ORNL-1294.txt

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ORNL- 1294

This document consists of 202 pages

Ty

Copy é@;&f0f1239» Series A.
Contract No. W-7405-eng-26
AIRCRAFT NUCLEAR PROPULSION PROJECT
QUARTERLY. PROGRESS REPORT

for Period Ending June 10, 1952

R. C. Briant, Director
- J. H. Buck, Associate Director

A, J. Miller; Assistant Director

. Edited by:

W. B. Cottrell

DATE 1SSUED

 

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ORNL-1294
Progress

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e
b
A

 
 

Reports previously issued

ORNL-528
ORNL-629
ORNL-768
ORNL-858
ORNL-919
ANP - 60
ANP-65
ORNL-1154
ORNL-~1170
ORNL-1227
ORNL-1234

Period
Period
Period
Period
Perio&
Perioa
Period
Period
Period

Period

in this series are as follows:

Ending
Ending
Ending
Ending
Ending
Ending
Ending
Ending
Ending

Ending

November 30, 1949

February 28, 1950
May 31, 1950
August 31, 1950
December 10, 195Q
March 16, 1951
1951
September 10, 1951
December 10, 1951

1952

June 10,

March 10,

Reactor Program of the Aircraft
Nuclear Propulsion Project

 
b

e ]
 

TABLE OF CONTENTS

PAGE

FOREWORD 1
PART I REACTOR THEORY AND DESIGN 3

SUMMARY AND INTRODUCTION , 5
1. CIRCULATING-¥FUEL AIRCRAFT REACTORS 6
Reflector-Moderated Circulating-Fuel Beactor : 6

- Spherical Heat Exchanger and Shield Design 7
2. CIRCULATING-FUEL AIRCRAFT REACTOR EXPERIMENT 9
Core and Pressure Shell 9
Stress analysis of the pressure shell 9

Fluid Circuit . . 10
Control 12
Operation | _ 12
Pretesting 12
Critical loading : 12

Power operation : 13
Shutdown 13
Building ' 13
Electrical power 13
Shielding by concrete pits 13

3. EXPERIMENTAL REACTOR ENGINEERING 15
Seals and Closures | 16

Seal tester 16
Stuffing-box seals 16

Bellows tests . _ . 17
Oval-ring seal | 17

 

viil
4»

viliil

 

Pumps
ARE centrifugal pump

Laboratory-size maintained-level gas-sealed pump

Worthite frozen-sodium-sealed pump

Durco convection-cooled frozen-sodium-sealed pump

Durco frozen-fluoride-sealed pump
Fluoride pumps with stuffing-box seals

Valves

Wide-open valve

Closed valve

Throttling valve
Canned-rotor-driven valve

Heat Exchangers

NaK-to-NaK heat exchanger

Tube-matrix spacer tests

Sodium-~to-air radiator
Fluoride-to-ligquid-metal heat transfer

Instrumentation

Pressure measurement
Liquid-level indicators

Fluid Dynamics of the ARE

Technology of Fluoride Handling

Fluoride production
Fluoride removal
Pretreatment of pipeline helium

system

Pickling procedures for fluoride containers

REACTOR PHYSICS

Oscillations in the Circulating-Fuel Aircraft Reactor

Calculations for the Circulating-Fuel ARE

Methods for Reactor Computation

 

17

17
17
18
18
18
19

19
20
20
20
20
20

22
22
24
26
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26
26
28

29

29
29
29
30

31
31
32
33
A e R AR R A A a R a e a BR A AP A

5. CRITICAL EXPERIMENTS

Direct-Cycle Reactor

Shelf-type reactor assembly
Box-type reactor assembly

Circulating-Fuel Reactor

Reactor materials
Beactor assembly

Graphlte Reactor

Crlrlcal mass

Control rod calibration

Gap ef fect

Flux and power distribution
Danger coefficient measurements

PART 11 SHIELDING RESEARCH
SUMMARY AND INTRODUCTION

6. BULK SHIELDING REACTOR ‘
Mockup of the Divided Shield

Air-Scattering Experiments

Shield arrangement

Radiation dosage at crew compartment

Attenuation of gamma rays by lead

Comparison of experiment with standard de51gn conditions

Determination of the Albedo of Neutrons and Gamma Rays
Irradiation of Animals

Irradiation of Electronic Equipment

7. GAMMA-RAY ATTENUATION EXPEBIMENTS IN THE LID TANK
63% Iron-37% Water Thermal Sh1eld

 

34

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38

38

38

39

39
39
39
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40

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43

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47
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47
49
49
52
54
54
54
54

1X
 

63% Iron—-37% Borated Water Thermal Shield
Solid Iron Thermal Shield in Water
Solid Lead Shadow Shields in Water

8. DUCT TESTS

Duct Tests in the Thermal Column

Straight ducts

Ducts with bends
Wall-scattering experiment
Comparison with theory

GE-ANP Annular Air Ducts

9. NUCLEAR MEASUBREMENTS
Cross-Section Measurements with Van de Graaff Accelerator

Time-of-Flight Neutron Spectrometer

PART III MATERIALS RESEARCH

SUMMARY AND INTRODUCTION

10. CHEMISTRY OF HIGH-TEMPERATURE LIQUIDS
Low-Melting-Point Fluoride Fuel Systems
NaF-KF-ZrF, -UF,
KF-ZrF, -UF,
NaF-ZrF, -UF,
NaF-RbF-ZrF, -UF,

Analyses of Fluoride Compounds

X-ray examination of solid, complex fluorides
Spectrographic analysis

Study of solid phases in the NaF-BeF,-UF, system
Petrographic examination of fluorides

 

35
55
55

69

69

69
71
71
72

72

76
76
76

79
81

82
83

83
85
85
85

85

85
86
86
88
 

Simulated Fuel Mixture for Cold Critical Experiment

Preparation of Pure Hydroxides

Sodium hydroxide
Lithium hydroxide
Potassium hydroxide

Coolant Development

NaF-ZrF,
KF-ZrF,
RbF-Zrk,
Nal-KF-ZrF,
NaF-RbF-Zrf,
NaF-ZnF,
KF-ZnF,
RbF-Znt,
NaF~-KF-ZnF,

Fuel Preparation and Liquid.Handling

Zirconium fluoride production
Preparation of pure fuel mixtures
Liguid handling equipment

Solubility of Uranium in Sodium Cyanide

11. CORROSION RESEARCH

Static Corrosion by Fluorides

Effect of additives
Temperature dependence
Effect of plating metals
Effect of cold-work

Corrosion of ceramic materials

Static Corrosion by Hydroxides

Effect of additives
Effect of plating metals

 

89
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99

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101

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103
103
104
104

105
105
105

X1
’llllll'llll‘lr"l.'l.l.l'iil'lll'llllfi

Dynamic Corrosion by Liquid Metals 107
Dynamic Corrosion by Fluorides 107
Corrosion by fluorides in seesaw tests 107
Corrosion by fluorides in thermal convection loops 109
Corrosion in rotating dynamic test rig 116
Dynamic Corrosion by Hydroxides 116
Corrosion by hydroxides in seesaw tests 116
Standpipe tests of hydroxide corrosion 118
Hydroxide corrosion in cold-finger thermal convection
apparatus 118
Modified thermal convection apparatus ~ TCA 118
Fundamental Corrosion Research 119
Interaction of fluorides and structural metals 119
Synthesis of complex fluorides 120
Examination of corrosion products 121
X-ray~-diffraction studies 123
EMF measurements in fused fluorides 124
Reactions in fused sodium hydroxide 124
Compounds resulting from hydroxide corrosion 125
Mechanism of fluoride corrosion 126
Free-energy relations of fluorides and structural metals 128
Solution of metals in molten halides 128
Fluoride corrosion phenomena 129
12. METALLURGY AND CERAMICS 131
Loose Powder Bonding 131
Sintering temperature 132
Sintering time 132
Fuel-componcnt particle size 132
Surface preparation 135
Sintering under load 135
Control Rod Fabrication 135
Safety rod slugs 135
Regulating rod slugs 135

 

xii SN PR,
 

Mechanical Testing of Materials 136

Stress-rupture tests in argon 136
Stress~corrosion tests : 136
Cone-Arc Welding | | 138
"Effect of welding conditions : 138
Heat exchanger assembly ' 139
Tests of Brazing Alloys _ 140
Nicrobraz ‘ _ 140
Manganese-nickel 140
‘Palladium-nickel ' ' 143
Ni-Cr-Si-Mn alloys 143
Silver-base alloys - 143
Ceramics 144
Ceramic coatings for metals ' 144
Ceramics laboratory program ' 146
13. HEAT TRANSFER AND PHYSICAL PROPERTIES RESEARCH 146
Viscosity of Fluoride Mixtures 146
Ber-bearing'fluorides 147
ZrF,-bearing fluorides - 147
Material from corrosion loops : 148
NaF-KF-LiF and NaF-KF-LiF-UF, mixtures : 148
Thermal Conductivity - | 149
Heat Capacity _ 149
Density : 150
Vapor Pressure of Fluorides . 150
Beryllium fluoride | 150
ZrF,-bearing fluoride mixtures 150
Physical Property Data | | 152

Heat Transfer in Sodium Hydroxide 152

 

X111

88O T R T 0,08 b A R R S e n s A s s s
)

et

Boiling Heat Transfer
Entrance Region Heat Transfer

Natural Convection in Confined Spaces with Internal Heat
Generation

Circulating-Fuel Heat Transfer Studies

14. RADIATION DAMAGE

Irradiation of Fused Materials
Pile irradiation of fuel
Cyclotron irradiation of fuel

Inpile Circulating Loops

Creep Under Irradiation

Radiation Effects on Thermal Conductivity

PART IV APPENDIXES
SUMMARY AND INTRODUCTION

15. ANALYTICAL CHEMISTRY

Analyses for Components of Fluoride Mixtures

Total alkalil metals

Sodium, potassium, and lithium
Beryllium

Zirconium

Analyses for Impurities in Fluoride Mixtures

Iron
Chromium

Analyses for Impurities in Fluoride Compounds

Sulfate and sulfide
Zirconium oxide and oxyfluoride
Water

 

156
156

158
159

159

160
160
163
163
164

164

167
169

170

170

170
170
170
170

171
171
171
171

171
172
172
Service Analysis
16. LIST OF REPORTS ISSUED

17. DIRECTORY OF ACTIVE ANP RESEARCH PROJECTS
Reactor and Component Design
Shielding Research
Materials Research

Technical Administration of Aircraft Nuclear Propulsion
Project at Qak Ridge National Laboratory

 

172
173
176
176
178
179

184
 

 

 

 

 
 
ANP PROJECT QUARTERLY PROGRESS REPORT

FOREWORD

The Aircraft Nuclear Propulsion Project at the Qak Ridge National Laboratory
i1s comprised of some 300 technical and scientific personnel engaged in many
phases of research directed toward the nuclear propulsion of aircraft. A
considerable portion of this research is carried out to provide support for
other organizations participating in the national ANF effort. However, the
purpose of the bulk of the ANP research at OBNL is the develdpment of a circu-
lating-fuel type of reactor. The nucleus of this effort 15 now centered on
the Aircraft Reactor Experiment mia 3-Mev, high-temperature prototype of a

circulating-fuel reactor suitable for the propulsion of aircraft.

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 of the Laboratory under its contract,
W-7405-eng-26. The report is divided into four parts; I. Reactor Theory and
Design; II. Shielding Research; I11. Materials Research; and IV. Appendixes.

"

Each part has a separate "Summary and Introduction.

 
 
SUMMARY AND

Analysis of the circulating-fuel
atrcraft reactor systems incorporating
intermediate heat exchangers has led
to the use of a spherical-shell type
of heat exchanger and shield ar-
rangement. This arrangement has
resulted in an engineered reactor-
shield assembly weighing about 8000 1b
less than the best arrangement previ-
ously examined. A most promisimg
design study has been undertaken of. a
circulating-fluoride-fuel ceactor
core in which a thick reflector
replaces the moderator. Such a
design, which results in a homogeneous-
fluoride reactor, has been shown teo
have a low critical mass and a neutron
lifetime more than adequate to permit
control. This compact, spherical-
core design is compatible with the
low-weight spherical-shell type of
heat exchanger and shield arrangements

(sec. 1).

The Aircraft Reactor Experiment to
be constructed by the Oak Ridge
National Laboratory was described in
the last report. Only minor modifi-
cations in the design have occurred
during the past guarter; the most
significant changes were the reduction
in number of fuel passes through the
core and the reversal of the inlet
and outlet fuel headers. The design
effort is concentrated on detailing
drawings of components of the reactor
and the procurement of these parts.
The building for the ARE was released
on June 6 to ORNL by the construction
contractor, and installation of
equipment 1is proceeding as rapidly
as the receipt of parts permits
(sec. 2). '

The emphasis of the research and
development program for reactor
plumbing and assocciated hardware
continues to be placed on the tech-
nology of high-temperature fluoride
mixtures (sec. 3). Both pumps and
valves have been operated successfully

INTRODUCTION

with molten fluoride mixtures for
extended periods of time at 1500°F.
In both instances the problem reduces
to that of sealing a moving shafg;
this has been satisfactorily effected
by both frozen seals and stuffing-box
seals. NaK-to-NaK and Na-to-a1ir
heat exchangers have beenoperated — the
former for over 3000 hours. - In
addition considerable effort has been
devoted to developing instruments for
measuring flow and pressure and for
liquid level indication and control
as required both by the ABE and the
experimental high-temperature-liguid
systems. Satisfactory technigues
have been developed for the manu-
facture, transfer, and loading of
fluoride mixtures without introducing

contamination. .

Studies of the kinetics of the
circulating-fuel aircraft reactor
have led to the conclusion that an
spite of the loss of some delayed
neutrons power oscillations i1n such
reactors will be strongly damped by
a mechanism associated with the actual
circulation of the fuel (sec. 4).
Physics research for the aircraft
reactor experiment has involved
numerous static calculations conse-
guent to significant, but more or
less minor, design changes.

A critical assemblyof the box-type,
direct-cycle reactor was investigated
for the GE-ANP program. Fission rate,
flux measurements, reflector effects,
and temperature response were de-
termined. Upon completion of these
me asurements, a mockup of the circu-
lating-fuel aircraft reactor will be

assembled - the components are now
on hand., ' The beryllium oxide machined
for the ARE will be used 1in the

critical assembly, tegether with a
simulated fluoride mixture. A resume
of the results from an earlier series
of experiments with a graphite-~
moderated reactor is also presented

(sec. 5).
ANP PROJECT QUARTERLY PROGRESS REPORT

1. CIRCULATING-FUEL AIRCRAFT REACTORS

A. P.

Fraas,

Several circulating-fuel aircraft
reactor systems employing closely
coupled i1ntermediate heat exchangers
were described in the last report.¢1!)
The extension of these design studies
led to consideration of a spherical-
shell type of heat exchanger and
shield arrangement that resulted in a
saving of about 8000 lb in engineered
shield weight over the best arrangements
previously examined.¢?? 1In addition,
a modified core design having all of
the moderator removed from the core
and added to the reflector now appears
to be an attractive possibility.
With the modified design, the critical
mass would not be large, the neutron
lifetime would be longer than with
previous designs {(which would ease
the control problem), and the structure
in the core would be eliminated.

REFLECTOR-MODERATED CIRCULATING-FUEL
REACTOR

Calculations made on the basis of
relations derived for thermal reactors
had i1ndicated that circulating-fuel
reactors employing a mixture of
fluoride salts as a homogeneous
working fluid would reguire a very
high critical mass and would be
difficult to control. Multigroup
two-region IBM machine calculations
made during the past quarter have
shown that this 1s not the case.
Removal of the moderator from the
core of a circulating-fuel reactor
does not result 1n excessive fuel
1f a thick reflector

In a large measure
this comes about because the removal
of the moderator means the
removal of the poisoning effect of a

requirements
1s employed.

also

 

(1)

Atrcreft Nuclear Propulsion Project Quarterly
Progress Report for Period Ending March 10, 1952,
ORNL-1227, p. 7.

Dieia., . 12.

ANF Division

substantial amount of stainless steel.
Critical mass calculations have been
made on the IBM machines for two core
diameters. With a 12-1a.-~thick
beryllium oxide reflector, critical
masses of 23 and 19 lb were obtained
for 16- and 32-in.-dia spherical
reactor cores. The structural sim-
plicity and attendant advantages
of this type of reactor are obvious.

At first 1t also seemed that these
unmoderated circulating-fuel reactors
would have a high median energy for
fission and hence might present a
serious control problem. However,
about one-~third of the fissions are
caused by thermal or near thermal
that have spent a large
fraction of their lifetimes 1in the
relatively poison-{free reflector.
Thus the effective neutron lifetime
from the control standpoint may
actunally be greater than in beryllium
oxide~moderated <circulating-fuel
reactors that have of necessity a
substantial amount of stainless steel
1n the core.

neuntrons

It was also feared that the power
density at the core-reflector interface
would be many times the average power
density. However, the attenuation
length for the fission-producing
neutrons from the reflector was about
10 ¢m, so the high power density
region was found to be quite thick.
Since approximately two-thirds of the
volume of a sphere lies between 1ts
surface and the surface of a smaller
concentric sphere, or "island," having
a radinus two-thirds that of the
larger sphere, much of the volume
of the core falls i1n the high power-
density region. As a result, the
ratio of peak-to-average power density
was about 2 for both of the
studied.

reactors
Further calculations are being
made "to determine the effects of a
spherical island of moderator in the
center. Since the extra structural
complication should not prove to be
too great, this may be very worth
while from the standpoint of both
uranium investment and power density
considerations.

SPHERICAL HEAT EXCHANGER AND SHIELD
DESIGN

The last report®?’ showed the large
saving in shield weight effected by
going from the tandem to the annular
heat exchanger arrangement. This
suggested that a further reduction in
weight might be realized by going to
a more nearly spherical configuration.
A heat exchanger in which the tube
matrix is laid out in the form of a
spherical shell was therefore devised
to form a basis for further designs.
The basic configuration envisioned is
shown 1n Fig. 1; note that it ia-
corporates the reflector-moderated
reactor discussed above. This spheri-
cal heat exchanger design reduced the
over-all weight of the reactor, heat
exchanger, and shield combination for
the design conditions®?? to only
90,000 1b for the divided shield and
120,000 1b for the mear unit shield
(a shield permitting 3 r/hr 50 ft
from center of reactor), a saving of
roughly 8000 1b over the annular heat
exchanger arrangement. (!} Both shields
are designed for 1 r/hr in the crew
compartment.

The heat exchanger tubes are grouped
in bundles and each bundle has a
rectangular cross section except where
the bundles terminate in circular-disk
headers. If the two pumps are at the
North and South poles respectively,
each tube bundle will lie on a vari-
able-pitch helix running from ap-

proximately the "Arctic" to the

FOR PERIOD ENDING JUNE 10, 1952

"Antarctic Circle." The angle between
tube centerlines and the "equator™
will be about 40 deg, whereas the
corresponding angle at the Arctic
Circle will be about 90 degrees.
This will make 1t possible to keep
the center-to-center spacing of the
tubes independent of "latitude.™
Although the tube bundle shape is
quite unusual, fabrication of the
heat exchanger should not be too
difficules. After jigs have
prepared for building up a tube bundle,
one tube bundle after another can be
assembled, brazed and/or welded into
i1ts header disks, pressure tested,
inspected, and assembledin the spheri-
cal, steel shell. A simple fillet
weld between the header disk and the
pressure shell should suffice to
complete the installation.

be eIl

In other respects the heat ex-
changer and shield design of Fig. 1

is essentially similar to the tandem
and annular arrangements described
in the previous report.¢!’ The 11-

in.-thick reflector of the annular
heat exchanger design was used again
as was the same type of shield con-
struction employed ‘in both of the
other designs. One difference is
the use of boron carbide in the baffles
between tube bundles instead of in

"dead!" tubes.

This reactor design is suggestive
of a homogeneous reactor, which 1t
really approaches qguite closely.
The thick beryllium oxide reflector
could be cooled by either metallic
sodium or a nonuranium-bearing fluoride
salt. It should be possible through
careful designof the moderator cooling
passages and the conical diffusing

vanes at the reactor core inlet to
keep the metal temperature at all
points below the average temperature

of the fuel leaving the reactor. Thus
the hottest metal anywhere in the
system would be the heat exchanger
tube walls at the fuel inlet end.
ANP PROJECT QUARTERLY PROGRESS REPORT

ARE—

PUMP DRIVE SHAFT DWG. 156486

  
   

\\\ L--NoK INLET

NaK INLET—w2,

 

 

 

 

      
 

 
 

PRESSURE SHELL
BORATED H,0

    
 

INNER PRESSURE 7
SHELL OF SHIELD g

 

   

—--HEAT EXCHANGER
TUBE BUNDLE

 

   

 

 

  

BORATED Hp0

  
 

FUEL PUMP

  

SELF SEALING RUBBER-—
\ NeK QUTLET

NeK QUTLET
PUMP DRIVE SHAFT

Spherical Heat Exchanger — Shield Arrangement for a Circulating-Fuel

Fig. 1.
Reactor.
FOR PERIOD ENDING JUNE 10, 1952

2. CIRCULATING-FUEL AIRCRAFT REACTOR EXPERIMENT

R. W.

The reactor system descriptions
given in the previous report{?) is
applicable to the current system and
will not be repeated. Furthermore,
detailed descriptions and illustrations
of this its com-
ponents are the topic of a separate
. (2) . . , . .
report. '~ The several changes made
duringthe last quarter and the progress
made

reactor system and

system and
procuring outside manufactured com-
ponents will be discussed,
a procedure that has ‘been outlined
for the operation of the reactor.

in detairling the

as well as

CORE AND PRESSURE SHELL

The core design has been revigsed
to provide 11 tubes per serpentine
pass 1instead of 13. This change was
dictated by the number of beryllium
oxide-moderator blocks available and
had no major effect on the functional
performance of the core. Calculations
indicated a fuel-fi1lm temperature
gradient of approximately 4C°F due to
moderator heat inflow. This maximum
gradient occurred near the core center
and, as the {fuel exited from this
region, was additaive to a 1300°F
datum, which caused a fuel~tube wall
temperature of 1540°F.° Accordingly,
the fuel flow direction was reversed
so that the 40°F gradient is additive
to a 1150°F datum (fuel inlet temper-
ature). The moderator heat inflow is
very small at the fuel ocutlet end
(core periphery), seo the maximum
wall temperature is expected to be in

the region of 1500 to 1510°F.

 

(1)

Aircraft Nuclear Propulston Project Quarterly
Progress Report for Period Ending Murch 10, 1952,
ORNL-1227, p. 13,

(2)

Recctor Program of the Adircrajt Nuelear

Propulsion Project, ORNL-1234, ‘to be issuecd.

Schroeder,

ANP Division

The core and pressure shell as-
sembly has been completely detailed.
The elevation section of the
shown in the last report®3? is ap-
plicable except for the described
changes 1n fuel 1inlet and outlet
headers and the number of fuel passes.
A plan section of the reactor showing
the revised presented 1n
Fig. 2. A better concept of the core
itself may be obtained from a photo-
graph of the beryllium oxide blocks
that have been sized and stacked
(Fig. 3). 1In the future these
blocks will be used for a criticality
test prior to their use iIn the ex-
perimental reactor. The pressure
shell will be welded by lLukenweld,
and its shipment within several weeks
is anticipated. The fuel tubes, as
well as all other core material, have
been ordered, and delivery of all
outstanding material is scheduled for
June, July, and August 1952.

core

core 1s

near

Stress Analysis of the Pressure
Shell (R. L. Maxwell and J. W. Walker,
Consultants, ANP Division). The
analysis of the stresses in the pres-
sure shell as a function of the
internal pressure has been completed.
At a preéssure of 30 psi the maximum

bending stress on the heads of the
shell 1s 2560 psi. This 1s approxi-
mately one-half the stress for a

creep rate of 0.10% an 10,000 hr
for Inconel at the operating temper-
ature of 1150°F. There will be stress
concentrations due to holes 1n the
head, but in all these
holes occur in areas of high stress
the holes have been reinforced. Wich-
out reinforcement the stress con-
centration factor would not be greater

cases where

 

(3 op. cir., ORNL-1227, p. 14,
ANP PROJECT QUARTERLY PROGRESS REPORT

 

 
  
 
 
 
  
 
 

TUBE COIL B
(INCCNEL )

TUBE COILL A
(INCONEL)~

........

 

  
 
 
 
  
  

 

 

 

 

 

 

 

Fig, 2.

than 2, so the creep rate would not
exceed 0.10% in 10,000 bhours. Also,
the stress concentration would be
relieved at operating temperatures.

The stresses on the sidewalls of
the shell as a function of internal
pressure have also been analyzed.
At a pressure of 30 psa
axlal stress 1n the cylinder 1s
2230 psi. This 1is somewhat less than
the maximum stress 1n the
head. The creep rate is less
one-half the creep rate of 0.10% 1n
10,000 hours.

the maximum

Fending
than

10

DWG, 15547

INSTRUMENT TUBE
(INCONEL)

REFLECTOR EDGE BLOCK
°\ ( BERYLLIUM OXIDE)
\\\\\\ - — REFLECTOR BLOGK
(BERYLLIUM OXIDE)

_———CORE BLOCK
[BERYLLIUM OXIDE)

-—— REFLECTCR COOLANT
TUBES (INCONEL)

—— CAN
{INCONEL)

 

----- CORE SLEEVE
(INCONEL)

 

SAFETY ROD GUIDE
SLEEVE (INCONEL)

SCALE INAIEégES

Plan Section of the Experimental EKeactor.

FLUID CIRCUIT

The fluid circuit remains as
described and 1llustrated in the
previous report and recent efforts
have been devoted to detailing and
ordering outside manufactured com-
ponents. Many detailed drawings have
been released for manufacture, in-
cluding the helium ducts for the main
heat disposal loop, the radicactive-
fuel dump tank, Of outside-
manufactured compenents, the Vickers
hydraulic drive systems for the main
heat disposal have been

ebtc.

circuits
FOR PERIOD ENDING JUNE 10, 1952

UNCLASSIFIED
Y-{12 PHOTO 6-4905

 

1l
ANP PROJECT QUARTERLY PROGRESS REPORT

and most of the other com-
ponents, including heat exchangers,
blowers, heaters, etc., are scheduled
for delivery in June, July, and
August 1952. Delivery of several
components, including heat exchangers
and valves, has been jeopardized by
the manufacturer’'s 1inability to
procure stainless steel, and it has
been necessary to supply material
from local stock.

received,

CONTROL

The interlock circuit for the ARE
has been completed in 1ts elementary
form and integrated with the process
instrumentation and operational plan
so that i1t will be possible to begin
the interconnection wiring diagrams
for the various control components.
Detailing of the control rods and
control rod actuators is 90% complete.
Fabrication of the shim-rod absorber
elements for one rod to be used 1in
the critical assembly has been com-
pleted by the X-10 Metallurgy Division
(cf., sec. 12).

OPERATION

An operational procedure has been
outlined in detail¢?) and will be
only briefly reviewed here. The
operation of the experimental reactor
may be divided i1nto four separate
stages; pretesting, critical loading,
power operation, and shutdown.

Pretesting. After the reactor
plumbing has been installed and the
system 1s ready for checking, both
the primary and secondary circuits
will be given a pneumatic test. In
addition to eliminating any leaks to
atmosphere, 1t 1s also important that
any cross-talk between the primary
and the secondary circuits be elimi-
nated. With helium in the primary
circuit and vacuum in the secondary

12

circuit, 1t will be possible to
determine both the presence and extent
of the cross-talk between the two.
After both circuits have been made
tight so far as the above tests are
able to reveal, they will be filled
with NaK, which will be circulated at
near design flow rates. 1In addition
to giving a further leak test, the
circulation of the NaK will permit
checking of pumps, tanks, and level
controls under ambilent temperature
conditions. The NaK will also further
clean the system and permit easy
repair, and up to temperatures of
about 1200°F it will not deteriorate
the piping. Upon completion of this
phase of the pretest, the hot Nak
will be removed from the system by
evacuating the system and then evapo-
rating the residual NaK under vacuum.
Then, with the vacuum maintained, the
fuel carrier (fluoride melt without
UF,) will be released into the evacu-
ated fuel system.

Critical Loading. For the critical
loading of the reactor, a fuel 1in-
jection system, which will be filled
with highly concentrated UF, in the
in the fuel carrier, will be employed
to inject the critical mass 1nto the
fuel circuit. The injection system
is so arranged that samples of known
weight can be i1introduced into the
fuel system at a controlled rate.
The injection system is also equipped
with two weighing devices and a
television pickup to monitor the
injection of fuel i1nto the reactor
fuel circuit. The UF,-rich fuel will
be charged to the reactor in distinct
fuel quanta and in such a manner as to
be equally distributed throughout the
reactor fuel circuit. As a safety
precaution attempts will be made to
measure the temperature coefficient
of reactivity when the reactor has
reached the value of kK = ~0.98. To
do this the fuel temperature will be
lowered from the equilibrium value of
1200°F to approximately 1175°F. If
the reactivity increases with this
lowering of the temperature, it will
show that the temperature coefficient
is negative and that the loading to
criticality may safely continue. At
this time an attempt will alsc be
made to obtain a rough calibration of
the regulator rod.

Power Operation. Operation at
power will commence as quickly as
possible commensurate with safety.
After the control rod has been cali-
brated, the temperature of the reactor
will be raised to 1300°F and the
regulating rod withdrawn sufficiently
to put the reactor om a positive
period of about 20 to 30 seconds.
The reactor will then be at a power
level of about 10 kw. At this point
the helium blowers will be started
and heat extraction will have been
initiated. The mean fuel temperature
of 1300°F will be maintained closely,
but the fuel inlet temperature will
drop and the reactor outlet temperature
will rise until at full power there
will be a temperature drop across the
reactor of about 350°F. The servo

regulating system will not be emploved

unless 1t 1s found to be necessary
because of an insufficiently large
negative temperature coefficient.

Shutdown. After the power run has
been made the reactor will be scrammed.
When temperatures have reached equi-
librium and after 1odine has had time
(6 to 8 hr) to decay into xenon, an
attempt will be made to go critical
again. If xenon has evolved as a gas
from the fuel, the reactor should go
critical at essentially the same rod
conditions, since the xenon poison
will not be present. If, however,
the reactor will not go craitical with
shim rod adjustment, instead of adding
fuel, further time will be allowed to
elapse and another attempt to go
critical will be made. If this attempt
1s successful, it may be assumed that
the xenon has had time to decay and

FOR PERIOD ENDING JUNE 10, 1952

that the poison preventing criticality
in the first attempt was the xenon
that was not evolved from the fuel.
Nonuranium-bearing fluorides will
then be introduced into the system to
flush the fuel 1into the radioactive-
fuel dump tanks.

BUILDING

The building for the ARE was
released on June 6 to ORNL by the
construction contractor, the Nicholson
Company. figure 4 1llustrates the
building exterior, and Fig. 5 1s an
interior view showing the reactor
dump vank pit, and heat
exchanger rooums.

test pirt,

Electrical Power. The building
electrical system has been outlined.
Detailed drawings of the heater
circuits, main conduits, etc.
being prepared. Requisitions have
been released for most of the major
components such as motor-generator
sets, motors, and motor controls,
but few orders have been placed. The
delivery of these i1tems will tend to
be critical and special follow-up will
be required. '

are

Shielding by Concrete Pits (H. L. F.
Enlund, Physics Division). The sources
of radiation from the aircraft re-
actor experiment may be divided into
two categories: the reactor 1tself
and the components of the system
external to the reactor. Further,
the sources must be considered from a
full-power as well as a shutdown
condition. The shutdown condition 1s
important because of the necessity
for servicing. Since the reactor
core is the originating point, either
directly or indirectly, of all the
radiation in the form of active
particles, particular attention was
given to the determination of gamma-~
radiation sources, neutron leakage,
and activity in the circulating fuel.

13
ANP PROJECT QUARTERLY PROGRESS REPORT

UNCLASSIFIED
PHOTO {0Q79

 

 

UNCLASSIFIED
PHOTO 40083

 

Fig. 5. Interior View of the ARE Building.

14
The reactor physics group furnished
core- and reflector-composition and
neutron-escape data based on multigroup
calculations for the experimental
reactor.

The shielding effectiveness of the
walls of the reactor pit was calculated
on the basis of the leakage data. It
was recommended that the 7 1/3 ft of
wall shielding be constructed of 4 to
5 ft of barytes-concrete blocks at the
control pit wall and 6 ft at the side
wall adjacent to the control instrument
bay instead of being entirely of
ordinary concrete blocks. This will
give factors of safety of about 8
based on a tolerance of 7.5 mr/hr.
The roof over the reactor pit was
found to give attenuation sufficient
to reduce the dosage to below labo-
ratory tolerance. The end leakage
from the reactor was less than from
the sides because of the greater
thickness of the Inconel shell.

In the vicinity of the proposed
reactor control 1nstrumenhts the

FOR PERIOD ENDING JUNE 10, 1952

shielding provided by the Inconel
shell, and to a much lesser degree
by the thermal insulation blanket,
reduces the radiation leakage to the
following at a distance of 30 ¢m from
the boundary of the side reflector
of the reactor:

Gammas
2.5 Mev = 9 X lOlO/cmz'sec
7.0 Mev = 4.7 X 1010/cm2'sec
Neutrons

Thermal = 4.7 X 10'%/cm? *sec
250-kv thermal = 8.6 X 1011/cm2'sec
Fast (above 250 kv) = 4.5 X lOlo/cm2'sec.

The side wall of the heat exchanger
room 1s sufficient to reduce the
activity from the equipment to below
tolerance levels, that is, about 0.1
mr/hr. Calculations are being made
for determining activity within the
heat exchanger space so that working
tolerances can be predicted for
personnel servicing equipment.

 

3. EXPERIMENTAL REACTOR ENGINEERING

H. W.

The primary effort of the Experi-
mental Engineering Group 1s the
development of major components for
systems circulating high-temperature
fluids (1500 to 1800°F) that will
operate reliably over extended periods.
These components consist of pumps,
valves, liquid-level indicating and
control instruments, pressure sensing
instruments, and flow measuring
devices. Other than certain design
considerations to compensate for
thermal distortion and thermal ex-
pansion, the pump and valve problems

Savage, ANP Division

resolve themselves into the single
problem of developing a zero-leakage
seal for the pump shaft or valve stem.
Considerable progress has been made
and 1t 1s now possible to report that
satisfactory means of sealing against
molten fluoride mixtures at tempera-
tures up to 1500°F are available. Both
pumps and valves have been operated
successfully at 1500°F and experience
in the design, construction, and
operation of high-temperature fluoride
and liquid metal systems 1s continuing
to accumulate. Investigations are

15
ANP PROJECT QUARTERLY PROGRESS REPORT

under way to evaluate the several
sealing arrangements that have been
developed, and new designs are being
worked out, constructed, and tested
in order to i1mprove on present sealing
methods.

Increased emphasis has been placed
on i1nstrumentation for flow and
pressure measurement and liquid-level
indication and control in high-
temperature systems. This work has
resulted in instruments that accomplish
these functions with reasonable re-
liability; however, continued effort
is being made to improve both accuracy
and reliabilaity.

Tests are being continued to
determine the over-all heat transfer
from fluorides to liquid metals and
liquid metals to air. Thin-walled
tube-finned radiators are being
investigated in sodium-to-air heat
transfer experiments to evaluate the
several design parameters,

Mockups of the ARE core and fluid
circults are being made for studying
the fluid dynamics of the circuit and
for the development of techniques for
filling and emptying high-temperature
fluid systems such as the proposed ARE.

Progress has been made in the
technology of fluoride handling to the
point where the manufacturing and
handling of most fluoride fuels,
including the filling of systems, has
become routine. Fuels composed of
sodium, potassium, lithium, and
uranium fluorides are being produced
in sufficient quantities to meet the
experimental engineering requirements,
and equipment 1s almost ready for
producing the zirconium-bearing fluo-
rides in like quantities.

16

SEALS AND CLOSURES

Several satisfactory sealing ar-
rangements have been developed for
sealing rotating shafts and valve
stems at high temperatures. Frozen
seals have proved reliable for both
sodium and fluorides at temperatures
approaching 1500°F. Stuffing-box
seals packed with Inconel braid and
nickel and graphite powder have
successfully sealed against 1500°F
fluorides in both valves and pumps.
A program is under way to test the
effectiveness of bellows as a valve-
stem seal in intimate contact with

1500°F fluorides.

Seal Tester (P. G. Smith, ANP
Division). A seal-testing device has
been designed and is presently being
constructed that may be used for
testing either packed, stuffing-box
seals or frozen seals for 1500°F
fluoride systems. The eguipment
includes a bearing housing and pump
volute mounting from a CG-1 Worthite
pump that can be easily converted to
pumping service when no longer needed
for sealing tests. The sealing section
has been designed to incorporate a
Calrod heater, which has proved
essential 1n stopping and restarting
frozen-fluoride seals. Shafts hard-
faced with Stellite No. 3, Stellite
No. 6, and chromium plating will be
tested.

Stuffing-Box Seals (H. R. Johnson,
ANP Division). Stuffing-box seals
with both rotating and reciprocating
shafts have been operating for several
hundred hours. The shafts are made
of type-316 stainless steel with
chromium-plated surfaces. The stuffing
boxes are packed with Inconel braid
and nickel and graphite powder, and
ligquid tricresyl phosphate is added
as a lubricant. This type of seal
operates satisfactorily for loag
periods against 1500°F fluorides so
long as frequent additions of tricresyl
phosphate are made. However, both the
rotary and the reciprocating shafts
show tendencies to seize if not
frequently lubricated. This condition
1s being corrected in actual pump
applications by the addition of a
forced-lubrication system that will
continuously add tricresyl phosphate
to the packing materials at a low
rate.

Bellows Tests (P. W. Taylor, ANP
Division). Thin-walled bellows made
of type-347 stainless steel have been
tested in 1300°F fluorides for possible
use as a valve-stem seal. Initial
tests 1ndicate that this type of
bellows will not be satisfactory when
exposed to high-temperature fluorides.
The bellows tested ruptured at the
convolutions within 100 cycles at
1/4-in. deflection.

Inconel bellows with thicker walls
will be tested as soon as such bellows
can be obtained. A bellows tester 1is
presently being designed in which the
bellows will operate in a cooled,
trapped—gas volume to eliminate ex-
posure of the bellows material to
high-temperature fluorides.

Oval-Ring Seal (W. G. Cobb, ANP
Division). A gasketed joint for
sealing fluoride salts at 1500°F has
shown excellent results in operation.
This joint was made with a Goetz
oval-ring gasket, size R-49 (6 3/4-in.
pitch diameter, 5/16 in. wide), made
of type-316 stainless steel. The
7/8-in. -thick flange contained twelve
5/8 N.C. bolts, on an 8-in. diameter,
that were tightened to 100 ft-1b
torque (in 20 ft-1b steps) with lubri-
cated threads.

The joint was operated at tempera-
tures of 1200 vo 1500°F for approxi-
mately 960 hr and then coocled to

FOR PERIOD ENDING JUNE 10, 1952

room temperature and reheated several
times to further increase the severity
of the test,

PUMPS

Mechanical pumps that have been
successfully operated for pumping
1500°F fluorides have demonstrated a
high degree of reliability. Pumps
with a variety of shaft sealing
arrangements have been operated in the
range of 1200 to 1500°F, and sufficient
experience has been obtained to
indicate that pumping at these tempera-
tures 1s not extremely difficult when
the pumps are carefully constructed
from sound engineering designs.

ARE Centrifugal Pump (A, G. Grindell,
Engineering and Maintenance Division).
Most of the components for the first
ARE centrifugal pump (gas-sealed
Model DA) have been fabricated and the
pump 1s nearly ready for assembly. The
second pump, Model E, incorporating a
frozen-fluoride seal with a closed
cooling circuit, has been completely
designed and is being fabricated. It
is expected that pump components for
the assembly of three ARE pumps will
be available to begin pump tests by
early summer. The components for
associated fluid circuits are presently
being fabricated.

The first of these pumps will be
fabricated for initial tests from
type~-316 stainless steel as a matter
of expediency because of long delivery
dates for Inconel. The components
exposed to high-temperature fluorides
in the final ARE pump will be con-
structed of Inconel.

Laboratory-Size Maintained-Level
Gas-Sealed Pump (W. G. Cobb, ANP
Division). The gas-sealed pump
previously reported{!’ as operating

 

(I)Aircraft Nuclear Propulsion Project Quarterly
Progress Report for Period Ending March 10, 1952,
ORNIL.-1227, p. 27.

17
ANP PROJECT QUARTERLY PROGRESS REPORT

1n an 1sothermal circulating loop was
restarted after repairs tothe loop and
installation of anew type of throttling
valve. Operation was contlnuous at
temperatures of at least 1200°F for
a period of 314.5 hours. Termination
was caused by a leak in the loop near
the valve. The exact cause of failure
is yet to be determined. The final
100 hr of operation was at 1500°F,
Some leakage from the pump body
occurred during the latter stages of
operation but the leak was essentially
"self-healing'" and pump operation was
not affected., The total operating time
for the pump was 964.5 hours.

The pump has been redesigned to
substitute a rotary mechanical seal
for the stuffing-box seal and to
relocate the parting faces of the pump
to avoid exposure to high-temperature
fluids. Most of the parts for two
pumps of this model have been received
and one pump is assembled and awaiting
test,

Worthite Frozen-Sodium-Sealed
Pump. The frozen-sodium-sealed pump
previously reported(!) has now logged
over 3000 hr of operation ina figure-8
loop. The cooling arrangement for the
seal has been altered so that the seal
1s now cooled by water i1nstead of
refrigerated kerosene, and as a result
the pump can be operated at higher
temperatures. The sodium temperature
at the pump has been increased from
approximately 800 to 1200°F. It is
expected that the pump will operate
with the fluid temperature near
1500°F; however, with the present
heater and insulation limitations on
the figure-8 loop, this is not possible
without a shutdown for a major heater
overhaul, which will be effected 1n
the near future.

The sealing annulus on the pump was
remachined to give approximately
0.035 1n. of radial clearance between
shaft and sleeve, and a solid shaft

18

was substituted for the original
hollow shaft. Neither of these changes
seems to affect the ability to form a
frozen scal around the shaft, and as
a result the shaft clearance over a
fairly wide range 1s not thought to be
a critical parameter in the design of
such a seal. Heat removed from the seal
with the temperature of the stream at
1200°F is approximately 2500 Btu/hr.

DPurco Convection-Cooled Frozen-
Sodium-Sealed Pump (W. R. Huntley, ANP
Division). A Durco pump (Model H24
MBR654) has been designed to incorporate
a frozen-sodium seal that is convection
cooled. The pump and associated fluad
circuit equipment are being fabri-
cated by the Y-12 shops. The seal
consists of a copper-finned sleeve
approximately 5 1/4 in. long with
3 5/8-1n.-dia fins. Heat traunsfer
calculations indicate that this seal
will operate effectively against
1500°F sodium. The pump will deliver
approximately 20 gpm against a 20-ft
head.

Durco Frozen-Fluoride-Sealed Pump
(P. G. Smith, ANP Division). The
Durco centrifugal pump (Model H34
MDVX-80) previously reported(!’ was
operated for more than 500 hr 1in an
isothermal loop to pump 1100°F fluo-
rides., It was then modified to 1in-
corporate a frozen-fluoride seal that
was cooled by blowing air across a
finned sleeve constructed of type-316
stainless steel. Diametrical clearance
between shaft and sleeve was 0.060 inch.
The parting faces of the pump were
sealed with a nickel ring gasket.

The operation of this pump was
limited to 1100°F because of differ-
ential thermal expansion between the
shaft and pump housing that caused
interference between the impeller and
the housing when attempts were made to
operate at higher temperatures.
Al though the test was terminated
because of a slow leak at the parting
face of the pump, the seal was com-
pletely successful during the more
than 500 hr of operation.

Upon examination of the pump shaft
and the i1nternal surfaces of the
finned sleeve it was found that there
was only a very slight amount of
scoring of the shaft, The scoring
was not considered to be detrimental
to the operation of the pump, and 1t
probably can be eliminated completely
by coating the shaft with a harder
grade of Stellite than was used. No
difficulty was encountered with the
outboard pump bearings that were cooled
by continuously circulating the
lubricating oil through an air-cooled
radiator.

A new pump 1is being fabricated
that incorporates an improved design
of the frozen-fluoride seal. This
design employs a Calrod heater for
heating the portion of the seal in
which the fluorides are normally
frozen. The frozen seal is backed
up by a stuffing-box seal so that the
pump can be operated with either the
frozen seal or the packed seal.
Additional clearance has been provided
between the impeller and the pump
housing and it is expected that this
pump will operate to pump fluorides
at temperatures approaching 1500°F.

Fluoride Pumps with Stuffing-Box
Seals (W. R. Huntley, H. R. Johnson,
ANP Division). A small coolant pump
(manufactured by Eastern Industries,
Inc.), driven by al/15-hp motor, was
operated for approximately 300 hr to
pump 1500°F fluorides. The shaft had
a diameter of 1/8 in. through the seal,
which was the conventional stuffing-
box type packed with Inconel braid and
nickel and graphite powder. The shaft
operated at approximately 5000 rpm,
and the pumping rate was approximately
5 gpm. The test was terminated when
the 1/8-in. shaft broke at the outside
end of the packing gland. Since the

FOR PERIOD ENDING JUNE 10, 1952

test was run for about 300 hr before
failure, 1t i1s significant in that it
demonstrates the feasibility of using
commerical pumps without extensive
modification for pumping high-tempera-
ture fluids.

A larger pump (Durco Model H-24
MBR-654) that has been modified to
incorporate similar packing will be
placed in operation at an early date.
The modification consists of machining
a suction plate of type-316 stainless
steel to improve the sealing arrange-
ment of the parting faces and increasing
the depth of the packing gland to
2 1/2 inches. Means for constant-
pressure lubrication of the seal with
tricresyl phosphate have been added.
It is expected that this pump will
operate with fluorides at temperatures
approaching 1500°F.

VALVES
D. R. Ward P. G. Smith
ANP Division

Three types of valves that may be
needed for handling molten fluorides
at 1200 to 1800°F in ARE-type reactors
are the wide-open valve (open except
for occasional closing), the closed
valve (closed except for occasional
opening), and the throttling valve
(requiring frequent adjustment). Extra
valve-stem seals may be desirable as
a precaution against loss of fluid in
case the primary seals fail. Leakage
of molten fluorides into the cool
region of the bonnet would form a
"frozen seal' between the stem and the
bonnet. Such leakage must be avoided
because the frozen seal thus formed
would immobilize the valve stem and
probably make the valve inoperable.

Certain valve features and components
can best be studied on an individual
basis rather than as a group 1in-
corporated in the same valve. Such
studies include (1) selection of stem

19
ANP PROJECT QUARTERLY PROGRESS REPORT

packing and lubricants, (2) selection
of anti-galling lubricants for threaded
members 1n the high-temperature region,
and (3) determination of the effects
of differential expansion between
stem and bonnet during heating and
cooling., Tests are under way to
assist in these studies.

Wide-Open Valve. One design of
a wide-open valve utilizes back-seating
of the valve in the open position as
the primary stem seal, which 1s
followed by a stuffing-box seal. With
this arrangement the packing will not
be exposed to the fluid until the valve
is closed and then only for a short
time.

Closed Valve. A closed valve may
be obtained by modifying the design
of the wide-open valve described
above. If the direction of flow
through the valve is chosen properly,
the fluid will not contact the packing
until the valve is opened.

Throttling Valve. The designs of
throttling valves are being examined.
One design, which 1s essentially
complete, consists of a 1 1/2 I.P.S.
globe valve machined from Inconel bar
stock. A stem packing is used in the
high-temperature fluoride region,
a finned temperature-gradient section
of the bonnet leads up to a bellows
seal located in the cool region,

and

Another throttling valve under con-
sideration employs an Inconel braid
aud graphite and nickel powder packing
in the high-temperature region and a
convectional stem packing in the cool
region, with provision for introducing
inert gas into the region between the
two seals. By carefully controlling
the gas pressure between the seals,
the tendency for the molten fluoride
to escape through the high-temperature
packing will be balanced by the
opposing gas pressure against the
packing.

20

A third design of throttling valve,
the poppet type, employs a bellows
seal in the cool upper region of the
bonnet and no auxilliary stem seals in
the lower valve-body region. This
valve will rely on a cone of trapped
gas to prevent the molten fluorides
from leaving the high-temperature
region and reaching the bellows.
Obviously, this valve will not be
suitable for use in systems requiring
wide variations in pressure or filling
while evacuated, and it will operate
in only one position with respect to
the direction of gravity,

Canned-Rotor-Driven Valve (W, B.
McDonald, A. L. Southern, ANP Division).
Tests are under way to determine the
feasibility of using a canned-rotor
drive for valves in high-temperature-
fluid circuits. The proposed design
1s shown in Fig. 6. Preliminary tests
indicate that a rotor constructed of
a permanent magnet can be operated in
a cooled, trapped-gas space with the
actuating-field coi1l separated from
the rotor by a stainless steel or
Inconel can up to 1/16 in. thick,
Complete control of the actuation
of such a valve may be obtained by
using amultiple-position rotary switch
to rotate the poles of a three-phase
induction-wound motor. One complete
rotation of the switch causes the
rotor to follow at the same speed 1in
the direction the switch 1s rotating.
Bearings could possibly be eliminated
by mounting the rotor on a shaftg
having screw threads on one side of
the rotor and a guide sleeve on the
other and thus permitting the rotor
to travel axially in the magnetic

field.

