ORNL-1154.txt

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o

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ORNL-1154

This document consists of 225 pages.
Copy §i of 203 . Series A

Contract No. W-7405, Eng-26

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

R. C. Briant
Director, ANP Project

C. B. Ellas
Coordinator, ANP Project

Edited by:

W. B, Cottrell

DATE 1SSUED:

  

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

 
 

ORNL-1154
Progress Report

INTERNAL DISTRIBUTION

1 G. T. Felbeck (C&CCC) 37 R. W. Stoughton
2-3 Chemistry Library 38 F. R. Bruce

4 Physics Library 39 H. W. Savage

5 Biology Library 40 W. K. Eister

6 Health Physics Library 41 A. S. Householder
7 Metallurgy Library 42 C. B. Graham

8-9 Training School Library 43 R. N. Lyon

10-13 Central Files 44 C. P. Keim

14 C. E. Center 45 W. R. Gall

15 C. E. lLarson 46 A. J. Miller

16 W. B. Humes (K-25) 47 R. W. Schroeder
17 W. D. Lavers (Y-12) 48 D. S. Billington
18 A. M. Weinberg 49 E. P. Blizard

19 E. H. Taylor 50 C. E. Clifford
20 E. D. Shipley 51 G. H. Clewett

21 £. J. Murphy 52 A. D. Callihanr
22 F. C. VonderLlage 53 R. S. Livingston
23 R. C. Briant 54 W. D. Manly

24 J. A, Swartout 55 J. L. Meem

25 C. B. Ellis 56 C. D. Susano

26 A. H. Snell 57 W. B. Cottrell
27 A. Hollaender 58 W. M. Breazeale
28 F. L. Steahly 59 W. R. Grimes

29 K. Z. Morgan 60 A. Brasunas

30 D, W. Cardwell 61 N. M. Smaith

31 M. T. Kelley 62 H. F. Poppendiek
32 E. M. King 63 F. C. Uffelman
33 C. E. Winters 64 D. D. Cowen

34 J. A. Lane 65 P. M. Reyling

35 J. H. Buck 66-75 ANP Library

36 J. P. G111l 76-81 Central Files (0.P.)

 
82-103
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142-145
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163-166
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189-203

 
 

 

ORNL-1154
- Progress Report

EXTERNAL DISTRIBUTION

Aircraft Nuclear Propulsion Project, 0Oak Ridge
Argonne National Laboratory

Armed Forces Special Weapons Project (Sandia)
Atomic Energy Commission, Washington

Battelle Memorial Institute

Brookhaven National Laboratory

Bureau of Aeronautics

Bureau of Ships

Carbide and Carbon Chemicals Company (Y-12)

Chicago Patent Group

Chief of Naval Research

duPont Company

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New York Operations Office

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Patent Branch, Washington

Savannah River Operations Office

University of California Radiation Laboratory
Westinghouse Electric Corporation

Wright Air Development Center

Technical Information Service, Oak Ridge

data as

of 1946.

 

111

 
iv

 

Reports previously issued in this series are as follows:

ORNL-528 Period Ending November 30, 1949
ORNL-629 Period Ending February 28, 1950
ORNL-768 Period Ending May 31, 1950
ORNL-858 Period Ending August 31, 1950
ORNL-919 Period Ending December 10, 1950
ANP-60 Period Ending March 10, 1951
ANP-65 Period Ending June 10, 1951

 

 

Ll
 

TABLE OF CONTENTS

SUMMARY _
Part 1. REACTOR THEORY AND DESIGN

1. THE AIRCRAFT RFACTOR EXPERIMENT .
Core Design |
Fluid Circuit
Heat-disposal circuit :
Contamination of helium system
Control of the ARE

Liquid-fuel control system
Solid absorber rod design
Electronic computor design

Electrical Circuit
Remo te-Handling Equipment
Building Facility for the ARE

2. EXPERIMENTAL REACTOR ENGINEERING

Liquid-Fuel Systems
Pump Development

Centrifugal pumps for figure-eight loops
~ARE pump design

Canned-rotor pump

Level tank pump

Electromagnetic pumps

Test Loops

Calibration loop
Sodium manometer loop

Seal Tests

Frozen-sodium seal
Bellows seal
Graphitar and tool steel seal

Instrumentation
Stress-Rupture Tests
Self-Welding Tests

Valve-Packing Experiments

 

17

PAGE
~NO.

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i4

14
15

15
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17

20

20
20

20

21
21
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23
23
24
24
 

Contamination of Liquid-Metal Systems

Cleaning of liquid-metal systems
Purification of liquid metals
Purification of inert gases
Analytical results with sodium

Investigation of Sodium Condensation

ARE Component Tests

Heat-Exchanger Tests

Building Modifications and Experimental Facilities

Alkali Metals Manual

3. REACTOR PHYSICS

IBM Calculations

Production report

The multiregion reactor problem

Cylindrical multigroup calculation

The effect of the borom blanket in the ANP reactor

Statics of the Aircrafc Reactor Experiment

Summary of calculations on the ARE
Estimated critical mass of the ARE
Control rod effectiveness

Kinetics of the Aircraft Reactor Experiment

Mass-reactivity coefficient
Fuel conductivity

Neutron lifetime

Start-up accident

Coolant temperature

Preparatory Physics Calculations

Correction to "Wigner formula for resonance escape" probability
Effect of "Wigner formula for resonance escape’™ correction
The transmission coefficient of the B,C curtain in the ANP

and ABE reactors
Heating in the boron carbide curtazin in the AMP reactor
Temperature savings in the ANP reactor

The Sodium Hydroxide Reactor

4. CRITICAL EXPERIMENTS

Critical Assembly of Air-Water Reactor

 

PAGE
ND.

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25
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30

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34
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48

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49

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36

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60
64

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66
67

70

19
79
 

Critical Assemblvy of Graphite Reacter

Part II. SHIELDING RESEARCH

BULK SHIELDING REACTOR

Reactor Operation
Mock-Up of the Unit Shield
Mock-Up of the Divided Shield

LID TANK
DUCT TEST

SHIELDING CALCULATIONS
Analysis of Bulk Shielding Reactor Neutron Data
Interpretation of Lid Tank Gafima-flay Data

Shielding Calculations for the ARE

Activation of nitrogen and helium in the ARE reactor pit

Activation of impurities in BeO :

Detection of leaks in the fuel elements by means of radiocactive
tracers ‘

NDA Divided-Shield Studies
Use of NH, as a Shielding Material

Mechanical requirements of an ammonia shield
Calculational procedure &
Comparison of shield weights using NH; and H O

NUCLEAR MEASUREMENTS

The 5-Mev Van de Graaff Accelerator
Measurement of the (n,2n) Reaction in Beryilium
Bremsstrahlung from Li8

Radiation detection equipment
Bremsstrahlung activity

Time~of-Flight Néutron Spectrdmeter

 

. PAGE

NO.

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83

B3
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94

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96

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98

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98

100

S vil
 

Part III. MATERIALS RESEARCH

10. CORROSICN RESEARCH

Static Corrosion by Fluoride Fuels

The pretreatment process
Corrosion of structural metals
Corrosion of platinum

Static Corrosion by Moderator Coolants

Capsule technique with hydroxides
Corrosion by sodium hydroxide

Corrosion by potassium hydroxide
Corrosion by other hydroxides

Corrosion by binary hydrogenous systems

Static Corrosion by Fluoride Coolant Mixtures
Static Corrosion by Sodium

Static Corrosion Test of a Reactor System
Mass Transfer Phenomenon in Static Corrosion

Chemical-reaction mechanism for mass traansfer
Experimental evidence

Dynamic Corrosion Tests in Thermal-Convection Loops

Corrosion by lithium and lead
Corrosion by seodium

Dynamic Corrosion Tests in Forced-Convection Loops

11. PHYSICAL PROPERTIES AND HEAT-TRANSFER RESEARCH

Investigation of Free Convection Within Liquid-Fuel Elements

Theoretical analysis of natural convection
Measurement of the fuel-element temperature distribution
Measurement of the fuel-element velocity distribution

Physical Properties

Heat capacity

Thermal conductivity of liquids
Thermal conductivity of solids
Falling-ball viscometer
Brookfield viscometer

Vapor pressure

Density

Heat-Transfer Coefficients

vili

 

PAGE
NO.

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104

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106
110

112

113
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118

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137
12.

13.

 

Heat transfer in fused hydroxides and salts
Heat transfer in boiling-liquid-metal systems
Heat transfer in molten lithium

METALLURGY AND CERAMICS

Welding of Inconel

Tensile tests

Fatigue tests

All-weld-metal tensile tests
Corrosion of welds

Special weld tests

Welding of Molybdenum
Creep of Metals in Controlled Atmospheres

Creep of 316 stainless steel
Creep of inconel :
Creep of niobium

Creep test of loaded inconel tube

Stress-Relaxation Tests
Fuel-Element Fabrication

Hot rolling

Mechanically formed matrix

Loose-powder sintering

Rubberstatic pressing

Compatibility of potential fuel-element materials

Control-Rod Fabrication
Metal Cladding of Beryllium Oxide
Refractory Metals

Ceramics Laboratory

CHEMISTRY OF HIGH-TEMPERATURE LIQUIDS

Fuel Development

Low-melting fluoride systems

Preparation of UF,
Homogeneous fuels

Moderator~Coolant Development

Preparation of pure sodium hydroxide
Preparation of other hydroxides

 

PAGE

137
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139

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165

1X
 

Decomposition pressures of hydroxides
Binary hydroxide systems
Hydroxide-fluoride systenms
Hydroxide-borate systems

Coolant Development

14. RADIATION DAMAGE

Creep Under Irradiation
Radiation Effects on Thermal Conductivity
Irradiation of Fluoride Fuels

Pile irradiation of fuel capsules
Cyclotron irradiation of fuel capsules

Corrosion of Iron by Lithium Under Cyclotron Irradiation

Liquid Metals In-Pile Loop

Part IV. ALTERNATE SYSTEMS

15. SUPERCRITICAL WATER REACTOR

Outline of a Specific Design
Fuel Elements and Assemblies
Reactivity

Stabilaty

Pressure Shell

Shield

16. CIRCULATING-MODERATOR-COOLANT REACTOR: HKF

Operational Characteristics
Reactor Characteristics
Core Design

Reactor Physics

17. CIRCULATING-MODERATOR-COOLANT REACTOR: ORNL

Fluid-Circuit Specifications

 

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188
 

Fuel-Element Design

Reactor Design

18. HIGH-TEMPERATURE POWER PLANT STUDIES

Sodium-Liquid-Vepor Cmmprassorljet

Helium-Cooled Reacters

Part V. APPENDIXES

19, ANALYTICAL CHEMISTRY

Analysis of Reactor Fuels

Spectrographic results
Development of colorimetric metheds
Determination of platinum

Oxygen in Sodium

Oxygen in Lead

Oxygen in Helium .

Oxygen and Nitrogen in Lithium

Carbon in Lithium

Uranium Trifluoride in Uranium Tetrafluoride
Identification of Residue 1in Lithium-Metal Coolant System

Analytical Services

20. LIST OF BEPORTS ISSUED

 

‘PAGE

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192

1923
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199

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201

201
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201
1202
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202

203
- 204

205

X1
3.4

3'5
8.1
10.1

10.2

10.3

10.4

12.1

12.2
12.3
13.1
13.2
13.3

13.4
13.5
13.6
13.7

x11

 

LIST OF TABLES

TITLE

Composition and Design Data of Core 93

Volume Fractions of the Materials in the Side and Bottom
Reflectors

Summary of Calculations on the ARE Core Having 25 1b of
Uranium and Reflector 520

Calculation of Uranium Requirement for the ARE Design of

10 June 1951
B4C Transmission Coefficients
Comparison of Shield Weights Using NH; and H,0

Summary of Fluoride-Corrosion Data Obtained in 100 hr at
1000°C (1830°F)

Summary of Corrosion Data Obtained in 100-hr Tests at 815°C with

PAGE
NO.

36

37

38

41
67
96

111

Mixtures of Sodium Hydroxide with Sodium Hydride, Sodium, or Water 119

Corrosion and Operational Data on Lithium-Containing Thermal-
Convection Loops

Corrosion and Operational Data on lead-Containing Thermal-
Convection Loops

Room-Temperature Tensile Properties of Inconel Tube-to-Header
"Pairs"

Inconel All-Weld-Metal Tensile Values

Results of Investigations of ZrO, as a Control Material
Summary of Promising Fluoride Fuel Systems

Solubility of Uranium in Hydroxides

Solubility of Uranium in Mixtures Consisting of Sodium
Hydroxide and Sodium Tetraborate

Purification of NaOH by Recrystallization from Ethyl Alcohol
Purification of NaQOH by Recrystallization from H,O
Decomposition Pressures of Ba(OH)z'(1.9)H20

Low-Melting Non-Uranium Fluoride Eutectics

 

125

126

142
144
152
156
161

162
164
164
166
169
 

TABLE

PAGE

NO. TITLE NO.

14,1 Results of 31-Mev Alpba Irradiation of Lithium in Iron Capsules 175
15.1 Summary of Reactor Design-Point Values 181
17.1 Design Coolant Condition for Maximum and Cruise Power 190
17.2 Temperature Throughout Sodium Hydroxide Core 190
19,1 Determination of Uranium Trifluoride by Two Methods 203
19,2 Backlog Summary | 204

X111

 
FIGURE
NO.

2.3
2.4
3.1
3.2
3.3

3.4
3.5
3.6
3.7
3.8

X1V

 

LIST OF FIGURES

TITLE

Fluid-Bearing (Canned-Rotor) Pump

Comparison of General Electric and ANP Electromagnetic Pump
Data (750°F, Sodium)

Centrifugal Pump with Frozen-Sodium Seal
Experimental Facilities

Schematic Drawing of ARE Core Arrangement
kefif vs. Uranium Weight for the ARE

Reactivity of the ARE Core Backed by Various Reflectors or
Additional Core Material

Reflector Savings of Various Reflectors Backing the ARE Core
Spatial Distribution of the Lethargic Average of the Fast Flux
Spatial Distribution of the Thermal Flux in the Core

Flux Spectrum in the ARE

Leakage Spectrum from the ARE Core to the Reflector

l.eakage Spectrum fromthe ARE Reflector

Spatial Power Distribution in the ARE

Radial Thermal-Flux Distribution in the ARE Reactor with
Seven Control Rods; Placement No. 1

Radial Thermal-Flux Distribution in the ARE Reactor with
Seven Control Rods; Placement No. 2

Radial Thermal-Flux Distribution in the ARE Reactor with
Seven Control Rods; Placement No. 3

Response of Flux to a Step Increase in Reactivity for Various
Reactivity Changes

Response of Fuel Temperature to Step Increase in Reactivity
for Various Reactivity Changes

Phase Plot of Flux vs. Fuel Temperature for Various Step
Increases in Reaectivity

Power Response of ARE to Step Reactivity Change of 0.009125
with Various Mass-Reactivity Coefficients

 

PAGE
NO.

18

19
22
29
35
37

39
40
42
43
44
45
46
47

418

48

48

50

51

52

53
FIGURE
NG.

3.18

3.19

3.21

3.22
3.23

3.24

3.25
3.26
3.27
3.28
3.29
5.1

502

6.1 -

7.1
1.2
8.1
9.1
9.2
10.1

10.2

 

TITLE

Response of Fuel Temperature to a Step Increase of Reactivity
of 0.009125 for Various Effective Mass-Beactivity
Coefficients

Phase Diagram of Flux vs. Fuel Temperature for Response of
ARE to a Step Increase in Reactivity of 0.009125

Response of Flux to a Step Increase in Reactivity of 0,002
for Various Average Neutron lLifetimes

Phase Diagram of Relative Fxcess Power (Flux) vs. Fuel
Temperature

Phase Plot of Flux (or Puwér) vs. ¥uel Temperature

Neutron Abscorption vs. Penetration with the B,C Layer of

the ARE

Abserption Spectrum for thfi 200-megawatt Sodium Hydroxide
Reactor

Leakage Spectrum for the 2fi0~megawatt Sodium Hydroxide Reactor
Fission Spectrum for the Bare Water-Moderated Reactor

Leakage Spectrum for the Bhre Water~Mederated BReactor

Fission Spectrum for the Rgflected Water-Moderated Reactor
Leakage Spectrum for the Reflected Water-Moderated Reactor

Preliminary Gamma-Bay Specfirum at 130 em from the Water-
Reflected Reactor |

Bulk Shielding Facility

Camparison of Shields for:3~ft Reactor

NeUtTOH Attennation in Water Areund Duct

§Water Patch Around Duct

;Typical Composite Shie]d!§H20~PthH3
Bremsstrahlung {rom Efl“pimfl Lithium Loop
Absorption of Bremsstrahl@ng from [.i? Beta Rays

Inconel Specimen from Pretreating Pot in Which Fluoride Bath
Mixture was "Deactivaced”™ at 950°C {for 100 br

Corrosion of Inconel by 3INa¥F-UF,

 

PAGE
NO.

54
55
57

58
59

68

73
T4
15
76
7
78

85
86
88
89
90
95
99
99
105
107

xy
FIGURE
NO.

10.3
10. 4

10.5
10.6
10.7
10.8
10.9

10.10
10.11

10.12

10.13
10.14
10.15
10.16
10.17

10.18
10.19
10,20
10.21
10.22

10.23

11.1
11.2

XVi

 

TITLE

Corrosion of 347 Stainless Steel by 3NaF-UF,

Heat-Treated Inconel Specimen

Corrosion of Inconel by Pretreated Fluoride Fuels

Heat-Treated 316 Stainless Steel Specimen

Corrosion of 316 Stainless Steel by Pretreated Fluoride Fuels
Corrosion of Inconel by Dehydrated Commercial Sodium Hydroxide

Corrosion of 316 Stainless Steel by Dehydrated Commercial
Sodium Hydroxide

Corrosion of Nickel A by Sodium Hydroxide

Corrosion of Inconel by Dehydrated Commercial Potassium
Hydroxide

Corrosion of 316 Stainless Steel by Dehydrated Commercial
Potassium Hydroxide

Corrosion of Inconel by Strontium Hydroxide

Corrosion of Iron by Strontium Hydroxide

Corrosion of 318 Stainless Steel by Strontium Hydroxide
Corrosion of Inconel by Pretreated Fluoride Coolant

Corrosion of 316 Stainless Steel by Pretreated Fluoride
Coolant

Corrosion of 316 Stainless Steel by Sodium

Metal Crystal Formation in Loop Containing Lead
Failure in 316 Stainless Steel Loop Containing Lead
Thermal -Convection Loop Operated with Lead

Cold Zone of 446 Stainless Steel Loop Showing Attacked
Surface and Metal Crystals Which Form in Lead (Dark Areas)

Cold-Zone Weld in 1010 Steel Loop Showing Metal Crystal
Formation Adjacent to Pipe Wall Covered with Iron Oxide,
Presumably Formed During Loop Fabrication

Schematic Diagram of Free Convection Apparatus

Thermal Conductivity of Copper

 

PAGE
NO.

107
108

108
109
109
114

114
115

115

116
116
117
117
120

120
121
127
128
128

130

130

133
136
 

FIGURE _ : fi : PAGE

NO. | - TITLE | NO.
11.3 Schematic Diagram for Determining Heat-Transfer Coefficients 138
11.4 Comparison of ORNL Lithium Heat- Transfar Data with Those of

Other Investigators | 140
12.1 Effect of Penetration of Fatigue Life on Inconel Tube-to-

Header Sp301mens 143
12.2 Effect of Cold-Working Stainless Steel-~U0, Cores 148
12.3 Effect of Cold-Working Irofi-U02 148
12.4  Effect of Particle Size of U0, 150
12.5 Die for Rubberstatic Pressing 151
13.1 The System NaF-LiF-UF, | | 157
13.2 The System KF-LiF-UF, | | 158
13.3 The System RbF-LiF-UF, 159
13.4 The System NaF-KF-RbF-UF, 160
13.5 Sodium Hydroxide Purification Apparatus | 163
13.6  The System Sr(OH),-Ba(OH), 167
13.7 The System NaOH-LiOH | 168
14.1 Composite Plot of Three Bench (o) and Three In-Pile (o) _

Cantilever Creep Curves _ 172
14.2 Liquid-Fuel Sample Container with Thermocouple Well and :

Pressure Fittings 173
15.1 Supercritical Water Reactor Flow Arrangement 180
15.2 Cross-Section of Fuel-Elemeat Assembly | 182
16.1 Schematic Drafiing of Reactor Core | 185
17.1 Sodium Hydroxide Cooled and Moderated Reactor 189
17.2 Temperature Throughout Sodium Hydroxide Core 190
17.3 Sodium Hydroxide Velocity and Film Coefficient Throughout Core 191

XV11

 
 

 
 

SUMMARY

This quarterly progress report
of the Aircraft Nuclear Propulsion
Program at the Qak Ridge National
Laboratory i1s divided into five parts:
- RBeactor Design, Shielding Hesearch,
" Materials RBesearch, Alternative
Systems, and Appendixes, each of which
15 discussed separately in this summary.

Part I.

REACTOR DESIGN

The Aircraft Reactor Experiment
{$ec. 1). The design of the high-
temperature Aircraft Reactor Experi-
ment has been modified to use the 30-
1n. core instead of the 36-in. core,
as proposed in the last report. All
reactor components are on order with
deiivery expected this coming quarter.
Design of the external fluid circuit
is essentially complete, and many of
the major components are on order.
Construction of the building facility
is on schedule.

Experimental Beactor Engineering
{Sec. 2)., Recent developments have
included a frozen-sodium seal for a
centrifugal pump, satisfactory cper-
ation of an electromagnetic pump and
of a canned-rotor pump with sodium,
operation of a liquid-metal valve at
1500°F without self-welding, and oper-
ation of a mock-up of the reactor
liquid-fuel system. Technigues have
been developed further for the handling
and purification of sodium and sodium
systems to minimize oxygen contami-
nation; this bas contributed appreci-
ably to the successful eoperation of
these systems at high temperatures.

Reactor Physics (Sec. 3). The
resctor physics calenlations have been
dz=voted largely tostatics and kinetics

 

of the ARE, although some analysis of
a sodium hydroxide reactor and a full
aircraft-size BeO reactor were com-
pleted, The uranium investment 1n the
ARE 1is 29.2 lb, the slight increase
over previous estimates being due
largely te increased structural and
material poisons in the completed
design. As the change in reactivity
for normal control is relatively
insensitive to the location of the six
cuter control rods, they have been
located to maintain the most unifeorm
flux distribution. Calculated kinetic
responses of the ARE to arbitrary
changes in reactivity, as a result of
possible accidents or failures, show
the system to be well damped. Pre-
liminary calculations of a sodium
hydroxide reactor designed at ORNL
have yielded data on leakage and
absorption spectra of the core and
indicate a critical mass of approxi-

mately 32 1b.

Critical Experiments (Seec. 4). The
preliminary reactor assemblies which
are being, or have been, investigated
include simulations of the air-water
aircraft reactor, proposed by General
Electric, and a graphite-uwranium
assembly to study power and flux
distributions. Measurements of this
latter reactor were still being made
at the end of the quarter. The two
modi fications of the air-water reactor
demonstrated the savings in critical
mass to be gained by decreasing the
thickness of the water lavers, ‘

Part Ii. SHIELDING RESEARCH

Bulk Shielding Reactor (Sec. 5).
The measurements on the mock-up of the
 

ANP PROJECT QUARTERLY PROGRESS REPORT

ideal unit shield have been completed
and indicate a weight of 1Z4,000 1b
for a 3.8-ft spherical reactor. Con-
struction of the mock-up of the divided
shield i1s essentially complete, and
preliminary shadow-shielding tests
have been completed.

Lid Tank (Sec. 6). Recent measure-
ments on the lead-—borated water unit
shield, i1n which the boron concen-
tration has been 1ncreased to 1.3 wt %,
result in a unit shield weight for a
3-ft spherical core of 101,200 1b,
6000 1b lower than that calculated by
the Shielding Board in ANP-53,

ago.

a year

Duct Test (Sec. 7). A practical
treatment for liquid-metal ducts 1in a
reactor shield has been demonstrated
by the use of an extra comical layer
of water around the duct as it emerges
from the reactor shield. The weight
of such 2 patch for 2 single 6-1n.
sodium duct surrounded by a l-in. air-
filled annulus is approximately 1000 1b.

Shielding Talculations (Sec. 8).
Analysis of Lid Tank data has yielded

"removal cross-sections" for lead,
boron, and water of 3.4, 2.0, and 0.8
barns, respectively. The theoretical

analvsis of the divided shield 1s
again being scrutinized very care-
fully. It seems likely that, when
some of the approximations are better
defined,

greater

the weight may be somewhat
than previously supposed,
although not prohibitively s0. A
weight advantage of several tons by
using ammonla i1n the divided shield
in preference to water has been demon-
strated. An analysis of some possible
radiation hazards associated with the
ARE has also been made.

Nuclear Measurements (Sec. 8). The
5-Mev Van de Graaff generator is inm
operation, and preliminary experiments
indicate its useful range of energies
to be from below 0.2 Mev up to 6 Mev,
An in-pile lithium loop has yielded
data on Bremsstrahlung intensities,
and measurements of the (n,2n) reaction
in beryllium were made, The instal-
lation of the neutron time-of-flight
spectrometer at the LITR has progressed
although fabrication of the assembly
has been seriously delayed during the
quarter,

Part II1I. MATERITALS RESEARCH
Investigation of the materials
problems of a high-temperature reactor
continues to comprise a major part of
the effort of the Aircraft Nuclear
Propulsion Project. 1In addition to
the empirical research on corrosion,
radiation damage,
cation,

materials fabra-
and reactor chemistry, this
program includes the determination of
the basic thermal and physical con-
stants associated with these materials
at reactor temperatures.

Corrosion Research (Sec. 10). Cor-
rosion studies have been expanded to
include corrosion by hydroxides and
possible fluoride coolants as well as
corrosion by the liguid metals and
fluoride fuels already under 1in-
vestligation for some tilme. The
extensive corrosion tests with sodium,
both statiec and in thermal convection
loops, conclusively demonstrate that
sodium causes negligible corrosion to
elther inconel or a number of stain-
less steels at 1500°F. Static cor-
rosion tests of the pretreated fluoride
fuel mixtures indicate that containing
of these liquids 1s likewise possible
under the same circumstances. Cor-
rosion attack by such pretreated fuels

 
 

FOR PERIOD ENDING SEPTEMBER 10,

averages about 1 mil on stainless steel
and 2 mils on inconel after 100 hr at
1500°F. Corrosion attack encountered
with hydroxides and with hydroxide-
bearing materials, however, 1s a great
deal more severe. Both commercial and
specially pure sodium hydroxide are
extremely coerrosive to inv@nel and to
the stainless steels. Corrosion by
potassium hyvdroxide 1is 5emewhat less
severe. Tests with barium and strontinm
hydr@x1fiea indicate these canstics to
be somewhat less corrosive than @h,“
of the alkali hydroxides, but sxili 50
severe as to preclude their use at
this time. Rigid controel of the
purity of the waterial, which was
necessary with the fluorides, has not
yet been realized but should reduce
the present hydroxide cerrosion rates.

Physical Properties and Heat-Trams-
Fer Research {Sec. 1i1). Of particula
interest in the design of the reactor
are i%n measuremenis of the physical
properties of probable reactor ma-
Epridis as well as the studies of
heat-transfer phenomena and assoriated
coefficients. Heat-capacity measure-
ments have been completed for 316
stainless steel, lithium, molybdenum,
and zircoeaium in the range 300 to
1006°C, The thermal conductivity and
density of the flvorade fuel have also
region of interest,
measurement of the

been defined in the
Apparatus for the

viscosity of high~temperature liguids
and the heat-transfer coefficieats of
hydroxides and fused salts is now
easentially complete. Investigation
of the free-convection mechanism within
quiescent liquid-fuel elements 1is
underway. ‘

Metallurgy and Ceramics {(Sec. 12).
The metallurgical processes involved
construction and assembly of a
in-

in the
high-temperature reactor core,

 

1951

cluding welding of tubular fuel
elements, fabrication of s0lid fuel
elements, creep of metals, fabrication
of control rods, and cladding of
beryllium oxide moderator, are cur-
rently being investigated. It is now
apparent that tube-to-header welds
having tensile strength comparable to
that of the parent metal may be resdily
gffected with znconsel by an electro-
magnetic "come-arc” technigue. Added
to the list of technigees by which
solid-fuel pfiwments may be fabricated
s that of "rubberstatic pressing."”
in all technigues the uss of 2 screensd
fraction of sintered UQ, secems to he
desivable. The creep laber atory and
stress-corrosion laboratory are now in
operatlon and are emphasizing tests on
incenel and stainless steel specimens.
A ceramic labersatery has besen set up
and 13 bheing equipped and staffed.

Chemistry of Gigh-Tenperature
Liguids (8sc., 133Y. The chemical
research on reactor fluids has been
extended to clude study of nen-
metallic liquids for use as moderators
and/er heat-travnsfer fluids, in ad-
dition to the development of liguids
for use as high-temperatnre
fuels. Nine fluoride fuel systems,
both ternary and quaternaty, are
singled out as covering a useful rangs
of uranium concentration
ing satis

reactor

and possess-
sfactery melting points. In
the development aof hmmfigenfl@ua reactor
fuels, solmtinnsfififi) in NaOH-Ne B, O,
show premise, in dd&l tien $o the
NaOH-1.10H system previously developed.
The recent investigations for moder ~
ator coolants have indicated that
several bimary hydrogenous systems
appear to be satisfactory as far as
liquid ranges, moderating ability, and
heat-transfer properties are concerned.
Preparation of sodium hydroxide of
greater than 99, 8% NaOH by weight was
 

ARP PROJECT QUARTERLY PROGRESE REPORT

required in the above and in the
assocliated corrosion studies. The
list of nonmetallic coolants now imn-
cludes 11 fluoride systems of usable
liguid range and low corrosiveness,
although their heat-transfer properties
are not sufficiently well known to
permit evaluation of their usefulness.

Radiation Damage (Sec, 14). Although
a number of fuel capsules have been
irradiated in both the pile and
cyclotron, complete data are not yet
available ewing to the residual activaty
of the capsule. From preliwinary
results 1t appears unlikely that
radiation will have a2 significant
effect on the fuel. On the other
hand, the thermal conductivity of
incoenel appears to decrease by a factor
of 2 iu less than a week of exposure
in the X-10 pile. The flux devendence
of this decrease has not been de-
termined, but a temperature anneal of
the effect has been demonstrated.
As regards creep under irradiation,
there 1s a definite reduction in
primary creep due to airradiation with
347 stainless steel. However, after
long periods of strain, 200 hr and
irradiated specimens exhibit
significantly greater elongations than

abeove,

control specimens. Twe preliminary
experiments on irradiating lithium ia
iron capsules showno appreciable added
corrosion ascribable to radiation
effects.

Part IV, ALTEBNATIVE SYSTEMS

The major effort of the Aircraft
Nuclear Propulsion Project at Qak
Ridge Natienal lLaboratory is directed
toward a 1500°F liquid-fuel reactor.
However, research is 1in progress both
here and at associated laboratories on
several other reactor systems which

show promise. Among these alternative

systems recelving consideration are
the supercritical water reactor and
two configurations for a cirvculating-
moderator reactor.

Supercritical ¥aisr Reactor (Sec.
15). The design of a supercritical
water reactor 1s being amalyzed by
Nuclear Develcpment Associates, Inc.
Their present concept of the reactor
and shield 1s a 2.5-ft square-cylinder
active core surrounded by an 11-ft-
diameter sphere of water. The water
makes two passes through the core
which contains fuel elements of the
sandwich design previously proposed.
Studies of reactivity show that uniform
thermal flux i1s simultaneously con-
sistent with minimum critical mass,
approximately 25 1b. The
designed to deliver 400 megawatts wath
a maxlmum wall temperature of 1290°F,

reactor 1S

Circulating-#oderator-Coolant
Reactor: HKF (Sec. 18). A circulating-
moderator-coolant reactor ewploying
circulating sodium hydroxide as both
moderator and coolant and a fixed
ligquid fuel of a mixture of fluorides
has been proposed by The H. K. Ferguson
Co. The reactor is designed to deliver
140 megawatts at the desigu point of
0.8 Mach and 35,000 ft in a modified
B-52 airplane, with a maximum power at
of 230 megawatts. The
uranium investment in this reactor is
high, 187 l1b. This 1s directly at-
tributable to the poor heast-transfer
characteristics of sodium hydroxide,
which necessitates a large amount of
inconel for heat-transfer surface.

sea level

Circulating—¥Moderater-Coelant
Reastor: ORNL (Sec. 17). A design
of a circulating-moderator-coolant
reactor has also been advanced by the
Reactor Design Group at Oak Ridge
Nationsal Laboratory. The working fluid

 
 

FOR

of this moderator-coolant reactor is
likewise specified as sodium hydroxide.
A 2.5-ft spherical core using liquid
fluoride fuel 1s expected to deliver
200 megawatts with a maximum wall
temperature of 1500°F. An essential
feature of this design is the use of
eannular fuel elements to attain the
high ratio of heat-transfer surface to
fuel volume necessary with the hy-
droxide coolant. '

High-Temperature Power-Plant Studies
{Sec. 18). North American Aviation,
Inc., has conducted an investigation
of high-temperature (above 1800°F)
helium and sodium liguid-vapor power
cycles with regard to their application
to a Mach 1.5, 45,000-ft-altitude
aircraft. They conclude that a helium-
cooled reactor cannot achieve the
supersonic propulsion of aircraft
even 1f reactor temperatures as high
as 3300°F and helium pressures in the
range 1000 to 2000 psi should be used.
On the other hand, a liquid-metal-
cooled reactor operating in conjunction
with a sodium liquid-vapor compressor
Jet system does appear feasible for
the supersonic aircraft.

 

PERIOD ENDING SEPTEMBER 10, 1951

Part V. APPENDIXES

Analytical Chemistry {(Seec. 18).
Chemical analysis is required in
almost every phase of the reactor
program. Although some of these
analyses are routine, the development
of many new analytical techniques is
required. In all, over 400 samples
were submitted for analysis during
the last quarter and over 1700 determi-
nations were made. The n-butyl bromide
me thod, which has been developed for
the determination of oxygen in sodium,
1s extremely accurate for oxygen
contaminations down to 0.015%.  The
oxygen content of argon and helivum 1s
now determined by a colorimetric
method which gives excellent precision
below 25 ppm. Methods are also being
developed for the determination of
metallic corrosion products in fluoride
fuels and metallic coolants and for
the determination of oxygen in lithium
and lead. '

List of Reports Issued (Sec. 23).
The reports issued during the past
quarter include scme fifty reports
on all phases of the ANP program.
 

 
 

 

 

 

 

 
 

1. THE AIRCRAFT REACTOR EXPERIMENT

W. M. Breazeale
G. A. Cristy

L. F. Hemph1ll
5. V. Manson

R. W. Schroeder
ANP Division

The Aircraft Beactor Experiment
(ARE) is a 3-megawatt reactor designed
to provide first-hand experience with
a high-temperature (1500°F) reactor.
Recent modifications of the ARE design
have been permitted for convenience
where this has not affected the operat-
ing temperature of the core. In
particular, the smaller core, proposed
in the last report, (') has been adopted
and consideration 1s being given to
the use of NeK since this coolant would
eliminate the need of preheaters.

The external-fluid, electrical, and
control circuits have been generally
established and detailing is now 1in
process. The design of the fluid
circuit is sufficiently well defined
to permit ordering of heat exchangers,
blowers, and associated tubing. The
reactor control signal will now be
obtained from both inlet and coutlet
coolant temperature. The reactor will
contain si1x control rods symmetrically
dispersed around a central safety rod.

Construction of the building facility
for the AREis proceeding as scheduled.
The excavation 1s complete and most of
the concrete foundation has been laid.

 

{I)Nd M. Smith, Jr., “*‘Becommendation on
Alternative Loading,’ Aircraft Nuclear Propulsion
Project (uarterly Progress Report for Period
%ggigg June 10, 1951, ANP-65, p. 55 (Sept. 13,

 

CORE DESIGN

The core arrangement has been re-
designed to provide for an active
lattice 30 in. in diameter and 34 1in.
in length, as suggested by the Reactor
Physics Group.{!? (The smaller core
and thick reflector favor critical
mass and power distribution.} The
original core-tube and fuel-tube sizes
have been retained, providing a maximum
fuel capacity of 31 1b of U?3%%, using
fused fluorides (NaF—KF-[W;)containing
150 1b of U?3® per cubic foot. The
current core—pressure shell assembly
provides for seven control thimbles,
located in the core, and two instrument

chambers, located in the reflector.

Material and component orders have
been revised to include the changes
and features referred to above. Detail
and assembly drawings for the entare
core-—pressure shell assembly are
being prepared. The orders for core
components which were placed during
the preceding quarter®?) are still
outstanding. Delivery of most items
is expected within the next twomonths.
After extensive negotiations with the
Norton Coempany, the Brush Company,
AFC, and Norris, it has been deczded
that all moederator and reflector
beryllium oxide blocks will be hot-
pressed by Norton and that the original
bilock si1ze {(3.75 flats?

in. BCross

 

(2)8, W, Schroeder, "Design of the Aircraft
Reactor Experiment’ ANP-65, op. cit., p. 10.°
 

ANP PROJECT QUARTERLY PROGRESS REPORT

will be retained. Initial shipments
of BeO have been made.

FLUID CIRCUIT

Detailed layouts of all aspects of
the external fluid circuit are currently
being made. The availability of the
various components from either at-hand
stock or commercial suppliers 1s being
determined, and in some cases the
procurement of these i1tems has been
initiated. The use of NaK as the
reactor coolant will eliminate the
need of preheaters in the coolant
system. The helium monitoring circuit,
the reactor room and pump-room space
cooling circuits, and the control-rod
cooling circuits are being reanalyzed
in an attempt to obtain dual usage of
some of the components.

Heat-Disposal Circuit. The heat-
disposal systems provide for heat
transfer from NaK to helium to ethylene
glycol to water. The NaK-to-helium
heat exchangers and the helium-to-
glycol heat exchangers have been
studied in detail, and the configu-
rations which appeared to be most
attractive have been reviewed with two
prospective suppliers, the Griscom-
Russell Company and the Vulcan Copper
and Supply Company. FEach of these
companies confirmed our basic approach
and have submitted formal gquotations
for the heat exchangers. As each of
the prospective suppliers has agreed
to quote on the basis of l-in. inconel
tubing in the NaK heat exchanger, the
tubing has been requisitioned to favor
the optimum delivery date. The afore-
mentioned companies also confirmed the
calculated helium flow rates and
pressure drops, enabling the release
of requisitions for the helium blowers.
Invitations to bid have been extended
to six blower manufacturers; the bids

10

‘released to the stack.

have been received and currently are
being evaluated. The glycol-to-water
heat exchangers have been found inY-12
surplus, and their custody has been
transferred to the ARE project.

It has been decided to revise the
NaK system to provide for upward flow
from reactor to an expansion tank,
then down from the expansion tank to
the heat exchangers and down from the
heat exchangersto the low point immedi-
ately above the dump tank, and from
there to the pump and the reactor.
This revision places the pump on the
low-temperature side of the heat ex-
changer and obviates the need for
employing the pump as a surge chamber.
It 1s planned to maintain pump level
control by means of a forced helium
flow through an orifice, the area of
which 1s adjusted by the NaK level.
It has also been decided to employ a
mechanical valve in the dump tank line
rather than to maintain the liquid
level by gas pressure as previously
contemplated.

Contamination of Helium System.
Failure of a fuel element 1s expected
to release xenon and krypton to the
NaK circuit and in turn to the helium
present at the various NaK free sur-
faces. Studies have indicated that
the helium so contaminated must be
held for several days before being
After several
possible scrubbing arrangements had
been investigated, 1t was decided to
make this helium circuit a closed
loop with a valved line to the stack,
permitting release after any desired
waiting time. This system will 1in-
clude a low-pressure reservolr, a
scrubber, a small compressor, amonitor,
aod a high-pressure reservoir.

 
 

FOR PERIOD FENDING SEPTEMBER 10, 1951

CONTROL OF THE ARE

F. 8. Bettis, Research
Director’s Division

The control of the ABRE will in-
corporate solid absorber rods 1in
addition to the liquid-fuel control

apparatus which decreases the fuel
volume in the core. The control
signal 1is now taken from both the

inlet and outlet coolant temperatures
rather than from the outlet tempera-
ture only. Detailed design of the
absorber control rods has been com-
pleted, anda rod and actuator assembly
is being fabricated. A high-tempera-
ture fission chamber has been designed,
and a satisfactory multiplier has been
the reactor

developed for dynanic

computor,

Liguid-~Fuel Control System. The
elementary diagram of the control
system has recelved major emphasis
quarter. A completed
elementary control diagram is now
ready . The control-room layout, in-
cluding operating console, relay and

during this

instrument racks,
conduits, has
pleted. Detailing of these items 1is
now being completed.

and interconnecting

been partially com-

A fundamental change in the control
the source of the
servo regulating signal was changed.

method was made when
This signal no longer originates from
the cooclant outlet temperature but 1is
obtained from both coolant ocutlet and
inlet temperatures., The equation for
this signal 1s given by

E = (T,

T

- T, + K(T, - T)

 

where

error signal actuating the

regulating rod

T, = coolant inlet temperature

T, = temperature of reactor core
circult at equilibrium condition
at start-up with reactor power
output essentially zero

K = constant determined by ratio
of outlet temperature rise to
inlet temperature drop from
start-up valuae T

T, ™ coolant outlet temperature

the
the

Using this equation a solution of
reactor kinetics is 1in process by
Nuclear Physics Group. :

The test rig, for testing the
operatien of bellows for moving high-
temperature liquids, has been completed
and 1s in the Experimental Engineering
building. It has not yet been put
into operation, and the tests will
not be started until some of the more
important loop tests have been com-
pleted.

The

been

80lid Absorber Rod Design.
number of absorber rods has
established, pending verificatvion by
the critical experiments. Present
designs show six rods symmetrically
disposed about the center with an
additional red at the center of the
reactor. Detailed designs of the
control rods, thimbles, and actuators
have been completed. The fairst work
order for one cemplete rod and actuator
assembly will go into the shop early
in the next quarter.

i1
 

ANP PROJECT QUARTERLY PROGRESS REPORT

Calculations on the heat generated
in the absorber rods show that these
cooled satisfactorily with

The rod design, which has
been completed, provides for the
coolant helium to escape i1nto the
reactor pit and provides a monitoring
atmosphere for the pit. Gas
adequate to cool B,C rods are being
provided even though the possibility
still

can be
helium.

volumes

of using hafnium absorbers
ex1sts.

