ORNL-1375.txt

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ORNL- 1375

This document consists of 180 pages.
Copy f?@fi of 237 copies. Series A,

Contract No. W-T405-eng-26

AIRCRAFT NUCLEAR PROPULSION PROJECT
QUARTERLY PROGRESS REPORT

for Pericd Fnding September 10, 1952

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

A. J. Miller, Assistant Director

EDITED BY:
W. B, Cottrell

DATE ISSUED

 

0AK RIDGE NATIONAL LABORATORY
Operated by
CARBIDE AND CARBON CHEMICALS COMPANY
A Division of Union Carbide and Carbon Corporation
Post O0ffice Box P
0ak Ridge, Tennessee

 
 

MARTINMARIETTA ENERGY SYSTEMS LIBRARIE

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3 445k 0353149 9

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

R

 
 

 
 

DRNL-1375

    
    
    
 
     
 
 
  
 
    
     
 
   

P. Smith {(consultant)
H. Snell _
I.. SBteahly

Clewett
Clifford

-

Progress
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39.

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vi

Reports previously i1ssued in this series are as follows:

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

ORNL-1234

ORNL.- 1294

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

Reactor Program of the Aircraft Nuclear
Propulsion Project

Period Ending June 10, 1952

 
 

FOREWORD ., . . . . .

PART 1.

SUMMARY AND INTRODUCTION .

1.,

e

CIRCULATING-FUEL AIRCRAFT REACTOR EXPERIMENT

Fluid Circuit

Pressure-Shell Stress Analy81q'

Beflector Coolant
Instrumentation

®

&

Off-Gas System . .

Reactor Control System

Shielding . . . .

EXPERIMENTAL REACTOR ENGINEERING

Pumps . . . . .+ .

Frozen-sodium-sealed pump in figure € loop

&

-

-

.

TABLE

REACTOR THEORY AND DESIGN .

]

a

OF CONTENTS

a

-

4

-

-

-

Durco frozen-fluoride-sealed pump.
l.aboratory-size maintained-level gas-sealed pump

Durco pump with shaft packing

ARE centrlfugal pump .

Valves . . . e

Bellows fype of valve stem seal

.

»

»

»

»

High- temperature valve-stem packings

Sel f-welding of seat materials
Canned-rotor-driven valve

Heat Exchangers .

-

»

®

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* »

a

" ®

»

*

»

*

-

Test of core element of sodium-to-air radiator
Bifluid heat transfer loop .

NaK-to-NaK heat exchanger

Instrumentat1on .

.

Rotameter type of flowmeter

Rotating-vane type of flowmeter

.

Mcore nullmatic pressure-measuring device

Diaphragm pressure-measuring device
Diaphragm pressure transmitters

Moving-probe level indicator .

Null-balance level control
Strain-gage level indicator

Fluid Dynamics
ARE core mockup

 

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Possible fuel-header manifold for the reactor core ,

Technology of Fluoride Handling ., . . . « « « + « « « &

FlUOFide prOduction . . . s 2 . s . e e 8 s ¢ e
Removal of fluorides from contaminated systems . . .
Purity of pipe line helium . . . . « + . . . s e

Fluoride plug-removal test from simulated ABE core .
Injection of NaK into a flowing fluoride stream .
GaS'linE‘plugging tests " s * 4 e 8 ® & v & 8 8 & »

-
.
-

Cooled-baffle vapor trap for zirconium-bearing fluoride fuel

Descaling and pickling tests . . + + « + ¢ « « « + &

3. REACTOR PHYSICS . . . ¢« v ¢ v ¢ v v« 0 v o o o« o o o o

Oscillations in the Circulating-fuel Aircraft Reactor .
Effect of Gaps on Reactivity . . .« « ¢« & ¢« v o « « o &
SlOWing'Down KETHEIS . ] . . . . » . . . . @ . . » . .

4., CRITICAL EXPERIMENTS . . . « & v v v o 4 v v 0 0« o o 4 s

Direct-cycle Reactor . . + & v ¢« v ¢« « v ¢ o 0 « v o

RBeflector studies « o « o o s & o & o s & o« o o o«
Poison rod calibrations .+ . « o« o + &+ s o o &+ & &

ARE Critical Assembly o + + v v & v & @ o o o & o « o

PART 11. SHIEIDING RESEARCH . . . .

SUMMARY AND INTRODUCTION . . . . + ¢« v v & o v o o o« v o«

5. LID TANK EXPERIMENTS . . . . . . « « v ¢« v ¢ o v v o &

Air Ducts . . . e e e e e e e e e e e
R-1 Reactor Gamma Shleldlng e e e e e e e e e e e
Other Lid Tank Experiments . . . . .+ « « « « ¢« + « &
Correlation of Neutron Attenuation Data « « « « o o« &

6. BULK SHIELDING REACTOR . . . . . . . « « .+ . « . .

Mockup of the Divided Shield . . . . . « « « + « . .
Air-scattering Experiments . ,

Reactor Power Determination . « + + ¢« v « + v o « o + &
Irradiation of Animals . . . . .+ « & « ¢ v « 4 4 0 .
Irradiation of Electronic Equipment . . . . . . . . . .
Capture Gamma-Ray Measurements ., . . . . . . . . . . .

7. TOWER SHIELDING FACILITY . . . . . + + « « o v o o v o

8. NUCLEAR MEASUREMENTS . . . . s e e e e 4 s e s e e

vilil

Fission Cross Section of U234 and U?3%¢, . . .

 

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Total Cross Section of N . . . . . . . ..
Time'flf'Flight Sp&ctrometer > L] ° » . - » » [ a »

PART 111. MATERIALS RESEARCH . . .

SUMMARY AND INTRODUCTION . « & & o v v v 0 o v o o o o

9, CHEMISTRY OF HIGH-TEMPERATURE LIQUIDS . . . . . . . . .

Fuel Mixtures Containing UF, . . . . . . . . . .
NaF-KF-ZrF, -UF, . . .« v v v o 0 o o v 0 0 o v o
NaF-ZrF, UF s e e e e e e e e e e e e e e e
NaF- Zr¥, BeF UE, o o s e s s e e e e e e e
ZnF —UF4 s e s s s s e e sa e s s a e e e

Fuel Mixtures Containing UF,. . . . . . + « + « « + « .
NaF-KF-LiF-UF; . . . .« . . « v « o v o o v« o o
KE-UF, . . « v o v 0 v« o o v 0w o o 0 s s v 0w s
NaP'UF e e s e e e s a e e s e e a s e e e e e
Preparatlon of cempounds of UP o 4 4 a4

Alkal: Fluoborate Systems . . . v ¢« « & & + & & « « o« &
Vapor pressures of commercial flucborates . . . . .
Thermal data for fluoborate systems . . . . . . .

Differential Thermal Analysis . . . . .

X-Ray Studies of Complex Plufirlde Sy%temq o 4 e s e
Na ZrF -Na UF, . . .« o o o o o v o b o o 0 v 0 o
NaQZrF Nd UF o & 3 e s m m e e 5 e e e s e s s e
NaZrF, NaUP s e e s s e e e e e e e e s

Spectrcgraphlv Analysws of Fluorldes O

Simulated Funel Mixture for Cold Critical Experiment ., .
Moderator Coolant Development . . . . . . « . +« « 4« 4
Coolant Development . . o . ¢« s & & @ ¢ v o o« « o o »

RbF-AIF, . . . . v o v v o v e v o e e e e
NaF-KF- A1F3 e b e e e s v e e s e e e a e e e
NaF-ZrF,-ALF, . . . « o o o0 v 0 v 0 v e s
NaF-ZrF, . . . 4 o o v s o oo v 0 o v v e e e e
ZrF BeF o . ’ s e e s e e e e
Hydrolybls and 0x1dat10n of Fuel MlXtUFEb ca e

Heat of solution and heat of formation of uranyl fluorlde
Free energy change for the reaction of uraninm tetrafluoride

with oxygen . . e e e
Fuel Purification Research . . . e e e e e s

Purification of molten tlu011deb e e e e e s e e s

Preparation of pure zirconium tetrafluoride
Preparation of pure aluminum fluoride . .

 

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1x
Reaction of Fuels with Alkali Metals . . + & + &« & « & + o o & o o o 93
Service Functions .+ .« o v o o t o s s s s e e 4 6 4 s 4w w4 e e 94

CORROSION RESEARCH + + + o v v v v e e e e e s s e s s e e s s s e 95

Parametric Studies of Fluoride Corrosion . . « ¢« v &+ ¢ v o o o v o« + & 96
Temperature of test . + « v v o ¢ ¢ ¢ v 4 e 0 e e e e s 4 e e e e 96
Length of test + « v v v &« o o o o o s o s o o s s o v 1 2 a2 & 4 96
Residual stresses in SPECIMEN +« « » o o o o s o &+ & o o o o s 1 o » 99
Carbon content of specimen . .+ + + ¢« « o ¢ ¢« « & o o 0 s 4 e e e 99
Corrosion inhibitoTs + « v « 4 « o o = o s o o« & o a o & s o « o » 100
Pretreatment .« o+ + « o o o o o o o o o o 0 o o 4 w0 e e o0 o0 owo. 101

Seesaw and Static Tests with Fluorides Containing ZrF, . . . . . . . . 103

Fluoride Corrosion in Thermal Convection Loops « . + « « ¢ & &« « « . » 103
Fluoride mixtures containing ZrF, . . . . . . .+ + « v « o « . . . 104
Effect of cleaning loops « « « v s 4 & o« o o o 0 o s v o+ + o« . 104
Corrosion inhibitors ..ce v ¢ 4 & & ¢ & o o o o o + o + + & + o« « . 108
Crevice COTTOSION « wsv o + o o o o o o o o & o o o o o o o o o o« 109
Temperature variations . + « + o & « o o o o « o o o o o« o o 2+« 110
Variations in alloy composition of the walls . . . . . . . . . . . 110
Postrun examination of fuels « . & « v ¢ v v v &+ o o o o & o & o . 112

Compatibility of Beryllium Oxide with Various Fluoride Mixtures . . . 112

Hydroxide COTroSion « « v o« o o o & & o o o o o & o o« & o o o+« + 4 114

e v e e e e e s 114

Temperature of test . + . + v v v v v 4 v e a0 v e

Residual stresses in specimen . .« 4 + & o + & o & o o« « o+ + « « o o 114
Corrosion inhibitors . . . + v ¢ & & &« & o« o & o s &« « = o + & o« 115
Hydrogen atmosphere . + + « « v v 4 o o o o o & o o o o o v o o o o 115
Liquid Metal Corrosion . . . e e e e e e e e e e e e e e e e e s 1T
Mass transfer in liguid lead e e e e e e e e e e e e e e e e 1T
Effect of temperature on lead corrosion . o+ « « « & & o« o o o« « o« » 118
Lead-sodium miXtUTes .+ &« & &+ & o + o o & o s « & & o o 2 « « & + . 118
Spinner tests with sodium ., . . . . . . . . e e e e e e e .. 119

Type 1020 chromized steel in low-melting- p01nt alloys . .. 119
Compatibility of beryllium oxide with sodium, NaK, and lead « .. 119
Fundamental Corrosion Research . . . . + « & « « « &« & o o o o « « « . 121
Interaction of fluorides with structural mwetals . . . . . . . . . . 121
Synthesis of complex fluorides . +« + « « o o« & o & o o &+ o & o o o 122
EMF measurements in fused fluorides . . . « + + « « « « « &« « &+ « . 123
Polarographic studies in sodium hydroxide . . . . . « « « . . . . . 124
Electrochemistry of sodium hydroxide . . . + « ¢« & « &« « « + + o » 124
Mechanism of fluoride COTTOSION « & &« + & « & & & o o o s o + + « o 126
Tripositive nickel compounds from hydroxide corresion . . . « . . . 127
Solutions of metals in molten halides . . . « « « « .+ « « « +« « + . 130
Fluoride corrosion phenomena . .« + + + « o o« o o « o o« o o+ o « « o+ 130

 
 

11, METALLURGY AND CERAMICS . . . & . v v ¢ v 4 s e s s e e v e o o » o o« 131
Control Rod Fabrication . + + +ie v o « o v 4 ¢ & = o o o a o 2« o . 131
Mechanical Testfng of Materials . 4 o « v v 4 4 ¢ « & 4 o « o « + « » 132

Physical testing in fluorides . . . « + + « & « « o 52 v o+ » o 132
Creep and stress-rupture tests in ArgoOn . « + &+ & « » 4 + o+ » » . 133
Welding . . . e e e e e e e e e e e e e e e e e e e e e e e e w133
Cone-arc weldlng e b s s 4 e e e ah e e h e 4w a e e e e e .. 135
Specifications for ARE weldlng e e e e v e e a e e s e o a s oa ow . 135
Tests of Brazing Alloys . -+ . .« v & o & & + 4 « & & 4 « o« » « o« o » . 138
Static corrosion tests of brazed joints . . . . . . + .+ . . . . . 138
Tensile strength of brazed joints . . . . + + ¢« « ¢« + + &« +« « » . 138
Brazing of molybdenum . . « . & + © 4 & v . 4 4 v s . e . e . . . 140
Dry-hydrogen brazing furnace . . . .+ « « & & o « o o » o + « » & . 140
Reduction of Molybdenum Disulfide . . . . . . < « « . « « v « « » » ., 141
Ceramics Research . . . . « .« . v o 4 o o 4 o v v v e s s e e e .. 143
Ceramic coatings for metals . . . « & v « & + 4+ « « 2 & o« « + « . 143
Ceramic reflector . . &+ & & & =« & o 5 « 4 s s o & s o o + & v « . 143

12, HEAT TRANSFER AND PHYSTCAL PROPERTIES RESEARCH . . . . . . . . . . . . 144
Viscosity of Fuel Mixtures . . & + & « « » & o o 5 o o o s o « » » o 145
NaF-ZrF,-UF, fuels . . . . . v . .« o o o o v v v v v o v o o v . 145
Capillary viscometer . + « « v « o » v 4 & s + « & 5 «. 6 « o+ + « . 145
Thermal Conductivity of Liquids . . . « + . & & 4 « &« + « & o« o o o« o 146
Heat Capacify v = « v v 4 v v v 0 o v n o o 4 e 4 e e e e e e e e . LEB
Density . . . . .o C e e e e e e e e e e e e e e el 14T
Vapor Pressure of Fuel Conqtltuents P e e e e e e e e e e e e e e . 147
Zirconium tetrafluoride . . . . 4 . « o . . ¢ o 4 e e s . oa e s . 147
Zirconium-bearing fluoride mixtures .+ « .+ » + + 4 « o« 2+ . . . 147
Convective Heat Transfer in Molten Sodium Hydroxide . . . + . . . . . 148
Boiling Heat Transfer in Mercury . . soe s e b e s e . . o, 149
Natural Convection in Confined Spaces w1th Volume Heat Generation . . 149
Heat and Momentum Transfer Analysis of the Thermal Convection Loops . . 152

13. RADTATION DAMAGE . . . . s e e e e s e e e e e e e . 152
Irradiation of Fused Materlals e e s e e e e e s w e a e s w w183
In-Reactor Circulating LooPs v v v v v o« « 4+ o o o o s o 4 o o & o+ . 154
Creep Under Irradiation . . . ., . s e e e e e a e e e e 4 e s .. 154
Radiation Effects on Thermal Conduct1v1ty O X1

PART IV. APPENDIXES , . . . . . . . . . . . . 157
SUMMARY AND INTRODUCTION . & v v v v e e e e e e e e e e e N ¥ 1

 
 

14, ANALYTICAL CHEMISTRY . . . . . . .. e e v e e e e e e e e e e .. 159
Analytical Studies of Components of Fluoride Mixtures . . . . . . . . 159
Alkali metals o & ¢« & v v v 4 v 4 4 e e e e e e e e e e w e o« . . 159
ZITCONIUM v v & o & o & & & + o s & o s & o o s o o 4 o o o o« o 160
Analytical Studies of Impurities in Fluoride Mixtures . . . . . . . . 160
Chromium « & v v & o & &+ & o « o & a & s o o o o s s o o s 2 o+ « » 160
Nickel . v . v v v e e e e e e e e e e e e e e e e e e . )6]
OXYEEN « v & v s & & & & & & o 4 &« 4 4 4 e 4 w4 e e e e e 161
Chloride . « « v & v 4 4 4 4 4 & 4 4 o & 4 o o 4 o o o o 2 o « . 161
Water o o o 4 v v 0 v v e e e e e e e e e e e e e e e e e e 162
Determination of Carbon in Zr¥, and Zr0,-NaF-C Mixtures . . . . . . . 162
Compatibility of Reactor Fuels and NaK . . , . . . . . . . . . . . . 162
Service Analysis + v 4 4 4 4 4 4 4 4 e 4 s s 4 4 4 4 s s e e e e o4 . 163

15, LIST OF REPORTS ISSUED . . . ¢ v v v v v v v v v s v v v s e v o v o o 164

 
 

ANP PROJECT QUARTERLY PROGRESS REPORT

FOREWORD

The Aircraft Nuclear Propulsion Project is 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
performed in support of other organizations participating in the national ANP
effort. However, the bulk of the ANP research at ORNl. is directed toward the
development of a circulating-fuel type of reactor. The nucleus of this effort
is now centered upon the Aircraft Reactor Experiment ~ a 3-megawatt high-
temperature prototype of a circulating-fuel reactor for the propulsion of
aircrafeu,

This quarterly progress report of the Aircraft Nuclear Propulsion Project at
OBRNL records the technical progress of the research on the circulating-{fuel
reactor and all other ANP research at the laboratory under its Contract W-7405-
eng-26. The report is divided into four parts: I. Reactor Theory and Design;
II. Shielding Research; III. Materials Research; and IV. Appendixes. Each
part has a separate Summary and Introduction,

 
 

 

   
 
SUMMARY AND

The over-all concept and design of
the Aircraft Reactor Experiment were
set forth in the two preceding re-
ports. The most significant modifi-
cations during the past quarter have
been those attendant to the specifi-
cation of NaK as the reflector coolant
and the NaF-ZrF, -UF, mixture as the
circulating fuel. Most of the equip-
ment for the experiment is now on order
and some has already been received and
installed in the ARE Building. The
safety aspects of the reactor are being
analvzed, particularly, the off-gas
disposal system and the effect of a
postulated fuel~tube rupture (sec. 1),

Valves, pumps, and instrumentation
for the fluid circuit of the Aircraft
Reactor Experiment are being developed
{sec. 2), Valves with both bellows
seals and packed sesls have been suc~
cessfully used with molten fluorides,
but the bellows seal appears to be the
more reliable., Water tests with the
maintained-level gas-sealed pump pro-
posed for the ARE have been unsatis-
factory. Since the packed seal 1s
satisfactory when the back end 1is
maintained at temperatures below the
fluoride melting point, a pump 1n-
corporating this seal has been specified
for the ARE., Tnstrumentation for the
ARE fluid circuits must be designed so
as to be unaffected by the high vola-
ti1lity and subsequent condensate of
the ZrF,-containing fuel, Accordingly,
flowmeters, pressure-measurement

INTRODUCTION

devices, and fluid-level indicators
and controls are being redesigned and
tested for this application, Data from
the NaK-to-air radiator tests corres-
pond exceptionally well with theoreti-
cal values. A hydrodynamic mockup of
the ARE core has served te illustrate
the problems associated with fuel
loading and draining. Also of signifi-
cance to the ARE is ap experiment in
which a fuel-tube rupture is simulated
so that NaK {(the reflector coolant) is
injected into the fuel stream.

Theoretical investigations of the
kinetics of the circulating-fuel
reactor indicate that all subseguent
oscillations will be moderate 1f the
survives the first, short,
The applicability
of multigroup calculations to reactors
with vastly different core and re-
flector materials 1s uncertain pending
critical experiments. An improved
method of computing the effects of
gaps on reactivity has been developed
(sec. 3),

reactor
maximum excursion,

The recent experiments on the
critical assembly of the General
Electric sir-cooled water-moderated
reactor {R~1) have included reflector
studies and poison-rod calibrations,
The preliminary assembly of the circu-
lating-fuel reactor for the ARE has
been critical with a uranium mass in
excellent agreement with that pre-
dicted (sec. 4},
ANP PROJECT QUARTERLY PROGRESS REPORT

1. CIRCULATING-FUEL AIRCRAFT REACTOR EXPERIMENT

E. S, Bettis

R. W. Schroeder

ANP Division

The main core design and the general
concept of the circulating-fuel air-
craft experiment are essentially the
same as reported 1in the preceding
quarterly report.t1) The few changes
made in the design and in the 1in-
strumentation were necessary because
of certain fuel characteristics; the
most probable fuel (NaF-ZrF,-UF,) has
higher vapor pressure and higher
viscosity than expected, and some
fluid-circuit design characteristics
have had to be altered accordingly.
NaK has been selected as the reflector
coolant after an examination of the
hazards introduced by a postulated
fuel-tube rupture. Although a fuel-
tube plug would almost certainly
result 1f the rupture were large,
studies on the simulator indicate that
the reactor could be safely scrammed
and drained. The off-gas system in-
corporating a vapor trap and a low-
temperature carbon absorber unit is
being designed. Final drawings of the
reactor assembly and fluid-circuit flow
sheet are shown in Figs. 1 and 2. A
detailed description and design draw-
ings of the ARE were published in the
ARE status report,(?’

FLUID CIRCUIT
G. A. Cristy, ANP Division

Some problems have arisen regarding
the design of the fluid circuit as a
result of the high viscosity of the
NakF-ZrF, -UF, fuel. Since the viscosity
of this circulating-fuel mixture 1s

 

(I)H. ¥W. Schroeder, Aircraft Nuclear Propulsion
Project Quarterly Progress Report for Period
Ending June 10, 1952, OBNL-1294, p. 9.

(Z)W. B. Cottrell, Reactor Pragram of the Air-
craft Nuclear Propulsion Project, ORNL-1234 (June
2, 1852),

twice as high as originally anticipated,
1t caused a reduction of one-half in
the Reynold’'s number within the heat
exchanger tube system, that is, from
approximately 8000 to approximately
4000, This, in turn, has caused a
reduction 1n the tube-side heat-
transfer coefficient, such that the
calculated minimum film temperatures
are in the vicinity of the freezing
point of the fuel mixture (940°F),
In order to re-establish the minimum
film temperature sufficiently above the
freezing point, two major changes were
made. First, the operating temperature
range was raised from between 1150 and
1500°F to between 1200 and 1500°F,
which i1ncreased the flow rate by a
factor of 7/6 and increased the minimum
bulk fuel temperature by 50°F. Also,
the heat exchanger tube arrangement
was modified., In contrast to the
previous arrangement with five tubes
in parallel, the present arrangement
has all the fluid passing through two
tubes in parallel and thence through
a second bank of three tubes 1in
parallel. By incorperating these two
changes, the minimum calculated film
temperature was 1increased to about
1050°F. The changes also increased
the system pressure drop and the
pressure level within the pressure
shell., The consequences of these
differences are discussed in the
following section,.

The gas-sealed pump originally
designed for use with Na or NaK has
been found to be unusable for fluoride
circulation without extensive modifz1-
cation of the seal. Since the surge
tank design is 1ntimately related to
that of the pump, details on the surge
tank will not be released prior to
resolution of the pump problem, All
other details of the fluid circuits
FOR PERIOD ENDING SEPTEMBER

REGUL.ATING ROD
ASSEMBLY .

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SAFETY ROD
GUIDE SLEEVE

SAFETY ROD ASSEMBLY e

   

1952

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are proceeding as rapidly as drafting
man power permits. The fuel-injection
system has been completely detailed,
and the egquipment is now being fabri-
cated in the local shops. The bellows
valves for use in the fluoride circuits
are being supplied from an outside
vendor. It was found that the welding
of these units could be done better by

 

    

SCALE IN INGRES

Experimental geactor (Elevation Section).

OBNI, and arrangements for doing this
welding for the vendor were completed.

STRESS ANALYSIS

J. W, Walker
ANP Division

PRESSURE -SHELL

B. L. Maxwell
Consultants,

For a pressure of 100 psi in the
pressure shell, the maximum bending
T oaRacN '
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o
 
stress 1n the heads of the shell
would be 8500 psi. At a temperature
of 1150°F this gives a creep rate of
about 1% in 10,000 hr and a factor of
safety of 1.6 based on stress-rupture
in 1000 hours. By plate reinforcing,
the maximum bending stresses are
reduced to 6000 psi. This 1s slightly
less than the stress for a creep rate
of 0,10% in 10,000 hours. As stated
in the previous report, all holes that
occur 1in areas of high stress are re-
inforced.

With added reinforcing, the stresses
in the side walls have also been
reduced to a maximum axial stress 1in
the cylinder of about 6000 psi. With-
out added reinforcing, the stress at
1006 psi is 7430 psi. The reinforced
side walls give a creep rate somewhat
in excess of 0,10% in 10,000 hr and a
factor of safety of 1.8 based on
stress-rupture in 1000 hours.

REFLECTOR COOLANT
E., S. Bettis, ANP Division

The fuel-carrier NaF-ZrF, was
originally considered as the most
probable choice for the reflector
coolant.,  However, recent dynamic
tests of the corrosion of the beryllium
oxide reflector by a similar fluoride
mixture, NaF-KF-ZrF,, showed that the
beryllium oxide was dissolved to an
extent that would prohibit the use of
that fluoride as a reflector coolant,

Two alternative secondary coolants
have been considered: NaK and the
fluoride eutectic NaF-BeF,. The choice
between these two coolants resolves to
an analysis of the hazards attendant
to a fuel-tube rupture that would
connect the fuel and reflector-coolant
circuits, Although i1t 1s unlikely that
a rupture will occur and allow the
coolant te leak into the core, the
possibility of such a leak must be

admitted., Furthermore, if such a leak
does occur, the reactor must and will
be shut down no matter what reflector
coolant is used.

The fluoride eutectic NaF-BeF, has
acceptable physical characteristics
and 1s compatible with both the
beryllium oxide reflector-moderator
and the fluoride fuel. An objection
to the use of this coolant 1s the
expense and time involved in producing
a sufficient quantaity for the ABE.
Another objection, which would net be
present 1f NaK were used, emerges from
the consequences of a fuel-tube leak.
With the relatively inert NaF-BeF,
mixture as the reflector coolant, in
the case of a fuel-tube leak there is
the possibility that sufficient fuel
could seep into the moderator inter-

stices to overcome all contrel rod
effects.
NaK, however, is known to reduce

fluorides such as ZrF, and UF, to
lower valence states of the metal.
Since these reduced compounds have
higher melting points than the original
material, a fuel-tube plug could
readily occur. That this would
actually happen with a sufficiently
Jarge rupture has been demonstrated
experimentally (cf., sec. 2, “Experi-
mental Engineering’ ),

The reactor system simulator was
used to study the effect of a leak of
NaK into the fuel channel of such
magnitude that the channel 1s com-
pletely plugged. When this occurs the
temperature of the plugged fuel passage
begins to rise at a rate of about
36°F/sec and will rise to an asymptotic
value of about BOO°F above the fuel
temperature at the time of the plug-
ging., This condition obtains by the
temperature coefficient alone 1f no
control rod correction is employed.
When the plugged tube begins to heat
the reactor goes on a 25-sec negative
period. The outlet temperature of the

11
ANP PROJECT QUARTERLY PROGRESS REPORT

five unplugged passages drops at the
rate of about 15°F per second. This
drop in temperature of five out of six
outlet fuel tubes provides a unique
symptom of plugging that will enable
the reactor to be shut down by scram-
ming and subsequent draining of fuel.

Since there appears to be some
additional safety to be realized from
the use of NaK and NaK has the most
favorable physical properties of the
coolants considered, 1t will be used
as the reflector coolant in the ARE.
The secondary heat exchanger and
associated gear are being designed for
use with this coolant.

INSTRUMENTATION
S. A. Hluchan,

Instrument Department
Some ARE instrumentation has had to
be modified because of the high vapor
pressure of the ZrF, component of the
fuel mixture. Sufficient ZrF, con-
densate would accumulate in the gas
space above the hot liquid to com-
pletely clog the open instrument lines
in a few hours i1f the liquid tempera-
ture was around 1200°F,
the instrument lines
contemplated for use 1in
liquid level, pressure,
flow had to be replaced by devices
with no open lines. Bellows were

Consequently,
previously
measuring
and liquid

substituted as pressure indicators and
a fluid-immersed inductance type of
instrument was substituted for measur-
ing flowand liquid level. A discussion
of these instruments appears ain the
‘Experimental Engineering’’ section of
this report (sec. 2).

OFF-GAS SYSTEM
T. Boseberry, ANP Division
A carbon absorber unit at liquad

nitrogen temperature 1s being used to
completely remove the fission gas

12

(Br,, I,, Kr, Xe) from the helium
stream coming from the NaK vapor trap
before it is released to the stack.
The unit consists of four 4-im.-IPS3S
stainless steel tubes connected 1n
series by 2-in.-IPS stainless steel
tubes. The first and second tubes
cool the gas stream and solidify the
Br, and I,, and the third and fourth
tubes, which contain approximately 700
in,? of 20-28 mesh CXA carbon, absorb
Kr and Xe.

With normal off-gassing (unit
designed for 10 fe*/hr; probably the
flow will be less and not continuous),
the carbon bed has a break-through time
of at least 800 hours., If it becomes
necessary to purify the helium atmos-
phere in the heat exchanger pit be-
cause of a hot leak, each of the two
units can take a flow rate of 100
ft? /hr for at least 70 hours. '

The break-through time of the unit
was obtained as follows:(?)

1 A%
t”? = 2.30 + 26.1<-~-~>
Q

where

[}

weight of carbon (g),

flow rate of helium (cc/min),

o~
i

time of break-through (hr),.

REACTORE CONTROL SYSTEM

E. P. Epler
Research Director’s Division
The reactor control system 1s
essentially the same as outlined in

previous reports. (%) Detailed drawings

 

(B)T. S. McMitllan and W, L. Johnson, Pissolver

Off-Gas Processing, ORNL-1309 {(to be published).

(4)R. ¥. Schroeder, op. cit., ORNL-1294, p. 12;
R, W. Schroeder and E. S. Bettis, Aircraft Nuclear
Propulsion Project Quarterly Progress Report for
Period Ending Marck 16, 1952, ORNL-1227, p. 10,
FOR

have now been completed for the
ionization~-chamber liner-and-shield
installation and for the ionization-
chamber gas system. Layout and wiring
diagrams for the relay cabinet are
being prepared. The construction of
ionization chambers, preamplifiers,
amplifier cabinets, relay cabinets,
contral console, and recorder panels
by the Radiation Counter lLaboratories
is on schedule for October delivery.

Components for a simplified servo
control have been assembled for testing
with the analog computor. The regulat-
ing rod speed has been established at
0.1% Ak/k per second, with a total of
0.4% Ak/k in a total rod travel of 12

inches.

The pilot model of the high-temper-
ature fission chamber 1s ready for
test. Two additional assemblies are
being constructed with completion ex-
pected in October.

SHIELDING
H. I_Ji P‘I

Enlund, Physics Division

The adequacy of the shielding and
the anticipated radiation levels in

PERIOD ENDING SEPTEMBER 10, 1952

various components of the ARE are being
studied. To date nothing untoward has
been brought to light.

The servicing of equipment in the
heat exchanger room will be governed
more by radiation that comes directly
from activated equipment and fission-
product deposition than that which
comes through the concrete wall from
the fuel dump tank or the shut-down
reactor. The rate of deposition and
the retention of fission products on
Inconel surfaces are not readily pre-
dictable, except by experiment. Tt
would appear that satisfactory removal
of fission products may be possible 1f
methods analogous to these for the
decontamination of solid-fuel process-
ing equipment were used.

An upper limit to the activation of
the Inconel by thermalized delayed
neutrons has been calculated by assum-
ing all are captured in the Inconel.

The cold trap in the helium scrubber
system is being investigated for the
heat generation caused by the retention
of radiocactive xenon and krypton on
the activated carbon,.

 

2. EXPERIMENTAL REACTOR ENGINEERING

H. W, Savage, ANP Division

Continued emphasis by the ex-
perimental engineering group on the
development of components for high-
temperature fluid-circulating systems
has resulted in improvements in pumps,
valves, seals, and instruments such as
those for indicating and controlling
liquid levels, measuring system
pressures, and measuring flow rates.

Valves for use in high-temperature
fluid systems, such as the ARE, must

operate with complete reliabiiity for
long periods over a temperature range
of from 1200 to 1500°F. The stem
seal 1s considered the most critical
problem in the operation of valves at
high temperatures; conventional sealing
materials are inadequate. Consequently,
the major effert of the valve progranm
has been on seal development, and
progress has been made in sealing
valves with both Inconel bellows seals
and stuffing-box seals packed with

13
ANP PROJECT QUARTERLY PROGRESS REPORT

Inconel braid and graphite and nickel
powders. Some time has been spent,
also, in considering several valve
configurations and 1n determining
whether self-welding of valve seating
materials may be encountered in the
presence of the ABRE fuel at operating
Lemperatures.

Seals are also the critical problem
in pumps. Gas seals, packed seals,
frozen seals, and combination packed
and frozen seals have been tested with
varylng degrees of success. The
combination of packed and frozen seals
appears to furnish the most positive
sealing action against high-temperature
fluoride fuel mixtures. The hydro-
dynamic tests of the first model
(Mcdel DA) of the ARE pump, with water
as the circulated fluid, indicated
that the pump sealing arrangement had
to be modified., Two avenues of

modification were then open: a sump
type of pump could be uwused in which
the fluid 1s not 1n contact with the

shaft seal but is separated by an
inert gas volume,or a
packed and frozen seal could be used.
The latter was selected for use in

the ARE.

combination

Heat transfer tests have continued
with sodium-to-air radiators with
varying fin spacing. The test results
correlate with the changes in fin
spacing made 1n the radiator configu-
rations tested and are in excellent
agreement with the theoretical heat
transfer. A test loop is near com-
pletion to make measurements of over-
all heat transfer coefficients of
fluorides to liquid metals.

Instrumentation for the ARE cannot
incorporate gas-sensing lines that
contact the liquid fluoride surface
because of the high-vapor pressure of
the zircouium fluoride. Accordingly,
flowmeters, pressure-measuring devices,
and fluid-level 1ndicators and controls
are being designed with closed liguid

14

surfaces. Two types of high-tempera-
ture fluid flowmeters are being
developed. A rotameter type of
flowmeter incorporating a tapered iron
core and an i1nduction coil as the
sensing element is being tested. A
rotating-vane flowmeter that uses a
permanent magnet and a pick-up coil
as the sensing element 1s being
prepared for test. Several pressure-
measuring devices are being developed.
These are divided i1nto two general
types in which the sensing element,
either bellows or diaphragm, operates
in a cooled, trapped-gas volume or is
completely submerged in the high-
temperature fluird. A float type of
level indicator and controller that
uses a tapered iron core and induction
coil to sense and control levels has
been developed that will reliably main-
tain two levels within 0.1 in. of each
other when used in connection with a
venturi for flow measurement. Other
level indicators are also being
developed.

Hydrodynamic tests have been con-
ducted with a glass mockup of the
ARE core. With 45-gpm flow 1t was
demonstrated that the system could be
filled with certainty when evacuated
to a pressure of 28 1in. Hg, but under
o circumstances can all the fluid
be removed from the core structure,
Later tests i1ndicate that with 50-gpm
flow,vacuum is not required for com-
plete filling.

A possible reactor header ar-
rangement has been developed that
gives an optimum flow pattern through
a reactor core that has single-pass,
parallel fuel tubes.
greatly reduces the amount of stagnant
fuel that would be held up in the
header as compared with that held up
in the conventional plenum chamber.

This design

The technology of fluoride handling
has progressed. Fuel production
equipment capable of meeting current
FOR

fuel needs is 1in operation. The
equipment for producing the ARE fuel
and moderator coolant is being designed.
Methods have been developed for
cleaning fluoride-contaminated systems
with high-pressure steam jets.

Tests were conducted 1in an attempt
to determine whether a plug of solid
fluoride fuel mixture could be remelted
in the ARE core-tube bends without
rupturing the tubes. Repeated cycles
of increasing severity failed to
rupture the tube under test, but a
slight swelling was noted in the bend
at the conclusion of the experaiment,

Fuel-flow stoppage resulted when
rapid additions of NaK, which 1s one
of the proposed ARE moderator-coolants,
were made to a stream of fuel, NaF-
Zr¥,-UF,, at 1500°F in a small, forced-
circulation loop. This test simulated
a large fuel tube rupture that, 1f
developed in the ARE fuel tubes, would
permit the moderator coolant to leak
into the fuel system.

The vapor pressure of NaF-ZrF, -UF,
(46.0-50,0-4.0mole %) is high compared
with other fuel mixtures considered,
and at ARE operating temperatures some
Zr¥, 1s sublimed from any {free sur -
faces. This sublimed material plugs
cool control-gas lines in dynamic
fluid systems and makes their operation
difficult. The vapor pressure problem
with this fuel 1s being studied, and
equipment that may alleviate the
problem is being developed,

PUMPS

Frozen-Sodium-Sealed Pump in
Figure 8 Loop (W. R. Huntley, ANP
Division). The Worthite frozen-
sodium-sealed pump test first re-
ported in ORNL-1154(!’ was terminated

 

(l)fi’. B, McDonald, Afrcraft Nuclear Propulsioen
Project Quarterly Progress Reporit for Period
Ending Septeaber 10, 1951, OBNL-1154, p. 21.

PERIOD ENDING SEPTEMBER 10, 1952

after more than 4000 hr of pumping
sodium at temperatures from 800 to
1200°F. During the major portion
of the test the sodium temperature
at the pump was approximately 1100°F.
A weld failure and a simultaneous
flange failure in the test loop caused
the termination of the test. :

Shaft speed during the greater part
of the test was 2000 rpm. Pump
operation was smooth during the test
except for a short period during which
there was interference between the
impeller and the pump bhousing. An
inspection of the pump shaft after
disassembly showed the shaft to be in
excellent condition except for a
build-up of nickel on the shaft at the
so0lid sodium-liquid interface. The
build-up of nickel was caused by mass
transfer from the nickel sealing ring
at the parting faces to the cold
region of the seal. Although this
build-up caused no operational diffi-
culty, it indicates that nickel waill
not be entirely satisfactory for use
as a sealing gasket for parting faces
for high~temperature pumps that are
to be operated for extended periods.

The loop is to be disassembled for
complete examination and oxygen
analysis of the sodium in the system
and in the cold stub that was placed
in the loop for trapping oxides from
the system.

Durco Frozen-Fluoride-Sealed Pump
(W. B. McDonald, W. G. Cobb, P. G.
Smith, ANP Division). The Durco
centrifugal pump (Model H34MDVX-80)
incorporating an improved design of
the frozen-fluoride seal has been
constructed and i1s being assembled
into a test loop. Previous tests of a
frozen-fluoride seal on a similar
pump, which operated for over 500 hr
at 1200°F, indicated that such a seal
will furnish a positive sealing action
against high-temperature fluorides.
Failure occurred at the parting-face

15
ANP PROJECT QUARTERLY PROGRESS REPORT

seal and some scoring of the shaft
occurred in the sealing area.

The operational characteristics of
the seal are considered to be the most
perplexing problem encountered. The
pump in the first test could not be
stopped for periods longer than 5 min
without considerable difficulty in
restarting. Upon stopping, the frozen
fluorides in the seal bonded the shaft
tightly to the sleeve and 1t was
necessary to apply sufficient heat to
melt the fluorides before operation
could be resumed. The improved design
incorporates a 2000-watt Calrod heater
by means of which the heat input for
melting of the fluorides in the seal
can be carefully controlled so that
the pump can be more easily restarted
after shutdown. The cooling fins on
this seal are baffled so that the air
flow across these seals can be ac-
curately controlled. This degree of
close control over the temperature at
which the seal is operated should
enable a better determination of its
optimum operating conditions. The
shaft for this pump is coated with
Stellite No. 3, which 1s harder than
the Stellite No, 6 used in the previous
test, to reduce shaft wear and scoring.

The weakest point of the pump is
the high-compression sealing joint at
the parting faces; nevertheless, it
1s expected that this pump can be
operated with the fluoride temperature
approaching 1500°F.

Laboratory-Size Maintained~-Level
Gas-Sealed Pump (W. G. Cobb, P. W.
Taylor, G. D. Whitman, ANP Division).
The redesigned gas-sealed pump re-
ported previously®?) has bheen con-
structed and placed in operation. The

 

(2)y, g, Cobb, Aircraft Nuclear Propulsion
Project Quarterly Progress Report for Period
Ending June 10, 1952, ORNL-1294, p. 17.

16

redesign consisted of the following
modifications:

1. The ring-joint parting-face
seal was moved from the region in
which 1t was in contact with high-
temperature fluorides to a higher
position where it now seals against
helium only.

2. Heat-radiation baffles were
placed between the high-temperature
liquid-gas interface and the top
flange of the pump.

3. The stuffing-box shaft seal was
replaced by a carbon-ring hardened-
tool-steel face seal.

The fluoride fuel NaF-ZrF,-UF, has
been pumped for over 250 hr at 1200°F
with this pump. The flow rate 1is
approximately 10 gpm, the shaft speed
is 2500 rpm, and a 16-ft head of fluid
is developed. Some operational diffi-
culties have been experienced which,
although not severe, indicate that
some further minor design changes are
needed. Turbulence of the free liquid
surface i1n the pump 1is rather violent
and prevents true level indication by
the spark-plug probes. The shaft
speed is limited to between 2500 to
3000 rpm because of this disturbance.
Some o0il leakage has occurred from the
bearing housing past the carbon-ring
face seal and/or the '0”-ring oil
seal 1nto the o0il catch basin im-
mediately below. Although this basin
1s drained continuously and no o1l 1is
permitted to drain into the fluorides,
the system 1s probably contaminated
by o1l vapors. The extent of this
contamination will be determined when
the pump is shut down and analysis of
the fluorides made. With only the
above difficulties, which are not
considered to be serious, this pump
continues to operate very smoothly
and 1ts success to date 1s very en-
couraging.
FOR PERIOD ENDING SEPTEMBER 10, 1952

Durco Pump with Shaft Packing
(W. R. Huntley, P. W, Taylor, H. H.
Johnson, ANP Division). Althoug
packed seals that will operate satis-
factoerily have been developed for
high-temperature pump shafts, such
seals must either be tightened at all
times to the point where shaft wear 1is
likely to Tesult or maintained at a
temperature below the freezing point
of the circulated fluid. During the
operation of a standard Durco pump
that was wodified to incorporate a
stuffing-box seal packed with Inconel
braid and nickel and graphite powders,
it was found that the pump could be
operated for long periods with zero
leakage of the stuffing-box seal only
when the back end of the seal was
permitted to operate at a temperature
below the melting point of the fluid
pumped. When only the metallic packing
is at a temperature above the melting
point of the fluid, a small amount of
fluid penetrates the packing inte the
cold regilion near the compression
member, where 1t freezes to seal the
pump (leakage rate approximately 1 g
of solid material in 24 hr). This
pump has operated for 450 hr, pumping
the fluoride fuel NaF-Zr¥F,-UF, at
temperatures from 1150°F to 1300°F.
The flow rate 1s 16 gpm, and an ap-
proximately 35-ft fluid head 1s de-
veloped at a shaft speed of 1500 rpm.

This sealing method was derived
because when the entire stuffing box,
including the compression member, was
heated to a temperature substantially
above the melting point of the fluid,
considerable leakage at a rapid rate
resulted. When the heat source was
remocved from the back end of the seal,
the leakage i1mmediately stopped and
the pump continued to operate smoothly
without leakage.

Operation of this Durco pump with
shaft packingis smoother than operation

o0f the Durco pump with the frozen
“seal.

When the pump operating power

1s measured, power surges of ap-
proximately 0.5 kw and about 1/2-min
duration are found to peccur at intervals
of 2 to 3 minutes. Similar power
surges occurred in operation with the
frozen seal that contained no packing,
but they were greater in magnitude
because of the longer frozen section.

The amount of wear on the shaft,
which 1s hard-surfaced with Stellite,
cannot be determined until the pump
1s shut down for examination; however,
the pump operation is sufficiently
smooth te indicate that the shaft
wear may not be severe. Two lubricants
(tricresyl phosphate-molybdenum
disulphide mixture and lead borate
glass) have been introduced into
the seal in an attempt to eliminate
the power surges. Either lubricant
reduces the amplitude of the power
surges temporarily but does not reduce
their frequency,. '

ARE Centrifugal Pump (W. G. Cobb,
A. G. Grindell, G. D. Whitman, ANP
Division). The Model DA gas-sealed
pump originally designed for the ARE
has undergone hydraulic tests 1in a
closed loop with water as the circu-
lated fluid. As designed, a combination
surge tank, de-gasser, and pump fluid-
level contrel has been placed in the
loop at the pump discharge. Figure 3
shows a schematic outline of this
circuit. The tank carries full flow
and 1n addition is connected to the
pump with liguid and gasequalizer
lines. The liquid connection leads
from the bottom of the surge tank,
enters the pump above the 1impeller
housing, and feeds a narvow, annular
volume surrounding the shaft to form
a liquid seal for the impeller shafrt.
The gas equalizer interconnects the
two gas volumes in the pump and surge
tank. '

Successful pumping has not been
obtained except when the gas equalizer
is valved shut. With this inter-
connecting line open as designed, the

17
ANP PROJECT QUARTERLY PROGRESS REPORT

GAS
SOURCE

  
    

MODEL DA CENTRIFUGAL PUMP7

 

 

Teo—VENT

UNCLASSIFIED
OwG, 16338

DEGASSER -AIR SPACE TO GAS CONNECTION

SURGE -DEGASSER TANK

H ’-)/ VENTURI
T [

 

 

   

ADJUSTABLE VARIABLE -
SPEED DRIVE

SUCTION TO GAS SPACE CONNECTION

DEGASSER-AIR SPACE TO LIQUID CONNECTION

THROTTLING VALVE

DRAIN VALVE

~FILL AND DRAIN VALVES

 

L G |

 

Fig. 3.

pump loses 1ts prime and, as observed
in a glass pipe section in the pump
discharge, the flow is almost wholly
gas. With the gas equalizer closed,
but with the liquid equalizer either
or closed, satisfactory pump
performance data have been obtained.
The performance data are presented in

Fig. 4.

open

LLiquid-level detection in the pump
has been attempted by means of a
sight glass, and satisfactory indicated
levels have been maintained with
separate gas-pressure supplies on the
pump and surge tank through a wide
range of flows. However, small,
rapid, pressure perturbations in the
system communicable to the pump will
cause the liquid level to rise or fall
unduly 1in the shaft annulus without
these changes being indicated on the
pump-level sight glass. Thus over-
filling of the pump bedy or gassing of
the liquid being pumped can occur.

18

el e

Loop for Water Test of ARE Pump.

UNCLASSIFIED
DWG. 16339

00

 

100 -

o
o

e am ] 50

HEAD (ft OF Hy0)

FRICIENCY (%)

o

 

 

 

0 .~ —_— o
0 50 100 150
FLOW (gpm)
Fig. 4. Water-Test Performance

Data for ARE Pump.
FOR PERIOD ENDING SEPTEMBER 10, 1952

Either overfilling or gassing would
result in flow stoppage if the fluid
involved was a fluoride.

Although 1t may be possible to
render perturbations of this type
harmless by some combination of
stationary vanes 1n part of the
annulus, by increased flow (or de-
creased head loss) from the surge
tank through the liquid equalizer
line, or by packing dampers 1in the
annulus, the development work required
appears to be considerable. In view
of the requirements of the ARE {for
high mechanical certainty of operation
and the progress on packed seals,
further development work on the
Model DA pump has been postponed
indefinitely.