HEAT EXCHANGERS

A heat exchanger with NaK i1n both
the primary and secondary circuits
operated at inlet temperatures up to
1500°F for over 3000 hr; the test was

voluntarily terminated. The agreement
FOR PERIOD ENDING JUNE 10, 1952

DWG. 15332

  
   

DIFFERENTIAL TRANSFORMER

PERMANENT MAGNET ROTOR

  
 

bDC SUPPLY
{ PHASE

 
 

STATOR - 3 PHASE

 
 
 

3 PHASE

   
   

ROTARY SWITCH

BALL BEARINGS

PACKING TO EXCLUDE VAPORS
FROM RCTOR CAN

LIQUID LEVEL AT
MAXIMUM SYSTEM PRESSURE
TRAPPED-GAS CHAMBER
(LARGE VOLUME IN COMPARISON TO

THE VOLUME ABOVE THE PACKING)

 

Fig. 6. Proposed Design for Canned-Rotor-Drivenm valve,

21
ANP PROJECT QUARTERLY PROGRESS REPORT

between the experimental heat transfer
data-awd*those calculated from Lyon’s
equation was poor at low flows but
good at high flows. The heat transfer
coefficient showed little change for
2000 hr and at completion of the
endurance run was within 12% of the
initial value. Pressure drops, both
through and around the tubes, showed
no particular trend with time.

Tests on avariety of heat-exchanger
tube~-matrix spacers showed that a
spacer friction factor, defined as the
ratio of the spacer head loss to the
incident-velocity head, was a unique
function of the Reynolds number in the
tube bundle. Data for the several
sets of spacers were correlated with
model parameters to obtailn a single
curve of friction factor against
Reynolds number.

The sodium-to-air radiator operated
for 330 hr at sodium inlet temperatures
up to 1600°F. The over-all heat
transfer coefficients were between
6 and 12 Btu/hr*ft?:°F, that is, about
20% below predicted values.

NaK-to-NaK Heat Exchanger (M. E.
LaVerne and A. P. Fraas, ANP Division).
The NaK-to-NaK heat exchanger previ-
ously reported¢?? has completed heat
transfer tests at i1nlet temperatures
from 800 to 1200°F and flows from 3.5
to 17 gpm and an endurance run of
2835 hr at an inlet temperature of
1500°F and a flow of 5.2 gpm. The
total operating time was 3006 hr and
the test was voluntarily terminated.

Results of the heat transfer tests
It 1s apparent
from the figure that a considerable
discrepancy exists between the results
reported here and the heat transfer
data computed from Lyon’s equation, ()

are shown in Fig. 7.

 

(2 ibid., p. 32.

(S)R. N. Lyon, Forced Convection Heat Transfer
Theory and Experiments withlLiquid Metals, ORNL-361,
p. 21 (Aug. 19, 1949).

22

particularly at low flows. However,
the convergence of the predicted and
experimental curves 1s such as to
indicate good agreement (within 10%)
for flows in excess of about 15 gpm.

An item of some concern at the time
the heat exchanger tests were begun
was the possibility of serious per-
formance deterioration with operating
time. It was noticed that changes
in the heat transfer coefficient were
negligible for the first 2000 hr - the
mean curve did not fall outside the
probable scatter banduntil 2200 hours.
The coefficient at completion of the
run was within 12% of that at the
beginning. Pressure-drop measurements
taken both through and around the tubes
during the endurance run displayed a
singular independence of operating
time. The observed scatter was of the
size expected with careful gage
readings and was as prone to be down
as up, with the result that there was
no discernible trend in pressure drop
with time.

A view of the inlets and exits of
the cut-up heat exchanger is shown 1in
Fig. 8; note the cleanness of the hot
end (operating at 1500°F) and the
extremely dirty appearance of the
"cool" end (operating at about 900°F).
These pieces are being examined
metallographically.

Tube-Matrix Spacer Tests (M. E.
LLaVerne and A. P. Fraas, ANPDivision).
Information on the flow characteristics
of spacers is of considerable 1interest
because the losses across spacers
of the type used 1n the NaK-to-NakK
heat exchanger may constltute an
appreciable portion of the total
losses, Since this information does
not appear to exist in the literature,
tests were run on a variety of spacers
with the expectation of obtalning a
general set of characteristics.

A model of the tubematrix comprising
a 10 by 10 bundle of 1/8-in.-da1a
FOR PERIOD ENDING JUNE 10, 1952

UNCLASSIFIED

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

o0 DWG. 15333
o
\ .
L 1T
S 90 o T~
': —~— - ] '
EIJJ Eg T %\fi\\
e = AR :
g 8 T T
o »
O
o ' :
20 ‘ CALCULATED USING
Nu=7+ 0.025 Pe©8
— D= EXPERIMENTAL. INLET
TEMPERATURE : 800 TO 1200°F

E 7000
ul
o
Lo ?
b L
S 6000 ——— &
& & 1 | 2T

£ ‘/
é ; 5000 e @
- - -7 @
e = Z/’//
W3 |~

4000 // _
-
z,é /
x 7°
= ' -
S 3000 .
2 4 6 8 10 12 14 16 18

No:K FLOW (gpm)

Fig. 7. Cooling Effectiveness and Over-all Heat Transfer Coefficient vs.
Flow in the NaK-to-NaK Heat Exchanger.

23
ANP PROJECT QUARTERLY PROGRESS REPORT

Y-i2 PHOTO
NO. 11305

 

Fig. 8. Inlet and Qutlet Headers of
the NaK-to-NaK Heat Exchanger.

stainless steel rods was built for
testing with air. The matrix was
clamped between top and bottom pieces
and sides made of wood. Pressure taps
along the top allowed determination of
pressure gradients along the model.
The head loss chargeable to the spacers
was taken as the difference in head
loss between a section containing a
set of spacers and a section of equal
length containing no spacers. The
spacer friction factor was then defined
as the ratio of the spacer head loss
to the 1ncident-velocity head. For a
given set of spacers, the friction
factor so defined was found to be a

24

unique function of the Reynolds number
in the tube bundle.

The data for the several sets of
spacers were correlated with the
spacer-thickness-to-width and spacer-
thickness-to-tube-diameter ratios as
parameters. Figure 9 gives the final
result; note that the two sections of
the curve have the same slopes (-1 and
~1/4 for the laminar and turbulent
regions, respectively) as the corre-
sponding curve for smooth pipe.

Sodium-to-Air Radiator (G. D.
Whitman, ANP Division). The sodium-
to-air radiator previously described(*?
operated for 330 hr before developing
a leak. During this time the average
temperature of the sodium entering the
radiator was 1200°F, including 140 hr
during which the i1nlet temperature was

between 1500 and 1600 °F.

The over-all heat transfer coef-
ficients obtained were between 6 and
12 Btu/hr+ft?2.°F. The maximum coef-
ficient was about 20% lower than the
predicted value. There was no marked
change in coefficient with increase
in fluid temperature, and there was
a slight decrease 1n the coefficient
with respect to time. Varying the
sodium flow between 1 and 3.5 gpm did
not affect the over-all coefficient
since the local sodium film 1s not
controlling. The maximum rate of
over-all heat transfer was approxi-
mately 100,000 Btu/hr. The air-
pressure drop across the radiator at
the maximum heat transfer rate was
11.8 in. of water.

The exact cause of the failure is
not yet known, but the radiator
section 1s now being examined metal-
lographically. One plausible ex-
planation of the failure is that excess
Nicrobraz at the joint between fin
and tube may have caused undercutting
or local reduction of tube wall

 

(4)op. cit., ORNL-1227, p. 31.
FOR PERIOD ENDING JUNE 10,

EWG. 15334

1952

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

20
T
w
. ° 0020 0056
@ 0032 0088
v ¢ 0032 0.157
v 0.05{ 0.139
&
10 B H—
9 % SPACER THICKNESS = 7
° g R SPACER WIDTH  =w
@ { TUBE DIAMETER = d;
o 7 BN
3 N
w3 6 Y -
= 8 "
o5 Y | @
A
aq \% ¢
WO~ a o % 0
g v 7. @
P e ez g v
3 ©
2
1 2 3 4 5 6 7 8 910 20 30

/VR(W/-)O.BOB (_f_)—O.BZO x 10—3

Fig. 9.

thickness during the brazing operation,
The failure resulted in a small
sodium fire in a tube at a point
between the top fin and a header
drum. This section of the tube was
on the sodium inlet side of the heat
exchanger and was not exposed to the
air stream. Aside from this failure
the radiator appeared to be 1in excellent
condition. There is evidence that the
leak started 40 hr before shutdown was
necessary. During this period the
system pressure at the pump discharge
was observed to fall from 15.9 to
14.9 psi,

The core section tested was a
continuous fin-and-tube two-pass
counterflow exchanger with 3/16-1in.-
OD tubes spaced 2/3 in. on a square

Modified Spacer Friction Factor vs.

ar
Modified Reynolds Number.

pitch. The 0.010-in.-thick fins were
spaced so that there were 10.5 per
inch of tube. The alr inlet face was
2 by 3 3/8 in. and the length in the
direction of air flow was 12 inches.
The air-side surface had a total area
of 12.78 ft?®, Heat transfer data were
taken at sodium inlet temperatures
between 700 and 1600°F and flow rates
between 1 and 3.5 gpm. The air flow
was varied between 0.06 and 0.16 1b/sec,
which gave a maximum Reynolds number
of 12,000, The maximum alr tempera-
ture rise was from 80 to 1100°F,

Additional tests are planned 1in
which the core element to be used will
have 16 to 18 fins per inch and will
be coated with a thin layer of ceramic
to reduce oxidation.

25
ANP PROJECT QUARTERLY PROGRESS REPORT

gluoride-to-Liquid-Metal Heat
Transfer System. A heat transfer
system has been designed and 1s under
construction that will transfer heat
from a fluoride to a liquid metal to
air. This system will be tested to
determine endurance, heat transfer
coefficients of fluorides, and the
effect of corrosion, 1f any, on heat
transfer performance and pressure
drop.

The system, which 1s being fabri-
cated from Inconel pipe or tubing,
has electric tube-furnace elements
for use as heaters and a double-pipe
intermediate heat exchanger. A
centrifugal gas-sealed pump will be
used for circulating the fluorides,
and an electromagnetic pump will
circulate the liquid metal. Fabri-
cation of the system 1s approximately
30% complete. Testing is planned over
a range of Reynolds numbers from
laminar to approximately 80,000 at a
maximum fluoride temperature of 1500°F.

INSTRUMENTATION

Pressure Measurement (P. W, Taylor,
ANP Division). Several techniques
are being tested in order to develop
a reliable method of measuring pressures
of high-temperature (1500°F) fluoride
systems. These techniques include a
pressure-sensitive diaphragm, two
null-balance systems and the use of
the Pourdon tube gage.

Bourdon Tube Gage. The PBourdon
tube gage consists of a pressure-pot
facsimile to which gages are attached
through stainless steel and copper
tubing. Experiments with this type
of gage should determine the cause of
gas-line plugging and also the re-
Jiability of this type of pressure
indicator. The fluid will be pumped
to the pressure pot at 1500°F, but the
gage will be at such a distance from

26

the pot that it can be maintained at
room temperature.

Pressure-Sensitive Diaphragm. A
pressure-measuring lnstrument con-
sisting of a pressure-sensitive.
Inconel diaphragm, 13 by 1/16 in.,
is being tested, The deflection of
the diaphragm is a linear function of
the pressure exerted upon it. At
60 psi and 1000°F the diaphragm should
deflect approximately 10 mils., The
deflection is sensed by a microformer
transformer that operates a servo dial-
indicating device. The accuracy of
this type of instrument 1s limited
by the temperature variation of the
sensing element; therefore, the
instrument must be maintained at the
temperature at which i1t is calibrated.

Nullmatic., A commercial instrument,
the Moore Nullmatic 1:1 pressure
transmitter 1is being tested. This
pneumatic instrument contains a stain-
less steel bellows and utilizes a
null-balance principle that should
eliminate all temperature effects.
However, since this device 1s not
rated for high-temperature service,
the purpose of the current test 1s to
determine 1ts performance at high
temperatures when subjected to corrosive
vapors.

G-E Null-Balance Device. A null-
balance transmitter with a corrugated
diaphragm as the pressure-sensitive
element has been built and tested by
the General Electric Company at 1000°F
with good results. This instrument
is now being tested to determine 1ts
performance in the 1500°F range.

Liquid-Level Indicators (A. L.
Southern, ANP Division). Several
different types of liquid-level
indicators are being developed for
use 1in high-temperature liquid sys-
tems. In addition to level indication,
the indicators can also be adapted for
level control.
Variaeble-Inductance Level Indi-
cators, The variable-inductance
level indicator consists of a lami-
nated, tapered core enclosed in an
Inconel tube. The pick-up coil 1s
wound on a form 1 in. 1n diameter and
1 in. long. The design of the coupling
rod between the float and the core
is such that the response of the
indicator is linear for only 10 in.
of the 18-1in, level change for which
it was designed. The accuracy of thas
indicator is approximately 0,735%,
An endurance run of 1500 hr gave
no trouble, nor was any change en-
countered in the operation of the
indicator. Level can be controlled
with this indicator to 0.5% of the
total indication by using limit
switches on the indicating millivelt
meter. In addition, by using two of
these indicators together a null-
balance system can be obtained that
will hold one level even with another
one.

Bubbler Type of Level Indicator,
A working model of the bubbler-type
of indicator described in the previous
quarterly report(s) has been designed
and tested to operate over a level
range of 18 inches.

Two tanks were constructed with a
transfer line between them to give
the required level change. The fluo-
ride, NaF-KF-LiF (composition in mole
%:11.5-42.0-46.5), was alternately
pushed from one tank to the other by
gas pressure and the rate of level
change could be varied. A well-type
manometer was used to get a constant
zero and then only one reading was
needed from the manometer. (No. 3
Merian manometer fluid was used in
the manometer.) A Nullmatic pressure
regulator was used to hold a constant
pressure over the ligquid, and the
temperature was varied between 1000

 

 

() rbid., p. 34.

FOR PERIOD ENDING JUNE 10, 1952

and 1200°F, The gas flow through the
dip tube was 1indicated by a Fisher
and Porter flowrator. Level control
was accomplished by a "Thermocap"
capacitance control that worked off of
the fluid in the manometer. The
level in the tank was checked by means
of a variable-inductance pick-up coil
in which inductance was varied by an
iron core mounted on a float,.

The accuracy of this type of level
indicator depends on the difference
in specific gravity of the manometer
fluid and the liguid. In this expera-
ment the ratio was 1:3 and the accuracy
of the level was determined to 0.04
inch., The experiment has operated for
140 hr under dynamic conditions, with
over 1000 cyecles of operation, and
742 hr under static conditions. As
a result of tests with this model, 1t
is concluded that the level in a
1500°F fluid system can be reifotely
indicated and controlled by this
technique.

Resonant-Cavity Type of Ligquid-
Level Indicator, A liquid-level
indicator that uses the principle of
a resonant cavity has been designed
and will be tested as soon as the
commercial equipment now on order 1is
delivered. The frequency used will
be in the S-band microwave region,
and all S-band microwave components
will be used. The control signal
be the energy absorbed by the
cavity. This signal will operate
a servo system which will 1n turn
contral the frequency of the Klystron
tube. Only simulated experiments
have been made with this indicator
and 1t will be between three and six
months before the actual tests are
made. This level indicator 1s well
suited for use 1n a reactor system
because all parts that will be within
the reactor shield can be made of metal
and thereby eliminate the danger of a
leak or radiation damage.

will

27
ANP PROJECT QUARTERLY PROGRESS REPORT

FLUID DYNAMICS OF THE ARE

L. A, Mann D. F. Salmon
ANP Division

Since criticality of the ARE reactor
depends, among other considerations,
upon the amount of fuel in the core of
the reactor and since the present
design of the core presents opportuni-
ties for entrapment of gas or vapor,
a full-size hydraulic model of the
fuel passages 1s being built (Fig. 10)
of transparent materials for obser-
vation of fluid-flow patterns, gas-
trapping characteristics, and for
testing filling, draining, and emptying
of the system.

A simple system that is similar
to the fuel system of the ARE core

FUEL INLEY
HEADER

 

FUEL TUBES

fiydraulic Model of the ARE.

28

 

has already been built and tested for
filling and emptying. The test system
consisted of six parallel sets of
six loops; each loop had 1/2-in. Tygon
tubing between the inlet and discharge
headers and an appropriate pump,
flowmeter, and tanks external to the
headers. Tests were made with (1) water
and air and (2) 56% zinc chloride 1in
water (sp. gr. = 1.65) and kerosene
(sp. gr. = 0,82), and the hydraulic
similarity to the reactor was based
on the Reynolds number. Test results
were as follows:

1. With a gradual increase in the
pumping rate, 17 gpm of water was
required to completely fill a system
originally filled with air.

2. With a constant pumplng rate,
12 gpm of water was required to fill
the system from all air to all water.

3. With a gradually increasing
pumping rate, 7 gpm of zinc chloride
solution was required to completely
replace the kerosene.

4, With a gradually increasing
pumping rate, 3 gpm of kerosene was
required to completely replace the
zinc chloride solution,

5. 15-psia air failed to remove
the water, kerosene, or zinc chloride
solution from a full system. It did,
however, remove the liquid from any
one circuit fairly completely 1f the
other five circuits were closed,

6, With a gradual increase of
pumping rate or gas pressure each of
the above removals was stepwise in
character., That is, after replacing
the fluid in one circuit, an increase
in pumping rate was required to replace
the fluid in another of the remaining
parallel circuits. The 1ncreases
required were approximately equal
increments for each additional circuit
cleared.
7. At the design flow rate for the
reactor all bubbles were swept through
the tubes, but at considerably lower
flow rates bubhles were observed to
cling to the 1nside radii of the
bottom bends. It was noticed that a
large bubble would sweep out a col-
lection of smaller bubbles.

TECHNOLOGY OF FLUORIDE HANDBLING

Fluoride Production (J, C.
ANP Division; F. F. Blankenship,
Materials Chemistry Division). The
need for large-—scale testing of
corrosion by fluoride fuels 1ia pump
loop and heat exchanger assemblies
has necessitated the preparation of
the pure fuelsinup to 100-1b batches,
The equipment for effecting hydro-
genation and hydrofluorination 1s
being installed in Building 9928, which

has been modified to receive 1t.

White,

The fluoride-production process
involves six steps {(cf., sec, 10 gives
the specific process conditions):

1. Charge the reaction vessel with
an intimate mixture of the components.

2, Treat the charge with HF and
with H, at temperatures below the
melting point.

3, Melt the charge under an inert
atmosphere or one of the treating
gases,

4, Treat the melt at temperatures
in the range of 1500 to 1800°F with
HF and H,.

5, Strip the treating gases from
the melt with inert gas.,

6. Filter into areceiver. Control
of batech temperature and atmospheric
pressure is available at all steps.

Mechanical agitation 1s available in

steps 3, 4, and 5. Treating and

FOR PERIOD ENDING JUNE 10, 1952

stripping gases are introduced through
a perforated bubbler. An electrically
driven stirrer will be used to mix the
material during passage of the gaseous
reagents.

The equipment will be able to
produce from 5 to 100 1lb of mixture in
a sealed, and easy to sample,
as needed, and should be in operation
by June 15, At that time sufficient
zirconium fluoride should be available
for large-scale productioun.

recelver

Fluoride Removal (D. R, Ward, ANP
Division). Tests were made to deter-
mine the best practical means for
removing the fluoride salts afrer
completing the operation of test
equipment. Specimens were prepared
by cutting rings containing solidified
Fulinak from the Inconel tubing used
for a previous test. The specimens
were then soaked 1n various chemicals
and the degree or ease of fluoride
removal was noted. No solvent for the
fluorides was found. Running water
appeared to be the best practical
fluoride softener, although up to
48 hr was required for softening in
some cases., Also, mechanical assistance
was usually needed., Removal of most
of the fluorides while they are molten
will simplify the final cleaning
problem.

Pretreatment of Pipeline Helium
(L. A, Mann, ANP Division). Pretreat-
ment of the bulk helium supply will be
accomplished by using the following
cquipment (in series as listed)
(1) dry-ice cold trap, (2) NakK
bubbler, (3) molten~fluoride bubbler,
(4) water-cooled cold trap. Item 3
was added to previous pretreatments
so that the gas will have at least a
short-time contact with the type of
fluoride material it 1s later to
protect, JItem 4 was added to remove
gross, condensible vapors., The equip-
ment 1s being constructed,

29
ANP PROJECT QUARTERLY PROGRESS REPORT

Pickling Procedures for Fluoride
Containers (D, C. Vreeland, E. E.
Ho ffman, R. B, Day, L. D, Dyer,
Metallurgy Division),., Some experiments
in which various pickling procedures
were used on type-321, -304 and -446
stainless steels and Inconel have
been carried out to determine the
efficiency of the pickling procedures
on these types of alloys. Three-inch
lengths of 1/2-1in. tubing were welded
at one end to subject them to the
oxidizing conditions that occur 1in
tube preparation prior to static
corrosion testing, After welding,
the tubes were cut in half, longi-
tudinally, to facilitate observation
of the interior surface of the tube
during and after pickling. Some tests
were conducted on 1 by 1 by 1/4-in,
specimens that had been oxidized in an
air atmosphere furnace at various
temperatures to test the action of
the pickling solutions on oxide
coatings that were heavier than could
be obtained on the welded tubing.
Before pickling, the specimens were
degreased by immersing in a 10%
solution of sodium phosphate at 93°C
for 15 min, and then rinsed in warm
water.

The pickling treatments tried are
listed below, Treatments that were
most successful in these tests were
3 (b) for types—321 and -304 stainless
steel and 4 for type-446 stainless
steel or Inconel. Treatment 4 for
type—-446 stainless steel and Inconel
was effective only for light-temper-
colored oxide films. Treatment 3 (b)
for types-304 and -321 stainless steel
appeared to be helpful on all types
of oxides tested, but 1t was most
effective on the thinner coatings.

30

1.

(¢) Immerse in Virgo for 15 main
at 615°C and rinse in cold water.
Immerse in 10% H,S50,, using thio-
urea as inhibitor, for 30 min at
65°C and rinse 1in hot water, cold
water, and again in hot water.
Immerse in a solution of 10%
HNO, plus 1 1/2% hydrofluoric
acid for 10 min at 60°C and rinse
in cold and hot water. Immerse
in 20% HNO,; for 1/2 hr at 60°C
and rinse in cold and hot water,
and dry.

(b) Substitute Hibitite for thio-
urea in 1 (a),

(a) Immerse in Virgo for 15 min
at 615°C and rinse in cold water.
Immerse in 10% H,80,, using thio-
urea as inhibitor, for 30 min at
65°C and rinse in hot water, cold

‘water, and again in hot water.

(b) Substitute Hibitite for thio-
urea in 2 (a),

(a) Immerse in 10% H,S0,, using
thiourea as inhibitor, for 30 min
at 65°C and rinse 1in hot water,
cold water, and again in hot water.
Immerse in a solution of 10% HNO,
plus 1 1/2% HF for 10 min at
60°C and rinse i1n cold and hot
water, Immerse in 20% HNO; for
1/2 hr at 60°C and rinse in cold

and hot water.

(b) Substitute Hibitite for
thiourea in 3 {(a).

Immerse in a solution of 10% HNO,
and 4 to 5% HF for 30 min at 30°C

and rinse i1n cold water.
FOR PERIOD ENDING JUNE 10, 1952

4. REACTOR PHYSICS

W. K. Ergen, ANP Division

The belief that the circulation of
the fuel causes damping of power
oscillations in all practical cases
has been strengthened further by
theoretical considerations. Further-
more, specific examples have been
computed on the differential analyzer
at the University of Pennsylvania, and
some of the examples show the damping
to be extremely powerful. The damping
will counteract any antidamping
tendencies that the circulating-fuel

reactor, like any other high-powered

reactor, might have for various
reasons.
The critical-mass and power-

distribution calculations for circu-
Jating-fuel aircraft reactors, although
incomplete, have been directed toward
determining how far it is possible to
go ~ in the interest of ease of fabri-
cation and mechanical ruggedness -~ in
increasing the fuel tube size and
decreasing the number of tubes. In
fact, the limiting case of a homo-
gengous fluoride reactor with all the
beryllium oxide-moderator in the
reflector was computed (cf., "Re-
flector-Moderated Circulating-Fuel
Reactor,"™ sec. 1). The critical mass
in this case is not far different from
the values obtained with multitube
reactors -~ a very gratifying result.
A possible difficulty may be a positive
temperature coefficient due to the
large number of fissions caused by
high-energy neutrons.

Work on the aircraft reactor experi-
ment was directed towards important,
but more or less minor, design questions.
In addition, the large number of ANP
Physics Division memorandums on
computational techniques and various
reactor designs are being reviewed with
the intention of assembling the in-
formation into systematic reports. A

glossary of reactor terms 1s also

being prepared.

GSCILLATIONS IN THE CIRCULATING-FUEL
AIRCRAFT REACTOR

S, Tamor, ANP Division

The kinetie equations of a somewhat
idealized circulating-fuel reactor,
given in the last ANP quarterly, (1)
were further investigated. At the
beginning of the reporting period 1t
was known that the power oscillations
are always damped if they are small to
begin with, and it seemed plausible that
even large oscillations are never
antidamped,(2+3)

Solutions of the kinetic equations
for representative values of the
parameters and initial conditions have
been obtained by Professor Cornelius
Weygandt and his staff on the dif-
ferential enalyzer of the Univerasity
of Pennsylvania. The data are not
expected to arrive at Oak Ridge before
the and of the present reporting
period, buta preliminary communication
confirms that none of the solutions is
antidamped; and, furthermore, under
some conditions, the damping due to
the circulation of the fuel 1s so
strong that only an aperiodic motion
without any "ringing" occurs,

As to the general theory of these
equations, it was shown that the power

always stays hbetween an upper and a

 

(1)Aircraft Nuclear Propulsion Project
Quarterly Progress Report for Period Ending
March 10, 1952, ORNL-1227, p. 41.

(2)F. G. Probammer, Note on the lLinear
Kinetics of the ANP {irculating-Fuel Reactors,
Y-F10-99 (April 22, 1952).

(3)W. K. Ergen, Physics
Circulating-Fuel Reactors,
1952).

Constiderations of

Y-F10-98 (April 16,

31
ANP PROJECT QUARTERLY PROGRESS REPORT

lower bound determined by the parameters
of the equation. In other words, there
are certainly no oscillations with
amplitude that increase to infinity,
The practical value of this statement
is somewhat impaired by the fact that
the bounds appear to be serious over-

estimates of the amplitudes, and while

the oscillations themselves may stay
within acceptable limits, if the
bounds were reached,1t might mean
destruction of the reactor. It was
shown that no periodic oscillation
other than the ones with period &/n
can persist,

CALCULATIONS FOR THE
CIRCULATING-FUEL ARE

C. B. Mills, ANP Division

The following results were obtained
fromcalculations assuming NaF-BeF,-UF,
fuel 1n the circulating-fuel ARE, but
the results are expected to hold for
the recently proposed ZrF,-bearing
fuels:

1. Tf the outermost 12 or 24 fuel
tubes in the core were eliminated,
leaving everything else constant, the
number of tubes could be reduced from
the originally contemplated 78 to 66
or 54. This would mean a decrease in
reactor size, but the reflector size
would i1ncrease. The net effect on
total uranium investment would amount
to an increase of only a few per cent
for each of the two groups of 12 tubes
eliminated. Hence, from the uranium
investment viewpoint, there 1is no
objection to some reduction of the
number of tubes. (%)

2. The fission heating of the ARE
and ANP reactors was investigated.(3)

(4)C. B. Mills, Optimization of Core Size for
the Circulating-Fuel ARE Reactor, Y-F10-96
(March 28, 1952).

(S)C. B, Mills, Fission Heating of Moderator
and Fuel-Coolant of the ANP and ARE Reactors,
Y-F10-90 (April 7, 1952).

 

32

3. An engincering simplification
in the reactor could be obtained, 11
the inert-salt coolant were eliminated.
If the fuel-coolant were circulated,
for cooling purposes, through the
reflector, 6% of the fissions would
occur in the reflector, and the neutron
leakage would increase and shift
somewhat toward higher energies.(8)

From calculations assuming the most
recently proposed ZrF,-bearing fuels,
the following results were obtained.(7)

1. Two fuels were compared; the
inert part (fuel minus UF,) of one
had the following composition {(in
mole %): 43% ZrF,, 37% NaF, 20% KF,
and the inert part of the other was
43% ZrF,, 5% NaF, 52% KF. The fuel
with the high KF content gave a
critical mass only 10% greater than
that of the low KF content fuel 1in
spite of the large K absorption cross
section.

2. The 2-in. Inconel layer around
the reactor does not act appreciably
as a reflector, since the neutron
spectrum is rather low in energy.

3. For the 20% KF content fuel,
the "best guess” for the total uranium
requirement is 124 Ib of 93.4% assay
U235, This includes the uranium in
the reactor and the external circuit,
with allowance being made for kef =
1.034,¢8) additional uranium for
control, and uranium in tube bends at
the ends of the core.

4, The important reactivity co-
efficients

(O/E)/ (Bu/m)

(6)C. B. Mills, ARE with Fuel-Coolant in the
Re flector, Y-F10-91 (Feb, 27, 1952).

(7)C. B. Mills, Statics of the ARE Reactor,
Summary Report, Y-F10-103 (May 8, 1952).

8)op. cit., DRNL-1227, p. 50.
(Ak/k)/(Dn/m)

ocderator

are 0.4 and 0.5, respectively - the
same as quoted in the last quarterly,(®’

5. About two-thirds of the fissions
arec thermal.

6. The approximate distribution of
the neutron absorption over the reactor
constituents in the reactor 1s given
in Table 1,

TABLE 1

Neutron Abserptions in the Constituents
of the Core of the ARE Reactor

 

 

 

CORE: NUMBER OF NEUTRONS NORMALIZED
consTiTuenT | TO 1 FISSION (2.5 neutrons)
PER CUBIC CENTIMETER OF CCRE

Uranium 1.113 o
Zirconium 9.9 x 10°3

FPotassium 0.171

Sodium 2.9 x 103

Fluorine 7.0 x 1073

Beryllium 2.49 x 1072

Inconel 0. 409

Oxygen 0

 

 

7. When the
experiment 15 completed,
coolant will be dumped into & tank
cooled by water in 2-in. tubing.
Assuming 100 1lb of U235 in the eqguiva-
lent of a 1.8-ft-radius sphere and
considering the self-shielding of the
uranium as well as of the water, the
tank will be suberitical. Furthermore,
boron will be added to the cooling
water as a safety factor,

aircraft reactor
the fuel-

 

(g)Ibid., p. ol.

FOR PERIOD ENDING JUNE 10, 1952

8. The loading of the ARE will be
accomplished by slowly adding a rich
solution of uranium to the hot, circu-
lating coolant. An estimate of the
criticality of 150 1b of U235 i5n
1.3 ft? of rich fuel solution gave a
value for kcff of 0.45, which is far

subcritical.

METHODS FOR REACTOR COMPUTATIGN
B. T. Macauley C. B. Mills
ANP Division

The Reactor Physics group 1s
attempting to expand its computational
techniques to cover reactors with
hydrogen in the core and/or reflector.
No entirely satisfactory method to
take into account the correlation
between scattering angle and energy
loss has been found, and other un-
solved problems exist., Nevertheless,
correspondence between calculations
and the few available experiments 1s
guite satisfactory, at least as far as
the critical mass is concerned. No
experimental data are available
regarding power distribution and
neuntron spectrum, and this, together
with the above objections on theoretical
grounds, prevents reporting these
attempts for general circulation at
present,

In order to simplify the survey of
variations of the ARE and similar
reactors, the results of the two-group
equations for the critical mass have
heen presented,(1?) with parameters of
the equations given numerically for
a beryllium oxide moderator. The
equations prove guite valuable, since
they agree within a few per cent with

the results of the machine calculations.

 

(IU)CE B, Mills, A Simple Criticality Relation
for Be Modzrated Intermediate Reactors, Y-F10-93
{Morch 10, 1952).

33
ANP PROJECT QUARTERLY PROGRESS REPORT

3. CRITICAL EXPERIMENTS

A. D. Callihan,

The investigation of the direct-
cycle aircraft reactor has continued
during this guarter with the use of
the critical assembly described in the
preceding report, Fission-rate and
neutron-flux measurements have been
made and further changes in reactivity
produced by reflector modifications
have been observed. A significant
rearrangement of the components that
was intended to reduce the neutron
channeling decreased the reactivity to
a degree requiring additional end
reflector to retain a critical system.
This new assembly has a temperature
response similar to that of the
earlier one.

The components of the low-temperature
mockup of the ABE are being assembled
and the fuel composition has been
modified to conform closely with the
present ARE fuel design. Zirconium
oxide is to be substituted for zirconium
fluoride since the latter is unavailable
at this time.

A resume of the results from an
earlier series of experiments with
a graphite-moderated reactor 1is
included,

DIRECT-CYCLE REACTOR
E. V. Haake D. V. P, Williams
Physics Division
R. C. Keen
Louisiana State University

W. G. Kennedy
Pratt and Whitney Aircraft Division

Dunlap Scott
ANP Division

The results reported here were
obtained in a program carried on by

34

Physics Division

Oak Ridge National laboratory at the
request of the General Electric Company
to provide design data for the water-
moderated, air-cooled aircraft reactor.
Two arrangements of the reactor loading
have thus far been made: one, now
referred to as the "shelf type,” has
been described completely elsewhere;(1)
and the second, called the "box type,"
will be described here. Data showing
the dependence of reactivity on various
composite reflectors have been reported
elsewhere, (2)

Shel f-Type Reactor Assembly. Studies
have been made of the power distri-
bution, neutron-flux distribution, and
self-shielding of a shelf type of
reactor assembly.

Power Distribution. A gross power
distribution throughout the core of the
reactor assembly shown in the pre-
ceding report(!) has been measured by
the usual "catcher-foil" wethod in
which aluminum placed in contact with
the metallic uraniumcollects recoiling
fission fragments, The resulting
activity 1s a measure of the fission
rate, This distribution, along the
axis of the reactor, i1s shown in
Fig. 11 in which the relative activity
is plotted against the distance from
the mid-plane. The distribution 1is
also shown in a direction parallel to
the axis at three locations on the
periphery of the core. Some fission
rate data were also obtained by count-
ing the induced activity in the urarnium
and the results are plotted. It is to
be noted that the two methods are
consistent,

Aircraft Nuclear Propulsion Project Quarterly
Progress Report for Period Ending Harch 10, 1952,
ORNIL-1227, p. 59-60.

A. D. Callihan, Preliminary Direct Cycle
Reactor Assembly, Part I, Y-B23-1 (Feb. 26,
1952); and Pgrt II, Y-B23-2 (May 21, 1952).
RELATIVE ACTIVITY {(NORMALIZED)

O ALUMINUM FOIL DATA
O CELL U=-i2

A CELL M-4
O CELL R-7

0 3 6 o

CELL M-4 AND U-{2
Lo
LL R-

 

FOR PERIOD ENDING JUNE 10, 1952

DWG. 15335

CELL M-1{2

GRAPHITE REFLECTOR

® URANIUM DISK DATA }CELL M-12

15 18 21 24

INCHES FRCM REACTOR MID-PLANE

Fig. 11.
End Reflector,

Neutron-Flux Distribution. The
neutron-flux distributions along a
radius vertically below the axis in
the mid-plane of the reactor and
extending into the beryllium reflector
are plotted in Fig. 12. The results
are from the activation of indium
foils, both bare and covered by 20
mils of cadmium, The difference
between the two experimental curves
shows the distribution of neutrons

Power pDistributions Extending from Reactor Mid-plane to Graphite

with energies below the cadmium cut-
off at about 0.5 ev.

Self-shielding. The uranium fuel
pieces are 0,01 in. thick, and a
measurement has been made of the
shielding imposed by the uranium on
fission neutrons. It was possible to
substitute five disks, each 0.002 in.
thick, for one of the 0.01-in.-thick
fuel pieces and to interpose aluminum

35
ANP PROJECT QUARTERLY PROGRESS REPORT

RELATIVE ACTIVITY

0 3 & 9 12 15

 

Dw!!. 15336

L.l
BERYLLIUM|
REFLECTOR

BARE INDIUM AGTIVATION 1
s

At
QSINE CURVE

‘7.

CADMIUM-COVERED INDIUM AGTIVATION

18 21 o4 27 30 33

iINCHES FROM AXIS OF REAGTOR

Fig. 12.
through M-22).

foil between adjacent disks. The
resulting fission fragment activities
as measured by the aluminum foils show
the variation of fission rate through
the fuel and indicate an average
fission rate of 88% of that on the
surface, As a measure of the energy
of the neutrons producing fission, the
fission rate of a fuel disk was
measured with and without a 20-mil
cadmium cover. It was shown that about

90% of the fissions are caused by
neutrons with energies below the
cadmium cut-of £,

Box-Type Reactor Assembly. In

another design of the direct-cycle
reactor; the alternate moderator and
fuel-coolant channels are annular. In
order to more nearly simulate this
design in the mockup discussed here,

36

Gross Indium Traverse Vertically Down in Plane of Interface (M-12

some of the elements(3) consisting of
plastic, stainless steel, and uranium
were rotated through 90 deg with the
intent of reducing the channeling in
the horizontal fuel layers of the
first assembly. The final loading
arrangement of this model 1s shown 1in
Fig. 13. The components are the same
as in the shelf type with two signifi-
cant exceptions: the plastic was
omitted from the central element, M-12,
to give better symmetry, and an
additional 2 in. of graphite reflector
(a total of 8 in.) were required on
one end to make the system critical.
It has not yet been possible to measure

(B)It is recalled that a fuel element is a
rectangular parallelepiped, 2 7/8 by 36 in., with
the long dimension horizontal. In cross section
it consists of plastic, 1 by 2 7/8 in., and an
open structure, 1 7/8 by 2 7/8 in., containing
thin stainless steel sheets, uranium metal, and
voids,

 
DWG. 15337

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

W N b WM

 

 

 

 

 

 

 

 

 

 

 

 

 

 

1 BERYLLIUM JAGKET REFLEGTOR, 36 in. LONG
FUEL ELEMENT, 36 in LONG

STAINLESS STEEL, URANIUM, AIR SPACE, 1% in. THICK
§ PLASTIC, fin THICK

 

Fig. 13. Loading Diagram.

the net loss in reactivity incurred by
the fuel element rotation, but it is
estimated at one dollar, that 1is, a
change in reactivity of the order of
the delay fraction of fission neutrons.

Control Rod Calibration. All
control rods have been recalibrated
and found to be comparable i1in value to
those of the shelf-type reactor except
the one at the center, which, as
pointed out above, does not contain
plastic. Its removal corresponds to a
decrease in reactivaity of 29 cents,
whereas each of the other rods corre-
sponds teo about 17 cents. This dif-
ference is due to the variation 1in
neutron distribution caused by the
change 1in plastic distribution. It 1s
interesting to note that the cali-
brations of the rods are unaffected by
a transverse gap up to 0.8 in. thick at
the reactor mid-plane.

FOR PERIOD ENDING JUNE 10, 1952

Temperature Effects. The dependence
of reactivity on reactor temperature,
referred to in Fig, 19 of the previous
report, (}) was again examined and found
to be greater for the present assembly.
The current value is 1.46 ¢/°F compared
with 1.25 ¢/°F in the shelf-type
reactor,

Variation of Reactivity with Plastic
Thickness. Also of interest was an
experiment in which the variation 1in
reactivity with plastic-layer thickness
was measured. The thickness of the
plastic 1n cells designated by column
K and rows 1 through 16 in Fig, 13 was
varied from 0.5 to 1.25 1n. and con-
comitant changes in reactivity were
noted. The void resulting from the
removal of the plastic was not filled
and in the last test one stainless
steel sheet was removed, It was
observed that the reactivity was a
maximum with the plastic 7/8 in. thick
and that it fell sharply with thickness
changes in both directions. A reduction
to 0.5 in., and an increase to 1.25 in.
decreased the reactivity about 10 cents,

Variation of Reactivity with
Separation at Reactor Mid-Plane. The
construction of the assembly 1in two
parts, one of which 1s movable by
remote control, permitted separating
the core along 1ts wid-plane 1n an
axial direction. The accompanying
changes 1n reactivity have been
measured and the results are shown in
Fig. 14 in which the separation of the
parts is plotted against the cor-
responding reactivity change. The
accuracy of the separation distances
is no greater than £0.02 inch. The
purpose of the experiment was to
ascertain the degree of operational
safety incorporated in the mechanism
for separating the parts of the
assembly.

37
ANP PROJECT QUARTERLY PROGRESS REPORT

 

»
Q

 

o]
Q

    

DECREASE IN REACTIVITY (cents)
o0
o
—_—

e

o
o
|
i

 

 

\ | | |
moL ........ L_ml“Jfi%_Lmfllmwmw_w

0 of 0.2 03 0.4 0.5 06 o7 0.8 09
’ SEFARATION (in)
Fig., 14. variation of Reactivity

with Axial Separation of Mid-plane.

CIRCULATING-FUEL REACTOR
B. Scott, ANP Division

A preliminary, experimental studyof
the nuclear design of the experimental
to be made during the
summer of 1952 at ORNL by building
some of the major components into a
system that will be made critical with
enriched uranium. The experiments
will be done at a power level of the
order of 1 w, maximum, and at room
temperature, and will yield information
on uranium requirements, control
problems, and power and neutron-flux
distributions, It is expected that
assembly of the mockup will commence
June 1 and that the system will be
made critical around July 15,

reactor 1s

Reactor Materials. The beryllium
oxide hexagonal blocks prepared for the
ABRE will be used as the reflector and
moderator in these experiments and
transferred to the ARE later. The
fuel is to be a dry-packed powder with
nuclear properties closely approxi-
mating those of the ARE fuel (cf.,
sec, 10). At present the ARE fuel is
designed to be a molten mixture of the

38

fluorides ZrF,, NaF, and UF,, with the
possible inclusion of some KF, In this
preliminary assembly the fuel is
expected to be a mixture of Zr0O,, NaF,
enriched UF,, and graphite. The Zr0O,
is being substituted for the ZrF,
because the latter is unavailable at
the graphite is being added
scattering cross

this time;
to give a total
section of oxygen and carbon equivalent
to that of the fluorine. The compo-
sition will be that required to give
the designed uranium density of the
ARE fuel. It is probable that the
ratio of zirconium to sodium fluoride
will approach ABE design, but the over-
all density will be somewhat lower.
The simulated reflector coolant of the
ABE to be used in this preliminary
study has the same composition as the
fuel, with the omission of the UF4.

These materials will be packed as a
dry powder 1in stainless steel tubes,
1 1/4 in. in diameter and 40 in. long,
which can be sealed to minimize moisture
pick-up. The stack of beryllium oxide
blocks will be surrounded by an Inconel
shell.

Reactor Assembly. The beryllium
oxide blocks will be stacked, with
axes vertical, in a right cylinder,
also with its axis vertical, 36 in.
high and 47 in. in diameter. The
fuel and coolant mixtures in stainless
steel tubes will be placed vertically
through the axial holes in the beryllium
oxide blocks. Some of the tubes will
be movable, by remote operation, and
will serve as control rods. A section
of tubing containing a nuclear poison
will be attached above other fuel
elements, The fuel elements will be
magnetically supported. Upon release
by the magnet, the fuel tubes will
fall out of the core and will be
replaced by the poison. This safety
mechanism will be operated both manually
and by a high-flux-level detection
instrument.
Provisions will be made for in-
serting neutron-detecting foils, such
as indium and gold, through the re-
flector and moderator. The usual
neutron- and gamma-sensitive detectors
will be placed adjacent to the assembly
for operational control and safety.
They will feed recordingand indicating

instruments in the contral room.

GRAPHITE REACTOR

E. L. Zimmerman, Physics Division

A critical assembly of enriched
uranium and graphite was described 1in
earlier reports,(4) and a summary of
the results obtained with this assembly
is given here, The core of the assembly
was essentially a cube 45 by 45 by
44.11 in. covered by a 3-in.-thick
graphite reflector on four sides.

Critical Mass. Although the quantity
of uranium available limited the load-
ing, the assembly became critical with
41.5 kg of U235, Measurements indi-
cated that little increase in reactivity
would have resulted had uranium been
included in the reflector lavyer.,
Therefore, for purposes of calculation

the reactor was considered to be un-

reflected, with dimensions 51.0by 51.0
by 44.1 in., and to contain 52.48 kg
of U235, The difference between the

two mass values 1s that reguired to
fill the 3-in., reflector to the same
uranium density as the core. A multi-
group calculation for the assembly, (5)
in which a value for the buckling, B?,
of 0.00185628 cm~? and an extrapolation
distance of 2 cm were used, gave an
effective multiplication of 0,9912.
The disadvantage factor due to lumping
of the fuel 1n 0.01-in.-thick disks
was measured and found to be 0.94 ¢
0.04, which 1s comparable to the value

(4)AircraftNuclearPropulsion Project Quarterly
Progress Reports for Periods Ending September 10,
1954, OBRKNL-1154, p. 80; June 10, 1951, ANP-55,
P. 82; Merch 10, 1951, ANP-60, p. 120.

(5)y. J. Nielson, Bare Pile Adjoint Salution,
Y-F10-18 (Oct. 27, 1950).

 

 

 

FOR PERIOD ENDING JUNE 10, 1952

of 0.95 used in the calculation. The
calculations gave values of 0.15 ev
for the mean fission emergy and 27.41%

for the fraction of fissions in the
thermal group.
Control Rod Calibration. Control

rods of the core-removal type were
calibrated by a “rod-drop" method and
by the observation of stable reactor
periods. The form of the control rod
calibration curves indicated a contri-
bution due to neutron streaming in the
void formed by withdrawing the rod as
well as the expected cosine-sguared
variation. A comparison using a flat
strip of cadmium as a poison rod, which
left essentially no void when removed,
showed good agreement with a cosine-
squared sensitivity curve,

Gap Effect.
as a function of a gap

The loss 1n reactivity
between the
assembly halves showed a loss of
multiplication of 0.00635 due to a gap
0.3 in, wide. This agrees fairly well
with a recent calcnlation, (%) but is
roughly double the value calculated by
an carlier method.(7)

Flux and Power Distribution. Bare-
indium and cadmium-covered-indium foils
were exposed in various parts of the
assembly to observe the flux distri-
bution both macroscopically and micro-
scopically. Power distributions were
observed by means of aluminum catcher
foils in contact with the uranium. An
experiment showed the catcher-foil
technique to be insensitive to the
neutron spectrum, Comparison between
activations of bare, cadmium-covered,
and cadmiunm-indium-covered fuel disks
gave values of the ratio

(Fuel activation) Bare

 

{Fuel activation) Cadmium covered

 

(6]5. Tamor, unpublished work.

(T)y. G. Goldberger, M. L. Goldberger, and
J. E. Wilkins, Jr., The Effect of Gaps on Pile
Reactiwity, CP-3443 (Feb. 20, 1946).

39
of between 2.1 and 2.6 compared with a
calculated average value of 2.72. The
values of the ratio

(Fuel activation) Bare
(Fuel activation) Cadmium and indium covered

 

varied between 3.4 and 3.9; the calcu-
lated value was 2,98. The variation in
observed ratios was due to the 3-1in.
graphite reflector on four sides of the

assembly.

Danger Coefficient Measurements.
Danger coefficients were calculated

40

for sodium, iron, nickel, and molybdenum
by using the neutron spectrum determined
by multigroup calculations and the
known cross-section data. The values
agreed substantially with the corre-
sponding observed values. As a further
experiment, a 3- by 2-in. hole extend-
ing from the center of the assembly to
the outside along a major axis was
filled step-wise with sodium. Plugging
the hole with sodium caused a slight
increase in multiplication, whereas
caompletely filling the hole gave a
small net loss.
 
 

 

   

 

 

 

   

 

 

 

 

   

   

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

: - i '
: -
:

  
s

SUMMARY AND

E. P. Blizard
Physics

Measurements of the energy and
angular distribution of gamma rays
from the divided-shield mockup around
the Bulk Shielding Reactor have
continued, and development of the
proton-reco1l counter for neutron
spectral measurements 1is progressing
(sec. 6). In another experiment with
this reactor the effect of radiation
from the side of the reactor shield
that scatters in air and penetrates
the crew compartment was measured,
These data for neutrons were reasonable,
but the gamma rays appear to be under-
shielded., The albedo of neutrons and
gamma rays on concrete has been
determined with artificial sources
to facilitate calculations for the
Tower Shielding Facility.

Both GE-ANP and NDA shield design
groups have come upon the difficult
problem of calculating secondary gamma
production from neutrons absorbed or
scattered in the thermal shields and
pressure shells of reactors, Since
these problems are insoluble with the

INTRODUCTION

J. L. Meem

Division

presently available nuclear constants
and would be very difficult even if
all data were well known, an extensive
survey was carried out in the Lid Tank,
This survey has enabled both groups to
proceed with their respective designs

(sec. 7).

The air-duct research in the Lid
Tank and Thermal Column Facilities
has continued. The Lid Tank research
has been concentrated on large-scale
mockups of annular designs, and the
Thermal Column work has been on a
series of experiments to provide
primarily a better understanding of
the important attenuation processes

(sec, 8).

The total neutron cross section of
Li% has been measured up to 4 Mev on
the 5-Mev Van de Graaff. The measure-
ments show only one resonance, which
is at 270 kv. The time-of-flight
spectrometer has been operating
satisfactorily with a full resolution
width of less than 1.2 usec/meter.