A design of a fission chamber for
use at elevated temperatures has been
completed and such a chamber is being
fabricated. This chamber will be
ready for testing 1n about six weeks.

Electronic Computor Gesign. The
first step toward the solution of the
computor problem has been completed.
A satisfactory multiplier for taking
the product of two analogue voltages
has been constructed. Work will
continue on the computor design at the
present level of endeavor.

ELECTRITAL CIRCUITY

Finalization of some of the major
fluid-circuit features has permitted
an estimate of electric-power reguire-
Detailed investigation of
following the failure of

ments.
conditions
external power has indicated that the
heat capacity of the system 1s very
large compared to the post-scram heat
generation. Accordingly, 1t 1s un-
necessary to provide for continued
operation of the main heat-disposal
circuit. It 1s planned to provide for
operation of the NaK pumps, the
monitoring and control rod coolant
hlowers, and the space cooling systems,

following failure of outside power.

12

On this basis it appears that the
maximum battery power requlirement will
be less than 10 kw. The a-c—d-c motor
generator set appsars to be governed
by the design point condition, which
requires approximately 50 kwd-c power.
Warm-up of the entire system from room
temperature to 1500°F, within 24 hr,
appears to require 100 kw a-c¢ and
10 kw d-¢. Circuit diagrams are being
prepared on this basis.

REMOTE-HANDLING EQUIPHENT

Conferences have been held with the
Timken Bearing Company regarding bear-
ing selection and other design features
of the remote cutting machinery. On
the basis of these discussions final
layouts are being made, and requisi-
tions have been released for the larger
procured components. FExperiments are
in progress involving welding, cutting,
rescarfing, and rewelding a 2-in.-
thick stainless steel plate. These
tests will be repeated with a 2-1n.-
thick inconel plate when the material
currently on order is received.

BUILDING FACILITY FOR THE ARE

The facility design has been com-
pleted by the Austin Company and the
work has been awarded to
At the present

construction
the Nicholson Company.
time the building excavation 1s com-
plete, the concrete pouring 1s near-
ing completion, and the building steel
and the concrete reinforcing steel are

on order. The Atomic Energy Commission

has suggested that certain portioms of
the

facility work that Oak Ridge

 
 

"FOR PERIOD ENDING SEPTEMBER 10, 1951

National [Laboratory had intended to
perform could advantageously be con-
tracted to the Nicholson Company in
the i1nterest of hastening the final
completion date. Accordingly, AFEC has

 

been advised of several services that
Nicholson can perform and that other
such items probably will become indi-
cated when ORNL facility engineering
has progressed further.

13
 

ANP PROJECT QUARTERLY PROGRESS REPORT

2. EXPERIMENTAL REACTOR ENGINEERING

H. W.

The ANP Experimental Enginesering
Group has the responsibility for
developing and/or testing materials,
methods, and components applicable to
the ARE and ANP reactors. Presently
under development are pumps, valves,
level and pressure indi-
preheaters, heat
etc., and, in addition,
studies are being made on static and
dynami¢ corrosion, methods for handling
liquid fuels, purification and handling
methods for ligquid metals and blanket-
ing gases, full-scale test facilities,
test
Information derived from
this program provides fundamental
engineering data for component design
and provides a basis for establishing
adequate operational procedures to
demonstrate the reliability of both
laboratory and full-scale systems.

flowmeters,

cators, seals, ex-

changers,

and operational techniques for
equipment.

Work has continued on liquid-{fuel
systems with the result that a non-
enriched sample of fluoride fuel
mixture has been filtered and trans-
ferred from one fuel container to
another for the first time. A frozen-
sodium seal had been in operation for
700 hr at the end of the period, seal-
ing agalnst 26 psi 1in a system which
contained sodium at 1500°F. A pump
incorporating this frozen-sodium seal
had also been put 1nto operation
svccessfully. In other tests, a
graphite gas seal showed promise.
Satisfactory operation was obtained
with a General Electric type G-3
electromagnetic pump during tests to
reproduce G. E. performance curves;
good agreement was obtained up to
750°F.

14

Savage, ANP Division

Stress-rupture and self-welding
tests were well underway, with early
resnlts indicating that moderately
stressed inconel and 316 stainless
steel suffered very little, i1f any,
lncrease 1n corrosion from sodium at
1500°F. Inconel tends to weld to
inconel only slightly in sodium at
1500°F;

steel.

the same 1s true for stainless
However, zirconium shows no
tendency to weld to stainless stsel.
Methods were developed by which aging
and double filtration reduced the
oxygen contenl of sodium to less than
200 ppm, and blanketing-gas purifi-
cation equipment reduced the oxygen
content to below 15 ppm. Specafi-
cations were drawn up and engineering
drawings are being prepared for full-
scale testing of ARE components, both
individually and as a unmit,

LIQUID-FUEL SYSTEMS

E. Wischhusen and D. R.
ANP Diavision

Ward,

Present plans call for the first
ARE reactor to employ fuel sealed
inside i1ndividual pins and inserted in
the reactor. A second rszactor core
may contain the fused fluoride fuel in
a fuel system consisting of perhaps
ninety fuel tube clusters in which tube
clusters would be filled and emptied
by means of differential gas pressure.
Full-scale mock-ups of portions of the
system are being made to reveal in-
formation on such i1tems as bubble
formation, filling and dumping charac-
teristics, valving, and operating
techniques.

 
 

- FOR PERIOD ENDING SEPTEMBER 10, 1951

tlock-up No. 1 was a crode-but-guick
system constructed froem glass tubes,
rubber stoppers, etc. and using colored
water to simulate the fuel. It served
primarily to indicate valving arrange-
ments, gage requirements, and operat-
ing techniques which would be needed
for working with mock-up No. 2.

Mock-up No. 2 was a carefully con-
structed all-glass full-size fuel
system employing tetrachlorvethylene
(sp. gr. 1.6) to simulate the liquid
fuel. Work with this system revealed
data on bubble formation, filling and
empiying rates, flushing methods, etc.,
all of which were applicable to mock-
up No. 3.

Mock-up No. 3, which bas been
assembled, 1is an all‘meta] system con-
structed from 316 stainless steel
3/16-in.-0.d. tubing and is capable
of operating at temperatures up to
1500°F. In a preliminary test using
tubing of this size the fuel mixture
of fluorides has been successfully
filtered and transferred through the
tubing in the molten state.

In conjunction with uranium fuel
transfer experiments, it was decided
to investigate the containing of
helium in equipment fabricated from
316 stainless steel and 1nconel of
0.030 in, wall thickness at tempera-
tures of 1500°F or greater. These
experiments were conducted on weld-
free type 316 stainless steel and
1nconel tubing at 1600°F under 54 psi
helium pressure differential and using
a Westinghouse mass spectrograph
helium leak detector for detecting
helium. It was foudd that no helium
diffused interstitially through the
tubing walls during 150 hr, the length
of the test. ‘

 

PUMP LEVELOPMENT

Pomps currently under development
for the ARE and experimental loops
include a variety of electromagnetic
and centrifugal pumps as well as the
unique canned-rotor pump and Mlevel-
tank” pump. Performance curves of two
electromagnetic punps have been obtained
up to 1000°F. At this temperatnre a
high rsontact resistance created an
open The centrifogal pumps
currently development are
distinguished by the type of shafe
seal. The centrifugal pump discussed
below ‘embodied a gas seal; other
promising pump seals are discussed
under "Seal Tests."” The canned rotor
pump has operated for 90 hr with NaK
at temperatures up to 400°F with no
discernable wear. Awong the pump
developments are a Duriron pump which
operates immersed 1n the to-be-pumped
fluid and the two-stage electro-
magnetic pump. Other pump information
1s mentioned under "Loop Tests™ and
"Seal Tests.™

circuilt.
under

A loop for testing either electro-
magnetic or mechanical pumps of ARE
size has been designed which contains
approximately 40 ft of 2J-in. pipe and
utilizes 7.5-kw tube furnaces for
heating the circulating medium up to
1500°F. The loop volume, including a
3-ft3 sump tank, 1is approximately
5 ft®. Space is provided for electro-
magnetic or venturi type flowmeter in-
strumentation.

Centrifugal Pumps for Figure-Eight
Loops (W. G. Cobb, ANP Division). The
centrifugal pump for liguid metals
described in the previous reportf?)

 

(1)w. G. Cobb, “Pumps,” Aircraft Nuclear
Propulsion Project Quarterly Progress Beport for
Period Ending June 10, 1951, ANP-635, p. 167,
esp. p. 168 (Sept. 13, 1951).
 

ANP PROJECT QUARTERLY PROGRESS REPORT

was 1nstalled in a figure-eight loop.
Teflon packing rings were used for
shaft sealing, and level indication
and control were taken from the liguid
surface in a tank mounted at the side
of the pump housing. Flow observed
during operation was 7.5 gpm at ap-
proximately 45 ft of sodium, but the
difficulties of simultanecusly con-
trolling the three liquid levels 1in
the system under dynamic conditions
inhibited smooth operation. The
initial run lasted for 3% hr at a
sodium temperature of 795°F before
failure of the shaft seal;
operation the final
% hr. A second run lasted 16 hr,
reaching a temperature of 900°F, but

however,
was smooth for

terminated when the pump became starved
and hence was unable to pump smoothly.

Examination of both these runs
after shutdown i1ndicated a failure of
the ligquid-level control system. In
the first run the pressure equalizer
line between the pumpand level control
tank became plugged, allowing sodium
to rise through the seal. This satu-

ation was corrected by additional

heaters and insulation on both the
ligquid and gas lines between the
pump and level control tank. The

explanation for the intermittent
pumping 1n the second test was failuve
of the level indication and control
equlpment to maintailn a satisfactory
liquid level in the pump. Subsequent
checks did reveal considerable variation
between the two liquid levels during
operation. Difficulties encountered
1n maintaining constant liquid level
in the surge tank and in the pump
indicated that this method of control
was impractical.

ARE Pump Design (W. G. Cobb and
J. F. Haines, ANP Division). A con-

tract has been let with Allis-Chalmers

16

Company for design and fabrication of
a sump type internal-bearing direct-
driven centrifugal pump for testing
and possible use with the ARE. This
and other pumps of ARE capacity being
designed by ANP Experimental Engincer-
ing personnel, are being guided by the
following considerations:

1. The pump is preferred to be
located on the cold side of the
reactor to minimize high-temper-
ature effects on pump structure.

2. Separate pump sump and surge
tanks are to be provided.

3. An 1individual level control is
to be provided for the pump.
Tests are being performed to
determine the feasibility of
controlling pump level from a
dynamic surface.

4. A gas seal for the pump shaft

1s to be used.

5. Flow direction through the pump
1s to be comnventional with
suction through the lower side
and discharge from a concentric
casing.

6. The to be carried
by an overhung vertical shaft
with no internal bearings.

impeller 1is

7. Pump cooling 1is to be ac-
complished by flowing helium,
which 1s to be monmitored for
coolant leakage.

8. Bearing lubricant 1s to be
circulated continuously; this
feature provides cooling for a
rotating face seal.

9. The controlled liquid surface

 
 

FOR PERIOD ENDING SEPTEMBER 10, 1951

in. the pump is to be as small
as compatible with stability.

10, The initial pump for testing is
to use a cast impeller with
casing and other parts to be
fabricated and machined.

Canned-Rofor Pump (A. R. Frithsen
and M. Richardson, Reacter Experimental
Fngineering Division). The 1%-hp
fluid-bearing pump (Fig. 2.1}, de-
scribed in the last report, (?) has
been operated with water and NaK
without detectable wear. The first
test was made in a water~circulating
system and continued intermittently
for a total of 1200 hr running time.
Examination of the pump parts after
dissassembly showed no visible or
measurable wear or corrosion. All
clearances were the same as before the
test was begun. :

In a subseguent test the diametrical
clearance between the motor rotors and
motor cans was reduced from 0.017 to
0.012 in., and the pump was then used
to circulate NaK for a total running
time of 90 hr. During most of this
period the temperature was held at
200 to 300°F, although short runs
were made at room temperature and also
as high as 400°F. No indication of
wear or corrosion was found on any of
the parts upon dissassembly, Lut a
considerable accumulation of alkali
oxides was found at the tabs of the
motor cans and close to the walls of
the pump volute where the liguid 1is
believed to have been relatively
stagnant. A Fischer-Porter Flowrator
was used for flow measurements and
operated with complete satisfaction
even at the higher (400°C) tempera -~
tures. :

 

(2)/-\. R. Frithsen and M. Richardson, “Canned
Rotor Pump,”” ANP-65, ep. cit., p. 170.

 

The limitation on operating temper-
ature for this pump 1s believed to be
determined by thermal damage to the
insulation in the motor windings. To
increase the permissible operating
temperature, an order has been placed
with the General Electric Company for
winding two motor stators with class H
wire which 1s capable of withstanding
a temperature of 500°F. 1In addition,
a proposal has been requested from
Allis-Chalmers Company for special
wire which can be used at TO0°F. With
the latter winding and by cooling the
liquid fed into the motor cans, it is
expected that liquid metals or salts
can be pumped at 1000°F or higher.

A 3-hp fluid-bearing pump has been
designed and has been partially fabra-
cated to test a number of revisions in
the design of the smaller pump and to
determine the practicability of seal-
ing up the smaller model. This pump
is expected to be ready for check ‘runs
with water about October 1. |

Level Tank Pump (W. B. McDonald,
ANP Division). A Duriron conventional
pump has been modified for operating
while immersed in liquid metals. This
pump has a capacity of approximately
12 gpm at a 40-ft developed head. A
conventional packing gland is used to
seal the shaft, and a labyrinth fitting
closely about the shaft minimizes the
amount of liquid metal by-passed to
the level tank. This pump will operate
while tmmersed in liquid metal in a
level tank and will be driven by a
motor on an overhung shaft mounted on
top of the tank. ‘Actual testing
awalts the availability of a test loop.

Electromagnetic Pumps (J. H. Wyld
and A. L. Southern; ANP Division;
A. G. Grindell, Engineering and
Maintenance Division). A General

17
UNCINS¥FIED

ANP PROJECT QUARTERLY PROGRESS REPORT

s

FLOWRATOR

 

 

 

Fig. 2.1 - Fluid-Bearing (Canned-Rotor) Pump.

" UNEEFIED
FOR PERIOD ENDING SEPTEMBER 10,

Electric a-c electromagnetic pump was
modified by Experimental Engineering
personnel to utilize a flexible second-
ary conductor. Preliminary tests gave
flows of approximately 16 gpm of sodium
at 5 psi and 890°F. In a combination
performance and endurance test, a G. E.
type G-3 electromagnetic pump was
operated in the calibration loop (see
below) for 173 hr. Performance curves
were obtained at temperatures of 300,
500, and 750°F for comparison with
data supplied by G. E. The pump

 

HEAD (psi)

 

: L o o n 50 voiis
@ | I
0

0 2 4 6 8

—_— Its
@~_-:itf4LT1r~o-.__“. oo
4 t— ““‘--— -0

T ——n

[———=H

1951

failed at 1000°F owing to a current
lug melting loose from the cell wall.
This failure occurred at 250 volts
input but after performance curves had
been obtained for that temperature.

General Electric supplied data for
their type G-1 pump whose performance
closely resembles that of the type G-3

pump procured by ANP. A graph compar-

ing operation of the G-1 and G-3 pumps
1s shown in Fig. 2.2.

The efficiency

UNCLASSIFIED
DWG, 12876

® ANP DATA, PUMP 3216680, G-3
B G-E DATA, PUMP 9159849, G- 1

 

0 12 14 16 i8 20

FLOW (gpm)

Fig. 2.2 -
(750°F, Sodium).

Comparison of General Electric and ANP Electromagnetic Pump Data

 

N I L0
ANP PROJECT QUARTERLY PROGRESS REPORT

of these electromagnetic pumps in-
creases essentially linearly from 0.5%
at flows of 2 gpm to about 3% at flows
of 14 gpm. Although performance curves
were taken at lower temperatures, the
larger part of the operation was con-
ducted at 750°F, which was the maximum
temperature recommended for the pump
cell in use. Operating temperature
was increased to 1000°F to obtain
performance curves and to determine
the upper limit of temperature allow-
able with the pump cell supplied.

Engineering drawings for the two-
stage electromagnetic pump were com-
pleted and submitted to the shop.
Fabrication is incomplete.

TEST LOOPS

A. G. Grindell, Engineering
and Maintenance Division

The "calibration" loop and the
"sodium manometer' loop, both of which
were designed to calibrate and other-
wise check flowmeters, have been
operated with liquid metals during the
past quarter. To date the more sig-
nificant application of these loops
has been in conjunction with pump
tests.

Calibration Loop. The calibration
loop, which was designed primarily to
check electromagnetic flowmeters
against venturi types and to check the
performance of electromagnetic pumps,
has been equipped with heaters, thermo-
couples, pressure measuring devices,
a lock valve for operating on a con-
stant weight of sodium, and a static
cold leg for continually removing
oxygen from the circulating sodium.

20

The system was degreased with per-
chloroethylene, evacuated, outgassed
at 800°F, and flushed twice with
sodium at 800°F. 1In this condition
it has operated satisfactorily as a
test loop for the G. E. electromagnetic
pump described above.

Sodium Manometer Loop. The sodium
manometer loop is an experimental loop
designed to afford means of developing
testing, and gaining experience on
flow-nozzle manometer type flowmeters
and simultaneously developing and
testing electromagnetic flowmeters.
The first attempt to pump sodium dur-
ing the period resulted in electro-
magnetic pump cell failure and sodium
fire during the start-up.

The loop was dismantled, repaired,
and restarted. In this test the
electromagnetic pump was provided with
a flexible secondary conductor, and
during the first 2 hr of operation at
constant power input, flow increased
from 2.5 to 9 gpm. This increase in
pumping rate was attributed to 1in-
creased wetting of the pump cell wall,
and increased pump 1nput power over-
heated the flexible secondary con-
ductor. Pump power was decreased, but
loop temperature was taken up to 890°F.
The pump failed after27 hr of operation
owing to the inadequate current capacity
of the flexible secondary conductor;
another conductor 2% times the cross-
sectional area of the first 1is now
being fabricated.

SEAL TESTS

Development of seals for pumps
handling liquid metals or other coolant
at 1500°F has received attention
throughout the period. Tests have

 
 

FOR PERIOD ENDING SEPTEMBER 10, 1951

been conducted simultaneously with
pump experiments in some instances,
while 1n others tests have been con-
ducted in simulated equipment. The
seal types i1nclude a frozen-sodium
seal, metal-to-metal seals, and metal-
to-metal seals at the end of a bellows.

Frozen~Sodium Seal (W. B. McDonald,
ANP Division). A seal employing the
principle of solidifying a column of
sodium around a rotating shaft is
under development. The testing device
consisted of a pot containing sodium
at 1500°F. A sleeve extended from
below the sodium level i1n the pot to
the outside, and a shaft rotated inside
this sleeve. Sodium was forced by gas
pressure into the annular space between
the rotating shaft and the sleeve. A
refrigeration coil, to cool the sodium
to below its freezing point, was wound
around the outside of the sleeve and
soldered to 1t to increase thermal
conductivity. The rotating shaft was
1% in. in diameter with a diametrical
clearance of 0.032
shaft and sleeve. By the end of the
period this seal had operated for
approximately 400 hr against a pressure
of 18 psi1. The coolant employed was
kerosene refrigerated by passing
through a coil immersed in a dry ice
and ethylene glycol bath.

in. between the

At the end of the period a second
test using sodium at 1500°F but with
the pressure differential increased to
26 psi was underway. Also, a con-
ventional centrifugal pump was modified
by adding a frozen-sodium seal and was
installed in a pump test loop (Fig.
2.3). The relatively short space
between the pump and the bearings 1is
expected to limit the temperature to
which the circulating sodium can be
raised; however, this pump was de-

 

livering 15 gpm of sodium at 650°F at
2000 rpm at the end of the period.

Bellows Seal (A. P. Fraas and M. E.
LaVerne, ANP Division). A Dureg stain-
less steel centrifugal pump(3): was
reworked to provide radial holes in
the rear face of the impeller to keep
the seal cavity dry and was operated
for 80 hr with sodium at temperatures
up to 1000°F. Tests with water showed
that the design is effective in
scavenging liguid that leaks into the
seal cavity through the labyrinth seal
on the rear face of the impeller. The
test with sodium was discontinued when
finely divided sodium was found in the
silicone 01l circulated over the outer
side of the bellows seal. Since only
gas should have been i1n contact with
the inner side of the seals, and since
the liquid level contrel had given
difficulty with sticking of a solenoid
valve, the rig was disassembled for
inspection., All parts were found in
good condition except the bellows
which found to have two axial
cracks on the periphery of two con-
volutions. The test will be repeated
with another bellows.

was

Graphitar and Tool Steel Seal
(W. B. McDonald, ANP Division). A seal
consisting of a Graphitar (U. S.
Graphite) ring sandwiched between a
stationary member and a rotating member
made of Ketos tool steel sealed suc-
cessfully against a gas pressure of 10
psi at a seal temperature of 350°F for
450 br. Shaft speed was approximately
2000 rpm. This test was termimated at
450 hr in order to make changes in the
test equipment to enable NaK vapors to
be i1ntroduced to the sealing members.

 

(3)A. P. Fraas, Progress Report on Stainless
Steel Acid Pump Reworked To Test Special Features
for Operation with Liguid Metals, ORNL, Y-12 site,
report Y-F15-6 (Feb. 1, 1951).

21
UNCE®FIED

ANP PROJECT QUARTERLY PROGRESS REPORT

 

Fig. 2.3 - Centrifugal Pump with Frozen-Sodium Seal.

” UNCPASSBIIED
 

FOR PERIOD ENDING SEPTEMBER 10, 1951

These tests were underway at the end
of the period.

INSTRUMENTATION

J. F. Bailey, Consultant to
ANP Division

The development of suitable instru-
ments for indicating flows, pressures,
levels, ete. 1s an 1mportant con-
sideration for the adequate testing of
liquid-metal systems and for the
successful operation of the ARE.
Hecently considerable thought has been
given to these instrumentation problems,
with the result that a study has been
made of all types of level control
devices to determine the advantages
and the disadvantages of each. The
types studied include electroresistance,
electromagnetic, electroreluctance,
gas flow, and float, and recommended
applications for each have been com-
piled. This work has been done with
the view to designing a vigorous re-
search program to develop adequate
instruments for application by both
the Experimental Engineering and the
ABE groups.

spark plug type
probes are being used to control
levels. This method is simple but
unreliable because of periodic short-
ing out of the probes. FElectro-
resistance probes for: level indication
are to be investigated 1n an experi-
mental system which is presently under-
going fabrication. FExperimental work
1s to determine the merits of this
type of probe as a possible replace-
ment for the spark plug type.

At the present,

A model of the gas-flow control has
been set up in the Liquid Models Test

 

Laboratory, and the operating charac-
teristics are being studied with water
as the liquid. Experimental work on
other types is in the planning stage.

During the gquarter two magnets for
use with électromagnetic flowmeters
were made by using two U-shaped Alnico
V pieces and cold-rolled steel spacers.
These magnets givea flux of 1500 gauss
across a 1l4%-in. gap. Using L-in.
stainiess steel tubiang with these
magnets, a flow of 20 gpm pgives an
induced voltage of 10.6 mv.

The low-flux magnets have advantages
in that their use permits existing
equipment to be used for vecording
high flows, and braking effect in the
tlow circuit 1s reduced. When {flows
are below 10 gpm, the magnet strength
should be above 2500 gauss to allow
the same recording egquipment to be
used with higher accuracy.

STRESS-RUPTURE TESTS
J. L. Gregg, Consultant to:
ANP Division

Stress-rupture tests a¥e being con-
ducted by the Experimental Engineering
Group to determine the strength of
materials at elevated temperatures.
For testing, sections of a tube of
type 316 stainless steel and one of
inconel were machined to 0.015 in.
wall thickness, pressurized to 110 psi
with argon, and immersed in sodium at
1500°F. The gas pressure produced a
hoop stress of approximately 1500 psi
in the thinned section of the tube.
Ry the end of the quarter, these tubes
had been under test for approximately
500 hr without failure, indicating
that attack of sodium on moderately

23
 

ANP PROJECT QUARTERLY PRCOGRESS REPORT

stressed inconel and stainless steel
1is rather small. Equipment is being
assembled for testing sheet metal 1in
sodium at elevated temperatures.

SELF-WELDING TESTS

J. L. Gregg, Consultant to ANP Division
A. P. Fraas and M. E. LaVerne,
ANP Division

Self-welding tests are being con-
ducted in sodium at 1500°F to provide
information pertinent to seals, valves,
hearings, etc. for the ARE. Preliminary
results show that inconel has only a
slight tendency to weld to inconel
under these conditions.
true for stainless steel.
shown no tendency to weld to zirconium
in any experiment conducted thus far.

The same 1s
Inconel has

Consideration of their atomic
structure had indicated that zirconium
should be as unlikely to weld to 1iron-
chrome-nickel alloys as any metal
availlable. In order to test this
supposition, a 304 stainless steel
standard globe valve was reworked to
provide a zirconium washer floating on
the valve stem hetween the seat and
the valve disk of 347 stainless steel.
The reworked valve was operated 1in
sodium for 142 hr, 48 at 1200°F and
and 94 at 1500°F. Operation con-
sisted 1in a number of closings and
openings, the required opening torque
being taken as a measure
tendency toward welding.

of any

No indication of welding was obtained
during the test, and visual examination
after completion of the test disclosed
no signs of welding between the zir-
conium washer and either the seat or
disk. A second test 1s being made

24

with three valves in parallel on the
same rig, one valve having a zir-
conium washer, one a molybdenum washer,
and one just the standard stainless

steel valve disk.

VALVE-PACKING EXPERIMENTS

W. C. Tunnell, ANP Division

Since the ABRE system as designed
requires valves to operate at 1500°F,
a program for determining valve design
characteristics 1s underway. If con-
ventional valves should prove un-
satisfactory, 1t 1s likely that a
bellows type valve would be a suitable
alternative. Such a valve would re-
quire packing to provide safety
features and also to permit temporary
use of the valve 1n case of bellows
failure. Experiments to determine
satisfactory packing materials for use
with sodium are in progress.

Following isa list of the materials
tested and the pressures and tempera-
tures reached during preliminary in-
vestigatlions:

1. Amosite asbhestos: held sodium
for 21 hr at 1500°F under 15 psi

before leaking.

2. Graphite powder: leaked sodium
at 600°F under 10 psi pressure.

3. Met-1.-X: held sodium for 2 hr
at 1500°F under 15 psi
leaking.

before

4. Lead-mill slag: held sodium for
10 days at 1500°F under 15 psi
pressure; no leaks had occurred
before experiment was terminated.

5. Nickel metal powder: held sodium
for 10 days at 1500°F under 15

 
 

FOR PERIOD ENDING SEPTEMBER 10, 1951

ps1 pressure; no leaks were
observed before experiment was
terminated,

6. Soda ash: leaked sodium at
1300°F under 10 psi pressure.

Since this experiment was considered
a preliminary screening operation for
selecting possible packing materials,
no effort was made to operate the
valve stem when the sodium reached
1500°F. Other egquipment is being
designed with smaller clearances be-
tween the sealing parts of the pack-
ing glands to afford more conclusive
tests of powdered or granular materials.

CONTAMINATION OF LIQUID-METAL SYSTEMS

The corrosion of metal containers
by liquid metals, in this case primarily
is attributed to the oxygen
Nominal

sodium,
contamination in the system.
precautions to exclude oxygen from
ligquid-metal systems, including the
use of 1nert-gas blankets and cleaned
systems, had been taken, but they did
not succeed 1in limiting corrosion to
the desired level. It became apparent
that even the small remaining oxygen
contamination was a fault, and conse-
quently provisions for its reduction
have been developed. The three main
sources of oxygen, barring leaks, are
from scale in the mechanical system,
from the sodium charged to the system,
and from the inert-gas atmosphere used
with the system. Oxygen contamination
from cach of these sources is minimized
by technigques now in use. |

Cieaning of Liguid-ymetal Systems
and H. R. Bronstein, ANP

Ligoid-metal systems re-

(K. Devenish
Division).
quire careful cleaning prior to being

 

put into operation to remove welding
scale, oxides, absorbed gases, etc.,
to reduce contamination of circulating
coolant. These systems are now being
degreased with hot perchloroethylene,
after which they are evacuated and
outgassed at approximately the temper-
ature of operation. An experiment 1is
underway 1in the calibration loop to
test the efficiency of preconditioning
a systemwith sodium prior to operation.
In this experiment, in addition to
degreasing and evacuating, sodium was
circulated at 800°F and removed, and
clean sodium was introduced. This
second batch of sodium was circulated
for 3 to 4 hr at 800°F also and then

dumped. Clean sodium was provided for
operational use. By the end of the
period, the loop had operated for

approximately 175 hr with no apparent
difficulty due to faulty cleaning
procedures.

Equipment has been installed for
descaling stainless steel alloys by
the "Virgo Salts" process promoted by
Hooker Electrochemical Company. 1In
this process the metal 1s immersed in
a molten caustic, which converts scale
to salts soluble in hot dilute hydro-
chloric acid. After these salts are
removed by hydrochloric acid, parts
are brightened and passivated by a
dip in nitric acid. No results show-
ing the efficiency of this process
were available by the end of the
quarter.

Removal of sodium from systems after
operation is a problem also receiving
attention during the period. Sodium
1s now removed from small tube passages
by immersing the entire assembly in a
water bath which is heated up to 100°C
by live steam. When the residual
sodium hés melted, 31t i1s forced from
the small passages by gas pressure.

25
l'lllli'ii'lh[

ANP PROJECT QUARTERLY PROGRESS REPORT

Of course, some burning and small
explosions occur which require that the
operation be conducted in the open,
but no danage to equipment has been
experlienced. Sump tanks are now
emptied by heating to above the melt-
ing point of sodium and flowing it
either by gravity or gas pressure into
an ash can immersed im a heated o1l
bath which keeps the sodium from pro-
jecting above the surface, thereby pre-
venting a combination oi1l-and-sodium
fire,

Purification of Liguid Metals
(.. A. Mann, ANP Division). Methods
of purifying liguid metals to reduce
the oxide content to acceptable levels
for use in convection loops, figure-
el1ght loops, pump test loops, etc.
have been 1mproved. In the method
now 1n use scdium or NaK 1s held at
approximately 250°F for at least 24 hr
to allow suspended oxide particles
time to agglomerate. The metals are
then filtered through a S5-p sintered
stainless steel filter into another
fill tank. From this tank the metal
is again filtered through a 5-u filter
into the operating system., Analytical
results indicate that double filtration
consistently reduces the oxide content
to below 200 ppm. The results also
indicate that no appreciable reduction
1s achieved with further filtration.

Purification of Inert Gases (L. A.
Mann, ANP Division). 1Inert gases used
for blanketing ligquid metals in operat-
ing systems require purification to
minimize oxXygen content. Argon gener-
ally contains on the order of 20 to 50
ppm oxygen, and helium cylinders have
shown as high as 65 ppm (chemical
analyses are reported in Sec., 19). A
purification apparatus has been put
into operation which removes oxygen
from blanketing gases by bubbling

26

through a column of NaK. Tests thus
far indicate that use of this system
consistently reduces the oxygen con-
tent of gases used in operating systems
to below 15 ppm, which is considered a
safe level, As an added precaution,
helium cylinders have been set aside
for the exclusive use of the ANP
Experimental Engineering Group, and
these cylinders are never permitted to
go below 100 psi to prevent air from
being introduced. These methods, which
are currently 1in use,
satisfactory.

are considered

Analytical Results with Sodium
(H. R. Pronstein, ANP Division). Sam-
pling techniques instituted with
existing equipment during the period
removed lnconsistencies 1n analytical
results, making possible the standardi-
zation of limits of detection of
impurities in sodium.
analytical results

Evaluation of
(see Sec. 19) on
sodium samples taken at the time of
fill and showed that
filtered sodium generally contained
approximately 0.02% oxygen, but the
oxide content upon termination appeared
to be a function of the size of the
operating equipment,
fill to termination,

termination

From the time of
the oxygen content
of sodium 1n thermal-convection loops
rose from 0.02 te 0.03%,
responding results from figure-eight
loops were 0.02 and 0.06%, respectively.
These findings indicated the need for
more thorough cleaning and outgassing

while cor-

procedures; consequently the technique
of repeated flushing of a system with
hot sodium prior
instituted,

to operation was

Design of a sampling device to take
repeated samples of a coolant directly
from the operating system at any
temperature up to 1500°F 1is virtually
completed. Two small loops for testing

 
 

FOR PERIOD ENDING SEPTEMBER 10, 1951

the reliability of this sampling device
prior to its installation with an
operating system are under construction.

INVESTIGATION OF SODIUM CONDENSATION

H. R. Bronstein, ANP Division
Condensation of sodium in gas-
pressurizing lines leading to operat-
ing systems was investigated further
during the quarter. Electrostatic
precipitation showed that plugging was
due to a solid aerosol phenomenon, and
that quantities of sodium removed by
the precipitator were of such a magni-
tude that a means of returning this
condensate to the main tank should be
sought. Consequently, the precipitator
was redesigned to allow for periodic
or continuous melting of the sodium
collected. The experiment was rerun
on a dynamic gas flow basis with traps
placed after the precipitator to col-
lect any sodium getting through it.
No sodium was detected beyond the
electrostatic precipitator.

NaK traps have been suggested as a
possible solution for the problem of
gas-line plugging. Models of several
types have been built, and their
operating characteristics are being
studied with air and water. From
information obtained from these tests,
a trap for use with operating systems
may be designed. '

A group fromthe MIT Practice School
investigated three methods of reducing
or preventing sodium plugging of
blanketing-gas lines during the period.
These methods were (1) electrostatac
precipitation, {(2) a sodium condenser
coil immersed in a het oil bath, and

{3) a NaK bubbler. Although all methods

 

tested proved to be partially effective,
results of work by this group indicated
that the oil-bath trap was the most
practical. Design improvements would
increase the efficiencyofall methods,
however.

The oil-bath condenser causes
particles to condense on the walls of
the tubing because of its long length
and curvature, and, upon condensation,
the sodium flows back into the main
tank by gravity since the oil tempera-
ture is maintained just above the
melting point of sodium. Additional
experiments to determine the effect
of variations in tubing length and size
and bath temperature were recommended.

ARE COMPONENT TESTS

H. P. Kackenmester and D. L. Salmon,
ANP Division

Specifications have been written
by the ARE Components Testing Committee
and engineering drawings are being
prepared for construction of facilities
for full-scale ARE components testing.
Five all-steel completely enclosed
test cells are to be erected, four
of which are 1dentical 1n design and
are to be used for measuring physical
and operating characteristics of
individual components such as pumps,
pipes, valves, instruments, heat ex-
changers, etc. prior to their use with
the ARE. The other test area, larger
in size, is to be used for testing
components 1in combination and will
include facilities for testing one
complete heat-exchanger system under
simulated operating conditions. A
gas-fired furnace delivering 3,000,000

27
 

ANP PROJECT QUARTERLY PROGRESS REPORT

Btu/hr is to be used to simulate the
reactor., Design of pilping systems and
test loops for individual equipment
items 1s progressing satisfactorily,
and procedures for components testing
are being established.

An apparatus has been set up for
determining the pressure drop of
different orifice configurations 1n
connection with coolant-tube orifice
investigations. A detail drawing has
been completed for lucite headers to
adapt an ARE fuel-pin cluster assembly
for a water flow test, and detail
drawings are heing prepared of a full-
si1ze reactor flow passage mock-upto be
used for measuring the pressure drop
and flow distribution with water.

HEAT-EXCHANGER TESTS

A. P. ANP Division

Fraas,

The heat-exchanger model designed
for testing in the figure-eight loop
has been tested with NaK at 800, 1000,
and 1200°F, and a curve of heat-
transfer coefficient as a function of
the flow rate has been obtained. The
experimental data give values as much
as 40% below those calculated from
Lyon's formula ¢(*) at low flow rates,
but check closely at high flow rates.
As soon as endurance testing 1s com-
pleted, a report will be 1issued on the
the flow and heat-transfer tests which
have been carried out on heat-exchanger
and fuel-element models.

 

(4)R. N. Lyon, Forced Convection fleat Transfer
Theory and Experiments with Liquid Metals,
ORNL- 361, p. 21, Eq. 34 (Aug.19, 1949).

28

BUILDING #ODIFICATIONS AND
EXPERIMENTAL FACILITIES

P. L. Hill, USAF

Building modifications and the pro-
curement of experimental eauipment and
facilities 18 progressing on a schedule
compatible with that of the ARE. The
120-ft hood has been completed and 1is
in use for testing intermediate-silze
liquid-metal systems. To allow maximum
flexibility, no fixed partitions are
provided inside the hood, bwt pro-
visions have been made for placing
barriers around equipment undergoing
test. The exhaust system 1s designed
to give a linear facial velocity of
100 to 150 ft/min. Four power-control
panels similar to those 1n for
thermal-convection-loop operation have
been installed i1n front of the hood
and are ready for use. Each panel
provides up to 24 kw of controlled
heater power for use on systems under-
going tests. An overall view of this
installation 1s shown in Fig. 2.4,

use

Areas for ligquid and air test
equipment have been completed, and
these facailities will permit the use
and/or air in testing the

flowmeters, and

of water
performance of pumps,
heat exchangers 1in so far as these
tests may be conducted using air and
water as working fluids. Design for a
one-man standard chemical laboratory
and a special laboratory for testing
high-temperature liquid metals has
been completed and equipment has been
ordered. The cleaning arca for de-
greasing and/or removing trace residues
of ligquid metals 1s designed and
partially completed. Bulk residues of
liquid metals or other hazardous ma-
terials are being removed in a facility
outside the building. This facility

 
I HNBEWION 1

Fig. 2.4 - Experimental Facilities.
120 ft. is across the top.

The pump test is

PHOTO 6 339

SoDIUM
o ;,‘,)‘M FIRES

"

ALT A

bl Y
U S0ODIM 1y
i “

'
THje
w xa &
e
“li
i

 

 

in the left foreground; the hood,

‘0T HAAWALdAS ONIANA dOoIydd ¥O0d4

IS61

AIGRRARION
ANP PROJECT QUARTERLY PROGRESS REPORT

is incomplete but was in partial
operation at the end of the period.

ALKALI METALS MANUAL

R. Devenish, ANP Division
P. L. Hill, USAF

During the quarter the Y-12 Alkali
Metals Guide, which sets forth safety

30

considerations to be observed when
working with alkali metals or liquid-
metal systems, was revised. The
revised guide 1s being prepared for
general distribution. In this con-
nection, colored movies have been
made of sodium fire extinguishing
operations and liquid-metals disposal.
The film has been edited, and a dupli-
cate is being made for showing to
interested groups.

 
 

FOR PERYOD ENDING SEPTEMBER 10, 1951

3. REACTOR PHYSICS

N. M. Smith, Jr.,

The principal subjects of activity
have been the calculation of the
static characteristics of the design
referred to here as the "ARE of 10
June 1951" and extensive calculation
of the kinetic behavior of the ARE by
machine tcchniqfies. The design of 10
June 1951 refers to an ARE of two-
thirds the core volume of that in the
design of 10 March 1951, Besides
reduction of core size, manufacturing
specifications and tolerances required
a reduction of the average demsity of
moderator in both core and reflector
and an increase in structure and
coolant volume, The effect of the
densi1ty changes overrides the effect
of reduction of core size, resulting
in a net 1ncrease 1n uranium inventory
requirement. Minor changes 1in the
static coefficients which resulted are
summarized
form,

in tabular and graphical

The uranium investment required is
estimated to be 27.8 lb +10%, -20%,
for the depleted, hot, poirsoned,
instrumented reactor with control rbd
thimbles in place. Allowance for
expansion of liguid fuel through the
boron curtain adds a requirement of
1.4 1b to the above figure, yielding
a total inventory requirement of 29.2
b of uranium. :

The effectiveness of mutual shadinhg
effects of a hexagonal array of seven
2-in, BR,C control rods has been calcu-
lated using three-group procedure
made equivalent to the 3Z2-group pro-
cedure for a homogeneous reflected
reactor. With one rod in the center,
and six on a ring surrounding the

 

Physics Division

center,. the total reactivity ef-
fectiveness is insensitive to changes
in the radius of the ring between 6.5
and 10.5 in. Absorption in the perman-
ently placed thimbles will reduce the
differential effect of the seven rods
to 0.22 1n Ak/k. Muatual shading to
the extent of a reduction of ef-
fectiveness of seven isolated control
rods by 14% 1s indicated by the calcu-
lations,

The kinetic response of the ARE to
arbitrary changes of multiplication
constant and to changes in inlet
coolant temperature for various values
of neutron lifetime and of fuel temper-
ature coefficient has been studied and
graphs are presented. The results
show that the fuel temperature responds
in an overdamped fashion to reactivity
changes well over prompt critical.

A start-up accident throwing the
reactor prompt critical at a low
(300-watt) level was studied, re-
sulting in a tolerable transient,

These calculations have been made
by starting with steady-state initial
conditions and by numerical integration
of the set of nonlinear differential
equations by IBM machine procedure.
Work bas continued on the development
of other IBM calculations: multiregion
problems, reactor solutions in cy-
lindrical coordinates, and the effect
of a boron layer between core and
reflector. | :

Variocus problems relating to the
physics of reactor calculations or
to the physics of the ARE arediscussed.
These include a proper formulation for

31
 

ANP PROJECT QUARTERLY PROGRESS REPORT

the age theory of the relation between
neutron flux and slowing-down density,
the transmission coefficient of a
boron curtain, and heating in the
boron curtain,

Calculations have been started on
the physics of hydrogeneous reactors,
and in particular for a NaOH moderated
and cooled reactor. Results are quoted.