Modifications for using the same
impeller and housing in both the sump
type of pump and packed-seal pump are
being carried out.

VALVES

Bellows Type of Valve-Stem Seal
(P. W. Taylor, ANP Division). One of
the most promising valve-stem seals
appears to be that constructed of
three-ply Inconel bellows, each ply
in the bellows being approximately
0.009 in. thick. Results of tests of
two such bellows have been so highly
satisfactory that this type of seal
has been incorporated in the design
of 1 1/2-in. throttling valves for
use in the ARE system and in pump
test loops.

One test was terminated for ex-
amination of the bellows after operation
for 1000 hr submerged in a bath of
fluorides at 1500°F, The bellows was
cycled once each 12 hr through 1ts
maximum rated travel of 9/32 inch.
Upon examination it was found that
only the first ply of the bellows has
been penetrated by the fluorides. A

second such bellows has been operated
under similar conditioens for over
1600 hr with no indication of failure.
This bellows has been cycled 127 times
during the test, which will continue
until failure. :

High-Temperature Valve-Stem Packings
(D. R. Ward, H. B, Johnson, ANP
Division; R. N. Mason, Engineering and
Maintenance Division). A program for
investigating high-temperature valve-
stem packing materials is under way.
Considerable operating experience has
been accumulated with valves packed
with Incenel and graphite powder and
lubricated with tricresyl phosphate-
molybdenum disulfide mixture. Such
valves have proved quite successful
in laboratory operation; however, some
stem seizure has been encountered that
required heating of the valve bonnet
to free the stem, and in some instances
leakage of fluoride around the stem
has occurred. At the present time
reliability of a valve packed in this
manner is not completely assured for
long~time maintenance-iree operation
with fluorides at 1200 to 1500°F.

It is possible that powdered packing
materials that are not wetted by the
molten fluorides may give better stem
sealing and better valve operation. A
test apparatus has been constructed to
investigate the wettability of variocus
powdered materials by the fluorides,
and tests are presently being con-
ducted.

Experience to date with high~temper-
ature packing materials indicates that
in almost every instance stuffing
boxes have required retightening
after reaching operating temperature.
For example, with powdered-graphite
packing a tightening motion of 0.030
in. per inch of packing length 1is
required Lo re-establish the original
degree of compression after the packing
temperature has been elevated to
1100°F. A packing-expansion testing

2

19
ANP PROJECT QUARTERLY PROGRESS REPORT

rig has been constructed and the
expansion rates of various packing
materials will be investigated in the
search for a powdered material that
will show minimum dimensional change
upon being heated.

Tricresyl phosphate has proved
valuable as an antigalling lubricant.
The application of this fluid teo
threads during assembly has permitted
greater ease of disassembly after
valves have operated for many hours

at 1500°F,

Self-%¥elding of Seat Materials
(D. R. Ward, ANP Division; R. N.
Mason, Engineering and Maintenance
Division). Preliminary tests were
conducted to determine whether self-
welding 1s encountered in the presence
of high-temperature fluorides with
the following combinations of materials:
type 315 vs. type 316 stainless steel,
type 316 stainless steel vs. Inconel,
and Inconel vs. Inconel. Test specimens
were clamped tightly together and
submerged i1n fluorides for 187 hr at
1300°F. Although visual examination
showed no evidence of self-welding in
any of the specimens, metallographic
examinatlon showed some evidence of
interdiffusion of the specimens of
similar materials. Investigation of
self-welding is being continued.

Cauned-Rotor-Drivea Valve (W, B.
McDonald, A. L. Southern, ANP Division).
Preliminary tests of the driving
elements of the canned-rotor-driven
valve(?®? indicate that an operating
mechanism can be built that will
operate such a valve satisfactorily.
A design of a canned-rotor-driven
valve has been completed, but the
priority given to the bellows valve
hasdelayed construction. The favorable
performance of the three-ply Inconel
bellows indicates that a canned-rotor
type of drive may not be required.

 

(3)!7. B. McbBonald and A. L. Southera, op. cit.,
ORNL- 1294, p. 20.

20

HEAT EXCHANGERS

Test of Core Element of Sodium-to-
Air Radiator (G, D. Whitman, A. P.
Fraas, M. E. LaVerne, ANP Division}.
The second performance and endurance
test of the core element of the
sodium-to-alr radiator was initiated
on June 6, 1952, and 200 hr of operation
was logged before failure. The average
sodium inlet temperature during the
test was 1200°F, and a maximum temper-
ature of 1600°F was held for 74 hr
before failure.

As 1n the first radiator core
test, failure occurred in a tube
between the top fin and header on the
sodium inlet si1de of the radiator.
Metallographic examination of both
core elements indicated failure at a
brazed tube-to-header joint because
of local porosity in the joint. In
both cases the failure was accelerated
by the attack of the sodium on the
outer surface of the tube.

When the second test had been 1in
progress (2 hr a building power-supply
failure occurred and the sodium froze
1in the radiator before 1t could be
drained. To resume the test, sufficient
preheat was obtained through normal
methods to restart sodium flow without
difficulty. FElectric strip heaters
were used on the outside of the air
duct.

The second radiator core element
contained Nicrobrazed Joints and was
identical in construction to the first
except that the fin spacing was
altered to give 15 fins per inch
instead of 10.5 fins per inch as on
the first core element.

Some fabrication difficulties have
been encountered in producing these
radiators, Since the furnace at Y-12
is too small to accommodate brazing
of air radiators for the turbojet
project, an attempt was made to
FOR PERIOD ENDING SEPTEMBER 10, 1952

es tablxsh
Jarger furnace located im the rolling
mill at X-10. The second core 1s an
excellent example of successful
brazingof an intricate design; however,
to complete the radiator as planned,
the header caps had to be heliarc
welded in place. From the results of
several uwnsuccessful attempts to do
this it may be concluded that the
heliarc method does not produce a
in the presence

the techmigue i1in a new and

sound welded joint
of Nicrobraz metal.

Heat transfer data from the first
and second cores (10.5 and 15 {fins per
inch, respectively) were plotted as
Nusselt number vs. Reynolds number.
When the physical properties of air
at the mean temperature of the free
calculate these

stregm were used tao

parameters, the points were segregated
on a temperature basis with a large
spread between the data taken with
sodium temperatures of about 500°F and
those taken with sodium temperatures
of 1500°F. Since NACA RM-No. EBLO3
recommends basing physical properties
of the air on the boundary layer
rather than on the mean free-streanm
temperature, the data were replotted.
(The mean sodium temperature in the
radiator was taken as being ecquivalent
to the ailtr temperature in the boundary
layer.) As v Fig. 5, this
gave excelleat correlation of the
data, it should be noted that 1in
calculating these data full allowances
were made for variations in fin
efficiency resulting from variations
in both heat transfer coefficient and
thermal conductivity of the fins.

shown

The air-pressure-drop data were
correlated on the basis of the mean
temperature of the free air |
As shewn in Fig. 5, this gave good
correlation of the data for each of
the two [1n spacings tested.

stLream.

A third radiator with interrupted
ting has been assembled and brazed

and 1s now being tested. The fins
have been cut through and slightly
offset every 2 in. {about 50 hydraulic
radii) in the direction of the air
flow. Preliminary data show that
interruption of the fins vields an
increase in heat transfer coefficient
of approximately 30% depending on the
Reynolds number. The concomitant
increase in pressure drop is only 10%.
Not only do the interrupted fins give
an increased heat transfer coefficient,
but 1t appears that they make the
heat transfer coefficient almost
independent of fin temperature at a
given air-weight flow rate instead of
falling off rapidly with an increase
in fin temperature, as was found to
be the case for the plain fins.

The use of pure wnickel in place of
stainless a fin material
should produce asubstantial 1mprovement
1n fin efficiency because i1ts thermal
conductivity is nearly three times as
high. A ceramic coating 1is desirable,
however, to protect the nickel {from
oxidationif 1t 1s used at temperatures

steel as

above about 1200°F, The ceramics
group have developed a suitable coating
(mainly chrome oxide) and are able to

get pood adherence 1f the nickel 1=
first bright-annealed ina wet-hydrogen
atmosphere. A test on a radiator of
this type 1s plarnned as
priorities permit.

soon as

Bifluid Heat Tramnsfer Loop (D. ¥,
Salmon, ANP Division). Fabricatien of
components for a lnop te measure heat
transfer from flucride fuel mixtures
to NaK has been completed, and the
components awalting assembly

(Fig. 6).

are

The tube side of the double-pipe
heat exchanger has been calibrated
with water. The friction factor was
determined over a range of Reynolds
numbers 3,000 to 90,000, @ An
average value of 0.032 was obtained.
The calibration of pressure drop vs.

from

21
ANP PROJECT QUARTERLY PROGRESS REPORT

S I
FIN SPACING

RS

. Na INLET [__FIN SP i
TEMPERATURE (°F) | 0.5 per in. | 15 per in

700 |

i

|

|

  
 

 

&
900

1oo ;
13C0 ‘

  
 
  

1500
1600

& O * =

  

 

 

h0
s
10 [—
|
5,
|
|
27 —_
I
1ol
0%
R,
Fig. 5.

0.010 in.

to center line square spacing.

thick;

stainless steel fins,

flow will be used as a secondary
method of measuring tube-side flow
rate. Venturis for flow measurement
1n both primary and secondary systems
have been calibrated with water.

A calculated prediction of the
heat transfer performance of the
system has been made to determine
operating characteristics. The original
calculations for the loop were based
on the fluoride mixture NalF-KF-TLaF
(11.5-42,0-46.5 mole %}). The fluoride
fuel NaK-Zr¥F,-UF, (46-50-4 mole %),

22

       
   

Performance of Core Element of Sodium-to-Air Radiator.

DWG 163540

 

 

 

|
— L
|
4 — + - 5
! | | o
. i
i
: T
! P
i |
L
S S S
|
|
|

b

 

|

w
5

Type 304

tubes on 2/3-in. center line

3/16-1in.

which has approximately 100% greater
density and viscosity, will now be
used for tests aund, consequently, the
range of operation will be reduced.
The pressure drop with the NaK-Zr¢F,-
UF, fuel will be greater for the same
flow rate than with the NaF-KF-LiF
mixture, and as a result the operating
pressures for the system will be
higher. The maximum flow rate probably
will be reduced so that the maximum
Reynolds number will be 1in the order
of 30,000 instead of 80,000 as origi-
nally reported.
FOR PERIOD ENDIING SEPTEMBER 10,

SURGE TANK

~=--— FLUORIDE SYSTEM
e _IQUID METAL SYSTEM

 

 

   

1952

UNCLASSIFIED
WG 1834

 

ELECTROMAGNETIC FLOWMETER

 

 

 

 

e COOLER
Lt
HEAT EXCHANGER ames \ )“/
FILTER~ ] |
-~ CENTRIFUGAL PUMP % Mt —
e e e A i I
i
(f”'_“ L e e e~ U
o N e e l
HEATER SECTION }Z l 1
( wmwaL{:}~h_L ~u4 AIR
o VENTURI SLOWER
e - FLOWMETER

  
 

VENTURI FLOWMETER

SUMP };]
~
Fig. 6.

NaK-to-NaK Heat Exchanger (M. E.
Il.aVerne, A. P. Fraas, ANP Division;
E. E. Hoffman, Metallurgy Division).
Metallographic examination of the

ction of the NaK-to-NaK heat ex-
changer described i1n the previous
quarterly reports®*’ has been completed.
Two tube failures were discovered in
the Nicrobrazed tube-to-header joint
that were not detectable from sepa-
ration. It was concluded that these
failures were caused by the vibration
in the system and the etching effect
the circulating NaK had on the tubes
in the hot zones. These failures were
encouraged by the embritctlement of
the areas near the joint by the brazing
alloy and the large grain size of the
16-mil-wall, type 304 stainless steel
tubes. No fractures were detected in
the 19.5-mil-wall, type 347 stainless
steel, Nicrobrazed, tube-to-header
loints.

Upeon sectioning the heat exchanger,
1t was found that the hot (inlet) end

was very bright, whereas the cold

 

(434 p, Fraas, Aircraft Nuclear Propulsion
Project Quarterly Progress Report for Period
Ending March 10, 1952, ORANL-1227, p. 32; M. E.
LaVerne and A. P. Fraas, op. cit., OBNL 1294,
p. 22, .

EL&CTROMAGNE? IC X

g:@dmp

Schematic Diagram of Bifluid Loop.

{outlet) end was covered with a dark,
powdery film. Spectrographic ex-
amination of the dark and bright
surfaces revealed the only detectable
difference to be a manganese content
present in the dark (cold) surface
that was two to three times the ex-
pected value for type 347 stainless
steel. The x-ray powder pattern of
the dark powder scraped from the
surface of a tube 1n the cold zone
showed Ni0O, Cr, and type 347 scainless
steel base material.

INSTRUMENTATION

RBotameter Type of Flowmeter (P. G.
Smith, A. L. Southern, ANP Division).
A rotameter type of flow-measuring
device has been designed for measuring
the flow of a fluoride stream at
1200 teo 1500°F. This instrument
consists of a conventional rotameter
float operating in a tapered barrel,
The float position is indicated by a
tapered iron core, chrome plated to
resist corrosion, which operates in a
cylinder of static fluorides extending
downward from the flowmeter (Fig. 7).
An inductance coil that is capable of

23
ANP PROJECT QUARTERLY PROGRESS REPORT

rfl~/13fig|nREF ’~—m~l¥fflw-
| |

134 in DRILL THROUGH '1|
n.

ONE WA LL~W~~\:S:fi

3/8 n.

      

 

|
[
X 3/, -in-1PS SCH. 40 PIPE 29 ¥4~ in-LONG -
|
i
l

—~0.250-in-D % 1¥%-in-LONG PIN

UNCLASSIFIED
CWG. DSK- 157314

Iy i, e

LAVA COil. FCRM
( FIRED AFTER MACHINING)

 

 

 

 

 

5V2ll’|,’§
n

 

 
  
 
 
   

 

;
_L-, TR RN
7
/
/
/

L34 inTHICK X 3% in-D ROTAMETER FLOAT

INCONEL PLATE TAPERED BARREL

Fig. 7.

being operated at a constant tempera-
ture of 1300°F is placed arocund this
cylinderin the vicinity of the tapered

core, and as the core moves 1n the
field of the inductance co1l, the
movement 1s sensed by suitable 1n-

strumentation. The core movement may
be calibrated to very accurately
indicate the flow rate of the fluid
through the rotameter barrel. The
initial test of this instrument 1in a
high-temperature dynamic fluid system
was somewhat disappointing. Thermal
distortion of the cylinder in which
the tapered core 1s located resulted
in binding of the moving parts and
preventing further operation of the
instrument. A new design, which
should eliminate this difficulty, has
been completed that allows greater
clearance between the tapered core

and the containing cylinder and
requires stress relieving of the
parts before assembly. The new

instrument 1is being constructed and
will soon be 1incorporated in a test
loop.

BRotating-Vane Type of Flowmeter
(W. B. McDonald, J. M. Trummel, ANP
Division; F. A. Anderson, Research
A rotating-vane type
manufactured by the

Participant).
of flowmeter

24

 

Yg=in-THICK X 3 ¥,-in-D PLATE L
3-in-IPS SCH. 40 PIPE 2 Va-in-LONG

3 -in-D x 12-in-LONG INCONEL BAR

      

Yy -in-THICK % 1 3%4-in-D
INCONEL PLATE —--
“TAPERED CORE (#g~in-Dx 15-in-
‘\ LONG CHROME~ PLATED MILD
\\ STEEL BAR)

INDUCTANCE COIL

Rotameter Type of Flowmeter with Variable Induciance Core.

Potter Instrument Company and modified
for expected high-temperature appli-
cation has been calibrated on water
and 1s awaiting test i1n a high-tempera-
ture fluid system.

This instrument consists of a
permanent magnet rotated in the fluad
stream by a small turbine. A pick-up
colil 1s placed outside the flowmeter
in close proximity to the rotating
This co1l, when connected to
a Hewlett-Packardelectrical tachometer
through a preamplifier furnishing
60 db of amplification, will very
accurately sense the rate of rotation
of the permanent magnet. At room
temperature this arrangement was
calibrated in terms of accurate flow
measurement with a pick-up co1l located
as much as 3 i1n. from the magnet.
It 1s expected that this instrument
can be operated at high temperatures
with the pick-up coil located as close
as 1/2 in. to the magnet. The limi-
tation of this flowmeter is expected
to be the curie point of the permanent
magnet, although some operational
difficulty may be encountered with
the Carboloy bearings on which the
magnet and the rotating vanes
Operation of this instrument
wi1ll be examined at fluid temperatures

to between 1250 and 1300°F.

magnet,

are

mounted.
FOR

Moore Nullmatic Pressure~Measuring
Device (P. W. Taylor, ANP Division).
The design of the Moore pressure
transmitter, reported previously, (%)
has been modified to prevent the high-
temperature fluid from contacting the

pressure-sensing bellows at high
pressures. The modified instrument
is rated at 0 to 60 psi at 1200°F

maximum temperature. The temperature
of the fluid stream at the point of
pressure determination may be as high
as 1500°F. This instrument incorporates
the trapped-gas principle used in the
previous design and must be operated
in a vertical position, Further
designs are under way that will in-
corporate three-ply Inconel bellows.
The trapped-gas space will be eliminated
and the bellows pressure-sensing
element will be completely submerged
in the high-temperature fluid. Such
an instrument may be located 1n any
pesition and at any point 1in the
system. '

Diaphragm Pressure-Measuring Device
(p. W. Taylor, ANP Division). The
diaphragm pressure-measuring device,
Fig. 8, utilizes a linear differential

 

(5)p. W. Taylor, op. cit., OBNL-1294, p. 26.

PRESSUREROD*\\

  

  
 

I's
Y

Wt

 

Yai
- Yain D
.

2% in.

 

 
 
     
  
 
    
  
 

‘/2 in.

 

\‘“DIFFERENTIAL
TRANSFORMER

SPRING
BLOCK

 

   
 
 

I-*3-'hint-J T

PERIOD ENDING SEPTEMBER 10, 1952

transformer to measure the deflection
of a 1/16-in.~-thick Inconel diaphragm.
The deflection will be calibrated 1in
terms of the applied pressure. This
instrument, which is being procured
from an outside vendor, incorporates
the trapped-gas~-space principle;
however, test results indicate that
the high-temperature fluid may be 1in
contact with the diaphragm without
resulting instrument failure. The
instrument is rated at 0 to 60 psi at
a maximum diaphragm deflection of
0.010 inch. The operating temperature
is limited to 1000°F at the diaphragm
to minimize the creep rate of the
Inconel; however, by using a riser
leg to the pot and diaphragm and by
cooling either the trapped gas or the
fluid te this temperature, the instru-
ment should accurately measure the
pressure of a fluid stream at 1500°F.

Diaphragm Pressure Transmitters
(P. W. Taylor, ANP Division). Six
pneumatic null-balance pressure
transmitters are being fabricated.
These pressure transmitters will be
constructed of Inconel and will have
0.014-1in. Inconel X diaphragms. The
instrument is rated at 0 to 100 psi
at 1000°F; however, since the pressures

UNCLASSIFIED
DWG. DSK~15608A

 
  
   

r\" DIAPHRAGM HOUSING
N,

\

  
 
    
    
       

  

Yy~in. ORIFICE

%mfi

    
   

    
 

Mein‘
3"15""

 

¥gin. D

7-’fain.

—2in.

—+— FLUCRIDE STREAM

 

134hgin.

 
  

  

WELD NCONEL

OIAPHRAGHM

Fig. 8. Diaphragm Pressure-~Measuring Device.

25
ANP PROJECT QUARTERLY PROGRESS REPORT

on opposing sides of the diaphragm
are balanced, 1t may be run at higher
temperatures without
sacrifice of accuracy.
mitters will be tested completely
filled with high-temperature fluid.

too great a
These trans-

Moving-Probe Level Indicator (P. W.
Taylor, A. L. Southern, ANP Division}.
A level indicator has been designed
that incorporates a conventional
spark-plug probe moving through a

high-temperature packed seal. For
the first test the probe level 1is
controlled manually by means of a

lead screw mechanism., If the packed
proves to be an effective gas
seal and 1f the first test 1is en-
couraging, an automatic control for
the probe will be aincorporated into

the design.

seal

Null-Balamce Level Control (P. W.
Taylor, A. L. Southern, ANP Division).
Fluid flow in high-temperature systems
is presently measured by means of a
venturi with trapped-gas pressure pots
to sense the pressures of the venturi
entrance and throat. Unequal levels
in the pressure pots result in con-
siderable flow measurement

error,
particularly with fluids of high
density. A null-balance level con-

troller (Fig. 9) has been developed
that equalizes the level 1n the two
pressure pots and greatly 1ncreases
the accuracy of flow measurement.
This level controller consists of a
tapered core operating in the field
of an induction co1l. The core 1is
attached to a float that rides on the
free of the fluid 1n the
pressure pots. This instrument will
maintain the levels 1n the pots within
0.1 in. of each other.

surface

In operation, when flow 1s started
in the system, fluid rises 1in the
venturl entrance and throat pressure

pots, with the fluid rising to a
higher level in the high pressure
pot. The level device causes the

26

voltage output of circuit No. 1 to be
greater than that of circuit No. 2.
This voltage differential appears
partially across variable resistor
(A) and 1s fed to a Brown controller.
The signal in the controller operates
a limit switch to open a solenoid
valve, which bleeds gas into the high
pressure pot and forces the level
down. When the levels equalize, the
voltages in the two circuits buck each
other and the signal from the con-
troller drops to zero. The solenoid
valve then closes and shuts off the
gas supply. This sequence 1s repeated
at any time the levels in the two
pots become unequal.

Strain-Gage Level Imdicator (A. L.
Southern, ANP Division). A level
indicator 1s being designed that
incorporates a three-ply Inconel
bellows to transmit pressures exerted
by rising levels to a high-temperature
strain gage. This device will be
limited to use in systems pressurized
within the limits of the bellows. A
questionable feature, which must be
tested, is the long-time reliability
of high-temperature strain gages,

The advantage of such an i1nstrument
is that the bellows sensing element 1is
completely submerged in the high-
temperature fluid and 1s not subject
to failure or loss of calibration as
a result of solid deposits sublimed
from the free surface of fuels having
a high vapor pressure, such as the
fuels containing zirconium fluoride.

FLUID DYNAMICS

ARE Core Mockup (L. A. Mann, ANP
Division). A full-scale mockup of the
fuel passages of the ARE core, as
reported and illustrated previously, (%)
was completed and tested for flow

(6)1.. A. Mann and D. F.
ORNL-1294, p. 28, Fig. 10.

Salmon, op. cit.,
FOR PERIOD ENDING SEPTEMBER 10, 1952

UNGLASSIFIED
OWG 16342

 

 

 

 

 

 

  

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

250 n IRON CORE 250 N
o
TVVVA A ~ AAAN~
D
_ b
H
D
\ P

FLOAT
HIGH LOW

CIRCUIT PRESSURE PRESSURE CIRCUIT

NO. 1 NO. 2

 

 

 

 

 

 

. SOLENOID
- GAS CONTROLS

 

 

 

BROWN
CONTROLLER

 

 

Fig. 9. Null-Balance Level cController.

27
ANP PROJECT QUARTERLY PROGRESS REPORT

characteristics, ease of filling, and
ease of emptying. The tubes were
assembled from l1-in. glass tubing and
bends with flanged joints, and the
headers were constructed of copper
sheet and tubing. A centrifugal
pump, a rotameter, a vacuum pump, and
a 50-gal reservoir were the major
pieces of auxiliary equipment. Both
water and a zinc chloride solution
were circulated 1in this mockup. The
zinc chloride was used to simulate
the fluorides because of 1ts density
(1.7 g/cc) and viscosity (10 cp) at
the operating temperature {(90°F) of
the mockup. The operating charac-
teristics of the mockup on both water
and zinc chloride are given in Table 1.

Table 1

PRESSURE AND FLOW IN THE ARK
CORE MOCKUP

 

 

 

TAP WATER | ZnCl, SOLUTION
Pressure at pump
(psi) 20 38
Pressure at core
entrance (psi) 10 18
Rotameter reading
(gpm) 43 >50

 

 

 

Evacuation of the core prior to
filling was a definite aid, as 1in-
dicated by the percentage of tests 1in
which all six parallel circuits were
filled completely with zinc chloride
solution at three core pressure

levels (Table 2).

When the core was first evacuated
to 28 1n. Hg and the bottom header
then opened to the liquid (ZnCl,
solution) supply with the vacuum
pump continuing to operate, all tubes
filled almost completely. When the
vacuum connection was shut off, the

28

Table 2

PERCENTAGE OF TESTS IN WHICH
CORE FILLED COMPLETELY

 

 

 

 

 

TESTS FILLED COMPLETELY
CORE PRESSURE
(%)
Atmospheric 80
26 in. Hg vacuum 95
28 in, Hg vacuum 100
top header opened, and the pump

started, the remaining gas was swept
out and the core filled completely
with liquid. In all tests with zinc
chloride solution, many small bubbles
remained in suspension in the liquid
throughout all test runs, with some
runs being as long as 20 min; however,
this egquipment did not utilize a surge
and degassing tank as incorporated in

the ARE design.

All attempts to completely empty
the core by blowing out the fluid with
air failed. When 20-psig air was
applied to the top header an estimated
15% of the liquid remained in the
tubes, unevenly divided hetween
circuits, The largest rematning
volumes were approximately 18 1in.
(of tube length) in some tubes. in
each test one or two tubes emptied
almost completely, and it appears
that no reasonable gas pressure will
remove all the fluid from all circuits,

Tests of evenness of flow in the
six parallel circuits, with both
water and zinc chloride solution,
were made by suddenly adding ink to
the fluid reservoir and observing the
relative flow rate through each of
the six circuilts. With water, no
difference in flow rate was observed,
With zinc chloride solution, one
circuit lagged 2 to 3 sec behind the

others. It 1s planned to install
FOR PERIOD ENDING SEPTEMBER 10,

greater pump and rotameter capacity
- for making further tests at 80- to
100~gpm flow rates. Surge tanks and
a water mockup of the ARE fluid
circuit are being constructed that
will be combined with the core mockup
for hydrodynamic tests of the combined
systems.,

Possible Fuel-Header Manifold
for the Reactor Core (W. C. Tunnell,
ANP Division)., A possible reactor
core for ANP use would comprise upper
and lower fuel headers joined by fuel
tubes in parallel. Such an arrangement
has the advantage of complete drain-
ability and allows some increased
flexibility 1n the external fuel
circuit,

The major problem of design 1is
reducing the header volume to avoid
large, relatively stagnant volumes of
fuel 1n regions adjacent to the high-
neutron flux of the core or in regions
subjected to delayed-neutron effects,
and at the same time providing for
optimum flow conditions im all the
parallel fuel tubes. Cptimum flow in
a right-cylinder reactor core will
provide uniform flow in all tubes 1f
the reactor flux is uniform throughout
the core volume or the flow conditions
vary inversely with the flux.

For purposes of investigation, a
half-scale plastic model of the bottom
half of the ARE core was built in
which all of the tubes were headered.
This was tested with water at flow
rates up to 125 gpm.

The header design started as two
plates. One of the plates served as
the tube header sheet and had grooves
milled in 1t to provide communication
to all tubes, and the other plate
provided the tube sheet closure and
had a circular milled groove in open
communication with the network of
grooves 1in the tube sheet. No location

1952

of inlet with respect to the tube-
sheet closure passage was f{ound that
resulted in optimum flow.

After a series of tests it was
found that the insertion between the
original tube and closure sheets of a
third plate with appropriately po-
sitioned small holes, like a salt
shaker, and a thin annular plenum
chamber in the closure sheet over all
the holes in the middle added plate
resulted in optimum flow conditions.
It is also apparent from the tests
that variations 1in the flow pattern
can be obtained by variations in the
arrangement of holes in the middle
plate. 3

As a point of interest, the volume
of fuel 1n an ARE-size header and
supply pipes would be approximately
1942 in.?, as compared to 2780 in.?3
of fuel within the tubes.

TECHNOLOGY OF FLUORIDE HANDLING .

Fluoride Production (G. Nessle,
Materials Chemistry Division). The
design of equipment capable of producing
3 kg of treated fuel was completed
July 1.

Briefly, the treatment consists of
bubbling hydrogen through molten fuel
for 1 hr followed by hydrogen fluoride
for 2 hr at 1550°F. The initial
heating and melting of the fuel mix
is done under a hydreogen fluoride
atmosphere. The equipment has operated
with only one failure, which was due
to faulty welding, since being placed
in operation on July 14.

Following the success of the 3-kg-
batch equipment, the design was scaled
up for construction of equipment to
produce 25 to 30 kg per batch (50 to
75 1b). Unlike the smaller apparatus,
the 25-kg equipment requiresa receiver

29
ANP PROJECT QUARTERLY PROGRESS REPORT

furnace that can be lowered away from
the receiver to facilitate removal of
the filled receiver without disturbing
the remainder of the apparatus. Since
some zirconium tetrafluoride (around
50 g in 5 hr) 1s evolved during the
treatment of the fluorides at 1500°F,
a trap containing copper servo has
been introduced to prevent clogging of
the hydrogen fluoride and helium purge
lines and damage to the valves 1in
those lines. The trap signal, which
is sized according to the size of the
production equipment,has been quite
satisfactory and requires only minor
attention.

Calrod heaters were used to heat
the 3/8-1in.-0D nickel tubing transfer
line, The treated fuel is transferred
from the treatment reactor to the
receiver by 1inert gas pressure.
Construction of this apparatus began
July 16 and the first run was suc-
cessfully made on July 25.

At the end of August the total
production was 128.3 kg of treated
fuel. Table 3 gives a break-down of
the production in the two units,
Although the equipment, which 1s
capable of producing 60 kg per week
1f needed, 1s not operating at maximum
capacity, production 1s well ahead of
the demand.

Equipment 1s now being designed to
produce the mixture NaF-ZrF,-UF,
(46-50-4 mole %) in the quantities
required to meet ARE fuel requirements.
The production of fluoride fuel is
dependent upon the production of
sublimed ZrF,. Arrangements have been
made to obtain sufficient hafnium-
bearing ZrF, to permit
production of fluoride fuel for
experimental needs. It i1s now possible
to produce 300 1lb of sublimed ZrF,
per week starting from the base
material ZrCl,.

adequate

30

Table 3

FLUORIDE PRODUCTION

 

 

 

 

SMALL APPARATUS LARGE APPARATUS
(3 kg) (25 kg)
1.869 21. 054
1.970 21.509
1.837 20,300
2,003 26.900
1.984 27.138
1.770

11.433 kg 116.901 ke

 

 

Every batch of ZrF, fluoride
produced 1s now being crushed and
sampled prior to use in the fuel
mixture. The Zr¥F, thus far obtained
has been of very good quality, but
it 1s grayish in color because of
entrained carbon. After September,
hafnium-free ZrF, will be available
for the production of ARE fuel.

Removal of ¥luorides from Comn-
taminated Systems (L. A. Mann, ANP
Division). High-pressure (140-psig)
steam jets are used successfully far
removing fluorides from contaminated
systems. Heavy fluoride deposits are
removed completely without damage to
the equipment. It has been found
desirable to select a steam-jet size
compatible with the size of the
equipment to be cleaned so that high-
velocity steam will always impinge
directly on the deposits to be re-
moved, and there will be sufficient
clearance for the spent steam to
escape from the container. 1In several
instances this method has reduced time
of cleaning from 48 to 4 or 5 hours.
In other cases parts have been cleaned
FOR PERIOD ENDING SEPTEMBER 10,

by this method that could not be
economically cleaned by methods used
previously. When systems or components
are contaminated with a thin fluoride
coating, steam jets furnish no par-
ticular advantage over washing with
water and scrubbing with a wire brush.

Purity of Pipe Line Helium (L. A,
Mann, ANP Division). Installation of
the helium line direct to the tank car
located outside the building is
complete. After the lines were purged
the oxygen content of the helium was
reduced to a level not detectable by
analyzing equipment sensitive to
1 ppm. Helium of this high purity 1is
now distributed to all major points
of usage in the experimental engineering
areas of Building 9201-3, and the
requirements for bottled helium and
argon have been greatly reduced. '

Fluoride Plug-Removal Test from
Simulated ARE Core (L. A. Mann, ANP
Division). Some operational experience
indicated that it would be very
difficult to remelt a plug of frozen
fluoride in a pipe line without
rupturing the pipe. Since the ARE
core as presently designed cannot be
completely drained, tests have been
coniducted to determine whether the
residual fluorides can be remelted
without damaging the core structure,
A type 316 stainless steel tube bent
to simulate an ARE core U-~tube was
filled with NaF-ZrF,-UF, fuel, which
was then permitted to freeze. The
assembly was placed in a high-tempera-
ture pot furnace and cycled three
times at a heating rate of 1R°F/hr
between 850 and 950°F. No failure
occurred. More fuel was then added
te the tube and insulation was placed
arcound the upper legs of the tube to
simulate the beryllium oxide in the
ARE core. The assembly was again
taken through eight temperature
cycles, and again no failure resulted.

1952

Additional cycles over the same temper-
ature range were then made at in-
creased heating rates, with the final
cycle being made at full furnace
power. Again no failure resulted.
Final examination of the tube showed
that it had expanded 0.030 in. at one
point and. 0.050 in. at another point.
These tests did not entirely duplicate
ARE conditions but were probably
just as severe.

Injection of NaK into a Flowing
Fluoride Stream (L. A. Mann, ANP
Division; F. F. Blankenship, Materials
Chemistry Division; T. N, McVay,
Consultant, Metallurgy Division).
NaK is one of the coolaunts being
considered for the ARE reflector
and moderator. Since the fuel tubes
will be immersed in the coolant, i1t 1is
necessary to determine the extent of
the NaK-fluoride reaction that would
result from a fuel-tube rupcure. A
test apparatus to determine the extent
of this effect consisted of a small,
forced-circulation locp with an
Fastern centrifugal pump to circulate
fluoride fuel, Nab-ZrF, -UF, (46-50-4
mele %), at 1500°F and a flow rate of
approximately 3 ft/sec. The effect
of both slow and rapid injection of
NaK into the fluid stream will be
measured. A NaK injection unit was
attached to this loop and when the
temperature of both the fluorides and
the NaK reached 1500°F, & quantity of
NaK (approximately 5% of the volume
of the fuel in the system) was injected
into the stream. After the injection
the pump was permitted to operate for
approximately 6 min, at which time all
heat was turned off, the pump was
stopped, and the entire assembly was
rotated through 90 deg to a horizontal
position and permitted Lo freeze.

According to the ‘exothermal re-

actions Na + UF, ——® NaF + UF; and
Na + ZrF,———®= NaZrF,, heat evolution

31
ANP PROJECT QUARTERLY PROGRESS REPORT

was expected and did occur. This
was indicated by an immediate tempera-
ture rise of 200°F that was measured
by a thermocouple immediately above
the NaK injection point.

Fxamination of the system showed
incomplete mixing of the NaK and the
fuel and indicated that the system was
almost completely plugged by the
reaction products., Possibly flow had
stopped completely. This test in-
dicates that
large fissure type of break) occurred
in an ARE core fuel tube during actual
operation and permitted a similar NakK
injection, the flow 1n the tube would
be completely stopped or considerably
slowed down. The fuel residence time
in the tube would thus be increased to
the point where very high temperatures
would result and perhaps bring about
complete rupture of the tube. The
possibility that the plug would heal
the leak and prevent the flow of fuel
into the interstices of the moderator
1s still a matter of speculation.

The loop was sectioned into seven
large pieces and most of the pieces
were subsequently subdivided as in-
dicated in Fig. 10. Also, the general
appearance and the complexity of the
products observed are suggested in
Fig. 10.

The extent and distribution of the
altered fuel suggests that complete
or nearly complete plugging of the
loop occurred very soon after the NaK
was injected. Material in the surge
tank had been attacked omly slightly
and even less reaction had occurred
on the side of the loop opposite the
injection point. The plugging and
consequent lack of circulation were,
of course, responsible for the system
being very far fromchemical equilibrium
and for the large differences observed
between this experiment and the static
systems described in the section on

32

1if such a failure (a -

“Chemistry of High-Temperature Liquids”
(sec. 9).

Samples of material from each
section of this loophave been examined
by x-ray diffraction, chemical analysis,
and the petrographic microscape.
Since the material found varies con-
siderably from the center to the edge
of each section, as well as along the
axis of the pipe, complete characteri-
zation of all the material has proved
virtually impeossible.

Such phases as the crystal solution
of NaUF, 1in NaZrF,, Na,ZrF., UF,, and
numerous complex compounds of KF
occur 1n abundance. In addition,
metallic zirconium occurs 1in sections
3B and 4 just above the NaK injection
point. At least two additional phases,
which are unknown, are recognized 1in
considerable quantity in this loop.
One 1s the dark-brown phase that is
pleochroic and has an average index of
refraction at about 1.556. The other
1s an orange-red phase generally
associated with, and perhaps formed
from, the brown phase. This mineral
has an average 1ndex of refraction of
about 1.588, with low birefringence,
and i1s probably monclinie. It 1s
probably a compound of UF,, since
some crystals are available that seem
to indicate transitions from this red
phase to pure UF,.

A second test will be conducted
during which NaK wi1ll be injected
into the fuel stream at a much slower
rate to determine the effects of a
very slow leak. It 1is thought that
fuel circulating at highvelocity will
perhaps sweep the reaction products
from a very slow leak out of the
reactor core.

Gas-Line-Plugging Tests (W, B.
McDonald, P. W, Taylor, ANP Division).
The vapor pressure of the fluoride
fuel NaF-ZrF,-UF, (46-50-4 mole %) at
FOR PERIOD ENDING SEPTEMBER 10, 1952

E.40U, .74 colg Hy , 37.6Zr, 41.9 Zr 349K

LEMON-YELLOW PHASES SIMILAR TO
NaUFg - NaZrFg SOLUTION, AND TANK
NapZrfg . TRACE UFy, A.

 

DWG. 16343

‘SURGE

 

 

 

68

6 A

 

    
  
 

SEE &8

 

A oot e

 

 

 

 

- _ CRYSTALS ON WALL: SOME
T _ 5 NaUFg -~ NaZrfg , Nap ZrFg OR

_ 17 K AMALOGS. LARGE CRYSTALS
zZ MUGH A 5 OF UF3. 19.3U, 9.65 cc/g Hy ,

7.8 U, 1.22 cc/g Ho = i | TRACE UFg = !

o 2 5 T 30.2 Zr, 37.4F, 228K

36.9Zr, 41.6F, 237 K— = | |5 :

& T }-NGZPF5(?) .
CONSIDERABLE B
BO2U, 140 cerg Hy, 2 tA ; Ns - ' wflj
369 7r, 41.4F, 22 K —— O 3 ol gy o
»alar, ez & ~1 TV—VERTWCAL VEIN OF A, PLUS CRYSTAL
0 : e e e - ~
SOLUTION OF NaZrFg~KZrFs SOME
L 1 ngSFB’ Neatrts Ky ZrFg OR NOEZPFES ”
2 1B 3 3c MUGH DARK BLUE-GRAY TO PURPLE,
- MOSTLY UFz, GRAINS SHOWING A-=UFj
b L 6410 245 onre 1 gae TRANSITION. SOME NaZrfs, SMALL
M . cc o - -
2 SOME A AND B8 i eI AMOUNT Zr METAL. VERY SMALL AMOUNT
~ lig] MN2irfe 2 2n, SREF, 32tk T YELLOW PHASE '
" NG UF 1)
% 35| — Nar
8.18U,1.28 ec/q H2 ) g e ) , .
381520,4L4F,255K'””/\ e EARLY AGRA 38a MAINLY NaZrFg, NO GREEN COLOR

 

 

 

     
 
 
   
 
  

N

APPEARING FUELL

e

8.30 U, .35 ccrg H,, 38.9 2
4B F, 054K e

B CHANGING: TO A
NagyZrfg OR KpZrfy

 

SOME A, UFy

BLEACHED NaZrfg -MNalifg
MUCH A, B AND A, NO UFg
~0.98K, 5.90 U, 1.60 cc/g vi,,40.2 Zr, 42 4F

  

A= RED-ORANGE UNKNOWN, AVG n = 1.588; PROBARBLY MONOGLINIC,

B= DARK BROWN UNKNOWN , PLEOCHRQOIGC, AVG n = 1.556 .
MAY CHANGE GRADUALLY TO A AS REDUCTION PROCEEDS
SINGE SOME GRYSTALS ARE BROWN ON ONE END AND RED ON THE OTHER

Fig. 10,
Fluoride Loop.

950°F, or slightly above, 1s 0.016 mm.
At 1250°F it is 1 mm and at 1500°F 1t
i3 13.26 mm. Tests were conducted to
determine the rate at which gas lines
are plugged by sublimed ZrF, rising
from the free surface of this fluoride
fuel at various temperatures. Gas
lines ranging from 1/8 to 1/4 in. in

Distribution of Reaction Products from Injection of MaK inteo

diameter were connected toc a pot
containing the fuel, which was heated
to 1500°F. High-purity helium was
introduced into the pot, swept across
the free surface, and exhausted threough
the gas lines at relatively slow flow
rates. All lines plugged solid in
less than 100 hr of operation.

L
(o8]
ANP PROJECT QUARTERLY PROGRESS REPORT

The fluoride temperature was
reduced to 1050°F and a second test
was started under conditions corre-
sponding to the first test, This
test has logged more than 1300 hr
with no evidence of line plugging.

Cooled-Baffle Vapor Trap for
Zirconium-Bearing Fluoride Fuel (W, B.
McDonald, J. M. Trummel, ANP Division;
F. A. Anderson, Hesearch Participant).
A cooled-baffle vapor trap was de-
veloped to condense and collect the
ZrV, vapor that evolves from the
Na¥F-ZrF,~UF, fuel. The vapor trap
consisted of a 3-in.-diameter pipe
24 in. long containing 28 baffles
spaced at 3/4-in. intervals, with
the openings in the baffles alternated
from one side of the pipe to the
other. A cooling coil through which
water was circulated was placed around
the complete length of the baffled
pipe. High-purity helium was passed
across the free surface of the 1500°F
fuel and exhausted through the Laffled
pipe. Filter paper was placed at the
exhaust port to collect any solid
particles not trapped by the baffles,
The flow rate of the helium across
the free surface was

15 cfh.

approximately

After slightly more than 50 hr of
operation the baffled pipe was so
completely plugged by the material
sublimed from the fuel that greatly
increased pressure was required to
maintain the specified gas flow rate.
This pressure 1ncrease resulted in the
bulging of the top and bottom of the
flat fuel container to such an extent
that the test was terminated. Ex-
amination showed that the baffle pipe
was completely plugged by a hard
crystalline deposit at the point of
entry into the fuel container. The
remainder of the pipe contained heavy
deposits of fine, powdery material.

34

This test indicates that a baffled
vapor trap 1s cntirely umnsatisfactory
for the preventionof gas-line plugging.
Other types of liquid vaper traps are
presently being tested. One of these
tests consists of passing the gas from
the free surface of zirconmium-bearing
fuel through a bath of Nak-KF-LiF
(11.5-42.0-46.5 mole %) at 1500°F,
and there 1s a similar test in which
from zirconium-bearing fuel
are passed through a bath of NaK at
1500°F, These tests are still in
operation and the results have not
yet been determined.

the vapors

Descaling and Pickling Tests (D), C.
Vreeland, E. E. Hoffman, R. B. Day,
L. D. Dyer, Metallurgy Division).
In connection with the problem of
removing welding scale from ARE
components some tests have been run
to check different methods of de-~
scaling and pickling oxidized Inconel.
Sodium and NaK were mentioned as being
two of the most promising descalers
(from a handling viewpoint) to be
tested. Oxidized Inconel specimens
were treated, 1n corrosion
test tubing, for various times and
temperatures with sodium and with NaK.
The results of these tests are best
summarized bv reference to Fig. 11.
It is quite apparent that in these
tests neither sodium nor NaK
efficient as an oxide scale remover
until the test temperature was railsed
to approximately 800°C. 1In order to
determine whether a dynamic NaK system
would be any more efficient as a scale
remover, an oxldized Inconel specimen
fastened i1nto the bottom of an
Inconel tube by crimping. The tube
was then half filled with NaK, sealed,
and run in the rocker furnace for 4 hr
at approximately 700°C. The oxide
scale was not removedby this treatment
and so, apparently, the temperature
of test is the deciding factor in
scale removal.

static

was

was
FOR PERIOD ENDING SEPTEMBER 10, 1952

UNCLASSIFIED
PHOTO Y-68993

  

INCONEL, AS OXIDIZED

             

    

  

     

     

 

Na
Lhr AT 200°C ~ {br AT 300°C fhr AT 500°C 5 hr AT 600°C
Ne
10 hr AT 700°C 3 hr AT 800°C
NgK
2 hr AT 300°C 2 hr AT 500°C 2 hr AT 700°C 2 hr AT 800°C
Fig. 11. Bescaling Properties of Na and NaK on Oxidized Inconel.
3. REACTOR PHYSICS
W, K. Ergen, ANP Division
The theoretical investigations of mechanical vibration. With this
the kinetics of the circulating-fuel simplified concept, a large temporary
reactor continued, still disregarding disturbance was assumed to be applied
for the present the possible coupling to the reactor, and it was found that

between “nuclear' oscillation and 1f the reactor suarvived the first

)
L
ANP PROJECT QUARTERLY PROGRESS REPORT

short maximums of power and temperature
the following power and temperature
maximums would be rather moderate.

The general properties of the
reflector-moderated circulating-fuel
reactor were discussed in the last
report. (') A number of multigroup IPM
calculations has since been performed.
However, this 1s the first time that
the method has been applied to reactors
with very different properties 1n core
and reflector and with absorptions that
are fairly large and also vary within
a lethargy interval. Hence, con-
firmation of the results by critical
experiment 1s necessary,

A calculation was performed regard-
ing the critical mass, power distri-
bution, and neutron spectrum of a
mockup of the ABRE for use in critical
experiments. So far only the critical
mass has been determined experimentally,
The agreement with the calculation was
very good (cf., sec. 4, "“"Critical
Experiments’’).

A previously known method of com-
puting the effects of gaps on reactivity
was refined and has yielded satis-
factory agreement with experiment.
In addition, several slowing-dcwn
kernels have been expressed
of kvnown functions,

in terms

OSCILLATIONS IN THE CIRCULATING-YUEL
AIRCRAFT REACTGR

S. Tamor, ANP Division
A summary report{?’ was issued re-
garding the work on the kinetic

equations(s) of a somewhat idealized

 

(1)Aircraft Nuclear Propulsion Project
Quarteriy Progress Report for Period Ending
June 106, 1952, OBRNL-129%4, p. 6 and 31.

