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ORNL-1170

This document consists of 191 pages.
Copy ~7 of 208  Series A.

Contract No. W-7405, eng. 26

AIRCRAFT NUCLEAR PROPULSION PROJECT
QUARTERLY PROGRESS REPORT
for Period Ending December 10, 1951

R. Co Bl‘iant
Director, ANP Project

Edited by:
W. B. Cottrell

DATE ISSUED

 

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ORNL-1170
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ORNL-1170
Progress

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iv

 

Reports previously 1ssued

ORNL- 528

ORNL-629

ORNL-768

ORNL-858

ORNL-919

ANP-60

ANP-65

ORNL-1154

Period

Period

Period

Period

Period

Period

Period

Period

in this series are as follows:

Ending November 30, 1949
Ending February 28, 1950
Ending May 31, 1950
Ending August 31, 1950
Ending December 10,~1950
Ending March 10, 1931
Ending June 10, 1951

Ending September 10, 1951

 
 

TABLE OF CONTENTS

FOREWARD

PART I REACTOR THEORY AND DESIGN

SUMMARY AND INTRODUCTION

1.

CIRCULATING-FUEL AIRCRAFT REACTOR

Airplane and Overall Arrangement
Reactor '

Engine

Radiators

Shielding

Accessory Systems

LIQUID-METAL-COOLED ATRCRAFT REACTOR EXPERIMENT

Core—Reflector—Pressure Shell
Fluid Circuit
Reactor Control
Control system
High temperature fission chamber
Reactor dynamic computer

Instrumentation

ARE Building Faciliaty
Remote-Handing Equipment

REACTOR PHYSICS

Reactor Calculations on IBM Equipment
Circulating-Fuel Reactors
Beryllium~oxide moderated reactor
Water-moderated reactor
Comparison of H,0- and Be0-moderated reactors
Effect of delayed neutrons on kinetic response
Circulating-Moderator Reactors
Errors in Reactor Physics Calculations

 

PAGE

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4.

vi

 

CRITICAL EXPERIMENTS

Preliminary Assembly of Direct-Cycle Reactor
Graphite Heactor

NUCLEAR MEASUREMENTS

The 5-Mev Van de Graaff Accelerator
Measurements of the (n,2n) Reaction in Beryllium
Time-of-Flight Neutron Spectrometer

EXPERIMENTAL REACTOR ENGINEERING

Pump Development
Centrifugal pumps for figure-eight loops
ARE centrifugal pump design
Canned-rotor pump
Frozen-sodium-seal pump
Two-stage electromagnetic pump
Electromagnetic pump cell development
Seal Tests
Frozen-sodium seal
Frozen-fluoride seal
Graphitar ring~--ketos tool steel gas seal
Test Loops
Calibration loop
Sodium manometer loop
Self-welding tests
Materials tests
Valve tests
Heat-Exchanger Tests
NaK to NaK heat exchanger
Sodium to air radiator
Liquid-Fuel Systems
Instrumentation
Level control and indication
Flow measurement
Pressure-measuring devices

Full-Scale ARE Component Test Facilities

 

PAGE

35

35
37

38

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47
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48
48
438
48
 

Fluoride Production

Cleaning of Fluorides from Systems
NaK Disposal '
Alkali Metals Manual

PART I1 SHIELDING RESEARCH

SUMMARY AND INTRODUCTION

7.

BULK SHIELDING REACTOR

Reactor Operation
Mockup of the Unit Shield
Mockup of the Divided Shield

DUCT TESTS

Air-Filled Duct Tests in Lid Tank
Cylindrical ducts
Annular ducts

Liquid-Metal Duct Test in Thermal Column

SHIELDING INVESTIGATIONS

Tower Shielding Facility Proposal
Circulating-Fuel Reactor Shields
NDA Divided-Shield Studies

PART 111 MATERIALS RESEARCH

SUMMARY AND INTRODUCTION
10. CHEMISTRY OF HIGH-TEMPERATURE LIQUIDS

Low Melting-Fluoride Fuel Systems
LiF-KF-UF,
LiF-NaF-KF-UF,
LiF-NaF-RbF-UF,
NaF-BeF,-UF,
LiF-NaF-BeF,UF,

Tonic Species in Fused Fluorides

 

PAGE

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11.

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Experimental procedures
Results of electrolyses
Homogeneous Fuels
Uranium solubility in hydroxide-borate mixtures
Uranium solubility in hydroxides with various additives
Moderator-Coolant Development
Preparation of pure hydroxides
KOH-LiOH
Hydroxide-fluoride systems
Coolant Development
RbF-LiF
NaF-Bel,
NaF-KF-RbF
KF¥-LiF-RbF
NaF-LiF-RbF
Ternary systems containing BeF,
Service Functions

CORROSTION RESEARCH

Static Corrosion by Fluorides
Static Corrosion by Hydroxides
Corrosion of uncoated metals
Corrosion of coated metals
Dissolution of metals in sodium hydroxide
Static Corrosion by Fluoride-Hydroxide Systems
Static Corrosion by Sodium Cyanide
Static Corrosion by Liquid Metals
Sodium on stainless steel
Lithium on coated stainless steel
Static Corrosion of Fuel Capsules in Sodium
Dynamic Corrosion in Thermal Convection Loops
Fluoride corrosion
Hydroxide corrosion
Fundamental Corrosion Research
EMF measurements in hydroxides
Polarographic study of sodium hydroxide
Survey of the mass-transfer phenomena

 

PAGE

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82
83
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106
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111
 

PAGE

12. HEAT-TRANSFER RESEARCH AND PHYSICAL PROPERTIES 115
Natural Convecticn in Liquid-Fuel Elements 115
Heat-Transfer Coefficients 116

Heat transfer in fused hydroxides and salts 116
Heat transfer in molten lithium ‘ 117
Entrance-region heat transfer in a sodium system 118
Heat transfer in a circulating fuel system 120
Heat Capacity 120
Thermal Conductivity 120
Thermal conductivity of liquids 120
Thermal conductivity of solids 122
Thermal conductivity of diatomaceous silica powder 122
Density of Liquids 123
Viscosity 123
Falling-ball viscometer 124
Zahn type viscometer 124
Viscosity of fluoride mixtures 126
Vapor Pressure of Liquid Fuels 126

13. METALLURGY AND CERAMICS 128

Solid-Fuel-Element Fabrication 128
Effect of U0, particle size 129
Effect of rolling temperature 130
Effect of elimination of capsule 130
Preparation of tubular fuel elements 132

Welding Techniques 132
Cone arc welding 132
Resistance welding 132

Brazing Techniques 135
Effect of brazing time 135
Brazing of clad fuel elements 135
Nickel-palladium brazing alloy 135

Creep and Stress Rupture of Metals 137
Operation of creep and stress-rupture equipment 138
Creep-rupture tests of inconel 139
Creep of nickel Z 140
Stress-rupture tests 140

ix

 
14.

 

Ceramics Laboratery
Equipment
Subcontract work

RADIATION DAMAGE

Irradiation of Fused Materials
Pile irradiation of fuel and KOH capsules
Cyclotron irradiation of fuel and KOH capsules
In-Pile Circulating Loops
Creep Under Irradiation
Radiation Effects on Thermal Conductivity

PART IV APPENDIXES

SUMMARY AND INTRODUCTION

15.

16.

SUPERCRITICAL-WATER REACTOR

Analysis of Supercritical Water Reactor by NDA
Stability
Start-up

Analysis of Supercritical-Water Reactor by ORNL
Compressor-jet cycle
Results

ANALYTICAL CHEMISTRY

Studies of Diatomaceous Earth
Determination of Ni, NiO, and O in Alkali Hydroxides
Available oxygen
Metallic nickel
Nickel oxide
Studies of Ternary Alkali Fluoride Eutectic
pH of aqueous solutions of alkali fluorides
Composition of the euntectic
Metallic impurities
Corrosion of Metal Containers by Hydroxide Solutions
Determination of Uranium Trifluoride
Determination of Oxygen in NaK

 

PAGE

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150

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155
 

PAGE

Determination of Oxygen in Helium 156
Preparation of Oxygen-Free Sodium Samples 156
Clarity of Borated Water in Concrete Tanks 157
Analytical Services 157
17. LIST OF BEPORTS ISSUED 158
18. DIRECTORY OF ACTIVE ANP RESEARCH PROJECTS AT ORNL 161
Reactor and Component Design 161
Shielding Research 164
Materials Research 165
Analysis of Other Nuclear Reactor Systems 172
19, TECHNICAL ORGANIZATION OF THE ANP PROJECT 173

 
10.

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x11

N e W N

 

LIST OF TABLES

TITLE PAGE

Features of the Circulating-Fuel Aircraft Reactor 8

Performance of Engines at Design Point, Mach 1.5 at

45,000 ft 9

Design Values for the BeO-Moderated Circulating-Fuel
Reactor 14

Design Values for the NaF-UF, - Cooled Water-Moderated
Reactor 17

Design Values for the Water-Moderated Circulating-Fuel
Reactors for Use with the NaF-BeF, and LiF-KF-NaF

Fuel-Coolants 22
Comparison of BeO- and H,0-Moderated Reactor 29
The NaOH-Cooled and —~ Moderated Reactor 31

Reactivity Calculation Results on the Alkali Hydroxide
Reactors 32

Critical Assembly Sizes, Composition, and Calculated
Multiplication Constants 33

Summary of Promising Fluoride Fuel Systems of Low Uranium

Content 30
The Pseudo-Binary System (LiF-NaF-KF)-UF, 81
The Pseudo-Binary System (LiF-NaF-RbF)-UF, 81
The Pseudo-Binary System (LiF-NaF-BeF,)-UF; 81
Solubility of Uranium in Hydroxide-Borate Mixtures 83
Effect of Various Additives on the Solubility of Uranium

in Hydroxides 84
Container Filling Services 87

 
TABLE

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10

11

- b Q2 N e

 

TITLE

Summary of Corrosion by Ba{OH), at 816°C for 100 hr
Summary of Corrosion by Sr(OH), at 816°C for 100 hr

Summary of Corrosion by Sodium Hydroxide at 815°C for
100 hr

Summary of Corrosion by Lithium Hydroxide at 816°C for
100 hr

Summary of Hydroxide Corrosion of Clad Metal Specimens at
816°C for 100 hr

Metal Content of Sodium Hydroxide at Function of
Temperature

Corrosion Data Obtained at 816°C Using Molten Sodium
Cyanide for 100 hr

Static Corrosion Tests in Sodium
Analysis of Inconel Fuel Capsules in Sodium at 800°C

Analysis of 316 Stainless Steel Fuel Capsules in Sodium
at 800°C

Analysis of Hydroxide From Nickel Thermal Convection
Loops After Plugging

Heat Capacities of Various Substances
Data on Several Fused Salt Mixtures

Vapor Pressure of the NaF-KF-UF, Eutectic
Tests on Standard NaF-KF-UF, and KOH

pH Values of Aqueous Solutions of Certain Alkali
Fluorides at 25°C

Summary of Service Analyses

 

PAGE

90
91

02
92
99
102

104
105
106 .

106

113
121
124
127
143

154
157

X111
 

LIST OF FIGURES

FIGURE TITLE PAGE

3.1 Total Uranium Inventory and Uranium in Core as a Function of
the Volume Fraction Fuel Coolant in the Core for Be(0-Moderated
Circulating-Fuel Reactors with 3-, 3 1/2-, and 4-ft-Diameter
Cores 15

3.2 Total Uranium Investment in BeO-Moderated Circulating-Fuel
Reactor as a Function of Reactor Core Diameter for Several
Assumed Holdup Volumes External to Core 16

3.3 Percent Thermal Fissions as a Function of the Fuel-Coolant
Volume Fraction in the 3- and 4-ft-Diameter Cores of the
BeO-Moderated Circulating-Fuel Reactor 17

3.4 Fission Spectrum vs. Lethargy for the BeO-Moderated Circu-
lating-Fuel Reactor with a 3 1/2-ft-Diameter Core 18

3.5 Absorption Spectrum vs. Lethargy for the Be(-Moderated Circu-
Jating-Fuel Reactor with a 3 1/2-ft-Diameter Core 19

3.6 Leakage Spectrum vs. Lethargy for the Be(C-Mcderated Circu-
lating-Fuel Reactor with a 3 1/2-ft-Diameter Core 20

3.7 Neutron Flux Spectrum vs. Lethargy for the BeO-Moderated Cir-
cuwlating-Fuel Reactor with a 3 1/2-ft-Diameter Core 21

3.8 Total Uranium Inventory and Uranium in the Core as a Function
of the Volume Fraction Fuel-Coolant ipn the 3-ft-Diameter-Core
Water-Moderated Reactor with NaF-UF4 Fuel-Coolant 22

3.9 Total Uranium Inventory and Uranium in Core as a Function of the
Volume Fraction Fuel-Coolant in the Core for the Water-Mod-
erated Circulating-Fuel Reactor with 2 1/2-, 3-, and 3 1/2-ft-

Diameter Cores 23
3.10 Fission Spectrum vs. Lethargy for the H,0-Moderated Circulating-
Fuel Reactor with 3-ft-Diameter Core 24

3.11 Absorption Spectrum vs. Lethargy for the H,0-Moderated Circu-
lating-Fuel Reactor with 3-ft-Diameter Core 25

 

X1iv
FIGURE

N~ Al =3 N e =2

. 12

.13

. 14

.15

. 16

~ o B W N

 

TITLE PAGE

Leakage Spectrum vs. Lethargy for the H,0-Moderated Circulating-
Fuel Reactor with 3-ft-Diameter (ore

Neutron Flux Spectrum vs. Lethargy for the H,O-Moderated Cir-
culating-Fuel Reactor with 3-ft-Diameter Core

Total Uranium Inventory and Uranium in Core as a Function of
Volume Fraction Fuel-Coolant in the Core for H,0-Moderated
Circulating-Fuel Reactors with 2 1/2-, 3-, and 3 1/2-ft-
Diameter Cores and LiF-NaF-KF-UF, Fuel-Coolant

Fractional Change in Neutron Flux as a Function of Time for
0, 67, and 80% Decrease in Delayed Neutrons

Effective Multiplication Cohstant as a Function of Volume
Fraction of Stainless Steel in the Core for the 2 1/2-ft-
Diameter-Core NaQOH-Moderated and - Cooled Reactor

Mid-Cross-Section of Second Mockup of Direct-Cycle Reactor
Sodium to Air Radiator

Gamma Radiation Intensity for the Unit-Shield Experiments
Thermal -Neutron Flux for the Unit-Shield Experiments
Fast-Neutron Dosage for the Unit-Shield Experiments
Dependence of the Weight of the Unit Shield on its Thickness
Installation of Divided-Shield Mockup with Reactor in Position
Shadow-Shield Experiment with Sodium Source

Bulk Shielding Facility Fast-Neutron Data

An Array of Four 2-in. Steel Conduit Ducts with Three 90°
Bends

 

26

217

28

30

32
36
46

36
57
o8
60
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Xv
FIGURE

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xvil

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SN e W N

 

TITLE PAGE

Lid Tank Duct Test D-915, X Traverse (Vertical) and Z Center-
line Measurements of Neutron Flux for Two Arrays of 2-in,
Steel Conduit

Lid Tank Duct Test D-10, Y Traverses in H,0O behind 54-in. Rubber
Conduit (2 in. 1.d.) with Two Bends of Variable Radius

Lid Tank Duct Test D-11, Y Traverses (Horizontal) in H,0 behind
52-in. Rubber Conduit (2-3/8 in. i1.d.) with Variable Bends

Lid Tank Duct-Test Annular-Duct Rectangular Cross-Section, 1/8-
in. Steel Plate Welded

Lid Tank Test D-12, Y (Horizontal) and Z Traverses behind Amnular
Duct

The System KOH-LiOH

The System RbF-LiF

The System NaF-BeF,

The System NaF-KF-RbF

The System KF-LiF-RbF

The System NaF-LiF-RbF

Surface of 304 Stainless Steel after 100 hr of Exposure to
Barium Hydroxide at 816°C

Surface of 446 Stainless Steel Specimen after 100 hr of Ex-
posure to Barium Hydroxide at 816°C :

Surface of Inconel Specimen after 100 hr of Exposure to Barium

Hydroxide at 816°C

Surface of Nickel Z Specimen after 100 hr of Exposure to Barium
Hydroxide at 816°C

 

67

68

69

70

71

85
85
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86
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87

93

93

94

95
FIGURE

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TITLE

 

Surface of Chromium Specimen after 100 hr of Exposure to Sodium
Hydroxide at 816°C

Surface of Inconel Specimen after 100 hr of Exposure to Lithium
Hydroxide at 80G°C

Surface of Inconel Specimen after 100 hr of Exposure to Rubidium
Hydroxide at 800°C

Effect of Exposure Time on the Corrosion of Inconel by Sodium
Hydroxide at 800°C

Effect of Exposure Time on the Corrosion of Inconel by Potassium

Hydroxide at 800°C

Effect of Temperature on the Corrosion of Inconel by Potassium

Hydroxide in 100 hr

Weight Loss of Inconel Specimen in Potassium Hydroxide for 100
hr as a Function of Temperature

Surface of 304 Stainless Steel Coated with 3-mil Nickel Electro-
plate after 100 hr of Exposure to Barium Hydroxide at 816°C

Nickel-Clad Inconel (Nickel Sheet) after 100 hr of Exposure to
Sodium Hydroxide at 816°C

Nickel-Clad Inconel (Nickel Powder) after 100 hr of Exposure
to Sodium Hydroxide at 816°C

Nickel Thermal-Convection Loop Walls with Potassium Hydroxide

Plugged Section Nickel Thermal Convection Loop with Sodium
Hydroxide Coolant '

Deposit of Silver Crystals by Mass Transfer in Silver Capsule

Containing Sodium Hydroxide

Temperature Ratio in a 3-mm Tube Filled with Brine in Which
Heat is Generated Uniformly

 

PAGE

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96

96

97

97

98

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101

102
108

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116

XxXvii
 

FIGURE TITLE

12.2 Flow Diagram for Lithium Heat-Transfer Experiment

12.3 Thermal Conductivity of Diatomaceous Earth

12.4 Viscosity Apparatus

13.1 Effect of U0, Particle Size on UQ, Distribution

13.2 Effect of Rolling Temperature on UQ, Distribution

13.3 Effect of Elimination of Capsule on UQ, Distribution

13.4 Seamless-Tube Fuel Elements Formed by '"Rubberstatic'" Pressing

13.5 Transverse Section of a 0.015-in.-Thick Inconel Sheet Spot
Welded to an 0.188-in.-0.d. Inconel Tube of 0.025-1n. Wall
Thickness

13.6 Section of a Spot Weld Joining Two Stainless Steel—Clad Fuel
Plates

13.7 Longitudinal Sections of a Type 316 Stainless Steel Tube-to-
Header Joint Nicrobrazed at 1120°C in a Dry Hydrogen Atmos-
phere

13.8 Transverse Section of a Stainless Steel-~Clad Fuel-Element Butt
Nicrobrazed to a Stainless Steel Sheet

13.9 Transverse Section of an Incomel Tube-to-Header Joint Brazed
with a 60% Pd - 40% Ni Alloy in Dry Hydrogen at 1270°C for
20 min

13.10 Stress-Rupture Time for Inconel Sheset

13.11 Time-Elongation Curve for Nickel Z

Xxviii

 

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137

138

139
140
 

FOREWORD

This is the quarterly progress
report of the Aircraft Nuclear Propui-
sion Project at the Oak Ridge National

Laboratory and summarizes the technical

progress on the project during the
-period covered. It includes not only
the work of the Laboratory under its
own contract, Y-7405, Eng 26, but such
research for the national ANP program

 

as 1s performed by Laboratory personnel.
The report is divided into four parts:
I, Reactor Theory and Design; 11,
Shielding Research; III, Materials
Research; and IV, Appendixes. As each
of these parts may be regarded as
a separate entity, each has a separate
"Summary and Introduction” which pre-
cedes the part in this report.
 
SUMMARY AND

The search for a nonoxidative high-
temperature fluid other than sodium
which would be suitable as a reactor
coolant has lead to the proposed use
of fused fluoride salts containing
uranium (Sec. 1). The resulting cir-
culating-fuel reactor would have the
important advantage of eliminating a
heat-transfer stage within the reactor
core. Preliminary design studies of
such a reactor indicate that a 3.5-ft-
“diameter beryllium oxide—moderated
circulating-fuel reactor will produce
around 350 Mw at amaximum temperature
of 1500°F, The design point of the
aircraft incorporating this reactor is
Mach 1.5 at 45,000 ft. Performance
and weights of the airplane, reactor,
shield, engines, and radiators are
being explored.

Although the Aircraft Reactor
Experiment (Sec. 2} was originally
intended as a prototype of the sodium-
cooled quiescent-liquid-fuel aircraft
reactor, substantial portions of the
~equipment would be applicable to the
circulating-fuel reactor. Consequently,
construction, design, and procurement
of equipment have proceeded on the
original schedule. The building
facility itself is 80% complete and
mest components are on their design or
delivery schedule.

Reactor physics calculations (Sec.
3) have been devoted primarily to the
staties, and to some lesser extent
the kinetics, of the circulating-fuel
reactor. The minimum uranium invest-
ment 1in the beryllium oxide—moderated
circulating-fuel reactor has been
determined as 69 1lb of which 26 lb 1is
in the core. (The uranium investment
in a water-moderated circulating-fuel
reactor is somewhat lower because of
the better moderating capacity of
water, but design considerations appear
to favor the use of Be0,) The percent

INTRODUCTION

thermal fissions of this BeO-moderated
reactor is 52, 1.e., the reactor lies
between epithermal and intermediate.
Kinetic studies, which have shown the
practicality of the conservation of
delayed neutrons, imply that the design
of the reactor must provide for as high
a fuel volume as practical, possibly
even at the expense of the uranium
investment.

A second mockup of the General
Electric "direct cycle” reactor has
been assembled (Sec. 4)}. The critical
mass of this assembly was 90 1b of
uranium.

The cross-sections of iron and
beryllium are being determined for
refined reactor physics calculations
(Sec. 5). Measurements of fast neutrons
in the 5-Mev Van de Graaff give an
average total cross-section for iron
which varies from 2.5 barns at 0.6 Mev
to 3.2 barns at 3.6 Mev. Upper limits
for the (n,2n) cross-section has been
determined in one case to be < 0.56
barn and in another to be < 0.26 barn.
Final adjustment are being made on the
time-of-flight neutron spectrometer
with which resolutions of at least
1.2 pusec per meter are anticipated.

The development of reactor plumbing
and associated hardware, while largely
concerned with that for a sodium-
cooled reactor, has been redirected
to the requirements of the circulating-
fuel reactor (Sec. 6). Valves, pumps,
seals, heat exchangers, and related
equipment are being developed for both
fluids for operation at reactor temper-
atures, i.e., up to 1500°F. However,
extensive development with the fluoride
fuels has first necessitated a program
for their manufacture, purification,
and handling; hence, to date the
experimental work with fluorides has
ANP PROJECT QUARTERLY PROGRESS REPORT

been somewhat limited. A frozen-
fluid-seal centrifugal pump which was
successful with sodium is being modi-
fied for use with the fused fluorides.
A centrifugal pump with a gas seal,
however, has already been used to
pump the fluorides satisfactorily for

short periods. Valve tests in sodium
systems show significant increase in
torque with time although several
combinations cf metals appear to resist
self-welding. A NaK to NaK heat
exchanger has now operated for 550 hr
with a maximum temperature of 1200°F.
'FOR PERIOD ENDING DECEMBER 10, 1951

1. CIRCULATING-FUEL AIRCRAFT REACTOR

R. W. Schroeder,

The objective of the Oak Ridge
National Laboratory in the national
Aircraft Nuclear Propulsion Program
is the development of high-performance
reactors, 1.¢., for supersonic propul-
sion. This implies the use of liquid-
conolant systems which require small
core sizes and have good heat-transfer
characteristics. The specific objective
of the ORNL-ANP project is, accordingly,
the exploitation of nonoxidative high-
temperature fluids. This line of
research avoids not only the high
pressures associated with some cycles
but alsc the oxidation inherent to
others. A sodium-cooled guiescent-
liquid-fuel reactor was the Laboraea-
tory’'s first considered proposal as a
reactor with potentialities for super-
sonic flight. However, the almost
prohibitively difficult task of
assuring the safety of a sodium~cooled
and water-shielded reactor, as well as
the limitations of a unifunctional
coolant, has enhanced the search for
a more versatile, and less inflammable,
coolant. As a result of this search,
OBNL has turned toward the use of
fused fluoride salts (generally a
ternary or quaternary sSystem composed
of uranium fluoride and a mixture of
two or three alkali1 fluorides or
beryllium fluoride) -—— not for just
the heat-transfer medium, however,
but as the reactor fuel as well. In
addition to the advantageous physical
properties of the fused salts (although
not so good as sodium from the stand-
point of heat transfer) and their
noninflammability in air and water,
an important advantage of such a cir-
culating-fuel reactor 1s that 1t
eliminates a heat-transfer stage
within the reactor core.

The advantage to be gained from
separating core and heat exchanger

ANP Division

cannot be overemphasized. Although
it is perhaps possible to conceive of
a small and at the same time a well-
designed 500 Mw heat exchanger, the
requirements of these entities are so
different that i1t is difficult te
design the heat exchanger into the
reactor without severely penalizing
both. Furthermore, whereas the sodium-
cooled reactor required considerable
departure from conventional design
practice to remove approximately 200 Mw
from a 3-ft core, it now appears
feasible, from a fabricaticnal and
fluid~flow standpoint, to remove
350 Mw from a 3%-ft core employing
circulating fuel.

The c¢irculation of fissionable
material external to the core dirscts
attention to design studies of the
entire system -— rveactor, radiator,
engine, etc. Although the circulating-
fuel system has not been studied suffi-
ciently long to ensure the performance
of the resulting aircraft, the more
outstanding problems of this cycle have
been appreciated. The preliminary
design studies which have been initiated
for the exploration of a supersonic
airplane application of a circulating-
fuel reactor are encoursaging. However,
performance and weights of the air-
plane, reactor, shielding, engines,
radiators, and other components are sco
interrelated that i1t was found necessary
to make studies of the overall arrange-
ment. The studies have not yet been
completed, and it is not possible to
to draw conclusions at this time.
However, the arrangement being studied
may be cursorily described. '

AIRPLANE AND OVERALL ARRANGEMENT

The airplane visualized is designed
for a speed of Mach 1.5 at 45,000 f¢,
ANP PROJECT QUARTERLY PROGRESS REPORT

with a gross weight of approximately
350,000 1b, an L/D ratio of approxi-
mately 6.5, and a wing loading of
approximately 70 1b/ft?, A divided
shield 1s employed with gamma and
neutron shielding about the crew com-
partment and neutron shielding about
the reactor. Six turbojet engines are
arranged 1n the fuselage in a circle
aft of the reactor-shield assembly,
and circulating fuel isducted directly
from the reactor to the engine radiators,
thereby eliminating the weight and
temperature loss associated with the
use of i1ntermediate heat exchangers.

REACTOR

Water-moderated and solid-moderated
reactors have been explored, and
tentatively, both types appear to be
feasible. However, the water-moderated
and -reflected reactor would, for
this application, require a water-
cooling system capable of disposing of
approximately 30,000 kw with a small
radiator temperature difference and a
small air temperature rise. The small
temperature difference would require
very large radiaters, and the small
air temperature rise would entail very
large cooling air flow requirements.
Further, the use of water as a moderator
involves a potential hazard associated
with rapid steam formation if a failed
fuel tube should permit abrupt mixing
of fuel and water. Accordingly, the
studies outlined here are based on the
use of BeO as amoderator and reflector.
Moderator heat is removed by the cir-
culating fuel and is employed usefully
in the propulsion cycle. By proper
allocation of coolant flows and
coolant-tube surface areas, it is
possible to maintain maximum coolant-
tube wall temperatures only slightly
in excess of the circulating-fuel
maximum temperature. To permit re-
moving reflector heat with a high
radiator temperature difference, the
design under study postulates cooling

the reflector with a non-fuel-bearing
fluoride mixture. This reactor con-
forms to the specifications given in
Table 1.1.

 

Table 1.1

Features of the Circulating-Fuel
Aircraft Rezactor

Size 3.5-ft sphere
Moderator BeO

Moderator ~65%

Fuel ~ 33%
Structural ~2%

Design point

Power ~ 325,000 Btu/sec
Reactor fuel inlet temp. "~ 1000 °F
Beactor fuel outlet temp. ~1500°F

 

ENGINE

A circulating-fuel type reactor
should be inherently capable of supply-
ing propulsive power by means of a
turbojet cycle or either of several
vapor cycles. The relative superiority
of these cycles has not yet been com-
pletely established, but, pending com-
parisons, the use of the turbojet
cycle has been presumed. Preliminary
optimization studies have indicated
that a compression ratio in the region
of 6.1, with a turbine inlet tempera-
ture of approximately 1250°F, 1is
desirable. Further increases 1n the
compression ratio would diminish the
radiator weight at the expense of
engine weight, and further increases
in turbine inlet temperature {(with
fixed reactor conditions), would favor
turbojet weight to the detriment of the
radiator log mean temperature differ-
ence and radiator weight. More rigorous
evaluations of these component weights
will be required before the turbojet
specifications indicated here can be
accepted with any degree of fimality.
FOR PERIOD ENDING DECEMBER 10, 1951

°

Installational convenience favors
the use of asmall number of relatively
large engines, whereas engine develop-
ment and availability considerations
favor the use of a larger number of
smaller engines., Tt is difficult to
predict at this time the size of engine
that could be made available when an
airplane of the type studied would
require these engines, but 1t appears
that the validity of the overall study
is relatively insensitive to the
number of engines presumed. Therefore
the use of six engines conforming to
the general specifications listed
in Table 1.2 has been assumed.

Table 1.2

Performance of Engines at Design Point,
Mach 1.5 at 45, 000 ft

Thrust 8200 1b
Maximum diameter 61 in.
Compression ratio 6.1
Turbine inlet temperature 1250°F
Equivalent sea-level airflow 665 lb/sec

Calculations have indicated a

specific impulse of approximately 30

and overall efficiency of 30%. This
latter 1s defined as the ratio of mnet
thrust horsepower to reactor thermal
hersepower.

RADIATORBS

Radiators with heat-transfer capac-

ity per unit volume 1n excess of the
best obtained to date will be desired

for any ligquid-cycle supersonic
nuclear-powered airplane. Arrangements
are being made to receive the advice
and consultation of established heat-
exchanger manufacturers relative to
the design of radiators for such appla-
cations. This advice will not be
available for at least several months,
however, and

1t was necessary to
achieve a preliminary radiator design

in order to permit related phases of
the study to continue. The design
conceived of, which may be far from
optimum, involves a radial grouping
of rectangular-shaped banks about the
engine centerline between the com-
pressor and turbine. Liquid-conveying
tubes are passed through, and normal
to, closely spaced sheet fins. The
flow pattern contemplated 18 counter-
current, as dictated by the tempera-
tures of the two fluids. The arrange-
ment entails the use of dividing
baffles between adjacent banks, and
the 1nstallation of by-pass valves in
these baffles will permit controlling
turbine inlet temperature while a
substantially constant liguid tempera-
ture 1s maintained.

SHIELDING

Shielding calculations currently
are 1n process, and it 1s not possible
to make a guantitative description of
the shielding at this time. Initial
studies have indicated (see "Circu-
lating-Fuel-Reactor Shields™ in Seec.
8), however, that the circulating fuel
in the radiators and the associated
plumbing may remain unshielded provided
that sufficient neutron and gamma
shielding is placed about the crew
compartment, and alsec provided that
the payload, and possibly other com-
ponents, are protected by local
shielding.

It currently i1s contemplated that
the reactor shield will include only
hydrogenous material,

ACCESSORY SYSTEMS

Accessory circuits are required to
permit cooling the reflector and shield
and to provide power for pumps and
general aircraft accessory demands.
As it appears that many of the acces-
sories are favored by the use of
variable-speed drives, the overall
ANF PROJECT QUARTERLY PROGRESS REPORT

accessory system contemplated involves
the use of individual pneumatic tur-
bines for power supply. The energy
1s supplied, and reflector cooling 1is
achieved, by passing compressor bleed-
of f air through the reflector radiators
and then to the accessory turbines and
to parallel propulsive nozzles. The
accessory turbines are controlled by
me ans of variable-area discharge
nozzles, which provide some net thrust
after imparting energy to the accessory
turbines. Preliminary analysis of
this system indicates that 1t not only
provides reflector cecling and acces-

10

sory power adeguately, but that the
reflector cooling system, when con-
sidered as an open Braytom cycle
power plant, has a favorable cycle
efficiency and power-weight ratio. The
use of 1ndividual accessory drive
turbines permits supplying the acces-
sories with power and speed control
without special mechanical, hydraulic,
or electrical transmission systems.

The shield-water radiator is cooled
by low-pressure compressor bleed-off
air, and the =2ir is then discharged
through a propulsive nozzle.
FOR PERIOD ENDING DECEMBER 10, 1951

2. LIQUID-METAL-COOLED AIRCRAFT REACTOR EXPERIMENT

R. W. Schroeder,

Aircraft reactor cycles other than
the sodium-cooled stationary liquid-
fuel cycle have been(!”? and are (Sec. 1)
under active study. A large portion
of design effort has, therefore, been
diverted to study of these alternate
cycles., These i1nvestigations have
indicated a high probability that
substantial portions of the equipment,
designed for the original sodium-cooled
reactor experiment, would still be
usable 1f the ARE were to be a prototype
of one of these other cycles. Ac-
cordingly, design and procurement of
the building and equipment have been
continued on the original basis.

At the present time the facility
under contract te the Nicholson
company is approximately 80% complete.
All major fluid-circuit components
have been designed and are on order
or under construction locally, and
all core and pressure shell components
are either scheduled for fabrication
or have already been completed. Instru-
mentation, electrical circuits, and
remote-handling equipment have been
partially designed, and 1tems requiring
long fabricational time spans have
been ordered.

CORE~ REFLECTOR—PRESSURE SHELL

Approximately 90% of the BeO
moderator and reflector blocks have
" been received., The inconel pressure-
shell billet has been shipped from
International Nickel Co. to Lukenweld,
and the shell is scheduled for December
rolling at Lukenweld. The fuel-

 

(. F. Hemphill, R. W. Schroeder, and H. R,
Wesson, “Circulating-Moderator-Coolant Reactor:
OBNL,” Atircraft Nuclear Propulsion Project Quar-
terly Progress Report for Period Ending September
10, 1951, OBNL-1154, p. 188 (Dec. 17, 1951).

ANP Division

element tubing has been drawn by
Superior Tube Company and the 1.235-1in.
coolant tubes are in process at Superi-
or. Miscellaneous sheet and bar
stock for the core have been received.

Detailed core manufacturing drawings
have been made and retained pending
material availabilicy.

FLUID CIRCUIT

The fluid-circuit design for the
original NaK system {(the use of Nak
was a modification, to eliminate pre-
heating, of the use of sodium) is
approximately BO0% complete, and all
major components have been released
for procurement or local manufacture.
These include (1) inconel pipe, (2)
dump tanks, (3) all heat exchangers,
(4) helium blowers for heat disposal
loop, (5) Vickers variable-speed
hydraulic drive system for helium
blower drive, (6} space-cooling—heat-
exchanger-——fan combinations, (7) con-
trol-rod cooling rotary blowers, and
(8) NaK purification assembly.

REACTOR CONTROL

E. S. Bettis

Research Director’s Division

Because of the proposed modifi-
cations in the ARE design, work on
the control system for this reactor
has necessarily slowed down. However,
satisfactory design of most of the
components of the control system had
already been attained.. The high-
temperature fission chamber failed at
T00°F and is being redesigned. Develop-
ment of the reactor dynamic computor
has continued.

11
ANP PROJECT QUARTERLY PROGRESS REPORT

Control System. Detailing of the
control rcom, console, 1nstrument
racks, and interconnecting cables has
continued. Discussion of the manu-
facture of the hardware incorporating
these designs was held with possible
outside vendors. Several companies
have indicated willingness to contract
for these 1tems,
have been made.

but no commitments

High-Temperature Fission Chamber.
The high-temperature fission chamber
was completed and tested. It failed
at about 700°F because the insulators
were inadequate. A new chamber,
correcting thisdifficulty, 1s presently
under construction and will be ready
for test in about six weeks. No
difficulty 1s anticipated in providing
a satisfactory fission chamber for use

in the ARE.

Reactor Dynamic Computer. The
reactor dynamic computer has progressed
to a point where 1t will be ready for
use in about six weeks. The converter,
from d-c to digital notation, and the
multiplier have been tested and
debugged.

INSTRUMENTATION

About 5% of the instruments required
for the ARE have been received, and
an additional 75% are on order and
should be received within four months.
Ten percent will probably be constructed
by the Laboratory because of unavaila-
bility from industrial concermns. The
remaining 10% are in a deferred status

12

pending the outcome of experimental
engineering investigations currently
in progress.

REMOTE ~-HANDLING EQUYPHMENT

Remote-handling equipment design 1is
in process, and some components have
been ordered. Demonstrations have
been made simulating installation,

removal, and reinstallation of the
core dome, involving a sequence of
welding, cutting, scarfing, and re-
welding. These demonstrations will

continue and will employ the automatic
cutting equipment intended for the

ARE.

Remote removal of the core dome
after power operation of the reactor
will be accomplished by means of
rubber-bonded abrasive cut-off wheels,
operating under water, driven by a
Vickers hydraulic system and guided
by peripheral tracks. This equipment
has been designed, some components
have been requisitioned, and other
components have been released to Y-12
shops for fabrication.

ARE BUILDING FACILITY

The building shell 1s virtually
complete; walls and roofing are 1in
place. Painting, plumbing, and internal
partitioning are in process. ORNL
engineering of special internal
features is approximately 80% complete
and should be ready for transmittal
to Nicholson via AEC in approximately
four weeks.
FOR PERIOD ENDING DECEMBER 10, 1951

3. REACTOR PHYSICS

Nicholas M. Smith,

Significant studies by the ANP
Physics Group have been the surveys
of the circulating-liquid-fuel designs.
These surveys have been conducted
using bare-reactor theory (i.e., a
condition under which space and
lethargy variables are separable)
and for several parametric variations.
For a given moderator these parameters
were core diameter and the volume
ratio of circulating fuel to that of
the moderator. The rather extensive
bare-reactor calculations were made
possible by having the calculation
set up entirely on the Card Frograming
Computor IBM equipment.

Considered first are the values

of these parameters leading to a

minimum fissionable material in-
ventory in the overall system. When
the core diameter 1s fixed, 1t 1is

seen that as the fuel-coolant volume
fraction increases, the critical mass
increases but that the uranium density
in the fuel-coolant at first de-~
creases, then increases. The result
is that there is a minimum in the
uranium investment in the overall
system. This minimum usually re-
sults 1n a reactor with about half
thermal fissions, 1.e.,
the borderline between the epithermal
and the intermediate.

It 15 known from kinetic studies
that the ratio of fuel in the reactor
to that in the overall system should
be as high as is practical 1in order
to conserve delayed neutrons. Thus,
in comparison to the condition of
minimum critical uranium mass, the
condition of minimum overall uranium
mass results in a higher reactor-to-
system volume ratio and favors the
kinetic behavior. The kinetic be-
havior can be favored to an even
greater extent by choosing a fuel-

a reactor on-

Jr., ANP Division

coolant volume ratio greater than
that resulting in a minimum uranium
inventory. Since the curves of total
uranium inventory vs. fuel-coolant
volume ratio havea very broad minimum,
it is possible to design for a greater
fuel-volume ratio without greatly in-
creasing the uranium requirement.

Thus it becomes advisable to design
the reactor with as high a fuel volume
ratio as practical — say, one in-
creasing the uranium requirement 5%
over that of the minimum —~ in order
to favor the kinetic response. The
conditions of near-minimum uranium
mass and of favorable kinetic behavior
thus result in an increase of the
neutron spectrum, yielding a reactor
just barely in the intermediate class.
The trend of the nuclear design to
one of the intermediate class is thus
inescapable.

Very satisfactory agreement within
theory and critical experiment 1s
reported for intermediate uranium-
graphite and uranium-water assemblies.
The agreement gives corroboration to
the explanation that the discrepancy
between theory and experiment for the
uranium-beryllium assembly is caused
by omission of a significant physical
phenomenon, possibly the (n,2n) re-
action.

REACTOR CALCULATIONSON IBM EQUIPHMENT

F. C. Uffelman W. C. DeMarcus
Phyllis Johnson

Uranium Control and Computing Department

The extensive bare-reactor calcu-
lations were made possible by having
the calculations set up entirely on
the CPC-IBM equipment. During the
past quarter the IBM equipment was

i3
ANP PROJECT QUARTERLY PROGRESS REPORT

also employed in the calculation of
cross-sections for cores and re-
flectors and kinetic calculations of
various reactors. The division of
effort 1s 1temized as follows:

1. Reactors. During the period

from September 1 to November
16, 1951, the IBM section com-
pleted calculations on 29 EPLA
(end-point linear approximation)
reflected reactors,
the total to 211 since the
start in February. Programing
was 1nitiated and completed for
twe series of bare-reactor cal-
culations, hydrogenous
nonhydrogenous. Twenty-nine
hydrogenous and 43 nonhydrog-
enous calculations were made.

bringing

and

2. Cores and Reflectors. During
this same period average cross-
sections were calculated for
117 cores and reflectors, and
"constant" variations were made
for two cores and reflectors.

3. Kinetic Calculations. Programing
completed for the dual
space-time varilable kinetic
calculations and three such
calculations were made, one for
time running to 15 sec and the
other two only up to 4 sec each.
One calculation, with time the
only variable, was also per-
formed.

was

CIRCULATING-FUEL REACTORS
C. B. Mills, ANP Division

The circulating-fuel reactor under
study consists simply of a spherical
moderating core with circular fuel-
coolant tubes penetrating the length
of the core into the reflector. The
reflector is a spherical shell of
core modearator material of 6 in. re-
flector savings. The moderator 1is
expected to be either beryllium oxide

14

If BeO 1s utilized
as a moderator, 1its cooling will be
provided by the circulating fuel
itself, making the "reflector" also a
neutron multiplying region. If water
is so employed, 1t will be 1nsulated
from the circulating fuel by a thin
layer of, say. Si0,, and the moderator
cooling will be accomplished by
circulating the water 1n a secondary
system, with 1ts separate heat ex-
changer.

or light water.

A series
calculations have
been made to determine optimum size
and composition. Reflector savings
were included by adding a constant
15.24 cm of core material plus 1.13 ¢cm
of extrapolation distance to the
variable core radius. A series of a
maximum of 27 calculations 1s suffi-
cient to determine the optimum size
and shape with the variable of re-
actor core diameter, volume fraction
of fuel-coolant or moderator,
mass, and total uranium
Thermal base effects, median energy
for fission, and neutron absorption,
fission leakage, and flux lethargy
(or energy) distributions are obtained
in the same group of calculations. The
reactor design values used 1n this
series are given in Table 3.1.

Re0-Hoderated Reactor.
of bare-reactor

uranium
inventory.

Table 3.1

Design Values for the Be@-Moderated
Circulating-fuel Reactor

 

 

CORE COMPONENTS VOLUME FRACTIONS

Fuel-coolant 0.10 0.20 0. 30 0.35 |0.40
(NaF-UF;)

Structure 0.015| 0.015] 0.015 1 0.015(0.015
(1inconel)

Moderator (BeD) 0.855] 0.755| 0.655 {0.650(0.555

Yoid (assumed) 0.03 | 0.03 { 0.03 |0.03 [0.03

 

 

 

 

 

 
FOR PERIOD ENDING DECEMBER 16, 1951

These volume fractions gilven 1in
Table 3.1 were used for threes reactor
core diameters: 3, 3%, and 4 {t.
The BeQ reflector thickness was 6 1in.
The uranium mass reactivity coefii-

cient, Nk/k T Am/m, was examined 1in
the process of the determination of
the minimum critical mass. The value

of the coefficient decreased by about
20% in changing the volume coolant
fraction of the coolant from C.1 to

uranium inventory as a fanction of
volume fraction fuel-coolant are
given in Fig. 3.1 for 3-, 3%-, and
4-ft reactor cores. The total uranium
inventory was a minimum of 66 1b at
0.28 volume fraction of fuel 1n the
4-fr-diameter core. The total uranium
inventory with the fuel-coolant
holdup in cubic feet 1n the heat ex-
changer as a parameter is given 1in
Fig. 3.2 as a function of reactor

0.4. The critical mass and total core diameter length.
SECRET
ANP-PHY-223
DWG, 13565
120 -

 

 

100 \

 

REFLEGCTOR SAVINGS =6in.
FUEL-COOLANT: NaF-UFg4
 EXTERNAL SYSTEM VOLUME

T T T T T

 

 

IS ASSUMED TEO BIE 15|fi3
:

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

LN
\\ - 3-ft DIAMETER TEJTAL
S 80 D - i URAN UM —
e ‘\\W\\ 3/é$f DIAMETER INVENTORY
W >
cqn S = 4-ft DIAMETER
= 60
<
=
=
<
€ 40
4(fleAMETER URANIUM
/ .
-3 /o-ft DIAMETER I N —
/fif’ 3-ft DIAMETER] CORE
e
O .
0 040 0.20 0.30 0.40 Q.50 0.60 0.70

VOLUME FRACTION FUEL-COOLANT IN CORE

Fig. 3.1, Total Uranium Inventory

and Uranium in Core as a Function of the

Volume Fraction Fuel Coolant in the Core for BeO-Moderated €Circulating-Fuel
Reactors with 3-, 3%-, and 4-ft-Diameter Cores.