IBM CALCULATIONS

Production Report (F, C. Uffelman
and P. Johnson, Uranium Control
Department). Reactors. During the
period ending September 1, 1951 the
IBM section completed calculations on
84 reactors. Included in these 84
reactors were four bare reactors
calculated by the end-point linear
approximation method and three adjoint
reactors. This brings the total number
of reactors calculated by the EPLA
method since calculations were begun
in February to 182 reactors,

Programing for the calculation of
reactors with an absorbing layer be-
tween core and reflector 1s very nearly
complete. PBecause the method{(!) used
consists in making two parallel calcu-
lations for flux distributions in the
reactor and combining the results of
these calculations for each group be-
fore going on to the next group, the
calculation of a reactor blanketed
with B,C will take between five and
six times as long as the calculation
of a regular reactor.

 

(1. B. Mills, IBW Procudure for B,C Layer
Between Core and Reflector by the Coveyou Method,
ORNL, Y-12 site, report Y-F10-61 (June 28, 1951).

39

Cores and Reflectors. During the
period ending September 1, 1951 average
cross-sections and constants were
calculated for 26 cores and reflectors,
and "constant'" variations were calcu-
lated for 34 cores and reflectors. Of
the cores calculated, 16 were hy-
drogenous and required calculation of
the following factors in addition to
those calculated for nonhydrogenous
cores:

TsH_ (1)
fUT ’

. Tr

BN = (2)

 

’
Ry + 0.71 (T/S,p)¥

 

 

 

N
82
———| (3)
3507 rp
B 1 N
z .fN = . - (4)
TR tan'l B/ZTR l/zTR

Z;;T? is calculated twice, once for a
constant B and once for the variable
B derived in formula (2) above,.

A new set of cross-section boards
and procedures 1s being set up for
calculation of special cross-sections
and constants, The new set-up will be
more flexible but slower in operation
than the specialized set-up used for
regular nonhydrogenous cores, so that
the old set-up will continue to be
used for regular cores and reflectors.

 
 

FOR

Kinetic
for 16 different sets of
The length

number of

Kinetic Calculations,
calculations
conditions were completed.
of the time interval, the

functions, and the initial starting

conditions were varied from set to
set. Boards and procedures are now
being set up for a kinetic calcu-

and

both

spatial variations are made,

lation in which temporal

The Multiregion Hezctor Pruhiem?Z}
(C. B, Mills, Physics Divistion). A
two-group calculation has shown that
the effect of material cutside the
reactor reflector is i1mportant for
the fast neutrons leaking out of the
reactor. A formula was developed(?’
for use with the IBM multigroup pro-
cedures to include the effects of
of these "outside” materials. A new
identxfy
a factor
eliminated in the development of the
presently used formula must be retained
as a factor in both the solution of
the homogeneous equation, A , and the
corrective term, Pn, This factor is
the ratio of the values of o* {(defined
in ANP-58)¢3) un each side of the
interface through which the neutron
being solved, to a
of reactor

index number 1s required to
the different regions. Alse,

flux equation 1is
power given by the number
space points between the first core-
reflector interface and the i1nteviace
Otherwise the eqguations
form to the two-

of 1nterest,
are i1identical 1n
region solution.

Cylindrical Multigroup Calculation
(N. Edwonson, Physics Division)., It

is desired to calcolate the neutron

 

(Q)Abstracted from the report by C. B, Malls,
The Multiregion Reactor Problem as Applied to the
Multigroup Method, ORNL, Y-12 site, report Y-F10-68
(Aug. 16, 1951;. :

3. K. Holmes, The Hultigroup Method as Used
by the ANP Physics Group, ANP-58 (Feb. 15, 19561).

 

PERIOD ENDING SEPTEMBER 10, 1951

and the effective multi-
plication constant k for a reflected
cylindrical reactor with bare ends by
transforming the partial differential
Fermi age equation.

discribution

The nettron flux ¢(r,u) and che
slowing-down density g(r,u) are assumed
symmetrical relative to the axis of
the cylinder. This assumption reduces
the variables to r and z. 1t 1is
further assumed that 1n the reactor
equation r and z may be
This the expression of the
z dependence by a factor

space

separated.
leads to

Ly

9(H + d)

 

where H 1s the half-cylinder length
and 4 is the linear extrapolation
length, 1t 1s assumed that d has the
same value for all lethargies and for
both the core and the reflector.

It 15 assumed that the neutron flux
@lr,u) is = 0 and isequal
to zetro at the extrapolated boundary

finite for r

of the reflector. The core-reflector
boundary conditions avre: (1) ¢{(r,u)
and (2) -D(op/Or) is

Work on this problem is

is continuous
cOontinuous.
being carcied on as rapidly as possible,
and the formulation is in process of

publication in report form.

The Effect of the Boron Blanket in
the ANP Reactor (C. B. Mills, Physics
Division}. The
a tveactor with an

method of solution for
energy~dependent
core and reflector(*?

been transformed) invto the system

absorber between

has

 

(A)RA R. Coveyou, Spherical Reactor with

Absorbing Interface. [I[, ORNL, Y-12 site,
veport Y-F10-52 {April 30, 1951).
 

ANP PROJECT QUARTERLY PROGRESS REPORT

required by IBM multigroup procedures.
This consists of computation 1in this
order:

1. Compute a flux distribution in
core and reflector with a source
term in the core only.

2. Repeat step (1) for a source
term in the reflector only.

3. Ceompute separately currents due
to sources and reflectors at the
core-reflector interface, Compute
four currents.

4., Compute the corrective term for
reflector flux due to a core
source and the term for core
flux due to a reflector source.

5. Add flux and correction 1in core
and reflector, and use the total
flux to determine the source
term for the next higher lethargy
group as in the regular multi-
group procedure,

STATICS OF THE AIRCRAFT

EXPERIMENT

REACTOR

Summary of Calculations on the ARE
(B. T. Macauley, USAF). Multigroup
reflected-reactor calculations on the
latest ARE design, hereafter referred
to as the design of 10 June 1951 and
1llustrated in Fig. 3.1, were made
during the past quarter. The core of
this design, having auranium inventory
of 25 1b, is referred to as Core 93.
The composition of the core and the
design data are given in Table 3.1.

Two series of reflected reactor
calculations were made; the first
consisted of a spherical core having
the properties listed in Table 3.1 and
surrounded by the side reflector
(Reflector 519), and the second series

34

of calculations consisted of a spheri-
cal core having the properties listed
in Table 3.1 and surrounded by the
bottom reflector (Reflector 520),
The constituents in both the side and
bottom reflectors are given in Table

3.2,

Calculations were made using the
compositions of the ARE core (Core 93)
except for varying the uranium con-
centration and using Reflector 520.
The results were plotted showing how
keff varies with uranium investment
(Fig. 3.2). Also, calculations were
made using Core 93 (25 1b uranium
investment) and each of Reflectors 519
and 520, varying the core thickness to
show how k varies with the reflector

thickness %gig. 3.3).

Graphs of &, § VS, reflector thick-
ness for the 3tft right cylindrical
corec design (Core 84; 1incomplete curves
previously reported in the last
quarterly), as well as this core re-
duced in volume by one-third (Core 91)
and one-sixth (Core 92) have been drawn
and show the same general relations
as indicated in Fig. 3.3 for the ARE
core (Core 93).

The reflector saving for the ARE
core (Core 93) is given in Fig. 3.4.

Numerical calculations were made to
determine various design and nuclear
coefficients, The results are summa-

rized in Table 3.3.

Figure 3.5 1s a plot of the spatial
distribution of the lethargic average

of the fast flux in the core, and

Fig, 3.6 is a plot of the spatial
distribution of the thermal flux 1in
the core. Plots of the flux spectrum

at various radii are shown in Fig. 3.7.
Figures 3.8 and 3.9 show the leakage

 
 

FOR

PERIOD ENDING SEPTEMBER 10, 1951

DWG. 12879

SECRET
jl

)

ANP~-PHY-210

 

 

?/IE

69 %

 

   

REFLEGTOR
!

 

   

PHYSICAL DATA
TOTAL

. : ZONE | ZONE 2 REFLECTOR
ZONE 2 % REFLECTOR 935 775 91.55
‘ % COOLANT | 6 8.20 2 a4
_ % STEEL 0.37 1.65 0.52
ALL DIMENSIONS ARE IN INGHES o, VOID 4.50 12.40 5.49

Fig., 3.1 - Schematic Drawing of ABE Core Arrangement.

35

 
 

ANP PROJECT QUARTERLY PRCGRESS REPORT

TABLE 3.1

Comgosition and Design Bata of Corc 23

Core geometry Cylinder
Core diameter (in.) 29,45
Core height (in.) 34.00
Core volume (ftd) 13.411
Core composition: VOLUME
FRACTION
VOLUME (£t%) (%)
Moderator (ReOQ) 10.872 81.10
Liquid fuel 0.185 1.38
Coolant (sodium) 0.621 4.63
Fuel-tube metal 0.130 0.97
Coolant-tube metal 0.078 0.58
Control-tube metal 0.221 1.64
Void 1.304 9,72
Fuel-element dimensions 225 1in. o.d. and
0.025 1in. wall
thickness
No. of fuel tubes 65
No. of control elements 1
Thickness of side reflector (in,) 8.75
Thickness of bottom reflector (in.) 8.00
spectrum from the core and the re- There results a critical mass of

flector, respectively, Figure 3.10 1is
the spatial power distribution with a
uranium investment of 25 1b,

Estimated Critical Mass of the ARE
(N. M. Smith, Jr., Physics Division,
and B. T. Macauley, USAF). The results
of the previous section have been
employed to estimate the critical mass
of the ARE design oif 10 June, The
calculations are summarized 1n Table

3.4.

36

27,8 1b of uranium. This 1s somewhat
higher than the value of 21.51b gquoted
in the last quarterly report (ANP-65)
for a reactor of two-thirds volume of
the design of 10 March 1951. This 1in-
crease has been brought about by the
engineering of the core of a more
of BeO block
clearances, and structural
densities., The design of 10 June
contains less Be( and more inconel and

realistic estimate

densities,

sodium,

 
 

FOR PERIOD ENDING SEPTEMBER 10, 1951

TABLE 3.2

Volume Fractions of the Materials in
the Side and Boettom Reflectors

 

 

CONSTITUENT (fe3) (%)

 

VOLUME | VOLUME FRACTION

 

Reflector 519 (Side Reflector)

 

Reflector {(Be() 19.676 94.89
Coolant (sodium) 0.41 1.98
Structure {(inconel) 0.074 0.36
Yoid G.590 2.85

 

Reflector 520 {(Bottom Reflector)

 

Reflector (BeD) 2.56 81.1
Coolant { sodiam) 0.220 6.98
Coolant-tube metal 0.0185 0. 58
Control metal ©0.0520 1.64

 

 

Void 0.3072 9.72

 

Contrel Rod Effectiveness (J, W.
Webster, Physics Division, and R, J.
Beeley, Oak Ridge School of BReactor
Technology). The last guarterly report
(ANP-65) gave the results and method
of calculation on the effectiveness
(Ak/k) of an axial 2-in. B,C control
rod. The heating due to the (n,a)
reaction and gamma absorption was also
reported,

This quarter it was decided that
seven 2-in. control rods would be
necded in the ARE tooffset the effects
of temperature and fission product
poisons with a margin of safety.
Caleulations were made by the Nordheim-
Scalettar Method(®"7) to determine the
best placement for these rods to obtain
the maximum effectiveness in k. The
physics work of the last six months on
the statics of the ABRE control system
has been written and distraibuted as
report Y-F10-71.¢%) "Included in this
report is a discussion of three-group
theory when, of the three solutions of

 

 

    
   

 

 

 

DWG. 12880
SEGRET
T 1t
{07 [ ANP-PHY-95 ]
REAGTOR G77, STANDARD
1,05 f—-
>
8 1,03 F—-
3 REAGTOR 1024
x -
x
0
o
Moo -
x
=
o
h .. A
Q1
ut
w
T
0499 t— e
t
[}
e .S ]
o
ur
o
[
W 0.97 -
@ { REAGTOR 1023
3
‘:- e ——
L.
Q
A
0gs e A
CORE RADIUS 43.807 om
REFLECTOR THIGKNESS 18.8C3 cm
Q.93 P REFLEGTOR 520 -1
81.10 % Be0, 2.22% INGOREL, 6.98% Na
!
o9l ) ' | J ."J J [ J [ B

16 8 0 22 24 26
URANIUM WEIGHT {10} :

¥Fig. 3.2 - keff vs., Uraaniom Weight
for the ARE.

(5)R. Scalettar and L. W. Nordheim, Theary of
Pile Control Rods, MDDC-42 (decl. June 17, 1946).

(6)C. BR. McCullough, Summary Report on Design
end Development of High-Tenperature Gas-Cooled
Power Pile, MonN-383, Appendix {Sept. 15, 1547);
Criticality and Control, NEPA-6, Appendix (Ove. 1,
1948).

(7)J. W. Webster, {ontrol-Rod Effect
Reactivity and Power Distribution, NEPA 1C-50-2-52
(Feb. 7, 1950). ,

(B)J. W. Webster and R. J. Beeley, Ph

Y
Calculations on the ARE Control Heods, ORNL,
¥-12 site, report Y-F10-71 (Aug. 29, 1951).

37
 

ANP PROJECT QUARTERLY PROGRESS REPORT

TABLE 3.3

Summary of Calculations on the ARE Core Yaving 25 1lb of Uraniomand Reflector 520

Nuclear coefficient of reactivity:

 

 

From start-up to normal operation Ak/kE = -0,007995
Temperature coefficient Ak/k/°F = 5.04 x 1076
. .. . Ak/k
Uranium mass coefficient of reactivity = +0.300
Am/m
Density coefficient for core materials: AL /T
Moderator f—lll = +0.4894
Do/ L) peo
Ak [k
Coolant [—“*—} = -0,00636
Do/ P ke
Ak /&
Structure [mmm*] = -0.1140
AVL/p inconel
Density coefficient for reflector materials: Ak/w
Moderator {—-fll = +0.2246
Do/p
BeO
(Ak /&
Coolant "fi““] = -0.00093
\Bo/o
Na
Ak/k
Inconel [ } = -0.03836
A¥Z/F) inconel
Lifetime (Core 93 with Reflector 519) 1.535 x 10°? sec
% thermal fissions 59.8

Spatial and lethargic average of fast

(0 < u < 18.6} flux, &&a“ 16.12 neutrons/cm?/sec per unit
of lethargy per fission per
second per cubic centimeter of
core

Spatial average of thermal (u = 18.6)
flux, ¢ 47.99 neutrons/cm?/sec per
fission per second per cubic

centimeter of core

thermnl

Integrated flux at 3 megawatts at central
position 1.682 x 10'* neutrons/cm?/sec

 
 

FOR PERIOD ENDING SEPTEMBER 10, 1951

 

 

  
      

 

 

OwG. 12881
. SECRET
AUGMENTED CORE RADIUS (em)
a4 48 52 56 80 &84 €8 T2 S 76 80 84 88 92
130 _ .
T rrTr Tt syttt
: - ANP-PHY- 196 |
126 CORE 93, kg = 1.664 : / : ]
1,22 fe-es : ' -
m BARE CORE ' : =
(1) 1BM GALCULATION e
118 = -
- “REFLECTOR 519 ™
e l— ST e
3.
e A0 - -t
3 .
o
. e —
-
w
g .
w108 P -
i
a2 Lo “~ REFLEGTOR 520 —_—
w . :
o
Mooz — ]
b
oo
wd
£ [~ ]
<
)u‘.
o 098 t— ]
w
b~
o
i - = =
o
'
o
S 094 — ]
&
R Lo ]
0.90 b COMPOSITION ~ VOLUME FRAGTION ]
L1QUID FUEL BeD NGONEL  Na
— GORE 93 - 0.0138 0.8110  CO3IS  £.0463 —
(W, = 0.0000716 % I07* atoms /ce)
086 . REFLEGTOR 519 0.9489 00036 00198 : =
/ : REFLECTOR 520 0.810  0.0222  0.0698
— ]
0.8 b -
0.78 ~ ‘ —
Cora _ 1 l l

 

o 5 1G 15 20 25 30 35 40 45 50
AUGMENTED REFLECTOR THICKNESS (cm)

Fig. 3.3 - Reactivity of the ARE Core Backed by Various Reflectors or Ad-
ditional Core Material.. '

39

 
 

DWG, 12882

 

 

 

 

 

" SECRET
[ 1+ 1 17 17 1 1T 71 | T I—"
L e ANP-PHY~-203 —]
20 — ]
REFLEGTOR 519
— -
6 — ]
g — — ]
z REFLECTOR 520
>
& g b ]
&
5 |
w COMPOSITION (VOL FRAGTION) T
E ol W, FUEL B8e0 INCONEL Na
CORE 93 0.716 x 10%° 0,038 0810 00319 0.0463 o
REFLECTOR 519 0.9489 0.0036 0.0198
—— REFLEGCTOR 520 0.8110 00222 00698 —
4t —]
b

0 4 8 12 5 20 24 28 32 36
AUGMENTED REFLECTOR THIGKNESS (cm)

40

Fig. 3.4 - Reflector Savings of Varions Reflectors Backing the ARE

Core,.

the buckling obtained, two are complex of six.
conjugates, instead of the uswual
situation where all three are real,one
positive and two negative. This case
was encountered in the control rod
studies,

31.3% for 9.5 in., and 29.3%

seven rods in each case was:
for a radial distance of 6.5 1in.,
31.1% for 7.5 in., 32.0% for 8.5 1in.,
for 10.5

The total Av/v (effect of
thimble not deducted) obtained for the
29.1%

in. In regard to the shadowing effect

of the rods on each other,
numbers are to be compared with
7 X 5.3 = 37.1%. Thus the minimum
shadowing effect of (37.1 - 32.0)/37.1

The discussion of the effectiveness
of the seven control rods given in

Y-F10-71 follows:

The total effect on reactivity of
the insertion of seven 2-in. rods was
investigated. The pattern was one
axial rod and a ring of six at the
same radius equally spaced around the
center rod, Five different distances
from the axis were tried for the ring

40

 

at a radial distance of 8.5 1in.
the axis and the net effect of the
seven rods, deducting the effect of
the thimble, is 22% in reactivity.

It follows from the results
the total reactivity effectiveness

these

= 13.7% occurs with the ring of six

of f

that
 

FOR PERIOD ENDING SEPTEMBER 10, 1951

TABLE 3.4

Calculation of Uranium Requiremént for the ARE Design of 10 June 1851

 

 

 

 

 

 

 

 

 

; RELATIVE WVEICHTED URANIUM MASS
) i : k R ib

CORE REFLECTOR POSITION YALU& eff eff (1b)
93 None Top 1/6 0.8045 0.1341

93 520 Bottom 1/6 1.0730 0.1788

93 519 Side 2/3 1.1585 0.7722

Weighted value of keff for clean homogeneous reactor 1.085

Seven control rod thimblés at 0.014,each:!ess 14% for

mutual shading ; ' -0.078
Net ARE, clean, with control-rod thimble 1.007
Total keff in clean reactor desired for fission poison
override, experiment, and control instrumentation 1.041
Additional & needed for above 1.034
T Teff Ak
Additionaj uranium needed for above, usihg —— = 0.300 _ 2.8
. . Ar/a

Uranium in Core 93 25.0
Total uranium in reacting volume (critical mass) ' 27.8
5% allowance nbove and inside B,C curtain 1.4
Total uranium inventory requirved 29.2 _
Estimated probable error +10%, -20%

 

 

 

15 not sensitive to rod placement in
the region investigated, 1.e., 6 to
10 in. from the axis. This 1s caused
by the fact that two opposing factors
atre acting. When the rods are put
close to the axis they are inserted
into a tegion of higher undisturbed
flux but the shadowing effect of rods
on each other is large; on the other
hand, when the rods are put farther
from the axis they are inserted into a
region of lower undisturbed flux, but
any one rod does not tend to feel such
a large flux depression caused by the

 

insertion of neighboring rods. The
two effects counteract to givea rather
broad region of good rod placement. It
is known from previous calculations (3
however, that if the rods are very
close to the axis or very far oput
toward the reflector, a sharp loss in
combined effectiveness is incurred.
Since the hexagon moderator blocks are
about 3% in. across flats, it develops
that the possible radii awvailable for
rod placement are 3%, 7.5, 11.235 in.,
ete. The closest to optimum place-
ment is 7.5 in., and this has been
adopted in the current ARE design.

41
=
8o

 

FLUX (neutrons/cmz/sec per unit of lethargy per fission)

3C.0

25,0

2C0

18.0

10.0

5.0

DWG. 12883

SECRET

 

 

e e e e

  
     
  
  
  

 

 

LATTICGE POINT, n

Fig. 3.5 - Spatial Distribution of the Lethargic Average of the Fast Flux,.

— ANP-PHY-204 —
L REACTOR 977 o
- COMPOSITION (VOL FRACTION) L
LIQUID Na
N FUEL BeD INCONEL COOLANT
T CORE 93 0716 x 10%° 00138 0.8110 00319 00463 ]
REFLECTOR 520 0.810 00222 00698 T
f b, ridr T
. . T _ space
__ Spatiai average of integrated fast flux, d)'fcas? = frzdr .
L = 16.12 neutrons/cme/sec per unit of lethargy _
= S, (v)du
n
| where =
$n =70 _
. For 3 megawatts, spatial avercge of integroted fast flux =0.904 X IOB neuTrons/cma/sec ]
thermal flux =1.086 X (02 neutrons/cmZ/sec
— Ratio of peak to agverage =1.884 —]
Ar =2.067 cm
e sy b ]
2 4 6 8 10 12 i4 5 i8 20 22

LY04AY SSTAU90¥d A THIALUVAD ID3If0oHd JdNV

 
- 75.0

70.0

65.0

60.0

55.0

 

45.0

FLUX (neutrons/cm</sec per fission per sec per cc of core)

40,0

35.0

b
o

DWG. 12884
SECRET

 

50.0

 

  

CORE 93
REFLECTOR 520

Spatial average of thermal fiux (at 18.6},

%,hem, = 48.00 neufrons/cmZ/sec per unit of lethargy per fission

b T T T T

ANP-PHY-206

   
  

REACTOR 977

GCOMPOSITION (VOL FRACTION) N

LIQUID Na
Ny FUEL  BeO INCONEL COOLANT
0.716 x10%°9  0.0138 0810 0038 00463

0.8110° 0.0222 0.0698

_ W, o -
P thermat 18.6 = (_fi_)ne.e

" Ratio of peak to average = 1563

| ]

 

bbb

 

2

4 6 8 10 i2 t4 18 i8 20
LATTICE POINT, n

Fig. 3.6 - Spatial Distribution of the Thermal Flux in the Core.

‘01 MIGWILdIS INIGNT QOI¥Ad ¥0d

ISel

 
 

ANP PROJECT QUARTERLY PROGRESS REPORT

FLUX (neutrons/cmzlsec per unit of lethargy per C.4 fission per sec per cc of corg)

44

40

28

24

DWG, 12885
SECRET

 

......

 

T T T T T T T T

ANP-PHY-206

REAGTOR 977
COMPOSITION (VOL. FRACTION)

LIQUID Na
N, FUEL Be0  INCONEL GCOOLANT
CORE 93 0.716 x102° 00138  0.810 0.0318 0.0463
REFLEGCTOR 520 0.8110 0.0222 0.0698

[fyb(u) (.‘fuj|(nw

): 256 neutrons/cm2/sec per 0.4 fission per sec per cc of core

Core volume = 3 424 X IO5 cec

For 3 megawatts,

Integrated flux(,.,)=1.682x 10" neutrons/cm?/sec

Thermal f|UX(,,=!) = 1.97 x 10" neutrons/cm?/sec

 

 

 

 

LETHARGY, v

Fig. 3.7 ~ Flux Spectrum in the ARE.

 

 
DWG. 12886
SEGRET

T T T T T T T T T T T T T T T T T T T

— ANP-PHY-207 —]

 

REACTOR 977
= , COMPQOSITION. (VOL FRACTION} —

LIQUID Na
010 — Ny FUEL  BeO INGONEL GOOLANT _
CORE 93 0716 109 0.0138  0.810 0.0319 0.0463
- REFLECTOR 520 0.8110 0.0222 0.0698 —

008 — —_

* Normalized to one neutron leaking or absorbed per ¢c of core —

0.06 — fi

 

   

0,04 S . —

 

NUMBER® _EAKING PER UNIT LETHARGY

 

0.02 L— 18,6 THERMAL

 

 

 

 

 

 

 

 

LETHARGY, v

‘0T HIGWAL4AS INIGNE QOI¥dd ¥O0d

e
e

IS61

Fig. 3.8 - Leakage Spectrum from the ARE Core to the Reflector.
e
N

 

NUMBER® LEAKING PER UNIT LETHARGY

0.ci8

0018

0.014

0.0i2

0010

0.008

0.006

0.004

0.002

DWG. 12887
SECRET

 

 

 

—

 

 

REASTOR 977
COMPOSITiON {VOL FRACTION]

L:QUiD Ne

Ny FUEL  BeQ INCONEL COOLANT
CORE 93 0.76 x 10%°  0.0138  0.81I0 0035 0.0453
REFLEGCTOR 520 0810 0.0222 0.0698

* Normaeiized 1o one neutron igaking or absorded per ¢c of core

 

 

 

 

b

b

ANP-PHY-208 —

 

 

 

 

 

JJLLf4—*—”1____L___J__J____L____ -—
16.6 THERMAL
L
\
Ly e e
13 3 14 2 o 8 & a4 2 Q0
LETHARGY, v
Fig., 8.8 - Leakage Spectrum frow ihe ARE Reflector,

LU0dd¥ SSHES0Md ATHALIVAD 1D3{0¥d 4NV

 
DWG. 12888

 

 
 
 
    
  
 
 
   
  

   

 

 

 

 

 

 

 

SEGRET
1.8 ; i E
ot
ANP-PHY =209
" REACTOR 977
S COMPOSITION (VOL FRACTION)
O ’ LiQUID Na
W - - | N, FUEL  Be0 INGCONEL GOOLANT |
- . CORE 93 0716 %10%° 00138 0.8110 00319  0.0463
G ™, REFLECTOR 520 0.8110 00222 0.0698
r 14— \ —
d
ll .
A * Normalized 1o cne source neutron
_3 | N per cc of core )
5 2
g‘:) .2 B =
a FISSION DISTRIBUTION CALCULATED FOR ARE -
g FROM ASSUMED DISTRIBUTION ] %
o oust
oz @
; =
Z 2
W &
O L ] o
« FISSION DISTRIBUTION ASSUMED AS A %
* SCURCE FOR THIS CALCULATION "
D 0.8 — ee
= - | 3
2 e CORE »{t REFLEGTOR ———»=] =
_— 7 &
; i | i ! i i | %
ol v =
0 4 8 2 16 20 24 28 32 =

LATT@E RflNTzfiyr;mAr;éyfi=206?cm”_

ISo6t

e
—~3

Fig. 3.:10 - Spatial Power Diséribution in the ARE.
 

ANP PROJECT QUARTERLY

     

 

 

 

 

 

 

OWG, 12889
50 .- SECRET
B T """""""""""""""
T T [
ANP-PHY-144
40 \ ]
2
08 .
4
x o /
o e
2 5 /
o -
20 |- © -~ —
o CONTROL-ROD
o PATTERN
o
o
10 a NORDHEIM ~SCALETTAR ]
x APPROXIMATION
Z
<
o
old 1 .
0 10 20 30 40 S0 S0 70
RADIAL DISTANCE FROM AXIS (em)
Fig. 3.11 - Badial Thermazl-Flax

Distribution in the ARE RBeactor with
Seven Comtrol Rods; Placement HNo. 1.

 

   
     

 

 

 

 

 

 

 

DWG. 12830
SECRET
I } | |
25 | /,m\\ ANP-PHY-145
p
- \
”
—
20 |-
L 15
>
-
"%
CONTROL -ROD
PATTERN
o1 4 -
g S
o
3 !
& 3
= ¥ | NORDREIV -SCALETTAR
o5 1 5 z APPROXIMATION -
© o
-]
o 11 1 I
0 10 20 30 10 50 g0 70
RADIAL DISTANCE FROM AXIS (cm)
Fig. 3.12 - Badial Thermal-Flax

Distribution iz the ARE Reactor with
Seven Control Rods; Placement No., 2.

Figures 3.11, 3.12, and 3.13 show
the distribution of the thermal flux
along a radius vector through a rod

and in between two rods for three

different placements corresponding to
I o
1n.,

of 6.5, 8.6, and 10.5

radiil

A8

PROGRESS REPORT

 

    
     

 

 

 

 

 

 

 

DWG, 12891
SECRET
25 L— ANP-PHY -146 |
20—
15 —
=<
3
T
GONTROL -ROD
ol PAT TERN
Q
Q o
x o
o @
o =}
a <
[ = &
0.5 5 2 | norRDHEIM-SCALETTAR 7
© o APPROXIMATION
L 5}
P
b L |
0 10 20 30 a0 50 50 70
RADIAL DISTANCE FROM AXiS (cm)
Fig. 3.13 - Radial Thermal-Flux

Distribution in the ARE Esacitor #ith
Seven Control Rads; Plaocement No. 3.

respectively, for the circle defined
by the centers of the six outer rods.

KINETICS OF THE ATRCERAFT
REACTOR EXPERIMENT

M. J. Nielsen, USAF
J. W. Webster, Physics Division

In the last qguarterly report 1t was
mentioned that solution of the non-
linear kinetic equations for the ARE
was in progress on the IBM machines,
Three graphs were included showing the
calculated response of flux (or power)
and fuel temperature toa step increase
in reactivity of 0,009125 (25% over

prompt critical).

This quarter the adaptation of the
kinetic equations to solution by IBM
machines was continued. Three types
of problems were comsidered: (1) the
response to astep change 1n reactivity
with the coolant inlet temperature held
constant, (2) the response to a change

 
 

in the coolant inlet temperature with
no control rod motion, and (3) the
response to a change in the coolant
inlet temperature with the control
rod being activated by the change in
coolant inlet and outlet temperature,

The equations and method of solution
have been written up in ANP-68,(9)
Solutions for the ARE (3-megawatt) and
ANP (200-megawatt) reactors and for
variations of parameters in the neigh-
borhood of those for the ABE and ANP
will be distributed in report form in
the near future.,

The results of kinetic calculations
completed to date by this method are
presented i1n the form of curves of
fuel temperature vs.
time, and phase plots of flux vs. fuel
temperature. The latter type of plot
has the advantage that the new asymp-
totic steady-state flux and fuel
temperature can be shown.

flux vs. time,

Most of the curves have to do with
the response of the flux and fuel
temperature to a sudden change of con-
trol rod position or to an accident
such as a deformation that might cause
a step increase in k. The changes in
reactivity inserted are 8k = 0.002,
0.004; 0.006, and 0.009125 where the
last represents a 25% increase over
prompt critical. The corresponding
flux and fuel-temperature response
are illustrated in Figs. 3.14, and
3.15. A phase plot combining these
two figures is given 'in Fig. 3.16.
Various parameters are changed to study
their effect, for example, the neutron
lifetime, the uranium mass-reactivity
coefficient which has to do partly
with the effectiveness of the B,C

 

(g}M. J. Nielzen and J. W, Webster, Solution
of Kinetic Equations of Cylindrical Liquid-Fuel
Reactors, ANP-68 (Sept. 18, 1951).

 

PERIOD ENDING SEPTEMBER 10, 1951

curtain at the top of the core, and the
conductivity of the fuel. The reactor
is assumed to be at its design power
of 3 megawatts at time zero except in
one calculation — investigation of
start-up accident -~ 1n which the
power 1s taken as 300 watts at t = (.

In none of the cases considered
does the energy release seem sufficient
to damage the reactor, and 1in all
cases the fuel temperature responded
in an overdamped fashion, The fuel
temperature for the case of &k =
+0.009125 stabilizes out at a new
temperature 186°C hotter than the
design value of 800°C when no counter
control rod action is taken. There
is, of course, some magnitude of 5k,
which would cause temperatures result-
ing in failure of the fvuel tubes or
fuel solution before the stabilization
occurred,

Mass-Reactivity Coefficient. The
investigation of the effect, on the
response, of assuming different values
for the uranium mass-reactivity coef-
ficient (Figs. 3.17 and 3.18) demon-
strated the strong sensitivity of the
response to this quantity. It 1is
clearly very important that the B C
curtain be as opaque to neutrons as it
is possible to make 1t. '

Fuel Conductivity. The increase 1in
the fuel conductivity by a factor of
10 had remarkably little effect on the
response to a ok_ (Fig. 3.19). The
explanation seems to:be that the heat
input to the fuel due to fission 1s
rather large compared toc the loss by
heat transfer to the coolant during
the time of the power pulse. Hence a
change in the coocling term {(which is
about a factor of 2 rather than 10
since the conductivity affects only

49
o
<

 

(¢—¢O)/gb0, RELATIVE EXCESS FLUX, OR POWER

200

8.0

6.0

14.0

2.0

100

8.0

6.0

40

2.0

 

DWG. 12892
SECRET
ol T I e ool
— ANP-PHY- 175 —
b REACTOR PROPERTIES (IDENTIFICATION, A-1) —
i GOMPOSITION (VvOL FRACTION)
FUEL
SOLUTION BeO Na INGONEL
CORE 84 0.01705 0.89000 0.07:80 0.02i80 T

e,
——————
—n

84=0.009125

  
  
  

8k=0006

Ek=0.004

REFLECTOR &I13 0.88910 0.006358 0.09i8C

CORE DIAMETER = LENGTH= 3 1

REFLECTOR THICKNESS = 5.57 in.

THERMAL FISSIONING =68%

AVG. FLUX, ¢,, AT STEADY-STATE DESIGN POWER =6.5 x 10"
DESIGN POWER =3 megawatts

STEADY-STATE FUEL TEMPERATURE AT DESIGN POWER =300°C
CONSTANT INLET COOLANT TEMPERATURE

NEGTRON LiIFETIME = 1.4 x 10™ % sec

84 =0.002

 

 

 

 

 

 

| ] T T : : I T T ; 7 1 T T

| | : i ] ! ] { ! | i I { |

 

 

2.0 2.4 2.8 3.2 36 4.0 4.4 4.8

TIME (sec)

5.6 6.0

Fig., 3.14 - Response of Flux toc a Step Increase in Reactivity for VYarious Reactivity Changes.

IHOdAH¥ SSAH90¥d ATHALYVAO 103alodd 4NV

 
 

 

 
  

   

 

 

 

 

DWG. 12893
0.0 SECRET
20, | 7 T T T T 1 Y 1 T
T T T T T T T T T T T T P oo T
L REACTOR PROPERTIES {iDENTIFICATION, A-1) o ANP-PHY-178 _|
COMPOSITION {VOL FRACTION)
8.0 — FUEL B
SOLUTION BeO Na INGONEL
GORE 84 0.01705  0.B9000 0.07160 0.02:80
w " REFLECTOR 513 0.88910 000836 0,09160
>3
3 180 i— —
>
= L ]
© CORE DIAMETER=LENGTH= 3 f1 ‘
W a0 REFLECTOR THICKNESS =557 in. |
w THERMAL FiSSIONING = 68%
g L : AVG. FLUX, ¢, AT STEADY-STATE DESIGN POWER =66 x i0" _
% DESIGN POWER = 3 megawatts
I .
s 120 = STEADY-STATE FUEL TEMPERATURE AT DESIGN POWER = 800°G —
< CONSTANT INLET COOLANT TEMPERATURE
H — NEUTRON LIFETIME = 1.4 x 07 % sec B - B
Z 100 ‘ 4 - _ — | : — -
£ o
= - 34 = C.009125 ] -~
& | u
o 80 F— o — e
= — e ==
o $#=0.006 st
- = S
- -
& - =
u:_J 60 t— o . : — T
' 3
s . T 8420004 - >
L =
© oo
2 4.0 - — A
= @
a e ]
8 o
2.0 — ’ 84 =0.002 — o
y - 2
— : i : P | ! ' z s
o1 | + | | | o b g
0 0.4 0.8 L2 1.6 2.0 24 2.8 32 38 40 4.4 4.8 5.2 56 6.0 o
TIME (sec)
-
=

_ Fig. 3.15 - Response of Fuel Témfietfitnré to Step Increase in Reactivity for Various Reactivity
LR
P Changes.

IS6t1
 

  
   
  
 
  
 
 
 
 
 
 
 
 
 
 
 

A
o DWG. 12894
SECRET
| T ¥ T
T T T T T T T T T T
16.0 —— ANP-PHY-I73 __]
015
L REACTOR PROPERTIES (IDENTIFICATION, A-1) ]
COMPOSITION (VOL FRAGTION)
14.0 — FUEL |
SOLUTION Se0 Na INCONEL
0.20 CORE 84 0.01705 0.89000 0.0760  0Q.02180
h REFLECTOR 513 0.86910 0.00636 0.0960
|20 —— —_—

CORE DIAMETER = LENGTH= 3 ft

REFLECTOR THICKNESS= 5.57 in.

THERMAL FISSIONING = 68%

06 AVS. FLUX, #,, AT STEADY-STATE DESIGN POWER=6.6 X 10"
' DESIGN POWER = 3 megawatts

STEADY-STATE FUEL TEMPERATURE AT DESIGN POWER=800°

T CONSTANT INLET COOLANT TEMPERATURE = 623°C —

NEUTRON LIFETIME =14 x 10 % sec

8.0

LY0ddY¥ SSAYIOHd ATHIALEVND 12Af0¥d 4NV

   

(¢>—¢0)/¢0, RELATIVE EXCESS FLUX, OR POWER

84 =0.009125 —]

0.40

  

20

 

 

 

0 20 40 60 80 100 120 140 i60 180 200
DEPARTURE OF FUEL TEMP. (°C) FROM STEADY~STATE DESIGN VALUE

Fig. 3.16 - Phase Plot of Flux vs. Fuel Temperature for VYarious Step Imcreases in Reactivity,
DWG. 12895

 

 

   

 

 

 

 

 

 

 

SECRET
100
| 17T 1 1T 1 ' f ] 1 T T ]
ANP-PHY- 129
90 }— ]
E  ap REACTOR PROPERTIES (DENTIFICATION, A=) ]
2  COMPOSITION {VOL FRAGTION)
e | FUEL
&, 70— %‘=oé_fl? SOLUTION BeO Na INCONEL ~ —
< / 7 CORE 84 0.01705 0.89000 007160 002180
b REFLECTOR 513 0.88910 0.00636 0.09/60
w,, 60 t— _—
-
<
wvs L . GORE DIAMETER=LENGTH=3ft_ o o
S x 50 — Bk _ o opp A REFLECTOR THICKNESS =5.57 in, ] 2
w < kYR, THERMAL FISSIONING = 68% -
ul
v a0l AVG. FLUX, ¢, , AT STEADY-STATE DESIGN POWER=6.6 x 10" _ rm
< CONSTANT INLET COOLANT TEMPERATURE =
&y Ak A Ak/k = FRAGTIONAL CHANGE IN EFFECTIVE MULTIPLICATION CONSTANT =
B 20 L— 21 —pi64 20 | S
& 3 k "m Am/m=FRACTIONAL CHANGE IN URANIUM INVENTORY —
< NO MOOERATOR EFFECTS &
& LK - pa6 A -
‘ 20 — k ‘ m — 2
&
=z A
..i:ojgaé_’? %
10— k — :3
S
I 1 i 4 4 . 4 £ z
O I [ 1 1 i 1 i T 3 oy r T g
O 04 08 {2 16 20 24 28 32 36 40 44 48 52 56 80 -
~ TIME (sec) o
=

Fig. 3.17 - Power Response of ARE to Step Reactivity Change of 0.009125% with
VYarious Mass-Reactivity Coefficients. '

g5
ICel
 

wn »
P DWG. 12895 %
SECRET
T T T T T T T T T T T T T 1T T 1T 1T T 17 L g
520 ANP-PHY-154 —]
S
REAGTOR PROPERTIES {IDENTIFIGATION, A-1] — <o
480 COMPOSITION (VOL FRACTION) ] (:3
FUEL
SOLUTION BeO Na INCONEL o -
CORE 84 001705  0.89000 C.07I60  0.02180 =
440 REFLEGTOR 513 0.889I0 0.00636 0.09160 — >
=
wi -—
> CORE DIAMETER = LENGTH= 3 #t d
£ 400 REFLECTOR THICKNESS = 5.57 in. ] g
- THERMAL FISSIONING = 689 =
2 AVG. FLUX, ¢, AT STEADY-STATE DESIGN POWER = 6.6 x 107 — !
b DESIGN POWER = 3 megawatts
w 360 STEADY-STATE FUEL TEMPERATURE AT DESIGN POWER = BOOD°C — ~
2 CONSTANT INLET COOLANT TEMPERATURE = 623°C g
o NEUTRON LIFETIME = 1.4 x 1079 sec T &
§ 300 g Ak/k =FRAGTIONAL GHANGE IN EFFECTIVE MULTIPLICATION GONSTANT ] =
= = .
o g Ak/k =0.082 Am/m Am/m= FRACTIONAL CHANGE IN URANIUM MASS IN CORE ™
th q "__,——" — ‘m
Q -
CE, I //’ wn
[ied 280 ~ — —
= = < =
& -~ —_
o |‘ o Aktk=0.164 A m/m g
& 240 7 —
2 ! s =
2 i =~
g ! / —
T e S 7
4 g f T Akzk=0.246 Am/m
2 |/ - _
o ’/ T — -
B 160 I ;/ e e T 7
s / v T —_
€ S e Aksk=0D.328 Am/m
E |20 !, //’ ey
& Ilf /'/
o ll_/' —
1
80 Ifif —
]
40 —
i ; : ; ; i ! | i { ' i i ,
0 L bbbt e b [ |
04 0.8 1.2 16 2.0 2.4 2.8 32 36 40 44 ag 5.2 56 6.0

 

 

 

 

 

 

TIME (sec)

 

Fig. 3.18 - Response of Fuel Temperature to a Step Increase of Reactivity of 0.009125 for
Various Effective Mass-Reactivity Coefficients.
DWG. 12897
SECRET

1T T 1 1 1 1 1 ' 1 1 7 T T T

o~ T _I 3
6 L ANP-PHY- 51 REACTOR PROPERTIES ({IDENTIFICATION, A1) B

 

   

 

 

 

COMPOSITION (VOL FRACTION)
- L ‘ ‘ ' " FUEL o o —
0.5 SOLUTION BeQ Na INGONEL
6 — GORE 84 0.01705 0,89000 O0.07i160 0.02180 _|
020 REFLECTOR 513 ' ' 0.88910 000836 0.09160
CORE DIAMETER = LENGTH = 3 1 ]
W4 — REFLECTOR THICKNESS =5.57 in. i
z 4  THERMAL FISSIONING = 68% ‘ I
i - AVG. FLUX, ¢, AT STEADY -STATE DESIGN POWER =6.6 x10'® ]
o STEADY-STATE DESIGN POWER = 3 megawatts
< 12— ASYMPTOTIC FLUX =20.4 x 10'? —
= 0.10 ’ ASYMPTOTIC POWER = 9,3 megawatts
. CONSTANT INLET COOLANT TEMPERATURE = 623°C —
@ 0 25 Y AVG. FUEL TEMPERATURE AT DESIGN POWER = 800°C
Q10— ) : ASYMPTOTIC AVG. FUEL TEMPERATURE = 1017°C ]
> g S
> & =
g [ |2 — -
o4 gl el
- T = -] W
o o
& o
~ & 0.30 ] <
5 | — S
= S
4 1005 ] 2
&
— 0.40 —
o A
o LI 0.60 : -
) ) .00 500 C o eed
| | - Z
L Ll by 3
O : rpy
0 20 40 60 80 oo 120 140 160 180 200 220
DEPARTURE OF AVG. FUEL TEMP (°C) FROM STEADY-STATE VALUE ' =
~Fig. 3.19 - Phase Diagram of Flux vs. Fuel Temperature for Response of ARE

to a Step Increase in Reactivity of 0.009125. Fuel conductivity arbitrarily
increased by a facter of 10, ' |

1S61

i
&
 

ANP PROJECT QUARTERLY PROGRESS REPORT

one term of several in the heat-
transfer expression) does not markedly
affect the shape or magnitude of the
power pulse,

Neutrom Lifetime. Variation of
neutron lifetime | was calculated only
for the case of 8k less than the
delayed neutron fraction (and with the
stabilizing effect of the fuel expansion
neglected in order to see more clearly
how the lifetime affects the growth of
flux). For 8k, larger than [, the
delayed neutron fraction (and no self-
stabilization), the flux grows, of
course, approximately like

5oc o PHeTA) N

When 8k _ is less than the fraction
of delayed neutrons, the flux also
increases very rapidly initially (as
if all fission neutrons were prompt)
with an initial slepe inversely pro-
portional to the lifetime., However,
after a time which 1is directly pro-
portional to the lifetime, the flux
curve flattens out with a rather sharp
"knee"™ and rises from then on with the
"stable'" reactor period {(when there
are no self-stabilizing features).
The stable reactor period is a function
of the 8k _ but not of the neutron
lifetime. The rapid initial rise 1in
the flux is due to the fact that the
rate at which delayed neutrons are
being returned to the population 1is
almost equal to the rate at which
delayed-neutronemitters are being
created, while the flux 1s still near
the steady-state value. The reactor
thus behaves momentarily as 1if all
fission neutrons were prompt. These
facts are borne out qualitatively 1in
comparing the flux response to a dk_ =
0.002 for three different assumed

56

lifetimes (Fig. 3.20). Equation
11.35.3 of TID-386¢1%) provides a
mathematical interpretation of the
flux behavior as obtained (Fig. 3.20).