S. Tamor, Note on the Non-Lineaer Kinetics

of Circulating-Fuel Reactors, Y-F10-109 (Aug. 15,
195%)3
3

W. K. Ergen, Aircrafit Nuelear Propulsion
Project Quarterly Progress Report for Period
Ending March 10, 1952, ORML-1227, p. 41,

36

In addition
to the results gquoted in the last
report, (%) it was found that rather
satisfactory limits can be set on the
second power and temperature maximum
following a disturbance of the resactor,

circulating-fuel reactor,

A large disturbance, such as intro-
duction of excess reactivity, will be
followed by large power and tempera-
ture maximums. However, these maximums
are of short duration and can, withan
limits, be tolerated, 1f it can be
ascertained that they will not be
followed by maximums of similar height.
Hence, it was regarded as important to
investigate the second power and
temperature maximums, The third and
following maximums are smaller than
the second because of the damping of
the reactor oscillations,

No general relationship has been
set up between the upper limit for the
second waximums and the other parameters
of the reactor. However, two specific
examples have been investigated by
numerical integration. Both inte-
grations refer to a reactor with the
following parameters, which fall an
the general range of contemplated ANP
design: temperatuvre coefficient of
reactivity, a, 10°*/°C; prompt gener-
ation time, 7, 10°% sec:; reciprocal
heat capacity of total fuel in the
reactor, 2.67 X 10°°%°C/watt sec;

reactor power, P, , 3 X 103 watts; fuel

transit time, &, through reactor,
1/8 sSsecC.,
One integration referred to the

case 1n which the power 1s four times
the average power, P,, for all times
t <0, At t > 0, the kinetic equations
of ORNL-1227¢2? apply. The other case
involved normal power, P , at all
times up to ome-tenth of a transit
time before t = 0, For a period of

 

(4)5. Tamor, Aircraft Nuclear Propulsion
Project Quarterly Progress Report for Period
Ending June 10, 1952, ORNL-1294, p. 31.
FOR PERIOD ENDING SEPTEMBER 160,

one~teth transit time duration before
t = 0, the power is 10 P,. As before,
at t >0 the kinetic equations of ORNL-
1227¢(2) valid. The power and
temperature as a function of time were
calculated through the second overswing.
The results are shown in Figs. 12 and
13, It can be seen that the extreme
values of P and T were nearly exactly
the same for the two cases, indicating
that they are close to the most extreme
values obtainable. |

are

For the reactor constants given,
it 1s found that after the first over-
swing has passed, P 1s bounded by
0.013 < P/P, < 2.1, and the bounds on
T are -50°C < T < +26°C.

P4
740

dat

POWER ( units of average power

 

1952

The similar shape and limiting
values of the curves in the two above
cases are easy to understand. After
the large initial disturbance, the
reactor almost shuts i1tself off. The
period of very low power is essentially
the same in both cases, about one
transit time, and 1t determines the
future time behavior of the reactor.
the following fic-
is plotted 1in

For this reason,
titious limiting case

Fig. 14. The power is zero for t < 0.
At t = 0, the power 1s suddenly brought
to P,, and thereafter the kinetic

equations of ORNL-1227(?) apply. Even
in this case, the power and temperature
oscillation is not much more violent

than in Figs. 12 and 13.

DWG .1€344

{

lrjlog F=-12.5 fl_P(;“M]-—U —-1] 7 &7

O

TEMPERATURE {units of r/af)

O 0.2 04 06 08 1.0 1.2 1.4 16 1.8 2.0 2.2 2.4
TIME (units of transit time)
Fig. 12. Power and Temperature vs. Time for P = 4 P,, t < 0.

- 37
ANP PROJECT QUARTERLY PROGRESS REPORT

DWG. 16345

£l 7 log P =425
F<~0 g 097 =-le A

=10 0

—01 <0

 

3
8 —
v o
5 S
5 2
2 E
£ 2
o
S #
if i
1
& G
o.
=
L)
'—...
0 0z 0.4 0.6 08 1.0 t.2 14 16 1.8 20 2.2
TIME (units of transit time)
Fig. 13. Poweyr and Temperature vs. Time for P = Py, t < =0.1; P = 10 P,
-B,1 < t < 0,

Of course, these calculations have stationary-fuel reactors as well and
to be gualified. It is necessary for are not uniguely problems of a circu-
the survival of the reactor that the lating-fuel reactor.
first temperature maximum stay within
tolerable limits. Furthermore, 1if
sufficient excess reactivity were EFFECT OF GAPS ON REACTIVITY
added for a prolonged period, the 5. T ] o
average temperature of the reactor 5. Tamor, ANP Division
zould 1ncrease_beyonqullowabfie lirjts As mentioned in the last report, (5)

cor A ]”N?g time, so, the a'zve calculations regarding the effect of a
calculations neglect any possible transverse gap through a reactor have

coupling between the “nuclear’ oscil-
lati1ons and mechanical vibrations. et e
All these qualifications apply to ) reida., p. 39,

 

38
FOR PERIOD ENDING SEPTEMBER 10,

3.0

 

1952

S e
;:;_y,fi‘-?f:_ T

DWG.1634¢6

1
g
2.6 77 109 P=~—12.5'/ [P(f—l-n—fl-—l]-q dn
0

2.2

POWER (units of average power)

0.6

0.2

 

TEMPERATURE (units of r/28)

    

O 0.2 0.4 0.6 08

 

TIME (units of tronsit time)

Fig., 14, Power and Temperature vs.

These calcu-
large reactor

been carried out.¢%’

lations refer to a bare,
and a gap that ts small compared with
the transverse reactor dimension,
They do not neglect, as in earlier
calculations(?? on the same problems,
the fact that neutrons preferentially
stream toward points of low neutron
density. The new calculations con-
stitute an improvement over the earlier
ones in cases where the gap is small.

 

The Effect of Gaps on Pile Re-

(6)5. Tamor,
1952).

activity, ORNL-1320 (July 14,

—4
~—8
1.0 1.2 14 1.6 18
Time for P = 0, ¢t < 0; P =P, t = 0,

comparison be-
based on the

Figure 15 shows a
tween calculations
earlier method, the calculation
according to the new method, and the
experiments on a 130- by 112- by
112-cm, rectangular, graphite-moderated
reactor with the gap perpendicular in
the long dimension and slightly off
center. The abscissa is essentially
the gap width,

 

(7)

J. E.
Beactivity,

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

39
ANP PROJECT QUARTERLY PROGRESS REPORT

e  DWG. 16347

100

| NS/
/

 

 

 

 

 

 

 

REACTIVITY LOSS {cents}

 

 

 

 

 

 

 

03

Reactivity Loss vs. Gap

Fig. 15.
Width.

SLOWING-DOWN KERNELS
R. D. Worley, USAF

The following kernels, of interest
to reactor theory, have been expressed

in terms of known and tabulated

functions: (&) (1) the Fermi age
kernel in the case of weak absorp-
tion, (2) the kernel describing, for

an infinite medium, the slowing down
of a neutron according to age theory
and the subsequent diffusion at thermal
energy, and (3) the kernel describing
the slowing down of a neutron in an
infinite space filled with a medium of
slowing down, and
except that a

uniform scattering,
absorption properties,
finite part of the space also con-
tains an additional absorber with an
absorption line at a definite energy.

(B)R. D. Worley, Slowing-Down Kernel with
Absorption and the Convolution of Various

Kernels, Y-F10-105 (to be issued).

4. CRITICAL EXPERIMENTS

A. D. Callihan, Physics Division

Measurements have continued on the
mockup of the G-E direct-cycle, air-
cooled, water-moderated reactor. Some
berylliumin the reflector was replaced
by a plastic and steel composite
reflector to compare the two arrange-
ments with respect toreflector savings
and reactivity., Loss 1in reactivity
incurred by the insertion of baron
carbide control rods was determined,

The preliminary assembly of the
fluoride-salt circulating-fuel air-
craft reactor experiment was completed,
and criticality was attained with
5.8 kg of U2?%% in the core and a total
investment of 6.5 kg of U??*%,  An
analysis of the results indicated
almost perfect agreement with theoreti-
cal caleculations.

41)

DIRECT-CYCLE REACTOR
R. C. Keen D. V. P. Willaiams

Physics Division

D. Scott, ANP Division

The study of a mockup of the G-E
direct-cycle reactor has continued at
ORNL. The critical assembly was fully
described 1n earlier reports,(1> and

the data summarized here have appeared

in more detail elsewhere.(?)

(I)E. V. Haake, D. V. P. Williams, ¥. G. Kennedy,

and D. Scott, Aircraft Nuclear Propulsion Project
Quarterly Progress Report for Period EndingMaerch
10, 1952, ORNL-1227, p. 5§9; V. Haake, D. V. P.
Williams, B, C. Keen, W. G. Kennedy, and D. Scott,
Aircraft Nuclear Propulsion Project Quarterly
Progress Report for Period Ending June 10, 18352,
ORNL- 1294, p. 34.

(2)A. D,

Reactor Assembly,
1952},

Callihan, Preliminary Direct Cycle
Part III, Y-B23-5 (June 18,
and Part IV, Y-B23-7 (June 30, 1852).
FOR

Réflector Studies. Some measure-
ments have recently been made with the
bervllium 1ina section of the reflector
replaced by a composite of stainless
steel and a hydrogenous plastic,
Boron, when placed between the com-
ponents of this composite reflector,
will reduce the neutron-induced gamma
radiation from the iron. The purpose
of this study was to compare the
reflector savings of the plastic and
steel composite reflector and the
neutron distribution in 1t with and
without the boron, The test section
was 5 3/4 in., thick, 14 3/8 in. wide
and 36 in. long. The beryllium re-
flector in these cells was replaced by
type 310 stainless steel except for a
5/16-in.-thick layer adjacent to the
cere, which was filled with Boral
strips 36 in. long. The Boral strips
were prepared from a boron carbide and
aluminum mixture that contained 35wt %
boren carbide. The mixture was
sandwiched between two layers of
aluminum sheets about 0.04 in. thick
and wrapped with masking tape. Thesys-
tem was made critical by the addition
of beryllium at locations remote from
the test section and by control rod
adjustment, and the loss in reactivity
was measured, By keeping the over-all
dimensions of the test section con-
stant, the thickness of the steel was
reduced and plastic was added between
the core and the Boral in a step-wise
manner until all the steel had been
removed, The changes in reactivity
that were incurred by these reflector
alterations, referred to the all-
beryllium reflector,
Fig. 16, An additional experimental
point on Fig. 16, designated as
“2 9/16-in. Plexiglas and 2 7/8-in.
stainless steel,”” shows the result of
removing the Boral from one plastic
and steel composite reflector section.
Another, measured with the Boral re-
placed by 5/16 in., of plastic, shows
that the effect of a 5/16-in. void
between the plastic and the steel is

are recorded 1in

PERIOD ENDING SEPTEMBER 10, 1952

small. The change in reactivity upon
removal of all material from the test
cell 1s also shown on the graph. It
is noted that with the hydrogenous
component 3 in, or more thick, the
presence of the boron does not greatly
reduce the reflector savings. Com-
parison of these data with earlier
results show, however, that the beoron
reduces the reactivity up to 30% for
plastic thicknesses approaching zero.

8&6. 16348

 

 

 

i |
REFLECTOR COMPOSED OF
~2%g~in. PLEXIGLAS AND

2% -in. STAINLESS STEEL

 

B0 e [

70} -

 

REACTIVITY LOSS {cents)
W o
< o

&
3

   

VOID 1IN TEST CELL

 

 

 

 

 

 

 

 

 

 

1300 ) . S RURNUORY RSPUOUUN: AUOSPUUES HEESOUOOIO NSO
0 f ? 3 4 5 6
PLEXIGLAS THICKNESS (in)

 

Fig. 16. Reactivity Losses vs.
Plexiglas Thickness for Stainless
Steel~Plexiglas~-Boral Composite Re-
flector.

Bare-indium and cadmium-covered-
indium traverses were made through

one of the above stainless steel,
Plexiglas, and Boral composite re-
flectors starting at a point 9 in.

inside the core periphery and termi-
nating at the edge of the reflector.
The reflector consisted of an outside
layer of 2 7/8 in. of stainless steel
separated from a 2 9/16~in, Plexiglas

41
ANP PROJECT QUARTERLY PROGRESS REPORT

inner layer by the 5/16-in. Boral
sheet, The data for the traverses
through the composite reflector when
the Boral strips were present in the
test cells are given in Fig, 17. Curve
1l is the bare-indium activation, curve
2 shows that obtained for the cadmium
covered indium, and curve 3 shows the
difference between curves 1 and 2, All
data were plotted as a function of the
distance from the outside of the test
sec&ion. Similar traverses were made
with the Boral strips omitted and the
space left empty. Figure 18 shows the
bare-indium and cadmium-covered-indium
activations and their difference
plotted also as a function of the dis-
tance from the ovtside of the test
section.

DWG. 16349

  
 
 

 

 

 

 

ACTIVITY {counis/min X 10%)

 

 

 

 

 

 

 

STAINLESS
| /' STEEL
0_5 R “\ CoTTT T o '_‘“ """ N ;_’1 """"""" T

i
O . } ‘ &‘::‘.fi—-f/’:'/:?
TS 2 9 6 3 o
DISTANGE FROM OUTSIDE OF REFLECTOR {in)

Fig. 17. Bare and Cadmium-Covered

Indium Traverses Through Stainless
Steel-Plexiglas~Boral Composite Re-
flector.

42

e O]
AT S

 

 

 

 

 

 

 

 

 

 

 

30 .

 

 

 

 

 

 

25—

 

 

ACTIVITY {counts/min X ’03)
— A R

 

 

 

 

 

 

  

 

 

 

 

 

 

 

 

 

 

\ 3 \\\\\\ |
e N _ N uf e
2.0 < [ | i
\\ / } J i
VoID !
\\ 2 t | i
15 e e S i s -
| | )E
li _
1Ok el o h STAINLESS_
‘ | STEEL |
‘ PLEXIGLAS
05 b \ \
' CORE-__ ¢
| | \
okf \ L =
15 12 9 6 3 0

DISTANCE FROM QUTSIDE OF REFLECTOR (in.)

Fig. 18. Bare and Cadmium-Covered
Indipm Traverses Through Stainless
Steel~Plexiglas Composite Reilector.

Poison Rod Calibrations. A measure-
ment was made of the reactivity
depression by each of three boron
carbide rods furnished by the General
Electric Co, for test as poison control
rods. The rods were 1/2 in. in diameter
and about 36 in. long and differed
slightly 1n 1internal structure; one
contained a plastic insert within the
boron carbide and the other rontained
aluminum. The loss in reactivity 1in-
curred by substituting the poison for
plastic within the reactor was found
to be about 20 cents‘?? at the center
of the core and about 11 cents at the

(S)One hundred cents is a rcactivity change
equivalent to the effective fraction of delayed

neutrons.
FOR PERIOD ENDING SEPTEMBER 10, 1952

edge. All the rods were of comparable

value.

ARE CRITICAL ASSEMBLY

D. Scott C, B. Mills
ANP Division

The preliminary assembly of the ARE
was described in the preceding re-
port.(4) The moderator and the re-
flector are beryllium oxide and the
fuel is enriched UF,. Hexagonal
beryllium oxide blocks with vertical
axial holes are stacked 1n a cylinder
36 in, high and 47 in. in diameter.
In the central blocks, a. section about
33 in. in diameter, are placed stain-
less stee)l tubes 1 1/4 in. in diameter
and 40 in. long containing a mixture
of 66 wt % Zr0,, 24 wt % NaF, and 10
wt % graphite to which was added
sufficient UF, to give a U?3% design
density of 0,16 g/cc. These com-
ponents, as dry powders, have a density
of about 1.9 g/cc. The same mixture,
without the uranium, has been packed
in 1/2-in.-dia tubes and placed in the
in the peripheral beryllium
The assembly

holes
oxide reflector blocks.
was made critical on August 21 with a
loading equivalent to slightly more
than 61 fuel tubes containing approxi-
mately 5.8 kg of U?3% 4in the core.
The total uranium investment, 1n-
cluding the sections of the tubes

 

1 i,

¢ )D. Scott, Aircraft Nuclear Propulsion Project
Quarterly Progress Report for Period fnding June
10, 1952, ORNL-1294, p. 38,

extending beyond the beryllium oxide,
is about 6.5 kg of U23%,

Before the critical experiment was
performed, a calculation of the
critical mass, power distribution, and
neutron spectrum had been carried
out.¢%) According to this calcu-
lation, the assembly should have gone
critical with about 58 tubes. However,
the following three effects, which had
not been included in the computation,
each add about one tube to the com-
puted critical mass so that there is
almost perfect agreement between ex-
periment and theoretical prediction.

1. A chemical analysis of the
Inconel actually used was received
after the computation was finished.
The analysis showed the Inconel to
be more of a poison than anticipated.

2. The computation did not consider
the loss of reactivity caused by voids
into which additional fuel tubes would
have been inserted 1f necessary.

3. Over and above the poison con-
sidered in the computation, some polson
was introduced by a stainless steel
piece in the control assembly.

No experimental data are available,
as yet, regarding power distribution
or neutron spectrum,

 

(S)C. B. Mills and D. Scott, The ARE Critical
Experiment, Y-F10-108 (Aug. 8, 1952).

43
 
SUMMARY AND

E. P. Blizard
Physics

The 1id tank facility was used
primarily for tests of the reactor
shield for the G-E direct-cycle design
(sec. 5). Extensive surveys have beén
made on the flux patterns in and around
large annular air ducts. In addition,
a study has been carried out on the
production of secondary gamma rays 1in
a metal layer near the reactor., The
attempted correlation between the
water-neutron data from the 1id tank
and the bulk shielding facilities
indicated that the data from the lid
tank facility give an attenuation that
1 low by a factor of about 2. Thais
is of particular importance because of
the number of shield designs based on
li1d tank data. |

The gamma~ray spectral measurements
on the divided shield mockup have been
completed 1n the bulk shielding facility
{sec. 6). The last measurements made
gave the energy and angular distri-
bution of the gamma rays streaming
around the edge of the lead shadow
shield, Approval has been obtained to
operate the bulk shielding reactor at
100 kw instead of 10 kw and experi-
ments have been started to remeasure
the air scattering from the divided

INTROBUCTION

J. L. Meem

Division

shield., An experiment is being carried
out to determine the amount of energy
released per fission, and a special
fuel element has been constructed for
this purpose. JTrradiation of mice,
rats, and rabbits by neutrons from the
Cockeroft-Walton accelerator was con-
tinued, and plans are being made to
expose monkeys to the radiation from
the bulk shielding reactor. A graphite
thermal column has been installed and
will be used with the gamma-ray spec-
trometer for measuring capture gamma
rays.

Construction of the tower shielding
facility has recently been approved,
and the design criteria have been
completed (sec. 7)., Preliminary
design work onmechanical and electronic
components 1s under way, and orders
have been placed for some of the
commercially available components.

Fission cross sections of U?%* and
U23% up to 4 Mev and the total cross
section of N** have been measured by
using the 5-Mev Van de Graaff neutron
accelerator (sec. 8). Some measure-
ments with the time-of-~flight spec-
trometer have been made on U?%® and Th.

47
ANP PROJECT QUARTERLY PROGRESS REPORT

3. LID TANK EXPERIMENTS

C. E, Clifford
L. S. Abbott
J. D. Flynn

G. T. Chapman
F. N. Watson
J. M. Miller

M. K. Hullings

Physics Division

C‘ [Jl

Experimental efforts inthe lid tank
facility have centered around the de-
sign of the shield for the G-E direct-
cycle reactor., Studies made on a
mockup of the outlet air ducts in the
GE-ANP initial engine have provided
estimates of the radiation that will
penetrate the G-E aircraft-reactor
shield., These estimates agree with
the approximate values obtained
theoretically by using assumptions of
isotropic scattering at the bends and
geometrical attenuation down the ducts,

The gamma shield for the side of the
G-E reactor was also mocked up in an
attempt to optimize the arrangement of
the components. In addition, the
contribution of a metal layer near the
reactor 1s being evaluated both in
terms of the 1ncrease in shield weight
and the reduction in radiation level,

Calculations have been made in an
attempt to correlate the neutron
attenuation data obtained in the lid
tank facility with those obtained in
the bulk shielding reactor. The lid
tank measurements seem to be lower by
a factor of about 2, depending on the
distance from the source., Since many
shield designs are based on lid tank
data, an attempt 1s being made to
improve this situation by altering the
lid tank geometry source,

AIR DUCTS

A mockup of the air outlet section
of the GE-ANP initial engine, com-
prising a transition section and duct

48

Storrs, General Electric Co.

(Fig. 19), has been examined in the
lid tank facility. Measurements were
made of the thermal-neutron intensity,
the fast-neutron dose, and the gamma
dose. Figures 20 and 21 give the
results plotted as isodose curves, and
Figs., 22 and 23 give fast-neutron
traverses as taken,

These measurements are primarily of
interest in estimating the radiation
that will penetrate the shield of the
G-E aircraft reactor., The geometry 1is
so complex that calculations of the
effect of the duct are difficult and
(with the present knowledge) not very
reltable. To a very rough first
approximation, however, the measured
transmission of neutrons agrees with
calculations based on the assumptions
of i1sotropic scattering at the bends
and geometrical attenuation down the
ducte. (1)

At the sides of the tank, where the
neutron flux varied laterally across
the counters, discrepancies appeared
as a result of different counter
diameters., This 1s attributed to the
fact that the effective center of
counting 1s not the geometric center
of the counters. Careful corrections
for this effect were not made, and as
a consequence the dosages given may be
in error by an amount corresponding to
a 1- or 2-cm displacement in the water,
The relative spacing of the isodoses,
however, is considerably more accurate.

 

(l)A. Simon and C. E. Clifferd, Simplified Duct

Theory (to be issued).
FOR PERIOD ENDING SEPTEMBER 10, 1952

L
OWG. 15405

 

 

 

 

 

 

 

 

 

 

 

42 in.

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

—% ;  _ ! = |
. ‘ S i ! Hj L~
o P 02" .
TRANSITION \<

SECTION
Fig. 19. G-E Qutlet Air Duct.

49
ANP PROJECT QUARTERLY PROGRESS REPORT

s THERMAL FLUX NORMALIZED TO NEUTRCN DOSIMETER

18Q
160
FIGURES ON CURVES ARE DOSE LEVEL IN mirep/hr

140 ; ST et

120

o @ &
o o Q

£, DISTAMCE FROM SQURCS (em)

o
o

2¢ '3

AIR-FILED
TRANSITICN SECTIO!
canlen

SOURCE /7,

 

1¢ 120 cOo 80 60 49 20 o 20 40 60 80 oo 120

Y, HORIZONTAL DISTANCE FROM SOURCE AXIS (cm)

Fig. 20. Neutron Isodose Measurementis with G-E Outlet Air Duct Mockup.

{cm}

ul

Z, DISTANCE FROM SOURC

50

DWG 154424
180

160

140

120

100

80

60

40

-~ TRANSITION SECTION

20

i SOURCE Yl

140 120 100 80 &0 40 20 0 20 40 60 B8O 100 120 140

 

Y, HORIZONTAL DISTANCE FRCM SOURCE AXIS {cm)

Fig. 21. Gamma Isodose Measurements on G-E Outlet Air Duct.
FOR PERIOD ENDING SEPTEMBER 10, 1952

2 - DWG. 154644
. 2X 10

 

{mrep/hr)

y=t047cm

1]
U3

e

NEUTRON DO

=75¢m

80 52
Z, DISTANCE FROM SOURCE (cm)

Fig. 22, Fast-Neutron Dosimeter Measurement on G-E Outlei Air Duct Showing
z Traverses for Various Values of vy.

51
ANP PROJECT QUARTERLY PROGRESS REPORT

DWI! 154654

2% 107

NEUTRON DOSE (mrep/hr)

=57cm

 

 

 

20 32 a4 68 80 g2 140
Z, DISTANCE FROM SOURCE (cm)

Fig. 23. Fast-Neutron Dosimeter Measurement om G-E Outlet Air Duct Showing
z Traverses for various Values of y.

52
FOR

%F1 the neutron data given were
obtained with the shutter between the
reactor and the source plate open.
The counting rate with the shutter
closed, that is, with only the back-
ground neutrons from the reactor, was
between 2 and 10% of the rate with the
shutter open. ‘

Most of the thermal-neutron data
were obtained by using an automatic
counting-rate plotter connected to a
counter that was moved at a constant
rate across the tank. Some isodoses
subsequently obtained directly
an automatic 1sodose plotter. The
obtained by the two methods are in
agreement.

were
with
data
good

The fast~-neutron dose was obtained
at several fixed positions with a poly-
ethylene-lined dosimeter,(%2) The
thermal-neutron fluxes can thus be
normalized to fast-neutron dose, to
which they should be directly pro-
portional, provided there are no major
perturbations in the neutron spectra,

The gamma background dose with the
shutter closed was of the order of 50%
of that with the shutter open and
therefore not negligible, Since,
furthermore, this ratio varies with
position, the difference was perforce
measured for all points to obtain the
attributable to the fission
source.,

dose

The distribution of neutron and
gamma dose around this duct was also
measured for various lengths of tran-
sition section. The gamma-dose distri-
butions for transition section lengths
of 0, 6, 12, and 18 in. have been
completed., For each geometry a center-
line traverse was made (Fig, 24) and
an isodose curve plotted (Figs. 25 and
26)., Figure 27 shows several isodose

 

(Q)F. M. Glass and &. 3.
Instruments 23, 67 (1952).

Hurst, Revw. Sci.

PERIOD ENDING SEPTEMBER 10, 1952

curves made with only water in the

tank.

The center-line measurement given
in Fig. 23 shows the difference be-
tween the readings with the shutter
open and closed. The 1sodose plots
have not been subtracted, since 1t 1s
apparent that the transition section
has no marked effect upon the dose.

The gamma isodoses are all auto-
matically normalized to alecal reactor
power monitor. Since this monaitor
reads less by a factor of 0.777 when
the shutter i1s closed, the corres-
ponding 1sodoses represent a dose
higher by 1/0.777 than the shutter-
open isodoses, as 18 noted on the
figures. In the case of the pure-
water isodoses, the resunlting curves
very nearly coincide, whereas with the
duct mockup in place there i1s a sepa-
ration of about 10 cm. This can be
attributed to the presence of gammas
from thermal-neutron capture 1in the
iron of the duct. The thermal-neutron
flux, of course, is considerably
affected by the shutter.

R-1 REACTOR(3) GAMMA SHIFLDING

The gamma shielding on the side of
the G-E aircraft reactor was mocked up
in the lid tank facility, and measures
were taken to determine its optimum
position in relation to the pressure
shell. The shield design under con-
sideration consisted of 4 in, of pure-
water reflector next to the reactor,
a l-in, steel pressure shell, and
additional iron and lead in borated
water, The guestionunder consideration
here was whether the additional iron
and lead could be placed next to the
pressure shell or whether 1t was

 

(B)The General Electric Co. direct~cycle air-
craft reactor, see Airecraft Nuclear Propulsion
Project Enginecering Progress Report No. 4, for
April 1, 1852 - June 30, 1552, APEX-~4.

53
ANP PROJECT QUARTERLY PROGRESS REPORT

GAMMA DOSE {(mr/hr}

 

 

DWG. 156334

12-in. TRANSITION SECTION

|

 

 

156

G-£ Outlet Air Duct.

10 i o
&0 72 84 96 108 120 132

Z, DISTANCE FROM SOURCE {cm)

Fig. 24. Gamma Center Line Measurements with

necessary to locate it farther out 1in
the borated water to reduce secondary
gamma production,

In the mockup used in the lid tank
facility, one 2.2-~cm iron slab was
placed 4 in. from the source to
represent the pressure shell., This
was followed by a 38B-in.-wide steel
tank, the inner wall of which (0.32 ¢cm)
is considered part of the pressure
shell., This tank was filled with

54

borated (0.6 wt % boron) water in which
the gamma shield wasmoved with respect
to the source. The gamma shaield,
sisting of one layer of i1ron and two
unequal layers of lead with each layer
separated by small thicknesses of
borated water, is shown 1n Fig. 28.
The gamma rays were measured 1in the
water behind the tank for six positions
of the gamma shield with both an 1o0n
chamber and an anthracene scintillation
counter. A seventh configuration was

con-
Fi it
180

160

140

120

1C0

80

60

Z, DISTANCE FROM SOURCE (cm)

40

20

Fig.

140

25.

120

100

80

60

 

40

FOR PERIOD ENDING SEPTEMBER 10, 1952

D\!! l !634.5

———NQ TRANSITION SECTION
— 18~in. TRANSITION SECTION

SHUTTER OPEN, 102 r/hr

SHUTTER CLOSED
1
1/0777 X152 r/hr

AIR-FILLED oucT

TRANSITION SECTION

AiR-FILLED DUCT

7

20 0 20 40 80 20 100 120 140

¥, HORIZONTAL DISTANCE FROM SOQURCE AXIS (¢m)

Gamma Isodose Curves with G-E Outlet Air puct Showing Effect of
Tramsition Section.

 

160

140

120

100

80

 

 

 

 

 

  
   

 

L LT

Ao , A_JWW
=Y
o SHUTTER OPEN,IO # r/hr

N p

 

 

 

 

  
 

   

 

\i
R
e PR

 

 

 

 

 

 

 

 

 

 

 

60

Z, DISTANCE FROM SOURCE (cm)

 

 

 

 

 

 

 

 

 

 

 

 

  

=== i2~in. TRANSITION SEC
s B-in. TRANSITION SECTION

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

  
 

 

 

TION

 

 

 

 

 

 
 

 

 

1ol

 
  

    

 

  

 

 

 

e w! = -
LIOTTT MO E pl b

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

  

40 4 — e mimabrnmmn e seds e il e e . _.,,_.;,:__,,_,_._L,__.
oo bl Ll L } |
L L
) RN |
140 120 100 80 & 40
¥, HORIZONTAL DISTANCE FROM SCURCE AXIS (cm)
Fig. 26. Gamma Isodose Curves with G-E OGutlet Air Duct Showing Effect of

Transition Section.

55
ANP PROJECT QUARTERLY PROGRESS REPORT

180 ”
- SHUTTER OPEN

160 e o

140 f—1 - -
SHUTTER OPEN
= bt DA , .-
=120 |
L UTTER CLOSED
e L L L S I S U
2 SHUTTER CPEN
100 -
=
o
T
5 SHUTTER CLOSED
o 80 SHUTTER GPEN
2 .
<
v
= 60
N
TTER CLOSED
40
20

140 120 100 80 60 40 20

 

DWG.15636A

10_11 r/hr P sefenees :

1
10”2 r/hr

10°% r/br

] 20 40 60 80 100 120 140

Y, MORIZONTAL DISTANCE FROM SOURCE AXIS (cm)

Fig.

also measured in which the gamma
shield was at the outside of the tank
(farthest from the source) and the
2.2-cm iron slab for the pressure
shell was removed. These data are
shown graphically in Figs. 29, 30,
and 31. Figure 32 shows the variation
in gamma dose at 70 cm from the source
as a function of gamma shield position,

Moving the gamma shield back 28 cm
from the pressure shell reduced the
gamma intensity by only 20% although
the weight of this component was in-
creased by 5000 pounds. This intensity
reduction can be achieved, on the
other hand, by adding only 1 cm of
iron 28 cm back from the main gamma
shield with a corresponding weight
increase of only 2000 pounds. The
iron slab 10,16 ¢cm from the source
reduced the radiation by about 20%,

whereas 1t would have reduced the

56

27, Gamma Isodose Curves with Pure Water.

intensity by about 33% if
neutron-capture gammas in the 1ron
itself had been eliminated.

g a mma

OTHER LID TANK EXPERIMENTS

Other work that has just been com-
pleted in the lid tank facility and
has not been published consists of
measurements on a mockup of the inlet
air duct¢*? of the G-E aircraft reactor
and a mockup of an array of round,
wavy ducts (Fig. 33), which arestill
considered to be a possible means of
carrying the cooling air to and from
the reactor., Work 1s now under way to
determine the effect on thermal-neutron
dosage of borating the water around
the exit air duct.

 

o

( )AircraftfluclearPropulsion Project Quarterly
Progress Report for Period Ending June 10, 1952,
ORNL-1294, p. 753, Fig. 41l.
SOURCE

FOR PERIOD ENDING SEPTEMBER 10, 1952

5 £ §5 EE

© BE an 2F

S ool ¥ o
Ad . I;M

DWG. 161554

31 35¢em
5t67cm

 

 

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: CONFIGURATION 2 KEY:

CONFIGURATION 3 SAME AS ABOVE EXCEPT W=1i0 cm

‘CONFIGURATION 4 SAME AS ABOVE EXCEPT: W =15 em

CONFIGURATION 5 SAME AS ABOVE EXCEPT W=20cm

PURE WATER
74 IRON
BORATED WATER

B LeaD

 

 

 

 

 

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CONFIGURATION 7

 

Fig. 28. Mockup of Gamma Shield for Side of GE-ANP Reactor.
ANP PROJECT QUARTERLY PROGRESS REPORT

DWG. 1633

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 
 

 

 

 

 

 

 

 

 

 

 

 

2 X 10° I s
CONFIGURATION 1
v
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- . CONFIGURATION § e e — ]
5 ,,,,,,,,,,,,,,,,,,,,,
E
E
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q e ——— el
W
5 — —
I
10'® 1ION CHAMBER
Lo 1, _
A Yp-in ANTHRACENE-CRYSTAL N
SCINTILLATION COUNTER &
]
?- ....... —_
10’
30 50 70 90 10 130 150 (70

Z, DISTANCE FROM SOURCE {cm)

Fig. 29. Gamma Measnrements Beyond Mockup of Side shield for GE-ANP Reactor.

Configurations 1, 3, and 5.

08
FOR PERIOD ENDING SEPTEMBER 10, 1952

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

o, DWG. 16352
2X10
& _CONFIGURATION 2
Tok &
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AN CONFIGURATION 4 -
5 \%\
= - ‘ X
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wi 2
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[ ]
<{
=
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>
10 \\
o 0 AN
& 10 10N CHAMBER LN
|4 Y-in ANTHRACENE - CRYSTAL. N\
SCINTILL.ATION COUNTER |
5 : : \{ \
N
R
2
A O
10°
30 50 70 90 110 130 150 170

Z, DISTANCE FROM SOURCE (cm)

Fig. 30. Gamma Measurements Beyond Mockup of Side Shield for GE-ANP Reactor.
Configurations 2 and 4.
ANP PROJECT QUARTERLY PROGRESS REPORT

)

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

150

2 DWG 16353
2 X105
CONFIGURATION 7
102 N N S _+.. _ _
””” - . CONFIGURATION 6 B
=
=
£
e 2 [ [ —_
el
w
2 8 Y,~in. ANTHRACENE - CRYSTAL
3 SCINTILLATION COUNTER
=
<1
v 10 T -
5 L i [ e
2 1 -
10° | |
30 50 70 90 Ho 130
Z, DISTANCE FROM SOURCE {cm)
Fig. 31. Gamma Measurements Beyond Mockup of Side Shield for GE-ANP Reactor.

Configurations 6 and 7.

60
FOR PERIOD ENDING SEPTEMBER 10, 1952

o e
DWG. 16354

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

2 X 10°
1 02 h e ——— — G,
e e .
O S _ L ;
_ 3 .
O - e .
=233
T .
5
Ll
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& _ .
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LLJ e o s -— e et
2 |
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< e e
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3
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- e ]
2 - ettt mt e e e e e ———————— e
Tox
O 5 10 15 20 25 30
BORATED WATER BETWEEN PRESSURE SHELL AND SHIELD {(cm)
Fig. 32. Gamma Dose as a Function ;of Borated Water Thickness in Side Shield
Mockup. '

61
¢9

Fig.

33‘

P

Wavy-Duct GE-ANP Mockup in Lid Tank.

e

PHOTO 10289

 

LHOdIY SSTY90Ud ATHALYVAO LJAl0¥d dNY
FOR

CORRELATION OF NEUTRON ATTENUATION DATA
E. P, Blizard

Physics Division

By the use of somewhat elaborate
transformations, the water-neutron
attenuation data from the lid tank
facility and the bulk shielding reactor
have been compared.(s) The 1id tank
data seem to be lower by a factor of
somewhat less than 2 at large distances
but more than 2 close to the sources.,
The reason for this discrepancy 1is

 

(S)E. P. Blizard and T. A. Welton, The Shield-

ing of Mobile Reactors - II., ORNL-1133 (to be
published in Reactor Science and Technology).

PERIOD ENDING SEPTEMBER 10, 1952

not understood, but several possible
explanations have been suggested.
Primary among these is the possibility
that the lid tank source plate 1is
actually operating at a somewhat lower
power than was indicated by early
measurements., The calculation of 1lid
tank source-plate leakage is compli-
cated by the rather difficult geometry
of the cylindrical, natural-uranium
slugs of which the source is made,
Installation of new, thin-plate U238
is being studied to determine how much
this might be expected to improve the
situation. Indications are that it
would be very desirable since so many
shield designs are based on lid tank
data,

 

6.
J. L. Meem
R. G. Cochram
M. P. Haydon
K. M. Henry
L. B, Holland

-

F.
G. M. McCammon

BULK SHIEFLDING REACTOR

Hungerford
. Johnson
Leslie

. Love

. Maienschein

0P xS m

Physics Division

The first series of studies of
gamma-ray spectra and angular distri-
butions with the divided shield mockups
have been completed. The final
measurements incorporated a lead
shadow shield and data obtained with
this arrangement will be used to
calculate the gamma-ray dose at the
crew compartment.

Plans are being made for using the
recoi1l-proton spectrometer to measure
spectra and angular distributions of
fast neutrons leaving the reactor part
of the divided shield mockup.

In order to resolve the uncertainty
of the power of the bulk shielding
reactor, an attempt 1s being made to
obtain a more accurate value of the

energy released per fission. A special
fuel element equipped with thermocouples
and water-flow tubes has been con-
structed so that the energy per fission
can be determined as a measure of
temperature rise and water flow.

Preparations are under way to
expose monkeys to radiation from
the bulk shielding reactor, and a
tentative program has been arranged.
Animals previously irradiated by the
Cockcroft-Walton accelerator are now
under observation.

A graphite thermal column has been
installed in the bulk shielding
facility to provide an intense source
of thermal neutrons. With this source,
the gamma rays produced by neutron
capture can be studied.

63
ANP PROJECT QUARTERLY PROGRESS REPORT

MOCKUP OF THE DIVIDED SHIELD

Gamma-ray spectral measurements
during the past quarter were made with
a lead shadow shield added to the
mockup of the aircraft divided shield.(!)
These measurements complete the first
series of studies of gamma-ray spectra
and angular distributions with the
divided shield mockup. The data
obtained will be used for a preliminary
calculation of the gamma-ray dose
received at the crew-compartment
location, which will then be compared
with the estimates of the Shielding
Board.(?)

The results obtained with the lead
shields are shown in a series of five
figures. Figure 34 shows a gamma-ray
1sodose curve obtained by using an 1ion
chamber without the lead disks. This
curve 1s very nearly a segment of a
circle with a radius of 168.2 cm and
a center near the reactor center. The
other curve in Fig. 34 shows the dose
along this same circle after addition
of the leak disks.

From these data it was decided to
examine the energy and angular distri-
butions at the angles ¥ = 0 deg and
Yy = 50 deg, where Y is the angle
between the aircraft axis and a line
connecting the pseudo reactor center
and the nose of the spectrometer
collimator. The energy spectra ob-
tained for various values of the
angle & are shown in Figs. 35 and 36
for Y = 0 and 50, respectively, where
fis the angle between the spectrometer
collimator and the line joining the
pseudo reactor center and the nose of
the spectrometer collimator. The
corresponding angular distributions
are shown 1n Figs. 37 and 38.

 

(I)Aircraft Nuclear Propulsion Project
Quarterly Progress Reports, ORNL-1227, p. 73
and ORNL-1204, p. 44.

(2 peport of the ANP Shielding Board, NEPA-
ORNL, ANP-53, Appendix C (Oct. 16, 1950).

64

Development of the recoil-proton
spectrometer 1s continuing. Plans
are being made to use the instrument
for measuring spectra and angular
distributions of fast neutrons leaving
the reactor part of a divided shield
mockup when the air-scattering ex-
periments are completed. A report on
the theory of this instrument has
recently been published.(3)

AIR-SCATTERING EXPERIMENTS

A memorandum(?%) giving the details
of the air-scattering experiments
described in the previous report has
been issued. Similar experiments at
100-kw reactor power are now getting
under way.

REACTOR POWER DETERMINATION

One of the largest uncertaintiles
in the determination of the power of
the bulk shielding reactor rests in
the lack of an accurate value for the
energy released per fission., A special
fuel element has been constructed to
be used in measuring this quantity
(Fig. 39). The center fuel plate of
the element i1s removable and disks
are punched out of the plate. The
gamma activities of these uranium-
bearing disks are counted after an
exposure to determine the number of
fissions that occurred in the fuel
element and then compared with the
gamma activity in a similar disk
previously calibrated in the standard
reactor.

The special fuel element is equipped
with thermocouples and water-flow
tubes. Special precautions against

 

(3)8. R. Gossick, General Principles of
Proton-Recoil-Fast-Neutron Spectrometer, ORNL-1283
(July 14, 1952).

)J. L. Meenm and H. E. Hungerford, Air
Scattering Experiments at the Bulk Shielding
Fecility, ORNL CF-52-7-37 (July 8, 1952).
89

 

   
  
  
   
   
  
 
 
  
  
     

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 
 
  

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

—GANMMA DOSE ALONG 1SOT0SE CURVE
~ WITHOUT LEAD SHADOW SHIELDS
LEFT SiDE B.49 X107 r/he/wett
SCALE DIAGRAM SHOWING LOCATION OF ISODOSE CURVE : roheswe
WTH RESPECT TO REACTOR SHIELD
= RIGHT SIDE:
_ S POLAR PLOT OF GAMMA DOSE
KEY TO SYMBOLS ) - 7R ALONS 1SODOSE CURVE WITH
& POSITIONS OF DETECTOR AT WHICK EXPERIMENTAL ARG SHADOW SHIELDS iN POSITION
DATA WAS O3TAINED - SN
= :SODOSE CURVE VALUES OBTAINED BY INTERPOLATION S "‘\\ o \\ \ © EXPERIMENTAL PCINTS
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| N . , b
| | - VoL
20! it - - vl
; T ‘ 7 REACTOR ,wlfq——rf

 

 

 

 

 

 

 

GAMMA DOSE (r/hr/watt)

 

 

 

Fig. 34. Gamma Isodose Curve Around Reactor Shield Without Shadow Disks and Dose Alomg Isodose
Curve with Shadow Disks im Position.

‘01 HAGWIALAAS ONIOANA GOTHAd HOod

cS6l
ANP PROJECT QUARTERLY PROGRESS REPORT

=0
8 = VARIABLE

T { gammas /cm2/sec/Mev/watt /steradian’

 

0 2 4 6 8 10
GAMMA-RAY ENERGY (Mev)

Fig. 35.
Lead Disks.

Gamma-BRay Flux Behind

spurious loss of heat are taken so
that the heat generated in the fuel
element 1s measured directly from
water flow and temperature rise.
Making allowance for the net leakage
of gamma radiation into or out of the
fuel element, the energy per fission
can then be calculated. The experiment
is well under way.

IRRADIATION OF ANIMALS

In cooperation with the Health
Physics Division and the USAF School
of Aviation Medicine, preparations
are being made to expose monkeys to
radiation from the bulk shielding
reactor. Special water-tight cages

66

r (gammas/cma/sec/Mev/wuH/ steradian)

 

0 2 4 6 8 10
GAMMA-RAY ENERGY (Mev)

Fig. 36. Gamma-Ray Flux Behind

Lead Disks.

are being constructed in which the
monkeys will be placed under water
near the reactor. The tentative
program for the irradiation of the
monkeys is presented in Table 4.

In addition to the animals ir-
radiated, 12 animals will be used
for control immersions in the bulk
shielding facility,

After completion of the exposures
the monkevys will be returned to the
Primate Laboratory, USAF-SAM, at
Austin, Texas. Particular attention
is to be devoted to examination for
eye cataract,
L9

DWG. 45757

\0
HYPOTHETIGAL SHIELD ¢5§¢
/ BOUNDARY Zé\% @ \ so0
. \(J
L AN
DIVIDED SHIELD ™~ s \\\ ‘ \
e \ S §§\\
\ 3 ’Q
¥ "‘b
TR \
OB\ s

    
 

DIVIDED SHIELD
AND REACTOR &

 

2
___________ Ao L)

Fig. 37. Gamma-Ray Flux as & Function of Angle for Energies as Shown Be-
hind Lead Disks (DSML, Y = 0°),

‘0T HAEWALJIS ONIONT qOoIddd HOA

csol
ANP PROJECT QUARTERLY PROGRESS REPORT

Wt e e Snanin —mmmn mmaan SRl e—— —— ma —rm—— ———

~ YPOTHETICA HI
DIVIDED SHIELD \\\///H POTHE L SHIELD

MOCKUP

     
  

DIVIDED SHIELD
AND REACTOR &

T st — ——— i A R, e s et

-30°

Fig. 38. Gamma-Ray Flux as a Function
hind Lead Disks (PSML, ¢ = 50°),

68

BOUNDARY

LEAD SHADOW
DISKS

of Angle for Energies as

60°

50°

40°

20°
ANGLE O
Shown Be-
FOR PERIOD ENDING SEPTEMBER 10,

1952

 

Fig. 39.
sion (Partly Disassembled).

Special Fuel Element for Measurement of the Energy Released per

Table 4

TENTATIVE PROGRAM FOR IRRADIATION OF MONKEYS IN THE BULK SHIELDING FACILITY

 

 

SERIES 1 SERIES 2
Exposure rate* (rem/hr) 1 0.25
Number of exposures 8 16
Time per exposure (hr) 16 8
Time between exposures (days) 7 7
Total exposure (rem) 128 32
Animals used 12 12

 

 

 

*One~-half of dose will be in ueutrons and one«half in gamma rays to the animals irradiated,

The initial irradiations of mice,
rats, and rabbits on the Cockcroft-
Walton accelerator have been completed.
The animals were subjected to total
fast-neutron doses of 14-Mev neutrons
ranging from 107 to 3 X 10!° neutrons
per e¢m?. Tt is expected that the
threshold for eye cataract will lie
in this interval and the animals are
now under observation. The results
will be published by members of the
Biology Division.

IRRADIATION OF ELECTRONIC EQUIPMENT

Additional electronic equipment
from the Air Force was exposed to
radiation frow the bulk shielding
reactor early in the quarter. No
appreciable damage was observed.
The exposures have been discontinued
temporarily because of conflicting
higher priority work.

69
ANP PROJECT QUARTERLY PROGRESS REPORT

CAPTURE GAMMA-RAY MEASUBEMENTS

Of considerable importance 1in
shielding calculation 1s a knowledge
of the gamma rays produced by neutron
capture in the various materials of
the shield. Upon completion of the
gamma~ray spectral measurements with
the divided shield mockup, the spec-
trometer(53) will be used for measuring
capture gamma rays from many materials
of interest. 7To provide an intense

source of thermal neutrons for the
production of such gamma Tays, a
graphite thermal column has been
installed in one corner of the pool.
A 4- by 6- by 7-ft stack of AGOT-grade
graphite has been installed, and the
spectrometer is being adapted to the
new program,

 

(S)F. C. Matenschein, Multiple-Crystal
Gamma-Ray Spectrometer, ORNL-1142 (July 3, 1952).

 

 

7. TOWER

C. E. Clifford

SRIELDING FACILITY

T. V. Blosser

Physics Division

The tower shielding facility, as
described in the recent proposal,(l)
has been approved for construction,
Calculations of structure and ground
scattering to be expected in the tower
shielding facility have been completed
and published in an ORNL report.(?)

Design criteria for the tower and
bui1ldings have been prepared for sub-
mission to an architect engineer who
will complete the detailed design.
Completion of the design contracts
should require three to four months,
and construction should be completed
in approximately nine months.

