin (C30H62) was used.
as follows:

The results are

Carbon measurement
oc = 0.84 b
Qil measurement (H20-H2C)
Oy 0 = 0,15 b
Oxygen removal cross section from oil measurement
Oy = 0.99 b
Paraffin measurement (3]H 0-C

3 100 -300c

30ts2)
= 6.07 b

Oxygen removal cross section from paraffin meas-
urement
g =

5 = 1.01b

In addition, an effective removal cross section

for nickel and a removal relaxation length for

 

(2)T. A, Welton and E., P. Blizard, The Shielding of
Mobile Reactors (Part [1}, TID-2002, p. 73 (Aug. 1952).
Inconel have been measured. The results are as

follows:

Nickel, o

Inconel,
The relaxation

wi = 1.85b,

A= 633 cm,
length obtained for
probably accurate for specifying the performance

il

Inconel is

of the alloy, but it is not adequate for determining
the removal cross section for chromium, since this
element constitutes only 13.5% of the alloy.

FACILITY MODIFICATION

During April, a new shutter and an iris made with
1/S-in. boral were installed behind the Lid Tank
source plate to replace those made of boron-painted

PERIOD ENDING JUNE 10, 1953

aluminum. This change was necessary because
the gradual sagging of the Lid Tank toward the
pile restricted the movement of the a-in.-fhick
shutter,

The iris, the purpose of which is to limit the
effective source to a circle, had previously been
unsatisfactory because some of the boron paint
had crumbled and partly filled the circular space
which should have been boron-free. The new de-
sign makes this occurrence impossible. The
neutron and gamma fluxes in the water of the tank
have been remeasured since the new shutter and
iris have been installed, and the new flux values
agree very well with the values obtained previ-
ously.

123
15. TOWER SHIELDING FACILITY

3. Y. Estabrook, ANP Division

C. E. Clifford
L. B. Holland

T. V. Blosser
M. K. Hullings

Physics Division

H. E. Stern, Consolidated Vultee Aircraft Corp.

Although no structural steel has been delivered
to the site of the Tower Shielding Facility, the
tower foundation and cable anchors are completed
and the construction may be said to be well under
way, A three-week delay occasioned by a strike,
while inconvenient, will probably not cause «a
setback in the deadline for final completion at the
end of 1953,

Actuating mechanisms for the control rods have
been built and are undergoing tests., The first
models ore unsatisfactory, but it is hoped that

124

they can be modified and that the defects can be
remedied.

A report to the Reactor Safeguard Committee is
essentially complete.!!? In the course of the study,
it appeared that there were no unusual hazards
for this particular reactor which could not be
minimized by careful design and prudent operation,

(Me. E. Clifford and L. Abbott, Tower Shielding
Facility Hazards Summary Reporf, ORNL-1550 (to be

issued).
 
REPORT NO.

CF-53-3-184
CF-53-3-210

CF-53-4-43
CF-53-4-195

ORNL.-1509

ORNL-1517

ORNL-1535

CF-53-3-125
CF-53 3231
CF-53-4-29

ORNL-1178

CF-52-12-225
CF-53-2-250
CF.53.3-81

ORNL-1526
CF-53-3-166

CF-53-4-67
CF-53-4-76
CF-53-4-108
CF-53-4-118
CF.53-4-140

16. LIST OF REPORTS ISSUED DURING THE QUARTER

TITLE OF REPORT

{. Reactor Design
Comments on the ARE Review Committee Criticisms

A Reflector-Moderated Circulofing.Fuel Reactor for
an Aircraft Power Plant '

Minutes of the Final Meeting of the ARE Design

Review Committee

Design Review Committee Report on Tower Reactor

Facility

Preliminary Design and Performance Studies of

Sodium~to-Air: Radiators

Fireball Reactor Moderating Cooling Sy stem

Thermal Dynomic ond Heat Transfer Analysis of
the ARE
i{. Reactor Physics
Multigroup Method Calculations
Kinetics of the Circulating Fuel Nuclear Reactor

Study of the Kinetics of the General Electric Air-

craft Reactor

Dynamics of an Aircraft Reactor

{1I. Shielding
Preliminary Direct Cycle Reactor Assembly - Part V.
The Shielding of Nuclear Aircraft

Radiation Cutqracf Meeting in Washington, D. C.,
February 28, 1953

A Neutron Timé-of-Fléghf Spectrometer

Attenuation by Lead of Fast Neutrons from a Fission

Source
Foil Exposures
ANP Engineering Test Facility Requirements
Fuel Elements for Bulk Shielding Reuctor
Fuel Elements for Tower Shielding Reactor

1953 Summer Shielding Session

AUTHOR(S)

d. He Buck
A. P. Fraas

We R, Gall
CQ .’a BDrkOWSki

A, P, Fraas
Mo Jo Stumpf
W. 5. Farmer
C. D. Whitman

£. P. Blizard
W, 5, Farmer
A. H. Fox

A. P, Fraas

C. B. Miils

B. L., Greenstreet

An H. FOX
W. K. Ergen

E. R. Mann
J. J. Stone

Nuclear Development

Associates, Inc,

Dixon Callihan et of.
E. P. Blizard
J- La Meem

G. S. Pawlicki

J. D, Flynn
G. T. Chapman

M. K, Hullings
E. P. Blizard

Jo L. Meem

C. E. Clifford
£, P. Blizard

DATE ISSUED
3.23-53
3.27-53
4.7.53
4.22-53

to be issued

to be issued

to be issued

3-16-53
3-30-53
4+7-53

4-22-53

12-15-52
2.27-53
3-11.53

4.9.53
3-10-53

4.10-53
4.13-53
4-17.53
41453
4.17-53

127
ANP PROJECT QUARTERLY PROGRESS REPORT .

REPORT NO.