Although the effect on reactivity
of the fuel expansion was not included
in these calculations, it is apparent
that for step changes in reactivity
less than the fraction of delayed
neutrons the peak power 1n the power
pulse (when fuel expansion is allowed)
will be roughly independent of the
neutron lifetime for lifetimes smaller
than 10°* sec.

Start-Up Accident. In regard to
the investigation of the response to a
start-up accident, the results (Fig.
3.21) indicate that a step change in
reactivity at low power will be less
serious than the same step change at
high power from the point of view of
fuel temperatures reached,

Coolant Temperature. The response
of the flux and fuel temperature to a
step increase i1n the inlet coolant
temperature of 100°C is shown in Fig.
3.22. Since the rise 1in the coolant
temperature as the cooclant passes
through the ARE core (at design power
of 3 megawatts) 1s 194°C, the drop
through the heat exchanger is roughly
the same. A rise in reactor inlet
coolant temperature of 100°C implies
a rise of heat-exchanger outlet temper-
ature of approximately 100°C. The
temperature drop through the heat
exchanger must have been 94 instead of
194°C., This implies that the external
power requirement was suddenly reduced
from 3 to (94/194) x3 = 1.45 megawatts,
Calculations on the reactor were made
holding the reactor inlet coolant

 

(IO)S. Glasstone and M. C. Edlund, The Elements
of Nuclear Reactor Theory,.
(November,

Part III. TID-386

1950).

 
 

DWG. 12898
SECRET
20 1 ; I f ! o ‘ i ’
0T . o R b L T T T T
- _ - _ _ S S - ANP-PHY-185 —
18— REACTOR PROPERTIES (IDENTIFIGATICN, A-1) —
COMPOSITION (VOL FRAGTION)
FUEL T
e b SOLUTION BeC Na INCONEL
CORE 84 0.0ITO5 083000 00760 002180 —

REFLECTOR 513 0.88810 0.0083e  QQ960

CORE DIAMETER={LENGTH=3 ¢
N REFLEGTOR THICKNESS = 5.57 in.
THERMAL FISSIONING=88%

2 — AVG. FLUX, ¢, AT STEADY-STATE DESIGN POWER =66 X o' —_
DESIGN POWER = 3 megawotts
T STEADY-STATE FUEL TEMPERATURE AT DESIGN POWER = 800°¢G T
CONSTANT INLET GCOOLANT TEMPERATURE .

   
 
  
  

- NEUTRON LIFETIME = 1.4 X 1077 sec

   

(¢ —dy)/,, RELATIVE EXCESS FLUX, OR POWER
o

‘0T HAGWILLAS ONIONH dOI¥Ad HOo4

NEUTRON LIFETIME = 0.4 %1077 sec —

NEUTRON LIFETIME =14 % 10" % se

[¢]

 

 

§ } i ' { | ! t ) i » : 1 ) ! i :
! P b b Lo e

2.8 e 44 4B 5.2 55 8.0
TIME {sec}

 

[

Ca
[
™
<
m
o
ot
i
o
M
4
i
m
)
0

Fig. 3.29 -~ Response of Flux to g Step Yncrease in Reactivity of 0.002 for
Various Average Neuiron Lifetimes. HReactivity change due to thermal expansion
3 of the fuel not included.

1861
o
o]

 

(p—d,)/¢,, RELATIVE EXCESS FLUX, OR POWER

DWG. 1289¢

 

 
 
  
    
   
 
 
 
  
 
 
 
  

 

 

 

SECRET
o s : s
6.0 — ANP-PHY-139
0.30 REAGTOR PROPERTIES (IDENTIFICATION, A-1)
— COMPOSITION (VOL FRAGTION) -
0.80 FUEL
50 — SOLUTION BeO Na INCONEL —
CORE 84 0.01705 0.89000 0.07160 ©.02i80
— REFLECTOR 513 0.88910 0.00636 0.09160 —
40 — CORE DIAMETER=LENGTH= 3 t1 ]
REFLECTOR THICKNESS =557 in.
. 0.75 THERMAL FISSIGNING = 68% __
AVG. FLUX, ¢, AT STEADY-STATE DESIGN POWER=6.6 x 10"
30 b— DESIGN POWER =3 megawatts —
AVG. FUEL TEMPERATURE AT DESIGN POWER =800°C
- CONSTANT INLET COOLANT TEMPERATURE = 623°C —
NEUTRON LIFETIME = 1.4 x 10™ % sec
2.0 |— —
10— —]
- 0.64 |
o
O . —
048
-1.0 — 0 —
o
-200 -180 - 180 -140 -120 -100 -80 ~80 -40 -20 0

DIFFERENCE OF AVG. FUEL TEMPERATURE (°C) FROM DESIGN VALUE

Fig. 3.21 - Phase Diagram of Relative Excess Power (Flux) vs. Fuel Tempera-
ture, For response of ARE to a step increase in reactivity of 0.009125 when
reactor power at 300 watts and initial fuel temperature are the same as the de-
sign coolant inlet temperature.

20

LY0dA¥ SSAY90Ud ATHALYVNO XLJAf0Ud dNV

 
 

FOR PERTOD ENDING SEPTEMBER 10, 1951

 

    

DWGE. 12900
: SECRET
+ .05 ’ I l I
0 ‘ , ANF-EHY-155
000 -
-005 ]
~0.10 -
t

o

L

=

2

& -0.5

x .

< REACTOR PROPERTIES (IDENTIFICATION, A~1)

_:)_, COMPOSITION (VOL FRACTION)

L. e

o —020 — FUEL ) ]

& : SOLUTION Be Na INCONEL

2 CORE 84 O.OI705 282000 007160 0.02180

L REFLEGTOR 513 0.839I10 000635 0.03160

Lul ; :

Zo-025

=

CORE DIAMETER =LENGTH =3 ft!

=
oL

 

 

 

G REFLECTOR THICKNESS = 8.57 in.
& THERMAL FISSIONING = 68%,
SL030 RIG.FLUX, ¢, AT STEADY-STATE DESIGN POWER=6:6 % 107
= DESIGN POWER = 3 megowatts
& NEUTRON LIFETIME = 1.4 2 107 * sec o
STEADY-STATE FUEL TEMPERATURE AT DESIGN POWER =800°C 7
o 5 ASYMPTOTIC FUEL TEMPERATURE AFTER CHANGE = 815.6°C ' /
035 e _ _ —
” ASYMPTOTIC POWER AFTER CHANGE = i 58 megawatts ////
e
=040 - // e
DOTTED PORTION OF GURVE ESTIMATED '/,"/
: -"‘\>//
_ -
~045 — —
- CALGULATED T
POINT e .
K. C. 1020
0 | | B | | | | L
O 10 20 30 40 50

DEPARTURE OF FUEL TEMPERATURE FROM STEADY-STATE DESIGN VALUE (°G)

Fig. 3.22 - Phase Plet of Flux (or Power) vs. Fuel Temperature, After 1in-
crease of 100°C in inlet coolant temperature, ' :

59

 
 

ANP PROJECT QUARTERLY PROGRESS REPORT

temperature at 100°C hotter than the
steady-state design value, and it was
found that the reactor would,
long time,

after a
steady out at a new power
of 1.575 megawatts, assuming no change
in control-rod setting. This is very
close to the new power demand, but,
even 1f the coolant circuit from heat-
exchanger outlet to reactor inlet was
long enough to allow the reactor to
settle to this new power, several more
circuits of the coclant accompanied by
small incremental temperature adjust-
ments would be required before the
power output settled to the
requirement.

new

Actually,
regulating control rod in motion by an
error signal which 1s such a function

it 1s planned to set the

of the 1inlet and outlet coolant temper-
atures that the point in the reactor
coolant pass which 1s at 1250°F under
design power remains at this tempera-
ture (this point is about one-third of
the way through the reactor). The
inlet coolant temperature, outlet
coolant temperature, and fuel tempera-
ture will adjust themselves to new
conditions when a new power reguire-
ment 1s made, but always with the
fulcrum point of the coolant at 1250°F,
The scheme ensures that the coolant
pumps which are located at the reactor
coolant inletendwill not be subjected
to high-temperature coolant {(approxi-
mately 1500°F) when the reactor is at
low power. This would occur 1f the
control rod was not moved.

An interesting feature of the curves
showing the response of the flux to a
step increase 1n coolant temperature
of 100°C is the "thermometer" effect
of the fuel-tube expansion.

This

60

expansion lets more fuel into the core
momentarily and the power increases.
The fuel 1tself soon feels the thermal
effect of the coolant temperature,
however, at which point fuel 1s lost
to the core owing to expansion above
the B,C curtain, and the power starts
down.

PREPARATORY PHYSICS CALCULATIONS

Correction to "Wigner Formuia for
Resonance Escape" Probability(!!)
(M. C. Edlund, Physics Division)., In
the present multigroup method the age-
velocity equations without a source
are made homogeneous in the slowing-
down density by assuming that the flux
per unit lethargy 1s related to the
slowing-down density by

q(u)
d(u) - .
£z, + 3, + DB?)

(1)

 

This expression, correct for hydrogen
moderator, does not hold for heavier
moderators. A somewhat better approxi-
mation than that given by Eq. (1) for
nonhydrogenous moderators can be
obtained from the integral represen-
tation of gq.

 

(ll)This procedure for obtaining a correction
to the Wigner formula, and in particular the
method of eliminating the d[Z (u)]/du from the
Taylor’'s series expansion of zqu(u') was polnted
out to the author by E. Greuling and G. Goertzel.
An NDA report by G. Goertzel is forthcoming.

 
 

FOR PERIOD ENDING SEPTEMBER 10, 1951

The slowing-down density in a mix-
ture of N elements 1s given by

 

 

Ii - 1~ . J {E%i Plu)
T
- .
i
N n ~
\ Zsi{ug) dlu’)
glu}) = e )
I - a. diza- PHlu)] ,
~ t A s - {ua'- u) (e "8 a.)du'.(3)
pnl ey du * '
[:”””—-ai du’ (2)
LA N Integrating,
here .
e Iy =8 25 dlu) + -1 Fa,
M. - 13?2
B L
4:1,1- - e e
.‘., I
Mi 1 fl.i:il l d[gsi (b(U)]
+ €, + (4)
. : 2 1 - a. an
M. = mass number of the 1th scatter- l

ing nucleus,

€. = In (1/@i)~

Each Z_. ¢pu') in Eq. (2) 15 ex-
panded as a Taylor's series about
lethargy u. 1If there was no capture
or leakage, 2_ ¢{u’) would be constant,
Hence if the capture cross-sections do
not vary much in a lethargy interval
¢, the variation of ES P{u’) mav be
given quite well by just the first two
terms of the expansion. The integrals
in the summation Fq. (2) after expansion
of ES H{u’) are of the form

 

61
 

ANP PROJECT QUARYERLY PROGRESS REPORT

Again expanding %_ ¢{(u’') aboutu and
taking only the first two terms, Eq.
(5) becomes

dg d 1
—=N 5. ) -
du 1-a.

=1

d[ZS Plu)]
+ -_—

(' - a)l  e® v gyt (6)

du

(6),

N
d ZE:? &, 2y elu)
i=1

Integrating Eq.

dq (7)

du du

and substituting Eq. (4) into Eg. (2),
we obtain
N
g ‘—5& &, 2y, Plu)
i=1
N
B
B; g; Plu) (8)

 

du

62

where

a.€. 1

 

2 l-—ai

Defining average £ and 8 as

N
7 E‘::i Zs 1
i=1
é’ e
2 s
(9)
N
} Si Zs t
T
Bz
2

where 2_is the total macroscopic scat-

 

 

tering cross-section, FEgs. (7) and (8)
become
dg g d[f 2. Plu)l
du du
d& _dlE; ¢(u}]
= 2, p(u) + & »(10)
du du

_ dIB %, ¢(u)]
g ~¢&2 o) - ——y
s du

 
 

FOR

or

df3

P

 

 

g =& Z, plu) ~—2 @lu),
du
_ dZ, plu)]
- B - . (11)
du

The derivative termS'hmis ¢(u) can now
be eliminated from Eqs.(10) and (11),
yielding

dq

 

B
du

S p). (12)

In a finite system, the rate of
change of slowing-down density with
lethargy is equal to the total rate of
loss of neutrons 1in the system; 1.e.,

d .
29 - (s + DB?) #(u)
du a

(13)

where DB? #(u) is the rate at which
neutrons leak from the system,

Equations (12) and (13) now de-
termine the relation between ¢(u) and
g{u). Eliminating dq/du, we obtain

 

- glu)
d)(u) T”ES + ')/(Za 4 DBZ)’

(14)

 

PERIOD ENDING SEPTEMBER 16, 1951

where

If all the Esi(g)’s vary the same way
with energy, df3/du and d&/du become
equal to zero, and Eq. (14) becomes

 

_ q(fi)
P T ETG,

+ DRz)

If we have a single moderator,

1 - a - ae ~aé/2

y o BN R (15)
1 - a - ae
recalling that &€ = 1 - ae/{(l-a) and
€ = -In a. Rewriting Egq. (15},

- a toa . 2 r
1l -atalna - {aln Q)/zn(lfi)

e - a In a

 

In the limiting case of hydrogen

moderator,

a = 0 and lim vy = 1,

a0
which checks with the rigorous result.

63
 

ANP PROJECT QUARTERLY PROGRESS REPORT

In the limit of infinite mass scat-
tering nuclei,

a = 1 and lim vy = 0.

a1

In this case there 1s no slowing down,
and from Eq. (14) ¢ = 0.

The resonance escape probabilaty in
an infinite medium of a single type
moderator may be obtained by eliminat-

ing ¢(u) from Egs. (12) and (13)
and integrating. In this case B? = 0,
and S(u), the neutron source, is taken

to be zero in the resonance region.
The resonance escape probability 1is

u

a

>
p(u) = exp ~| —m-—
&x, t yz

a

du.(17)

0

A. M. Weinberg has calculated the
correct resonance escape probability
for the case of constant cross-
sections {1?) the result is

p(u) = e-,u.u

For carbon moderator and EG/ZS = 0.2,
(= 1.118817. According te Eq. (17),

1

y_: =

E — = 1.116024
sgs/za Ty

 

(12)A. M. Weinberg and L. C. Noderer, Theory
of Neutron Chain Reactions, ORNL CF-51-5-98, p. 37
(Aug. 10, 1951).

for the corresponding case. If the
resonance escape probability 1s given
by the Wigner formula,

=

zfl.
p(u) = exp - | ————— du,
. E(Z, + 2))
_ 1
Lo — = 1.056401,

 

E(Z /5, + 1)

and, if the weak capture formula 1s

used,

1) zd

du,

 

plu) = exp -
£
0

p = E fES = 1.267681.

It is apparent that for the regions of
heavier capture the best result 1is
given by Eq. (17) for the case 1in which
the absorption cross-section variles
only slowly with energy.

Effect of the "Wigner Formula for
Resonance Escape" Correction (C. B.
Mills, Physics Division). It has been
shown by Edlund in the preceding
section, using a method of Greuling and

Goertzel, that a more correct relation
 

FOR PERIOD ENDING SEPTEMBER 10, 1951

 

 

 

 

 

 

 

between the neutron slowing-down DBr+'Za : og(u)
density g and the flux ¢ is — ; glu) +
{ * gzs oy, DB?) ' du
| g(u)
Plu) = — ‘ = fission source.
EE_ + y(E_ + DB?) |
where ~ : The corrective term y(Z_ + DR?) mus t
| : be inserted into the solution of the
usual recursion formulas. Thus the
average values required for multigroup
B 1 - a - ae - a€2/2 constants are :
y E-m 1l -a - a€
N
for a single moderator., Here 1
€ = 1n l/a, ‘ . : 3§zsztr ¥ 7(32aztr + BY
M-1y? | | N
a = [-~] where M is the atomic ' zb
M+1 number, : ' 2
fis + y{DB% + Za)
£ = average loss in log of energy |
per collision, ‘
: N
¥ = scattering cross-section (cm™1), z,
s : : : -
2, = absorption cross-section (cm™ 1), ' gzt vy + DB?)
D = diffusion constant,
B? = geometrical buckling.

The values of & = & and v for the

Introduction of this definition into beryllium-moderated critical experi-
the bare reactor eguation{!3) gives . ment are

Uy
I

0.2078,

 

(13)M. J. Nielsen, Bare Pile Adjoint Solution,
ORNL, Y-12 site, report Y-F10-18 (Oct. 27, 1950). Y

il

0.1427.

&5

 
 

ANP PROJECT QUARTERLY PROGRESS REPORT

A hand calculation was made 1n the
usual way for keff' The values are

Corrected kcfif = 0,91617,

Uncorrected kqfi = 0.90347.

The effect of the correction is seen
to be small but significant,

The Transmission Coefficient of the
B,C Curtainin the ANP and ARE Reactors
(C. B. Mills, Physics Division). A
boron carbide curtain 1s part of the
design of the ANP and ARE reactors.
The transmission coefficient as a
function of lethargy of this curtain
is required for determination of
criticality coefficients and the
neutron lethargy and space distribution.

The B,C curtain is 2 in. thick in
the ANP and 1 in, thick in the ARE
reactor. In both cases i1t 1s a dense
layer thickly studded with coolant-
tube holes. For purposes of calculation
the transmission of a solid slab is
first computed and then a constant
correction due to the total aperture
of the holes is added. The transmission
coefficient is almost unity for the
high-energy neutrons. For thermal and
near-thermal neutrons, the holes
provide the only leakage through the
curtain.

The transmission 7y of a layer of
thickness T and macroscopic absorption
cross-section 3_ is given by the
integral

e~ Y

 

y = 2(2,T)?

66

This function has been integrated
numerically,(!*) The leakage through
the holes i1s evaluated by assuming a
cosine distraibution of neutron ve-
locities (as for ) with a probability
p{x) of being x cm off the axial line
at each hole being 27x. A projection
of the solid angle subtended by the
reflector side of the hole permits
evaluation of the probability of a
neutron traversing the opening. A
table of the fraction of the neutrons
penetrating the B,C curtain at each
lethargy group is given in Table 3.5.
This table will be used for a trial
solution for the neutron flux in the
vicinity of the absorber, An iteration,
requiring new values of v to fit the
more accurate form of the flux, may be
required,

Hieatinmg in the Boron Carbide Curtain
in the ANP Reactor (C. B. Mills,
Physics Division). A solution of the
reflected ANP reactor, including the
effect of the boron carbide curtain
intended to shield the fuelreservoirs,
is not yet complete. The possibility
of some secondary effect preventing
the use of this curtain 1s of some
interest prior to a more exact calcu-
lation., One such possibility is the
heating in the curtain due to absorption
by the B!'%(n,a)Li’ reaction, with a
release of 2.88 Mev.

The model used for an approximate
value of total heating per square
centimeter and power distribution is a
bare reactor with the ANP power distri-
bution, The boron curtain side of the
ANP reactor will be essentially bare
because of the high absorption and low
reflection.

 

(14)B. R. Coveyou and J. E. Bradley, Tabulation
of F-Functions, &1-1629 (May 1944).

 
 

?Gfi PERTOD ENDING SEPTEMBER 160,

1951

 

 

 

 

 

 

 

 

 

 

 

TABLE 3.5
B,C Transmission Coefficients
ANP (2 in. B,C) ARE (1 in. B,C)
v (holes} : v (holes) _

GROUP | v (slab) | (%-in. dia.) |y (total) v{slab) (%-in. dia.) y {total)

1 0.95671 0.01024 0.967 0.97800 0.00211 0.980

2 0.94504 0.01301 0.958 0.97200 0.00268 . 0.975

3 0.93051 0.01643 0.946 0.96428 0.00342 0.968

4 0.91235 0.02073 0.933 0.95464 0.00434 . 0.959

5 0.88990 0.02603 0.916 0.94244 0.00551 0.948

6 0.86228 0.03257 0.895 0.92720 0.00697 0.934

7 0.82874 0.0300 0.839 0.90828 0.00878 0.917

8 0.78802 0.0250 0.813 D.86492 0.01101 0.896

9 0. 57498 0.0150 0. 590 0.64038 0.01610 0.656
10 0. 14577 0.0065 0.151 0.17710 0.02150 0.1986
11 0.00580 0.00539 0.0112 0.05238 0.02610 - 0.07858"
12 0.00012 0.00539 0.00551 0.00524 0.03276 0.0380
13 0 0.00539 0.00539 0.00028 0.03276 - 0.0330
14 0 0.00539 0.00539 0. 60002 0.03276 0.03276
15 0 0.00539 0.00539 0 0.03276 - 0.03276
16 0 0.00539 0.00539 0 0.03276 0.03276
31 0 0.00539 0.00539 0 0.03276 0.03276
91 0 0.00539 0.00539 0 0.03276 ° 0.03276.
92 0 0.00539 0.00539 0 0.03276 . 0.03276
03 0 0.00539 0.00539 0 0.03276 | 0.03276
The absorption distribution was thickness of the B,C layer. The powér

determined for each of the 31 lethargy
groups by the usual current attenuvation
method expressed by the F-function

where b = Z d; Z

absorption cross sectlon and d is the

15 the macroscopic

 

dlstrlbutlon was determlned'byafldltlon
of the contrlbutlon_of each lethargy
group at several points through the
B C curtain. The total power was
46 watts/cm? for the ANP reactor. A
similar calculation gave 0.69 watt/cm?
for the ARE reactor, The power distri-
bution is given in Fig. 3.23. Most of
this power is concentrated near the
core surface of the curta1n. ‘

Temperature Savings in the ANP

Reactor (R. J. Beeley, Oak Ridge School .
of Reactor Technology, and C. B. Mills,

67
 

ANP PROJECT QUARTERLY PROGRESS REPORT

 

 

 

 

 

 

DWG. 12901 -
SECRET
100 §— —
- ANP-PHY-137 ]
5 50 ||— .
(0 /
] — tharr,-1) — -
a. ABSORPTION = 2—70---—— B
o H— j.—:o X ]
L
m
e @
8 1 ? ]
@ dil,~1) 210()() g2
< ot e | e
&n WHERE e 7 \7 3 a7
é x/L
-
!_—) 0 i~ ]
% i X = DISTANCE FROM CORE SIDE OF THE B,C LAYER ]
~_ 1 N . = NEUTRON MEAN FREE PATH FOR ABSORPTION 7
N 5 I\ Io= [o(M)=INCIDENT CURRENT INTENSITY AT ]
IR SURFACE OF B,C LAYER -
< B TOTAL POWER REPRESENTED BY THIS ABSORPTION |
o ASSUMING SAME CURRENT LETHARGY DISTRIBUTION | ]
=N FOR BOTH ARE AND ANP REACTORS:
= - ARE, 0.69 wafis/cmz; ANP, 46.0 wafi's/cmz J
= THIS CURRENT WAS COMPUTED ON SAME BASIS
8 OF | FISSION PER cubic centimeter PER second
@
l_
O 10—
o [-—f
z " -
9 .
5 ol
© 05— .
a |
<I
z b
O
o
‘_._
0
L — —_—
=
S L

 

 

0 0.5 1.0 1.5 2.0 25
X, DISTANGE FROM CORE SIDE OF 84(3 CURTAIN {(cm)

Fig. 3.23 - Neutron Absorption vs. Pepnetration with the B4C Layer of the ARE.

68

 
 

FOR

Physics Division). One of the limit-
ing factors in reactor power density
1s the high temperature at which
the structural and heat-transfer
metals must operate., The highest
temperature to which reactor metals
are exposed 1is at the inner surface
of the fuel tubes. A rough estimate
of the maximum value of this tempera-
ture 1s made below. This is compared
to the temperature that would be re-
quired for the same total heat transfer
1f the temperature at the inner sur-
face of the reactor fuel tubes is held
constant over the length of the reactor,
An axi1al line is chosen for maximum
temperature evaluation.

The axial reactor power distribution
is given, by curve fitting to four
points, by

Q(z) = 1.76996 + 134.57~f

2 3

380.36 |—| + 211.63 |2
36 ]+ 2ues [

for radial heat flow from am axial
fuel tube, where Q(z) 1s the calories
per centimeter per second and h 1is
the reactor length, 98.4 cm.

The fuel temperature, T,(z), is
given by

z

d
Tz =T o+ |24z
3 %3 MH
0
where
T, = 704°C, the inlet temperature,
3

 

PERIOD ENDING SEPTEMBER 10, 1951

M= 27.9, the mass of sodium per
second (g/sec) passing z,
H= 0.32, the heat capacity of

sodium (cal/g/°C).

The temperature, T (z), at the inner
wall surface of the reactor is given
by the relation

kAl
(Q(z) + hAZTS(z)] 1+

htd,

- hAsz(z)

-3

 

(z)=
RA,/t

The physical constants required to
evaluate Tl(z) are:

E= 0.0670 cal/sec/cm/°C {(heat

conductivity of fuel tube),

h = 2.3 cal/sec/cm?/°C (sodium film
transfer coefficient), '

tube area:

R
o
i

= 1.2 em?/cm,

-
H

2 1.57 cm?/cm.

Tube radii:

0.159 em (inner radius),

r, = 0.222 cm (outer radius),
t =r,-r, = 0,063 cm (tube wall
thickness),
7 = 0,191 ¢m {average radius),
Therefore,

T, (z) = 1.1237 Q(z) + 0.99987 T,(z),

69
 

ANP PROJECT QUARTERLY PROGRESS REPORT

and

948°C at the outlet end of
the reactor for a heat flux
of 1990 calories per coolant
tube per second.

T _

1 (max)

The corresponding uniform tempera-
ture 1s obtained by requiring the
power distribution to be proportional
to the coolant temperature in such a
way that a surface between fuel and
coolant has a constant temperature

throughout the reactor. Thus
thd, + kA,
T, = T,(z) + ———" Q(z),
khA A,

where T,(z) is the coolant temperature,
T, is the temperature of the inner
surface of the fuel tube wall, Q(z) 1is
the power per unit length of the fuel
tube, and the constants are those
appropriate to the situation. For the
same total heat flux as above,

T, = 927°C.

The temperature saving 1is thus

approximately

AT = 21°C = 38°F

for the case of coolant entrance into
the B,C end of the reactor. For the
symmetrical case of coolant exit from
Fhe B,C layer the temperature saving
1s only

2°C ~ 4°F.

1

AT

70

THE SODIUM HYDROXIDE BEACTOR”

C. B. Mills,

Physiecs Division

Tt is necessary to determine the
critical mass of an almost spherical
reactor, 1.25 ft in radius, con-
taining sodium hydroxide as a coolant,
moderator, and reflector, The fractions

are

vol %
Sodium hydroxide moderator 75.00
316 stainless steel structure 20.23
Fuel, uranium, 50 1h/f¢3 4.77

This problem is essentially different
from those previously solved by the
ANP Physics Group because of the way
in which neutron diffusion and slowing
down in hydrogen must be treated. Most
methods of solution which treat hy-
drogenous materials correctly are nat
correct for this reactor because of
the large absorption and slowing-down
effect of the
stainless

oxygen, and
steel, An admittedly
approximate method, (1) first used
by this group to study the effect of
water on criticality on the ARE
reactor, (?%) which sums the effects of

sodium,

continuous slowing down of neutrons in
heavy materials with the discontinuous
slowing down 1in hydrogen, was applied
to this problem. This method of
solution has been discussed in some
detail separately,(17)

 

*
See Sec. 17 of this report for design features
of this reactor.

15)g, Goertzel, Criticality of Hydrogen-
fiode;ated Reactors, ORNL, ANP, TAB-53 (July 25,
950).

(16) 5, w. Webster, Effect on Reactivity of ARE
of Flooding Coolant Channels with Borated Water,
ORNL, Y-12 site, report Y-F10-67 (Aug. 14, 1951).

(17)J. W. Webster, Numerical Technique for
Criticality Calculations on Hydrogen-Moderated
Reactors, ORNL, Y-12 site, report Y-F10-66
(Aug. 20, 1951).

 
 

'FOR PERIOD ENDING SEPTEMBER 10, 1951

A convenient proof of the signifi-
cance of the results of the method
used 1n the sodium hydroxide reactor
can be found in the application of the
same method to the critical experi-
ments conducted at K-25 in 1949, (%)
Both reflected and bare, large and
small reactors were constructed and
criticality data were obtained. A
small bare and a large reflected
reactor with high and low uranium
concentration, respectively, were
chosen for the check calculation. ' In
for one calculation the

slowing-down term was

addition,
continuous
neglected for a check on the relataive
importance of this term for a known
reactor, The derivation of the re-
cursion relations used for thais will
serve to indicate the general method
used in the above hydrogenous-reactor
calculations,

The space-independent form of the
Fermi age theory used for bare reactor
multigroup calculations 1s

DB? + o og (u)
glu) + ———
§O&1 Qu
Yip
f
Sufy || L g d’ fop, bl @)
&TT
and
9eh = (DB* + Cra)qbth’

U8)c p Beck, A. D. Callihan, J. W. Morfitt,
and R. L. Murray, Critical Mass Studies. Part
117, C&CCC report K-343 (Apr. 19, 1949).

 

"in which all the terms are familiar

from previous discussions., Addition
of the contribution at each lethargy
{u) by hydrogen from all lower lethar-
gies involves the use of the term

0
nu) = e¥ [ $lu’) USH(u'} e’“f du’,

u

This can be expressed as

s

7] -
o, T osyd

~ {3)

and Eq.
there

Using this as one equation,
(1) plus 7m(u) as the second,

result the two simultaneous equations
9q
(A+ plg +t ™ =7 + g, (4)
du
om
n ot pg, (5)
ou
where
g = qlu),
N = (o, t DB®)/éog,

p - O—SH/‘fO-T:
q = o ¢,

g = fission source term.

Here T g does not include the hydrogen
scattering cross-section Ogy (to avoid
slowing down one neutron in both ways).

71
 

ANP PROJECT QUARTERLY PROGRESS REPORT

Integrating Egs. (4) and (5) over
the lethargy range u, - u, = U, after
setting 0¢/%u = 0, so that 7 is the
only neutron source at each lethargy,

there results
= %'Un t vz, (6)

- Ty (7)

Solving for ¢ in terms of Mp., and for
7, in terms of ¢ and 7 _,

 

n
fln_i Vzn
v,/2 + 1
In = - ,  (8)
e
Un o+ 0
n

v,/2 + 1

Pl v./2 -1
(O —q, - n . (9)

u,/2 + 1 u./2 + 1

These two recursion equations, (8) and
(9), are used in the usual way for a
determination of the multiplication
constant k .

eff

The effects of hydrogen and heavy
mass moderation in the sodium hydroxide
recactor are separated 1n absorption
and leakage spectra (Figs. 3.24 and
3.25). Neutron leakage and fission
spectra are shown for the two cases

72

dq/ou = 0, dg/3u # 0.

For the first, hydrogen moderation
only is 1included, and, for the second,
both heavy atomand hydrogen moderation
are i1ncluded. The same comparison 1s
made for the water-moderated reactor
in Figs, 3.26 and 3.27,. Fission
spectra and neutron leakage for the
reflected water-moderated reactor are
given in Figs. 3.28 and 3.29,.

The analysis of the applicabilaity
of the Goertzel method 1s not yet
complete (primarily because of the
difficulty with which croess-section
data were obtained and the multigroup
constants formed), but certain tentative
statements can be made:

1. Hydrogen is respoansible for
most of the neutron moderation
in the sodium hydroxide reactor.

for the ef-

fective multiplication constant
with and without the continunous

Two calculations

slowing-down effect give results

different by only 4.28% for Ak/k.

2. The multiplication constant 1n
the calculated water-moderated
reactor 1s lower than the
critical-experiment value by
3+ 1%.

3. The critical mass of the 200-

megawatt sodium hydroxide
moderated and cooled reactor is

in the vicinity of 45 1b of
uranium (enriched 93.2%). In-
control

cluding an 8% kef for
u{d

purposes, 48 1b wo be required,

 
 

-3
A

ABSORPTION PER UNIT LETHARGY FOR 2.5 FISSION NEUTRONS, 4%/ v

0.28

Cc24

0,20

0,08

O
O
I

DWG. 12902

 

 

 

   

 

 

 

 

 

 

 

 

 

 

SECRET
17 1 117177 7177 "t 171t 17 1t 1°1 1T T}
— ANP -PHY-171 —
vOL % o
““““““““““ NaOH COOLANT AND MODERATOR 75.0 _
316 STAINLESS STEEL, STRUCTURE AND FUEL TUBES  20.23
FUEL, U235 50 1b/4t3 4.77 —
MODERATION BY ALL GONSTITUENTS
e e MODERATION BY HYDROGEN ONLY
18.6 THERMAL |
N\
5
\\
\ —t
) e
\ —_—t
\ bl et I T =S
|8 6 14 2 o 8 5 4 2 0
LETHARGY,

Fig, 3.24 - Absorption Spectrum for the 200-megawatt Sodium Hydroxide Reactor.

‘0T YAAWALJAS SNIANAd aOoI¥idd HOd

TS61

 
 

 

 

 

   

 

 

 

 

 

 

~ DWG. 12903 2
SECRET -
0.8 g
I L e ’.—.F”E %
—  ANP-PHY-172 | — —
2 ! =
, 016 — = -
> {
W - : ] =
- >
0 i =
=
o4 |— —
: : 3
: - B
W VoL % B ~
z 012 — NaCH, COOLANT AND MODERATOR 75.0 =— =
& B 316 STAINLESS STEEL, STRUGTURE AND FUEL TUBES  20.23 L S
Z FUEL, UZ*° 150 Ib/£+3 4.77 r- | e
=
00,10 — f —
3 ] &
% r | ;.__ s
v MODERATION BY ALL CONSTITUENTS | ! =
> L —
@ 008 — MODERATION BY HYDROGEN ONLY | 2
o — |
X — -
I—
L
- 0,06 — ]
=
Z - - _
=]
o
W 004 |— . .
tad
o L. e
<l
=
w 0,02 —
—
— 18.6 THERMAL .t T
o LI Ve
8 i6 14 12 10 8 6 4 2 0

LETHARGY, v

Fig. 3.25 - Leakage Spectrum for the 200-megawatt Sodium Hydroxide Reactor.
 

-3
wn

FISSIONS PER UNIT LETHARGY FOR 2.5 FISSION NEUERONS, £Yu?

o
&

C.i2

0.10

0.08

0.06

.04

0.02

o

DWG. 12904

 

 

 

 

 

SECRET
R e e e e
ANP-PHY-{92 —
MODERATION BY ALL CONSTITUENTS
__________ - e MODERATION BY HYDROGEN ONLY |
- —

-

CYLINDRICAL CORE, DIAM, 25.4 ¢m, HEIGHT 34,0 cm

N . LT
‘8 _NO OF H ATOMS PER cc =509

 

“ATOMIC RATIO
Nye3s  no OF U239 ATOMS PER oo

 

8.6 THERMAL

 

 

pos ava. . auseny

 

 

 

 

 

 

 

 

LETHARGY, v

Fig. 3.26 - Fission Spectrum for the Bare ¥ater-Moderated Reactor.

‘0T HMAGWALdAS INIONI 00I¥dd HO4

IS61

 
~3
o~

 

   

 

 

 

 

 

 

>
2
s
e
s
o
DWG. 12905 :
SECRET -
2 7 T : : =
X P N -
~
< 070 — o - — -
3 ANP=PHY -9} ;
¢ [ 1 g
© 060 (— | =
= !
2 — ———— MODERATION BY ALL CONSTITUENTS — =
& 050 — ————— MODERATION BY HYDROGEN ONLY _ &
2 =
i — - o
o W
N 040 . -
& CYLINDRICAL CORE, TIAM. 25.4 ¢cm, HEIGHT 34.0 cm %
O
o — ATOMIC RATIO N, /N 235 = 52,9 g
>
© 030 |— =3
<
T
’_‘  ——
Led
-
L 0.20 —
=
D  —
x
ul
& 010 —
0
v
<I I
= 18.6 TH®
w 0 —— | ndi i ] I [ ! i
18 16 14 12 10 8 6 4 2 0
LETHARGY, v

Fig. 3.27 - Leaksge Spectrum for the Bare ¥Water-Modersted Reactor.
DWG, 12906

 

 

   

 

 

 

 

 

 

 

 

 

SECRET
0.036 ,
T T T T T 7T T T T
. o o | | o -~ o | - ANP=PHY-I95 —
> 0032 — —_
3
o L ]
W
(é}h o028 — ]
S CYLINDRICAL CORE, DIAM. 25.4 cm, HEIGHT 22.4 om
5 | ATOMIC RATIO Ny /N ,235 =328.7 - o
L
Zz 0024 — —
=
o
v - —
o
“ 0020 — ]
Wy
i - ~ S
tad
& =
0,016 — —
> =i
T - }_ __________ 1 ] g
2 ; FRACTION THERMAL ! —
- i FISSIONS =(.813 { &
wooolz — e 4 — =
z — \ _ ™
- \ =
008 ] >
o \\ a
w3 : Y,
L —] w
& \ 4
b - g
=
1 {
. ! O O S T O T =
I8 16 14 2 o
o=
LETHARGY, v .

Fig. 3.28 - Fission Spectrum for the Reflected Water-Moderated Reazctor.
Moderation by all constituents,

LL
IS61
-3
oc

 

LEAKAGE PER UNIT LETHARGY PER 2.5 FISSION NEUTRONS, Vo

o
~
<

C60

0.50

0.40

0.30

G.20

0.10

O

DWG.

12207

SECRET

 

 

ool

CYLINDRICAL CORE, D!AM. 25.4 ¢m, HEIGHT 22.4 cm

L ATOMIC RATIO N /N je3s = 328.7

 

 

ANP-PHY-193 —

 

 

 

Moderation by all constituents,

— 18.6 THERMAL
] l —— ] = i 1 ] i |
8 i6 |14 12 10 8 6 4 2 0
L ETHARGY, u
Fig. 3.29 - Leakage Spectrum for the Reflected Water-Moderated Reactor.

JHOJHY SSAU90Hd XA'MILYVNO LDAT0Ud ANV

 
 

FOR PERIOD ENDING SEPTEMBER 10, 1951

4. CRITICAL EXPERIMENTS

A. D. Callihan, Physics Division

The Critical Experiment Group,
which is responsible for the experi-
mental investigation of preliminary
reactor assemblies, has completed the
investigation of one assembly and is
now taking data on another. The
first mock-up was that of the aircraft
reactor, proposed by General Electric,
which employs water as a neutron
moderator and air as a heat-transfer
medium. - Two modifications of this
reactor with the same hydrogen-to-
carbon-to-uranium atomic ratlo were
considered, and the distribution of
thermal neutrons throughout the com-
ponents was compared. The second 1in-
vestigation, still in progress, 1s a
study of the power and flux dis-
tributions through a graphite-uranium
reactor with carbon-to-uranium atomic
ratio of 990. '

CRITICAL ASSEMBLY OF AIR-WATER REACTOR

The study of the proposed air-water-
cvele aircraft reactor, initiated at
the request of the General Electric
Company and executed with the co-
operation of some of their personnel,
has been completed for the present.

In this reactor metallic uranium
of high U??® enrichment was inter-
spersed among blocks of graphite which
simulated the proposed uranium—silicon
carbide core material, and a meth-
acrylate plastic served as an adequate
substitution for water. The plastic
was in 1%-in.-thick horizoental layers
between 4%-in.-thick layers of uranium-

bearing graphite. The critical mass

 

of the assembly was about 29 kg.
Measurements of the thermal-neutron
distribution through a unit cell
showed a high concentration in the
hydrogenous material. The thermal
flux in the plastic was 1.8 times the
average in the graphite and 3.6 times
that at the surface of the uranium.