 

(Vg p. Blizard, Proposal for a Divided Shield
Testing Facility, ORNL CF-52-4-85 (Apr. 17, 1952).

(2)A. Simon and R. H., Ritchie, Background
Calculations for the Proposed Tower Shielding
Facility, ORKL-1273 (to be issued).

70

Orders have been placed for a
majority of the reactor control com-
ponents and instrumentation, and an
investigation 1s being made to deter-
mine the best solution to the rather
difficult problem of rigging the
cables for this configuration. Orders
for components of the experimental
instrumentation that are commercially
available are being placed, and pre-
liminary design of both the mechanical
and electronic components of the remote
instrumentation is in progress.

A supersensitive fast-neutron
dosimeter 1s being developed with the
assistance of G, S. Hurst of the Health
Physics Division, who developed the
original instrument. The first approach
will be to try to make use of multiple
sensitive chambers operating a single
electronic recording circuit,
FOR PERIOD ENDING SEPTEMBER 10,

8. NUCLEAR

1952

MEASUREMENTS

A. H. Snell, Physics Division

Fission cross-section curves of %34

and U23% relative to U??® for incident
neutron energies up to 4 Mev have been
determined by using the 5-Mev Van de
Graaff neutron accelerator. The N!*
total cross section has been measured
from 1.6 to 4 Mev on this accelerator.
Some energy levels in thorium and
resonance values in U?3% were observed
with the time-of-flight spectrometer.

FISSION CROSS SECTION OF U?3* anp y23é
H. B. Willard, Physies Division

The shapes of the fission cross-
section curves for U2?34 and y23é
relative to U23%% have been determined
by using the 5-Mev Van de Graaff
accelerator.(!) The statistical error
was 2 1/2% when taking readings about

 

(1)

R. W. Lamphere, The Fission Cross Sections
of Urenium-234 and Uranium-236 for Incident
Neutron Energies up to 4 Mev, ORNL-1312 (July
15, 19§2).

3.0 e ettt e et

 

W)

 

 

2.0

TroTAL {barns}

 

 

 

 

 

50 kev apart up to 4 Mev, A tritium-
gas target that gave an rms energy
spread varyving from 60 to 100 kev was
used. Owing to foil uncertainties,
however, the ordinate scale factors
are in doubt to a probable error of
14% for U?** and 9% for U?3®, Thres-
holds were found to be 0.37 and 0.69
Mev for U??* and u??$, respectively,
Two minimums were found im each curve
at different neutron energies for the
two isotopes, which indicates that
they are true minimums for U%3* and
U236 rather than maximums in U235,

TOTAL CROSS SECTION oF N!?
H. B, Willard,

Physics Division

The total cross section of N'? has
also been measured with the 5-Mev Van
de Graaff accelerator. Resolution of
35 kev was obtained between 1.6 and
4 Mev., The graph of these results 1is
shown in Fig. 40.

UNCLASSIFIED
BWG. 16355

 

 

 

 

 

 

15 20 25

3.0 35 40

NEUTRON ENERGY (Mev)

Fig. 40,

Total Cross Section of N4

(35-kev Resolution).

71
These measurements corroborate the
mean free path values used in air-
scattering calculations for the divided
shield.(?) Eventual extension of
these data up to 8 Mev will also be of
interest, although it 1s unlikely that
the extended data will indicate any
gross differences from the extrapolated
values, The data are sufficiently
detailed for IBM calculations of air
scattering as far as incident neutron
energy is concerned, but there 1s still
considerable uncertainty regarding the
energy and angular distribution of
scattered neutrons. When the latter
data are available, and when the energy
and angular distribution of neutrons
leaving the reactor shield have been
measured in the bulk shielding facility,
it will then be possible to make
machine calculations of neutron air

 

(2) heportof the ANP Shielding Board, NEPA-ORNL,
ANP-53 (Dct. 16, 1950).

72

These calculations should
provide valuable information prior to
operation of the tower shielding
facility.

scattering.

TIME-OF-FLIGHT SPECTROMETER

G, S, Pawlicki E. C. Smith
P. E. F. Thurlow

Physics Division

The time-of-flight spectrometer has
been used to measure the transmissions
of thick samples of thorium oxide and
depleted uranium oxide (U,0, with less
than 7 ppm U??%). Energy levels were
observed in thorium at 23.5, 35, 71,
127, 260, and 870 electron volts.
Resonances were also observed in U?38
at 6.8, 21, 39, 62, 110, 200, and 1700
electron volts., Further details will
be found in the next Physics Division
quarterly report,
 
SUMMARY AND INTRODUCTION

The research on high-temperature
ligquids has been directed primarily
toward the production of a fuel for
the ARE. The longer range work 1is
principally concerned with studies of
fluoborate systems and the purification
of hydroxides (sec. 9). The ARE fuel
is in the system NaF-ZrF,-UF,, and the
most probable composition is 46 mole %
Na¥F, 50 mole % ZrF,, and 4 mole % UF,,
The melting point of this composition
is 510°C and its vapor pressure and
viscosity are tolerable. A satis-
factory loading technigue involving
the addition of ZrF,-UF, toan NaF-ZrF,
base has been proposed. Numerous
tluoride systems, with and without UF,,
have been examined, and systems con-
taining UF, have been prepared. The
study of the reaction of fluorides
with alkali metals has indicated that
large quantities of NaK may be added
to the system NaVF-ZrF, -UF, before
producing any free UF,.

Most of the corrosion research
during the past guarter has centered
around the determination of corrosion
characteristics of fluoride mixtures
containing ZrF, (sec. 10). Both static
and dynamic tests 1indicate that this
class of fuels and coolants 1is less
corrosive than the previously tested
fluoride fuel systems, an improvement
which may result from better material
production and testing technigues,
Static tests and dynamic tests have
been run with fluorides and hydroxides
to determine the effects of such
time, temperature,
and environment,

variables as
additives, stress,
A curious result was the apparent de-
crease of fluoride attack with in-
creasing temperatuve. This may have
been the result of the formation of a
protective oxide coating., Aside from
the reduced attack that may have
resulted from improved preparation
techniques, the most significant

advance during the last guarter came
from the addition of ZrH, to the
NaF-KF-LiF-UF, mixture. To supplement
corrosion tests, considerable effort
has been applied to fundamental studies
of the corrosion mechanism, including
synthesis and identification of cor-
rosion products, The hypothesis that
fluoride corrosion of Inconel depletes
the chromium from the metal lattice and
the resulting voids precipitate has
been supported by further tests and
observations.

The metallurgical methods for the
construction and assembly of the Air-
craft Reactor Experiment, including
welding and brazing, fabrication of
control rods, and ceramic coating of
radiator fins, have been developed
(sec. 11). Specifications have been
established on procedure and qualifi-
cations for inert-arc welding of
Inconel pipe and fittings for highly
corrosive applications., The B,C-Fe
and the Al,0,-B,C inserts for the ARE
control and regulating rods have been
pressed, An apparently satisfactory
high-temperature ceramic coating has
been applied to nickel sheet for use
in a liquid metal-to-air radiator. In
addition, the creep and stress of
Inconel are being determined in air
and argon, and the mechanical and
corrosion-resistance properties of
brazed joints are being investigated.
Tensile tests have shown brazed-joint
efficiencies as high as 92%.

Heat transfer and physical proper-
ties measurements on various fluorides
have resulted in lessening the effort
on liquid metal and hydroxide measure-
ments (sec. 12). The viscosity of the
ARE fuel, NaF-ZrF,-UF, (46-50-4mole %},
ranges from 20 to 7 centipoises he-
tween 580 and 830°C. The vapor
pressure of this fuel increases from

12 to 84 mm Hg between 807 and 940°C.

75
ANY PROJECT QUARTERLY PROGRESS REPORT

Thermal conductivity, heat capacity,
and density measurements have been
made on several high-temperature
liquids. The experimental heat
transfer data for sodium hydroxide may
be represented by an equation that can
be evaluated to within 9% of the
values normally used for ordinary
fluids. A heat-momentum-transfer
analysis of a thermal convection loop
indicates a circulation velocity of
about 0.1 ft/sec.

The radiation damage program in-
cludes irradiation of fluoride fuel
mixtures, measurements of the effect
of radiation on creep and thermal

conductivity, and operation of in-
reactor loops (sec. 13). Reactor
irradiations of a mixture containing
ZrF, indicate that no significant
radiation-induced corrosion will occur
at ARE intensities., An experiment to
determine the rate of diffusion of
Xe!®® from the irradiated static fuel
mixture NaF-BefF,-UF, indicated that
almost all the xenon will remain in
the fuel unless flushed out. The
1700°F annealing temperature proposed
for the ARE fuel tubes appears to be
of some consequence in minimizing creep
and thermal conductivity changes under
irradiation. No radiation-induced
corrosion was observed in the sodium
in-reactor loop.

 

9. CHEMISTRY OF HIGH-TEMPERATURE LIQUIDS

W. B. Grimes, Materials Chemistry Division

Research in the ANP chemistry group
has been concerned almost entirely with
studies of fluoride mixtures for use
as fuels and coolants for an aircraft
reactor. Although the major effort
has necessarily been directed to
problems of immediate concern to the
ARFE, a number of longer range studies
is being carried out.

The program in research and pilot-
scale production of liquid fuels of
high purity isstill actively followed,
although responsibility for production
of the large quantities of such ma-
terials needed for engineering evalu-
ation 1s now shared with the experi-
mental engineering group,

Efforts continue on definition of
the optimum composition of the ARE
fuel and on studies of correosion of
metals by the various high-temperature
liguids being considered. The previous
attempts to identify chemical species
in the cooled melts before and after
corrosion testing have continued.

76

A study of the possible reactions
of the fuel mixture with alkali metals
at elevated temperature has been
initiated to ascertain in quantitative
fashion the effect of a leak between
the fuel and moderator circuits.
Detailed study of the system after
reduction of UF, and/or Zr¥, has
indicated that the mixture of products
is quite complex even if chemical
equilibrium is obtained. Some indi-
cations of equilibrium among U¥F,, UF,,
ZrF,, and lower valence fluorides of
zirconium suggest, however, that
addition of small amounts of alkali
metals may be beneficial as far as
corrosion 1s concerned.

Detailed explanation of the systems
obtained by addition of NaK to the
various fluoride fuels has necessitated
reactivation of the study of the
preparation and the properties of UF,.
Preparation of the pure material has
been accomplished on a moderate scale,
and studies of its chemical properties
and phase eguilibriums in systems of
which i1t is a component are under way.
I'OR

The preparation of a simulated fuel
for the ARE critical experiment has
been concluded, and the tubes have
heen filled with powder by the Y-12
Production Division. Satisfactory
progress 1is being made on the only
remaining problem ~ filling one tube
with a solid rod of fuel.

A small program dealing with the
purification of alkali hydroxides and
the determination of the high-tempera-
ture properties of these materials has
been maintained.

FUEL MIXTURES CONTAINING UF,
R. E., Traber, Jr.

Barton

L:NM? Bratcher
C. J.

Materials Chemistyry Division

Several binary systems have been
studied because of the need for thas
information 1in construction on the
three-component systems containing
them. The binary and termary mixtures
containing aluminum fluoride that have
been examined show little promise of
successful application.

The four-component system NaF-KF-
Zr¥,-UF, appears to yield no melting
points below 510°C at the 4 mole %
UF, level. Since this temperature can
be achieved in the NaF-ZrF, -UF, system
at this uranium level, more attention
has been paid to the simpler three-
componenti system,

The mixture containing 50 mole %
ZzF,, 46 mole % Nal’, and 4 mole % UF,
appears at present to be the most
promising of the fuel mixtures, This
material, which melts at 510°C, shows
a partial pressure of ZrF, of 13 mm at
1500°F, and its viscosity seems satis-
factory for the ARE (c¢f., “Heat
Transfer and Physical Properties
Research,” sec. 12). Slightly lower
(8 mm) vapor pressure and lower
viscosity values are available with

PERIOD ENDING SEPTEMBER 10, 1952

the similar mixture containing 46
mole % ZrF,, 50 mole % NaF, and 4
mole % UF,. This mixture, which
appears to freeze to a single solid
solution of NaUF; 1in NaZrF{T melts
sharply at 520°C., The final choice
between these mixtures has not been
made. Their preparation, handling,
and corrosion problems appear very
similar,

Start-up of the reactor by filling
the machine with Na¥F-ZrF, (corres-
ponding to either of these fuels) and
adding the NaF-UF, eutectic (26 mole %
UF,) to bring the fuel to criticality
would not be an easy operation, since
a mixture melting as high as 700°C
could be produced by local concen-
trations and inadequate mixing.

However, on the line representing
50 mole % NaF the melting point de-
creases gradually from NaUF, (m.p.,
710°C) to NaZrFg (wm.p., 510°C). All
intermediate mixtures appear to be
solid solutions of the two compounds.
The mixture containing 25 mole % UF
and 75 mole % ZrF, melts at 610°C and
should contain about 105 lb of uranium
per cubic foot at 800°C., This mixture
could, presumably, be used to bring
the reactor to criticality at tempera-
tures above 625°C,

NaF-KF-ZrF,-UF,. Data presented
previously({1) showed that at the 4
mole % UF, level two compositions in
the NaF-KF-ZrF,-UF, system yielded
melting points below 500°C. Re-
examination of the system 1ndicates
that these determinations were 1n
error. The lowest melting point
verified in this system is at 505°C and
corresponds to 38.4 wole % ZrF,, 18.7

 

mole % NaF, 38.9 mole % KF, and 4.0
m(}le % UF4 .
(I)L. 8. Bratcher, R, %, Traber, Jr., and C. IJ.

Barton, Aircreft Nuclear Propulsion Project
Quarterly Progress Report for Pertiod Ending June
10, 1952, ORNL-1294, p. 84.

77
ANP PROJECT QUARTERLY PROGRESS REPORT

Material containing 53 mole % KF,
5> mole % NaF, 40 mole % ZrF,, and 2
mole % UF, melts at 410°C, and it would
prove to be a very satisfactory fuel
for a future model with sufficient
volume to contain a critical mass of
this composition,

NaF-ZrF,-UF,. Tentative contours
for the NaF-ZrF,-UF, system are shown
in Fig. 41, Although some difficulty
was experienced in reproduciang data in
this system, possibly because of the
ease of oxidation and hydrolysis of
the guadrivalent fluorides, repeated
measurements have indicated that the
contours are not 1in serious error,

especially in the low-melting-point
region, It should be possible to use
up to 8 mole % UF, should this be
necessary with melting points below

550°C,

NaF-ZrF, -BeF,-UF,. Systems con-
taining BeF, have previously been
shown to possess high viscosities in
the 600 to 700°C range, but a number
of such systems have usefully low
melting points, Some attempts have
been made to decrease the melting
point of the NaF-ZrF,-UF, system by
addition of BeF,. Additions of BeF,

to a mixture containing 50 mole % NaF,
46 mole % ZrF,,

and 4 mole % UF,

DWG. 16356

 

 

 

Ko
/ | Q

Na3ZrF7

NaoF

Fig. 41.

18

 

 
  

INDETERMINATE

ZF‘Fq,

The System NaF-ZrF4-UF4.
FOR

redwcé&dthe melting point to 487°C,
Mixtures with 10 and 15 mole % BeF,
apparently melted at this temperature.

Addition of ZrF, to a ternary mix-
ture of NaF-BeF,-UF, (76-12-12 mole %,
respectively; m.p., 480°C) caused an
initial rise in melting point. At
10 mole % ZrF,, the mixture melted at
670°C; at 20 mole % ZrF,, the melting
point was 570°C,

mixtures
study of
will be

I1f the viscosities of these
are usefully low, additional
the four-component system
attempted.

ZnF,-UF,. Thermal data were obtained
with a number of mixtures in the ZnF, -
UF, system., The data indicate compound
formation but do not show definitely
the composition or melting point of
the compound. The lowest melting point
observed was 730 + 10°C. Since none
of the alkali fluoride-zinc fluoride
systems showed low melting points,
further work on this system 1s not
presently contemplated. '

FUEL MIXTURES CONTAINING UF,

V. 8. Coleman W. C. Whitley
Materials Chemistry Division

Study of the phase relationships of
uranium trifluoride with other fluorides
is of interest for several reasons.,
This compound may be a product of
radiation damage to fuels containing
uranium tetrafluoride. Tt has a lower
vapor pressure than uranium tetra-
fluoride and might produce fuels with
‘more suitable liquid-temperature
ranges than uranium tetrafluoride.
Since i1t 1s a much less powerful
oxidant, uranium trifluoride may be
less corrosive to container material
than uranium tetrafluoride. Uranium
trifluoride has been observed 1in the
product after NaK has been added to
fuels containinguranium tetrafluoride,

PERTOD ENDING SEPTEMBER 16, 1952

About 3000 g of UF, has been
prepared by the method described in a
previous report.(?’) This material,
the analysis of which is higher than
99% UF,, has been used in all the
studies reported below,

Uranium trifluoride is quite stable
at room temperature but oxidizes
rapidly at elevated temperatures in
the presence of traces of water vapeor
or other oxidizing agents. It is
therefore necessary to protect UF,
from all oxidizing agents when it 1is
involved in high-temperature studies.
This 1s accomplished in these experi-
ments by use of the apparatus dia-
grammed in Fig. 42, This apparatus
contains a stuffing-box seal(?)
packed with an oil-free graphite-
asbestos product, This seal permits
the removal of air and water vapor by
means of a vacuum pump; an 1inert
atmosphere is then maintained while
the sample i1s studied at high tempera-
ture,

NaF-KF-LiF-UF;. Some preliminary
experiments were performed with the
eutectic mixture of NaF, KF, LiF, and
varying amounts of UF, up to 10 mole %.
Jt was observed that the freezing
point of the eutectic mixture was
depressed by an amount up to 13°C.
However, the molten material diffused
through the graphite crucible and
stirrer and cemented them quite fairmly
to the stainless steel container,
Examination of the product showed that
some of the UF; had been oxidized,
some had been converted into alkali
fluoride complexes, and some remained
as unreacted UF,. '

The reaction of NaF-LiF-K¥ eutectic
plus UF,; with the graphite crucibles

 

(2)W. C. Whitley and C. J. Barton, Aircraft
Nuclear Propulsion Project Quarterly Progress
Report for Period Ending Septermber 10, 1952,
ORNL-1154, p. 159, '

(3)Deuigned by W, C.

Pivision.

Tunnell of the ANP

79
ANP PROJECT QUARTERLY PROGRESS REPORT

SHAFT (CONTAINING
THERMOCOUPLE WEL.L)'“*-—-\“k

 

<

 

 

 

- THERMOGOUPLE DWG. 16357
_——PULLEY
<
1
]

 

 

PACKING GLAND ——-

  
 

COOLING FING————_

 

 
 

SN

CrA e RRABT N NN
e e
L

 

 

 

. ——GRAPHITE-ASBESTOS SEAL

 

 

V

 

 

 

A

 

 

  
 

TUBE FOR
ADDING SAMPLE——

 

 

 

COPPER GASKET
= | |
S X s e-—""UPPER FLANGE
SIS RS - L OWER FLANGE
VACUUM
HELIUM

 

_ér,w- -------------- —THERMOGOUPLE WELL

 

SAMPLE——~w~nH%¥

 

 

 

 

 

 

 

 

- GRAPHITE OR STAINLESS
STEEL STIRRER
FEENTT - GRAPHITE OR STAINLESS
Vo STEEL CRUGIBLE
"

 

 

 

 

fFig., 42.

80

Apparatus for Phase Studies of Fluoride Systems Containing UF,.
FOR PERIOD ENDING SEPTEMBER 10,

a

is a matter of interest
This reaction does not occur with the
eutectic when UF, is used instead of
UF;., Three pieces of graphite exposed
to 5 mole % UF, in the NaF-LiF-KF
eutectic at temperatures up to 900°C
were found to contain 11.8 wt % uranium.
The pronounced odor of acetylene when
the mixture 1is exposed to moisture
suggests that a uranium carbide 1s
produced. Since the lowest recorded
temperature for the reaction of uranium
metal and carbon was 1200°C and most
workers reported 1700°C or higher, the
uranium carbide is being formed at a

in 1tself.

much lower temperature than would be
expected. -‘Examination of the broken
graphite with binoculars magnifying
30X shows the presence of red-orange,
green, and white crystals. Inter-
granular penetration of the carbon 1s
obvious. Since the reacted mixtures
smell strongly of acetylene when
exposed, 1t appears that the carbide
compound diffused throughout the melt.

KF-UF;. Attempts to study the
simpler binary system, KF-UF;, showed
that the mixture containing 5 mole %

UF; yvielded breaks in the cooling
curve at 824 and at 711°C, However,
the molten material again diffused

through and reacted with the graphite
crucible to such an extent that further
work on this system could not be
attempted in graphite containers.

NaF-UF,;., Studies are now being
made on the NaF-UF, system. The
molten material in this system does
not diffuse through a graphite crucible.
Cooling curves have been taken on two
mixtures, and in each case two breaks
were observed:

UF3 FIRST BREAK SECOND BREAK
{mole %) (°c) {(°c)
5 agy 588
10 922 588

1952

These few studies suggest that a
eutectic exists at 588°C, but the
location is as yet indefinite., These
studies will be continued.

Preparation of Compounds of UF;.
Attempts have been made to prepare
compounds of UF; of known composition
that would aid in i1dentification of
unknown compounds in phase equilibrium
studies. For this purpose samples
corresponding to NakF-UF;, 2NaF-UF,,
KF-UF,, and ZrF -UF; have been mixed
and added to reaction tubes that were
then evacuated, flushed, filled with
helium, sealed, and heated to 900°C
for about 1 hour. The materials pro-
duced were examined by the technigues
of x-ray diffraction and chemical
Microscopy.

Although the products are not all
completely characterized, it appears
that NaUF, is prepared in reasonable
yvield by heating the first three
mixtures. The samples corresponding
to NaFfiUF3 show some unreacted UFS; a
suggestion of a second compound, but
no NaF. The 2ZNaF-UF, samples show
material that 1is presumably NaUF, with
unreacted NaF, no UF,, and some poorly
de fined crystals that may be an
additional compound.

The KF-UF, sample shows a dark-red
cubic compound with a refractive index
of 1.556 and essentially complete
reaction, as revealed by failure to
detect the starting materials. It 1s
likely that the red material is KUF, .

The UF;-ZrF, sample appeared to be
mossy-green when in large crystals and
mustard~yellow when finely ground, A
new phase is evident, with 15 to 20% of
the UF; remaining and with some ZrF,
and UF,. The refractive index of the
material is in the range 1,56 to 1.58,
Incomplete examination of the material

81
ANP PROJECT QUARTERLY PROGRESS REPORT

does not permit evaluation of the
reaction

UF3 + ZrF4 ey UF4 + ZrFS

or additional
these materials.

complex compounds of

ALKALT FLUOQBORATE SYSTEMS

J. G. Surak R. E.
C. J. Barton

Materials Chemistry Division

Moore

interest in fuel
with melting points and
viscosities lower than those under
consideration for the ABE has prompted
examination of the alkali fluoborate
systems. It is obvious that separation
of boron isotopes would be required to
make such a fuel feasible and that
appreciable pressures of BF; would
exlist over such a fuel in the tempera-
ture range proposed for aircraft
reactors, Since neither of these
obstacles would appear to be 1nsur-
mountable, the study of phase relation-

The long-range
mixtures

ships and decomposition pressures of
some alkali fluoborate systems was
started. To date, only the commercially
available fluoborates of sodium and
potassium have been used in this study.

Figure 43 shows a schematic diagram
of the apparatus employed for measure-
ment of decomposition pressures of
these materials. Graphite liners were
tried 1nitially as containers for the
fluoborates, but they were discarded
1n favor of stainless steel liners
because of the adsorption of gases by
the graphite, Nickel and Monel
thermocouple wells proved unsatis-
factory because of corrosion and
embrittlement i1n a few hours of ex-
posure., Stainless steel wells are,
apparently, quite satisfactory. It
should be noted that the corrosion
described resulted from contact with
commercial NaBF, and KBF,, which are

82

known to contain appreciable amounts
of i1mpurities.

The apparatus used for thermal
analysis of fluoborate systems is the
same as that used for phase studies
with UF,. The preliminary studies have
shown that the techniques and apparatus
are satisfactory., Extension of this
research with pure NaBF, and KBF, and
with mixtures of these materials with
alkali fluorides will be attempted.

Vapor Pressures of Commercial
Fluoborates. The vapor pressures of
the commercial NaBF4<4) and KBF4(S)
and one mixture of these two materials
have been measured, The data may be
expressed by the equation:

log P (mm Hg) = "w5—+ B

°K
Table 5 gives the values for the con-
stants A and B together with the values
reported in the literature, which were
presumably obtained with purer ma-
terials. The vapor pressures observed
with the commercial NaBF, are slightly
lower than the values calculated from
the equation in the literature, The
pressures observed with the commercial
KBF, are much higher than the literature

values. Since the NaBF, contained
about twice as much silicon as the
KBF,, the discrepancy must be due to

other volatile impurities in the KBF,.

Thermal Data for Fluoborate Systems.
Data obtained from cooling curves of
the commercial NaB¥, and some mixtures
are given in Table 6.

The cooling curves obhtained on
NaBF, -KBF, mixtures give no indication
of a eutectic in this system; the
melting points deviate only slightly

— T

General Chemical Co. Code 2240, Tech, Lot

No. G291,
(S)General Chemical Co. Code 2136, Tech. Lot
No. GG63,
FOR PERIOD ENDING SEPTEMBER 10,

1952

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

TO UNCLASSIFIED
ARGON DWG, 163568
TO
VACLIUM
TO PUMP
VACUUM
PUMP
% FULTON 10 !
INSTRUMENT fl;fiifi? " &
U CVALVE ~, 1
RUBBER COPPER po—
‘ggumme //ruTusmm
[‘ \____ . -
THERMOCOUPLE FULTON
TO RECORDER =>i/TEéTRUMENT
— (= VALVE
i sTaINLESS [} (" COPPER KOVAR
STEEL TU&NG { SEAL
TRAP— .‘\i —
(LIQUID Np) N/ T0 w/
COOLING GLASS
TER
WATE STAINLESS
fflmm STEEL J
. REACTOR
————STAINLESS
m— ="y " gTEEL
] LINER
— !
| v
.
HOSKINS ~4
POT | | | S~ GRAPHITE
FURNACE BLOCK

 

 

Fig. 43.

from a straight line connecting the
melting points of NaBF, and KBF,.
The thermal data for the commercial
NaBF, check with the literature values
gquite well, but the melting point of
the commercial KBF, is considerably
higher than the literature value.

Thermal data on the two mixtures of
NaBF4 and NaUF, that have been examined

 

Dissociation Pressure Apparatus.

are rather inconclusive. It appears
that a mixture of 10 mole % NaUF¢ in
NaBF, has a melting point apprec1ab1y
greater than that of the NaBF,

A 95 mole % NaBF4—5 mole % NaUF,

‘mixture was examined with the petro-

In addition to
green, crystal-
amounts than

graphic microscope.
NaBF,, an unidentified,
line phase in larger

83
ANP PROJECT QUARTERLY PROGRESS REPORT

Tablie 5

VAPOR-PRESSURE CONSTANTS FOR ALKALI FLUOBORATES

 

 

 

 

 

 

 

 

 

 

VALUES FOR CONSTANTS
MATERTAL This lLaboratory Literature
A B A B
NaBF, 3820 6.68 3650 6.63
KBF, 6660 9.065 6317 8.15
90 mole % NaBF,-10 mole % KBF, 3870 6.64
80 mole % NaBF,-20 mole % KBF, 3895 6.55
10 mole % NaBF,-90 mole % KBIg 6330 8.99
Table 6

THERMAL DATA FOR FLUOBORATE SYSTEMS

 

 

THERMAL EFFECTS

 

MATERTAL "

Melting Point (°C)

Solid Tramsition (°C)

 

 

General Chemical Co. NaBF,
General Chemical Co. KBF,
90 mole % KBF4—10 mole % NaBF4
80 mole % KBF,-20 mole % NaBF,
70 mole % KBF,-30 mole % NaBF,

60 mole % KBF,-40 mole % NaBF,

 

50 mole % KBF4—50 male % NaBF4

381

563

535

515

498

470

450

 

238

277

242

233

232

215

205

 

would be expected from the amount of
NaUF, added and an unidentified,
colorless, crystalline phase were
present., On the basis of this infor-
mation, the existence of a complex
fluoborate compound containing uranium
may be postulated.

84

Rl

J.

DIFFERENTIAL THERMAL ANALYSIS
J. Sheil

R. E. Traber, Jr.

Barton
Materials Chemistry Division

The

need for a more

sensitive

thermal analysis procedure, especially
FOR PERIOD ENDING SEPTEMBER 10,

for careful examination of important
samples, has been evident for some
time, Two different experimental
arrangements have recently been used
for differential thermal analysis,
with promising results. It is expected
that increasing use of equipment of
this type will be made., FEquipment for
accurate control of heating and cooling
rates will be incorporated as soon as
possible; manual adjustment of these
rates, however, seems to lead to use-
ful information.

For the smaller apparatus, a nickel
block was drilled to accommodate the
1- to 2-g samples, the Al,0, reference
material, and a thermocouple. The
block, the differential thermocouple,
the temperature recording couple, and
the metal tube for introduction of
inert gas are mounted together and
inserted into a ceramic tube within a
vertical l-in. tube furmnace. The
output of the platinum-platinum-rhodium
couples is fed to a preamplifier and
to a Brown recorder.

Tracings were obtained with two
NaF-ZrF, -UF, fuels of the compositions
50-46-4 mole % and 46~50-4 mole %.
The latter curve is shown in Fig., 44,
Neither fuel exhibited thermal effects
in these tracings above the stated
melting points of the fuels. However,
both materials showed thermal effects
in at least two regions at lower
temperatures, |

An apparatus that handles larger
samples and in which the sample is
stirred has also been devised that has
the sensitivity of the differential
method. Although only a few measure-
ments with this apparatus have been
performed, i1t appears that use of
iron-constantan couples fed directly
to a Micromax recorder will be satis-
factory., This apparatus will be used
in the study of difficult systems such
as those containing BeF,.

1952

A

!!! !6359

™

DIFFERENTIAL
TEMPERATURE {mv)
Q —

 

100 2C0 300 400 5C0 B8CO
TEMPERATURE (°C)

Fig. 44. Differential Temperature
Curve of NaF-ZrF,-UF, (46-50-4 mole %)
vs. Al,0, in Nickel-Cup, Pt-Pt, Rh
Thermocouples.

X-RAY STUDIES OF COMPLEX
FLUODRIDE SYSTEMS

P. Agron, Chemistry Division

The binary NaF-ZrF,, NaZrF,-NalUF,
systems, and several systems 1nvolving
the trivalent uranium fluoride and
possibly a reduced form of ZrF, have
been examined on the x-ray spectrometer.
The trivalent uranium fluoride systems
are at an early stage of development,
and their comprehensive treatment will
require further experiments and study.

Fused mixtures analogous in compo-
sition to compounds found in the NaF-
UF, system were examined. Interpre-
tation of their x-ray patterns gave
the crystal structures listed in
Table 7.

Na,ZrF,-Na ,UF,. The structure of
NayZrF, is isomorphous with Na,UF,
(tetragonal; a = 5.45, ¢ = 10.90). It
may safely be inferred from the
similarity of the twocrystal structures
that their solid solutions should
exist. This is confirmed by the
observation that the Na,UF, goes 1into
the lattice of the Na,ZrF, with the
expected expansion of the crystal
structure of the latter. '

85
ANP PROJECT QUARTERLY PROGRESS REPORT

‘Na,Zr¥,-Na ,UF,. The results from
the x-ray study of melts approximating
the formulation of Na,ZrFg, shown in
Table 7, have been grouped together
since detailed diffraction analysis is
lacking at present, Compositions
ranging from 28.5 to 33 mole % ZrF,
show some similarities in their dif-
fraction patterns. At 33 mole % ZrF,,
certain differences 1in the x-ray
pattern appear when samples are prepared
either vnder an inert atmosphere, in a
sealed tube, or in large batches 1n
the hydrofluorinator. At 28 to 29
mole % Zrk,, additional dissimilarities
are evident when the sample is formed
under an inert atmosphere or obtained
commercially. It seems probable that
solid solutions of NayZrF, and NaZrky
in the Na,ZrF, structure are to some
extent responsible for the variations
observed.

NaZr¥,-nNaUF.. The crystal structure
of NaZrFg is also isomorphous with
that of NaUF, (hexagonal; a = 14,72,
c = 9,76), These compounds should
also form solid solutions with each
other, Detailed inforwmation as to
the extent of the mutual solubilities

in the NaZrky-NallF, system 1s of
interest both for the i1nitial loading
and the operation of a reactor using
this type of fuel,

A seriles of compositions corres-
ponding to 4, 8, and 12 mole % NalUFy
was made by melting NaF, ZrF,, and
UF, in their proper concentrations.
In addition, fused mixtures were
prepared from NaZrkFy and NaUF, cor-
responding to 3NaZrF,-NalUF,, NaZrF,-
NaUF,, and NaZr¥;-3NaUF,. The x-ray
diffraction patterns of the above
materials indicate that the distri-
bution of NalUF, in the lattice of
NaZrF, causes a uniform increase with
concentration in its lattice up to a
region lying between 12 and 25 mole %
NaUF,. The diffraction pattern also
indicates that the major phase present
at 25 mole % is the solid solution of
NaUF,; 1in the NaZrF,; crystal. At the
50 mole % composition both the expanded
lattice of NaZrF; and the contracted
lattice of NaUFg are observed in major
concentrations, demonstrating that two
solid solutions are present in the
solidified melt., The pattern for the
NaZrFg-3NalUF, melt shows that the

Table 7

X-RAY ANALYSIS OF NaF-ZrF, COMPOSITIONS

 

 

COMPOSITION CRYSTAL LATTICE Q}MENSIONS CHEMICAL

' T STRUCTURE () FORMULA*
25 mole % ZrF,-15 mole % NaF Tetragonal a = 5.3, ¢ = 10.7 Naj,ZrF,
28.5 mole % ZrF4w71.5 mole % NakF**

33 mole % ZrF,-67 mole % NaF Unknown (Nazszfi)***
50 mole % ZrF,-50 mole % NaF Hexagonal a = 14,1, ¢ = 9.6 NaZrFg

 

 

 

 

.
The analogy with crystal forms observed in the NaF-UF4 dystem is the criterion used for assignment
of formulas,
*re
The range of tompositions probably involve solid solutions with resultant distortions of the
Na,ZrF, lattice.
2 6
e w
A definite compound in this region is supported by thermal analysis.

86
FOR PERITOD ENDING SEPTEMBER 10, 1952

predominant phase 1s the solid so-
lTution of NaZrF; 1n the NalUFy lattice.
In samples with lower concentrations
of NaZrF¢ in NaUFg (made by fusion of
NaF, UF,, and ZrF,), however, the
presence of unreacted UF, would indi-
cate that complete equilibrium had not
been attained. It is apparent from
these studies that the fluoride fuel
NaF-ZrF,-UF, (50-46~4 mole %) may be
represented as a solid solution of
approximately 8 mole % NaUF, in the
lattice of NadrF..

SPECTR@GRQPHIC ANALYSIS 0O0F FLUORIDES
C. R, Baldock

Stable Jsotope Research and Production
Division

The mass spectrometry group 1s now
engaged 1n making both gqualitative and
quantitative examination of reactor
fuels. The qualitative examination 1is
intended to vield only a rough estimate
of the elements and simple compounds
that are volatilized from the fuel
material at temperatures obtainable 1in
the mass spectrometer ovens, Ton
sources currently in use can heat a
sample of 1 to 2 mgtoabout 1500°C, or
larger amounts, up to 50 mg, to about

700°C. A more recently developed
source oven 1s satisfactory up to
2100°C. Tons of the base metals are

easily obtained, and it 1is anticipated
that the temperature will be adeguate
to produce measurable guantities of
ions from such refractory materials as
the oxides of uranium.

The spectrometer has recently been
emploved in the determination of the
contamination of UF, with higher
valence compounds of uwranium that
yield UF; epon heating, A 2- to 3-mg
sample of UF,, which was
making fuel studies, was examined at a
constant temperature a time
sutficiently lomg to vaporize the
sample completely. The continuously

used 1in

for

recorded peak heights were then
corrected for mutual interference of
UF, and UF,, and the resulting in-
tensities were graphically integrated
over the period of observation. The
ratio of these integrated intensities
over the period of observation yields
the percentage of UF¢ in UF,. This
resulted in a value of approximately
1%, which is in close agreement with
the chemical estimate for the sample.
The sensitivity of this type of
analysis is very high, especially when
1isotope dilution is empleyed. The
development of higher temperature ion
sources will greatly expand theutility
of this type of analysis.

Qualitative examination of plug
materials removed from the cold legs
of loops in which NaF-KF-LiF-UF, had
been ci1rculated the known
compenents chromium, iron, and KNa,CrF,

showed

amounts - of
should: be

and also evidence of trace
BFs, SiF,, X, and Hg. Tt
noted that
samples, namely,
KNa,CrF,, were not observed with the
700°C source because the temperature
obtainable has been too low to produce
adequate vaporization. '

the base materials of these

iron, chromium, and

SIMULATED FUEL MIXTURE FOR COLD
CRITICAL EXPERIMENT

{Cuneo

D. R. L. . Overholser

Materials Chemistry Division

The fuel composition for the cold
critical experiment has been based on
a tentative choxrce of NaF-ZrF,-U¥F,
(46-50-4 mole %) for the ARE fuel.
The powder fuel base was therefore
established as 68 wt % Zr0O, and 21. 4
wt % NaF, ‘with 10.6 wt % carbon added
to mock up the moderating effect of
the missing flvoride ions. A large
portion of this work was done in
cooperation with the Y-12 Chemical
Production Division. '

87
ANP PROJECT QUARTERLY PROGRESS REPORT

The chemically pure NaF used con-
tained about 0.,1% water and was used
without drying. The hafnium-free (less
than 50 ppm Hf) Zr0O, contained about
0.2% water, The carbon was ground in
a micropulverizer before mixing. The
carbon powder passed 200-mesh screen
and the Zr0, passed 100-mesh screens
before the blending operation, Blend-
ing of the fuel base was accomplished
in two batches of about 200 1b each
by rolling the mixture in a stainless
steel drum for at least 72 hours.
Chemical analysis for Zr(O,, Na, and C
in samples from various locations in
the drum was used to estimate the
degree of homogeneity. The moisture
content of the final fuel base was
less than 0,3%, well below the tolerance
value of 0, 5%,

About 225 1b of the fuel mixture
was prepared by addition of enriched
UF, to small amounts of the fuel base
to bring the uranium concentration to
10.19 lb/fts, after the fuel was
packed into stainless steel tubes 1.1
in. ID and 40 in. long. The powder
density of 1.8 was uniform in these
tubes. The 70 fuel elements thus
prepared were sealed by screw caps of
stainless steel,

Sixty coolant tubes of stainless
steel 1/2 in, ID and 40 in. long were
packed to a density of 1.8 with the
fuel base, These tubes were sealed by
inserting stainless steel plugs and
then welding. The remaining fuel base
mixture has been retained for use if it
should be necessary to change the fuel
concentration for another trial.

It is still necessary to load the
one experimental tube with a mixture
of NaF-Zr¥F,-UF, of high density and
with proper ratios of Na to Zr to U,
All necessary materials are on hand
for this operation, which will be
accomplished by casting fuel slugs to
fit the tube or by packing the tube

88

with sized powder prepared from a
fused mixture of these materials.,

MODERATOR COOLANT DEVELOPMENT

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

Most of the work done during the
past gquarter has involved the purifi-
cation of sodium and potassium hy-
droxides by methods previously re-
ported, and virtually no experimental
work was done on the development of
new methods of purification.

Production of purified sodium
hydroxide, by removal of Na,CO, from a
50% agqueous solution of NaOH and
subsequent dehydration under vacuum at
450°C, has averaged about 1 lb per
week, The purified hydroxide continues
to be of the same order of purity as
previously reported,¢%¢7) with the
weight per cent of Na,CO, and of H,0
being less than 0.,1.

A batch of lithium hydroxide has
been purified by the method previously
described, (%7 and the purified hy-
droxide was found to contain signifi-
cantly less carbonate and sodium than
batches previously purified.

Several batches of Sr(OH), have
been purified by the method described
and previously used(%7) for both
Sr(OH), and Ba(OH),. The SrCO; content
of the purified Sr(OH), was found to
be 0.2 to 0.3%, which is higher than
the values for batches previously
purified by the method.

The purification of potassium hy-
droxide by removing the carbonate as

 

(6)p. R. Cuneo, E. E. Ketchen, D. E. Nicholson,
and L. G. Overholser, Aircraft Nuclear Propulsicn
Project Quorterly Progress Repoert for Period
Ending March 10, 1952, ORNL-1227, p. 104.

(7)E. E. Ketchen and L. G. Overholser, op. cit,,
OBNL-12%94, p. 90.
FOR

in an agueous medium and de-

BaCO,
hydrating the filtrate under vacuum at
475°C has continued to yield material
containing about 0.2% K,CO0; and an
equivalent amount of barium, The
lowest value for K,CO, attained so
far 1s 0.1%. Because of the continual
high values obtained for carbonate,
this method has been shelved 1n favor
of the method using potassium. This
method, described previously, (7} jin-
volves the interaction of water with
pure potassium and dehydration of the
potassium hydroxide solution.

Four
runs have been made

since a lavger
vacuum dry box became available for
trans ferring the potassium to the
reaction vessel. Approximately 1/2 1b
of potassium hydroxide has been pre-
pared per ‘run, and the K,CO; content
has varied from 0.05 to 0,15%. It
appears possible to prepare potassium
hydroxide containing 0.1% or less of
both K,C0, and H,0 by this method.
The only impurity present in determina-
ble amount in the potassium used 1s
sodium, which occurs to the extent of

0. 2%.

CDOLANT DEVELOPMENT

I. M. Bratcher R, E. Traber, Jr,
C., J. Bartpn -

faterials Chemistry Division

Investigation of nonuranium fluoride
systems of interest as possible coolants
or as a base for preparing fuel mix-
tures has continued. Although ZrF,
was a compounent of a number of the
systems being considered, the field of
interest was broadened by the addition
of BeF, and AlF; to the list of com-
porents., Zinc fluoride received very
little attention., Few of the systems
have been investigated thoroughly
enough to warrant definite conclusions
on their usefulness, but preliminary
tests have not, for the most part,
been very encouraging.

PERTOUD ENDING SEPTEMBER 1¢, 1952

RbF-ALF,. RbAIF (%) and Rb,AlF, (%’
have been described in the literature,
A melting point of 985°C was reported
for the Bb;AlF, compound; the mixture
corresponding to this compound, pre-
pared 10 this laboratory, showed a
melting point of 925°C, The fused
mixture has not been analyzed; so the
reason for the difference is not known
at present. The data obtained on thisg
entire system, although incomplete,
indicate a eutectic melting at 525°C,
Cooling curves for mixtures containing
about 40 mole % AlF, also show a
thermal effect, which 1s as yet un-
explained, at approximately 475°C,
RbAl¥, appears to melt at 450 + 10°C.

NaF-KF-A41F,. Partial phase diagrams
for the binary systems NaF-AlF, and
KF-AlF; are given in the literature.
Meither diagram extends beyond 50 mole
% AlF, because of the volatility of
AlF,. The fact that the KF-AlF, system
was reported to have a eutectic melting
at 565°C led to a study of the ternary
system, No low-melting-point compo-
sitions bave been found among the few
mixtures tested thus far, but the
study 1s continuing, |

NaF-ZrF,-AlF,;. FEven though the use
of two volatile components ina ternary
system seems rather unpromising, a
number of compositions in the NaF-ZrF, ~
AlF; system has been subjected to
thermal analysis., The addition of 5
mole % AlF, to the NaF-ZrF, eutectic
and of 5 mole % Zr¥, to the NaF-AlF,
eutectic apparently raised the melting
point. The cooling curves give no
indication of the existence of a
eutectic melting at lower than 460°C
in this system. |

NaF-Zr¥F,. Na,ZrF; preparedin an HF
atmosphere melted at 625 £ 10°C. The

 

8 ' .
( )X-Ruy Diffraction Patterns, A.S.T.M.
Standards, First Supplement, August 1945, :

(Q)International Critical Tables of Numerical
Data, McGraw-Hill, New York {1926-33}. '

89
ANP PROJECT QUARTERLY PROGRESS REPORT

addition of NaF raised the melting
point of the Na,ZrF, very sharply to a
maximum of 830°C near 25 mole % ZrF,,
which indicates the probable existence
of the compound NazZrF,. The prepa-
ration containing 57 mole % Nal and
43 mole % ZrF, melts at 480°C and
appears to be the lowest melting
composition of this system. This
material expands sufficiently on
solidification to break graphite
crucibles. Neither Na,ZrF, nor NaZrF,
has been found to act this way.

ZrF,-BeF,. Melting points of mix-
tures 1in the ZrF,-BeF, system were of
interest in connection with the NakF-
ZrF,-BeF, system. Three mixtures were
prepared containing 25, 50, and 75
mole % ZrF,. Their melting points
were 650, 740, and 825°C, respectively.
These three points, together with the
melting point of BeF, (543°C), made
almost a straight line. The melting
point of ZrF, has not yet been de-
termined. The extrapolation of the
liguidus curve to 100% ZrF, gives a
value of about 910°C,

HYDROLYSIS AND OXIDATION OF
FUEL MIXTURES

R. P. Metcalf

Materials Chemistry Division

In connection with the oxidation of
fuel mixtures, the reaction of uranium
tetrafluoride with oxygen 1s of
Previous 1nvestigators of
(10, 11) hayve attempted

interest,
this reaction
to estimate the free energy change 1in
the reaction

2UF, + 0, ——> UF, + UO,F, .

S. Fried and N. R. Davidson, The Reaction
of UF4 with Dry 02; A New Synthesis of UFG‘
AECD- 2981 (May 1945},

11)5. S. Kirslis, T, S. McMillan, and H. A.

Bernhardt, The Reaction of Uraniur Tetrafluoride
with Dry Oxygen, K-567 (March 15, 1950),

90

The availability of an alternate method
of estimating the heat of formation of
UO,F, has prompted a new calculation
using an extrapolation of recently
published data on UF, and UQ,F,.

Heat of Solution and Heat of For-
mation of Uranyl Fluoride. For the
calculation of the free energy change
for the reaction, the heat of for-
mation of uranyl fluoride was required,
To aid in estimating this quantity,
the heat of solution of uranyl fluoride
was measured,

A glass Dewar flask was used as the
calorimeter, with 0,5 M perchloric
acid as the solvent., The determi-
nation of the heat capacity of the
calorimeter was based on the known
heat of solution of sodium nitrate,
The result obtained for the heat of
solution was

UO, F, (crystal) + HY — UO,F* + HF,

AH298 = ~-5.3 £ 0.5 kcal/mole .

From an estimate of the heat of
ionization of the complex ionU02F+,(12)
together with known data, the heat of
formation of uranyl fluoride was calcu-
lated to be

U(crystal) + 0,(gas)

+ F,(gas) ——> UOQ,F,{crystal) ,

AN = ~-391,3 + 3 kcal/mole .