 

6. BULK SHIELDING REACTOR

J. L. Meem
H. G, Cockran E. B. Johnson
M. P. Haydon J. K. Leslie
K. M. Henry T. A. Love
I.. B. Holland F. C. Maienschein
H. E. Hungerford T. N. Roseberry

Physics Division

The requirements for divided-shield
researchwere explored quite completely
during the past quarter. A set of
experiments at the Bulk Shielding
Reactor was executed to examine  the
possibility of testing daivided shields

on a limited basis, that is, testing
the radiation that penetrates the
si1de of the reactor shield, air-
scatters, and enters the side wall
of the crew shield. The experiments
were necessarily so unclean that 1t

43
ANP PROJECT QUARTERLY PROGRESS REPORT

is difficult to extrapolate from the
radiation measured to that to be
expected i1n the aircraft, This fact
1is much more meaningful than the
actual results, which on a crude
interpretation seemed to indicate that
the divided shield described in
ANP-53¢1) was properly designed for
shielding from neutrons but that in
regard to gamma rays it was somewhat

undershielded.

When 1t became apparent that such
partial experiments would not supply
adequate information, constiderable
effort was expended 1n designing a
new facility that would be satis-
factory., To facilitate design work,
other experiments were carried out to
determine the scattering to be ex-
pected from the ground, and calcu-
lations were made to determine the
minimum requirements for power and
altitude. A proposal describing a
divided-shield testing facility has
been prepared and submitted to the
Atomic Energy Commission for review
and possible approval. The study
revealed that if the reactor operates
at 100 kw and the fast-neutron do-
lmeter sensitivity 1s increased by a
factor of 10, full-scale divided-
shield mockups can be tested and
optimized. The tower configuration
shawn in the last quarterly report
continues to be the favored design.

Measurements of the energy and
angular distribution of gamma rays
from the divided-shield mockup have
continuned, and development of the
proton-recoil counter for neutron
spectral measurements 1s progressing.
Other activities 1involving the Bulk
Shielding Reactor i1nclude partial
participation 1n an experiment on the
determination of the threshold for
eye cataract formation for the Biology

 

(I)Report of the Shielding Board for the

Aircraft Nuclear Propulsion Prograea, ANP-33
(Oct. 16, 1950).

44

Division and irradiations of electronic
equipment for the Solid State Division
and the Wright Air Development Center.

MOCKUP OF THE DIVIDED SHIELD

Additional gawmma-ray spectral
measurements with the divided-shield
mockup(2?) have been obtained and a
report on the spectra measured to date
is being prepared.(?? Table 2 lists
the spectrometer positions for the
me asurements, and Fig. 15 shows spectra
at various distances from the reactor
with and without the divided-shield
mockup. For all the measurements in
Fig. 15 the spectrometer was at zero
degrees, that 1s, facing the center
of the reactor. From this data
relaxation lengths as a function of
energy have been determined that are
not inconsistent with shielding theory.
An angular distribution has been
plotted from the data obtained at
147.6 ¢cm from the reactor, and the
gamma-ray flux at this position has
been converted to dosage, which may
be compared with ion chamber readings.
Details will be given in the report
being prepared by Maienschein.(3)
A lead shadow shield has been added
to the mockup(*’ for the gamma-ray
spectral measurements now being made.

Centerline measurements of the
thermal-neutron flux, fast-neutron
dose, and gamma~ray dose are nearing
completion and will be reported at a
later date.

Development of the proton-recoil
counter for neutron spectral measure-
ments 1s progressing. The instrument
will measure the spectrum of an

(2)Aircraft Nuclear Propulsion Project Quarterly
Progress Report for Period Ending March 10, 1952,
ORNL-1227, p. 73.

(e, .

Msienschein, Gemama-Ray Spectral
Heasurements with the Divided Shield Mockup,
Part II (in preparation]).

(4)pigure 20, op. cit., OBNL-1227, p. 74.

il
LY

FOR PERIOD ENDING JUNE 10, 1952

TABLE 2

Spectrometer Positions for Gamma-Ray Spectral Measurements
with the Divided-Shield Mockup

 

 

 

ANGLE BETWEEN REACTOR
ARBRANGEMENT DISTANCE FROM REACTOR CENTER LINE AND SPECTROMETER
(cm) COLLIMATOR (deg)
DSM* (borated) 96.6 0
20
DSM (unborated) 06.6 0
Water 6.6 0
20
40
DSM (borated) 147.6 0
10
14
20
25
30
40
60
DSM (unborated) 147.6 0
Water 147.6 0
14
20
40
DSM (borated) 207.6 0
Water 207.6 0
DSM (borated) 267.6 0
Water 267.6 0

 

 

 

*Divided- shield mockup.

arti1ficial neutron source quite satis-
factorily, as will be described i1n a
Physics Division guarterly progress
report; (3) however, the feasibility
of collimating the neutrons for a
measurement of the reactor shield
spectrum is still to be demonstrated.

 

(S)J. L. Meem et al,, * Bulk Shielding Reactor)’
Physics Division Quarterly Progress Report for
Period Ending June 20, 1952 (in preparation).

Research on this problem 1s being done
with the Cockcroft-Walton Accelerator.

Me asurements have been completed
on the power distribution of the
beryllium oxide-reflected reactor used
for the divided-shield-mockup ex-
periments. A report will be prepared
during the next quarter.

45
ANP PROJECT QUARTERLY PROGRESS REPORT

6 DWG. 14736
10

OPEN WATER
——— DIVIDED SHIELD MOCKUP

 

 

 

GAMMA-RAY FLUX, I (gammas/cm?/sec/Mev/watt / steradian)

 

O 2 4 6 8 10 12
GAMMA-RAY ENERGY (Mev)

Fig. 15. Gamma-Ray Flux as a Function of Energy and Distance from the Bulk
Shielding Facility Reactor.

46
L

AIR-SCATTERING EXPERIMENTS

During preparation of the Tower
Shielding Facility Proposal, the
guestion arose as to whether a meaning-
ful experiment on a divided shield
that would include the effect of air
scattering i1nto the crew box could
be carried out in the Bulk Shielding
Facility. An experiment was devised(®)
to determine the attenuation of the
radiation leaving the sides of the
reactor shield, scattering in air,
and penetrating the side wall of the
crew shield. The results of this
experiment were compared with the
standard divided-shield design of the
ANP Shielding Board. (1)

Shield Arrangewent. The standard
divided-shield design 1s shown 1in
Fig. 16, and the experimental ar-
rangement is shown in Fig. 17. The
water level of the pool was lowered
to a predetermined distance above the
Bulk Shielding Reactor to mock up a
given reactor side shield thickness
for the standard design. Some of the
radiation leaving the surface of the
water above the reactor was scattered
in air and reflected down into the
water about 5 meters away where
measurements were made. By lowering
a counter under the water to a gilven
depth, the attenuation of the side
wall of the crew shield could be
simulated.

Three lead slabs with a total
thickness of 4.5 in. were placed 1in
front of the reactor so that the tops
of the slabs were at an angle to the
center of the Bulk Shielding Reactor
that corresponded to the angle of
interception of the gamma rays by
the lead in the divided-shield design.
The lead slabs plus the approximately
5 meters of water essentially blacked
out the direct beam of both neutrons

 

(6).]. L. Meem and H. E,
Scattering Experiments at the Bulk Shielding
Facility (in preparationm).

e e T A e e T = e A R e R

Hongerford, Air

FOR PERIOD ENDING JUNE 10, 1952

and gamma rays toward the point of
me asurement.

Radiation Dosage at Crew Com-
partment. The experimental results
are shown in Figs. 18 and 19, Data
were taken at 30, 60, and 90 cm above
the active core of the reactor. In
Fig. 18 the neutron-dosage rate 1is
plotted against the distance of the
counter from the water surface. As
the counter went below the water
surface a sharp break in each curve
was noted and the break was taken to
define the position of the surface of
the water. Similar data for gamma-ray
dosage is shown in Fig. 19. The
uncorrected data are given else-
where, (9

As can be seen 1n the standard
design (Fig. 16), the thickness of
the reactor shield side wall was
113 cw, which 1s equivalent to 116 cm
of water if corrections for the fast-
neutron cross sections of beryllium
carbide, 1ron, and gasoline are made.
Since it was necessary to run the
reactor at its maximun power to obtain
data with 90 cm of water over the
reactor, the neutron data were ex-
trapolated to 116 c¢m, as shown 1in
Fig. 18. At the position of the
detector below the water corresponding
to the thickness of the side wall of
the crew shield the dose was 8 X 10°8
mrep/hr/w.

For the gamma-ray measurements the
effective thickness of the reactor
shield side wall was obtained from
the pt(density times thickness)of the
standard design, which was 142 g/ cm?
(Fig. 16). Since this is equivalent
to 142 ¢m of water, the gamma-ray
data were extrapolated to that value
(Fig. 19). The value observed through
17.8 ¢m of water at the crew position
was 5.4 X 107 r/hr/w without any
lead. Referring to Fig. 16, 1t can
be seen that the standard design
called for 1.54 cm of lead on the

47
ANP PROJECT QUARTERLY, PROGRESS REPORT

SIDE WALL MATERIAL
178 cm PLASTIC

- SIDE WALL MATERIAL

SE—
OWG. 14687

pt
36
24
50

32
t42

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

15 cm ’BGZC'C
1.54 cm LEAD 3 ¢m Fe
50 cm HZO
45 cm GASOLINE
113 cm TOTALS ©
7 ',/7'.77'77'7; -
LEAD GLASS y ” Sl L0 / / // // / / / / /
I 188R.
223
————— - ] B -
£ DOSIMETER
S //”/// NG
L-—-m—-——— ------------ 412 »L
L -—'\N—iSOO*’\N—
LEGEND : CREW SHIELD

ALL OIMENSIONS IN CENTIMETERS

R IRON

 

UPLASTIC

FZAFUEL

BB REFLECTOR

[E=] WATER OR GASOLINE AS NOTED

Standard Divided Shield Design,

 

REACTCR SHIELD

 

 

 

 

 

 

OWG. 14666
----- VERTICAL TRAVERSE MAOE ALONG THIS LINE
SUPPORT
- 510 ]
e -
| ~DETECTOR WATER LEVEL LEAD SLABS 4 5—-&59
| — T
‘ = HEIGHT OF

. —— . O, 25 ||| 120

 

 

POOL FLOOR ~__

REACTOR (‘ORE

   

| WATER  ABOVE
REACTOR CORE

60—

 

 

ALL DIMENSIONS IN CENTIMETERS

Fig. 17. Arrangement for Air-Scattering Experi

48

ment.
......

crew shield.

Attenuation of Gamma Rays by Lead.
A separate experiment was performed
to determine the attenuation of the

1071

¥=DEPTH OF WATER ABOVE
TOP OF REACTOR

FAST-NEUTRON DOSE (mrep/hr/watt)

gxio™8
10°7 EXPERIMENTAL DOSE AT
CREW POSITION =

 

10-10
+40 +30 +20 10 0O 10 -20 -30 -40

DISTANCE ABOVE WATER SURFACE {cm)

Fig. 18. Air-Scattering Experiment
Fast-Neutron Dosimeter Readings at
Crew Position for Various Reactor
Shield Thicknesses,

FOR PERIOD ENDING JUNE 10, 1952

gamma rays by the lead. An 18-in.-dia
water-tight container for an anthracene-
crystal gamma-ray counter was con-
structed. The container had 3 in. of
lead on the sides, 1 in. of lead on
the bottom, and a thin aluminum top.
Sheets of lead varying in thickness
up to 1 in. were placed on top of the
container, and for each lead thickness
the assembly was lowered 17.8 c¢m
below the surface of the water at the
simulated crew position. The resulting
curve for lead absorption is shown in

Fig. 20.

Comparison of Experiment with
Standard Design Conditions. The
standard design specified(!) that
21.5 mrep/hr of neutrons and 0.25
r/hr of gamma rays would be received
through the side walls of the crew
compartment with the reactor at 200
megawatts of power. In order to
compare the experimental data with the
standard design specifications, it
was necessary to calculate a scale-up
factor.

10-3 DWG. 14665

Y=DEPTH OF WATER ABCVE
TCP OF REACTCOR

SAMMA DOSE (r/nr/watt)

EXPERIMENTAL DOSE AT CREW = 7
5 POSITION ( WITHOUT LEAD) :

 

10-7
+40 +3C +20 A0 0 -0 -20 -30 -40

DISTANCE ABOVE WATER SURFACE (cm)

Fig, 19. Air-Scattering Experiment
Gamma-Ray Dosage Readings at Crew
Position for Various Reactor Shield
Thicknesses (No Lead at Crew Position).

49
ANP PROJECT QUARTERLY PROGRESS REPORT

A

DWG. 15072

 

0.4

Ul
w
O
(-]
<I
=
=
<
o
el
> AMOUNT OF LEAD IN
- CREW COMPARTMENT ~—
- AS SPECIFIED BY ANP-53
Qc

0.0¢

N\
N 534cm
\
N
0.0022—-—+\
™~
AMOUNT OF LEAD INDICATED
AS NECESSARY BY THESE EXPERIMENTS
0.001
0 1 0 3 4 5 6

THICKNESS OF LEAD (cm)

Fig. 20, Air-Scattering Experiment Gamma-Ray Attenuvation by Lead at Crew Com-~
partment,

50
If

¢ = neutron dose, mrep/hr/w, as
., measured 1n the experiment,

" = gamma dose, ¥/hr/w, as measured
in the experiment,

D, = airplane neutron dese, mrep/hr
at 2 X 108 w,

 

 

 

 

 

Dy = airplane gamma-ray dose, r/hr
at 2 X 10% w,
then
p, - gpmie|le, O
" Ay L, ®d,
where
Aa
= ratio of effective source
Ay areas exposed (explained
below) = 16,
La
2 = ratio of leakages per watt = 0.2,
“b

= correction factor for the
leakage from a cylindrical
shield surface instead of a

 

 

plane surface = 1.8,
P, = aircraftreactor power = 2% 108w,
db
7 = correction, based on single

a scattering i1n air, for the
fact that the separation
distance in the experiment ts
5 meters instead of 15 meters
as in the airplane.

A 1s the area of the cylindrical

a
side surface of the airplane reactor,

and 4, 1is the area of the top of the

Bulk Shielding Beactor. In the ex-

periment the only radiation escaping
upward through the surface of the

FOR PERIOD ENDING JUNE 10, 1952

water 1s assumed to have originated
on the top areca of the reactor. This
is to be compared with the radiation
emerging from the cylindrical side
sur face of the airplane reactor (not
including the front and rear surfaces).
As discussed in more detail else-
where, (%) the ratio of
found to be 16.

areas was

In the unit shield report¢’? the
ratio of the leakages per watt for
the airplane reactor and the Bulk
Shielding Reactor was found to be
about 0.2. r_ 1s the aircraft reactor
radius and r_ i1is the aircraft reactor
shield radius. The factor vr /r_ is a
correction to take 1nto account the
fact that the airplane shield 1is
cylindrical, whereas the Bulk Shielding

Reactor shield is flat.(%> Applying
all the corrections,
db
8
e T 3.8 x 10° .
da

 

This should be an upper limit. The
best estimates of the amount of
scattering from the walls of the pool,
reactor bridge structure, etc., in-
dicate that wall-scattered radiation
was 2 or 3 times the ailr-scattered
radiation for the experiment for both
neutrons and gamma rays. A lower
limit on the scale-up factor is then
1.3 % 10® with a most probable value
of around 2.1 % 10%, By using the
latter value for the calculations:

D, = (2.1 x 10%) ¢,

and since ¢ was measured to be 8§ X 10°8
mrep/ hr/w,

D = 16.8 mrep/hr.

n

 

(7)J. L. Meem and H. E. Hungerfard, The Unit
Shield Experiments at the Balk Shielding Faecility,
ORNL-1147 (April 390, 1952),

ol
ANP PROJECT QUARTERLY PROGRESS REPORT

This compares reasonably well with
the specified dose of 21.5 mrep/hr.

In converting the experimental
gamma-ray dose, [, to the airplane
dose, Dy, an additional factor, f(Pb),
must be introduced for the attenuation

 

 

of the lead inside the crew shield:
D -—rA“L“J—r—iP if(Pb)
7 Ab Lb " a ¢ a ,
D, =T (2.1 % 108) f(Pb).

The experimental gamma-ray dose,
I", was 5.4 X 10°7 r/hr/w through 17.8
cm of water, and 1f the specified
value of 0.25 r/hr 1s used for Dy, the
factor by which the lead should
attenuate the gamma rays can be found;
that 1s,

D, 0.25

[ (2.1 % 10%) 5.4%x10°7 x 2,1%x108

 

1

f(Pb)

2.2 x 1073,

i

Inspection of Fig. 19 shows that
5.34 cm of lead 1s required to at-
tenuate the gamma rays by this amount.
The standard design specified 1.54 cm
of lead, which would give an
tenuation of 5.9 X 10°2%. The gamma
dose 1n the standard design would
therefore be too large by a factor
of about 27. The additional lead
required 1f 1t were placed inside
the crew compartment is 3.8 em. This
much lead would weigh approximately
25,000 pounds. Perhaps considerable
savings 1n welght could be made by
placing the lead 1in the reactor
shield. The experiment gave no in-
formation concerning such a dis-
position of the lead.

at -

The approximations made in con-
verting the experimental dosage rate
to that expected 1n the airplane

52

configuration are admittedly crude.
Furthermore, the effect of the walls
of the pool and building in scattering
the radiation back to the point of
detection has not been determined.
However, the disconcerting fact re-
mains that the neutron data 1s 1n
rough agreement with the calculations
of the Shielding Board, whereas the
gamma-ray dosage appears to be high
by a factor, which, while less than
50, seems to be at least as highas 15.
Further experiments, at 100 kw reactor
power, are contemplated to verify or
correct this high factor.

DETERMINATION OF THE ALBEDO OF
NEUTRONS AND GAMMA RAYS

To aid in the calculations of
backgrounds at the proposed Tower
Shielding Facility several short
experiments¢®’ were performed to
determine the reflection of neutrons
and gamma rays. DBy using a neutron
source and the fast-neutron dosimeter
shown i1n Fig. 21, measurements were
made with the source and counter at
varying heights above the concrete
floor in the Bulk Shielding Reactor.

By using the relation(®’

alN,
F =,
107d?
where F is the scattered flux, N; is
the source strength, and d 1s the
distance above the floor, the albedo,

a, for neutrons was found to be 0.12
for concrete. A similar experiment
in which an ion chamber and a Co®°
source with lead shielding was used
gave the reflection coefficient of
gamma rays on concrete as 0.04.

 

(S)H. E. Hungerford to J. L. Meem, Some Ground
Scattering Experiments Performed at the Bulk
Shielding Facility, ORNL CF-52.4-99 (April 16,
1952),

g}A. Simon to E. P, Blizard, Estimate of
Background at Tower Shielding Facility, ORNL CF-
51-12-185 (Dec. 17, 1951).
1 "fif'

  

Ceen LHTA T ewe weeTia g fasaty,
. *
setr oy e a e . ’ ‘ol et
Q. ¢ « . e
.
-

FOR PERIOD ENDING JUNE 10, 1952

DWG. 14697

TO0 CRANE

   
 
  
 
 
 
  

PARAFFIN BLOCK

FAST-NEUTRON
DOSIMETER

Po-Be
NEUTRON SOURCE

“ b % M Eoaey e . .
T eat i d et e st p e

Sl e e LR e s A s a. :'.;"
e P T T . : Lt Tttt e R
sl e CONCRETE FLOOR . <" " o 1. .4
., . .'. . . . . . . Lo . -. st . . . . . . . "
T . - et "- . 0 = A i -~ ., - a . . . . ‘“ *
. fi'. . . . : -~ s . - . - ¢
. . .« . .

Fig. 21. Arrangement of Apparatus for Albedo Experiment.

53
ANP PROJECT QUARTERLY PROGRESS REPORT

IRRADIATION OF ANIMALS

In the 1fiterest of determining the
threshold for eye cataract formation,
time on the Cockcroft-Walton ac-
celerator has been furnished to members
of the Biology Division for exposures
of rats, mice, and rabbits to source
intensities of 10% neutrons/sec with
14-Mev energy. Approximately 60 hr of
irradiation time on the machine has
been used for these experiments
during this quarter. HResults will be
published by members of the Biology
Division. This irradiation program
will continue.

IRRADIATION OF ELECTRONIC EQUIPMENT

In cooperation with the Solid
State Division and the Wraight Air
Development Center, the Bulk Shielding
Reactor is being used on week ends to
irradiate electronic equipment.
Reactor operators furnished by the
Air Force have been trained by ORNL
personnel.
ment,
posed for 100 hr to a fast-neutron
flux of between 107 and 10® neutrons/sec
with no appreciable effect. Detailed
results will be reported by the Solid
State Division.

The first piece of equip-

a radio compass, has been ex-

 

7. GAMMA-RAY ATTENUATION EXPERIMENTS IN THE LID TANK

C. E. Clifford
T. V. Blosser
J. D. Flynn

M. K. Hullings
L. S. Abbott
M. C. Marney

Physics Division

A series of short experiments on
the gamma-ray attenuation of heavy
materials (iron and lead) in regions
close to the source have been completed
in the Lid Tank. The experiments
planned to aid in the design of both
thermal and shadow shields for various
aircraft reactors were necessary
because of the difficulty of calculating
the production of neutron-induced
secondary gamma rays 1n these regions.

63% IRON-37% WATER THERMAL SHIELD

Specific shield configurations
requested by Goldstein of NDA for
possible use in the supercritical-water
reactor are shown as part of Figs. 22
through 26, The i1ron-water mixtures
were placed at various distances from
the source to simulate the varying
water-reflector thicknesses present 1in
the reactor design; although the
reactor is cylindrical, 1t 1s sur-
rounded by a spherical pressure shell
and thermal shield.

24

The shield configurations were
mocked up with one to six 2.22-cm iron
slabs separated by 1.3 cm of water,
and they were placed at various
distances from the source. In each
series the sixth slab was moved forward
to touch the preceding slab to simulate
a pressure shell, as shown in Fig. 26.
The measurements were made with both
an air-filled 1onization chamber and
an anthracene scintillation counter.

Figure 27 indicates the gamma-ray
dosage distribution in water, and
Figs. 22 through 26 give the gamma-ray
measurements behind the i1ron-water
thermal shields. A curve of the dose
at a constant z (distance from the
source) as a function of the water
thickness between the source and the
nearest iron slab (referred to as
reflector thickness) is also shown.
The tabulated "z measurements to back
of last slab" (inset in each figure)
are probe measurements of the distance
from the source to the back of the
last slab in each array. As would be
expected, the data indicate that for
thick thermal shields the secondary
gamma-ray production is predominant
up to rather large reflector thickness

(i.e., 20 cm).

63% IRON-37% BORATED WATER
THERMAL SHIELD

A stfi%%fibf the effect of borating
the water in the region of the thermal
shield was made to determine the
resulting reduction of secondary gamma-
ray production (Figs. 28, 29, 30).
The configurations were chosen to
match the neutron and gamma-ray leakage
in the Lid Tank with that of the
reactor on the basis of a calculation
by Goldstein. A scale-up factor was
also calculated and used to determine
the required Lid Tank dosage beyond
the thermal shield. The 2.22-cm iron
slabs were separated by the borated
water, and since sufficient attenuation
was obtained with a four-slab (27%
borated water) thermal shield, attenu-
ation of the pressure shell (two
adjacent slabs) was then measured

(Fig. 30). The measurements were made
with the 10'% jonization chamber
(air-filled).

A comparison of attenuation of the
borated-water thermal shield with that
of the unborated-water shields previ-
ously measured indicates that at this
reflector thickness (20 cm) a reduction
in intensity to 70% of the iron-water
case (at z = 80 cm) was obtained. The
reduction of intensity as the water
reflector thickness was increased to
30 c¢cm indicates that secondary gamma-
ray production 1s not entirely sup-
pressed by the 1% boration of the
thermal shield water. This 1s not
surprising 1in view of the large volume
percentage of i1ron in this region.
There may also be some 1nelastic
scattering of gamma rays, which would
of course not be affected by the
boron.

FOR PERIOD ENDING JUNE 10, 1952

SOLID TIRON THERMAL SHIELD IN WATER

As a continuation of the thermal
shield investigation, the gamma-ray
attenuation of a 9-cm iron thermal
shield was measured for the GE-ANP
design group. Four 2.2-cm iron slabs
were clamped together and placed 1in
the Lid Tank at various distances from
the source, as indicated in Fig. 31,
Since the iron slabs are not flat,
water (approximately 0.3 cm) was
present in the space between adjacent
slabs.,

The measurements presented 1n
Fig. 31 were obtained with the usual
air-filled ionization chamber, which
is, unfortunately, neutron-sensitive
owing to the nitrogen (n,p) reaction,
This invalidates the measurements
close to the i1ron, but since the
neutron relaxation length in water
behind solid iron is short (4 to
6 cm), the neutron effect should be
negligible beyond about 30 c¢cm of
water. Measurements that indicate
the magnitude of the neutron effect
and confirm the fact that 1t 1s
negligible beyond 30 cm have been
made on similar mockups by using CO,
in the i1onization chamber.

SOLID LEAD SHADOW SHIELDS IN WATER

Measurements were taken in the Lid
Tank at the request of GE-ANP to
determine the effect of various water-
reflector thicknesses on the gamma
attenuation of solid lead shadow
shields, Since the experiment was
performed in water, a large amount of
secondary gamma-ray production reduced
the effectiveness of the lead by a
factor greater than 40 in the worst
case. As the thickness of the water
layer between the source and the lead
was increased, the secondary gamma-
ray production was, of course, reduced
at a rate equivalent to the neutron
attenuation of the water.

55
ANP PROJECT QUARTERLY PROGRESS REPORT

The gamma-ray measurements for the
shadow shields consisting of one, two,
and three adjacent 3.8-cm lead slabs
are given in Figs, 32, 33, and 34.
The "z measurements to back of last
sl ab"™ are probe measurements and
include water gaps caused by warpage
of the slabs. 1In each case the water-
reflector thickness was held constant

by wooden spacers 1inserted between
the source and the first slab,

The gamma measurements were made
with the 10'° ionization chamber.
During the course of the experiment
CO, was substituted for air as the
ion-chamber gas to eliminate the nitro-
gen (n,p) reaction, since measurements
were required in a high neutron flux.

DWG. 14716R1

10 0.9-cm WATER REFLECTOR

5 GAMMA DQSE VS._. .. ,
REFLECTOR THICKNESS
AT z=39 cm

{mr/hr)
N

GAMMA DOSE
o

WATER Z MEASUREMENT
REFLECTOR TO BACK OF LAST
(cm) SLAB
{cm)
2 fo . 3

4.7
7.2

ONE 2.22-cm IRON SLAB
2 4L

S i
b°°°° -

SOURCE |-~

O 20 40 60

Fig. 22. Thermal Shield

56

 

WATER REFLECTOR (09 to 5.0 cm)

80 100
Z, DISTANCE FROM SOURCE (cm)

Measurements
Iron Slab with Various Water-Reflector Distances.

120 140

- Gamma-Ray Dose Beyond One 2. 22-cm
FOR PERIOD ENDING JUNE 10, 1952

DWG. 1464 5R1

0.9-cm WATER REFLECTOR

WATER Z MEASUREMENT
REFL.LECTOR TO BACK OF
(cm) LAST SLAB
(cm)

 

6.6

@O NO
NOOHNOW

!
{
2

=
>
E
L
wn
O
O
<1
=
=
<I
w
GAMMA DQSE VS. REFLECTOR
THICKNESS AT 2z =80 ¢cm
WATER REFLECTOR (0.9 to 25.4 ¢m)
TWO 2.22-cm IRON SLABS (SEPARATED BY 1.3 cm OF WATER)
SOURCE

 

0 20 40 60 80 100 120 140
z, DISTANCE FROM SOURCE (cm)

Fig. 23. Thermal Shield Measurements - Gamma-Ray Dose Beyond Two 2, 2Z-cm
Iron Slabs (Separated by 1.3 cm of Water) with Various Water-Reflector Distances.

57
ANP PROJECT QUARTERLY PROGRESS REPORT

DWG. 14717R1

= 1.4-cm WATER REFLECTOR

 

WATER 7z MEASUREMENT
REFLECTCR TO BACK OF
LAST SLAB
{em)
14 2
~ 18 .5
A
= 23.7
i 29.0
- 38.6
L
w
C
(48]
<{
=
=
T
j 4]
GAMMA DOSE VS. REFLECTOR
THICKNESS AT 2z =80 c¢cm
WATER REFLECTOR (%.4 10 25.9 ¢m)
FOUR 2.22-cm IRON SLABS
( SEPARATED BY 1.3 cm OF WATER)
SOURCE

 

O 20 40 60 80 100 120 140
z, DISTANCE FROM SOURCE {cm)

Fig. 24. Thermal Shield Measurements -~ Gamma-Ray Dose Beyond Four 2,22-cm
Iron Slabs (Separatedby 1.3 cmof Water)with Various Water-Reflector Distances.

58
FOR PERIOD ENDING JUNE 10, 1952

DWG. 14718R!

0.0-cm WATER REFLECTOR

WATER Zz MEASUREMENT
REFLECTOR TO BACK OF
(cm) LAST SLAR
(cm)

 

12.8
25.7
30.6
35.6
45.6

£

=

£

L}

wn

Q

0O

<I

< AMMA DOSE VS. REFLECTOR

g THICKNESS AT 7z =80 cm

WATER REFLECTOR (0.0 to 25.8 cm)

SIX 2.22-cm IRON SLABS
(SEPARATED BY 1.3 cm OF WATER)

SuR

 

0 20 40 60 80 100 120 140
Z, DISTANCE FROM SOURCE (cm)

Fig. 25. Thermal Shield Measurements - Gamma-Ray Dose Beyond Six 2.22-cn
Iron Slabs (Separated by 1.3 cmof Water) with Various Water-Reflector Distances.

59
ANP PROJECT QUARTERLY PROGRESS REPORT

DWG. 1471SR1

WATER z MEASUREMENT
REFLECTCR TO BACK OF
{(cm) LLAST SLAB
(cm}

 

18.8
21.5
24.6
29.5

{mr/hr)

GAMMA DOSE VS. REFLECTOR
THICKNESS AT 7=80c¢m

GAMMA DOSE

WATER REFLECTOR (0.3 to 25.9 cm)

FOUR 2.22-cm IRON SLLABS
(SEPAI’?ATED BY 1.3 cm OF WATER)

| |
PRESSURE SHELL (TWO 2.22-cm IRON SLABS )}

 

 

20 40 60 80 100 120 140
z, DISTANCE FROM SOURCE (cm)

Fig. 26. Thermal Shield Measurements -~ Gamma-Ray Dose Beyond Six 2,22-cm
Iron Siabs (Four Slabs Separated by 1.2 cm of Water, Two Slabs Effecting a
Pressure Shell) with Various Water-Reflector Bistances.

60
GAMMA DOSE {mr/hr)

N

9
W

R

10

RS

20

Fig.

27,

FOR PERIOD ENDING JUNE 19,

1952

Sl
DWG. 14715R1

 

40 60 80 100 120

Z, DISTANCE FROM SOURCE (cm)

’

Gamma-Ray Dose Beyond Water Shield.

140

61
ANP PROJECT QUARTERLY PROGRESS REPORT

62

DWG. 147 70R2

  

 

10° T
5 [ N
2
SIX SLABS
0P T e N e
e T meEVEN oL AR TN SN
t ______
Eosb b NN e
we T T T TR HT Ql ARG T e NN
w
o ..............
[
a P L NN N
b
=
<I
@]
10 S S——
5 ....................
> IRON TANK, 0.32-cm WALLS | | =
~ ™~
v |
{ L5 SIX TO EIGHT 2.22-cm IRON SLABS SEPARATED BY— oo
" 1.3 ¢cm OF BORATED WATER (1% BORON SOLUTION)
e E e~ N -
5 z oo T L
= DY b
2 _ 2 I
soet 3 |
SOURCE = - -
ot oo |7 T BORATED WATER
a0 o T

Q ° o Q| - -
—~10~!

O 20

40 60 80 100 120 140
Z2, DISTANCE FROM SQURCE (cm)

Fig. 28. Gamma-Ray Dose Beyond Iron-~Borated Water Thermal Shield with

20-cm Water Reflector.
FOR PERIOD ENDING JUNE 16, 1952

DWG. 14905R1

FOUR 2.22-cm IRON SLABS SEPARATED BY
1.3 cm OF BORATED WATER

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

59 3 a3 &

Ny ('_i) o M M
S et ==

| T LT z . \_fi

il — g:%fi% I

NN
ONAN —=f-BORATED WATER

= |-

SOURCEX \A’ WATER IN ALUMINUM
TANKS

 

 

“-IRON TANK, Q.3-cm WALLS —
FOUR 2.22-cm IRON SLABS SEPARATED

BY 1.3 cm OF BORATED WATER-

TWO 2.22~-cm IRON SLABS (ADJACENT)

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

§o S { o o
Ny O aJ nl”) wn
T UNONNE
VN
ol NUNA. . |
\ /TONON
o\ A - — IANANZANM—, —
SOUF{CEl WATER IN ALUMINUM ~BORATED WATER
TANKS IN IRON TANK

“~IRON TANK,C.3-cm WALLS

FOUR 2.22-¢m IRON SLLABS SEPARATED

BY 1.3 cm OF BORATED WATER

TWO 2.22-cm IRON SLABS

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

— ( ADJACENT)
5o 0 3 5 5
N O -— 0 'R‘i‘ w
¢ o == o~ */g"-- —.#IT.—_ & "?/I;'""Q_,f:’, ¢ :_“‘“—1’— —_
— E T =oNONNE= =
ol é§é§é i
Sl
7z
1 %’“ w;\ 7/7 ? Z%fi% A
SOURCE‘X ZWATEF.’ \Z—WATER’ IN ALUMINUM ZBORfl\TED WATER
TANKS IN IRON TANK
IRON TANK, 0.3-cm WALLS
Fig. 29. Schematic Diagrams for Gamma-Ray Measurements Beyond

Iron-Borated Water Thermal Shield.

See Fig. 30 for plotted data.

63
ANP PROJECT QUARTERLY PROGRESS REPORT

4 Fe, 20.0-cm
WATER REFLECTOR

c

o

£

1

w

< Fe, 20.0-cm WATER REFLECTOR;
< PRESSURE SHELL
=

=

<

o

6 Fe, 30.5-cm WATER REFECTOR, PRESSURE

0 20 40 60 80 100
z, DISTANCE FROM SOURCE {cm)

Fig. 30. Gamma-Ray Dose Beyond Iron-Borated Water Thermal
Solution). See Fig. 29 for schematic diagrams.

64

 

DWG. 14906R1

120 140

Shield (1% Boron
o

U

{mr/hr)

GAMMA DOSE
N

10

10

FOR PERIOD ENDING JUNE 10, 1952

4 0-cm WATER REFLECTCR

REFLECTOR
(cm)

DWG. 14694R1

Z MEASUREMENT
TO BACK OF

LAST SLLAB
(cm)

 

GAMMA DOSE VS. REFLECTOR
THICKNESS AT 2=79.3 ¢m

WATER REFLECTOR (4.0 TO 24.0 cm)
FOUR 2.22-cm IRON SLABS (~ 0.3 cm OF WATER BETWEEN SLABS)

 

20 40 60 80 100

z, DISTANCE FROM SQURCE (cm)

13.7
18.¢
23.7
33.8

120 140

31. Gamma-Ray Dose Beyond Solid TIron Thermal Shield.

65
ANP PROJECT QUARTERLY PROGRESS REPORT

DWG. 14914R! -

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

  

 

 

 

 

 

  
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

    
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

10° N T T T e
5 ]
2 | -
a ]
10 ]
5 [710.0-cm, AR Y7 e ]
18.0-cm, CO—=> NN N | .
i
15.0-cm, AIR ——XQ\
2
~
3 . .29:07em, Co——= "N\ N\ %rfem AR
10 RN NN N T T T
= T .
N _ AN
g 5\\ """"""" . 34.8-cm, CO, \
O \
< 2 ‘H% e e
z GAMMA DOSE_VS. REFLECTO
I THICKNESS AT 7z =79.3 ¢m
© 102 TR J\ . !_
| WATER z MEASUREMENT
7| REFLECTOR TO BACK OF
5 ——|  {cm) LAST SLAB
eeerenrensd {cm)
0.7 4.5
) 5.0 8.8
- 10.0 13.8
15.0 8.8
25.0 28.8
10 34.8 386 —- ]
T T ] T T o
. | jWATER REFLECTOR (07 TO 35 cm)———
| / (ONE 3.8-cm LEAD SLAB | S . ]
gfiat' :2 : ;7_77_.3 """""""""" 1 —J
SOURCE |~~~ .
© 060 ‘:O e — |

0 20 40 60 8O 100 120 140
Z, DISTANCE FROM SOURCE (cm)

Fig. 32. Thermal Shield Measurements -~ Gamma-Bay Dose Beyond One
3.8-cm Lead Slab with Various Water-Reflector Thicknesses.

66
b'&

{0

9 [5.0-cm, AIR

 

FOR PERIOD ENDING JUNE 10, 1952

CWG. 14915R1

WATER Z MEASUREMENT
REFLECTOR TO BACK OF

{cm) LAST SLAR
(cm) %

 

8.7

. 13.1
10, 18.1
15. 23.4
25.0 33.0
348 42.7

% ACCOUNTS FOR WATER GAPS

10.0~cm, AIR 0.7-cm WATER REFLECTOR,
AIR-FILLED ION CHAMBER
15.0-cm, AIR
WATER
15.0-cm
3 9
10
£
S~
£ 5
3)) 25.0-cm, CO,
o
0
b 2
=
3
& 5 34.8-cm, CO
10
5
GAMMA DOSE VS, REFLECTOR
THICKNESS AT 2=79.3 cm
2
10
5 WATER REFLECTOR (0.7 TO 34.8 cm )
SOURCE TWO 3.8-cm LEAD SLABS
<@
0 20 40 60 80 100 120 140
z, DISTANCE FROM SOURCE {cm)
Fig., 33. Thermal Shield Measurements - Gamma-Ray Dose Beyound Two

3.8-cm Lead Slabs with Various Water-Reflector Thicknesses.

67
ANP PROJECT QUARTERLY PROGRESS REPORT

68

10

   

{mr/hr)

GAMMA DOSE

10 t=——— THICKNESS AT 2= ?93cm .....
I ,,,,,, l
e —TWATER REFLECTOR (O? TO 348cm) -
>
[ 7 THREE 38 cm LEAD SLABS
_ Y 7
g ¥ g o — ——=— -
oq°02 T T iAled
SOQURCE|
qao oo R R
-
0 20 40 60 80 100 120
z, DISTANCE FROM SOURCE {(cm)
Fig. 34. Thermal Shield Measurements - Gamma-Ray Dose Beyond

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

T 411

Ay
DWG. 149t6R1
X — O I
s N—f—— S R — WATER 7 MEASUREMENT [
X ~—— REFLECTOR TO BACK OF
\ ] (cm) LAST SLAB
N S e (cm) *
\
10.0-cm WATER REFLECTOR,
Y \« AIR-FILLED ION CHAMBER 0.7 131
. 5.0 17.6
A | 10.0 22.8
0.7-cm, AIR 15.0 28.4
25.0 38.2
oo X\ »\\ 1 .'50 C_T_. AR 34.8 47.3
X\ [ —F % ACCOUNTS FOR WATER GAPS

 

 

 

 

Q‘?‘\ _______ - -

 

 

 

 

 

 

—]

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

GAMMA DOSE VS. REFLECTOR

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

140

Three 3.8-cm Lead Slabs with Various Water-Reflector Thicknesses.
FOR PERIOD ENDING JUNE 10, 1952

8. DUCT TESTS

F. J.
A. Simon

Muckenthaler

C. E. Clifford
M. K, Hullings

Physics Division

Duct work in the Thermal Column¢1)
has continued to obtain further experi-
mental corroboration of the simplified
duct theory(?) of neutron transmission
through cylindrical air-filled ducts
in water. Measurements have been made
on ducts with two and three bends to
supplement the data previously reported
on straight ducts and ducts with one
bend. Some effort has been made to
determine what portion of the flux
measured by the counter 1s due to
neutrons suffering at least one col-
lision in the walls of a straight duct;
the measured contribution appears to
be somewhat higher than was predicted
theoretically, Agreement with the
theory has remained good; the
value of the constant A, con-
cerns the albedo at the corners and
other factors assumed to be constant,
had to be redetermined. '

however,
which

An extensive survey of the radiation
transmitted by the annular air ducts
designed by G.E. has been started. The
1id Tank will be used for this survey
for approximately six weeks.

DUCT TESTS IN THE THERMAL COLUMN

Straight Ducts. Various lengths of
a straight, 4 1/4-in.-d1a aluminum-wall
(1/8 in. thick) duct were measured. A
circular fission source (4 3/8 in. in
diameter) was used for all the 4 1/4-
in,-dia duct measurements. Typical
curves for the neutron transmission
measured along an extension of the
duct center lines are givenin Fig. 35.

 

(1)
Progress Report for Period Ending March 10,
ORNL- 1227, p. 79.

¢. E. Clifford and A. Simon, Simplified
Theory of Neutron Traensmission Through Air-
Filled Cylindrical Ducts in Water {in preparation).

Air‘craft Nuclear Propulsion Quarterly
1952,

o’
DUGT: 4 Yg—~in., , ALUMINUM WaLls {BENY DUCTS

5 HEVE TWD 7O FOUR £0~-in. STRAIGHT SECTVONSD)
¥ 2B-in. STRAIGHT SECTION

@ 4g-in. STRAIGHT SECTION
: + Bp-in STRAGHT SECTION
2 & 3-dog BEND
' 7 43-dey BEMD
08 0 66-deg BEND

¢ 9u-deq BEND
S TWO 45-deq BENDS
5 ferfoens ¢ THREE as-deq BENDS
~ SOURCE: CIRCULAR, 4%~ in. DIA

w04

o

[

RELATIVE THERMAL NEUTRON FLUX N WATER{ 125‘2-in. BF COUNTER RESPONSE, counts /min)

 

o2
0

© 20 30
z, COUNTER DISTANCE FROM END OF DUCT {cm)

Fig. 35. Neutron Traosmission
Through Water in Cylindrical Ducts
with variable Bends.

69
ANP PROJECT QUARTERLY PROGRESS REPORT

7O i005C

UNGCLASSIFIED

PHO

70

 

Y

~

 

E

|

i
90w

~

4"

o

,wfiifiAOO

 

 

 

SOURCE

s —— om:||||||L

and 45-deg Bends

Straight Sectionmns,

-in,

Configurations of 4 1/4-in.-ID Air Ducts, 20

36.
{Aluminum Walls).

Fig.
-

Ducts with Bends. Neutron measure-
ments were made on the 4 1/4-in.-~dia
ducts with one to three bends formed
by 20-1in. straight sections joined at
45 deg angles. Further experimental
data for the angular correlation were
obtained by using three ducts with
single sharp bends of various angles
(31, 65, and 90 deg). Typical duct
configurations are shown in Fig. 36.

To prevent neutrons from traveling
the full length of the ducts in air
without scattering at the bend, the
smallest angle of bend afforded more
than a one~-diameter displacement of
the center of the duct., Measurements
of the neutron transmission along the
duct center lines are given in Fig. 37.

Wall-Scattering Experiment. The
duct transmission theory that has been
used ignores all wall-scattering except
at the bends, so the attenuation of
each straight section 1s proportional
to the square of its length. If
single scattering in the walls of the

104 DWG. 15340

D N X W

DUCT: B-in. I0 BY 36 in, PLASTIC WALL

COUNTER: 8-in. BFy CHAMBER AT UPPER
END OF DUGCT ¢

COUNTER RESPONSE {counis/mind
A

 

103 ‘
20 an 60 80 100

HEIGHT OF WATER ABOVE SQURCE PLATE (cm)

Fig. 37. Measurement of Thermal-
Neutron Flux as a Function of the
Level of the Water Surrounding the
Duct. ’

FOR PERIOD ENDING JUNE 10, 1952

straight sections is considered, a
component attenuated: according to the
cube of the length should be added.
Therefore, an axperiment was carried
out to investigate the magnitude of
the inverse-cube component.

A 6~in.-dia by 36-1n. duct (plastic
wall) and a fission source {6 1in. in
diameter) were mounted in the Thermal
Column water tank so that the level of
the water surrounding the vertical
duct could be varied. A counter was
mounted 10 cm from the upper end of
the duct and surrounded by a small
volume of water, as shown in Fig. 38.
The neutron flux was then measured as
a function of the depth of water
around the duct, '

A second set of measurements was
made with the upper two-thirds of the
duct filled with water. This prevented

FHOTS 10048 §

 

Fig. 38. Duct Configuration for.
Wall-Scattering Experiment.
ANP PROJECT QUARTERLY PROGRESS REPORT

the neutrons that would ordinarily
travel the length of the duct 1n aar
from reaching the counter. The daf-
ference of the two curves should give
the flux contributed by scattering in
the walls.

According to calculations®?) the
ratio of the flux contributed to the
counter by the neutrons that make one
collision in the walls to those which
travel the full length of the duct
directly 1in air 1s approximately equal
to 7r /2L, where r 1s the radius of the
duct and L 1s the length. For a 6-in.-
dia by 36-in. duct the ratio should be
approximately 0.15, which 1s about a
factor of 3 lower than the ratio found
in the experiment. The curve in Fig.
37 shows that the "contribution 1s not
linear but peaks at the end of the
duct farthest from the source.

Comparison with Theory. The results
of the measurements with both the
straight and bent aluminum-wall ducts
indicate good agreement with the theory

(Fig. 39). 1If agreement were perfect,
all the curves 1n Fig. 39 would
coincide. The spread indicates the

errors in predictions of attenuations,
which in some cases are as much as 10.
A new value for A, the constant
independent of duct geometry, has
been obtained that gives a better
prediction of the transmitted dose.
The redetermination of A was necessary
hecause of the two- and three-bend
experimental results.

of the
wall-scattered components are higher
than was predicted by the theory. The
discrepancy is for the most part the
result of degradation in energy of the

Preliminary measurements

Since thermal measurements
the degraded neutrons were

neutrons.
were made,

72

counted efficiently. The position of
the counter was approximately 10 cm
from the end of the duct to corre-
spond to the 10-cm point used in the.
previous duct measurements., Calcu-
lations of the scattered components
for these points werealso high compared
with the theory. I1f, however, the
20-cm points of the previous duct
measurements are compared with the
theory, the agreement 1s markedly
improved, which is in accordance with
the above reasoning.

Except for the three points nearest
the source, the shape of the curve
shown in Fig. 37 1is about what would
be expected. Failure of the initial
points to agree with the theory is not
understood, and plans are being made
to repeat this experiment under
different conditions,

GE-ANP ANNULAR AIR DUCTS

Mockups of the inlet and outlet air
ducts that penetrate the shield in the
G-E"direct-cycle design have been
supplied by G.E. fora lLid Tank experi-
ment. Measurements will be made to
provide complete neutron and gamma-ray
isodose plots in the water shield sur-
rounding the duct. Such plots will
indicate points of excessive leakage
that may require additional shielding.

A schematic diagram of the outlet
air duct is shown in Fig. 40. The
steel section between the source and
the duct consists of alternate 1 1/2-
in. air and water layers perpendicular
to the source and simulates the
transition region (perforated re-
flector) in the reactor design. A
photograph of the inlet air duct and
transition region is shown in Fig, 41.
FOR PERIOD ENDING JUNE 10, 1952

103 OWG, 15344

DUCT: 4% —in. 1D, ALUMINUM WALLS (BENT DUCTS
HAVE TWO TO FOUR 20—in. STRAIGHT SECTIONS)

28-in. STRAIGHT SECTION

40-in. STRAIGHT SECTION

52—in. STRAIGHT SECTION

31—~deg BEND

- 45~-deg BEND

- 65-deg BEND

90-deg BEND

TWO 45-deg BENDS

THREE 45-deg BENDS

SOURCE: GIRCULAR, 43%gin. DIA

RELATIVE SOURCE STRENGTH (neutrons/cm? - sec)

4+ WX ood o

 

0 10 20 30
EQUIVALENT CENTIMETERS OF WATER FROM END Of DUCT

Fig. 39. Comparison of Calculated Effective Source for Cylindrical pucts in
Water,

73
ANP PROJECT QUARTERLY PROGRESS REPORT

DWG. 15105

 

 

 

 

 

 

 

 

 

 

  
  

 

 

 

 

 

 

 

 

  

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

  

p

 

 

 

 

 

| | MH )
. |

)

TRANSITION \<
SECTION

Fig. 40. Schematic Diagram of GE-ANP Mockup of Outlet Air Duct.

w7 TN

 

74
Sl

Fig.

41.

 

GE- ANP Mockup of Inlet Air Duct.

 

 

PHOTO 10115

f [.n

30 in.

‘0T ANNL ONIANAI doIddd Hod

¢Sol
ANP PROJECT QUARTERLY PROGRESS REPORT

9. NUCLEAR MEASUREMENTS
A, H. Snell, Physics Division

The total neutron cross section of
Li® has been measured up to 4 Mev on
the 5-Mev Van de Graaff. The measure-
ments show only one resonance, which
is at 270 kv, The time-of-flight
neutron spectrometer has been operating
satisfactorily with a full resolution
width of less than 1.2 usec/meter.

CROSS-SECTION MEASUREMENTS WITH
VAN DE GRAAFF ACCELERATOR

H. B. Willard, Physics Division

The total neutron cross section of
Li® has been measured on the 5-Mev
Van de Graaff by a transmission experi-
ment. In this experiment the lithium
sample was 89% Li®, but the resulting
data has been corrected for the 117.
The cross-section curve up to 4 Mev is
shown in Fig. 42. Only one resonance,
which is at 270 kv, was observed in
this energy range.