The thickness of the individual
uranium pieces was 0.01 1in.; this
introduced some self-shielding {rom
incident neutrons. In an experiment
to evaluate the extent of this effect,
a standard piece of uranium was re-

placed by five pieces, each 0.002 ia.
thick, with aluminum foil separating
adjacent ones. In this technique

recoiling fission fragments collect on
the aluminum, and their activity is a
measure of the fission rate in, es-
sentially, the surface layer of the
uranium adjacent to the aluminum. A
comparison of the activities on the
successive aluminum surfaces gives,
relatively, the fission rates at
0.002-1in. intervals throughout a fuel

piece. In the reactor under study the
average fission rate throughout a
0.01-1in, fuel piece was 84% of that

occurring at its surface.

The observed large thermal-neutron
concentration in the plastic implied
that the moderator-fuel inhomogeneity
was causing a significant increase 1in
the uranium requirement. Conseguently,
the plastic thickness was reduced to
% in. and was placed between the
graphite-uranium layers, maintaining
the same H:C:U atomic ratio as before.
because of the

Tt was

necessary,

79
 

dimensions of available materials, to
alternate the ¥%- plastic layers
between 3- and 1%-in.-thick graphite
layers, thereby probably not achieving
optimum fuel-moderator homogeneity.
The critical mass of this modification
of the original assembly was 18.5 kg,
and a more uniform distribution of
thermal neutrons throughout the com-
ponents resulted.

The data are reported more fully
elsewhere. (1)
CRITICAL ASSEMBLY OF GRAPHITE REACTOR

The inaugural program for the
Critical Mass lLaboratory was the study

 

(1), Hunter, Report on Critical Experiments
for a Water Moderator {CA- 2), G.E. ANP Report
DC-51-9-11 (September, 1951)

80

of simple reactors of good geometry
for cowmparison with theoretical pre-
dictions. The first of these, de-
scribed in the-twoe preceding ANP
guarterly reports (ANP-60, ANP-65),
was a bare cube having a beryllium
moderator. Further simple reactor
studies were postponed in order to
carry out the air-water-cycle reactor
investigation described in the above
paragraphs., On completion of these
studies a graphite-uranium reactor was
assembled, consisting of the 0,01-1n,
thick uranium metal disks separated by
4-1n.-thick blocks of graphite, the
C:U atomic ratio being 990, It 1is a
rectangular parallelepiped 45 by 45 by
44 1in. Insufficient materials were
available to make the system critical
without a reflector so 3 1in. of
graphite surrounds four sides of the
core. The loading 1s about 45 kg of
uranium. Measurements of power and
flux distributions are being made.

 
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 
 

.

 
 

5. BULK SHIELDING REACTOR

J. L. Meem, Physics Division

The mock-up of the unit shield has
been completed with the water-reflected
Bulk Shielding Reactor. Preliminary
analyses of the data indicate that an
ideal unit shield for a 200-megawatt

reactor with a 3-ft right square

cylindrical core will weigh 124,000 1b.

The construction of the divided shield
is nearly completed, and preparatory
shadow shielding measurements are
reported. The spectroscopic instru-
ments required for the divided-shield
measurement appear to be satisfactory.
This reactor; formerly known as the
"swimming pool" or the "Shield Testing
Reactor," is now identified as the

Bulk Shielding Reactor (BSR).

REACTOR OPERATION

All measurements during the past
quarter, including those on the unit
shield, have been made with the water-
reflected reactor.('? Because of the
urgency of the unit shield measure-
ments, the check on the power of the
reactor by measuring the heat rise of
the water through the fuel elements
has not been completed.

Preliminary measurements using the
gamma-ray scintillation spectrometer
(see below) indicated that to measure
the spectrum at the face of the reactor
a complete set of cold fuel elements
would be required. These cold fuel
elements have been obtained. '

 

(I)J. L. Meem and E. B. Johnson, Determination
of the Power of the Shielid-Testing Reactor. 'I.
Neutron Flux Measurements in the Water-Reflected
Reactor, ORNL-1027 (Aug. 13, 1951).

 

MOCK-UP OF THE UNIT SHIELD

During the past quarter the measure-
ments on the mock-up of the unit shield
have been completed and a report is
being written. Preliminary analysis
of the data indicates that an 1deal
unit shield around a 3.8-ft-diameter
spherical reactor (approximating a
3-ft right square cylinder) operating
at a power of 200 megawatts will weigh
about 124,000 1b. Final evaluation of
this weight will be given in the above-
mentioned report. The limits of error
of all the measurements will be
analyzed so that the weight for this
mock-up can be definitely bracketed.

Measurements were made of gamma-ray
ionization, thermal-neutron flux, and
fast-neutron dosage along the center-
line beginning at the face of the
shield and proceeding outward into 'the
water. HRuns were made with different
concentrations of borated water up to
0.4% boron by weight.

With the borated water in the
shield, it was observed that the
neutron-flux distribution in the
reactor was altered. Since this would
change the fast-neutron leakage,
neutron-flux measurements with gold
foils were made throughout the reactor
following the procedure in ORNL-1027.

MOCK-UP OF THE DIVIDED SHIELD

The construction of the divaided
shield mock-up by General Electric 1is

- 83
 

ANP PROJECT QUARTERLY PROGRESS REPORT

nearing completion. A suggested pro-
gram for the measurements and calcu-
lations has been outlined. ?’ The
development of the spectroscopic in-
struments has progressed to the point
where the measurements appear quite
feasible. Using an underwater model
of a multiple-crystal scintillation
spectrometer, (®’ a preliminary gamma-
ray spectrum at 130 cm from the water-
reflected reactor has been obtained.®
The spectrum 1s shown in Fig. 5.1. For
details of the measurement see RBefer-
ence 4,

An additional experiment, pre-
liminary to the divided-shield measure-
ments, has been completed. (%) This was
a shadow shield test in which two lead

 

(Q)J. L. Meem and R. H. Ritchie, Suggested
Program for Divided Shield Measurements and
Calculation o{ Air Scattering, OBANL CF-51-8-7

(Aug. 1. 1951
(3)5. K. Bair, F. C. Maienschein, and W. B.
Baker,

Multiple Crystal Gamma-Ray SEectroscopy
Using Nal-ThI Crystals, NEPA-1701 (Fekh. 3, 1951).

(4)F. C. Maienschein and R. H. Ritchie, Pre-
liminary Gamma-Ray Spectral Measurements at the
to be

Bulk Shielding Facility, ORNL CF report No.

assigned.

84

slabs, each approximately 6 ft long,
5 ft wide, and 1% in. thick, were
placed symmetrically along the center-
line north of the reactor. The rear
of the first lead slab was 17% 1in.
from the reactor with 2 in. between
the slabs. Vertical traverses were
made behind the slabs with a gamma
ionization chamber. The same traverses
were repeated after removing the lead
slabs. The gamma traverses behind the
lead are shown in Fig. 5.2.

A reporton the nuclear plate camera
for fast-neutron spectroscopy has been
issued.(®? For further details of
spectroscopic instrument development
work, see the Physics Division Quarterly
Progress Report for the quarter ending
in September.

 

(S)H. E. Hungerford, Jr., Experiment 5 at the
Bulk Shielding Facility — the Shadow Shield,
ORNL CF-51-8-252 (Aug. 20, 1951).

(6)J. .. Meem and E. B. Johnson, A Nuclear
Plate Camera for Fast Neutron Spectroscopy at the
Bulk Shielding Facility, ORNL-1046 (in declas-
sification).

 
 

FOR PERIOD ENDING SEPTEMBER 10, 1951

 

 

 

 

 

 

| : DWG 12468
5 \ . SEGRET
10° — | —
- | ¥ COMPTON DATA 0’ ]
o © PAIR DATA 0 o
2T8S .
— ¥ \ X COMPTON DATA 30 1
g & ‘ o
f : fi\@ + PAIR DATA 30
oL\ 238
0% — : -
. - x\! \ 5 —
s -\ 233 ~
~. I P ]
s b X 645
v [ % | o
5 \ 1
':,-, [ f 1 /A T
~
x \ . ~
o BT
2 3 oA
07 b \ % ]
O o i T
JiH] - L —
2 - I B ]
o L— / i
e [ o —
s 1: -
= . \r ~
e A
- - IT L
5 qu
o \t r
2107 |— \ =
: L N <> -
2 N ¢
L + \T -]
| “I\]’J —
1 ]
- 4N
10 — | f POINTS WITH O COUNT RATE ]
+
| P T 7 v
| | | | l l
0 2 4 6 8 10 12 14

GAMMA -RAY ENERGY (Ey, Mev)

Fig. 5.1 - Preliminary Gamma -Ray Spectrum at 130 cm from the Water-Reflected
BReactor. ‘

85

 
 

ANP PROJECT QUARTERLY PROGRESS REPORT .

SECRET
DWG. 12253 R~

As,TrmRe

55 t— A

 

 

 

  
 

 

 

1IZONTAL DISTANGE FROM REAR OF SHADOW SHIELD (cm)

 

HOR

 

 

CURVE 1—-TRAVERSE S5cm FROM SHIELD

CURVE 2-TRAVERSE |5em FROM SHIELD

CURVE 3-TRAVERSE 35cm FROM SHIELD

CURVE 4-TRAVERSE €0cin FROM SHIELD

CURVE 4'-CURVE 4 PLOTTED TO SAME SCALE
BASE AS OTHER CURVES

 

 

 

 

 

 

 

 

 

 

 

q SGALE TO RIGHT OF EAGH GURVE GIVES
2 ¢ INTENSITY IN r/hr/wat!
I
o
g 50cm LINE A-LINE CONNEGTING POINTS OF MINIMUM
17%in. S [ S s INTENSITY
% L L5 em LINE B- LINE GONNECTING POINTS OF MAXIMUM
S \ 5 cm . INTENSITY
o SCALE DIACRAM OF SUADOW= | | INE G- SHADOW LINE OF SHIELD FROM
?rj A VERTICAL T;"IAVERSES WERE CENTER OF REAGTOR
) MADE AND THE RELATIGN OF
LINES A,8, AND G TO REAGTOR
G A /B

 

 

 

 

A I R

0 20 40 60 30 100 120 14C 160 180 200 220 240 260
VERTICAL DISTANCE FROM GCENTERLINE (cm)

 

Fig. 5.2 - Bulk Shielding Facility. A pictographic view of gamma radiation
intensities existing along vertical traverses measured behind the shadow shield.

86

 
 

FOR PERIOD ENDING SEPTEMBER 10,

1951

6.  LID TANK
E. P. Blizard J. D. Flynn
C. E. Clifford M. Marney
T. V. Blosser R. Burnett
T. Hubbard
Physics Division
The unit shield of iead and borated been reached in the unit shield,:and

H,O kas been optimized for a 3i{t
spherical reactor using double the
boron concentration, now 1.3% boron: hy
weight, of the shield presented
ANP-53, (1) Preliminary Lalculatlgns
now indicate that this results in: an
ideal shield of 50.6 tons, represent-
ing a weight saving of about 3 tons
over the corresponding shield weight
of 53.6 tons, using the data reported
in ANP-53, Appendix B, corrected to a
J-ft reactor. This new low weight
value for a unit shield differs from
that in the preceding section because
of the difference in the core size as
well as in the boron concentrat1on of
the water,

The saving in weight is primarily
the result of the addition of boron
which, by further reducing the number
of high-energy capture gamma rays pro—
duced in the shield, allows the lead
to be placed at smaller radii for the
same attenuation., It is doubtful that
the optimum concentration of boron has

 

‘Vpepors of the ANP Shleldtng Board, NEPA-
ORNL, ANP-53 (Oct. 16, 1950).

therefore work with higher concen-
trations will be continued using
potassium metaborate. With this salt
concentrations of up to 10% boron by
weight can be investigated,

The compositions of the present
shield, the shield of ANP-~53, and the
1dealized shield for halt dose are
presented in Fig. 6.1. The shield had
a nearly constant B2l and therefore
should be very close to. the minimum
weright for this concentration of
boron. (%’ Approximately ten con-
figurations were measured before
achieving the constant R%1 It is
of 1nterest that an additional 6 fions
in shield weight (i.e., 56.5 tons) will
further attenuate both the gamma rays
and neutrons by a factor of 2 (gémma
rays from 1.10 to 0.55 x 10°° r/hr and
neutrons from 3.6 to 1.8 X 10°* mrep).

 

(2)g, P Blizard, ‘‘A Method for Experimental
Shield Optxmlzatlon,’ Aircraft Nuclear Propulsion
Project (QQuarterly Progress Report for Period
fi;gé?g August 31, 1950, ORNL-858, p. 17 (Dec. 4,

(B)Appendlx B of ANP-53,:op. cit.
0
oo

 

{a)

LEAD-0.6 % BORATED Hy0 SHIELD
{FOOTNQTE 3}

3—-ft REACTOR WITH 6.6-¢tm
STRUCTURAL Fe

LID-TANK GAMMA DOSE=1.ix10"

5r/hr

NEUTRON DOSE =3.6x10"* mrep
SHIELD WEIGHT = 53.6 TONS

{b)

LEAD-4.3% BORATED Ho0—-PURE H,0
OPTIMIZED SHIELD

3—f: REACTOR WITH 6.6-cm
STRUCTURAL Fe

LiD-TANK GAMMA DOSE = 1.1x 102 r/hr

NEUTRON DOSE = 3.6 x 10" mrep
SHIELD WEIGHT = 50.6 TONS

{c)
LEAD-1.3% BORATED H»0-PURE H,0
OPTIMIZED SHIELD
{HALF STANDARD DOSE}
3~f1 REACTOR WITH &.6-cm
TRUCTURAL Fa

L1D- TANX GAMMA DOSE = 0.55x 1072 t/hr

MEUTRON DOSE = 1.8 x10°* mrep
SHIELD WEIGHT = 56.5 TONS

2.54-cm Pb SLABS {8}

 

SECRET
DWG. 12909

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

¢ 1.00-cm Pb SLAB B, O3 SAT. SOLN. -
REACTOR —2.22-cm Fe SLABS{3) f /
(457—cm RADIUS) N\ % 3 a4 & e {0.6 % BORON BY WEIGHT)
0 ® R NN N N T R AN S e
OOO O OOOOO Oo Q \\k \“ : \Q s -
e Oy 8, O’ ce A R N BN ARy N t. . '
OO OO O /"\.‘.' K%\ - AL - : . Wt
'OO' O O N R RN RN NN . ot
088 05880 NN AN N N - T
O.OO ol NN oy b - : - ! ' .
: | |
o b 1
o M@ g o Mmoo 0 QN ~ ~
RB B T B £ a g & § & @
SCALE IN GENTIMETERS
0.3-cm Fe Fe TANK WALLS {(2)~— Fe TANK WALLS {2)
TANK WALL 3.8-cm Pb om Fe TANK Wa
REAGTOR 2.2-cm Fe SLABS / 7 SLABS i5) Kz 8904 SOLN. O-sem :URE .0 -
(457 cm RADIUS) /g 3 3 / 5 {1.3 % BORON) J ‘2
AT Q' D e o e
OOO . O Ll T L -l L S . Y ]
°O O oy Se0 e e § SR LN L
OO Oo" o 00 e IIRANT Y. N = '
0?0 @00 0s AV RN SN
o) o-o 895300 ¢ O AN < I AR N L
| . P | |
© e~ N W M o o ~
g 88 3 - e & 8 @
SCALE IN CENMTIMETERS
_ 3.8-cm Pb SLABS(4)
0.3-cm Fe 2.2-cm Fe SLABS (3) j./ Fe TANK WALLS {2) 0.3-cm Fa TANK
TANK WALL / @ Ko B. 0. SOLN WALL
REAGTOR i 2 54-cm Pb Fe TANK , 259294 .
(45.7-cm RADIUS) Wi /G)/\ SLABS WALLS ¢ {1.3% BORON BY WEIGHT) .~ PURE H,0
.0. o0 ¥ 7 N-"_".:'- E-. . . I ,_./
38 020 e Pe NN Y N NN
20 ///‘ ] \\‘-..\‘.- E . - > . ST
O Op oo° O OMANININ: R RN CO TN ; L
00 o O oo j %\ ‘ : s S ‘.' :
s e 2 Ry i RN Ll
O O Q OOOOO.O fi\%\ ‘.::-: i N ol
J o i o
o NN NG~ e o N 2 S
¢ 858 8 2w & 2 & 2 @
SCALE iN CENTIMETERS
Fig. 6.1 - Comparison of Shields for 3-ft Reactor.

LUO4AY SSTUO0Ud ATHALMVAO LDALOUd dNV

 
 

FOR PERTIOD ENDING SEPTEMBER 10,

1951

7. DUCT TEST

E. P. Blizard
C. E. Clifford

M. K. Hullaings
M. Marney

Physics Division

A simplified design for the patch
required to prevent excess neutron
leakage from ducts perforating a
divided (water) shield can be made on
the basis of the duct test measurements
completed to date. The weight of a
patch needed for a single 6-in.
aluminum- filled duct surrounded by a
1-in. air-filled annulus was found to
be approximately 1000 lb exclusive of
duct or container. '

The calculation for this patch
weight was based on the data and con-
figuration presented in Figs., 7.1 and
7.2. The duct 1s straight and extends
beyond the shielda distance sufficient
to attenuate the neutrons to the same
level as 105 cm of water, which is
given in ANP-53¢!) as the thickness of
the divided shield on the rear of the
reactor. The extension is surrounded
by a cone of water which attenuates
nentrons scattered radially to the
appropriate level. This level is the
intensity at the edge of the shield

multiplied by the ratio of the area of
the shield covered by the cone to the
area of the cone.

Some assumption had to be made with
regard to the attenuation radially
from the duct since experimental
limitations prevented measurement of
the thickness required for such a
patch, The attenuation of water for
the Lid Tank source was used as an

 

(l}Be ort of the ANP Shzeldmg Board, NEPA- OINL
ANP-53 (Oct 16, 1950).

 

upper limit for extending these measure-
ments. The water curve is normalized
at Z = 0 in Fig. 7.1 to the curve of
the attenuation along the duct axis.
The ratioc of the heights of these two

SECRET

 

vo? DWE. 12910
[ [ ] ! | | i | | | |

5 o ATTENUATION 8Y BENT OucT o]
A ATTENUATION BY STRAIGHT DUGT
# NORMALIZED ATTENUATION IN Hy0

21 FOR 28-in. FISSION BGURCE ]
8 & NORMALIZED TRAVERSE, AT RIGHT

15% f-- ANGLE TO DUCT AXIS ]

51— ]

_~EDBE OF DLCT
2 - ]
10 20 3D cm FROM DUGCT AXIS
ol b ST -]

   
    

RELAXATION LEKGTH = 12.1 cm -

  

RELATIVE THERMAL NEUTRON FLUX ON DUCT tOcm FROM END OF DUCT

 

 

103 |— -
5 |~ -
er (3505100 -

10% [— ]
5 —q
2 _ -

10 L4t 1 ol

 

0 20 40 &0 BO GO 20 140 160 118G 200 220 240

TOTAL LENGTH OF DUCT AXiS {cmi

Fig. 7.1 - Neutron Attenuation in
Water Around Duct. 6-in., aluminum-
filled duct with 1-in. air-filled
annulus. '

89
 

ANP PROJECT QUARTERLY PROGRESS REPORT

curves at the edge of the shield
(105 cm), corrected by the ratio of
the areas, 1s taken to be the radial

attenuation required of the patch.

The traverse measurements close to
the endsof the ducts are approximately

a7 5 CM =~

the same shape for all configurations
measured, and this shape was fitted to
the Lid Tank water curve in a region
where the slopes were similar over the
30 cm measured. The measured traverses
were then extrapolated to the desired
attenuatiaon,

SECRET
OWG. 12911

WEIGHT OF PATCH = 960 Ib

 

CENTERLINE OF DUCT
PASSES THROUGH CENTER
OF SPHERICAL REACTOR;
PATCH CONSTRUCTED
SYMMETRICALLY ABOUT

CEN_IERLINE Y I

 

SCHEDULE 20 STEEL

1-in. ANNULUS {AIR-FILLED)
SCHEDULE 40 STEEL

 

   

{Ocm

 

A¢(SURFACE AREA OF
TRUNCATED CONICAL PATCH)=
29,450 cm?

A5 (AREA GOVERED BY PATGH ON
SPHERE) = 16,570 cm?

SPHERICAL SHELL OF SHIELDING
SURROUNDING REAGTOR

Fig. 7.2 -

90

Water Patch Around Duct.

 
 

'FOR PERIOD ENDING SEPTEMBER 19,

Analysis of the data from shield

1951
8. SHIELDING CALCULATIONS
being made. The data(l) for HzO,

testing facilities has permitted the
semiempirical establishment of im-
portant radiation attenuation relations.
An expression was derived for attenu-
ation of neutrons from a point source
in water by analysis of data taken at
the Bulk Shielding Reactor. Similarly,
analysis of Lid Tank data has yielded
an approximate formula for the total
centeriine gamma dose which has been
applied to subsequent experiments,

In the continued investigation of
radiation hazards in the ARE, further
calculations were made on the acti-
vation of the Be(O moderator, the
detection of leaks in the fuel elements
by means: of radioactive tracers, and
the activation of nitrogen and helium
in the reactor pit.

Calculations show the divided-
shield weight to be higher than origi-
nally estimated, owing chiefly to
symmetry and angular distribution
considerations, Some new effects have
been computed relative to the shadow-
shield theory which may prove that
this device 1s much less useful than
previously believed.

A shield using ammonia in place of
water has been considered which would
provide a decrease in shield weight of
about 7% at takeoff and 13%at landing.

ANALYSIS OF BULK SHIELDING REACTDR
NEUTRON DATA

S. Podgor, Physics Division

An analysis of fast-neutron dosimeter
data of the Bulk Shielding Reactor 1is

 

V(May 7 1951);

Pb-H,0, and Fe-H,0 were treated by a
method ‘similar to the one used by
Welton®?) in the analysis of Lid Tank
dosimeter data. A point source of
neutrons was there calculated to be
attenuated in water as follows:

A » .
P(S) = == &% o-BS e'zJa?S‘
4nS*?

where P(S) 1s the uncollided dose at a
distance S from the source, A is a
normalizing constant to be adjusted,
and B8 is the oxygen “removal" cross-
section in reciprocal centimeters which
is assumed constant and to be adjusted.
a comes from the flSSlon spectrum
representation

N(E) g e 0-75E

where N{E) is the number of neutrons
of energy E Mev per neutron emitted
per Mev, and @ = 0.75. ¥ comes from
the hydrogen cross - section repre-
sentation

0.73
o(E) = 2738 -
E + 1.66

where v ¥ 0.735. These two repre-

sentations are taken from a paper by
Blizard and Welton.(3)

 

(1)3.:6. Cochran et al. to J. L. Mecm, Fast-
Neuvtron Dosimeter Measurements, ORNL CF-51-5-61
OBNL CF-51-5-73 {(May 11, 195]1).

Report Of the ANP Shteldtng Board, NEPA-ORNL,
ANP 53, p. 120 (Oct. 16, 1950). '

(3)E P. Blizard and T. A. Welton, “ The Shle]d—
ing of Mobile Remctors. I,)” Reactor Science and
Technology (to be publlshed)

91
 

ANP PROJECT QUARTERLY PROGRESS REPORT

The source was taken to be the
leakage from the side of the reactor
facing the detector. To simplify the
analytical work the source is here
assumed to be circular instead of
rectangular. With these assumptions
1t was possible to fit the shape of
the experimental attenuation curves by
using removal cross-sections for lead
and 1ron of 3.4 and 2.0 barns, re-
spectively, which values are the same
as those obtained in the analysis(*%)
of Lid Tank thermal data. The oxygen
cross-section neecded was 0.8 barn.
The neutron build-up was assumed to be
that for pure water, as given 1n pre-
liminary form in an earlier quarterly
report. (5)

INTERPRETATIGN OF LID TARK
GAMMA -RAY DATA

A. Simon, Physics Division

An approximate formula for the
total centerline gamma dose 1in the
Lid Tank has been applied to several
experiments. The variation in gamma
dose with the number of slabs in a
uniform lead-water shield was re-
produced satisfactorily for several
positions of the detector. In addition,
the variation of gamma dose with
degree of boration of a uniform lead-
water shield was also reproduced.
Differences between experimental and
theoretical normalization factors can
be explained 1n terms of neglect of
gamma build-up in water.

 

(4)g, Podgor, Analysis of Lid Tank Neutron
Data for Lead and Iron, OBNL-895 (Jan 23, 1951).

(S)g, Podgor and T. A. Welton, ‘““Neutron Build-
up in Water,” Aircraft Nuclear Propulsion Project
Quarterly Progress Report for Period Ending March
10, 1951, ANP-60, p. 169 (June 19, 1951).

92

It is planned to apply the method
to the computation of an iron-water
shield and to observations 1in the
BSF. A complete report will be pre-

pared in the near future.

SHIELDING CALCULATIONS FOR THE ARE

W. K. Ergen, Physics Division

Consideration of the radiation
hazards associated with the ARE has
led to the analytical evaluation of
several such possible hazards. The
instantaneous release of the nitrogen
and helium in the reactor pit could
possibly, but not probably, expose
personnel at nearby sites to a dose
equal to 20 min exposure to the
accepted "maximum permissible concen-
tration" of C'? in air.(® A means of
detecting leaks in the ARE fuel elements
by the use of radiocactive tracers has
been proposed. A re-evaluation of the
activities of the impurities in BeO,
based upon a later analysis of the
impurities and a revised Co(n,7y) cross-
section, still indicates no handling
problem after shutdown.

Activation of Nitrogen and Helium
in the ARE Reactor Pit.¢’) It is pro-
posed to fill the ARE resctor pit with
nitrogen during operation as a pre-
cautionary measure in the event of a
sodium leak. Assuming certailn average
operating conditions, about 1.3 curies
of C" would be formed by the N*&,p)Cct?
reaction. This activity, released 1in-
stantanecusly through a 100-ft stack

 

(G)K, Z. Morgan, Maximum Permissible Concen-
centrations of Radioisotopes in Air, Water, and
the Human Body, Subcommittee on Internal Dose of
the National Committee on Radiation Protection,
report to be published.

(T)Abstracted from the report by W. K. Ergen,
Activation of Nitrogen and Heliunm
Reactor Pit,
(July 13,

in the ARE

ORNL, Y-12 Site, report Y-F20-16

1951).

 
 

FOR

under extremely unfavorable and very
rare meterclogical conditions, would
give personnel at X-10 and the HRE site
a dose equal to a 20-min exposure to
the sccepted maximum permissible
concentration of C'*
parts

in air, Twenty
per million (by volume) of argon
in the nitrogen would give the same
biological dose as the C'*. The other
probable impurities of nitrogen as well
as the N'%(n,¥)N'® reaction, eld
negligible

If the pit were filled with pure
well helium, we would get, by the
Hed (n, p)H® reaction, 5.6 % 1072 times
the biological dose computed above for
nitrogen filling.

activities.

Activation of Impurities in Beo.(®)
Since Memo Y-F20-11 was issued on this
subject, a more accurate chemical
analysis of the BeO for the ARE showed
the presence of greater amounts of
impurities; also, the cobalt activation
cross-section has been found to be 34
barns, and not 22 barns, as assumed 1in
¥Y-#20-11 on the basis of then avail-
able data. Nevertheless, the acti-
vation of the iwmpurities in BeO wall
not cause handling problems after the
shutdown of the ARE, under the con-
ditions specified.

Betection of Leaks in the Fuel
Elemenls by Means of Radigactive
Tracers. (%) A gamma-emitting sub-
stance added to the fuel elements 1in
the ARE could be used as a means of
detecting leaks of fuel into the coolant
before the ARE is operated. In
to just get a definite detection,

2ven
order

 

(B}Absti«cLed from the report by W. K. Frgen,
Activation of Impurities 1n Be), ORNL, Y-12 S:ite,
report Y-F20-14 (May 21, 1981).

\g}fiastractfid from the report by W. K. Frp
Detection of Leaks in the Fuel Elements by
of Rudicactive Tracers, OBRNL, Y-12 Siie,
Y-i20-15 (July 10, ]951}.

DIl

Heans
report

 

PERIOD ENDING SEPTEMBER 10, 1951

if any one fuel tube should break 1in a
system of 500 fuel tubes, a totval
gamma activity of about 10 mc
quired.

is re-
How much additional radiation
precaution would have to be assumed in
using this method 1s not cervain, but
the possibility of leak detection
without bleedingoff coclant makes this
method seem attractive. Investigation
1s also being made of a2 photofluori-
metric method of leak detection.

NDA DIVIDED-SHIELD STUDIES

Nuclear Development Associates, Inc.

An
assumptions 1involved in calculations
of the split shield. A sazmplified

examination is being made of the

geometry has been assumed thus far,
with the reactor a point source sur-
rounded by a spherical shield. The

represented by
a spherical cap of 20° half angle.
The following are tentative conclusions

gamma shadow shield 1is

emerging out of the calculations:

ANP-53
scatter-
scatter-
ing with penetration normal
the crew compartment
conservative. Jt appears likely
that the scattered neutron dose 1s
one-half or less of the figure
there obtained,.

1. The assumptions made 1in
in regard to neutron air
ing (isotropic, elastic
to
sides) are

2. Gamma rays can be scattered 1in
the plastic at the crew com-
partment 1n such & manner as to
increase greatly their subse-
guent penetration through the
fead. Gamma shielding by the
plastic 1s hence sharply reduced.
It may be advisable to move the

tic back to the re

plas actor

93
94

ANP PROJECT QUARTERLY PROGRESS REPORT

shield, a change which will not
affect the weight of the ANP-53
neutron shield.

The first-scattered dose at the
crew, with any appreciable
amount of lead crew shielding,
is nearly independent of initial
source energy down to around
1 Mev. Below that it drops off
very rapidly owing to the in-
creased lead absorption. The
SFAC calculations by the National
Bureau of Standards indicate
that the radiation leaving the
reactor shield does not contain
any extremely large proportion
of its energy 1n low-energy
photons. The current practace
of assuming the source energy
flux to be concentrated at 3 Mev
thus does not lead to error.

Scattering in the lead at the
crew shield has been calculated
using build-up factors computed
by The Rand Corporation. The
resulting increase in dose 1s a
factor of 2.5 for 1.55 cm of
lead and about 4 for 2.5 cm of
lead.

It has usually been assumed that
in the zone not shielded by the
shadow cap the gamma rays emerge
radially from the reactor shield.
However, i1f they deviate from
the radial direction, the effect
of the shadow is partly nullified.
Thus, i1if the emerging rays are
uniformly distributed in angle
from the radial out to 30°
zero thereafter, the advantage
gained by the shadow shield
drops (in a typical case) from
around 16 to 2. This appears to
be one of the most significant

and

effects uncovered so far.

6. Streaming of photons around the
shadow cap so as to emerge 1in
regions assumed to be covered
has been examined. There 1s some
indication, both theoretical and
experimental, that this i1s not
so Important as the deviations
from radial discussed above,

7. In the absence of the effects
mentioned in 1tems 5 and 6, the
shadow shield is quite effective
in decreasingthe first-scattered
dose. However, it has little
effect on the second and higher
scatterings. Some preliminary
calculations 1indicate that the
second scattering can then be
quite significant,

USE OF NH, AS A SHIELDING MATERIAL (1)

E. P. Blizard and H. L. F.
Physics Division
* %

Wyld, ANP Division

Enlund,

J. H.

An analysis has been completed of
the weight savings realized by re-
placing water in a reactor shield by
anhydrous ammonia., Although the number
of hydrogen atoms per cubic centimeter
is about the same for ammonia and

water, the ammonia density 1s less
than 70% that of water. The shield
weights for a 4-in.-diameter reactor

delivering 200 megawatts were calcu-
lated for a divided shield employing
water as the hydrogenous material and
for a similar shield with ammonia. The
composite (ammonia) shield is about 7%

 

(IO)Abstracted from the report b
H. L. F. Enlund, and J. H.
Anmonia as a Shielding Materzal
(in preparation)

cm loan from Gibbs and Cox, Inc.

On loan frem Reaction Motors,

P. Blizard,
Yd The Use of
ORNL CF report

Inc.

 
 

FOR

lighter at takeof{ and, owing te the
evaporation cof the boil-off ammonia,
is some 13% ligheer on landing. The
welght savings from use 6f ammonia
thus appear to be appreciable, but
whether this advantage would be offset
by the obvious disadvantage of carrying
a large volume of a hard-to-handle ma-
terial is still a question.

Mechanical Requirements of an
Ammonia Shield. Anhydrous ammonia has
a vapor pressure of 1 atm at -28°F and
must therefore be refrigerated to take
care of inleakage of heat from the
atmosphere as well as that due to
neutron and gamma radiation from the
reactor. The mechanical requirements
for maintaining an ammonia shield are
refrigeration equipment, a pressure
shell to maintain the liquid at 120°F
in case the refrigeration fails, and
insulation to reduce heat influx. Since
the heat flux into the shield due to
nuclear radiations from the reactor
is quite large, a higher temperature
inner shield layer of water, con-
ventionally cooled, and lead 1is pro-
vided. The lead in this inner shield
is inserted as a "heat trap" to lower
the heat flux into the ammonia.
Approximately 1 ft of water 1s pro-
vided to remove neutron heat as well
as to reduce secondary gamma production
in the lead heat trap. The heat pro-
duced in the ammonia is calculated from
the gamma flux 1t will receive. A
section of this shield assembly is
shown 1in Fig. 8.1.

Calculational Procedure. The total
heat energy (H, inBritish thermal
units) entering the ammonia from the
gamma heating of the iron and lead
heat trap during a 25-hr flight may be
expressed in terms of the thickness
(T, in centimeters) of the lead {(assum-

ing a 3-in. ircon pressure shell):

PERIOD ENDING SEPTEMBER 10, 1951
SECRET
DWG. 12812

  
    
   
  
  
   
  

 

A

-3-in. STEEL SHELL

   
  
  

12 in,

20+n. LEAD SHELL

 

l/8“'!1'\. STEEL '.-.
< 1-in. INSULATION §

 

i, _:
CORE /a-in. STEEL
—1.375 in

43 375-n. R

79.0-n.R.__
“H,0 '

 

Vgrin, STEELY  J7
tin. INSULATION | ",

  

NHs

 

Fig. 8.1 -
HZ{}-Pb-NHE .

TypicaiCump@site Shield,

H = 2,36 x 107 ¢-T7/0.95

Since 589 Btu may be removed by evap-
orating 1 1b of ammonia, the required

 
 

ANP PROJECT QUARTERLY PROGRESS REPORT

"boil-off" of ammonia (B, in pounds)
to dissipate the above heat is:

B = 4,01 x 10% ¢~T7/0.95

The lead heat trap is assumed to
replace some of the lead (for gamma
shielding) on the rear crew shield
disk. But since increasing the thick-
ness of the heat trap increases the
total weight and also decreases the
amount of NH, required for boil-off
refrigeration, the shield must be
optimized with respect to weight for
these two design variables., The
thickness of ammonia neutron shield*
was first calculated to be 35.6 1in.,
and the expression for the equivalent
in the two lead

When this in-
formation is combined, the total weight
of the ammonia shield, which for the
purposes of this calculation includes
the lead heat trap, the rear crew
shield, and the boil-off ammonia, may
be expressed in terms of the thickness
of the lead heat trap. The weight of
the shield is then minimized by taking
the first derivative of the variable
weights with respect to the thickness
of the lead trap. The thickness of
this trap becomes 1.875 in., and the
thickness of the rear crew shield

3.36 1in.

gamma attenuation
shields is derived,

Comparison of Shield Weighis Using
NE, and H,0. A comparison of shield

 

*In a divided shield the neutron shielding is
most effectively accomplished by hydrogenous
material at the reactor.

96

TABLE 8.1

Comparison of Shield Weights Using
NH, and H20

 

 

 

 

3
H,0 (1b) NH; (1b)
Water 79,800 6, 520
Ammonia 42,300
Lead heat trap 15,450
Crew disk 9, 400 8,730
Stee]l shells 6,200 8, 300
Insulation 600
Total (fixed) weight 95, 400 82,900
Ammonia boil-off radiation 5, 570
Thermal conduction 380
Savings, takeoff condition 6,550 1b
Landing condition 12,500 1b

 

 

weights as calculated above, with
water in one case and anhydrous ammonia
and associated equipment in the other,
is given in Table 8.1. It will be
noted that the composite shield 1is
6550 1b lighter at takeoff and, owing
to the evaporation of the boil-off
ammonia, it will be 12,500 1b lighter
on landing.

The appeal of the boil-off scheme
for refrigeration lies in its sim-
plicity and in 1its low weight for
landing, the most critical period of
the flight. 1In addition, 1t 1s possi-
ble that the vapors boiled off could
be burned in the engines for added
thrust, On the other hand, 1t 1s
entirely possible that the refrigeration
could be accomplished by mechanical
means with considerably less weight.
The Air Force at Wright Field has
agreed to investigate this possibilaty.

 
 

FOR PERIOD ENDING SEPTEMBER 10, 1951

9. NUCLEAR

Continued efforts on the 5-Mev Van
de Graaff accelerator research program
have satisfactorily succeeded in
lowering the machine background,
providing an energy calibration accu-
rate to a few kilovolts, and improving
the ease of operation below 1.5 Mev.

Data taken from a neutron source
with and without beryllium surrounding
the source are being analyzed to de-
termine the (n,2n) reaction in beryl-
lium.

~ An in-pile lithium loop operating
at 1000°F for one week yielded data on
Bremsstrahlung caused by beta particles
from the Li’(n,B8)Li® reaction. The
average intensicvy of the Bremsstrah-
lung was about 2 X 10° Mev/cc/sec with
an average energy of only 0.06 Mev.

THE 5'MEV VAN DE GRAAFF ACCELERATOR
H. B. Willard, Physics Division

Both the neutron and gamma-ray
backgrounds of the 5-Mev Van de Graaff
accelerator have been reduced by the
extensive use of only tantalum in the
vicinity of the beam. Additional
tantalum slits have been provided
during the past guarter in order to
further collimate the proton beam and
reduce these backgrounds to a satis-
factory level. In addition, the
X-ray intensity coming from the reverse
electron current up the tubes has been
found sufficiently low to permit
unshielded operation of a Nal-Thl
crystal counter in the target area.

Gamma-ray yield curves resulting
from the proton bombardment of fluorine

 

MEASUREMENTS

and beryllium have been obtained.
These data provide a large number of
energy calibration points in addition
to new 1nformation regarding levels in
the compound nuclei involved. By
proper manipulation of the machine, 1t
was found possible to cbtain data for
proton bombarding energies as low as

0.158 Mev.

In preparation for neutron gross-
section work, the Li7(p,n)Be’ yield in
the forward direction has been extended
to above 5 Mev. The response of the
Bonner type neutron counter has been
measured. Neutron yvield curves for
protons on boron and beryllium have
been obtained from threshold to above
5 Mev.

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

E. D. Klema, Physics Division
David Martin and David Ott, Oak
Ridge School of Reactor Technology

An experiment has been carried out
in which the neutrons from an {(a,n)
source are detected by means of aglong
counter. The neutron source is sur-
rounded by a sphere of beryllium, and
the counting rates with bare source
and source surrounded by the sphere
are measured. Similar measurements are
made with a carbon sphere to estimate
the neutron absorption in the bervyl-
lium sphere.

Measurements have been carried out
with polonium-boron and pelonium-~
beryllium reutren sources. Beryllium
spheres of 4 and 9 in. diameter and
carbon spheres of 4 and 5% in. diameter
have been measured.

97
 

ANP PROJECT QUARTERLY PROGRESS REPORT

The data obtained are being reduced
to the (n,2n) cross-sections for the
neutron spectra used, and the as-
sumptions made 1n the calculations are
being considered in detail in order to
set limits on the values of the average

(n,2n) cross-sections for the two
neuntron spectra.
BREMSSTRAHLUNG FROM Lis
C. D. Baumann W. W. Parkinson

R. M. Carroll 0. Sisman
Chester Ellas
Physics of Solids Institute

Lithium at 1000°F has been circu-
lated in the X-10 reactor for one week
in a 316 stainless steel loop and sub-
sequently examined. The primary purpose
of this experiment 1s to study the
Bremsstrahlung activity produced by
beta decay of lithium. Li?, created
by the reaction Li7(n,B)Li8%, decays
with a 0.88-sec half-life, giving a
12.7-Mev beta ray. This beta ray 1is
absorbed in the lithium stream and the
container walls (1/16-1in. stainless
steel ), producing the Bremsstrahlung
activity. Preliminary analysis of the
data indicates the average energy of
the Bremsstrahlung to be about (.06 Mev
with an average intensity of 2 x 10°
Mev/cc/sec. Actual operation of the
loop 1s discussed in this report in
the section on radiation damage, where
it may be noted that there was no
significant amount of corrosion,.

Radiation Detection Equipment. The
1on chambers used 1n this experiment
were cylindrical in shape (4 1in. in
diameter, 4 in. high) with one of the
bases a thin-walled aluminum "window."
The chamber was filled with air at
atmospheric pressure, and its walls
were maintained approximately 150

volts above ground. The chambers and
their amplifying and recording equip-
ment were bench-tested with Co®? of
various strengths so as to simulate
the line sources of the circulating
lithium in the external pile loop.
A lead brick with a 2-1n.,-diameter
hole was placed between the source and
the counter to act as a collimator for
radiation coming into the counter.
The in situ calibration curves were
approximately linear and of the form
S = AR - b where S 1s the source
strength per unit length of pipe, R 1is
the recorder reading (the amplified
chamber response), A i1s a constant
inversely proportional to the distance
of the source from the counter, and b
is the background correction which
varies directly as pile power level,
the distance of the
and includes
the natural background. The data have
nct yet been corrected for the ef-
ficiency of the counter for radiations
of this energy range, and the neutron
flux 1s still in the process of being
determined.

inversely as
counter from the pile,

Bremsstrahlumng Activity. The
Bremsstrahlung activity, uncorrected
for counter efficiency and neutron
flux, at the three ionization chambers
(or three different distances from the
active lattice) for various velocities
of the lithium stream is shown in Fig,
9.1. By extrapolating back to infinite
velocity we find the Bremsstrahlung
activity for zero decay to be 2 x 10°
Mev/sec per cubic centimeter of natural
lithium (at a flux of about 2 x 10'!
neutrons/cm?/sec), Figure 9.2 shows
absorption curves for the Bremsstrah-
lung using iron and copper absorbers.