208

Free Energy Change for the Reaction

of Uranium Tetrafluoride with Oxygen.
of the thermodynamic
for O, and UF, at high
temperatures, calculated from spec-
troscopic data, are 1n the literature.
Values of the functions for UF, and
UO,F, were estimated by extrapolation

The values
functions

 

(12)

K. A. Kraus, perscnal communication.
FOR PERIOD ENDING SEPTEMBER 10,

of published heat capacity data at
lower temperatures. Results of the
calculations are given in Table 8.

Table 8

STANDARD FREE ENERGY CHANGE AND EQUI-
LIBRIUNM CONSTANT OF THE REACTION

2UF, + 0, —> UF, + UO,F,

 

 

 

T (°K) AF (kcal) K
298.16 -10.7 7 x 107
500 -10.6 4 x 10*
800 -9, 4 4 x 10°
900 -8.4 1 x 107

1000 -7.1 4 x 10

1100 -5.6 1 x 10t

1200 -3.7 5

1300 -1.3 2

 

 

 

For temperatures below 1000°K, the
principal uncertainties in the calcu-
lations probably arise {from the
experimental errors in the heats of
formatien of UF,, UFg, and UO,F,. The
resulting uncertainties in AF are of
the order of +8 kcal,

At temperatures above 1000°K,
uranyl fluoride 1s known to be un-
stable, with one of the principal de-
composition products being uranium
tetrafluoride.('?? Tt has been pre-
dicted that uranium hexafluoride would
also be unstable above this tempera-
ture, (%) Thus, the present calcu-
lations are not expected to have much
significance for temperatures higher
than 1000°K. Subject to these limi-
tatiouns, the calculations confirm the

(IB)F. Vaslow and A. 5. Newton, Chemistry of
Plutonium Report for Period April 10 - May 10,
1944, CK-1498, p. 10-11,

(14)L, Brewer, .. A. Bromley, P. W. Gilles,
and N. L. Lofgren, The Thermodynamic Properties
and Equilibria at High Temperatures of Ureniunm
Helides, Qxides, Nitrides, and Carbides, MDDC-
1543 {(Sept, 20, 1945).

1952

earlier conclusion that the oxidation
proceeds at low temperatures except
where the reaction rate is too slow.
The calculations alse indicate that
AF is somewhat more negative than the
previous estimates and that the
reaction proceeds at temperatures
200 deg higher. However, the dif-
ference is still within the limits of
error of the heats of formation
employed.

FUEL PURIFICATION RESEARCH

F. F. Blankenship F. P. Boody
B. E. Thoma, Jr. C. M., Blood

Materials Chemistry Division

Purified of fluoride
mixtures have been prepared satis-
factorily in a routine manner by the
hydrogenation-hvdrofluorination tech-
nique previously described, (1%)
Purification of small samples of
enriched fuels for radiation-damage
testing has been accomplished by a
modification of this technique.

samples

Purification of pure zirconium
fluoride from oxide-bearing material
by sublimation has been discontinued,
except for special samples, since the
Y-12 production facility 1s now 1in
positicn to supply the pure material,
Preparation of ZrF, in a melt con-
taining NaF by ligquid-phase hydro-
fluorination appears to be very
promising.

Purification of Molten Fluorides.
A previous document®!®) from this
laboratory described the apparatus and
techniques used in the high-temperature
treatment of fluoride melts with
hydrogen and HF. By application of
this procedure on a routine basis, a
total of about 80 kg of purified
fluorides was produced during the

 

(IS)C. M. Blood, F. P. Boody, A, J. Weinberger,
and G. J. Nessle, op. c1t., OBNL-1294, p. 97.

91
ANP PROJECT QUARTERLY PROGRESS REPORT

guarter, This material was, for the
most part, mixtures containing both
UF, and ZrF,, but a few
fuel-solvent materials
possible moderator-coolant mixtures
containing BeF, were included,

uranium-free
and some

Although replacement of the trans-
fer lines and sintered nickel filters

must occasionally be made, the nickel
apparatus has performed 1n a very
satisfactory manner., Results from

this program have justified con-
struction of similar apparatus of the
same size, as well as equipment to
produce 50-1b batches of fuels by this
technique,

Corrosion of the equipment by the
fluorides has
the reactor life,
of some stainless steel sections 1n
the reactor below the liquid led to
constiderable contamination of three
batches of product with 1ron, chromium,
and nickel, Samples prepared in all-
nickel equipment generally show less
than 100-ppm nickel, Careful exami-
nation of the product by chemical
microscopy rarely reveals the presence
of oxides in detectable amounts.

not seriously limited
Accidental inclusion

The removal of sulfur by reduction
of oxidized compounds and evolution of
H,S appears quantitative; chemical
analysis reveals no trace of this
element, Although no analytical data
are availlable, 1t appears likely that
the traces of chloride present in the
ZrF, presently being used are also
removed.

Preparation of Pure Zirconium
Tetrafluoride. FEase of hydrofluori-
nation of Zr0, by gaseous HF at
elevated temperatures 1s a function of
previous history of the Zr0,. Zir-
conium hydroxide may be hydrofluori-
nated quite easily; Zr0O,, which has
been calcined at 800°C or higher,
however, reacts incompletely with HF
at temperatures as high as 700°C even

92

when reaction times of the order of
20 hr are used.{!®) Unfortunately,
the hafnium-free product from the Y-12
production plant is purified by pre-
cipitation with phthallic acid and
must be calcined at high temperature
to remove the last traces of carbon.
Accordingly, plans have been made to
produce the hafnium-free ZrF, by
hydrofluorination of ZrCl, prepared
from an oxide by the Bureau of Mines.

The difficulty in obtaining com-
plete reaction of Zr0O, with gaseous HF
is probably due to a combination of
several factors. Tt is difficult to
obtain good contact of the gas with
the particulate solid in most apparatus.
The reaction product tends to coat the
surface of the unreacted oxide and
prevent reaction. In addition, high
concentrations of HF are achieved only
at elevated pressures.

Jn an attempt to improve the con-
version of ZrQ, to ZrF,, a series of
experiments in which the reaction 1is
performed in the liquid state has been
performed. The Zr0,, which had been
calcined at 800°C, was charged, along
with NaF, into a graphite-lined nickel
apparatus similar to the fuel purifi-
cation apparatus previously described.
The HF was admitted at atmospheric
pressure and the temperature was slowly
raised until liquid phase was present,
Evolution of HF accompanied by water
of reaction occurred at temperatures
above 110°C, and the temperature was
slowly raised over a period of 3 to 4
hr until HF evolution ceased at about
500°C. At this stage, HF was bubbled
through the fused-salt mixture con-
taining suspended Zr0O, while the
temperature was raised to 800°C and
maintained there for 2 to 3 hours. By
this technique, good contact and high
HF concentrations are assured at the

J. L. Williams and B. 5. Weaver, Prepara-
tion of Zirconium Tetrafluoride Progress Report
No. 1, Y-619 (June 15, 1950),.
FOR

low-temperature range. In the high-
temperature range the Zr¥, produced
dissolves in the melt and leaves the
Zr0, free to react.

The optimum conditions for smooth
and complete conversion of the Zr0,
have not yet been demonstrated. Jt
has been demonstrated, however, that
complete conversion of l-kg charges
can be accomplished in 8 to 10 hr and
with three to four times the stoichio-
metric quantity of HF. These experi-
ments are being continued; if the
present optimistic results are verified,
the technique may be extended to
preparation of cryolites and Bel,
mixtures,

Preparation of Pure Aluminaum
¥luoride. Pure, anhydrous AlF,; has
been prepared by dehydration of com-
mercially available AIF,-H,0 under a
flowing atmosphere of HF in nickel
equipment at 300 to 600°C for 6 hours.
Several batches of material, totaling
7 1b of anhydrous product, have been
made available for phase~equilibrium
studies.

REACTION OF FUELS WITH ALKALI METALS

R. J. She1il E. E. XKetchen
N. V. Smith D. C. Hoffman
¥. F. Blankenship

Materials Chemistry Division

P. Agron
Chemistry Division

The reducing action of alkali
metals on UF, and ZrF, is well known,
and under proper conditions the free
heavy metals can be produced readily
by this method. However, when rela-
tively small amounts of alkali metal
are added to fluoride mixtures the
reaction proceeds not to the metal but
to a complex series of intermediate
reduction products, including UF,.

PERIOD ENDING SEPTEMBER 10, 1952

On the basis of available thermo-
dynamic estimates, the reduction of
UF, to UF; is more likely than that of
ZrF, . However, when the UF, is dis-
solved in NaF, where strong com-
plexing may occur, and when a large
excess of ZrF, 1s present, the re-
duction of UF, is not necessarily the
first step in the process. In fact,
preliminary evidence indicates that,
in the presence of excess ZrF,, up
to 40% of the NaK theoretically re-
quired to reduce all the UF, present
may be added without preoducing detecta-
ble amounts of free UF,,

Although much remains to be learned
about the reaction products of NakK
with ZrF,-bearing fuels, the following
conclusions, typicel of material con-
taining 30 mole % Na¥F, 46 mole % ZrF,,
and 4 mole % UF,, are of interest,

Slow heating of sealed capsules
containing the powdered, purified fuel
with a few weight per cent of NaK shows
that the rapid exothermic reaction
takes place at about 250°C, In a
typical case, 1 g of NaK with 50 g of
fuel yielded a sudden temperature in-
crease from 240 to 286°C, The reaction
will, presumably, be very rapid at all
temperatures above 250°C,

Although there is evidence from
other sources that UF, forms compounds
with NaF, KF, and other fluorides, the
bulk ef this material formed such
capsule tests appears on examination
of the cooled melts as pure UF, at the
bottom of the capsules. The solubility
of UF, in the reaction melt at temper-
atures from 600 to 800°C is not vyet
known, but it appears to be relatively
low at temperatures well abeve the
freezing point of the mixture.

in

1f one equivalent of NaK is de-
fined as the guantity required to
reduce the UF, present toUl,, addition

93
ANP PROJECT QUARTERLY PROGRESS REPORT

of less than 0.4 equivalent produces a
yellowish tinge in the green, solidified
melt., The yellow color replaces the
normal green color of the fuel by the
time 1 equivalent has been added.
Segregated UF; in a yellow matrix
results when 2 equivalents are used;
the solid solution of NaUF; in NaZrF,,
typical of the unreacted fuel, is
considerably altered, and a yellow,
fine-grained phase, some of which
forms a cubic crystal of n = 1,476,
has appeared.

When 4 equivalents are added, the
vellow, cubic phase, a colorless cubic
phase of n = 1,430, considerable UF;,
an opaque phase that may be metal, and
Na,ZrF, appear in the melts. When 8
equivalents are added, the yellow
phase and the solid solution are not
detected. Some Na,ZrF. and the opaque
material are present, and all the
uranium seems to be present as UF,.
The material appears to the eye as a
gray-white matrix containing UF,.

Examination of the cooled and ground
melts for production of H, on reaction
with dilute acids shows that the re-
ducing power is much less than expected
from the amount of NaK added. This is
surprising since UF; yields the theo-
retical volume of H, after similar
grinding and handling in air.

X-ray diffraction studies have indi-
cated that prominent but unidentified
lines have disappeared from spectra
run on the same sample at a later date.
It is possible that a low-valence
compound of zirconium that 1is unstable
when handled in air may be responsible
for this behavior.

Although it 1s evident that asnpdden
large leak of NaK at elevated tempera-
tures into the fuel stream would have
serious consequences, it is not evident
that a leak small enough to permit

94

chemical equilibrium to be maintained
would be disastrous,. If reducang
conditions are desirable in the fuel,
1t may be possible, on the basis of
these data, to provide up to 0.2
equivalent of reducing power (fluorine
deficiency) without precipitation of

UFy,. It may be that the corrosive
behavior of these fuels could be
further improved in this fashion., A

considerable effort will be maintained
on this complex problem.

SERVICE FUNCTIONS

J. Truitt C. J. Barton
Materials Chemistry Division

The need for preparation of small
batches of fluoride mixtures for
standards in x-ray diffraction or

petrographic microscope work, for
enriched uranium fuels, and for other
purposes still exists. About 30

batches of fluoride mixtures ranging
in size from 10 g for enriched uranium
fuels to about 200 g were prepared
during the last quarter. In addition,
various containers were filled with
fluoride mixtures, as shown in Table 9,
Fluoride mixtures were removed from
406 corrosion-testing tubes by melting
in an inert atmosphere; the recovered
melts were prepared for analysis by
grinding to a fine powder,

Table 3

CONTAINER-FILLING SERVICES

 

 

 

TYPE OF CONTAINER NO. OF CONTAINERS
FILLED
Cyclotron tubes 59
Corrosion tubes 286
Radiation-damage capsules 5
Heat-capacity capsules 4
Vapor-pressure pots 3

 

 
FOR PERIOD ENDING SEPTEMBER 10,

10.

1952

CORROSION RESEARCH

W. B. Grimes, Materials Chemistry Division

W. D. Manly, Metallurgy Division

H. W. Savage, ANP Division

The emphasis in dynamic corrosion
testing i1in thermal convection loops
has shifted from testing the fuel
NaF-KF-LiF-UF, to testing various
coolants and fuels containing zirconium
fluoride. The coolants and fuels
containing zirconium fluoride are not
so corrosive to Inconel as the NaF-KF-
LiF-UF, fuel. This is probably due
to the more careful preparation and
handling prior to the corrosion tests.
With the zirconium fuels, the corrosion
in the hot leg is approximately 5 mils
after 500 hr of operation compared
with the 10- to 15+-mil depth of
corrosion that was previously ex-
perienced.

The effect of different cleaning
procedures on the corrosive action of
the fluorides on Inconel 1is being
studied. Static tests and modified
dynamic tests have been used to de-
termine the effect of time, tempera-
ture, and various additlives on the
corrosion behavior of the fluorides.
It has been found that most of the
cerrosion occurs in the first 500 hr
of operation. Strange results were
obtained in corrosion testing at
extremely high temperatures; the amount
of corrosion appreciably decreased.
This was probably due to the formation
of a thin film of uranium oxide on
the surface of the test capsule. The
additionof strong getters has markedly
decreased the corrosive action of the
fiuorides. The most striking example
was the addition of zirconium hydride

to the NaPF-KF-LiF-UF, fuel that was

circulated in a thermal convection
loop. Upon examining the thermal
convection loop after 500 hr at 1500 °F,
it was found that the loop suffered
the least amount of attack of any
thermal couvection loop run with the
fluorides; the maximum depth of attack
was 0,5 mil.

Tests of chemical compatibility of
several possible moderator coolants
with the beryllium oxide moderator
material have shown that Zr¥, -bearing
melts are not useful in this connection.
Sodium, NaK, and molten lead appear
to be satisfactory, as do fused salts
containing beryllium fluoride.

Tests on the mass transfer properties
of lead have been conducted 1n which
additions of 3, 5, and 50% sodium
were made to lead. These additions
had a very marked effect on the mass
transfer characteristics and the
corrosive action on the container
materials.

Additional static and dynamic tests
were run in an attempt to find in-
hibitors for the mass transfer and
corrosion exhibited in the metal-
hydroxide systems. Additional tests
have shown that the mass transfer
experienced in nickel-hydroxide
systems can be stopped by the use of
purified hydrogen over the system.

In additien to the program of
empirical corrosion testing, a number
of studies designed to explain the

65
ANP PROJECT QUARTERLY PROGRESS REPORT

corrosion mechanismhas been conducted.
Careful examination by physical and
chemical means has been made of the
corrosion products formed during
large-scale dynamic corrosion tests,
and studies have been made of the
reactions of structural metals with
high-temperature liquids in simple
systems and the possible reactions of
molten fluorides and hydroxides under
applied potentials. A considerable
number of complex fluorides of the
structural metals has been prepared
and characterized to assist in the
identification of corrosion products,
A study of the compounds formed by
the interaction of nickel in the
hydroxides has shown the appearance
of two new compounds, NaNi0O, and
LiN10,. These compounds, which have
been named sodium nickelate{(III) and
lithium nickelate(II1), have been
1identified by chemical analysis for
nickel and the alkali metals; by
several chemical reactions, including
a quantitative determination of the
oxidizing power; by the reaction that
gives a known trivalent nickel com-
pound; and by complete crystal structure
analysis.

PARAMETRIC STUDIES OF FLUORIDE

CORROSION
A. des Brasunas E. E. Hoffman
I.. S. Richardson R. B. Day
D. C. Vreeland I.. D. Dyer

Metallurgy Division

F. Kertesz

Materials Chemistry Division

The efforts to understand and
minimize fluoride corrosion have
logically led to a study of the many
parameters that enter into corrosion
phenomena and corrosion tests, In-
cluded among these parameters are
temperature of test, length of test,
effect of residual stresses in the

96

corrosion specimen, carbon content of
the specimen, effect of additives,
and effect of pretreatment, Each of
these parameters, except for the
carbon content of the specimen,
affected the corrosion in some measur-
able manner. Additives were both
good and bad, pretreatment was bene-
ficial, and corrosion decreased with
increasing time or temperature.

Temperature of Test (A, des Brasunas,

L. S. Richardson, D. C. Vreeland,
E. E. Hoffwman, R. B. Day, and L. D.
Dyer, Metallurgy Division). Inconel

was tested at a series of temperatures
for 200-hr periods in molten fluoride
NaF-KF-LiF-UF, (10.9-43.5-44,5-1.1
mole %). A rather surprising effect
was noted; at the higher temperatures
void formation became less prominent,
as shown by Fig. 45 and by the data
of Table 10.

The analyses of the fluoride baths,
after the tests, were 1n agreement
with the observations given in Table 10;
in the test at 800°C, 1800 ppm chromium
was detected in the bath, whereas the
fluoride of the test at 1300°C yielded
only 65 ppm chromium. The nature of
the film detected on the Inconel
surface of the high-temperature tests
has been identified as U0, by x-ray
diffraction,

Length of Test. A series of tests
was made to determine the extent of
corrosion as a function of time.
Inconel test specimens were prepared
for testing at several time intervals
ranging up to 3000 hr in the seesaw
test apparatus, 1n which the cold
zone was at about 800°C and the hot
zone at about 650°C,

The test data are summarized 1in
Table 11 and plotted in Fig. 46.
These data indicate that most of the
corrosion occurs in the first 500 hr;
beyond this time the additional amount
of attack 1s quite small.
FOR PERIOD ENDING SEPTEMBER 10, 1952

UNCLASS I F IED
v.7183 =

 

in

 

jo>
‘.'

UNCLASSIFIED

UNCLASS I FIED
Y.7184 :

Y.7185

 

£

AN

Fig. 45. Depth of Veid Formation as a Function of Temperature in Inconel
by NaF-KF-LiF-UF,. {(a) 900°C. {(b) 1000°C. (¢) 1100°C. {(d) 1200°C. Un-
etched. 250X '

.97
ANP PROJECT QUARTERLY PROGRESS REPORT

Table 10

CORROSION OF INCONEL BY NaF-gKF-LiF-UF, AS A
FUNCTION OF TEST TEMPERATURE

 

 

 

 

 

© TEST TEMPERATURE DEPTH OF VOID FORMATION THICKNESS CHANGE
(°C) (mils) OF SPECIMEN
800 20 None
900 15 None
1000 10 None
1100 10 None
1200 0.5* None
13060 0.2* None

 

 

 

*Thin film of UO2 observed on metal surface.

Table 13

EFFECT OF TIME ON THE DEPTH OF FLUORIDE ATTACKX ON INCONEL
AS DETERMINED BY THE SEESAW TEST

 

 

METALLOGRAPHIC OBSERVATIONS

 

L
TEMPERATURE ( °C Depth of Void
TIME NO., OF e (‘ml Formation in

 

 

TEST (hr) CYCLES Hot Zone COLD Z20NE
Hot Zone Cold Zone Average Maximum
(mi]s% {mils)
S5F-3 66 18, 500 g12 618 0.5 1 No evidence of
reaction
SSF-4 115 30, 200 780 600 1 2 No evidence of
reaction
SSF-5§ 210 §5, 200 780 630 2 5 No evidence of
reaction
SSF-8 500 132,000 750 610 3 7 Metallic deposit

0.5 mil thick

SSF-8 750 198,000 816 620 3.5 6 Metallic deposit
| 0.5 mil thick

SSF-17 3000 186,000 300 560 4 13 Metallic deposit
0.5 mil thick

 

 

 

 

 

 

 

 

 

98
FOR PERIOD ENDING SEPTEMBER 10,

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

tp—
P
{2 e o e
D J,/
0 ~w§§¢fifl ----------------------------
l‘I" gk Ll e el L et e -
% Q0
o o L .
&
BYERAGE
et
b
500 1000 1500 TTEOGD T 3500 3000
TE (hr)
Fig. 46. Corrosion vs. Time for

Inconel in NaF-KF-LiF-UF, from Seesaw
Tests at 800°C.

Residual Stresses in Specimen.
From quench-annealed flat stock a
series of cold-worked Inconel speci-
mens were prepared that ranged from
0 to T1% reduction 1in thickness.
These specimens were then corrosion
tested in NaF-KF-LiF-UF, for 100 hr
at 815°C to determine possible differ-
ences 1in corrosion behavior. The
test results listed in Table 12 show
that there is a small, but never-
theless perceptible, effect.

1952

There appears to be maximum cor-
rosion in the range of 5 to 25%
reduction in thickness. Such results
are not surprising, since hole for-
mation, which is believed to be caused
by the diffusion of chromium from the
metal into the fluoride bath, is
accelerated by the presence of cold
work. The recrystallization tempera-
ture range for Inconel is about 1500°F,
which 1s the test temperature, and
hence heavily cold-worked specimens
recrystallize in a relatively short
time. Therefore specimens that have
not been cold worked and specimens
that have been heavily cold-worked
may be essentially stress free for
most of the period of the corrosion
test, and the intermediate stress
range may not result in recrystal-
lization throughout the exposure
period.

Carbon Content of Specimen. To
establish the effect of carbon content,
1f any, on the corrosion behavior of
face-centered-cubic metals in NaF-KF-
LiF-UF,, two seesaw test specimens
were run simultaneously so that
identical time and temperatures would

Table 12

VARIATION IN SUBSURFACE VOID FDRMATION AS A RESULT OF COLD WORK

 

 

 

REDUCTION IN THICKNESS | WEIGHT LOSS | DEPTH OF stfn%s]iJsBFACE VOIDS |  mpicKNESS CHANGE
0 0.0020 1 None
5.5 : 0.0043 2 None
11.5 0.0036 1.5 None
26.0 0.0036 1.5 None
50.0 0.0023 1 None
71.0 0.0028 1 None

 

 

 

 

99
ANP PROJECT QUARTERLY PROGRESS REPORT

prevail. One specimen was a regular,
type 304 stainless steel (containing
0.08% C), and the other was an extra-
low-carbon type 304 stainless steel

(0.006% C).

The temperature conditions for both
specimens were (20°C in the hot zone
and 600°C in the cold zone. The
duration of the test was 190 hr,
during which time the specimens were
subjected to 51,150 cycles.

Examination of the specimens after
testing gave 1dentical results,
namely, 5 mils of void formation in
the hot zones of both. No attack or
deposition was detected in the cold
zones.

Corrosion Inhibitors, Static and
dynamic corrosion tests at 816°C for
100 hr have been continued with
various agents being added to check
the possibility of the additions
acting as corrosion inhibitors,
Particular attention was given to
additions of chromium because of the
part chromium plays in the fluoride
corrosion of Inconel (cf., paragraph
on ‘Mechanism of Fluoride Corrosion,”
this section). Tests have been run
with 0.17% additions of sodium (previous
addition tests were with higher
percentages ot sodium) and also ad-
ditions of MnO,, CrF,, NiF,, and FeF,.
In the case of the sodium addition
and also in the previous tests with
additions of lithium, potassium,
calcium, titanium, and manganese, 1t
was thought that these elements might
possibly act as getters to reduce the
available amount of oxygen in the
system and thus help to reduce cor-
rosion. The MnO, additions were made
to determine whether oxygen added to
these tests in this form would in-
crease corrosion,

100

" The metal fluorides (and also
sodium, lithium, and potassium) were
added with the thought that perhaps a
reaction of the following typemight be
the corrosion mechanism:

2¢F + [ — BF, + 2a ,

where a could represent sodium,
lithium, or potassium and S could
represent iron, nickel, or chromium.
If this were the case, then additions
of iron, nickel, or chromium fluorides
and also sodium, lithium, or potassium
should tend to drive the above re-
action to the left and help reduce the
corrosion. However, additions of the
iron, nickel, and chromium fluorides
had no effect in reducing corrosion.
It should be mentioned that the purity
of the salts used is not known. Ad-
ditions of sodium, lithium, and
potassium do seem to reduce corrosion,
but, as mentioned above, their action
may be that of getters rather than
that of participators in the reaction.

In Table 13, the additions (in both
the previous and the most recent
tests) are classified according to the
results of the tests. It should be
emphasized that the classifications
are based on tests that were run with
certain percentages of these addition
agents, and 1t 1s conceivable that
different amounts might result in
different corrosion results, Table 14
shows the results of the most recent
tests in which addition agents were
used. The beneficial effect of the
addition of 1/4% chromium to the
fluoride bath i1s demonstrated in
Fig, 47, which shows the corrosion of
Inconel by NaF-KF-LiF-UF, with and
without the chromium additive. The
absence of subsurface voids when
chromium is added agrees with the
hypothesis that the voids are caused
by chromium depletion of the metal
by the fluorides.
FOR PERIOD ENDING SEPTEMBER 10, 1952

Tab}e 13

RELATIVE EFFECTS OF ADDITION AGENTS ON
CORROSION BY FLUORIDES

 

 

 

 

 

 

AGENTS AGENTS AGENTS WITH LITTLE
REDUCING INCREASING OR NO EFFECT oN
COBROSION CORROSION CORROSION

Na MnO3 CrF,
L1 NiF, NiF, + CrF, + FeF,
K Fel, C
Ca Nal Mo
Ti KI Si
Mn Lii0, Zn
Be KC1 W
Al NaCl Fe
U KBr Zn
Cr Mn
Cu Ag
NaH
Table

Pretreatment. In an attempt to
evaluate the influence of the pre-
treatment procedure, parallel dynamic
tests of material containing NaF-KF-
ZrF,-UF, (36.6-14.0-45.6-3.8 mole %)
were made with samples prepared 1in
graphite and in nickel containers
without hydrofluorination and in
nickel containers with hydrogenation
and hydrofluorination. The material
that had been hydrofluorinated was
somewhat superior, although in these
tests no marked differences were
noted.

In other tests the fluoride mixture
NaF-KF-LiF-U¥F, was treated by bubbling
hydrogen through it for 2 hr while the
molten fluoride was held at a tempera-
ture of approximately 980°C. There
appeared to be no significant re-
duction in corrosion of Inconel,
type 321 stainless steel, or 4 nickel
when the hydrogen-pretreated fluoride
was used.

14

EFFECTS OF VARIOUS ADDITIONS T0 NaF-KF-LiF-UF, IN STATIC CORROSION
TESTS FOR 100 hr AT 816°C IN VACUUM

 

 

 

MATERI AL ADDITION R e METALLOGRAPHIC NOTES
Inconel 0.17% Na No attack
Type 304 stainless steel 0.17% Na No attack
Inconel 10% Cer 4 Subsurface voids
Inconel 10% NifF, 7 Subsurface voids
Inconel 10% Fek, 9 Subsurface voids
Inconel 3 1/3% CrF,
3 1/3% NiF,
3 1/3% Fel, 2 Subsurface voids
Inconel 10% MnO, 9 - Subsurface voids and
intergranular penetration
Type 309 stainless steel 10% MnO, 4 Subsurface voids and
intergranular penetration

 

 

 

 

 

101
ANP PROJECT QUARTERLY PROGRESS REPORT

IR (i ASSIFIED §
I v_7180

 

{0/ . L | - -w—‘k‘- - . ( jn.‘ '. i cfi..s

S UNCLASS I FIED @
e .00 B

 

- ., . . ' - . - - . * T v - a ¢ . : -y
» - . . - i ~: e .\ e L!f e e ) . “ ’ ~ ] . .
« 0 Y - =T - ! - - v T
* . - - . « . - S
CE i , AU S oo TN .
- ' ‘ - . i T - .l : ¢ - .
‘ ’ 3 [ . . - v : - 4 a .
{5/ ¢ oY - . " - - NS L s - .
: .o .- P . e . N . L
» » ’ - a4 Y s P “ g
- - e R . - . . +

Fig. 47. Corrosion of Inconel in Seesaw Tests by NaF-KF-LiF-UF4 With and
Without Added Chromium After 200 hr at 780°C. (a) No chromium added,.
(6) Chromium added, 1/4%. 250X

102
FOR

SEESAW AND STATIC TESTS WITH FLUORIDES
CONTAINING Zr¥,

C. R. Croft N. V. Smith
R. E. Meadows H. J. Buttram

Materials Chemistry Division

Ten different Zr¥F, -bearing fused
salts ranging from 58 to 44 mole %
ZxF,, 0 to 52 mole % NaF, 0 to 46
mole % KF, and 0 to 4 mole % UF4 have
been tested in sealed capsules of
type 316 stainless steel and Inconel
under static and dynamic conditions.
The static tests were maintained for
100 hr at B00°C. The dynamic tests
were performed with the tilting
furnace previously described;(!) the
hot-zone temperature of 800°C and the
cold-zone temperature of 650°C were
maintained for 100 hr, during which
time the fuel made 24,000 cycles.

Comparison of results from the two
test procedures substantiates previous
statements that corrosion, as measured
by penetration, is significantly
greater in the dynamic test and that
mass transfer of small quantities of
metal is often observed in the dynamic
tests. However, even the dynamic
tests are not sufficiently severe to
permit evaluation of small differences
in corrosion behavior.

The zirconium~-bearing fluoride
mixtures were shown to be less cor-
rosive than previously tested NaF-KF-~
LiF-UF4 mixtures. Variation of
corresiveness with composition over
the concentration ranges listed above
could not be ascertained with accuracy.
In general, corrosion in each case
(as 1ndicated by penetration data)
was scattered and the depth of attack
ranged from 0.5 to 2 mils on Inconel,
with slightly higher values for

 

(Dp. c. Vyeeland, R. B. Day, E. £. Hoffman,
and L. D. Dyer, Aircreft Nuclear Propulsion
Project Quarterly Progress Report for Period
Ending March 10, 1952, ORNL-1227, p. 120.

PERIOD ENDING SEPTEMBER 10, 1952

type 316 stainless steel. Weight
changes of the specimens after test
were slight and metal dissolution
in the fuels was less than with fuel
mixtures tested previously. Slight
penetration or surface roughening was
found in the hot zones of the capsules,
and occasional small metallic deposits
were observed in the cold zones. The
corrosion observed appeared independent
of the UF, concentration over the
range studied.

In a series of static exposures of
Inconel to NaF-ZrF,-UF, mixtures for
100 hr at 1000°C the only corrosion
observed was a slight surface deposit
similar to that observed at lower
temperatures. Tests at higher tempera-
tures are under way.

FLUORIDE CORROSION IN THERMAL
CONVECTION LOOGPS

G. M. Adamson, Metallurgy Division

The emphasis in dynamic corrosion
testing has shifted from testing
NaF-KF-LiF-UF, to testing various
coolants containing zirconium fluoride.
As a class, the zirconium fluoride
coolants are less corrosive to Inconel.
The method of attack appears to be the
same for the two types. The zirconium
fluoride coolants do have the dis-
advantage of a high vapor pressure
that results 1n the sublimation of
zirconium fluoride i1n the expansion
pots of the loops.

Some work 1s still being done with
the mixture NaF-KF-LiF-UF, because it
1s more readily available;
1t 1s more corrosive, trends are
easlier to discern. A portion of
the work has been on the addition of
possible 1nhibitors to the coolants.
From the results obtained with a
single loop, zirconium hydride will
reduce corrosion, but the effect of
NaK 1s questionable. Corrosion from

also, since

103
ANP PROJECT QUARTERLY PROGRESS REPORT

the fluorides appears to increase if
crevices are present in the loop walls.

Tables 15 and 1,5, which summarize
the results obtained from the thermal
convection loops operated during this
period, are continuations of Tables 12
and 13 in the previous report,¢?)
Even with the additional fluids being
tested, none of the data in these
tables conflact with the conclusions
presented in the last report.

Fluoride Mixtures Containing ZrF4.
As shown in Table 15, four additional
fluoride mixtures containing zirconium
fluoride have been tested in Inconel
loops. The hot-leg attack 1in every
one of these loops i1s less than that
normally found with NaF-KF-LiF-UF,
fuel. Since loop 234 was cleaned
only by degreasing, which previously
was the standard procedure, it can be
compared with the standard loops.
The maximum attack in this loop was
only 5 mils rather than the 10 to 15
mils normally found with the earlier
fuel composition., The amount of
attack was also greatly reduced, as
shown in Fig. 48. Some of the other
loops in which the zirconium fluoride
fuels and coolants were circulated did
show a maximum attack greater than
10 mils., As discussed below, the
attack seems to be more a function of
the cleaning procedure than of the
mixture.

There does not appear to be much
variation in the attack by the various
zirconium-base mixtures, since all
produced about the same attack under
comparable conditions. The attack
was about the same as that shown 1in
Fig. 54 of the previous report,(3)
which 1llustrated attack by a mixture

 

[2)G. M. Adamson and K. W. Reber, Aircraft
Nuclear Propulsion Project Quarterly Progress
Report for Period Fnding June 10, 1959, ORNL-1294,
p. 112 and 113,

(S)Ibid., p. 111,

104

containing ZrF,. Even though the
number of loops run with these mixtures
1s limited, the attack mechanism
seems to be the same as the process of
leaching and diffusion of chromium,
which was discussed previously(*) in
connection with the NaF-Kb-LiF-UF,
fuel,

The expansion pots of twoe Inconel
loops 1n which NaF-ZrF,-UF, was
circulated for 500 hr were cut open
and over 200 g of sublimed zirconium
fluoride was recovered from each,
The material was found both on the
insulated sides and exposed top of
this chamber. The area of the surface
from which this material escaped was
only 7.4 in.?, and the surface was at
a temperature of about 1300°F. 1In
general, large crystals were found on
the sides and small ones on the top.
This vaporization and subsequent
sublimation of the ZrF, fluoride is a
major problem in the handling of these
mixtures (cf., “Experimental Engi-
neering,” sec. 2).

Effect of Cleaning Loops. The
standard procedure for cleaning thermal
convection loops before operation has
been vapor degreasing and flushing
with a fresh organic solvent. This
procedure was felt to be more desirable
than pickling because of the variables
introduced by pickling, even though
it does not remove the thin oxide
layers at the welds. Heating 1n a
dry-hydrogen atmosphere seemed to be
a possible method for reducing these
oxides without attacking the metal.
This operation was tried on several
Inconel loops in which the zirconium-
bearing fluorides and the earlier
nonzirconium-bearing fluorides were
circulated. The corrosion results
were all comparable, no matter which
fluoride mixture was later circulated.
From this 1t 1s apparent that the

 

(4)G. M. Adamson, A. des Brasunas, and L. S,

Richardson, op. cit., ORNL-1294, p. 126.
S01

Table 135

CORROSION DATA FROM INCONEL THERMAL CONVECTION LOOPS CONTAINING VARIOUS FLUORIDE MIXTURES

 

METALLOGRAPHIC NOTES

 

 

1 E ) s p
. B . TIME OF REASON HOT LEG
1;\%01}' e PLUORIDE ! CTRCELATION FOR TEMPERATURE CHEMICAL NOTES
NO. 1 PROCEDURE MIRTURE (inx) TERMTNATION {°F) ‘Hot Leg Cold Leg
591 | Degreased NaF-KF-ZIF4-UF4 500 Scheduled 15G0 Light to moderate inter- Rough surface; thin noxne Sgme increase in Cr, decrease in
{4.8-50.1-41,3-2.8 mole %} granulaer pictting, 3 to 7 metallic layer Ni, smell increase in Fe, de-
mils crease in U
222 | Degreased | NaF-KF-LiF-UF, 500 Scheduled 1650 Intergranular attack, up Lo Thin deposit Cr snd Fe bolh decreased; ko
{16.9-42.5-44.5-1.1 mole %) 18 wmils; pitting, § mils systematic variations
228 } Hydrogen NaF-KF-Zol «UF, 500 Scheduled 1500 Light intergranular pitting, Rough surface; thin de- Small increase in Cr, decrease in
(4,8-50,1-41.3-3.8 mole &) 3 to 13 mils, average 3§ posited layer Ni, U variahle
mils
224 | Depreased | NaF-KF-Lif-UF, + NeK ! 500 Scheduled 1500 Intergranuiar pitting, up to | Thin surface lever with Cr increased only siightly, Fe
{10.9-43.5-44.5-1.1 mole %)! 10 »ils some adhering nonme- decreased
tullic parcicles
230 | Hydrogen NaF-KF-ZrF, 500 Scheduled 1500 Light intergranular pitting, No deposited layer Gr increessed slightly, Fe de-
{36.0-18,0-46.0 mole %) average 4 mils with maxi- creasad, others vary
mum 7 mils; grains were
not outlined
226 ! Hydruogen NaF-KF-21F, -UF, 500 Scheduled 1500 ¥idely scattered inter- Both metallic apd non- Cr increased, Fe and Ni de-
(36.6-14.G-45,6-3.8 mole %) grannlar pitting, average wetallic deposits creascd, others vary
T omils with maximum of 10
mils
229 | Degreased | NaF-KF-Lif-UF, 500 Scheduled 1500 Moderate to heavy inter- Thin deposit ut least Large increase ia Or, Fe de-
{10.9-43.5-44.5-1. 1 mole %) granular pitting, up te 18 partially nonmetallic creased, small decresse inUand F
Mot treated mils, average 8§ mils
227 | Degreased HaF-RF-LiF.UF, 500 Scheduled 1500 Moderate to heavy pitting, Yery thin metsllic layer | Cr increased, Fe decreased, U
{10.9+43.5-44.5-1. 1 mole %) § to 16 mils; mainly guite uniform
intergranular
234 | Degreased | NaF-ZrF -UF, 480 Pover 1500 Light seartered intergranular! Both mevallic and nen- Cr increased, large decresse in
{46,0-50,0-2.0 mole %} failure pittiag, up to 5 wils, mevallic crystals on Fe and Ni, U increased
average 2 mils; dirty the walls
fnconel used
236 : Hydroger NaF-Zrl, -UF, 462 Pewer 1560 Moderute intergranular pitting, | Both metallic and non- Cr increased, Fe ond Ni de-
{46.0-50.0-4.0 mole %) failure average B mile with maximum of metellic particles on creased; LU increased
10 wils; scme pils 1n grains wall
225 | Degreased | NaF-KF-LiF-UF, + 1,2% ZrH, 500 Scheduled 15090 Very little attack; surface Metallic layer, 9.2 Cr, Fe, U, and Ni all decreased
{19.9-43,5-44,.5-1.1 mole %} rough, few pits, 0.5 mil; mil; layer was guite
possibly some atrtached adherent
i crystals; precipitate
i formed in crystal boundaries
231 | MNaK Nab-ZvF, -UF 500 Scheduled 1500 Moderate general pitting, up { Both metallic and non- Cr increased, Fe, Ni, and U all
{46.0-50.0-4.0 mole R) to 9 pile metallic particles decreased
deposited on wall
237 | Na Naf -ZrF  -UF 560 Scheduled 1500 Tvpical Inconel pitting, Rough surface with both
4 P
{50,0-46.0-4.0 mole %) maximum penetrution 5 mils metallic and non-
with average 3 mils metullic crystals ad-
hering
238 | Nuk NaF-KF-LiF-LUF, 500 Scheduled 1500
(10.9-43.5-44.5-1.1 male %)
223 | NaK Nal-KF-LiF  -UF 500 Scheduled 1500%

 

 

£10.9-43,5-44.5-1.1 mole %)

 

 

 

 

 

 

 

 

“Hot leg emnealed 1/2 hr at 1750°F,
**Lap joint in hot leg.

‘01 HAGWILJES ONIGNd dOI¥dd HOd

c461
901

Table 16

CORROSION DATA FROM STAINLESS STEEL THERMAL CONVECTION LOOPS CONTAINING VARYOUS

FLUORIDE MIXTURES

 

 

i
1

 

 

 

 

 

 

TIME OF

CIRCULATION
{qr)

500

309

igg

62

91

500

43

COF
Loop | (AYPE OF FLLUORIDE
. STAINLESS LRE
0. STEEL MixTl
123 316 NaF-4F-LiF-UF |
(10.9-43.5-44.5-1. 1 mole %);
126 316* | NaF-KF-LiF-UF, i
(10.9-43.5-44.5-1. 1 mole %)
H, treated :
1
i
124 3156 NaF-XF-ZrF, -UF,
{4.8-50,1-41.3-3.8 mole %)
1
120 | 316 Nab-KF-LiF-UF,
(10.9-43.5-44.5-1.1 mole %)
Not treated
125 316 NabF-KF-LiF-UF,
(10.9-43.5-44.5-1.1 moie %)
1238 316 NaF-ZrF, -UF, §
(46.0-50.0-4.0 mole %)
127 ELC NaF-KF-LiF-UF,
i6 {10.9-43,5-44.5-1. 1 mole %)

 

 

 

REASON

METALLOGRAPHIC NOTES

 

 

 

 

 

 

 

 

 

 

 

HOT LEG
] kQH , TEMPE@ATURE Hot Leg i Cold Leg CHEMICAL NOTES
TERM INATION (°F) :
1
Scheculed 1650 Intergranular attack, up to | Metallic layer, 1 mil Fe, Cr, and U all decreased; hot
I 13 mils; some grains re- thick feg highest in all
i moved
Plug 1500 ‘ Heavy intergranular attack, MNonuni form metallic i Cr increased, Fe decreased, U
up to 14 mils; some grains deposit; few non- § varied, but little change, Ni
removad metallic crystals in same
top pertion; slighc
attack in seme areas
l.eak 1500 Intergranular attack, usually | Some evidence of a thin : Cr up slightly, Ni decreased, Fe
S mils but occasionally up metallic deposit i varied but is unchanged, U lower
to 11 mils
Plug 1500 Heavy intergranular attack, Thin surface deposit with| U increased, all other elements
up to 12 mils; some grains both metallic and non- varied
removed metallic crystals ad-
hering
Piug 1650 Heavy intergranular attack, Metallic deposit with ve and Cr varied, Ni decreased
up to il mils some crystals not at-
tached to wall
Scheduled 1500 Heavy intergranular attack, Surface rough; heavy Cr increased, Fe and Ni de-
up to 8 mils; some grains layer of nonmetallic creased, U increased
removed crystals in which some
metallic crystals were
embedded; some non- ;
i metallic crystals were |
i encased in metai ;
i i
Plug 1500 i Heavy intergranular attack, Metallic deposit with i Cr increased, Fe decreased, others

 

up to li mils

 

some metallic crystals
not adhering

varied

 

*Hydrogen fired; hydrogen atmosphere.

LHOdA SSHHI0Ud A THALYVYNO LJAT0Md dNV
FOR PERIOD ENDING SEPTEMBER 10, 1952

 

Fig. 48.

cleaning cycle is just as important
as the choice of fluoride mixture.
Comparison of these loops with similar
loops that have only been degreased
shows that the amount of attack, as
indicated by the number of holes, is
reduced. The attack 1s concentrated
in a few grain boundaries and actually
extends to a greater depth. This is
shown by Fig. 54 in the previous
report.(3) In cleaning with hydrogen,
temperatures in excess of 1850°F are
necessary to provide a minimum temper-
ature above 1750°F. These temperatures
cause grain growth and possibly concen-
tration of impurities in the grain
boundaries, Since hydrogen firing did
cut down the amount of attack, a
method of removing the oxide without

DT e

.
%

. W
"
Jwts
"

.y
. e s gr

Corrosion of Inconel Thermal Convection Loop by NaF-ZrF4-UF4
(46-5%0-4 mole 9) After 500 hr at 1500°F.

Etched with aqua regia. 250X

the resulting grain growth would be
desirable.

It was found that NaK at 1500°F
would remove weld scale from Inconel.
Therefore several Inconel loops were
cleaned by heating the hot leg to
1600 °F and the cold leg to 1500°F and
circulating NaK for 4 hours. The NaK
was drained while still hot, and the
loops were then heated and evacuated
to remove any remaining NaK. The one
loop examined showed attack as deep
as would be obtained by degreasing,
or even deeper. It seems likely that
when the NaK was drained, sodium or
potassium oxides were left behind in
cold spots, or possibly some oxides
were deposited in the cold leg by the

107
ANP PROJECT QUARTERLY PROGRESS REPORT

100°F temperature difference. This
method of cleaning will be checked
further and other methods will be
tried.

Corrosion Inhibiiors. In the
static corrosion studies 1t was shown
that the addition of zirconium metal
to NaF-KF-LiF-UF, reduced the attack.
Because of the great affinity of
zirconium metal for the various gases
and the difficulties i1n obtaining
adequate mixing, 1t was not practical
to add the zirconium metal to a loop
either before or after heating.
Zirconium hydride was also tried as a
possible inhibitor because 1t does
not decompose until slightly below
operating temperature, and the nascent
hydrogen liberated during decomposition
might helpin cleaningup the fluorides.

 

Fig. 49.

with 1/2% Added ZrH, After 500 hr at 1500°F.

108

The NaF-KF-LiF-UF, with 1/2% ZrH,
added was circulated in an Inconel
loop for 500 hr at 1500°F. After
circulation, the hot leg was found to
have a roughened surface but the
attack extended only to a depth of
0.5 mil; the surface 1s shown 1n
Fig. 49. The attack was the least
found in any loop in which fluorides
have been circulated. 1In the hot
leg, some precipitated crystals were
observed just under the surface and
in the grain boundaries. These
crystals do not seem to be continuous
enough to act as a protective layer.
An adherent metallic layer about 0.2
mil thick was found in the cold leg.
The lack of corrosion in this loop
was further evidenced by the very low
amounts of impurities in the fluorides.
This test is now being repeated, and

IPHOTO T- 1764

.
s
Voo L ; .

Corrosion of Inconel Thermal Convection Loop by NaF-KF-LiF-UF4

Etched with aqua regia. 250X
FOR PERIOD ENDING SEPTEMBER 10,

Tt

the effect
addition to a zirconium-base fluoride
fuel is also being studied. *

It was pointed out in the previous
report{3) that a small NaK addition
reduced the corrosion obtained when
neonuranium-bearing coolants were used.
As an extension of this work, NaK was
added to two other Inconel loops.
l.oop 224 was used to circulate Nal-KF-
LiF-UF,, and Loop 241 to circulate
NaF-ZrF, -UF,. About 8 cc of NaK was
added to each loop. In both loops
the NaK was placed in the pot and the
fluorides were added to 1t. Loop 224
was started without difficulty and
operated satisfactorily for 500 hours.
The maximum hot-leg attack measured
10 mils, which is lower than the
usual average but not enough to show
a definite effect. However, since the
metallic impurities found in the
fluorides after circulation were also
Jower than normal, it seems possible
that some actual improvement took
place. Loop 241 1s now operating
satisfactorily, but there was con-
siderable difficulty 1in start-up,
which may have been due to a reaction
that yielded some solid precipitates.
The cold-leg temperatures were lower
than normal and kept decreasing.

 

(S)Adamson and Reber, Ibzd., p. 109,

1952

Another Inconel loop, in which chromium
metal powder was added to the circu-
lating NaF-KF-Li¥F-UF, mixture, 1s
operating normally.