Preliminary measurements of the
fission cross section of U2%3* have

76

been made up to 4 Mev., In addition
detailed information on the T(p,y)He*
reaction, the B"(p,n)" angular distri-
bution, the C!¥(p,n)N'? yield, and- the
(p,n) thresholds in neon is reported
in the Physics Division quarterly
progress report, (1)

TIME-OF-FLIGHT SPECTROMETER
G. S. Pawlicka E. C. Smith

Physics Division

The neutron time-of-flight spec-
trometer has been operating satis-
factorily with a full resolution width
of less than 1.2 psec/meter., Isotopic
assignment of the levels of indium
has been reported.¢!? Preliminary
measurements indicate levels 1in copper
at 700 £ 100 and 2200 600 ev.
Measurements of the cross sections of
the uranium isotopes are in progress.

 

(I)Physics Division Quarterly Progress Report
for Pertiod Ending March 20, 1952, ORNL-1289

(in press),.

K
FOR PERIOD ENDING JUNE 16, 1952

AENED
OWG.15342

 

 

 

 

 

MEASURED BY A TRANSMISSION EXPERIMENT
USING 89 % Li® AND CORRECTING FOR Li'

 

 

 

{ barns)
o

 

Gy

 

 

 

| lll
|
2 m

i
I !l’l,n,l!,mn'

 

s ity
i ™ .’"llfllufimfifi,'fi*"'“*" =L
”,m’mmMWMH'l
lH'I'“l“[I“Il’l'l,IIIlllliiilldl!l'mi" In My
I:

 

 

 

 

 

 

 

 

0 . e
0 | 2 3 a
. NEUTRON ENERGY ( Mev)

Fig. 42. Total Neutron Cross Section of Li6.

77

.........................................................................................
 
SUMMARY AND INTRODUCTION

The research on high-temperature
liquids has been devoted almostentirely
to the development of a satisfactory
fluoride. fuel for the aircraft reactor
experiment {(sec. 10), The phase
diagrams of numerous binary, ternary,
and quaternary fluoride systems have
been examined, and a number of compo-
sitions in the NaF-ZrF -UF, and
NaF-KF-ZrF,-UF, systems with as much
as 4 mole % UF, have been shown to
have melting points around 500°C.
Although 1t 1s not certain that either
of these fluoride mixtures can be used
with the envisioned loading technique,
preliminary data are encouraging. A
substantial program of research and
pilot~scale production of liguid fuels
of haigh purity and a study of high-
temperature transfer methods have
recently been initiated.

The corrosion research effort was
divided among static and dynamic tests
of fluorides, hydroxides, and liquid
metals, with the fluoride corrosion
tests predominating (sec. 11). Static
tests and the modified dynamic tests
(seesaw tests) that have been developed
are useful for screeninga large number
of samples but are inadequate for
precisely predicting the corrosiveness
of a circulating ligquid. However,
the tests have shown that small
additions of sodium, potassium,
manganese, and calcium are beneficial
in minimizing fluoride corrosion.

Various fluoride mixtures were
circulated in thermal convection loops
during the past quarter. Stainless
steel loops generally plug after short
periods, whereas Inconel loops con-
sistently operate for upto 1000 hours,
Some reduction in dynamic fluoride

corrosion has been obtained as a

result of improved fluoride- and loop-
preparation techniques. Fluoride
attack on Inconel i1s believed to bhe
the result of chromium diffusion out
of the metal lattice. A considerable
program of corrosion research om both
hydroxides and fluorides has been
undertaken.

The metallurgical processes involved
in the construction and assembly of a
high-temperature reactor, including
fabrication of control rods, welding
and brazing, and fabrication of solid
fuel elements, are being successfully
developed (sec. 12). Brazing alloys
were tested for flowability, corrosion
resistance, and joint strength, and
the 60% Pd-40% Nialloy proved superior
in a fluoride environment. Powder
mixtures of B,C with Fe and with Al 0O,
are being hot pressed for the safety
and regulating rods, respectively, of
the ARE. Loose-powder sintering has
been investigated as a technique for
the fabrication of solid fuel elements.

Heat transfer and physical property
measurements on the various fluorides
together with some
measurements on hydroxides {(sec. 13).
Determinationsof viscosity, thermal
conductivity, density, heat capacity,
and vapor pressure have been made on
several fluoride mixtures. It as
apparent that the physical properties
of the ZrF -bearing fuels are compatible
with reactor design requirements. In
particular, the viscosity, which was
so high ‘in the NaF-BeF, -UF, fuel
mixture as to necessitate 1ts rejectiron,
is less than 10 centipoises for the
NabF-KF-Zr¥, -UF, fuel mixture at all
reactor temperatures. Mathematical
analyses pertaining to circulating
fuel heat transfer systems have been
developed. A measurement of the héat

have continued,

81
ANP PROJECT QUARTERLY PROGRESS REPORT

transfer coefficient of molten sodium
hydroxide showed 1t to be similar to
that of ordinary fluids rather than
liquid metals.

Pile irradiations of proposed fuel
mixtures 1n Inconel capsules comprised
the greater part of the effort on the
radiation damage program, although
measurements have continued on inpile
creep and thermal conductivity of
metals (sec. 14). Irradiation of
Inconel capsules containing the
NaF-KF-UF, fuel mixture at power

densities considerably 1in excess of
that to be found 1n the ABRE caused an
attack on the
similar capsules

rate of
However,

increased
container.
containing BeF,-bearing fuels showed
ne positive evidence of radiation-
induced corrosion. An increase in the
secondary creep rate of irradiated
Inconel similar to that previously
observed 1in nickel and stainless steel
was found. No detectable change was
observed in the thermal conductivity
of irradiated specimens of specially
heat-treated Inconel or nickel.

 

10.

CHEMISTRY OF HIGH-TEMPERATURE LIQUIDS

W. R. Grimes

Materials Chemistry Division

Researchon high-temperature liquids
has been concerned almost entirely
with their development for use as
fuels for an aircraft reactor. In
addition, some effort has been devoted
to the purificationof alkali hydroxides
and the determination of the high-
temperature properties of these
materials. The principal research
work has been the phase equilibrium
studies necessary to define the optimum
concentration of the fuels, and, 1in
conjunction with others, the de-
termination of the physical properties
and corrosiveness of these high-temper-
ature systems. It has also been
necessary to intensify the efforts to
identify chemical species in the
cooled melts by x-ray diffraction and
other techniques.

Research on liquid fuels is still
directed toward development of low-
melting-point solutions of UF, in
mixtures of permissible fluorides.
The experimental reactor (the ARE)
will require about 16 1lb of U?3% per
cubic foot of fuel solution, so 1t
will be necessary to incorporate about

4 mole % of UF4 1n the fluoride fuel.

82

A number of compositions 1in the
NaF-7ZrbF,-UF, and NaF-KF-ZrkF,-UF,
systems with as much as 4 mole % Uk,
have been shown to melt below 550°C.
In addition, it has been demonstrated
that the viscosity of typical mixtures
of these materials 1s sufficiently
low for successful operation (cf.,
13). For easy startup, 1t 1s
necessary to fill the reactor with a
dilute (subcritical) fuel and add
small increments of a concentrated
solution to bring the system to
criticality. This that
solutions of widely varying uranium
content and melting points considerably
below the operating range be available
and that no high-melting-point com-
pounds be formed at intermediate
concentrations. It is not yet certain,
although preliminary data are en-
couraging, that the NaF-KF-ZrF,-UF,
system meets this reguirement.

sec.

requires

A substantial program of research
and pilot-scale production of liguid
fuels of high purity and a study of

high~temperature transfer methods
have recently been 1nitiated. A
smaller program for the study of
.t

chemical reactions of high-temperature
fluorides and hydroxides is presently
concerned with reactions that may
help to explain the corrosion behavior
of these liquids. Toward this end, a
guenching apparatus for studying
solid phases in the NaF-BeF,-UF,
system has been placed in operation,
and x~ray examination of complex
mixtures in the NaF-KF-ZrF, system
has yielded valuable information as
to species present.

Experimental preparation of simu-
lated fuel for the cold critical
experiment has been continued. Speci-
fications for the powdered mixture
can be completed as soon as the
actual ARE fuel composition is fixed.

LOW-MELTING-POINT FLUORIDE FUEL SYSTEMS

I.. M. Bratcher R. E. Traber, Jr;
C. J. Barton ‘

Materials Chemistry Division

The high viscosity of the melts
has virtually eliminated the BeF,-
bearing salt mixtures from consideration
as ARE fuels, and the emphasis 1s now
being placed on systems containing

ZrF4.

In the previous report®!? it was
indicated, on the basis of very
limited data, that the addition of
1 to 5 mole % UF, to systems con-
taining ZrF, affected the melting
point only slightly. Further in-
vestigation has shown that the ad-
dition of 1 or 2 mole % UF, depresses
the melting point of most alkali
fluoride—-zirconium tetrafluoride
mixtures slightly but that further
additions produce an increase 1n
melting point.

 

(I)Aircraft Nucleaor Propulsion Project Quarterly
Progress RBeport for Period Ending March 10, 1952,
OBNL-1227, p. 101,

FOR PERIOD ENDING JUNE 10, 1952

The first separation of a solid
phase from these fused mixtures 1is
frequently accompanied by a small
thermal effect that makes detection
of melting points by thermal analysis
difficult. In several instances a
second break, probably corresponding
to a eutectic temperature, has been
mistaken for the melting point. Some
difficulty has been experienced with
poor reproducibility of data. This
may be due in part to sensitivity of
the melting point to variances in the
ZrF, concentration in the fused
mixtures. Loss of ZrF, may result
from both volatilization and con-
version to zirconium oxide., These
phenomena will be investigated in
detail as soon as possible.

NaF-KF-ZrF -UF,. Most of the
effort on ZrF,-bearing fuels has been
concentrated on the NaF-KF-ZrF, -UF,
system, since preliminary studies
indicated that it was most likely to
produce a suitable fuel. Heating
and cooling curves have been run on
a large number of mixtures containing
4 mole% UF, and varying concentrations
of the other three fluorides. The
rather small thermal effect mnoted
when the first solids crystallize
from these systems has encouraged
the use of larger samples (200 to
300 g) in these studies. Zirconium
tetrafluoride purified by vacuum
sublimation was used in all the ex-
periments reported.

The data in Table 3 show the ex-
pected melting points when 4 mole %
UF, is added to various mixtures 1in
the NalF-KF-ZrF, system. The data are
not sufficiently complete to state
the optimum fuel composition at
present, It 1s obvious, however,
that fuels with melting points below
550°C can be obtained over fairly
wide ranges of NaF-KF and ZrF, con-
centration. These studies are being
continued to define the optimum
concentration.

83
ANP PROJECT QUARTERLY PROGRESS REPORT

TABLE 3

Melting Points of NaF-KF-ZrF4-UF4
Mixtures Containing 4 mole 9 UF4

 

 

 

 

FIRST BREAK
COMPOSITION (mole %) | rpvbre)s TURE

NaF KF ZrF, (°c)
7.7 | 40.3 | 48.0 668

8.2 | 44.6 | 43.2 540

4.8 | 50.0 | 41.2 540

9.1 | 48.5 | 38.4 490 (?)
9.6 | 52.8 | 33.6 585
15.4 | 32.6 | 48.0 615
17.3 | 35.5 | 43.2 565
18.7 | 38.9 | 38.4 575
20.2 | 42.2 | 33.6 555
22.9 | 22.6 | 50.5 620
24.9 | 25.6 | 45.5 565
27.8 | 27.4 | 40.8 490
29.8 | 30.2 | 36.0 535
30.1 | 15.4 | 50.5 605
33.7 | 16.8 | 45.5 550
34.6 | 17.3 | 44.1 545
36.7 | 18.5 | 40.8 515
39.8 | 20.2 | 36.0 5490
36.5 | 13.5 | 46.0 540
52.8 9.6 | 33.6 545

 

 

 

 

The data in Table 4 show the
effect of large additions of UF, to
a single, ternary NaF-KF-ZrF, mixture.
It appears from these data that the
start-up operation could be safely
conducted 1f the temperature were
maintained at 700°C and if concen-
trations of more than 25 mole % UF,
were avoided at all times. In a
similar study UF, was added as NaF-KF-
UF, eutectic (46.5-26.0-27.5 mole %)
to a NaF-KF-ZrF, mixture (33-16.1-50.9
mole %) in several proportions. No
high melting points were observed at
the intermediate compositions. These
studies are encouraging, but they
must be repeated when the final fuel
composition 1s chosen.

Should 1t become
operate the ARE with a
content than now seems
the low-melting-point
position previously reported(?)
(4.9 mole % NaF, 51 mole % KF, 42.1
mole % ZrF,, and 2 mole % UF,) would
be gquite i1mportant. It does not
appear likely that compositions, of

possible to
lower uranium
to be needed,
(410°C) com-

 

(2)

Ibid., p. 102,

TABLE ¢4

Effect of UF4 on the Melting Point of 2 NaF-KF-ZrF4 Mixture

 

 

 

 

COMPOSITION {(mole %) BREAK TEMPERATURES*

NaF KF ZrF, UF, (°c)

36 18 46 0 440, 425, 395

34.6 17.3 44.1 4.0 545, 493, 412

32.4 16,2 41.4 10.0 553, 505, 465, 405
30.6 15.3 39.1 15.0 585, 575, 500, 395
28.8 14. 4 36.8 20.0 626, 580, 480, 400
27.0 13.5 34.5 25.0 663, 575, 495

 

 

 

 

 

*Highest break temperature considered most reliable indication

of melting point,
melting
in this

aily uranium concentration,
hbelow 400°C will be found

system.

KF=-2rF, -UF . Only a few com-
positions in the KF-ZrF,-UF, system,
all with 4 mole % UF,, have been
tested. Since all these mixtures
showed melting points at about 600°C
(about 150°C higher than the corre-
sponding KF-ZrF, binaries), it is not
likely that a satisfactory fuel
exists in this ternary system.

NaF-ZrF,-UF,. The data obtained
on the NaF-Zr[,-UF, system, shown 1in
Table 5, indicate that fuels con-
taining up to 4 mole % UF, and melting

TABLE 5

Melting Points of NaF-ZrF, Mixtures
Containing 4 mole % UF,

 

 

 

 

 

 

 

FOR PERIOD ENDING JUNE 10, 1952

around 500°C can be obtained with
this system.

It may prove feasible to start the
reactor with a suitable NaF-ZrF,
binary and build the fuel to criti-
cality by addition of the NaF-UF,
binary eutectic (27 mole % UF,;
melting point, 620°C). The difference
in behavior of the Nal-ZrF -UF, and
KF-ZxF,-UF, systems 1s certainly due
to the strong tendency of KF to form
stable, high-melting-point compounds
with UF,.

NaF-RbF-ZrF,-UF,. A few mixtures
have been investigated in the NaF-
RbF-ZrF,-UF, system to determine
whether substitution of RbF for KF
would result in lower melting points.
Data obtained with such mixtures
(Table 6) indicate that RbF has no
decided advantage over KF as a com-

 

 

 

 

ponent of a ZrF, fuel as far as
COMPOSITION (mole %) MELTING POINT mMelting point is concerned.
Na¥F ZrF, (°C)
Y ~y D D
44.0 &9 0 597 ANALYSES OF FLUORIDE COMPOUNDS
50.0 46.0 513 X-Ray Examination of Solid, Com-
46.0 50.0 505 plex Fluorides (P. A. Agron, Materials
Chemistry Division). To assist 1n
. . 1 ) - . .
52.8 43 2 52 the 1dentification of the solid,
57.0 39.0 335 complex fluorides encountered in phase
studies of fuel mixtures, the technique
TABLE 6
Effect of UF, on Melting Point of NaF-RbF~ZrF, Mixtures
COMPOSITION (mole %) FIRST BREAK TEMPERATURE
NaF RbF ZrF, UF, (°C)
35.0 20.0 45,0 0 445
34.3 19,6 44,1 2.0 480
34.0 19.4 43.6 3.0 540
33.6 19,2 43.2 4.0 575
10.90 45,0 45.0 0 445
9.6 43,2 43.2 4,0 485

 

 

 

 

 

85
ANP PROJECT QUARTERLY PROGRESS REPORT

of x-ray-diffraction examinatioen on a
high-angle spectrometer was adepted.
Analyses of the binary complexes
occurring 1n cooled melts from the
equilibrium phase studies indicate
the presence of the compounds and
polymorphic forms listed in Table 7.

The crystal structures of the two
polymorphic forms of K,ZrF, have not
vyet been established. An examination
of other concentrations in the KF-ZrF,
and the NaF-ZrF,

made.

systems 1s being

Spectrographic Analysis (Bussell
Baldock, Stable Isotope Hesecarch and
Production Division). Efforts to
develop a mass spectrometer method
for 1nvestigation of aircraft fuels
has been delayed by the discovery
that the molecular fragmentation
patterns of the fuel i1ngredients were
more conmplex than was anticipated.
In particular, the UF, dissociation
pattern was largely masked by the
superposition of the dissociation
fragments from UF ;. It is now believed
that the UF; comes from UO,F,, which
is a contaminant in the UlF,, and
arrangements are being made to use
some 1sotopically enriched ingredients
to establish the origin and magnitude
of the UF,. The presence of UF, as
a contaminant in fuel mixtures could
give rise to considerable corrosion.

The 1nvestigations of elemental
compounds are summarized 1n the
following statements. Uranium tri-
fluoride dissociates at temperatures
above 700°C into UF, and uranium
metal, 1n agreement with findings
reported in the literature, but at
lower temperatures than heretofore
reported. No evidence of the subli-
mation of UF;, as such, could be
found at temperatures up to 750°C
with the mass spectrometer or up to
1000°C in sublimation tests. Even
UF, prepared by the reduction of

86

UF, with uranium metal shows
evidence of higher valence compounds
of uranium that give rise to UF, on
heating. All unheat-treated UF,
examined has shown pronounced evidence
of UF5 contamination when examined 1in
the mass spectrometer,

some

Study of Solid Phases in the
NaF-BeF,-UF, System (A.G. H. Andersen,
ANP Division, and C. J. Barton,

Materials Chemistry Division). A
study of the solid phases 1in the
NaF-BeF,-UF, system was initiated at
the time beryllium fuel was being
considered for use 1in the ARE. Since
a fuel composition in this system may
be suitable for use 1n a reactor
design not requiring a low viscosity
fuel, this study has been continued.

A furnace has been set up for the
heat treatment of fluorides or fluoride
mixtures sealed 1n quartz tubes under
a vacuum. The furnace 1s arranged
so that a number of samples can be
heat-treated simultaneously and
quenched by dropping into liquid
nitrogen or some other quenching
medium, The quartz capsules are
heated to a definite temperature
(£10°C) for intervals ranging from
16 to 60 hr and then quenched or
cooled slowly. The samples are then
examined with a petrographic micro-
scope and by x-ray diffraction.

In the UF,-BeF, system, three
crystallinephases have been 1dentified:
UF,, BelF,, and UO,. In addition, a
BeF, glass 1s present in mixtures
high in BeF,. This glass can be
converted into crystalline BeF, by
long heating at 300 to 600°C. The
UO, present 1s produced either by
hydrolysis or oxidation of the UF,
during the original sample preparation.
An unidentified phase appeared in a
90 mole % BeF,-10 mole % UF, mixture
heated at 300°C for 60 hours.
FOR PERIOD ENDING JUNE 10, 1952

TABLE 7

Phase Studies of Complex Fluoride Compounds

 

 

 

. - Lattice Dimensions(?)
A (b)
Compound Crystal Form _ (A) D, (g/ce)
a-K,UF, Cubic a = 9,21 4,12
a'-K,UF, Tetragonal ' a; = 9.20 4,13
ay, = 18.40
51*K2UF5 Hexagonal a, = 6,54 5.10
@, = 3.76
B,-K,UF, Hexagonal | a, = 6.53 4,77
_ ' a; = 4,04 |
KUF, Rhombohedral _ a = 9,387 ' 5. 38
a = 107° 157
KU,F, Orthorhombic | a, = B8.68 ' 6.49
' a, = 7,02
as = 11.4‘4‘
‘ Na,UF, Tetragonal - a, = 5.448 4,49
| ' | a, = 10.896 |
B,-Na,UF Hexagonal e, = 5.94 f 5.74
a, = 3.74
NaUF, Rhombohedral a = 9,08 | 5.81
' a = ]107° 56
K ZcF, Cubic a = 8.951(¢)
. {(d
A~K22rF6( )
B-K,ZrF (4]

 

 

 

 

(Q)W. H. Zachariasen, The Crystal Structure of Nn3UF7, AECD-1798 (Mar., 3, 1948); The Crystal Struc-
ture of 7-NaZUF6 yAECD-2089 (June 29, 1948); New Crystal Structure Results., Part J., AECD-2093 (Jpne
28, 1948); The Crystal Structure of Alpha-Phase Coapounds ApXFg and AXF,, AECD-2162 (July 19, 1948);
The Crystal Structure of Beta-Phase Compounds A XFg and AXF4, AECD-2163 (July 20, 1948).

(b)Theoretical density calenlated from Xerayediffraction data.

()G, C. Hampson and L. Pauling, J. An. Chem. Soc. 60, 2702 (1038).
(44

2ZrF6 is the Jow-temperature, stable polymorphic form, and B’Kzszfi is the high-temperature
form. : '

87
ANP PROJECT QUARTERLY PROGRESS REPORT

In the NaF-BelF, binary system,
only three samples have been prepared:
one with 40 mole % NaF and two with
50 mole % NaF. An unidentified phase,

NaBek,, appeared i1n these mixtures.
3 pp

Only one composition in the NaF-
BeF,-UF, system has been examined:
47 mole % NaF, 51 mole % BeF,, and
2 mole % UF, (composition 17). 1In
each sample of this mixture, held at
temperatures ranging from 250 to
500°C, either crystalline or glassy
BeF, has appeared. Some UO, was
present but UF, was not positively
identified. An unidentified phase
appeared 1n this mixture also.

Petrographic Examination of Fluo-
rides (1T. N. McVay, Consultant,
Metallurgy Division). Several hundred
examinations have been made of binary
mixtures of KF-ZrF,, NaF-ZrF., NaF-UF,,
Bel,-UF,, and K¥F-UF, as well as
ternary and quaternary mixtures of a
number of the fluoride components.
In addition, numerous samples of fuels
taken from capsules and loops have
been examined.

The optical properties of a number
of fluoride compounds not found in the
literature have been determined. All
index of refraction determinations
are probably accurate to 10.003.

Li,CrF,
Biaxial negative
2V = about 40 deg
Alpha = 1,444
Gamma = 1.464
K,CrF, Cubic
n = 1.422

Na,CrF, Cubic
n = 1.411

88

KUng Orthorhombic

Biaxial negative
2V = 10 deg
Alpha = 1.544
Gamma = 1.588

UF, Monoclinic
Biaxial negative
2V = 75 deg
Alpha = 1.500
Beta = 1.585
Gamma = 1.598

Anomalons interference colors

dark green

z

x = light green

The following systems and indices
were determined by previous 1n-
vestligators:

NazBeF4 Orthorhombic

Y || C, low birefringence

Average indices ~ 1.303

NaUF, Hexagonal

Uniaxial negative
0 = 1.520
E = 1.512

KUF, Hexagonal

Uniaxial negative

0 = 1.512

UF, Hexagonal

Low birefringence

n = 1,73

Anomalous interference colors
Bel

, Crystallized

Anisotropic
Low birefringence

n o= 1,328

The optical properties of several
other compounds have been determined,
but the compounds have not as vet been
identified.

Jt was found in the examination of
capsules containing fuel 21 that the
oxygen present reacted with the ZrF4
to form Zr0,. The evidence to date
is that the oxygen present will react
with Zr¥, to form well-crystallized
ZrQ,. Zirconium tetrafluoride appears
to be a more potent getter for oxygen
than UF,.

SIMULATED FUEL MIXTURE FOR COLD
CRITICAL EXPERIMENT

D. R. Cuneo L. G. Overholser
Materials Chemistry Division

The lack of complete definition of
the ARE fuel has prevented preparation
of final specifications for the
simulated fuel for the ARE critical
experiment. Since the ARE fuel will
probably contain ZrF,, NaF, KF, and
UF,, experimental studies have been
confined to mixtures containing these
fluorides.

Lack of an adequate supply of
hafnium-free ZrF, has made necessary
the substitution of Zr0, in the powder
mixture. Carbon is also added to
the mixture to compensate for the
di fference in moderating power of the
Zr(G, and ZrF,.

Calcination of the hafnium-free
Zr0, at 800°C is sufficient to reduce
the water content to below 0.1%.
Sodium fluoride has been shown to be
sufficiently pure and dry as received,

FOR PERIOD ENDING JUNE 10, 1952

but some samples of KF may require
vacuum drying.

By grinding UF,, Zr0,, NaF, KF, and
activated carbon in a small ball mill
and by packing the mixed powder imto
stalnless steel tubes, 1t has been
possible to ebtain a bulk density of
about 1.8. The powder so packed can
be made to show a uniform uranium
density of the proper value, and no
detectable segregation of the UF,
occurs with normal handling. By
use of reasonable precantiouns, such
as drying of the Zr0, and KF and
subsequent handling of the powders in
a dry box and a sealed ball mill,
the water content of the final mixture
can be kept below 0.2%. The hydrogen-
to-uranium ratio that may be tolerated
corresponds roughly te 0.5% H,0 1in
the mixture.

Grinding, blending, and canning of
the fuel and the preparation of the
large quantity of uranium-free base
material for the plenum chamber will
be started soon after the final ARE
fuel composition is chosen.

PREPARATION OF PURE HYDROXIDES

E. £. Ketchen .. G. Overholser
Materials Chemistry Division

The only experimental work on
moderator-coolants during the past
quarter has been concerned with
purification of several of the alkala
hydroxides. Sodium hydroxide with
less than 0.1% Na,CO; has been con-
sistently prepared, and a practical
technique for the removal of carbonate
and sodium i1ons from Li0OH has been
developed. Pure KOH has been prepared
both by removing the carbonate from
and by the re-
action of petassium and water. The
latter technique is expected to
ultimately produce the purest products;
however, the best material prepared

commercial material

89
ANP PROJECT QUARTERLY PROGRESS REPORT

to date by either method has con-
tained 0.10% K,O,.

Sodium Hydroxide. Because of
continued interest in purified NaOH,
additional batches have been prepared
by the method in which Na,CO,; is
removed by filtration from a 50%
aqueous solution of NaOH and the
clear supernatant liquid is dehydrated
under vacuum at 450°C. The material
has assayed 100.0% NaOH, and the
Na,CO; content has been less than
0.1% in all runs.

Lithiusm Hydroxide. A limited
amount of work has been devoted to
the purification of LiOH by using
the apparatus previously described(3)
for the purification of Ba(OH),.
Since the method proved to be ef-
fective in removing the two major
impurities, sodium and carbonate ions,
it appears to be a practical procedure
for the preparation of very pure
LiOH when the demand warrants such
production.

Potassium Hydroxide. 'The major
effort has again been devoted to the
preparation of pure KOH by removing
the carbonate as BaCO; in an aqueous
medium and dehydrating the filtrate
under vacuum at 475°C. Determination
of the solubility of BaCO,; as a
function of KOH concentration 1in-
dicated that the minimum carbonate
content could be obtained i1f a KOH
concentration of 45 to 50% were used.
Under these conditions 1t appeared
possible to prepare a product con-
taining carbonate equivalent to 0.07%
K,CO,. To date,
given a product corresponding to
0.10% K,CO; and containing 0.11% Ba.
The low value for carbonate is due
in part to a redesigned dehydration
vessel that eliminates contamination

only one run has

 

(3)

Aircraft Nuclear Propulsion Project Quarterly
Progress Report for Period Ending December 10,
1951, ORNL-1170, p, 84,

90

of the KOH by atmospheric CO,; this
preparation probably approaches the
lowest value possible by this method.
Substitution of Ca(OH), for Ba(OH),
to precipitate the carbonate proved
unsatisfactory.

One run has been made in which
pure potassium was reacted with excess
water and the resultant solution of
KOH was dehydrated by heating under
vacuum at 500°C. The apparatus used
allowed very slow addition of water
and removal of both hydrogen and
heat, thus the reaction occurred under
carefully controlled conditions. The
apparatus was design so that the
dehydration could be accomplished 1n
the same vessel in which the water
reacted with the potassium. Material
from this first runm contained 0.10%
K,CO;. A vacuum dry box 1n which a
very pure atmosphere can be attained
will be used in future work, and the
dehydration system will be modified
to improve its vacuum performance.
With these refinements it 1s expected
that a purer product will be obtained.

COOLANT DEVELOPMENT

I.. M. Bratcher R. E. Traber, Jr.
C. J. Barton

Materials Chemistry Division

The studies of uranium-free fluo-
ride systems to determine thear
possibilities as coolants as well as
fuel solvents have been confined to
systems containing ZnF, and ZrF, for
which preliminary data have been
presented previously.(*%) It now
appears that the previous data on
systems containing ZrF, was influenced
by the presence of Zr0, or ZrOF, 1in
the samples studied. Sublimed ZrF,
of high purity became available 1in
quantity during the past qguarter and
has been used in all the studies

 

4)op. cie., ORNL-1227, p. 104,
reported here. The simpler thermal
data obtained with the highly pure
material have helped to clarify the
phase relationships to some extent.
It should be emphasized, however, that
the applicationof the thermal analysis
technique to the study of alkali
fluoride-zirconium fluoride systems
has not yet produced completely
satisfactory phase equilibrium dia-
grams. The diagrams presented here
should be regarded as tentative and
subject to revision after further
study. Thermal studies are being
supplemented by examination of solid
phases by means of the petrographic
microscope and x-ray-diffraction
equipment. The results of these
studies are given elsewhere in this
report. '

FOR PERIOD ENDING JUNE 10, 1952

The two- and three-componentsystems
containing ZnF, have not shown melting
points sufficiently low to be of
interest at present; however, they
may be of value as constituents of
more complex mixtures. :

NaF-2ZrF,. Cooling curves for the
NaF-ZrF, system have been obtained
from a large number of mixtures con-
taining from 5 te 60 mole % ZrF,,
but the data obtained do not give a
satisfactory phase diagram. The
compound NaZrF, melts at 510 * 10°C
and thereis a eutectic at approximately
43 mole% ZrF, that melts at 480 t 10°C,

KF-ZrF,. Thermal breaks for KF-ZrF,

mixtures containing from 5 to 65
mole % ZrF, are shown in Fig. 43.

OWG. 14623

 

 

 

 

 

 

 

 

 

 

l l | ] T 1 :
KstF?
200 e
800 -—
©
2.
w 7Q0 .
4ol
>
—
<
1
i
=
= 600 —
-
500 —
4 00— .
s00 L | | | L
KF 10 20 30 40 . 50 60 70 80 30 .ZrF4

ZrF (mole %)

Fig. 43.

The System KF-ZrF4.

91
ANP PROJECT QUARTERLY PROGRESS REPFORT

The melting point of the eutectic
at 13 mole % ZrF, (765°C) and of the
compound K§ZrF7 (915°C) were reported
earlier. () The data indicate ad-
ditional eutectics at 42 and 55 mole
% ZrF, with melting points at 430 and
440°C, respectively. The compound
KZrF, appears to melt congruently at
445°C. The compound K,Zrk, probably
exists, but efforts to determine 1its
melting point with fused mixtures of
KF and ZrF, and with commercially
obtained material have been un-
snccessful. Data on mixtures con-
taining more than 65 mole % ZrF, are
unreliable because of the volatility

of ZrF,.

 

S rhia., p. 105,

RbF-Zr¥,. The phase equilibrium
diagram for the RbF-ZrF, system 1s
shown in Fig. 44. The melting point
of the eutectic at 6 mole % ZrF, 1is
725°C and that of the compound Rb;ZrF,
is 880°C. The eutectic at about 40
mole % ZrF, appears to melt at 400+ 10°C.
There 1s probably a compound RbZrF,
melting at 425 + 10°C. The data at
55 and 60 mole % ZrF, do not show
whether the compound melts congruently.
There 1is some evidence for the ex-
istence of the Rb,ZrF. compound, but,
as 1n the case of K,ZrF., 1t was not
possible to determine its melting
point by the conventional thermal
analysis techniques with mixtures of

RbF and ZrF4.

NaF-KF-ZrF,. Tentative contour
lines for the NaF-KF-ZrF, system

-
DWG.14624

 

1000 [ ]

S00

700

TEMPERATURE {°C)

500

400

 

|

 

 

 

 

200 | | | |

 

 

RbF 10 20 30 40

50 60 70 80 90 ZrFy

Zrfg {mole To)

rig. 44.

62

The System RbF-ZrF4.
are shown in Fig. 45. There appears
to be a low-melting region between
40 and 50 mole % ZrF, that extends
almost all the way across the diagram.
The lowest melting region lies close
to the KF-ZrF, eutectic. The data
give no definite indication of ternary
compound formation. ‘No attempt has
been made to obtain data on mixtures
containing more than 55 mole % ZrF,
because of the high volatility of ZrF,.

 

FOR PERIOD ENDING JUNE 10, 1952

NaF-RbF-ZrF,. The tentative
contours for the NaF-RbF-ZrF, system,
shown in Fig. 46, differ to a con-
siderable extent from those for the
NaF-KF-ZrF, system. The lowest melting
region appears to occur at higher
ZrF, concentrations except for the
region very close to the RbF-ZrF,
eutectic. On the basis of the data
available at this time, this system
shows no definite advantage over the

DWG. 14622R1

 

 

Na¥F

 

 

995°C

Fig. 45.

KF
850°C

Tentative Contours for the System NaF-KF—ZrF4.

- 93
ANP PROJECT QUARTERLY PROGRESS REPORT

NaF-KF-ZrF, system, and no further
work on it 1s contemplated at present.

Na¥-ZnF,. The equilibrium diagram
for the NaF-ZnkF, system is shown in
Fig. 47. Only one compound is formed
in this system: NaZnF;, which has a
melting point of 748 + 10°C. The two
eutectics are at approximately 32.5
and 69 mole % ZnF, and melt at 640 and
685°C, respectively.

Zrky

  

NoZrFS

INDETERMINATE

KF-ZnF,. The data for the KF-ZnF,
binary system, as shown in Faig. 48
give some evidence for the formation

of K,ZnF, and Kink,. The KZnF,
melts congruently at 850 + 10°C, and
the K,ZnF, appears to melt (with

at 720 + 10°C. The

two eutectics at about 21 and 80
mole % ZnF, melt at 670 and 740°C,

respectively.

decomposition)

DWG. 15346

  

 

510°C \ u ' »

e,

—

‘ = 600
N

NaF

\\\\\Qk\J

 

\

 

RbF

 

995°C

Fig. 46.

94

795°C

Tentative Contours for the System NaF-RhF-ZrF4.
FOR PERIOD ENDING JUNE 10, 1952

Owe. 14627

 

 

 

 

 

 

 

 

1000 1 T I [ l I
®
300 +—
o
06 BOOF— G
— NoZnFB
|31 | .
& :
%3 700 @ @ ®
= @
Lt
b o @ .. ® &
® @ © ® ©
6 00— _ ~
500/ . _‘
soo | { | | | L 1 | |
NaF 10 20 30 40 50 &0 70 80 20 ZnFy

ZnfF, (mole %)

Fig. 47. The System NaF-Zan.

OWG. 14626

 

£4000 { ~ T T | 1

200

800

 

 

 

 

 

 

 

 

 

S
N 700
x
D
&
o
i
.
£ 600 |- ' . ]
-
500 |— ' _ —
400 | : ' —
|
300 l I ! J + l L 1
KF 10 20 30 40 50 60 ° 70 80 90 ZnFy

_ . ZnFy (mole %)
Fig. 48. The System KF~ZnF2.

95
ANP PROJECT QUARTERLY PROGRESS REPORT

RbF-Zn¥F,. Strong evidence for the
existence of the Rb,ZnF, compound 1is
found in the equilibrium diagram for
the RbF-ZnF, system, as shown in
Fig. 49. This compound melts 1in-
congruently at 620 + 10°C, whereas
RbZnF, melts congruentlyat 730 * 10°C.
The lowest melting eutectic in this
system 1s at approximately 20 mole
% ZnF, and it melts at 595 %+ 10°C,
The other eutectic, which 1s near 70
mole % ZnF,, melts at 650 & 10 °C.

NaF-KF-ZnF,. Not enough datahave
been obtained for the NaF-KF-ZnF,
ternary system to define the location
of contour lines accurately. The
lowest melting point found was 626°C
for the mixture containing 25 mole
% IZnF,, 30 mole % NalF, and 45 mole?% KF.
The cooling curves do not a1ndicate
that any lower melting compositions
exist in the system, so no further

work on it is contemplated at this

time .

FUEL PREPARATION AND LIQUID HANDLING
F. F. Blankenship

Materials Chemistry Division

Because of the need for liguad
fuel samples for chemical examination,
physical property evaluation, and
corrosion testing, a program ifor
preparation of the molten liquids on
research and pilot-plant scales has
been started. The need for materials
of high purity for the final reactor
fuel, as well as the indications that
corrosion by the liquids 1s strongly
dependent on purity of the materials,
has directed the research program to
a study of feasible methods for

SR
DWG. 15349

 

1000
1

200

(°c)

=
=

700

MPERATUR

TE

600

 

500 (—

 

 

 

|

 

 

 

 

400 L
RbF 10 20 30 40

50 60 - [4¢] 80

InF, (male %)

Fig. 49.

96

The System RbF-Zan.
preparation of very pure liquids and
methods for handling the liquids to
minimize contamination.

Preparation of pure ZrF, has been
accomplished by vacuum sublimation of
impure commercial preparations and
incompletely hydrofluorinated materi-
als available in this laboratory.
The preparation of this material will
be handled in the future by the Y-12
Production Division; a supply of up
to 50 lb per week of pure, sublimed
ZrF, will be available after about
June 15.

The preparation of pure liguid
fuels of various types has, apparently,
been accomplished on a research (5-1b
batch) scale. Installation of similar
e quipment to turn out 10- to 100-1b
batches for larger scale testing
should be complete by the time the
ZrF, production permits their use.

Zirconium Fluoride Production
(C. M. Blood, J. E. Eorgan, G. J.
Nessle, Materials Chemistry Division).
The sudden large demand for ZrF,,
which developed when mixtures con-
taining BeF, were shown to be un-
satisfactory, could not be met without
delay. Delivery from commercial
suppliers was slow, and their products
contained up to 20% of ZrO, and ZrOF,.

The commercial material has been
purified satisfactorily by vacuum
sublimation in nickel equipment
capable of handling 1 to 1.5 kg of
material per charge. The sublimers
are 4-1in. cylinders, 18 inches 1in
length, containing disk-shaped baffles
arranged to provide a tortuous path
for the vapor. An air-cooled finger
suspended from the gasketed lid
serves to condense the pure ZrF,

- Thermocouples and inert gas and vacuum

connections are introduced through
the laid.

The output of the sublimers has
increased, along with the supply of

FOR PERIOD ENDING JUNE 10, 1952

raw material, to about 10 kg per
week. The best index of purity of
the material has been examination of
the product with the petrographic
microscope to ascertain uniformity
and the presence of not more rhan
trace quantities of ZrD

With the equipment and charge
material available at present 1t
appears that the sublimation of 1500-g
portions is best conducted at 800°C
for about 4 hr at a pressure of 450
microus. Under these conditions
virtually all the ZrF, is volatilized
and about 80% of the material 1s
deposited on the cold finger. The
rest is deposited on the upper portion
of the cylinder walls and 1s added
to another batch to be sublimed.

Under optimum conditiens the product
is snow-white. However, a slightly
green coloration appears, which 1in-
creases with time, temperature, and
decreasing pressure. This discoloration
seems to be from nickel fluoraide
formed by attack by HF, which 1is
formed from the hydrolysis of ZrF, by
the trace of water it contaains.

Arrangements have been completed
for the Y-12 Production Division to
supply 50:1b per week of pure, sublimed
ZrF, beginning in mid-June. The
plant for hydrofluorination of hydrous
zirconium oxide and sublimation of
the product is scheduled to be finished
June 1. This plant, which should be
capable of double the design figure
1f necessary, is also adaptable to the
use of hafnium-free ZrO, when this
becomes desirable.

PREPARATION OF PURE FUEL MIXTURES

C. M. Blood A. J. Weinberger:
F. P. Boody G. J. Nessle

Materials Chemistry Division

Several impurities are known to be
present in the materials that are

97
ANP PROJECT QUARTERLY PROGRESS REPORT

incorporated in the fuel preparations.
The sodium and potassium fluorides
are hygroscopic and contain up to
0.4% H,0 and about 200 to 300 ppm
of sulfur, mostly as sulfate. The
ZrF, even after sublimation contains
traces of ZrO, and probably ZrOF,.
The commercial UF, contains small
amounts of water, up to 1% of hexa-
valent uranium probably as UO,F,, and
traces of UO,. Sulfur compounds are
known to be quite corrosive expecially
to nickel and its alloys and hexu-
valent uranium is a strong oxidant.
Water reacts at elevated temperatures
with UF, and ZrF, to yield U0, (and
Zr0,) and HF. Accordingly, a purifi-
cation technique has been developed
for the preparation of small batches
of fuel for laboratory research
purposes. This technique has been
adapted to equipment capable of the
production of 10- to 100-1b batches
(cf., "Fluoride Production™in sec. 3).

It appeared that a sequence of
operations that would reduce SO4= and
U02++ to lower valence states followed
by high-temperature hydrofluerination
of the liquid would be desirable. The
sulfur, which had been reduced to S°,
would be eliminated as H,5, and the
oxides and oxyfluorides of uranium and
zirconium formed by hydrolysis would
be reconverted to fluorides.

The apparatus used for this purpose
is shown schematically 1n Fig. 50.
The heated copper and titanium traps
for removing oxygen from the 1inert
gas and the cold trap assembly for
removing water from the inert gas and
hydrogen are not shown.

The fuel constituents are mixed as
powders and charged to the vessel.
The air is removed and an HF atmosphere
1s introduced by repeated evacuation
and flushing. The melting operation

UNCLASSIFIED
DWG.15350

 

GAGE

GAGE . S

 

 

 

 

 

 

 

 

 

 

 

     

 

 

 

;%6
g fl O:i"é 2
QlL VACUUM
TRAP PUMP
b U
X5

2} SODA-LIME
TRAPS

 

 

 

 

 

 

H,0 AND
LIQUID HF (
Dot
REACTOR RECEIVER
&
& T 5
< =
T
Fig. 50. Apparatus for Fuel Hydrofluorimation.

98
is then conducted under an HF atmosphere
to minimize the hydrolysis produced
from the small amounts of water
adsorbed on the powders. Hydrogen
is bubbled through the molten liquid
for 2 hrat 600°C to reduce the uranium
to the quadrivalent state. Hydro-
fluorination of the liquid 1s carried
out for at least 1 hr at 600°C or
higher to insure conversion of the
oxides. The HF is& removed by stripping
with an inert gas at 800°C and the
fuel is forced through a sintered-
nickel filter to the recetiver. After
the sample has cooled to room tempera-
ture under an inert gas the container
can be detached and sealed without
exposure of the material to an un-
controlled atmosphere.

The apparatus and all associated
lines exposedto HF at high temperature
are made of nickel. Lines carrying
HF at low temperatures are made of
copper, and copper 1s also used as
the gasket material. The valves
seats are Monel or Fluorothene. The
major difficulties at present are
those associated with maintaining a
high-temperature system (600 to 800°C)
that will contain HF without leaks.
The receiver must be connected and
disconnected for each run, and the
connections are somewhat vulnerable
to the extreme conditions of exposure.

Approximately twenty 2-kg batches
of fuel have been prepared during the
quarter. The present production rate
is 3 batchés per week.

The hydrogenation and hydrofluor-
ination treatment seemsto be effective
in removing sulfur compounds. It is
significant that hydrolysis is mini-
mized, and re-hydrofluorination seems
quite successful, since no UD, or
Zr(Q, has appeared in or on the nickel
filter.

Liquid Handling Equipment (C. M.
Blood, F. P. Boody, A. J. Weinberger,

FOR PERIOD ENDING JUNE 10, 1952

G. J. Nessle, Materials Chemistry
Division). Purification procedures
have been perfected to the point that
the handling of the fuel materials
after purification can be the major
source of contamination. However,
equipment has been developed to ac-
complish transfer of fuels without
exposure to contaminating atmospheres.

In this equipment, molten fluoride
from a large reservoir is transferred
by inert-gas pressure to a small
container equipped with an overflow
so that a sample of constant size 1is
caught. This sample 1isthen transferred
to the capsule or other receiver by
pressure of inert gas. There are no
valves in the ligquid lines and transfers
from the container to the metering
volume and thence to the receiver are
controlled by valves in the inert
gas—vacuum manifold. Except for the
gas and vacuum lines, the whole system
must be maintained at about 600°C
while transfers are 1n progress.
Access to vacuwnm, HF, and H, manifolds
is provided; the receiving container
can be hydrogenated, hydrofluorinated,
or otherwise treated as required.

Trials to date have shown that the
method is feasible and indicate that
the apparatus can be operated on a
routine basis after a few modifications
to eliminate some of the minor daffi-
culties. Tt 1s worthy of note that
apparatus of this type should be
quite useful for bringing the reactor
slowly to criticality by injection of
a concentrated solution of UF, in the
alkali fluorides. :

SOLUBILITY OF URANIUM IN
SODIUM CYANIDE(®)

Sodium cyanide has been proposed
for use as a fuel carrier because of
its stability at high temperatures

 

(6)op. cit., ORNL-1170, p. 103

99
ANP PROJECT QUARTERLY PROGRESS REPORT

and its nonoxidizing characteristics.
Accordingly, several tests have been
run 1n an effort to determine the
solubility of uranium in this cyanide.
It was considered necessary to filter
the melten NalCN after a soaking period
at high temperature to eliminate any
particles of uranium that might
possibly be present. For this reason
the test was conductedin the following
manner. Sodlum cyanide contalning
uranium turnings was sealed into the
lower end of the test container,
which consisted of a piece of 1/2-in.-
OD type~-1035 steel tubing with an
iron filter welded in at the center.
After holding for 100 hr at 816°C,
the tube was inverted and held foT
24 hr at 816°C to allow the molten
NaCN to filter through into the other

It 1is believed that the relatively
hi1gh percentage of uranium in the
first test might be due to a leak
between the filter that was welded
1nto the tube and the tube wall. A
small amount of free sodium was found
in the upper part of the tubes at the
conclusion of the tests. Since NaCN
1s hygroscopic, 1t is possible that
much of the uranium may have united
with the water that was present
instead of dissolving in the NaCN.
Therefore it is planned to run similar
tests with dehydrated NaCN to determine
whether the solubility of uranium can
be 1ncreased i1n this manner.

TABLE 8
Solubility of Uraniom in
Sodium Cyanide

 

 

 

 

 

 

 

end of the tube. When the filtering

period had ended, the tube was allowed TEST URANTUM (%) IRON (%)
to cool and the NaCN was removed and

analyzed for uranium. The analyses 1 0.026 0.01
that were reFeived from the;e tests 9 0.001 0.096
are listed in Table 8. Since the

test was run 1n a type-1035 steel 3 0.001 0.038
tube, an 1iron analysis was also made.

11. CORROSION RESEARCH

W. D. Manly, Metallurgy Division
W. R. Grimes, Materials Chemistry Division

H. W.

Dynamic corrosion tests in thermal
convection loops have been conducted
with uranium-bearing fluoride mixtures
and nonuranium-bearing fluoride
mixtures in an effort to find a
suitable container for the fluoride
fuels to be used in the aircraft
reactor experiment, Thirty loops have
been run with fluoride as a circulating
medium, The container materials tested
have included the 300- and 400-series
stainless steels, Nimonic, and Inconel.
The stainless steel loops operated
only a short period of time before
pPlugging, whereas the Inconel loops

100

Savage,

ANP Division

operated for 500 and 1000 hr success-
fully. However, the Inconel suffered
severe corrosion to a depth of 10 to
15 mils with the formation of voids.
The formation of these subsurface
holes at the hot zone was found to be
the result of chromium diffusion to
the molten bath., Lattice vacancies
resulting from such diffusion merely
"precipitate"” to form the observed
voids, Static and modified dynamic
tests are being made in a concentrated
effort to ascertain possible inhibitors
for fluoride corrusion. Additions of

sodium, potassium, manganese, and
calcium metals have been shown to be
guite beneficral in minimizing the
corrosive action of the fluorides on
the container material. Several static
tests on the compatibility of various
fluoride mixtures with beryllium oxide
and other ceramics have been performed.

Other static and modified dynamic
tests were conducted in an effort to
find a possible corrosion and/or
mass-transfer inhibitor for the metal-
hydroxide systems., Preliminary experi-
ments have shown that the corrosion and
mass transfer usually associated with
the metal-hydroxide systems can be
appreciably reduced by the removal of
oxygen from the system with purified
hydrogen. In the study of the reaction
products of container materials and
the hydroxides, a new compound has
been found that is believed to be
NaNiO,. This is the first time that
NaNi0O, has been shown to exist. A
similar compound, NaFeOZ, has also
been found as a reaction product of
iron with sodium hydroxide, but this
compound has been identified by
previous investigators. Two simplified
methods for studying dynamic corrosion
have been developed. Preliminary
tests using these methods for studying
the metal-hydroxide mass-transfer
phenomenon have shown that the methods
operate successfully.

In the belief that a better under-
standing of the hydroxide and fluoride
corrosion mechanism will ultimately
lead to the reduction 1f not the
elimination of corrosion, considerable
effort is being devoted to fundamental
corrosion studies. The work includes
not only the synthesis and determi-
nation of corrosion reaction products,
but also such problems as emf measure-
ments of mass-transportfphenomena and
determination of the free energy of
postulated reactions.