The variation 1n activity with these
copper and 1ron absorbers,
between the tube containing the circu-
lating

interspersed

lithium and No. 2 counter,

 
 

FOR

UNCLASSIFIED

PSl—A~208

s DWG. 12178RI
3x10

9
10 o Chamber No. |

X Chamber No. 2
v Chamber No.3

BREMSSTRAHLUNG ACTIVITY [ Mev/ {ceXsecl |
5
wm

 

0 0.l 02 03 0.4
1/ VELOCITY (sec/ f1)

Fig. 9.1 - Bremsstrahlung from In-
Pile Lithium Loop.

gives the half-thickness of copper to
be 0.025 in. and that of the iron
0.030 1n. Using these figures, in
conjunction with the mass absorption

 

PERIOD ENDING SEPTEMBER 10, 1951

UNCLASSIFIED
PSl-A—209

Lo DWG ~— 12IT9RI

1,

0
-

o Copper Absorber, Velocity 5.533 fi/sec
x: Copper Absorber, Velocity 4.98 ft/sec
Velocity 3.38 fi/sec

v iron Absorber,

 

0 ©o002 004 Q.06 008 0.10
THICKNESS OF ABSORBER (in)

Fig._9.2 - Absorption of Bremsstrah-
lung from Li® Beta Rays.

coefficient for X-rays!’? and ignoring
the scattering cross-section, the
average energy of the Bremsstrahlung was
found to be 0.063 and 0.058 Mev,
respectively, Thus the average energy
of the Bremsstrahlung arising from the
circulation of lithium in type 316
stainless steel tubes 1is of the order
of 0.06 Mev. The radiation is identi-
fied as that from Li® since it shows
the 0.88-sec half-life by its decay
while moving from one chamber to the
next.

During the week of operation of the
loop, the velocity of the lithium
stream was maintained most of the time
at approximately 3 ft/sec in the
external loop (approximately 1% ft/sec
in the 1n-pile loop because of the
larger diameter tubing). The velocity
was changed only during the periods
when runs of activity vs. velocity

{1)W. S. Snyder and J. L. Powell, Absorption of
¥ Rays, 0H¥L~4§1, Supplement 3 (March 14, 19501,

99
 

were made and when the above absorption
curves were determined. The velocity
increased for the absorption
measurements 1n order to increase the
starting activity. The curves of
activity vs. velocity were taken
periodically during the entire week of
operation, but, because no significant
change occurred,
1s presented.

wa s

only one set of curves
Since the intercepts
and slopes of these curves showed no
significant variation with time, 1t
may be concluded that there was no
appreciable corrosion of the tube by
the circulating lithium. That there
was no significant amount of corrosion
1s further shown by the
counting the

results of
Ilithium from a 9-1n.
section of the tubes adjacent to the
active pile lattice. The total count
as obtained 1n a "100% geometry"
counter was only twice background.

TIME-OF-FLIGHT NEUTRON SPECTROMETER

G. S. Pawlacki, Oak Ridge Institute
Of Nuclear Studies

F. C. Smith,

Physics Division

A time-of-~-flight neutron spec-

trometer for operation 1n the energy

100

range up to several thousand electron

volts is being constructed. ?? Final
assembly has been delayed by diffi-
culties associated with the fabri-

i.e., brazing, of the rotor.
Preparations for the installation of
the spectrometer at the LITR will be

completed early in October.

cation,

Serious delays occurred 1n the
fabrication of the rotor because of
a brazing operation 1in assembly. The
use of copper brazing was abandoned
in favor of silver-copper eutectic
alloy. The brazing was successfully
completed i1in an 1mprovised hydrogen-
atmosphere furnace. The brazed sub-

assembly 1s being finished-ground by

an outside contractor who previously

ground the shroud ring for the rotor.
When the subassembly is returned, the
rotor will be assembled and balanced
at the laboratory. The present esti-
mate of completion date 1is the end of

October 1951.

 

(Q)G. S. Pawlicki and E, C, Smith, "Time-of-
Flight Spectrometer,”” Aircraft Nuclear Propulsion

Project Quarterly Progress Report for Pertiod
Ending June 10,
1951).

1951, ANP-65, p. 129 (Sept. 13,

 
 
 

 

 
 
 

 
 
 
 
 

16.

CORROSION RESEARCH

‘W. D, Manly, Metallurgy Division
H. W. Savage, ANP Division

"W. BR. Grimes and F. Kertesz, Materials Chemistry Division

The corrosion program has been ex-
panded to include corrosion by hydrox-
1de and fluoride coolants as well as
the corrosion by liguid metals and
fluoride fuels already under investi-
gation for some time. Although the
majority of the corrosion data are
from static tests, a moderate number
of data on liquid metals are available
for thermal-convection loops. Forced-
convection loops are being developed,
but this development has not yielded
many significant corrosion data as yet.

In general, corrosion of incornel
and stainless steel by pretreated and
carefully handled fuel mixtures in the
absence of radiation is not sufficiently
severe to cause failure in operation
for several hundred hours at 800°C.
Limited experience at higher tempera-
tures suggests that corroesion at 1000°C
will not be excessive. Platinum,
contemplated as thermocouple shield
material, 1s not appreciably attacked
by the fuel mixture contained in:a
platinum or an inconel capsule. Welded
or brazed joints of platinum to inconel
are attacked by the molten eutectic.

Experience with fluoride coolants
indicates that corrosion, at least 1in
the absence of radiation, will not be
a serious problem with either inconel
or the stainless steels.

Corrosion problems encountered with
hydroxides and with hydroxide-bearing
materials, however, are a great deal
more serious. Sodium hydroxide, eather
commercial or specially pure material,

 

1s extremely corrosive to stainless
steel or inconel. Copper, monel, and
nickel are much more resistant although
mass transfer occurs with these ma-
terials. The corrosion by commercial
potassium hydroxide which has been
freed from excess water 1s noticeably
less severe than that by sodium hydrox-
ide but is sufficiently bad to preclude
use of inconel and stainless steel at
the preéent. Some tests with barium
and strontium hydroxides conducted with
dehydrated commercial material have
that these materials are less
corrosive than the dalkalies. These
results are preliminary, and the lack
of reproducibility in multiplicate
tests 1s a serious problem, :

indicated

While the results obtained to date
with hydroxide-bearing systems are
not encouraging, it should be noted
that (1) only in the case of sodium
hydroxide has pure hydroxide been
available for use; (2) inert, rather
than reducing, atmospheres have been
used: and (3) the metal walls have
been in the as-received condition,
While it is not obvious that changes
in technigue will result in improved
corrosion resistance, 1t 1s suggested
that many variables remain to be tested
before 1t can be stated that corrosion
of stainless steel and inconel by the
alkalies is necessarily so severe as
to preclude their use.

An extensive number of corrosion
tests, both static and dynamiec, have
conducted with sodium. It 1s now
apparent as a result of this work that

been

103
 

ANP PROJECT QUARTERLY PROGRESS REPORT

sodium may be contained with little if
any corrosion. Recent static corrosion
tests with sodium at 1000°C in 316
stainless steel for periods up to 1000
hr revealed no surface reaction and
did not indicate any significant weight
change. Operation of thermal con-
vection loops of inconel and several
of the stainless steels with sodium
for 1000 hr at 815°C produced very
little corrosion,. Indeed, 1t now
appears that little more can be learned
of sodium corrosion by operating
additional loops of inconel or stain-
less steel with sodium. It remains
to be demonstrated, however, that a
circulatory failure will not occur with
saodium in a forced-convection loop 1in
which flow rates are higher and temper-
ature changes greater.

Metallographic examination of several
lithium- and lead-containing thermal-
convection loops, operated during the
past two quarters, 1indicates severe
corrosion of the hot leg and somewhat
less severe corrosion of the cold leg.
Only one of nine such loops completed
the scheduled 1000-hr test; all but
one of the other eight failed because
of an obstruction in the cold leg of
the loop. This obstruction is a build-
up of corrosion products,

STATIC CORROSION BY FLUORIDE FUELS

evaluation of
metals

The experimental
static corrosion of structural
by fluoride fuel mixtures has all
been conducted by the sealed-capsule
technique described in previous re-
ports.{173) This technique has been
demonstrated by experience to be rapid

 

(l)fl. E. Moore, G. J. Nessle, J. P. Blakely,
and C. J. Barten, ‘“Low-Melting Fluoride Systems)”
Airecraft Nuclear Propulsion Project Quarterly
Progress Report for Period Ending December 10,
1950, OBNL-919, p. 242 (Feb. 26, 1951).

104

and completely satisfactory for
handling the fluorides, since mixtures
thus treated have consistently demon-
strated low corrosion rates on inconel
and a wide variety of stainless steels.
The penetration of inconel by a typical
fuel mixture in 100 hr at 800 and

1000°C does not excezed 4 mils. The
penetration of the stainless steels
(304, 310, 316, 330, 347, 446) in 100

hr at 800 and 1000°C did not exceed
2 mils and was generally less than 1
mil., The corrosion of platinum, in-
troduced in thermocouples, has been
shown to be negligible in the fluorade
fuel at these temperatures,

The Pretreatment Process (G. J.
Nessle, H. S. Powers, and C, J. Barton,
Materials Chemistry Division). 1In a
previous report(3?’? the improvement in
corrosion which resulted {from treat-
ment of the fluoride fuel with pieces
of stainless steel and incomel at B850
to 900°C prior to filling of the cor-
rosion capsules has been described.
While no mechanism for this improve-
ment has been demonstrated, it appears
likely that this treatment removes
traces of hydrogen fluoride which
result from hydrolysis of some of the
UF, in the fuel. A photograph of an
inconel sample which was used to
“"scavenge' the active 1mpurities 1in a
fluoride bath is shown in Fig. 10.1.
The fluoride thus freed from theactive
impurities, and containing some 1iron,
nickel, and chromium 1n solution, was
used for making the static fluoride
corrosion tests,

 

(Dp, . Hagelston, “Static Corrosion by
Fluoride Melts,” Aircraft Nuclear Propulsion
Project Quarterly Progress Report for Period
%;gi?g March 10, 1951, ANP-60, p. 212 (June 19,

(S)F. Kertesz, F. A. Knox, H. J. Buttram,
S. D. Fulkerson, and J. A. Griffin, “Static Cor-
rosion by Fluoride Melts on Metal Containers,”
Aircraft Nuclear Propulsion Project Quaerterly
Progress Report for Period Ending June 10, 1951,
ANP-65, p. 148 (Sept. 13, 1951).

 
 

FOR PERIOD ENDING SEPTEMBER 10,

  

 

Fig. 16.1 -

1951

UNCLASSIFIED

{ v 4686

Inconel Specimen from Pretreating Pot in Which Fluoride Bath

Mixture was "Deactivated" at 950°C for 100 hr. The smallest divisions offthe

scale in the photograph are 1/100 in.

A comparison test has shown that
corrosion by untreated fluorides which
had been vigorously outgassed and then
sealed with a helium atmosphere enclosed
was not appreciably different from
that obtained when the capsules were
sealed with an air atmosphere. :

Pretreatments of the metal surface
has not been successful in minimizing
corrosion., Pickling, oxidation, and
decarburization of the metal surface
have seemed to result in increased
corrosion., The oxide film on the
surface as received seems to have
reasonably good corrosion resistance
properties, ‘ ‘

It has been found that some of the
fuel samples, whether pretreated with
structural metals or not, do contain
appreciable quantities of uranium
oxide. Sedimentation experiments in

sealed capsules have shown an in-
crease in uranium content in the lower
section of the capsule without a cor-
responding change in the fluoride

- In

content of the wvarious sections.

 

Note very severe intergranular attack.

addition, filtration of the molten
fuel through sintered stainless steel
has removed significant amounts of
black uranium oxide. However, uranium
oxide did not increase the corrosion
rate. ' '

Pretreatment of the ligquid fuel
with structural metals serves to
introduce appreciable quantities of
iron, chromium, and nickel 1nto the
fuel melt., Chemical analysis of such
fuel mixtures indicates the presence
of 600 to 1200 ppm of irom, 200 to 1000
ppm of chromium, and 30 to 200 ppm of
nickel after pretreatment although
the compounds actually involved have
not been identified. The variation in
concentration observed indicates that
not all these impurities are dissolved.
This is borne out by the fact that some
sedimentation of the iron seems to
occur during the I00-hr corrosion
tests, It also seems evident that the
nickel content of the pretreated fuels
is reduced during the subsequent cor-
rosion testing while the chromium

105
 

ANP PROJECT QUARTERLY PROGGRESS REPORT

content 1s not, Further studies are
required, but 1t appears that con-
siderable chromium is dissolved in the
fuel and that the nickel present after
pretreatment may be electrolytically
deposited on the corrosion specimens
at the expense of chromium from the
test metal, The amounts involved are
small enough to make demonstration of
this phenomenon difficult. It seems
likely, however, that pure chromium
may have some advantages as the pre-
treatment metal,

Corrosion of Structural Metals
(H. J. Buttram, N. V. Smith, C. R.
Croft, and J. A, Griffin, Materials
Chemistry Division; A. D. Brasunas,
L. A. Abrams, and E. E. Hoffman,
Metallurgy Division). It appears from
studies to date that corrosion of
inconel and stainless steel by care-
fully handled fluoride fuels is not a
serious factor. Furthermore, improve-
ment of the pretreatment process and
of handling techniques for the pre-
treated material should result in
further improvements. Three different
fuel mixtures, the binary system
NaF-UF, (75-25 mole %), the ternary
system NaF-KF-UF, (46.5-26-27.5 mole
%), and the ternary system NaF-Bel, -UF,
(76-12-12 mole %), have each been
tested at 800 and 1000°C with no
significant difference in corrosion.
In general, the corrosion of inconel
specimens in 100 hr with each of these
fuels was of the order of 2 to 4 mils;
that of several stainless steels was
around 1 to 2 mils.

By NaF-UF,. Several static cor-
rosion tests have been completed using
the NaF-UF, fluoride mixture with a
composition 75 mole % NaF and 25 mole
% UF,. Tnconel and 347 stainless steel
were exposed for 100 hr at 1000°C
(1830°F) to the fluoride cutectic

106

which had been pretreated with these
metals for 150 hr at 150°C,

The attack by treated fluoride on
inconel, in the form of subsurface
was noted to a depth of 0.004
in. as shown in Fig. 10.2. The 347
stainless steel was less severely
attacked as may be seen in Fig. 10.3,
where the subsurface attack was limited
to 0.002 in., The nature of the cor-
rosion product has not been determined
An analysis of the fluoride bath after
testing gave the following results:

volds,

 

 

 

 

) AMOUflT (%)
IRON NICKEL CHROMIUM
Inconel 0.25 Not detected 0.03
347 stainless 0.2 Not detected 0.05
steel (%)

 

 

 

By NaF-KF-UFé. The results of
corrosion testing of stainless steel
and inconel with the NaF-KF-UF, fluoride
fuel have been very satisfactory. The
pretreatment technique has reduced the
corrosion of these metals to mnearly
negligible proportions with very little
intergranular penetration of the
specimens, The photomicrograph re-
produced as Fig. 10.4 shows a heat-
treated incoenel specimen and Fig. 10.5
shows the type of corrosion te be
expected from the NaF-KF-UF, eutectic

in 100-hr exposures at 800°C. Figure
10.6 shows a heat-treated 316 stain-
less steel specimen, and Fig. 10.7

shows the corrosion in stainless steel
after 100 hr exposure to NalF-KF-UF,
eutectic at 800°C, The inconel speci-
men shown, attack toadepth of 2 mils,
is by no means the best specimen of
this metal tested. The stainless steel
specimen, attacked to a depth of 1 mil,
is typical of the 304, 310, 316, 330,

and 347 stainless steel samples tested.

 
 

FOR PERIOD ENDING SEPTEMBER 10, 1951

 

 

Fig. 10.2 - Corrosion of Inconel by 3NaF-UF,. 250X.
of exposure at 1000°C to 3NaF-UF,.

fluoride attack to depth of 0.004 in.

| Specimen after 100 hr
Voids adjacent to surface represent

        

E .40

 

 

 

N . ~
- #
. .
- «
. 4
" - -
“ ' L
>
= i
LAY -
R -‘.x e « - -
. A - - &
. “ o .
o B cY L - . 5 .
£ i ° - . L ! R
<. ¥ ‘ RS v 5 L
s - c. : - i
., ok, . . . ..
o sy . 67
w - i LA
e L s
ff ’ - ! >
. ~ Ty 5 -,
. - ; . . . , .
Ry o 3 :
. s R ’ * <
A, \ 3 A . . T,
L e i - & 3
Y - y 7 w”"? Ty ~
: S T 4 AT i
« L ; . . o
- gy : 8
. -
. / ¥ L 4
- - ¢ -~

Fig. 10.3 - Corrosion of 347 Stainless Steel by 3NaF-UF‘.

of specimen after 100 hr of exposure at 1000°C to 3NaF-UF,.

250X, Surféce
Note formation
of voids to depth of 0.002 in.

107

 
 

ANP PROJECT QUARTERLY PROGRESS REPORT

 

s T W T, ..
L . _ 'Y ‘ &
P W ® ey e
- - -”*, S . a # oy . " . 9 & *

Fig. 10.4 - Heat-Treated Inconel Specimen. 250X.

 

T " f"-l' Lo A ' -
Fig. 10.5 - Corrosion of Inconel by Pretveated Fluworide Fuecls. 250X,
Specimen after 100 hr at 800°C in NaF-KF-UF, eutectic; slight attack, to depth

of 2 mils.

108

 
 

FOR PERIOD ENDING SEPTEMBER 10, 1.951

P
* e ,,‘ o / ’ ~ -
"‘,/ - - / -
. W - »
o~ ¥ ’ - -
” \ -
> P "~
r .
* "
CE - L T e
/ c _ o M
3 : - % @ Tt
;,‘"‘ | e Y Yo R
. < - ’ -
. « " P ot A &
] 3 » .
oy [ {/ ‘\b %
R - P L
. - ek o
. : x
S
" i - ¢ i
. - - . . .
Lot e N ol
-~ ¥ e ¥ "

   
   

5

- - .
- 5 o ~ st ¥
o % . . . i
5 . X 1 ¥ .
- kT e
a - _"’ [ A - o+ . -
. Y 7 4 B e ha s
. - x - :
* ¥ e o - ! & ¥ o
' 4 e a * ' .
st L ¢ A
PO . "
. w D ly,n - * \1 . £ . S
i ¢ ';- { g " i f IS
i » 1 .
Y * f 4 . . ._,r a 4
Pos > fi #
$ * . - “E . x -
oy T - ~ L - -
- 1 \"4\ : e - * \
* .. 4
LN 4 ¥ e .

Fig. 10.6 - Heat-Treated 316

Fig. 10.7 - Corrosion of 316 Stainless Steel by Pretreated Fluoride Fuels.
250X. Specimen after 100 hr at 800°C in NaF-KF-UF, eutectic; no appreciable

attack.
109

 
 

ANP PROJECT QUARTERLY PROGRESS REPORT

In multiplicate experiments 1inm
which individual capsules were removed
at intervals for examination 1t has
been shown that the major fraction of

the corrosion observed occurs in the
first few hours. Little additional
effect 1s apparent after 1000 hr of

exposure.

By NaF-BeFQ-UFé. The ternary
fluoride NaF-BeF,-UF, (76-12-12 mole %)
was also used 1n corrosion testing.
A 100-hr test with 310 stainless steel
at 1500°F (815°C)
lographic evidence of attack, although
an analysis of the fluoride bath indi-
cated 0.2% iron, 0.02% nickel, and no
chromium to be present. The presence
of nickel and absence of chromium 1is
the reverse of the findings listed
above for the binary fluoride mixture,
NaF-UF4. However, these differences
may be 1llusory because tests described
below indicate that gravity segregation
may exist in the container, and the
composition of the fluoride mixture
will depend on the location from which
the test sample was obtained.

showed no metal-

A number of fluoride-corrosion
tests were made on commercial alloys
using a mixture of 10.7 mole % BeF,,
68.0 mole % NaF, and 21.3 mole % UF,.
The materials tested are listed 1in
Table 10.1 together with notes taken
while the exposed specimens were being
studied on the metallograph and results
from bath analyses. The attack in all
cases was not severe. Some decar-
burization was observed, but this 1s
not a highly objectionable type of
surface instability.

Weight losses are undoubtedly caused
by solution of the metal sample into
the fluoride bath as indicated by bath
analyses which are also reported in
Table 10.1. Weight gains, 1in spite of

110

some loss by solution, must be in-
terpreted as absorption of one or more
bath constituents (fluoride or metal)
into the metal specimen. Comparison
of results of analyses of the fluoride
bath at various levels of the tube
after test with results of the original
fluoride analysis indicate a definite
beryllium, a slight de-
and very little, af

decrease 1n
crease 1n sodium,
any, decrease 1n the uranium concen-
tration. These changes 1in bath
analysis may be directly associated
with the weight gains observed on
several test specimens.

Corrosion of Piatinum (D. G, Hill,
Consultant, Materials Chemistry
Division). Thermocouples are to be
inserted in a number of ARE fuel tubes
to determine the center temperatures
of these structures. Since platinum
was suggested as the material for the
tubular thermocouple wells, 1t has
been necessary to examine the resist-
ance of this metal and of some welded
joints to the fluoride melt.

Platinum slugs immersed in the raw
or pretreated fuel, NaF-KF-UF4, in
platinum capsules showed noappreciable
corrosion as measured by weight changes.
Microscopic examination of prepared
specimens revealed occasional shallow
pitting which may have been due to
local impurities. 1In a similar test
using liquid sodium metal instead of
the fluoride fuel mixture, the platinum
metal was completely consumed. These
test data will be reported in detail
later. At this time 1t appears that
pure platinum will prove satisfactory
although perhaps nickel plating of the
material will be desirable. The brazed
Joints must certainly be kept well
above the liquid level.

 
TABLE 16.1

Summary of Fluoride-Corrosion Data Obtained in 100 hr at 10600°C (1836°F)
(10.7 mole % BeF,, 68 mole % NaF, 21.3 mole % UF,)

 

 

BATH ANALYSIS AFTER TEST (ppm) | WEIGHT CHANGE
MATERIAL TESTED METALLOGRAPHIC NOTES _ TOP BOTTOM {g/in.?)

 

 

Nickel A Only evidence of attack was precipitation of fine | Ni 350 Ni 40 +0.002
partlcles along surface to depth of 0.002 1

304 stainless steel |Shallow surface voids observed to depth of <0.001 -0.005
: ste in. , , . , .
316 stainless steel Very few shallow surface voids observed to depth Fe 5,500 te 1, 500 0.000
of <¢.001 in. Ni 420 Ni 40
Cr 560 Cr 390
347 stainless steel |[Surface voids observed to depth of §.001 in.; no Fe 10,900 Fe 9, 800 -0.003
other visible evidence of attack Cr 2,400 Cr 1,300 S
310 stainless steel |DPecarburized to depth of 0.002 in.; few scattered -.003

voids in this area

   

330 stainless steel |Surface voids to depth of (.001 in. fe 4,500 Fe 7,200 -0.003
Ni 20 Ni 800
Cr 520 Cr 320
446 stainless steel |{Sigma-like precipitate throughout ferritic matrix | Fe 1,440 Fe 1,410 -0.9008
except at surface zone 0.00% . deep; few voids | Cr 3G Cr 190
observed to depth of 0.001 in.
Hastelloy B Decarburized to depth of 0.005 in.; lightly Fe 5,800 Fe 1,200 +0.003
etched phase persistent throughout Ni 310 Ni 70

Mo <10 Mo <ig

Inconel {as-received){Surface voids observed to depth of approximately

‘01 YIGWALJAS INIONT dOTHAd HOA

 

 

 

 

G.0062 in.
Inconel Surface voids observed to depth of 0.003 in. Fe <10 Fe i,810 +0. 003
{decarburized} Ni §00 N1 20
Cr <10 Cr 350
et 347 stainless steel |Surface roughened, voids chserved to depth of -0.126 &
:: in inconel tube 0.006 in. &2

 
 

ANP PROJECT QUARTERLY PROGRESS REPORT

Platinum specimens exposed 1n
inconel capsules in a manner such that
the metals were in contact under the
melt level showed similar and some-
what more severe pitting; pretreated
fuels caused 2 slight deposit to form
on the platinum. This deposit has
been tentatively identified as nickel
perhaps electrochemically transferred
from the i1nconel walls., Platinum
wires suspended so that they were 1n
direct electrical contact with the
inconel capsule above the melt only
and were partially immersed in the
fuel mixture showed only minor pitting
when examined by metallographic tech-
niques. The wires seemed stiffened
and slightly roughened and may possibly
have had some deposited nickel. This
type of construction, which approxi-
mates that proposed for the special
fuel tubes, will probably prove
satisfactory.

Brazed joints of platinum and in-
conel, prepared with pure silver,
copper, and nicrobraz, were shown to
be completely unsatisfactory when
immersed 1in the fuel mixture in an
inconel capsule. Both the platinum
and the inconel were seriously attacked;
the platinum virtually disappeared from
the nicrobrazed samples. Chemical
analysis, however, revealed no dis-
solved platinum in any of the fluoride
melts. Tt appears that the attack on
the welded platinum must have resulted
in finely divided dispersed metal or
electrodeposited platinum. Tt will
probably be necessary to design the
special tubes so that the inconel tube
extends beyond the active lattice and
through the coolant header, with the
brazed joint, therefore, outside
core and well above the coolant
fuel levels.

the
and

112

STATIC CORROSION BY MODERATOR COOLANTS

H. J. Buttram C. R. Croft
N. V. Smith J. A, Griffin

Materials Chemistry Division

Brasunas L. A. Abrams
E. E. Hoffman
Metallurgy Division

A. D.

Virtually all the work on moderator
coolants has been confined to studies
of corrosion by alkali and alkaline
earth hydroxides. These materials

have been studied under various con-

ditions 1n copper, monel, nickel,
inconel, and stainless steel Nos. 316,
321, and 347. Thus far corrosion by

sodium hydroxide has been examined at
some length, Preliminary data on
corrosion by potassium hydroxide,
barium hydroxide, strontium hydroxide,
and lithium hydroxide are also pre-
sented.

The corrosion data have all been
obtained in 100-hr tests at 800, 815,
or 1000°C. 1In general, the corrosion
by hydroxides is more severe than that
by the fluoride fuels. Sodium hy-

droxide, in particular, 1s about the
most corrosive of the hydroxides
studied; the surface penetration of

inconel and several stainless steels
varies from 7 to 16 mils, Only nickel,
monel, and copper of the metals tested
with sodium hydroxide show better (mno
evidence of attack) corrosion resis-
tance. FEach of these metals, however,
1s subject to mass transfer. Cor-
rosion by potassium hydroxides 1is
severe — around 6 to 7 mils — but not
nearly sosevere as by sodium hydroxide.
The attacks by strontium and barium
hydroxide are of about the same order
of severity, 1.e., 5 to 7 mils,

 
 

FOR PERIOD ENDING SEPTEMBER 10, 1951

To date the commercial grade and
the dehydrated commercial grade hy-
droxides have been used in most of
these experiments, and further re-
duction in corrosion rates may be
expected as purer quality hydroxides
become available. Weight changes of a
metal cofrosion specimen 1n pure
(99.8%) sodium hydroxide was less by a
factor of 3 to 4 than those obtained
with commercial grade. '

Capsule Technique with Hydroxides.
The capsule technique used so success-
fully in fuel and coolant testing has
required considerable modification for
this study. Most of the molten hy-
droxides, regardless of state of purity,
tend to crawl on the capsule surfaces
and make welding of the closed capsule
virtually impossible., Tt has been
necessary to add the powdered hy-
droxide 1n a manner such as to avoid
contact with the upper portion of the
wall and to seal the capsuvle without
melting of the material or to use a
double-capsule technique in which the
inner or test capsule is unsealed.
Neither of these technigues 1is com-
pletely satisfactory and a better
method is needed.

Corrosion by Sodium Hydroxide.
Commercial sodium hydroxide attacks
stainless steels and inconel strongly.
Monel, copper, and nickel are much
more resistant; nickel slugs exposed
to the alkali resemble electropolished
specimens. The corrosion problem with
these metals is, however, complicated
by the phenomenon of mass transfer for
which the mechanism is still in doubt.

The corrosion tests made with sodium
hydroxide. using inconel and stainless
steel specimens and container tubes
gave evidence of appreciable inter-
action. Photomicrographs of inconel

 

and 316 stainless steel specimens
exposed for 100 hr at B00°C to de-
hydrated commercial sodium hydroxide
are shown in Figs., '10.8 and 10.9,
respectively. These should be compared
with Figs. 10.4 and 10.6. A very
heavy oxide layer which contains some
metallic particles is formed on the
exposed surfaces. Purified sodium
hydroxide containing less than 0.15%
each of water and Na,CO, shows asimilar
but considerably less marked behavior
with stainless steel and inconel.
Weight changes observed on the metal
specimens in pure sodium hydroxide are
less by a factor of 3 or 4 than those
obtained in similar tests with com-
mercial material,

A photomicrograph of anickel speci-
men exposed to sodivm hydroxide is
shown in Fig, 10.10. There is no
evidence of a corrosion reaction,
although some metal loss due to solution
may have occurred. The possible
solution-deposition phenomenon
commonly referred to as "mass trans-
fer," remains the major deterrent to
the use of nickel as a hydroxide
container. An intensive study of this
phenomenon has been started, and
apparatus 1s being constructed teo
study solution deposition in a thermal
gradient under static and dynamic
conditions,

Corrosion by Potassium Hydroxide,
Potassium hydroxide tested only as the
dehydrated commercial product also
tends toattack the chromium in inconel
and stainless steel. Typical photo-
micrographs of such tests with inconel
and 316 stainless steel as shown in
Figs. 10.11 and 10.12 indicate that
the corrosion is very severe but is
not so violent as for NaQH. Nickel,
monel, and copper containers are also

attacked less severely. It seems

113
UNCLASSIFIED

ANP PROJECT QUARTERLY PROGRESS REPORT

 

e

T T ' 3 et il M
,,-."‘..- t - "; prad . ‘. - A k- JP -
tu - .‘ oy N ! ; P ; ;- \ PSS 07 Ly i .

0.012 in.

 

 

e

Fig. 10.8 - Corrosion of Inconel by Dehydrated Commercial Sodium Hydroxide.
Original photograph taken at 250X; reduced 19% in reproduction, Specimen
after 100 hr at 800°C in dehydrated sodium hydroxide; severe attack, to depth
of 0.012 in. Compare with Fig. 10.4.

 

 

0.011 in.

 

 

Fig. 10.92 - Corrosion of 316 Stainless Steel by Dehydrated Commercial
Sodium Hydroxide. Original photograph taken at 250X; reduced 19% in repro-
duction. Specimen after 100 hr at 800°C in debhydrated sodium hydroxide;
severe attack, to depth of 0,011 in. Compare with Fig. 10.6.

e UNCLASSIFIED
UNCLASSIFIED

FOR PERIOD ENDING SEPTEMBER 10, 1951

 

G.009 in.

 

 

Fig. 10.10 - Corrosion of Nickel A by Sodium Hydroxide. 250X, Surface

apparently unattacked after 100 hr of exposure to sodium hydroxide at 815°C
(1500°F). (

 

 

) é . ' , \J s A
el "{’\ S
. ¢ . TG
’ o “. "{N:fi\

Corrosion of Inconel by Dehydrated Commercial Potassium
Specimen after 100 hr at 800°C in dehydrated potassium

hydroxide; considerable grain growth and penetration to depth of 0.005 in.

Uj‘;; PR . L
ON/ ' .
- . S - . A \,/f

Fig. 10.11 -
Hydroxide. 250X,

Compare with Fig. 10.4.
115

UNCLASSIFIED
UNCLASSIFIED

ANP PROJECT QUARTERLY PROGRESS REPORT

 

3.007 in.

 

 

Fig. 10.12 - Corrosion of 316 Stainless Steel by Dehydrated Commercial
Potassium Hydroxide. 250X, Specimen after 100 hr at 800°C in dehydrated
potassium hydroxide; intergranular penetration and leaching to depth of

0.007 1n. Compar?figétb Figiglgféj._w

        

  

 

[3

Fig. 16.13 - Corrosion of Inconel by Strontium Mydroxide. 250X. Surface
of specimen after 100 hr of exposure to strontium hydroxide at 1000°C. Note
0.002-in. grain boundary penetration beneath 0.006-in. corroded layer.

116 UNCLASSIFIED
 

0.017 in,

 

 

 

Fig. 10.14 - Corrosion of Iron by Strontium Hydroxide. 175X. Surface of
specimen after 100 hr of exposure to strontium hydroxide at 1500°F. Note
formation of several oxide layers; approximately 0.005 in. of metal consumed
to form 0.017-1n. oxide layer. ' : ' '

 

 

Fig. 10.15 - Corrosion of 318 Stainless Steel by Strontium Hydroxide.
150X. Surface of specimen after 100 hr of exposure to strontium hydroxide at
1500°F. Note retention of tiny metallic glebules in oxade; 0,003 in. of metal
consumed to form 0.006-in. oxide layer.
 

ANP PROJECT QUARTERLY PROGRESS REPORT

possible that potassium hydroxide in a
high state of purity might be useful,
at least at somewhat lower tempera-
tures. Tests with pure potassium
hydroxide will be repeated as soon as
the material 1s available.

Corrosion by Other Hydroxides.
Corrosive attack by strontium and
barium hydroxides appears to be less
severe than the attack obtained in
tests with sodium hydroxide. Nickel
and copper appear moderately resistant
while inconel and stainless steel are
strongly attacked. Figure 10.13
illustrates the attack on inconel
obtained in 100 hr at 1500°F with
strontium hydroxide. This should be
compared to the sodium hydroxide attack
on inconel shown in Fig. 10.8. The
attack by strontium hydroxide on iron
and on 318 stainless steel at 1500°F
in 100 hr is illustrated in Figs. 10.14
and 10.15, respectively. The oxide
layer formed is quite heavy, especially
in the case of 1ron in which several
oxide layers (probably FeO and Fe,0,)
are evident, The eutectic mixture of
strontium and barium hydroxide did not
differ appreciably from the con-
stituent hydroxides. It should be
noted that the dehydrated commercial
materials have been used in these tests,
It is possible that the pure materials
will be less corrosive,

Lithium hydroxide tested as the
relatively pure dehydrated commercial
material seems to resemble sodium
hydroxide in its corrosive charac-
teristics. A thick layer of corrosion
product 1s formed on inconel and stain-
less steel; copper and nickel are
reasonably resistant,

Corrosion by Binary Hydrogenous
Systems. A few static corrosion
experiments have been made in which

118

varying amounts of NaH, Na, and H,0
have been mixed with NaOH. Corrosion
experiments with these mixtures on 316
stainless steel and inconel for 100 hr
at 815°C are summarized in Table 10,2,
No significant decrease in the cor-
rosion of NaOH was ohserved.

STATIC CORROSION BY FLUORIDE
COOLANT MIXTURES

H., J. Buttram C. B. Croft
N. V. Smith J. A. Griffin

Materials Chemistry Division

Uranium-free fluoride mixtures,
which are
coolants,

under consideration as
have been investigated 1inm
static-corrosion studies using the
capsule technique previously described.
Photomicrographs of a typical rum inm
which inconel ahd 316 stainless steel
were exposed at 800°C for 100 hr to
the NaF-KF-BeF, eutectic are shown as
Figs. 10.16 and 10.17, respectively.
While the number of specimens so far
studied is not large, it appears that
this mixture and the NaF-KF-LiFeutectic
will both be satisfactory in so far as
corroston 1is concerned. The beryllium
fluoride mixture needs careful handling
to avoid hydrolysis of this compound.
The alkali fluoride cutectic 1s much
more stable in this respect.

STATIC CORROGSION BY SODIUM

A. D. Brasunas .. A. Abrams
E. E. Hoffman
Metallurgy Division

l.Liguid-metal corrosion tests now
are confined mainly teo sodium attack.
In an effort to determine the effect

 
TABLE 10.2

Summary of Corrosion Data Obtained in 100-hr Tests at B15°€C with Mixtures of

Sodium Hydroxide with Sodium Hydride, Sodium, or Water

 

 

 

   

 

 

 

 

 

WEIGHT CHANGE
MATERIAL HYDROXIDE (g/in.2) REMARKS METALLOGRAPHIC NOTES
Inconel NaQH t+ 3% NeH +0.10 None Sample severely oxidized to depth
of 0.013 in.: container tube com-
pletely attacked

)
Inconel NeGH + § drops of H,0 +0.12 Tube failed 0.015-in. oxidized layer %
<
e
) =
Inconel NaOH + 3% Na +(.08 None 0.012-in. oxidized layer S
. ) . ] g2
316 stainless steel | NaOH + 3% NaH +0.011 None 0.003-in. oxide layer; 0.001-1in. %
intergranular attack beneath o
2
@
316 stainless steel | NaOH + 30% Na +0.033 None 0.003-in. oxide layer; 0.002-in. z
m
=
=
316 stainless steel | NaOH + 5 drops of H,0 Data not Tube burst from 0.003-in. oxide layer; 0.002-in. =
available internal pressure| intergranular attack beneath ok
=

1S6T

ot
ot
D
UNCLASSIFIED

ANP PROJECT QUARTERLY PROGRESS REPORT

 

Fig. 10.16 - Corrosion of Inconel by Pretreated Fluoride Coolant. 250X.

Specimen after 100 hr in NaF-KF-BeF, eutectic; no appreciable attack. Compare
with Fig. 10.4.

 

Fig., 10.17 - Corrosion of 316 Stainless Stee]l by Pretreated Fluoride
Coolant. 250X. Specimen after 100 hr at 800°C in NaF-KF-BeF, eutectic; no
appreciable attack., Compare with Fig, 10.6.

0 UNCLASSIFIED
 

FOR

of time on sodium attack of 316 stain-
less steel at 1000°C, tests were made
for1, 5,10, 25, 100, 400, and 1000 hr.
Gravimetric determinations indicated
no significant weight change 1in any -of
these tests. Metallographic exami-
nation of sectioned specimens revealed
no surface reaction. Visual ekami-
nation revealed a velvet-like frosted
appearance after the test (the speci-
mens were highly polished prior to
testing). High magnification indicated

this to have been caused by a re-

 

Fig.

10.18 - Corrosion of 316 Stainless Steel by Sodium.
face of specimen after 400 hr of exposure to sodium at 1000°C.

PERIOD ENDING SEPTEMBER 10, 1951

organization of surface metal. Grain
boundaries, twin lines, and crystal
faces were prominently displayed, as
shown in Fig. 10.18. This surface
is probably more stable than the
original polished surface and merely
readjusted i1tself either by surface
diffusion or by the limited solubility
of the metal components in the sodium.

Testéare now in progress to evaluate
the behavioer of Timken alloys 16-25-6
and 16-13-3 (Cr-Ni-Mo) in sodium at

UNCLASSIFIED

¥

y.3824

250X. Outer sur-
Note develop-

ment of crystal faces on certain grains and grain boundary attack.

 

121
 

ANP PROJECT QUARTERLY PROGRESS REPORT

815°C (1500°F) and 1000°C (1830°F).
The possibility of surface carburi-
zation of inconel, 310 stainless steel,
and 316 stainless steel 1s being de-
termined by using sodium to which a
small amount of charcoal has been
added. The indications are that
carburization and embrittlement has
occurred. The corrosion resistance of
molybdenum-coated steels will also be
evaluated when they are procured.

STATIC CORRBOSICN TEST OF A
REACTOR SYSTEM

H. W. Savage W. C. Tunnell
E. M. lLees

ANP Division

Static-corrosion experiments to
determine the extend of combined
damage (*) to capsules containing
fuel and i1mmersed in sodium at 1500°F
These tests are con-
incorporating a

have continued.
ducted in a vessel
series of wells or "standpipes™"™ filled
with liquid sodium. During the quarter
a 316 stainless steel standpipe was
used to test fuel-containing 316 stain-
less steel capsules. Tests were com-
pleted on the 100-, 200-, 400-, 600-,
900-, and 1000-hr capsules, but ex-
amination of these capsules has not
been completed. A similar experiment
is underway with inconel, and tests
on the 100-, 200-, and 400-hr capsules

have been completed.

MASS TRANSFER PHENOMENON IN
STATIC CORROSION

D. G. Hill, Research Participant
R. A. Bolomey, Materials
Chemistry Division

The mechanism by which nickel 1is
removed by sodium hydroxide from the

 

Sodium-Inconel -Fluoride

147.

(4L‘Static Corrosion:
Fuel,’ ANP-65, op. cit., p.

122

in many
and

parent metal of the container,
cases 1n a very uniform manner,
redeposited on certain areas of the
container or in discrete particles 1in
the liquid 1s referred to as "mass
transfer.” It 1s not certain at
present whether the phenomenon is shown
by all the other hydroxides or whether
metals other than nickel and copper
exhibit the phenomenon. There 1s also
some gquestion as to whether 1t occurs
in completely isothermal systems,

Mass transfer may be explained 1in
nonisothermal systems by assuming a
high thermal coefficient of solubility
of nickel in the fused alkali. Such
an explanation 1s acceptable 1n the
case of mass transfer in liguid metals,
but true solution of unreacted nickel
in fused alkali is somewhat harder to
accept. If the transfer can be shown
to occur 1n completely isothermal
systems, then another explanation 1is
patently required.

Chemical-Reaction Mechanism for
Mass Transfer., Jt is possible that
the mechanism of mass transfer can be
explained on the basis of chemical
reactions. For purposes of this dis-
cussion nickel 1s used as the example.
Other metals might, obviously, be
chosen.

The first stepin the reaction chain
may be either

Ni + Na,0 -——> NiO + 2Na + H,
or

2Ni + 2NaOH —> 2Ni10 + 2Na + 1,

Experimental evidence for these re-
actions is found in the work of Villard
who demonstrated the distillation of

 
 

FOR

sodium and of H, by heating a nickel
boat containing sodium hydroxide in a
Sodium and H,,

quartz tube. oT some -

times Nall, could be collected at the
cooled end of the tube, the products
depending on the conditions. Thus

the second reaction shown should in-
clude the possibility that the actual
product is NaH, This work demonstrates
that, although the free energy change
is unfavorable to the reactionsg,
reaction does occur so that, 1f a
further process displaced the equi-
librium, continued attack on the metal
might take place.