Crevice Corrosion. Crevice cor-
rosion has been observed but requires
further 1nvestigation. In an Inconel
pump loop operated with NaF-KF-LiF
€11.5-42.0-46.5 mole %) very little
attack was found in most of the loop,
but heavy attack was found 1n two
different types of crevices, One type
of ecrevice resulted from sleeves
joining pipe sections and the other
from a thermocouple well. This loop
had operated for about 1000 hr at
1250 °F and below and 100 hr at 1500°F.

Several other loops were then
checked for crevice corrosion by
examining the top welds. The corrosion
of crevices found i1n the hot legs of
loops 219, 229, and 234 1s compared 1in
Table 17 with the corrosion of the
wall directly opposite the crevice.
Figure 50 shows a section through the
joint of Inconel loop 219. It should
be pointed out that the attack in
this portion of the hot leg 1s con-
siderably less than that given in
Table 15, since the sections were at
slightly different temperatures. One
additional piece of evidence for the
existence of crevice corrosion 1s the
fact that the attack at the mouth of

Table 17

CREVICE CORROSION IN THERMAL CONVECTION LOOPS

 

 

 

 

 

 

METALLOGRAPHIC NOTES

 

LOOP MAXIMUM ATTACK (mils)
FLUORIDE MIXTURE :
NO. Hot Leg Crevice
219 | NaF-KF-LiF-UF, | 3 6
229 NaF-KF-LiF-UF, | = 4 6
234 | NaF-ZrF-UF, 3 5

 

 

 

 

 

Two to three times more attack in the crevice
Many times more attack in the crevice

Very little attack in hot leg with moderate
amount in the crevice; tendency for attack to
follow grain boundaries

 

109
ANP

PROJECT QUARTERLY PROGRESS REPORT

5 Ll W o M . . .-
3 L kT RA] A / F
t VRS dfi%‘;i;'f : f;\x L r"Pdor T-1802 °
P e RN R T
vl } o o - L4

  

'5’
¢ q‘-u‘
o ‘ ‘.. 1. i.
’i‘ "i "‘ ‘..‘ 0 \, "‘0J
: sV of _5%; flflkfi?
. ‘,,h:i ™ -u Y f"(o _
CREVWE

Ng"- "i,f .
) * g *fi A ..
re .- ‘ K.‘.;‘gf*;
::_ - @ '# ‘ o
L% ‘5 ° 4 .
b
RN
-* "'\w ,".f..
* .‘ ,‘i:ayn-,
.
- »
Fig. 50.

After 500 hr at 1500°F.

the crevice 1s less than that found
within. The mechanism of this attack
is not yet known. Further tests will
be made to determine just how serious
this effect will be.

Temperature Variations, To de-
termine the effect of temperature
on the corrosion of Inconel, two
loops (218 and 222) were run at 1300
and 1650°F, respectively. It
apparent that the attack increased
slightly with temperature. The type
of attack was the same at each temper-
ature,

was

A type 316 stainless steel
(123) was operated
NaF-KF-LiF-UF,.
300-series

loop
at 1650°F with
This was the first
stainless steel loop to

110

   
     
    

*’fw//‘”w

Effect of a Crevice on Corrosion of Inconel by NaF-KF-LiF- UF ,
Etched with aqua regia.

250X

operate 500 hr without plugging.
The attack in the hot leg was inter-
granular and extended to a depth of
13 mils. The test was repeated
(loop 125), but the loop plugged in
only 91 hours. The insulation in the
second loop was not so good as in the
first and the cold leg operated at a
slightly lower temperature. Further
work 1s necessary to determine whether
1t could be the small temperature
difference that caused the plugging.

Variations in Alloy Composition of
the Wwalls. Tables 18 and 19 present
the results of chemical analyses of
successive layers drilled from the
inside of two loops (78 and 121) for
comparison with similar data obtained

from two other loops (210 and 219) and
FOR PERIOD ENDING SEPTEMBER 10, 1952

Table 18

CHEMICAL ANALYSES OF SUBSURFACE LAYERS OF INCONEL WALLS OF LOOP 78 IN WHICH
: NaF-KF-Li¥F WAS CIRCULATED

 

 

 

 

DEFTH OF | COVPOSITION (%)
LAYER : : :
(mils) Fe | N | Cr | Fe/Ni | KF* | LiF* | NeF* | Other nga;njgé Total
5 7.03 | 78.3 | 12.0 | 0.089 | 0.09 | 0.3 | 0.07 | 1.68%* - 97.4 99.5
10 8.66 | 77.5 | 11.8 | 0.112 | 0.03 | 0.15 o.o? 1. 46 - 98.0 99.6
15 6.93 | 77.6 | 13.6 | 0.090 | 0.02 | 0.1 | 0.02 | 1.18 98.1 99. 4
20 6.81 | 77.7 | 13.7 | 0.088 | 0.02 | 0.1 | 0.02 | 1.30 98.3 99.6
25 6.66 | 77.9 | 14.3 | 0.086 | 0.06 | 0.2 0.07 | 1.78 - 98.9 101.§
Exterior | 7.15 | 77.4 | 15.2 | 0.092 : 1. 48%ss 99,8 161.0

 

 

 

 

 

 

 

 

 

 

 

These values calculated from spectrographic analysis of metals by assuming them to be present as
fleorides.

- : :
Total of spectrographic analysis for Al, Co, Cu, Mn, Mo, 51, and Ti.

5 :
Average of values above.:

Tabie 19

CHEMICAL ANALYSES OF SUBSURFACE LAYERS OF TYPE 316 STAINLESS STEEL WALLS
OF LOOP 121 IN WHICH NaF-KF-LiF-UF, WAS CIRCULATED

 

 

 

 

DEPTH OF | :  COMPOSITION (%)
LAYER . - e o Total Fe, ‘
(mils) Fe Ni .Cr Mo KF* Lif Nak* | Other Ni, Cr, Mo Total
2 60.4 | 16.9 | 14.5 | 2.8 | 1.5 | 3.7 | 1.1 | Avg. 1.6 94.6 102.5
= : Co, Cu, : -
5 60.8 | 16.1 | 147 | 2.7 | 0.9 | 3.7 1.1 Mn, Si 95.2 101.6
10 63.5 | 14.8 | 15.1 | 2.5 | 0.5 | 2.2 | 0.5 95,9 160.7
15 64.1 | 14.6 | 15.7 | 2.5 | 0.3 | 1.1 | 0.5 96.9 100. 4
20 0 65.2 | 14.3 | 16.2 | 2.5 | 0.2 | 0.7 | 0.5 98.2 101.1
25 65.1 | 14.0 | 16.8 | 2.4 | 0.3 | 0.7 | 0.5 - 99.3 101. 4
Exterior | 66.0 | 13.3 | 16.8 | 2.4 | | 98.5 100.1

 

 

 

 

 

 

 

 

 

 

 

* : :
These values calculated from spectrographic analysis of metals by assuming them to be present as
fluorides. ‘ :

111
ANP PROJECT QUARTERLY PROGRESS REPORT

presented in the previous report, (%)
[.oop 78 was constructed of Inconel
and circulated NaF-KF-LiF for 500
hours. In the first loop{®’ the
variations 1n chemical composition
appeared to be the same as those
found in loop 78, but in loop 78 they
were much smaller. Chromium was
depleted to a depth of at least 25 mils,
Since the fluoride content was low and
constant, 1t was evident that there
was practically no penetration.

In the type 316 stainless steel
loop, chromium was reduced slightly,
and the iron was the major material
leached out. The nickel and molybdenum

contents were essentially unchanged.

In this loop the fluoride content was
much higher than in the other loops
and decreased gradually with depth.
Evidently the fluorides are trapped
in crevices. A loop in which zirconium-
bearing fluoride mixtures will be
circulated 1s being prepared.

Postrun Examination of Fuels (D, C.
Hoffman, F. F. Blankenship, Materials
Chemistry Division). Examination of
sections from a number of thermal
convection loops operated by the
experimental engineering group 1n
which zirconium-bearing fluoride
fuels were circulated in loops of
type 316 stainless steel and Inconel
has been accomplished by the method
and techniques previously described.(7?
The results of these investigations
are briefly summarizedin the following
statements.

The zirconium-bearing fluoride
fuels show considerably less corrosion
transfer than the fuels
tested previously. No NaK,-CrF_ has
been discovered in the loops run with
these fuels. Inconel seems generally
superior to stainless steel.

and mass

 

(6)rpia., p. 115, Table 1f.

(7)[). C. Hoffman and F. F.
op, cit,, ORNL-1294, p. 121.

Bl ankenship,

112

In loops of type 316 stainless
steel, a small amount of magnetic
material showing an x-ray pattern
resembling an iron-chromium alloy
appears 1in and above the trap at the
bottom of the cold leg.

Cne Inconel loop that had been
hydrogen fired showed somewhat more
mass-transferred metal in the cold leg
than a similar section of an Inconel
loop that had simply been degreased.
The metal gave an x-ray diffraction
pattern for iron i1in which the lines
were shifted slightly.

In one type 316 stainless steel
loop in which a mixture containing
NabF-ZrbF,-UF, (46-50-4 mole %) was
circulated for 500 hr, a brown un-
known material (average index of
refraction = 1.556) was found in
the vertical hot section. A con-
siderable amount of metallic deposit
was found in the trap of this loop.
Identification of the deposit is
1ncomplete.

COMPATIBILITY OF BERYLLI1IUM OXIDE
WITH VARIOUS FLUORIDE MIXTURES

R. E. Meadows

Materials Chemistry Division

H. J. Buttram

In the present ARE design a liquid
heat transfer agent will be pumped
slowly through the moderator space to
maintain a uniform temperature in the
beryllium oxide. The flow rate of
this liquid past the beryllium oxide
blocks should be at most a few feet
per minute. The number and irregular
shapes of the beryllium oxide speci-
mens would seem to preclude canning
or electroplating of the pieces., One
of the essential characteristics of
the moderator coolant therefore is
chemical compatibility with the
beryllium oxide.
FOR PERIOD ENDING SEPTEMBER 10, 1952

Small specimens sawed from hot-
pressed beryllium oxide blocks typical
of those proposed for use in the ARE
have been used in these studies.
These pieces have been exposed for
100 hr in sealed capsules of type 316
stainless steel to molten alkali
fluorides with and without ZrF, or
Bng.(s) ‘In all static tests the
temperature was maintained at B800°C;
in all the:dynamic capsule tests the
capsule was tilted every 15 sec, with
the cold end at 650°C and the hot end
at 800°C. ’ :

Preliminary studies of the bebavior
of beryllium oxide in a static bath
of molten NaF~KF'ZrF4 (5-52-43 mole %)
indicated that the rate of reaction
was appreciable but that i1t decreased
with time. JIdentification by x-ray
diffraction and petrographic microscopy
of a film of ZrO, on the beryllium

 

(B}Addibional beryilium oxide compatibility
vests, with Na, NaK, and Pb, are reported in
this section under * lLiquid Metal Corrosion.”

oxide blocks suggested thata protective
film of ZrO, might make it possible
to use a coolant containing zirconium.
Exposure of the blocks for longer
times in static tests and for 100-hr
intervals in dynamic tests, however,
clearly demonstrated that zirconium-
bearing mixtures are not suitable in
contact with beryllium oxade. llata
shown in Table 20 are representative
of those obtained. The attack seems
to be due to a uniform dissolution of
beryllium oxide, since the corners
and sharp edges of the specimens
retained.

are

However, as the data in Table 20
indicate, systems of alkali fluorides
containing Bel, are compatible with
beryllium oxide. Microscopic and
x-ray examination of specimens crushed
afrer dynamic testing show BelF, in the
body of the beryllium eoxide block.
The penetratien, which appears to do
no damage, is undoubtedly responsible
for the weight gain observed. Further
studiesof the Be0-BeF,-alkali fluoride
systems are being made.

Table 20

REACTI0N OF BERYLLIUM OXIDE WITH FUSED SALTS CONTAINING ZfF4 O/ Bek,
UNDER DYNAMIC CONDITXIONS

 

 

 

 

 

. WEIGHT CHANGE
COOLANT B . 5 REMARKS
: In % In mg/dm®/day
NaF-KF-ZrF, ~66 ~5879 Specimen in hot end of tube
(5-52-43 mole %) | | |
. -16 ~-1432 Specimen in cold end of tube
~60 -4907 Specimen dissolved uniformly®
NaF-KF-BeF, +4.1 +362 Specimen appeared unchanged
(30-5-65 mole %) 4 fi : :
: 6,9 +594 Salt apparently penstrates blocks

 

 

 

 

Original dimensions, 0.508 by 0.253 in.; final dimensions, 0.410 by 0.176 in.
ANP PROJECT QUARTERLY PROGRESS

HYDROXIDE CORROSION

F. Kertesz
Materials Chemistry Division

A. des Brasunas E. E. Hoffman
G. P. Smith

Metallurgy Division

The effectiveness of a hydrogen
atmosphere in inhibiting hydroxide
corrosion was demonstrated previ-
ously(?) in at least two independent
tests. Recent tests 1n a thermal
gradient apparatus have substantiated
the results. Furthermore, addition of
the reducing agent, sodium hydride,
has reduced mass transfer 1n an
Inconel system tested with potassium
hydroxide. A series of temperature-
dependence tests indicated that the
corrosionof Inconel by sodium hydroxide
is negligible up to 450°C; more ex-
tensive tests at 600°C gave evidence
of severe attack. Corrosion tests of
a stressed specimen 1indicated in-
creased corrosion,

Temperature of Test. A series of
temperature-dependence testsof Inconel
in sodium hydroxide has been completed.
From the results listed in Table 21,
it would appear that in 100-hr tests,
corrosion of Inconel 1s negligible
unti]l the test temperature 1s raised

to above 450°C.

 

(Q)Brasunas and Richardson, op. cit., p. 116.

REPORT

Nickel was tested similarly 1in
sodium hydroxide at a series of hot-
zone temperatures ranging from about
400 to 800°C. This was done in the
hope of finding a threshold tempera-
ture below which no mass transfer
occurs. However, the results indicated
that as the maximum temperature was

diminished, the amount of nickel
crystals in the cold zone was di-
minished; mass transfer was not

eliminated.

More extensive corrosion data are
available at 600°C, at which tempera-
ture a variety of materials was
subjected to 100-hr exposures to
sodium hydroxide. These data are
summari zed in Table 22. Crystal-bar
zirconium was heavily attacked even
at this temperature, although it 1s
possible that the canning in type 347
stainless steel may have adversely
affected its behavior. Of the alloys
tested, Monel showed the least cor-
rosion. It 1s obvious that the use of
these alloys with sodium hydroxide
under an inert atmosphere 1s not
possible.

Residual Stresses in
Tests have been run to check the
effect of residual stresses on the
corrosion resistance of Inconel 1n
sodium hydroxide at 816°C. The stresses
were induced by cold rolling. The
tests were first run for 50 hr, but

Specimen,

Table 21

TEMPERATURE-DEPENDENCE TESTS OF INCONEL IN SODIUM HYDPROXIDE FOR 100 HOURS

 

 

TEMPERATURE OF TEST WEIGHT CHANGE
o .2 METALLOGRAPHIC NOTES
(°C) (g/in.%)
350 None No attack
450 None No attack
550 +03. 0050 3 to 4 mils of intergranular penetration
593 +0.0060 3 to 4 mils of intergranular penetration

 

 

 

114
FOR PERIOD ENDING SEPTEMBER 10, 1952

Table 22

STATIU CORROSION TESTING MATERIALS Ifl SODIUM HYDROXYIDE AT 606°C FOR 100 H@URS

 

 

 

- WEIGHT CHANGE. -
Crystal-bar zirconium _ ~1199 Specimen retained its original shape and
{14.5% of specimen) appeared to be uniformly dissolved
~8787 Specimen retained its original shape and
(88.4% of specimen) appeared to be uniformly dissolved
Inconel +459 Surface oxidized to depth of 6 to 7 mils
with a very thin layer on outside of
attack; attack proceeded intergranularly
Type 316 stainless steel -79 Surface oxidized to T mils deep; attack
proceeded evenly
Type 321 stainless steel -182 Surface oxidized te 6 mils deep; attack
proceeded intergranularly
+13 Surface oxidized to 6 mils deep; sttack
proceeded intergranularly
Menel +8 Light intergranular attack 1 to 2.5 mils
deep
+10 Light intergranular attack 1 to 2.5 mils
deep '
Hastelloy B ~350 General pitting, 4 mils deep; specimen
very brittle

 

 

 

the specimens were all corroded
throughout their entire thickness.
The time was then cut to 8 hours.
The results of the 8-hr tests, listed
in Table 23, seem to indicate that
residual stresses of this nature
do have a deleterious effect on the
corrosion resistance of Tnconel 1in
sodium hydroxide. Similar tests were
run on A nickel for 50 hr, but there
was no evidence of attack even on the
most severely cold worked (71% re-
duction in area) specilmens tested.

Corrosion Inhibitors. Inconel was
tested in potassium hydroxide with a
2% NaH addition. Very little evidence
of mass transfer was evident after
100 hr (51,200 cycles) with a hot-zone
temperature of 750°C and a cold-zone
temperature of 550°C. The tempera-
tures were measured by spot-welded
external thermocouples.

Hydrogen Aimosphere. Earlier(1!?)

 

(10)e p. Swith, J. V. Cotheart, snd W, H.
Bridges, op. cit,, OBNL-1227, p. 127.

115
ANP PROJECT QUARTERLY PROGRESS REPORT

Table 23

EFFECT OF RESIDUAL STRESSES ON CORROSION OF INCONEL BY
SODIUM HYDROXIDE FOR 8 hr AT 816°C

 

 

 

 

 

 

 

REDUCTION IN ORIGINAL THICKNESS WEIGHT CHANGE , .
THICKNESS (%) (mils) (&/in.?) METALLOGRAPHIC NOTES
71 10 -0.0011 Attacked throughout
50 17.5 +0.0088 9 mils of penetration,
2 to 3 mils of unattacked
material
26 26 10.0188 7 to 8 mils of penetration,
12 nils of unattacked
material
11 31 10.0171 4 to § wils of penetration,
23 mils of unattacked
material
5 33 10,0203 3 to 5 mils of penetration,
25 mils of unattacked
material
0 35 +.,0210 3 to 5 mils of penetration,
27 mils of unattacked
material
studies of mass transfer with small melt. 1In 40 hr a copilious quantity of

thermal convection loops 1indicated
that a hydrogen atmosphere had an
appreciable effect i1in reducing nickel
mass transfer. This effect was more
closely investigated with the thermal
convection apparatus described in the
last report.{?!) A number of ex-
periments has been performed with
this apparatus. The results are
1llustrated by two typical tests.
In both tests the temperature of the
nickel cup was about 725°C, whereas
the temperature of the inner nickel
tube was 525°C, a temperature difference
of 200 degrees.

The first test was conducted with
a static vacuum over the hydroxide

 

(ll)J. V. Cathgart, W, H., Bridges, and G. P,
Smith, op., cit,, ORNL-1294, p. 118,

116

dendritic nickel crystalshad encrusted
the itnner nickel tube and extended
more than half way across the hydroxide
melt. This 1s a case of normally
expected nickel mass transferin sodium
hydroxide {for a large temperature
difference. In the second test a
purified hydrogen atmosphere replaced
the static vacuum. In 170 hr, as
compared with 40 hr for the first
test, the only evidence of mass transfer
obtained was a tiny band of nickel
crystals on the inner nickel tube at
the gas-liquid interface. This band
was about 1 mm wide and 0.1 mm thick.

Confirmatory evidence for the
inhibiting effect of hydrogen was
also obtained with the tilting-furnace
apparatus. The test with nickel and
FOR

sodium hydroxide at a hot-zone temper-
ature of about 750°C and a cold-zone
temperature of 600°C gave very little
evidence of mass transfer after a
100~hr test when a hydrogen atmosphere
of about 1 atm was maintained. This
represents a very appreciable re-
duction 1n mass transfer when compared

with tests made under similar con-
ditions in vacuum.
LIQUID METAL CORROSION
A, des Brasunas G. P. Smith

D. C., Vreeland

Metallurgy Division

F. Kertesz

Materials Chemistry Division

The desirability of lead as a
coolant is largely offset by 1ts
susceptibility to mass transfer and
1ts attack on most commercially
available heat-resistant alloys.
The use of a lead-sodium mixture
would be advantageous i1f 1t possessed
predominately the corrosion character-
1stics of the sodium and the radio-
active characteristics of the lead.
Tests indicate that the corrosiveness
of lead 15 diminished with the ad-
dition of increasing amounts of
sodium. In other corrosion tests
with lead, the mass transfer phenomenon
and the effect of temperature have
been examined further. Sodium cor-
rosion was measured in the “spinner”
test 1n a new, revolving, corrosion
apparatus. Testsof type 1020 chromized
steel 1in various low-melting-point
alloys show that this steel cannot be
recommended as a container for the
alloys tested. In addition, the
compatibility of beryllium oxide with
such probable moderator coolants as
lead, sodium, and NaK i1s being in-
vestigated.

PERIOD ENDING SEPTEMBER 10, 1952

Mass Transfer in Liguid Lead.
Studies are being undertaken of the
mass transfer of molybdenum and Inconel
in liquid lead. Preliminary to the
corrosion studies, research has been
conducted on the handling and purifi-
cation of liguid lead, and a few mass
transfer measurements have been made.
Since molvbdenum 1s difficult to weld
and oxidizes excessively at high
temperatures, the molybdenum corrosion
tests were conducted on specimens of
the metal in convection loops made
of quartz or pyrex. These materials
are resistant to corrosion by lead(!?)
and do not present the problem of a
bimetallic system,

Studies were made of the de-
oxidation of high-purity lead. From
the quantity of oxide that collected
on the surface of lead melted under
purified hydrogen, it was concluded
that the original material had con-
tained a considerable amount of
occluded oxide. A number of ex-
periments was performed in which
hydrogen was passed in fine bubbles
formed by a fritted pyrex disk through
molten lead at 450 to 500°C for 20
hr, and the lead was then filtered
through a fritted pyrex disk. There
were no visible oxide films on the
surface of the filtered lead, but a
small amount of oxide was found on
the filter. This indicated that the
hydrogen reduction had not been
complete. Further studies of the
conditions required for complete
reduction by hydrogen will be made.

No results of the corrosion of
either molybdenum or Inconel in the
convection loops have yet been ob-
tained because of difficulties as-
sociated with the operation of such

 

12 ' '
( )R. N. Lyon {(ed.), Liquid Yetals Handbook,
NAVEXOS P~T733, p. 105 {(June 1, 1950},

117
ANP PROJECT QUARTERLY PROGRESS REPORT

loops. However, a preliminary ex-
periment was performed to determine
whether there 1s mass transfer of
molybdenum during temperature cycling
in the manner reported{!'3) for copper
in bismuth, A mechanically polished
molybdenum specimen, contained in lead,
was cycled every 12 min for 42 hr
between 500 and 510°C. At the end of
the test period the lead was filtered
at 500°C. No mass transfer crystals
were found. This result should be
taken as preliminary because the
molybdenum was undoubtedly coated with
an oxide film that may have prevented
contact with the molten lead.

 

(IS)G. P. Smith, and J. V. Cathcart, Metallurgy
Division Quarterly Progress Report for Period
Ending Jenuary 31, 1952, OBNL-1267, p. BRS.

Effect of Tempevature on Lead
Corrosion. Temperature-dependence
tests of several materials i1n molten
lead have been run. The only attack
that was noted 1n these tests was at

700°C and above. The results of the

tests, which were run for 100-hr
periods, are given in Table 24.
Lead-Sodium Mixtures, Seesaw

corrosion testsof lead-sodium mixtures
were very encouraging 1n that both
corrositon and mass transfer were
minimized. The tests in type 310
stainless steel were made with ad-
ditions of 0, 3, and 50 wt % sodium to
the lead. The test capsules had a
hot-leg temperature of 780°C and a
cold-leg temperature of 475°C, The

Table 24

TEMPERATURE-DEPENDENCE TESTS OF VARIOUS MATERIALS IN MOLTEN LEAD

 

 

 

 

 

TEMP(%P::A;TURE MATERI AL METALLOGRAPHIC NOTES
400 Inconel No attack
Type 446 stainless steel No attack
Type 309 stainless steel No attack
500 Inconel No attack
Type 446 stainless steel No attack
Type 309 stainless steel No attack
600 Inconel No attack
Type 446 stainless steel No attack
Type 309 stainless steel No attack
700 Inconel Specimen attacked to about
1/2 mil
é Type 446 stainless steel No attack
| Type 309 stainless steel No attack
\
816 Inconel Specimen attacked 5 to 9 mils
Type 446 stainless steel Slight roughening of surface

 

 

118
FOR

tests were for 190-hr periods, during
which time the capsules were cycled
51,000 times. '

Examination of the specimens after

test indicated: {1) the usual mass
transfer 1n the cold zone of the
sodium-free test, (2) reduced mass

transfer when 3% sodium was added to
the lead, (3) no mass transfer when
50% sodium-50% lead was used. The
attack in the hot 2zone was reduced
from 6 mils in the case of the sodium-
free lead ©to 1 mil in the case of the

50% sodium~50% lead mixture.

Spinner Tests with Sodium. The
spinner test, designed to supply
what might be called *“high-velocity”
corrosion data, has been used for two
tests with molten sodium.
consists of a large pot of type 347
stainless steel for holding the molten
bath and a rotating shaft with speci-
men holders that extend into the pot.
The spinner i1s equipped with a gas-
tight cover so that an inert gas
atmosphere can be contained during
test. The entire assembly is lowered
into a pot furnace to attain the
required test temperature. |

The spinner

The first spinner test was conducted
with Incenel specimens in sodium.
During the test the specimens were
immersed in sodium first at 450°C,
but not in motion, for 189 hr; next
at 450 to B816°C, in motion, for 11 hr;
and finally at 816°C, in motion, for
72 hours. The total time of the test
272 hours. The spiuning speed
400 ft/min; a helium atmosphere
was used; and 8.9 1lb of sodium was 1n
the test container. The specimens
were three 1 3/4-in. lengths of 1/2-
in.-0D Inconel tubing with the leading
end flared to 3/4 in. in diameter.
Afver the test, a very light, crystal-
line deposit approximately 1/2 mil in
depth was noted on the three specimens
and all had weight gains of approxi-
mately 1%.

Was
was

Under a microscope 1t was

PERIOD ENDING SEPTEMBER 10, 1952

apparent that beneath this deposit
there was a layer of voids approxi-
mately 1 mil deep in the Incomel.
Some of the crystalline deposit was
scraped from the tubing and sent for
chemical analysis. The analysis
reported was 64% nickel, 27.5% chromium,

8.3% 1ron, 0.15% aluminum, 0.23%
copper, and 0.12% titanium. There was
an 1ncrease in chromium and i1rton
content and a decrease 1in nickel

content as compared with the normal
Inconel composition (80% nickel, 15%
chromium, 5% i1rom). Since Lhe spinner
pot is constructed of type 347 stain-
less steel, 1t i3 not known whether
the surface layer and voids
a velocity effect or were due to the
presence of the two dissimilar metals,
the Inconel of the specimen and the
type 347 stainless steel of the con-
tainer. [t is planned to run
using type 347 stainless steel speci-
mens to obtain a moncometallic system.

wWeT e

a test

Type 1028 Chromized Steel in Low~
Melting-Point Alloys. Tests of type
1020 chromized steel have been run in
various low-melting-point alloys.
In some cases 1t was difficult to
estimate the exact extent of attack
because of the adherence of bath metal
to the specimen. 1n all these tests,
however, attack appeared to be suf-
ficiently severe to preclude the use
of type 1020 chrowized steel as a
container material for the low-melting-
point alloys tested. However, the
chromium layer on these materials
appeared to be quite porous 1in nature.
The tests were for 100 hr at 816°C,
and except 1in one case both the speci-
mens and chromized steel.
Results of the tests are presented in

Table 25.

tubes were

Compatibility of Beryllium Oxide
with Sodium, Nak, and Lead. Metallic
lead would be a satisfactory moderator
coolant from the standpoint of com-
patibility with the fuel and with
the beryllium oxide. As the data in

119
ANP PROJECT QUARTERLY PROGRESS REPORT

Table 25

RESULTS OF TESTS OF TYPE 1020 CHROMIZED STEEL IN VARIOUS
LOW-MELTING-POINT ALLOYS FOR 100 hr at 816°C

 

 

BATH

METALLOGRAPHIC NOTES

 

67.8% Sn-32, 2% Cd

67.8% Sn-32.2% Cd

43% Sn~57% Ba

60% B1-40% Cd

 

Test of chromized specimen in Inconel tube;
specimen apparently unattacked; however,
it was only partially covered by bath metal

Specimen completely penetrated in places
Erratic attack to 10 mils

Erratic attack to 8 mils

 

Table 26 show, insignificant weight
gains without appreciable change 1n
appearance and dimensions result when
beryllium oxi1de 1s tested with lead
in static systems. Dynamic tests have
as yet not been run with lead, but 1t
appears likely that the conclusions
stated above will be substantiated.

As the data in Table 26 indicate,
the beryllium oxide specimens often
crack during treatment with NaK. The
liberation of considerable hydrogen
for intervals up to 1 hr when the
specimens are placed 1in water after
test suggests that the NaK 1s present
in considerable amounts 1in fine cracks

Table 28

CORROSION OF BERYLLIUM OXIDE BY NaK AND LEAD

 

 

 

 

WEIGHT CHANGE
TYPE OF EXPOSURE METAL USED METALLOGRAPHIC NOTES
In % In mg/dm?/day
Static Pb +3 Slight coat of green material
Pb +8 No attack visible
Pb +7 No attack visible
NaK +1. 45 +128 No attack visible
NaK +0.98 +83 Specimen cracked
Dynamic NaK 1.0 +79 Specimen cracked
NaK 1.3 +99 Specimen cracked

 

 

 

 

 

120
FOR PERIOD ENDING SEPTEMBER 10, 1952

in the structure. Tt 1s possible that
prior to heating the NaK enters fine
cracks that are caused 1n the sawing
operation and that subsequent expansion
of the trapped NaK enlarges the
fissures until theyare easily visible,
Attempts to crack the blocks by thermal
shock, which might occur because of
the superior heat transfer properties
of NaK, have not been successful when
the capsules contained only beryllium
oxide and helium,

The spinner test was also run with
molten sodium and specimens of hot-
pressed beryllium oxide such as will
be used in the ARE. After the test
i1t was noted that some spalling had
particularly on specimen
the data given 1in
the weight losses and

taken place,
No. 1. In
Table 27,

test

Tablie 27

DATA FROM SPINNER TEST OF BERYLLIUM
OXIDE IN SODIUM

 

 

 

SPECIMEN 1 SPECIMEN 2
Weight Data
Before test {g) 21. 2583 23.7735
After test (g) 20,0749 23.7465
Loss (g} 1. 1834 0.027¢
Loss {%) 5.6 0.11
Mmensional Data
Length {(in.)
Before test 1,443 1. 545*
After test 1.427 1. 545¢%
Width (in.)
Before test 0.545 0. 548
After test 0.528 0. 546
Thickness {(in.)
Before test 0. 568 0.556
After test D.556 G.553
R U

 

 

 

 

*Surface very irregular, acturate measurements
difficult to obtain,

dimensional changes are due to this
spalling. It was also noted that
sodium had apparently penetrated the
specimens, so they were placed 1in
alcohal and then water for 24 hr,
that is, until the absorbed sodium
ceased reacting. In the test the
spinning speed was 400 ft/min, and the
specimens were submerged in sodium,
first at 450°C, but not spinning, for
31 hr; next at 450 to 816°C, spinning,
for8 hr; and finallyat 816°C, spinning,
for 100 hours. The total time of the
test was 139 hours. Test data are
summarized in Table 27.

FUNDAMENTAL CORROSION RESEARCH

In addition to the program of
empirical corrosion testing,
siderable effort has been devoted to
a number of studies designed to assist
in the discovery of the corrosion
mechanism. These studies include the
synthesis and identification of cor-
rosion products and studies of re-
actions in fluorides and hydroxides
under applied potentials. Of particular
significance is the mass of experi-
mental data supporting the postu-
lated(!*) fluoride corrosion of
Incomnel, that is, chromiuw depletion
of the metal and precipitation of the
resulting voids.

con -

Interaction of Fluorides with
Structural Metals (H. Power, J. D.
Redman, L. G. Overholser, Materials
Chemistry Division). FExamination of
the reaction between iron, chromium,
and type 316 stainless steel with a
fluoride melt containing [iF, NaF, and
KF under an inert atmosphere has been
extended to cover a range ol tempera-
tures and reaction times. lo this
study the finely divided metal has
been contacted with the melt under an
atmosphere of helium at 600 to 800°C

 

(14)Adamson, Brasunas, amd Richardson,

op., cit,, ORNL-1294, p¢. 126.

121
ANP PROJECT QUARTERLY PROGRESS REPORT

for various periods followed by
filtration at high temperature. The
concentration of structural elements
in the filtrate 1s then determined by
chemical analysis.

Previous examination(!%) indicated
that little (20 to 100 ppwm) chromium
and nickel and large amounts (1500 to
3000 ppm) of iron were solubilized by
a 6-hr treatment of type 316 stainless
steel with the NalF-KF-LiF eutectic at
800°C. Parallel tests have shown that
exposure for 72 hr raised the chromium
and nickel values slightly (40 to
400 ppm) without perceptible 1ncrease
in iron concentration. The amount of
reaction required to produce soluble
products appears similar when structures
of graphite or nickel as the
reaction vessel. There 1s some in-
dication that the values for nickel
and chromium are slightly higher when
filtration is performed at 600 rather
than 800°C. It was reported previously
that when the NaF-KF-LiF eutectic
contained added UF,, the iron content
of the filtrate roseto 6000 to 7000 ppm
and the amount of soluble chromium

serve

increased to 500 to 2000 ppm. When
metallie 1ron 1s exposed to the
eutectic mixture at 800°C, 1500 to
2500 ppm of 1ron appears 1n the
filtrate. The dissolved 1ron seems
to be trivalent in all cases. When
chromium powder is used, 150 to 450

ppm of the element 1s solubilized.
Examination of the cooled filtrate
indicates the presence of NaK,CrF.
However, when materials that
exposed 1n circulating-loop
corrosion tests arerecovered, filtered,
and analyzed, a different pattern 1s
revealed.

have
been

The NaF-KF-LiF eutectic containing
2 mole % UF4 when filtered after

exposure 1in equipment of Inconel or
stainless steel shows 100 to 300 ppm

 

(l‘q}H. Powers, J. ND. Redman, and L. G.

Overholser, op. c¢it., ORNL-1294, p. 119.

122

Fe, 20 to 100 ppm Ni, and 100 to
2000 ppm Cr. The decreasein solubility
of iron without an appreciable in-
crease 1in solubility of the other
elements may, perhaps, be due to the
deposition of iron by the mass transfer
mechanism. Experiments of this type
will be continued.

Synthesis of Complex Fluovrides

(B. J. Sturm, L. G. Overholser,
Materials Chemistry Division). There
has been no indication of simple

fluorides of the structural elements
in the fluoride mixtures following
interaction of the melts with the
containers, However, the presence of
NaK,CrF, has been confirmed, and it
is not unlikely that more sensitive
means of detection will show the
presence of other complex fluorides.
The preparation of complex fluorides
has continued as an aid in identifying
the various materials formed during
static and dynamic corrosion tests,
The complex fluorides synthesized
in these studies have been identified
or characterized by x-ray diffraction
patterns and/or by optical crystal-
lographic data.

It has been demonstrated that
NaK,CrF, is the only stable complex
fluoride of sodium, potassium, and
trivalent chrowium, Attempts to
prepare complexes of other proportions
resulted 1n mixtures contalning

NaKzCrF6 and either NaF or KF.

Fusion of various fluorides 1in
ratios corresponding to LiK,CrF_,
Li,KCrF,, Na,LiCrF,, and NaLi,CrF,
yielded products with characteristic
x-ray diffraction pattern, indicating
that each 1s probably a definite
compound. In addition, materials
that should correspond to the com-
pounds RbK,CrF_. and CsK,CrF, have
x-ray patterns similar to that of
NaK,CrF,, but they show slight shifts
of the lines that indicate somewhat
different lattice constants.
FOR PERIOD ENDING SEPTEMBER 10, 1952

K,FeF, has been precipitated by
slow addition of KF in aqueous so-
lution in less than stoichiometric
quantity to an aqueous solution of
FeCl, acidified with HF. K,;FeF. may
be prepared by fusion of K,FeF, with
KHF, or by fusion of KF with NH FeF, .
The optical properties of these
preparations agree with those of
K;FeF,, which is prepared by addition
of aqueous FeF, to aqueous solutions
of KF, as described 1n the litera-
ture.{18) Fusion of K,FeF, with NalF,
yields NaK,FeF,, which appears to be
isomorphous with NaK,CrF.

Hydrated KNiF, is formed when
proper quantities of aqueous KF so-
lutien are added to NiF2 in agueous
hydroflueric acid solution. Fusion of
a mixture of KHF, with hydrated NiF,
and subsequent extraction with water
vields a yellow powder that corre-
sponds, by chemical analysis, to
K,NiF,. The analogous complex com-
pounds of LiF and NaF are prepared in
a similar manner. The x-ray data from
this series of materials are in-
complete,

Manganic fluoride forms a series of
purple to reddish-violet products
upon fusion with the alkali fluorides.
Compositions corresponding to K MnkF,
and K;MnF, yield x-ray patterans typical
of pure compounds after fusion. Other
materlials prepared in similar fashion
have not yet been i1dentified.

Since only NaK,CrF, has been
positively i1dentified in the products
of corrosion and since many other
compounds are readily formed from the
corresponding fluorides, it might be
surmised that complex compounds of the
other metals are unstable in the
presence of chromium. Fusion of

NaK,FeF, (m.p., 970°C) with metallic

 

(1)  Minder, z. Krist. A96, 15-19 (1937).

chromium for 18 hr in aw inert atmos-
phere demonstrated that the reaction

Cr + NaK,FeF, ——s Fe + NaK,CrF,

takes place almost quantitatively.
The melting points of the pure com-
pounds have been shown to be: K,CrF,,
1055°C; NaK,CrF,, 1000°C; and NaK,FeF,,
970 °C.

EMF Measurements in Fused Fluorides
(L. E. Topol, L. G. Overholser,
Materials Chemistry Division). De-
composition potential measurements
inmolten fluorides have been continued,
Whereas previous work dealt with the
electrolysisof KF atnickel electrodes,
subsequent work deals with the effects
of additions of small amounts of NiF,
(approximately 0.3 wt %) to the KF.
Containers of graphite, magnesia,
and alumina were used. Graphite 1is
advantageous in that 1t 1s fairly
stable 1n the presence of fused
fluorides but 1s disadvantageocus in
that 1t is an electrical conductor.

The plot of £ vs., I was a straight
line through the origin and yielded a
resistance of 0.7 ohm when magnesia
was used as the container. In this
test neither electrode changed in
appearance, but the amount of nickel
formed in the cell was greater than
conld be accounted for by simple
electrochemical processes. The
magnesia vessel absorbed the molten
fluoride and cracked in numerous
places.

Norton Alundum crucibles (RA 7232)
are the most satisfactory of the
containers tried despite the slow
dissolution of the Al,0, by the fused
fluorides. The degree to which this
effects electrochemical results is
unknown. Breaks in the voltage-
current curves were noted at 0.4 and
0.8 to 0.9 volts. Both electrodes had
undergone attack in this test, and the

123
ANP PROJECT QUARTERLY PROGRESS REPORT

presence of nickel throughout the melt
and of NiF, near the anode was noted.
The significance of the breaks 1in the
voltage-current curves 1is difficult to
ascertain because the dissolution of
Al,0;, complicates the picture . and
because there 1s evidence that the
NiF, thermally decomposed at the
temperatures used (850 to 900°C).

The thermal decomposition of NiF,
appears impossible based on thermo-
dynamic considerations (AF = 120 kcal
and the equilibrium partial pressure
of F, = 10°23 atm(t7)), However, the
nickel formation 1in the
electrolysis experiments suggested
that decomposition was occurring and
led to experiments that verified it.

excessive

A series of experiments was run 1in
which various stock samples of NiF,
were heated for 4 to 5 hr under helium,
first with KF and then with NikF,,
alone, 1n cups of nickel, graphite,
alumina, and porcelain. In all cases
with temperatures from 675 to 900°C,
the compound, which 1s somewhat
volatile, decomposed and yielded
various amounts of nickel. Neither
the introduction of a CuO train,
heated at 600°C, into the helium line
nor the use of an activated charcoal-
liquid nitrogen trap had any effect.
This indicates that reducing substances
in the helium were not responsible
for the reduction to nickel, which
leaves possible impurities 1n the
NiF, as the only cause for the nickel
formation other than thermal de-
composition. Experiments with K,NiF,
gave similar results.

In studies of other materials it
was found that FeF, heated at 860°C
showed some magnetic properties,
whereas CrF, at the same temperature
appearecd to have been converted to

CLTXp L. Quill (ed.), The Chemistry and
Metallurgy of Miscellanecous Materials: Thermo-
dynamics, McGraw-Hill, New York, 1950.

124

CrF,. Nickel oxide that was heated
above 800°C showed some decomposition,
but not so wuch as NiF,. Samples of
NiO studied in the mass spectrometer
have shown dissociation into nickel
and the diatomic gas molecule at
temperatures of 1000 to 1050°C,
Tentative results with NiF, also
indicate that dissociation occurs at
lower temperatures.

Polarographic Studies in Sodium
Hydroxide (R. A, Bolomey, Materials
Chemistry Division). The collection
of polarographic data on the sodium
hydroxide—~platinum system has con-
tinued. The data have been analyzed
in terms of the location of the
maximum and half-wave values as
affected by temperature. Many of the
polarographic waves were found to be
too poorly defined to obtain reliable
half-wave-potential values. In
general, however, the half-wave
potentials found to closely
parallel the maximum values. Data
obtained in the past year have been
analyzed and a status report has been
written,

were

Electrochemistry of Sodium Hydroxide
(A. R, Nichols, Materials Chemistry
Division). Measurements of thermo-
galvanic potentials were made to study
the mechanisms of corrosion and metal
transport in fused hydroxide systems.
These measurements were made by de-
termining the potential between two
nickel electrodes maintained at
different known temperatures 1n a melt
of sodium hydroxide 1n the apparatus
described previously.(!%) Since the
voltage developed by the thermo-
galvanic cell depends upon the temper-
ature difference between the electrodes
and also upon the part of the tempera-
ture in which the cell 1s
operated, the results are conveniently
expressed by plotting the guantity

scale

 

(18)
p. 124.

A, R. Nichols, Jr., op. eit,, ORNL-1294,
FOR PERIOD ENDING SEPTEMBER 10, 1952

Q¢ = E/AT, in millivolts per degree,
against the mean of the electrode
temperatures. The hotter electrode
was negative in these trials.

Although curves showing some
features in common were obtained under
an atmosphere of purified helium,
erratic changes 1in potential were
observed, whether hydrogen-fired or
untreated nickel electrodes were used.
However, when a hydrogen atmosphere
was used, the results were fairly
reproducible.

Figure 51 shows the result of a
trial in which the charge consisted of
sodium hydroxide and approximately
0.2 mole % nickel oxide. It 1s noted
that a maximum value of Q of between
0.8 and 0.9 1s maintained over a range
from approximately 400 to 600°C, with
a gradual decrease to about 0.2 at a
temperature above 750°C. These values
do not change appreciably with time;
it 1s possible to retrace the curve
through a cooling and a second heating
and cooling, with good agreement.
Trials with a helium atmosphere, in
addition to giving less smooth and
reproducible curves, showed gross
changes with time and gave generally
higher @ values in the higher temper-
ature range. '

!afi,ls_?iq

o

@ {mvideg)

 

 

360 400 500 eno TR T ano
TEMPERATURE (°C)

 

 

 

Fig. 51. Potentials in Molten
Sodium Hydroxide as a Function of
Temperature Gradient.

In addition to the trials in which
nickel oxide was added, several trials
were made with sodium hydroxide alone.
With hydrogen-fired electrodes and a
helium atmosphere, it was found that
the valueof Q was erratic and actually
reversed 1ts sign occasionally, which
indicated that the cold electrode
became negative. When this experiment
was carried out in hydrogen, the value
of Q became negative after an initial
positive interval, during which,
presumably, traces of oxide were
removed from the electrodes, and
remained slightly negative during
successive heating and cooling cycles.

It is evident that if some form of
dissolved nickel, represented in these
experiments by the added nickel oxide,
is present in molten sodium hydroxide,
which 1s 1n contact with nickel
surfaces at different temperatures, a
potential will develop that is of the
proper sign to correspond to oxidation
(attack) of the nickel metal at the
hot surface and to reduction (de-
position} of nickel at the cooler
surface.

The process may be represented by
the following equations:

Ni (metal) ——3> (Ni**) + 2 electrons

(at hot surface) ,

(Ni**) + 2 electrons——> Ni (metal)

(at cold surface) .

Since little is known of the actual
state of the nicke}l in these melts,
the term (Ni*") must be interpreted
as being some 1on containing divalent
nickel in a form capable of electro-
chemical equilibrium with metallic
nickel.

The experiment with sodium hydroxide
containing no added nickel oxide and
a hydrogen atmosphere showed that
under these conditionsno concentration
of dissolved 1onic nickel develops

125
ANP PROJECT QUARTERLY PROGRESS REPORT

and, hence, no potential corresponding
to metal transport is observed. The
potential of opposite sign that was
observed hasnot been clearly accounted
for. It would appear to be dependent
upon some process involving the hy-
drogen and sodium hydroxide, with the
nickel electrodes serving only as
carriers for the hydrogen gas and thus
acting as hydrogen electrodes.

The measurement of the thermo-
galvaniec potential appears to offer a
fundamental method of studying the
sodium hydroxide system apart from
any participation by the electrode
metal 1tself. It 1s significant in
relation to the corrosion and metal
transport processes that the use of a
hydrogen atmosphere cleans up traces
of nickel oxide and leaves the melt
free of dissolved nickel and hence
incapable of developing the potential
corresponding to metal transport.

The work described establishes the
existence of thermogalvanic potentials
between nickel electrodes in fused
sodium hydroxide containing dissolved
nickel oxide. The potentials have
been shown to be of such sign and
magnitude as to be capable of accounting
for transport of metallic nickel
in nonisothermal nickel systems in
which fused sodium hydroxide 1s
circulated,

Mechanism of Fluoride Corrosion
(A. des Brasunas, L. S. Richardson,
Metallurgy Division). A number of
tests 1ntended to give a better
understanding of fluoride corrosion
has been completed. The postulate
in the previous report(!®) that the
subsurface voids encountered during
fluoride corrosion are caused by the
outward diffision of chromium atoms,

 

(lglAdamaon, Brasunas,
op. c¢it., OBNL-1294, p. 126.

and Richsardson,

126

which leaves the metal enriched with
vacant lattice sites that “precipi-
tate” to form voids, has been sub-
stantiated repeatedly. It has been
demonstrated when chromium is re-
moved from the metal lattice by either
high-temperature oxidation or high-
temperature vacuum treatment. Con-
versely, corrosion tests made with
fluoride containing small amounts of
chromium powder have shown that void
formation can be suppressed. The
basis for this conclusion may be
briefly summarized as follows:

1. Even with careful nonaqueous
polishing, it has never been possible
to retain particles in the porous
surface region.

2. Chemical analysis of the at-
tacked metal surface has revealed a
substantial loss in chromium (from 15
to less than 5%).(29)

3. The chromium content of the
fluoride abnormally high 1n
comparison with the nickel and iron
contents, which indicated a preferential
solution of chromium from the alloy.

was

4. Calculations of void area
anticipated from change in chromium
content are 1in accordwith observations.

5. Depth of void formation bears
a direct relationship to the amount
of chromium in the bath.

6. Attacked specimens show weight
losses but no dimensional changes.

7. Chromium loss from the alloy
caused by either high-temperature
oxidation or high-temperature vacuum
treatment has also resulted in similar
void formation,

 

(20)Brasunns and Richardson, op. cit.,
ORNL-1227, p. 124.
FOR

8. Such voids have been observed
previously(21+-22) in diffusion ex-
periments in which diffusion in one
direction exceeded the extent of
diffusion in the opposite direction.