15
ANP PROJECT QUARTERLY PROGRESS REPORT

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

SECRET
ANP-PHY-224
DWG. 13566
1 ! T T T ]
- 40 AN FUEL-COOLANT VOLUME FRAGCTION=035
uly30 ™ FUEL-COOLANT: NaF -UF4
= N
o ]
1ip] R,
zZ \\\; \\N‘Nmm —~
\\ 30 112
= _
5100 <~
- 3
Z 90 \\_\ - 25 ft HOLDUP VOLUME _|
a -
x e OUTSIDE THE
|, 80 20 f1° REACTOR GCORE
2 of —
O Nh""""*—-«....,_,_____ ] 3 T
o 15 f17
v 1l
50 *
30 3.25 35 3.75 4.0 4.25 4.5 4.75
REACTOR CORE DIAMETER (ft)
Fig. 3.2. Total Uranium Investment in Bed-Moderated Circulating-Fuel Reactor

as a Function of Reactor Core Diameter for Several Assumed Holdip Volumes

Externa}l to Core,

The percent thermal fissions as a
function of fuel-coolant volume per-
centage 1s shown in Fig. 3.3 for the
3- and 4-ft-diameter cores. Apparently
the ANP reactors are in the epithermal
to intermediate energy range of
neutron energy distribution. Fis-
sioning, absorption, and leakage
spectra for the 3%-ft-diameter core
are given in Figs. 3.4, 3.5, and 3.6,
respectively. The percent thermal
fissions for this reactor 1s 59, the
percent thermal absorption 1s 41.6,
and the percent thermal escape 1s 8§,
all at lethargy of 18.6. The neutron
flux distribution (Fig. 3.7) is high

16

and rather uniform over all lethargy
intervals above thermal.

The poisoning effect of stainless
steel for this reactor been
estimalted as 6% change in k__. . for a
change in volume fraction of stain-
less steel of 0.01 in the vicinity of
a total volume fraction of 0.028.

has

The set of data 1s not yet suffi-
ciently complete to do more than
specify the region of design interest.
Apparently the reactor will have
approximately 35% fuel-coolant and
65% BeD moderator, with a structure
FOR PERIOD ENDING DECEMBER 10, 1951

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

SECRET
ANP-PMHY-225"
OWG. 135667
100
90
Y 8o
§ -\~§§: /p~4‘fl-COREIMAMETER
“ 70 P
® ~
“ 60 S~
o NN
<L \‘?\h\\\
% 50 I
g 2-f+ CORE mAMEreéx \\
= 40
-
Zz 30
i
=
w 20
o FUEL~COOLANT : Naf -UFg
10
0
0 0.40 0.20 0.30 0.40 0.50
VOLUME FRACTION FUEL ~COOLANT
Fig. 3.3. Percent Thermal Fissions

as a3 Function of the Fuel-Copolant
Volume Fraction in the 3- and 4-ft-
Biameter Cores of the Be(O-Moderated
Circulating-Fuel Reactor,

and void volume percent tobe specified
by design considerations.

Water-Moderated Reactor. The first
calculations on the water-moderated
circulating-fuel reactor were made
with the NaF-UF, fuel-coolant since
density values on other possible
fluoride mixtures were not available.
The appearance of some data on the
new fuels NalF-BeF-UF, and NaF-LiF-KF-UF,
for use with circulating-fuel re-
actors subsequently permitted 1nitia-
tion of a more complete set of cal-
culations. Critical mass was deter-
mined as a function of reactor size
and percent fuel-coolant for all
fuel-coolants. The thickness of
reflector for saturated reflector
savings was estimated in all cases,
and the actual core radius was 1in-
creased by a fraction of this amount
{(plus augmentation distance) to obtain
the radius of the equivalent bare
reactor. This reflector thickness 1in

each case was such that a 6-in, re-
flector was still unsaturated. The
assumption that the reflector is
equivalent to added core material is
thus sufficiently accurate. All
calculations in the first survey have
been done on this equivalent bare
reactor.

Naf-UF, Fuel-Coolant. This series
of bare-reactor calculations were
made on a design series of the same
type as that used with the BeO~-moderated
reactor. Cores 2%, 3, and 3% ft in
diameter were used, with volume per-
cents of H,0 varying from 10 to 80%.
The first set of calculations was
made with the NaF-UF, fuel-coolant,
since density values for the two new
coolants were not initially available.
A reflector savings of 6 in. was in-
c¢luded i1n the usual way. Volume
fractions for the NaF-UF, —~cooled
water-moderated reactor are given 1in
Table 3.2. The core diameter for this
set was 3 ft. Calculations of uranium
mass vs. k_,. for this core indicated
that the critical mass i1is a very
slowly varying function of the percent
fuel-coolant, since a change from

15 to 35% results 1n an increase of
only 8% in critical mass in the
vicinity of 18 lb. Figure 3.8 1in-

dicates the total uranium inventory
and critical mass as a function of
fuel-coclant fraction for 15 ft* in
the heat exchanger. With the fuel-
cocolant NaF«UF4 and a 3 ft core, the
uranium inventory will be near 75 1b.

Table 3.2

Design Values for the NaF-UF‘-Cooled
Water-Moderated Reactor

 

CORE COMPONENTS VOLUME FRACTIONS

 

Fuel-coolant (NaF-UF,) 0.15 0.25 0.35
Structure {(inconel) 0.015 0.015 0.015
Moderator (Hzo) 80.5 70.5 60.5

Void (assumed) 0.03 0.03 0.03

 

 

 

 

17
81

FISSION PER UNIT LETHARGY PER v, INITIAL NEUTRONS

048

0.16

O
>

O
N

Q
O

0.08

0.06

0.C4

0.02

O

SECRET
ANP-PHY-226
DWG. 13568

COMPOSITION BY VOLUME FRACTION

 

FUEL-COOLANT (NaF-UR, )
STRUCTURE (INCONEL)
MODERATOR (BeO;
VOID

59% THERMAL
FISSION AT
v=18.6

18 16 14 12 10 8
LETHARGY {v)

0.300
0.C15

0.655
0.030

 

Fig. 3.4. Fission Spectrum vs. Lethargy for the Be0D-Moderated Circulating-

Fuel Reactor with a 3k-ft-Diameter (ore.

LY0d34 SSHUI0Md ATHILVVYAD 123af0Md JINV
61

ABSORPTION PER UNIT LETHARGY PER 1, INITIAL NEUTRONS

0.36

G.32

0.28

0.24

0.20

0.46

0.12

0.08

0.04

SECRET
ANP-PHY -227
DWG, 13569

 

COMPOSITION BY VOLUME FRACTION

FUEL-COOLANT (NaF-UF ) 0.300
STRUCTURE (INCONEL) 0.015
MODERATOR (BeO) 0.655
VOID 0.030

44697 THERMAL
ABSORPTION
Al w=18.6

18 16 14 12 10 8 6 4 2 0

LETHARGY ()

Fig. 3.5. Absorption Spectrum vs. Lethargy for the BeO-Moderated Cir-
culating-Fuel Reactor with a 3%-ft-Diameter Core.

IS6T ‘0T HAGWADHAG INIGNT AOIWAd HO4
0

0.18

0.6

0.4

02

040

0.08

0.06

0.04

0.02

LEAKAGE PER UNIT LETHARGY PER v, INITIAL NEUTRONS

Fig. 3.8.

Leakage Spectrum vs.

COMPOSITION BY VOLUME FRACTION

FUEL-COOLANT (NaF-UF,) 0.300
STRUCTURE (INCONEL) 0.015
MODERATOR (BeO) 0.655
VOID | 0.030

i8 16 14 12 10 8 6

LETHARGY ()

Fuel Reactor with a 3%-ft-Diameter Core,

SECRET
ANP - PHY-228
DWG, 13570

 

Lethargy for the BeO-Moderated Circalating-

L1H043Y SSIU90Hd ATHRLYVND 1JAfodd dNV
ié

SECRET
ANP-PHY-229
DWG. 13574

COMPOSITION BY VOLUME FRACTIONS

FUEL-COOLANT (NaF-UF,) 0.300
STRUCTURE  (INCONEL) 0.015
MODERATOR  (BeO) 0.655
37 VOID 0.030

RMAL

AT LETHARGY
‘ u =86

NEUTRON FLUX ¢ {(¢) PER 7, INITIAL NEUTRON

18 e i4 12 {0 8 6 4 2 O

LETHARGY ()

Fig, 3.7. Neutron Flux Spectrum vs. Lethargy for the BeO-Moderated Cir-
culating-Fuel Renctor with a 3%-ft-Diameter Core.

 

IS6T ‘O HIAWFDAG SNIONZ QOIVAd HOd
ANP PROJECT QUARTERLY PROGRESS REPORT

160

SECRET

ANP—PHY~23!
DWG,. 13572

 

|

 

1

LIGHT WATER , DENSITY=09667 g/cc
140 \\\EXTERNAL VOLUME =15 ft.

|

FUEL-COOLANT : NaF-UF,4

|

|

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

120 \
o N
o 100
& N TOTAL URANIUM INVENTORY
= \\
80
s
=
P
< 60
(el
o
40
{
URANIUM IN THE CORE
20
0
O {0 20 30 40 50
VOLUME FRACTION PERCENT COOLANT IN CORE

Fig. 3.8. Total Uranium Inventory and Yranium in the Coreas a Function
of the Volume Fraction Fuel-Coolant in the 3-ft-Diameter~Core Water-
Moderated Reactor with NaF-UF4 Fuel-Coolant,

BeF,-NaF-UF, Fuel Coolant. Re-
actor design values for this coolant
in the water-moderated core are given
in Table 3.3 and are applicable to
each of the 2%-, 3-, and 3Y%-ft-dia-
meter cores.

The uranium mass reactivity coeffi-
cient, Ak/k + Am/m, decreased by only
30% for values of volume fraction of
fuel-coolant in the core from 0.3 to
0.8. The critical mass and the total
uranium inventory 1in the core as a
function of the volume fraction fuel-
coolant are given in Fig. 3.9 {for a
2%-, 3-, and 3%-ft-diameter reactor

22

Table 3.3

Design Values for the Water-Hoderated
Circulating-Fuel Reactors for Use
with the Na¥F-Bel, and LiF-KF-NaF

Fuel-Coolants

CORE COMPONENTS

 

 

VOLUME FRACTIONS

 

Fuel-coolant
Moderator (H20)

Structure
(inconel)

Void (assumed)

0.30
0.6038
0.0100

0.0862

 

0.45
0.4681
0.0150

0.0669

 

 

0.60
0.3325
0.0200

0.0475

 

0.80
0.1255
0.027

0.0495

 
FOR PERIOD ENDING DECEMBER 16, 19531

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

SEGRET
ANP—PHY-235
120 DWG. {3573
EXTERNAL SYSTEM VOLUME 15 #°
FUEL-COOLANT Be F2~NGF-UF4
WATER TEMPERATURE 1i83° F
100 LIGHT WATER ; DENSITY=0.9667 g/cc
24 ft |
80 :\\‘1‘ B =~ : TOTAL URANIUM |—
~ 4
£ \ L3V ft INVESTMENT
N 23 &
D Vel "
w \\‘~mw_"1j;/7 ™
=
2
=
= 3l ¢
> 40 x 420 12 1 DA,
-~
.~ |43 ft DA URANIUM
- // le2' 1t DA, j N CORE
20 e
R B oy
O :
O 0.20 .40 060 0.80 100 1.30 .50
VOLUME FRAGTION FUEL~ COOLANT
Fig. 3.9. Total Uranmium Inwentnry and Urapium in {ore 28 a3 Function of the

Volume Fraction ¥Fuel-Coolant in the Core for the Water-Moderated Circulating-

Fuel Reactor with 2%-, 3-,

core. The total uranium inventory
was a minimum of 60 1lb at 0.60 volume
fraction of fuel in the 3-ft core.
The minimum critical mass with water
moderation 1s realized with a smaller
core than that required with beryllium
oxide moderation, primarily because
water 15 a better moderator. '

absorption, and leakage
the 3-{t diameter

Fissioning,
spectra for all

and 3%-~-ft-Diameter Cores.

water-moderator cores with 0.60 volume
fraction fuel-coolant are given in
Figs. 3.10, 3.11, and 3.12, respec-
tively. The percent thermal fissions
for this reactor 1s 86.4, the percent
thermal absorption 1s 86.4, and the
percent thermal escape 1s 14.6, all
at a lethargy of 19.6. This 1s,
therefcre, an epithermal reactor.
The neutron-flux distribution (Fig.
3.13) has a high peak in the high-
energy (low-lethargy) range.
Ve

FISSIONS PER UNIT LETHARGY PER 7, FISSION NEUTRONS

018

016

012

010

0.08

0.06

0.04

0.02

SECRET
ANP—PHY —244
DWG, 13574

OMPOSITION OF CORE BY VOLUME FRACTION

FUEL-COOLANT (NaF - BeF, - UF,) 06000

MODERATOR (H,0) 0.3325
STRUCTURE (INCONEL) 0.0200

VOID | 0.0475

864 %
THERMAL
FISSIONS AT
- v=196

 

18 16 4 12 10 8 6 4 2 O

LETHARGY ()

Fig., 3.10¢. Fission Spectrum vs. Lethargy for the flzo-Moderated Circulating-

Fuel Reactor with 3-ft-Diameter Core.

1H043Y SSAUT0Ud ATHALEVNO LDALOUd dNV
Se

ABSORPTIONS PER UNIT LETHARGY PER Ve FISSION NEUTRONS

SECRET
ANP—-PHY-242
0.36 DWG, 13575

032
OMPOSITION OF CORE BY VOLUME FRACTION
0.28 FUEL-COOLANT (NaF - BeF, - UF,) 06000

MODERATOR (H,0) 0.3325
STRUCTURE (INCONEL) 0.0200

VOID 0.0475

0.24

0.20

0l6

012

87 %

THERMAL
008 ABSORPTION
AT v=19.86

0.04

18 6 4 12 {0 8 6 4 2 O

LETHARGY (u)

Fig. 3.11. Absorption Spectrum vs. Lethargy for the H20=Moderated Cir-
culating-Fuel Reactor with 3-ft-Diameter Core,

 

IS6T ‘0T WIIWADHG INIANT aoryiad Ho4
9¢

LEAKAGE PER UNIT LETHARGY PER v, INITIAL NEUTRONS

0.18

0.16

G114

Q12

010

0.08

C.06

0.04

C.02

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

SECRET
ANP—-PHY-243
DWG, 13576
! T fi P J T Y
| | 7 L |
EEEEEEEEE
| *% 4‘ _I J 1 { T %
COMPOSITION OF CORE BY VOLUME FRACTION
ZL FUEL-COOLANT (NoF~BeF - UF )Y 08000
; MODERATOR (H,0) 0.3325
| |STRUCTURE (INCONEL) 0.0200
- WOoID j 0.0475
| T %
| ) |
| | |
- ; [ |
L | -
— — =
+ i
[ I S i' —
| '
46 % - =
THERMAL ]
ESCAPE | — S | | —
ATu=196-4 | ! —— — |
T R ! |
i8 16 14 12 {0 8 4 2
LETHARGY ()
Fig., 3.12. Leakage Spectrum vs. Lethargy for the HQQ-MO&erated Circulating-
Fuel Reactor with 3-ft-Diameter Core,

140438 SSTUO0HA ATHILYYNO LDAf0Ud JNV
La

NEUTRON FLUX ¢ (¢) PER UNIT LETHARGY

PER 7, FISSION NEUTRONS

SECRET
ANP-PHY-244
DWG, 13577

COMPOSITION OF CORE BY VOLUME FRACTION

FUEL— COOLANT (NaF -BeF, - UF,} 06000

MODERATOR (H,0) 0.3325
STRUCTURE (INCONEL) 0.0200
VOID 0.0475

 

26 %
THERMAL

FLUX AT
v =196

 

18 16 14 12 10 8 6 4 2 0

LETHARGY (¢)

Fig. 3.13. Neutron Flux Spectrum vs. Lethargy for the Hzo-Moderated Cir-
culagting-Fuel Reactor with 3-ft-Diameter Core,

TS6Y ‘01 HIAGWADAA ONIOGNA GOIVAdL HO4
ANP PROJECT QUARTERLY PROGRESS REPORT

The fuel-coolant density value
used for this designset was 2.1 g/em’.
Evaluation of the fuel-coolant density
reactivity coefficient as well as of
the reactivity coefficients for
structure and moderator will be com-
pleted soon. An estimate of the
water-moderator temperature coeffi-
cient, made to determine the effect
of water cooling stability on re-
activity, was found to be:

The stainless steel poisoning effect
was determined as a loss of 2.55%
in k_., per 0.0] increase in volume
fraction of stainless steel. A change
from 0.015 to 0.030 in stainless steel
volume fraction changed k_,, 1in the
water-moderated reactor with 35%

fuel-coolant from 1.024 to 0.987.

Naf-KF-LiF-UF, Fuel Coolant. The

reactor design values for this coolant

in the water -moderated reactor core

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Ak/k i k :
“Zé" =-7.8 X 10°% per degree Fahrenheit.zf: ;?;i same as those given in Ta-
SECRET
ANP~PHY— 239
DWG. 13578
200 S I B
LiF-NaoF-KF-UF4 FUEL COOLANT
VOLUME OF THE SYSTEM OUTSIDE | .
THE CORE: 5.0 ft°
160 b—n LIGHT WATER: DENS!TY=Q.9/667 e
g/cc i
_ Z3 > ft DIAMETER TOTAL
o ] —.
= ~ e e _4«-2%ft DIAMETERURANIUM
w T~ *:_-;-q;‘:i.m:_; o “‘"\ INVEN TORY
2120 3 ft DIAMETER Y |
= | |
= ‘ -
=
% 80
> 3% ft DIAMETER
AT URANIUM
bt _ 1 3 ft DIAMETER | N
40 ST o 2% ft DIAMETER CORE ™
e Ffl_flfl_fl,,+w-” 2
O
0.20 0.30 040 050 060 Q70 080 090
VOLUME FRACTION FUEL-COOLANT IN GCORE
Fig. 3.14. Total Uranium Inventory and Uranium in Core 28 2 Fuanction of

Volume Fractionm Fuel-Coclant in the Core for Hzfi-MQGerated Circulating-Fuel
Reactors with 2%-, 3-, and 3%-ft-Diameter Cores and LiF-NaF-KF-UF, fuel-Coolant.

28
FOR PERIOD ENDING DECEMBER 10, 1951

Figure 3.14 presents the summary
graphs for the system, showing critical
mass values for the cores and in-
dicating the minimum uranium require-
ments for the systems to be 127 lb for
the 2¥%-ft-diameter core, and 122 1b
for the 3- and 3¥%-ft-diameter cores.
The optimum volume fraction for fuel-
coolant in the core is mear 0.50, with
a very small sensitivity to this
parameter indicated by the small rate
of change between 0.30 and 0.60 volume
fractions.

This reactor has approximately 80%
thermal fissions.

Comparison of H,0- and BeD-Moderated
Reactors. These reactor series serve
to indicate the regions of interest
for possible designs. No attempt has
been made to compute any one design
in detail because of the sensitivaty
of reactivity calculations to shape
and constituents. The relative ac-
curacy of the results depends not only
on the usuvual assumptions but also on

the proximityof the estimated physical
constants {(such as the fuel-coolant
density) to their actual values, as
well as on the structural details.

Some of the differences between the
water-moderated and the Be0O-moderated
reactors are given in Table 3.4. More
detailed comparisons may be made
directly from the various spectra
that are presented for both types of
reactor.

In addition to the i1tems compared
in Table 3.4, the poisoning effect of
stainless steel is somewhat worse for
the BeO-moderated reactor. On the
other hand, there is a great deal
more escape of fast neutrons from the
H,0-moderated reactor. However, no
important "physical"difference between
the reactors has appeared to date, so
that design considerations must be
given the greater weight 1in choice
of a reactor type. Attention 1is
called to the large increase in
uranium inventory caused by the use

Table 3.4

Comparison of Be(- and Hzo-Moderated Reactor

 

 

 

 

 

 

ITEM COMPARED BeO REACTOR H,0 REACTOR
Fuel NaF-UF, NaF-BeF,-UF,
Core diameter (ft) 3.5 to 4.0 3
Fuel-coolant volume fraction 0.36 0.71
Critical mass(® (1b) 26 to 30 26
External fuel volume (estimated) (ft3) 15 15
Total uranium inventory(®) (1bh) 69 to 72 63
Ratio of fuel in core to total fuel 0.37 to 0.42 D.41
Percent thermal fissions 52 to §5 69
MOSEZ:E?r dg;sity-temperaturg reactivity coeffi- 0.6 x 10°° 7.8 x 10°°
At 183°F -9.93 x 10°°F -14 x 1075
At 1500°F -9.85 x 10°°

(e necessary heterogencity of structure will increase these values sowewhat.

 

B3
N3
ANP PROJECT QUARTERLY PROGRESS REPORT

of large amounts of potassium in the
fuel-coolant, It is also very im-
portant to minimize the structure
volume fraction in any of the above
reactors.

Effect of Delayed Neutrons on
Kinetic Response (C. B. Mills, ANP
Division). The importance of delayed
neutrons for controlling the speed of
response of a stationary-liqguid-fuel
ARE (Aircraft Reactor Experiment) was
determined by repeating a calculation
in which an increase in inlet coolant
temperature of 25°C is 1introduced
stepwise and then maintained and a

 

control rod is actuated to 1insert or
remove 0.00025 unit of reactivity per
second. Two-thirds and four-fifths
of the delayed neutrons were removed
from the kinetic response equations,
and the twe curves for (¢ - ¢0)/¢0 vs.
time were compared with the curve for
no loss of delayed neutrons.

The results in Fig. 3.15 1ndicate
a rate of change of flux three times
as great as normal when the fraction
of delayed neutrons is decreased by
80%. For a 67% decrease the rate of
change of flux 1is almost twice as
great as normal.

SECRET
ANP—PHY-246
g 3600 DWG. 13579
@ NITIAL SLOPE OF THE FLUX VS. TIME CURVES
— ¢-¢
d 0 PERCENT LOSS
X 3200 4, OF DELAYED
g 1 NE UTRONS
E:¢ 2800 00 80
< O 0.07 67
=i 0.03 0
- . 2400
™ wJ
L 5 80% LOSS OF
. = 2000 DELAYED
v > NEUTRONS 7% LOSS OF
D
ooz DELAYED
- © 1600 NEUTRONS
<L N
A
W O
o 1200
wl
>
- 800 0% LOSS OF
< DEL AYED
tl 400 NEUTRONS
-
T e
- 0
1.0 2.0 3.0 40
TIME (sec)

Fig. 3.15.
67,

30

Fractional Change im Neutron Flux
and 88% Decrease in Delayed Neutrons.

as a Function of Time for 0,
FOR PERIOD ENDING DECEMBER 10, 1951

These results were then extrapolated
to the case of kinetic response of
the circulacing-fuel reactor. The
factors i1im the kinetic calculations
described above areabout the same for
the stationary and the circulating
fuel reactors with the single ex-
ception of fuel expansion in the fuel
tubes. However, for times less than
0.5 sec there 15 very little daif-
ference in the rate of change of flux
between the case with fuel expansion,
as it occurs in the stationary fuel
and the case with no expansion. 1t
appears therefore thatif a disburbance
such as the one specified above is
introduced, the kinetic response
within 1 sec thereafter 1s not sensi-
tive to fuel expansion and the circu-
lating-fuel-reactor flux will follow
a rate-of-change curve very similar
to that shown on Fig. 3.15. The loss
0of delayed nentrons from the core
up to 80% of these there for the
stationary-fuel case should not in-
crease the kinetic response rate of
the liquid fuel reactor by more than
a factor of three times that for the
reactors with stationary fuel for
which kinetic response data are
available.

CIRCULATING-MODERATOR REACTORS

B. T. Macauley C. B. Mills
ANP Division

Tentative designs for alkali hy-
droxide ~~moderated reactors
been computed to determine the re-
actor criticality constants. These
are the 200- and 400-Mw sodium hy-
droxide —moderated and -— cooled
reactors. The design data are given

in Table 3.5.

have

Reactor criticality constants for
the 200~ and 400-Mw sodium hydroxide
reactor and for a 200-Mw potassium
hydroxide reactor are given in Ta-
ble 3.6.

Table 3.5

The HadB-CTooled and ~ Moderated

Reactor
(July 12, 1951)
Reactor power (Mw) 200 400
Reactor core diameter {ft) 24 K4

RBeactor core material
{volume fractions)

Stationary liquid fuel 0.0477 0.0148
(NaF-UF4¥
Fuel~-element structure 0.1766
0.1252
Baffle structure 0.0257
Moderator~-coolant 0.7500 0.8600

Apparently a larger reactor will
be required if 1t 1s necessary to
use KOH. However, the critical mass
for a 3k-ft-diameter KOH-umoderated
reactor is only 70 1b. The higher
critical mass of the smaller core
(see table} i1s mecessitated by the
greater leakage of that core. More
complete data and analysis of these
hydroxide reactors will be presented
in the next report.

Figure 3.16 shows the sensitiveness
of this reactor to the stainless steel
used for structure. Evidently stain-
less steel 1s an 1mportant reactor
poison.

ERRORS IN REACTOR PHYSICS CALCULATIONS
C. B. Mills, ANP Division
The value of the present method¢(!'??

of reactor calculations lies in the
large amount of accurate information

 

{1)p, K. Holmes, The Muitififigug Method as
U;gd)by the ANP Physics Group, -38 (Feb. 15,
1951).

M. J. Nielsen, Bare Pile Adjoint Solu-
téggj ORNL, Y-12 site, Beport Y-¥F10-18 {Qct. 27,
19507,
ANP PROJECT QUARTERLY PROGRESS REPORT

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Table 3.6
Reactivity Calculation Resulis on the Alkali Hydroxide Reacters
NaOH, 200 Mw NaOH, 400 Mw KOH, 200 Mw
Core diameter (ft) 24 3% 2%
Critical mass (1b) 52 77 220
Ne/k
S .47 .48 0.065 (at 60 1b mass)
(Dm/m ) 235
Percent thermal fissions 55.5 17.1 42.9
Median energy for fission (ev) 0.04 Thermal 0.15
Ak/k
- . 209
(Ap/p) moderator
Dk/E
—_—rT T -0.221
(AP/P) struc ture
Ak/fk
- = radius) 0.5347
AR/R
SECRET obtained, 1its speed, and 1ts rel-
ANF—-PHY 245 .
| 20 OWG. 13580 atively small cost. Almost any re-
' \\\\ actor can be computed by this method
\\\ 1if the cross sections of the con-
stituents are known.
110 - ™ \
\\\ Results of a comparison between
S uncorrected bare-reactor calculations
.00 - : | .
and experiment show a surprising
consistency for the very simple

 

COMPONENTS BY VOLUME FRACTION

 

 

 

EFFECTIVE MULTIPLICATION CONSTANT

 

 

 

 

 

 

 

 

 

 

0307 LIQWD FUEL (NaF-UFg) 0.0477 ~
FUEL ELEMENT STRUCTURE (5.5)0.1766
[~ BAFFLE STRUCTURE (S.S.) 0.0257 |
MODERATOR=- COOLANT 0.7500
0.80
0 005 040 015 020 0.25

VOLUME FRACTION STAINLESS STEEL IN THE GORE

Fig. 3.16. Effective Multiplication
Constant as a Function of Volume
Fraction of Stainless Steel in the
Core for the 2%-ft-Diameter-Core
NaQH-Moderated and - Cooled Reactor.

32

multigroup method used. Consistency
of multiplication constant deter-
mination is about *1% with an average
error of -2% for water-moderated
reactors of any size of uranium con-
-1% for the
moderated reactor experiment, and
-10% for the beryllium-moderated
critical experiment. The error 1n
the beryllium calculation is beang
investigated in some detail and seems
to be due to the omission of a signif-
icant physical effect [the (n,2n)
reaction, for instance] and to known

centration, carbon-
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

FOR PERIOD ENDING DECEMBER 10, 1951
Table 3.7
Critical Assembly Sizes, Composition, and Calculated
Multiplication Constants
Water-Moderated Reactors {(a Right Cylinder, with Extreme Variation in NH/NU)
CALCULATION DIAMETER HEIGHT MASS OF U233 k.,
{core No.) {cm) (cm) Ny/Ny23s (kg) CALCULATED
Reflected 106 25. 40 22. 40 328.7 0.893 0.970
RBeflected 107 38.10 44.30 999.90 1.31 0.988
Bare 1089 25.40 34.00 52.9 7.90 0.984
Bare 110 25.40 32.30 43.9 8.80 0.985
Carbon Critical Experiment (a Cube)
CALCULATION AREA LENGTH | . MAsS OF U*®® k5
{core No.) (in. x in.) (in.) Ny2as | N ”;1 (kg) CALCULATED
Bare 146 51 x 51 44.111 0.725 770 | 36.8 44.6 0.991
Beryllium Critical Experiment
CALCULATTON AREA LENGTH | Njjas (Ng, |Na, | Mass oF u?%° kg
(core No.) (in. x in.) (in.) (kg) CALCULATED
Bare 60 21 x 21 23.22 3.08 1074 .36.5 19.3 0.903
*x 10°%° per cec.

simplifications in the conservation
equations. Reactors that have had
both critical-experiment test and
bare-reactor calculations are de-
scribed in Table 3.7.

The calculations on the four water-
moderated reactors were based on the
theory of Goertzel and Selengut(?®)
described in the last quarterly re-

 

port.¢*> Two bare and two reflected

(3)G. Goertzel, Criticalit { drog en-
Moderated Reactors, ORNL, ANP, TAB-53 uly 25,
1950) .

43¢, B. Mills, * The Sodium Hydroxide Re-
actor,” Aircraft Nuclear Propuls;on Project
Quarterly Progress Report for Period Ending
?; {fmber 10, 1951, 0 L-1154, p. 70 (Dec. 17,

critical experiments were computed
using the modified age equations
described in detail elsewhere.(%?
The two reflected-reactor calculations
used a correction on the core radius
equal to the slowing-down length of
neutrons in water.

Bare-reactor multigroup calcula-
tions were made on the four reactors
with the multiplication constant
(keff) resulting as shown in Ta-
ble 3.7. Note that these reactors

were selected to give the worst

 

533, w. Nebster, Numerical Technique for
Criticality Calculations on Hydrogen-Moderated
fleacters, OBNL, Y-12 site, Beport Y-F1l0-466
(Aug. 20, 1951).

33
ANMP PROJECT QUARTERLY PROGRESS REPORT

possible case for results, as 1s an-
dicated by the large changes in
critical mass and atomic density of
U?35, The experiments are described

in detail in report K-343.(%®)

It was necessary to use a variable
buckling [B?*(u)] in the age equation
of the small size of the

Constant buckling results

-0.0208.

because
reactors.
were lower by Ak

The relative importance of hy-
drogen moderation was determined by
ignoring moderation by all other

(%¢. b. Beck, A. D. Callihan, J. W. Merfite,
and R. L. Murrey, Critical ¥ass Studies. Part
III, C & OCC Beport K-343 (Apr. 19, 1949},

34

atomic constituvents. This lowered

k,,. of Ok = -0.0247.

A Fermi age equation solution of
this problem without consideration
of the effects of hydrogen moderation
separately results

k .o of Ak ., = 0.192.

in an error 1n
ff

The details of the leakage and
fissioning lethargy distribution were
presented previously.(7) Note that
the percent thermal fission values
change from 94% for core 107 to 41.5%
for core 110, 1.e., from thermal to
intermediate reactor fission distri-
butions for equally good results.

(7)Figures 3.26 and 3.27, ORMi.-1154, op. cit.,
pp- ¥5 and 76.
FOR PERIOD ENDING DECEMBER 10, 1951

4. CRITICAL EXPERIMENTS
A. D. Callihan, Physics Division

A second wmockup of the General
Electric direct-cycle (air-water
cycle) reactor was completed, and the
critical mass of the assembly was
found to be 90 1b of uranium. The
inhomogeneities of the mockup are
serious, however, and the extrapo-
lations of the critical mass of this
assembly to that of the aircraft
reactor are not yet reliable.

Considerable data have been taken
on the uranium graphite assembly, but
the results have not been reduced.

PRELIMINARY ASSEMBLY OF
DIRECT-CYCLE REACTOR

At the request of the General
Electric Company and with its co-
operation, studies are being made of
the proposed direct-cycle nuclear
reactor for aircraft propulsion. This
reactor, as presently conceived, will
be air cooled and water moderated,
with stainless steel as one of the
principal structural materials,
Beryllium will be the reflector. One
series of experiments was outlined in
the preceding report(!) and has been
described more fully elsewhere,{(?)

During the period reported here a
second mockup has been assembled and
made critical. The core is an approxi-
mate right cylinder, 36 in. long and
51 in. in diameter, the periphery of
which i1s not smooth because the

 

(1)a. p. Callihan, “Critical Experiments,”’
Aircraft Nuclear Propulsion Project Quarterly
Progress Report for Period Ending September 10,
1351, ORNL-1154, p. 79 (Dec. 17, 1951). :

(233, A. Hunter, Report of Critical Experi-
ments for a Water Moderator, CA-2, G.E. Report
DC-51-9-11 (Sept. 12, 1951).

smallest structural unit is 1% in.
square. The axis of the cvlinder is
horizontal. The mid-cross-section is
shown in Fig. 4.1. The hydrogenous
material, a methacrylate plastic, is
placed in horizontal layers, 1 in.
thick. These layers are separated by
an open structure of stainless steel
and uranium about 2 in. thick. This
structure consists of six sheets of
steel, each 0.017 in. thick and 36 in.
Jong. The edges of each piece are
bent under in order to separate
adjacent stacked sheets by about 0.25
in., simulating the air ducts of the
reactor. The upper three of each
group of six vertically stacked pieces
are inverted, 1.e., the plane surfaces
of the center pair are in contact. The
3-in,- diameter 0.010-in.-thickenriched
uranium metal disks are placed, with
planes horizontal, in the interspace
between the central steel sheets.
{In some instances, because of insuf-
ficient large disks, four 1¥%-in.-
diameter disks are used.) The center-
to-center spacing between disks 1s
3.6 1n. 1n the direction of the
cylinder axis and 3 in. in the hori-
zontal direction perpendicular to the
axis, the latter being fixed by the
dimensions of the square aluminum tube
matrix into which the reactor com-
ponents are built. The lateral sur-
face of the core is enclosed by a
6-1in.-thick reflector of beryllium;
the ends of the reactor, including the
beryllium jacket, have a graphite
reflector 6 in, thick, Insufficient
available beryllium necessitated the
substitution of graphite as the end
reflector, The void fraction in the
core is 0.57 and the H/U?*3° atomic
ratio is 225. The fuel and moderator
inhomogeneities are seriocus. The
critical mass 1s 41 kg of uranium or

38 kg of U?3%,

35
ANP PROJECT QUARTERLY PROGRESS REPORT

 

o -

Iy

==
——— S
==
o— -

By
;fl;g

 

Fig. 4.1. Mid-Cross-Section of Second Mockup of Direct-Cycle Reactor.

36
FOR PERIOD ENDING DECEMBER 10, 1951

GRAPHITE REACTOR

In the time available prior to the
inception of the program described
above, a series of measurements was
made on the uranium-graphite assembly

described in an earlier quarterly

The results, obtainable
have not been re-

report, (3)
from these data,
alized.

 

(3icritical Assembly of Graphite Reactor,”
ORNL-1154, op. cit., p. B80.

37
ANP PROJECT QUARTERLY PROGRESS REPORT

9. NUCLEAR MEASUREMENT

Data for gamma-ray and neutron
yields from proton bombardment of some
materials has been extended using the
5-Mev Van de Graaff accelerator. This
accelerator has also been used to
measure the total cross-section of
iron from 0.6 to 3.6 Mev and (p,n)
thresholds for F'? and Na??. Upper
limits have been set for the (n,2n)
cross-sections 1n beryllium in an
experiment i1n which a neutron source
was surrounded witha beryllium sphere.
Final adjustments are being made on
the time-of-flight spectrometer for
neutrons.

THE 5-Mev VAN DE GRAAFF ACCELERATOR

H. B. Willard, Physics Division

The gamma-ray and neutron yields
from the proton bombardment of Li7,
Be®, and F!9 have been extended to
approximately 5.4 Mev. These curves
are to be printed in the Physics
Division quarterly report for period
ending December 20, 1951.

The (p,n) thresholds for F!% and
Na2? have been determined to be
4,253 £ 0.005 and 5.09]1 £ 0.010 Mev,

respectively.

The total cross-section of ordinary
iron for fast neutrons has been
measured from 0.6 to 3.6 Mev with a
resolution of 35 kev. Unresolved
levels occur every 100 kev, with
indications that with better resolution
many more would be observed. The
average cross-section varies from 2.5
barns at 0.6 Mev to 3.2 barns at
3.6 Mev. These total cross-section
data have not yet been reduced to
specific cross-sections, which are
used inshield calculations.

38

MEASUREMENT OF THE (n,2n) REACTION
IN BERYLLIUM

E. D. Klema, Physics Division

The data on the (n,2n) reaction in
beryllium discussed in the previous
quarterly progress report{!) have been
reduced by G. B. Arfken to an average
(n,2n) cross-section for the spectra
of the two sources used. The calcu-
lations were made under the assumption
that a neutron which makes any sort of
collision [(n,n), (n,a), or (n,2n)] in
the beryllium sphere is lost as far as
producing the (n,2n) reaction 1is
concerned, The absorption of neutrons
by the beryllium sphere was estimated
from the data obtained with the carbon
spheres.

The upper limits for the (n,2n)
cross-sections as determined by this
experiment are as follows:

o.(n. 2n)

Po-a-B source,

< 0.25 barn

Po-a-Be source,

Sn,2n) < 0.56 barn
The upper limit of this cross-section,
although appreciably lower than that
reported by Houtermans, (2) is still
so high as to be capable of introduc-
ing a significant error in the critical
mass of a reactor assembly employing
beryllium, In reality, this cross-
section is the suspected cause of the
discrepancy in the calculated (disre-
garding the cross-section altogether)
and actual critical mass of the

 

(I)E. D. Klema, ‘“Measurement of the (n,2n)
Reaction in Beryllium,’ Aircraft Nuclear Propulsion
Project Quarterly Pragress Report for Period
Ending September 10, 1951, OBNL-1154, p. 97
(Dec. 17, 1951).

(DF, 6. Houtermans, *“On the (n,2n) Reaction
in Beryllium with Neutrons of a Polonium-Beryllium
Source,” Nachr. Akad. Wiss. Gottingen, Math. -
physik. Klasse IIb Abt. p. 52 (1946), translated
by W. K. Ergen, April 20, 1951 (Y-F20-12).
FOR PERIOD ENDING DECEMBER 16, 1951

beryllium-reflected graphite-uranium
critical assembly.(3)

. TIME-OF-FLIGHT NEUTRON SPECTROMETER

G. S. Pawlicki, ORINS
E. C. Swith, Physics Division

The spectrometer 1s essentially
completed, and calibration aund adjust-
ment of the i1nstrument are in progress.
The best resolution to be esxpected

with this rotor is 1.2 pusec per meter,
which 1s comparable to that of the
best operating spectrometers. Another
rotor has been ordered whiech should
improve this resolution by a factor of
2. Details of the apparatus will be
found in the Physics Division quarterly
reperts for June 20 and September 20.

(3)(3. B. Mills and N. M. Swith, Jr., ‘“Calcu-
lation for the Critice) Experiment,” Aircraft
Nuclear Propulsion Project Quarterly Progress
Repart for Period Ending June 10, 1951, ANP-65,
p. 71 (Sept. 13, 1951).
ANP PROJECT QUARTERLY PROGRESS REPORT

6. EXPERIMENTAL REACTOR ENGINEERING
H. S. Savage, ANP Division

The activities of the Experimental
Engineering Group now center around
the vigorous pursuit of the technology
of both sodium and the fused fluorides
to provide fundamental engineering
data through the development of actual
components: valves, pumps, seals,
heat exchangers, etc. Consideration
of the advantages of a circulating-
fuel-coolant reactor for aircraft
application necessitated a program for
manufacturing fluorides and for
determining their effects on container
materials and reactor components.
Approximately 400 1b of eutectic
LiF-KF-NaF has been produced for test
purposes.

A pump incorporating a gas seal has
satisfactorily pumped fused fluorides
up to 1300°F for 35 hr. Also, fluoride
transfer through small tubes has been
repeatedly demonstrated, as has
successful filtration through a 10-pu
sintered stainless steel filter. Thus
far, fluorides appear to reduce oxides
inside the system and to be severely
corrosive only in the presence of
oxygen. Equipment is being assembled
as rapidly as possible for conducting
self-welding and pump-shaft-seal
experiments with fluorides.

Continued work with ligquid-metal
systems during the quarter resulted in
a pump incorporating a frozen sodium
seal operating approximately 450 hr
with sodium at temperatures up to
1100°F with a pressure across the seal
of 25 psig. In other equipment, a
seal test rig operated with sodium at
1500°F sealing successfully against
107 psig. Electromagnetic pump-cell
development reached the stage at which
sodium was pumped at temperatures up
to 1300°F for approximately 160 hr;
test-loop heater limitations prevented

40

the attainment of higher sodium temper-
atures. Self-welding tests were
continued with sodium, and equipment
1s being modified to continue work of
this nature with fluorides.

Valve testing was continued to
evaluate certain commercially available
varieties for use in liquid-metal
systems. In some cases instrumentation
needs for fluorides are radically
different from those required for
liquid-metal systems; this program has
thus undergone change compatible with
the changes 1in ARE. Null-type pressure-
measuring devices were belng tested
with both ligquid-metal and fluoride
systems at the end of the quarter.
Preliminary work on heat exchangers
and analytical methods for sodium were
continued during the period, as were
cleaning procedures for both fluoride
and ligquid-metal systems. Planning of
full-scale ARE component testing
facilities were being continued along
only such lines as would allow the
facilities to be used for testing
either liquid-metal or fluoride
systems.

PUMP DEVELOPMENT

The centrifugal pump employing a
gas seal of graphite set against a
flat sealing face has operated with
NaK and subsequently with the fluoride
eutectic LiF-KF-NaF. The latter test
system was operated for 35 hrat 1300°F
and was terminated because of a leak
outside the pump. The design of the
centrifugal pump of the above type for
the liquid-metal (NaK) ARE has been
completed. The 1V-hp canned-rotor
pump, as rebuilt with high-temperature
(500°F) windings, has been satis-
factorily pretested with water, but
some difficulty has arisen in canning
FOR PERIOD ENDING DECEMBER 16, 1951

the rotor. The 6-hp canned-rotor pump
operated poorly because of an inef-
ficient suction inlet, which is being
redesigned. The centrifugal pump with
a frozen-sodium seal operated with no
indication of seal failure for 450 hr
with fluid temperatures up to 1100°F,
Limited development of electromagnetic
pumps has continued, and two new
design cells have been tested.

Centrifugal Pumps for Figure-Eight
Loops (W. G. Cobb, ANP Division). The
centrifugal pump for liquid metals(1)
was operated on two occasions with NaK
as the pumped fluid. 1In the first
test pumping was continuous for 12 hr
and the NaK temperatures reached 820°F
be fore a leak
necessitated a

in a welded joint
shutdown. The second
test also was conducted with NaK, and
temperatures reached 940°F during the
12-hr pumping period. Excessive gas
leakage through the shaft packing and
gasketed joint under the stuffing box
necessitated termination of this test.

The pump and system were cleaned,
reassembled, and provided with heaters
and insulation to allow preheating to
approximately 1000°F. The system was
then filled with the eutectic LiF-KF-
NaF. Flow, metered by a sharp-edged
orifice with pressures being indicated
by gas trapped in bulbs connected to
the orifice taps, was approximately
4.8 ft/sec (about 6 gpm). Level
control was accomplished by means of
electrical probes and gas pressure
directly in the pump body. The test
system ran continuously for 35 hr,
pumping at temperatures up to 1300°F.
Considerable encouragement was derived
from the fact that gas leakage through
the shaft seal was negligible through-
out the run, and a2 small but definite

 

(1)“C€ntrifugal Pumps for Figure-Fight Loops,”
Aircraft Nuclear Propulsion Project Quarterly
Progress Report for Period Ending Junz 10, 1951,
ANP-§5, p. 168 (Sept. 13, 135i).

signal was observed from the electro-
magnetic flowmeter. Moreover, smooth
operation was obtained with a gas-held
level control. Although several small
leaks in the system developed, omne of
which eventually caused termination,
the pump itself operated satisfactorily
throughout the run.