ORNL-1438

CF-53-5-21

ORNL-1537

ORNL.-1550

CF-53-2-165
CF-53-3-142

CF-53-5-47

ORNL-1538

CF-53-1-256

CF-53-3-100

CF-53-3-183
CF-53-4-46

ORNL-1514

ORNL-1540

MM-35

BMI-788

BMI-787

BMI-792

BMI-794

128

TITLE OF REPORT

Determination of the Power of the Bulk Shielding
Reactor ~ Part {l. MNeutron Flux Measurements
in Several Beryllium Oxide-Reflected Critical

Assemblies

Energy per Fission and Power of the Bulk Shielding

Reactor

The Determinationof the Power of the Bulk Shielding
Reactor — Part i1

Tower Shielding Facility Safeguard Report

1V. Chemistry
Analytical Chemistry ANP Program for 2/15/53

Fused Salt Systems, Chapters 6.1, 6.2, 6.3, and 6.4
of Reactor Engineering Handbook

Preparation of Enriched Uranium for Fireball Critical

Experiments

The Determination of Traces of Hydrogen Fluoride

in Inert Gases

V. Metallurgy ond Ceramics

Fluoride Corrosion Tests of Type 347 Stainless
Steel Weldments

Examination of As=-Received 4-Ply Inconel Bellows

Fabrication of Spherical Particles

Evaluation of Properties of Solid Fuel Elements

Cone-Arc Welding

Behavior of Chronomal

Quarterly Progress Report on Production of Sound
Ductile Joints in Molybdenum

The Effect of Purification on Welds in Molybdenum
Evaluation of Oxidation-Resistant Ceramics for
High-Temperature Reactor Elements

Properties of Some Columbium Oxide-Basis Ceramics

Continued Studies of Corrosion by Fused Coustic

AUTHOR(S)

E. B. Johnsen
J- L.o Meem

Jn L.a Meem
R. E. Mungerford
E. B. Johnson

Je Lo Meem

C. E. Clifford

J. C. White

W, R. Grimes
D. R. Cuneo
W. B. Cotirell

F. F. Blankenship

A. D, Callihan

D. L., Manning
J- Cn Whii‘e

D. C. Vreeland

G, M. Adamson
Ro Sc Crouse

Au Levy

A. Levy
E. 5. Bomar

P. Patriarca
Go Ms S‘OUQ"]“Q&'

Wc Jo Sfurm
Ro Jl Jones

Bottelle Memorial
Institute

Battelle Memorial
Institute

Baottelle Memorial

Institute

Battelle Memorial

Institute

Battelle Memorial

Institute

DATE ISSUED

4-21.53

4-25-53

to be issued

6-9-53

2.20-53
3-20-53

5.7-53

4-28-53

1-24-53

3-4-53

3-19-53
4.1-53

3-26-53

to be issued

1-30-53

12-1-52

12-1-52

12.15-52

12-18-52
REPORT NO.

BMI1-804

MM-101

CRO-79

1G~800-5-7

1G-800-5-6

CF-53-3-176
CF-53-3-261

CF-53-3-259

CF-53-5-103

CF-53-5-113

CF-53-5-149

NAA-SR-208

CF-53-2-246
ORNL-1515

CF-535-234

TITLE OF REPORT

Protective Coatings
Alloys

Progress Report No. 2 on Gas Plated Coatings on

Metals and Alloys

Coatings for the Protection of Low Carbon Steel at

Elevated Temperatures

Seventh Bimonthly Progress Report on Development
of a High Temperature Strain Gage Jan, 1, 1253 to

Feb. 28, 1953.

Sixth Bimonthly Progress Report on Development of
a High Temperature Strain Cage. November 1, 1952
to December 31, 1952,

High-Temperature Metal

PERJOD ENDING JUNE 10, 1953

AUTHOR(S)

Battelle Memorial
Institute

The Commonwealth
Eng. Co. of Dhio

University of

Alabamao

Cornell Aeronautical

L aboratory

Cornell Aeronautical

L aboratory

Vl. Heoat Transfer and Physical Properties

Minimum Weight Analysis for an Air Radiator

Physical Property Charts for Some Reactor Fuels,

Coolants, and Miscellaneous Materials, 3d ed.

Preliminary Surface Tension Measurements of the

ARE Fuel (Fluoride Mixture No, 30)

Heat Capacity of the Eutectic of Lithium Hydroxide
and Sodium Hydroxide

Meat Capacity of Fuel Composition No, 14

Preliminary Study of the Electrical Conductivity of
Strong Electrolytes for Possible Application in
Yolume Heat Source Experiments,

Cyclotron Irradiation of Fused Flucrides

ANP Information Meeting of February 18, 1953

Aireraft Nuclear Propulsion Project Quarterly Prog-
ress Report for Period Ending March 10, 1953.

AMNP Information Meeting of May 20, 1953

Miscelloneous

W. S, Farmer

ANP Physical
Properties Group

So 1. Cohen
T- Nl Jones

W. D. Powers
G. C. Blalock

W. D. Powers
G. C. Blalock

N. D. Greene

North American

Aviation, inc,
w. B. Cthel!
W. B. Cottrell

J. M. Cisar

DATE ISSUED

12.15.52

4-14-53

11-52

3-19-53

1-12-53

to be i ssu_ed

3.20-53

3-27-53

5-18-53

5-18-53

5-19-53

1.26-53

2.26-53
4.16-53

to be issued

129