FOR PERIOD ENDING JUNE 10, 1952

STATIC CORROSION BY FLUORIDES

D. E. Vreeland R. B. Day
E. E. Hoffman .. D. Dyer
‘Metallurgy Division

Static corrosion tests, while not
conclusively indicative of the cor-
rosion resistance of a material under
dynamic conditions, are useful for the
preliminary screening of a Jarge number
of specimens. In particular, the
effect on corrosion of various additives
in a fluoride mixture and the effect
of fluoride corrosion of plated metals
have been examined by the static-
capsule technique. Neither the
additives nor the plating was signifi-
cantly beneficial, and when an imper-
fection existed 1n the plating the
resulting attack of the base metal
was quite severe. Measurements of the
static corrosion resistance of Inconel
and stainless steel have shown little
or no correlation with temperature

between 538 and 1000°C.

Effect of Additives. Static cor-
rosion tests of several materials in
fluoride mixture NaF-KF-LiF-UF,
(10.9-43.5-44.5-1.1 mole %) with
approximately 10 wt % additions of Zr,
Na, and U and also approximately 5wt %
additions of L1, K, Ca, Li, aund Mn
have been run for 100 hr at 816°C.
These additions appeared to have some
effect i1in inhibiting corrosion 1in
these tests. Inconel, type-321
stainless steel, and A nickel were the
metals tested., Specimens and tubes
from the tests with Zr and U additions
had developed surface layers. A
nickel tested with the 5% Na additive
also showed some evidence of developing
a surface layer. Inconel developed
surface layvers in the Li, Ca, and Ti
addition tests. Type-321 stainless
steel developed surface layers in the
Na and Ca addition tests. In those

101
ANP PROJECT QUARTERLY PROGRESS REPORT

tests in which no surface layers were
observed, i1t 1s, of course, possible
that a surface layer may have been
inadvertently removed during stripping.
test with type-309
the same fluoride

In a previous
stainless steel 1in

a surface

This layer

mixture with a Zr addition,
layer was also developed.
was examined by x-ray diffraction and
reported to be composed of UO, and
Zr0,. The resnlts of the most recent
tests are summarized in Table 9.

TABLE 9

Effect of vVvarious Additives on the Static Corrosion of Several Metals

 

 

 

by NaF-KF-LiF-UF, After 100 hr at 816°C
MATERI AL ADDITION METALLOGRAPHIC NOTES

Type-321 stainless steel 10% Zr 1/2-mil surface layer on specimen and tube, no
attack beneath layer

Inconel 10% &r 1 1/2-mi] surface layer on specimen and tube, sub-
surface voids to 1 mil under surface layer

A nickel 10% Zr 2 1/2-mil surface laysr, no attack under layer

Type-321 stainless steel i0% Na No attack

Incenel 10% Na No attack

A nickel 10% Na 1/2-mi] surface layer on tube only, no attack

Type-321 stainless steel 10% U 2- to 3-mil surface layer, no attack under layer

Inconel 10% U 1- to 3-mil surface layer, voids to 1 wmil under layer

A nickel 10% U 2-mi]l surface layer, mno attack under layer

Inconel 5% Li 1-mi] surface layer on specimen, tube has traces of
surface layer; specimen attacked less than 1/2 mil

Type-321 stainless steel 5% 1.1 No attack

Inconel 5% K No attack

Type-321 stainless steel 5% K No attack

Inconel 5% Ca 1/2-m1l surface layer on tube, no attack beneath
tube; specimen lost

Type-321 stainless steel 5% Ca Tube has surface layer 1/2 to 1 mil thick; specimen
was attacked and has a surface }layer abont 1/2 mil
in depth

Inconel 5% Ti Surface layer of 1/2 mi]l present, no attack beneath
Yayer

Type-321 stainless steel 5% Ti No attack

Inconel 5% Mn No attack

Type-321 stainless steel 5% Mn Subsurface voids to 1/2 mil on specimen and 1 mil on
tube

 

 

 

1062
Temperature Pependence. Temperature
dependence tests 1in NaF-KF—LiF»UF4
(10,9-43.5-44.5-1.1 mole %) have been
completed, Inconel and types-430,
-304, and -321 stainless steel have
been tested for 100 hr at 538, 704,
and 1000°C, The maximum penetrations
of both containers and specimens are
listed in Table 10. The results of
previous tests of these materials at
816°C are also included for comparison.
It would appear from these results
that in statlc corrosion tests with
these materials there is little or no
effect on the extent of corrosioen that
can be traced to varying the tempera-
ture within the range mentioned above.
Apparently the phenomenon of sensiti-
zation occurring in type-304 stainless
steel and Inconel has little if any
effect on corrosion by these fluoride
mixtures under the testing conditions
employed. The temperature range for
sensitization is usually accepted as

400 to 850°C,

The first three test temperatures
in Table 10 are within the sensitization
range (538, 704, and 816°C). The
final test temperature of 1000°C is
well above the sensitization range, and

FOR PERIOD ENDING JUNE 10, 1952

yet no significant reduction of cor-
rosion can be noted.

Effect of Plating Metals. A speci-
men of type-304 stainless steel that
had been electroplated with approxi-
mately 8 mils of nickel was lightly
cut along one face with a hacksaw in
order to:expose the base metal, This
specimen was then tested inNaF-KF-LiF-
UF, (10.9-43.5-44.5-1.1 mole %) for
100 hr at 816°C under vacuum. The
nitckel plating was apparently un-
attacked, although many voids, which
may have occurred during plating, could
be noted, At the sawcut the type-304
stainless steel was more severely
attacked than any other steel corrosioen
specimen that had been tested in the
static fluoride mixtures. There was a
layer approximately 12 mils thick of
either corrosion product or unremoved
fluoride, beneath which was a region
of severe intergranular attack 15 mils
deep (Fig. 51). :

The results of this test might be
interpreted as giving some indication
that an electrochemicdl type of attack
had taken place. Tt 1s well known in
aqueous corrosion, in which it has been

TABLE 10

Depth of Attack of Metal Specimens in NaF-KF—LiF-UF4 for 100 hr at
Various Temperatures

 

 

 

 

 

DEPTH OF ATTACK (mils)
METAL 538°C 704°C  816°C 1000°C
Specimen Container | Specimen | Container | Specimen | Container |Specimen | Container
Inconel Slight Slight 4 1 1/2 1 1/2 3 ' 3 3
roughening{ roughening
Type-430 stain- <1 <1 1/2 1/2 /4 /4 Specimen | No attack
less steel lost
Type-304 stain-
less steel <1 <1 2 11/2 2 2 11/2 1
Type-321 stain- : _
less steel <1 <1 <1 <1 1/2 1/2 11/2 11/2

 

 

 

 

 

 

 

 

 

 

 

103
ANP PROJECT QUARTERLY

 

Fig.
Stainless Steel,

51, Nickel-Plated Type-304
with Exposed Base
Metal, Static Corrosion Tested in

NaF-KF-LiF-UF4 for 160 hr at 816°C.

established that electrochemical attack
takes place, that if a metal is coated
with a more noble metal and then the
coating is brcken and exposes a
relatively small area of the anodic
material beneath the cathodic coating,
the small exposed area will suffer from
accelerated corrosion attack. It would
appear that a similar process had
taken place in this test with molten
fluoride.

Effect of Cold Work. Previous
static corrosion tests in NaF-KF-LiF-
UF, (10.9-43.5-44.5-1.1 mole %) on
Inconel that had been cold worked
approximately 20% prior to testing
showed no i1ncrease 1n corrosion over
as-received material, Tests have now
been completed on Inconel with 52 and
also 74% cold working prior to test,.
There appeared to be no increase 1in
corrosion even after this severe cold
working. In these tests for 100 hr
at 816°C both the specimens and tubes
were attacked to a depth of approxi-
mately 1 mil -~ the tubing was not cold
worked,

Corrosion of Ceramic Materials

(C. R, Croft, N. V, Swmith, R, Meadows,
H. J. Buttram, Materials Chemistry

104

PROGRESS REPORT

A number of refractory
were tested 1in

Division).
ceramic materials
various fluoride mixtures. The main
purpose of this study was to find
container materials suitable for
electrochemical studies. In addition,
the behavior of beryllium oxide in
contact with several fluoride mixtures
was especially investigated because of
the importance of this material as the
moderator for the ARE,

The time of exposure for these
exploratery tests was reduced to 25 hr
from the usual 100 hours. Comparison
of results in NaF-KF-LiF-UF_ fuel and
in the corresponding nonuranium-bearing
mixture at 800°C showed that the attack
is heavier in the fuel since the UF,
appears to react with the oxides to
form a coating of UO,. In both
mixtures, vitrified beryllium oxide
was found to be the most satisfactory
material; 1t showed weight gains of
7% in the uranium-bearing mixtures and
1% in the uranium-free mixtures but

The materials
various grades

no dimensional changes.
tested included zircon,
of aluminum oxide, wagnesium oxide,
and sapphire, Also, a number of hot-
pressed refractory carbides were
tested in the uranium-bearing mixture.
Silicon carbide was penetrated, but
the carbides of titanium, tantalum,
and columbium did not show visible
signs of attack., The penetration of
silicon carbide might have been due to
the porosity of the sample.

A special recrystallized aluminum
oxide, Morganite, satisfactorily
resisted attack by the NaF-KF-LiF
mixture but gained over 90% when UF,
was present because of the formation
of a heavy layer of UO, on the surface.
A similar deposit but a much lower
weight gain (18%) occurred when a

NaF-Ber-UF4 fuel was used,

The resistance of vitrified beryllium
oxide to uranium-free ZrF,-bearing
mixtures was followed in a series of
experiments using exposure times up to
500 hours, The weight gains ranged
from 0.3% after 100 hr to over 11%
after 500 hours. X-ray-diffraction
studies showed a layer containing
considerable amounts of ZrQ, on the
surface of the specimen, whereas the
zirconium content of the liguid
decreased., It 1s possible that the
ZrO2 layer will protect the body of
material from further attack.

Beryllium oxide specimens tested in
a NaF-BeF, mixture at 800°C showed no
apparent attack aon the surface; the
specimens increased in weight by 0.7%
after 100 hr, 1.5% after 250 hr, and
8.7% after 500 hours.

STATIC CORROSION BY HYDROXIDES

D. C. Vreeland ‘R. B. Day
E. E. Hoffman L. D. Dyer
Metallurgy Division

Static screening tests have also
been conducted on the effect of various
additives on hydroxide corrosion.
Although there was some indication
that manganese addition had an in-
hibiting effect on corrosive attack,
the corrosion was still gquite severe,
The nickel plating of hydroxide cor-
rosion speclmens is not yet an entirely
reliable technique of reducing hy-
droxide corrosion, but it 1s most
effective when the plating lavyer is
thick, that is, much greater than 4
mils. *

Effect of Additives. There was
some indication in several tests that
when Inconel is tested inNaOH additions
of manganese may be beneficial inm
inhibiting attack. TInconel, when
tested 1n regular static corrosion
tests with NaOH, is severely attacked;
the entire thickness of a 35 mil

FOR PERIOD ENDING JUNE 10, 1952

specimen is often affected. In order
to check the possible inhibiting
charactéristics of manganese, static
corrosion tests were run with approxi-
mately 4% and also 1 1/2% additions of
manganese powder and a 1 1/2% addition
of electrolytic manganese. The
manganese additions did seem to have
some inhibiting effect on corrosive
attack, with the manganese powder
additions being the most effective;
however, corrosion was still quite
severe even with these additions,

A nickel-zirconium alloy (1/4%
zirconium) was also tested in NaQH.
This alloy has attracted attention
because of its better physical proper-
ties at high temperatures as compared
with pure nickel, The results of the
static corrosion test {110 hr, 816°C,
vacuum) were encouraging. The only
attack that could be noted was 1/2 mil
of light intergranular penetration.

Effect of Plating Metals. A series
of tests in NaOH of some specimens of
Inconel and Inconel X that were plated
with different thicknesses of nickel
by the International Nickel Company
has been completed. All the specimens
with platings under 4 mils were
severely attacked, but several of the
specimens with thicker platings were
unattacked even though the nickel
plating appeared to have many voids
(Fig. 52). Nickel plating as a means
of protecting materials against attack
by hydroxides does not appear to he
entirely dependable., When protection
1s afforded, 1t is complete and ex-
cellent; but when this protection 1s

not complete, attack 1s very severe.
The plated specimens used were vacuum-

diffusion heat treated before testing
for 100 hr at 816°C under vacuum. A
summary of metallographic observations
15 presented in Table 11.

105
ANP PROJECT QUARTERLY PROGRESS REPORT

t UMCLASSIFIED §
g Y-5743

 

(a) 2 mils of nickel plate on Inconel X. 75X,

| UNCLASSI F{ED
| v-.s747

L] :1_9* ‘l*

 

(b) 12 mils of nickel plate on Inconel, 100X.
rig. 52. Effect of Thickness of Nickel Plate on Static Corrosion by Sodium
Hydrexide after 100 hr at 816°C.

106
FOR PERIOD ENDING JUNE 10, 1952

TABLE 11

gorrosion of Nickel-Plated Materials Tested in NaOH at 816°C for 100 Hours

Base metal:

Inconel

 

 

THICKNESS OF
NICKEL PLATE (mils)

METALLOGRAPHIC NOTES

 

1 45 mils of oxide bet%een
creased from 241 to 180

2 Most of specimen covered

no attack on plating or

plating and specimen; unattacked material de-

mils

with 10 to 35 mils of oxide on base material;

base material along part of specimen

3 Corrosion product up to 50 mils between plating and specimen

4 Base material. unattacked; no thickness change; some voids apparent in

nickel plating

8 No apparent attack on plating or base material; no thickness change
12 No apparent attack; some voids apparent in nickel plating
26 51 mils of oxide on base material; unattacked base material decreased

 

frofi 241 to 182 mils; many voids in plating

 

DYNAMIC CORROSION BY LIQUID METALS

A. D. Brasunas L. S. Richardson
Metallurgy Division

Two tests on the dynamic corrosion
of liquid metals were performed with
the seesaw corrosion apparatus previ-
ously described, (!} The tests were
made at a 1500°F hot-zone temperature
and a 950°F cold-zone temperature; the
specimens were made of type-310 stain-
less steel. One specimen contained
lead and the other contained lead plus
3% sodium. After 187 hr (50,000 cycles)
the specimens were sectioned for
examination., Metallographic exami-~
nation of the specimens showed inter-
granular corrosion at the hot zone and
metal-crystal deposition at the cold

 

(I)AircraftfluclearPropulsifin Project Quarterly
Progress Report for Period Ending March 10, 1952,
ORNL-1227, p. 120.

zone (Fig. 53). Both corrosion 'and
mass transfer were less severe in the
test specimen containing 3% sodium.
A similar test is being made with an
appreciably greater addition of sodium.

DYNAMIC CORROSION BY FLUORIDES

Corrosion by Fluorides in Seesaw
Tests (C. R. Croft, N. V. Smith, R.
Meadows, H. J. Buttram, Materials
Chemistry Division; A. D. Brasunas,
.. S. Richardson, Metallurgy Division).
All studies performed to date have
justified the former conclusion that
when carefully prepared fluoride
preparations are tested in sealed
capsules the corrosion on Inconel and
stainless steel can be tolerated at
temperatures above 1500°F. Recent
experiments have also verified that
when the fuel contains less than about
10 mole % UF, pretreatment of the

107
ANP PROJECT QUARTERLY PROGRESS REPORT

N 1 TR, TR s
N -
D

P UNCLASSIFIE

 

(a) Lead with no sodium addition. 150X,

. : 3 A
UNCLASSIFIED E

Y-6592

UNCLASS | F I
Y.6639

 

. s
o

(b) Lead with 3% sodium addition. 250X.
Cold End (950°F) Hot End (1500°F)

Fig. 53. Effect of Sodium Additive on the Corrosion of Type-310 Stainless
Steel by Lead After 187 hr at 1500°F in a Seesaw Test.

108
stainless steel or Inconel does not
appreciably improve the corrosion
behavior. :

Addition of UO, to various fuel
preparations has not resulted in in-
creased corrosion by the mixtures;
addition of hexavalent uranium as U,0,
or as UOiFg, however, results 1in
considerably increased corrosion. 1t
still appears likely, therefore, that
oxidation of the fuel components during
preparation 1s more important than
hydrolysis in so far as the corrosion
behavior is concerned.

Addition of sulfate, as Na2804, to
various fuel samples has resulted in
large increases The
increase ln penetration and weight loss
of the specimens seems to be pro-
portional to the sulfate added. Tt
does not appear likely, however, that
the small gquantities of sulfate
present in the commercial fluorides
used can be responsible for the cor-
rosion routinely observed.

1n COrTOS10N.

The evidence to date justifies the
conclusion that the ZrF,-bearing fuel
mixtures are somewhat less corrosive
to Inconel and stainless steel than
any of the mixtures previously tested.
Static testing has indicated that
negligible corrosion results when
these materials are tested in stainless
steel and Inconel. Studies with the
tilting furnace seem to indicate the
superiority of Inconel to stainless
steel as a container material for the
zirconium-bearing fuels. These experi-
ments are being continued and extended
to recently developed fuel compositions
as raptdly as possible,

Seesaw tests with NaF-KF-LiF-UF,
(10.9-43.5-44.5-1.1 mole %) fuel have
been made with type-304 and type-304
ELC stainless steel for 190 hr with a
hot-zone temperature of 715°C and a
cold-zone temperature of 575°C. The
tests revealed no difference 1in

FOR PERIOD ENDING JUNE 10, 1952

behavior between these two specimens,
and confirmed the theory that the
carbon content plays an exceedingly
small role, i1f any, in fluoride cor-
rosion under these conditions. The
hot zone revealed voids to a depth of
4 mils; whereas a film slightly less
than 1 mil thick was observed on the
cold-zone surface. Similar tests with
Inconel have repeatedly shown about
5 mils of voids in the hot zone and a
small amount of crystal deposition at
the cold zone. Various additions are
being made to the fuel in an effort to

minimize or eliminate this attack.
Thus far, three additions, namely,
chromium, magnesium, and sodium, have
met with success i1n single tests.

These results should be confirmed by
either static or additional dynamic
tests. However, 1t 1s felt that
chromium additions should be expected
to be effective in suppressing void
formation, since it has been demon-
strated that the voids are caused by
chromium depletion.

Corrosion by Fluorides in Thermal
Convection Loops. (G. M. Adamson, K. W.
Reber, Metallurgy Division). The
programof dynamic testing with thermal
convection loops has received greater
during the past quarter.
zirconium fluoride

emphasis
Tests with new
fuel have produced optimistic results.
Although it appears that the fluoride
fuel NaF-KF-LiF-UF, (10.9-43.5-44.5-
1.1 mole %) will not be used as the
coolant, 1t is the only possible one
now available in large quantities, so
the preliminary work with it has been
continued, Considerable corrosion has
been measured with all container
materials in which this fluoride was
circulated. The same combination of
fluorides, but without the uranium, 1is
not nearly so corrosive, and the
addition of small amounts of NaK
reduces coOrrosion even more, The
attack, at least in Inconel, appears
to be a leaching out of the chromium

109
ANP PROJECT QUARTERLY PROGRESS REPORT

in the metal and a gathering of the
voilds,

Two Inconel loops and one type-316
stainless steel loopwere operated with
NabF-KF-ZrF,-UF, (4.8-50.1-41.3-3.8
mole %) at 1500°F. One Inconel loop
and the type-316 stainless steel loop
(No. 124) were cleaned by passing dry
hydrogen through them at 1950°F,
whereas the other Inconel loop was
degreased. In the degreased loop the
attack was very similar to that found
with the NaF-KF-LiF-UF, fuel but only
about half the number of pits was
present (Fig. 54). The depth of
pitting varied from 3 to 7 mils. In
the hydrogen-cleaned loops the number
of pits was reduced further and the
attack was more nearly confined to a
few grain boundaries; however, 1in a
few cases the attack extended ‘to a
depth of 13 mils (Fig. 54). The 1in-
crease in depth of attack 1s thought
to have been caused by the large grain
size produced by the hydrogen faring,
which would not be present with other
cleaning methods. Very thin deposits
were found i1in the cold legs of both
loops.

The other special coolant tested
was NaF-BeF,-UF, (47.0-51.0-2.0
mole %)}). This mixture was circulated
in two Inconel loops at 1500°F.
Intergranular pitting that varied from
heavy to moderate was found 1in the
hot leg. The attack varied in depth
from 6 to 13 mils. Thin deposits
were present in the cold legs of both
loops. Coolant samples from both
loops showed slightly lower uranium
concentration in all sections than
reported for the original batches.
This may be an indication that some
uranium precipitated from the system,

As mentioned above, since the
NafF-KF-LiF-UF, mixture was the only
fuel available 1in quantities, most of
the work has been done with 1t. The
data from the Jloops in which this

110

mixture has been circulated are tabu-
lated in the following tabkles. The
results from a few of these loops were
listed in the last report, but they
are repeated here for comparison.
Table 12 summarizes the results from
the Inconel loops, Table 13 those from
the 300-series stainless steel loops,
and Table 14 those from the remaining
loops. A study of the data presented
in these tables leads to the following
conclusions:

1. Tf NaF-KF-LiF-UF, 1s used as
the circulating media, all stainless
steel loops plug in a relatively short
time. The 400-series steel loops plug
in a shorter time than the 300 series
loops.

2. Inconel loops have not plugged
when any of the fluorides have been
con-

circulated in them; however,

siderable corrosion has been evident.

3. By comparing either Inconel or
stainless steel loops in which NaF-KF-
LiF has been circulated with those 1n
which NaF-KF-LiF-UF, has been circu-
lated, it is evident that the presently
available UF, accelerates both plugging
and corrosion.

4, When NaF-KF-LiF-UF, 1s circu-
lated in a loop in which chromium 1s
one of the alloying elements, the
chromium concentration of the mixture
increases and the iron concentration
decreases., Little, 1f any, change
takes place in the nickel and uranium
concentrations., This 1s discussed
further below.

5. When NaF-KI-LiF 1s circulated,
no large changes are found 1in the
concentration of the 1mpuraities.

6. A thin metallic layer 1s de-
posited on the walls of the cold leg.

This deposit is usually thickest 1n
the top, or hotter, part of the cold
leg.
FOR PERIOD ENDING JUNE 10, 1952

 
 
  

.t ~ "" . . : o o . a-" ‘.‘.' -~ N l' . 7? *

k "~ a C . » ~‘ o . e e e ¥ ‘ T et . T a
o S « - - . . o e y - -

. - » :: :: » . . ’q\ “-u‘__“ ' B,‘ . ' N o ‘_r.'f"' . /,r/ ‘

  
    

 

-
-

+
®
»
1

 

G - « . \’*\:Jf;’,,’%‘. # 5
. . " : %
’ vt ¥

I 2 . *
) -
e . " N
-~ < ey -

(b) Hot leg of hydrogen-flred loop.
Fig. 54.

corrosion of Inconel Thermal Convection Loop by NaF-KF-ZrF —UF
After 100 hr at 1500°F,

Showing Effect of Hydrogen Flring. Etched with aqua
regia, 250X,

111
Il

TABLE 12

Corrosion Data from Inconel Thermal Convection Loops Containing various Fluoride Mixtures

 

 

 

 

 

 

 

 

 

 

 

TIME OF HOT {EG METALLOGRAPHIC NOTES
1 - .
,{EXO)p COOLANT* CIBCULATION RLASOE\‘ FOH TEMPERATURE CHEMTCAI. NOTES
NG, (hr) TEPMINATION {°F) Hot Leg Cold Leg
73 | NaF-KF-1if 1000 Scheduled 1500 fayer of pits, 5 mils; intergranular No attack; possibly a very thin deposit Cr increased, Fe decreased; Cr slightly
attack, 10 mils; considerable grain higher in cold leg
growth
214 | NaF-KF-LiF + 500 Scheduled 1500 Slight intergranular attack, ! to 3 mils Penetration, | mil; no metal deposition All impurities low; Ni slightly higher
Nak than normal
219 § NaF-KF-LiF-UF, 500 Scheduled 1500 layer of pits, 10 mils, with maximem of Slight roughening Large increase in Cr, Fe decreased, Ni
15 mils; possible surface layer, 2 mils variable; Cr higher in cold leg
21} | NaF-KF-LiF-UF, 524 Scheduled 1500 Intergranular pitting attack, 4 to 8 Metallic deposit, 0.5 mil; no attack Cr increased, Fe decreased; even distri-
mils; some pitting in grains bution
212 NaF-KF-LiF—UF‘ 37 leak 1500 Intermittent layer of intergranular No atrack Cr, Fe, and Ni all show large increases
pits, 5 mils in both hot and cold leg
213 | NaF-Kf-LiF-UF, 500 Scheduled 1500 Intergranular pitting, 2 to 11 mils; Thin, continuous metallic iayer with an Large increase in Cr, Fe decreased; no
some pitting in grains intermittent naonmetallic layer on it systematic distribution; H, fired, H,
atmosphere
218 | NaF-KF-LiF-UF, 500 Scheduled 1309 Intergranuiar pitting, 6 to 10 mils Metailic deposit, 0.1 to 1 mil; appeared Cr increased, Fe decreased, %I decreased,
to be two layers; the second layer was ne systematic distributinn
not continucus and contained inclusions
219 | NaF-KF-Lif-UF, 480 Heater 1560 Heavy pitting, 5 to 13 mils Wall rough with thin layer; bottom layer | lLarge increase in Cr, Fe decreased; cold
failure metallic with discontinuous nonmetallic leg high in Fe and Cr
layer (particles) on top
216 NaF-BeF2-UF‘ 574 Scheduled 1500 Heavy to moderate intergranular attack, No uttack; some nonmetallic particles on Large increase in Cr, Fe decreased, siight
6 to 13 mils; an hot herizonta! leg wall decrease in U
there was attack only on one side
217 NBF-BEF!-UF‘ 500 Scheduled 15490 Heavy to moderate intergranular attack No attack; light metal deposit as layer Increase in Cr, decrease in Fe and U
up to 13 mils and crystals
220 | NaF-KF-LiF-UF, 2 Scheduled 1500 Light pitting, 2 1o 4.5 mils Scattered crystal deposit Cr decreased slightly in cold leg and
considerably in fiot leg, Fe decreased
slightly in both legs
221 } NaF-KF-ZrF -UfF 500 Scheduled 1500 f.ight to moderate intergranular pitting, Rough surface; thin nonmetallic layer
4 4 f g P
3 te 7 mils
222 | Ma¥-KF-Lif-UF, 500 Scheduled 1650
228 NnF—KF-ZrF.-UF‘ 500 Scheduled 1500 Light intergranular pitting, 3 to 13 Rough surface; thin deposited layer
mils

 

 

+ - -
Composition of coolants:

mole %.

NaF-KF-LiF - 11.5-42.0-46.5 mole %;

NaF-KF-LiF-UF, - 10.9-43.5-44.5-1.1 mole %;

Naf-Bef,-UF, - 47.0-51.0-2,0 mole %

NaF-KF-Zr¥ -UF, - 4,6-50,1-41.3-3.8

LHOdIY SSHYI0Ud X THALHUYNO I2af0odd dNV
€11

Corrosion Data

 

TABLE 13

from Stainless Steel Thermal Convection Loops Containing vVarious

Fluoride Mixtures

 

 

 

 

 

 

 

 

 

 

 

TYPE OF TIME OF | peasow pop | HOT LEG METALLOGRAPHIC NOTES
LOOF | STAINLESS coDLANT(2) | CIRCULATION! pppuinarion | TRNPERATURE CHEMICAL NOTES
* STEEL (hr) {°F} Hot Leg Cold Leg
111 316 NaF-KF-LifF i74 Leak 1508 Intergranuiar atteck, 10 mils; some wall No attack; continuous layer, | mil; some Large increase in (r, Fe variable; gas
reduction; tremendous grain growth grain growch leak around spark plug
116 316 NaF-KF-LiF 500 Schedulegd 13906 Some intergrenular pitting, 2 to 4 mils; Thin, continuons metaliic deposit with in- No impurities increased; no =zystematic
surface layer shows grain growth clusions variations
119 316 NaF-KF-LiF + NaK 500 Scheduled 1500 Surface rough with some deprassions, 2 No attack or deposit All impurities very low - much lower
o . wily; no pitting or intergranulsr attack than at start
112 318 NaF-KF-LiF-UF, 82 Plug 1500 Intergranular attack, 8 wils; grain growth Intermittent thin layer; grain growth Large increcase in Cr, Fe decreased
113 316 NaF-KF—LiF'UF4 123 Plug 1500(6) lntergranular attack, 8 mils; some pitiing | Thin deposited layer Large increass in Cr, decrease ia Fe
but irregular, {§ higher in cold leg
118 316 MaF-KF-LiF-UF, 147 Plug 1509 Intergranuiar attack, 5 te 12 mils; sowe Rough surface with 1 mil layver Cr increased, Fe decreesed, highest in
grains removed cold horizontal leg; Hz ficed, #,
atmosphere
121 3t NaF-KF-LiF*UF‘ 153 Plog 1360 Heavy intergranular attack, # to 10 mils; Duposited laver, 0.5 to 1.5 ail; irregular Increase in Cr, decrease in Fe; no
some grains removed in thickness systematic distribuvion
275 347 MNaF-KF-LiF-UF, 39 Leak 1500 Ynnergranuiar attack, 8 to 13 fiils;. Metallic dcfiosit, 0.5 to 1 mil; contains " 'Cr increased with no large varisvioh;
tremendous grain growth inclusionsg Fe decreased irregunlarly
274 347 NuF-KF-LiF-UF, 123 Plug 15008 Intergranular attack, 2 to 4 mils Bough surface with = thin depesited layer Cr increassd, Fe decrcased; bath
slightly higher in cold leg
251 316 NaF-KF-LiF-UF, 75 Plug 1500 Yery heavy intergranular pitting, 8 to 15 General nonmetallic deposit with thin Cr increased, Fe decrsased; cold leg
mils; surface rough, trewendous grain metallic layer; some crystals attached to siightly higher
grawth surface
252 310 NaF-KF-LiF-UF, 368 Plug and 1548 Very heavy intergranular pitting, up to Thin deposited laver; large crvstals Chemical snalyses irregular and no
instrument 20 mils ettached to surface definite trends were discernabla;
failure - - : . metnllic plug found in lower cald lag
127 116 NaF-KF-LiF-UF‘ 87 Teak 17490 Heavy intergranular attack, 3 to 7 mile; Metallic layer 2 mils thick
sSome grains removed
1232 316 NaF-KF-LiF-UF, 500 Scheduled 1650
40 418 Raf-KF-Lif-UF, 9 Plog 1500 No pitting or intergranular attmck; up to Meraltic deposit with inclusions; slight Cr increased, Fe decreased; batk
10 mils removed; considerable martensitic attack; Erain growth higher in cold leg
structure
43 410 NaF-KF-LiF-UF, 12 Plug 156G No pitting or usual intergranular attack; Metallic deposit with inclusions; grain Cr increased in cold leg but only
however, many grains are }ooee; con- growth slightly in hov leg, Fe decrsused but
siderable martemsitic structure varivd; wetullic crystals found in
cold leg
a3 430 Nab-KF-LiF-UF, 8 Plag 1500 Heavily pitted, 2 mils; general rTemoval of § Merallic deposit with inclusisns; surface Cr increased, Fe decressed; both higher
about 2 mils rough under the deposit in hot leg
14 43¢ NaF-BF-LiF-UF, 9 Plug and 1500
lrak

 

(G)CompositiOhs of conlants:

NaF-KF-LiF — 11.5-42.0-46.5 mole %; NaF-K¥-LiF-UF, - 10.9-43.5-44,5-1.1 mole %.

{b}Operaced far 72 hr at 1500°C and at 1600°C for remaining time,

‘0T ANNT ONIANA AoXdiad 404

2861
PIT

TABLE 14

Corrosion Data From Miscellaneous Thermal Convection Loops Containing NaF-KF-LiF-UF4

Composition of NaF-LiF-UFq:

100 9‘43- 5'44- 5'1| l mOle %

 

 

 

 

 

 

 

 

 

TIME OF REASON HOT LEG METALLOGRAPHIC NOTES
LOOP | MATERIAL | CIRCULATION FOR TEMPERATURE CHEMICAL NOTES
0. {hr) TERMINATION (°F) Hot Leg Cold Leg
104 Nickel 560 Scheduled 1500 Surface polished: no No attack; layer of Fe and Cr unchanged, Ni
pitting or inter- crystals, 1 to 3 mils higher in cold leg; some
granular attack; up to evidence of plugging;
9 mils removed; some crystals found in all
Erain growth sections
107 Nickel 1690 Scheduled 1500 Surface polished; no Crystal deposit
intergranular attack
or pitting; from § to
10 mils removed; large
gTains
340 Monel 117 Leak 1500 No pitting or jpter- No attack; no deposit No appreciable change in
granular attack impurities
365 Nimonic 500 Scheduled 1500 Intergranular pitting, | No attack; thin de- Cr increased, Fe decreased,
B to 13 mils, a few j posited layer siight increase in Ni
| spots up to 15 mils ;
| |
341 Monel 31 Leak J 1500 Ne intergranular attackj No attack; no deposit
or pitting; up tc 14 Q
! mils removed :
| |
44 Iron 25 Plug ; 1350 Rough surface with no g Rough surface with no E
i i large attack evident E attack or deposit 5
| | | |
: | ‘
375 | Hastelloy 80 Plug ‘ 1500 | | |
B | I i |

 

 

 

 

LYOdIY SSIY908d X THALYVNO 1DAL0¥d JNV
7. The addition of NaK to NaF-KF-
LiF reduces corrosion in both Inconel
and type-316 stainless steel and
suppresses formation of the metallic-
layer deposit.

8. The mechanism of plugging in
the loops has not been determined.
Definite crystal masses were found
during sectioning and melting out of
the fluorides in only three loops
(Nos. 43, 252, end 104; see Tables 13
and 14). In the other loops either
no metallic crystals or only a few
scattered ones were found. The
chemistry group has checked melting
points and viscosities of both hot-
and cold-leg material without finding
changes from the original values.

FOR PERTOD ENDING JUNE 10, 1952

9, The attack is about the same
at 1300°F as it 1s at 1500°F.

As an additional step in the under-
standing of corrosion in Inconel loops,
successive samples were drilled from
the inside of a section of hot-leg
pipe and analyzed. The results are
reported in Table 15. Since the pipe
from which these samples were drilled
was not perfectly round, the figures
in Table 15 should only be used for
discerning trends. The poor sampling
is one reason for the apparent presence
of fluorides at considerable depths.
In spite of the discrepancies 1t 1is
obvious that chromiumwas being removed
from the wall and that iron and nickel
were remaining essentially unchanged.

TABLE 15

Chemical Analyses of Subsurface Layers from an Inconel Thermal (onvection
Loop in Which NaF-KF-LiF-UF, Was Circulated

 

 

 

 

CEPTH OF CUT COMPOSITION (%) |
(in.) Fe Ni cr | Fe/Ni | KF(® | Liv(®) | Nar(®) | UF,(?) | Total(®
0.002 7.33 80.84 8.74|0.091 | 0.45 0.55 0.18 NDCE) 98.09
0.005 7. 42 80. 42 8.73 | 0.092 | 1.50 0.55 0.36 ND 98.98:
0.010 7.46 79.60 9. 39 0.094 | 1.20 0.55 0.36 ND 93,56:
0.015 7.60 79.98 | 10.63| 0.095 | ©.30 0.30 0.18 ND 98.99
0.020 T.48 79.88 | 11.08| 0.095 | 0.30 0. 30 0.18 ND 99.22:
0.025 7.15 77.57 | 13.18| 0.092 | 0.15 0.15 0.07 ND 98. 27
0.030 7.20 76.90 | 14.15 | 0.094 | 0.15 0.15 0.07 'ND 98.62‘

Exterior 7.08 | 76.46 | 15.850.093 99. 39

Inconel

specifications| 7 77 15 0.091 99

 

 

 

 

 

 

 

 

 

 

 

(a)

that all the wetal was present as the fluoride.

(b)
would add am ‘additiomal 0.6% to 0.75% to the total.

(G)ND = none detected.

These values were calculated from the spectrographic analyses for

the metal with the essumption

In addition to the items repofted, the percentages of cobalt, copper, silicen, and manganese

115
ANP PROJECT QUARTERLY PROGRESS REPORT

The chromium was being removed from
depths that appear by metallographic
examination to be unattacked.

As a further study of this attack,
the metallographic section was asked
to determine whether the apparently
unconnected spherical holes ohserved
actually connected to the surface. The
section was examined and photographed
while successive 1-mil layers were
being removed. By examining the
photographs and trying to follow the
holes, 1t was determined that they did
not connect to the surface,

Corrosion in Rotating Dynamic Test
gig (W. C. Tunnell, ANP Division). As
mentioned in the last quarterly re-
port, {1 the rotating dynamic test rig
1s an attempt to determine corrosion
effects more conveniently than 1m the
thermal convection loops. The apparatus
was calibrated for temperature measure-
ments with lead and appeared to be
accurate within 5°F. The rig was
charged with NaF-KF-LiF (11.5-42.0-
A6 .5 mole %)} contalned 1in nickel
liners. The mixture was held below
the melting temperature (550°F) and
under S5-micron pressure for several
days to dehydrate the fluoride,
Following this pretreatment the tem-
perature was raised to 1500°F and
held for one day. Raising the temper-
ature did not affect the vacuum, so 1t
was assumed that the material was dry,
The rigs were then allowed to cool
and the stuffing-box seal assembly was
installed with the Inconel specimens.
The temperature was then raised to

1500°F and the specimens were rotated
for 100 hours.

The Inconel specimens were examined
metal lographically and the fuel mixture
was subjected to chemical analysis,
The fluoride mixture before rotation
showed 1340 ppm of Ni, 35 ppm of Cr,
and 170 ppm of Fe, but after rotation

116

one flvoride sample showed less than
20 ppm of Ni, 160 ppm of Cr, and 75
ppm of Fe and the other sample showed
less than 20 ppm of Ni, 45 ppm of Cr,
and 80 ppm of Fe, Photomicrographs of
the Inconel have not yet been made,
but preliminary examination on the
metallograph does not show an appreci-
able change.

DYNAMIC CORROSION BY HYDROXIDES

Hydroxides in Seesaw
L. S. Richardson,

Corrosion bhy
Tests (A.D. Brasunas,
Metallurgy Division). Since oxygen,
as Ni0 or as 0,, added to NaOH has
resulted in greatly accelerated mass
transfer of nickel in nickel tubes, a
series of oxygen getters have been
added to seesaw tests. These have 1n-
cluded Na, NaH, Be, Mg, Al, and Zr,
None has shown beneficial results with
the possible exception of Zr, which
has given very erratic results, A
summary of the seesaw test results 1s
given in Table 16, The temperatures
listed in Table 16 are not the liquid
temperatures but merely the tempera-
tures measured with a thermocouple
spot welded to the metal tube. Pre-
liminary experiments show that the
temperature gradient is probably an
error by 100°C.

The data in Table 16, 1in con-
junction with previous experimental
work, lead to several general con-
clusions,

1. Mass transfer is accelerated by
the presence of relatively large
guantities of oxygen.

2. Apparently there 1s a low
temperature limit belcw which little
or no mass transfer occurs.

3. Additions to the NaOH have
little or no beneficial effect on mass
transfer.
L11

TABLE 16

Results of various Test Conditions on Mass Transfer of Nickel in Nickel Tubes Containing NaOH

 

 

TEST CONDITIONS

 

AMOUNT OF Ni DEPOSITED

 

: Hot-Zone Cold-Zone
Time No. of Environment Temperature Temperature AT COLD ZONE
(hr) Cycles (°C) {(°C)
117 14,000 Vacuum 797 505 Light to moderate
100 12,000 Yacuum 600 371 Practically none
125 45,000 Vacuum 645 475 Light
72 19,600 Oxygen atmosphere 750 540 Very heavy
147 39,500 Hydrogen atmosphere 740 520 Moderate
147 39, 500 Hydrogen atmosphere 765 515 Moderate
259 69,000 Argon 715 560 Light to moderate
8 2, 000 Vacuum + 5% Na 715 5G0 Heavy
8 2,000 YVacuum + 10% NaH 750 500 Heavy
0.2 56 Vacuum + 30% NaH 750 300 Failed immediately
102 27,000 Vacuum + 2% Nafi 700 555 None
142 38,000 Vacuum + 1/4% Zr02 720 540 Very light
104 28,000 Vacuum + 1/4% Al 750 660 Light to moderate
104 28,000 Vacuum + 1/4% B 750 665 Light
104 28, 000 Vecuum + 1/4% C 170 660 Moderate
104 28, 000 Vacuum + 1/4% Mrn 770 650 Very heavy
168 45,000 Vacuum + 1/4% Be 730 575 Heavy
114 31,000 Vacuum + 1/4% Ag 710 360 VYery heavy
114 31,000 Vacuum + 1/4% Li 790 635 Moderate

 

 

 

 

 

 

‘0T aNAl INIGNA aolddd 404

cS61l
ANP PROJECT QUARTERLY PROGRESS REPORT

Standpipe Tests of Hydroxide
Corrosion (A. D. Brasunas, L. S.
Richardson, Metallurgy Division). A
standpipe test run with A nickel and
NaOH vyielded deposition at a 700°C
zone surrounded by higher temperature
zones and solution at another 700°C
zone surrounded by lower temperature
zones. These results indicate that,
as observed many times, mass transfer
can occur 1n a so-called "static'™
system 1f a temperature gradient or
temperature cvycle 1s present.

Hydroxide Corrosion in Cold-Finger
Thermal Convection Apparatus, (J. V.
Cathcart, W, H. Bridges, G. P. Smith,
Metallurgy Division), A new type of
thermal convection apparatus, desig-
nated as theicold -finger" apparatus,
beern developed for use 1n the
corrosion testing of hydroxides. In
principle, the cold-finger apparatus
consists of a small-diameter tube that
1s sealed off at one end and inserted

has

into a pot containing molten hydroxide,
The inside of the small tube is cooled
by a jet of air to create a tempera-
ture differential between the inner
tube, or cold-finger, and the walls of
the hydroxide pot. As a result of the
temperature difference convection
currents are set up in the hydroxide.
This new apparatus has several ad-
vantages over the thermal convection
loops commonly used
research:

in corrosion
(1) few welds are required
1n i1ts comnstruction; (2) it
easily operated and repaired; and (3)
it has a relatively large area of
contact between hydroxide and blanket-
ing atmosphere. It is probable that
the effective rate of flow in a thermal
convection loop 1s greater than that
in the new apparatus, but the rate of
flow in the cold-finger apparatus is
sufficiently large to cause copious
mass transfer under conditions that
were knownm to produce mass transfer in
a loop,

1S more

118

With the apparatus completely
assembled except for the gquartz jacket,
the hydroxide pot was filled with c.p.-
grade NaOH pellets, which were then
dehydrated at 500°C under a vacuum of
approximately 0.04 wm Hg for 48 hours.
The temperature was then increased
until the outside of the hydroxide pot
reached the desired temperature. At
the same time the air jet was turned
on and regulated so that the cold
finger remained at a temperature of
approximately 500°C.

In the first of two runs in the
apparatus the hydroxide was blanketed
with a hydrogen atmosphere. The pot
walls and the cold finger were main-
tained at temperatures of 625 and 525°C,
respectively, for a period of 117
hours., Neither loose dendritic
crystals nor mass transfer of any sort
was noted during this run, The second
run was carried out under a vacuum of
about 0,04 mm Hg, and the temperature
was adjusted so that the pot walls and
the cold finger were at 725 and 500°C,
respectively. Heavy mass transfer
occurred between the walls of the
dehydration pot and the cold finger.
Large quantities of dentritic crystals
were observed on the surface of the
cold finger.

It was shown that at least under
the conditions specified for the first
run, a blanketing atmosphere of hy-
drogen effectively stopped mass
transfer in a nickel-sodium hydroxade
system,

Modified Thermal Comvection Appa-

ratus - TCA (J. V. Cathcart, W. H.
Bridges, G. P. Smith, Metallurgy
Division). A second, modified, thermal

convection apparatus, representing a
simplification of both the thermal
convection loop and the cold-finger
apparatus, has been constructed and
given preliminary testing. Because of
its simplicity, the modified apparatus
is easy to erect, requires a small
amount of laboratory space, and can
be put into operation in a short
period of time. The apparatus 1is
designated as the TCA, or thermal
convection apparatus. Jt consists of
a 1-ft length of l-in.-dia nickel
tubing with a 1/8-in. plate welded on
the bottom. A Kovar seal is attached
to the top end. A nickel vane about
3 in. long 1s posationed inside the
tube near the bottom. There i1s suf-
ficient space below the vane to permit
circulation. The nickel tube 1is
placed between 4-in., split-core,
semicircular heaters and the system is
connected to hydrogen and vacuum lines.
The desired blanketing atmosphere is
then introduced, and one of the semi-
circular heaters 1s removed to expose
one side of the apparatus to the
atmosphere. The power in the other
heater may be increased until the
wall temperature on the hot side of
the apparatus reaches 800°C. This
procedure results in the formation of
a temperature differential between
opposite sides of the vtube and sets up
convection currents in the hydroxide.

In the two runs to date with sodium
hydroxide the walls in the hot zone in
the vicinity of the heater were highly
polished whereas those in the cold zone
had a roughened appearance. Massive
deposits of crystals were found only
at the hydroxide-vacuum interface. On
the basis of these results it is felt
that the TCA represents an apparatus,
which, because of its simplicity and
ease of operation, might be well suited
to preliminary corrosion testing in
new systems. 1t is planned to continue
the development of the equipment,

FUNDAMENTAL CORROSION RESEARCH

In addition to the program of
empirically testing corrosion of
structural metals by fused salts,
considerable effort has been devoted

FOR PERIOD ENDING JUNE 10, 1952

to a number of studies designed to
assist in the discovery of the cor-
rosion mechanism. These studies have
included careful examination by
physical and chemical means of cor-
rosion products formed during large-
scale cyclic corrosion tests, exami-
nation of reactions of structural
metals with high-temperature liquids
in sample systems, and studies of
possible reactions of molten fluorides
and hydroxides under applied po-
tentials., A considerable program of
preparation of complex fluorides of
the structural metals has been con-
ducted to assist in the identification
of corrosion products. Preliminary
results have been obtained by several
independent groups, but the corrosion
phenomena and mechanisms proposed are
not vet entirely consistent.

Interaction of Fluorides and
Structural Metals (H. Powers, J. D.
Redman, .. G. Overholser, Materials
Chemistry Division). The interaction
of fluorides and structural metals at
high temperatures is being studied 1in
an attempt to define the mechanism of
attack on containers and to assist 1n
development of measures to control the
corrosion observed.

In general the studies have been
concerned with the NaF-KF-LiF eutectice,
with and without UF,. This fluoride
mixture is heated in contact with
type-316 stainless steel at 800°C for
about 6 hr under an inert atmosphere
and then cooled te room temperature.
After representative samples are taken
for analysis, the remainder 1s re-
heated under an 1inert atmosphere, and
samples are removed by application of
vacuum to the filter in which the
porous diaphragm 1s sintered nickel
or a thin membrane of graphite,

As the data in Table 17 indicate,
the reaction of the NaF-KF-LiF mixture
under these conditions tends to
solubilize some iron but virtually no

119
ANP PROJECT QUARTERLY PROGRESS REPORT

TABLE 17

Reaction of LiF-NaF-KF-UF4 Mixtures with Stainless Steel at 800°C

 

 

 

 

 

 

 

 

STRUCTURAL METAL IN LIQUID (ppm)
UF, CONTENT Before Filtration © After Filtration
(mole %) T T Re T Cr Ni Fe cr | N
0 4600 1600 600 3500 20 20
0 6 400 1200 470 2600 20 30
0 6300 1500 500 2400 20 30
0 8500 2200 1300 2300 20 30
2 8300 2000 800 7400 1700 20
2 9200 2400 1600 7100 500 700
2 8100 2400 500 6000 1400 100
2 8600 2000 1100 7000 1600 180

 

 

 

 

 

 

 

 

chromium or nickel, Although the
iron passing the filters shows con-
siderable variation from test to test,
at least 2000 ppm are solubilized by
this technique. Tt is significant to
note that the use of nickel filters
(Micro-Metallic Corporation), which
are considerably more porous than the
graphite, does not change the values
for soluble iron and chromium com-
pounds; also, when nickel filters are
used, the nickel content of the
filtrate rises to about 300 ppm., In
general the agreement of results from
the two filters indicates that the
material found in the filtrate is dis-
solved rather than suspended and that
reaction with the filter is probably
not significant in so far as
and chromium are concerned.

iron

When UF, 1s a constituent of the
mixture, however, significant changes
are apparent. The iron content of the
filtrates rises considerably (2 to 3
fold) and the chromium obviously
becomes solubilized to a large extent.
The nickel content of the filtrates

120

also rose significantly, except in one
case (as yet unexplained) in which the
soluble nickel low
level.

remained at a

These studies have not yet yielded
sufficient concentrations of the
soluble corrosion products for identi-
fication by x-ray diffraction or other
means, However, 1t 1s believed, in
view of data presented below, that the
soluble materials are complex fluorides,
Similar and somewhat more detailed
studies are planned for the more
complex NaF-ZrF,-UF, systems in the
near future.