The further reaction may be the

disproportionation reaction
2N10 = Ni + Ni0,

Under the more common conditions of
aqueous solution and strong acids, the
equilibrium in this type of reaction
is far to the left. Tt is considered
possible that, in the extreme con-
ditions of high temperature and very
strongly basic solvent, the equi-
librium 1s shifted to the right. Such
reversals of equilibrium waith a change
in conditions are well known in organic
reactions, The scanty information now
available on reactions in fused sodium
hydroxide does not forbid the theory
for this unexplored mediunm. :

further reactions are certainly
possible:

NiQ, + 2NaOH = Na,NiO; + H,0

NiQ, + 2Na = NiD + Na,O

The water produced in the first of
these possibilities would react with

metallic sodium, regenerating sodium

 

PERIOD ENDING SEPTEMBER 10, 1951

hydroxide for further corresion, while
the Na,0 produced in the second would
be immediately available for reaction
with more nickel.

Thermodynamically the total process
is only the transfer of nickel from
impure wall metal to very pure nickel
found in the deposit; such a process
operates with a decrease of free energy
even at a constant temperature. If the
temperature varies from place to place
or from time to time, the differing
temperature coefficients of the several
reactions might operate in such a way
as to increase the rate of the process,

One factor
the experimental
observation that NiO remains inert at
the bottom of a platinum crucible under
fused NaOH at 350°C, but at 750°C it
1s either suspended or more probably
dissolved in the melt. If a melt of
sodium hydroxide in platinum is treated
with nickel at 750°C and the contents
of the crucible are gquenched rapadly
Lo room temperature,

Expeyimental Evidence.
worthy of note 1s

the melt gives

off a small amount of a gas on treat-
ment with water. On analysis this gas
proves ta be O,. This is very daffi-

cult to explain on any basis other than
that suggested here, since 1t 1s known
that either Ni0O, or*Na,NiQO,
with water to give oxygen.

reacts

The same type of reaction may apply
to all metals that are known to exhibit
mass transfer 1n sodium hydroxide.
Fven the noble metals silver and copper
should react toaminute extent to . give

sodium, and only a very small amount
may be needed to permit the dis-
proportionation to operate., Work is

in progress to study the reactions by
several weans, particularly by polarc-
graphic technigues and by potential

measurements,

123
 

ANP PROJECT QUARTERLY PROGRESS REPORT

DYNAMIC CORROSION TESTS IN
THERMAL-CONVECTION LOOPS

E. M. Lees, ANP Division
J. L. Gregg, Consultant, ANP Division
R. B. Day, Metallurgy Division

During the period June 1 to August
31, 1951, the ANP Experimental Engi-
neering Group has continued testing
dynamic corrosion of liquid metals
on various materials in conjunction
with the Metallurgy Division. Tests
of this nature conducted in thermal-
convection loops afford a practical
means of evaluating the usefulness of
certain materials with liquid metals
or other circulating coolant, supply
fundamental information on the suita-
bility of fabricating techniques,
afford opportunities to test certain
instruments under development, and
provide experience from which adequate
techniques for operating larger scale
dynamic systems may be formulated.

Metallographic examination of six
lead-containing and three lithium-
containing loops operated during the
preceding quarter¢®’ have been com-
pleted. Operation of all but one of
these loops was prematurely (before
1000 hr) terminated, and all but one
of these failures was due to internal
obstruction in the cold zone of the
loops. In general, the corrosive
attack of both lead and lithium on the
variety of steels and alloys tested
1s most severe in the hot zone of the
loop with some attack usually apparent
in the cold zone. The solid particles
released to the circulating lead and
Jithium streams are apparently "trapped”
in the cold lead, accumulate, and
eventually plug the loop.

 

(5)"Dynamic Corrosion by Liquid Metals,”
ANP-65, op. cit., p. 150.

124

During the past quarter, primarily
sodium was tested in the thermal-
convection loops. JIn no case was
there a failure with a sodium-con-
taining convection loop in which the
failure was due to sodium corrosion.
The successful operation of loops with
sodium at 1500°F and preliminary re-
ports of metallographic examination on
loops sectioned after 1000 hr of
operation indicate quite strongly that
inconel and types 310 and 316 stain-
less steel suffer very little cor-
rosion by slowly flowing sodium at
temperatures up to 1500°F. The
successful operation of a 310 stainless
steel loop fabricated from Y%-in.-o.d.
tubing with a wall thickness of only
0.020 in. may be cited to indicate the
relative inertness of stainless steel
to sodium at 1500°F.

Corrosion by Lithium and Lead. A
number of thermal-convection loops
which were operated with lead and
lithium were examined after failure.
The hot-leg temperature was approxi-
mately 1500°F, although much higher
temperatures were obtained momentarily
when internal plugging obstructed the
natural convective flow. Failure
occasionally occurred at these higher
temperatures, not from internal cor-
rosion by the molten metal but rather
from external causes, such as arcing
or high-temperature air oxidation. As
flow restriction, due to corrosion,
was quite frequent with both lead and
lithium, so were the i1ntervals of
excessive temperature, and this
probably contributed appreciably to the
failure rate with these loops.

The life of the various loops tested
with lithium and lead, the compositions
of the loops, and metallographic notes
are given in Tables 10.3 and 10.4,
respectively. The metallographic

 
TABLE 16.3

corrosion and Operational Data on Lithium-Containing Thermal-Convection Loops

 

 

 

 

 

LOOP TEMP. (°F) ANALYSIS OF LOAD AFTER TEST (%)
HOT COLD i LIFE OF HOT ZONE COLD ZONE
LOOP ZONE ZONE [LOOP (hr} REMARKS METALLOGRAPHIC EXAMINATION Cup cup
No. 20, 1100 1020 34.5 Loop plugged (?) Intergranular attack observed Fe 0.0025 Fe 0.0075
1010 steel during operation; | in hot and cold zones te
failed in hot depth of 0.020 in.; hot weld
zone region attacked to depth of

0.040 in.; no metal crystals
observed in cold zone

   

 

-
S
=
~
No. 35, 1500 1375 280 Loop plugged (?) Intergranular attack noted to Fe 0.01 e
Haynes internally; depth of 0.0608 in. in hot- Ni  0.05 3
Alloy failed in hot zone pipe, and to depth of Cr 0.75 g
No. 25 zone 0.016 in. in the hot-zone Co 0.75 ,
weld; only 0.005 in. attack Man 0.02 %
evident in cold-zone pipe, —
none in cold-zone weld =
<
N
o
No. 38, 0.0 Loop failed at Some decarburization noted Fe 0.08 Fe 0.04 3
V-36 start of test Ni  0.0025 Ni 0.011 ;
because heater Cr (.001 Cr 0.9624 o
element arced Co <0.0025| Co 0.0035 =
across hot-zone
. e
pipe &~

 

 

 

 

 

 

 

1S61

o
o
T
ot
N
N

TABLE 10. 4

Corrosion and Operational Data on Lead-Containing Thermal-Convection Loobps

 

LOOP TEMP. (°F)

 

ANALYSIS OF LOAD AFTER TEST (%)

 

 

 

HOT COLD LIFE OF HOT ZONE HOT ZONE COLD ZONE COLD ZONE
LooP ZONE ZONE | LOOP (hr) REMARKS METALLOGRAPHIC EXAMINATIOM cup PIPE cup PIPE

No. 19, 1010 1100 885 1000 Temperature gradient | Hot and cold zone tube walls showed very little Fe 0.0049] Fe 0.013) fe 0.0081 Fe 0.0009
steel in tube increased evidence of attack; metal crystuls formed in

gradually, indicat- cold rones
ing partial plugging

No. 18, 1010 1400 62 Loop plugged in- Tube walls showed very little evidence of attack; Fe 0.39 Fe 0.017 fe 0.0079
stee] ternally during meta) crystals formed in cold zones {Fig 10.23) Cr 92.071
{(enamejed aperstion
exterior)

No. 32, 316 1500 1300 7 Loop plugged in- Intergranular attack to depth of 20 mils noted in Fe 0.057 Fe 8.8 Fe 0.C14
stainless ternsliy in cold hot zone; none noted in cold zone; metal crystals Ni 0.064 Ni 1.5 Ni 0.35
steel zone; faijure in observed in cold rone (Fig. 10.20) Cr 0.022 Cr 3.6 Cr D.026

hot zone due to ex- Mo 0.020
ternal causes Mn 0.01 Mo 0.13 Mn C.11

Ne. 26, 316 1500 1390 138 Loop plugged in- Hot zone much wore severely attacked than cold Fe 0.15 {Fe 0.04 Fe 4.6 Fe 11.2
stainless terpally during tone; metal crystals observed in cold zone Ni 0.05 |Ni 0.03 Ni 9.6 Ni 5.0
ateel operation Cr 0.001|Cr 0.001 Cr 1.39 Cr 0.03

Mo 0.012 { Mo 0.004 Mo 0.005 | Mo 0.78
G, 0.048 } O, 0.005

No. 28, 446 1500 1300 62 Loop plugged in- 0.00% in, intergranuler attack obaerved in hot Fe 0.03 |Fe 0.045 Fe 7.8 Fe 5.1
stainless ternally during sone; 0.002 in. observed in cold zone; wmetal Cr <0.001 jCr <0.001 Cr 2.92 Cr 4.64
stee] operation crystals observed in portions of coid leg

(Figs. 10.21 and 10.22)

Mo. 39, ¥-36 1500 1200 34 Loop plugged in- Genera! attack in hot zone only; cold rone showed Fe ¢.025 F= 0.0014 Fe 0.046

Alloy ternally during evidence of metal crystal formation {Fig. 10.19) Ni 0.21 iNi ¢.11 Ni 0.39
operation Cr 0.22 {Cr 9.022 Cr 0.81

Mn 0.004B{Mn 0.0018 Mn D.01T

Co 0.52 |Co 0.08 Co 0.79

No. 3§, Haynes 1500 1290 699 Loop plugged in- Hot zone revealed more general attmck than cold Fe 0.0036 Fe 0.0083|Fe 0.016

Alloy No. 25 ternally; feiled zone; mets]l crystals observed in cold zone Ni 90.073 Ni 0.055 {Ni 0,067
in hot zone Cr 0.018 Cr 0.031 !Cr 0.030
Mn 0.0039 ¥Ma 0.0018iMm 0,0059

Co 0.14 Co 0.09 |[Co 0.15

 

 

 

 

 

 

 

 

 

 

1044 SSTYI0™d ATHALAVAO LDATOMd dINV

 
 

FOR

notes were made after examination of
random samples of the hot leg, cold
leg, and welded sections. The extent
of corrosion was dependent upon the
materials involved, the temperature,
and the time of interaction. It should
be kept in mind that the temperature
was not actually controlled but merely
attained by initially selecting a
certain power input to obtain approxi-
mately 1500°F at the beginning of the
test, when convective flow was un-
obstructed, When partial or complete
plugging developed as a result of
metal crystal formation in the cold
leg, as shown 1in Fig. 10.19, heat
losses were reduced and the hot-leg
temperature was increased. This was
corrected manually and the test was

 

Fig.
divisions on

10.19 - Metal Crystal Formation in Loop Containing Lead.
the scale are 1/100 in.

PERIOD ENDING SEPTEMBER 10, 1951

allowed to proceed until the loop was
completely plugged or until it actually
failed., A point of failure 1s shown

in Fig. 10.20.

Althongh it is accepted that the
corrosive attack in the hot zone is
appreciable, there is a general mis-
conception that attack in the cold
zone is negligible or even nonexistent,
Figure 10.21 1llustrates an attack 1in
the cold zone of a 446 stainless steel
loop by lead which appears to be inter-
dendritic 1in nature. Such attack
releases solid particles to the circu-
lating lead stream and may cause or
accelerate plugging in this manner. A
mass of metal crystals observed in the
cold leg of this 446 stainless steel

UNCLASSIFIED

Y. 3945

¢

B LS A

The smallest
{(z) Transverse and (b} longitudinal

sections showing formation of metal crystals adjacent to cold leg wall of V-3¢
thermal-convection loop containing lead after 84 hr of operation.

 

127
UNCLASSIFIED

ANP PROJECT QUARTERLY PROGRESS REPORT

 

Fig. 10.20 - Failure in 316 Stainless Steel Loop Containing Lead. 3X, Hot
leg failure after 7 hr of operation at 1500°F of 316 stainless steel thermal-

convection loop containing lead. Note that failure resulted from attack on
outer surface of tube wall,

 

. - . -

« W

Fig. 106.21 - Thermal-Convection Loop Operated with Lead. Original photo-
graph taken at 1000X, reduced 31% in reproduction. View of internal surface
of 446 stainless steel cold leg from thermal-convection loop operated with
lead (area at approximately 1300°F). Note that interdendritic attack releases
solid particles to lead stream,

UNCLASSIFIED

128
FOR PERIOD ENDING SEPTEMBER 10, 1951

loop is shown 1in Fig. 10.22. These
crystals need not necessarily be
bonded directly to the tube walls but
may merely be trapped there. The
opxide layer shown in Fig. 10.23
separates the metal crystals from the
loop material, making bonding im-
possible. '

An attempt was made to determine
the composition of the metal crystals
by quantitative analysis. Neglecting
lead, the following elements were
detected: |

In the cold zone of a 316 stainless
steel (18% Cr, 10% Ni, 3% Mo) loop
after 138 hr of operation: 67% Fe,
6% Cr, 24% Ni, 3% Mo.

In the cold zone of a 446 stainless
steel (26% Cr) loop after 62 hr
operation: 63% Fe, 37% Cr.

The composition varied somewhat and
these values are averages of analyses
made in different areas of the cold
zones.

Corrosion by Sodium. Materials
tested with sodium during the quarter
included inconel and types 310, 316,
and 347 stainless steel. FEighteen
loops were taken out of service, 14
of which completed the scheduled 1000-
hr tests with sodium at 1500°F. Three
stainless steel (two 347, one 310)
%-in. thin-walled tubing loops failed
after less than 150 hr of operation
owing to leaks in the welds, and one
inconel loop failed after 350 hr owing
to a plugged gas line 1n the liquid
level control system; thus none of
these failures are directly attributa-
ble to the use of sodium. Included in
the 14 loops completing 1000-hr tests
was one 310 stainless steel thin-

 

walled loop. On the three loops
operating with sodium at the end of
the period, one 1inconel loop was
operated at 1600°F, and one each of
inconel and 316 stainless steel at
1500°F. The increase in successful
runs over the preceding guarter is due
largely to (1) improved methods for
purifying and introducing liquid metals
into the systems (2) improved methods
for degreasing, cleaning, and fabri-
cation, and (3) improved methods for
heating and temperature control.

The sodium-containing loops are now
being examined. Preliminary metal-
lographic examination of some of these
loops indicates very little corrosion,
Small test specimens of similar and
dissimilar composition were suspended
in the hot and cold legs of these loops.
Examination of these specimens has
indicated that although weaght changes
are small, the hot-zone specimens lost
weight but the cold-zone specimens
gained weilght,

Tt is believed that little can be
learned by operating additional thermal
loops with sodium except for the pur-
pose of testing a particular joint or
component. Although it hasbeen proved
that thermal-convection loops can be
operated without plugging and without
appreciable attack of the loop material
by sodium, it remains to be proved
that a forced-convection loop can be
operated with sodium in one section at
1500°F and 1000 to 1200°F in another
section without plugging. Therefore
future tests in figure-eight forced-
convection loops will determine whether
there will be sufficient attack on
inconel or stainless steel to result
in plugging in addition to testing
simulated components or flow channels.

129
 

Fig. 10.22 - Cold Zone o9f 445 Sizinless Steel Loop Showing Attacked Surface
and Metal Crystals %hich Form in Lez# (Dark Areas). 150X. The crystals
restrict convective flow.

 

- ;“"n  H; :'.. - . - fixf

Fig. 10.23 - Cold Zone Weld in 1910 Sigsl Loop Showing Metal Crystal For-
mation Adjacent to Pipe Wall Covered with Yroi Oxide, Presumably Formed During

Loop Fabrication. 175X,

130 UWE

  
 

FOR

DYNAMIC CORBOSION TESTS IN FORCED
CONVECTION LOOGPS

W. B. McDonald, ANP Divisions

Figure-eight sized loops are inter-
mediate in size between the thermal-
convection loops and full-scale ARE
components and afford opportunities
to test cleaning, handling, operating,
and pumping techniques developed from
smaller systems, in addition to supply-
ing useful data on corrosion. Ex-
perience gained from intermediate-
sized systems 1s necessary for develop-
ing adequate technigques and methods
for assuring acceptable performance
testing of full-scale ARE components.
Operation of these loops to date,
however, has afforded very little
opportunity to observe corrosion
effects.

During the past quarter one figure-
eight loop comnstructed from 347 stain-
Jess steel was tested with a down-1flow
centrifugal pump designed by the Experi-
mental Engineering Group for circulat-
ing sodium. The test was terminated

 

PERIOD ENDING SEPTEMBER 10, 1951

after 7 hr of operation by failure of
the pump shaft seal. The seal was
repaired, but during the following
start-up period external arcing from
electrical heaters burned a hole in
the main heater section of the loop.
The damage caused by the resulting
fire was sufficient to require com-
plete dismantling of the loop. Total
operating time for this experiment was
insufficient to test revised cleaning
procedures which i1ncluded repeated
flushing with hot sodium until the
oxide content remained low,

In an effort to prevent repetition
of loop failure being caused by
auxiliary equipment in future tests, a
General Electric electromagnetic pump
has been tested in a separate loop to
determine its reliability prior to its
installation in the figure-eight loop.
To prevent future failures due to
external electrical arcing, specifi-
cations have been drawn up for a
70,000-Btu/hr gas furnace for heating
figure-eight loops, and these specifi-
cations have been submitted to the
Purchasing Department for bids or
proposals.

131
ANP PROJECT QUARTERLY PROGRESS REPORT

11.

PHYSICAL PROPERTIES AND HEAT-TRANSFER RESEARCH

H. F. Poppendiek, Reactor Experimental Engineering Division

The influence of the free-convection
mechanism within liquid-fuel elements
upon the fuel temperature structure 1is
being investigated both experimentally
and theoretically. Mock-up systems
which simulate the actual fuel-element
systems have been studied. Volume
heat sources were produced by means of
resistance heating, and temperatures
were measured with thermocouples. An
apparatus has been built which will be
used to study the velocities of the
free-convection cells. A theoretical
analysis of the free-convection heat
transfer existing in the fuel element
has been completed for the laminar
flow case; the turbulent flow analysis
is currently being completed. These
solutions are to be used to predict
temperature distributions in liguid-
fuel-element systems.

Five Bunsen 1ice calorimeters for
the measurement of heat capacity are
now in operation., Experimental work
and analyses have been completed for
316 stainless steel, synthetic sapphire,

lithium, molybdenum, and zirconium 1in

the range 300 to 1000°C,

An improved thermal-conductivity
apparatus (Deem type apparatus No, 2)
for ligunids has been constructed, A
value has been obtained with the
equipment previously deseribed(!)
{Deem type apparatus No. 1) on fuel
salt mixture No. 2 (46.5 mole % NaF,
26.0 mole % KF, 27.5 mole % UF,) 1in

 

(1)L. F. Basel, “'Thermal Conductivity of
Liquids,' Aircraft Nuclear Propulsion Project
Quarterly Progress Report for Period Ending June
10, 1951, ANP-65, p. 161 (Sept. 13, 1951).

132

the range 643 to 700°C as 0.53
Btu/hr-ft-°F., The longitudinal-flow
apparatus for the determination of
thermal conductivity of solids is in
operation and additional conductivity
measurements on various densities of
boron carbide have been obtained. The
radial -flow apparatus has been com-
pleted.

Two pieces of eguipment have now
been constructed for the measurement
of liquid densities. The density of
the NaF-KF-UF, eutectic has been
determined in the range 535 to 1000°C.
The apparatus for the measurement of
viscosities is scheduled for com-
pletion by September. Work has been
retarded by delays in fabrication of
parts.

An apparatus which will be used to
determine the heat-transfer coef-
ficients of fused hydroxides and salts
has been nearly completed, These
fluids are potential reactor coolants.
Some preliminary heat-transfer data
for a boiling-mercury system have been
determined and the figure-eight lithium
system has yielded some data.

INVESTIGATION OF FBEE CONVECTION
WITHIN LIQUID-FUEL ELEMENTS

Theoretical Analysis of Natural
Convection (D. C. Hamilton and H. F,
Poppendiek, Reactor Experimental
Engineering Division). An analytical
solution for the temperature distri-
bution i1n the parallel-plate system
has been derived. 1In this system a

 
 

FOR PERIOD ENDING

fluid with a uniform heat source 1s in
laminar flow between two parallel and
infinitely long plane bounding sur-
faces. In the system it was postulated
that heat was generated uniformiy and
that the long convection cells were in
the laminar flow regime. A report
based on this analysis 1s being pre-
pared. A similar analysis wz2ll be
made for a system in turbulent flow.
These solutions will also be duplicated
for cylindrical coordinate systems.
The solution indicates that, for a
system in which the distance between
the parallel plates 1s 0.18 in., the
axial temperature gradient is 100°F/ft
and the physical properties are identi-
cal te the measured values for the
NaF-KF-UF, eutectic; the temperature
difference which would exist 1f all
the heat were transferred by con-
duction is reduced by 25.5%. This
result is based on a value of 1 centi-
poise for the dynamic viscosity.
If the viscosity is greater than 1
centipoise, the reduction in tem-
perature difference will be smaller

than 25.5%.

Measurement of the Fuel-Element
Temperature Distribution (F. E. Lynch
and M. Tobias, Reactor Experimental
Engineering Division). A new apparatus
has been devised for measuring the
effects of free convection on the
temperature distribution in tubes
filled with fluid in which heat 1is
being generated by electrical means.
Because of the fragility and com-
plexity of previously constructed
devices, considerable difficulty has
been encountered in making a series of
without interruptions due to
A diagram of the present
apparatus is shewn in Fig., 11.1.
Thermocouples are arranged in the
central tube as described in previous

runs
breakage..

 

SEPTEMBER 16, 1951

UNGLASSIFIED
DWG 12938

POTENTIAL LEADS

 

 

9% in,

CAN

     

TEST
SECTION

o

4in,

 

 

 

 

BELL

 

 

 

 

Fig. 11.1 - Schematic Diagram of
Free Convection Apparatus.

reports,(z's) Aluminum wires arve

sealed into curved tubes which are

 

(2JF._E. Lynch and P. €. Zmola, “‘Study of Free
Convection in Liquid-Fuel Elements,” ANP-65
op. c¢it., p.

(3YF. E. Lynch and P. C. Zmola, “Study of Free
Convection in Liquid-fuel Elements,’ Aircraft
Nuclear Propulsion Project Quarterly Progress
Report for Period Ending March 10, 1351, ANP-60,
p. 232 (June 19, 1951). :

»

133
 

ANP PROJECT QUARTERLY PROGRESS REPORT

Joined to the ends of the straight
test section and serve as potential
measurement leads. The electric
current which generates the heat in
the central tube 1s led in by an
electrode through the large curved
tube at the top and passes along the
test section and out through the bell
shaped base. The entire apparatus is
placed i1nside a stainless steel can
which 1s filled with liguid to just
below the top of the large curved
tube. The central section 1s held
vertical by a retaining clamp at the
top and by a swall well at the base.
The return electrode is clamped to the
When used with brine, or
of low freezing

can 1itself,
other substances
point, the system is cooled by wrapping
a coil of a swall thermostatically
controlled refrigerating unit around
the can., A stirring motor assists 1in
maintaining a uniform temperature 1in
the can. -

Tests will be run using mercury and
brine. Design of a new apparatus which
will be used for the fuel salt itself
is now being considered., The chief
difficulty 1s finding a contailner
material which will resist attack by
the salt, 1s structurally strong, and
is a nonconductor of electricity,

Measurement of the Fuel-Element
Velocity Distribution (R, Redmond,
Reactor Experimental Engineering
Division). An apparatus has been
designed and constructed for the
purpose of obtaining preliminary in-
formation about the nature of the
free convection patterns 1in the volume
heat source system; temperature distri-
butions are also to be measured. A
flat plate cell which contains dilute
acid 1s to be heated by electric
current. The acid will have the
generated heat removed at the walls of

134

the flat-plate cell by water cooling.
Velocity patterns will be determined
by tracing the paths of particles
suspended 1n the solution, and the
distribution will be
thermocouple measure-

temperature
determined by
ments.

PHYSICAL PROPERTIES
A, R, Frithsen, USAF

M. Tobias, HReactor Experimental
Engineering Division

Heat Capacity (William D. Powers,
Reactor Experimental Engineering
Division). The heat capacities of

type 316 stainless steel, syathetic

sapphire, lithiuw, molybdenum, and
zirconium have been determined. The
values which follow give the heat

capacity as a function of temperature
(Cp is the heat capacity at T°C in
calories per gram per degree centi-

 

 

 

grade) in the indicated temperature
range.
TEMPERATURE Cp
MATERTAL RANGE (°C) (cal/g/°C)
Type 316 150 - 1000 {0.11 + 5.7 x 10”57
stainless
steel
Synthetic 300 - 900 |0.26 + 4.4 x 10°°T
sapphire
Lithium 950 - 1100 [1.0 + 2.8 x 10757
Molybdenum 200 - 1100 0.068
Zirconium 150 - 1050 [0.070 + 3.6 x 10757

 

 

 

 
FOR

Heports are forthcoming on each of
these materials. The substances now
under investigation -are sodium hy-
droxide, nickel, sodiuvm flmoride,
petassium fluoride, uranium fluoride,
potassium hydroxide, and fuel mixtures.
Work will be started soon on lead,
bismuth, and lead-bismuth eutectic,

A modification of the original ice
calorimeter has been made and included
in the five calorimeters. The heat
content of 2 sample is measured by
noting the change in volume of a pure
ice-water mixture. The mixture is
contained in a lucite shell in which =
hollow finned copper cyvlinder 1is
sealed. The copper cylinder receives
the sample at a known temperatnre,.
Jece i1s melted in the lucite shell and,
by noting the change 1in volume i1n the
ice-water mixture, the heat content of
the sample way be calculated. To
prevent heat from getting into the
system from any source other than the
sample, the lucite shell was surrounded
by ice. However, impurities in the
surrounding i1ce lowered 1ts melting
point encugh te cause noticeable
freezing of the water in the lucite
shell., The lucite shell was therefore
surrounded by 2 in, of S:1-0-Cel
insulation contained in a steel shell.
The steel shell was kept at 0°C by
surrounding 1t by 1ce. The heat
leakage into the lucite shell was
small.

An appreciable error can occur in
the measurement of the actual tempera-
ture of the sample. At high tempera-
tures, variations of 5 to 10°C have
been found on the surface of capsules
placed in the most uniform tempera-
ture position in the furnace. It 1is
felt that tube furnaces longer than
those now in use will reduce thzs
error. Manufacturers are being con-

 

PERIOD ENDING SEPTEMBER 10, 1%51

sulted to see if furnaces of the
desired characteristics can be fur-
nished.

Thermal Conductivity of Liguids
(M. Tobias, A. R. Frithsen, and L.
Basel, Reactor Experimental Engineering
Division)., The thermal conductivity
of the fused salt NaF-KF-U¥F, (com-
position 46,5, 26, and 27.5 mole %,
respectively)} has been measured using
the OBNL Deem type apparatus,¢!? The
average conductivity of four measure-
ments was 0.53 Btu/br fo-°F at temper-
atures ranging from 643 to T700°C.
These data were urgently needed and
no check runs were made with sodium.
As several imperfections an the design
of the apparatus have become apparent,
measurements with this apparatus have
been discontinued.

A new apparatus, the ORNL Deem type
No. 2, which has a bellows seal instead
of the Pb-Bi seal used in apparatus
No., 1, is now approximately 95% com-
plete and is scheduled for cperation
in September, A few check runs will
be made on lead, after which data will
be obtained on BeF -NaF-¥F salt
mixtures. A hood is being installed
over the apparatus to protect operators
from escaping BeF, vapor.

Thermal Conductivity of Solids
(M. Tobias, Reactor Experimental
EFngineering Division). The thermatl
conductivity of copper has been
measured nup to 600°C by the long:-
tudinal flow method.{*? The data
(Fig. 11.2) agree to within 4% wath
the values given in the International
Critical Tables.

 

(4)y. Tobias, “Thermal Conductivity of Salids,”
ANP-65, op. cit,, p. 162,

135
 

ANP PROJECT QUARTERLY PROGRESS REPORT

 

 

 

 

 

 

 

 

 

 

 

 

 

UNCIASSIFIED
250 DWG, 12939R1
=
Z 20
3 . [GT GOPPER DATA
& b~ ] !
& . m""""'*-:-—-l_...f_h__“‘_&_ L .
>, : . Tt
= 400 o 3
0
2
a
&
U
~ 150
3
oL
w
=
=
0 100 200 300 400 500 00 0
TEMPERATURE (°0)
Fig. 11.2 - Thermal Conductivity of
Copper,

Improvement of the preliminary
data on the thermal conductivity of
boron carbide by the longitudinal flow
method 1is continuing. In addition,
experiments have been started using
the radial flow apparatus to measure
thermal conductivity.

Falling-8all Viscometer (S. T.
Kaplan, Heactor Experimental Engineer-
ing Division). The falling-ball
viscometer 1s now being checked out
with water. High-temperature checks
with bismuth are contemplated. Delivery
of the electronic eguipment to measure
the rate of fall of the radiocactive
ball was overdue and has only recently
been placed i1n satisfactory operation.
The dropping valve of the apparatus(®?
has been modified to accommodate balls
up to % in. diameter. Tantalum-plugged
stainless steel balls, 0.25 1.d. with
a densityof 12.2 g/cm?® at room tempera-
ture, as well as 0.20-1n.-diameter
cobalt-coated pyrex balls with a room-
temperature density of 2.6 g/cm’® are

 

(SL‘Viscosity—Measuring Tube,” Atrcraft Nuclear
Propulsion Project Quarterly Progress Report for
Pertiod Ending December 10, 1950, ORNL-919,
Fig. 7.8, p. 200 (Feb. 26, 1951).

136

Additional
balls of various densitlies are
prepared,

presently available. test

being

Brookfield Viscometer (F. A. Knox
and I, Kertesz, Materials Chemistry
Division). Some approximate values
for viscosity of typical fuel specimens
and for some fluoride mixtures of
interest as coolants have been obtained
by a wmodified Brookfield Synchro-
Lectric viscometer fitted with a
special spindle., The 1nstrument was
and
at Toom

calibrated 1n distilled water

water-glycerol solutions

temperature and in molten sodinum

chloride at 830 to 1000°C.

The viscosity of the NaF-KF-UF,
eutectic was found to be about 32 to
35 centipoises at 600°C and 12.5 to 16
The NaF-BeF,-XI

under consideration as a

centipoises at 1000°C.
eutectic,
coolant,
Since these samples

also showed high values.
both showed con-
siderable guantities of suspended
{uranium oxide in the fuel
and beryllium oxide in the coolant},
these values may be greatly in error.

matter

The viscosity of the NaF-KF-1aF
eutectic mixture was 1n the expected
low range of 2.2 centipoises at 500°C
and 1.5 centipoises at 800°C.

Brookfield Engineering laboratories
are designing a high-temperature
rotational viscometer for this work.
Preliminary designs were submitted by
them for inspection, and modifications

are now under study.

Vvapor Pressure (R. E. Moore and
C. J. Barton, Materials Chemistry
Division}). The temperature of the
fluoride fuel at the center of the ARE
fuel pins i1s still a subject of some
uncertainty but will almost certainly
 

FOR

be higher than 2000°F. While the
components of the ligquid fuel are
thermally stable to temperatures well
above this figure,
determine

1t 1s necessary to
the vapor pressure of
suitable fuel mixtures in this temper-
ature range, Preliminary results
indicate that the vapor pressure of
the KF-NaF-UF, eutectic 1s less than
3 mm at 1000°C.

The apparatus and procedure employed
are those reported by Rodebush and
Dixon¢®? and moditied by Fieck and
Rodebush, ¢(7) The prepared fuel
mixture 1s placed in a cylindrical
vessel of stainless steel which may
be heated in a resistance furnace,
The vessel i1s connected by one tube
to a manometer and by a second tube
to an intermittent pump and toe a
The two tubes
outside the furnace,

source of inert gas.
are connected,
by a differential manometer containing
a light liguad.

With the container and sample
maintained at the desired temperature,
the inert gas pressure 1s reduced 1n
small intervals by the pump. The
residual inert gas pressure 1s equiva-
lent to the vapor pressure of the
sample at the highest pressure suf-
ficient tomaintain a permanent pressure
difference in the differential ma-
nometer. This action depends on the
fact that when the i1nert gas pressure
is equal to or less than the vapor
pressure of the ligquid, diffusion of
the inert gas 1s the only mechanism
for equalization of the pressure 1in

 

(6)W. H. Rodebush and A. L. Dixon, ‘“The Vapor
Pressures of Metals; a New Experimental Method,”
Phys. Rev. 26, 851 (1925).

(7)E. F. Fiock and W. H. Rodebush, “*The Vapor
Pressures and Thermal Properties of Potassium and
Some Alkali Halades,” J. Am. Chem., Soc. 48,
2522 (1926). '

 

PERIOD ENDING SEPTEMBER 10, 1951

the two tubes. TIf the tubes are of
sultable diameter this diffusion is
very slow.

Considerable difficulty has been
encountered 1n mailntaining temperature
control in the resistance furnaces at
the high temperatures required. It is
certain, however, that the total vapor
pressure of the NaF-KF-UF, eutectic
mixture 1s well below that of pure
UF, at temperatures in the range 0 to
1050°C. These experiments are being
continued at present.

Pensity (S. I. Kaplan, Reactor
Experimental Engineering Division).
Following test runson sodium at 200°C,
a determination of the density of
NaF-KF-UF, eutectic was made over the
range 535 to 1000°C (46.5, 27, 26.5
%, respectively) using the
buoyancy apparatus described in a
previous report.(®> The data can be
represented by the equation

mo le

o = 4.702 - 0.00115T,

where £ is the density in grams per
cubic centimeter and T is the tempera-
ture in degrees centigrade. The test
material was blanketed in an atmosphere
of argon purified by bubbling through
NaK. Determination of the density of
various coolant and fuel salt mixtures
is projected for the next quarter,

HEAT-TRANSFER COEFFICIENTS

Bealt Transfer in Fused Hydroxides

and Salts (H. W, Hoffman, Reactor

 

(8)5. I. Kaplan, “Density of Liguids,’ ANP-60,
op. cit., p. 246. '

137
ANP DIVISION QUARTERLY PROGRESS REPORT

Experimental Engineering Division).
The construction of the apparatus for
the determination of the heat-transfer
coefficients of fused hydroxides and
salts 1s approximately 80% complete.
Operation of the system 1s anticipated
sometime within the next month. The
first material to be tested in the
apparatus will be the eutectic mixture
of NaF-KF-LiF having a melting point
of 454°C, Future fluids to be tested
include NaOH and the eutectic mixture

of Ba(OH)z-Sr(OH)Q.

A schematic diagram of the experi-
mental system 1s given in Fig. 11.3.
The charge is first melted under vacuum
in the melt tank and then moved by
applying argon gas pressure to tank A.
In an experimental run the molten fluid
1s pushed by gas pressure through the
heat exchanger and i1nto tank B. Tank
B rests on a scale, and cumulative
welght readings are made to determine
the fluid flow rate, Temperatures of
the outside surface of the heat ex-
changer are obtained by means of
chromel-alumel thermocouples spark-
welded to the tube wall. An electric
current 1s passed through the wall of
the heat exchanger tube, and the heat
input 1s determined from the measure-
ment of this current and the voltage
drop along the heat exchanger. The
inlet and outlet temperatures of the
test fluid are measured by thermo-
couples on the tube wall in the iso-
thermal regions immediately before and
after the test section. The 1nside
wall temperature is calculated from
the outside wall thermocouple readings.
At the conclusion of a run, lasting
approximately 30 min, the fluid 1is
pushed through the return line 1nte
tank A, The system 1s then ready for
another experimental run. Temperature
level and flow rate 1n the system can
be varied within reasonable limits,

 

Heat Transfer in Boiling-Ligquid-
Metal Systems (W, 5. Farmer, Reactor
Experimental Engineering Division).
The boiling-heat-transfer investi-
gations have been continued during
this period on the boiling-mercury
experiment using a flat plate and on
the boiling-sodium and -potassium
experiments using a horizontal tube.

only were
boiling-mercury

Preliminary results
obtained for the
experiment owing to condenser failure
in the high-frequency induction heater
employed as a heat source. However,
using pure mercury, temperature dif-
ferences of approximately 100°F were
observed between the metal surface
and the boiling fluid for a heat flow
rate of about 50,000 Btu/hr ft2?, The
large temperature difference observed
in this case was probably due to non-
wetting of the heat-transfer surface
by the boiling mercury. In addition,
considerable temperature fluctuations
were observed in the boiling liguid,
accompanied by bumping. The apparatus
has been moved to a new location to
utilize another induction heater, and
further results will be forthcoming.

UNGLASSIFIED
CWG. 12940

FILTER

. HEAT EXCHANGER | {1 TANK
+ {5

™

POWER
SUPPLY

 

RETURN LINE

Fig. 11.3 - Schematic Diagram for
Determining Heat-Transfer Coefficients.
 

FOR PERIOD ENDING SEPTEMBER 10,

of the horizontal
tube boiler has been accelerated during
this past quarter. The erection and
fabrication of all tanks and piping
are complete, Only the duct work and
electrical construction remain to be
done. This latter work should be
completed and preliminary sodium and
potassium runs should be possible this
coming quarterly period,

The construction

Heat Transfer im Molten Lithium
(C. P. Coughlen and H. C. Claiborne,
Reactor Experimental Engineering
Division)., During the past quarter
the figure-eight lithium system was
placed in operation, and preliminary
heat-transfer data were obtained. To
remedy pump and instrumentation dif-
ficulties whichwere encountered during
operations, extensive alterations have
been made and a new Einstein-Szilard
pump has been obtained on loan from
Allis-Chalmers Manufacturing Company.
The new pump will permit operation
over a

much larger range of velocities
in the

test heat exchanger.

The experimental results obtained
with a concentric-pipe heat exchanger
are preliminary because of the 1im-
perfect operation of the auxiliary
squipment which prevented the attain-
ment of complete equilibrium and
because of the low flow rates obtainable
(1140 to 1600 lb/hr in the annulus and
980 to 1150 1b/hr in the core pipe).

 

1951

Separation of the i1ndividual heat-
transfer coefficients was not possible
under these conditions.

The overall heat-transfer co-
efficients determined experimentally
ranged from 4100 to 5160 Beu/hr-ft?-°F,
Heat balances checked to within 24%
with an average deviation of 12.5%.
The major errors in the heat-balance
data undoubtedly resulted from in-
complete equilibrium and the un-
certainty itn the flow measurements.

A comparison of the Nusselt modulus
computed from the experimental data
with that of Lyon(?) and that of
Isakoff and Drew('?) is shown in Fig.
11.4. The experimental values are 16
to 80% higher than those found by
Isakoff and Drew when using mercury

for an L/D = 38. Saince L/D % 30 for

the lithium heat exchanger, the
Nusselt modulil are expected to be
somewhat higher than those for a

system with L/D =58, Tt is ewmphasized
that the results shown in Fig. 11.4 are
based on a rough separation technique
of the fluid thermal resistances. A
more detalled discussion is given in

ORNL CF-51-8-32.

 

(9)R. N. Lyon, “Liquid Metal Heat Transfer
Coefficients,” Chem. Eng, Progress 47, 75 (1951).

(10)5. E. Isakoff and T. B, Drew, ‘““Heat and
Momentuim Transfer in Turbulent Flow of Mercury,”
to be published.

139
 

ANP PROJECT QUARTERLY PROGRESS REPORT

UNCL ASSIFIED

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

2 DWG. 11913
10
ISAKOFF AND DREW,
Ha, L/D = 58
A
o '
\~
Q@@ e
45 /
21 4 Ve
S EXPERIMENTAL POINTS FOR /
s LITHIUM, £/0 =30
—
- O
wn
wn
2 o LYON EQUATION (NaK)
OO
& o
10 4
10° 10

PECLET MODULUS

Fig. 11.4 - Comparison of ORBNL Lithium Heat-Transfer Data with Those
of Other Investigators.

 
 

'FOR PERIOD ENDING SEPTEMBER 10, 1951

_ 12. METALLURGY AND CERAMICS

W. D. Manly, Metallurgy Division

The installation of eguipment in

the creep laboratory, powder metallurgy
faboratory, and arc-welding laboratory
iz essen tiaily complete. The work on
the stress-rupture laboratory to test
meta is in a ii1quid-metal medium i1s
progressing, and the laboratory should
e 12 nhshal %y ctober i. In addition,
2 csramics laboratorv has been eos-

tablished and 1s mow 1o pavtial oper-
4 C1ON,

Welds of the tube-to-header type
have been made using nicrobraz, cone
arc, and manuai arn; thase welds were
evaluated using tensile tftests and
wodiafied fatigue tests. Strengths
comparable to the parent inconel stock
have been found 1n these jmint51 in
creep testing the condition of Lhe
specimen is being corvelated wal? the
creep resistance of the material.
The effect of h@a% treatment of inconel
on its creep resisbance is considerable,

Fabrication of fuel elements usiug
technigues previously discussed’ '’ has
sontinvued, A new method of fabrication,
rubberstatic pressing, has shown
promise in fuel-element fabrication,
Work on control-red fabrication has
been started using the dafferent
powder metallurgy techniques,

 

g, A Adamson, “Fuel-Element Fabrication,”™
Aircraft Nuclear Propulsion Project Quarterly
Progress Report for Period Ending June 10, 1857,
ANP-65, p. 181 (Sept. 13, 1951). :

 

WELDING OF INCONEL

P. Patriarca F. W. Dresten
G. M. Slaughter
Metallurgy Division

The bulki of work to date has been
confined rte basic evalustion ot the
guality of inconel tube-to-hesader
welded joints described in the laszt
report. ) A limited amcunt of fabri-
cation 0f tvypical tube-to-header
assemblies has been undertaken, Tensile
tests have shown that tube-te-header
welds can be made which compare faver-
ably 1n both strength and ductility to
the parent inconel tubing. Fatigue
tests of walded Joints exhibit astrong
dependennsy ou the degree of penetvation
of the weld, 2 condition not evident
in tenstle data. Two manuwally ineve-
arc-welded inconel tube-to-header
pairs were Lested in ligquid sodiuw at
1000°C for 100 hr to determine the
effect on the room-temperature tensile
strength. The only effect nbsarved
was a slight lowering of the tensile
strength and an appreciable increase
in elongation due to annealing.