ARE Centrifugal Pump Design. (W. G.
Cobb, L. V., Wilson, and J. F. Haines,
Consultant, ANP Division). The design
and detailing of a centrifugal pump
employing NaK for the ARE has been
completed, General design features
are compatible with those listed
previously(?? with the exception of
level-control features. Avoiding the
necessity for simultaneously develop-
ing a satisfactory pump and a level
control device appeared advisable;
this pump is therefore to be located
in the fluid circuit on the same level
as the expansion tank. Therefore, the
ligquid level in the pump 1s determined
by the level in the expansion tank,
and sufficient shaft overhang has been
incorporated to compensate for level
variations occurring in the surge
tank. A gas seal of graphite against
a flat sealing face has been in-
corporated in the design, and tempera-
ture control of bolted joints and pump
body above the casing is to be effected
by circulating gas. Also, provisions
for internal cooling of the shaft by
circulating gas have been included.
All bearings are external to the gas
space above the coolant being pumped,
and a circulated lubricant serves the
bearings as well as the graphite seal.
All parts except the impeller are to
be of fabricated and machined con-
struction. Impellers of Worthite have
been received from Worthington Pump
Corporation,

. G. Cobb and J. F. Haines, “ARE Pump
Design,” Aircraft Nuclear Propulsion Project
Quarterly Progress Report for Period Ending
Sep{;mber 10, 1951, CORNL-1154, p. 16 (Dec. 17,
1951).

i
i—l
ANP PROJECT QUARTERLY PROGRESS REPORT

Canned-Rotor Pump (M. Richardson,
Reactor Experimental Engineering
Division), The NaK test loop discussed
in the previous report 1is being
mod1fied to include a 30-kw heater and
a vertical double-tube heat exchanger
in which water flows as a falling film
down the 8-ft length of the inside
2-in. tube and the sodium-potassium
alloy rises in the annulus formed by a
3-in. pipe. The two canned ¥%-hp
motors, wound with 500°F Class H
windings, have been tested in water
and found to operate satisfactorily.
Attempts are being made to can the
to protect them from
high temperatures.
Stainless steel has been used and
found to operate satisfactorily, but
the increase in gap caused by the
0.010-1n. wall of stainless steel
causes the current load to become
excessive, Some magnetic material
will be tried next which will protect
the rotor and reduce the stator-rotor

gap.

rotor elements
COTTrosi1on at

The 3- to 6-hp canned-rotor pump
was operated for a short period but
the performance was poor. It 1s
believed that this poor performance
was caused by the turbulence and
thrust load which, in turn, were
caused by the poor design of the
suction inlet to the pump. This 1is
being corrected by straightening out
the sharp bend at the suction face and
placing a baffle at the inlet to reduce
the pressure on the shaft at this
point, A sleeve bearing will be
introduced within the motor can which
will protect the can from damage
caused by misalignment but will allow
the rotor bearings
draulically.

to operate hy-

A new wire 1nsulating material
which is manufactured by the Bentley-
Harris Insulating Company of Coshocton,
Pa., would allow the canned-rotor

pumps to operate at 1200°F. This

42

insulation will be 1included in future
pumps. Investigations are underway to
determine the possibility of making
the rotors of suitable material, which
wi1ll make canning unnecessary.

Frozen-Sodium-Seal Pump (W. B.
McDonald, ANP Division). Testing the
frozen-sodium seal under dynamic
conditions was effected in an 1sothermal
test loop containing a modified
Worthington centrifugal pump. This
test ran approximately 450 hr at sodium
temperatures up to 1100°F, alimitation
imposed by silver-solder pump-casting
repairs. Flows up to 40 gpm were
attained, and the pressure differential
across the seal was 25 psig. No
indication of seal failure was observed
during the test.

Two-Stage Electromagnetic Pump (J.
H. Wyld and A, L. Southern, ANP
Division). Some parts of the two-
stage electromagnetic pump have been
received from the main shop; however,
the increase in emphasis on fused
fluorides has necessitated postponing

plans for immediate testing.

Electromagnetic Pump Cell Develop-
ment (A. G. Grindell, Engineering and
Maintenance Division). Two types of
electromagnetic pump cells designed by
ANP personnel for high-temperature
liquid-metal applications were given
preliminary tests during the period.
The first of these featured integral
construction of the throat and elec-
trodes from type 316 stainless steel.
Initial tests indicated that severe
local overheating occurred. The
second cell, employing nickel lugs
inserted into the stainless steel
pump throat and welded in place with
inconel filler, operated satisfactoraly
for approximately 160 hr, pumping
sodium at temperatures up to 1300°F.
Test-loop heater limitations prevented
raising the sodium temperature above
the 1300°F value.
FOR PERIOD ENDING DECEMBER 10, 1951

SEAL TESTS
W. B. McDonald, ANP Division

Many problems arising in connection
with pump development for liquaid
metals and fluorides may be more
properly classified as seal problems.
A program is underway in which various
types of seals are evaluated on pumps
as well as in simulated eguipment.
Seals currently being developed are
the frozen-sodium seal, the frozen-
fluoride seal, and the graphitar
ring-—tool steel gas seal. The frozen-
sodium seal has been successfully
tested for 1150 hr with pressure across
the seal of up to 107 psig. Only
limited success has yet been ex-
perienced with the two other seals.

Frozen-Sodium Seal. During this
quarter one phase of testing the
frozen-sodium seal was successfully
completed., Using the seal-testing
device described in a previous re-
port,¢?? the frozen-sodium seal
operated for a total of 1150 hr at
sodium bath temperatures up to 1500°F.
Maximum pressure across the seal with
no indication of failure was 107 psig.
Just prior to termination, however,
the pressure was raised to 150 psig,
resulting in a small amount of sodium
being forced out of the system around
the shaft. Pressure was further in-
creased to 200 psig in an effort to
make the seal fail completely, but the
leakage observed appeared to be no
worse than at 150 psig. Leakage was
at no time sufficient to cause a
serious sodium fire; rather, it was
characterized by sparks emerging from
around the shaft.

A design has been completed in which
the refrigeration coil, used in the
tests above to freeze the sodium, heas
been replaced by copper cooling fias
attached to the stainless steel sleeve.

 

(3)%'. B. McDonald, “Frozen-Sodium Seal,”
ORNL-1154, op. cit., p. 21,

Calculations indicate that a section
approximately 5 in. long and contain-
ing 3%-in.-diameter fins spaced 1/8 in.
apart are sufficient to freeze sodium
around the shaft by convective cooling
only. This radiant-cooled seal has
not been tested.

Frozen-Fluoride Seal. Design and
assembly have been completed on a
device with which to determine the
feasibility of a frozen-fluoride seal.
The principle involved is identical to
that of the frozen-sodium seal in
which the liquid is solidified in an
annulus between a sleeve and a rotating
shaft. (3)

The first test has been completed
in which the shaft, extending through
a finned cooling sleeve into a pot
containing eutectic NaF-KF-LiF-UF, at
1200°F, was driven by a 5-hp motor.
Pressure across the seal was 20 psig.
The test lasted for 19 hr hbefore a
heater failure and a leak caused
termination.

Postrun inspection revealed that
material was deposited on both the
shaft and sleeve, and some scoring
at the bottom end of the shaft was
observed. Whether the attack on the
shaft 1s chemical or mechanical was
undetermined, but the material deposits
were found to be magnetic.

Graphitar Ring—Ketos Toel Steel
Gas Seal. Testing of a graphitar
ring—Ketos tool steel gas seal was
continued during the period. Thas
type of seal, consisting of a graphitar
ring sandwiched between a stationary
and a rotating member both made of
Ketos tool steel full hardened, was
tested in the presence of air and of
helium to secure control data for
comparison with runs made in the
presence of NaK vapors. Successful
sealing was obtained in these i1nstances
with differsntial pressures of 5 psig.

43
ANP PROJECT QUARTERLY PROGRESS REPORT

Three tests were conducted with the
seal operating in the presence of
vapor from NaK bath at 600°F, a seal
temperature of 500°F, a spindle speed
of 1300 rpm, and a differential
pressure of 5 psig. The first test
failed after 24 hr, and postrun ex-
amination indicated that the graphitar
had been attacked. Subsequent tests
operated under similar conditions with
no apparent attack on the graphitar
ring for 70 and 170 hr,
Failure in both cases appeared to be
due to eccentricity in the seal test
device. As the period ended, a fourth
test had been in progress for 220 hr
with no indication of failure. Ap-
parently the graphitar in the first
test was faulty,

respectively.

TEST LOOPS

A. G. Graindell, Engineering
Maintenance Division

A calibration loop for checking
electromagnetic pumps and flow measur-
ing devices is now being used routinely
for this purpose. The sodium manometer
loop for measuring flow rates has not
proved successful.

€alibration Loop. The calibration
loop was designed for testing electro-
magnetic pumps and flow measuring
devices. During the previous period
this loop was used for reproducing
manufacturer’s performance curves for
a type G-3 General Electric electro-
magnetic pump. This work was con-
tinued i1nto the present period with
the result that two electromagnetic
pump cells designed by ANP personnel
were tested. HResults of these tests
are included in the section "Electro-
magnetic Pump Cell Development™ (see
above).

Sodium Manometer Loop. During the

previous quarter, the sodium manometer
loop, designed for developing and

44

testing flowmeters, was used for test-
ing an electromagnetic pump. First
attempts to determine sodium flow in
the system resulted 1n plugging of the
small tubes connecting the flow nozzle
to the primary manometer. Efforts to
prevent this plugging by using NaK in
place of sodium as the working fluid
failed because NaK shorted electrical
probes 1n the level control device.
De-emphasis on liguid metals systems
in favor of fused salts caused this
project to be placed on a standby
basis.

SELF-RELDING TESTS

The self{-welding of metal to metal
contacts in a high-temperature fluid
stream, as might occur in reactor
pumps or valves, is being examined.
Tests of five pairs of metals under
static load in sodium for 100 hr have
been completed. Stellite against
stellite and inconel against zirconium
exhibited no welding. Test of valves
in sodium systems indicated a signifi-
cant increase 1n opening torque with
time. The valve had a stellite face
and a 316 stainless steel seat.

Materials Tests (G. M. Adamson,
Metallurgy Division)., Several ma-
terials were examined to determine
their self-welding properties at high
temperatures in the presence of liquid
sodium. A series of three cylinders
with diameters of 1/8, 1/4, and 3/8
in. were machined from one material to
be tested, and three corresponding
cylinders 1/2 in. in diameter were
machined from either the same or
different material. After being
polished, pairs of these were placed
end to end in the desired liquid bath
under a load of 75 1b for 100 hr. The
results, obtained in sodium at 1500°F,
were:

Few points of welding
but broke during
handling

Inconel-inconel

(120 grit)
FOR PERIOD ENDING DECEMBER 10, 1951

Inconel-inconel Welding hed just started
(polished) but pieces could be
hand] ed

Inconel - zirconium No welding

Few points of welding on
two smallest cylinders

347 stainless stee]l —
347 stainless steel

Stellite-stellite No welding in 65 hr
Valve Tests (W, C., Tunnell and K.
W. Beber, ANP Division). Valve testing
experiments were begun during the
gquarter to determine the suitability
of commercially obtained valves for
use in liquid-metal systems. The
first test was conducted on a Y%-in.
Powell bellows valve with a stellite
face and a 316 stainless steel seat,
The valve was installed in the hot leg
of a thermal-convection loop which was
filled with sodium and heated to
1500°F. The valve was to be left open
for 24 hr and then closed for 24 hr
with a torque of 20 ft-1b. The torgue
required for opening the valve varied
widely during the 323 hr of the test,
although the valve was operable at all
times. Postrun inspection of the
valve awaits availability of suitable
inspection equipment. :

A second test was started in which
two %-in. Powell bellows valves were
placed in parallel]l in the test loop
in order that the valves could be
alternately opened and closed on 24-hr
cycles, thereby permitting convection
fiow of the sodium at all times. The
valves in this test also had stellite
faces and 316 stainless steel seats
and were closed with 35 ft-1b torque.
Torgques required to open each ranged
from 20 to 53 ft-1b. The time of
closure was increased from 24 to 48 hr,
and the torque required to open the
first valve was 76 and 44 ft-1b after
the first two cycles, while the second
valve required 46 ft-1b after one
cycle, Valve tests were still under-
way at the end of the period.

The study of bellows type valves,
with the bellows surrounded by con-

ventional packing material, has been
suspended owing to the successful
operation of the conventional valves
and the emphasis on fluorides.

HEAT-EXCHANGER TESTS

A. P. Fraas and M. E. LaVerne
ANP Division

A heat exchanger with NaK as both
the primary and secondary fluid has
operated for 550 hr at temperatures up
to 1200°F. The agreement between
actual and expected heat transfer 1is
good at high (11 gpm) flow rates but
is poor at low (5 gpm) flow rates. A
heat exchanger (radiater) with sodium
as the primary fluid and air as the
secondary fluid has been designed and
built.

NaK to NaK Heat Exchanger. The NaK
heat exchanger figure-eight loop has
completed 550 hr of operation at
temperatures up to 1200°F and flows
of 5 and 11 gpm. Increased tempera-
ture could be achieved only by reducing
the flow rate. Present plans call for
continuing testing until 1000 hr of
operation has been completed.

Thus far there has been indicataon
of neither any increase in pressure
drop across the heat exchanger nor any
loss in heat-transfer performance.
This has been reassuring since the
heat-exchanger matrix was made of 1/8-
in.-o.d. tubes spaced 0.041 1n. apart,
and it had been feared that clogging
might be a preblem. All experimental
features of this loop have proved quite
satisfactory; for example, the nicro-
brazed electromagnetic pump cell,
by-pass filter, venturi, and null-type
pressure units have given no trouble.

Heat-transfer data at high flow
rates agree well with values computed

from Lyon’s formula, but data at low

45
ANP PROJECT QUARTERLY PROGRESS REPORT

CLASS I FIED

{
N

=
=
—

46

 

ir

Rad

 

iator.
FOR PERIOD ENDING DECEMBER 10, 1951

flow rates give heat-transfer co-
efficients as much as 40% below the
theoretical values,

Sodium to Air Radiator. A high-
performance air-radliator core section
suitable for use 1n a J-53 turbojet
engine has been designed and con-
structed (Fig. 6.1). The core matrix
consists of 0.010-1n.-thick stainless
steel fins spaced 15 to the inch on
3/16-in.-0.d. stainless steel tubes.
The latter have been placed in a square
spacing on 2/3-in. centers. Parts for
the core-element test section have
been completed and are ready for
nicrobrazing, Preliminary testing 1is
to be effected with air supplied by a
blower extracting the heat which 1s
supplied by electrically heated Nak
pumped through the radiator.

LIQUID-FUEL SYSTEMS

E. Wischhusen and D. R. Ward
ANP Division

The fluid dynamics of the ligquid-
fuel system of the liquid-metal-cooled
reactor have been studied from mockups
of the fuel system. The first two
mockups(*’ were primarily glass systems
with a colored fluid and yielded use-
ful qualitative information on such
problems as filling and emptying rates,
valving requirements, bubble formation
and migration, and flushing methods.
A third mockup of the static-fuel
system was intended, but the design
change to a circulating-fuel system
and the difficulty in transferring
molten fluorides has directed attention
to their handling.

The fuel-transfer apparatus consists
essentially of two tanks, one in which
the raw material 1s introduced and a
second to which the molten fluorides

 

g, L. Meem, ‘““Bulk Shielding Reactor,”
ORNL-1154, op. cit., p. 83.

are transferred through a filter by
gas pressure, Fuel lines are 347
stainless steel while all other com-
ponents are of 316 stainless steel.
The filter is of 5-u stainless steel.
After the first transferral, a viewing
window was provided in the charge tank.

The fuel has been successfully
filtered and transferred from one tank
to the other., The unfiltered fluid is
generally dark, the filtered fluid
rather bright green and uniform in
appearance. At the transfer tempera-
ture, 1400°F (the melting point of the
fuel, NaF-KF-LiF-UF_, is 960°F), the
fuel has a watery consistency and a
concave meniscus. Postrun examination
of the equipment revealed a black
layer of uranium oxides on the bottom
of the supply tank and on the inlet
side of the filter. Portions of the
system wet by the fuel mixture exhibited
a rather bright satinlike appearance,
indicating the removal of welding
scale and oxides. Subsequent transfer
resulted in a leak in a connecting
fitting and after that a plugged
transfer line. Examination of the
filter now shows the presence of
bundles of dendritic metal crystals
which are being analyzed.

INSTRUMENTATION

J. F. Bailey and P. W. Taylor
ANP Division
A. G. Grindell, Engineering
and Maintenance Division

Instruments capable of giving
adequate indication and control of
either liquid-metal or fluoride systems
must be developed., Work is being
carried out on the development of
level-control, flow-measurement, and
pressure-measurement devices for the
ARE. Spark plugs appear satisfactory
for level control sincemolten fluorides
conduct electricity. Venturi methods

47
ANP PROJECT QUARTERLY PROGRESS REPORT

appear to be best suited for the
measurement of the flow of fluorides.
The null-balance type pressure-measur-
ing devices are suited for both liguid-
metal and fluoride systems.,

Level Control and Indication. Work
on instrumentation for level control
and indication for liquid-metal systems
centered around the testing of resist-
ance type probes. An experimental
loop for testing level control and
indicating devices was fabricated in
which a resistance type probe immersed
in sodium was connected to a Brown
potentiometer through a Wheatstone
bridge. As the sodium level fluctuated,
the resulting resistance change of the
probe produced a signal which both
indicated and controlled the level.
This method proved unsatisfactory
since the probe resistance constantly
changed as a result of sodium buildup.

[imited experience with fused
fluorides indicates that sparkplug
type level controls are adequate for
laboratory-scale systems. Solid-
state fluorides are nonconductors,
which eliminates the problem of elec-
trical shorting of probes previously
encountered with sodium,.

Flow Measurement. The shift in
emphasis to fused fluorides caused a
change in flowmeter development.
Electromagnetic flowmeters had received
major consideration for use with liguid
metals, but the comparatively poor
electrical conductivity of fused
fluorides as compared to container
materials made this means for detect-
ing flow appear impractical. On the
other hand, venturi methods of flow
measurement appear to be quite practi-
cable, having virtue in their sim-
plicity of construction and small head
loss under operation. Since this method
involves differential pressure measure-
ment, null-balance type pressure-
measuring devices are being examined.

48

Pressure-Measuring Devices. Seek-
ing suitable methods for measuring
pressure i1n liquid-metal systems led
to the modification of null-balance
type devices to meet specific needs.
The device in use, as modified,
operates on the principle of balancing
pressure on either side of a bellows
and employing electrical methods for
establishing approximate bellows
The bellows
1s essentially filled with
trapped gas which transmits circu-
lating-fluid pressure. This pressure
1s, in turn, balanced by gas pressure
applied external to the bellows. The
vent valve enables the pressure of the
system to be "bracketed" between two

expansion or position.
volume

values within 0.1 psi, the pressure
required to expand the bellows suf-
ficiently to make electrical contact
which 1s indicated by a signal light
and/or a buzzer.

Although these pressure-measuring
devices were designed and fabricated
primarily for use with liquid-metal
systems, the only foreseeable objection
to their use with systems containing
fused fluorides 1is the methods used
for joining the bellows to bellows
heads. The search for suitable methods
for making these connections centers
around heliarc welding, resistance
seam welding, and nicrobrazing. Thus
far, two devices have been assembled
using nicrobrazing for joining bellows
and heads and have been installed on
the gas-seal fluoride pump.

FULL-SCALE ARE COMPONENT
TEST FACILITIES

H. P. Kackenmester and G. A, Cristy
ANP Division

Plans for full-scale testing
facilities for ARE components to be
located in Bldg. 9201-3 were reviewed
in relation to providing them by late
winter or early spring, 1952, For
FOR PERIOD ENDING DECEMBER 16, 1951

the present, utilities and services
are being provided sufficient to handle
fluorides and other equipment capable
of being safely operated without test
cell enclosures.

Plans and specifications have been
developed for construction of an
equipment train in which fluid-flow
measuring instruments can be calibrated
and pumping characteristics can be
determined by weight-rate measurement
of fluid flow. Equipment for the
calibration system will be capable of
handling either liquid metals or molten
salts. Jnquiries to manufacturers
indicate that low-frequency (60-cycle)
induction heating apparently offers a
versatility along with higher heat-
input rates to vessels contalning
either liquid metals or molten salts
than can be achieved by radiant- or
conductive-heating equipment. Also,
investigations of fume-collecting
equipment show that several types of
relatively inexpensive commercial
equipment are available which will
collect 99%% of condensed dispersoids
of liquid wetals or liquid salts 1in
large guantities (20,000 to 50,000
cfm) of air in temperature ranges of

0 to 2000°F.

FLUORIDE PRODUCTION

L. A. Mann, ANP Division

The decision to make engineering-
scale tests on eutectic salts of
LiF-NaF-KF activated earlier plans for
assembling and fabricating production
type equipment and storage contalners
for this fluoride mixture. Liaison
with the Liguid Fuels Chemistry Group
was maintained in the effort to ensure
adequate quality and purity of product.

Early runs were made by charging
c.p. salts 1nto a melt tank 1n proper
proportions for producing the eutectic

mixture, purging air from the system,
and melting at 1825°F under 5 psi of
helium. Material was forced from the
melt tank into the receiver through a
5-4 sintered 316 stainless steel
filter by helium gas pressure. After
three melts, succeeding runs were made
uvnder approximately a 500-u Hgvacuum,
Also, 316 stainless steel scrap was
put into the chamber to "precondition"
the mixture., Thus far, seven melts
have been made, totaling approximately
400 1b of eutectic fluorides. The
addition of desired amounts of UF, to
certain batches has been accomplished
in the receiving tank. The entire
contents are agitated by bubbling
helium through the mixture.

CLEANING OF FLUORIDES VFROM SYSTEMS
R. Devenish, ANP Division

With the advent of fused fluorides,
the problem of removing the residue
from containers in asafe and efficient
manner arose. Sodium and potassium
fluoride constituents are comparatively
water soluble and hence present less
difficulty; however, lithium fluoride,
being only slightly soluble in water,
adhered to the container walls. The
best method found thus far for removing
this material i1is with hot 25 vol %
nitric acid solution. Aluminum
nitrate for this application proved
unsatisfactory.

In collaboration with the Health
Physics group the following pre-
cautions have been temporarily es-
tablished for working with fluorides:

1. Dust respirators should be worn
by personnel when charging the
fluorides of lithium, sodium,
and potassium i1nto production
furnace pots.

2. Assault masks withM-11 canisters
should be used when working with

49
ANP PROJECT QUARYTERLY PROGRESS REPORT

molten fluorides in open con-

talners.

NakK DISPOSAL
R. Devenish, ANP Division

A method was formulated for dispos-
ing of used NaK which previously had
presented difficult storage problems.
In this procedure NaK was mixed with
graphite powder and kerosene, and the
resulting slurry was oxidized off with
steam. A continuous fire was observed
until all the NaK was consumed.

50

ALKALY METALS MANUAL
R. Devenish, ANP Division

The Alkali Metals Manual setting
forth safety considerations to be
observed when working with alkali
metals was distributed during the
period. Also, a Y-12 plant procedure
covering handling of liguid metals was
prepared by joint effort of the Y-12
Alakli and Liguid Metals Safety
Committee and the Plant Procedures
Group. This procedure was submitted
to the Y-12 plant superintendent for
approval,
 
 
SUMMARY AND INTRODUCTION

E. P, Blizard, Physics Division

The aircraft unit shield design as
specified over one year ago by the Lid
Tank optimization experiments has heen
verified in the Bulk Shielding Facility
(Sec. 7). The calculations are now
complete and show a slightly lighter
shield (127,700 1b actual weight) than
was originally predicted. The experi-
ments in this shielding reactor have
been started on the divided shield.
A long period of detailed measurement
will ensue, however, before a divided
shield weight will be available.

The most significant progress in
the shielding art has been the air-
duct theory, which grew directlyout of
a series of tests in the Lid Tank
(Sec. 8). Measurements of the neutron
at tenuation in air ducts agree to
within 20% of the calculated dose.
For these duct sizes (2 in. 1.d.) with
bends, no correction 1s necessary for
streaming and the net effect of the
ducts on the shield weight is of the

order of 1 to 2 tons. This favorable
result with air ducts implies that
liguid-metal ducts would have a
negligible effect on total shield
weight.,

The divided shield mockup, mentioned
above, cannot provide shield weights
with the degree of accuracy possible
with the unit shield because of the
uncertainty associated with the trans-
mission of gamma rays in air. Conse-
quently, a proposal has been submitted
to the AEC for a "Tower Shielding
Facility” with which air scattering,
as well as the effect of this scat-
tering on the crew shield, may be
determined (Sec. 9). Shielding weights
and thicknesses have been estimated
for the circulating-fuel reactor and
do not appear to be significantly
different from those for the sodium-
cnoled reactor in spite of the circu-
lating fissionable material.

53
ANP PROJECT QUARTERLY PROGRESS REPORT

. BULK SHIELDING REACTOR

7

L. Meem
G. Cochran

. P. Haydon
M. Henry
B. Holland

H. E. Hungerford
. Johnson

. Leslie

. Maienschein

G. M. McCammon

E.
J.
F.

OR®

T. N. Roseberry
Physics Division

During this quarter themain efforts
at the Bulk Shielding Facility have
been devoted to analyzing and reporting
the results of the measurements on
the unit shield mockup and inpreparing
for and installing the divided shield
mockup.

REACTOR OPERATION

Preliminary measurements on the
temperature rise of the water flowing
through one of the reactor fuel elements
have been made, and the results look
reasonable. However, to obtain an
accurate value for the heat released
per fission, an especially insulated
fuel element is being ordered and the
measurements will be repeated.

Since the presence of the borated
water in the unit shield mockup altered
the power distribution i1n the reactor,
it was necessary to redetermine the
power with neutron-flux measure-
ments. {!? The total power of the
reactor against the borated shield was
only 3% less than the value found
previously with the reactor im water
without the shield. However , the power
generated in the front row of fuel
elements was 16% less with the reactor
against the shield than with no shield.

These new power measurements were
use to calculate(?’ the leakage from

 

(1)g. B. Johnson, Power Calculations of the
l{géi)Shidd Reactor, ORNL CF-51-9-112 (Sept. 18,

(2)]. L. Meem and H. E. Hungerford, Calculations

E; geaka ¢ from the Bulk Shielding Reactor, OBNL

1-10-94 (Oct. 5, 1951).

54

the front face of the reactor into the
shield. The leakage L from a reactor
is glven by

z

L =F P(Z) e %/N 4z

where

F is a factor converting from the
power produced to the appropri-
ate type of radiation dosage
escaping,

Z is the distance into the reactor

from the surface,

P(Z) is the power per unit volume as
a function of Z,

Z, an@ Z, are'appropriate limits of
integration, and

A is the relaxation length of the
escaping radiation.

P(Z) was determined from the above-
mentioned experiment and a value of
A= 9,2 cm. The factor F need not be
evaluated as will be seen later.

The leakage from the Bulk Shielding
Reactor was found to be

L =F x 0,735 watts/cm?

with reactor operating at a power of
10 kw. Taking into account the attenu-
ati1on of a small amount of water and
aluminum between the reactor face and
FOR PERIOD ENDING DECEMBER 10,

the inside surface of the shield, the
leakage from the reactor into the unit
shield was

L, = F x 0,607 watts/cn’

at 10 kw power.

MOCKUP OF THE UNIT SHIELD

The results of the unit shield
experiments are presented in detail
in ORNL-1147.¢?) The experimental
curves for gamma rays, thermal neutrons,
and fast neutrons are shown in Figs.

7.1, 7.2, and 7.3.

In one experiment (Exp. 3) measure-
ments were made starting from the
surface of the last lead layer and
proceeding outward in the water of
the pool along the centerline. In
another experiment (Exp. 4) a tank of
borated water was placed around the
shield as described in an earlier
pProgress report,(q) and measurements
were made ocutward from the wall of the
tank. As the boron concentration was
increased, gamma measurements were
repeated, as shown in Fig. 7.1, until
a concentration of 0.4% was reached.

The weight of the unit shield as
specified by the Shielding Board(®’
was 134,000 1b for a 200-Mw reactor
in the shape of a 3-ft right square
cylinder with a dose of 1 r/hr al-
lowed at the crew, 50 ft away.

The power density of the aircraft
reactor was assumed to be constant,

 

(3}J. L. Meem and H. E. Hungerford, Unit Shield
Exfieriments at the Bulk Shielding Facility,
ORNL-1147 (in preparation)

(4)A. S. Kitzes, “Mock-up of Unit Shield,”
Aircraft Nuclear Propuls;on Prafifct Quarterly
Progress Report for Period Endlng arch 10, 1951,
ANPZ50, p. 158 (June 19, 1951

(siafl?art of the ANP Shteld:ng Board, NEPA-
OFNL, 53 (Oct. 16, 1950).

1951

and accordingly the leakage is calcu-
lated to be

L =

a

F x 3,07 x 10* watts/cm?

with reactor at 200 Mw.

It 1s convenient for the purpose of
these calculations to define a dosage
unit D as follows: The D unit 1s the
maximum dosage rate of nuclear radia-
tions that may be taken by military
persennel during a 25-hr flight of a
nuclear-powered aircraft. Accordingly,

1D =

1 r/hr of gamma radiation or

0.1 rep/hr of fast neutrons

The tolerance dosage of 1D is not to
be exceeded in any combination of fast
neutrons and gamma rays.

If 1D is the tolerance dosage at
the crew compartment, then the allowed
dosage, D_, at the surface of the
shield measured at the Bulk Shielding
Reactor 1s

 

 

where

S is the crew separation = 50 ft,

R 1s the outside radius of shield
required,

L and L  are the leakages defined
above,

. is the radius of the unit shield
reactor = 33 cm, and

rajfi the radius of the aircraft
reactor = 58 cm.

55
93

GAMMA RADIATION INTENSITY, R/HR/WATT

1073

0™

o7

 

BORATED WATER

WITHIN TANK : ALUMINUM TANK
EXP 4 ONLY

( ) (EXP. 4 ONLY)

EAR OF
UNIT SHIELD
MOCK-UP CURVE 4

; GURVES 2,3
WITHOUT ALUMINUM '3
TANK 4,AND 5

EXP.3 EXP 4

50 cc STD 10'2 GHAMBER o .
900 cc 10'0 CHAMBER A

900 ¢c 10'® CHAMBER 0

G-M GOUNTER

10 130 150 170 190 2i0
DISTANGE FROM REAGCTOR, cm

SECRET

DWG. {2817

y —INTENSITIES BEHIND Pb—H,0
UNIT SHIELD MOCK-~UP (EXP. 3)

y —INTENSITIES BEHIND Pb-B- H,0

UNIT SHIELD MOCK-UP WITH VARIOUS
CONCENTRATIONS COF BORON AS
INDICATED (EXP. 4)

230 250 270

Fig. 7.1. Gamma Radiation Intensity for the Unit-Shield Experiments.

 

290

LHOdAY SSAUN0Ud A THILYVAD LDAfoUd dNV
10

103

Toka

0—4

THERMAL NEUTRON FLUX, NV, /WATT

 

 

 

 

Fig. 7.2.

 

80

FOR PERIOD ENDING DECEMBER 10, 1951

SECRET
DWG. 12818

REAR OF ALUMINUM TANK

REAR PERIPHERY OF EXP 4 ONLY)
UNIT SHIELD MOCK-UP

BORATED WATER
WITHIN TANK
“(EXP 4 ONLY) -

~ THERMAL NEUTRON FLUX-—1"~
““BEHIND Pb- Ho0 UNIT SHIELD

 

THERMAL NEUTRON FLUX -

"BEHINDG Pb-0.4% B-Hp,0 -

UNIT SHIELD MOCK-UP~~
(EXE 4)

2N, FISSION COUNTER

INDIUM FOIL. MEASUREMENTS

8 IN. BF; COUNTER

2 IN. BF, SINGLE CHAMBER COUNTER
2 IN. BF, DOUBLE CHAMBER COUNTER

100 120 140 160 180 200 220 240 260
DISTANCE FROM REACTOR, cm

Thermal -Neutron Flux for the Unit-Shield Experiments.

57
ANP PROJECT QUARTERLY PROGRESS REPORT

58

FAST NEUTRON DOSAGE, MILLIREP/HR/WATT

10

S,

5]
W

g,
A

10

SECRET

DWG. 12816

 

|

L L]

 

__REAR PERIPHERY OF
UNIT SHIELD MOGK-UP

/REAR OF ALUMINUM
TANK (EXP. 4 ONLY)

]

 

 

 

 

=
)

 

! T
—~ BORATED WATE

 

 

 

WITHIN TANK

 

 

TT(EXP. 4 ONLY)

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

o~ DATA TAKEN BEHIND Pb-H,0 B e e
— UNIT SHIELD MOCK~UP (EXP 3) T N

DATA OBTAINED BEHIND Pb-04%B-
HoO UNIT SHIELD MOCK-UP (EXP. 4)

 

 

 

 

 

\O\ e |
NPT FAST NEUTRON oo -
N#“ DOSAGE BEHIND ~F— S -
N UNIT SHIELD (EXP 3) ~ffo oo - ]
p\%\ & """""""""""
N
|5y
""""""""" ' *\\* T = ]
\\ : Tl . ]
"""""""" N B
_______ e, - O\_O_
TN .
o

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

I . e -
\O

u_;fikf ''''''''' i o Frrie ] \E\_

I T FAST NEUTRON DOSAGE .. 1 e

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

+— —UNIT SHIELD (EXP. 4)
|

 

 

 

 

 

 

 

 

 

 

T

 

- BEHIND 0.4 9 BORATEVV";A\_W.-

 

 

 

 

80

Fig.

7.

3.

100

120 140 160

DISTANCE FROM REAGTOR, cm

180

Fast-Neutron Dosage for the Unit-Shield Experiments.
FOR PERIOD ENDING DECEMBER 10, 1951

For these calculations the 3-ft square
cylindrical aircraft reactor isclosely
approximated by a 3.8-ft sphere. The
significance of each term in this

expression 1s:

(S/R)? is the
attenuation due to reactor-
crew separation,

inverse-square-law

Lu/La is the ratio of leakages of
the two reactors, and

ru/ra is a factor which represents

the geometrical attenuation

in a shield with spherical

surfaces.

Using this formula and the data of
Fags, 7.1, 7.2, and 7.3, the shield
thickness was determined and the weight
of the aircraft shield was found to be
129,200 1b. Since the insertion of a
layer of lead in water has anegligible
effect on the neutron attenuation,; and
since the effective relaxation length
for gamma rays in lead at this position
in the shield has been established as
3 cmat the Lid Tank, 1t 1s interesting
to calculate the effect of adding or
substracting lead in the last layers of
the shield. This 1s done by holding
the total neutvon plus gamma dosage
constant but varying the fraction of
the dosage taken in either neutrons or
gamma rays. When lead is subtracted,
the apportioned gamma dosage goes up
and a thicker outside layer ol water
muist be added to bring the apportioned

neutron dosage down, and vice versa.

The results of these calculations
are shown inFig. 7.4, where the shield
weight 1s plotted against both the
shield thickness and the outside radius
of the shield around the Bulk Shielding
Reactor. The percentage of the dosage
taken in neutrons is also shown. In
Shield No. 1, the shield
actually measured as weighiang 129,200
1 (mentioned above), 44% of the dose
is taken 1n neutrons and the rest in

which was

gammas. For ease of construction of
the mockup all lead layers were made
with l-in. thicknesses. Actually, the
shield specified by the Shielding
Board was only 0.4 in. thick in the
last lead layer. If this excess lead
is peeled off and water added, the
weight goes down to 127,600 1b, as shown
under Shield No. 2 in Fig. 7.4. This
is the minimum weight of shield which
is possible with the present lead
spacing and boron concentration. This
shield is identical with that predicted
by the Shielding Beoard except it has
6 cm less water. The agreement in
weight 1s within 5%. All the combined
errors in measurement and calculation
from the bulk shielding data do not
add up to an error of 5% in shield
weight. It 1s estimated that these
combined errors would amount to less
than a ton.

The important result of these
exper iments 1s not that an aircraft
shield will weigh 127,600 1b, since
this is an ideal shield. The necessary
engineering will undoubtedly increase
1ts weight, while current Lid Tank
experiments iadicate that the addition
of more boron to the water with a
subsequent reoptimization of the lead
spacing may decrease the weight by as
much as 10%. The real significance of
these tests 1s that the methods of
calculation used by the Shielding
Board have been confirmed. Given the
proper specifications about the reactor
and airplane, it is probable that the
weight of an engineered unit shield
can be calculated from existing data
to within a few tons.

MOCKUP OF THE DIVIDED SHIELD

The divided-shield mockup has been
installed in the pnol and measurements
are now underway. A photograph of the
installation with the reactor in place

2%
ANP PROJECT QUARTERLY PROGRESS REPORT

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

DWG 12962
THICKNESS OF UNIT SHIELD, cm SECRET
120 124 128 132 136 140 144 148 152 156 160 164 168 172
‘ MINIMUM THEORETICAL SHIELD THICKNESS
f.—{NEUTRON DOSE =100% TOTAL DOSE)
{ } REMOVE Pb FRGM NEXT
l ADD Pb TO | INNER Pb LAYER
tOUTSIDE Pb LAYER . - | (LAYER 17) - 100
N 57— | REMOVE Pb FROM OUTSIDE f’ W
]\ i Pb LAYER (LAYER 19) SHIELD WEIGHT —el /. 90 Q
I VARIATION WITH DISTANGE 3
FOR 1 R/hr TOTAL DOSE =
2\ /N S
148 i l \, {USE SCALE TO LEFTj//// 80 o
AN 70 &
l \ N //7 o
144 1 p 4 60 4
\ N | NEUTRON DOSE L/ g
| N 44% OF TOTAL DOSE A 50 W
o \ 5| FoR SHIELD NO. 1 7 &
LJ
% l _____ . :‘\ . J l A1 /1 e 40 <
v ~ w
NEUTRON DOSE
0
z ~| 25% OF TOTAL DOSE o ;7// — 30 g
1 \\ ‘\(%Ef/’ FOR SHIELD NO. 2 V4 S
o
§es YMFE\ L 20 3
= : : THE PREDIGTED SHIELD — = 0 =
z l (SAME AMOUNT OF Pb ‘; e L b L] W
2132 -eh——1  AS SHIELD NO. 2) ] 4 _ Tl g =
z | J WT.= '34’300 ib PERCENT OF TOTAL NEUTRON DOSE
3 1 + B t 1= VARIATION WITH DISTANGE :
= SHIELD NO. | .y | [ {USE SCALE TO RIGHT)
$ 128 - (ME ASURED) etk ] }-- —
{ R/hr TOLERAMNCE /’/{-
| wr=t29200 | t—7T 1 1 | } 4+ | 1 | 4+ i ot b ] ]
<"SHIELD NO.2 (OPTIMIZED)
| { R/hr TOLERANGE
124 T WT. =127,700 1b
[t
|
120 l
178 182 186 190 194 198 202 206 210 214 218 222 226 230

OUTER RADIUS OF SHIELD, cm

Fig. 7.4. Dependence of the Weight o

is shown in Fig. 7..5. A report on the
gamma-ray spectrometer being used for
the gamma-ray measurements 1s being
prepared.(fi)

In the last quarterly report{’? an
exper iment was described in which the
gamma radliations from the reactor were
used as a source to observe the effec-
tiveness of a lead slab as a shadow
shield. The objective was to obtain
information which would be of help in

(G)F. C. Maienschein, Multiple Crystal Gamma-
Ray Spectrometer, ORNL-1142 (in preparation).

(7)epy1k Shielding Facility,” Aircraft Nuclear
Propulsion Project Quarterly Progress Report for
Period Ending September 10, 1951, ORNL-1154,
Fig. 5.2, p. 86 (Dec. 17, 1951).

60

f the Unit Shield on its Thickness.

designing the lead shadow disk within
the hydrogeneous reactor shield. This
problem is common to many divided
shield designs. The reactor plus the
capture gamma rays 1t produces 1in the
water proved to be too diffuse a source
for definitive conclusions. Conse-
quently, the experiment has been
repeated(8> during this quarter using
a can of radioactive sodium which 1s
small enough to simulate a point
source. The layout of the experiment
and the shadow patterns observed behind
the lead slabs are shown in Fig. T7.6.

(S)H. E. Hungerford, Experiment V-A at the
Bulk Shielding Facility — Shadow Shield Measure-

ments with aNae>? Source, ORNL CF-51-11-95 (Nov.
15, 1951).

?
FOR PERIOD ENDING DECEMBER 10, 1951

 

Fig. 7.95. Installation of Divided-Shield Mockup with Reactor in Position.

61
ANP PROJECT QUARTERLY PROGRESS REPORT

 

 

 

 

SECRET
DWE. 1298

T T 1717 1 1 1 1 1 117 17T 1771117 11 1T 71T
1073

@
@

60 & 10—4._-
@

I
1

3
|
|

10-2 —

&

&
l

 

 

8
N
5
|

@
25
@
20 p
15 >

 

HORIZONTAL OISTANCE FROM REAR OF SHADOW SHIELD, cm
w
o

3

 

o
5 J/‘ /4/ |0-3 —
7 ®
0 —
Pb SLAB Iwz in. J

2in, / A PICTOGRAPHIC VIEW OF GAMMA RADIATION INTEN—
SITIES EXISTING ALONG VERTICAL TRAVERSES

 

 

 

 

 

 

 

 

 

 

 
  
 

 

H
Pb SLAB xl,/z n. MEASURED BEHIND THE SHADOW SHIELD
GURVE 1 VERTICAL TRAVERSE 5cm FROM SHIELD
CURVE 2 VERTICAL TRAVERSE 15cm FROM SHIELD
CURVE 3 VERTICAL TRAVERSE 35cm FROM SHIELD
CURVE 4 VERTIGAL TRAVERSE 60cm FROM SHIELD
CURVE 4 CURVE 4 TO SAME SCALE BASE AS GURVE 3
SGALES TO RIGHT OF EACH GURVE GIVE RADIATION
N INTENSITIES IN r7hr
L -I—’UP LINE A LINE CONNEGTING POINTS OF MINIMUM
A 5 INTENSITY
P c LINE B LINE CONNEGTING POINTS OF MAXIMUM
z F——=< 60 cm INTENSITY
a3 yay 4,1 LINE C SHADOW LINE OF SHIELD FROM GENTER
Qs /. OF Na SOURCE
15¢cm
<[ E &9 2
cl o 5cm
e
e m
v SCALE DIAGRAM OF SHADCW
23 SHIELD EXPERIMENT WITH
2 SODIUM SOURGE
B SHOWING WHERE TRAVERSES WERE
MADE ANG THE RELATION OF LINES

 

A,B, AND G TO SOURCE

 

 

 

 

SOURGE
@ /et e bl
O 20 40 60 8¢ 100 120 140 160 180 200 220 240

VERTICAL DISTANCE FROM GENTERLINE, ¢m
+*

Fig. 7.6. Shadow-Shield Experiment with Sodium Source.

62
FAST FLUX {A/sec/cm@/watt )

10 5 LTI

boo—e

 

   

FOR PERIOD ENDING DECEMBER 10, 1951

SECRET
OWG. NO. 13194

 

T T T T

.......

 

 

 

B UP® FAST FISSION GHAMBER, NG BORON OR e -

CADMIUM LINER, NO Pt SHIELD "
,,,,,,, S~ —
= e U FAST FISSION CHAMBER WITH BORON-CADMIUM T =
S LINER, 3 INGHES OF Pb 10 cm IN FRONT OF GOUNTER N
T @ DOSIMETER —
| A SULFUR THRESHOLD DETECTOR
— | ]
e b

20 40 60 80 100 120 140 160 180 200

DISTANGE FROM REACTOR {cm)
Fig. 7.7. Bulk Shielding Facility Fast-Neutron Data.

63
ANP PROJECT QUARTERLY PROGRESS REPORT

. 9
Measurements are now available! )

on the fast-neutron flux from the
Bulk Shielding Reactor using the
dosimeter, a U?3% fission chamber,
and sulfur threshold detectors. The

(®JH. E. Hungerford and R. G. Cochran, Fast
Neutron Mecsurements at the Bulk Shielding Facil-
ity, ORNL CF.51-11-96 (Dec. 10, 1951).