9. The addition of small amounts
of chromium powder to saturate the
fluoride bath with chromium has
prevented void formation under con-
ditions that otherwise favor void
development. |

10. Fluorine 1s not detected
below the metal surface,

The porous surface region of an
Inconel metal specimen before and
after exposure to molten fluoride 1s
shown in Fig. 52. Many of the voids,
especially those not in grain bound-
aries, have definite shapes that bear
a relationship to the metal lattice.
The inclusions shown in Fig. 52 are
convenient sites for the deposition of
vacancies; two such sites may be seen.
The corrosion of Inconel by air 1is
known to form a chromium-rich oxide
layer at elevated temperatures. In
a 200-hr exposure to air at 1200°C
(2192°F), the region beneath the oxide
layer of an Inconel specimen was found
to contain voilids similar to those
observed in fluoride-attacked Inconel

(Fig. 53).

The high vapor pressure of chromium
relative to that of iron or nickel
suggests that high-temperature vacuum
treatment would also be a suitable
means of removing chromium. Inconel
and an 80% Ni-20% Cr alloy were exposed
to a vacuum of 0.1 mm Hg for 42 hr at
1375°C. Both samples showed many
subsurface voids; those in Inconel
appeared to be spherical, whereas

those i1n the 80% Ni-20% Cr alloy were

 

L. C. €. da Silva and R. H. Meh!l,
of Metals 191, 155 (1%951).

(zz)uanthly Technical Progress Report for
October 20 to November 20, 1951, SEP-82.

Journal

PERIOD ENDING SEPTEMBER 10, 1952

angular, as shown in Fig. 54. Chemical
analysis of the entire Inconel speci-
men indicated a drop in chromium
content from about the usual 15% to
9. 2%.

The mechanism of void formation in
Inconel is therefore believed to
occur in two stages. The first
involves a loss of chromium atoms by
diffusion to the metal surface and
then into the fluoride bath. The
resulting large number of vacant
lattice sites 1s reduced to a more
stable number by “precipitation” as
1llustrated 1n Fig. 55. For ease of
illustration only six vacancies are
shown to constitute a void. The
number of vacant atoms involved in
the voids shown in previous photo-
micrographs is of the order of 10%2
atoms.

Tripositive Nickel Compounds from
Hydroxide Corrosion (J., V. Cathcart,
W. H. Bridges, L. D. Dyer, B. Borie,
Metallurgy Pivision). Two new tri-
positive compounds of nickel have been
identified as corrosion products of
nickel attacked by sodium hydroxide
in the presence of oxygen. The purpose
of this investigation was twofold.
First, in corrosion and mass transfer
studies with the hydroxides, a variety
of black and colored products has been
obtained that do not correspond to
any known nickel compounds. The
occurrence of these products has been
somewhat of a difficulty in the efforts
to ascertain the mechanismof hydroxide
corrosion. It was desired therefore
to try to identify some of the more
unusual corrosion products and de-
termine some of their physical and
chemical properties to aid in future
identification and to give some clue
as to how they might have been formed.

The second purpose of this in-
vestigation was to determine whether
nickel ions of valence greater than
2 might occur in hydroxide melts,

127
ANP PROJECT QUARTERLY PROGRESS REPORT

UNCLASSIFIED
Y.6795

UNCLASSIFIED
Y.6796

o

 

corrosion of Inconel by Fluorides.

Fig. 52. Voids BResulting from the
2000x. (b) Porous surface

(a) Inconel specimen prior to fluoride attack.
region of fluoride-attacked TInconel showing geometric-shaped voids and larger
2000x.

irregularly shaped voids in grain boundaries and around inclusions.

128
FOR PERIOD ENDING SEPTEMBER 16, 1952

These higher valence ions are required
in the very ingenious mechanism of
nickel mass transfer proposed about a
year ago.(?3) At that time the only
nickel compounds of valence greater
than 2 for which there was anything
but the most tenuous evidence were
four trivalent nickel oxvhydroxides
and a hydrous oxide of gquadrivalent
nickel.

 

 

 

 

 

 

Briefly, the most important feature
of the work was the preparation of two
compounds whose empirical formulas are
NaNiO, and LiNiO,. These compounds,
which are named sodium nickelate(III)
and lithium nickelate(III), have been
identified by chemical analysis for
nickel and for the alkali metals; by
several chemical reactions, including

 

 

 

 

 

 

 

(23)p, G, Hill and R. A. Bolomey, Aireraft

; . : : Nuclear Propulsion Project Quarterly Progress

Flg" 53. Voids F"‘.”“e" Dourlng Alr Report for Period Ending September 10, 1851,
Corrosion of Inconel at 1200°C. 250X, ORNL- 1154, p. 122,

‘LASSIFIED § B R o TUNCLASSIFIED §
y.6826 | K y v o, . X.6827

 

Fig. 534. Voids in Inconel and an 80% Ni-20% Cr Alloy After Vacuum Treat-
ment at 1375°C. (a) Inconel. (&) B80% Ni-20% Cr Alloy. 50X

129
ANP PROJECT QUARTERLY

C CHROMIUM ATOM
F IRON ATOM
N NICKEL ATCM

   

BEFORE LEACHING

Fig. 585.
Chromium Atoms.

a quantitative determination of the
oxidizing power; by a reaction that
gives a known trivalent nickel com-
pound; and by complete crystal struc-
ture analysis.

Solutions of Metals in Molten
Halides (M. A, Bredig, J. W. Johnson,
H. R. Bronstein, Chemistry Division).
Considerable effort has been directed
toward i1mproving the experimental
technique so that more accurate
determinations can be made of the
solubility of metals tn the melts of
their halides. With the improved
technique, which does not appear to
be subject to the previous objections,
a solubility of 2.87 mole % potassium
in the molten ternary eutectic LikF-
NaF-KF at 880°C was obtained. The
discrepancy with the value of 1.6
mole % reported previously, which 1s
not too serious 1in view of the ex-
perimental difficulties, 1is thought
to be due mainly to removal of potassium
metal 1n the earlier experiments by
leaching the salt phase with alcohol.
The leaching may not, as intended,
have been confined to the dissolution
of metal mot dissolved in the melt;
the alcohol may also have attacked the
dispersed in the solidified
salt phase. Separation by alcohol
leaching has been abandoned. Satis-
factory separation of molten halide

metal

130

 

VACANCIES CAUSED BY LEACHING

PROGRESS REPORT

UNCLASSIFIED
DWwG. 18322

 

VACANCIES PRECIPITATED

Sketeh Illustrating Formation of Subsurface Voids by Removal of

and molten metal phases 1s now being
achieved in capsules of 1mproved
design that incorporate a ball check
valve between the lower part containing
the salt phase and the upper part
containingthe metal phase and portions
of the salt phase.

Among other halidesystems, chlorides
are recelving continued attention,
In view of the possibility of chlorine
isotope separation, such dystems may
even be of direct interest for reactor
applications. In the sodium metal-
sodium chloride system, an increasing
solubility of the metal from amounts
of the order of 1 mole % near the
melting point of sodium chloride to
amounts of 20 mole % 200°C above the
melting point was observed. The
disagreement of such comparatively
high solubility with the solubility
predicted by a current theory of
metal~metal halide solutions 1s being
studied. Possible relationships
between changes in structure and
volume upon melting of the molten
halides and the solubility of metal
in them are beiling investigated.

Fluoride Corrosion Phenomena
(M. A. Bredig, H. R. Broustein, J. W,
Johnson, Chemistry Division). A
misleading statement appears in the

previous report (ORNL-1294, p. 130,
FOR PERIOD ENDING SEPTEMBER 10,

fourth line from the
“The heat of for-
mation of ZrF,, AF = 445 + 30 kg/cal,
1s 10 that of KF, and the
equilibrium concentration of FelF, at
1100°K, which is from 10°% to 1077,
is one~thousandth that of KF,” 1t
should read, “The free energy of

right column,
top); 1nstead of,

times

1952

formation of ZrF,, AF ¥ -445 + 30 keal,
yields for theequilibrium concentration
of FeF, at 1100°K (1500°F) a value
between 10°° and 10°°. The higher
of these limits 1s ten times larger,
and the lower one one-thousandth the
Fer concentration 1in the case of
corrosian by KF.” '

 

11.

W. D. Manly

METALLURGY AND CERAMICS

Jo M. Warde

Metallurgy Division

A systematic study of the basic
welding variables associated with the
cone-arc welding process is being made,
The effects of arc-current magnitude
and arc-current duration omn cone-arc
welds are described 1n this report.
The following welding specifications
have been prepared for the joining of
Inconel pipe and fittings for use with
highly corrosive materials: procedure
specifications for d-~c¢ inert-arc weld-
ing of Inconel pipe and fittings, and
the operator’s qualification specifi-
cations for d-¢ inert-arc welding of
Inconel pipe and fittings.

The mechanical properties of high-
temperature brazing alloys have been
investigated, and the joint efficiency
has been calculated by comparing the
tensile strength of the brazed joint
with the tensile strength of a speci-
men of the parent metal that has
undergone the same temperature cycle
as that used in brazing. |

The preparation of the components
for use in the ARE control and safety
rods has been completed., Sixty of the
B,C~-Fe inserts and 29 of the Al1,0,-R,C
inserts were made and canned by Nicro-
brazing. The control rods and safety
rods have been delivered to the ARE
staff for calibration.

The effect of small differences 1in
test temperature on the stress-rupture
life and creep rate of Inconel has
been studied. The creep properties of
Inconel when tested in argon and azir
have been measured, and it was found
that Inconel tested in an air atmos-
phere has a much longer stress-rupture

life,

A satisfactory high-temperature
ceramic coating has bheen applied to
nickel sheet to be used in a ligquid
metal-to~air radiator. The possibility
of the development of a glass-bonded
beryllium oxide material for use as a
reflector moderator 1s being investi-
gated,

CONTROL ROD FABRICATION

E. S, Bomar J. H. Coobs
Metallurgy Division

The preparation of components to be
used 1n the ARE regulating and safety
putlined pre-
been completed., A

on work
has

based
(1)

rods,
viously,

(1)E. 5. Bomar and J). H, Ceobs, Adircraft

Nuclear Propulsion Project Quarterly Progress
Report for Period Ending Jaurne 10, 1952, ORNL-
1294, p. 135; and Metallurgy Division Quarterly
Progress Report for Period Ending April 30, 1952,
ORNL-1302, p. 78.
ANP PROJECT QUARTERLY PROGRESS REPORT

total of 60 inserts composed of boron
carbide and 1ron were pressed; 54
inserts were needed for three safety
rods. Thirty-three cylinders composed
of dilute mixtures of B,C and Al,0,
were fabricated for use in two control
rods with different B,C investments,
These two rods required 29 inserts,
The surplus parts were prepared for
use in case of breakage,

Only minor difficulties were
encountered 1in producing the inserts,
The first cylinders prepared cracked
along their length during the cooling
cycle following hot pressing. This
difficulty was avoided by ejecting the
graphite mandrel from the center of
the slugs while they were still at
temperature., Another difficulty in-
volving differences 1in expansion co-
efficients was encountered when the
first insert was sealed
brazing.

in a can by

The clearances originally set up
were not adequate to allow for the
expansion of the inner stainless steel
liner when the assembly was heated to
1120°C for brazing. Consequently, the
cylinder broke
excesslive 1nternal loading.
the inside diameter
cylinders by 0,015 1in.

because of
Increasing
of the pressed
prevented this

ceramic

type of failure during the canning
operation, The percentage of B,C in
the B,C~Al,0; cylinders was 1ncreased
to compensate for the decreased
volume accompanying the increase
in inside drameter. Densities of 80%
of those of the theoretical mixtures
were obtained 1in pressing full-size
parts of the two types of materials.,
The B,C-Fe cylinders were pressed with
2500 psi of pressure at 1520 to 1540°C,
The A1,0,-B,C inserts were compacted
at the same pressure but were heated
to 1650°C., In both instances graphite
dies surrounded by bubble alumina
insulation and an 1nert atmosphere of
argon were used,

132

MECHANICAL TESTING OF MATERIALS

R. B. Qliver
D. A, Douglas G. M, Adawmson
K. W. Beber

Metallurgy Division

J. W, Woods

Tests have been initiated to deter-
mine the effect of stress corrosion by
molten fluoride salt mixtures on the
creep rate and stress-rupture life of
structural materials., The creep and
stress-rupture tests 1n argon reported
in the previous report(?} have con-
tinued, and new data for Inconel
obtained.

were

Physical Testing in Fiuorides (G. M.
Adamson, K., W, Reber, Metallurgy
Division)., In the physical testing
program of the joint experimental
engineering section of the ANP Divisiocn
and the Metallurgy Division, NabF-KF-
LiF-UF, continued to be the fluoride
mixture tested. It will not be possible
to test any of the zirconium-base
fluorides in the present apparatus
because of their high vapor pressures,
When the emphasis was shifted from
sodium to fluorides 1t was necessary
to use the sodium-testing apparatus,
since speed was a prime consideration
of this program. The sodium-testing
apparatus was constructed from types
316 and 347 stainless steel. 1Inconel
was the material most likely to be
used 1n the ARE, so 1t the first
material tested. The photomicrographs
from the initial tests
and they show that
in every case 1n which the material
in the test for an appreciable

was

of specimens
have been received,

was
mass-transferred
This
is true of both the stress-rupture and
tube-burst speciwens, All work waith
Inconel has been stopped until TInconel
pots and sample holders are obtained
from the shop.

length of time a
layer appeared on the specimen,

 

(2)R. B. Oliver, D. A. Douglas, . W. Weaver,
and J. M. Woods, op. cit., ORNL-1294, p. 136,
FOR

Jnconel stress-rupture specimens
have been tested in Nab-KF-LiF-UF, at
1500°F under stresses of 6000 and
7000 psi. A plot of the rupture times
of 131 and 94 hr falls on the line
through previously obtained values,
These tests were run in type 347 stain-
less steel pots. Stress-rupture tests
of type 316 stainless
are now being set up.

steel samples

Several tube-burst tests have been
run with type 316 stainless steel
specimens. The rupture times obtained
from these tests did not agree well
and were much lower than expected.
The tube-burst tests have consistently
praoduced low results with both Inconel
and type 316 stainless steel. Tests
now being set up with better
control to deterwine

are
temperature
whether this 1s the difficulry.

An apparatus was designed for
carrying out self-welding tests i1n the
fluorides. The final design was an
which, by

the present

means of a
stress -

apparatus 1in
simple adapter,

rupture equipment could be used. The
apparatus will allow the two tesct
specimens to be held apart until the

fluorides have time to condation the
surfaces. The specimens can then be
brought together under any compressive
load Self-welding will be
determined upon completion of the test
by applying a tensile load while the
apparatus 1s still hot., The
necessary to separate the samples will
be a measure of the amount of self-
welding. This procedurewonld alleviate

desired.

stress

PERIOD ENDING SEPTEMBER 10, 1952

the difficulties encountered 1n remov-

ing and cleaning the samples.

Creep and Stress-Rupture Tests 'in
Argon. The stress vs., creep curves
for fine-
sheet tested at

and coarse-grained Inconel
1500°F in argon pre-
sented 1w a previous.report(a) have
been revised and are presented in
Figs, 56 and 57.

It has been suspected that a small
difference in the test temperature
could have a noticeable effect on the
rupture time and perhaps also on the
the
two

creep rate. Table 28 summarizes
data on tests now 1in progress at
temperatures.

The Inconel sheet specimens tested
1n argon exhibit a shorter rupture
life than reported by the International
Nickel Company for similar tests con-
ducted 1in air. Two specimens are
being tested 1n air to resolve this
inconsistency. The data from these
tests are summarized in Table 29,

WELDING

P. Patriarca G. M. Slaughter
Metalluregy Division

The systematic study of the cone-
arc welding process is continuing with

emphasis during the last quarter on

 

(313. B, Oliver, J, W. Woods, and C. W, Weaver,
Aircraft Nucleor Propulsion Project Quarterly
Pragress Report for Period Ending March 10, 19592,
OBNL-1227, p, 149 and 150, Figs. 46 and 47,

Table 28

CREEP AND STRESS-RUPTURE

TESTS ON INCONEL IN ARGON

 

 

 

TEMPERATURE STRESS TOTAL ELONGATION AT CREEP RATE
(°C) (psi) 1000 hr (%) (% per hr)
815 2000 0.5 0.0007
821 2000 2.0 0.0040

 

 

 

 

133
ANP PROJECT QUARTERLY PROGRESS REPORT

UNCLASSIFIED
DWG.15684 A

CREEP RATE (% perhr)
G 1 10

0.000 0oet oo
20,000 =g oo

\ ‘
0% | 0.5% 10% 2% 5% RUPTURE | | |
% [EXTENSION AT Ti

   

YT

10,000
T‘é
%5000 g b LT L
uJ
o
=
w

2,000 ———t L bl b IS NN

hooo[w“—n»

ol 02 05 10 2 5 10 20 50 100 200 500 1000 2000 5000 10,000
TIME (hr)

Fig. 36. Stress vs. Creep Rate for Fine-Grained Inconel Sheet Cold Rolled

at 1650°F and Tested in Argon at 1500°F. Grain size: approximately 90 grains
0.105 mm in diameter. Times were measured for 0.1,

per square millimeter,
Extension measured

0.5, 1, 2, and 5% extension and rupture vs. Stress.
optically.

UNCLASSIFIED

 

 

 

|

 

 

 

 

 

   

 

 

 

 

     
  
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

CREER RATE (% per hr) EWG 156854
Q000! .00 00 o1 10
20000 | T | ‘ \l ! ‘ ; T C T T T
| i;l o | ] ‘1’
L] | | | |
Q, 1 o, o, I 1 i |
0 1%, 05% | /ni 2% D% 33 % (EXTENSION 4T TIME OF Rup*rur‘c‘i)‘ ‘ ; l i ‘
. B ! | i ( | !
10000 gt — b - 5 RUPTURE 2’ L ll Ll . f P
e B e S T e L) . 4 !
hn.!ljh m“m”-mL,=1fl_mn
P~ . [ 315
5 . |
&
W
O 5000!—
[vel
=
o)
2000t - [~ :
w
1000 J ) H |
500 1000 2000

 

 

 

 

ol
TIME {br)

Creep Rate for Coarse-Grained Inconel Sheet Annealed
Grain size: approximately
Times were measured
Extension was

Fig. 57. Stress vs.
for 1 hr at 2050°F and Tested in Argon at 1300°F.

15 grains per square millimeter, 0.250 mm in diameter.
for 0.1, 0.5, 1, 2, and 5% extension and rupture vs. stress.

measured optically.

134
FOR PERIOD ENDING SEPTEMBER 10, 1952

Table 29

COMPARISON OF RUPTURE TIMES OBTAINED FOR INCONEL SHEET

 

 

 

 

RUPTURE TIME (hr) _ '
STRESS ‘ - OBNL TEST IN AIR.
(psi) OPNL. Test in International Nickel Co. {time to date, hr)
Argon Test in Air ‘
3500 700 - 6000 2475
4000 400 3000 1711

 

 

 

 

the effect of the initial surge on the
resulting weld. Specifications have
been prepared for the welding of com-
ponents for the ARE,

Cone-Arc Welding., The feasibility
of the application of the cone-arc
welding process to fabrication of
tube-to-header heat exchangers was
determined by experiment in previous
work. (%) Tt has been recognized, how-
ever, that the systematic determination
of the behavior of the basic welding
variables associated with the cone-~arc
welding process would contribute toward
a better understanding of possaible
applications and limitations. The
effects of arc-current magnitude and
duration on the inside and outside
diameters (Di and DU) of the weld and
on the minimum weld penetration (P)
have been determined. The initial arc
current i1s generally of the order of
100% greater than the steady-state
value and is the subject of another
study now 1n progress.

As in the previous work, the header
configuration and material sizes con-
sisted of 0,100-in.-0B, 0,010-3in. -
wall, type 316 stainless steel tubing
and 0,125-in.-thick, type 316 stain-
less steel header sheets containing 19
tube holes drilled one diameter apart
at the apex of an equilateral triangle,
which was the basic pattern. All

 

(4 i
' )P. Patriarca and G. M. Slaughter, op. cit

ORNL-1294, p. 138,

tubes were set flush with the header
surface prior to welding. A series of
welds were made using the conditions
outlined in Figs. 58 and 59, the
variables being arc current and arc
time, respectively.

Examination of Fig, 58 will reveal
that the predominate effect of in-
creasling arc current 1s to 1ncrease
the value of D . This, coupled with
the relatively small increase in weld
penetration P and the equally small
decrease 1in Di’ indicates that the
major portion of additional energy
introduced as increasing arc current
served to increase the 1ineffective
weld area. It may be noted, however,
that cone-arc welds may be.made over a
relatively wide range of conditions,
but that operation in the higher
current range should be avoided since
greater header distortion becomes
evident with higher energy input,

The results of a similar study are
presented in Fig. 59, which presents
arc time as the variable. The sima-
larity of Figs. 58 and 59 1s evident.
Analysis of the curves in Fig., 59 will
reveal that arc time 1s not a critical
variable within the limits of 1.5 to
4.2 sec under the other welding con-
ditions indicated,

Specifications for ARE Welding.

The following specifications have been
prepared for use in the fabrication of

135
¢l

WELD DIMENSION (in.)

0.20

O
o

MINIMUM CURRENT FOR
ACHIEVEMENT OF COMPLETE
PERIPHERAL WELD

UNCLASSIFIED
DWG. 6380

|0, OVERLAP|
! AND

\i (01 SENSITIVE |

PERMISSIBLE OPERATING RANGE RA

NGE

|

HEADER HOLE CLOSED

0,

P

40 50 60 70
ARC CURRENT {D-C amp)

Fig. 58, Effect of Arc Curvent on Cone-Arc Weld Parameters Do. Di.

Material: tubes, type 316 stainless steel, 0.10 in. OD, 0.010-1q. w
header, type 316 stainless steel, 0.125 in. thick
tube hole separation, 0.10 in. (one tube diameter}

Welding Conditions: arc time, 1.5 sec.
arc distance, 0.05 in.

high-thoria tungsten filament, 0.0625 in. in diameter {pointed)

argon gas fiow, 30 cfh

8C

and P.

all

 

LAT6Ud dNV

LHOdAY SSAUs0Hd A'HALEVNI0
WELD DIMENSION (in.)

LET

UNCLASSIFIED
DWG, {638t

MINIMUM ARC TIME FOR
ACHIEVEMENT OF COMPLETE Dq —
0.20 ___ PERIPHERAL WELD HEADER HOLE CLOSED

|

D, OVERLAP RANGE

D, SENSITIVE

PERMISSIBLE OPERATING RANGE _, __ RANGE

 

{ 2 3 4 5
ARC TIME (sec}
Fig. 39. Effect of Arc Time on Cone-Arc Weld Parameters D_, D, and P.

Material: tubes, type 316 stainless steel, 0.10 in. OD, 0.010-in. wall
header, type 316 stainless steel, 0.125 in. thick
tube hole separation, 0.10 in. (one tube diameter)

Welding Conditions: arc current, 5.9 amp d.c.
arc distance, 0.05 in,
high-thoria tungsten filament, 0.0625 in. in diameter (pointed)
argon gas flow, 30 cfh

‘0T HAGWALJAES INIANT aoI¥dd HOoA

cSo6l
the ARE: procedure specification for
d-c¢ inert-arc welding of Inconel pipe
and fittings for high-corrosion appli-
cations; operator’s gualification test
specification for d-c¢ 1inert-arc weld-
ing of Inconel pipe and fittings for
high-corrosion applications.

TESTS OF BRAZING ALLOYS

P. Patriarca G. M. Slaughter
Metallurgy Divaision

The tensile strength and static
corrosion tests of brazed joints are
continuing, (%) The corrosion tests
are being extended to include static
tests of brazed joints in moist air
and tests of nickel joints 1n hy-
droxides. Brazing efficiencies as
high as 92% for butt-brazed joints
have been demonstrated in tensile
tests, The brazing of molybdenum 1is
being i1nvestigated, An additional
brazing research facility has been
provided by the recent installation of
a large G-E brazing furnace for dry-
hydrogen bhrazing operations,

Static Corrosion Tests of Brazed
Joints. The static corrosion testing
of brazed joints for the high-tempera-
ture brazing alloy evaluation program
is continuing. The 16,5% Cr-10.0%
Si~2.5% Mn~71,0% Ni alloy developed by
the General Electric Company 1s attacked
rather severely by the fluoride mix-
ture NaF-KF-LiF-UF, when brazed both
on Inconel and type 316 stainless steel,
Photomicrographs of these corroded
joints are shown in Fig. 60,

nickel has been found to be
relatively unattacked in molten sodium
hydroxide, static corrosion test
specimens of this material brazed with
several of the promising alloys have
been

Since A

prepared. Examination of the

 

D rbid., p. 140.

138

samples has not been completed, but
several photomicrographs of the as-
brazed joints, particularly the joint
brazed with 60% Pd, 40% Ni, and
Nicrobraz, show excellent wettability
and flow characteristics.

The resistance of brazed joints to
oxidation at high temperature 1is an
important factor, so tests are being
conducted on brazed jolnts to deter-
mine the amount of deterioration.
Butt-brazed joints of Inconel, type
316 stainless steel, and A nickel have
been prepared by using several of the
brazing alloys being investigated,
Many of these have bheen tested in an
atmosphere of moist air at 1500°F, but
metallographic examination of these
jJoints 1s 1in the preliminary stage.
Effects of oxidation at higher tewmper-
atures will also be investigated, but
the tests are presently being run to
provide a source for the screening of
the less desirable alloys,

Tensile Strength of Brazed Joints.
In order that an approximate tensile
strength of joints butt-brazed with
the 60% Pd-40% Ni alloy could be
obtained, a few preliminary tensile
tests were made on these joints prior
to a more thorough i1nvestigation,
pending the arrival of an appropriate
high-temperature extensometer, Standard
tensile bars 0,505 in. in diameter were
machined from butt-brazed type 316
stainless steel tensile blanks., Three
tensile bars were tested at room
temperature and three were tested at
1500°F, Test bars of type 316 stain-
less steel, which were subjected to
heating cycles similar to the brazing
cycles, were also pulled as check
samples on the strength of the parent
metal. The results of these tests are
summarized in Table 30,

The fracture of the butt-brazed
joint at room temperature occurred
along the diffusion zone of the joint,
R 070 v-66298

 

   

 

&

 

 

Fig. 60. Joints Brazed With 16.5% Cr~198.0% Si-~2. 9% Mn~731.9% Ni Alloy and
Tested in NaF-KF-LiF-UF, (10.9-43.5-44.5-1.1 mole %) for 100 hr at 150D°F.
(e¢) Type 316 stainless steel joint showing moderate attack of the brazing
alloy. (b) Inconel joint showing severe attack of the alloy. Unetched. 75X,

139
ANP PROJECT QUARTERLY PROGRESS REPORT

Table 30

TENSILE STRENGTH OF TYPE 316 STAINLESS STEEL BRAZED WITH 60% Pd-49% Ni ALLOY

 

 

 

 

CHECK BARS BRAZED JOINT
TEMPERATURE BRAZING EFFICIENCY
(°F) Tensile Strength | Elongation | Tensile Strength | Elongation (%)
(psi) (%) (psi) (%)
Room 75,400 40.0 82,000 75.0 92
1500 22, 400 18.4 25,600 87

 

 

 

 

 

 

The fracture of the joint brazed at
1500°F resulted in a more ragged
surface and occurred frequently along
the braze metal-base metal interface.
However, all these data suggest that
the 60% Pd-40% N1 alloy is very
promising, although it is sti1ll ex-
tremely desirable to strive for lower-
ing of the melting point by alloying.

Only one brazing alloy has been
investigated by using the larger
0.252-1n,-dia butt-brazed tensile bars.
A high-temperature extensometer de-
with this size of
and work on this

si1gned for
specimen has arrived,

use

phase of the brazing alloy evaluation
program should proceed much more
rapidly. A series of tensile tests at
room temperature and at 1500°F were
completed for Inconel brazed with the
75% Ag~20% Pd-5% Mn alloy. The Inconel
check bar has not yet been tested., At
room temperature, the average tensile
stress of the brazed joint was 55,900
ps1, and the average elongation 1in a
l-in. gage length was 9%. At 1500°F
the average tensile stress of the
brazed joint was 19,700 psi, and the
average elongation was 2.5%. These
results apparently mean that the
brazed joint decreases 1n strength
with 1ncreasing temperature more
rapidly than does the base Inconel.
The elongation of the base metal is
quite large at room temperature,
whereas at 1500°F there is very little
elongation.

140

Brazing of Molybdenum. In view of
the interest shown in the fabrication
of molybdenum, it was decided to
conduct a limited program on the
brazing of this metal with the various
high-temperature alloys presently being
investigated. The wettability charac-
teristics of the various brazing alloys
on bright-annealed molybdenum sheet in
a dry-hydrogen atmosphere are listed

in Table 31.

A further indication of the strength
of molybdenum brazed joints should be
obtained from the tensile test speci-
mens to be prepared in future work.
If a small guantity of ductile mo-
lvbdenum can be obtained, tensile
tests will be made on this material,
since morereliable data should result,

Dry-fiydrogen Brazing Furnace. A

24-kw, 312-amp, G-E brazing furnace
has been i1nstalled for use 1in dry-
hydrogen brazing operations. A

“Square D" stainless steel muffle,
welded shut at one end, was built 1into
this furnace for the purpose of ob-
taining the atmospheres required for
many furnace-brazing problems, The
physical dimensions of this muffle are
such as to accommodate a sample of the

following dimensions: width at base,

6 in.; height at center, 5 in,; length
to be heated at one time, approximately
4 ft (hot zone of furmace); total

length, approximately 9 ft {(includes a
5-ft cooling .hamber). A stainless
FOR PERIOD ENDING SEPTEMBER 10, 1952

Tablie 31

WETTABILITY CHARACTERISTICS OF VARIDUS BRAZING ALLOYS ON BRIGHT-ANNEALED
MOLYBODENUM SHEET IN A DRY-HYDROGEN ATMOSPHERE

 

 

BRAZING ALLOY

WETTING PROPERTIES

 

60% Pd-40% N1

Nicrobraz

16.5% Cr-10.0% S1-2.5% Mn-71.0% Ni

50% Pd-40% N1-10% Mn

40% Pd-40% N1~20% Mn

60% Pd-37% N1~-3% Si

73.5% Ni-16.5% Cr-10,0% Si
75% Ag-20% Pd~5% Mn

64% Ag-33% Pd-3% Mn

Excellent
Fair
Good
Very good
Gond
Excellent
Excellent
Excellent

Excellent

 

 

steel pipe was welded into the sealed
end of the muffle to permit the entry
of dry hydrogen.

A system, Fig, 61, for providing a
dry-hydrogen atmosphere is attached to
this furnace. Included in this system
is a nitrogen cylinder (to purge the
hydrogen lines of air), a Deoxo
Duridryer,{®) and a Lectrodryer
dehumidifier.¢’? The dryer converts
oxygen to water and the dehumidifier
removes the water., Dew points of
~120°F can be obtained, although -60°F

is satisfactory for most applications.

When the need arises for the furnace
brazing of assemblies larger than the
stainless steel muffle permits, 1t 1s
convenient to use canning techniques.
Such techniques consist of enclosing
the assembly in a stainless steel can,
which 1s then welded shut, Stainless
steel entrance and exit tubes are
provided in the can for the hydrogen

 

“”Baker and Co,, Newerk, N. J.

7
¢ )Pitcflburgh Lectrodryer Corp., Pittsburgh, Pa,

flow.
located entirely in the stainless can,
any furnace of proper size and temper-
ature limits can be used as the source
of heat. The sodium-to-air heat ex-
changer, Fig. 62 (approximate di-
mensions, 12 by 6 by 8 in.) is typical
of the assemblies that can be brazed by
this canning technique.

With the controlled atmwosphere

Nicrobraz powder was placed on all
of the joints, including approximately
2500 tube-to-baffle plate joints.
Good hydrogen flow over the entire
assembly was needed for proper flow of
the brazing alloy, and 1t 1s expected
that a series of baffle plates 1inside
the can will be required to attain
this flow for brazing of similar, but
larger, heat exchanger units,

REDUCTION OF MOLYBDENUM BISULFIDE
G, P. Smith, Metallurgy Division
Simple thermodynamic calculations

have been made in an effort to deter-
mine 1in a preliminary fashion the

141
ANP PROJECT QUARTERLY PROGRESS REPORT

UMCLASSIFIED

 

 

 

 

 

 

 

 

 

 

 

 

 

  

 

 

 

 

 

 

 

 

 

 

 

DWG. 16537
DEOXQ - - =
PURIDRYER H2 FLOWMETER N2 FLOWMETER
N2 H(2 H2 L_-._.l.,.--.J
ST TSBURG TO MOLYBDENUM
LECTRODRYER FURNACE WINDINGS
TYPE BAC—0
BOTTLED GAS MANIFOLD
(mxn :
OUTLET FOR
DEW-POINT MEASUREMENTS
CEE) = TWO-STAGE REGUL ATOR TG FURNACE MUFFLE
Fig. 61. HManifold System for Dry-Hydrogen Furnace Brazing.
SR R R R AR, S : 2 S . .
UNHA%gIlED:, SR e possibility of reducing molybdenum

disulfide with hydrogen or zinc. These
calculations show that flowing hydrogen
should be capable of reducing molybdenum
disulfide at temperatures 1in the
neighborhood of 1000°C, Solid zinc is
thermodynamically capable of reducing
molybdenum disulfide and would probably
work as well as molten zinc,

The reduction of molybdenum di-
sulfide by hydrogen would be ac-
complished by the following reaction:

 

MoS, + 2H, = Mo + 2H,S ,

from which

Fig. 62. Heat Exchanger Test Unit .
Nicrobrazed by Camning Techmigque.

142
FOR PERIOD ENDING SEPTEMBER 10, 1952

where p 1s the equilibrium pressure
and K 1s the appropriate equilibrium
constant, as implicitly defined by the
above equation, Values of K are given
by Kelley. (%)

For a hydrogen pressure of 760 mm
the equilibrium pressure of H,S at
910°C 1s 6.65 mm, and at 1100°C 1t is
14.5 mm, These are guite appreciable
pressures, and i1t should be easy to
reduce molybdenum disulfide at these
temperatures if the reaction rate is
at all reasonable. Tt would be neces-
sary to remove H,S from the reacting
mixture, but this could be accomplished
with a current of hydrogen.

The reduction of molybdenum di-
sulfide by zinc can be represented by
the following equation: '

MoS, + 2Zn - > Mo + 2ZnS .

The standard free energy change for
this reaction in the case of solid
zinc may be obtained from the data

given by Kelley.,(®)
AFC = 28,340 - 2,567 log T + 2.40 x 107312

+92.198 x 10737 + 9,121 .

Thus at 25°C, AF° = ~26.6 kcal/g-atom
of MoS,. The standard free energy
equation does not exist, and calcu-
lations of AF° at elevated temperatures
would guantitatively be somewhat
deceiving., However, the reduction 1s
thermodynamically possible at 25°C,
and the AF® is sufficiently sub-
stantial to warrant consideration of
this methed of reduction at elevated
temperatures.

 

(8)K. K. Kelley, Contributions to the Data on

Theoretical Metaliurgy, Bureau of Mines Bulletin
406 (1937), p. 53, '

(9 ybid., p. 54, 66.
Eq. 305 onm p. 54 is incorrect
multiplied by T.

The I factor as given in
and should be

CERAMICS RESEARCH

J. M. Warde, Metallurgy Division

A ceramic coating to inhibit
oxidation at elevated temperatures has
been successfully applied to nickel to
be used in the construction of a
radiator for the ANP. Beryllium oxide
shapes for use in ANP reactor experi-
ments have been made and fired at the
U, 8. Bureau of Mines Electrotechnical
Laboratory at Norris, Tennessee., Pre-
liminary work on the development of a
glass-bonded beryllium oxide material
for use as a reflector-moderator for &
liguid-cycle aircraft
completed,

reactor was

Ceramic Coatings forMetalis. Further
work was carried out on the application
of a ceramic coating to nickel to
provide oxidation resistance at
elevated temperatures. Successful
application of such a coating would
make it possible to use nickel in a
radiator. It was found that by
annealing specimens of pyrometallurgi-
cal nickel, 10 mils thick, 3In an
atmosphere of wet hydrogen for 65 hr
at 1000°C, good adherence was obtained
with National Bureaun of Standards
ceramic coating A-418, 1 to 2 mils
thick, applied at 1700°F., The frig
composition of this coating has been
described in a previous report,(!?)
Specimens coated with the A-418 com-
position showed no signs of oxidation
in a 9-day oxidation test at 1500°F,
It i1s proposed that in the near future
this coating be applied to actual
radiator parts fabricated from nickel.

Ceramic Reftlector. To satisfy the
requirements for an efficient and
economical material for use as a
reflector-moderator for a ligquid-cycle
aircraft reactor, a glass-bounded
beryllium oxide material has been
proposed, This material would have

143
ANP PROJECT QUARTERLY PROGRESS REPORT

nearly zero porosity and would contain
a minimumof 75 mole % beryllium oxide.
The glass forming the bond would also
contain a small amount of beryllium
consistent with viscosity re-
Simulated
mixes using soda-lime-silica glass and
silica sand were prepared on a small
scale, and it was determined that such
a material could be made successfully,
The density of a selected, glass-bonded
beryllium oxide, reflector-moderator
material was determined as 2.6 g/cc.
It is proposed that the reflector be
cooled by metal pipes embedded in the

oxide,
quirements for fabrication.

glass~bonded beryllium oxide. The
pipes would contain flowing sodium as
a coolant, and once the glass-bonded
beryllium oxide had been ponrred into
the shell and had flowed around the
cooling pipes, the reactor shell would
not be allowed to cool below a temper-
ature just above the annealing range
of the glass bond. Further refinements
and engineering tests will be made,
should this material be of interest,
Among these would be the determination
of the devitrification tendency of the
glassy matrix and the corrosion effect
of the glass with selected metals,

 

12.

HEAT TRANSFER AND PHYSICAL PROPERTIES RESEARCH

H. F. Poppendiek, Reactor Experimental Engineering Division

Viscosity measurements have been
made for two zirconium-bearing fluoride
fuel mixtures. For the probable ARE
fuel NaF-ZrF,-UF, (46-50-4 mole %) the

viscosities range from 20 centipoises

at about 580°C to 7 centipoises at
about 830°C, and for the fuel Nalk-
ZrF,-UF, (50-46-4 mole %) the vis-

cosities range from 17 centipoises at
about 580°C to 6 centipoises at about
830°C. A capillary viscometer is
being assembled and will be used to
obtain data to supplement the data
from the existing viscometers,

Some preliminary thermal conductivity
experiments on the fuel NaF-KF-LiF-UF,
(10.9-43.5-44.5-1.1 mole %) have
vielded the conductivity value 2.2 £
0.2 Bru/hr:ft? (°F/ft) over the temper-
ature range 500 to 750°C, The vapor
pressures of Zr¥F, and ZrF, -bearing
fuel mixtures are being determined.
Zirconium tetrafluoride is the most
prevalent vapor phase above Zrl,-
bearing fuel to the extent of about
10 mm Hg at 800°C, Enthalpy and heat
capacity have been determined for
lithium hydroxide.

144

Experimental, forced-convection,
heat transfer data for moelten sodium
hydroxide flowing turbulently in
circular tubes were represented by
the equation

Nu = 0,021Re®:8pyp0.4

over the Beynolds number range of
6,000 to 12,000, This equation 1s
within 9% of the one normally used to
describe convective heat transfer of

Information on
lengths was also

ordinary fluids.
thermal entrance
determined,

Some experimental, pool-boiling
data for a mercury system have been
obtained for several pressure levels
and two different kinds of heat-
trans fer-surface materials,

Heat and momentum transfer analyses
have been made for the ANP thermal-
convection harps for predicting the
liquid circulation velocities. This
information is needed in the interpre-
tation of the harp corrosion results.
In the one case examined, the circu-
lation velocity was approximately

0.1 ft/sec.
FOR PERIOD ENDING SEPTEMBER 10,

VISCOSITY OF FUEL MIXTURES

The viscosities of two ZrF,-bearing
fuel mixtures have been determined on
the rotational viscometer. Some
viscosity measurements are now being
made for NaOH and NaNQO; by using the
efflux viscometer., In addition, the
importance of viscosity of fuel and
moderator-coolant mixtures in design
of the ARE components has prompted
study of additional methods for de-
termination of this property at high
temperatures, An improved capillary
viscometer has been constructed. Also,
a new structure has been completed
that will permit the operation of the
Brookfield viscometer, the efflux
viscometer, and the density device
when working with toxic materials,

NaF-ZrF,-UF, Fuels (R. F. Redmond,
D. F., Smith, T. N. Jones, Reactor
Experimental Engineering Division).
Viscosity measurements have been made
with the Brookfield rotational vis-
cometer for the zirconium-bearing fuel
mixtures NaF-ZrF, -UF, (46-50-4 mole %)
and NaF-ZrF,-UF, (50-46-4 mole %), and
they have been plotted in Fig. 63.
The viscosities of both mixtures are

SECRET
DWG 16366

-ZrF, */UFq
NaF-7rE, -UF, (46-50-4 mole %)

(50~-46-4 mole

VISCOSITY (centipoises)

Naof ~KF -LiF-UF,
(10.2-42.5-44 5-1.1 mole %)

 

750 BOO 850 900 950 1000 1050 1100
TEMPERATURE ' (°K)

Fig. 63. viscosity of Na¥F-ZrF,-UF,
Salt Mixtures.

1952

approximately twice as high as that of
the NaF-K¥-LiF-UF,.

Capillary Viscometer (F. A. Knox,
N. V. Smith, F. Kertesz, Materials
Chemistry Division). The method
employed in the capillary viscometer(?)
consists of forcing the liquid through
the measuring tube with the help of a
vacuum pump, which establishes a
pressure differential between the
outside and inside liquid surfaces.
The pressure difference is automati-
cally stabilized at predetermined
levels through a circuit that ‘has
leads into a mercury manometer, - The
ligquid travels upward through the
calibrated tube and reaches a wider
tube in which three electrically in-
sulated metal probes are immersed.
The time elapsed between reaching a
lower and a higher level 1s measured
with an electric timer connected to
these leads. Upon completion of the
time measurement, a relay closes the
valve to the vacuuw pump and opens a
valve to compressed argon, the pressure
of which can also be regulated, The
liguid is then forced back to 1its
original position and the apparatus
is ready for another measurement., The
temperature 1s regulated and recorded
during the experiment with a Micromax
recorder, A 5-gal container 1is used
te reduce the pressure fluctuations of
the vacuum pump. During the whole
operation the surface of the melt 1s
protected with a blanket of argon from
contamination by the air,

Preliminary trials with the apparatus
have indicated satisfactory performance.
Calibration studies utilizing a number
of fused salts are now 1n progress.
In the only experiment to date on
material of interest, the viscosity of
the mixture containing 50 mole % NaF,
46 mole % ZrF,, and 4 mole % UF, was

found to be about 13 centipoises at

 

(l)H.
Proc. Roy. Soc.

Bloom, B. 5. Harrap, and E. Heywmann,
{London) 194A, 237 (1948},

145
ANP PROJECT QUARTERLY PROGRESS REPORT

600°C and about 4 centipoises at 800°C,
These data must be considered tentative
until the calibration studies
completed and evaluated.

are

THERMAL CONDUCTIVITY OF LIQUIDS

L. Cooper S. J. Claiborne

W. D. Powers R. M. Burnett

Reactor Experimental Engineeraing
Division

Preliminary thermal conductivity
values have been obtained for the fuel
NaF-KF-LiF-UF, by using apparatus
described previously,(2?) The values
obtained and some of the data on the
fused salts studied previously are
given in Table 32, From the data of
Table 32 1t appears that the thermal
conductivity increases as the per-
centage concentration of the heavy UF,
compound 1s decreased in the salt
mixture. The ZrF,-bearing fuels are
currently being studied.

A thermal conductivity device (a
modification of apparatus previously
described¢3) for use with solids) that
1s particularly useful for studying
liguid metals has been used to check
values for liguid sodium; the measure-
ments obtained were within 20% of the

 

(Z)L. F. Basel and M. Tobias, Aircraft Nuclear
Propulsion Project Quarterly Progress Report for
Period Ending December 10, 1950, ORNL-919, p. 196,

(B)M. Tobias, Aircraft Nuclear Propulsion

Project Quarterly Progress Report for Period
Ending March 10, 1551, ANP-60, p. 243,

values given in the literature. Pre-
liminary data for the lead-bismuth
eutectic have been obtained.
is currently being studied.

Lithium

A transient hot-wire method 1s be-
ing studied in connection with thermal
conductivity measurements, This
method, which i1nvolves 1imposing a
transient temperature daistribution on
a wire surrounded by the fluid to be
investigated, appears tobe well suited
to the study of the molten fluorides
and hydroxides.

HEAT CAPACITY
W, D. Powers C. G, Blalock

Reactor Experimental Engineering
Division

The enthalpy and the heat capacity
of lithium hydroxide have been deter-
mined by Bunsen ice calorimeters for
the temperature range of 500 to 900°C
and are given by the equations

H (liquid) - Hyo (solid) = 0.85T + 110 , (1)

c, = 0.85 % 0.06 , (2)

where H i1s in cal/g, T in °C, and ¢
. P
in cal/g°C.

Some of the 12-1in,
for the calorimeters have been re-
placed by 24-in. furnaces that have
yielded more uniform capsule tempera-
tures; this modification has reduced

heating furnaces

Table 32

THERMAL CONDUCTIVITY VALUES FOR VARIOUS FLUORIDE SALT MIXTURES

 

 

TLUORIDE SALT MIXTURE

COMPOSITION (mole %)

THERMAL CONDUCTIVITY, Btu/hr- ft? (°F/ft)

 

NaF-KF-UF, 46.5-26.0-27.5
NaF-KF-LiF-UF, 10.9-43. 5-44.5-1.1
Nat'-KF-LiF 11.5-42.0-46.5

 

0.53 at 550°C < ¢t < 750°C
2.0 to 2.5 at 500°C < ¢t < 750°C
2.6 at 500°C < t < 750°C

 

 

146
FOR PERIOD ENDING SEPTEMBER 10,

the scatter of the enthalpy data.
Several fluoride fuels with Sr(OH),
added and the LiCl-KCl eutectic are
currently being studied.

DENSITY
D, F. Smith

Reactor Experimental Engineering
Division

Preliminary density data on Nal-
Zr¥,-UF, (46-50-4 mole %), NoF-ZrF, -
UF, (50-46-4 mole %), NaNO;, and
molten NaOH are being obtained. A
mean value for the density of the
50-46-4 mole % mixture at 640°C < ¢ <
870°C is 3.3 g/ce. The BeF,-bearing

mixtures are to be studied next.