Synthesis of Complex Fluorides
(B. J. Sturm, L. G. Overholser,
Materials Chemistry Division). Analyses
of the various fluoride fuels after
static or dynamic corrosion testing
have shown that significant amounts of
the container (chiefly iron, nickel,
and chromium from Inconel or stainless
steels) 1nteract with the fused
fluorides. Tdentification of the
chemical combinations in which these
g

structural elements occur 1is an
important part of the study of the
reaction mechanism. It was assumed
that the iron, nickel, and chromium
are present as simple or complex
flvorides 1in the mixture of alkali
fluorides; consequently, the prepa-
ration of various fluorides containing
iron, nickel, or chromium was under-

takeno

The dehydration of most fluorides
can be accomplished without hydrolysis
only if an atmosphere of HF is main-
tained over the material. Dehydration
of KF, NaF, and LiF by'heating under
vacuum or inert gas results in hy-
drolysis of the alkali fluorides; the
production of a free base may easily
be detected by measuring the pH of a
solution of the dehydrated alkali
fluoride. Melting of hydrated nickelous
and chromic fluorides results 1in
extensive hydrolysis unless the ma-
terials are kept under an atmosphere
of HF while heating, Ammonium bi-
fluoride may be added as an alternate
to provide the protective atmosphere,

Anhyvdrous nickelous, chromic, and
manganic fluorides have been prepared
from the commercially available
fluorides., Ferrous fluoride has been
prepared by the decomposition of
NH,FeF, at 800°C; the ferric fluoride
results from decowposition at a lower
temperature, The NH,FeF, may be
prepared by reacting anhydrous FeCl
with NH,HF, at 350°C.

Compounds corresponding to K3CrF6,
Na,CrF,, and (NH,),CrF, have been
prepared by heating the proper molar
gquantities of the corresponding acid
fluorides with chromic fluoride.
Samples of Li CrF, were prepared by
heating LiF with (NH4)3CrF6. NaK, CrF,
has been prepared by fusing a mixture
of fluorides corresponding to this
composition, Attempts to prepare
NaQKCrF6 by a similar method yielded
NaK,CrF, contaminated with excess NaF

FOR PERIOD ENDING JUNE 10, 1952

and some unidentified chromium com-
pound, Preparations of complex
fluorides of iron and nickel with the
alkali fluorides are under way.

Examination of Corrosion Products
(D. C. Hoffman, F. F. Blankenship,
Materials Chemistry Division). Loop
tests in which ZrF, -bearing fuel
mixtures have been circulated by
thermal convection have not shown any
restriction to flow in tests up to
500 hr, however, stainless steel loops
in which NaF-KF-LiF-UF, mixtures have
been circulated nearly always plugged
after intervals of about 100 hours.
During the past quarter a concerted
effort has been made to determine the
nature and location of these plugs.

The loop sections to be examined
were selected, 1n general, to contain
the portion indicated by radiographs
to be the most dense. The apparent
density differences, however, seem to
be a result of contraction and re-
arrangement of the melt on cooling and
do not necessarily reveal the presence
of a metallic phase, The presence of
large metallic deposits could be
deduced from the radiogram in the few
cases where they have been shown to
exist, but this technique was com-
pletely unreliable for the more finely
dispersed metallic erystals. Exami-
nation of representative sections
throughout the loops demonstrated that
the trap and the vertical cold leg
contained most of the deposited metal
if there were any,.

One section from the hot leg, two
sections from the cold leg, the trap,
and any additional sections showing
radiographic deposits were customarily
examined., Material was removed from
the sections, examined under the
microscope, and separated by means of
a magnet before and after selective
leaching with aluminum nitrate or
ammonium oxalate solutions, The
magnetic fractions and any others that

121
ANP PROJECT QUARTERLY PROGRESS REPORT

could be separated by the selective
leaching procedures were submitted for
examination by chemical microscopy,
x-ray diffraction, and spectrographic
and chemical analyses.

Locops in which NaF-KF-LiF was
circulated in type-316 stainless steel
and loops in which NaF-KF-LiF eutectic
plus 2 mole % UF, was circulated in
types-310, -316, -347, -410, and -430
stainless steel, nickel, Inconel, and
Nimonic were examined, In addition,
one Inconel loop in which NaF-KF-Zr¥F, -
UF, was circulated has been examined.

Metal "plugs" in quantity sufficient
to stop the flow were observed 1in
loops of types-410, -430, and possibly,
-310 stainless steel. In all other
cases, regardless of whether the loops
had plugged or not, only scattered
metal crystals, if any, were found,
therefore it does not seem possible
that physical plugging of the tubes
by these small quantities of material
could have occurred.

In the loops of types-316, -347,
-410, and -430 stainless steel the
metallic phase was highly magnetic and
all examinations indicated that the
material was nearly pure iron. The
deposit in the type-310 stainless
steel loop was only weakly magnetic
and showed a high chromium content:
the x-ray-diffraction pattern indi-
cated the presence of a noncubic
crystal i1n addition to a material that
was presumably a chrome-iron with a
shifted iron lattice. No evidence of
deposits of metallic nickel has been
found in loops of any of these alloys,
although in nickel loops copious
quantities of nickel crystals were
obtained.

Unidentified sulfides often seem to
concentrate in the vicinity of the
metallic crystals, as revealed by the
evolution of H,S during leaching with
acidified aluminum nitrate solution,

122

A green, crystalline substance that
was sparingly soluble in aluminum
nitrate solutions was found associated
with UO, and the metallic crystals in
many of the loops examined. In
addition, layers of the same material
were observed 1in containers in which
NaF-KF-LiF-UF, fuels were prepared and
stored. Spectrographiec analysis of
samples from these sources indicated
the presence of Na, K, and Cr but
virtually no iron. X-ray-diffraction
studies by Agron and microscopic
examination by McVay indicated the
material to be a cubic erystal with
an index of refraction close to 1.422.
By comparison with complex fluorides
prepared by Sturm and Overholser, the
material has been positively identified
as K,NaCrF, . This material, which
appears quite soluble in the molten
fuels at temperatures above 1000°F,
seems to be the major product of
attack on chromium by the fluorides,

Since the chromium in this compound
1s trivalent, whereas Cr*? would be
expected from the reaction

Cr® + 2U%*3 —— 9u*3 + (Crt?

it appears likely that reduction of
U*? does not explain the occurrence of
this compound.

Examination of the Inconel loop,
which circulated NaF-KF¥-ZrF,-UF, for
500 hr with a hot-leg temperature of
1500°F and showed no signs of plugging,
revealed only very slight guantities
of metal deposition in the trap. These
metal crystals, which have not vet
been identified, grew to a length of
about 1/8 in. in a band perhaps 3/16
in, wide, No other metallic deposits
or crystals were observed. No H,S
odor was detectable on leaching the
material from this loop with acidified
aluminum nitrate solution. Since the
fuel mixture used was treated with H,
and HF prior to transfer to the loop,
the sulfur should have been at a low
level prior to the run,

ot
Chromium has been shown by other
groups on the project to be selectively
removed from Inconel and stainless
steel containers by the molten fluo-
rides; however, chromium was rarely
found in metallic deposits from these
experiments. Quantitative analysis
generally revealed the presence of
large amounts (about 3000 ppm) of
chromium in the fluoride melts from
the loop tests; the chromium was
generally uniformly distributed in
both hot and cold legs of the loops in
a manner such as to suggest that it
was dissolved in some fashion.

Chromium removal 1is still being
studied, but it 1s suggested that it
may be explained by the reaction

3UF, + 2Cr,0,——> 300, + 4CrF, .

If thais then all
active oxygen 1in the system would
eventually be converted to UO, at the
expense of solubilizatioen of the
chromium. The fact that U0, 1is a
constant component of material from
corrosion tests, whereas UF3 has not

is the reaction,

been positively identified, would
seemto lend support to this hypothesis,

It 1s certain that difficulty from
the reaction

2UF, + Cr —> 2UF, + CrF,

would be experienced. It seems, how-
ever, that a large portion of the cor-
rosion presently observed may be due
to the other mechanism and that
elimination of oxides and active
oxygen will remove a major portion of

this trouble.

X-Ray-Diffraction Studies (P. Agron,
Materials Chemistry Division)., Stand-
ard, x-ray-spectrometer, diffraction
patterns were obtained for Zr¥F,,
ZrOF,, ZrOz; UF,, UF,, and UD, to

facilitate their identification in

FOR PERTIOD ENDING JUNE 10, 1952

fuels before and after corrosion
tests., The patterns of FeF,, FelF,,
FeO, Cr¥F,, and Cr203 were also con-
sidered i1in the examination of test
samples.

The indexed lines of known com-
plexes, such as K3FeF6, Na, FeF_, and
KNiF,, were compared with unknown lines
found in corrosion studies. None of
the latter materials were identified.
However, synthesis of the two new
compounds K,CrF, and K,NaCrF, (prepared
by B. J. Sturm, Materials Chemistry
Division) showed that the latter
double-alkali complex was a residue
commonly found in corrosion tests with
the NaF-KF-LiF-UF, mixture. Table 18
lists the crystal structure of the
complexes discussed above.

Examinations of synthesized com-
pounds of iron and chromium complexed
with sodium and lithium fluorides are
in progress. An attempt is being made
to synthesize K,NaFeF.

X-ray-diffraction patterns have
been obtained for several fuel com-
positions and two uranium-free eutectic
mixtures. Further study of these

TABLE 18

Complex Fluoride Compounds Formed by
Fluoride-Container Reactions

 

 

 

 

 

_ LATTICE

compounp{ @) DIMENSION, | DENSITY (g/ce)

a (&) D, D,

KNiF, 4.02

K FeF_ 8.62 2.96

K,CrF ) 8.53 2.99
(b

K, NaCrF (%) 8.27

Na FeF, 7.95

 

 

 

(2)Fach of these compounds had a cubiccrystal
form.

(b)Isomorphous with KaFeFG.

123
ANP PROJECT QUARTERLY PROGRESS REPORT

materials will be required to establish
whether solid complexes and/or solid
solutions occur,

EMF Measurements in Fused Fluorides
(L. E. Topol, L. G. Overholser, Ma-
terials Chemistry Division). Research
on the electrochemistry of fused
salts has been undertaken in order to
learn more regarding the fundamental
aspects of corrosion. The current
approach to the problem involves the
decomposition-potential measurement of
various fluorides (pure and 1n
tures). From these decomposition
voltages and the egquation

mix-

E; = Egy ~——1Ina,

some knowledge should be obtained
about the activities of the components
in these mixtures. In addition, a
relative, chemical-activity series of
metals 1n fused fluorides can be formed
analogous to that in agqueous solutions,
This series can be compared to one
determined from emf measurements if a
suitable reference electrode can be
found. A cell assembly employing
nickel as the reference electrode, of
the type

TABLE 13

Reversible Decomposition Voltages of
various Fluorides

 

 

 

 

 

 

 

 

 

 

DECOMPOSITION VOLTAGE (v) AT
FLUORIDE VARIOUS TEMPERATURES
298, 1°K | 500°K | 1000°K | 1500°K
NaF 5.60 5.39 4.86 4.45
KF 5.52 5.28 4.72 4.31
LiF 6.05 5.88 5. 40 4,95
Several electrolyses have been made

with KF at 885°C by using nickel
electrodes and a helium atmosphere,
In all cases there was a definite
break in the voltage vs. current curve
at about 1.2 volts,

Two different crucibles have been
used as contalners. Morganite, a
special recrystallized aluminum oxide,
did not hold up at all and virtually
disintegrated after a few hours of
contact with the molten salt. How-
ever, the Norton alumina RA 7232
crucible, which had a much greater
wall thickness, survived contact with
the molten salt fairly well, although

 

Ni | Nin(cl) in a fluoride, or fluoride mixture | MF,(c,) in same fluoride composition | M,

where the concentrations ¢, and c, are
in mole % and are approximately egual
and M = Cr, Fe, etc., would vield the
desired results. Finally, from the
temperature coefficients of the
measured potentials, i1f reversible,
useful thermodynamic information may
be obtained. (The accuracy of these
data depends on the current efficiency

and nature of the electrodes used.)

From known free energy data (Quill’s
Thermodynamics), the reversible de-
composition voltages of NaF, KF, and
LiF have been calculated for various
temperatures (Table 19).

124

 

1t 1s doubtful whether 1t can be used

again,

Additional work is planned with KF
with and without added NiF, in which
crucibles of BeO, MgO, and
Ni and electrodes of Ni, and
graphite will be used.

2
graphite,

Pt,

Reactions in Fused Sodium HYy-
droxide¢?) (A. R. Nichols, Jr., Ma-
terials Chemistry Division). Morganite
has been found to withstand fused NaOH
at T00°C sufficiently well to permit

 

(2 rpid, p. 135.
1ts use as a vessel for potentiometric
and electrolytic experiments. Because
of difficulties in obtaining repro-
ducible results from the concentration
cells previously described, efforts
have been made to use the measurements
of decomposition potentials as an
approach to the evaluation cfelectrode
potentials; however, attention is
stil]l being given to the relationship
of temperature to potential. Pre-
liminary measurements indicate that- a
nickel electrode placed i1in a hot
region of NaOH melt becomes negative
with respect to one in acooler region.
This is consistent with present 1ideas
of the processes occurring during mass
transport. |

Voltage-current curves have been
obtained for the electrolysis of fused
NaOH with and without added NiO by
using nickel electrodes. For the
sizes and spacings of electrodes used,
currents ranged up to 2 ampat voltages
of up to 2 volts. With only NaOH,
extrapolation of the linear part of
the curve gave an apparent decompo-
sition potential between 1.2 and 1.4
volts, whereas in the melts containing
added Ni0O, an additional electrode
process having an apparent decompo-
sition potential of 0.7 to 0.8 volt
was indicated. The melts containing
Ni0O gave a large deposit of metallic
nickel at the cathode but little or no
attack on the anode. Further experi-
ments will be necessary to obtain more
quantitative information about these
processes.

Various phenomena observed during
the electrochemical experiments de-
scribed have led to experiments con-
cerned with reactions occurring in
fused NaOH in the absence of applied
electric fields. In each case 1imn
which NaQOH and NiO were 1in contact at
temperatures of 700°C or higher, a
fine, crystalline deposit of metallic
nickel was observed when the melt was
cooled and leached with water. This

FOR PERIOD ENDING JUNE 10, 1952

has been observed whether the container
used was nickel, recrystallized
alumina, or corrundum, The nickel: may
be presumed to be formed by the same
process as that taking place as the
second step of the mass transport
phenomenon. ' :

When N10O and NaOH are held at 700°C
in a nickel vessel for several hours
and then slowly cooled, the soliditfied
melt is found te contain wiry, black
needles or fibers. These fibers are
from 0.001 to 0.003 in. in diameter
and up to 1/4 in, in length. Their
chemical analysis, 44% nickel and 29%
sodium, does not correspond well with
any simple formula. X-ray-diffraction
studies of this material indicate that
it is crystalline but not identical
with any compound for which data are
available. When this material 1is
allowed to stand 1in water, 1t slowly
changes from black to dark green in
color without changing noticeably 1in
crystalline form. These washed needles
contain no sodium and have a nickel
content corresponding fairly closely
to the formula Ni(OH)Z'ZHZO. X-ray
diffraction shows lines characteristic
of Ni(OH)z. However, its index of
refraction of approximately 1.652 1s
well below that of Ni(OH)z. Studies
of the nature of these products are
continuing.

Compounds Resultiting from Hydroxide
corrosion (G. P. Smith, L. D. Dyer, B.
Borie, Metallurgy Division)., Studies
are being made of reactions that can
occur in fused hydroxide media. In
making these studies, compounds have
been prepared that are believed to be
NaNiQ, and NaFeO,. This is the first
time that NaNi0O, has been shown to
exist. In preparing the NaNi10,, c.p.-
grade NaOH was heated in a nickel tube
at 700°C and oxygen was bubbled through
the melt for 48 hours. The tube of
hydroxide was then quenched in o1l and
the top portions of the melt were
immediately placed in absolute ethanol.

125
ANP PROJECT QUARTERLY PROGRESS REPORT

When the hydroxide had dissolved,
black crystals were obtained. These
crystals were washed with absolute
ethanol until the washings failed to
turn a phenolphthalein solution red.
The product was vacuum dried over
Drierite and then washed with water.
The compound obtained by washing with
water was i1dentified by x-ray dif-
fraction as Ni,0,;°H,0, which is a
known compound. The compound obtained
by washing with alcohol gave an x-ray-
diffraction pattern that could not be
identified; ,however, this compound
reacted with water to form Ni O,°H,O.
The product obtained by washing with
absolute alcohol was analyzed for
nickel, sodium, and carbon., The
results of the analysis are given 1in

Table 20.
TABLE 28

Analysis of Hydroxide Corrosion Product

 

 

 

FOUND THEORETICAL FOR
ELEMENT (%) NaNiQ, (%)
Nickel 51.02 51.6
Sodium 21.2 20.2
Oxygen 28.2
Carbon 0.1 0.0

 

 

 

In a2 similar experiment using a
Globe iron tube as the container for
the molten hydroxide, NaFeO, crystals
were formed. This compound was identi-
fied by x-ray diffraction.

Mechanism of Fluoride Corrosion
(G. M. Adamson, A. D. Brasunas, L. S.
Richardson, Metallurgy Division). As
mentioned in the last report, (3)
magnetic changes were noted in stain-
less steels after exposure to fluo-
rides. Similar changes have also been
noted in the hot legs of the Inconel
loops. As an additional check on the

 

(3)roid. p. 136.

126

corrosion mechanism, a microscopic,
magnetic method for determining the
presence of these transformed layers
has been developed. Figure 55 shows
that the transformation in an Inconel
loop follows a fairly straight-line
interface and extends to a depth of
approximately two-thirds the depth of
the holes. This change of Inconel to
a magnetic material may be explained
by the removal of chromium, which
leaves an iron-nickel alloy. The fact
that complete removal of chromium 1s
not necessary for this transformation,
explains the observation that the area
of apparent attack extends deeper than
the transformed layer.

A study of these facts leads to the
conclusion that the corrosion mechanism
of fluorides in Inconel is a leaching
of chromium from the surface of the
Inconel followed by diffusion of new
chromium from the interior of the pipe
to the surface from which 1t in turn
is also leached. As the chromium
diffuses to the surface, very small
voids are left behind. These voids
seem to concentrate in grain boundaries
and at similar areas of weakness faster
than nickel or iron can diffuse inward
to fill them,

A number of other experiments have
been made that strongly indicate that
these are true voids caused simply by
chromium depletion of the
regions. It has already been es-
tablished that the chromium content of
fluoride 1s high after corrosion test-
ing. Such preferential solution would
necessarily result in reduced chromium
at the surface of the metal. In one
test, metal crystals precipitated from
the fluoride contained 4.9% Cr.(%?

surface

Calculations of void volume based
on such chromium depletion are in good
agreement with the density of voids

 

M rpia, p. 124,
 

FOR PERIOD ENDING JUNE 10, 1952

 

T e ; A P s
sl 4 . - . o - - :
T o i T .ot s sl il L : L\ Lo s -~
* ] S . 4 T, e . ., . . . Loy # ‘o , e
e £ ~;\ . i i ’g cf & 3 9 "fl N L R Fe . § ..-\ B
. Low L e . ba . ST e, . - - o
i ; . o N . £ . » . .
P . e rm , . - . \ :,Aff‘; - [ o P .
. L S \ - e . ,
- - T -k .. . " .

¥ o TH .
“{‘T S L C s . :
LR ;s . s [

magnetic test,

 

P ad - Tty oY A o
. "“- - C e kS S : o T
- . Foa

(6) Transformed layer of hot-leg section.

Magnetic Susceptibility of Inconel Thermal Convection Loop After

. % Sy
At e e

Fig. 55.
166 hr at 1500°F with NaF-KF-LiF~UF4.

127
ANP PROJECT QUARTERLY PROGRESS REPORT

actwally observed; hence, such voids
may be expected. Similar vacancies
have been reported¢®'%) jin diffusion
experiments by other investigators.

It would be reasonable to expect
the formation of voids if chromium
depletion were accomplished by other
means. Therefore two tests were made.
(1) Inconel was oxidized at 1200°C for
200 hours. Chromium was selectively
oxidized and left a chromium-depleted
area., (2) Inconel and an 80% Ni1-20%
Cr alloy were heated to 1375°C for 42
hours. The vapor pressure of chromium
was appreciably greater than that of
iron or nickel.
both cases

The voids formed in
appeared to be 1dentical
with those formed in fluoride media.

As a further verification of the
theory that holes are caused merely by
chromium depletion, a test was made
using NaF-—KF-LiF-UF4 (10.9-43.5-44,5-
1.1 mole %) in Inconel at 850°C for
100 hours. (This test has always
resulted in hole formation.) Abont
10% of -300 mesh chromium powder was
added to saturate the fluoride with
chromium, and, as expected, no voids
were formed.

These data and observations are in
complete agreement with the theory that
the holes observed in fluoride-attacked
alloys are true voids caused by selective
solution of an alloy component. The
mechanisms discussed above apply to
Inconel, but further work is necessary
to determine whether they also hold
for fluoride corrosion of stainless
steels.

Free-Energy Relations of gFluorides
and Structural Metals (R. C. Briant,
M. E. Lee, L. A. Mann, ANP Division).
Considerable effort is being devoted
to determining the causes and mechanisms

 

(S)H. Buckle and J, Blin, J. Inst. Metals 88,
Part 7, p. 385 (19052).

{6)A. Smigelskas and E. Kirkendall, Trans, A=,
Inst. Kining Met. Engrs. 171, 130 (1947).

128

of corrosion of structural materials
by ligquid fluoride salts, Free
energies of formation of a considerable
number of fluorides, oxides, and
miscel laneous pertinent compounds have
been calculated and plotted against
temperature as basic data for de-
termining the likelihood of proposed
chemical mechanisms as well as the
concentrations of resulting products.

Solutionof Metals in Moltenm Halides
A. Bredig, J. W. Johnson, H. F.
Chemistry Division). A

(M.
Bronstein,
general survey of the subject of so-
lutions of metals 1n their molten
halides and some preliminary experi-
mentation were carried on during 1951(7)
following a suggestion made in 1950(8)
that melts containing "subhalides" or,
in other words, "halogen (fluorine)
deficient" melts may behave more
favorably in regard to corrosion than
halides of normal stoichiometry,

Although the general aspects of the
subject, i1ncluding
phase eqguilibria,

thermodynamics,
and atomic and
electronic structure of such melts,
continue to be studied on systems
contalning a varlety of melts and
halogens, a more specific investigation
of fluoride melts of particular 1interest
tothe ANP work has been started during
the past quarter. The results obtained
may be summarized as follows: (1) 3
mole % potassium metal dissolved at
860°C in the molten binary eutectac
of KFand NaF; (2) 1.6 mole % potassium
dissolved at 860°C in the molten
ternary eutectic KF-NaF-LiF; (3) sodium
and lithiuwm metal added to these
eutectics would reduce considerable
amounts of potassium fluoride and some
sodium fluoride and thus change the

composition and melting points of the

 

(T}ChEIistry Division Quarterly Progress
Reports for Periods Ending June 30, 1951, ORNL-
1116, p. 65-67; September 30, 1951, OBNL-1153,
p- 81; December 34, 1551, ORNL-1260, p. 167-108.

(S)M. A. Bredig to A. M.
communication, QOctober 1950,

Weinberg, private
molten mixtures; (4) the solubility
of potassium metal at lower tempera-
tures is expected to be considerably
smaller, perhaps by a factor of 10 at
500°C, and is now being measured.

It appears as a result of thermo-
dynamic considerations that even small
concentrations of potassium metal added
and dissolved in the fluoride melt, of
the order of 0.1%, can be expected to
have a considerable beneficial effect
on corrosion,

For details the Chemistry Division
quarterly progress report for the
period ending March 31, 1952 (ORNL-~
1285) should be consulted.

Fluoride corrosion Phenomena (M. A.
Bredig, J.W. Johnson, H. F. Bronstein,
Chemistry Division). The observation
that alkali metals dissolve 1n appreci-~
able quantities in the melts of their
halides is of significance in several
respects. In the absence of other
corrosive agents such as oxides, cor-
rosive mass transfer under "dynamic"
conditions and the absence for practical
purposes of corrosion under "static"
conditions may be explained. The
equilibrium in corrosion reactions of
the type Fe + 2KFT=== Fe¥, + 2K may
be favorably influenced by the additioen
of alkali metal to the fluoride
mixture, ‘

Thermodynamic estimates of equili-
brium constants for reactions of this
type, based on the most recent data
available in the literature, (%) indi-
cate concentrations of Cer, Fer, or
NiF, at 1500°F to be of the order of
10-4-8 , 1()-5 . 8 , and 1()—6 . 3 , réfi—
spectively, At 1200°F the figures are
1076, 10°7, and 10”8, respectively,
that 1s, for each metal the concen-
trations are less by more thana factor

 

(9)yg, Brewer, L. A. Bromley, P. W. Gilles, and
N. L. Lofgren, The Chemistry and Hetallurgy of
Miscellaneous Matertials: Thermodynamics, N.N.E.S.,
Div. 1Y, Vol. 19B, p. 76 {19590), :

FOR PERIOD ENDING JUNE 10, 1952

of 10 than at 1500°F. Thus the con-
stituents of the structural metal
dissolved as fluorides at 1500°F can
crystallize out as metallic elements
at 1200°F as a result of almost com-
plete reversal of the reaction occurring
at 1500°F. Assuming rapid establish-
ment of equilibrium for 2000 g of
molten NaF-KF-LiF (11.5~-42.0-46.5
mole %) in both legs of the thermal
loops and a flow rate of 1 cycle per
minute, a quantity of 6000 »x 2000 x
10°%8 or approximately 15 g of metallic
iron would deposit in the cold leg for
6000 cyeles in 100 hours., For chromium
and nickel, these quantities are

approximately 100 and 6 g, respectively.

These are very crude estimates that.
are probably correct to only within a
a factor of 10 except for the relations
between Fe, Cr, and Ni, which appear
qualitatively of the right order. ' The
effect of the formation of complexes

suchzu;K3CrF5, for which thermodynamic
data are not available at present,

remains to be considered 1n more

accurate calculations.

It is clear that the establishment
of equilibrium in a static test, that
is, without thermal gradients, would
not, with a mole fraction of the erder
of 107% to 107° of CrF,, FeF,, or
NiF,, produce any considerable cor-
rosion, : In fuel mixtures containing
UF,, the free energy of reactions of
UF,, such as Fe + 2UF,z=——= 2UF; +
FeF, (AF o ~6 kg/cal), is much less
positive than for the reactions dis-
cussed thus far for KF (AF = ~80
kg/cal)., The comparatively small
increase in cerrosion and mass transfer
observed with NaF-KF-LiF-UF, compared
with NaF-KF-LiF, in spite of the large
disparity of the free-energy changes
of the reactions, is explained by the
much lower concentration of the UF4.

It has been possible to rather
dramatically demonstrate the oc-

currence, under suitable conditions,
of the reaction Fe + 2KF =====FeF, + 2K,

129
ANP PROJECT QUARTERLY PROGRESS REPORT

A mixture of iron powder and potassium
fluoride was heated for several hours
at 920°C in an evacuated stainless
steel tube that had an air-cooled
section and a water-cooled section,
After the test it was found that most
of the potassium fluoride had evaporated
and redeposited at the air-cooled
portion., A very small guantity of
black particles were found that were
insoluble on dissolution of the fluo-
ride in water, The particles, which
were metallic 1ron, were probably
produced by the partial back reaction
FEFz(g) + 2K(g)€“‘"::"::}' Fe(s) + QKF(S)-
At the water-cooled part, approximately
1 g of potassium metal was collected,
which was the portion that could not
react backwards with FeF, because
of rapid precipitation of the latter
at the air-cooled part of the walls,

It 15 impossible to state at
present with certainty that the cor-
rosive mass transfer observed 1in
thermal loops must be entirely or
partly due to fluoride reactions of
this sort rather than to oxide reactions
as long as the possibility of the
presence of oxide i1n some form cannot
be ruled out. Eguilibria of the type

or

Fe + UO,F, w===FeF, + UQ,

obvicusly may also produce corrosive
mass transfer of metal. It seems that
recent observations on the decrease or
absence of corrosive mass transfer in
fuel mixtures containing ZrF, instead
of LiF or BeF, have been attributable
ta improved purification and decreased
oxygen content. Although the removal
of oxygen appears to be a mnecessary
condition for a low corrosion rate,

130

the possibility of a fluoride reaction
cannot be neglected., Sufficiently
accurate thermodynamic data for Zr¥F,
are not available., The heat of for-
mation of ZrF,, AF = 445 t 30 kg/cal,
is 10 times that of KF, and the
equilibrium concentration of FelF, at
1100°K, which is from 105 to 1079,
is one-thousandth that of KF. If the
figure 105 were correct, it would be
necessary to find an explanation for
the low rate of corrosive mass trans-
fer actually observed. Observations
on the peculiar behavior {(the apparent
self-welding) of nickel in fluoride
melts containing 7ZrF,, which might
perhaps be connected with some alloying
of nickel with zirconium produced by
the reaction ZrF, + 2Ni === 2NiF, +
Zr, may indicate a possible way of
reconciling the large figure of 107°
for the concentration of FeF, (zf 1t
is correct) with the low corrosion
rate observed in the thermal loop.
Zirconium from the above reaction
might be thought to plate out on, or
form an alloy with, the steel or the
Inconel, This could reduce the cor-
rosion rate by at least two independent
mechanisms: (1) formation of a pro-
tective surface film containing zir-
conium and (2) absence of transport of
zirconium to the cold leg for reversal
of the equilibrium there, which
reversal alone would make new attack
by KF in the hot leg possible.

It is obvious that much more work
is required to completely elucidate
the mechanism of corrosive mass trans-
fer, However, the possible occurrence
of zirconium plating or alloying may
lead to the development of methods
other than the use of fluorine-deficient
melts that may permit general control
of corrosion and mass transfer 1in
fluoride melts,

,l
AR D b T A AT R 000, P N P e e e

FOR PERIOD ENDING JUNE 10, 1952

12. METALLURGY AND CERAMICS

W. D. Manly

J. M. Warde

Metallurgy Division

High-temperature brazing alloys
have been studied for possible use in
the heat exchanger and reactor as-
semblies. The tests conducted on
brazing alloys include flowability of
the braze, determination of the
physical properties of the resulting
joint, and static corrosion tests of
the brazed joint in reactor fluids.
The 60% Pd-40% Ni alloy was found to
be the best brazing material to use
in contact with fluoride mixtures.
Additional experiments have been con-
ducted on the cone-arc techniques to
obtain an understanding of the variables
of the process. Some work has been
completed on the practical application
of cone-arc welding to fabrication of
the tube-to-header joints of heat
exchangers,

l.oose powder sintering has been
studied as a method for the production
of solid fuel elements. This technique
involves sintering a loose powder of
the uranium-bearing mixture to a solid
backing plate. The variables studied
have included sintering temperature,
sintering time, fuel-component particle
srze, cold working and resintering,
surface preparation, and sintering
under load. Powder mixtures of boron
carbide and 20% iron by volume have
been hot pressed to make slugs for the
safety rod of the ARE. Hot-pressed
mixtures of aluminum oxide and boron
carbide are being studied for use in
the regulating rod of the ARE.

An oxidation-resistant ceramic
coating has been successfully applied
to stainless steel, and similar
coatings are being developed for
nickel and for a stainless steel
radiator intended for use with the
experimental reactor.

LOOSE POWDER BONDING

E. S. Bomar - J. H. Coobs
Metallurgy Division :

In the development of fuel elements
utilizing a solid fuel, a variety of
possible configurations has been
suggested for incorporating the fuel
with a structural material. One method
of fabrication would be to bond a
fuel-bearing powder to an appropriate
backing material by sintering the
powder either with or without appli-
cation of pressure., Preliminary
experiments to prepare flat sheets
covered on one side with a sintered
fuel-bearing powder were carried out
in a furnace at the Micro-Metallic
Corporation in New York. This work
has been reported elsewhere,(!'+?)

A furnace with a suitable reducing
atmosphere has been made available at
ORNL and used for further work on this
problem. Several variables of the
bonding technique, including sintering
temperature, sintering time, fuel-
component particle size, surface
preparation, and sintering under load,
have been investigated. Samples
processed to include these variables
were also cold worked and resintered,
and they were examined metallographi-
cally for evaluation of results. Only
the sintering temperature and the
fuel-component particle size had
appreciable effect on the bond. Both
the higher temperatures and larger
particle sizes tested were shown to
be beneficial., ' ‘

(l)MetalZurgy Division:Quarteriy Progress
Report for Period Ending April 30, 1951, ORNL-
1033, p. 54.

Metallurgy Division Quarterly Progress
feport for Period Ending July 31, 1954, ORNL-1108,
p. 37 _

 

131
ANP PROJECT QUARTERLY PROGRESS REPORT

The materials used i1n these bonding
experiments included type-302 stain-
less steel (-325 mesh) as the metallic
component in the powder mixture and
UO, fired to 2100°C in a hydrogen
atmosphere as the fuel component,
Standard U,S., screens were used for
sizing particles. A Globar heated,
ceramic-tube furnace was used for all
the runs excepting those requiring
loading of the samples. These were
carried out in a larger, molybdenum-
wound furnace equipped with an Inconel
tube. The furnace atmosphere in every
instance was hydrogen that had been
passed over a platinum catalyst and
through an activated-alumina drying
column, The type-316 stainless steel
stock was annealed i1n the hydrogen
atmosphere prior to coating with the
powder mixture.

Sintering Temperature. The trend
was toward improved strength of bond
between the backing sheet and powder
with 1ncreasing sintering tempera-
tures, After sintering at 1300°C and
lower, bonding of the 30% by volume
(-200, +325 mesh) U0,~70% (-325 mesh)
type-302 stainless steel mixture used
was very poor, whereas the samples
sintered at 1360°C or higher showed
considerable tenacity even on bending.

The four samples shown in Fig. 56
were sintered at 1300, 1325, 1360,
and 1385°C, respectively, The powder
mixture was placed on the backing and
leveled to approximately 0,030 in. in
thickness, Retention of the powder
became successively poorer as the
sintering temperature was lowered. In
an effort to consolidate the powder
layer, the composites were rolled to
a 40% reduction in area and heated to
1250°C for 30 minutes, These samples
are also 1llustrated in Fig. 56,

Sintering at high temperatures
caused transformation of some of the
austenite to delta ferrite, which
appeared as needle-like crystals in

132

the type-302 stainless steel matrix
and as a band about individual grains
of the type-316 stainless steel.
The latter distribution of the ferrite
might be expected to lead to a pref-
ferential grain boundary attack 1f the
backing were subsequently exposed to
one of the coolants of current interest.
The presence of the magnetic phase 1n
the type-316 stainless steel was made
evident by placing adrop of an aqueous
suspension of a colloidal 1ron com-
pound on the surface of the metallo-
graphic specimens and then exposing
the samples to a magnetic field.

Sintering Time. The sintering-time
study indicated there was no appreci-
able increase 1n the quality of the
bond obtained with samples heated for
times up to 2 hr over that for a sample
held at temperature for 30 minutes.
Duplicate samples from this group were
also given a 40% cold reduction and
reheated to 1250°C for 30 minutes. In
every case the rolled and sintered
samples showed well-consolidated
structures with relatively small
amounts of porosity. The U0, retention
in several instances was poor and the
UO, could be removed by polishing.
The poor retention is thought to be
caused by partial crushing of the
oxide particles during rolling.

Fuel-Component Particle Size. The
quality of bond between the metallic
particles in the powder layer and
backing sheet showed general improve-
ment as the particle size of the UQ,
was 1ncreased, Particle sizes ranging
from -100,+140 mesh to =325 mesh were
used. The better continuity of
metallic particles with the larger
sizes is shown in Fig. 57. Both the
material sintered at 1300°C for 45 min
and its the rolled and resintered
counterpart are illustrated in the
figure. Cold rolling followed by
reheating at 1250°C resulted in
considerable compacting, as shown in

Fig. 57.
tel

     

_Fig. 56.
sintered.

O - A d’jvusfi&@!

 

%*

  

I 260°C <
(250 X).

#*

,§.'

 

   
 

 
 
 

fi385°c  A4__
(200 X}

g

Effect of Sintering Temperature on Loose Powder Bonding. Sintered 45 minutes.
(b} Cold rolled to 40% reduction in area and resintered 30 min at 1250°C.

 

   
    

(a) As

¢S6T ‘0T 3NNT ONIGNH doTIvdd HOA
VET

 

 

Y-5617 4

 

(=100, + 140 MESH)

’ SRR ‘
WY 5508
.

(-140,+200 MESH)

(- 200, + 325 MESH)

 

(—325 MESH)

 

(g),,,. | I - (b)A

Fig. 57. Effect of Fuel Component Particle Size on Loose Powder Bonding. Sintered 45 min at 1300°C.

{a) As sintered,

{b) Cold rolled to 40% reduction in area and resintered 30 min at 1250°C. 200X.

LHOdAY SSHYI0Hd ATHALYVNO LOAT0Ud dNV
w

Surface Preparation. Contrary to
the results obtained with samples
sintered at the Micro-Metallic Corpo-
ration plant, surface preparation of
the type-316 stainless steel sheet
had little effect on the degree of
bonding to the powder laver. Some
small benefit may have been obtained
in roughening by sand blasting or by
using annealed as-received stonck as
opposed to electropolished sheet.

Sintering Under Load. Samples were
loaded during sintering with steel
weights to obtain unit pressures of
1/2 to 5 psi. Cross sections of these
samples i1ndicated that no gross
improvements were effected by the
maximum loading used.

CONTROL ROD FABRICATION

Several important decigiocuns have
been reached concerning materials to
be used in these rods, and fabrication
of the components has begun. After
considering the availability of stock
for the cans and the time scheduls,
1t was decided to make the cans from
the types-316 and -304 stainless steel
tubing on hand. These steels are
admittedly not the best materials for
compatibility with boren carbide and
perhaps not the best from the self-
welding viewpoint; however, the delay
necessary to obtain the more desirable
type-430 stainless steel tubing is
prohibitive. The cans are being
fabricated by the Y-12 Shops, and they
will be loaded with slugs and brazed
with Colomonoy Nicrobraz by the Welding
Group. ' ‘

Safety Rod Slugs. The safety rod
slugs are being fabricated by hot
pressiog an iron-boron carbide mixture
containing 56% by weight (80% by
volume) boron carbide. The mixture
was selected as an alternate to pure
boron carbide because the latter offers
speciral fabricatiocu preoblems and is

A A e 8RB B b R R S s n A mm mmm e

FOR PERIOD ENDING JUNE 1¢, 1952

considerably more exvensive. The
slugs are prepared from boron carbide
that has been milied 16 hr in a steel
mill with steel balls and blended 8 hr
with the requisite amount of -325 mesh
iron powder, The mixture i1s then
fabricated by hot pressing at 1520°C
and 2500 psi. Enough slugs to prepare
one safety rod have been pressed with
little difficulty., Cracking of the
samples because of the difference in
thermal expansion of the iron~-boron
carbide compact and the graphite
mandrel ‘'was overcome by ejecting the
mandrel at temperature, The pressed
parts have densities of 2.80 g/cc
{about 80% of theoretical) and will
be brought within acceptable dimen~
sional tolerances by grinding. :

Four slugs less than full size have
been pressed, canned, and brazed: two
in type-304 and two in type-316 stain-
less steel, One can of each material
was sectioned and examined after the
brazing cycle., There appeared to be
no appreciable interaction between the
cermet insert and the containecr. The
other two samples were held for 100 hr
at 815°C before cpening for exami-~
nation. In each of these a thin layer
of a white, crystalline compound that
was slightly soluble in hot water was
found on the inside surface of the
container, The white, crystalline
compound was present 1n greater cone-
centration at points wet by the
Nicrobraz. Efforts are being made to
establish the identity of the compound
by x-ray-diffraction technigques.

Regulating Rod Slugs. The require-
ments for the regulating (shim) rod
are somewhat different., For this
purpose a small amount of beron (3 to
12.5 g) wust be uniformly dispersed
in a material relatively transparent
to neutrons, such as Al,0,., A l-in.
by 1/2-in,-dia slug was made by hot
pressing a mixture composed of 99%
grade 38-500 Al,0, plus 0.17% B,C at
at 1750°C in a graphite die under a
ANP PROJECT QUARTERLY PROGRESS REPORT

pressure of 2500 psi. Metallographic
examlination revealed that the two
materials are compatible. A second
slug containing 0,74% by weight of
-325 mesh B,C, prepared by the same
technique, had a density of 3.45 g/cc
(86% of theoretical) and possessed
satisfactory physical properties. The
composition of this compact corresponds
closely to that of material to be used
in the "high-B,C content" regulating
rad,

MECHANICAL TESTING OF MATERIALS

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

The installation of facilities for
stress-rupture testing in liqguid
metals and fused salts is complete.
The testing machines were calibrated
with a standard-sheet test specimen
on which SR-4 strain gages had been
mounted, and this specimen i1n turn was
calibrated agailnst a standard proving
ring. The testing chambers were
originally designed to contain sodium
or other low-melting-point materials.
With the change in the program and the
need for data from tests using the
higher melting fused fluorides, a few
minor design changes had to be made.
Containers to charge and empty the
testing chambers have been designed
and are being constructed of Inconel,
These containers are designed to also
fit the apparatus used to produce the
fluoride mixtures., Testing will be
started shortly after the first lot
of fluoride mixture 1s received,

The primary purpose of these tests
will be to study the effect of stress-
corrosion on the creep rate and stress-
rupture life of these materials. Both
Inconel and type-316 stainless steel
are being tested in the new machines
in an argon atmosphere in the same
stress ranges as they were previously

136

tested in the old machines. A com-
parison of these two sets of data will
establish the effects of the inherent
variables of the new testing systems,
and thus the results of the corrosive
action can be studied.

Stress-Rupture Tests im Argon. Both
fine- and coarse-grained Inconel
specimens are being tested at stresses
from 1500 to 4500 psi in argon at
815°C., The expected test duration in
this stress range 1s from 800 to
3000 hours. Upon completion of these
tests, revised design curves will be
compiled. A short series of tests at
815°C in air are being initiated to
obtain data for a comparison of the
effects of oxidation and fluoride
attack on the creep rate and rupture

1ife.

Type-316 stainless steel specimens
are being tested 1in argon in both the
old and the modified test chambers at
5300, 5800, 6300, 6800, and 7300 psi.
These duplicate sets of data will
serve to correlate the work 1n the
two types of testing chambers and
will furnish data for a design curve
for type-316 stainless steel tested
1n argon.

Stress-Corrosion Tests (G. M,
Adamson, K. W. Reber, Metallurgy
Division)s. Preliminary experiments
to determine the effect of fluoride
corrosion on the physical properties
of structural metals have continued
in the crude apparatus currently
available pending the installation of
more sensitive equipment. Both stress-
rupture and tube-burst tests were
carried out during this period.

The tube-burst tests have been
described in previous reports; however,
it should be noted that these tests
are designed to yield hoop stress-
rupture data. Table 21 is a summary
of the data obtained by immersing the
specimen under the desired stress 1n
NaF-KF-LiF-UF, (10.9-43.5-44.5-1.1
mole %) at 1500°F.

The stress-rupture apparatus trans-
mits a dead-load stress to the specimen
by means of a lever arm and a bellows
arrangement, A thin flat specimen is
subjected to a tensile stress while
immersed in a bath at the desired
temperature. For all these tests a
bath of NaF-KF-LiF-UF, held at 1500°F
was used. The data obtained are
tabulated in Table 22, The third

TABLE

FOR PERIOD ENDING JUNE 10, 1952

column i1n this table presents data
obtained 1n argon atmospheres.

It 15 readily apparent that a
contradiction is present in these
data. At high stresses the two sets
of data check, whereas at stresses
below 10,000 psi there is no corre-
lation. In such testing, as the
stresses are lowered the relative
effects of errors become larger. It
is apparent that the corrosion-stress-
rupture apparatus i1s not sufficiently
sensitive at the very low stresses.

21

Tube-Burst Data for Structural Metals Tested in NaF-KF-LiF-UF,

 

 

 

 

 

 

 

 

 

 

 

 

STRESS HOURS TQ RUPTURE INCREASE IN
(psi} In Inconel In Type-316 Stainless Steel DIAMETER (%)
1200 1000* L7
1600 - 1000+
2135 6588 .

2030 - 880
24990 128 2.2
*Did not rupture.
TABLE 22
Inconel Sttess-aupture Data
, HOURS TO RUPTURE
STRESS (psi) In NaF-KF-LiF-UF, In Argon Atmosphere
12,500 6.5 6
10,000 20 19
7,500 31 80
21.5
14.5
5,000 57504) 270
4,000 661¢8) 430

 

 

 

(a)Test apparatus fajiled before rupture.

(b

Specimen ruptured in weld area of grip rather than in test section,

137
ANP PROJECT QUARTERLY PROGRESS REPORT

CONE-ARC WELDING

P, Patriarca G. M., Slaughter
Metallurgy Division

It was found desirable to determine
the applicability of cone-arc welding
techniques to fabrication of tube-to-
header joints in heat exchangers.
This study was given precedence over
the long-range approach of determining
the individual and combined effects
of the various cone-~arc welding
process varliables. However, the
variables, such as arc time, arc
current, and electrode gap distance,
have been extensively tested on
materials and geometries applicable
to heat exchanger designs. One heat
exchanger was fabricated, and the
fabrication of test assemblies by
cone-arc welding will continue.
However, studies to evaluate the
fundamental effects of the cone-arc
welding variables will receive con-
current attention in future work.
Eguipment used for cone-arc welding
experiments was described previ-
ously, (3

Effect of Welding Conditiouns. A
series of headers were fabricated
to conform to a close-packed tube and
hole arrangement in a circular header.
The header material used was type-304
stainless steel sheet 1/8 in. thick.
Nineteen tube holes were drilled to
receive type-316 stainless steel
tubing, 0,100 in. OD with 0.010-1in.
wall thickness. The basic pattern
was an equilateral triangle with a
tube hole at each apex. Each tube
was 0,100 in., or one diameter, from
its nearest neighbor. Tube-hole to
header-edge distances were initially
chosen as 0.050 in. but were increased
in subsequent experiments to 0.100

 

(3)

Airceraft Nuclear Propulsion Project Quarterly
Progress Reports for Periods Ending December 10,
1951, ORNL-1170, p. 132; March 10, 1952, ORNL-
1227, p. 142.

138

inch. Suitable welding conditions
were chosen individually by experiment
be fore being applied to a 19-hole
header assembly for a consistency
evaluation,

Figure 58 illustrates a series of
typical tube-to-header consistency
determinations., Each header was
fabricated by using adifferent welding
condition, as set forth in Table 23,
to illustrate the flexibility of the
cone-arc process.

Cone-arc welds within the limits of
the material and joint design of this
investigation may be made over a range
of welding conditions, Although the
welds 1llustrated by Fig. 58 and
Table 23 are not representative of a
systematic study, theilr examination
reveals trends to be verified by future
work, It appears from comparison of
these welds that the primary effect of
shortening the arc distance 1s to
recess the cone-arc weld. Although

welds a, b, and ¢ illustrate the com-
parable net effect of increasing
. R Y. 6667
848 S hee o gy
2pH  » Q'.‘;j 2L $a 9
Qfifi@@iii&-@g ¢ o990y
&!l!}fi et e L 2 bee
‘ ‘yl .‘:‘m * ' ’ . \ ‘ ‘ ‘
o) (&) {¢)

 

! i l i {

Fig. 58. Typical Cone-Arc Welds.
For description of the welding con-
ditions for each of these samples see

Table 23.
FOR PERIOD ENDING JUNE 10, 1952

TABLE 23

Cone-Arc Welding Conditions for Tube-to~Header Joints

 

 

 

HEADER DESIGNATION | PISTANCE FROM ELECTRODE | ppc cURRENT | ARC TIME
(Ref. toFig. 58) |  THPTOPLANE OF WORK (amp) (sec)

a 0.020 64 1.5

b 0.050 60 1.8

¢ 0,040 70 1.2

d 0,100 70 ' 2.0

e 0.100 84 1.4

f 0.050 60 1.8

 

 

 

 

current with a corresponding decrease
1n arc time, 1t 1s not believed possible
to determine these effects on variables
such as weld penetration without
thorough metallographic examination.
Welds d and e illustrate the combined
effects of further increasing arc
length, arc current, and arc time - a
condition is reached that indicates
superfluous welding, This is apparent
by observing the relatively large welds
and the tendency to form a rosette
pattern on the header edges. The heat
flow pattern, the uniformity of which
seems to be an important variable in
cone-arc welding, appears to have been
distorted on prolonged heating. It may
be noted that the outer holes of headers
e and f do not contain welded tubing.
The tube-periphery to header-edge
distance in this case was 0,050 inch,
This distance was found to upset the
uniformity of the heat flow pattern and
welding conditions were not found which
would give satisfactory welds.

There 18 a tendency for an are that
is terminated instantaneously without
a current taper to leave a crater.
Examination of the periphery of cone-
arc welds reveals a small spot that is
characterized by a small amount of

scale. The extent of these craters
indicates that porosity due to cratering
1s not a serious problem. However,
the presence of a single pronounced
crater on the cone-arc weld periphery
tends to confirm the belief that the
cone-arc may be a single, rotating-arc
beam. : *

Heat Exchanger Assembly. Cone-arc
welding as applied to a heat exchanger
assembly is illustrated in Fig. 59.
This photograph shows one end of an
Inconel counterflow heat exchanger
consisting of 52 tubes, 0.150 1in.
OD with 0.,025-in, wall thickness,
cone-arc welded into a cylindrical
Inconel header, 3 in, ID with 1/8-in.
wall thickness. A total of 104 welds
(the other end of the heat exchanger
is identical to the one illustrated),
were made by using the following
conditions: :

Arec current 102 amp

Are time 2 gec
Electrode 3/3%-in,-dia thoriated-

tungsten, pointed
Distance from elec-
trode to plane of
work : 0.165 in.

Shielding gas -
argon 30 e¢fh

139
ANP PROJECT QUARTERLY PROGRESS REPORT

 

Fig. 59. One End of A cone-Arc-~
Welded Inconel Heat Exchanger As-
sembly.