64

results of the three types of measure-
ment are shown in Fig. 7.7. "Fast-
neutron flux" is in itself a nebulous
quantity and 1s defined differently
for each of the three detectors used.
The agreement obtained was nevertheless,
quite gratifying and gives added
confidence to all previous results
using the fast-neutron dosimeter.
FOR PERIOD ENDING DECEMBER 19,

1951

8. DUCT TESTS

C. E. Clifford, Physics Division

Following the decision of General
Flectric to build an air-cooled
reactor, 1t was decided that the Lid
Tank could do some useful experiments
to aid in the design of the air ducts
which penetrate the water-reactor
shield. As these experiments pro-
ceeded, the results indicated that the
problem of penetrating a water shield
with ducts, even though only filled
with air, was not so difficult as had
been previously supposed. Once a
theoretical evaluation of the experi-
ments has been made, the prediction of
the transmission through sodium-filled
only minor

ducts would i1ntroduce

complications.

AIR-FILLED DUCT TEST IN LID TANK

A. Simon J. D. Flynn
T. V. Blosser

Physics Division

Two types of air ducting are under
consideration: (1) small round pipes
with various bends, and (2) annular
ducts with various bends. In the
annular ducts the flow is confined to
the region between two large cylinders.
Bends are effected by displacing the
center of the ducts i1in fractions of
1ts diameter with both ends remaining
approximately fixed.

Cylindrical Ducts. Considerable
work was done on the round pipes since
they were believed to be the most
favorable arrangement, The size of
the ducts investigated was chosen on
the basisof pressure-drop calculations
made by the G.E. group. This indi-
cated that the pressure drop due to an
array of approximately 200 small
circular ducts on the order of 2 to 3
in. in diameter would not be excessive.
Design studies by the G.E. ANP group
have indicated that the ducting system

1s not radically changed i1n contem-
plated supersonic applications,

The measurements were concentrated
on the 1nlet air ducts, since these
ducts would face the crew compartment
and therefore would probably have the
greatest effect on the shield weight
due to the larger attenuation required
in that direction. After some experi-
mentation it became apparent that an
array of 2-1n. ducts on a 3.7-1n.
triangular lattice, four of which are
shown in Fig. 8.1, could be used with-
put seriously increasing the radiation
transmitted by the shield. This array
was built up of 2-1in.-1.d. electrical
conduit, the walls of which were
1/16-in. steel. Standard 90° elbows
were used which were bent onm a 10-1in.
radius. Each elbow had two 5-1n.
straight sections, so that when two
were joined by means of a standard
coupling, a 10-in. straight section
resulted. The assembled elbows were
held in place by wooden spacers and the
whole array was mounted in the Lid Tank
on a plywood table.

Neutron Transmission. The effect
of this duct array on the transmitted
radiation is shown in Fig. 8.2. This
shows the thermal-neutron-flux distri-
bution in the water beyond an array of
15 ducts, three rows high and five
rows wide. It was possible to measure
only thermal neutrons in these tests
because of time limitations and the
necessity of taking a large number of
data. However, 1t is felt that these
thermal measurements can be readily
interpreted in terms of fast-neutron
dose by correlations obtained from
measurements made 1in water of both
thermal -neutron and fast-neutron dose
under the same conditions. All that
is required is that at some point 1in
the thermal neutron measurements the

65
ANP PROJECT QUARTERLY PROGRESS REPORT

8 SOURCE PLATE

‘ SECRET |
B FHOTO 11054

_284n.mA.

VR e chi 3 2
2in-1D STEEL CONDUIT

| (O.4-in. WALLS)

g

Fig. 8.1. An Array of Four 2-in.

in excess of 6 cm after
correction for geometrical attenuation.
Measurements were taken both with two
elbows joined and with three elbows
joined. The two-elbow array penetrated
approximately 40 in. of the shield and
the three-elbow array penetrated

60 in. of the shield.

slope be

An integration of the vertical
traverse for the two-bend case should

66

 

3 90° BENDS

Steel Conduit Ducts with Three 90° Bends,

give the transmission of an infinite
array. The measured dose agrees
within 20% with the dose calculated by
taking account of only the reduced
density due to the voids. The infinite
array 1s the only geometry in which
the true reduced density effect can be
calculated. For this duct size 1n
these bends no correction 1s necessary
for streaming, a very encouraging
result. As a consequence the effect
FOR PERIOD ENDING DECEMBER 10, 1951

SECRET
DWG. 13013 R1

 

21, ,=9.224 x 104 FOR 7=113.8 cm

O-o0

S
o

O
o

O
o

/

x TRAVERSES MADE AT z=#{3.8cm
o 2 90° ELBOWS
® 3 90° ELBOWS

z CENTER LINE MEASUREMENTS AT y=0
® (¢ IN WATER

© ¢ IN WATER BEHIND DUCTS WITH
2 90° ELBOWS

/

RELATIVE THERMAL NEUTRON FLUX, counts/min (121/2'in. BF; COUNTER)
o
Mo

¢ OF DUCT ARRAY WITH
\ RESPECT TO «x

50 60 80 100 120 140 160 180
Z OR x (cm)

10

Fig. 8.2. Lid Tank Duct Test D-915, X Traverse (Vertical) and Z
Centerline Measurements of Neutron Flux for Two Arrays of 2-in. Steel
Conduit.

67
ANP PROJECT QUARTERLY PROGRESS REPORT

RELATIVE THERMAL NEUTRON FLUX, counts/min (42‘/2-in. BF3 TWIN COUNTER, z=153.8 cm)

-100 -80 -60 —-40 -20

 

SECRET
DWG. {3016R1

STRAIGHT
TWO 30°
TWO 45°
TWO 60°
TWO 90°

DUCT CROSS SECTION

20 40 60 BO {00 120

yi{cm)

Fig. 8.3.
Conduit (2 in.

Lid Tank Duct Test D-10,
i.d.) with Two Bends of

of the ducts on the shield weight is
very small — of the order of 1 to 2
tons — since the only additional
shielding required is a conical patch
along the periphery of the ducting
system.

Transmission with Variable Bends.
From the point of view of pressure
drop it was desirable to determine the
minimum amount of bending required in
the duct to limit transmission to the

68

Y Traverses in H20 behind 54-in. Rubber
VYariable Radius.

simple reduced density effect. To
this, experiments were

in the Lid Tank with
tubing of 2 i.d.
o.d.

determine

undertaken
flexible
2-3/8 1in.

in. and

Figure 8.3 gives the results obtained
by introducing two bends of various
degrees in a 54-in. section of rubber
conduit. The data indicate that even
two bends of only 30° are more than
sufficient to give a geometrical
FOR PERIOD ENDING DECEMBER 10, 1951

133.8¢cm)

RELATIVE THERMAL NEUTRON FLUX, counis/min (1244 in. BF; TWIN COUNTER, ~

-80 -60 - 40 -20

Fig.
52-1in.

8§.4. Lid Tank PDuct Test D-11,
Rubber Conduit (2-3/8 in, i.d.)

attenuation down the duct that 1s
larger than the attenuation of the
surrounding shielding material (water).

Since large duct sizes and fewer
bends reduce the pressure drop, the
next experiment was done ona 2-3/8-1in.
rubber conduit 52 1in. long with only
one bend. Results of this experiment
are presented 1n Fig. 8.4. The experi-
ment was conducted by measuring first
the transmission of the straight duct
in the water beyond the end of the

 

SEGRET
DWG. 13015 Rt

STRAIGHT DUCT
136 in. BEND
23/g in. BEND
33, in. BEND
7 Y% in. BEND
10" in. BEND
14, in. BEND

o 20 40 60 80 100

Y Traverses {(Horizeontal) in #,0 behind

with Variable Bends.

duct, and then by measuring the trans-
mission as the center of the duct was
displaced by multiples of 1ts diameter
from the centerline of the straight
duct. As the center of the duct was
displaced, the transmitted intensity
was reduced. The first measurement,
in which the center was displaced by
one-half the diameter, gave a reduction
of approximately a factor of 2, cor-
responding closely to the reduction 1in
the area of the source which could be
seen from the center of the other end

69
ANP PROJECT QUARTERLY PROGRESS REPORT

SOURCE PLATE

7’6 cm

 

 

{6 cm

—]

 

 

      

SECRET

 

 

 

 

 

 

 

 

DWG. 13021 Ry
AIRPL ANE .
V REACTOR ///A gom —=| -
///////// i
HEADER ,_,/»—x\
CROSS SECTION OF 61em
CYLINDRICAL SHELl——=
f 105.6¢cm
/
6icm
0.5 ¢cm - '
457
glem

x) ‘]L;~-4?.8 cm—s

 

 

 

T~

 

 

 

| 12 5 crn

CURVATURE NOT IN
MOCK-UP AS TESTED

 

 

2’yx160m

Fig. 8.5. Lid Tank Duct-Test Annular-Duct Rectangular Cross-
Section, 1/8-in. Steel Plate Welded.

70
FOR PERIOD ENDING DECEMBER 1¢, 1951

SEGRET
DWG, 13012R1

Zz TRAVERSES
y =0
¥y =623
¥y =90
¥ =100

y TRAVERSES
zZ =22.4
zZ=1470.4

FAST NEUTRON DOSAGE BY
DOSIMETER MEASUREMENTS -
z=1{75, y=+16
8.666 x 1072 mrep/hr
SOURCE PLATE (NOT PLOTTED)

RELATIVE THERMAL NEUTRON FLUX, counts/min {8-in. BFz; COUNTER])

 

-0 O 20 40 60 80 100 {20
CENTER LINE DISTANCE (cm)

Fig. 8.6. Lid Tank Test P-12, Y (Horizontal) and Z Traverses Behind
Aonular PDuct,

71
ANP PROJECT QUARTERLY PROGRESS REPORT

of the duct. When the center was dis-
placed onediameter, entirely eliminat-
ing a line of sight transmission down
the duct, the largest decrease was
noted, approximately a factor of 40.
Further displacement of the center
gave smaller decreases until, finally,
beyond a 30° bend no further decrease
was noted. Calculations for this
bend indicated that the intensity was
about what would be expected for the
reduced density shield.

Theoretical Duct Attenuation. On
the basis of these dataa simple theory
was derived which would predict the
geometrical attenuation of the duct to
within 20 to 40% for the various
configurations tested. The theory is
based on the picture of virtual sources
created at the bends of the duct with
the neutrons emerging with an isotropic
distribution from the walls. Since
these sources are small, the geometri-
cal attenuation is Jarge for duct
lengths that are many times the diameter
of the duct (in comparison to the
exponential attenuation of the sur-
rounding water). A fuller account of
this theory will be reported after
further experimental correlation.

Annular Ducts. One quick experi-
ment was done on a mockup fabricated
by G.E. to simulate a segment of an
annular duct (Fig. 8.5). The measure-
ments taken on this mockup are pre-
sented in Fig. 8.6. These data have
not been subject to analysis as yet,

12

but seem to indicate that theattenu-
ation of high-energy neutrons {(which
contribute more effectively to the
dose) is greater than that of the
lower energy neutrons. An analysis by
G.E. of these data indicates that such
a duct could be used, although it
results in a sizable increase in the
shield weight, The principal reason
for using this type of duct is to in-
crease the accessibilityof the reactor
faces for control mechanisms.

Liquid-Metal Duct Test in Thermal
Column (F. J. Muckenthaler and M. K.
Hullings, Physics Division). With the
thermal column facility (previously
referred to as the "Duct Test" or

" "Duct Test Facility") the measurements

carried out over the last year, on the
6- and 8-in. liguid-metal coolant
ducts supplied by KAPL for all the
interesting combinations, have been
completed. The analysis of these data
is in process and a complete report
will be written. The experimental
program in this facility is now ailmed
at obtaining further experimental cor-
relation with the simplified theory of
air ducts. Neutron transmission will
be measured for a set of small,
flexible, air-filled ducts 1n water,
in which the length, the diameter, the
degree of bending, and the source size
will be varied. Preliminary measure-
ments have been completed on a 3-1n.
duct 60 in. long for five bends from

0 to 90°.
FOR PERIOD ENDING DECEMBER 10,

9. SHIELDING

E. P. Blizard,

Present shielding facilities do
not permit a divided-shield mockup
which incorporates the effects of
crew-compartment shielding and air
scattering on the radiations emerging
from the reactor shield. Recause of
the importance of this effect on air-
craft shield weights, a proposal has
been submitted to the AEC for a new
shielding facility, called the "Tower
Shielding Facility," with which air
scattering and a crew shield mockup
can be effected. This facility could
be constructed at an estimated cost of

about $694,650.

Shielding weights and thicknesses
for circulating fuel reactors have
been estimated for application to
configurations as they are developed,
In general, these shields are pertur-
bations of the divided shield for the
sodium-cooled reactor. Divided-shield
design studies are continuing at NDA
in their effort for optimization of
this shield weight.

TOWER SHIELDING FACILITY PROPOSAL

E. P. Blizard H. L. F. Enlund
C. E. Clifford J. L. Meenm
A. Simon

Physics Division

It was recognized in the Shielding
Board Report{!) that the calculations
leading to divided-shield weight were
not on so firm a ground as those
leading to the weight of unit shields.
Unit shields have subsequently been
mocked-up in both the Lid Tank and the
Bulk Shielding Facility, and the

 

(1)Report of the ANP Shielding Board, NEPA-
ORNL, ANP-53 (Oct. 16, 1950).

1951

INVESTIGATIONS

Physics Division

resultant weights were in excellent
agreement with those predicted by the
Shielding Board. A divided-shield 1is
now being mocked-up in the PBulk
Shielding Facility. However, the
information to be obtained from this
mockup 1s only the intensity and
direction of the radiations emerging
from the reactor part of the divided
shield. One of the larger uncertainties
in divided-shield weights 1s the
assumed scattering of the complex
spectra emerging from the reactor
shield and the effect of thais radiation
at the crew shield after having been
air-scattered.

Since an increase of 1 1b in shield
weight entails an increase of 2% 1b in
aircraft weight,{(2) considerable
effort is justified in specifying
shield weights as closely as possible.
Full-scale shielding experiments,
including air-scattering and crew-
shield effects,
guickest,
optimizing the divided shield.
comparison,

offer the cheapest,
and most accurate method of
(By
an aircraft design based
on calculations would be appreciably
heavier because of the reguired
conservatism.) Consequently, a
proposal(3) for a Tower Shielding
Facility to effect full-scale shield-
ing experiments has been submitted to
the AEC by the Oak Ridge National
Laboratory.

The proposal 1s to mount a small
MTR type water-cooled reactor and
shield mockup in a tower with a crew
shield mockup supported 50 ft above

the reactor. The tower would be tall

 

(2)Report of the Technical Advisory Board to
the Technical Committee of the Aircraft Nuclear
Propulsion Programr, ANP-52 (Aug. 4, 1950).

(3)E. p. Blizard, C. E. Clifferd, and A. Simon,
Proposal for Divided Shield Expertiments, ORNL
CF-51-11-158 (no date).

73
ANP PROJECT QUARTERLY PROGRESS REPORT

enough so that ground scattering would
be unimportant, A preliminary cost
estimate indicates that the equipment,
including tower,
security fence,

about $694,650.

reactor,
and roads,

building,
would cost

CIRCULATING-FUEL REACTOR SHIELDS

E. P. Blizard and F. H. Murray
Physics Division

The current designs of circulating-
fuel reactors present new problems in
divided-shield design. These are being
attacked on a cursory basis i1n order
to approximate shield weights for the
power-plant configurations as they are
developed.

A preliminary estimate of the
required shielding thickness has been
derived for the Reactor Design Group.*)
Tentative conclusions based upon the
layout of the preliminary design show
that the crew shield will require
about 13 in. (see footnote 5) of
plastic on all sides (somewhat less
on the front) for shielding of delayed

 

Mg, p. Blizard, Preliminary Estimate of CFR
Shielding, ORNL CF-51-11-139 (Nov. 26, 1951).

(S)Subsequent more detailed calculations
indicate this dimension to be closer to 18 in.

714

neutrons, There appears to be no
reason that some of this shielding
could not be removed from the reactor
shield, but some other modification
may be reqguired, for example, the
location of gamma shielding around the
crew,

No calculations have been made
concerning the gamma shielding, but
for the present it seems safe to assume
that the exposed fuel will not consti-
tute a source much worse than some
cases already considered, namely, the
sodium coolant discussed in ANP-53,(1!)
It 1is also apparent that the divided
shield for the circulating-fuel reactor
is not fundamentally different from,
but merely a perturbation of, the
shield design for the sodium-cooled
reactor.

NDA DBIVIDED-SHIELD STUDIES

Nuclear Development Associates,
Inc., are continuing their analysis of
divided-shield design. The ANP-53

appear to be adequate 1n
although the details of design
The final

include

designs
weight,
will undoubtedly change.
report of their work will
specific recommendations for the full-
scale divided-shield experiments now
contemplated by ORNL.
 
 
SUMMARY AND

Recent interest in the high tempera-
ture circulating-fuel reactor has
emphasized the need for liquid fuels
or low (below 450°C) melting point with
relatively low (4 to 8 1b/ft®) uranium
concentration. Five fluoride systems
fulfilling these requirements have
been studied. Research on the chemistry
of these and other high-temperature
liquids, such as, the development of
liquid moderators, self-moderating
fuels, and the investigation of fluoride
mixtures as possible reactor coolants,
are reported im Sec. 10. Several
ternary systems of alkali fluorides
and beryllium fluoride are shown to be
suitable coolants and are also of
potential value as solvents for UF,
in the development of fuels.,

Static corrosion tests of fluoride
fuels i1indicate that these fluids are
readily contained at 1500°F by inconel
and any one of a number of stainless
steels (Sec. 11). Dynamic corrosion
tests in inconel have been similarly
successful, but the two stainless steel
loops in which the fluorides have been
circulated have plugged, presumably
as a result of mass transfer. The
static and dynamic corrosion of these
fluoride mixtures, which are of primary
interest to the circulating-fuel
reactor, are, except for the dynamic
test 1n stainless steel loops, reason-
ably satisfactory. Considerable
research has been undertaken in attempts
to contain the alkali hydroxides in
nickel and the structural metals.
Static-corrosion tests of hydroxides
have shown, however, that only nickel,
copper, and the more noble metals will
withstand the corrosive action of
these media. Furthermore, dynamic-
corrosion tests of hydroxides in
nickel have indicated severe plugging,
also, as with the fluoride in stainless
steel, the result of mass transfer.
Fundamental studies of this phenomenon

INTRODUCTION

indicate that oxygen may play an

important role in mass transfer, at
least in hydroxide-nickel systems.
Heat-transfer research (Sec. 12)

has been concerned primarily with the
determination of heat-transfer coeffi-
cients of various systems, including
those of the fluorides, hydroxides,
and liquid metals. Although satis-
factory data have not yet been obtained,
mathematical solutions have been
derived for the temperature structure
in forced-convection fluoride systems.
Excellent agreement has been obtained
between theory and practice for the
natural convection i1n a brine solution
simulating the fuel elements; the peak
temperature 1s substantially reduced
from that computed assuming conduction
only. HRoutine measurements are now
being made on the heat capacity of
materials and the density of high-
temperature fluids. Some data are also
available on the thermal conductivity
of materials and the vapor pressure
and viscosity of the fluoride salts.

The metallurgical processes involved
in the construction and assembly of a
high-temperature core, including
welding of fuel tubes, fabrication of
solid fuel elements, and creep and
stress rupture of metals, are currently
under investigation {Sec. 13). The
fabrication of solid-fuel elements
has been further refined by the evalu-
ation of the effects of the particle
size of the UQ, powder, rolling temper-
ature, and elimination of the capsule
during hot rolling. Satisfactory spot
welding and brazing of these fuel-
plate laminates has also been demon-

strated. The creep and stress~rupture
testing of reactor fuel tubes in
sodium has been initiated. Stainless

steel tubes have withstood hoop stress
up to 2600 psi for 1000 hr at 1500°F
in sodium, whereas similarly stressed
inconel tubes failed.

[
ANP PROJECT QUARTERLY PROGRESS REPORT

Although no pronounced radiation
damage effects have yet been observed
in irradiation of the constituents of
the aircraft reactor in the X-10
graphite pile or the Y-12 cyclotron,
the one irradiation of a fuel capsule
in the higher-flux LITR showed an
increase of corrosion products in the
fuel, an increase in the decomposition
of the fuel, and an 1ncrease 1n

78

corrosion of the capsule (Sec. 14).
It is not yet possible to draw any
conclusions regarding the significance
of radiation damage to the aircraft
reactor until confirmatory experiments
are completed. A second irradiation
of inconel 1in the X-10 pile at temper-
atures up to 575°C has not confirmed
the decrease in thermal conductivity
reported earlier.
FOR PERIOD ENDING DECEMBER 190,

10.

Warren Grimes,

Research in the ANP Chemistry Group
has been concerned almost entirely
with studies of high-temperature
liguids for use as fuels, moderators,
and/or heat-transfer fluids for an
aircraft reactor. The general prop-
erties required of such liquids along
with progress 1in development of such
materials have been described 1n
previous reports (see p. iv for list).
The research effort i1s at present
concerned with fuel and woderator
development with somewhat less emphasis
on the heat-transfer fluid (coolant)
development program.

Hecent interest in a high-tempera-
ture circulating-fuel reactor has
emphasized the need for liquids of
extremely low melting point with
relatively low uranium concentration
(4 to 8 1b of uranium per cubic foot),
Five systems fulfilling these require-
ments have been studied, but the
program for the development of self-
moderating fuels has not produced
materials more satisfactory than those
previously described.!) Several
moderator-cooclant systems have been
examined, but this program has not
been vigorously pursued pending the
demonstration, 1n design studies and
corrosion research, of the appli-
cability of these systems. Investi-
gation of fluoride mixtures as possible
reactor coolants have shown several
ternary systems of the various alkal1
fluorides and beryllium fluoride to
be suitable.

 

(Dy, p. Blakely, G. J. Nessle, L. Bratcher,
and C. J. Barten, “Phase Studies of Fluoride
Systems,” Aircraft Nuclear Propulsion Project
guarterly Progress Report g:r Period Ending

une 10, 1951, ANP-65, p. 84 (Sept. 13, 1951).

1351

CHEMISTRY OF HIGH-TEMPERATURE LIQUIDS

Materials Chemistry Division

LOW MELTING-FLUORIDE FUEL SYSTEMS

J. P. Blakely L. M,
C. J. Barton
Materials Chemistry Division

Bratcher

Phase equilibrium studies utilizing
the technique of thermal analysis(?+3)
were directed initially toward develop-
ment of fluoride mixtures with high
uranium content and melting polnts
below 550°C. A previous report(*’
listed nine systems which promised to
fit these gqualifications. However,
only three of these systems show low
melting points in the range 0 to 2.5
mole % UF, and promise to be of use
in circulating fuel reactors. This
relative scarcity of suitable fuel
mixtures along with the desirability
of lower melting liquids for such
reactors has prompted additional study
of systems containing low concentrations
of UF,. In addition 1t has been
necessary to re-examine 1n greater
detail the low-uranium regions of some
of the known systems.

At present the five systems listed
in Table 10.1 are those which appear
promising as circulating fuels. The
examination of several of these listed
systems is far from complete, and 1t

 

(2)¢c, J. Barton, R. E. Moore, J. P. Blakely,
and G. J. Nessle, “Low-Melting Fluoride Systems ~--
Thermal Analysis,” Aircraft Nuclear Propulsion
Project Quarterly Progress Report for Period
{ggé?g August 31, 1950, OBNL-858, p. 110 (Dec. 4,

(3)R. E. Moore, G. J. Nessle, J. P. Blakely,
and C. J. Barton, "Low-Melting Fluoride Systems,”
Aircraft Nuclear Propulsion Project Quarterly

Progress Report for Period Endinf December 10,
1950, ORNL-919, p. 242 (Feb. 26, [951).

Oy, R Grimes, “Chemistry of High-Temperature
Liquids,” Aircraft Nuclear Propulsion Project
Quarterly Progress Report for Period Ending
fggffuber 10, 1951, OBNL-1154, p. 154 (Dec. 17,

19
ANP PROJECT QUARTERLY PROGRESS REPORT

Table 10.1

Sunmary of Promising Fluoride Fuel Systems of Low Uranium Comntent

 

 

 

 

 

 

 

LOWEST

URANIUM CONC. OF | MELTING

SYSTEM MELTING AT POINT COMPOSITION OF LOWEST

OR BELOW 500°C FOUND MELTING MIXTURE

COMPONENTS (mole % UF,) (°c) (mole %)

LiF-KF-UF4 0 - 71 470 40 LiF, 55 KF, 5 UF4
LiF-NaF’-KF‘UF4 0 -6 450 97 (LiF-NaF-KF), 3 UF4
LiF-NaF-RbF-UF, | 0 - 4 425 97.5 (LiF-NaF-RbF), 2.5 UF,
NaF-BeF,-UF, 0 - 12 330 47 NaF, 50 BeF,, 3 UF,
LiF-NaF-BeF,-UF, 0 - 15 255 30 LiF, 20 NaF, 49 BeF,, 1 UF,

 

is possible that continued study will
result in compositions of lower melting
point. In addition, it should be
emphasized that the search for such
fuels was initiated quite recently
and it 1is likely that the list will be
extended considerably in the future.
As i1t is probable that four-component
systems will be required for melting
points below 350°C, such research will
consequently bhe relatively slow,

Examination of Table 10.1 indicates
that BeF, is a component of the lowest
melting systems. Indeed, 1t seems
certain that BeF, and LiF will be
components of any fluoride system
melting below 350°C.

The various fuel systems studied
are described briefly under appropriate
headings below.

LiF-KF-UF, Re-examination 1in
detail of the low uranium concentration
region of this system indicate that
the diagram previously presented(s)

 

(5)Figure 13.2, ORNL-1154, op. cit., p. 158.

80

1s substantially correct. The lowest
melting point found in this system was
470 + 10°C at 5 mole % UF,, 55 mole %
KF, and 40 mole % LiF while the melting
point at l mole % UF, would be approxi-
mately 485°C.

LiF-NaF-KF-UF, A few mixtures of
the pseudo- blnary system obtained b¥
treating the LiF-NaF-KF eutectic(®:
as the single component have been
studied. The data, shown in Table 10,2,
indicate that the addition of several
mole percent of UF, to this eutectic
does not raise the melting point
appreciably. This system will be of
definite interest and will receive
additional study 1f it appears that
melting points of the order of 450°C
can be tolerated.

LiF-NaF-RbF-UF,. A limited number
of studies in which this system has
been trecated as a pseudo-binary with

 

(8)F, P.Hall and H. Insley, Phase Diagrams for
Ceramists, American Ceramic Society, Inc., 1947,

(M)A, G. Bergman and E. P. Dergunov, “Fusion
Diagram of LiF-KF-NaF,” Compt. rend. acad. sct.
531 753 (1941); Chem. Abstr. 37, 823,
FOR PERIOD ENDING DECEMBER 19,

Table 10.2

The Pseudo-Binary Systen(“’

 

 

 

(LiF-NaF-KF)-UF,
UF, (mole %) MELTING POINT (°C)
0 455
1.1 450
5.0 485
10 597
15 615
20 625
30 630

 

 

(2)42 mole %LiF, 11.5mole % NaF, 46.5mole % KF.

the alkali fluoride ternary eutectic
as one component have been performed.
The data in Table 10.3 show that small
additions (up to 2.5 mole %) UF, do
not appreciably affect the melting
point, but additions of larger amounts
markedly 1ncrease this property.
While the melting points obtained are
slightly lower than those of the
system immediately above, 1t is not
likely that the slight benefit will
Jjustify the use of RbF.

Table 14.3

The Pseudo-Binary System(fl)
(LiF-NaF-RbF) -UF,

 

 

UF, (mole %) MELTING POINT (°C)
0 444
2.5 425(%)
5 615
10 7124
20 660
30 547
40 600

 

 

(242 mole % LiF, 6 mole % NaF, 52 mole ZRbF.

(®)This mixture showed signs of supercooling.

1951

NaF-BeF,-UF,. The equilibrium
diagram for this system has been
presented in a previous report.(®’ A
re-examination of the NaF-BeF, system
(see section on Coolant Development
below) and additional data at low
uranium concentrations has resulted
in some changes in the contour lines
in this portion of the diagram. This
system is still under study and a
revised diagram will be presented
later.

Difficulty in establishing true
melting points at high BeF, concen-
trations was encountered as before as
a result of the extensive supercooling
of such mixtures. The minimum melt-
ing point so far established 1is
330 + 10°C at 3 mole % UF, and 50 mole
% BeF,.

LiF-NaF-BeF,-UF,. Some low-melting
mixtures from the LiF-NaF-BeF, ternary
system (see section on Coolant Develop-
ment below) have been treated as
pseudo-binary systems with UF,. The
most encouraging of these results are
are shown in Table 10.4.

Table 10.4

The Pseudo-Binmary System(“)

 

(LiF-NaF-BeF,) -UF,

 

 

UF, (mole %) MELTING POINT (°C)

0 245-265

1 255

2.5 245-270

5 405

10 490

20 612
30 675

40 720¢¢8)

 

 

(4)30 mole % LiF, 20 mole % NaF, 50 mole %BeF,.

(6) his sample showed definite supercooling.

 

(8)Figure 4.4, ANP-65, op. cit., p. 91l.

81
ANP PROJECT QUARTERLY PROGRESS REPORY

These data show that this system
offers the lowest melting point avail-
able to date. Although some of these

mixtures were stirred at tempevratures

well below 300°C, it is possible that
extensive supercooling occurred. It
1s not possible to give any estimates
of the viscosity at these low tempera-
tures although such studies are being

made.

IONIC SPECIES IN FUSED FLUORIDES

M. T. Robinson

Materials Chemistry Division

Determination of the electrolytic
transference numbers in fused mixtures
of UF, and alkali fluorides has been
attempted in an effort to determine
the nature of the 1onic species 1in
these systems. In advance of the
experiments i1t was thought that the
uranium existed in solution as U*’ or
as some species of negative complex
ion, e.g., UF~ Electrolytic trans-
ference experiments should distinguish
clearly between these two possibilities
but would be less likely to determine,
with certainty, details of the nature
of species differing sllghtly from
those mentioned, e.g., UF*’ or UF,

Experimeantal Procedure. The
electrolytic vessel consists of a
graphite cup acting as the anode 1in
which 1s suspended a graphite rod
cathode. The anolyte and catholyte
are separated by a diaphragm of
platinum sponge. The apparatus 1is
sealed, evacuated, provided with an
inert atmosphere, and maintained at
temperature 1n a molten lead bath.
The electrolytic current provided by
two 6-v storage batteries is maintained
at about 0.3 amp for 90 min, after
which the apparatus is cooled. The
anolyte and catholyte are remcoved
separately, dissolved in HNO,;-A1(NO,;),
solution, and analyzed for U, F~, and
alkali metal ions.

82

Electrolyses have been performed
at both 750 and 900°C on a binary
mixture containing 28 mole % UF, and
72 mole % NaF. From the chemical
analyses before and after the experi-
ment, and from a knowledge of the
quantity of electricity passed, 1t is
possible to derive the transference
numbers of the various i1oms present.

Results of Electrolysis. It
appears that both sodium and uranium
are deposited at the cathode and that
fluorine is evolved at the anode. No
uranium (from anodic formation of UF)
has been detected in the vacuum lines
of the apparatus. A deposit of NaF has
been observed 1in the quartz tube con-
taining the electrolysis cell. This
is probably due to volatilization of
sodium and subsequent reaction with
evolved fluorine. The apparent
deposition of both metals 1s 1n agree-
ment with the reported partial reduction
of UF, by metallic sodium.(?’

The data indicate that no appreciable
quantity of uranium is transferred into
or out of either compartment. If com-
plex ions of the type UF:' are assumed
to be present, only the assumption that
n = 4 yields results i1in which the
transference numbers of all i1ons are

However, the
4 does not mean
that uranium 1s present as un-ionized
UF,, but simply that no detectable
transfer of uranium takes place. This
result is 1n excellent agreement with
the theory of Frenkel('%) of the
electrical conductivity of fused salts.
Briefly, Frenkel states that the
conductivity of a fused salt may be
taken as due only to the smallest 1on
present. The 1onic radii of the Na®,

F~, and U** ions are 0.98, 1.33, and

between 0 and unity.
conclusion that n =

(19)5. Frenkel, Kinetic Theorg of Liquids,
pp. 439-445, Oxford New York

(937, J. Katz and E, Rab1now1tch, The Chemistry
of Uranium, Part 1, N.N.E.S., Div. VIII, Vol. 3,
p. 125, McGraw- Hill, New York, 1951,
FOR PERIOD ENDING DECEMBER 10, 1951

0.97, respectively.(!?) Since the
sodium and uranium(IV) ions are about
the same size, 1t appears that uranium
is complexed with fluoride in some
manner, although the precise nature of
the complex canmot be clucidated from
the presently available data.

It 1s 1intended to extend the work
to another solution of UF, in NaF and
to two solutions of UF, in KF. Since
the radius of the potassium ion 1s
about the same as that of the fluoride
ion, it may be possible to throw more
light on the nature of the uraniferous
ion from measurements on mixtures
containing KF.

HOMOGENEOUS FUELS

J. D. BRedman and L. G. Overholser
Materials Chemistry Division

The solubility of uranium trioxide
in mixtures of sodium hydroxide and
sodium tetraborate with and without
additional boric acid has been previ-
ously reported.‘!?? It was found that
the presence of borate increases the
solubi1lity of uranium 1n sodium
hydroxide but that the concentration
of borate required to give a useful
solubility is so high that the melting
point of the resulting mixture 1is
above 500°C. Additional studies have
been made utilizing other hydroxides
and borates in an attempt to find a
mixture which will dissolve sufficient
uranium and still possess a sufficiently
low melting point. Nelther the
addition of borates nor of any of
several other materials appears likely
to yield a hydroxide mixture melting
below 600°C and containing 2 to 4 wt %
dissolved uranium. All experiments

 

(113, Glasstone, Textbook of Physical Chem-
istry, p- 375, Van Nostrand, New York, 1940.

(12)). p. Bedman and L. G. Overholser,
ORNL~1154, op. cit., p. 16

were performed at 650°C with 5 wt %
uranium added as uranium trioxide.

Uranium Solubility in Hydroxide-
Borate Mixtures. The results of some
of the experiments in which borate-
hydroxide mixtures were used as
solvents are summarized in Table 10.5.
These data 1ndicate that the solubility
of uranium at 650°C is too low to be
of any interest. It does not appear
likely that any combination of hydroxide
and borate will be found having a
melting point below 500°C and dissolving
2 to 4% uranium at 600°C.

Table 18.5

Soelubility of Uranium in
Hydroxide-Borate Mixtures

 

 

 

 

CONSTITUENT IN MIXTURE SOLUBILITY
(we %) OF URANTUM
NaOH KOH .1 OH B203 (we %)
85 15 0.3
70 30 1.3
95 5 0.3
90 10 0,7
85 15 1.6
80 20 1.5
75 25 (a)
99 8 0.3
84 16 0.9
80 20 0.8
48 A7 5(b) 1.0
43 49 15 1.7

 

 

 

 

 

(“)Me]ted above T00°C.

(b)yhen 10 wt %cesium fluoride was added to this
mixture the solubility was found to be 1.1%.

Uranium Solubility in Hydroxides
with Various Additives. Some addi-
tional experiments have been performed
to determine the effect of other
materials on the solubility of uranium
in sodium and lithium hydroxides. The

83
ARP PROJECT QUARTERLY PROGRESS REPORT

results of some of the runs made at
650°C are given in Table 10.6. None
of the materials listed shows an
appreciable beneficial effect on the
solubilivy.

Table 10.6

Effect of Various Additives om the
Solubility of Uranium in Hydroxides

 

 

 

 

 

 

 

 

 

 

CONSTITUENT IN MIXTURE SOLUBLLITY
(we %) OF URANIIM
LiCH | NaOH LiF Na28203 RLOH (wt %)

95 5 0.6
48 47 5 0.3
48 47 5 1.0
45 45 10 1.2
40 40 20 0.8

The addition of BRbOH to mixtures of

sodium and lithium hydroxide appears
to have a slight detrimental effect
on the solubility of uranium. Uranium
has been shown to dissclve in RbOH to
the extent of about 0.1 wt % at 650°C.
Addition of uranium trioxide to RbOH
does, however, produce a slurry very

similar to that obtained with NaOQH®3:14)

It should be noted that the RbOH used
contained about 5% Rb,CO;.

MODERATOR-COOLANT DEVELOPMENT

Development of moderator-coolants
is still concerned primarily with
preparation of pure alkali and alkaline
earth hydroxides and with phase
equilibrium studies of various mixtures

 

(1335, p. Redman, D. E. Nicholson, and L. G.
Overholser, “Suspensions of Uranium Compounds in
Sodium Hydroxide,” Aircraft Nuclear Propulsion
Project Quarterly Progress Report {or Period
figg{?g March 10, 1951, ANP-60, p. 135 (June 19,

(14)y,. G. Overholser, D. E. Nicholson, and
J. D. Bedman, “Suspensions or Solutions of Uranium
in Molten Hydroxides,” ANP-63, op. cit., p. 96.

84

of these materials withother compounds.
Since 1t 1s apparent that the most
difficult problem 1n use of these
materials 1s control of the corrosion,
major emphasis in the program 1s in
that field (see Sec. 11, Corrosion
Rescarch).

Relatively pure samples of the
bydroxides of sodium, potassium,
barium, strontium, and lithium have
been prepared. Study of one bimnary
hydroxide system and several hydroxide
fluoride systems were completed during
the past quarter. Pending encourage-
ment from corrosion studies no further
studies of this kind seem justified
at present.

Preparation of Pure Hydroxides
(L. G. Overholser, D. E. Nicholson,
E. E. Ketchen, and D. R. Cuneo,
Materials Chemistry Division). Sodium
hydroxide has been purified in pound
batches by removing sodium carbonate
from a 50% aqueous solution of NaOH
followed by dehydration. The residual
sodium carbonate concentration 1is
about 0.5 wt %. Attempted methods for
purifying potassium hydroxide have
included recrystallization of this
material from isopropyl alcohol and
precipitation of carbonate as barium
carbonate in aqueous medium. While
not completely satisfactory, the
former method has proved the bhetter,
yielding potassium hydroxide with a
total alkalinity greater than 99% and
with less than 0.5 wt % potassium
hydroxide. Approximately 10 1b of
barium hydroxide has been purified by
removing the carbonate 1n aqueous
solution, recrystallizing barium
hydroxide octahydrate, and dehydrating
under carefully controlled conditions.
The resulting samples have contained
less than 0.4 wt % contaminant 1n the
form of either the oxide or carbonate.
Strontium hydroxide is being purified
in an analogous manner. Commercial
lithium hydroxide monohydrate has been
found to be quite pure on dehydration.
FOR PERIOD ENDING DECEMBER 198,

Kou-LioH (J. P. Blakely, L. M.
Bratcher, and C. J. Barton, Materials
Chemistry Division). This system
. closely resembles the NaOH-LiOH system
previously described.('3) The eutectic
point at 70 mole % KOH is at 225°C
while the incongruently melting com-
pound appears unstable above 315°C.
The series of halts at 245°C on the
KOH side of the eutectic, as shown 1in
Fig. 10,1, is probably due to the
solid transformation of KOH reported
by von Hevesy(!®) at 248°C,

 

 

 

 

 

 

600 F T | T T
- 500 -
D \
e
w400 - . 399
q
& 300 7 © o |
a ~o. :
§ M .
200 @ o _]
100 L < L L
0 10 20 30 40 50 &0 70 'O 30 100
i KOH
LicH KOH (mole %)

Fig. 18.1i. The System KOH-LiOH.

Hydroxide-Fluoride Systems (J, P.
Blakely, L. M., Bratcher, and C. J.
Barton, Materials Chemistry Division).
The relatively few determinations made
in the system RbOH-RbF indicate that
these materials, like the analogous
sodium and potassium compounds, form
solid solutions with the melting point
of the mixture rising sharply up to
about 50 mole % RBRbF and then somewhat
less steeply in the range 50 to 100
mole % RbF.

The pseudo-binary systems NaOH-
(LiF-NaF-KF) and KOH-(LiF-NaF-KF) in
which the alkali fluoride composition
represents the ternary ecutectic have
been studied to ascertaln whether

 

(15)g  A. Allen and W. C. Davis, “Binary
Hydroxide Systems,” ORNL-1154, op. cit., p. 1

(163G, . Hevesy, “Uber Alkalihydroxyde. I,”
Z. physik. Chem, 73, 667 (1910).

1551

low-melting liquids of low hydroxide
content can be prepared. It is
apparent that up to 15 mole % NaOH or
up to 30 mole % KOH can be added to
the alkali fluoride eutectic without
elevation of the melting point.

CGOLANT DEVELOPMENT
J. P. Blakely L. M. Bratcher
C. J. Barton
Materials Chewmistry Division

Investigation of low-melting
fluoride mixtures of possible appli-
cation as coolants has been continued
during the past quarter. While a few
binary mixtures have been examined,
largely to fill gaps in the available
literature, most of these studies have
been concerned with termary systems of
the various alkali fluorides and
beryllium fluoride. In addition to
the value of such data in coolant
development, any low-melting composi-
tions which are found are of potential
value as solvents for UF, 1in the
development of fuels of low uranium
content.

BbF-LiF. This binary system has
been examined i1n order to complete the
ternary systems containing these com-
pounds. As indicated in Fig. 10.2,
this system forms a simple eutectic

melting at 470 % 10°C at 42 mole % LiF.

RESTRICTED
D'AG 12528

 

 

 

 

 

sco T i T T T T
_ 8OO
g
W 700 - @
D
=
W 600
=
w
=
500 /
T € g @
400 ) 1 b 1 I i I L
0 0 20 30 40 S0 &0 70 A0 30 W00
RbF LiF {mole %) LiF
Fig. 10.2. The System RbF-LiF,

85
ANP PROJECT QUARTERLY PROGRESS REPORT

NaF-BeF,. The phase diagram from
the open literature(®) has been used
in construction of termary diagrams
involving these materials. Various
di fficulties with some of these
diagrams has, however, prompted re-
examination of this system. Figure
10.3 shows the results obtained.
Below 50 mole % BeF, the published
diagram was essentially confirmed
although some minor differences were
noted. In the region above 50 mole %
BeF, thermal data are apparently un-
reliable. The dotted curve in this
region shows the quenching data of
Roy, Roy, and Osborne. (17)

RISTRICTEG
LWG.13589

L

 

©
© -
Ll
&
2
<
g TN e x
o > K :
= X o ©
i -0 ¢ e o
%
= : o
e
e o o a

 

 

 

60 vo 8O S0 100

BeF
Naf BeF, (mole %) ¢

Fig. 10.3.

The System NaF-BeFQ.

NaF-KF-RbF. The lowest melting
point found in this system 1s at
621 + 10°C at 74 mole % RbF and 21
mole % NaF. The failure to find a
low-melting region on this diagram
(Fig. 10.4) is probably due to the
high melting points of the solid
solutions of RbF and KF.

KF-LiF-RbF. The equilibrium diagram
shown 1n Fig. 10.5 indicates that LiF
is considerably more effective than
NaF in lowering melting points 1in the
KF-RBbF system. The data on this
system are still far from complete and

 

(17)8. Roy, D. M. Roy, and E. F. Osborne,
J. Am. Cerem. Soc. 33, 85 ¥1950).

86

RbF RESTR-CTED
795°( CwG 13530

 

KF
850°C

 

 

The System NaF-KF-RbF,

RbF RESTRICTED
5

795°G OWG 13531

 

S o o\oolo 1
3 ° §5g/8
[
g $
o @
KF / l LiF
850°C 285 G 845°C

 

Fig. 10.5. The System KF-LiF-RbF,

further study may change the contours
considerably. The lowest melting

point so far discovered 1s 440 = 10°C
at 27.5 mole % RbF and 40 mole % LiF.

NaF-LiF-BbF, The incomplete data
on this system have been used to plot
the diagram shown in Fig. 10.6. Since
all three binary systems form simple
eutectics, the melting point available
with this system is considerably lower
than that with the two 1mmediately
preceding. The ternary eutectic
appears to melt at about 425°C and
"

FOR PERIOD ENDING DECEMBER 10, 1951

to be near 42 mole % LiF and 6 mole %
NaF.

 

 

RAbF RESTRILTED
795°¢ .9
750
100
875°C 650
ey
) § B 4T0°G
%
%
® &£
%, &3
&9 @ S$
B 5L
®, 7
% &
NaF \ ” 1 LiF
995G 650°C 845°C

Fig. 10.6. The System NaF-LiF-RbF.