VAPOR PRESSURE OF FUEL CONSTITUENTS
R. E. Moore

Materials Chemistry Division

The vapor-pressure data for solad
Zr¥, and for two typical ZrF,-bearing
fuel mixtures have been obtained during
the past guarter by use of techniques

4 e (4,5,6)
and apparatus previously described.

Zirconpium Tetrafluoride. Since
salt mixtures containing ZrF, are
under consideration as fuels, and
since ZrF4 is the only important
component of the vapor phase above
these molten salt mixtures, a determi-
nation of the vapor pressure of solid
zirconium tetrafluoride was made 1n the
temperature range 722 to 854°C, The

(4)R. E. Moore and (. J. Bartom, Aircraft
Nueclear Propulsion Project Quarterly Progress

 

Report for Period Ending September 10, 1931,
OBNL-1134, p. 136,
(S)R. E. Moore, Aircreft Nuclear Propulsion

Project Quarterly Progress Report for Perieod
Ending December 10, 1951, ORNL-~1170, p. 126,

(6)H. E. Moore, Aircra)ft Nuclear Propulsisn
Projeet Quarterly Progress Report for Period
Ending June 10, 1852, OBNL-1294, p. 158.

1952

data, given in Table 33,
represented by the equation

may be

11320
T(°K)

+ 12.46 ,

 

log P(mm Hg) = -~

The heat of sublimation is 52 kcal/mole
and the sublimation point 1s 908°C, as
calculated from the equation. The
value for the sublimation point 1s 1n
fair agreement with the estimate
(927°C) found in the literature,’’’

 

 

Table 33
YAPOR PRESSURE OF ZIRCONIUM
TETRAFLUORIDE
TEMP ERATURE OBSERVED PRESSURE
(°c) {mm Hg)
729 13.5
741 21
743 22
754 27
776 43
777 48
778 45.5
854 278

 

 

Zirconium-Bearing Fluoride Mixtures,
In a previous report %) the vapor
pressure of a mixture containing 42
mole % ZrF,, 51 mole % KF, 5 mole %
NaF, and 2 mole % UF, was reported,
Since then interest has centered on
the system ZrF, -NaF-UF,, and vapor
pressure data for two compositions in
this system have been obtained.

The data for the solution containing
50 mole % Zr¥F,, 46 mole % Nal, and 4
mole % UF, are given in Table 34. The

efquation

8095

10g P{mm Hg) = - ?zgfii

 

7). .

( )La L. Quill {ed.), The Chemistry and
Metallurgy of Miscellanecus Matertials: Thermo-
dynamics, McGraw-Hill, New York, 1956,

147
ANP PROJECT QUARTERLY PROGRESS REPORT

best represents the data. The calcu-
lated heat of vaporization 1s 37

kcal/mole.
Table 34
VAPOR PRESSURE OF MIXTURE CONTAINING

530 mole % ZrF4, 46 mole % NaF, and
4 mole % UF,

 

 

TEMPERATURE OBSERVED PRESSURE
(°C) (mm Hg)
807 i2
842 21
876 31
907 45
936 72
940 84

 

 

Vapor pressure data for the solution
containing 46 mole % ZrF,, 50 mole %

 

NaF, 4 mole % UF, are presented in
Table 35. The eguation
7630
log P(mm Hg) = - — + 7.988
T{°K)

The heat

of vaporization as calculated from the
equation is 35 kcal/mole. An increase

was obtained from the data,

of 4 mole % ZrF, content results 1in a
considerable increase in vapor pressure,
It seems likely that, since ZrF, 1s by
far the most volatile component of the
mixtures being considered, the vapor
pressure will, in general, increase with
increasing ZrF, content, although
probably not in a regular fashion.

Table 35

VAPOR PRESSURE OF MIXTURE CONTAINING
46 mole & ZrF4, 50 mole % NaF, and
4 wole % UF,

 

 

TEMPERBATURE OBSERVED PRESSURE
(°C) (mm Hg)
807 8
858 17
8§72 22
906 33

 

 

148

CONVECTIVE HEAT TRANSFER IN MOLTEN
SOPIUM HYDROXIDE

H. W. Hoffman J. Lones
Reactor Experimental Engineering
Division

Final results have been obtained
from the turbulent forced-convection
heat transfer experiment in which
molten sodium hydroxide flowed through

tube. The apparatus was
in detail 1n previous re-

a circular
described

ports.(®) The experimental data for
the region of fully established flow
are correlated by the equation

Nu = 0.021Re® 8pyp0-*

for a Reynolds modulus range of 6,000
to 12,000 and temperatures between 700
and 900°F, This equation is compared
with the McAdams’' equation for the
heating and cooling of ordinary fluids
(fluids other than the liquid metals)
in Fig, 64. The curve lies 9% below
the McAdams’ correlation.

 

(B)H. ¥. Hoffman and J. Lones, op. cit.,
ORNL-1294, p. 152, and Aircraft Nuclear Propulsion
Project Quarterly Progress Report for Pertod
Ending March 10, 1952, ORNL-1227, p, 161,

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

UNCILASSIFIED
100~ — e g — DWG. 16361
m”__L;j%_mr_rmm{ e
s80+———— A4 R -
| — - McADAMS B a
- THIS WORK }
60 e 1 - s -
| = S
40F-—— - 44/
s j ;%V
£ /gj
g . fi ) -
20— |
‘ |
10, b |
2,000 10,000 30,000
REYNOLDS NUMBER
Fig. 64. Experimental Heat Transfer

Coefficients for Sodium Hydroxide.
FOR PERIOD ENDING SEPTEMBER 160,

Thermal entrance lengths were
obtained for each run. This length is
defined as the distance from the
entrance of the test section, in terms
of tube diameters, at which the local
heat transfer coefficlent, or con-
ductance, reaches a value within a
given percentage of the established
value. A 10% value was used here,
Figure 65 presents the thermal entrance
length as a function of the Peclet

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

modulus (Beynolds modulus times
Prandtl modulus),
UNCI_ASS]IFIED
50, DWG, 15262
- I
B b ek L L LA b
20 'L'""“’"“"
’C-:;fl)
o |
{Q frrmmmmrememem s d e L — -
9 i L { I e -
8 e e 38 wma .|
7 e s e e a oL e
Bfmmree s e i ]
o L e
0,000 100,000
PECLET MODULUS
Fig. 5. Thermal Entrance Length

for Molten Sodium Hydrexide.

A new apparatus to be used to
determine the convective conductances
for the NaF-KF-LiF mixture is now be-
ing assembled, This system is similar
to the one used in the sodium hydroxide
experiment, Owing to the corrosive
nature of the NaF-K¥-LiF mixture, all
system components except the test
section are constructed of Incomnel.
The test section will be made of
nickel tubing because the proper size
of Inconel tubing 1s not available.

. temperature incremexls,

1952

BOTLING HEAT TRANSFER IN MERCURY

W. 5. Farmer
Reactor Experimental Engineering
Division

Further measurements have been
obtained on free-convection hoiling of
mercury from horizontal surfaces. The
experimental heat transfer vs. temper-
ature difference data are presented in
Fig., 66. ¥From this plet the mean heat
transfer coefficients shown in Fig. 67
were computed. The increase 1n
pressure accompanying a change in
boiling temperature from 350 to 550°F
produced an approximately threefold
increase in the heat transfer co-
efficient for any given heat flux
level., On the basis of these results,
beiling and condensing mercury should
prove to be a good medium for heat
transfer in high-temperature power
cycles, A large difference between
the heat transfer coefficient data for
the mercury-copper {(wetted)} and
mercury-chromium (nonwetted) systems
may be observed. There are insufficient
data at this tiwe to establish accu-
rately the specific interface~-fluid
temperature difference at which boiling
is initiated, as well as other boiling
However, 1t
boiling co-

curve characteristics.
does appear that the
efficients rise rapidly at the higher
as in the case
of boiling water.

NATURAL CONVECTION IN CONFINED SPACES
WITH VOLUME HEAT GENERATION

D. C. Hamilton F. E. Lynch
Reactor Experimental Engineering
Division

Preliminary data obtained with the
flat-plate apparatus described in

149
ANP PROJECT QUARTERLY PROGRESS REPORT

100,000

80,000

60,000

40,000

20,000

HEAT FLUX, @/4 (Btu/hr-{)

10,000 |-

8,000

6,000

4,000
3 4 6 8 10

 

UNGLASSIFIED
CWG. 16363

COPPER SURFACE -———
BOILING TEMPERATURE 550°F

COPPER SURFACE
BOILING TEMPERATURE 350°F

CHROMIUM SURFACE —— v
BOLING TEMPERATURE 350°F

20 40 60 80 100 200

SURFACE~FLUID TEMPERATURE DIFFERENCE (°F)

Fig. 66.
Horizontal Surfaces to Mercury.

previous reports‘®) appear to be in
general agreement with the temperature
profile predicted by the laminar flow
analysis, (%) Repeated short circuits
from the electrolyte through the
delicate, quartz, traversing-thermo-
couple assembly to the thermocouple
have prevented obtaining other than
preliminary data. A new technique for
joining the assembly appears to have
alleviated this problem.

 

(Q)D. C. Hamilton, F. E. Lynch, L. Palmer, and
R. F. Redmond, op. cit.,, ORNL-1227, p. 156;
D. C. Hamilton and F. E. Lynch, op. c1it., ORNL-
1294, p. 158,

(IO)D. C. Hamilton, H. F. Poppendiek, and L. D.
Palmer, Theoretical and Experimental Analyses of
Natural Convection Within Fluids in which Heat Is
Being Generated, ORNL CF-51-12-70 (Dec. 18, 1951).

150

Experimental Data for Free-Convection Boiling Heat Tramsfer from

The circular annulus apparatus has
been constructed and assembled. The
cooling systems for the upper electrode
and the test section have been tested
for leaks, and the system will be
ready to operate when the flat-plate
data have been obtained. This apparatus
was designed to operate in the high-
laminar and low-turbulent flow regions.
The upper limit is set by the 10-kw
maximum power now available. The
validity of all previous data was 1n
question because it was not possible
to directly measure the wall tempera-
ture. Since the temperature gradient
is maximum at the wall, 1t is diffa-
cult to determine the wall temperature

from “near-the-wall” measurements.
FOR PERIOD ENDING SEPTEMBER 10,

10,000

8,000

6,000

4,000

 

1952

UNGLASSIFIED
OWG. 16364

40 60 80 100 200

SURFACE~FLUID TEMPERATURE DIFFERENCE (°F)

2,000
[
o
~,
£
.
2
@
< 1,000
80O
600
e COPPER SURFACE
BOILING TEMPERATURE 550°F
400 | ® COPPER SURFACE
BOILING TEMPERATURE 350°F
A CHROMIUM SURFACE _
BOILING TEMPERATURE 350°F
200
5 6 B 10 20
Fig. 67.

The present apparatus consists es-
sentially of a 1/2-in.-ID, thick-
walled, copper tube that is electri-
cally insulated on the inner surface
by a baked enamel coating, In this
design it is possible to install a
thermocouple in the wall to measure
the wall temperature, This represents

Free-Convection Boiling Heat Transfer Coefficients for Mercury.

a great improvement over previous

designs.

The cross.section view in Fig. 08
shows the method of installation of
the center and wall thermocouples.
The center thermocouple is located 1in
a gquartz capillary tube.

151
ANP

UNCLASSIFIED
CWG. 16365

--UPPER ELECTRODE

     
  
 
 
   
  
   

 

= LIQUID MERCURY
LEVEL

——--COPPER TEST
SECTION

— COPPER COOLANT
JACKET

 

~— COOLANT OUT

 

 

 

< =i
=M P T ) S
: — =
’ SRR
s o —--WALL THERMOCQUPLE
;g f——

——----CENTER THERMOCOUPLE

 

4——QUARTZ TUSE

 

1%in
e

 

-------- CERAMIC COATING

 

 

 

 

 

 

 

 

 

 

——-=COOLANT IN

 

 

4~ LOWER
=) ELECTRODE

 

 

 

 

 

 

 

 

 

 

 

 

 

Fig. §68. Annulas
Convection Apparatus.

Type of Free-

PROJECT QUARTERLY PROGRESS REPORT

HEAT AND MOMENTUM TRANSFER ANALYSIS OF
THE THERMAL CONVECTION LOOPS

H. F. Poppendiek [.. Palmer

Reactor Experimental Engineering Division

Mean circulation velocities should
be knowo i1n order to properly interpret
corrosion results in thermal convection
loops. At present, ANP convection
harps are instrumented sothat approxi-
mate wall-temperature distributions of
the cold legs and power inputs to
electrical heaters on the hot legs are
measured. This information can be
used to estimate some maxlimum and
minimum circulating velocities that
How -

to obtain more exact velocities

represent limiting situations,
ever,
it 1s necessary to solve the combined
heat and momentum transfer problem
numerically. Sowe incomplete numerical
analyses indicate that the calculated
circulation velocities fall within the
estimated range close to the minimum
value., Thus 1t appears that the
circulating velocities 1n the con-
vection harps are approximately 0,1
ft/sec when the fluid is NaF-KF-1iF
(11.5-42.0-46.5 mole %). It should be
possible to i1ncrease these velocities
by heating the hot legs and cooling
the cold legs 1n such a way that uni-
form wall-temperature distributions
exist in those legs.

 

13.

RADIATION DAMAGE

D. S. Billington, Solid State Division

A, J. Miller, ANP Division

The radiation damage studies of
materials exposed in the ORNL graphite
reactor, the LITR, and the 86-1in.
cyclotron have continued., Further
experiments have been carried out on
the effects of 1rradiation on the
fused fluoride fuels, the performance
of liquid metal loops, the creep of
metals, and the thermal conductivity
of metals.

152

As 1n the case of the beryllium-
bearing fuels previously reported, the
evidence to date indicates that there
1s no significant radiation-induced
corrosion with zirconium-bearing fuels
undergoing fission at the rates
that will occur during operation of
the ARE. Apparatus is currently being
assembled and inserted 1in the MTR to
provide an exposure facility for the
FOR

fused fluoride fuels in which fission
rates comparable to those that would
occur in an aircraft reactor can be
achieved. Measurements made in the
graphite reactor on the escape of
Xe!?5 froma capsule of molten fluoride
fuel showed that only a small fraction
of the xenon diffused out of the
liquid.

Examination of an Inconel loop
through which sodium had been circu-
lated at a peak temperature of 1500°F
in the graphite reactor disclosed no
positive evidence of radiation-induced
corrosion. '

In-reactor cantilever creep measure-
ments were made on Inconel specimens
subjected to the 1700°F anneal proposed
for the ARE fuel tubes. Under some
conditions of stress and flux there
was no increase 1n the creep rate due
to the irradiation, such as was
previously reported for specimens
annealed at Jlower temperatures. Tt
was also found that the thermal con-
ductivity of Inconel given the ARE-
specified heat treatment was un-
affected by irradiation in the LITR.

Additional details on radiation
damage studies are contained 1in the
quarterly reports of the Solid Stat
Division. '

IRRARIATION OF FUSED MATERIALS
G. W. Keilholtz D. F. Weeks
J. G. Mergan M.  T. Robinson
H. E. BRobertson D. D. Davies
C. C. Webster A.  Richt
P, R, Klein W. J. Sturm
M. J. Feldman

Solid State Division

R, J. Jones R. L. Knight

Electromagnetic Research Division
B, W. Kinyon
ORNL. Engineering Division

Three samples of the zirconium-
bearing reactor fuel, NaF-K¥F-ZrF,-UF,

PERIOD ENDING SEPTEMBER 106, 1952

(4.,8-50,1-41.3-3.8 mole %), were
irradiated at 1500°F for 140 hr in
Inconel capsules in the LITR. Energy
was dissipated in the fuel at a rate
of 125 watts per cubic centimeter of
fuel, which is 1n the range of the
peak power dissipation expected during
operation of the ARE. As with the
beryllium-bearing fuels previously
irradiated, chemical analyvses of the
zirconium-bearing fuels and metal-
Jographic examinations of the capsules
showed no evidence of radiation-
induced corrosion when compared with
two out-of-reactor control runs.

Bombardment of zircomium-bearing
fuel in an Inconel capsule for 8 hr in
the cyclotron with a beam of 22.5-Mev
protons dissipating 2500 watts per
cubic centimeter of fuel caused a
in the corrosion of
the Tncenel. This 1s similar to the
result reported previously for a
lithium-bearing fuel.,(?)

slight increase

An experiment was performed:to
determine the rate of diffusion of
Xe!3% from the fused flueride fuels
under essentially stagnant or
turbulent conditions,
made 1in the OBRNL graphite reactor at
1300°F with NaF-BeF,-UF, (25-60-5
mole %) fuel containing natural uranium
in Inconel capsules. In the first run
xenon was flushed out by bubbling
helium through the melt, whereas in
the other the gas was simply swept
over the surface. The xenon removed

non-
Two runs were

in the sweeping experiment was eqguiva-
lent to only 2.5% of that removed with
the flushing procedure.

Additional woerk 1n progress on
irradiation of the fused fluorides in-
cludes melting-point determinations of
irradiated fuels, rocking-capsule
tests to simulate the ARE cycling, a

 

(1){;. ¥. Keilholtz, dircraft Nuclear Propulsion

Projeet Quarterly Progress FReport for Period Ending
June 10, 1952, ORNL-1294, p. 160,

153
ANP PROJECT QUARTERLY PROGRESS REPORT

study of the relative importance of
ionization effects in
the proton beam, and longer cyclotron
beam exposures, The MTR exposure
facility for irradiation of fuel
capsules at 2,500 to 10,000 watts per
cubic centimeter of fuel has been
completed. The 1initial experiment at
2500 watts 1s expected to be run

duraing September,

“knock-ons"” vs.

IN-REACTOR CIRCULATING LOOPS

0. Sisman R. M. Carroll
W. W, Parkinson C. D, Bauman
J. B. Trice C. Ellis
A. S. Olson W. E. Brundage
M. T. Morgan D, T, James

F. M. Blacksher

Solid State Division

Metallographic examination of the
in-reactor and out-of-reactor portions
of the 1500°F socdium loop described
previously(?) showed no evidence of
radiation-induced corrosion of the
Inconel.

A loop with i1mproved welds and a
by-pass circuit for filtering the
sodium 1is being constructed for
additional experiments i1in the graphite
reactor. Considerable progress has
been made on fabrication of the stress-
corrosion sodium loop for operation in
the LITR. In addition, an in-reactor
system for circulating fluoride fuel
1s being designed.

CREEP UNDER IRRADIATION

W, W. Davis J. C.
J. C. Zukas

Solid State Division

Wilson

Increasing the annealing tempera-
ture from 1650 to 1700°F appears to

(2)
163.

 

0. Sisman et al., op, cit., ORNL-1294, p.

154

reduce the adverse effect of neutron
bombardment on the creep of Inconel at
1500°F and 1500 psi reported previ-
ously. ®> The bench result and an in-
reactor result after annealing at
1700°F are shown in the two lower
curves in Fig. 69; the upper four
curves, reported previously, () are
shown for comparison. Although the
difference in behavior is attributed
to the heat treatment, there was a
difference i1n the manner of annealing
that could have been partly responsible.
The 1650°F anneals were carried out
in sttu and only the gage length was
exposed to the desired temperature, so
some long-time metallurgical changes
in and beyond the fillets could have
occurred. During the 1700°F anneals
the full length of the specimen was
heated in a large furnace,

The possibility of some metallurgi-
such as grain growth,
increased creep rate
following the 1650°F annecal
unlikely in Inconel. The sluggishness
of the recrystallization process 1in
Inconel when only small amounts of
cold reduction have preceded the
anneal 1s well known. Although the
metallurgical history of the plate
from which these specimens were made

cal change,
causing the
1s not

is unknown, hardness measurements
indicate that it received about a 5%
cold reduction following the penultimate
anneal, Evidence of structural changes
over long periods was found in the
marked increase 1n intensity of all
x-ray diffraction lines following a
1000-hr soak at 1500°F after the usual
anneal of 2 hr at 1700°F, Diffraction
patterns of the outer fibers of both
compression and tension sides of all
the Inconel bench test specimens have

been made, More data are required for

 

(35, €. Wilson, J. C. Zukas, and ¥. W. Davis,
Solid State Division Quarterly Progress Report for
Period Ending May 10, 1852, ORNL-1301 {in press).
FOR PERIOD ENDING SEPTEMBER 10, 1952

SSD-4-472
DWG. 162754

 

Q.15

 

l ’ INDICATES REACTOR SHUTOOWN

—
1

 

Q
3
TN

EXTENSION (%}

  
 

SPECIMENS A, B,C, AND D ANNEALED 2hr AT {650°F -
SPECIMENS £ AND £ ANNEALED 2hr AT 1700°F

 

 

 

0.05 /

|
A, B, AND F ORNL GRAPHITE REACTOR TESTS AT A FAST FLUX
oF 4x10'¢ n/em? - sec

|
C,D, AND £ BENCH TESTS

 

 

 

 

 

 

 

 

 

 

 

 

C
I e —— . F
T 1 H
o .
0 200 400 600
TIME (hr)
Fig. 69. (Cantilever Creep Tests of Inconel at 1500°F and 1500 psi.

a detailed analysis, but it has been
noted in all cases that the (111) line
is much more intense on the compression
than on the tension side. No change
in the lattice constant of Inconel was
found within the accuracy of measure-
ment, approximately 1 part in 3000,

Strain-time curves for Inconel -
annealed at 1700°F for 2 hr have been
obtained in the graphite reactor and
the LITR at 2000 psi and 1500°F., The
data indicate that a number of ad-
ditional experiments are necessary in
order to fully explore the variables,

155
RADIATION EFFECTS ON THERMAL
CONDUCTIVITY

A. Foner Cohen L. C. Templeton
Solid State Division

Absolute thermal conductivity
measurements were made on an ARE heat-
treated specimen on Inconel by using a
radial-heat-flow method at tempera-
tures up to 1240°F before irradiation,
at temperatures up to 1500°F during
irradiation in hole HB-3 of the LITR,

and at various temperatures immediately

156

following irradiation. The fast flux
in the HB-3 hole is of the order of
10'? neutrons/cm?*sec. The thermal
conductivity values calculated, so
far, from the results of this experi-
ment show that before and after
irradiation the thermal conductivity
values are the same within the accuracy
of the measurements, which is of the
order of a few per cent. The ‘during-
irradiation’ data show no change in
the conductivity as a function of time
at constant temperature, and they
agree with the out-of-reactor measure-
ments,
 
 

 

 

 

 
SUMMARY AND

The analytical chemistry program
in support of the materials research
and equipment development programs
included the routine analysis of 1179
individual samples, as well as the
development of new analytical procedures
(sec. 14). 1In particular, new methods
have been developed for the quantitative
determaination of the constituents and
impurities in fluoride fuels. The

INTROBUCTION

work is concerned with the
improvement and simplification of
these analytical techniques.

current

The list of reports that have been
issued by the project during the last
quarter includes 17 formal reports and
56 informal documents (not including
internal documents) on all phases of

ANP research at OBNL {sec. 15).

 

14.
C. D.

ANALYTICAL CHEMISTRY

Susano

‘Analytical Chemistry Division

Progress has been made in improving
and simplifiing the methods of analyses
of reactor fuels. A volumetric method
has been developed for the determination
of alkali metals that depends upon
separation of the elements by ion
exchange and their indirect determi-
nation by titration of an equivalent
quantity of chloride ion. The in-
vestigation of a volumetric method for
the determination of zirconium in
fuels has continued., Zirconium in
fuels canbe determined by a volumetric
method, 1n the absence of uranium,
with an accuracy of about 1%. Tt
appears that this determination can
also be performed in the presence of
uranium, and this phase of the problem
is still being investigated. The use
of silver peroxide as an oxidant in
the determination of chromiuw by
means of diphenylcarbazide has been
studied, and a new and improved method,
based on this study, has been adapted
for the determination of chromium.
Attempts to determine nickel i1n reactor
fuels by means of nitroso salicylic
acrd were not successful,. '

A method similar to: the Dean and
Stark method¢!? has been developed
for the determination of trace amounts

of water 1n finely divided solzid
materials. This method 1s particularly
valuable for use with some fluorides
with which other methods fail. The
products of the reaction between NaK
and reactor fuels have been studied.
Tt was found that uranium{IV) is
reduced to UF,, and at least one other
constituent of the fuel,
be zirconium{1V),

which must

15 also reduced.

A review of the service analyses
performed during the
presented.

quarter 1is

ANALYTICAL STUDIES OF COMPONENTS
OF FLUGRIDE MIXTURES ‘

R. Rowan, Jr. €. M.
J. C. White W.
C. K. Talbott

Analytical Chemistry Division

Boyd
J. Boss

Alkali Metals. The emphasis on
ternary mixtures of fluoride salts
containing only a single alkali metal
fluoride has greatly simplified the
problem of the determination of alkali

 

(D oy,
Chen,

Dean and D. D.

Stark, J., Ind.
12, 486 (1920). '

Eng.

159
ANP PROJECT QUARTERLY PROGRESS REPCORT

metals and has reduced the need for
methods of separation of alkali metals
from each other. An indirect method
was, developed for the
separation and determination of sodium
and potassium that involved the
titration of the chloride equivalent
tothe alkali metal 1on, The separation
1s accomplished by means of 10n-
exchange resins. Although the sepa-
ration is time-consuming, by using
several columns the time for each
determination 1s far less than 1s
required by the conventional techniques
for this separation,

however,

Zirconium. A volumetric method for
the determination of zirconium that
makes use of the fluoride complex was
suggested by Sawaya and Yamashita.(2)
In essence, the procedure used in this
study 1s to precipitate zirconium
hydroxide by the addition of sodium
hydroxide, adjust the pH of the slurry
to 8.5, add a 15- to 20-fold excess
of potassium fluoride, neutralize the
potassium hydroxide formed with excess
standard nitric acid solution, and
titrate the excess acid to a pH of 7.5
with standard sodium hydroxide so-
lution. The reactions involved are
not stoichiemetric, and as a result
the method must be standardized
empirically.

Data gathered thus far have shown
that by standardization against a
known quantity of zirconium 1t 1is
possible to titrate the element with
an error of less than 1%, under the
conditions described, for a range of
zirconium concentration of 15 to 60 mg
in a volume of 200 ml. The error for
larger amounts has not been determined
with any degree of certainty, as yet.

The additionof uranium to zirconium

sulfate solutions results 1n some

 

inter ference, since uranium forms a
slightly soluble hydrous oxide.
{Q)T. Sawaya and M. Yamashita, J. Chem. Soc.

(Japen) 72, 414-16 (1951}); C. A. 46 1911 (1972},

160

Compensation for this interference was
made on the basis that 2.22 moles of
sodium hydroxide(3?’ 1is consumed to
precipitate each mole of uranium,
The results on samples containing
3, 6, and 12 mg of uranium per 60 mg
of zirconium agreed within 1% after
the application of the necessary
correction factor for uranium. The
application of this method to sulfate
solutions of reactor fuels showed that
the results for zirconium by the
volumetric procedure (with uranium
correction) were generally 3 to 5%
higher than those obtained gravimetri-
cally by the phenylarsonic acid
precipitation method, The study of
the application of this method 1s
continuing.

ANALYTICAL STUDIES OF IMPURITIES
IN FLUORIDE MIXTURES

Chrominm, The use of silver
peroxide¢*) as an oxidant for chromi-
um(III) in the colorimetric determi-
nation of chromium by diphenylcarbazide
has been investigated, and a procedure
in which silver nitrate and potassium
persulfate solutions are employed in
the oxidation step was adapted for the
determination of chromium 1n reactor
fuels. 'The precision of the method
1s equal to that obtained when any
other oxidant i1s used, and the method
is probably the most rapid of any
currently 1n use. An additional
feature is that the chromium(VI)
formed 1is stable for at least 24 hr
after oxidation, which 1s not true in
the case of a number of the other
oxidants commonly employed in this
determination. This work is described
elsewhere(5) in greater detail.

 

(S)B. Kunin, The Potentiometric Titretion of
Tuballyl Ion with Alkali, A-3294 (May B, 194°7).

(4)M. Tanaka, Bull. Chem. Seec. Japan 23,
165~8 (19°0).

(S)J. C. Yhite, C. M. Boyd, W. J. Ross, and
C. K. Talbott, " Analytical Studies of Reactor
Fuels and Their Components,” Analyticael Chemistry
Division Quarterly Progress Report for Period
Ending June 26, 1952, ORNL-1361 (in press).
FOR

Nickel. The commonly used di-
methylglyoxime method, with certain
modifications in procedure to adapt
it to use with fluoride fuels, 1s
employed for the determination of
nickel. Because the quantity of
nickel appearing as a corrosion
product in the fuels currently being
tested i1s generally smaller than has
been the case in the past, the lower
limit of determination of nickel has
become more critical.. A practical
limitation is imposed on the sample
size used because of the presence of
zirconium and the high acid concen-
tration necessary to prevent its
hydrolysis. The lower limit of de-~
termination of the present method is
10 ppm, and the use of absorption
cells of 5~cm path length is necessary
to achieve this figure,

A brief investigation was made of a
method for the determination of nickel
that involves extraction of nickel at
a pH of 5 to 6 with nitroso salicylic
acid.{®) The obvious advantage of
this method is that the color can be
developed in slightly acidic solution
instead of the strongly basic
lution requiredin the dimethylglyoxime
procedure. Unfortunately, the zirconium
concentration is so high that the
large quantity of complexing agent
(oxalate, tartrate, or fluoride)
retards the formation of the colored
complex. It appears that the use of
5-cm absorption cells will be necessary

S0 -

in most cases to obtain satisfactory
absorbancy measurements for nickel 1in
the range of less than 50 ppm.

0xygen, The determination of
oxygen in reactor fuels 1s a problem
of considerable importance, since the
presence of oxygen 1is generally con-
sidered to be a contributing facter
in the corrosionof container materials.
Bromine trifluoride i1s known to react,

(6)y. H. Perry and E. J.

Serfass, Anal. Chen.
22, 565-67 (19%50). '

PERIOD ENDING SEPTEMBER 10, 1952

with the liberation of oxygen, with
virtually all the oxides that can
possibly be present in these reactor
fuels (U0,, ZrO,, Fe,0;, Cr,0;, Ni0),
and 1t 1s believed that this reaction
can be made to serve as a basis for a
determination of oxygen.

An apparatus in which BrF, will be
reacted with the fuel at elevated
temperatures of the order of 325°C
is being fabricated at the present
time. The reaction bomb and the high-
pressure system are constructed of
nickel. The oxvgen liberated from the
reaction will be determined by a
simple gasometric measurement of the
oxygen pressure 1n an evacuated system
at a reduced temperature. This method
appears favorable because of 1its
simplicity and the relatively short
time required for each determination.
A pyrex unit to perform this operation
has been fabricated and leak-tested,
If unexpected difficulties arise,
however, other methods for the de-
termination of the oxygen evolved in
the reaction are available.

Chiloride. The zirconium tetra-
fluoride used in the preparation of
reactor fuels 1s prepared by the
hydrofluorination of zirconium tetra-
chleride. Since chlorine 1s an
undesirable impurity in fuels because
of 1ts rather high neutron-capture
cross section, it 1s necessary to
determine chloride 1in
tetrafluoride to
concentration 1s
limits.

zirconium
assure that 1ts
beleow tolerable

Several attacks were made on this
problem. A semiguaptitative test
was devised in which a turbidimetric
comparison was made between a hydro-
fluoric acid, silver nitrate solution
of the unknown sample of zirconium
tetrafluoride, and standard quantities
of chloride in similar solutions, As
little as 0.01% of chloride imparted
an easlly visible cloudiness to the

161
ANP PROJECT QUARTERLY PROGRESS REPORT

solution. It 1is significant that none
of the samples of zirconium tetra-
fluoride tested so far have produced
a visible cloudiness, which indicates
a chloride content less than 0.01%
for these samples.

A second method of attack was to
oxidize the chloride present to
chlorine and determine it iodometri-
cally. By using a sample of suitable
size, 1t was possible to determine as
little as 0.05% chlorine. Potassium
permanganate, ceric sulfate, and lead
dioxide were tested as oxidants,
Permangante 1s generally unsatis-
factory; 1t yields results averaging
20% haigh. This 1s believed to be
caused principally by decomposition of
permanganic acid during refluxing.
Ceric sulfate gave low results 1in
most cases, as did lead dioxide, which
was only briefly tested.

The semiquantitative test yields
results of sufficient accuracy for
the present purposes.

Water. A method for the determi-
nation ol trace amounts of water 1n
fluorides, oxides, and other solids
in a finely divided state has been
developed., Such a method was princi-
pally required for the analysis of
the mixture to be used for cold
critical experiments and 1ts components,
Water 1s extracted from the solids by
refluxing with xylene by an adaptation
of the Dean-Stark method.(!) The
xylene-water azerotrope is distilled
into methyl alcohol and the water 1is
titrated with Karl Fischer reagent.
This method was adopted in preference
to the direct titration of the solid
with Karl Fischer reagent partly
because of sample size limitations
and partly because investigation
showed that the completeness af the
reaction was uncertain in the direct
method. Tgnition methods for the
determination of water were unsuccessful

due to the hydrolysis of fluoride

162

salts that takes place at elevated
temperatures.

The extraction method has proved
to be completely satisfactory. Samples
of the order of 20 g can be handled,
1f necessary. The blank, although
accounting for 50 to 75% of the total
titration 1n some samples, has been
reproducible to within less than 1%.
The method has been tested against
standard sodium tartrate, with the
recovery of water averaging mnearly
90%. In addition to NakF, Zr0,, C,
UF,, and KF, the method has been used
successfully on fluorides such as

NH,F, NH,HF,, NiF,, KHF,, and K,FeF,.

DETEEMINATION OF CARBON IN ZrF, AND
270 ,-NaF-C MIXTURES

In determining carbon in zirconium
fluoride and i1n Zr0,-NaF-C mixtures
by combustion, the fluorides interfere.
This difficulty has been overcome
by removing the fluorides by dis-
solution. Zirconium fluoride 1s
dissolved by oxalic acid and the
alkali flvorides by warm water. The
residue 1s collected 1n a swmall,
gunartz, filtering crucible and the
carbon content i1s determined by
combustion. Since large samples may
be used, the accuracy of the carbon
determination 1s increased.

The fluoride in the filtrate from
the ZrO,-NaF-C mixturesis precipitated
as lead chlorofluoride, dissolved, and
titrated with mercuric nitrate. The
fluoride and carbon values, which can
be determined rapidly by the method
outlined, give an indication of the
degree of homogeneity attained in the
preparation of large batches by
mechanical mixing.

COMPATIBILITY OF REACTOR FUELS AND Nak

A number of experiments has been
conducted to determine the reaction
FOR PERIOD ENDING SEPTEMBER 10, 1952

products of reactor fuels (NaF~ZrF¢-
UF,) and NaK under static and dynamic
conditions., The conclusions drawn
from the chemical analyses were:
(1) NaK reduces U{IV) to UF,, which
separates from the molten mass and
settles to the bottom of the container;
{2) the greater the concentration of
NaK, the more complete the reduction;
(3) a nearly linear relationship
exists between the hydrogen evolved
from the fuel upon acidification and
the amount of NaK added to the fuel:
(4) NaK, 1in large excess over that
required to reduce quantitatively the
UF, to UF,, reduces other components
of the fuel, which form compounds that
alse evolve hydrogen from acidic
solutions; (5) less than 50% of the
reducing power of the NaK added was
recovered in any test, '

SERVICE ANALYSIS

H. P. House L. J. Brady
W. F. Vaughn

Analytical Chemistry Division

The analyses of zirconium fuels
and of zirconium fluoride have received
major attention during the quarter,
Application was made of the methods

adaptedby the research and development
group for determining sulfide, sulfur,
carbon oxide, water, and chloride in
fused zirconium-bearing fluoride salts.
In addition to zirconium samples,
many samples of alkali fluorides 'and
quaternary eutectics composed of
alkali fluorides and uranium tetra-
fluoride were analyzed. Unrelated
miscellaneous samples were tested: a
number of metals and alloys;
fluorides of iron,
nickel; 1nert gases;
bonding cement.

complex
chromium, and
ceramics; and

0Of a total of 1179 samples analyzed
during this quarter, 80% was for the
reactor chemistry group, 15% was for
the experimental engineering group,
and the remaining 5% was for the ANP
critical experiments group, the
electromagnetic research group, and
the maintenance shops.

A summary of the service analyses

for ANP follows:

 

 

Semples on hand May 10, 1952 198
Number of samples received 1192
Total number of samples 1390
Number of samples reported 1179
Backlog as of August 10, 1952 211

163
o

ANP PROJECT QUARTERLY PROGRESS REPORT

REPORT NO.

CF-52-4-191

ORNL- 1234

Y-F12-7

Y-F15-10

ORNL~ 1241

Y-F17-16

Y-F17-17

Y-F17-18

Y-F17-20

Y-F17-23

Y-Fl0-101

Y-Fl0-102

Y-F10-103

GRNL- 1320

Y-FI10-106

Y-F10-109

164

15. LIST OF REPORTS ISSUED
TITLE OF BREPORT AUTHOR(S)
I. GENERAL DESIGN
Hydrodynamics of Homogenecous Reactors G. F. Wislicenus

(Presented at Second Fluid Fuels Con-
ference, Osk Ridge, Tenn., April 1952)

Reactor Program of the Aireraft Nuclszar
Propulsion Project

Proposed ARE Design
Three Reactor~Heat Exchanger~Shield

Arrangements for Use with Fused Fluoride
Circulating Fuel

I1X. EXPERIMENTAL ENGINEERING

Fuel Flow Studies in ARE Fuel System Mockup
Study of Fluid Flow Measuring Devices
Moore Nullmatic Pressure Transmitter Test

Line Plug Test No. 1 - Fuel at 1500°F

Tank Car Helium as a High Purity Helium

Scource

Production of Fluorides in Buildings 9211
and 9201-3

III. REACTOR PHYSICS
Values of Resonance Integrals
An QOutline of the General Methods of Re-
action Analyses Used by the ANP Physics
Group
Statics of the ARE Reactor, Summary Report
The Effect of Gaps on Pile Reactivity
Reacteor Theory Terms

Note on the Non-Linear Kinetics of

Circulating-Fuel Reactors

™

B. Cottrell

F. Haines

. P. Fraas

H. Ward
A. Anderson
W. Taylor

W. Taylor

A. Maon

W. Savage

S, Wilson

B. Mills

B. Mills
Tamar

B. Mills
B. Arfken

Tamor

DATE ISSUED

4-30-52

6=-2-52

6-10-52

6-30-52

1-9-52

71-7=-52

71-10-52

1-10-52

B-4-52

8-6-52

4~30-52

5-16-52

5-8-52

8-14-52

7-16=-52

8-15-52
REPORT NO.

ORNL- 1304

Ck-52-6~99

ORNL-~1142

OBNL- 1312

Y-895

Y-B23-2

Y-B23-5

Y-B23-7

Y-F10-108

Y-B23-9

CF-52~3-125

CF-52-3-146

CF-~52-4-99

ORNL- 1147

CF-52-6~54

FOR PERIOD ENDING SEPTEMBER 10, 1952

TITLE OF REPORT

IV. NUCLEAR MEASUREMENTS

Thermal Neutron Capture Cross Sections of
Isotopes

Energy Absorption of Capture Gammas

Multiple-Crystal Gamma-Ray Spectrometer

The Fission Cross Section Qf Uranium- 234
and Uranium~ 236 for Incident Neutron
Energies up to 4 Mev :

Crosa Sections for Carbon and Water in the

Energy Range from 2.3 Mev to .025 ev. A
Literature Search

V. CRITICAL EXPERIMENTS

Preliminary Direct Cycle Reactor Assembly-
Part 11

Preliminary Direct Cycle Reactor Assembly-
Part Y111

Preliminary Direct Cycle Reactor Assembly-
Part [V

The ARE Critical Experiment

Loading of ARE Critical Experiment Fuel
Tubes

VI.  SHIELDING

OBNL “Tower Shielding Facility” Preliminary
Proposal

Health Physics Instruments Recommended
for ASTF

Some Ground Scattering Experiments Performed
at the Bulk Shielding Facility

The Unit Shield Experiments ‘at the Bulk

Shielding Facility

Data on ¥fnriched Fuel E]ements for the Bulk
Shielding Facility

AUTHOR(S)

H. Pomerance

H. L. F. Enlund
F. C. Maienschein

R. W. Lamphere

Frances Sachs

A. D. Callihan
A, D. Callihan
A. D. Callihan

and assoclates

C. B. Mills
‘D, Scett

Dunlap Scott

T. H. J. Burnette

H. E. Hungerford

. Meem

H. E. Hungerford

Cunningh am

DATE ISSUED

6-6~52

6-11-52
7=3=52

7-15-52

8-4-52

5-21-52
6~-18-52
5-30-52
B-8~52

8-12-52

3-14-52

3-20-52

4-16-52
5= 14-52

6~10-52

165
ANP PROJECT QUARTERLY PROGRESS REPORT

REPORT NO.
CF-52-7-150

CF-52-8-38

Ck-52-6-145

ORNL.-1283

ORNL=-1273

CF-52-6-158

ORNL~ 1133

Part 2

CF-52-6-T4

CF-52-6-158

CF-52-6-165

CF-52-7T-1

CF-52-7-37

CF-52-7-71

CF-52-7-83

CF-52-5-40

CF-52-5-41

CF-52-5-163

CF-52-5-183

166

TITLE OF REPORT
Uranium Penetration of Graphite

Gamma Ray Spectral Measurements with

Divided Shield Mockup. Part TII

Basic Research in Shielding

General Principles of a Proton Recoil-Fast
Neutron Spectrometer

Background Calcul ations for the Proposed
Tower Shielding Facility

Neutron and Gamma Dose Distribution in Water
Surrounding the GE Outlet Air Duct

The Shielding of Mobile Beactors-II

Re: Design of Inexpensive Shield for Small
Reactor

Neutron and Gamma Dose Distribution in Water

Phenomenological Theory of the Attenuation
of Neutrons by Air Ducts in Shields

Preliminary Data on GE Qutlet Air Duct

Air Scattering Experiment at the Bulk
Shielding Facility

Gamma Ray Spectral Measurements with the
Divided Shield Mockup, Part II

Gamma Measurements on the GE Qutlet Air
Duct as a Function of Length of Transition
Section

Gamma Dose Behind Iron-Water Thermal Shield
of Various Thicknesses as a Function of
Water Reflector Thickness.
for NDA)

{Investigation

Gamma Dose Behind Iron-Borated Water Thermal
Shield with 18-, 20-, and 30-cm Water
Reflector

(zamma Attenuation in [.ead-Water Shields

Neutron and Gamma [Mstribution in Water

Containing Lead, Steel and Air

AUTHOR{(S)

. K. Hullings

C. Maienschein

. P. Blizard

R. Gossick

Simon

H. Ritchie

E. Clifford
I.. Storrs, Jr.

. P. Blizard

A. Wel ton

Ralph Balent

C.

E. Clifford

Simon

E. Clifford

E. Clifford

—
T

. Meem

E. Hungerford

C. Maienschein

E. Clifford

E. Clifford

E. Clifford

E. Clifford

E. Clifford

DATE ISSUED
7-31-52

8-8-52

6-25+52

8-14-52

9/1952

6-26-52

6-30-52

6=-26-52

6-25-52

7-1-52

71-8-52

7-8-52

7-16-52

5-7-52

5=T7-52

5-20-52

5-26-52
REPORT NO,

CF-52=-5-1
Part 8

CF-52-6~120
Part 1

CFa52-6-120

CF-52-8-166

ORNL~ 1370

CF-~52-6-148

CF-52-7-138

ORNE-1291

CF-52-6-76

ORNL-1163

CF-52-6-~127

ORNL.- 1286

Y-B31-354

FOR PERIOD ENDING SEPTEMBER 10, 1952

TITLE OF REPORT

Nentron and Gamma Dose Distribution Beyond
Beryllium Slab {28.5 cm, Density 1.23 gm/ec)

Sources of Radiation, Chapter 2.1 of Reactor
Handbook :

Geometry, Chapter 2.7 of Reacitor Handbook

VII.

A Review of the Literature on Heat Transfer

in Noncircular Ducts and Annuli for Ordinary

Fluids and Liquid Metals

Turbulent Forced Convection Heat Transfer in
Circular Tubes Containing Molten Sodium
Hydroxide

Liquid Metals Handbook, Chapter 6, ‘Liquid
Metal Heat Transfer”

Viscosity of Fulinak

Viscosity of Fuel Salt Mixtures No. 27 and

No. 30

VITI. MATERIALS CHEMISTRY
General Information Concerning Hydroxides
Complex Fluoride Fuel Studies - X-ray

The Preparation of Thin and of Thick Targets
to b: Bombarded by Positive Particles

Petrographic Examination of Fluoride Fuels

IX. ANALYTICAL CHEMISTRY

The Determination of Oxygen in Sodium

Analytical Chemistry - ANP Program Quarterly
Progress Report for Period Ending May 26,
1952

AUTHOR(S)
C. E. Clifford
E. P. Blizard
F. C. Maienschein

HEAT TRANSFER AND PHYSICAL PROPERTIES

H. C.

H., W
R, N
H. F
B. F
R. F
Mary
Paul
R. A
T. N
J. C.
w. J

R. Rowan,

M. T.

Blizard

Claiborne

. Hoffman

. Lyon
. Poppendiek

. Redmond:

Redmond

E. Lee

Agron

Bolomey

. McVay

White
Ross

Jr.

Kelley

DATE ISSUED

f-6=52

6-23~52

6+ 2452

to be i1ssued

to be 1ssued

9-1-52

6~19-52

7-23-52

4+21-52

6~ 16-52

6~3-52

6~20-52

4-30-52

5~26-52

167
REPORT NO.

OBRNL- 1307

MM 2

MM=-6

MM-T7

Cr-52-7-124

CF-52-3~123

Y-872

OBRNL- 1114

Y-889

OBRNL-1294

Y-F26-40

168

TITILE OF REPORT AUTHOR(S)

L. McCutcheon
D. Susano

Effect of Fluoride on the Gravimetric R.
Determination of Zirconium C.

X. METALLURGY

Cone Arc Welding P. Patriarca
MetalJographic Examination of 347 Stainless E, E, Hoffman
Steel Heat Exchanger from Forced Convection
l.oop which Operated for 3,000 Hours in NaK
Metallographic Examination of Nicrobrazed E. E. Hoffman
Stainless Steel Sodium-Air Heat Exchanger
Following Failure During Test
Beta Treatment of Alpha Uranium R. J. Gray
A Simplified Apparatus for Making Thermal A. DeS. Brasunas
Gradient Dynamic Corrosion Tests (See-Saw
Tests)

Selected Physical Properties of Lead in the Frances Sachs
Temperature Range 100-1000°C,

Search

A Literature

Stress+Strain-Time Phenomena in Mechanical A. G. H. Andersen

Testing

Selected Physical Properties of Mercury in Frances Sachs
the Temperature Range 100-1000°C. A

literature Search,

XI. MISCELLANEOUS

Aircraft Nuclear Propulsion Project Quarterly W, B. Cottrell
Progress Report for Period Ending June 10,

1552
W. B. Cottrell

ANP Information Meeting of August 20, 1952

DATE ISSUED

6-17-52

7-15-52

7-16-52

T-24-52

3-14-52

5-7-52

5-9-52

7-10-52

- 8-5-52

8-29-52