TESTS OF BRAZING ALLOYS

P, Patriarca G. M. Slaughter
Metallurgy Division

The primary objective of the high-
temperature brazing alloy investigation
during the past few months has been to
screen brazing alloys for aircraft
reactor applications by using flow-
ability and static corrosion tests 1m
sodium hydroxide and NaF-KF-LiF-UF,
fuel mixture. After these pre-
liminary investigations, promising
alloys can be further studied in
physical property tests such as butt-
tensile tests (at both room and elevated
temperatures), elevated-temperature
creep tests, dynamic corrosion tests,
and room- and elevated-temperature
ductility tests. Such screening
procedures seem appropriate 1n view
of the large number of alloys now
under consideration,

In Table 24 the results of pre-
liminary investigations of eight high-
temperature brazling alloys are re-
corded. Statlc corrosion tests on

140

joints brazed with some of these alloys
are being made, and the results will
be reported later. The melting points
of the various alloys are only approxi-
mate and will vary somewhat since
chemical analyses of various heats of
the same alloy differ slightly,
Although most of the brazing alloys
tested flowed satisfactorily, none
exhibited satisfactory corrosion
resistance 1in sodium hydroxide.
Nicrobraz, 60% Pd-40% Ni (with and
without a 3% Si additien), and 64% Ag-
33% Pd-3%Mn were the only alloys not
severely attacked by the fluoride fuel
NaF-KF-LiF-UF, (10,9-43,5-44.5-1.1
mo le %).

At the present time, static tests
on Anickel joints brazed withpromising
alloys are being conducted. It 1is
hoped that better resistance of the
brazing alloys to corrosion by sodium
hydroxide can be obtained, since
nickel 1s attacked only slightly by
NaOH ~ this may not hold true with
the introduction of a bimetallic
system,

Nicrobraz, Static corrosion tests
of Nicrobrazed joints on Inconel and
type-316 stainless steel 1in both
NaF-KF-LiF-UF, fuel mixture and NaOH
for 100 hr at 1500°F have been com-
pleted. Photomicrographs of corrosion-
tested, Nicrobrazed, Inconel joints
were presented in aprevious report. (%)
Nicrobrazed specimens of type-316
stainless steel tested in both the
fluoride mixture, NaF-KF-LiF-UF,, and
NaOH are shown in Fig, 60, The
fluoride bath had a relatively minor
corrosive effect onthis joint, whereas
the sodium hydroxide was much more
severely corrosive.

Manganese-Nickel Alloy. The 60% Mn-
40% N1 alloy was found to be extremely
brittle and many brazed joints made
with it were subject to severe cracking,
This brazing alloy was found to have

 

op. cit., ORNL-1227, p, 145-146.
1¥T

TABLE 24

-

nC
A

Summary of Flowability and Corrosion Tests of Various Brazing Alloys

 

 

STATIC CORROSION RESULTS

 

 

APPROXIMATE STATIC CORROSION RESULLTS
BRAZING ON BRAZED INCONEL ON BRAZED
BRAZING ALLOY “fi;:fi:s TEMPERATURE J::Qg:ifi;gii TYPE-316 STAINLESS STEEL
(°F) (°F) In Fluoride In Fluoride
No. 14 In NaOH No. 14 In NaOH

Nicrobraz 1850 2050 Excellent Slight at- Extremely Moderate Severe

(70.17% Ni-13.95% Cr-5.86% tack with severe

leaching

Fe-~4,59% Si~4.92% R) of boron

60% Mn-40% Ni 1859 1960 Poor flow, Severe Severe
. L ..cracking . |. o
prevalent
60% Pd-40% Ni 2260 2320 Excellent No apparent Severe
attack

{60% Pd—40% Ni) + 3% Si L2150 2200 -Excellent Moderate: - Severe Moderate - Severe
16.5% Cr-10.0% Si-73.5% Ni 2100 2200 Moderate
16.5% Cr-10.0% Si-2.5% 2100 22060 Moderate Severe Severe Moderate Severe
‘Mn—71‘0% 8@
64% Ag-33% Pd-3% Mn 2130 2200 Excellent Moderate Severe VYery slight Severe
5% Ag-20% Pd-5% Mn 2100 2200 Excellent Severe Severe

 

 

 

 

 

 

 

 

cset ‘0T ANNL ONIaNd d0I¥dd 404
ANP PROJECT QUARTERLY PROGRESS REPORT

 

(a) Tested in NaF-KF-LiF-UF,

#2< =@#@.'”’ *‘@' '

vy J‘ /—*V\w 'Z - . e

 

(b) Tested in NaQH

Fig. 60. Type-316 Stainless Steel Joints Nicrobrazed and Static Corrosion
Tested for 100 hr at 1500°F. FEtched with agua regia. 200X.

142
low resistance to corrosion ian the
media in which 1t was tested; 1t was
severely attacked by the fluoride
mixture, NaF-KF-LiF-UF,, and NaOH (as
shown in the previous quarterly
report(3))., The poor corrosion re-
sistance coupled with relatively poor
flow properties seem to make this
alloy unsuitable for c1rculat1ng~fue1
reactor applications., ~

Palladium-Nickel Alloy. The 60% Pd-
40% Ni brazing alloy has excellent
flowability properties, and the rather
limited experimental data obtained to
date indicate that 1t has better than
average resistance to corrosion by the
fluorides, The insignificant attack
of the fluoride mixture, NaF-KF-LiF-UF,
on an Inconel joint brazed with this
alloy 1is shown in Fig. 61. It can be
seen that i1ntimate bonding with the
base metal 1s obtained during brazing,
since Lhe grain boundaries of the

.......

as the parent Inconel grains.

The 60% Pd-40% Ni alloy, with 3%
silicon added for 1ts effect upon
melting-point lowering, was also in-
vestigated. The s1licon addition
caused phenomenal flow that traversed
the entire: length of a 6-in. joint and
even flowed up the scratches., However,
the room-temperature tensile strength
of type-316 stainless steel joints
butt-brazed with the alloy was low,
and the brazed joint was very brittle.
In addition, the resistance to corrosion
in the fluorides of brazed Inconel
joints seems to have been lowered
somewhat by the addition of the silicon
to the 60% Pd-40% Ni alloy.

Further attempts to lower the
melting point of the 60% Pd-40% Ni
alley by the addition of a third
component are being made. A 1% beryl-
lium addition and a 10% manganese

G rbea., p. 147,

 

FOR PERIOD ENDING JUNE 10, 1932

addition are being studied, but more
work has to be done before the desir-
ability of such additions can be
determined.

Ni-Cr-Si-Mn Alloys. Two brazing
alloys of the Ni-Cr-51-Mn type have
been investigated in preliminary
experiments., The 16.5% Cr~10.0%
Si~73.5% Ni and the 16.5% Cr~10,0%
S1-2.5% Mn~-71.0% Ni alloys have
desirable flowability characteristics,
but they are apparently attacked to
some extent by the fluorides and rather
severely by sodium hydroxide. Static
corrosion tests for 100 hr at 1500°F
on Inconel joints brazed with the
alloy containing a small amount of
manganese have been completed. Typical
sections of joilnts show moderate
attack in the fluoride, NaF- KF—LiF—UF‘,
and rather severe attack in sodium
hydrox1de‘

Silver~-Base Alloys. Two silver-base

~brazing alloys of the compositions

64% Ag-33% Pd-3% Mn and 75% Ag-20% Pd-
5% Mn, which melt at approximately
2100°F, have also been investigated.
The flowability of these alloys 1s
very good, but the static corrosion
results look promising in only a few
instances, Inconel and type-316 stain-
less steel joints brazed with the
64% Ag-33% Pd-3% Mn alloy were static
corrosion tested for 100 hr at 1500°F
in both NaF-KF-LiF-UF, and NaOH. The
fluorides were relatively noncorrosive
(Fig. 62), particularly on the stain-
less steel, whereas the sodium hy-
droxide was severely corrosive.

Inconel joints brazed with the
75% Ag-20% Pd-5% Mn alloy have been
tested in NaF-KF-LiF-UF4 and NaOH for
100 hr at 1500°F. The resistance to
corrosion of both joints in these
media was poor, as evidenced by severe
attack, particularly in sodium hy-
droxide. '

143
ANP PROJECT QUARTERLY PROGRESS REPORT

CERAMICS
J. M. Warde, Metallurgy Division

National Bureau of Standards ceramic
coating A-418 has been successfully
applied to samples of stainless steel,
and 1t showed good resistance to
oxidation at 900°C in a 100-hr test.
Work is in progress to apply a ceramic
coating to nickel and to develop a
technique for coating a complete
assembly of a radiator intended for
the ARE. A study of the effect of
ceramic coatings on the thermal effi-
ciency of a radiator is under con-
sideration. Moderator blocks for use
in the ARE have been fabricated from
beryllium oxide at the U. S. Bureau of
Mines Station at Norris,

Ceramic Coatings for Metals. Work
was continued during the past quarter

on the application of ceramic coatings
to metals for possible ln a
radiator. National Bureau of Standards
coating A-418 applied to about 2 mils
thickness to three stainless steels
(types-302, -316, -347) successfully
retarded oxidation of the metals in a
100-hr test at 900°C. The frit com-
position of this coating has been
described in a previous report. (%
An attempt to apply a ceramic coating
to nickel is now in progress and a
technique for coating the complete
radiator assembly 1s under con-
sideration. It is proposed to construct
a thermal conductivity apparatus for
the purpose of determining the effect
of ceramic coatings on the thermal
efficiency of the radiator,

use

 

 

Fig. 61. Inconel Joint Brazed with a 60% Pd-40% Ni Alloy and Static Corrosion

Tested in NaF-KF-LiF-UF4 for 100 hr at 1I500°F,

144

Etched with agqua regia. 100X,
FOR PERIOD ENDING JUNE 106, 1952

 

(b) Type-316 stainless steel joint. Unetched.

Fig. 62, Joints Brazed with a 64% Ag-33% Pd-3% Mn Alloy and Static Corrosion
Tested in NaF-KF-LiF-UF, for 100 hr at 1500°F. 100x.

145
ANP PROJECT QUARTERLY PROGRESS REPORT

Ceramics Laboratory Program. Among
the projects planned for the Ceramics
Laboratory that are of particular inter-
est to the ANP are the investigation
of ceramic metal combinations (cermets)
for reactor components, radiation

damage studies of ceramic materials, and
a program of measurement of physical
properties of ceramic materials for use
in reactor design, These programs are
developing as rapidly as equipment can
be completed and installed.

 

13.

HEAT TRANSFER AND PHYSICAL PROPERTIES

RESEARCH

H. F. Poppendiek

Reactor Experimental Engineering
Division

The viscosities of NaF-KF-LiF-UF,
(10.9-43.5-44.5-1.1 mole %) and two
BeF,-bearing fluorides have been
determined with a rotational vis-
cometer. The viscosities of NaF-KF-
LiF-UF, range from about 9 centipoises
at 530°C to about 3 centipoises at
750°C. The viscosities of both BeF,-
bearing fluoride mixtures were so
high, greater than 12 centipoises
at 725°C and greater than 25 centi-
poises at 600°C, that BelF, was eliminated
from*further consideration as a
potential fuel constituent for the
ARE. Preliminary viscosity measure-
ments of the NaF-KF-ZrF, -UF, fuel
mixture (5-51-42-2 mole %) have been
obtained with the efflux viscometer.
The viscosities, which were from about
10 centipoises at 570°C to about 4
centipoises at (50°C, appear satis-
factory for the ARE.

Thermal conductivity measurements
on NaF-KF-LiF (eutectic) have in-
dicated a mean value of about 2.6
Btu/hr-ft? (°F/ft) over the tempera-
ture range 500 < t < 750°C. Heat
capacity data on barium and potassium
hydroxide have been determined. The
density-temperature relations of
several of the zirconium fluoride
salt mixtures have been determined,.
In addition, the vapor pressures of
BeF,- and a ZrF,-bearing fuels have
been measured. A table summarizing
some of the physical properties of
materials of interest to the ANP
Froject is included.

146

Experimental determinations were
made of the heat transfer coefficients
of molten sodium hydroxide flowing
turbulently in a heated tube. Results
indicated that sodium hydroxide heat
transfer was similar to convective
heat transfer of ordinary fluids.
Fundamental experimental information
on heat transfer of turbulently flowing
mercury in thermal entrance regions
has been obtained; the experimental
results are 1n agreement with previous
theoretical determinations. Mathe-
matical analyses pertaining to circu-
lating-fuel heat transfer systems
have been developed. In one analysis
a study of the thermal structure 1in
the entrance region was made, and
another anmalysis considered non-
isothermal heat transfer. Some ex-
perimental velocity information on
natural convection in an idealized
liguid fuel has been obtained.

VISCOSITY OF FLUORIDE MIXTURES

J. M. Cisar, ANP Division

F. A. Knox F. Kertesz
Materials Chemistry Division

R. F. Hedmond T. N. Jones
Reactor Experimental Engineering
Division

Study of the viscosity of fluoride
mixtures has been continued with both
the rotational (Brookfield) and
efflux viscometers that were modified
tosebtain control over the melt.
In the case of the efflux viscometer,
the whole system has been housed in a
small dry box so that determinations
can be made in inert atmospheres.
The efflux viscometer and the can-
tainer for the test liquid are instru-
mented with several thermocouples so
that a reasonably accurate value of
the liquid temperature can be obtained
while making determinations. Tempera-
ture errors due to thermal radiation
losses have been reduced in the
system by reducing the radiation-
configuration factor. The dry box
is divided inteo two sections; one
contains the furnace, viscometer, and
balance, and the other can be used
to load containers with test materials.
The Brockfield apparatus was similarly
enclosed.

With the modified apparatus, the
viscosities of several BeF,-bearing
fluorides, one ZrF, -bearing fluoride,
and several other fluoride mixtures
were determined. The two BeF ,-bearing
fluorides had viscosities greater
than 12 centipoises at 725°C and
both approached 25 centipoises at
650°C. These viscosities were too
high toe permit satisfactory heat
transfer in the experimental reactor,
and as a result a search is being
made for a lower viscosity fuel.
The one ZrF,-bearing fluoride measured
has a satisfactory viscosity, that is,
8 centipoises at 600°C, which de-
creases to 4 centipoises at 750°C.

BeF,-Bearing Fluoride Mixtures,
The viscosities of two Bel,-bearing
mixtures were determined with the
Brookfield viscometer in the range
600 to 800°C. For these measurements
1t was necessary to minimize the
formation of beryllium oxide by
bubbling HF gas through the molten
liquid prior to the run and maintain-
ing an atmosphere of HF over the melt
during the measurement. The first
mixture contained 31 mole % BeF,,

FOR PERIOD ENDING JUNE 10, 1952

47 mole % NaF, and 2 mole % UF, and
showed viscosities from 12 centipoises
at 800°C up to about 25 centipoises
at 650°C. Another preparation of
considerably lower Bel, concentration
contained 12 mole % BeF,, 76 mole %
NaF, and 12 mole % UF, and was con-
siderably less viscous at the high-
temperature end of the range. Iowever,
the viscosity (5 centipoises at 810°C)
rose rapidly as the temperature was
lowered, reached 13 centipoises at
725°C, and approached 25 centipoises
at 650°C, :

These findings are in agreement
with the fact that BeF, is a glass
formed with directionally bonded
atoms; 1ts viscosity increases with
decreasing temperatire considerably
more rapidly than the viscosity of
other ionically bonded salt mixtures.
These high viscosities have virtually
eliminated BeF, from the list of
potential compounds for the ARE fuel
mixture. '

ZrF4~Bearing Fluoride Mixtures,.
Preliminary viscosity measurements
have been obtained for a ZrF -bearing
fuel mixture, Nal' ~KF-ZrF, -UF, (5-51-
42-2 mole %), on the efflux viscometer.
The results are shewn in Fig. 63.
Although the viscosity is satisfactory
for the present ARE, this mixture
does not contain sufficient uranium
for criticality. However, the in-
crease in the uranium concentration
that would be required should not have
a significant effect on the viscositvy.

Jn an attempt to determine possible
variations of viscosity with various
compositions of the mixture, measure-
ments have been made with the rotational
viscometer on a mixture of 55 mole %
ZrF -45 mole % KF after additions of
varying guantities of NaF, The
viscosity of the ZrF, -KF binary was
guite low: 3 centipoises at B00°C and
8.5 centipoises at 600°C, Four
successive additions, which brought

147
ANP PROJECT QUARTERLY PROGRESS REPORT

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

R
DWG. 15351
17¢ (oK)
0.0009 0.0010 0.0011 0.0012 0.0013 0.0014 0.0015
10 ,
| | T 1 |
- P
g8 | // /4
NaF - KF —Zrfg ~UFg — s }
(5.0 ~51.0-42.0-2.0 mole % ); ~ NaF- KF — LiF —UF,
6 (ClsAR) 1 (109-435-44.5-1.1 mole %)l
[ 7 / (REDMOND AND JONES)
2 T
Ly
~§ 4 f/ //////
> v
|-»,_
& /
o /
> ——— NoF - KF - LiF
H (11.5 -42.0-46.5 mole % )
g 2 . (KAPLAN AND TOBIAS)
7))
m
<
1 L | | | |
800 700 600 500 400
7 (°C)
Fig. 63. Absalute Viscosity of Several Fluoride Mixtures as a Function of

Temperature.

the NaF content up to 20% by weight,
resulted in no detectable change in
the viscosity.

Material from Corrosion Loops.
A number of dynamic corrosion loops
of type-316 stainless steel have
shown signs of plugging during circu-
lation of NaF-KF-LiF-UF, mixtures.
The melting point of the fluoride
mixture remained unchanged during
the runs and, i1n some cases, no
physical evidence of a plug was
found. Repeated determinations have
shown that the viscosityof the mixture

148

after circulation was identical,
within experimental error, with that
before circulation. Measurements
made immediately after melting the
mixture were also identical with
those made after keeping 1t molten
under a blanket of inert gas for

24 hours.

NaF-KF-LiF and NaF-KF-LiF-UF,
Mixtures. The falling-ball apparatus
has been used to obtain the viscosity
of NaF-KF-LiF (11.5-42.0-46.5 mole %)

at several temperatures in order to
check earlier results by other methods
and to study behavior of the apparatus
~ the results checked satisfactorily.
The modified Brookfield viscometer
has been used to measure viscosities
in the 10-centipoise range at high
temperatures. Data on the viscosity
of NabF~KF-LiF-UF, (10.9-43.5-44.5-1.1
mole %) have been obtained with this
instrument. The results are shown
in Fig. 63.

THERMAL CONDUCTIVITY

W. D. Powers
B. M. Burnett:

L. Cooper
S. J. Claiborne

Reactor Experimental Engineering
Division

Three separate experimental de-
terminations of the thermal conductivity
of NaF-KF-LiF (eutectic) are presented
in Table 25. These determinations
were made with a modified, Deem-type
apparatus that had previously been
checked by making measurements of
fluids of known conductivity. An
error analysis of these thermal con-
ductivity measurements indicated that
no abnormal deviations existed. [t
1s next planned to measure the con-
ductivity of NaF-KF-LiF-UF, and the
ZrF,-bearing fuel mixtures. :

TABLE 25

ThQrmal Conductivity of

 

 

NaF-KF-LiF
TEMPERATURE RANGE THERMAL CONDUCTIVITY
(°C) [Beu/hr £t (°F/ft)]
634 to T47 2.9
490 to 590 2.6
567 to 670 2.5

 

 

The longitudinal heat flow ap-
paratus has been adapted to measure
the thermal conductivity of ligquid

FOR PERIOD ENDING JUNE 10, 1952

metals contained in thin-walled
capsules. A numerical analysis of
the heat flow in the apparatus in-
dicated unidimensional heat flow 1in
the region where measurements are
made. The apparatus is designed to
minimize convective heat flow. Measure-
ments on liquid sodium agree with
the values given in the literature
to within about 10%.

HEAT CAPACITY
W. D. Powers G. C. Blaleck

Reactor Experimental Engineering
Division

The enthalpies and heat capacities
of the following hydroxides have been
determined by the use of Bunsen 1ce
calorimeters:

Barium hydroxide (470 to 900°C)

Hy (liguid) - H o, (solid) = 0,20 T + 7
€y = $,20 £ 0,01

Potassium hydroxide (400 to 950°C)

HT (liQUid) - Hooc {s0lid) = 0.35 T + 58
°, = 0,35+ 0.02

where H 1s in cal/g and T 1imn °C.
At present the heat capacities of
LiOH, Sr(OH),, a NaF-KF-LiF mixture
with and without UF,, and a NakF-KF-
ZrF,-UF, mixture are being determined.

The following empirical heat
capacity relation has been found to
approximate the fused fluoride and
hydroxide data:

¥
Cp“;“z 9,
where ¢, 1s the specific heat 1in
cal/g*°C, M is the average molecular
weight, and m i1s the average number
of species present {the metal and

149
ANP PROJECT QUARTERLY PROGRESS REPORT

fluorine atoms and the hydroxyl
radical being considered as separate
species). This relation follows from
Kopp’s rule and holds to within 20%
for the liguids investigated.

DENSITY

J. M. Cisar, ANP Division

The densities of several fuel
mixtures have been measured 1n the
new dry-box density system designed
to obtain control of the atmosphere
over the melt. The results for
several fuel compositions are given
in Table 26. Further density measure-
ments on the ZrF,-bearing fuels are
being made 1n order to obtain the
uranium concentration vs. density
information necessary to make the
final fuel selection.

VAPQOR PRESSURE 0F FLUORIDES

R. E. Moore

Materials Chemistry Division

The vapor pressures of pure Bef,
and of a typical Zrf,-bearing fuel
mixture have been determined during

the past qgquarter by using the ap-
paratus and procedure previously
described, (12

Beryllium Fluoride. Since Bel,
had earlier been considered as a
possible fuel compomnent,
nation of 1ts vapor pressure was
made. Considerable scatter of data
in the low-temperature region was
observed. At high temperatures
(above 920°C), however, reasonable
reproducibility was obtained. The
vapor pressure data are shown 1in
Table 27 and may be represented by
the equation

a determi-~

8476.6

T(°C)

 

 

log P (mm Hg) + 8.50.

The heat of vaporizationwas calculated
to be 39 kg-cal/mole, and the ex-
trapolated boiling point 1s 1240°C.

LTrF,-Bearing Fluoride Mixtures.
Since fluoride mixtures containing
ZrF, show considerable promise as
fuels, vapor pressure measurements

 

(l)AircraftNuclearPropulsion Project Quarterly
Progress Report for Period Ending September 10,
1951, ORNL- 1154, p. 136,

(2)Aircrafth’uclearPropulsion Project Quarterly
Progress Report for Period Ernding December 10,
1951, ORNL-1170, p. 126.

TABLE 26

Densities of Sowme NaF-KF«ZrF4nUF4 Mixtures

 

 

 

MOLE % (o - g/DcECN_SItTY:—_ ocy TEMPEHA(TOUCH)E RANGE
5-51-42-2 p = 3.78 - 1,09 x 103 ¢ 430 < ¢t < 600
4.8-50,1-41.3-3.8 o= 4,27 - 1.63 x 10°3% ¢ 750 < t < 900
34,7-17.4-44,4-3.5 p = 3.78 - 0.909 x 1073 ¢ 580 < t < 900
35.0-17.6-49.9-2.5 P = 3.65 ~ 0.800 x 103 ¢ 580 < t < 900

 

 

 

150
26, 851 (1925).

TABLE 27

Vapor Pressure of Beryllium Fluoride

 

 

FOR PERIOD ENDING JUNE 10, 1952

TABLE 28

Vapor Pressure of NaB‘~KF~ZrF4-UF4.

 

 

 

 

 

 

TEMPERATURE OBSERVED PRESSURE TEMPERATURE OBSERVED PRESSURE

(°C) (mm Hg) (°¢) (mm Hg)

828 8.7 768 5.0

838 10.6 772 5.8

846 - 10.1 866 13.1

851 12,4 866 13.4

889 13.8 866 13.8

891 13.6 866" 14.5

893 14.5 901 15.7

893 17.0 901 19.3

896 17.1 306 _ 16.9

899 17.8 911 18.1

901 22.5 916 19.3

901 19.5 926 22.7

903 22,5 929 23,0

920 24. 4 931 23.0

925 . 28.5 954 - _ 31.4

935 30.5 956 ' 30.8

966 44,5 961 31.8

966 44.0

984 59.5

987 . 62.5 zation is 24.6 kg-cal/mole, and the
1019 - 89.5 extrapolated boiling point is 1530°C.
1019 88.5 :

 

for a typical composition have been

- performed over the temperature range

768 to 961°C.

The mixture contained

- 42 mole % ZrF,, 51 mole % KF, 5 mole

% NaF, and 2 mole % UF,. The data

- are given in Table 28, and may be
- represented by the equation

5352. 1

70°C) + 5.842.

log P (mm Hg) = -

~In determining the best straight line

on the plot, the higher values were
favored because it is believed that

. some of the low values resulted from

unsaturation. (3)

(3)w

The heat of vapori-

 

Ro&ebush and A, L. Dixon, Phys. Hcv;

Analyses of material distilled out
of the apparatus during the measure-
ments showed that the vapor above the
liquid at the temperatures employed
was virtually pure ZrF,.

At 927°C, the estimated boiling
point of ZrF4,(4) the vapor pressure
of the mixture is only 24 mm Hg as
compared with the 320 mm Hg to be
expected from Raocult’s law. The
considerable negative deviation from
ideality i1s not unexpected since
compounds such as Kj;ZrF,, K,ZrF.,
and KZrF, are known to be quite
stable. - :

 

(4)L. Brewer, The Cheaistry and Metallurgy of
Miscellaneous Materials: Thersodynomics, N.N.E.S.,
Div, IV, Vol. 19B, p. 193 (1930). :

151
ANP PROJECT QUARTERLY PROGRESS

PHYSICAL PROPERTY DATA

Physical Properties Group

Reactor Experimental Engineering
Division

Summaries of pertinent available
data on the physical properties of
fluoride salts and hydroxides
presented in Tables 29 and 30. Most
of the data were obtained by the
Physical Properties Group at ORNL;
however, pertinent physical property
measurements obtained by other organi-
zations are also included. The ORNL
physical property measurements are
being made to quickly supply the ANP
Project with data of reasonable
accuracy. The general accuracies
assigned to the data presented in the
tables are listed below.

are

1. Melting point: within about +10°C.

2. Heat capacity: within about 110%.
3. Thermal conductivity:
checks with the thermal con-
ductivity device for liguids
(made by using molten lead)
dicated that the

recent

1n-
data fell within

+15% of the known values, and an
error analysis of the system
(containing platinum, platinum-

rhodium thermocouples) also
suggested that the data are
probably within 115% of the true
values.

results so far are
so a definite ac-
not yet be assigned.

4. Viscosity:
preliminary,
curacy can

5. Density: within about *5%.

Some preliminary physical property
data are presented; other preliminary
data are merely referred to and are
to be presented in the future.

152

REPORT

HEAT TRANSFER IN SODIUM HYDROXIDE

H. W. Hoffman

Reactor Experimental Engineering
Division

J. Lones

Experimental determinations were
made of the heat transfer coefficients
for molten sodium hydroxide flowing
by forced convection through a heated
tube. Measurements were made 1n the
Reynolds modulus range of 5,000 to
12,000. Results show that molten
sodium hydroxide may be considered an
ordinary fluid - that is, any fluid
other than the ligquid metals - so far
as heat transfer 1s concerned.

The system used to make the study
consistedof two heated tanks separated
by a long tube, part of which was the
heated test section. The test section,
a 24-in. length of the A nickel tube
(0.1875 in. OD, 0.1175 in. ID), was
heated by passing a high-amperage
electric current through the tube
wall. Chromel-alumel thermocouples
(36 gage) were welded to the outside
of the tube wall. The test fluid was
moved by inert-gas pressure from one
tank to the other through the test
section and the fluid flow rate was
determined by weighing the fluid as
it left the test section.

The heat transfer coefficients
were calculated from the experi-
mentally obtained tube-surface and
fluid temperatures and the heat 1input
by the equation:

9
A

t -t
s m

3

where g/Ais the heat fluxin Btu/hr-ft?
through the tube wall based on the
heat gained by the fluid in passing
through the test section, t_ 1s the
€ST

 

TABLE 29

Physical Properties of Fluoride Salt Mixtures

 

 

 

 

 

 

 

 

 

 

 

 

MELTING RANGE
°C)

HEAY CAPACITY
{caljoC-g)

THERMAL CONDUCTIVITY
[Brushe-fe2 (°F/fc}]

 

VISCOSITY
{centipoise)

 

NaF-KF-ZrF4-UF4
5-51-42.2 mole 2

4.8-50,1-41.3-3.8 mole %

 

34.7-17.4-44.4-3.5 nole %

35.0-17.6-44.9-2.5 mole %

NaF-KF-LiF
11.5-42-46.5 mole %

NeF«KF-LiF-UF4
10.5.43.5-44.5-1,1 mole 3

 

450 to 510(0?

450 vo 510(8)

255(¢)

a55ic)

 

 

 

 

 

 

| Prelimipary

0.46 at
500°C <t < ggp°C

results:

 

 

 

=

2.6 at 500°C < i < 750°C

 

, 3 at 750°C

Preliminary resunlts

available(fl)

Preliminary results:

10 at 530°C
3.2 at 700°C
2 at gogog(d)

Preliminary resules:

9 at 530°C
6 at 510°C

 

{Q)J‘

Cisar, memo to be issued.

()1, Ciaar, ORNL CF-52.5.209 (May 2B, 1952).

'(CJC, J. Barton, OBNL Materials Chemistry Division

COmMUB: cation,

» personal

(M. Tobias and S. I. Kaplan, "The Me

 

 

 

 

DENSITY YOLUME EXPANSION

(glec) COEFFICIENT
3.78-1.09 x 10°% ¢ acf 3.29 x 10°% ac 430°C
430°C < ¢ < g00°Cla) | 3,49 x 1o-4 4p gogocis)
4.27-1.63 % 1073 ¢ at] 5.35 x 104 at 756°C
T50°C < ¢ < 960°Ct®) | 5. 83 x 3074 4¢ gggogie)
3.78-0.909 % 1673 tac| 2.79 x 10°% ar 580°C
580°C < ¢ < 900°Ctb) | 3,07 x 19-4 ar gogoc(d)
3.65-0.800 x 1073t ar| 2.51 x 194 a¢ 586°C
580°C < ¢ < 900°CtY) | 9 73 x 1974 4¢ gpgoc(h)
2.39-0.59 x 1073 ¢ ot} 2.84 x 10-4 ar 530°C
530°C < ¢ < 850°Cle) | 3,12 x 1071 4¢ gspoc
2.65-0.90 % 187% ¢ ac| 4.15 x 10"* ar 530°C
530°C < ¢ < 850°C5) | 4,79 x 104 av gsgoC

 

 

 

 

 

February 2, 1952 {personal communication),

Cedy. Civar, ORNL CF:S1-12-91 (Dec 14, 19513 .

(45, Cisar, OBNL CF-51-11-198 (Noy, 30, 1951).

asurement of the Viscosity

of Flinak,”

0T ANAL INIGNT dOoX¥3d Hod

§

GSO6L
ST

TABLE 30

Physical Properties of Hydroxides

 

 

NaOH

Ba(OH)2

KOH

 

Approximate melting
point(“) (°C)

Heat capacity
{(cal/°C-g)

Thermal conductivity

Bru/hr*fe? (°F/fe)]

Viscosity(f)

(centipoise)

Density (g/cc)

 

323

Liquid, 0.49 at
340°C < t < g9p°ci®?

0.81 at 520°C{e)

4.0 at 350°C
2.2 at 450°C
1.5 at 550°C

.786 at 320°C
L7711 at 350°C
.746 at 400°C
722 at 450°C)

o

 

395

0.30 £ 0.01 at
470°C < t < 90p°cte’

 

360. 4

0.35 + 0.02 at
400°C < t < 950°ctd?

.3 at 400°C
at 450°C
at 500°C
at 550°C
at 600°C

O e b e R
o S w =3

141.6 - 0.035 t at
790°F < t < 825°Fle)

 

(Q)C. J. Barton, ORNL Materials Chemistry Division,

personal communication,

{b)y  D. Powers, ORNL CF-51-11-195 (Nov. 30, 1951).

{¢}y. p. Powers and G. D. Blalock, ORNL CF-52-4-186.

(d)y  p. powers and G. C. Blalock, OBNL CF-52-3-229.

(E)H. R. Deem, Battelle Memorial Institute, personal

communication,

{f)Arnét and Ploetz, Z. physik. Chem. 121, 439-455 (1926).

(g)International Critical Taebles 3, 24.

190dA¥ SSAMI0¥d X T9ALUvA0 1D3f0dd dNV
temperature of the tube wall at the
metal~fluid interface, and t, is the
mixed-mean fluid temperature. The
pertinent data corresponding to that
part of the tube beyond the entrance
region are tabulated in Table 31.
The heat transfer coefficients for
sodium hydroxide are presented 1in
Fig. 64 and compared with the Dittus-
Boelter correlation for the heating
of fluids other than the liguid
metals. '

The variation of h with distance
from the entrance of the test section
is shown in Fig. 65 for a particular
set of experimental conditions (Re =
10,000 and Pr = 4.3) -~ note that for
these conditions the local coefficient
lies 10% above the established value
within 35 diameters of the entrance.

The experimental system is being
modified so that 1t can be used to
study heat transfer in the NaF-KF-LiF

FOR PERIOD ENBING JUNE 10, 1952

of type-316 stainless steel tanks
with Inconel liners. Flanges are to
be used 1in place of all-welded con-
struction to yield a meore flexible
experimental system.

UNGLASSIFIED
DWG. 15352

200 o e w,,,,”,,iu(,(,fl,, s
+ EXPERIMENTAL VALUES
e DITTUS - ROELTER CORREL ATION

| "

|
1
et e o e e e b e g

    

  
  
 
 

 

  
  

o b L
YO0OC

Fig. 64. Experimental Heat Transfer

 

 

 

 

mixture. The new system will be made Coefficients for Sodium Hydroxide.
TABLE 31
Experimental NaOH Heat Transfer Results

* * A t —t t  (avg) : |

RUN 9 %2 %/ ) f ot ‘. £ Re Nu Pr

(Btu/hr) (Btu/hr) (Btu/hr: ft*) ("F) ("F) .
1 4891 4500 73, 100 22.6 757 8,450 | 51.7 | 5.82
2 4920 4540 73,750 2.1 774 10,200 | 48.9 | 5.46
3 8038 7739 125,837 33.2 773 12,170 | 6l.8 | 5.33
4 9064 8823 143, 431 39.5 832 10,024 | 59.3 | 4.53
5 90 44 8293 134,823 40.0 830 9,916 | 55.0 | 4.55
6 7520 6166 100, 241 35.8 853 10,036 | 45.7 4.25
7 8968 6756 109,830 61.0 876 5,124 | 29.4 | 3.97

 

 

 

 

 

 

 

 

 

*
g4 is the total heat generation in the tube based on electrical measurement, and 9, is the heat

gained by the fluid in passing through the test section.

terna}l heat loss.

The difference between the two i3 the ex-

155
ANP PROJECT QUARTERLY PROGRESS REPORT

UNCLASSIFIED
IWG. 15353

ST T T
A

‘Re=KLOOO%fi
\

R
.

J‘F’ : ,,Jj;;..__‘. e

    

. _{_U_:_J[ { o

  

 

LOCAL HEAT TRANSFER CCEFFICIENT,

0 b B S . e
20 40 B0 80 100 120 140 160 180 200
PIPE DIAMETERS FROM TUBE ENTRANCE X/D

Fig. 65. Local Heait Transfer Coef-

ficients for Sodium Hydroxide Flowing
in a Tube under Uniform Heat Flux Con-
ditions.

BOILING HEAT TRANSFER

W. S. Farmer

Reactor Experimental Engineering
Division

Experiments using water as the
heat transfer medium have been con-
ducted on the horizontal-tube boiling
apparatus. Heat transfer coefficients
were computed from temperature measure-
ments taken at several circumferential
positions in the tube wall. The
resulting values exhibited the same
variation with position as found for
nonboiling heat transfer. The coef-
ficients measured on the sides of
the tube were much higher than those
found on the top surface. During an
experiment at a heat transfer flux
level of 600,000 Btu/hr-ft?, vapor
binding occurred on the tube surface
and caused a rapid rise of several
thousand degrees in the metal tempera-
ture and subsequent melting of the
heat transfer tube. A new tube has
been fabricated, and consideration is
now being given to operation of the
boiler 1n 1ts present position with
boiling sodium. A test 1s to be
conducted to evaluate the safety
hazard associated with operating a
tank of sodium vapors at high tempera-
ture under vacuum.

156

The flat-plate experimental boiler
has been revised and reassembled for
operation. The heat transfer of
boiling mercury to a copper-plated
surface will be the first test made
with this system. Amalgamation
between the copper and mercury will
result in complete wetting of the
heat transfer surface. The results
will then be compared with previous
data obtained for mercury boiling on
a chromium-plated surface that was
only partially wet by the mercury.

ENTRANCE REGION HEAT TRANSFER

W. B. Harrison

Reactor Experimental FEngineering
Division

Several months ago, experimental
work was 1nitiated on heat transfer
to sodium 1n an entrance region.
After considering a number of possible
explanations for rather scattered and
inconsistent results, 1t was suspected
that sodium was not wetting the copper
heat transfer surface. In order to
obtain heat transfer through a surface
which was wetted by the medium, mercury
was used 1n contact with copper.
Data that resulted from the use of
three different test sectlions are
shown in Fig. 66, and compared with
predictions from conduction solutions.
The range of operating conditions for
the tests 1is indicated in Table 32.

The experimental results tend to
corroborate the predictions for the
thermal entrance region as well as
the suspicion that sodium was not
wetting copper. The i1mplication 1is
that 1f sodium were used in a wetted
test section in the same range of
conditions heat transfer coefficients
as high as 400,000 Btu/hr-ft?*°F

should be achieved.
Nugye

100

FOR PEBIOD ENDING JUNE 10, 1952

UNCLASSIFIED
DWG. 15354

UNIFORM VELOCITY

/. POWER LAW
DISTRIBUTION

 

Fig.

PARABOLIC VELOCITY
DISTRIBUTION

SODIUM DATA REGION;

66. Heal Transfer to Mercury in a Thermal Entrance Region.

157
ANP PROJECT

QUARTERLY PROGRESS REPORT

TABLE 32

Entrance Region Heat Transfer to Mercury

 

 

TEST SECTION

LENGTH (in.)

DIAMETER (in.)

REYNOLDS MODULUS

h (Btu/hre*ft?+°F)

 

 

 

 

 

 

 

 

1 1/8 1/8 26,200 vo 197,000 10,30C to 42,600

2 1/16 1/16 32,100 to 89,400 24,600 to 66,300

3 1/16 1/8 24,100 to 100,500 14,750 to 48,000
NATURAL CONVECTION IN CONFINED SPACES

WITH INTERNAL HEAT GENERATION

FNCLASSIFIE O]
HOTQ 6-3681

Hamilton

D. C. F. E. Lynch

Reactor Experimental Engineering Tt
Division fifi%%@éé&'

The flat-plate natural-convection
apparatus, Fig. 67, is now operating
successfully. The test section 1is
defined by the two parallel copper
plates and the plastic end walls,
The copper plates serve as electrodes
for the electrolyte solution 1n the
test section. The heat is generated
by electrical resistance heating 1in
the volume of the liquid electrolyte.
The test section i1s approximately
0.75 by 6 by 20 inches. The generated
heat 1s carried away by the heat ex-
changer brazed to the copper plates.

The velocity traverse was obtained
by visually timing the movement of
suspended particles of bentonite.
The temperature traverse was obtained
by a traversing thermocouple assembly.
Qualitative velocity measurements were
taken in the laminar flow regions.
The mass flow rate and the velocity
distribution nbserved compare favorably
with theoretical analyses referred to
in previous reports. Temperature
traverses are currently being obtained.

 

 

Fig. 67. Flat-pPlate Natural-Convec-
tion Apparatus.

158
The cylindrical-annulus free-
convection apparatus has been con-
structed and will be operated after
the flat-plate studies have been
completed.

CIRCULATING-FUEL HEAT TRANSFER STUDIES

_m$$LiF\ Poppendiek .. Palmer
Reactor Experimental Engineering

Division

The evaluation of the mathematical
analysis of forced-convection heat
transfer in a pipe system with volume
heat sources within the fluids{3? was

 

(SJAircraftNuclearPropulsion Project Quarterly
Progress Report fer Period Ending March 10,
1952, ORNL-1227, p. 159-161.

FOR PERIOD ENDING JUNE 10, 1952

completed for turbulent flow for
Prandt]l moduli greater than 1 and
less than 10. An evaluation of the
solution for the low-Prandtl-modulus
region (liquid metals) has been
initiated.

Several thermal analyses have been
completed for laminar flow in entrance
regions (short tubes); numerical
methods were used to obtain these
solutions. The specific systems
studied corresponded to the ARE
reactor. :

A mathematical heat transfer
solution for a laminar-flow circu-
lating~-fuel system has been outlined
that pertains to fluids whose vis-
cosities vary with fluid temperatures.

 

14.

RADIATION DAMAGE

D. S. Billington, Solid State Division
A, J. Miller, ANP Division

Studies of radiation damage have
continued in the X-10 graphite pile,
the LITR, and the 86-in. cyclotron.
Previously reported experiments on
the effects of irradiation on the
fused fluoride fuels, the operation
of liquid metal loops, the creep of
metals, and the thermal conductivity
of metals have been extended and
supplemented. Preparations are under
way to continue these studies in the

higher flux MTR.

An increased rate of attack on the
container was caused by irradiation
in the LITR of Inconel capsules con-
taining the NaF-KF-UF, type of fuel
mixtures, whereas capsules containing
beryllium-bearing fuels showed no
signs of radiation-induced corrosion.
The evidence to date indicates that
the increased rate of corrosion by
the NaF~-KF-UF, fuel mixtures was due
to the radiation field, but that the

mechanism probably involved the im-
purities in the fuel rather than any
inherent instability of the fused
fluorides. ‘

The inpile experiments on cantilever
creep were extended to Inconel. As in
the previously reported experiments
with stainless steel and nickel,
the graphite~pile irradiation at the
temperatures and stresses applied
caused an increase in the secondary
creep rate.

Thermal conductivity studies in the
graphite pile and LITR on stabilized
Inconel and high-purity nickel demon-
strated that there was no detectable
effect due to irradiation,

Additional details on radiation
damage studies are contained in the
quarterly report of the Solid State
Division,

159
ANP PROJECT QUARTERLY PROGRESS REPORT

IRRADIATION OF FUSED MATERIALS

G. W. Keilho ltz

Materials Chemistry Division

The study of the radiation stability
of fused fluorides 1in contact with
Inconel has continued in the LITR and
the 86~in. cyclotron. Experiments on
beryllium-bearing fuel samples in the
LITR indicate that they are stable
during exposures of 140 hr in fields
producing power dissipations up to
550 watts/cc of sample. Cyclotron
experiments of short duration have
shown that the lithium-bearing fuel
samples are stable to proton beams
dissipating 750 watts/cc of sample.
However, a slight amount of radiation-
induced corrosion 1is produced by beams
dissipating 4500 watts/cc.

Pile Irradiation of Fuel (J. G.
Morgan, P. R. Klein, C. C. Webster,
M. S. Feldman, M. T. Robinson, H., E.
Robertson, Solid State Division;
B. W. Kinyon, X-10 Engineering Di-
vision), Experiments have been con-
ducted during the past quarter on
beryllium- and zirconium-bearing fuels
irradiated in the LITR., The results
of all the experiments on beryllium-
bearing fuels are shown 1n Table 33.
A photomicrograph of an irradiated
Inconel specimen that contained the
beryllium-bearing fuel 1s shown 1in
Fig. 68. The final runs 1in the
graphite pile and the entire data from
the LITR irradiations of the NaF-KF-UF,
fuel mixture are also shown in Table
33. A photomicrograph of an irradiated
Inconel specimen that contained this
fuel is shown in Fig. 69. Analytical
results are not yet available on
irradiated capsules of zirconium-
bearing fuel.

FEquipment similar to that to be
used for testing fuels in the MTR was
successfully operated in the LITR.
A concentrated NaF-UF, mixture was
successfully irradiated at a power

160

dissipation of 2450 watts/cc of fuel.
Bench tests indicate that the equip-
ment 1s suitable for a power dissipa-
tion of 3500 watts/cc. Equipment has
been constructed for carrying out
preliminary experiments on the escape
of xenon from inpile samples of the
fused fluorides.

§Y-6!44

 

A7
Unirradiated control sample.

250X,

(a)

 

(b)

Inconel
neutrons
power density of 554 watts/cc of

sample irradiated by
with an equivalent

fuel. 300X,

Fig. 68. Effect of Neutroom Irradia-
tion on the Static Corrosion of Inconel
by NaF~BeF2-UF4 (26-60-15 mole %) affer
135 hr at 1500°F.
Table

FOR PERIOD ENDING JUNE 10, 1952

33

Pile Irradiation Tests on Fused Fluoride Fuels in Inconel

 

 

 

 

 

 

 

 

 

 

 

" .
\ INCONEL C E
FUEL COMPOSITION | TEMPERATURE | TIME | IRRADIATION | pugp AFTEfi“fi?ZT“fs I?
- (mole %) (°F) (hr) | (wates/ec) et CAPSULE CONDITION
. Ni Cr : Fe
NaF-K€-UF,
46.5-26-27.5 1500 458 65 194 194 1,030
1500 458 65 262 20 1,630
1500 458 as 20 188 1,130
1500 472 0 160 125 2,630 | Very slight corrosion
1500 299 65 93 232 1,530
* 1500 299 65 1438 532 2,340 Very slight éorrosion
1500 . 299 65 183 146 - 622
¥ 1500 334 0 98 186 548
123 207 311 (Original salt analysis)
1500 - 115 8§00 26,800 8870 16,300 | Considerable corrosion; pressure test
u@%fikfi?5h- ' showed no evidence of buildup;
” T see Fig., 69
1500 © 115 Bench 1,030 754 1,276 | See Fig. 69
1500 164 800 45,100 950 6,41¢ | Considerable corrasion
1500 161 Bench 1,100 | 645 | 2,650 | Very slight corrosion
* 824 136 800 1,380 160 1,220 | Very slight corrosion
solid
NaF -BeF , -UF,
25-60-15 1500 126 554 Very slight corrosion
1500 131 Bench 1,540 7380 2,300 | Yery slight corrosion; pressure test
showed no evidence of buildup;
see Fig. 68
1500 139 554 1,490 1060 1,200 Very slight corrosion; pressure test
showed no evidence of buildup;
- see Fig. 68
1500 137 554 Very slight corrosion
1500 137 554 450 | 1810 780 | Very slight corrosion
47-51-2 1500 26 84 5,610 | 3900 1,430 | Pressure test showed no evidence of
: buildup :
1500 136 Bench 1,120 3300 811 | Pressure test showed no evidence of
; : buildup
1500 143 84 58,500 | 2450 | 3,100 | Very slight corrosion: pressure test
. showed no evidence of buildup
1500 115 B4 1,880 1880 2,690 | Very slight corrasion; pressure test
: : : showed no evidence of buildup
1500 145 Bench 60 | 4450 445 | Very slight corrosion
50 330 540 (Original salt analysis}

 

161
ANP PROJECT QUARTERLY PROGRESS REPORT

 

* o *

(a) Unirradiated control sample. 300X,

 

(b) Inconel sample irradiated by neutrons with an equivalent power density of

800 watts/cc of fuel. 300X,

Fig. 69. Effect of Neutron Irradiation of the Static Corrosion of Inconel
by NaF-KF-UF, (46.5-26-27.5 mole %) after 115 hr at 1500°F.

162
Cyclotron Irradiation of Fuel (W. J.
Sturm and M. J. Feldman, Solid State
Division; R. J. Jones and R. L. Knight,
Electromagnetic Research Division).
Fuel of composition NaF-KF-UF, (46.5-
26-27.5 mole %) was irradiated at
1500°F in Inconel capsules with 20-Mev
protons for up to 8 hr with a power
dissipation of 365 watts/cc of fuel.
Chemical analysis of the fuel and
metal lographic examination of the
capsules showed no evidence of radia-
tion-induced corrosion.

With the newly developed, small
Inconel capsules, fuel of composition
NaF-KF-LiF-UF, (10,9-43.5-44.5-1.1
mole %) was irradiated up to 6 hr with
a power dissipation of 425 watts/cc of
fuel. However, the fraction of fuel
irradiated in these capsules is so
small that such a power dissipation
is actually 4000 watts/cc of irradiated
fuel. In the single experiment at
this intensity, slightly enhanced
corrosion of the capsule was found in
the region of proton bombardment.
In other runs on lithium-bearing fuel
of 6-hr irradiation at 75 watts/cc,
2 hr at 150 watts/cec, and 1 hr at 125
watts/cc there was no evidence of
radiation-induced corrosion.

INPILE CIRCULATING LOOPS

R. M. Carroll.

0. Sisman

W. W. Parkinson C. D. Bauman
J. B. Trice W. E. Brundage
A. S. Olson C. Ellis |
M. T. Morgan D. T. James

Solid State Division

As reported last quarter, sodium
was circulated at a velocity of 1
ft/sec in the X-10 graphite pile
through a loop of Iuconel for 50 hr
at 1000°F and 115 hr at 1500°F. An
electromagnetic pump was used, and the
temperature of the material in the
pump cell was 1000°F. The flow rate

of sodium through the loop gradually

FOR PERIOD ENDING JUNE 10, 1952

decreased and eventually the loop
could not be operated.