Ternary Systems Containing BeF,.
Three ternary systems containing BeF,
are under investigation at present,
Study of these systems is difficult
both because of the precautions
necessary in handling the Bel, and
because of the supercooling and
generally poor thermal behavior of the
melts. The lowest melting points so
far observed are at 345°C in the
NaF-KF-BeF,, 325°C in the LiF-KF-BeF,,
and 220°C (possibly considerably
supercooled) in the LiF-NaF-BeF,
systems. Study of these materials is
continuing because of their possible
use in very-low-melting fuel systems.

SERVICE FUNCTIONS

G. J. Nessle V. C. Love
H. S. Powers C. J. Barton

Materials Chemistry Division

The Fuel Preparation and Service
Group was called upon to fill a wide
variety of containers during the past
quarter. The amount of fuel required
ranged from 0.5 to 3900 g. The 3900-g
batch of material required the con-
struction of special equipment for
melting and filtering the fused salt.
This equipment has been used several
times successfully. The filtering
material 1s sintered stainless steel,
Table 10.7 shows the number of con-
tainers filled for different types of
experiments,

Table 10.7

Container Filling Services

 

 

 

NO. OF
CONTAINERS

TYPE OF EXPERIMENT FILLED
Static corrosion tests 112
Cyclotron bombardment 31
Radiation damage 23
Loop tests 2
Circulating system 1
Density measurement 1

 

 

In addition to the container-
filling services, about fifty batches
of fused fluoride mixtures have been
prepared for various uses. This group
has also started tests designed to
elucidate the mechanism of fluoride
fuel stabilization and to indicate the
best method of carrying out this
stabilization.

87
ARP PROJECT QUARTERLY PRGGRESS>REPBRT

11.

CORRCSION

RESEARCH

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

During the past quarter a consider-
able change in emphasis occurred in
the corrosion testing program. While
fluoride fuels and coolants continued
to receive some attention, a major
effort was placed on attempts to
contain the alkali hydroxides in
nickel and the structural metals.
Corrosion tests have been run at
800 to 815°C in various nonmetallic
media, including the molten hydroxides
of barium, stromium, sodium, lithium,
and potassium and molten sodium
cyanide. Some tests were run on the
liquid metals sodium and lead to
check previous results.

In the corrosion tests with the
hydroxides 1t was found that only
nickel, copper, and the more noble
metals silver, gold, and platioum will
withstand the corrosive action of
these media. Therefore coatings of
copper and nickel on structural mate-
rials, stainless steels and inconel,
were tried with variable results.
Tests are underway to determine the
minimum coating thickness needed.
The hydroxides exhibit considerable
metal transport {mass transfer) in
tests when there 1s a temperature
gradient across the tube and the
hydroxides are flowing. Sodium,
potassium, and lithium hydrdxides,
circulated in nickel thermal-convection
loops at 1400°F, each provoked suffi-
cient mass transfer to plug their
respective loops. This mass transfer
phenomenon 1s being studied with
thermal-convection loops of two
different designs and with two addi-
tional pieces of equipment, a thermal
gradient standpipe furnace and a

88

"seesaw" furnace in which the hydrox-
ides are sloshed back and forth from
the hot to the cold portion of a tube.

In addition, work has been started
on a fundamental approach to the
problem of the reaction of the hydrox-
ides with metals at high temperature.
Some of the first tests tobhe run in the
NaOH-nickel system demonstrate the
role oxygen plays 1in the metal trans-
port phenomenon; in thermal-convection
loops the presence of oxygen will
increase the amount of mass transfer.
Results from thermal gradient stand-
pipe tests have shown the effect of
hydrogen atmosphere onmetal transport.
Considerable research will be needed,
however, before the transport phenomenon
will be understandable.

There i1s no evidence of unusual
problems concerning the behavior of
fluoride mixtures 1n static tests with
structural metals. Such tests have
demonstrated that the pretreated
fluoride mixtures can be contained in
capsules of inconel or 316 stainless
steel. Dynamic-corrosion tests of
fluorides in inconel convection loops
have been completely satisfactory
although similar tests 1w stainless
steel loops have shown appreciable
mass transfer. There is evidence that
the iron and nickel content of the
pretreated fuel decreases after 100 hr
of exposure 1n the structural metal
while the chromium content of the fuel
increases. This phenomenon, however,
is not completely certain and 1s being
investigated further because of the
light i1t may shed on the pretreatment
problem.
FOR PERIOD ENDING DECEMBER 10,

STATIC CORROSION BY FLUOGRIDES

H. J. Buttram C. R. Croft
N. V. Smith J. M. Didlake

Materials Chemistry Division

No information was obtained which
would change the previously reported(!’
conclusions that properly treated
fluoride mixtures present no serious
corrosion problems with structural
metals, such as inconel and stainless
steel, 1n static tests for periods up
to several hundred hours. This applies
also to the low-uranium-content fuels,
as well as to the uranium-free alkali
fluoride mixtures. Exposure at temper-
atures up to 900°C for 100 hr also
yielded satisfactory results.

Although the fluorides were pre-
treated in the majority of cases,
untreated fluorides did not seriously
attack inconel test specimens. Pre-
treatment with a single metal rather
than by alloys appears to be feasible.
The mechanism of the pretreatment
process 1s not completely clarified
yet, but it 1s receiving further
attention.

Mixtures containing lead fluoride
(60 mole % NaF, 23 mole % PbF,, 17
mole % UF,) were subjected to the
standard corrosion test. Materials
containing PbF, are unsatisfactory.
The mixture was found to be unstable
and the metallic lead which formed
alloyed with the container metal. No
further tests are planned with mixtures
containing lead fluoride.

Stabilization of capsules of stain-
less steel and inconel by pretreatment
with elemental fluorine was attempted
wlthout significant effect. No
difference was observed in corrosion

 

(s gtatic Corrosion by Fluoride Fuels,” Air-
craft Nuclear Propulsion Project Quarterly Progress
Report for Period Ending September 10, 138514,
OB§L91154.

1951

resistance to fluorides of these
materials as compared with untreated
specimens.

STATIC CORROSION BY HYDROXIDES

W. D. Manly, Metallurgy Division
F. Kertesz, Materials Chemistry Division

A number of corrosion tests have
been made at 800 and 816°C for 100 hr
in nonmetallic media which might find
possible application as reactor
moderators or coolants. These include
the molten hydroxides of barium,
sodium, strontium, lithium, and potas-
sium. JThe media were dehydrated
commercial products and especially
pure sodium, potassium, and barium
hydroxides. Purified sodium hydroxide
appeared somewhat less corrosive than
did the dehydrated commerical material;
similar benefits do not appear from
purification of potassium and barium
hydroxides.

Without reference to change in
mechanical properties it would appear
that potassium hydroxide 1s less
corrosive on inconel than is sodium
hydroxide. Changes 1in mechanical
properties of the test specimens have
not been examined. In these tests,
however, nickel and copper resisted
chemical reactionm with most of the
hydroxides. Stainless steels coated
with copper and nickel were therefore
a logical development to combine
strength with corrosion resistance.
It has been found, however, that thin
coatings are not effective 1n suppres-
sing corrosion whereas heavier coatings
do show promise. Tests are underway
to determine the minimum coating
thickness required and the factors
which influence the adherence of these
coatings.

The temperature dependence of the

attack was also investigated; at 500°C
potassium hydroxide showed only a very

89
ANP PROJECT QUARTERLY PROGRESS REPORT

slight penetration which increased
somewhat at 600°C and became much
stronger when temperatures of 700 and
800°C were reached. As far as time
was concerned, 1t appeared that most
of the attack occurred during the first
few hours of the exposure.

Corrosion of Uncoated Metals (A.
D. Brasunas, D. C. Vreeland, E. E.
Hof fman, and R. B. Day, Metallurgy
Division). Several metals were tested
inmolten hydroxides of sodium, lithium,
potassium, barium, and strontium for
100 hr and at temperatures up to 816°C.
These tests were conducted in evacuated

containers of the same composition as
the metal specimen., The results with
barium, strontium, sodium, and lithium
hydroxides are summarized in Tables
11.1, 11.2, 11.3, and 11.4, respec-
tively. Most metals and alloys are
susceptible to hydroxide corrosion 1in
varying degrees, the notable exceptions
being nickel, copper, and monel. :

These vresults were obtained with
especially purified sodium hydroxide,
while the potassium hydroxide was a
dehydrated commercially pure product.
There was evidence of decreased
corrosion i1n sodium hydroxide which

Table 11.1

Summary of Corrosion by Ba(OH), at 816°C for 100 hr

All specimens tested in evacuated capsules of like materials

 

 

 

 

 

 

DEPTH OF
METAL
WT. CHANGE AFFECTED
MATERIAL (g/in.z) {(mils) METALLOGRAPHIC NOTES
304 stainless steel 5.5 0.010-in. oxide layer
310 stainless steel +0, 046 3.5 0.005-in. oxide layer
316 stainless steel 9.0 0.010-in. oxide layer
318 stainless steel -0.061 4.5 0.006-in. oxide layer
321 stainless steel -0.740 4.5 0.005-in. oxide layer
347 stainless steel -0.083 4.7 0.006-in. oxide layer
446 stainless steel 5.0 0.008-in. oxide layer (Fig. 11.5)
Copper (O,F.H.C.) -0.003 0.5 No evidence of attack
Iron (Globe) -0.994 8.5 Heavy oxide formed
Zirconium -0.221 9.0 0.001-in. gray oxide layer;
voids observed to depth of
0.004 1in.
Hastelloy B -1.109 9.5 Heavy oxide formed
Inconel X 6.5 0.020-in. oxide layer (Fig. 11.6)
Nickel Z 6.0 Erratic attack,0.002 in. average,
(Fig. 11.7) 0.006 in. maximum
Monel -0.005 0.0 No evidence of attack

 

90
FOR PERIOD ENDING DECEMBER 10, 1951

Table 11.2

Summary of Corrosion by Sr(0H), at 816°C for 100 hr

All specimens tested in evacuated capsules of like material except where noted

 

 

 

DEPTH OF
METAL
AFFECTED
MATERIAL (mils) METALLOGRAPHIC NOTES
304 stainless steel 12.0 Heavy oxide layer; specimens cracked on
bending
310 stainless steel 4.7 Very brittle oxide layer
316 stainless steel 3.0 0.003-in. oxide layer
318 stainless steel 2.9 0.006-in. oxide layer
321 stainless steel 4.5 0.006-in. oxide layer
347 stainless steel 7.0 0.020-in. oxide layer
446 stainless steel 4.0 0.011-in. oxide layer
Iron (Globe) 2.0 Heavy oxide layer, 0,017 in., on speci-
men (Fig. 11.8)
Inconel X 3.0 0.007-1in. oxide layer
Hastelloy B Specimen heavily oxidized and embrittled
Copper (O.F.H.C.) 0.0 No evidence of attack
Nickel A No evidence of attack
Nickel Z 0.002-in. layer of attacked zone
Zirconium™ 7. 0.001-in. oxide layer

 

 

 

*Tested in nickel capsule.

could be ascribed to purification.
Potassium hydroxide did not show any
improvement when especially purified.

The depth of attack on the different
stainless steels tested varied from
2.9 mils [Sr(OH), on 318 stainless
steel] to 24.5 mils (LiOH on 304 stain-
less steel). Typical oxide layers are
shown in Figs. 11.1 and 11.2 which
1llustrate the attack by barium hydrox-
ide on 304 and 446 stainless steel,
respectively. Inconel 1s attacked to
a depth of 6.5 mils by barium hydroxide
with the formation of a 20-mil oxide

layer (Fig., 11.3). The attack by
barium hydroxide on nickel Z was
erratic, with a maxipum depth of
attack of about 6 mils (Fig. 11.4).
Chromium, while suffering a uniform
surface solution to a depth of 3 mils
in sodium hydroxide, showed no evidence
of this attack (Fig. 11.5). The
attack by lithium hydroxide on inconel
seems to be of a different nature, as
can be seen in Fig. 11.6. The penetra-
tion does not follow the grain bound-
aries but shows a wavelike pattern.
On the basis of the tests performed to
date, dehydrated RbOH attacks inconel

91
ANP PROJECT QUARTERLY PROGRESS REPORT

Table 11.3

Summary of Corrosion by Sodium Hydroxide at 816°C for 100 hr

Specimens tested in evacuated capsules of like material except where noted

 

 

 

 

 

 

DEPTH OF
SPECIMEN
WT. CHANGE AFFECTED
MATERIAL (g/in.?) (mils) METALLOGRAPHIC NOTES
304 stainless steel 24.5 0.025-in. oxide layer
Chromium -0.1 3.0 No evidence of attack; uni-
form solution only (Fig. 11.9)
Zirconium® -2.0 82.0 Very severe attack; specimen
almost completely consumed
Nickel A +0,001 0.0 Slightly attacked
Nickel Z 0.6 20.0 Severely attacked
Al,0, (sapphire)* 2.3 Uniform solution; attack not
severe
*Tested in a nickel capsule.
Table 11.4

Summary of Corrosion by Lithium Hydroxide at 816°C for 188 hr

All specimens tested in evacuated capsules of like material

 

 

 

 

DEPTH OF
SPECIMEN
WT. CHANGE AFFECTED
MATERIAL (g/in. 3 (mils) METALLOGRAPHIC NOTES
304 stainless steel -0.185 4.5 0.014-1in. oxide layer
Inconel +0,043 4.0 0.006-1n. oxide layer
Nickel A -0.001 0.0 No evidence of attack

 

 

 

 

more severely (Fig. 11.7) than KOH
under the standard test conditions.¢?’

(H. J.
C. R, Croft,

Effect of Exposure Tinme
Buttram, N. V. Smith,

 

(2)F; gure 10.11, ORNL-1154, op. cit., p. 115.

92

and J. M. Griffin, Materials Chenistry
Division). In order to have a standard
to which subsequent results might be
compared, two series of experiments
were initiated. They consisted of a
study of the attack of sodium and
potassium hydroxides on inconel as a
function of exposure time, using 800°C
FOR PERIOD ENDING DECEMBER 10, 1951

UNCLASSIFIED
Y-4696

 

Fig. 11.1. Surface of 304 Stainless Steel after 100 hr of Exposure to
Barium Hydroxide at 816°C.

UNCLASSIFIED
Y-4701

 

Fig. 11.2. Surface of 446 Stainless Steel Specimen after 100 hr of Exposure
to Barium Hydroxide at 816°C.

93
ANP PROJECT QUARTERLY PROGRESS REPORT

r’
/
i.

Wmmwmfix‘.uw

 

11. 3.

Fig.
posure to Barium Hydroxide at 816°C.

as the reference temperature. The
periods of exposure were 1, 25, 50,
100, 250, 500, and 1000 hr. These
tests are complete except for the

1000-hr runs.

Figure 11.8 demonstrates the
corrosion of inconel by sodium hydrox-

94

 

s S o M ol

UNCLASSIFIED
Y-4700

Surface of Inconel Specimen after 100 hr of Ex-

ide at 800°C as a function of time.
After 1 hr the metal is penetrated to
a depth of 3 to 4 mils; this increased
after 50 hr to 3 to 10 mils. After
100 hr a nonmetallic layer is visible
under the corroded metallic layer.
Little additional attack is evident
after 250 hr.
FOR PERIOD ENDING DECEMBER 10, 1951

UNCLASSIFIED
Y-4732

 

Fig. 11.4. Surface of Nickel Z Specimen after 100 hr of Exposure to Barium
Hydroxide at 816°C.

 

Fig. 11.5. Surface of Chromium Specimen after 100 hr of Exposure to Sodium
Hydroxide at 816°C.

95
ANP PROJECT QUARTERLY PROGRESS REPORT

UNCLASSIFIED
T-497

 

Fig. 11.6. Surface of Inconel Specimen after 100 hr of Exposure to Lithium
Hydroxide at 800°C.

UNCLASSIFIED )
T.347

 

. 2 .

o . . \ N
% .’ T o 4 . o
% .
4 P C - = : . .

¥ . - . A - * * < * .1 'L - . ’ i

: . < .
.

L8 0 b : &
, /o :
» s . - e T
. - - " ,
s . -I .‘ - \ . )... . "k‘« A\ 2 b - R ’

g i -
A T A e & f\ * 3 r e e e, ML B ¢ T IES

Fig. 11.7. Surface of Inconel Specimen after 100 hr of Exposure to Rubidium
Hydroxide at 800°C.

96
FOR PERIOD ENDING DECEMBER 10,

1951

UNCLASS | F IED

 

{ hr 50 hr

Fig. 11.8.
Hydroxide at 800°C.

ORIGINAL SURFACE

   

£ :a . - ". . . . &
: ; 3
. -.' " . . = -,y ¢ .
P " -
)W el
™o : i .I_ & e - * -
A Tal N 8
- . L 1 {:" ‘ &
Vet [ .j. i X~Hk?
i{a‘n v 2y i - ‘\ 2
» o »
s g 5 * 4 L
."‘ ," -j! .
{ hr 50 hr
Fig. 11.9.

Hydroxide at 800°C.

Figure 11.9 shows the analogous
attack on inconel at 800°C by potassium
hydroxide. After 1 hr the penetration
is only 1 to 2 mils, and relatively
slight additional attack occurs during
succeeding test periods. Measurements
indicate 3% weight loss of the specimen
after 500 hr. Apparently KOH is less

100 hr 250 hr

Effect of Exposure Time on the Corrosion of Inconel by Sodium

UNCLASS I FIED

100 hr

250 hr

Effect of Exposure Time on the Corrosion of Inconel! by Potassium

corrosive than NaOH under these
conditions.

Effect of Temperature (H. J. Buttram,
N. V. Smith, C. R. Croft, and J. M.
Griffin, Materials Chemistry Division).
Of equal importance is the effect of
the test temperature. It was desirable

97
ANP PROJECT QUARTERLY PROGRESS REPORT

to determine the temperature at which
only negligible attack occurs during
the usual period of 100 hr. Figure
11.10 shows the behavior of potassium
hydroxide in inconel at 500, 600, 700,
and 800°C. At 500°C there is no
visible attack. At 600°C there is
localized attack. In the sample shown
a penetration of 5 to 6 mils can be
observed. This is not found to be the
case 1n other portions of the test
specimen and 1s pictured here to
illustrate occasional erratic be-
havior of the hydroxide-metal systems.
At 700°C there is some pitting and
penetration which does not always
follow the grain boundaries. The
attack at 800°C is considerably worse.
Weight loss data (plotted in Fig.
11.11) indicate a strong increase with
increasing temperature.

A series of thermal standpipe tests
1s 1n progress using potassium hydrox-
ide at 538 to 815°C in an attempt to
determine the rate of reaction at these
temperatures and the effect of the
types of atmosphere above the hydroxide
on the rate. Vacuum, air, and hydrogen
are being used. Preliminary results
in air and vacuum indicate that both

 

600°C

500°C

Fig. 11.10. Effect of Temperature
Hydroxide in 100 hr.

98

316 stainless steel and inconel are
attacked. The depth of attack at
538°C is of the same order of magnitude
as that noted i1n a static corrosion
test at 816°C, i.e., around 0.006 in.
In hydrogen, however, no corrosion
product was noted on either 316 stain-
less steel or inconel at 538°C. At
704°C shallow corrosion had occurred.
Verification of the above preliminary
results is being sought.

Corrosion of Coated Metals. Nickel
and copper have been found to be among
the few metals which are resistant to
molten hydroxides. Therefore these
materials have been applied as protec-
tive coatings on stainless steels for
use 1n contact with the hydroxides.
Preliminary results indicate that
corrosion protection can be achieved
in this manner.

Type 304 stainless steel was found
to be severely attacked by sodium
hydroxide at 815°C in a 100-hr test.
The depth of metal showing oxidation
by the molten caustic was 25 mils.
Consequently, it was decided to evaluate
the protective coatings of nickel and
copper by depositing them on this
steel.

UNCLASSIFIED

  

o l; i i
® G . <
o o8 . . .
VaRieE . S e . = s
R e s O A
,\1._-.": \'c -(_' ;" -) .l : C‘ o
. o g % .," e \ » -
- 2 B © g - te.. '
Jj wee EEL O
* .., 23 ? _"_'fl '.“.‘ in
3 g Ve
" : \ ;;' { n .
By L et SRy
‘ o . .~::‘" : l':'. o
t . .’ ",
: b .Y
- .
o :
o o
0O C 800C

on the Corrosion of Inconel by Potassium
FOR PERIOD ENDING DECEMBER 10, 1951

UNCLASSIFIED
DWG. 43629

 

l 1 T ?
!
{
— i
/-
/
100 | ) 1.0
/ i
90_? // —403
/
80 t— 0.8
n}}— 0.7

mq /dm>/day

60

50 0.5

   

40 0.4
mg /dmz/duy

30 Q.3

20 — 1
500 600 700
TEMPERATURE ({°C)

 

 

800

Fig. 11.11, Weight Loss of Inconel
Specimen in Potassium Hydroxide for
100 hr as a Function of Temperature.

Electroplated Specimens. A 3-nmil
copper-plated coating on 304 stainless
steel reduced attack by sodium hydrox-
ide to 5 mils in 100-hr tests at
816°C. A 3-mil nickel plate also
resulted in a reduced attack of 9 mils
on the underlying stainless steel.
Copper-plated specimens were tested
in copper capsules, and nickel-plated
specimens were tested in nickel
capsules. Electroplated specimens
tested in sodium hydroxide appeared to
have poor plate adherence while those
tested in barium hydroxide were very
adherent. The data are summarized in

Table 11.5. 2

In tests with barium hydroxide for
100 hr at 816°C the 3-mil nickel plate
on 304 stainless steel offered good
protection. The 10-mil oxide layer
which was obtained with the uncoated
specimen (Fig. 11.1) is almost entirely
eliminated although a small amount of
oxidation, as shown in Fig. 11.12, was
observed. With the 3-mil copper plate
the specimens were severely attacked.
with the exception of the corners,
which were relatively unattacked.
This may possibly be attributed to
the heavier electroplate which occurs

Table 11.5

Summary of Hydroxide Corrosion of Clad Metal Specimens at 816°C for 100 hr

 

 

 

 

STAINLESS DEPTH OF METAL WT. CHANGE
STEEL COATING BATH ATTACKED (in.) (g/in. D) REMARKS

304 None NaOH 0.025 (uniform) Uniform attack

304 3 mils of Cu NaOH 0.005 (uniform) +0.094 [Cu plate not adherent

304 3 mils of Ni NaOH 0.009 (uniform) -0.094 {Ni plate not adherent

304 None Ba (OH) , 0.029 (uniform) -1.0 ‘Uniform attack

304 3 mils of Cu Ba (OH) , 0.094 (maximum) -1.92 Adherent Cu plate
0.027 (average)
0.001 (minimum)

304 3 mils of Ni Ba (OH) , 0.001 (uniform) +0.002 Adherent Ni plate

 

 

 

 

 

 

99
ANP PROJECT QUARTERLY PROGRESS REPORT

 

UNCLASSIFIED
Y-4815

 

Fig. 11.12.

fication 500X; reduced to 96%.

at the corners. Heavier copper
coatings will be tried in order to
test this possibility.

Nickel-clad 316
stainless steel and nickel-clad inconel

Clad Specimens.

specimens were prepared using the
"picture frame" technique and tested
in sodium hydroxide for 100 hr at
816°C. The nickel coating was approx-
imately 5 mils thick at the sides and
appreciably heavier at the ends. The
thin sections of nickel were attacked
but the thicker end sections were not
(Fig. 11.13). There is no satisfactory
explanation for this behavior at this
time. On inconel and 316 stainless
steel a heavier nickel cladding (10

100

Surface of 304 Stainless Steel Coated with 3-mil Nickel Electro-
plate after 100 hr of Exposure to Barium Hydroxide at 816 °C.

Original magni-

mils) produced from nickel powder gave
excellent protection from corrosion 1in

NaOH at 816°C for 100 hr (Fig. 11.14).

No evidence of oxidation 1s apparent.

Dissolution of Metals in Sodium
Hydroxide (C. R. Croft and N. V. Smith).
Attempts have been made to determine
the amounts of various metals dissolved
in sodium hydroxide under 1inert
atmospheres as a functionof temperature
and exposure time. The sodium hydrox-
ide is melted under an argon atmosphere
in a container fabricated from the
test metal, or of nickel in the case
of hard-to-fabricate metals such as
chromium. Filings of the test metal
FOR PERIOD ENDING DECEMBER 10, 1951

 

11.13.

Fig.
Sodium Hydroxide at 816°C. 50X

were present in all cases to increase
the rate at which equilibrium could
be established. The sample was main-
tained at temperature for 4 hr before
a sample was drawn from the container
through a filter of graphite to avoid
the dispersed but undissolved metal.

 

UNCLASSIFIED
Y-4948

 

ATTACKED Ni
COATING ON
OUTSIDE

316 ON INSIDE

 

Nickel-Clad Inconel (Nickel Sheet) after 100 hr of Exposure to

The metal content of the hydroxide
was then determined by analysis.

Table 11.6 lists the results obtained
to date. The oxidation state of the
dissolved metal and the mechanism of
process are as yet unknown.

101
ANP PROJECT QUARTERLY PROGRESS REPORT

 

i /A -
\ | L w
Fig. 11.14. Nickel-Clad Inconel (Nickel Powder) after 100 hr of Exposure
to Sodium Hydroxide at 816°C. 100X
Table 11.6

Metal Content of Sodium Hydroxide as Function of Temperature

 

 

 

 

 

 

 

 

 

METAL CONTENT (wt %)
4 HR AT 4 HR AT 4 HR AT 24 HR AT 24 HR AT
METAL 400°C 600°C 700°C 400°C 600°C
Copper 0.15 0.19 0.6 0.16 0.54
Nickel 0.14 0.15 0.26
Chromium 0.40 1.94 0.60
Iron 0.17 1.07 1.17
STATIC CORROSION BY with moderating properties. Results

FLUORIDE-HYDROXIDE SYSTEMS

H. J. Buttram C. R. Croft
N. V. Smith J. M. Griffin
Materials Chemistry Division

Mixtures of sodium hydroxide—sodium
fluoride and potassium hydroxide—
potassium fluoride were studied because
of their potentialities as coolants

102

obtained show some reduction of corro-
sion on inconel and stainless steel,
but it appears that the improvement is
mostly due to the dilution of the
hydroxide by the fluoride. This
fluoride addition cannot be great for
otherwise the melting point will be
above the desired range. The systems
are mutually soluble in each other
without evidence of eutectic formation.
"

FOR PERIOD ENDING DECEMBER 10, 1951

STATIC CORROSION BY SODIUM CYANIDE

A. D. Brasunas R. B. Day
E. E. Hoffman D. C. Vreeland
Metallurgy Division

Numerous molten salts have been
considered for use 1in a reactor, but
cyanides do not appear to have been
mentioned. This nonoxidizing salt
is quite stable at high temperatures
and has been used extensively in the
heat treating and carburization of
metals. Pure sodium cyanide melts at
564°C (1047°F); however, this tempera-
ture can be lowered by suitable addi-
tions. The specific heat of sodium
cyanide is reported as 0.25 Btu/lb
(solid) and 0.40 Btu/lb (liquid); the
heat of fusion is 135 Btu/lb. Uranium
is appreciably soluble in sodium
cyanide at 816°C. A number of static
corrosion tests on a variety of
metals and alloys were run at 816°C
for 100 hr using evacuated capsules.
The results are given in Table 11.7.

The cyanide reaction consists in
carbon and/or nitrogen absorption by
the metal. This frequently results
in embrittlement as determined by a
simple bend test at room temperature.
There is, of course, the possibilaity
that embrittlement does not occur at
high temperatures but is merely a
room- temperature phenomenon.

The superiority of nickel and high
nickel alloys (inconel and nichrome V)
in resistance to carburization is well
known. Iron also appeared to be quite
inert to the molten cyanide, while
stainless steels showed varying
tendencies for reaction. The results
obtained with iron were somewhat
surprising. The lack of carburization
and subsequent embrittlement could be
caused by the absence of oxygen which
is essential to this reaction.

 

(3)4.5.M. Handbook, p. 694, 1948.

Silicon additions are known to be
very potent in suppressing carburi-
zation of commerical alloys that are
ordinarily susceptible to carbon
adsorption.

STATIC CORROSION BY LIQUID METALS

A. D. Brasunas R. B. Day
E. E. Hoffman D. C. Vreeland
Metallurgy Division

Corrosion tests of up to 1000 hr
of sodium on 315 stainless steel and
lithium on coated 304 stainless steel
have been completed. The attack by
sodium was negligible even at tempera-
tures up to 1000°C and did not appear
to increase after the first 100 hr.
Molybdenum and chromium coatings on
stainless steel have somewhat, though
not yet sufficiently, reduced the
attack by lithium.

Sodium on Stainless Steel. Long-
time tests in liquid metals initiated
some time ago have recently been
evaluated. Tests have been run with
316 stainless steel in contact with
sodium up to a thousand hours at
816 and 1000°C. The attack was
negligible at both temperatures.
Static corrosion tests have also been
made on 347 and 316 stainless steel
and inconel in sodium at 650, 704,
and 816°C for 100 hr. Attack in these
tests was also negligible, although
some embrittlement was noted. The
results are summarized in Table 11.8.

Lithium on Coated Stainless Steel.
Type 304 stainless steel specimens
were thinly coated with molybdenum and
chromium by the Linde Air Products
Company. These specimens were tested
in lithium at 1000°C for 100 hr in the
usual manner. The attack of the under-
lying metal, although somewhat mini-
mized, was not eliminated as much as
had been hoped. It is believed that

103
ANP PROJECT QUARTERLY PROGRESS REPORT

Table 11.7

Corrosion Data Obtained at 816°C Using Molten Sodium Cyanide for 100 hr

 

 

 

DEPTH OF
METAL
AFFECTED
METAL (in.) EMBRITTLED REMARKS

Iron (Globe) 0.000 No No evidence of attack

Nickel A 0.000 No No evidence of attack

Nickel Z 0.005 Yes 0.002-in. outer layer; gray
constituent formed to depth
of 0.005 in.

Inconel 0.003 No Integranular penetration of
gray constituent to 0,003 in.

Hastelloy B 0.020 Yes 0.002-in. outer layer, nitride
needles throughout

Hastelloy C 0.003 0.003-in. outer layer only
(nitride?)

Nichrome V 0.005 No 0.00l-in. outer layer; carbu-
rized to 0,005 in,

Uranium 0.01 Appreciable solution

Beryllium 0.002 Gray layer formed to depth of
0.002 1in.

405 stainless steel 0.002 Yes 0.002-1in. layer of scattered
voids

430 stainless steel 0.020 Yes Completely carburized; voids to
depth of 0.002 1in.

446 stainless steel 0.003 Yes 0.001-in. outer layer; 0.002-in.
carburized layer beneath

304 stainless steel 0.011 Yes 0.001-in. outer layer; 0.0ll-1in.
carburized layer beneath

316 stainless steel 0.007 Yes 0.002-in. outer layer; 0.007-in.

’ carburized layer beneath

347 stainless steel 0.007 Yes 0.002-in. outer layer; 0.007-in.
carburized layer beneath

310 stainless steel 0.012 Partly 0.002-in. outer layer; 0,012-in,
carburized layer beneath

 

 

 

 

thicker coatings would probably be more planned in order to check this possi-
effective and additional tests are Dbility,

104

Ly
Static Corrosion

FOR PERIOD ENDING DECEMBER 106, 1951

Table 11.8

Tests in Scdium

 

 

 

 

 

 

 

 

 

TEMP., TIME
OF TEST ¥T. CHANGE
MATERIAL (°C) (hs) (g/in.2) REMARKS

316 stainless steel | 1000 1000 +0, 001 No attack; no subsuarface
phase

316 stainless steel 815 1000 +0.013 Thin film; no cracks on
bending 90°; subsurface
phase to depth of 1 mil

316 stainless steel 815 100 +0.014 Thin film cracked on bending
90°; specimen contains a
subsurface phase to a
depth of 1 mil

316 stainless steel 704 100 +0, 009 No attack; no evidence of
cracking on 135° bend

316 stainless steel 650 100 +0, 006 No attack; cracked slightly
on 90° bend

347 stainless steel 815 100 +0,016 No attack; no cracking on
135° bend

347 stainless steel 704 100 +0.394 Thin surface deposit; no
attack but trace of sub-
surface phase present;
cracked on 90° bend

347 stainless steel | 650 100 +0,010 No attack; no cracking;
trace of subsurface phase

Inconel 815 100 +0, 004 No attack; cracked om 135°
hend

Inconel 704 100 +0.003 No attack: broke om 45° bend

Inconel 650 100 +0, 002 No attack; cracked om 135°
bend

STATIC CORROSION OF FUEL period. These tests included static

CAPSULES IN SODIUM

H. W. Savage and W. C. Tunnell
ANP Division

Static-corrosion tests on fuel-
containing capsules to determine the
combined corrosion effects of the
fluoride fuel NaF-BeF,-UF, and sodium
at 1500°F were completed during the

runs on inconel fuel~contzining cap-
sules in sodium of 600, 800, 900, and
1000 hr duration. In all, omne complete
series of tests has been rum on inconel
and on 316 stainless steel fuel cap-
sules. Fach series specified that
seven capsules be tested 1n sodium at
1500°F for periods of 100, 200, 400,
600, 800, 900, and 1000 hr. Each

capsule contained a slug of the capsule

105
ANP PROJECT QUARTERLY PROGRESS REPORT

material in addition to the fuel. The
analysis of the capsule and fuel are
reported in Tables 11.9 and 11.10 for
the inconel and the 316 stainless steel
capsule tests, respectively. In
general, the effect of either corrosive
fluid was no greater than that obtained
in tests 1n which the other fluid 1is

DYNAMIC CORROSION IN THERMAL
CONVECTION LOOPS

G. M, Adamson, Metallurgy Division
During this period the major effort

with thermal-convection loops has been
divided into corrosion studies with

not present.

fluorides and hydroxides.

Table 11.9

Analysis of Inconel Fuel Capsules in Sodium at 880°C

Yarious -

 

 

 

 

WT. CHANGES'¢) WT. CHANGES'?®’
RUN TIME OF CAPSULE OF SLUG ANALYSIS OF FUEL (ppm)
(hr) (mg/dmzfday} (ug/dm?/day) Fe Ni Cr

0 (start) 0 0 125 1065 20
100 -14.3 +1.0 160 110 4180
200 -7.5 -6.9 120 66 2880
400 -5.8 ~-1.6 215 180 1880
800 -2.6 +0,7 400 870 2550

 

 

 

 

 

 

(°)Weight changes

(b)Weight changes of slug are due to fuel corrosion.

Table 11.10

of capsule are due to sodium corrosion frem the outside.

Anzlysis of 316 Stainless Steel Fuel Capsules in Sodium at 800°C

 

 

 

 

WT. CHANGE(#’ WT. CHANGE'?®’
RUN TIME OF CAPSULE OF SLUG ANALYSIS OF FUEL (ppm)
(hr) (mg/dm?/day) (mg/dn?/day) Fe Ni Cr

0 (start) 0 0 125 1065 20
200 -4,32 -11.15 85 <15 3650
400 -8.78 -5,88 1750 300 6640
600 -2.05 -4,44 90 <20 630
626 -1.5 -5.97 110 <20 1820
900 -2.21 -0,848 140 <20 2310
1000 -4.97 -1.58 150 35 1110

 

 

 

 

 

 

(a)Wéight changes of capsule are due to sodium corrosion from outside.

(6)Wéight changes of slug are due to fuel corrosion.

106
FOR PERIOD ENDING DECEMBER 10, 1951

‘hydroxides have been run in nickel
and in every case excessive
‘mass transfer has taken place. Fluoride
:salt mixtures have been successfully
‘circulated in . inconel but have plugged

loops,

in stainless steel.
Fluoride Corrosion.
to be the most desirable material for

circulating fluorides.
loop has now been operating at 1500°F

for 240 hr with the fluoride fuel
and
29.2 wt % LiF to which enough UF, has
) Since
no falling off is being found in the.
-cold-leg temperatures, mass transfer.
~does not seem to be taking place.
‘Another inconel loop has been operating
‘with the above fluoride mixture with
UF, for 570 hr, again without any sign
The 316 stainless
steel loop containing the same mixture
The loop
has been X-rayed and shows both light
"and dark patches in the hot leg. Which
‘constituent caused the plugging and
‘the exact location of the plug could
A 316 stainless
steel loop was terminated after 173 hr'
of operation with the uraniumless
This loop developed
~a leak about the center of the vertical
‘portion of the hot leg. While the leak
was responsible for the termination,
‘the cold-leg temperature had dropped
55°F, which indicates that some mass
‘transfer was also taking place. X rays
of this loop also showed several dark
areas in the hot leg. Whether these are
segregation or gas pockets formed on
~cooling can be determined only when the
| However, since the
they are

Metal-
lographic examinations have not yet

11,7 we % NaF, 59.1 wt % KF,

‘been added to make 1.1 mole %.

of mass transfer.

plugged completely in 82 hr.

‘not be determined.

“fluoride mixture.

loop is sectioned.
. cold leg does not show any,
most likely to be segregation.

"been made on these loops.
Hydroxide Corrosion.

‘potassium, and lithium hydroxides were
‘circulated in nickel with a hot-leg

On the basis
of preliminary tests, inconel appears

An inconel

‘been partially dissolved.

Sodium,

temperature of 1400°F. All these
loops failed by plugging in the hot
leg as a result of mass transfer. The
operating times were: sodium hydrox-
ide, 54 hr; potassiumw hydroxide, 51
hr; and lithium hydroxide, 317 hr.
The lithium hydroxide loop has not
yet been completely examined but
appears to be like the other two which
are almost identical,

Photomicrographs of both the hot
and cold leg of the KOH loop are shown
in Fig. 11.15. Although the surface
of the hot leg appears to be well
polished, examination under the micro-
scope shows that this is a false surface
which has been pulled away from the
original material. Under the micro-
scope, the cold leg surface appears
guite rough with a layer of equi-axis
crystals on the surface. Held in the
rough spots on this surface is a con-
siderable quantity of a nonmetallic
constituent. Dendritic crystals were
found lightly attached to the wall in
all sections of the loop except the
hottest part of the hot leg. The
walls in all sections of the loop have
‘ In the
sodium hydroxide loop about 0,002 in.
had been dissolved from the cold leg
and 0,007 in. from the hot leg. The
figures are even higher with the
potassium hydroxide loop. From X rays
and dissolving out the hydroxides,
which leaves the dendritic masses in
place, the plugs in all loops have
been found to be located in the
horizontal section and lower part of
the vertical section of the hot leg.
Although dendrites were found attached
to the wall in the cold leg, no large
crystal masses were found in this
section., Figure 11.16 is two views
of the plug in the horizontal section
of the sodium hydroxide loop. The
top view is as cut while in the bottom
the hydroxide has been washed out.
The metallic masses are high-purity
nickel. From X rays of the plug the

107
ANP PROJECT QUARTERLY PROGRESS REPORT

B UNCLASSIFIED
T.522

 

 

(b)

Fig. 11.15. Nickel Thermal-Convection Loop Walls with Potassium Hydroxide.
(a) Hot Leg. (b) Cold Leg. 250X

108
FOR PERIOD ENDING DECEMBER 10, 1951

UNCLASSIFIED
Y-4884

UNCLASSIFIED
Y-4883

 

(b)

Fig. 11.16. Plugged Section Nickel Thermal Convection Loop with Sodium
Hydroxide Coolant. (a) As Cut. (b) Metallic Constituent.

109
ANP PROJECT QUARTERLY PROGRESS REPORT

crystals appear to grow from the bottom
of the pipe and slant into the coolant
stream. No intergranular type of
attack was found i1n any portion of
these loops.

Inconel loops were terminated after
1000 hr with sodium at 1600°F and with
NaK at 1500°F. Neither loop developed
serious amounts of mass transfer or
corrosion. A small amount of pitting
was found in the hot area, but no
intergranular type of attack was found.

FUNDAMENTAL CORROSION RESEARCH

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

Reported failure of the empirical
approach to furnish a satisfactory
container material for molten hydrox-
ides has emphasized the need for
fundamental understanding of the
mechanism of corrosion and mass trans-
fer of metals by these liquids. At
present 1t 1s not possible to state
with certainty the nature of the chemi-
cal reactions taking place; attempts
to minimize or control the corrosive
action are difficult to plan system-
atically.

Efforts to understand the nature
of the reacting species are, at present,
directed along two parallel lines.
Electrochemical studies of two different
types are attempting to define the
ionic species involved in corrosion
and mass transfer and are intended as
useful tools for study of these com-
plex phenomena. In addition, a careful
study of mass transfer, as well as
other aspects of corrosion such as
diffusion of various atoms from liquid
phase into the solid phase, and the
effects of various atmospheres and
other environmental factors on corro-
sion has been undertaken. One of the

110

primary aims of this work will be
concerned with the development of a
suitable inhibitor for corrosion in
the chemical sense and for inhibition
of metal crystal (mass transfer)
formation. These programs have been
initiated only recently, and results
to date are tentative and incomplete.
Sufficient progress has been made,
however, to indicate that useful data
will result from these approaches.

EMF Measurements in Hydroxides (A,
R. Nichols, Jr., Materials Chemistry
Division). Apparatus and materials
are being assembled for the measurement
of electrode potentials i1n fused
hydroxides. The apparatus will permit
measurements involving a range of
temperatures (325 to 1000°C) of
solvents (fused hydroxides), of solutes
(NiO, Cr,0,, FeO, etc.), of electrode
materials (Ni, Fe, Cr, Ag, Pt, C,
alloys), and of atmospheres (H,, He,
vacuum). A nonmetallic cell container
and diaphragm will be used over as much
of the temperature range as possible.

An attempt will be made to develop
reversible electrodes which correspond
to reactions of the type involved 1in
the corrosion and mass transport
phenomena. If such electrodes can be
devised, it will be possible to deter-
mine free energies and temperature
dependencies. This may provide an
understanding of the mechanisms of
these processes and hence point the
direction for their prevention.

Measurements now in progress involve
the concentration cell

Ni, NiO (dissolved in NaOH at c,) ||
NiO (in NaOH at ¢,), Ni

to determine whether the Ni-dissolved-
NiO half-cell is reversible.

Polarographic Study of Sodium
Hydroxide (R. A. Bolomey, Materials

.
FOR PERIOD ENDING DECEMBER 10, 1951

Chemistry Division). Techniques for
the study of polarographic curves to
_indicate the presence of contaminants
~in fused caustic are under investi-
gation. Because of the high conductivity
"of molten NaOH, a rather elaborate
_compensating circuit 1s required in
~order to distinguish the polarographic
waves. As presently operated, using
~a Brown recorder to trace the curves,
- the position of the waves is indicated
- by sharp peaks 1in an otherWISeascendlng
" current-voltage curve.

Sufficiefit reproducibility has not
not yet been achieved, but the results
~are encouraging. JIrials with a
- stationary platinum electrode showed
- two peaks with NaOH contained in =
platinum crucible under a nitrogen
~atmosphere. These peaks, at -1 and
-1.4 volts, may have been attributable
"to platinite and platinate ionmns.
. The instrument was modified and it was
~ found advisable to operate the polaro-
- graphic cell in a vacuum te avoid
~erratic behavior apparently due teo
convection currents caused by surface
cooling in the presence of a gas.
Hydrogen was an ountstanding offender
in this connection. It will be of
~interest to modify the apparatus so
that gaseous products from the mixture
~can be analyzed. So far it seems
~improbable that the peaks can be
assoclated thh Na* or NaH.

Experiments with NaOH contained in
~silver.give peaks occurring at different
~voltages than those encountered in
" platinum; however, the experiment is
~ being repeated under improved condi-
tions. The solidified NaOH contained
a mat of fine metallic needles which
appeared to have been freely suspended
in the upper part of the melt. These
are pictured in Fig. 11.17 but have

not yet been identified.

Snrveytlfthe Mzss-Transfer Pheneme;
non {W. D. Manly, Metallurgy Division;
F. F. Blankenship, and R. P. Metecalf,

Materials Chemistry Division). The
phenomenon of mass transfer of nickel
in systems comntaining caustic has
been reported from a number of 1nstalla-
tions which have experimented with
these materials under a variety of
conditions. The experiments which
have definitely shown this phenomenon
are descrlbed briefly helow. :

Summary of Experimental Observe-
tions. Experimental Engineering
thermal-convection loops have shown
adherent wall deposits in colder
regions and Spongy dendritic plugs in
hotter regions in three experlments
using loops of Y%-in. nickel pipe
approximately 90 in. in circumference.
Velocities of about 25 ft/min and a
temperature differential of 760 to
680°C prevailed in these experiments.

NACA. Temperature gradient loops
with centrifugal circulation utilizigng
a temperature differential of 815 to
800°C with velocities of 15 ft/sec gave
massive wall deposits in cooler regioas.
Mass transfer was accelerated by the
addition of sodium metal. '

BMI. Corrosion tests 1in a nick§1
pot having a thermocouple well, gas
inlet tube, and specimen support
immersed in the melt showed heavy
deposition of nickel at the liguid
level around metal walls and at culder
portions.

NACA. Corrosion tests 1in sealed
crucibles with a temperature differ-
ential of 815 to 730°C from top to
bottom gave rise to a crystalllne
deposit in the colder portion.

X-10. Asingle low-velocity thermal-~
convection loop locaded under hydrogen
utilizing a temperature differentizal
of 700 to 425°C developed a scattered
deposit on the cold wall with small
evidence of dendritic plugging in
in 300 hr of operat1on. ,

111
ANP PROJECT QUARTERLY PROGRESS REPORT

 

Fig. 11.17.
Containing Sodium Hydroxide.

Y-12. Corrosion tests in sealed
tubes with temperature cycling showed
small amounts of deposited metal after
12 cycles of 8 hr each from 800 to
200°C using helium gas inside sealed
tubes.

Instances in which directly observ-
able deposits of metal were not apparent
were as follows:

Y-12. Isothermal corrosion tests
(25 experiments) in capsules sealed
under helium and heated in a vacuum
at 800°C for 100 hr showed erratic
weight losses but no significant mass
transfer.

X-10. Temperature-gradient corro-
sion tests with hydrogen covering the
melt and surrounding a vertical tube
17 in. long, 400°C at top, 800°C at
middle, and 600°C at bottom, gave no

112

Deposit of Silver Crystals by Mass Transfer

UNCLASSIFIED
T«511

_

in Silver Capsule

evidence of mass transfer and very
little oxidation product in 100 hr.

Chemical Examination of Plugged
Convection Loops. In view of the
failure of mass transfer to develop
in certain cases 1n which 1t was
expected, particular attention has
been paid to the contents of plugged
loops obtained from the Experimental
Engineering experiments mentioned
above and described in greater detail
elsewhere in this report.

A mapping of the extent and general
appearance of deposited metal was
carried out. No correlations were
achieved, largely because of the lack
of uniformity of the deposit on colder
walls. There rough sparse deposits
were blotched with dull surfaces and
with regions containing fine dense
crystals. Occasionally the line of
FOR PERIOD ENDING DECEMBER 10,

demarcation between sharp, bright
needlelike crystals and a dull surface
was surprisingly sharp. Such lines
occurred in both the upper and lower
arms, running longitudinally along
the tube as though tracing a liquid
level., Uniformly, the hotter sections
beyond the dendritic plugs were highly
polished. Surfaces in the immediate
vicinity of plugs were visibly pitted
but bright.

Hydroxide from various portions of
the loops was submitted for analysis
with the results shown in Table 11.11.
A careful study of the oxidation
products from sections of the KOH loop
is still in progress. On being
leached with water, the hotter portions
gave rise to a gelatinous black
precipitate which dried to a brownish
powder. Colder regions yielded brown

1951

powder, and the trap contained a
mixed green and black powder. Tiny
metallic crystals were also found in
each case. X-ray-diffraction methods
showed only NiO and nickel in varying
proportions in the residues obtained
from nine different regions in the

KOH loop.

An extensive black deposit from the
cold leg of the LiOH appeared to be
NiO from the standpoint of X-ray
diffraction, but the chemical analysis
showed only 49% Ni. There was much
less discoloration of the surface of
the hydroxide as found in the filling
chamber in the case of LiOH,

Spectrographic examination of the
pipe of which the nickel loop (con-
taining NaOH) was constructed showed

Table 11.11

Analysis of Hydroxide From Nickel Thermal Convection Loops

After Plugeging

 

 

 

 

LiOH NaQH KOH
Ni NiO Ni NiO Ni10O2 Ni NiO
REGION (%) (%) (%) (%) (%) (%) (%)
Hot leg 1.02 1.35 3,79(a) | 0,0 0.01 } 23.3%) | 0,01¢(%)
Cold leg 1.31 15,402 | 0,06 0.01 | 0.06
Top arm 0.08 1.58 400 ppm | 0,01
Bottom arm 9.45 0.01 24.4 0.01
Darkest crust from
fill chamber 0.3 1.2 6,07 0.83 0.20 1,92 20.2
Clean hydroxide
from fi1ll chamber | 1 ppm 100 ppm |Total Ni

 

 

 

 

 

 

 

 

(“)Sample contained some metallic dendritic sponge.

(b)x-ray-diffraction analysis of this sample showed 40 to 70% NaOH, 20 to 40% Ni, and (probably

present) 5 to 15% NiO.

(C)A lar

case of LiOH was unique.

amount of oxide adhered to the wall near the bottom of the cold leg.

In this respect the

113
ANP PROJECT QUARTERLY PROGRESS REPORT

1.3% Fe, 0.6% Co, 0.8% Cu, and 0.04%
Mn. The dendritic sponge contained
0.04% Fe, 0.3% Co, 0.8% Cu, and 0.01%

Mn, indicating that cobalt and copper
tend to be transported to a greater
extent than iron and manganese.

Conclusions. Regardless of differ-
ences in proposed mechanisms, the
oxidation of nickel 1s an essential
part of all plausible explanations of
mass transport., There 1s reason to
believe that, even with large tempera-
ture gradients, mass transfer would

114

not be a problem in nickel—-molten
hydroxide systems operated under
conditions such that the oxidation of
nickel does not occur.

A small step toward this end has
been the attempted development of
nickel vessels suitable for carrying
out the final stages of hydroxide
purification, and of such design that
the molten hydroxide could be trans-
ferred to experimental apparatus
without exposure to air or other
sources of contamination.
FOR PERIOD ENDING DECEMBER 10, 1951

12. HEAT-TRANSFER RESEARCH AND PHYSICAL PROPERTIES

H. F. Poppendiek, Reactor Experimental Engineering Division

Experimental temperature measure-
ments in a simulated liquid- fuel
element, using a brine solution as the
hcat-generation medium, indicated
significant reduction of fuel-element
center temperatures at high heat
fluxes as a result of free convection,
These results are in general agreement
with the theoretical work which has
been carried on.

The fused-salt heat-transfer
apparatus has been completed, and
test-section heat-loss calibrations
are currently being made. Upon com-
pletion of these calibrations heat-
transfer coefficient measurements for
the NaF-KF-LiF eutectic are to be
obtained. Progress has been made in
ligquid-metal heat-transfer work on
lithium and sodium systems. Analysis
of data on entrance-region heat
transfer in sodium is presented, along
with mathematical solutionms. Mathe-
matical solutions have also been
derived for forced-convection heat
transfer (laminar and turbulent flow)
in long smooth pipes containing fluids
with uniform-volume heat sources.

Data are being obtained on heat
capacity, density, and vapor pressure
of materials., Some data have been
obtained on viscosity and thermal
conductivity, but further measurements
awailt the completion and testing of
additional equipment, Heat-capacity
data have been determined for uranium
fluoride, a lead-bismuth alloy, a
fluoride-fuel mixture, nickel, and
sodium hydroxide. Measurements on
thermal conductivity of diatomaceocus
earth have been obtained by use of a
radial-flow apparatus, while other
equipment for measuring this property
in ligquids and in solids 1s nearaing

completion, Data on density and

vapor pressure of fluoride salt
mixtures have been obtained, from
which coefficients of expansion and
heat of vaporization have been calcu-
lated. Some values for the viscosity
of fluoride salt wixtures have been
obtained, while other viscosity
apparatus remains under construction.

NATURAL CONVECTION 1IN
LIGQUIB-FUEL ELEMENYS

F. E. Lynch, ANP Division

D. C. Hamilton, R. F. Redmond, and
M. Tobias, Reactor Experimental
Engineering Division

Preliminary data have been obtained
on a new simulated fuel-element system,
described in the previous guarterly
report.{!) A 25% solution of NaCl in
water was used as the heat-generation
medium, and cooling was effected by
natural convection from the wall of a
3-mm quartz tube to a stirred bath.
The results plotted in Fig. 12.1 are
for six series of tests at coolant
temperatures of 8, 12, 17, 29, 32, and
66.5°F, respectively. The ordinate
9/9c 1s the ratio of the measured
temperature difference between the
wall and the axis of the tube to the
temperature difference computed from
the conduction equation for the systen.
The data are in general agreement
with analytical (laminar flow) so-
lutions for values of the scource term
that are less than 110 watis/cm?;
at this point the ordinate decreases
sharply. This decrease is attributed

 

(D E. Lynch and M. Tobias, ‘““Measurement of
the Fuel-Element Temperature Distribution,”
Aircraft Nuclear Propulsion Project Quarterly
Progress Report for Period Ending Septesmber 10,
1951, ORNL-1154, p. 133 (Dec. 17, 1951).

115
ANP PROJECT QUARTERLY PROGRESS REPORT

UNCLASSIFIED
DWG. 13593

 

 

 
  

 

 

Sttt T 17 T 17 1 1
1.0 —
© o —— — —
o
0.8 — ]
o 0.6 | —
~ | § =0 CENTER ~ @ WALL ]
® 8.=8 FOR CONDUCTION ONLY
0.4 |— —
0.2 }— ~
o ey e
0 20 40 80 80 100 120 140 160 180
q'"' (WATTS PER CUBIC CENTIMETER)
Fig. 12,1, Temperature Ratio inp a 3-mm Tube Filled with Brine in which Heat

is Geperated Uniformly.

to a change in the flow from the
laminar to the turbulent regime.

_ A coolant jacket is being added to
the quartz-tube system so that the
cooling will be accomplished by forced
flow parallel to the quartz tube. The
flow rate and the axial temperature
gradient of the coolant will be
measured.

Four analytical solutions which
relate radial temperature distributions
to several dimensionless moduli,
including such variables as pipe
diameter, volume heat source, physical
properties of the fuel and coolant,
and coolant fluid flow rates for
free-convection systems, have been
obtained for the laminar flow region.(®

 

(2)Three memorandums on this are to be
issued around January, 1952.

116

Similar solutions for the turbulent-
flow regime are being sought. A new
parallel-plate apparatus in which the
velocity distribution will be measured
is being constructed which will employ
an electrolyte as the heat-generating
mediuvm.

HEAT-TRANSFER COEFFICIENTS

Heat Transfer in Fused Hydroxides
and Salts (H. W. Hoffman and J. Lones,
Reactor Experimental Engineering
Division). The apparatus for the
determination of the heat-transfer
coefficients of molten salts and
hydroxides has been completed. Heat
loss calibrations of the test section
are being carried out. Heat-transfer
measurements using a fluoride mixture
will be made upon completion of these
calibrations,
™

FOR PERIOD ENDING DECEMBER 18, 1951

Since the last report the experi-
mental system has been fully instru-
mented, and all tanks and flow lines
have been cleaned with 2 nitric acid
solution and filled with approximately
200 1b of the NaF-EF-LiF cutectic.
During the filling operation several
cold spots were noted i1n the system,
and the auxiliary heating circuit was
altered to eliminate these regions.
The system has been further modified
to include mixing pots of a helical-
coil design immediately preceding and
following the test sectiom in order
to determine more accurately the
entering and leaving mixed-mean
temperature of the test fluad.

Heat Transfer in Moltem Lithiun
(H. C. Claiborne and G. M. Winn,
Heactor FExperimental Engineering
Division)., It was previously re-
ported¢?’ that the modified figure-
e1ght system for the determination of
lithium heat-transfer coefficients
had been successfully operated, and
that preliminary heat-transfer data
had been obtained. To facilitate
accurate separation of the individual
thermal resistances from the experi-
mental overall heat transfer coef-
ficients, a large pump (approximately
40 gpm at 100 psi) 1is required. Since
a pump of the required capacity
appeared to be unavailable 1n the
next few months, 1t was decided to
hold the figure-eight system in stand-
by status until the regquired capacity
pump is obtained and to build a smaller
system utilizing a heat-transfer
section that is resistance-heated by
an electrical current. The latter
method obviates fluid thermal-resistance
separation techniques and has the
additional advantage that it allows
the measurement of the actual heat
flux distribution along the tube
length,

 

(3)0. P. Coughlen 2nd [[. C. Claiborne, ““Heat
Tra?§fer in Molten Lithium,’ OBRNL-1154, op. cit.,
p- 9.

Construction of the new system
is about 95% complete. Data are
expected to be obtained by Janunary,
1952, for Reynolds meoduli up to 40, 000.

The design gqualifications for the
test section are:

1. About 95% of the heatis generated
in the tube wall,

2. The average temperature differ-
ence betwsen the wall and the
fluid mean temperature is at

least 10°F.

3. The axial mean temperature rise
is at least 10°F,

4., At least 75% of the thermal

resistance 1s 1n the lithium.

5. There is a maximum of 3000 amp
to the test section.

6. The length-to-diameter ratio of
the test section is 100.

Qualifications 1 and 2 require a metal
having high thermal and electrical
conductivities. Copper was selected
as the material for the test section,
A 10-hr static corrosion test(*) at
450°F indicated that a polished copper
specimen lost only 0.02% in weight.
Therefore it appears that copper will
resist attack by lithium long encugh
(about 20r 3 hr) to take the necessary
data if the maximum operating tempera-
ture does not exceed 450°F.

A flow diagram of the system 1is
shown in Fig. 12.2. The dimensions of
the test section are: length, 11¥% in.;
1.d.,, 0.1175 1in.; and o.d., 0.1715 in.
Flow is produced by two electromagnetic
pumps 1in series (previous operational
data indicate that this arrangement
will give 1.7 gpm at 40 psi). Flow
measurements will be made with an

 

(4)1,. A. Abrams, personal communication to
W. D. Manly, Oct. 5, 1951.

117
ANP PROJECT QUARTERLY PROGRESS REPORT

—a— GAS INLET

 

 

 

 

 

 

UNCLASSIFIED
DWG. 13594

 

 

 

 

 

 

 

 

 

 

 

 

 

MICROMETULLIC a-MELT AND FILTER TANK COPPER COOLING COILS
FILTER
Ei 1 COOLING WATER E{ E{
GAS !NLET7 = Y
E( l BUS BAR "L"
—[ T -= BY PASS
MIXING CUP—" il
SURGE TANK-» T
TEST SECTION —=
EXPANSION JOINT~\H‘1::i;:i
ELECTROMAGNETIC EZ }i
FLOW METER T -
ELECTROMAGNETIC
PUMPS + I
g
[:] ~~—GAS INLET
DRAIN AND CALIBRATION TANK —=
Fig. 12.2. Flow Diagram for Lithium Heat-Transfer Experiment.

electromagnetic flowmeter that will
be calibrated in place. The bypass
is used for testing the circuit and
for obtaining approximate equilibrium.
This prevents unnecessary exposure of
the copper test section to lithium.
Before data on the test section are
obtained, the bypass will be frozen
and disconnected to prevent short-
circuiting of the heating current to
the test section. The tube can be
easily replaced in case of failure.

It 1is felt that the described
system will produce satisfactory

118

lithium heat-transfer data and, with
modification of the test section,
will allow the heat-transfer charac-
teristics of other liquid metals to be
determined in the same apparatus.

Entrance-Region Heat Transfer in a
Sodium System (W. B. Harrison, Reactor
Experimental Engineering Division).
The system built for obtaining entrance-
region heat-transfer coefficients to
molten sodium has been described in
previous guarterly reports (see p. 1v
for list of previous reports). Briefly,
the test section consists ofa circular
FOR PERIOD ENDING DECEMBER 10, 1951

copper plate having a hole in the
center anda tube around the periphery.
Sodium flows through the center hole,
and a cooling or heating medium is
circulated through the tube at the
periphery., The plate is insulated to
promote radial conduction, and the
temperatures at different vadial
positions were measured in order to
determine the copper surface tempera-
ture at the copper-sodium interface
and the rate of heat flow through the
plate.

A total of 48 experimental runs
have been made to date. Plates of
1/16 and 1/8 in. thickness have been
used 1n conjunction with hole diameters
of 1/16, 5/64, and 3/32 in. Values of
P, D/L ranged from 140 to 585, where
P_is the Peclet modulus, D is hole
diameter, and L 1s the plate thick-
ness, All data were taken with the
sodium in turbulent flow at tempera-
tures between 250 and 310°F. The
data have been extremely erratic
and very low when compared with the
predictions based on the postulates of
slug flow (uniform velocity distri-
bution} with heat transfer by molecular
conduction only. In a number of
cases the data were low even when
compared with predictions based on the
postulate of laminar flow (parabolic
velocity distribution).

Following are several possible
explanations for the low data:

1. Peculiar velocity distribution,
This might be caused by creep of
the gasket upstream from the
plate. It was found that some
creep actually does take place.
However, the effect was practi-
cally eliminated by giving the
gaskets a preliminary compression
set at the operating temperature.

2. Nonwetting of the copper by the
sodium. There are few data
available on wetting of surfaces

4.

by sodium, but it appears that a
clean surface is wet by sodium
without difficulty, even at the
low operating temperature, The
surface in the test section was
degreased with a detergent and
electrochemically polished. No
significant changes resulted in
the data.

Precipitation of oxide on the
heat-exchange surface. This was
considered to be a possibility
during the runs in which the
sodium was being cooled. The
apparatus was equipped for heat-
ing the sodium without signifi-
cant improvement of the data.

Contamination of the sodium such
that its properties have been
altered; particularly, the
thermal conductivity has been
reduced, The chief contaminant
is the oxide, with some iron
particles also in suspension,
This appears to be the wmost
likely cause for the erratic and
low data. The system is presently
being cleaned and modified so as
to minimize contamination. There
has been a fairly consistent
decrease in the heat-transfer
coefficients with time, implying
that the condition is being
aggravated by a buildup of
contaminant., It is believed
that particles of oxide and 1iron
have been suspended in the
stream, such that the extent of
contamination exceeds the equi-
librium solubility of sodium
oxide in sodium. The oxygen
comes from three sources: (a)
residual oxygen in the system
prior to loading the sodium;
(b) residual oxygen in the form
of oxide on the surfaces of
sodium bricks which were loaded
into the system; and (c) oxygen
present in the argoun used in the

system., A filter is being in-

‘stalled in the sodium line so as

119
ANP PROJECT QUARTERLY PROGRESS REPORT

to remove oxide 1in excess of the
equilibrium solubility. In order
to eliminate the oxide introduced
during loading of bricks directly
into the system, a separate melt
tank will be used. From this
tank the sodium will be trans-
ferred into the system through a
filter,. In order to reduce
contamination from the gas,
helium will be passed through a
NaK scrubber befere it reaches
the system. These changes in
the system are being incorporated
at the present time. Jt 1s
hoped that the system may be
started up again by the middle
of December,

Heat Transfer in a Circualating rFuel
System (H. F. Poppendiek and .. Palmer,
Reactor Experimental Engineering
Division)., Mathematical solutions
have been derived for temperature
structure in the case of forced-
convection heat transfer (laminar and
turbulent flow) in long smooth pipes
containing fluids with uniform-volume
heat sources; heat i1s transferred to
or from the fluids at the pipe wall.
Some specific evaluations of the
radial temperature distributions for
the case of no heat transfer to or
from the fluid at the pipe wall have
been made for laminar and turbulent
flow. Diwmensionless temperature
profiles for the case of turbulent
flow are presented in terms of Reynolds
modulus, Prandtl modulus, the volume
heat source, and the boundary hesat
transfer.

HEAT CAPACITY
W. D. Powers G. C. Blalock
R. M. Burnett, Reactor Experimental
Fngineering Division

The heat capacities of nickel,
sodium hydroxide, uranium fluoride, a

lead-bismuth alloy, and a fuel mix

120

have been determined by means of
Bunsen ice calorimeters.(3* %) The
equations in Table 12.1 give the heat
capacity as a function of temperature;
H, - Hyo, is in cal/g, ¢, is 1in
cal/g*°C, and T 1is in degrees centi-
grade.

THERMAL CONDUCTIVITY

A Deem type apparatus for the
measurement of thermal conductivity of
liguids has been completed and 1is
currently being tested with lead. A
longitudinal flow apparatus for
measuring thermal conductivity of
solids 1s mearing completion. Some
measurements of thermal conductivity
and sintering have been made on
diatomaceous earth, a material antici-
pated for use as insulation in the
circulating-fuel —~water-moderated
reactor.

Thermal Conductivity of Ligquids
(S. J. Claiborne and M. Tobias,Reactor
Experimental Engineering Division),.
A Deem type apparatus for the measure-
ment of the thermal conductivity of
liquids has been modified as indicated
in an earlier report.{?’ The apparatus
is now complete and 1s being tested
using lead. Some difficulty has been
encountered in keeping thermocouple
wells leak-free, but this problem has
been solved, at least for lead as the
test liguid. Replacement of the
stainless steel bellows with one of
nickel-plated copper or brass is now
contemplated, because the present
bellows is very stiff and causes a
certain amount of distortion in the

apparatus. An additional apparatus 1s

 

(5)A. R. Frithsen, “Physical Properties,”
OFNL.-1154, op. cit., p. 134,

(6)a. R. Frithsen, “Physical Properties,”
Aircraft Nuclear Propulsion Project Quarterly
Progress Report for Period Ending June 10, 1951,
ANP-65, p. 159 (Sept. 13, 1951).

(7). Tobias, A. R. Frithsen, and L. Basel,
“Thermal Conductivity of Liquids,” ORNL-1154,
op. cit., p. 135,
121

Nickel

NaQk

UF

Pb-Bi alloy (55.3 mole % Be)

Fuel mixture {46.5 mole % NaF,
26.0 mole % KF, 27.5 mole %
UF4}

Table 12.1

Heat Capacities of vVarious Substances

250 - 1000°C  H, (solid) - Hyo, (solid) = 0.26 + 0.119T +1.53 x 10”572

340 - 1000°C  H, (Jiquid) - Hyo, (solid) = 69 + 0.497T

300 - 930°C Hp (solid) - H o (solid) = &.0 + 0.073T + 3.1 x 107572

200 - 950°C HT (liquid) - Hyo. (solid) = 5.8 + 0.0345T

240 - 535°C H, (solid) - Hjo. (solid) = -0.3 + 0.15T

535 - 1000°C HT (liquid) - Hjo. (solid) = -13.4 + 0.237

H

1t

0.12 + 3.1 x 10°°T

0.49 ¢ 0.02

0.073 + 6.1 x 10°°T

0.034 £ 0.0015

0.15 + 0.01

0.23 £ 0.01

IS6T ‘0T HIGWADIA ONIGNT GOIUAS NMOJ
ANP PROJECT QUARTERLY PROGRESS REPORT

being build so that delays caused by
failure of equipment parts may be
reduced.

Thermal Conductivity of Solids
(W. D. Powers, Reactor Experimental
Engineering Division). Construction
of an additional logitudinal flow
apparatus for the thermal conductivity
of solids 1s 90% complete and will be
checked with Armco iron shortly. This
apparatus can be used in a vacuum or
in an 1nert atmosphere and will permit
more accurate measurements than the
original(®) apparatus.

Thermal Conductivity of Diatomaceous
Silica Powder (D. F. Salmon, ANP
Division). Research on the circu-
lating-fuel—water-moderated reactor
has led to a search for an insulating
material for the fuel containers. Thais
insulation would allow the water to be
maintained at a considerably lower
temperature than the fuel mixture,
thereby simplifying many of the
problems associated with high-tempera-
ture high-pressure systems, Diato-
maceous silica powder seemed feasible;
hence tests were made to determine
thermal conductivity and sintering
effect at reactor temperatures under
atmospheric and reduced pressures.
The material used was a Johns-Manville
product, Celite.

The apparatus consisted of two
annuli formed by concentric tubes
around a standard 750-watt 115-v
tubular heating element. The annulus
next to the heater element contained
diatomaceous silica while the outer
annulus was a water passage, Experi-
mental parameters included heater
power, sheath temperature, density to
which insulation was packed, water
flow rate, and water inlet and outlet
temperatures.

(8)M. Tobias, “Thermal Conductivity of Solids,”
Aircraft Nuclear Propulsion Project Quarterly
Progress Report for Period Ending March 10, 1951,
ANP-60, p. 243 (June 19, 1951).

122

The first series of tests was made
at atmospherzc pressure. The heater
element sheath temperature was main-
tained at 1500°F for 75 hr at 1675°F
for 100 hr, and at 1800°F for 100 hr.
During this entire series the tempera-
ture rise of the water never exceeded
10°F., Figure 12.3 shows the variation
in thermal conductivity with tempera-
ture both from the literature values(®)
and from the experimental ones. In-
spection at the conclusion of these
tests showed that a thin layer of the
powder next to the heater element had
darkened considerably, probably as a
result of oxidation of the inconel
sheath., Apparently no significant
change in physical properties had
occurred, since the thermal con-
ductivity remained consistent through-
out the tests.

A second series of tests was begun
in which the powder annulus was evacu-
ated to determine the effect of the
presence of air on thermal conductivity
and sintering. The insulation annulus
was loaded to a packing densaity of
15.7 1b/ft? and evacuated to approxi-
mately 0.17 wm Hg. Other experimental
conditions were unchanged.

At a heater sheath temperature of
1800°F, an apparent conductivity of
0.025 Btu/hrft*°F was calculated,
representing a reduction of some 60%
over that at atmospheric pressure.
After 24 hr of operation, however, the
conductivity began to increase as
evidenced by a decrease in sheath
temperature, while the power on the
heater element remained essentially
constant; however, insufficient time
had elapsed to establish the extent of
this increase. Examination of the
powder after testing indicated that
a layer approximately 1/16 in. thick
around the heater had sintered and
darkened appreciably in color.

 

()G, B. Wilkes, Heat Insulation, Wiley, New
York, 1950.

3y
vh

FOR PERIOD ENDING DECEMBER 190,

1951

UNGLASSIFIED

 

 
 

 

 

 

 

D¥6G. 13595
" [ I ! T B | ! I | !
0.140 |— _
CALCULATED DENSITY i
o OF POWDER TESTED, 17.2 ib/ ft
= AVERAGE WATER FLOW RATE
Jx oazof— 135 1b/ hr —
~le° EGEN
oo A-
L 0- } VALUES FROM LITERATURE
N © --
0.100 - —
> o~ EXPERIMENTAL VALUES AT ATMOSPHERIC PRESSURE .
s - EXPERIMENTAL VALUES AT REDUCED PRESSURE 2o b/
}_
|5 ,..-v"‘/
L ]
§ 0080 flw&”““ 187Ib/fl
5
O THEORETICAL ,4r"’”’wfl ’flflflM{}mflfwf~;w_flflaw~‘
., o
% | m,:::::ji::::::d,,,flv~w~*”” 12.5 Ib/ ft3
i o
I
l_
0.040 ONE POINT _
AFTER 12hr
*
e B S
0.020 L - -
0.000 | 1 I [ [ bt [ ‘ , |
0 j00 200 300 400 500 60O 700 800 900 1000

MEAN TEMPERATURE (°F)

Fig. 1Z.3.

No reason has been advanced to
explain why diatomaceous silica
sintered at reduced pressure but
failed to sinter at atmospheric
pressure. Further tests at reduced
pressure are underway.

DENSITY OF LIQUIDS

J. M. Cisar, ANP Division

Density measurements have been made
on three more fused salt mixtures.
Compositions, melting points, and
temperature ranges over which data
were taken are given in Table 12.2.
The equations representing the data
within 5% are indicated, where
p = density and T = temperature (°C),

The data on the NaF-KF-LiF-UF,
mixture, although consistent in them-

Thermal Conductivity of Diatomaceous Earth.

as the
salt was accidentally allowed toremain
exposed to air for a 3-hr period
before the runs were made. A sample
1s being analyzed to determine the

selves, may not bhe reliable,

effect of hydrolysis during thais
period, The apparatus used for these
determinations 1s based on ligquid
buoyancy principles; it was described
in a previous quarterly report, (1%
In addition, a second density apparatus
of this type has been built and is now
in operation,

VISCOSITY

The falling-ball viscometer has
been modified and tested with satis-
factory results. Approximate results

 

(1035 J. Kaplan, “Density of Liquids,” ANP-60,
op. cit,,p. 246.

123
ANP PROJECT QUARTERLY PROGRESS REPORT

Table 12.2

Data on Several Fused Salt Mixtures

 

 

 

 

COMPOSITION MELTING POINT TEMPERATURE RANGE
COMPOUND | (mole %) °c) (°c) EQUATION

NaF 48.92 558 625 - 890 p = 4.54 - 0.0011T
KF 26.8

UF, 25.0

NaF 10.9 430 - 440 525 - 850 p = 2.647 - 0.00090T
KF 43.5

LiF 44.5

UF, 1.1

NaF 11.5 455 550 - 850 p = 2.385 - 0.00059T.
KF 42.0

LiF 46.5

 

 

 

 

 

are expected with the use of a Zahn
type viscometer now hearing completion,
Some values for fluoride salt mixtures
have been obtained using the modified

Brookfield Synchro-Lectric viscometer
(Fig., 12.4).

Falling-Ball Viscometer (S. I.
Kaplan and T. N. Jones, Reactor
Experimental Engineering Division),
The falling-ball viscometer has been
tested at room temperatures, using
SAE 10 lubricating oil as the test
fluid. The viscosity of the o0il was
checked with a Brookfield viscometer
and found to agree with the result of
the falling-ball instrument to within
about 10%. After high-temperature
testing 1is complete, the instrument
will be used for investigations of
molten salt mixtures,

Recent modifications of the falling-
ball viscometer include removal of the
solenoid valve at the bottom and
substitution of an externally heated
straight pipe extension and instal-

124

lation of a removal cap to permit
cleaning.

Zahn Type Viscometer (M. Tobias,
Reactor Experimental Engineering
Division). A Zahn type viscometer has
been designed to give rapid, approxi-
mate values for fuel and coolant-salt
mixtures. The viscometer will consist
of a stainless steel cup, at the base
of which 1s a hole 1 in. long and 1 mm
in diameter. The cup will be dipped
into the melt whose viscosity 1s
desired. The melt container will then
be dropped away from the cup and the
time of efflux of 5 cm?® of molten
material from the cup will be measured.
Since the apparatus will be inside a
furnace, the discharge time will be
obtained by weighing the cup erther
continuously with a dial-indicating
balance or by noting when the weight
of the cup plus its contents falls
below a predetermined amount., To
minimize corrosion, the apparatus will
be operated inan inert gas atmosphere.
The device will be calibrated using

c K
FOR PERIOD ENDING DECEMBER 10, 1951

RESTRICTED
PHOTO 1048

VISCOME TER

MANIPULATION PORT

oo ]

-INERT GAS QUTLET

THERMOCOUPLE
ORLE+

e FILLING FUNNEL
; GUARD PLATE

FURNACE AND |
CONTAINER — 1 - L
' FURNACE LEADS

INERT GASFNLETSV?’

 

Fig. 12.4. Viscosity Apparatus.

125
ANP PROJECT QUARTERLY PROGRESS REPORT

materials of known viscosity, both at
room temperature and at high tempera-
tures, The equation which an apparatus
of this type obeys 1s of the form

v = At - B/t

where v is the kinetic viscosity of
the liquid, A and B are constants
characteristic of the apparatus, and t
is the time of efflux,

Viscosity of Fluoride Mixtures
(F. A. Knox and F. Kertesz, Materials
Chemistry Division). Preliminary
viscosity determinations of the
various fluoride mixtures were con-
tinued. Although trials were made
with a Saybolt type apparatus, 1t was
found desirable to continue the use of
a modified Brookfield Synchro-Lectric
viscometer. In order to eliminate at
least some of the disadvantages of the
earlier techniques, the furnace was
enclosed in a metallic shell, covered
with a gas-tight cover which contained
the measuring instrument, a glass-
covered viewing hole, and an opening
for a rubber glove to permit certain
manipulations while the experiment
was in progress. The enclosed unit
was under slight argon pressure. In
this way oxidation of the material
during measurements should be reduced
very considerably, Details of the
apparatus can be seen in Fig. 12.4.

The need for such an essentially
gas-tight apparatus was apparent from
the results obtained with beryllium-
bearing fluoride mixtures. In a
poorly sealed system the mixture be-
came cloudy while the measurements
were 1n progress, indicating the
formation of insoluble beryllium oxide.
The viscosity values at a given temper-
ature continued to increase as repeated
measurements were made, resembling the
behavior of a suspension with rheopectic

126

properties rather than a Newtonian
liguid.

In order to establish the effect of
the uranium concentration on the
viscosity, the viscosity was determined
for fluoride salt mixtures contalning
various amounts of uranium fluoride.
The latest values for the viscosity of
the NaF-KF-LiF ternary eutectic range
around 3 centipoises at 800°C. Results
with 2 and 15 wt % UF, in the eutectic
show that these mixtures have es-
sentially the same viscosity as the
uranium-free composition.,

VAPOR PRESSURE OF LIQUID FUELS

R. E. Moore, Materials Chemistry

Division

The apparatus and procedure employed
in this work have been described in a
previous report,(!?) The cylindrical
vessel of 316 stainless steel con-
taining the NaF-KF-UF, eutectic mixture
was heated in a monel block 1na pot
furnace wound with Kanthal A wire,
Temperatures were measured by chromel-
alumel thermocouples attached to the
outer wall of the vessel. A hand-
controlled Powerstat was used to
regulate the temperature of the
furnace.

A preliminary test of the apparatus
and procedure was made in which vapor
pressures of mercury were determined
at several temperatures using a ma-
nometer containing Hyvac oil feor
pressure measurement, The results
showed this method was satisfactory;
hence measurements with fused salts
were 1nitiated,

The results of the vapor pressure
determination of the NaF-KF-UF,

 

(11)p E. Moore and C. J. Barton, *'Vapor
Pressure,;’ ORANL-1154, op. cit., p. 136.

@y e
FOR PERIOD ENDING DECEMBER 10, 1951

eutectic mixture are given in Table
12.3. The values at 1199 and 1267°C
were obtained with a meréury manocmeter
and are not so accurate as the others,
which were obtained with a wanometer
containing Hyvac oil. These data are
represented by the equation

logln P__ He - -9500/T + 7.234.
The heat of vaporization as calcu-
lated from the equation is 43.5
kecal/mole.

Table 12.3

VaporPressure(fl?theNaF°KF~vE¢Eutectic

TEMPERATURE PRESSURE, OBSERVED,

(°c) {mm Hg)

1073 1.43

1131 2. 82

1138 3.34

1199 6.5

1242 9.02

1267 12.0

127
ANP PROJECT QUARTERLY PROGRESS REPORT

13.

W. D.

METALLURGY AND CERAMICS

Manly, Metallurgy Division

T. N. McVay, Consultant

Solid-fuel-element fabrication has
been studied in order that the effect
of several variables on the metal-
lurgical bond between the fuel-bearing
core and the protective cladding and
the distribution of the UO, in the
metallic carrier in the core may be
understood. Previous reports have
outlined this effect by such variables
as percentage hot reduction, per-
centage UO, present, composition of
cladding, particle size of the metallic
carrier 1n the core, and degrees of
cold-working of the flat plate after
rolling. In this report the following
variables have beemn investigated:
effect of UO, particle size range,
effect of rolling tempesraturs, and
elimination of the capsule used during
hot-rolling.

Preliminary experiments have shown
that fuel-plate lawminates may be spot-
welded together. Fuel-tube bundles
can be held together by spot-welding
bundle straps to the individual tubes.
A study of high-temperature brazing
indicates Nicrobraz to be guite useful
in joining two fuel-plate laminates
without floating the U0, to the sur-
face and for joining a fuel plate to a
stainless steel sheet in a T joint.
A new brazing alloy containing 60% Pd
and 40% Ni has been found which has a
favorable nuclear cross-sectionaswell
as good high-temperature brazing
properties.

The creep and stress-rupture labo-
ratory for the testing of materials 1in
an inert gas atmosphere has been
completed. Results are presented which
show the effect of grain size on the
creep strength of inconel and of heat
treatment on the creep strength of
Duranickel,

128

Installation of equipment for the
ceramics laboratoryis still in process
although the laboratory has been in
partial operatiocn since early fall,
The work to date has included ceramac
coating of radiators and the hot
pressing of alumina,

SOLID-FUEL-ELEMENT FABRXCATION

E. S. Bomar and J. H. Coobs
Metallurgy Division

Of the three basic approaches to
preparing a solid-fuel element out-
lined in earlier veports (see p. 1v
for list), the e¢mz upen which primary
emphasis has been placed to date makes
use of picture frames and cladding
plates, The objectives of these in-
vestigations are twaofold: (1) to
abtain a metallurgical bond between
a fuel-bzaring core and protective
cladding, and (2) to create a distri-
bution of the solid fuel 1n 2 metallic
carrier which will give acceptable
heat-transfer conditions. The results
of completed experiments indicate that
flat or curved fuel plates can be
fabricated with 300 series stainless
steel cladding on a UD, ceramel core,

Earlier reports listed the effects
of the following variables:

1. Percent hot reduction.
2, Percent UQ, present.

3. Composition of cladding.
4

. Particle size of metallice

carrier 1n core.
5. Coarse U0, particles.

6§, Cold-working of flat plates
following hot-rolling.
FOR PERIOD ENDING DECEMBER 19, 1951

To these we may now add: ' Effect of U0, Particle Size. Vari-

3 ~ ationindistribution of U0, with three

1. FEffect of 1o, particle size. ; different sizes of U0, particles is

j § : shown 1n Fig. 13.1. The mesh grades

. 2. Effect of rolling temperature. | employed supplied particles measuring

. _ ‘ : from 250 to less than 10 4, the distri-
3. Elimination of capsule {origi- bution improving inversely as the size

nally used duvring hot-rolling). of the parvicles. This range of particle

_ ; sizes is of interest since it is
Preparation of tubular sections is  believed to cantain the minimum size
more difficult and a satisfactery end which will suppress radiation damage

 

 

product is not yet assured, f to a tolerable level.
SECRET
e AR ;
T GO o
TR CORE G
0% UQ, 44200
MESH
?Qo?.:- 3 e { B ‘,S’u:s:_
( STAINLESS | [ofn
STERL j Vi
8Y VOL.
CORE |
CORE 200
. 0T U, deums
f MESH
?QG& -:5\. ,_‘ 7 -
STaniEss {0228
Greny o |MESH
BY VOL,
CORE :
i MESH
- PO 302 1 L.
STAINLESS <« 250
STEEL g
BY VOL.
fe)

Fig., i3.1 a, b, c. Effect of U, Particle Size on UG, Distribution. 175X,

12¢
ANP PROJECT QUARTERLY PROGRESS REPORT

  
  
   
  

¥Ry

 

[ ?'._'
{

A

 

 

oy
Q%fi
w

4 W
g

ROLLING TEMP
1335 °C

 

(b)

Fig, 13.2 a, b. Eifect of Rolling Temperature ca Uo, Distribution. 175X,

Effect of Rolling Temperature. capsule during hot-rolling.¢!*2) This
Dependence of oxide distribution on step, however, proved to be one of the

rolling temperature is shown in Fig. more time-consuming 1tems 1in the
13.2. The obvious benefits of the fabrication schedule, Production of
higher temperature are somewhat offset fuel plates in large numbers would be

by the deeper penetration of surface facilitated by working out a schedule

ox1dation of the samples during rolling. omitting the capsulating step.

 

. . . (])flm T Fi 1 PI 2 tal L.
ype Fue ate,”” Metallurgy Division
Effect of Eliminatiom of Capsule. Quarterly Progress Report for Period Ending

The first sample fuel plates were January 31, 1951, OPFNL-987, p. 57 (June 7, 1951).

prepared by enclosing a stainless steel (2)Hot-Rolled Clad Fuel Plate,” Hetallurgy

"picture frame,”" fuel-bearing ceramel,
and cladding plates i1n an evacuated

130

Division Quarterly Progress Report for Period
€3gi?g April 30, 1951, ORNL-1033, p. 56 (Oct. 9,

w“l
FOR PERIOD ENDING DECEMBER 10, 1951

Two samples were processed without
 protective capsules with results whlch
- were quite encourag1ng.

The first laminate was heliarc-
~welded around its outer edges in a dry
box containing a purified helium
atmosphere. The second laminate was
prepared by a combination sintering
and hot-forging operation. A thin

PICTURE /
FRAME f

 

 

PICTURE
FRAME

 

(b)

Effect of Elimination of Capsule on Uflé Bistrihutién.

Fig.

13{3 a, b.
150X, :

 

 

 

band of stainless stéel powder was
painted around the ocuter periphery of
the picture frame using Nicrobra:z
cement as a binder; cladding plates
were spot welded on and the assembly
sintered overnight at 1250°C. The

sintered laminate was next hot-forged
at 900 to 1000°C by pressing at 5 tsi.
Photomicrographs of specimens taken
from the samples after hot-rolling are
shown in Fig. 13.3.

302 ’
: STAINLESS

i SfiflfiET
R -mmz

 
  

302
STA*NLESS
STEEL - |
U0,

GCORE

 

i3l
ANP PROJECT QUARTERLY PROGRESS REPORT

Preparation of Tubular Fuel Elements.
Short segments of tubular geometry have
(1) cold-

into semi-

been prepared in two ways:
forming of flat plates

circles and joining with heliarc welds,
and (2) "rubberstatic" pressing of
fuel bearing mixtures 1inside a seamless
cladding tube.

A series of 12 specimens was pre-
pared, 11 by the first method and one
by the second, cold-drawn at Superior
Tube Company, and examined at QOak
Ridge National laboratory.

Three metal powders were used as
carriers for the UO, in the core —
302 stainless steel, nickel, and iron,
however, the clad-
type 316 stainless steel.
Metallographic examination revealed in
every specimen a tendency toward a

In every instance,
ding was

stringer type of UO, distribution,
varying from moderate in the 1iron
cores to severe 1n the 302 cores.
Cladding-to-core bonding was 1in most
instances satisfactory after the cold-
drawing. Further work will have to be
done to determine conditions of draw-
ing under which more acceptable uo,
distribution results. Twc of these
seamless-tube fuel elements formed by
"rubberstatic™ pressing at 40 tsi with
302 powder are shown in Fig. 13.4.

WELDING TECHNIQUES

P. Patriarca and G. M. Slaughter
Metallurgy Division

Cone Arc(3) welding. Cone-arc
apparatus which incorporates a fixed
magnet nozzle on an inert-arc welding
torch has been used with moderate
success to fabricate typical inconel
tube-to-header assemblies. The inconel

 

(D, R. Mann, Means for Meking Uniform
Circular Heliarc Welds by Deflecting the Ion
Beam Continucusly, ANP-63 (Apr. 9, 1951).

132

tubes were 0.188 1in. o.d. with a
0.025-in. wall thickness. Headers
were stamped from 0.0625-in.-thick

inconel sheet,

Although the feasibility of tube-
to-header welded construction was
demonstrated by these experiments,
results were somewhat inconsistent.
Apparatus has been completed which
incorporates a Selsyn stator as part
of the magnetic inert-arc welding
torch nozzle as described in a previous
report.¢?? Jt is felt that the proper
control of the additional welding
variables introduced by this method
will improve the consistency of
results.

An investigation of the effect of
these parameters on the type and
quality of welded tube-to-header
joints will be conducted concurrently
with fabrication of full-sized tube-
to-header assemblies for testing by
the Experimental Engineering Group.

Resistance Welding. Experiments
have been conducted using available
equipment which has demonstrated the
feasibilaty of joining tube bundle
straps to fuel tubes by spot welding.
A typical spot weld 1s presented in
Fig. 13.5 and shows a 0.016-in.-thick
inconel strap spot-welded toan 1inconel
tube 0.188-1n. o.d., 0.025 in. wall
thickness, It may be seen that
penetration and soundness of the spot
weld 1s excellent,

A preliminary investigation has
been conducted to determine the
feasibility of spot-welding clad fuel
elements together. The fuel elements
consisted of a mixture of UD, and 302
stainless steel powder. A sound spot
weld could be made which apparently
bonded the two sheets together without
any gross macroscopic movements of the
powder layers.

Fig. 13.6.

This 1s 1llustrated in
FOR PERIOD ENDING DECEMBER 10, 1951

‘, ~160
30% U0, +200
i - MESH

STAINLESS 4 7220

0% 302 {
STEEL
PRESSED AT
40781
SINTERED

12509¢

316 TUBE

- =100
30% U0, 3§ +200
. | MESH
70% 302 :_f 135

STAINLESS { ~22
STEEL | MESH

PRESSED AT
A0TS! |
SINTERED

1310°C

316 TUSE

 

{B)

Fig. 13.4. Seamless-Tube Fuel Eléments Formed by "Rubberst:atic'" Pressing.
{a) 2530x. (b) 175X. ~

133
ANP PROJECT QUARTERLY PROGRESS REPORT

 

Fig. 13.35. Transverse Section of a 9.815-in.«~Thick Inconel Sheetl
Spot Welded to an 0. 188-in.-0.d4. Inconel Tube of $.025-in. Wall
Thickness. FEtchant, aqua regia and glycerin. 175X,

SECRET 8

 

Fig. 13.6. Secticnef a Spot Weld Joining Two Stainless Steel—=—Clad
Fuel Plates. Etchant, agqua regia and glycerin. 60X,

134
FOR PERIOD ENDING DECEMBER 10, 1951

A 20-kva combination spot and pro-
~jection welder with suitable electronic
controls has been ordered for further
~experiments. It is expected that the
production of amultitude of resistance
" welds with consistent results will
' require precise mechanical and elec-
- tronic control, :

BRAZING TECHNIQUES

P. Patriarca and G. M. Slaughter
Metallunrgy Division

An introductory study has been
directed to the subject of furnmace
brazing under a controlled hydrogen
atmosphere. The use of a commercial
high-temperature brazing alloy
(Nicrobraz) has been investigated, and
the effects of some of the brazing
From the
it appears that Nicro-
brazing 1s an entirely feasible
technique of fabricating reactor
components. However, since Nicrobraz
contains on the order of 4% boron, the
use of this brazing alloy for fabri-
cation of reactor components 1is some-
~what limited. Tests performed using
a modified Nicrobraz in which beryllium
- was substithted for boron have shown
- little promise. The wetting properties
of this modified alloy were poor.
Further research i1s necessary to
completely evaluate thls approach to
the problem.