When the activity of the loop had
decayed sufficiently, an examination
was made to determine the condition
of the loop and the cause of flow
stoppage. It was found that there
were several constrictions caused by
penetration of welds. One of the con-
strictions in the cold part of the
loop was almost completely closed by
grey, powdery material. X-ray exami-
nation showed the powder to be largely
nickel, and 1t 1s now being anaryzed
spectrographically.

At various times while the flow of
the loop was halted, radicactive decay
curves were obtained for the part of
the loop outside the pile. An analysis
of these curves showed that the sodium
contained no long-lived corrosion
products from the Inconel tube walls.
Specimens of the Inconel from the in-
pile and out-of-pile portions of the
loop are being examined metallo-
graphically. |

A second sodium loop was operated
in the X-10 graphite pile for 95 hr at
1500°F; operation was discontinued
because of a leak in the pump cell.
An examination of this loop 1s being
made. Another sodium loop containing
several engineering improvements 1s
being constructed for operation in the
graphite pile. Design and construction
of the sodium loop for operation 1in
the LITR is under way, and a fused-
fluoride~fuel loop 1s being designed
for operation in the MTR.

The neutron spectrum and gamma-ray
heating in hole HB-3 in the LITR are
being determined. This horizontal
beam hole has been used for a large
number of experiments by the GE-ANP
project and is expected to be available
for a considerable amount of experi-
mentation by the ORNL-ANP group in: the
near future,

163
ANP PROJECT QUARTERLY PROGRESS REPORT

CREEP UNDER IRRADIATION

J. C., Wilson J. C. Zukas
W. W. Davis
Solid State Division

Three cantilever creep tests on
Inconel have been completed i1n the
graphite pile; the results are shown
in Fig. 70, It appears that Inconel
as it 1s now heat treated (1650 or
1700°F anneal) is an unstable material
with respect to 1ts creep properties
under irradiation. This 1s analogous
to the resvlts previously reported for
type-347 stainless steel and nickel.
A cantilever creep test on Inconel
with the standard, ARE, 1700°F anneal
is in progress in the higher flux LITR.

The first model of a bellows-loaded
tensile creep apparatus for irradia-
tion experiments in the MTR has been
designed and construction has begun.
The specimen to be tested 1s1/2-1in.-00,
0.035-in. wall Inconel tubing. Several
models of the furnace that will reside
inside the specimen have been bench
tested. An attempt 1is being made to
minimize temperature gradients.

Bench testing is 1n progress on an
extensometer that utilizes a Bourdon
tube and an electrical contact. It 1is
believed that this type of extensometer
will be operable in the MTR because
of its low mass and simple geometry.
A type-5734 BCA transducer tube has
been placed in the can with the
cantilever creep test in the LITR so
that the effect of neutron radiation
on its static and emission character -
istics can be studied. Approximately
0.003-in. movement of the sensitive
element of the tube gives a change 1in
output voltage of about 40. The low
mass of the tube should permit 1t to
be used in high-flux reactors without
excessive gamma-ray heating.

A piece of natural quartz, 0.18 1in.
square and 1.3 in. long, was irradiated

164

in the LITR for a one-week period to
determine whether irradiation produced
dimensional changes, since the thermal
expansion of quartz along the Z axis
appears usable for calibrating ex-
tensometers in a reactor. A slight
discoloration of the crystal was noted
after irradiation, but, within the
precision of measurement (0.05%), no

dimensional changes were faund.

RADIATION EFFECTS ON
THERMAL CONRUCTIVITY

A. F. Cohen .. C. Templeton
Solid State Division

Recent experiments 1in the X-10
graphite pile and LITH on heat-stabi-
lized Inconel and high-purity nickel
show that the i1rradiation has no
measurable effect on thermal conduc-
tivity. In previously reported ex-
periments on Inconel at high tempera-
tures some loss of conductivity was
suffered by the inpile specimens. In
light of the recent results, this now
appears likely to be the result of
phase changes in unstabilized material
induced or accelerated by the irradia-
tion.

Two fabricated Inconel samples were
stabilized by heat treating for 1/2 hr
at 2050°F, slowly cooled to 1550°F
where they were held for 24 hr, and
then furnace cooled to room tempera-
ture. This heat treatment caused
precipitation of chromium carbides
during the slow cooling so that the
specimen would be 1in the most stable
condition from room temperature to
1500°F. One of these samples was used
for a relative type of experiment 1n
the graphite pile and the other for
an absolute conductivity experiment

in the LITR.

The relative experiment 1in the
graphite pile was in a fast-neutron
flux of approximately 3 x 10'!, The
specimen was held at lower temperatures
PERGENT EXTENSION

015

0.10

0.05

FOR PERIOD ENDING JUNE 10, 1952

DWG. 14942R
SSD~A~-430a

 

e STRESSED TO 2000 psi

SPECIMENS A,B,D, AND E:
ANNEALED 2 HOURS AT 4650°F
STRESSED TO 1500 psi.

SPECIMEN G: ANNEALED 2 HOURS
AT {700°F ‘

  

 

A,B,AND C: IN-PILE TESTS X-10
GRAPHITE REACTOR
D AND E: BENCH TESTS

|

b INDICATES PILE
 SHUTDOWN

 

 

 

 

   
  

 

 

 

 

 

 

 

 

Fig.

200
TIME {hr)

400

70. Cantilever Creep Tests of inconel at 1500°F,

600

165
and then at 1500°F for five weeks. No
change in ggpductivity greater than
l to 2% was observed.

The absolute thermal conductivity
specimen was a long cylinder with an
internal heater in which the heat
flowed radially outward. The specimen
was placed in the hole HB-3, and the
maximum power level of the LITR during
the experiment was 1000 kw. Changes
in thermal conductivity were sought
as a function of irradiation time at
a given flux level and a given specimen
temperature after thermal equilibrium
was reached. The time required to
reach equilibrium 1s of the order of
2 to 4 hr after a change in reactor
power level or a change in the heater
current. The highest temperature at
which measurements were made was
1300°F. No decrease in the conduc-
tivity was observed at 1300°F over a
period of 24 hours. At lower tempera-

166

tures no decrease 1n conductivity was
observed during a four-week period.

One absolute conductivity experi-
ment, mostly at low temperatures, was
carried out in the LITR on an ARE heat-
treated Inconel specimen. Preparations
are being made for two more such
experiments at elevated temperatures.

A cylindrical, high-purity nickel
specimen was used i1n an absolute thermal
conductivity experiment in hole HB-3 of
the LITR. A series of experiments were
performed at various LITR power levels
and at various temperatures up to
1475°F. Half of the 17-day total
irradiation time was at a power level
of 1500 kw. No changes in conductivity
as a function of irradiation time were
found. The constancy of the variables
involved 1is such that a 5% change 1in
conductivity would bhe very easily
detected.
 
 

 
SUMMARY AND INTRODUCTION

The analytical chemistry program
regquired tn support of the materials
research program included the routine
analysis of 1316 samples as well as
the development of new analytical
procedures when necessary (sec. 16).
In particular, the substitution of
zirconium fluoride for beryllium in
the fuel mixture has necessitated a
review of the existing methods of fuel
and corrosion-product analysis.

The list of reports that have been
1ssued by the project during the past

quarter includes 13 formal reports and
33 informal documents on all phases of
the ANP Project {(sec, 17).

The directory of research projects
of the Aircraft Nuclear Propulsion
Project of the Oak Ridge National
Laboratory is given in sec. 18. In-
cluded are the research projects of
the Laboratory's subcontractors on
the ANP Project and the research
projects being performed by ORNL for
the ANP programs of other organiza-
tions. |

 

13.

ANALYTICAL CHEMISTRY

C. D. Susano
Analytical Chemistry Division

The substitution of zirconium
tetrafluoride for beryllium in the
fuel mixture has necessitated a review
of existing methods for the deter-
mination of alkali metals, uranium,
and the major corrosion products - iron,
nickel, and chromium ~ to determine
their applicability to the new fuel
mixtures.

A rapid and accurate method for the
determination of total alkali metals
has been developed. In this method,
the hydrogen ion displaced from: a
cation-exchange resin by the alkalx
metals 1s titrated. A cation-exchange
column 1s also being used successfully
for the quantitative separation of
sodium and potassium. :

A volumetric method for the deter«
mination of zirconium is‘being studied,
This method, which depends upon the
titration of the base liberated when
a fluoride i1is added to zirconium
hydroxide, appears promising, and 1t
1s expected to be more rapid than the

gravimetric method presently being
used. ' : :

Improvements that eliminate the
interference of zirconium have been
made 1n the method for the deter-
mination of iron. In the deter-
mination of sulfide 1i1n fluorides and
eutectics, the colorimetric method
that depends upon the formation of
methylene blue has been adapted. A
method is currently being studied for
the determination of total sulfur in
these materials by reductioen of all
sulfur compounds and subsequent deter-
mination as sulfide by the same method.

It has been found that zirconium
oxide can be determined 1n zirconium
tetratfluoride because of i1ts 1nsolu-
bility in oxaliec or tartaric acid.
However, a successful method for the
determination of zirconium oxyfluoride
1s not yvet available. A method 1s
being sought for the determination of
water in eutectic components to a

lower laimait of 0.05%.

169
ANP PROJECT QUARTERLY PROGRESS REPORT

ANALYSES FOR COMPONENTS OF
FLUORIDE MIXTURES

Jo. C. White W. J. Boss
C. M. Bovd C. K. Talbott

Analytical Chemistry Division

Total Alkali Metals. A rapid and
accurate method for the determination
of total alkali metals in fluoride
eutectics has been developed. In this
method, whiech makes use of an 1o0on-
exchange resin, a sulfate solution of
the alkali metals, separated from
uranium and other cations, 1s passed
through a cation-exchange column. The
alkali metals are retained by the
resin and replaced by equivalent
amounts of hydrogen ions, and the
sulfuric acid formed is eluted from
the column with water, Two titrations
are required: the first to determine
the free-acid content of the starting
solution and the other to determine
the total acidity of the eluted
sulfuric acid solution, The difference
in acidity is equivalent to the alkala
metals present.

The success of this method depends
upon the removal of all cations except
the alkali metals, since other cations
will also replace hydrogen 1ons on the
resin., In the case of uranium, the
separation 1s effected by passing the
solution through an anion-exchange
column 1n which uranium is retained as
an anionic sulfate complex. The alkali
metal ions, which are not retained,
are gquantitatively eluted with dilute
sulfuric acid. The resulting alkali
metal sulfate solution 1s then ready
to be passed through the cation-
exchange column., The method has so
far been applied to fuel mixtures
containing only uranium 1n addition
to alkali metals, The same technique
for separating uranium and alkali
metals should, however, be applicable
to solutions contalning zirconium,
since zirconium also forms anionic
complexes in IN sulfuric acid; this

170

will be investigated. Approximately
1% precision in the range 35 to 90%
can be obtained with this method of
determining total alkali metals.

Sodium, Potassium, anmd Lithium. In
viewof the desirability of determining
individual alkali metals in the presence
of each other, attention has been given
to methods of separating the con-
stituents. Lithium can be separated
from sodium and potassium by extraction
of the alkali chlorides with Z2-ethyl
hexanol.{!) The use of ion-exchange
columns is being studied for separating
sodium and potassium. By means of a
70 by 2 cm column of Dowex-50 (50 to
100 mesh), sharp separations of sodium
and potassium 1n quantitlies of the
order of 100 mg can be achieved by
using 2N hydrochloric acid solution
as the elution agent. Although the
procedure is extremely time-consuming
(approximately 8 hr per separation),
it has the advantage that very little
attention by the operator is regquired.
Standard methods are used for the
determination of the separated alkali
metals.

Beryllium. The volumetric method
of determining beryllium proposed by
McClure and Banks(?’ hasbeen used with
The separation of beryllium
uranium 1n

success.,
from interfering cations,
this case, by means of anion-exchange
resins has been described 1in aprevious
report. ¢3)

Zirconium. The addition of zirconium
salts to the fuel mixture necessitated
a number of changes in the methods of
dissolving and analyzing these ma-
terials. A mixture of aqua regia and

 

(1)E. R. Caley and H. D. Axilrod, Ind, Eng.
Chem., Anal. Ed. 14, 242 (1942).

(2)J. H. McClure and C. V. Banks, An Empirical
Titrimetric Method for the Determination of
Beryllium, AECU-812.

(S)Aircraft Nuclear Propulsion ProjectQuarterly
Progress Report for Period Ending March 10, 1952,
ORNL-1227, p. 176.
sulfuric acid proved to be satisfactory
for dissolving the samples and for
volatilizing the fluoride. Because of
the interference of zirconium, the
method previously used to determine
iron could not be applied to the
analysis of the new fuel samples. An
alternate method employing hydroquinone
as the reducing agent, ammonium acetate
to adjust the pH and prevent the
precipitation of zirconium, and
orthophenanthroline as the chromogenic
agent is mow in use.  Zirconium is
determined by gravimetric methods in
which phenylarsonic acid(*’ or mandelic
acid(®? are used as the precipitating
agents. This method has proved satis-
factory with respect to accuracy and
precision.

When large numbers of determi-
nations are required, however, volu-
metric methods are generally more
satisfactory; hence, investigations
are now under way to determine the
feasibility of a volumetric method
that would increase the output of
zirconium determinations., Sawaya and
Yamashita®®? indicated that a fluoride
complex could serve as a basis for a
volumetric method. Zirconium 1s first
precipitated as Zr(OH),, KF is added
to form the zirconium-fluoride complex
ion, and the liberated base is titrated.

ANALYSES FOR IMPURITIES IN
FLUORIDE MIXTURES

Iron. The method previously used
for the colorimetric determination
of iron in fluoride eutectics has not
been satisfactory when employed for
zirconplum-containing salt mixtures
because of the precipitation of
zirconium hydroxide. To avoid the

 

(4)U.S- Steel Corporation, Sempling and Analysis
of Carbon and Alloy Steels, p. 246, Reinhold, New

York, 1938.
(5)¢. A. Kumins, Anal. Chea. 18, 376 (1947).
(6)T Sawaya and M. Yamashl J. Chem, Soc.
{Japan) T2, 414.16 (1951).

FOR PERIOD ENDING JUNE 10, 1952

necessity for removing zirconium from
the solution, the method was satis-
factorily modified by employing ammonium
acetate, which prevents hydrolysis of
the zirconium ion. Hydrogquinone is
used as the reductant for iron, and
orthophenanthroline 1s used as the
chromogenic agent.

Chromium. The diphenylcarbazide
method for the determination of
chromium, 1in which perchloric acid 1s
used as the oxidant, has proven satis-
factory in the presence of zirconium.

ANALYSES FOR IMPURITIES IN
FLUORIDE COMPOUNDS

A study 1is being made of the
impurities in the individual components
of eutectics The i1mpurities are
alkali metal, beryllium, zirconium,
and uranium(IV) fluorides. Determi-
nations of traces of silicon, nitrate,
chloride;, sulfate, sulfide, and oxides
have either been made.or are currently
in progress. Details of methods for
the determinations of the above
elements, with the exception of
sulfate, sulfide, and zirconium oxide
in zirconium tetrafluoride, have been
described in a previous report, (3’

Sulfate and Sulfide. The presence
of traces of sulfide in fluoride
eutectics is believed to be a cause of
corrosion of metal containers (stain-
less steels and Inconel) at high
temperatures. During the cleaning
(with acidic solutions) of containers
used for the preparation of the
eutectics, a distinct odor of hydrogen
sulfide has been noticed. Colorimetric
determinations of sulfide by the
methylene blue method(?’) have shown
that traces of sulfide are present.
Development work is currently in
progress on methods for the determi-
nation of the sulfur content of

(7)C J. Rodden (Editor~in-Chief), Aralytical

Chemistry ofthe MHanhattan Praject, p. 311, MeGraw-
Hill,New York, 1950,

171
ANP PROJECT QUARTERLY PROGRESS REPORT

individual components of fuels, since
this 1s probably the only possible
source of sulfur contamination. The
method currently under study involves
the reduction of all sul fur compounds
(sulfate, sulfite, and thiosulfate)
to sulfide with a reducing mixture
composed of hydriodic, hydrochloric,
and phosphorous acids, and the deter~
mination of sulfide colorimetrically.
Preliminary tests have shown this
method to be applicable to fluoride
salts, It will also be extended to
eutectlcs.,

Zirconium ¢xide and Oxyfluoride. A
method for the determination of the
oxide 1n the fluoride has been de-
veloped. This procedure i1s based on
the fact that ZrO, i1s insoluble 1in
saturated solutions of oxalie acid
(pH = 0.6) or tartaric acid (pH = 0.7),
which readily dissolve ZrF,. No
soluble zirconium was detected in
tests 1n which Zr0, was refluxed for
40 hr with oxalic acid solution or
tartaric acid solution. Unfortunately,
this method does not separate all the
oxygen-contalning materials, since the
oxyfluoride 1s nearly as soluble as the
the fluoride 1n oxalic or tartaric
acid solutions. Attention 1is being
given to chemical means for determining
all the oxygen 1n zirconium fluoride.

Some lots of ZrF, that were tested
contained as much as 9.5% of ZrO,.
Examinations of sublimed ZrF, by the
oxalic acid procedure have shown no
detectable residue.

Water. The water content of alkali
metal fluorides and zirconium oxide
must be taken 1nto account in calcu-
lations connected with cold critical
experiments., For this reason, and
also because traces of water may
contribute to the corrosive properties
of the fuel, 1t 1s important that
methods be available for the deter-
mination.

172

Two methods were tested. One, a
modi fied method of Feibig and Warf(¢?®)
in which sodium carbonate 1s used to
react with any HF liberated during
combustion, yielded results which
showed that the lower limit for the
determination (0,05%) could not be
attained under the conditions used.
Indications are that the size of sample
must be increased topermit the weighing
of significant amounts of water.
Preliminary tests are in progress 1in
which a large sample 1s dried 1n a
vacuum OVen.

The second method, that of Karl
Fischer, with certain modifications in
iodine concentration, may be used for
the determination of traces of water 1in
alkali metal fluorides. However, there
is some uncertainty in this method when
applied to solid materials,
part of the water may be trapped in
the solid phase and not come in con-
tact with the Karl Fischer reagent.

since a

SERVICE ANALYSIS

H. P. House L. J. Brady
J. W, Robinson

Analytical Chemistry Division

Fxpansion of the work of the ANP
Reactor Chemistry group during the
last gquarter resulted in more than a
twofold increase in the analytical
services required. Approximately 85%
of the samples analyzed were from
recactor chemistry studies of various
fluoride salt mixtures. During the
latter part of the period there was a
decrease in the number of fuel samples
containing beryllium, but fuel mixtures
in which zirconium fluoride is a major
constituent have become 1increasingly
important.

 

(8).]. G. Feibig and J. C. Warf, Determination
of Water in Fluorides, CC-2939 (June 29, 1945).
Attempts were made to supplement
information, gained by metallographic
me thods, concerning preferential
leaching of the components of the
container wall by analyzing samples of
consecutive thin layers taken from
the 1nner surfaces of tubes that had
been in contact with the molten salt,
Because of lack of uniformity of the
surface of the tubes only small amounts
of sample could be collected and thus
limitations were imposed on the
analytical results that could be
obtained.

Most of the remaining 15% of the
samples analyzed during this period
were alkali fluoride fuels derived
from tests conducted by the ANP Experi-
mental Engineering group. The Experi-
mental Engineering group has also been
interested in determining preferential

FOR PERIOD ENDING JUNE 10, 1952

corrosion of the components of the
alloys usedto fabricate the test loops,
A limited number of samples were also
derived from the following projects:
ANP Critical Experiments, ANP Radiation
Damage Research, and studies with the
5-Mev Van de Graaff. ' A summary of the
backlog of analyses is presented in

Table 345

TABLE 34

Backlog Summary

Samples on hand February 2 213
Number of samples received 1302
Total number of samples 1515
Number of samples reported 1316
Backlog as of May 10 199

 

16. LIST OF REPORTS ISSUED

REPORT NoO. TITLE OF REPORT AUTHOR(S) DATE ISSUED
Iksnm

ORNL- 1234 Reactor Program of the Aircraft Nuclear " W. B. Cottrell (to be issued)
Propulsion Project

ANP - 64, The Technical Problems of Aircraft Reactors~ C. B. Ellis 5-13-52

Part 11 Part II '

Y- F26- 33 ARE (CF-SM) Design Data W. B. Cottrell 4-9-52

ORNL- 1255 Basic Performance Characteristics of the - A, P. Fraas 5-14-52
Steam Turbine.Compressor-~Jet Aircraft . G. Cohen
Propulsion Cycle

Reactor Physics

Y-F10-93 A Simple Criticality Relation for Be- C. B. Mills 3-10-52
Moderated Intermediate Reactors

Y-F10-96 Optimization of Core Size for the Circu- C. B. Mills 3-28-52
lating-Fuel ARE Reactor

Y-F10-98 Physics Considerations of Circulating-Fuel W. K. Ergen 4~ 16-52

Reactors.

(Lecture at Second Fluid Fuels

Devel opment Conference, April 17, 1952)

R A R SR R 408,88 R A A n kA m e

173
ANP PROJECT QUARTERLY PROGRESS REPORT

REPORT NO.

Y-F10-99

Y-B23-1

ORNL-1246

CF-52-1-12

Y-F10-103

CF-52-4-37

CF-52-3-20

ORNL-1248

OBRNL-914

CF-52-3-230

CF-52-3-229

CF-52-4-186

CF-52-5-209

CF-51-10-70

C¥-51-10-70

174

TITLE OF REPORT

Note on the Linear Kinetics of the ANP
Circulating-Fuel Reactor

Preliminary Direct-Cycle Reactor Assembly-
Part 1T

Study of the Kinetics of the Direct-Cycle

Aircraft Reactor

Control of Nuclear Reactors

Statics of the ARE Reactor

Heat Transfer Research

Experimental Heat Transfer Coefficients
for Molten Sodium Hydroxide

Bubble Behavior Within Liquids Flowing in
Tubes Containing “U"” Bends

A Critical Review of the Literature on
Pressure Drop in Noncircular Ducts and

Annuli

Forced~Convection Heat Transfer in Thermal
Entrance Regions -~ Part 1II

Physical Properties
Densities of Certain Salt Mixtures at Room

Temperature

Heat Capacity of Potassium Hydroxide

Heat Capacity of Barium Hydroxide

Densities of Fuel Mixtures Nos. 25 and 25a

Shielding Research

Introduction to Shield Design. Part I: The
Scources of Radiation, the Attenuation
Processes, and Geometry of Shielding

Introduction to Shield Design. Part II:

Comparison Method of Shield Design

AUTHOR(S)
F. G. Prohammer
A. D. Callihan
J. J. Stone
J. D. Trimmer
J. H. Jordan
C. B. Mills
H. W. Hoffman
H. F. Poppendiek
H. C. Claiborne

H. F. Poppendiek
L. D. Palmer

M. Tobias

S. I. Kaplan

S. J. Claiberne
W. D. Powers

G. C. Blalock
W. D. Powers

G. Blalock

J. M. Cisar

E. P. Blizard
E. P. Blizard

DATE ISSUED

4-22-52

2-26-52

3-24-52

1-3-52

5-8-52

4-2-52

3-3-52

5-6-52

5-26-52

3-26-52

3-31-52

4-30-52

5-28-52

1-30-52

3-7-52
REPORT NO. -

CF-52-3-180

CF-52-3-219

CF-52-4-85

CF- §2- 4-99

CF-52-4-124

CF-52-4-126

CF~52-4-157.

ORNL- 1147

CF-52~5-40

CF-52-5-41

CF- 52-5- 163

ORNL-1252

Y-B31-336

ORNL- 1279

OBRNL- 1286

FOR PERIOD ENDING JUNE 10, 1952

TITLE OF REPORT
Copies of Pdwe]]-Snyder Gamma Cross Sections

Transformation from Disk to,Point-Source

Geometry

Proposal for a Divided Shield for Testing
Facilitcy

Some Ground-Scattering Expefiments Performed
at the Bulk Shielding Facility

Fission Cross Section of U236

Gamma Dose Behind Solid Iron Thermal Shield
as a Function of the Water~Reflector
Thickness

Shielding Requirements and Heat Generation,
etc.

The Unit Shield Experiments at the Bulk
Shielding Facility

Gamma Dose Behind Iron-Watef Thermal Shie]d
of Various Thicknesses as a Function of
Water-Reflector Thickness

GGamma Dose Behind Iron-Borated Water Thermal
Shield with 18-, 20-, and 3-cm Water Re-
flector

Gamma Attenuation Behind 3.8-7.6- and 11. 4-
cm Solid Lead Shadow Shield in Pure Water
as a Function of Water-Reflector Thickness

Chemistry

General Information Concerning Fluorides

Clarity of Borated Water

Methods of Determination of Uranium

Tetrafluorade

The Determination of Oxygen in Sodium

AUTHOR(S)

., P. Blizard

P. Blizard

P. Blizard

E. Hungerford

W. Lamphere
E. Clifford
Stephenson
.. Meenm

. E. Hungerford

E. Clifford

E. Clifford
E. Clifford
E. Lee

. P. House .
. D. Susanq

L. Manning
K. Miller

Rowan, Jr.

C. White
J. Ross
Rowan, Jrx.

DATE ISSUED
3-19-52

3-21-52
4-17-52
4-16-52
4-23- 52
4-23-52
4-28*52

4- 30~ 52

5-7-52
5-7-52

5-20-52

2-19-352

3-5-52

5-22-52

5~23-52

175
ANP PROJECT QUARTERLY PROGRESS REPORT

REPORT NO.

CF-52-4-128

CF-52-4-130

CF-52-4-131

OBNL-1114

Y-854

ORNL- 1227

Y- F26-36

CF-52-3-172

CrF-52~3- 147

TITLE OF REPORT

Metal lurgy

Liquid Metal Corrosion by Sodium, Lithium,
LLead at Elevated Temperatures

The Structure of L.iquid Bismuth and Liquid
Lead

Creep and Stress-Rupture Facilities at ORNL

Stress-Strain-Time Phenomena in Mechanical
Testing
Miscellaneous
Radiation Effects on Inorganic Liquids. A

Preliminary Literature Search

Aircraft Nuclear Propulsion Project Quarterly

Progress Report for Periond Ending March 10,
1952

ANP Information Meeting of May 21, 1952
Induced Activity in the Cooling Water - ARE

Health Physics Instruments Recommended for

=

AUTHOR(S)

des Brasunas

Smith

Oliver

H. Andersen

. Carter

Cottrell

Cottrell

J. Burnett

J. Burnett

DATE ISSUED

4-25-52

4-25-52

4-24-52

5-9-52

3-12-52

5-7-52

6-6-52

3-24-52

3-20-52

the ARE

 

17. DIRECTORY OF ACTIVE ANP RESEARCH PROJECTS(V
June 1, 1952
I. BEACTOR AND COMPONENT DESIGN
A. Aircraft Reactor Design
1. Survey of Circulating-Fuel Cycle 9704-1 Fraas
B. ARE Reactor Design
1. Core and Pressure Shell 9201-3 Hemphiil, Wesson

 

2. Fluid Circuit Design 9201-3 Cristy, Lawrence,
Jackson, Eckerd,
Scott, Bussard

3. Pressure and Flow Instrumentation 9201-3 Hluchan, Affel,
Williams

4, Structural Analysis 9201-3 Maxwell, Walker

5. Thermodynamic and Hydrodynamic Analysis 9201-3, [ubarsky, Greenstreet,

9704-1 Longyear
(I)Thia directory was previously issued as Directory of Active ANP Rescarch Projects at ORNL, by W. B.
Cottrell, Y-F26-35, June 1, 1952.

176
D.

FOR PERIOD ENDING JUNE 10, 1952

6. RBRemote-Handling Equipment
7. Electrical Power Circuits

8. FElectrical Power Circuits

ARE Control Studies
1. High-Temperature Fission Chamber

2. Control System Design

3. Control Rod Design '
ARE Building Facility
1. Construction

2. Internal Design

Reactor Statics
1. 1IBM Calculations for the ORNL ARE and ANP Proposals

Summary Reports on Multigroup Technique
Analysis of Critical Experiments
Temperature Coefficients of Beactifity
Bol tzmann Equation on Computing Machines

Review of Cross-Section Data

~N N W B W N

Odd Beactor Geometries

Reactor Dynamics

l. Kinetics of Circulating-Fuel Beactors

Critical Experiments
1. ARE Critical Assembly

2. Air-Water Reactor Critical Assembly (GE)

Pump Development
1. Gas-Seal Centrifugal Punmp

2. Frozen-Seal Puwp for Sodium and Fluorides

ARE Pump Design and Development

Electromagnetic Pump

Rocking-Channel Sealless Pump

Seals for NaOH Systems

Seals for High-Tempefature Systems

Packed-Seal Pump for High-Temperature Fluorides

Therma] Cyc]ing Tests

S W O N R g W

ol
*

Inpile Fluoride Pump

9201-3
9201-3
1000

2005
2005

9201-3

7503
1000

9704-1,
9766-1A

9704-1
9704-1
9704~1
2068

9704-1
9704-1

9704-1

9213

9213

9201-3
9201-3

9201-3
9201-3
BMI

BMI

9201-3
9201-3
9201-3
9201-3

Hutto
Eh]und
Walker

Hanauer

Epler, Kitchen, Ruble,
Green

Estabrook, Alexander

Nicholson Company

Browning

Mills, Ufifelman,
DeMarcus, Johnson

Mills,
Affken

Holmes

Osborn, Ikenberry
Edmonson .
Wil son

Worley

Tamor, Thompson,
Coveyou

Callihan, Zimmerman,
Williams, Haake, Scott,
Kennedy, Keen

Callihan, Zimmerman,

Williams, Haake, Scott,
Kénnedy, Keen

Cobb

McDonald, Cobb, Smith,
Huntley

Cobb, Grindell, Southern
McDonald, Southern:
Dafton

Simons, Allen

Ward

Cobb, Huntley, Johnson
Cobb |

Cobb

Johnson,

177
ANP PROJECT QUARTERLY PROGRESS

1.

KI

178

REPORT

Valve Development

I.

Sel f-Welding Tests

Bellows Tests at High Temperatures

Valves for High-Temperature Systems

Canned-Botor-Operated Valve for High«Temperature
Fluid Control

Heat Exchanger and Radiator Development

1. NaK-to-NaK Heat Exchanger

2. Sodium-to-Air Radiator

3. Boeing Turbojet with Na Radiator

4. Fuel-to-Liquid Heat Exchanger

5. Fuel-to-Gas Radiator

6. Radiator and Heat Exchanger Design Studies

Instrumentation

1. Flow Measurement Devices for High-Temperature Fluids

2. Pressure Measurement in High-Temperature ¥Fluid
Systems

3. Level Indication and Control in High-Temperature
Fluid Systems

4. Leak Detection for Fluoride Systems

5. High-Temperature Fluid Flow Measurement

II.

Cross-Section Measurements

8.

Neutron Velocity Selector
Analysis for He in Irradiated Be
Total Cross Sections of N4 and ol® (GE)

Elastic-Scattering Differential Cross Section of
N4 and 06 (GE)

Total Cross Sections of Li%, Li’, Beg, pio, Bl2, cl2
Fission Cruss Sections
Cross Sections of Be, C, W, Cu

Inelastic Scattering Energy Levels

Shielding Measurements

1.
2.

Divided Shield Mockup Tests (GE)

Bulk Shielding Reactor Power Calibration

SHIELDING RESEARCH

9201-3
9201-3

9201-3

9201-3

9201-3
9201-3

9201-3
9201-3

9201-3

9201-3

9201-3

9201-3

9201-3
9201-3
9201-3

2005
3026
9201-2
9201-2

9201-2
9201-2
3001

9201-2

3010
3010

Adamson, Petersen, Reber,

Ward
McDonald, Taylor
Ward, Johnson, Southern,

McDonald

McDonald, Southern, Ward

Fraas, LaVerne, Petersen

Fraas, lLaVerne, Whitman,
Petersen

Fraas, LaVerne, Whitman

Frass, Whitman, Bailey,
McDonald, Salmon

McDonald, Bai]eiri Salmon,

Whitmen, Tunne

Fraas, Bailey, Salmon,
Whitinan

McDonald, Bailey, Taylor

McDonald, Taylor

McDonald, Southern, Bailey

McDonald, Southern, Taylor

McDonald, Anderson
Pawlicki, Smith

Parker

Willard, Bair, Johnson

Jobnson, Bair, Willard,
Fowler

Johnson, Willard, Baxir

Lamphere, Willard

Clifford, Flynn,

Blosser, Chapman, Watson

Willard, Bair, Kington

Meem and group

Johnson, McCammon
-~ Gy R W

FOR PERIOD ENDING JUNE 10, 1952

Bulk Shielding Reactor Operation
Heat Release per Fission

Na Brémsstrahlung Measurement

Air Duct Tests ~ Single Pipes (GE)
Air Duct Tests - Large Mockups (GE)

Thermal Shield Measurements {NDA)

Shielding Theory and Calculations

1.

~ O o s W B
. . . : . .

81

Survey Report on Shielding

Shielding Section for Reactor Technology
Theory of Neutron Transmission in Water
Interpretation of Pb~H20 Lid Tank Dafia
Divided-Shield Theory and Design

Air Nuct Theory (GE)

Shielding Section for Reactor Handbook

Consultation on Radiation Hazards (GE)

Shielding.Instruments

1.

2
3
4.
5
6

Gamma Scintillation Spectrometer
Neutron Dosimeter Development

Proton Recoil Spectrometer for Neutrons
He® Counter for Neutrons

Lil Crystals for Neutrons

Neutron Spectroscopy with Photographic Plates

Shielding Materials

1.

2.

Preparation of High-Hydrogen Rubber

Development of Hydrides for Shields

3010
3010
3025
3001
3001

3001

3022
9204-1

3022
3022
2005
NDA

3001
3022

2001

3010
3010
3010
3010
3010
- 3006

Goodrich
Company

. MHI

IX1. MATERIALS RESEARCH

Liquid Fuel Chemistry

1.

Phase Equi]ibrium Studies of F]uorideé
Prepargtion of Standard Fuel Samples
Special Methods of Fuei Purification
Evaluation of Fuel Purity

Thermodynamic Stability and Electrochemical
Properties of Fuel Mixtures

9733-3

9733-3

9733-3

19733-3

9733-3

Holland, Leslie, Roseberry
Meem and group

Sisman

Muckenthaler, Hullings

Clifford, Flynn, Blosser,
Hullings, Muckenthkaler

Clifford, Flynn, Blosser,
Hullings, Muckenthaler

Blizard, Welton

Blizard

B]iiard, Enlund

Simon

Goldstein

Clifford, Simon

Blizard, Hungerford, Simon,
Ritchie, Meem, Lansing,
Cochran, Maienschein,

Burnett

Morgan

Maienschein

Hurst, Glass, Cochran
Cochfan, Henry, Hungerford
Cochran

Matienschein, Schenck

Johnson, Haydon

Born

Banus

Barton, Bratcher, Traber,
Truitt

Nessle, Truitt, Morgan,
Love '

Grimes, Blankenship,
Nessle, Blood, Weinberger

Overholser, Sturm

Overhol ser, Topol

179
ANP PROJECT QUARTERLY PROGRESS REPORT

6.
7.
8.
9.
10.

11.
12.

13.

14,
15.
16.
17.

Hydrolysis and Oxidation of Fuel Mixtures
Simulated Fuel for Critical Experiment
Stability of Slurries of U0, in NaOH

Phase Equilibria Among Silicates, Borates, etc.
Fuel Mixtures Containing Hydrides

Chemical Literature Searches

Solution of Metals in Their Halides
Preparation and Properties of Complex F]uorides
of Structural Elements

Reactions of Fluorides and Metal Oxides

Preparation of UF,

Preparation of Pure ZrF,

Preparation of Specimens for Radiation Damage

B. Liquid Moderator Chemistry

1.
2.

L.

ol W N

=)

1.

2.

180

Preparation and Evaluation of Pure Hydroxides

Electrochemical Behavior of Metal Oxides in
Molten Hydroxides

Moderator Systems Containing Hydrides

lfydroxide-Metal Systems

€orrosion by Liquid Metals

Static Corrosion Tests in Liquid Metals and Their
Alloys

Dynamic Corrosion Research in Harps
Effect of Crystal Orientation on Corrosion
Effect of Carbides on Liquid Metal Corrosion

Mass Transfer in Molten Metals

Diffusion of Mclten Media into Solid Metals
Structure of Liquid Pb and Bi

Alloys, Mixtures, and Combustion of Liquid Sodium

Corrosion by Fluorides

Static Corrosion of Metals and Alloys in Fluoride
Salts

Isothermal Static Corrosion Tests in Fluoride Salts

Fluoride Corrosion in Small-Scale Dynamic Systems

Dynamic Corrosion Tests of Fluoride Salts

Magnetic Susceptibility Due to Fluoride Corrosion

9733-3
9733-3
BMI
BMI
MHT
9704-1
3550

9733-3
9733-3
9733-3
9733-2
9733-2

9733-3

9733-2

3550

2000
9201-3
2000
2000
2000

2000
2000
2000

2000
3766

9766,
2000
9201-3

9201-3,
2000

Blankenship, Metcalf
Overholser, Cuneo
Patterson

Crooks

Banus

Lee

Bredig, Johnson,

Bronstein

Overholser, Sturm

Bl ankenship, Hoffman
Coleman
Nessle, Ergen

Truitt

Overhal ser, Ketchen

Bolomey, Nichols
Banus

Bredig

Vreeland, Day, Hoffman
Adamson, Reber

Smith, Cathcart, Bridges
Brasunas, Richardson

Brasunas, Richardson

Richardson
Smith
Bridges, Smith

Vreeland, Day, Hoffman

Kertesz, Buttram,
Smith, Meadows, Croft
Kertesz, Buttram, Croft,

Smith, Meadows,
Brasunas, Nicholson
Adamson, Reber

Tunnell, Brasunas, Gray
FOR PERIOD ENDING JUNE 10, 1952

6.  Fluoride Corrosion in Rotating Cylinder Apparatus

7. Beaction of Metals with Fluorides and Contaminants

8. FEquilibria Between Flectropositive and Transition
Metals in Halide Melts

9. (Corrosion by Fluorides in Standpipes.

Corrosion by Hydroxides

1. Static Corrosion of Metals and Alloys in Hydroxides

2. Mass Transfer in Molten Hydroxides

3. Physical Chemistry of the Hydroxide Corrosion
Phenomenon

4. Static Corrosion by Hydroxides

5. Static Corrosion by Hydroxides

Physical Properties of Materials
Density of Liquids

Viscosity of Liquids

Thermal Conductivity of Solids
Thermal Conductivity of Liquids
Specific Heat of Solids and Liquids
Viscosity of Fluoride Fuel Mixtures

. Vapor Pressure of Fluoride Fuels

W =2 N o e W N e

. Vapor Pressure of BeF,

Heat Transfer

1. Convection in Liquid Fuel Elements

2. Heat Transfer in Circulating-Fuel Reactor

e

Heat Transfer Coefficients of Fluoride and
Hydroxide Systems

e

Heat Transfer Coefficient of Lithium Systems
. Boiling-Liquid-Metal Heat Transfer

Sodium Heat Transfer Coefficients in Short Tubes

-~3 [N e

Heat Transfer in Special Reactor Geometries

Fluoride Randling
1. Production of Fluorides for ARE
2. Design of Special Fluoride Handling Equipment

Pretreatment of Fluoride-Containing Systems
Inspection of Components of Fluoride Systems

Filling Techniques for Fluoride Systems

o o W

Scale- and Full-Size Systems

2201-3
9733-3

3550

9766

2000
2000

2000
9766
BMI

9204-1
9204-1
9204-1
9204-1
9204-1
9766
9733-2
Bvmi

9204-1

95204-1

9204-1
9204-1
9204-~1
9204-1
9204-1

9201-3
9733-3

95201-3
9201-3
9201~3
9201-3

Tunnell, Coughlen

Overhol ser, Redman,
Powers, Sturm

Bredig, Johnson,
Bronstein

Kertesz, Buttram, Croft

Vreeland, Day, Hoffman
Brasunas, Richardson,
Smith, Cathcart

Cathcart, Smith
Kertesz, Croft

Jaffee, Craighead

Cisar

Ci&ar, Redmond, Johes
Powers, Burnett
Claiborne, Cooper
Powers, Blaleck
Kertesz, Knox
Barton, Moore

Patterson, Clegg

Hamilton, Redmond,
Lynch

Poppendiek, Palmer

Hoffman, lones
Claiborne, Winn
Farmer

Harrison

Claiborne

White, Blankenship

Grimes, Blankenship,

Nessle

Mann, White, Nessle
Reber, Mann |
Mann, White, Cougfi]en

Tunnell, Salmon

181
ANP PROJECT QUARTERLY PROGRESS REPORT

Decontamination of Fluoride Systems
Fluoride Salvage and Disposal
Tanks and Piping for High-Temperature Fluid Systems

Small-Scale Handling of High-Temperature Liquids

I. Liquid Metal Handling

1.

2
3
4.
5

Equipment Cleaning Techniques
Sampling Techniques

LLiquid Metal Salvage and Disposal
Liquid Metal Safety Fquipment

Blanket-Gas Purification

J. Dynamic Liquid lLoops

Operation of Convection Laops
Operation of Figure-Eight Loops
UQ,-NaOH Slurry Loop

Operation of Thermal Convection Loops

Fluoride-to-Liquid Metals Heat Exchanger System
Fluid Circuit Mockups

Reactor Mockups

K. Materials Analysis and Inspection Methods

1.

W

o ~I O ot

10.
11.
12.
13.

Determination of Metallic Corrosion Products in
Fluoride Eutectics

Analysis of Fluoride Eutectics

Determination of Trace Impurities in Inorganic
Fluorides

Chemical Methods of Fluid Handling
Metallographic Examinations

Identification of Compounds in Solidified Fuels
Preparation of Tested Specimens for Examination

Identification of Corrosion Products from Dynamic
Loops

Assembly and Interpretation of Corrosion Data from
Dynamic Loop Tests

Liquid and Gas Quality Control

High-Temperature Mass Spectrometry

X-Ray Study of Complex Fluorides

Petrographic Examination of Fuels

L. BRadiation Damage

1.

182

Liquia Compound Irradiation in LITR

9201-3
9201-3
9201-3
9733-2

9201-3
9201-3
9201-3
9201-3
9201-3

9201-3
9201-3
BMI1
2000

9201-3
9201-3
9201-3

9201-2
9201-2

9201-2
9201-3
2000

9733-2
9733-2

9733-3,
2000

2000,
9201-3,
9733-3
9201-3
9735
3550

9766

3005

Mann, Coughlen
Mann, White, Coughlen
Mann

Nessle, Boody,
Weinberger

Mann

Mann, Blakely
Devenish, Mann
Devenish, Mann

Marnn

Adamson, Reber
Coughlen

Simons

Cathcart, PBridges,
Smith

Salmon
Mann

Tunnell

White, Talbott

White, Boyd, Ross

White, Ross
Mann, Blakely

Gray, Krouse, Roeche

Barton, Anderson
Truitt, Didlake
Hoffman, Blankenship,

Smith, Borie, Dyer

Smith, Vreeland,
Adamson, Blankenship,
Bl akely

Mann

Baldock

Agron

McVay

Keilholtz, Morgan,
Webster, Robertson,
Klein, Kinyon, Feldman
FOR PERIOD ENDING JUNE 10, 1952

2. Liquid Compound Irradiations in Cyclotron
3. Liquid Compound Irradiations in MTR

4. Flvoride Fuel Irradiation in Berkeley Cyclotron

5. Liquid Metal Corrosion in X-10 Graphite Pile Loops
6. Stress Corrosion and Creep in LITR Loops

7. Creep of Metals in X-10 Graphite Pile and LITR

8. Thermal Conductivity of Metals in X-10 Graphlte
Pile and LITR

9. Diffusion of Fission Products from Fuels

10. Neutron Spectrum of LITR

Strength of Materials
1. Creep Tests in Fluoride Fuels

2. Creep and Stress-Rupture Tests of Metals in Vacuun
and in Fluid Media

3. High-Temperature Cyclic Tensile Tests
4. Tube-Burst Tests |

5. Tube-Burst Tests

6 Relaxation Tests of Reactor Materials
7 Creep Tests of Reactor Materials (GE)
8. High-Temperature Strein Gauges
g

High-Temperature Strain Gauges

Metals Fabrication Methods

1. Welding Techniques for ARE Parts
Brazing Techniques for ARE Parts
Mol ybdenum Welding Research

2

3

4. Molybdenum Welding Research

5. Resistance Welding for Mo and Clad Metals
6

Welds in the Presence of Various Corrosion Media

~3

Nondestructive Testing of Tube-to-Header Welds

8. Basic Evaluation of Weld Metal Deposits in
Thick Plates

9. FEvaluation of the Cone-Arc Welding Technique
10. Development of High-Temperature Brazing Alloys

11. Evaluation of the High'Temperature Brazing Alloys

9201-2

3025

NAA
3001

3005

3001,
3025

3001,
3005

3001
3005

9201-3
2000

2000
9201-3
2000
2000
2000
BLH
Cornel]

" Aero Lab.

2000
2000
BM1
MIT
RPI
2000

2000

2000
2000

Wall-Colmonoy

2000

Keilheltz, Feldman,.
Sturm, Jones, Knight

Keilholtz, Klein, Kinyon,
Morgan, Robertson

PearIman

Sisman, PBauman, Carroll,
Brundage, Parkinson,
Ellis, Olsen, Jones,
Morgan

Sisman, Bauman, Carroll,
Brundage, Parkinson,
Ellis, Olsen, Jones,
Morgan

Wilson, Zukas, Davis
Cohen, Templeton

Keilholtz, Robinson

Sisman, Trice, Lewis

Adawson, Reber

Oliver, Woods, Weaver

Oliver, Woods, Weaver
Adfimson, Reber
Oliver, Woods, WeaQer
O]iver, Woods, Weaver
Oliver, Woods, Weaver
Tatnall ‘

Puffer, Grey, Donovan

Patriarca, Slaughter
Patriarca, Slaughtcr
Parke

Wulff

Nippes, Savage

Vreeland, Patriarca,
Slaughter

Patriarca, Slaughter

Patriarca, Slaughter
Patriarca, Slaughter
Peaslee

Pafriarca, Slaughter

183
ANP PROJECT QUARTERLY PROGRESS REPORT

4
5.
6
7

2
3.
4
5

L=t

PROJECT DIRECTOR
ASSOCIATE DIRECTOR FOR ARE J.

ASSISTANT DIRECTOR FOR COORDINATION

STAFF ASSISTANT FOR PHYSICS

New Metals Development
1.
2.

Mo and Cb Alloy Studies

Heat Treatment of Metals

Solid Fuel Element Fabrication
1.
2.
3.

Selid Fuel Element Fabrication
Diffusion-Corrosion in Solid Fuel Elements

Determination of the Engineering Properties of
Solid Fuel Elements

Electroforming Fuel Tube-to-Header Configurations
Electroplating Mo and Cb

Carbonyl Plating of Mo and Cb

Rolling of Fuel Plate Laminates (GE)

Ceramics and Metals (Ceramics
1.

BeQO Fabrication Research
Metal Cladding for BeQO
B4C Control Rod Development

Hot Pressing of Tungsten Carbide Bearings

Hot-Temperature Firing of Uranium Oxide to Produce

Selective Power Sizes
Development of Cermets for Reactor Components

Ceramic Coatings for Stainless Steel

Ceramic Valve Parts for Liquid Metals and Fluorides

IV.
AT OAK RIDGE NATIONAL

Administrative Assistant

Project Editor

Shielding Research
Reactor Physics
Critical Experiments

Nuclear Measurements

 

184

*Dual capacity,

> o = M=

2000
2000

2000
2000

2000

Gerity Mich.
Gerity Mich,

2000

3012,
2000

Gerity Mich,
Gerity Mich,

2000
2000

2000
9766
9766
OSU

LABORATORY

R, C. Briant*

H. Buck*

A, J. Miller*

L. M. Cook
W. B. Cottrell

Bomar, Coobs
Bomar, Coobs
Bomar, Coobs
Bomar, Coobs
Bomar, Coobs
Graaf
Graaf
Bomar

Bomar, Cunningham,
Leonard

Graaf

Graaf

Bomar, Coobs
Bomar, (Coobs
Bomar, Coobs
Johnson, Shevlin
White

Shevlin

TECHNICAL ADMINISTRATION OF AIRCRAFT NUCLEAR PROPULSION PROJECT

PROJECT DIRECTORY

SECTION NO.
K. Ergen*
P, Blizard 11 A,B,C,D,E
K. Ergen* I E,F
D. Callihan I G
H. Snell IT A
STAFF ASSISTANT FOR BRADIATION DAMAGE

Radiation Damage

STAFF ASSISTANT FOR GENERAL DESIGN

General Design

STAFF ASSISTANT FOR ARE
ARE Operations
ARE Design

STAFF ASSISTANT FOR ENGINEERING RESEARCH

Heat Transfer and Physical Properties
Ceramics
Experimental Engineering

STAFF ASSISTANT FOR METALLURGY
Metallurgy

STAFF ASSISTANT FOR CHEMISTRY

Chemistry
Chemical Analyses

FOR PERIOD ENDING JUNE 10,

o >

=

R.

H.
J

H.

JI

= =

=}

R.
Bl

° DI

Miller*

. Billington

. Fraas*

. Fraas*

Buck*
Bettis

Schroeder

PBriant*

Poppendiek

. Warde
. Savage

.. Manly#*
. Manly

Grimes*

Grimes*®
Susano

11y

111
111

111

111

I11
111

1952

L
A
C,D
B,D
F,G
Q
H,1,J,K
D,H,1,J,K
C,D,E,J,K,M,

N,0,P,Q
A,B,D,E,F,H
K

185