~ variables have been studied.
work to date,

Effect of Brazimg Time. The in-
fluence of brazing time on Nicrobraz
" joints made in type 316 stainless
steel can be seen in Fig. 13.7. The
- joint held for 5 min at the brazing
temperature shows somewhat less
diffusion of the brazing alloy 2nto
the base metal than that held at the

brazing temperature for 30 min, indi-
cating that time at temperature is not
critical within these limits. A

similar Nicrobrazed joint on 304

stainless steel was struck by a sharp
hammer blow at 1500°F. The high-
temperature ductlllty of the joint
was excellent.

Brazing of Ctad Fuel Elements. A
preliminary investigation has been
conducted on the brazing of clad fuel
elements., From the rather limited
amount of work that has been done on
this subject, 1t seems. entirely
feasible that these elements can be
brazed either to a stainless steel
sheet or to anether clad fuel element,
Figure 13.8 shows a clad fuel element
Nlcrobrazed toa stainless steel sheet.
As can be seen, the brazing alloy wets
the stainless steel very well and
apparently has little tendency to
float any of the fuel powders to the
surface. A Nicrobrazed edge-to-edge
joint of two clad fuel elements was
formed, and, although stainless steel
in this case 1s also wet very well by
the brazing alloy, the powder appeared
to be recessed in each sheet, This
probably resulted from the preliminary
grinding operation, since no brazing
alloy was observed in the recessed
area and the powder did not appear to
be contaminated by the brazing alloy.

Nickel-Palladium Brazing Alloy. In
another attempt to find a brazing alloy
with a more favorable nuclear ecross-
section, consideration has been given
to the nickel-palladium system with
minor additions of silicon for melting
point control. Figure 13.9 is a
photomicrograph of an inconel tube-to-
header joint using 40% nickel —60%
palladium alloy brazed at 1270°C in
dry hydrogen for 20 min. As can be
seen, the flow characteristics and
wetting properties are apparently
good. Room temperature tube-to-header
tensile strengths were of the order of
80,000 psi. Further consideration will
be given this alloy system pending
results of corrosion tests being con-
ducted in molten coolants and fuels,

135
ANP PROJECT QUARTERLY PROGRESS REPORT

136

 

E SECRET
| Y-4779

    

.-
s =,

T % N
‘ - /j?-:::?;‘,‘ ,<;,$:';;‘_:_,?
(b)

 

Fig. 13.7. Longitudinal Sections of a Type 316 Stainless
Steel Tube-to-Header Joint Nicrobrazed at 1120°C in a Dry
Hydrogen Atmosphere. Etchant, aqua regia and glycerin. 75X.
(a) Nicrobrazed for 5 min. (&) Nicrobrazed for 30 min.
FOR PERIGD ENDING DECEMBER 10, 1951

 

‘ Fig. 13.8.
‘Steel=-Clad Fuel-Element Butt Nicrobrazed to a Stain-

less Steel Sheet.
CREEP AND;STRESS RUPTURE OF METALS

R. B. Oliver G. M. Adamson
C. W. Weaver
Metallurgy Division ‘
The laboratories for creep and
stress-rupture tests and for the

Etchant, none.

Transverse Section of a Stainless

40X.

testing of metals immersed in liquid
media are essentially complete.
Extension measurements in the forwer
have been modified for greater accuracy
and recalibrated. A stress-rupture
curve has been obtained for both
"coarse' and "fine" grained inconel at

137
ANP PROJECT QUARTERLY PROGRESS REPORT

 

13. 9,

Fig.
with a 60% Pd - 40% Ni Alloy in Pry Hydrogen at 1270°C for 20 min.

aqua regia and glycerin. 75X,

815°C. Several elongation curves were
obtained for a variety of heat-treated
nickel Z (Duranickel) specimens,

Operation of Creep and Stress-
Rupture Equipment. The creep and
stress-rupture testing laboratory was
placed in operation without having
been properly instrumented in order
to train technicians and to avoid the
delay associated with the development
of an extension-measuring system,
During the past quarter it was neces-
sary to revise the existing measuring
system as 1t showed the creep of the
entire specimen, the threaded con-

138

Transverse Section of an Inconel Tube-to-Header Joint Brazed

Etchant,

nections, and pull rods, and the
seating of the knife edges of the
lever arm, as well as the differential
expansion of the entire assembly and
an effect from the sealing bellows.

The extension is now measured by
attaching a scribed platinum strip
extensometer across the gage length of
the speeimen. Marks are selected on
the center strip, and either of the
side strips and the distances between
the pairs of marks are measured with a
micrometer microscope having a least
scale division of 0.00005 in. The
Instrument Development Group has built
FOR PERIOD ENDING DECEMBER 10, 1951

and 1s testinga recording extensometer
designed around a motor-driven microme-
ter screw. This system was selected
because 1t has an optimum combination
of range, sensitivity, and stabality.

The lcading of all machines was
checked with a 65-mil sheet specimen
on which were mounted two SR-4 strain
gages connected in series. This
standard specimen im turn was cali-
brated on the Baldwin tensile-testing
machine using a Baldwin type L strain
indicator.

The effective area of the bellows

brated specimen in the furnace chamber:
then at several pull-rod locads the
change in specimen load was observed
for vacuum and several positive
pressures, The area so measured
averaged 3.47 in.?; for example, under
vacuum the bellows delivers a com-
pressive load of 51 1b,

Creep-Rupture Tests of Inconel.
Seventeen creep-rupture tests on
inconel sheet have been run or are in
process. These tests were made at
815°C in an argon atmosphere. Figure
13.10 18 a plot of the logarithm of
stress vs. the logarithm of rupture

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

was evaluated by enclosing the cali- time. The percentage figures placed
UNCLASSIFIED
DWG. Y-5304
e
(\
20,000 \\\\\\\
"~
N
~ FINE GRAIN
~ Nt
~ e
\"'--..
~ 23%“;\\\\
10,000 : N x\\\ A
I
‘:‘__" \\_ P e L e e
& ~. | K] 5% ,
@ R e seanl LTI YT 7
£ L I s
P S S
COARSE GRAIN — 15%
® 5,000 \/’ .-
< 3%
g ::xm\ I
g \\\\\,\ I
N X\
8% \ \..\
~ I~
\'\
e
2,000
1,000
10 100 1,000

Fig.

13. 10.

RUPTURE TIME (hr)

Stress-Rupture Time for Inconel Sheet.

139
ANP PROJECT QUARTERLY PROGRESS REPORT

on the chart are the elongations
determined from measurement of the
specimens before and after testing,
One set of points relates the stresses
and rupture times for specimens from
the annealed sheet and is indicated
as fine grained. The other set of
points refers to specimens that were
heated to 2050°F for 1 hr and air
cooled prior to testing;
1s designated as

this curve
coarse grained.
These grains were about 16 times
larger than the grains in the annealed
sheet,

Creep ©f Nickel Z. Five tests were
conducted on Nickel Z (Duramnickel)
sheet:
are presented

the elongation-vs.-time curves
in Fig. 13.11. The
elongations indicated were measured
with a dial gage on the pull rod and
include from several
These were conducted
at 815°C in argon atmosphere, and, as
indicated, several of the specimens

hence
sources.

erTrors
tests

 

were in the cold-rolled temper and
others were quenched and aged. It is
interesting to note that the effect of
heat treatment is retained at the
testing temperature,

Lack of
data for
materials at elevated temperatures
under nonoxidizing conditions has

Stress-Rupture Tests.
adequate stress-rupture

required conducting tests to provide
this information, Values obtained in
this laboratory 1n molten sodium,
fluorides, and other materials will be
compared with those obtained elsewhere
in vacuum and inert atmospheres.
Because many of the stresses found in
reactor auxiliary systems are hoop
stresses, these were the first type to
be investigated. Test specimens were
prepared by machining tube samples to
desired over a Z-1n.
section, welding one end closed,
applying helium pressure inside the
tube, and then iwmersing the tube 1n a

thickresses

UNCLASSIFIED
Y-5303

 

  

 

 

 

 

 

 

 

R, — E— E
| | | |
030 _
SPECIMEN DIMENSIONS: 0.024-in. THICK
- Y -in. X 3-1n, TEST SECTION —
. 815°C IN ARGON ]
o X - — AS RECEIVED - 2750 psi
I 024 f--- DEAD LOADED - LDAD ADJUSTED FOR CHANGES —
3 N BELLOWS SPRING LOAD
z ]
= —
<
s
£ 018 .
£
~ — ]
£
z x -
S / AS RECEIVED
2 o2 — J 1500 psi ]
o «—AS RECEIVED ~ 1950 psi QUENCHED AND AGED %ifl_f/,—"E'DBCONTWUED
5 - * / / 2750 psi e |
o A /___.A——“
/ . .@/O e ]
A}f::::::lk,,,—~fflé'“ QUENCHED AND AGED
rf‘__,a,@.—i:::f;afi 1950 psi
. ——/‘___;_/___,,_J_er o e
=t P DISCONTINUED |
° -e @
- J
400 600 800 1,000 1,200
TIME IN TEST (hr)
Fig. 13.11. Time-Elongation Curve for Nickel Z.

140
FOR PERIOD ENDING DECEMBER 10, 1951

bath of molten material for 1000 hr
or unti]l 1t ruptured.

Two tubes of type 316 stainless
stee]l lasted 1000 hr in 1500°F sodium
with hoop stresses of 1900 and 2600
psi, respectively.  Examination
revealed little or ne stress cor-
rosion or attack on either specimen.
On the other hand, two inconel samples
stressed at approximately 1500 psi
failed 1n less than 1000 hr, these
failures apparently resulting from
intergranular attack inside the tube
rather than from exterior sodium
attack. An inconel sample was later
prepared from a new lot of seamless
tubing and stressed to 1200 psi in
sodium; it failed in 465 hr through a
small area which seemed to be high in
impurities. Again, the rupture was
due to interior attack although
exterior attack up to 0.003 in., deep
was noted.

As a supplement to the above tests
a program for determining stress-
rupture values by applying a dead load
to flat thin (0.020-1in.) plates of
test materials immersed in sodium is
being carried out, Approximate values
obtained for inconel stressed at
13,000 and 10,000 psi and for type 316
stainless steel stressed at 7000 pSi
indicate that sodium at 1500°F has no
marked effect.

Both the tube-burst and stress-
rupture machines are undergoing modifi-
cation to permit use ofmolten fluorides
as bath materials., |

CERAMIC LABORATORY

T. N. McVay, Consultant
Metallurgy Division

Active work in placing the laboratory
in operation was started in August and

Pre-
liminary work on protective coatings
for the radiator of the ARE has been
initiated at Ohio State University.
lLaboratory facilities for carrying on
this work at OBNL are now virtually
complete,

1s now practically complete.

Since Cr-UQ, cermets are being
considered for fuel elements,
ment for processing and
cermets is being designed and should
be in operation within four months.
Some work has been done on the hot
pressing of alumina, and it is planned
to proceed with a study of a Al,0,-U0,
material for fuel elements.

eguip-
testing

Fquipsent. A vacuum dilatometer
and a Smith vacuum furnace for phase
studies has been completed, and the
dilatometer i1s in operation. Work has
started on a vacuum induction furnace,
and apparatus for tensile strength
tests at high temperatures has been
ordered. Egquipment for the hydro-
static pressing of ceramic bodies; a
high temperature X-ray camera, a
molybdenum wound tube furnace, and
specific heat and conductivity eguip-
ment are 1in the design stages.

Subcontract Work. A working agree-
ment has been negotiated between
Carbide and Carbon Chemicals Company
and the Electrotechnical Laboratory of
the U. S. Bureau of Mines, and work on
beryllia crucibles and other special
refractory shapes has started.

Thomas Shevlin, of the Ohio State
Engineering Experiment Station, is
working on Ni-BeO cermets for valve
seats to hold liquid metals. Also,
hot-pressed beryllia 1s available at
the Experiment Station for various
refractory shapes including valve
parts. |

141
ANP PROJECT QUARTERLY PROGRESS REPORT

14.

D. S. Billington,
A. J. Miller,

Studies have continued on the
effects of radiation on the physical
and chemical stability of the con-
stituents of the airplane reactor.
Considerable data have been obtained
from experiments in the X-10 graphite
pile, and emphasis has shifted to
preliminary experiments in the higher
flux LITR facility. 1In all phases of
the work preparations are being made
for further experimentation in the
LITR and tests in the MTR. Additional
radiation damage experiments have been
carried out with the Y-12 and Berkeley
cyclotrons.

The work has been concerned to a
large extent with the stability of the
fused fluoride salt mixture (NaF-KF-UF,,
46.5-26-27.5 mole %) proposed as fuel
for the sodium-cooled reactor. In the
X pile a series of runs was made in
which approximately 65 watts was
dissipated for 450 hr in each cubic
centimeter of U?3%_enriched fuel, With
the Y-12 cyclotron energy dissipations
up to 415 watts/cm® for 1 hr were
achieved with 20-Mev protons. In both
cases no evidence of radiation damage
was observed. Some radiation damage
appeared to take place in the single
preliminary experiment wade in the
LITR at about 1000 watts for 115 hr,
but a more exacting control run and
additional experiments are necessary
before a conclusion can be reached 1in
the matter.

The low uranium content of the
fused fluorides proposed for use 1n
the circulating-fuel reactor would
reduce the power production during
pile irradiation by a factor of 15 or
20. In this case only in the MTR and
cyclotron can power dissipations of
aircraft reactor intensity, about

142

RADTATION DAMAGE

Physics of Solids Institute

searc 1rector’s Jivision
He rch Director’s 1) 1

3000 watts/cm®, be achieved. Prepa-
rations for experiments with circu-
lating type fuels and materials with
related compositions are underway.

Additional information on the fuel
stability studies and i1nformation from
experiments on fused KOH stability,
thermal conductivity of metals, and
creep are contained in the following
sections of this report, Complete
details onall radiation-damage studies
are contained 1n the Physics of Solids
Institute quarterly report for period
ending October 31, 1951.

IRRADIATION OF FUSED MATERIALS

G. W. Keilholtz, Materials

Chemistry Division

The effects of radiation on the
stability of the fused fluorides and on
fused KOH have been under investigation
using the X pile, the Y-12 cyclotron,
and the LITR. In experiments with fused
fluorides in inconel at 1470 to 1500°F
no evidence of radiation damage was
found in the X-pile or cyclotron run.
Some evidence of what appears to be
radiation damage was found in the
single preliminary LITR experiment,
which still requires control and check
runs, Data from the single experiment
on escape of xenon from the melt was
inconclusive, Pressure build-up
measurements on capsules of fused KOH
in the X pile and LITR indicated no
instability to radiation,

Pile Irradiation of Fuel and KO
capsules (J. G. Morgan, H. E. Robertson,
C. C. Webster, P. R. Klein, and B.
Kinyon, Physics of Solids Tnstitute).
For reasons of safety, capsules con-
taining the fused materials were first
FOR PERIOD ENDING DECEMBER 10, 1951

checked in the pile for pressure
evolution at low fluxes, and then at
higher fluxes. The capsules were
pressurized with helium and the
pressures were continuously measured
by means of a strain gage to an
accuracy of #0.5 psi. The standard
NaF-KF-UF, fuel and KOH were tested as
shown in Table 14.1 with no evidence
of pressure increases due to irradi-
ation.

When pressure tests on the fused
fluoride mixture were completed in the
X pile, 0.5 g samples of the material
enriched in U??® were subjected to a
thermal flux of 10'? at 1472°F wall
temperature in sealed inconel capsules.
After several preliminary tests, five
capsules of 0.223 1.d. were irradiated
for 300 to 450 hr at 1472°F and one at
pile ambient temperature., Chemical
and metallographic analyses, which are
mostly completed, have showed no in-
crease of iron, chromium, or nickel in
the fuel, decomposition of the fuel,
or observable effect on the inconel as
compared to bench tests. The capsule
in Table 14.1, which was used in the
LITR pressure test, with a power
dissipation of about 1000 watts/cm?,
showed increased damage in all three

Table

Tests on Standard NaF-KF-UF, and KOH

respects. However, a control sample
prepared simultanecusly is yet to be

bench-tested and examined.

In order to determine 1f xenon is
evolved from the fluoride melt under
irradiation, the following experiment
was conducted. A sample of enriched
urenium metal was placed in a flux of
10'? neutrons/cm® and irradiated for
the same length of time as a sealed
microcapsule of melted salt. Both
were removed and the microcapsule was
opened., When they were counted, no
significant difference between the
twe was noted, which seemed to indi-
cate that the xenon formed did not
escape from the melt. This test is
not considered conclusive and a more
significant experiment is planned.

Cyclotrom Irradiation of Fuel and
KOH Capsules (W. J. Sturm and M. J.
Feldman, Physics of Solids Institute;
R, J. Jones, J. S. Luce, and C. L.
Viar, Electromagnetic Research Di-
vision). Sixteen inconel pins, 0.052
and 0.100 in., i.d., containing the
standard fuel were bombarded with 20-
Mev protens. Thermocouples on the
irradiated surface of the pins indi-
cated average run temperatures between

14.1

 

 

 

MATERIAL FREE SPACE APPROXIMATE |
WEIGHT | CAPSULE [CAPSULE | TOTAL | MATERTAL | THERMAL FLUX | TIME
MATERI AL (g) | MATERIAL | (em®) | (ew®) | TEMP. (°F) | PILE | (n/cm® - sec) | (hr)
Fael
Normal uranium 10 Inconel 6 17 200-1472 (X pile| 8.5 x 10! 12
93. 4% U335 10 Inconel 6 17 1472-1560 | X pile| 8.5 x 10 {170
0.5 Inconel 2 13 1370-1500 |LITR 1.6 x 10*3 | 116
KOH 5 316 Stain- 6 17 805 X pile| 8.5 x 10! 16
less steel :
3 Inconel 3 13 805 LITR 1.6 x 103 68

 

 

 

 

 

 

 

 

 

143
ANP PROJECT QUARTERLY PROGRESS REPORT

1200 and 1475°F. Pins irradiated with
low beam currents were usually radi-
ation-cooled, while those irradiated
with high currents were attached to a
cold-water coil. In the case of the
water-cooled pins there was probably a
layer of solid fuel at the water-
cooled face., Five of the 1nconel
cases melted owing to local over-
heating, which probably occurred in
all pins to some extent. Energy
dissipations between 30 and 415 watts
per cubic centimeter of fuel were
achieved in various runs, usually of
1 hr duration. Analyses of the fuel
and metallographic examinations of the
capsules indicated no radiation-
induced damage when comparison was
made with the from eight
bench-tested Several
irradiations by deuterons of fuel 1in
inconel containers have been made by
the North American Aviation group with
the Berkeley 60-in. cyclotron. The
results are being analyzed.

resnlts
controls.

Stainless steel pins contalning
fused KOH have been irradiated with
protons, and analyses of the results
are in progress. Capsules containing
dilute circulating type fuel are being
prepared,

IN-PILE CIRCULATING LOOPS

W. E. Brundage
R. M, Carroll

C. D. Baumaan

C. Ellis
Physics of Solids Institute

0. Sisman
W. W. Parkinson
A. S, Olson

The general results of circulating
lithium at 1000°F through hole 58N of
the X pile were reported last quar-
ter.¢!) A detailed metallographic

 

(D¢, b Baumann, R. M. Carroll, Q. Sisman,
W. W. Parkinson, and . Fllis, “*Liquid Metals
In-Pile Loop,” Aircraft Nuclear Propulsion Project
Quarterly Progress Report for Period Ending
gs%gsnbcr 10, 1951, OBNL-1154, p. 174 (Dec. 17,

144

examination now has beena made of
various sections of the 316 stainless
steel loop. No evidence was found of
increased intergranular penetration or
other damage due to radiation,

An inconel loop for circulating
sodium in the X pile at 1500°F has
been designed and partially con-
structed., Design of equipment 1is
nearing cowpletion for stress cor-
rosion and creep tests on inconel 1in
circulating sodium at 1500°F in the
LITR.

CREEP UNDER IRRABIATION
J. C. Wilson J. C. Zukas
W. W. Davis

Physics of Solids Institute

Work during last quarter(?) on
cantilever creep showed that X-pile
radiation at a flux of 4 x 10!° fast
neutrons/cm? caused an increase 1in
total creep strain of about 20% 1in
347 stainless steel after about 250 hr
of exposure, the duration of the tests.,
Extrapolation of the bench and in-
pile curves to longer times indicated
that the difference between them in-
The first of the
current tests should supplement the
above observations. The time has been
extended to 500 hr under the same
conditions (1500°F at 1500 psi), but,
because the expected beam deflection
would exceed the range of the micro-
former, the transducer has been
omitted and fiducial marks on the
loading beam and baseplate will permit

creased with time.

measurement of the total extension
after the experiment has been with-
drawn from the reactor. The deflection
will then be compared with that
observed in a bench test at the same

 

(2)J. C. Wilson, J. C. Zukas, and W. W. Davis,
‘Creep Under Irradiation,”’ ORNL-1154, op. cit.,
p. 170.
FOR PERIOD ENDING DECEMBER 190,

temperature and stress levels. The
irradiation has been completed, and
the activity of the apparatus is being
permitted to decay to a safe level for
handlaing. Metallographic and X-ray
comparison of bench and irradiated
samples will be made.

A second test is still in the pile.
It was stressed to B8.000 psi and for
its first 500 hr has been cperating at
1200°F. The temperature will be
subsequently raised in steps to observe
the temperature dependence of strain
rate under irradiation for comparison
with a companion bench test. Prepa-
rations for tensile type tests and for
experiments in the LITR and MTR are in
progress.

 

(A, F. Cobhen, *“Radiation Effects on Thermal
Copductivity,” ORNL 1154, op. cit., p. 171.

1951

RADIATION EFFECTS ON THERMAL
CONDUCTIVITY

A. Foner Cohen, Physics of
- Solids Institute

In a preliminary relative thermal
conductivity experiment on inconel,
reported last quarter,(3) a large
decrease 1in thermal conductivity was
observed upon three davysofirradiation
at approximately B825°C in the X pile.
To check this result, a carefully
heat-treated inconel specimen has been
irradiated at 575 and 250°C with ne
apparent effects due to the radiation,
Testing at 825°C is 1n progress.

In addition to the above experiment,
an absolute measurement of the thermal
conductivity of a 316 stainless steel
specimen at 100 to 200°(C is being
carried out in the LITR,

145
 
SUMMARY AND

The survey of the supercritical-
water reactor by the Oak Ridge National
Laboratory’s subcontractor Nuclear
Develepments Associates, Inc., is
essentially complete and has confirmed
the feasibility of aircraft nuclear
propulsion using that reactor. Although
ORNL will not further pursue this
reactor cycle because of the prevailing
belief in the greater potentialities
of low-pressure liguid-coolant re-
actors, NDA will continue its work,
together with Pratt and Whitney, under
direct AEC contract. The new results
in the NDA survey of the cycle for
ORNL are outlined in Sec. 15, together
with a brief cycle analysis that was
undertaken at ORNL.

Alarge analytical chemistry program
is required to support the materials
research program now being undertaken
by the ANP project. This includes not
only routine service analysis ~~ 688
individual samples were analyzed during
the past quarter = but 1n many cases
the development of new analytical
procedures. This development, as
reported in Sec. 16, is largely con-
cerned with the determination of
oxygen in gases and fused mixtures and
similar studies of hydroxides and
fluorides. '

INTRODUCTION

The "List of Reports Issued,” Sec.
17, includes seven laboratory reports
and 39 informal documents on ‘all
phases of the ANP project.

A directory of the research preojects
of the Aircraft Nuclear Propulsion
Project of the Oak Ridge National
Laboratory is given in Sec. 18. The
research projects of the Laboratory’s
subcontractors to its ANP project are
iisted, as well as the research now
in progress at the lLaboratory. In
additzon, such research as is being
performed by OBNL for the ANP programs
of other organizations 1is included
and marked as such.

A chart of the technical organi-
zation of the Aircraft Nuclear Propul-
sion Project at OBNL is included
(Sec. 19) to identify the personnel
and emphasis associated with the
various phases of the project. There
are now about 2ZB5 scientific and
technical people employed on the
project and about 26 active consultants,
Although contractural changes at the
start of this fiscal year shifted
several former Laboratory subcontrac-
tors to the AEC, there are, in all,
seven allied laboratories performing
research for the ANP project, where
ORNL has either the direct contract
or technical supervision.

149
ANP PROJECT QUARTERLY PROGRESS REPORT

15.

For the past year (see p, 1v for
list of previous reports) Nuclear
Development Associates have been
studying the supercritical-water
reactor first proposed in Wash-24.,(1)
A first survey analysis by NDA 1s
essentially complete and a final
report is being written. This report,
when 1ssued, will complete the sub-
contract work by NDA for OBNL on the
supercritical-water reactor. NDA will,
however, continue i1ts work in this
field under subcontract with Pratt and
Whitney Aircraft Division,

In addition to the above-mentioned
survey of the supercritical-water
reactor, an analysis of the cycle was
undertaken at OBRNL. This analysis,
which will also soon be 1ssued as a

separate report, is summarized below.

ANALYSIS OF
WATER

SUPERCRITICAL
REACTOR BY NDA

The only new results since the last
quarterly report pertaining to the
supercritical-water reactor system are
in the control aspects. The reactor
has been shown to be stable for small
oscillations, and a start-up technique
has been outlined in which the control
mechanism is derived from change of
water density in the reactor.

Stability. A calculation has been
carried out¢?) on the stability of
the design point reactor (400,000 kw,
2.5-ft square cylinder core, flux flat
radially) connected between constant-
pressure reservoirs, The system was
studied with equilibrium xenon con-
centration, Under the assumptions
made, the stabilizing density effects

 

(1)Aircraft Reactor Branch of USAEC, Appli-
cation of a Water Cooled and Moderated Reactor to
Aircraft Propulsion, Wash-24 (Aug. 18, 1950).

(2)NpaA Quarterly Report on ANP Activities,
June 1 to August 31, 1951, Y-¥5-55.

150

SUPERCRITICAL-WATER REACTOR

on reactivity were found to compensate
the unstabilizing effect of the xenon;
the reactor i1s stable for swmall
oscillations., The longest two decay
periods were calculated as 5.9 min and
22 sec.

The response to a sudden step-
function i1ncrease in reactivity 1is
such as to cause the power level to
overshoot 1ts equilibrium increase by
a factor of 10. When the reactivity
is raised linearly during an interval
of 2 sec, the overshoot is by a factor
of 2; when 10 sec is taken to accomplish
the linear reactivity increase, there
is no overshoot.

Start-up. The start-up problem has
been considered in a preliminary way
with the reactor connected to a simple
heat-exchanger external systemand with
no control mechanism other than that
afforded by changes of water density
in the machine.

The envisaged start-up procedure 1is
as follows: (1) Fill the system with
high-temperature high-pressure water
whose density i1s too low to make the
machine reactive; (2) keeping the
system at constant pressure, clrculate
and slowly cool the water; the cooling
rate is adjusted(3®) so as to bring the
reactor power safely up to some
fraction (say, 3%) of rated output,
the circulation rate being large enough
so that the water density is still
substantially uniform throughout the
reactor; (3) adjust the cooling rate
so as to hold the reactor power con-
stant at this 3% level, and slowly
reduce the circulation rates through
the moderator chamber and fuel tubes,
while the density variation along the
fuel tubes builds up to the normal
operating pattern,

 

(3)After the nmeutron flux has become suf-
ficiently strong, the flux can be employed for
automatic control of this adjustment.
FOR PERIOD ENDING DECEMBER 10, 1951

Questions concerning the external
equipment called for and how and when
the reactor can be switched to connect
to the power plant system have not
been 1nvestigated.

ANALYSIS OF SUPERCRITICAL-WATER
REACTOR BY ORNL

One of the most widely advocated
methods ofutilizing the supercritical-
water reactor for aircraft nuclear
propulsion is to use the compressor-
jet propulsion cycle. A brief study
of this fundamental cycle to discover
the relations existing among the
various parameters involved 1s de-
scribed,

Compressor-Jet Cycle. In this
cycle energy is added to the airstream
in a low compression ratio blower and
then in a radiator. The supercritical
steam from the reactor expands through
a turbine which operates the air
blower and the water-return pump. The
turbine exhaust steam is then con-
densed, and 1t transfers its heat of
vaporization to the airstream flowing
through the radiator.

The particular parameters that
have been considered - condenser air-
inlet face area, condenser weight, and
were evaluated per
terms of the
temperature and pressure:of the steam
leaving the reactor and the condenser

airflow rate -—
pound of thrust 1in

pressure. This was accomplished for
flight at 45,000 ft altitude at Mach
numbers of 0.9 and 1.5.

Reasonable values were assumed for
component efficiencies and actual test
data used for the condenser perfor-
mance. The results should give
attainable performance of the funda-
mental cycle at the two design points
investigated. Somewhat better per-
formance can probably be obtained
through a program of optimization of
the cycle and equipment., However, the
data presented here should give per-

formance not far from the optimum, and
hence should be useful for preliminary
design studies of the system.

Results. The results of this in-
vestigation indicate that for reactor
outlet steam conditions ranging from
1000 to 1500°F and pressures from
5000 to 10,000 psi, the following

statements can be made:

1. The specific impulse is low 1in
all cases (from 15 to 20 1b per
pound of air per second at a
flight Mach No. of 0.9 and from
9 to 14 1b per pound of air per
second at a flight Mach No. of
1.5).

2. Increasing the reactor ocutlet
steam temperature or pressure or
condenser pressure effects some
improvement in all cases and in
all parameters considered, 1.e.,
specific impulse, specific heat
consumption, and specific
radiator weight and frontal
area.

3. Increasing steam condenser

pressures above 400 psi1 gives

relatively littlé improvement
in performance.

reactor steam
from 5000 to
relatively

4. Increasing the
outlet pressures

10,000 ps1 gives
litctle improvement in perfor-
mance.

5. Cycle performance is insensitive

to the relative amounts of energy
put into the air by the turbine-
compressor and by the condenser.

In general, the obvious advantages
of this cycle are the use of water, a
familiar and fairly noncorrosive
substance, for both coolant and
moderator., The disadvantages include
an inherently low specific impulse and
the necessity of developing an entirely
new type of aircraft engine.

151
ANP PROJECT QUARTERLY PROGRESS REPORT

16.

ANALYTICAL CHEMISTRY

C. D. Susano, Analytical Chemistry Division

Since the use of diatomaceous earth
as an insulating agent in the aircraft
reactor is being contemplated, studies
are presently being made of the compo-
sition of these materials with partic-
ular attention to elements of high
neutron-capture cross-section, of which
boron and the rare earths are the most
important. Preliminary results on two
samples of diatomaceous earth indicate
that these materials may contain
several hundred parts per million of
boron, which i1s higher than can be
tolerated.

It has been mecessary to develop or
adapt methods for the determination of
metallic nickel, nickel oxide (Ni0),
and available oxygen in alkali hydrox-
ide melts in connection with tests in
which these hydroxides are circulated
through nickel loops at 1400°F. These
methods appear to be satisfactory with
the exception that, if the nickel
occurs 1n a massive crystalline form,
difficulty is encountered in separating
it completely from the oxide.

As a part of the corrosion test
program for the evaluation of fluoride
eutectic coolants, methods are under
study for the determination of iron,
nickel, chromium, molybdenum, and
copper in these materials. Methods for
the precise determination of the major
constituents are also under consider-
ation. A study of the pH values of
aqueous solutions of one of the alkali
fluoride cutectics and its components
is reported.

Tests were made to determine whether
or not alkali and alkaline earth
hydroxide melts could be removed from
metal tubes by dissolution in water
without appreciable corrosion of the
container material during the disso-
lution step. Monel, inconel, nickel,

152

and stainless steels 316 and 347 were
attacked to only a negligible degree
and copper was corroded only slightly
more.

The development of two methods for
the determination of uranium trifluo-
ride has been completed. It appears
that the method involving measurement
of the hydrogen evolved on reaction of
of the UF, with acid is more precise
than the method which depends upon the
total oxidizing power of the sample.
Although some additional work remains
to be done, 1t appears that oxygen can
be determined in sodium-potassium alloy
(NaK)} by a modification of the n-butyl
bromide method for the determination
of oxygen in sodium.

A survey of the results obtained 1in
the determination of oxygen in tanks
of Burcanof Mines helium 1s presented.
Approximately 85% of the cylinders
tested contained less than 25 ppm of
oxygen and were, therefore, acceptable
for use by the ANP Experimental
Engineering Group.

Tests are being made for the purpose
of producaing a borated water solution
(1% boron) at minimum cost which will
retain 1ts clarity 1n contact with
concrete, 1nconel, and 1iron.

Summaries of the service analysis
work, which indicate the distribution
with respect to sample type and groups
originating the samples, are presented.

STUDIES OF DIATOMACEOUS EARTH

J. C. White and W. J. Ross
Analytical Chemistry Division

The use of diatomaceocus earth as
an insulating agent in the aircraft
FOR PERIOD ENDING DECEMBER 10, 1951

reactor 1s contemplated, and studies
are being made of the composition of
these materials with particular atten-
tion to elements of high cross-section
for thermal-neutron capture, chiefly
boron and the rare earths.

Diatomaceous earth, which consists
almost entirely of the siliceous
skeletons of minute marine animals,
is composed essentially of 5i0, but
may vary significantly in minor and
trace constituents. An effort is being
made to find an infusorial earth of
suitably low boron content (less than
10 ppm). '

Spectrographic analysis of two
types has been completed todate: one,
a type used primarily as a paint
pigment, showed 600 ppm of boron and
no detectable rare earths; the second,
a "fresh-water™ earth with the trade
name "Sil-o-cel,” showed 300 ppm of
boron and ne detectable rare earths.
Since these concentrations are higher
than can be tolerated, the possibility
of reducing the boron content of these
materials by washing with hydrochloric
acid and other solutions is under
investigation.

A new method for the determination
of boron is under study.(!’ This
method depends upon the color complex
between boron and 1,1°-dianthrimide
which is formed in sulfuric acid 1n
aqueous medium. Preliminary tests
indicate that the method is extremely
sensitive (0.0l ug of boron per milli-
liter).

DETERMINATION OF Ni, NiO, AND O
IN ALKALI HYDROXIDES
J. C. White
Analytical Chemistry Division

A series of tests has been conducted
in which lithium, sodium, and potassium

 

(1)G, M. Ellis, E. G. Zook, and O. Baudisch,
“"Colorimetric Determination of Boron Using 1,
1’-Dianthrimide,” Anal. Chem. 21, 1345 (1949).

hydroxides were circulated through
nickel loops at 1400°F. The loops
were operated for various periods,
ranging from 51 hr for KOH to roughly
300 hr for LiOH, before the runs were
terminated because of the formation of
plugs. The loops were X-rayed in order
to determine the locationof the plugs,
cut in sections, and submitted for both
chemical and metallographic exami-~
nation. Some development effort was
necessary to arrive at suitable methods
for the determinationof nickel, nickel
oxide, and available oxygen in these
plugs.

Available Oxygen. A method similar
to that employed for the determination
of available oxygen in lead peroxide
is used in this case. The amount of
iodine liberated from an acidic iodide
solution is titrated with standard
sodium thiosulfate. 1In only one case
was an appreciable amount of available
oxygen found in the samples submitted.

Metallic Nickel. Since the plugs
were formed from large, lustrous nickel
dendrites, metallic nickel was deter-
mined directly. Samples from other
sections of the loops were characterized
by the presence of much more finely
divided nickel, and i1n these cases the
displacement technique was applicable.
This method depends upon the displace-
ment from solution of a metal lower
than nickel in the electromotive
series. The excess of the added 5%
copper sulfate reagent can be deter-
mined and is a measure of the amount
of metallic nickel originally present.

Nickel Oxide. Work is underway to
separate nickel and nickel oxide when
the metal exists largely in a massive
crystalline form. The use of i1odine
to convert the metal to Nil, will be
investigated. In samples 1in which the
nickel 1s more finely divided, the
oxide is separated out by filtering
the inscluble nickel oxide and copper
metal from the solution used for the

153
ANP PROJECT QUARTERLY PROGRESS REPORT

determination of nickel by the dis-
pl acement method. The residue is
digested with acid, and the nickel 1is
determined by the dimethylglyoxime
precipitation.

Studies of the higher oxides of
nickel are being made in an attempt
to gain an understanding of the
reactions which are taking place 1n
the molten hydroxides under the
conditions of these tests.

STUDIES OF TERNARY ALKALI FLUORIDE
EUTECTIC

J. C. White
Analytical Chemistry Division

A proposed coolant for the ARE
reactor is the ternary alkali fluoride
eutectic composed of 11.7% sodium
fluoride, 59.1% potassium fluoride,
and 29.2% lithium fluoride by weight.
This eutectic melts at 450°C. VWork
on this eutectic is confined at present
to study of the pH of aqueous solutions
of the eutectic and its components,
of the composition of the eutectic,
and analysis of the eutectic for
metallic impurities.

pll of Agueous Solutions of Alkali
Fluorides. The alkali fluorides will
hydrolize completely at temperatures

Table 16.1

pH Values of Aquecus Soclutions of
Certain Alkali Fluorides at 25°C

 

 

 

 

MOLAR

SALT CONCENTRATION pH
LiF 0.05 5.3-5.4
LiF (fused) 0.05 8.3-8 .4
NaF 1.0 7.2-7.3
NaF (fused) 1.0 8.9
KF 7.95 9.2
KF (fused) 7.95 9.9

 

154

around 1000°C but not significantly
at room temperature. Hence, a deter-
mination of the pH of an agueous
solution of the cooled fused salt at
room temperature will indicate the
degree of hydrolysis which has taken
place at the high temperature required
for fusion (pyrohydrolysis) and is a
me asure of the moisture present at
these temperatures. Table 16.1 shows
typical values for the higher pH of
the fused salts as compared to com-~
parable concentrations of unheated
c.p. reagent grade salts. The following
comments may be made:

1. The increase inbasicity on fusion
is very likely a consequence of
pyrohydrolysis, the water being
already present in the salt.
Potassium hydroxide, despite its
hygroscopic natuvre, appeared
more stable to pyrohydrolysais
than the other alkali metal
fluorides. Agueous solutions
of this fluoride are more basic,
however, than solutions of
either sodium or lithium fluoride.

2. The alkali fluorides may be
classified as salts of strong
bases and fairly weak acids (the
ionization constant of hydroflu-
ric acid is 7.7 x 10°*), so that
aqueous solutions of these salts
should be slightly basic. How-
ever, aqueous lithium fluoride
tested acidic and sodium fluo-
ride nearly neutral. This 1s
believed due to the presence of
some free hydrofluoric acid or
acid fluoride salt in the reagent.

3. pH is relatively independent of
concentration except possibly in
the case of unfused anhydrous
potassium fluoride, the reason
for this 1s under current investi-
gation.

The pH of a saturated solution of
the eutectic, determined from three
FOR PERIOD ENDING DECEMBER 10, 1951

lots prepared in various container
materials, was 9.8, roughly that of a
solution of fused potassium fluoride.
A fourth lot, prepared using hydrated
potassium salt dried overnight under
vacuum, had a pH of 6.9, which is
nearly neutral. Evidently some
basicity may be removed by this drying
process. '

Composition of the Eutectic. Pre-
liminary studies have been made to
compare gravimetric and flame-photo-
metric methods for the determination
of i1ndividual alkali metals in the
eutectic. It appears that flame-
photometric methods will not be suffi-
ciently accurate but that total alkali
and total fluoride determinations will
provide all the information required.
Total alkali can be determined gravi-
metrically and total flunoride will be
determined by pyrohydrolysis.

Metallic Impurities. As a part of
the corrosion test program, the trace
impurities iron, nickel, chromium,
molybdenum, and copper will be deter-
mined in the eutectic. Colorimetrac
methods are being developed for this
purpose.

CORROSION OF METAL CONTAINERS BRY
HYDROXIDE SOLUTIONS

J. C. White
Analytical Chemistry Division

The present method for removing
alkali and alkaline earth metal
hydroxides from metal containers in
preparation for the determination of
metal constituents involves dissolution
of the hydroxide melt with water. This
procedure exposes the container to a
warm, saturated hydroxide solution
for as long as 2 to 3 hr in some cases;
hence, a study was made to determine the
extent of corrosion of the container
metal from this source. Results
indicated that this procedure has

little effect, and that the present
method 1s suitable for all containers
with the possible exception of copper.

DETERMINATION OF URANIUM TRIFLUORIDE

W. K. Miller and D. L. Manning
Analytical Chemistry Division

In a previous report{?’ two methods
for the determination of uranium tri-
fluoride were described. It was
reported that the ceric sulfatemethod
appeared more promising than the
hydrogen evolution method, although it
was believed that the precision of the
latter could be improved by reducing
the volume of the apparatus. This
modification has resulted in a marked
improvement in the results by the
hydrogen evolution method, which 1is
now considered more precise. Although
no standard sample 1s available for
a direct test of the two methods, the
agreement between the methods indicated
that sither can be satisfactorily used
to determine trivalent uranium.

DETERMINATION OF OXYGEN IN Nak

J. C. White and W. J. Hoss
Analytical Chemistry Division

Efforts are currently being directed
toward the application of the n-butyl
bromide method(?? for the determination
¢f oxvygen in sodium to the determi-
nation of oxygen in sodium-potassium
alloy (NaK), a eutectic composed of
22% sodium and 78% potassium. The
principal physical and chemical
properties of the alloy which are of

 

(D%, K. Miller and D. L. Manning, “Uranium
Trifluoride in Uranium Tetrafluoride,” Aircraft
Nuclear Propulsion Project Quarterly Progress
Report for Period Ending September 10, 1951,
ORRL- 1154 p. 202 (Dee. 17° 1981)

(3)5. C. White and W. J. Ross, “Determination
of Oxygen imn Sodium,” Aircraft Nuclear Propulsion
Project Quarterly Progress Report forPeriod Ending
March 10, 1951, ANP-60, p. 336 (June 19, 1951)

155
ANP PROJECT QUARTERLY PROGRESS REPORT

significance in this particular study
are: (1) It is a free-flowing liquid
above 15°C, (2) it has a rather high
coefficient of expansion upon freezing,
and (3) it is extremely reactive
toward oxygen. Owing to the high
reactivity of NaK with n-butyl bromide,
the reagent must be added in small
increments rather thanmn in one single
portion. The reaction 1s complete
within 1 hr for 1- to 2-g samples of
NaK.

The results obtained on test samples
have shown somewhat higher oxygen
content (about 0.25%) than has usually
been the case for sodium samples.

DETERMINATION OF OXYGEN IN HELIUHX

J. C. White and W. J. BRoss
Analytical Chemistry Division

Data for the determination of
oxygen 1in helium by the Brady method
have shown that about 85% of the United
States Bureau of Mines cylinders
tested have oxygen contents below
25 ppm, and hence, according to the
present standards, are acceptable for
use. No attempt was made to determine
the average oxygen content, as this
method 1s unreliable for concentrations
greater than 50 ppm.

Because of 1ts wider working range,
a modification of the Winkler method,
as modified by Pepkowitz,¢*’ has been
considered as a substitute for the
Brady method. It has been i1mpossible
to determine the precision of this
method since acceptable low- or zero-
oxygen standards have not been obtained.

PREPARATION OF OXYGEN-FREE
SODIUM SAMPLES

H. R. Bronstein, ANP Division

Widely varying results on sodium
samples taken from operating systems

156

pointed out the unreliability of
analytical methods for determining
quantitatively the oxygen in sodium.
Unreliability was felt to be largely
due to lack of suitable standards by
which analytical methods could be
evaluated. Consequently, a search was
begun to find suitable methods for
producing samples containing a known
amount of oxygen.

A literature search revealed a
method believed suitable for producing
standard samples. DBriefly, the method
consists in 1mmersing a highly evacuated
glass bulb containing a filament in a
bath of molten sodium nitrate. An
electrode (anode) is also immersed in
the bath external to the evacuated
bulb. The filament i1s heated by a
220-v alternating current; at the same
time, a 220-v direct current is placed
on the filament and anode. Sodium ions
are pulled out of the glass, neutralized
by electrons from the filament, and
replaced in the glass by sodium 1ons
from the bath. This method allows
pure sodium to be plated out on the
inner side of the bulb.
deposited may be determined by either
weighing the bulb before and after
electrolysis or calculated by Faraday’s
law of electrolysis. Standard samples
may be produced by adding known
amounts of oxygen to known amounts of
sodium by high-vacuum techniques.

Amounts

During the guarter equipment was
assembled, and preliminary experiments
were conducted to evaluate the equip-
ment and method. Both appear to be
adequate, and sufficient experience
has been gained to proceed with
preparing actual samples.(®’

 

L. p. Pepkowitz, private communication.

)T, E. Willmarth, “Microscopic Study of a
Submitted Sample of Diatomaceous Earth,”” ORNL
CF-51-11-12 (Nov. 1, 1951).
- FOR PERIOD ENDING DECEMBER 10, 1951

CLARITY OF BORATED WATER IN
CONCRETE TANKS(6)

H. P. House
Analytical Chemistry Division

It has been proposed that the con-
taining concrete tank of the ARE be
filled with borated water (at least
1% boron by weight) after shutdown,
while subsequent disassembly operations
are performed remotely. Hence, several
solutions of boron salts have been
tested for prolonged clarity while in
contact with concrete. Saturated
borax solution met the requirement for
clarity, but contained only 0.8% boron
by weight. Potassium tetraborate met
both requirements, but its higher cost
discourages its use. Mixtures cf the
two salts in agueous solution form a
precipitate on contact with concrete.
Final studies are being made te
determine the optimum borated solution.

 

(S)H. P, House and €. D. Susano, Clarity of
Borated Water in Concrete Tanks, CRCCC Y-12 Memo
¥B-31-273 (July 24, 1951).

ANALYTICAL SERVICES

H. P. House J. W. Robinson
L. J. Brady

The bulk of the analytical work
carried out during the quarter for the
ANP Project consisted of determinations
of (1) corrosion products in reactor
fuels (fluoride salt mixtures), (2)
purity of components used in compounding
reactor fuels and for coolants, {(3)
corrosion products in sodium, potas-
sium, and barium hydroxides, and (4)
oxygen and corrosion products in sodium

and NakK.

A summary of service analyses per-

formed this quarter is shown in Table
16.2.

Table 16.2

Summary of Service Apaiyses

Samples on hand 8/10/51 151
No. of samples received 686
Total no. of samples 837
No. of samples reported 688
Backlog as of 11/2/51 149

157
ANP PROJECT QUARTERLY PROGRESS REPORT

17. LIST OF REPORTS ISSUED
REPORT NO. TITLE OF REPORT AUTHOR(S) DATE ISSUED
Reactor Physics

CF 51-10-83 Enlargement of Cross-section Program C. E. Larson 10-12-51

CF 51-11-92 Pile Simulator Study of Flux, etc. S. Hanauer 11-12-51

Y-F10-73 Suggested Correction to Age-Diffusion R. R. Coveyou 11-6-51
Equation as Used by the ANP Physics Group B. T. Macauley

Y-F10-69 Numerical Integration of Differential R. R. Coveyou 8-20-51
Equations; Multi-Point Boundary Problems

Y-F10-71 Physics Calculations on the ARE Control Rods R. J. Beeley 8-29-51

OBNL- 1099 The Elements of Nuclear Reactor Theory, S. Glasstone No date
Part [ M. C. Edlund

Y- B4- 39 Nuclear Properties of U?3*. A Literature E. P. Carter 9-12-51
Search

Y-F10-64 Heating in the B,C Curtain Due to Neutron C. B. Mills B8-16-51
Absorption and the B!%(n,a)Li” Reaction

Y-F10-67 Effect on Reactivity of ABE of Flooding J. W, Webster §-14-51
Coolant Channels with Borated Water

ANP-68 Solution of Kinetic Equations of Cylindrical M. J. Nielsen 9-18-51
Reactor J. W. Webster

Y-F10-74 Note on the Doppler Effect R. R. Coveyou 11-26-51

Y-F10-56 Some Results of Criticality Calculations on J. W. Webster 10-15-51
BeQ and Be Moderated Reactors 0. A. Schulze

Shielding Research

ORNL-1130 Analysis of Bulk Shielding Facility Neutron S. Podgor 11-26-51
Dosimeter Data

CF 51-9-112 Power Calculations of the Unit Shield Reactor E. B. Johnson 9-18-51

CF 51-10-70 Introduction to Shield Design E. P. Blizard 10-12-51

CF 51-10-94 Calculations of Leakage from the Bulk Shield- J. L. Meem 10-5-51
ing Facility Reesctor

CF 51-10-203 Tentative Comparison of Ionization Chambers R. H. Ritchie 10-31-51

CF 51-10-212 Application of a Scintillation Counter to F. K. McGowan 10-16-51
Gamma Ray Dosimeter C. E. Clifford

CF 51-11-95 Experiment V-A at Bulk Shielding Facility; H. E. Hungerford 11-15-51

158

Shadow Shield Measurements with a Na24 Source
REPORT NO.

CF 51-11-96

CF 51-11-139

Y-F5-57

CF 51-11-168

Y-F12-6

Y-F17-9

ANP-72

ANP-71

CF 51-11-23
CF 51-11-67
CF 51-11-72

Y-B4-38

Y-B4- 41

Y-811

CF 51-11-78
CF 51-9-64

CF 51-10-178

CF 51-11-47

CF 51-10-178

FOR PERIOD ENDING DECEMBER 10, 1951

" TITLE OF REPORT

Fast Neutron Measurements at Bulk Shielding
Facility

Preliminary Estimate of Circulating-Fuel
Reactor Shielding

The Divided Shield

Proposal for Divided Shield Experiments

Component of Reactor Systems
Status of Reactor Coolant Pump Program

Performance Characteristics for a General
Electric G-3 Electromagnetic. Pump

Containment of Helium in Stainless Steel and
Incone] at the 1500°F+ Range

Thermal Conductivity of Steel Wool and Some
Granular Solids

Metallurgy
Hydroxide Corrosion
Fission Product Analysis of ANP Material
High Temperature Mechanical Properties

Bismuth ~ Selected Physical Properties in
the Temperature Range 100 to 1000°C

Selected Physical Properties of Stainless

Steel in the Temperature Range 100 to 1000°C

Y-12 Alkali and Liquid Metal Safety Committee

AUTHOR(S)

R. G. Cochran
H. E. Hungerford
E. P. Blizard
L. A, Wills

E. P. Blizard
C. E. Clifford
A. Simon

H. L. F. Enlund
J. L. Meen

J. F. Haines

A. G. Grindell

=

S.

G.

Wischhusen
F. Salmon
F. Bailey

. D. Manly
A. Reynolds
H. Boss

Martha Wilson

Frances Sachs

p.

L. Hill

Physical Properties and Heat Transfer Research

Density of One Mixture of NaF-KF-UF,
Heat Capacity of Fuel Mix

Temperature Distribution in Thin Walled
Reactor Passages

Review of Air Cycle Heat Transfer Analysis

Temperature Distribution in Thin Walled
Reactors Having Noncircular Flow Passages

. M. Cisar

D. Powers

S. Farmer

. B. Harrison

S. Farmer

DATE ISSUED

11-20-51

11-26-51

9-17-51

No Date

10-26-51

10-24-51

10-16-51

11-93-51

11-5-51
11-14-51
11-15-51

9-14-51

10-1-51

8-13-51

11-14-51
9-13-51

10-23-51

11-9-51

10-23-51

159
ANP PROJECT QUARTERLY PROGRESS REPORT

REPORT NO.
ORNL- 1087

Y-F30-3

Y-B31-305

CF 51-10-194

CF 51-11-129

Y-F26-25

CF 51-11-159

Y-F26-23

160

TITLE OF REPORT

Heat Cepacity of Molybdenum

Forced Convection Heat Transfer in a Pipe
System with Volume Heat Sources Within the
Fluids

Miscellaneous

Analytical Chemistry — ANP Program Quarterly
Progress Report for Period Ending November
20, 1951

The H,0 Moderated, Salt Cooled Heterogeneous
Aircraft Reactor

NDA Quarterly Report on Supercritical Water
Reactor Work, September 1 to November 23,
1951

Directory of Active ANP Research Projects
at ORNL

Review of the Feasibility of the Air Cycle
Nucl}ear Reactor

ANP Information Meeting of Nov. 14, 1951

AUTHOR(S)

&=

A. Redfield

. H. Hil}

F. Poppendiek
Palmer

D. Susano

M. Weinberg

Gale Young

W.

J.

W.

B. Cottrell

F. Lane

B. Cottrell

DATE ISSUED

9-24-51

11-20-51

11-20-51

10-8-51

11-21-51

12-1-51

12-29-51

11-21-51
FOR PERIOD ENDING DECEMBER 10, 1951

18. DIRECTORY OF ACTIVE ANP RESEARCH PROJECTS AT OBNL(1)

December 1, 1951

I. REACTOR AND COMPONENT DESIGN

A. Aircraft Reactoer Design

1. Core and Pressure Shell 9704-1 Waislicenus
9201-3  Schroeder
2. Heat Exchanger and Radiator 9704-1  Fraas
9204-1 Hamilton
3. Pumps and Plumbing 9204-1  Wyld
Haines
4, Control 9704-1 FErgen
9201-3 Bettis
5. Shielding 9704-1 Ergen
3022 Blizard
6. Nuclear Physics 9704-1 Ergen
Mills
B. ARE Reactor Design
1. Core and Pressure Shell 9201-3  Hemphill
' Wesson
2, Fluid Circuit Design 9201-3 Cristy
. Lawrence
Jackson
Eckerd
3. Pressure and Flow Instrumentation 9201-3 Reese
4. Structural Analysis 9201-3 Maxwell
5. Thermodynamic and Hydrodynamic Analysas 9201-3 Lubarsky
6. Remote Handling Equipment 9201-3  Hutto
Alexander
7. Hazards Analysis 9704-1 Ergen
8. Monitoring Equipment for Na Leaks K-1005 Cameron
McKown
9. Electrical Power Circuits 3500 Owens
Beler
C. ARE Control Studies
1. High Temperature Fusion Chamber 2005 Hanauer
2. Control System Design 2005 Epler
Kitchen
Ruble

D. ARE Building Facility

1. Construction 7501 Nicholson Co.
2. Internal Design 1000 Browning

 

(Dmngs directory has been printed separately in a document by W. B. Cottrell Directory of Active ANP
Research Projects ¢t ORNL, Y-F26-25 (Dec. 1,(1951).

161
ANP PROJECT QUARTERLY PROGRESS REPORT

F.

G.

162

Reactor Statics

1.

U
- .

-3

11.

12'

13.

14.

Statics of the Circulating Fuel Reactor

Parametric Studies of H,0 Moderated Circu-
lating Fluoride-Fuel Reactor

Statics of the Critical Experiments

Statics of the NaOH Moderated Reactor

Final Report on Sodium Cooled Stationary
Liquid Fuel Reactors

Investigation of Simplified Calculations
Preparation of Reactor Calculations and
Cylindrical Coordinates for the IBM
Investigation of Errors in Multigroup
Procedures

Problem of Minimum Critical Mass

Energy Distribution of Thermal and Epi-

Thermal Neutrons
IBM Calculations for the ORNL ARE Proposals

IBM Calculations for the GE-ANP Proposals

Age Calculations of Hydrogen-Moderated
Reactor (GE)
Simplified Reactor Theory

Reactor Dynamics

1.

2@

3.

Kinetics of the Circulating-Fuel Reactor

Perturbation Calculation of Kinetics of
Circulating-Fuel Reactor
Kinetics of the Stationary lLiquid Fuel Reactor

Critical Experiments

1.

Graphite Critical Test Assembly

9704-1

9704-1

9704-1

9704-1

9704-1

9704-1
9704-1

9704-1
9704-1
9704-1
9704-1

9704-1

9704-1

9704-1

9704-1

9704-1

9704-1

9213

Mills
Macauley
Smith
Mills
Macauley
Smith
Mills
Smi th
Holmes
Mills
Mills

Prohammer
Edmonson

Coveyou

Coveyou
Coveyou

Macauley
Uffelman
Johnson
Macauley
lLeeth (GE)
Johnsaon
Macauley
Leeth (GE)
Thompson

Mills
Smith
Macauley
Ergen
Smith

Mills
Smith

Macauley

Callihan
Zimmerman
Williams
Haake
Scott
2.

3.

FOR PERIOD ENDING DECEMBER 10, 1951

Air-Water Critical Assembly (GE)

ABE Critical Experiments

H. Pump Development

1.

2.

Centrifugal Pump-Gas Seal
Centrifugal Pump-0il-Graphite Seal

Centrifugal Pump for ARE
Flectromagnetic Pump

Canned Rotor Pump
Frozen Sodium Seal Pump
Frozen Fluoride Seals
Meghanical Seals

Rocking Channel Sealless Pump
Seals for NaOH

I. Valve Development

1.

2.

Self-Welding Tests

Bellows Tests at High Temperatures

J. Heat Exchanger and Radiator Development

1.

2.

4!

Ligquid to laiguid

Liquid to Aar

Boeing Turbojet with Na Radiator

Fuel to Liquid

9213

6213

9201-3

9201-3

9201-3

5201-3

9201-3
9201-3
9201-3
9201-3

BMI
BMI

9201-3

9201-3

9201-3

9201-3

9201-3

9201-3

Callihan
Zimmerman
Williams
Haake
Scott
Callihan
Zimmerman
Williams
Scott
Haake

McDonald
Cobb
McDonald
Grindell
McDonald
Cobb
Haines
McDonald
Southern
Wyld
Richardson
Blalock
McDonald
Huntley
McDonald
Smith
McDonald
Huontley
Dayton
Simons

Allen

Adamson
Petersen
Reber
Reber

Fraas
Wyld
LaVerne
Fraas
Whitman
LaVerne
Fraas
LaVerne
¥raas
Whatman

163
ARNP PROJECT QUARTERLY PROGRESS REPORT

164

5. Htadiator Design

Finid Dynamics

1. Full Scale ARE Test Facilities
2, Fuel Manifold Mockups

IT¥. SHIELDING RESEARCH

Cross-Section Measurements

1. Neutron Velocity Selector
2. Be (n,2n) Cross-Section
3. Analysis for He in Jrradiated Be

4. Total Cross-Sections of Fe
5. Total Cross-Sections of N (GE)

6. Cross-Sections for BeQ and C

Shiclding Measurements

1. Divided Shield Mockup Tests (GE)
2. Gamma Shadow Shield Experiment (GE)
3. Bulk Shielding Reactor Power Calibration

4. Bulk Shielding Reactor Cperation
5. Heat Release per Fission
6. Metal Duct Tests

7. Li? Bremsstrahlung Measurement

8. Air Duct Tests {(GE)

Shielding Theory and Calculations
1. Survey Report on Shielding
2. Shielding Section for Reactor Technology

3. Calculations of Removal Cross Sections
4, Theory of Neutron Transmission in Water

9201-3

9201-3
9201-3

2005

2005

3026

9201-2

9201-2

3001

3010
3010
3010

3010

3010
3001

3025
3001

3022
9704-1
3022
3022
3022

fFraas

Kackenmester
Wischhusen

Ward

Pawlicki
Smith
Klema
Arfken
Parker
Ergen
Willard
Bair
Kington
Willard
Bair
Kington
Clifford
Flynn
Blosser

Meem, etc.
Hungerford
Johnson
McCammon
Holland
Leske
Roseberry
Meem
Hullings
Mockenthaler
Sisman
Clifford
Flynn
Blosser

Blizard
Welton
Blizard
Blizard
Blizard
Enlund
FOR PERIOD ENDING DECEMBER 10, 1951

5. Interpretation of Pb-H O Lid Tank Data 2005 Simon

6. Divided Shield Calculations 3022 Murray

7. Divided Shield Theory and Design NDA Goldstein
Feshbach

B. Design of Liquid Ammonia Unit Shield 3022 Blizard

. Enlund

Wyld

9. Air Duct Theory (GE) 3001 Clifford
Simon

Pp. Shielding Instrusments

1. Gamma Scintillation Spectrometer 30140 Maienschein
2. Neutron Dosimeter Improvement 3010  Hurst
Glass
Cochran
3. Proton BRecoil Spectrometer for Neutrons 3010 Cochran
Henry
4. He® Counter for Neutroas 3010 Cochran
9. L1 I Crystals for Neutrons 3010 Maitenschein
| Schenck
6. Neutron Spectroscopy with Photographic Plates 3010 Johnson
‘ 1 Haydon
yHoneycutt

E. Shielding Materials

 

1. Preparation of High Hydrogen Rubber Goodrich Davidson
' Co.
2. Development of Hydrides for Shields MH1 Banus

IXI. MATERIALS RESEARCH

A. Corrosion by Liquideetals

1. Static Cerrosion Tests in Ligquid Metals 2000 Vreeland
' Day
Hoffman
2. Static Corrosion Tests in Liguid Metal Alloys 2000 Vreeland
Day
Hof fman
3. Static Corrosion Tests in l.ow Melting Point 2000 Vreeland
Alloys Day
Hof fman
4. Dynamic Corrosion in Na-Isothermal 2000 Vreeland
| Trotter
5. Dynamic Corrosion Research in Harps 2000 Adamson
6. Effect of Crystal Orientaticn on Corrosicn 2000 Smith
' Cathcart
Bridges

165
ANP PROJECT QUARTERLY PROGRESS REPORT

166

7‘
8.
9

10.
11.

Corrosion by

1.

2.

Effect of Carbides on liguid Metsl Corrosion
Mass Transfer in Mclten Metals
Diffusion of Molten Media into Solid Metals

Structure of Ligquid Pb and Ba
Alloys, Mixtures and Combustion of Ligquid Sodium

Flvorides

Static Corrosion of Metals in Fluoride Fuels

Static Corrosion of Special Alloys in Fluoride
Fuels

Static Corrosion Tests in Fluoride Salts

Mechanism of Fluoride Corrosion

Dynamic Corrosion Tests in Fluoride Corrosion
Effect of Flow Velocity on Fluoride Corrosion

Effect of Contaminants on Fluoride Corrosion

Corrosion by Hydroxides

i.

-3
.

Static Corrosion of Metals

Static Corrosion of Special Alloys in Hydroxides
Mass Transfer in Molten Hydroxides
Physical Chemistry of the Hydroxide Corrosion

Phenomenon
Static Corrosion by Hydroxides

Static Corrosion by Hydroxides

Mechanism of Hydroxide Corrosion
Anodic Protection of Metals in Hydroxides

2000
2000
2000

2000
2000

2000

2000

9766

9766

9201-3

9201-3

9201-3

2000

2000

2000

2000

9766

BMI

BMI
9766

Brasunas
Richardson
Brasunas
Richardson
Richardson
Smith
Bridges
Smith

Vreeland
Day

Hof finan
Vreeland
Day
Hoffman
Kertesz
Buttram
Smith
Meadows
Kertesz
Buttram
Smath
Adamson
Coughlen
Adamson
Coughlen
Adamson
Coughlen

Vreeland
Day
Hoffman
Vreeland
Day
Brasunas
Richardson
Cathcart
Smith
Kertesz
Buttram
Croft
Smith
Meadows
Jaffee
Craighead
Pray
Kertesz
Buttram
Smith
9.
10.

FOR PERIOD ENDING

Dynamic Corrosion Tests in Hydroxides
Corrosion. by Homogeneous Fuels

Physical Properties of Materials

10

2'

10.

11.

Density of Liquids

Viscosity of Liquids

Thermal Conductivity of Solids
Thermal Conductivity of Liqqids
Specific Heat of Selids and Liquids
Thermal Diffusivity

Wetting

Electrical Resistance of Fluoride Salts
Viscosity of Fluoride Salts

Vapor Pressure of Fluorides

Vapor Pressure of BeF,

Strength of Materials

1.

2.

Creep Tests in Fluoride and Hydroxides

Creep and Stress Rupture Tests of Metals in
Controlled Atmospheres

Creep and Stress Rupture Tests of Materials an
Liquid Media

High Temperature Cyclic Tensile Tests

Tube Burst Tests
Tube Burst Tests

Relaxation Tests of Beactor Materials

DECEMBER 190,

9201-3
9766

9204- 1
9204-1
9204-1
9204-1
9204-1
9204-1

9201-3

9201-3
9766

9733-3

BMI1

9201-3

2000

2000

2000

9201-3
2000

2000

1951

Adamson

Kertesz
Buttram
Croft

Kaplan
Cisar
Tobias
Kaplan
Tobias
Powers
Chandler
Claiborne
Powers
Jones

Tobias
Cisar
Wischhusen
Ward

Affel
Kertesz
Knox

Barton
Moore
Patterson
Clegg

Adamson
Coughlen
Oliver
Woods
Weaver
Oliver
Woods

Weaver

Oliver
Woods
Weaver
Adamson
Oliver
Woods
Weaver
Olzver
Woods

Weaver

167
ANP PROJECT QUARTERLY PROGRESS REPORT

168

8.
9.

Creep Tests in Thermal Convection Loops
Creep Tests of Materials (GE)

Hetals Fabrication Methods

1.

N
»

Oy oo W
. .

].O.

11.

12.

New

1.
2.

Welding Techniques for ARE Parts
Brazing Techniques for ARF Parts

Molybdenum Welding Research

Molybdenum Welding Research

Resistance Welding for Mc and Clad Metals

Welds in the Presence of Various Corrosion Media

Nondestructive Testing of Tube to Header Welds

Basic Evaluation of Welds Metal Deposits in
Thick Plates
Evaluation of the Cone Arc Welding Technique

Evaluation of the High Temperature Brazing
Alloys
Development of High Temperature Brazing Alloys

Welding in Presence of Hydroxides

Wetals Development

Mo and Cb Alloy Studies
Heat Treatment of Metals

Ceramics and Metals Ceramics

1.

2.

BeO Fabrication Research

Metal Cladding for BeQ

B,C Control Rod Development

Hot Pressing of Tungsten Carbide Bearings

Hot Temperature Firing of Uranium Oxide to
Produce Selective Power Sizes

Devel opment of Cr-UO2 Cermets for Fuel Elements

Ceramic Coatings for Stainless Steel
Valve Parts for Liquid Metals and Fluorides

9201-3
2000

2000
2000
BMI

MIT

RP1

2000
2000
2000
2000
2000
Wall

Colmonoy
9766

2000
2000

Gerity
Mich.

Gerity
Mich.

2000

2000
2000
9766

0766
9766

Adamson
Oliver
Woods
Weaver

Patriarca
Slaughter
Patriarca
Slaughter
Parke
Wulff
Nippes
Vreeland
Patriarca
Slaughter
Patriarca
Slaughter
Patriarca
Slaughter
Patriarca
Slaughter
Patriarca
Slaughter
Peaslee

Kertesz
Buttram
Croft

Bomar
Bomar

Coobs

Graaf
Graaf

Bomar
Coobs
Bomar
Coobs
Bomar
Coobs
Johnson
Shevlin
White
Shevlin
9.
10.

FOR PERIOD ENDING

Application of Ceramic Materials to Reactors
Crucible Development for High Temperatures

I. Solid Fuel Element Fabrication

1.

2.

-~ N

Sclid Fuel Element Fabrication
Diffusion-Corrosion in Solid Fuel Elements

Determination of the Engineering Properties of
Solid Fuel Elements
Electroforming Tube to Header Configurations

Electroplating Mo and Cb

Carbonyl Plating of Me and Cb
Rolling of Fuel Plate Laminates (GE)

J. Liguid Fuel Chenistry

10

S O O~
*r & & =

Phase Equilibrium Studies

Preparation of Standard Fuel Samples

Mechanism of Fuel Pretreatment

Reaction of Fluoride Fuels with O, and H,0

Ionic Species in Fluoride Fuels
‘evelopment of Homogeneous Fuels

Stability of Slurries of U03 in NaOH

Phase Equilibria Among Silicates, Borates, etc.
Fuel Mixtures Containing Hydrides

Chemical Literature Searches

K. Liquid Moderator Chemistry

1.

Preparation of Pure Hydroxides

DECEMBER 10, 1951

9766

Norris

Elect.

iab.

2000
2000
2000

Gerit
Mich)

Gerity
Mich,
2000
3012
2000

9733-2

9733-2

9733-2

9733-2

9733-3
9733-3
BMI
BMI

MH1
9704-1

9733-3

Johnson
Wilson
Doney

Bomar
Coobs
Bomar
Coobs
Bomar
Coobs
Graaf

Graaf

Bomar
Cunningham
Bomar
[l.eonard

Barton
Blakely
Bratcher
Barton
Nessle
Powers
Love
Barton
Nessle
Powers
lLove
Barton
Nessle
Powers
Love

Barton
Robinson
Overholser
Redman
Patterson
Crooks
Banus

Lee

Overholser
Nicholson
Cuneo

169
ANP PROJECT QUARTERLY PROGRESS REPORT

4’.
3.

Thermal Stability of Hydroxide Mixtures

Solubility of Metals in Hydroxides

Moderator Systems Containing Hydrides
Hydroxide-Metal Systems

L. Liguid Coeclant Chemistry

1.

3.

Phase Equilibrium Studies

Preparation of Standard Cooclant Samples

Reaction of Fluorides with Alkali Metals

M. Radiation Damage

1.

o
.

11.

170

Liquid Compound Irradiations in LITR

Liquid Compound Irradiations in Cyclotron

Liguid Compound Irradiations in MTR

Fluoride Fuel Irradiation in Berkeley Cyclotron
Liquid Metal Corrosion in X-Pile loops

Stress Corrosion and Creep in LITR Loops
Creep of Metals in LITR and X-Pile

Thermal Conductivity of Metals in LITR and
X-Pile

Diffusion of Fission Products from Fuels

Neutron Spectrum of LITR

Irradiation of Water (GE)

9733-3

9766

MHI
3550

9733-2

9733-2

9733-3

3005

9201-3

3025

NAA
3001

3025

3025
3001

3005
3001
3001
3005

3550

Overholser
Ketchen
Nicholson
Kertesz
Croft
Smith
Banus
Bredig

Barton
Blakely
Bratcher
Nessle
Powers

Love
Blankenship

Keilholtz
Morgan
Webster
Robertson
Klema
Kinyon
Keirlholtz
Feldman
Sturm
Jones
Keilholtz
Klein
Kinyon
Pearlman
Brundage
Parkinson
Ellis
Olsen
Sisman
Bauman
Carroll
Sisman,
Wilson
Zukas
Davis
Cohen
Templeton
Keilholtz,
Sisman
Trice
Lewis

Taylor

etc,

etc,
;FOR PERIOD ENDING

N. Materials Analysis and Inspection Methods

9.
10.

11.

12.

13.

14.

Determination of Oxygen in Sodium and NaK
Determination of Oxygen and Nitrogen in Lithium

Determination of Trace Impurities in Sodium by
Colorimetric Methods

Analytical Studies of Fluoride Eutectics

Determination of Oxygen in Inert Gases
Composition of Diatomaceous Earth
Trace Quantities of Boron by Colorimetric Methods

Determination of Nickel and Nickel Oxides in
Alkali Metal Hydroxides

Clarity Tests of Borated Water Selutions
Chemical Methods of Fluid Handling

Metallographic Examination

Coordination of Chemical and Analytical Data
on Loops *

Reaction Products of Chromium and Alkali Metal
Hydroxides | ‘ '

lLead in Fluoride Eutectics and in Alkali Metal
Hydroxides

D. Heat Transfer

1.

1.

Convection in Liquid Fuel Elements

Heat Transfer in Circulating Fuel Reactor

Heat Transfer Coefficients of Fluorides and
Hydroxides
Hgat Transfer Coefficients of Lithium

Boiling Liquid-Metal Heat Transfer
Sodium Heat Transfer Coefficients in Short Tubes
Heat Transfer in Special Reactor Geometries

Fluoride Bandling

Fluoride Production

DECEMBER 10,

9733-2
9733-2

9733-2
9733-2

9733-2
9733-2
9733-2
9733-2
9733-2
9201-3
2000

9733-3
9733-2

9733-2
9204-1

9204-1

9204-1
9204-1
5204-1

9204-1
9204-1

9201-3

1951

White
Ross
White
Ross
White
Ross
Baxter
White
Baxter
Druschel
White
Brady
White
Ross
White
Ross
White
Baxter
House
Bronstein
Gray
Krouse
Roeche
Blankenship
Metcalf
White
Baxter
White

Druschel

Hamilton
Redmond-
Lynch
Tobias
Poppendiek
Palwer -
Hamilton
Hof fman
Lonas
Claiborne
Winn
Farmer
Harrison
Claiborne

Kackenmester

Mann

171
ANP PROJECT QUARTERLY PROGRESS REPORT

172

2. Fluoride System Cleaning 9201-3
3. Fluoride Salvage and Disposal 9201-3
4. Preheating of Fluoride Systems 9201-3
5. Fluoride Pressure and Flow Measurements 9201-3
6. Experimental Joints for Hot Ligquid Systems 9201-3

Liquid Metal Handling

1. Equipment Cleaning Techniques 9201-3
2. Continuous and Batch Sodium Purification 9201-3
3. Sampling Techniques 9201-3
4, Sodium Vapor Trapping 9201-3
5. Liguid Metal Salvage and Disposal 9201-3
6. Liquid Metal Safety Equipment 9201-3
7. Blanket Gas Purification 9201-3

Dynamic Liquid LooODPS

1. Operation of Convection Loops 9201-3
2. High Flow-Rate Convection Loop 9201-3
3. Operation of Figure-Eight Loops 9201-3
4, Bi-Fluid Systems 9201-3
S. Fluoride Fuel Flow Transfers 9201-3
6. UO;-NaOH Slurry lLoop BMI

7. Operation of Thermal Convection Loops 2000

IV. ANALYSIS OF OTHER NUCLEAR REACTOR SYSTEMS
Supercritical Water Reactor
1. Cycle Analysas 9704-1
2. Cycle Analysis NDA

Helium-Cooled System

1. System Analysis NAA

Na-Vapor Compressor-Jet System

1. System Analysis NAA

Air-wWater {ycle

1. Survey of Air-Water Cycle 9204-1

Mann
Mann
Affel
Coughlen
Bailey
Taylor
Reber
Wyld

Mann
Mann
Mann
Mann
Devenish
Mann
Devenish
Mann
Mann

Adamson
Tunnell
Coughlen
Coughlen
Wischhusen
Ward
Simons
Cathcart
Bridges
Smith

Fraas
Cohen

Gruber, etc.

Schwartz, etc.

Schwartz, etc.

[Lane
Noderer
 

 

 

 

 
 

 

 

 

 

 

  
  

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

OFFICIAL USE ONLY 19_Chart of the Technical Orgaonization of
) THE AIRCRAFT NUGLEAR PROPULSION PROJECT
ANP DIVISION DIRECTOR
R. C. BRIANT, 8iDG. 9704.1
D. MILYER. SEC,
ANP COORDINATOR
¢, B. ELLIS NOTE: This churt shews only the lines of technical coordination of
L. BOND. SEC the ANP project, The verious individvals end greups of pacple listed
’ ' ’ ANP LIBRARY 5 ch and dasign which
BLDG. $704.1 is coordinated fer the banefit vi the ANP projact in the monwer indis
ADMINISTRATIVE ASS|STANT : ' *d coted en the thart. Eoch group, hewaver, is oluc responsible to it
X group p
L. M. COOK M. CARDWELL Division Dirsctar for the dutailad pregress of Its reses and for
P. HARMAN® SEC M. BROWN wdministrative matrers, Peesennsi fram 13 differeny Divi of ths
. . E. CARTER Oak Ridge Nationol Laboretary and Enginesring secrisns of the Y-12
§. REAGAW Plont ara Included on the chart without speciic indication of divie
COORDINATING STAFF PROJECT EDITOR E. WEBSTER sionsl tines.
w. B. COTVTRELL
P. HARMAN® SEC.
STAFF ASSISTANT ARE FROJECT CHIEF STAFF ASSISTANT STAFF ASSISTANT STAFF ASSISTANT
FOR RADIATION DAMAGE Ww. M. BREAZEALE FOR EXPERIMENTAL ENGINEERING R, C. GRIANT FOR METALLURGY FOR CHEMISTRY
€. 8. ELLIS A. J. MILLER R. WiLL!AMS,. SEC. E. S5, BETTIS ¥. D, MANLY . R. GRIMES
SHIELD NG RESEARCH REACTOR PHYSICS GENERAL DESIGN GROUP RADIAT I ON DAMAGE ARE DESIGN EXPERIMENTAL ENGINEERING HEAT TRANSFER RESEARCH METALLURGY CHEM} STRY CHEMICAL ANALYSIS
E. P. BLIZARD, 8LDG. 3022 W. K. ERGEM, BLDG. 9704-1 C. 9. ELLIS, BLDG. 9704-1% 6. S. BILLINGTON®, BLDW. 3025 R. W. SCHRDEDER. BLDG. 9201-3 H. W. SAVAGE, BLDG. 9201-3 H. F. POPPENDIEK. 8LDG. 9204-3 W. D. MANLY. BLDG. 2000 W. R. GRIMES, BLDG. 3733-3 C. D. SUSANG, BLDG. 9733.2
J. t. MEEM
R. R. COVEYOQU A. G. H, ANDERSEN C. D. BAUMANN A. W. ALEXANDER G. M. ADAMSON® . J. M. CISAR® G. M. ADAMSON® C. J. BARTON L. J. BRADY
LiD TANR N. EDMONSON E, S&. BETTIS"® W, £, BRUNDAGE /. K. BROWNING R. G. AFFEL H. €. CLAIBORNE E. S. DOMAR 4. P. BLAKELY H. P. HOUSE®
C. E. CLIFFORD, BLDG. 3001 8, 1. MACAULEY. USAF J, M. CISAR"® R. M, CARROLL G. A. CRISTY" M. B, BRIDGES L. CCOPER A. D. BRASUNAS F. F. BLANKENSHIP W, K. MiLLER
€, H. MILLS A. P. FRAAS® A. F. COHEN G. W. ECKERD H. R, HRONSTEIN W. S. FARMER W. H. BRIDGES R. A. BOLOMEY J. W, ROBINSON®
T ;' :'-3555“ %. K. OSBORN M. E. LEE W. W. DAVIS J. Y. ESTABROOK K. E. BURMASTER D. Ot HAMILTON J. V. CATHCART b. R, CUNED R. ROWAN®
;. C. LYNN F. G. PROHAMMER® K. M. MARTIN CHESTER ELLIS B. L. GREENSTREET w. G. COBB W. B. HARRISON J. H, COOBS E, E. KETCHEN J. €, WHITE
L. MARMEY  ec S. TAMOR A. S. THOMPSON® J. T. HOwE® L. F, HEMPHILL €. P. COUGHLEN H. W, HOFFMAN R. B, DAY R, P. METCALF | AND OTHERS
L. GORDON. . E. S. WILSON, USAF G. F. WISLiCENUS. CONS. R. H. KERNGHAN E. L. HUTTO G. A. CRISTY® 5. 1. KAPLAN L. D. DYER R. E. MOORE B. YOUN&®. SEC.
TECHNICIAN R. D. WORLEY, USAF A. S. DLSON J. P. JACKSON R. DEVENISH F. E. LYNCH || E. E. NDFFMM} G. ; NE(S::;ES
1 A NI
J. L. HULL J. WATKINS. SEC. W, W. PARKINSON M. C. LAWRENCE A. G. GRINDELL L. D. PALMER® R. B. OLIVER .
COMPUTERS GENERAL DESIGN CONTRACTOR O, SISMAN® A. B. LONGYEAR R. HELTON W. 0. POWERS P. PATRIARGA D. E. NICHOLSON CONSUL TANT
L c. TEMPLETON AEROJET ENG. CORP. 8. E. HILL R. F. REDMOND L. 5. RICHARDSON L. G. OVERHOLSER H. H. WILLARD, UNIV. MICH,
DUCT TEST FACILITY P. BROWN NUCLEAR DEVELOPMENT J. B. TRICE 8. LUBARSKY, HACA w. R. HUNTLEY M. RICHARDSON® 6. M. SLAUGHTER H. POWERS
5, EZZELL ] ASSOCIATES, INC. J. C. WILSON w, L. SCOTT D. T. JAMES M, TORiAS G. P. SMITH J. D. REDMAN
€. E. CLIFFORD, BLDG. 3001 A, FORBES 1. H. MENKE 4. C. IUKAS J. W. WALKER® H. R. JOHNSON T. SUTTON. SEC. D. C. VREELAND M. T. ROBINSON SPECTROGRAPHIC ANALY3IS
F. J. MUCKENTHALER £, LAWSON G. YOUHNG J. WEST, SEC. H. R. WESSON H. P. KACKENMESTER TECHNIC1ANS J. ¥. WOODS D. CALDWELL, SEC. J. R. MCNALLY®, BLDG. %734
M. K. HULL INGS M. TSAGARIS 6. H. GOERTIZEL F. SHORT. SEC. W. G. LAMBERT c. 5. BLALOCK J. THOMAS, SEC. TECHNICI&NS AND OTHERS
R. WEAVER A. R. GRUBER E. M. LEES R M. GURNETT TECHNICTANS L. M. BRATCHER
AND OTHERS LI1QUID-FUEL RADIATION DAMAGE R. E. MACPHERSON s
BULK SHIELDING REACTOR o v RerLmoLTZ CONSULTANTS L A MANN S. J4. CLAIBORNE G. D. anaovm . ¥. C. LOVE
4 L. MEEM, BLDG. 3010 1BM MACH INE COMPUTATIONS - *. R, L. MARWELL® R. N. MASON I ':{-még"“ I : :;sga o
. L. . . M. J, FELDMAN W, R. CHAMBERS, UNiV. TENN, W. B. McOONALD . S
R. G. COCHRAN F. C. UFFELMAN®, BLDG. 9706-14 € w. KINYON oo u;gzm G. M. WINN G. M. SONZALEZ CONTRACTORS
M. P. HAYOON ¥. C. DeMARCUS P. R. KLEIN w. R. OSBORN J. D. HUDSON BATTELLE MEMORIAL INSTITUTE
K. M. HENRY $. C. JOHNSON 3. 6. MORGAN 3 & eatTon R, w, JOHNSON (AEC CONTRACTOR) CORROS 10 LABORATORY
L. B. HOLLAND AND OTHERS H. E. ROBERTSON €. B. PERRIN A, T. KINSER J. w. CLEGG
H, E. HUNGERFORD W. J. STURM 6. F. PLTERSEN ¥. &. LANE . u bAvTON F. KERTESZ, 9LDG. 9766
E. B. JOHNSON C. C. WEBSTER X. W. REBER B. E. McCARTY B ONEFEE
J. K. LESLIE CONSULTANT o REESE 1. E. POPE H. A. Jarr |
F. T. MAIENSCHEIN N, M. SMITH, OFERATIONS 86. INCH CYCLOTRON REACTOR CONTROL F. SALMDN CERAMICS RESEARCH C. E. SHUBERT .ANI.J P nERS H. 5. BUTTRAM
G. M. McCAMMON RESEARCH OFFICE P& SMITH o L. R. TRUTTER €. R. CROFY
T, N, ROSEGERRY . F THE ARMY RADIATION DAMAGE E. 5. BETTIS® - & T. N. McVAY®. BLOG. 3788 €. K. THOMAS FoAl KNOX
. . A L. N - k. . AL
A BULLARD. SEC. DEPT. 0 R. S. LIVINGSTON® BLDGS. 9201-3 AND 2005 At 32':;::& o W, WEAVER METAL HYDRIDES, INC. Ne GaiTH
TECHNICIANS OFFICES: @LDG. 9204.3 P W TavioR M. D, BANUS
CYCLOTRON: BLDG. 3201-2 T. E. COLE. R. % - 1. £, THOMAS L. M. DOREY METALLOGRAPHY AND OTHERS
S. B. DEHART ) D. &INDER E. P. EPLER* - €. THO S. D. FULKERSON
K. HONEYCUTT A. L. BOCH® S. H. HANAUER W, C. TUNNELL J. R. JOHNSON ] R. J. GRav® CONSULTANT
H. J. JARVIS - b M. S. KITCHEN D. R. WARD G. W. WHITE R. 5. CROUSE®*
p. J. KIRBY R, J. JONES E. R. MANN® J. F. WARNER 1 aE D. 5. HiLL, DUKE UNIY.
J 5' MINIARD CRITICAL EXPERIMENTS H. XEESE®" J‘ n. RUBLE J. C. WHITE TECHNIC I AN . K. ROEC
R. M. SIMMONS A. D. CALLIHAN®, BLDG., 9213 R. L. KNIGHT ' G. D. WHITMAN J. A. GRIFFIN TECHNICIANS
g. ; LEéE!:SENRINb L. V. WILSON N. M. ATCHLEY
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E. V. HAAKE ] R swiTH 5 :'SE::‘;“":"! CONSUL TANT E. R. BOYD
THEOREY LCAL ANALYSt!S O. XK. HOLMES* - - B . J. C. GOWER
c TECHNICTANS D. ALEXANDER, SEC. T. s. SHEVLIN. 05U E. P. GRIGGS
H. L. F. ENLUND D. scoTT b. STOREY. RECORD CLERK 5. ¢. LESLIE
F. H, MURRAY D, V. P. WILLIAMS C. A. GAULTY D. HARRIS. STEN T oa
A. SIMON® E, L. ZIMMERMAN 4. V. HILYER SPECIAL ARE PROJECTS . . . :. : ::fs:g:nv
L. GORDON, SEC. M. HALE. SEC. pG. 9704-1 s
MASS SPECTROMETRY 8L TECHNI1C1ANS
RUSSELL BALDOCK*, BLODG. 9735 'J‘ : ;l;l:'l‘:(:: CONSULTANTS
CONSULTANTS A. P. FRAAS® ol 3. F. CHARLES E. CREUTZ, CARMEGIE INST.TECH.
. .
H. A. BETHE, CORNELL UNIV, CONSULTANTS v : S:Lt::: CONS. G. S. CHILTON N : g:::; N RNELL UNiv
H. GOLDSTEIN, NDA B, R. GOSS51CK. PURDUE UNIV. 3 b omENS® I. E. CRABYREE T RIPPES. RPI .
NUCLEAR MEASUREMENTS N. J. GRANT. wmiT C F. wesT 4. M. CUNNINGHAN 4. SHOLUCHOWSK |
H. G. MACPHMERSON, NAT. CARBON F. A. DOSS IARNECIE INST. TECH
CONTRACTORS A. H, SMELL*, BLDG. 2005 L. P. SMITH, CORNELL UNIV. J. R. DUCKWORTH E. E, STANSBURY ‘UNIV "Tenn
METAL HYDRIDES. INC, R. SMOLUCHOWSKI, C. J. GREEN b . WRIGHT. UNIYV. OF ALA.
M. D. BANUS CARNEGIE INST. TEGH, 1. L. GREGORY s : ' :
AND OTHERS NEUTRON VELOCITY SELECTOR D, F. WEEKES, TEXAS A A ® F. M. GRIZZELL
G. 5. PAWLICKL, ORINS, BLDG. 2005 : ’J‘ 22:':'5:“ CONTRACTORS
B'(:'Ecsggg:;gzl‘g:v. E. €. 3MITH CONTRACTORS R. 6. WILEY BATTELLE MEMORIAL NSTITUTE REVISED DECEMBER 10, 195)
.. DAVIDSON NORTH AMERICAN AVIATION, INC. 4 {AEC CONTRACTOR)
3 AEC CONTRACTOR) CONSULTANTS - R. M. PARKE
AND OTHERS 6 MEV VAN DE GRAAFF ¢ K., SALLEY
H. P. YOCKEY J. F. BAILEY. UNIV. TENN. 2. SCHOFIELD
NUCLEAR DEVELOPMENT H. B. WILLARD, BLDG. 9201.2 K. PEARLMAN W. R. CHAMBERS, UNIV. TENN. "AND OTHERS
ASSUCIATES. INC, 1. K. BAIR AND OTHERS
H. GOLDSTEIN® €. W. JOWNSON — RENSSELAER POLYTECHNIC
. . 0. KINGDON
R. MELA J. 0 © INSTITUTE
H. FESHBACH® ENGINEERS E. NIFPES
AND OTHERS F. P. GREEN ANO OTHERS
R. W, LAMPHERE
GERITY MICHIGAN MPARY
TECHNICIANS 6. sanrr o
W. 7. NEWYON R. 5. — Dak Ridgs Schoo! of Resctor Tachnology "ARD OTHERS
W, w. WERNER * Part-time
MASSACHUSETTS INSTITUTE
OF TECHNOLOGY
BERYLL UM CROSS-SECT|ONS ). WULFF
E. D. KLEMA, BLDG. 2003 AND OTHERS
G. ARFKIN® OFFICIAL USE ONLY

 

 

 

 

 

 

 

 

173