Teasile Testx. The results of
room-temperature tests of inconel
tube-to-header tensile pairs made
using manual-inert-arc, cene-arc, and
braziong techniques are summarized in
Table 12.1. It may be noted that the
strength and ductility of manually and
cone-arc welded pairs were quite
similar and compared favorably to the

 

{Q)P. Patriarca, “Welding of Inconel,” ANP-65,
ep, cit. p. 191

141
ANP PROJECT QUARTERLY PROGRESS REPORT

TARBLE 12.1

 

 

 

Room—-Temperature Tensile Properties of Inconel Tube-to-Header '"Pairs"
NO. OF |AVERAGE TENSILE | AVERAGE ELONGATION,
*
DESCRIPTION TESTS |STRENGTH (psi) (% 1n 3 1in.)

Menual-inert-arc weld with complete pene- 3 93,000 23.8

tration; tested as welded
Cone-arc weld with complete penetration; 3 96, 500 21.5

tested as welded
Cone-arc weld with average penetration of 4 95, 500 26.9

70% of 0.062-in. header sheet; tested as

welded
As-received inconel tubing 2 105, 000 30.8
Nicrobrazed 2 82,200 28.9
Manual-inert-arc weld with complete pene- 2 89, 500 36.9

tration; tested after treatment in liquid

sodium at 1000eC for 100 hr

 

 

 

 

*Ténsi]e strength based on tube dimensions of 0.187 in. o.d.; 0.025-in. wall except for pairs treated
in liquid sodium; tubing used was 0.188 in. o.d.; 0.030-in. wall.

final pair length --

initial pair Jength

 

** .
Percent elongation =

x 100.

initial pair length

strength of the as-received inconel
tubing used in the fabrication.

It is interesting to note that
welded joints with average penetration
of 70% of the header thickness of
0.062 in, exhibited strength and
ductility comparable to welds with
complete penetration. Failures in both
types of joints occurred in the heat-
affected zone immediately adjacent to
the weld, as would be expected since
this area 1s that of minimum cross-
section. Partial penetration requires
considerably less critical control

of the variables of the cone-arc
method, Complete penetration 1is more
142

difficult to achieve with consistency,
and 1t is expected that even more
difficulty will be encountered when
experiments with thinner header sheets
and smaller tubing are attempted.

Strength values for joints brazed
with nicrobraz in preliminary experi-
ments are included in Table 12.1. The
lower tensile strength and somewhat
higher elongation of these specimens
may be attributed to the annecaling
received in the brazing operation,
Brazing of basic joints for evaluation
and comparison with welded joints will
recelve major consideration in future
work.
FOR PERIOD ENDING SEPTEMBER 10, 1951

fatigue Tests, In order to evaluate
further the quality of tube-to-header
joints, a rotating tubular beam fatigue
test was developed. The header plate
is immovably clamped with the tube
extending into a rotating eccentric
cam. The test is effectively torsion-
less since the eccentric cam, within
which the tube end rides, i1s equipped
with roller bearings, Specimens were
5.8 1n, long and were rotated at 1125
rpm at a pre-set deflection. '

A limited number of data are pre-
sented as an S-N scatter band curve in

Fig, 12.1. Further tests are necessary
for conclusive evaluation. It may be
noted, however, that welds incorpo-
rating complete penetration, whether
manual inert arc, cone arc, or brazed,
exhibited superioer fatigue life to
those joints with partial penetration.
This condition was not evident on ex-
amination of tensile data. The lower
fatigue life of these welds may be
attributed to the lack of penetration,
and, as a result, the presence of a
"notceh."™ The evaluation of fatigue,
tensile, and corrosion properties of
partially welded cone-arc joints with

‘UNCLASSIFIED
DWG~Y-4638-RI

 

1 T 1 I I

[TTl

&6C

50

 

[op]
Q
> 40 —
——
o
2
- -
0
Lel
* 30 -
n
20 -
To) i 1 _I_lLl!I! ] I 11||11|' ]

&® © O

   

T ‘ 1

MANUAL INERT-ARC WELD,
1000% PENETRATION

CONE-ARC WELD,
70% AVERAGE PENETRATION

NICROBRAZ

CONE-ARC WELD,
100% PENETRATION

 

 

10 100

000

NUMBER OF CYCLES X {0°

Fig. 12.1 - Effect of Penetration of fFatigue Life on Inconel Tube-to-

Header Specimens,

 

143
ANP PROJECT QUARTERLY PROGRESS REPORT

the notch eliminated by brazing will
be the subject of future work.

All-Weld-Metal Tensile Tests. The
properties of weld deposits in inconel
plate are of interest since the ARE
reactor shell and other components
will be fabricated from thick plate,
A limited number of tests have been
conducted using Y-in. inconel plate
and various filler metals to provide
all-weld metal 0,357-1n. tensile
specimens., The remaining weld deposits
were reserved for quantitative analysis
and static corrosion tests in liquaid
sodium and a fluoride fuel mixture.

TABLE

The results of room-temperature
tensile tests are presented in Table
12.2. 1t may be noted that the physi-
cal properties of the inert-arc and
metal-arc weld metal were comparable
with one exception. The weld deposit
of the Inco No. 42 welding rod was
definitely inferior. This rod was
designed for oxyacetylene welding;
however, chemical analysis being con-
ducted may explain these relatively
inferior physical properties.

Inspection of welds of the tube-to-
header type involving a relatively
intricate geometry and numerous small
components 1s difficult. Radiography

12.2

Inconel All-Weld-Metal Temnsile Values

0.357-1in.-diameter test specimens; 2-in,

gage length

 

 

 

 

 

 

 

 

TENSILE
SPECIMEN STRENGTH ELONGATION YIELD STRENGTH,
NO. DESCRIPTION {psi1) (% in 2 in.) 0.2% OFFSET (psi)
1 Inert-arc welded; 1/8-in. inconel wire 100,000 40.5 55,000
filler metal
2 Metal -arc welded; 1/8-in. Inco No. 132 rod 85, 500 31 47,200
filler metal
3 Metal-arc welded; 1/8-in. Inco No. 132 rod 97, 400 37.5 58,000
filler metal
4 Inert-arc welded; 1/8-in. Inco No. 62 rod 98,000 44 55, 500
filler metal
5 Inert-arc welded; 1/8-in. Inco No. 62 rod 101, 400 43.6 51,000
filler metal
6 Inert-arc welded; 1/B-in. Inco No. 62 rod 92,000 36 48,000
filler metal
7 Inert-arc welded; 1/8-in. Inco Ne. 42 rod 78, 600 33 33,000
filler metal
144

 
 

FOR PERTOD ENDING SEPTEMBER 10, 1951

has been attempted with single speci-
mens, and deliberate flaws which were
located on the radiographs were much
more evident with the naked eye., Tt
15 expected that the use of Dy-Chek
(a compound used to check for flaws
in welds), followed by microscopic
examination of weld surfaces, hydraulic
pressure testing of assemblies, and
helium leak testing will prove useful
in detecting major flaws., This prfiw
cedure will be set up as partof
future work as a preliminary approach
to the problem,

Corrosion of Welds. The effect of
the tensile stremgth of
inconel tube-to-header speci-

sodium on
we lded

mens has been indicated previously.

Samples of the weld deposit from
specimens 1 and 2 of Table 12.2 were
treated in liguid sodium at 1000°C for
100 hr and in a ligquid sodium-—po-
tassium—uranium fluoride mixture at
816°C for 100 hr, There was no sig-
nificant loss or gain in weight of auny
of the sawples. Metallographic ex-
amination indicated a weld immunity to
attack by these liquid metals during
the interval of test,

A recent prob-
because of de-

Special Weld Tests.
Jem which has arisen
parture from the tube-to-header type
of reactor involves sealing off of the
inconel tubing under a positive helium
pressure, Preliminary experiments have
been performed which indicate that
pressures of 25 psig of helium can be
held within inceonel tubing by appli-
cation of mechanical pressure to crimp
the tubing. While still under mechani-
cal pressure, the tubing is sheared
from the pressure source and is then
welded shut, Pressure-tight tubes with
gages brazed on one end for verifi-
cation have been subjected to swaging

 

experiments in an attempt to accomplish
a uniform cylindrical closure more

suitable to the newly proposed ARE
design,

Another recent problem involves the
use of a hydroxide as a cooling medium
in preference to sodium. Negotiations
are 1in progress with the International
Nickel Company for cladding of welded
inconel tube-to-header pairs with
various thicknesses of nickel. These
test specimens will be evaluated for
corrosion properties as plated and
atter diffusion treatment, Plating
experiments with nickel and copper
will also be conducted at Oak Ridge
using the facilities of the Research
Shops at X-10 to supplement the work
at INCO.

WELDING OF

i

A contract with Bensselaer

MOLYBDENUM

Poly-
technic Institute for an investigation
of the welding of molybdenum has been
granted, Jnitilal research will invelve
electric resistance flash welding and
w1ll be underway during the next
quarter,

CREEP OF METALS IN CONTROLLED
ATMOSPHERES

R. B. Oliver J. W,
C. W, Weaver
Metallurgy Division

Woods

The installation of equipment in
the creep laboratory is essentially
complete, The installation in the
stress-rupture laboratory for testing
in liquid metal 1s about 7T0% complete
and 1t 1s hoped that testing can be
started in September.
ANP PROJECT QUARTERLY PROGRESS REPORT

The creep-testing program was
started early in June. The initial
emphasis was to traln nine new tech-
nicians as well as to determine proper
operating conditions for the equip-
ment, Preliminary results from creep
tests on 316 stainless steel, inconel,
and niobium have shown that the labo-
ratory is functioning properly. These
early results are affected by some
inaccuracies which are being overcome.

Creep of 316 Stainless Steel.
Duplicate sheet specimens of 316
stainless steel (0.500- by 0.065-in.
reduced section) have been loaded to
3000, 4000, 6000, 7000, and 8000 psa
at 815°C in an argon atmosphere. There
1s considerable scatter of the results,
and this work 1s being repeated with
closer attention to details. The
scatter probably reflects the struc-
tural i1nstability of the stainless
steel.

Creep of Inconel. Sheet specimens
of 1inconel in the as-received temper
(0.500- by 0.065-in. reduced section)
have been loaded to 3000, 5000, 6000,
and 8000 psi at 815°C in an argon
atmosphere., Duplicate tests check
very well, and the results are close
to the published data. Also, speci-
mens were heated to 1120°C in vacuum
for 2 hr and air-cooled; the treatment
increased the grain size by a factor
of about 16. Following this anneal,
specimens were loaded to 5000 and 8000
psi. At B00O psi there was close
agreement between duplicate tests, but
the rupture life was much shorter than
With

loading the elongataion

with the fine-grained specimens.
the 5000-ps1
vs., time curves for the coarse- and
fine-grained materials were nearly
identical. Loadings of coarse- and
fine-grained material will be made at
3000 psi; it 1s expected that the

coarser grained material will have a
considerably longer rupture life at
this and lower stress levels.

Severe intergranular cracking has
been observed on the inconel sheet
specimens. These cracks appear to
originate at the edge. To see if this
is truly an edge effect, one inconel
0.505-in. bar specimen has been loaded
to 3000 psi. Unless the same type of
cracks develops on the round bar, it
can be assumed that this type of crack-
ing is an edge phenomenon and need be
considered only in the rare case 1in
which exposed and stressed edges occur.

Creep of Niobium. Two niobium
specimens were loaded to rupture at
815°C in argon. Rupture occurred at
26,000 psi with about 20% elongation.
One specimen loaded to 6000 psi has
been in test for 650 hr with a negli-
gible elongation. Work on niobium has
been temporarily suspended pending
delivery of additional sheet stock.

Creep Test of Loaded Inconel Tube,
A study is being made of the effects
of multicomponent stress systems on
creep. An inconel tube % in. in o.d.
with the wall thickness reduced to
0.010 in., for a length of 3% in. was
sealed at one end and loaded to 66.6
psi internal pressure with argon. The
tube was heated to 1800°F, and the
pressure was maintained until the tube
ruptured, the failure being longi-
tudinal. This loading gave the follow-
ing stresses:

Tangential 1500 psi (tension)

I.ongitudinal 750 psi (tension)

Radial 35 psi (compression)

 
 

FOR PERIOD ENDING SEPTEMBER 10, 1951

Rupture occurred in approximately
400 hr with the following deformations:

Length 0.8% increase

Circumference 2.8% increase

Internal cross-
sectional area

- 5.7%

6.5% increase

increase

Volume

A similar sheet specimen loaded in
tension to 1500 psi at 1800°F ruptured
in 200 hr with about 6% elongation.

STRESS-RELAXATION TESTS

R. B. Oliver, Metallurgy Division

Eight simulated stress-relaxation
tests for type 316 stainless steel
were conducted at stress levels of
5000, 10,000, and 20,000 psi using the
Baldwin creep-rupture machine. All
tests were made at a temperature of
1500°F. After 16 hr the stress was
about 4600 to 4800 psi regardless of
the initial stress level. Stress-
relaxation curves are also to be
obtained for inconel and nickel A,

Design work for stress-relaxation
tests on inconel bolts in sodium at
1500°F has been completed, but con-
struction of the equipment has not
been finished.

FUEL-ELEMENT FABRICATION

G. M. Adamson £. S, Bomar
J. H. Coobs

Metallurgy Division

The work on fabrication of solid
fuel elements has progressed along the

 

lines discussed in the last report. (!
An additional fabrication technique,
rubberstatic pressing, is now also be-
ing i1nvestigated., With all fabri-
cation technigques the use of a screened
fraction of sintered;UO2 seems to be
desirable,

Hot Rolling. After a plate has
been fabricated by hot-rolling it is
desirable that it be drawn to final
size, Since a suitable drawbench is
not available, cold-rolling was sub-
stituted, A set of plates was fabri-
cated by the procedure outlined in
OBNL-987. They were hot-rolled at
1100°C to a reduction of 75%, Six
plates, each containing 30% U0, , were
made; three had matrices of iron and
three of 302 stainless steel. Plates
from each set were then given additional
cold reductions of 1CQ, 25, and 50% by
hand-rolling., The plates with iron
matrices rolled very well, showing
only minor evidence of segregation
of the UO,. However, the units with
matrices of stainless steel suffered
considerable segregation and striation
of the UO,. 1In neither case was the
bond between the core and cladding
material altered. Figures 12.2 and
and 12.3 are representative of this
work.

To determine the effect of using
coarse U0, in the fabrication of cores
a high-temperature induction furnace
was assembled, A toctal of 330 g of a
special high-purity UQ, was sintered
at 2100°C in a hydrogen atmosphere.
The average weight change amounted to
a loss of 0.5%. After being sintered
the coarse fragments of U0, were
uniformly gray in color and quite hard.
The apparent density had been increased
by a factor of 3. The sintered product
was then ground and divided into
various size ranges by screening.

147
u

ANP PROJECT QUARTERLY PROGRESS REPORT

Y.4110

 

Fig. 12.2 - Effect of Cold-Working Stainless Steel—U0, Cores, 175X,

‘ Y.4227

 

Fig. 12.3 - Effect of Cold-Working Iron—-—-uo2 Cores. 175X .

 
FOR PERIOD ENDING SEPTEMBER 10, 1951

The -100, +200 mesh size fraction
of the U0, was mixed in two batches
with -100 and -325 mesh 302 stainless
steel. Compacts were prepared in the
usual manner and rolled at 1120°C to
reductions of 50, 75, and 90%. Me-
tallographic examination indicated
segregation of the coarse U0, (Fig.
12.4a) to be less semsitive to the
particle size of the matrix stainless
steel powder than is normal UQO,
(Fig. 12.4b). However, a tendency to
agglomerate into localized groups for
reductions greater than 75% was noted.
There was alsoc some evidence of the
U0, particles penetrating into the
cladding layers.,

To conserve on stainless steel an
attempt was made to use mild steel
tubing instead of stainless as the
protective covering. The samples
were rolled at 1100°C; however, they
were heavily oxidized before receiving
the first pass through the mill. On
metallographic examination it was
evident that the oxidizing conditions
had penetrated to the compact and had
prevented adequate bonding. Further
tests will be made, and it is hoped
that a satisfactory atmosphere may be
obtained by changing the adjustment on
the gas dissociator.

Mechanically Formed Matrix. The
work in the X-10 Shops on developing
a method for punching molybdenum
plates and obtaining a high percentage
of open area, has been held up waiting
the delivery of carbide punches., Other
methods of forming a suitable mechani-
cal matrix are being placed on an in-
active basis until the accompanying
bonding problems have been solved. A
low-temperature induction furnace 1is
being assembled for these solid-phase
bonding studies., Bonding will be
tried both with and without load. The

 

furnace 1is nearly completed and 1is
expected to be placed 1in operation. as
soon as the induction furmaces are

ready.

Loose-Powder Sintering. Work on the
sintering of loose powders 1is:
being carried out 1n our own furnaces.
The first run was to be one of a series

now

to determine optimum sintering temper-
ature and was to be run at 1250°C.
Instead of the intended as-received
UD,, a size fraction {-100, +325 mesh)
of the high-density sintered UOQ, was
used by mistake. With 30% coarse U0,
in 320 stainless steel a fair bond was
found at 1250°C., JIncreasing the
temperature to 1280°C did not give a
definite improvement in the bond.
Another sample using regular UO, was
prepared and sintered as above; how-
ever, very little bond was obtained.

Rubberstatic Pressing. Rubberstatic
pressing has been used during this
quarter as an additional fabrication
technique. A drawing (Fig. 12.5)
shows how this procedure may be used
to press a powder layer onto the inner
surface of a tube. Pressure applied
to the steel punches is transmitted
through the rubber and presses the
powder tightly onto the inner side of
the tube. This 1s followed by a
sintering operation. This technique
is being used as a fabrication method
on both 'solid fuel elements and con-

‘trol rods.

In the first specimen tried in the
fuel-element work the lower half of
the powder layer was 302 stainless
stee}l, and the upper half was electro-
lytic iron. After being pressed under
40 tsi and sintered at 1250°C, the
stainless steel was well bonded
but the iron had pulled away during
sintering, A series of samples was

- 149
 

ANP PROJECT QUARTERLY PROGRESS REPORT

Y.4428

 

Fig. 12.4 - Effect of Particle Size of U0, (a) Coarse vuo, . 175X, (b)
Normal U02. 250X,

150

 
 

FOR

UNCLASSIFIED
DWG-IZTI?_

 

N
§

| ——STEEL PUNCH

 

STEEL BACKING TUBE

 

AN

AN

 

b

 

SOSOSOSNY

  

 

NS

 

 

o

 

 

 

STEEL DIE
STEEL MANDREL

 

RUBBER CORE

AN

I in.

_—

 

 

 

 

Fig. DPie for Rubberstatic

Pressing.

1205 -

high-density coarse UOQ, _
particle sizes in -325 mesh 320 stain-
less steel matrices, All the compacts
were pressed at 30 tsi and sintered at
1250°C. The sample with -325 mesh U0,
was not bonded, while another with
~200, +325 mesh U0, was bonded only
superficially. However,
using -100, +200 mesh and -60,
mesh U0, were well bonded.

Compatibility of Potential Fuel-
flement Materials. To determine the

 

PERIOD ENDING SEPTEMBER 10, 1951

compatibility of UQ,-iron or U0, —302
stainless steel cores in contact with
316 stainless steel, cladding sections
of hot-rolled plates were sealed 1in
evacuated tubes. The tube and ‘i1ts
contents were held at 1000°C for
100 hr. Metallographic examination
disclosed a diffusion layer of less
than 0.00]1 in. between the 1ron cores
and the 316 cladding and none between
the 316 cladding and the 302 cores.
Both of these results were in agree-
ment with earlier tests,

CONTROL-ROD FABRICATION

G. M. Adamson E. S.
- J. H. Coobs

Metallurgy Division

Bomar

This group i1s now starting work on
the fabrication of control rods for
the ARE. While hot-rolling and rubber-
static pressing are both being in-
vestigated as possible fabrication
procedures, the most likely method, at
present, seems to be a simple canning
of powders. Materials now being in-
vestigated as possible control ma-
terials are B4C and HfOz, but the
investigation 1s being expanded to
include development of refractory
boron, hafnium, or cadmium compounds,

As a preliminary investigation ZrQ,
was used as a substitute for Hf02, and
some plates were prepared with the
standard hot-rolling technique. While
these two compounds are similar chemi-
cally, there is a considerable differ-
ence in their density; however, it is
not likely to affect this preliminary
work. Cores with 30% by volume of
ZrOzijlnickel, iron, and 302 stainless
steel were fabricated. They were clad
with nickel or 316 stainless steel by
hot-rolling at 1100°C. Table 12.3 is
a summary of the results of this in-
vestigation.

151
 

AND

PROJECT QUARTERLY PROGRESE REPOGHRT

TABLE 12.3

Remnlts of Iovestigaiions of Zro

 

25 2 Contvel Material

 

 

 

 

 

 

 

 

2
BOND OF CLADDING
CORE (vol %) CLADDING REDUCTION (%) TO CORFE REMATHS
30 Zz0,, TG Ni | Nickel 56 Very good Distribution of Zz0, good
80 Very good Distribution of Zr{)2 good
30 ZrOz, 70 N1 | 316 stainless steczi 92 Good Tendeney teward clumping of
oxide particles
30 ZrOzv 70 Fe | 316 stainless steel 53 Fair Distribution good
T4 Fair Slight tendency toward clumping
30 Zz0,, 70 Fe 316 stainless steel g2 Good Extreme clustering of Zi(
particles; sows penstration of
ZyQ, into cladding
30 Zr0,, 70 302| 316 stainless steel 71 Very good Distribution very gocd
stainless stee!

 

METAL CLADDING OF BERYLLIUM OXIDE
Gerity Michigan Company of Adrian,
Michigan, has been attempting to
developa plating procedure for cladding
beryllium oxide. The only oxide
readily available at the beginning of
this investigation was some high-fired
material left from the Daniels
and several crucibles of low-density
porous material. The high-faired
material had a dark film on the sur-
face which prevented any bonding. The
porous material introduced considerable
difficulty from gases and/or solutions

reactor

becoming entrapped in the pores and
then forming blisters during subsequent
heating., If this problem can be
solved, the Gerity investigators are
gquite optimistic that a successful
cladding may be plated on. However,
two samples which they gave us and
which appeared to be satisfactory did
not have enough mechanical bond to
allow mounting for metallographic

152

study. They have been asked to give
this phase of the work a2 low prioraity
and to push their electroforming

investigation.

REFRACTORY METALS

A survey of the literature on
molybdenum and niobium has been under-
taken in preparation for future work
with these metals. Simple tests are
being run on available specimens to
confirm or disprove some of the
questionable data.

CERAMICS LABORATORY

T. N. McVay, Consultant

The ceramics laboratory located at
Y-12 has been established as part of
the Metallurgy Division. The group
will consist of nine men under the

direction of Dr. John M. Warde. The

 
 

FOR PERIOD ENDING SEPTEMBER 16, 1951

work of the laboratory is teo be of
three types, namely, basic or long-
term ceramic research, engineering
development, and service research for
other divisions of GCak Ridge National
Laboratory. A considerable amount
of equipment has been received, and it
is expected that a large part of the
laboratory will be in operation by
October 1. ‘ '

Work has already been started on
and the first
teo be studied are alumina-—uranium
oxide compositions. Some work has

ceramic fuel elements,

 

also been done on the use of hot-
pressed beryllia for valve parts to
handle 1ligquid sodium, as 1t has been
found that beryllia hot-pressed at
Argonne National Laboratory was re-
sis tent to sodium at 1500°F.

The Metallurgy Division has a work-
ing agreement with the Electrotechnical
Laboratory of the Bureau of Mines,
Norris, Tennessee, whereby the latter
will fabricate crucibles and other small
refractory parts for OBNL and will
Carry on some engineering development
projects 1n the nuclear energy field.

153
 

ANP PROJECT QUARTERLY PROGRESS REPORT

13. CHEMISTRY OF HIGH-TEMPERATURE LIQUIDS

W. R. Grimes, Materials Chemistry Division

Research in the ANP Chemistry
Group has, as indicated by previous
reperts in this series, been con-
cerned almost entirely 1n the past
with development of liquids for use
as aircraft reactor fuels. While
such studies stil]l constitute an
appreciable fraction of this effort,
the program has been broadened 1nmn
recent months to include study of
nonmetallic liquids for use as
moderators and/or heat-transfer
fluids (coolants) for such a re-
actor. These researches have 1in-
cluded a small number of physical
property determinations, as well as
the phase-equilibrium and thermal-
stabi1lity studies necessary to
demonstrate the chemical properties
of the liquids.

Nine fluoride-fuel systems, both
ternary and quaternary, may be singled
out as covering a useful range of
uranium concentration {up to about
200 1b of uranium per cubic foot)
and with satisfactory melting points
(below 550°C). UF,;, which might be
used as a fluoride-fuel component,
has been prepared with only 0.8 wt %
of impurities. In the development
of homogeneous reactor fuels, solu-
tions of UO; in NaOH-Na,B,0, show
promise, together with the NaOH-L10H
system previocusly developed.

The moderator-coolant program
requires a stable (at 800°C) hy-
drogenous liquid with good heat-
transfer properties. Yhile sodium
hydroxide, a satisfactory moderator,
1s an acceptable heat-transfer
medium, the existence of a corrosion-

resistant container for this fluid
at this temperature has not been
demonstrated. For these studies
quantities of sodium hydroxide
assaying better than 99.8% NaOH by
welght are now routinely prepared.
In addition, potassium hydroxide,
barium hydroxide, strontium hydroxide,
lithium hydroxide, rubidium hydroxide,
and several binary hydrogenous
systems are being considered for this
application.

The development of nonmetallic
coolants has been largely confined
to consideration of non-uranium-
bearing fluoride systems 1n con-
Junction with the fuel development
program. Eleven fluoride systems of
usable liquid range and low cor-
rosiveness are listed, although thear
heat-transfer properties are not
sufficiently well known to permit
evaluation of their usefulness.

FUEL DEVELOPMENT

Research on liquid fuels 1s still
directed toward development of low-
melting solutions of UF, 1in alkala
and alkaline earth fluorides and
toward development of self-moderating
fuels, with the former program having
the major emphasis. Research on the
fluorides has progressed so that 1t
is possible to list nmine liquad
systems, covering a wide range of
uranium concentrations, which have
satisfactory melting points. Ul,,
which 1s a possible fuel component,
has beenr prepared in batch samples

which assay 99.2% UF; by weight,
 

FOR PERTIOD ENDING SEPTEMBER 10,

Solutions of UQ, in NaOH-LiOH and in
NaGH~N32B407both show some promise
as homogeneous reactor fuels.

Low-Melting Fluoride Systems
(J. P. Blakely, L. M. Bratcher, and
€. J. Barton, Materials Chemistry
Division; D. G. Hill, Consultant).
Phase equilibrium studies of systems
containing uranium tetrafluoride have
been continued by the technique of
thermal analysis previously de-
scribed.¢?*®)  Studies of the ternary
systems have been extended to include
four component systems  in an effort
to define systems in which low-
melting regions were available in a
wide range of uranium concentrations.
The system NaF-KF-RbF-UF, is con-
sidered the most promising of the
four—component systems investigated
to date.

At present the nine systems listed
in Table 13.1 below are those of
definite promise as fuels. In ad-
dition, at least three systems con-
taining PbF, have been shown to have
suitable melting points at useful
uranium concentrations. Jt 1s mnot
likely, however, that Pb¥, can be
shown to be stable 1n contact with
structural metals.
in small-scale separation of the

 

(l)C. J. Barton, R. E. Moore, J. P. Blakely,
and G. J. Nessle, ¢ LOW»Me}tlng Fluoride Systems
— Thermal Ana]y31s " Aireraft Nuclear Propulsion
Project (uarterly Progress Report for Period
indigg fugust 321, 1950, OBNL-858, p. 110 (Dec.

(z)B. E. Moore, G. J. Nessle, J. P. Blakely,
snd C, J. Barton, “Low-Melting Fluoride Syscems,”
Aircraft Nuclear Propulsion Project Quarterly
Progress Report for Period Ending December 10,
1950, ORNL-919, p. 242 (Feb. 26, 1951).

(3)5. p. Blakely, R. E. Moore, G. J. Nessle,
and C. J. Barton, “Phase Studies of Fluoride
Systems,”’ Aircraft Nuelear Propulsion Project
uarterly Progress Report for Period Ending
March 10, 1851, ANP-60, p. 127 (June 19, 1951).

Lapl :
I'he recent success

1951

isotopes of lithium‘*? prompted the

investigation of systems containing
lithium fluoride. From an inspection
of Table 13.1 1t 1s apparent that
lithium fluoride would be a valuable
addition to the list of possible
fuel components. The most valuable
mixtures without LiF would seem to be
the first three listed in the table.

The various fuel systems studied
during the period are discussed
briefly under appropriate headings
below.

NaF-LiF-UF, Previous studies(S$)
had indicated the existence of a
low-melting region in this ternary
diagram around 31 mele % UF, and
16 mole % NaF. Recent data have
permitted construction of the equi-
librium diagram, as shownin Fig. 13.1.
The low-melting region covers a wide
area on the diagram; two separate
eutectic points apparently exist
inside the 550°C contour.

KF-LiF-UF, Substitution of KF
for NalF resulted in higher melting
points in nearly all parts of the
diagram. The areas showing melting
points below 525°C are very small
although the 550°C contour encloses
a large area of the diagram. The
Pqulllbrlum diagram for this system
1s shown 1in Fig. 13.2,

RbF-LiF-UF, The equilibrium
diagram for this system i1s shown in
Fig. 13.3. The 550°C contour in-
cludes uranium concentrations from
20 to 42 mole % while the 500°C

contour includes an area nearly as

 

(4JL. P. Twichell, “Lithiuf~lsotope Sépa-
31.

ration,” ANP-~65, op. cit., p.

(SL‘NaF?LiF*UF4 System,” OPNL~919, op. c¢if.,
p. 246. '

 

~" - 155
 

ANP PROJECT QUARTERLY PROGRESS REPORT

TABLE 13.1

Summary of Promising Fluoride Fuel Systems

 

 

LOWEST MELTING COMP. OF LOWEST
RANGE OF UF, CONC. OF MIXTURE POINT FOUND (°C) MELTING MIXTURE
COMPONENTS MELTING AT 550°C OR LOWER (mole %) (each 110) (mole %)

 

NaF-P\bF-UF4 24-41 500 32.0 UF
23.0 BRbF
45.0 NaF

UF
NaF
RbF
KF

NaF-KF-RbF-UF, 18-41 500 25.
56.
15.

~N o W o

NaF-KF-UF, 26- 30 530 27.5 UF,
26.0 KF
46.5 NaF

KF-LiF-UF, 0-41 520 35.0 UF,
57.0 LiF

NaF-LiF-UF, 15-39 450 31.0 UF,
16.5 NaF
52.5 LiF

RbF-LiF-UF, 21-42 456 35.0 UF,
9.0 RbF
56.0 LiF

LiF-UF, 24-32 480 26.5 UF
73.5 LiF

NaF-BeF,-UF, 0-21 480 10.0 UF,
15.0 Ber
75.0 NaF

NaF-KF-Bel,-UF, 0-25 478 10.
54.
22.
13.

UF4
NaF
Ber
KF

Lty © 9

 

 

 

 

156

 
 

FOR PERIOD ENDING SEPTEMBER 10, 1951

 
 

OFFICIAL USE ONLY
DWG. 12947

 

 

NalF

 

9925°C

 

Fig., 13.1 - Tie System NaF°L1F~EF4.

great. The minimum-melting mixture
(456 + 10°C) convaians 82 wt % UF,
and 7 wt % RbF.

Naf -KE-PbF ,~UF,. This system was
studied only in a preliminary fashion
because of the reaction of PbF, with
metal contaimers at elevated tewpera-
tures., The system appears to be less
valnable than the NaF-KF-UF, system
in so far as high uvrasium concentra-

 

tions are concerned. The lowest
melting point so far demonstrated is
at 493 £ 107C av 20 mole % UF, and
48 mole % NaF, although there is
evidence for another eutectic at
470°C of as vet undiscovered com-
position.

Nef -KF-BeF ,-UF,.  This system has

been studied 1n some detail up to
JO mole % UF, where the melting

157
ANP PROJECT QUARTERLY PROGRESS REPORT

730 °C

KUFg 780°C

750°C A<—750-

UF, OFFICIAL USE ONLY
1035 °C DWG. 12948
1000

 

K3 UF; 970 °C

J/

 

 

900
N 200
735°C 5
130
~800 \)

KF \\\

850°C 490LC 845°C
Fig. 13.2 - The System KF'LI'F*UF‘.

points have become too high for NaF -KF-RbF-UF,. This four-

effective use. The 550°C contour
surface in the pyramidal diagram
extends from 0 to 25 mole % UF,. At
the 10 mole % UF, level the 550°C
contour area enclosesa large fraction
of the triangle, permitting wide
variation 1n the proportions of
NalF, KF, and BeF,. This system may
be of considerable interest 1f low
concentrations of uranium are re-
quired.

158

component system has been studied
in somewhat more detail than any of
the others since 1t provides low-
melting mixtures containing up to
40 mole % UF,. A sketch from the
model of the contour area enclosing
all compositions melting below 550°C
is shown in Fig. 13.4. For this
purpose the solid model was con-
structed using triangular prism
coordinates and was viewed from the
 

FOR PERIOD ENDING SEPTEMBER 10, 1951

UF, OFFICIAL USE ONLY
Gam oG DWG. 12949

 

 

 

720°C
RbF / LiF
795°C 845°C

Fig. 13.3 - The System BRbF-LiF-UF,.

NalF edge of the model. A study of
the data i1ndicates that mixture
melting below 550°C may be prepared
with uranium content varying between
18 and 41 mole %. The lowest meltaing
mixture found in this system (500°(C)
contains 25, 56.3, 15, and 3.7 mole %
UF,, NaF, RbF, and KF, respectively.

Preparation of UF, (W. C. Whitley,
Research Participant; C, J. Barton,
Materials Chemistry Division).

 

Uranium trifluoride 1s of i1nterest
to the Fuel Research Group primaraly
as a possible fuel component but also
as a possible product of radiation
damage to UF,. Since the compound
is5 not readily available, a method
for its preparation 1in a high state

of purity has been developed.

0Of the procedures which have been
described for preparation of this
material, the method using reduction

159
 

ANP PROJECT QUARTERLY PROGRESS REPORT

 

 

 

 

 

 

 

 

 

OFFICIAL USE ONLY
DWG. 12950

 

 

 

 

 

 

e e m‘k‘“"""fi =,
‘-——"‘"‘M %M"""%—-m_.‘__‘
e -
50 — 1~ T80
40 ST b T T | a0
2 S
@ 2
3 o
E &
D .
10 h““‘h—vw_._.__.___ ‘__‘._,__J-—"-“'""gojo
50 ] i
0 30 o b T 30 4O
1O 0 10
RbF (mole %) KF (mole %)
Fig. 13.4 - The System NaF-KF-RbF-UF,.
of UF4 by uranium metal(®) at tem- as yield of UF, 1s concerned. From

peratures above 1050°C seemed to be
best. While the product of that
reaction had been characterized as
UF; by X-ray-diffraction methods, no
determinations of the purity of the
product were made.

Preliminary experiments served to
demonstrate that temperatures
excess of 1000°C were not superior
to the range 850 to 900°C in so far

in

 

(6).]. C. Warf, Uranium Trifluoride — A
Summary Report, AE(D-2523 (March 10, 1949).

160

these preliminary studies products
of varying purity were obtained for
study. Typical material from these
runs was shown to be a black material
of specific gravity 8.83 (X-ray
density 9.06) which was inscluble 1n
all except oxidizing solvents.
Chemical analysis showed the material
to be about 90 wt % UF,.

Experiments 1n cooperation with
personnel from the Mass Spectrometer
Group of the Stable Isotopes Division
showed that 10 to 12 wt % UF, was
left unreacted i1n the product. It

 
L]

 

FOR PERIOD ENDING SEPTEMBER 10, 1951

was shown that at 750°C UF, is stable
under 10°° mm Hg but that at 800°C
the disproportionation into UF, and
metallic uranium 1s significant; the
disproportionation rate increases
rapidly with temperature.

Since 1t appeared that low-tem-~
perature reaction for prolonged
periods with proper precautions to
ensure homogeneity of the reaction
mass and to prevent build-up of re-
action products on unreacted UF,
offered the optimum conditions for
preparationof pure UF,, the following
procedure was adopted: Pure UF, was
vacuum~dried at 450°C for several
hours and ground te pass a 200-mesh
seive. This UF, was charged into a
10-in.-long by 3~in.~diameter stain-
less steel container with the stoi-
chiometric quantity of clean uranium
turnings and several stainless steel
balls. The reagents were again
vacuum-dried at 450°C for several
hours befere reduction of the uranium
to powder by repeated formation and
decomposition of UH;. After final
decomposition of the hydride, the
container was evacuated and sealed.
The contents were homogenized by
grinding with the steel balls. The
reaction was allowed to proceed 16 hr
at 900°C, and, after cooling and
grinding of the contents, the sealed
container was maintained at 900°C an
additional 8 hr.

The product of this reaction has
been shown by chemical analysis teo
be 99.2 wt % UF;. This material is
of sufficient purity for use 1in
phase equilibrium studies. It seems
likely that this material is of
considerably higher purity than any
which has previously been produced
elsewhere. A report on this research
will be i1ssued in the near future.

 

Homogeneous Fuels (J. D. Bedman
and L. G. Overholser, Materials
Chemistry Division). The high
solubility of uranium as UO; in
mixtures of sodium and lithium hy-
droxides has been discussed in a
previous report.¢’’ This solution
might well be of interest as a
homogenecus fuel 1f separated lithium
isotopes were available and 1f the
corrosive liguid could be contained.
A small amount of research has been
done to ascertain whether similar
solubility of uranium can be demon-
strated with other hyvdroxide-bearing
materials.

The data in Table 13.2 indicate
that mixtures of barium hydroxide
with lithium or sodium hydroxide are
of little value in this connection.
It 15 established that, at moderately
high temperatures, the solubility of
uranium 1ncreases with temperature
and with increasing concentration of
lithium hydroxide; the solubility at
750°C, however, never reaches a value
high enough to be of interest as a
fuel.

TABLE 13.2

Solubility of Uranium in Hydroxides

 

 

AMOUNT OF HYDROXIDE
USED {(wt %) TEMP. SOLUBILITY OF

 

 

LiOH | Ba(OH), [ NaOH | (°C) | URANIUM (wt %)
25 75 650 0.1
25 75 750 0.2
50 50 650 0.1
50 50 750 0.3
75 25 650 0.4
50 50 750 0.1

 

 

 

 

 

 

(7)“Sol1ibility of Uranium in Mixtures of
SodigT and Lithium Hydroxides,” ANP-65, op. cit.,
p. .

151
 

ANP PROJECT QUARTERLY PROGRESS REPORT

The solubility of UO; 1n mixtures
of sodium hydroxide and sodium
tetraborate with and without ad-
ditional boric oxide 1s indicated by
the data 1in Table 13.3. It 1s
obvious that considerable amounts of
the tetraborate ion must be present
1f significantly large guantities
of uranium are to be dissolved. The
solubility again shows a sharp tem-
perature dependence. Unfortunately,
the additionof borate in amounts more
than § wt % increases the melting
point of the mixture considerably;
mixtures containing 15 wt % melt at
about 500°C. These studies are to be
extended to include other hydroxide-
borate systems.

TABLE 13.3

Solubility of Uranium in Mixtures
Consisting of Sodium Hydroxide
and Sodium Tetraborate

 

 

AMOUNT OF CONSTITUENT

IN MIXTURE (wt %) TEMP. | SOLUBILITY OF

 

 

Na(QOH NazB407 B,0, (°C) URANIUM (wt %)
95 5 650 0.1
95 5 800 0.1
85 15 800 0.4
80 20 800 0.5
55 45 800 5.0
70 15 15 650 0.9
10 15 15 800 2.7
60 15 25 800 4.7

 

 

 

 

 

MODERATOR-COOLANT DEVELOPMENT

To be effective as a modera-
tor—heat-transfer fluid for the
aircraft reactor the ligquid must
obviously have good heat-transfer
properties, must be thermally stable
above 800°C, and must contain hy-
drogen. Hydrogen-bearing compounds

162

with the required thermal stability,
even without regard to probable
radiation stability, are not numerous.

Development of moderator coolants
1s concerned at present primarily
with preparation of very pure alkali
and alkaline earth hydroxides and
with phase equilibrium studies of
various mixtures of these materials
with other compounds. In so far as
ligquid ranges, moderating power, and
heat-transfer properties are con-
cerned, several of the materials and
mixtures tested could be considered
satisfactory. In addition to the
pure hydroxides several binary hy-
drogenous systems, such as hydroxide-
hydroxide, hydroxide-fluoride, and
hydroxide-borate, are being examined.

Preparation of sodium hydroxide
assaying better than 99.8% NaOH by
weight has been placed on a routine
basis. Pure potassium hydroxide has
not yet been prepared in guantity
although small-scale preparations
using recrystallization from isopropyl
alcohol have been accomplished. Pure
barium hydroxide, prepared by re-
crystallization from aqueous solution
followed by dehydration of the re-
sulting octahydrate, should soon be
available 1n sufficient quantity for
corrosion and physical property
testing. Methods for purification
of lithium and rubidium hydroxides
are sti1ll under study.

Preparation of Pure Sodium Hy-
droxide (I.. G. Overholser, D. E.
Nicholson, F. A. Vingiello, Research
Participant, and C. W. Harrill,
Materials Chemistry Division). Pure
sodium hydroxide has been purified by
a modification of the method used by

 
 

FOR

  

TO APPROPRIATE
COLD TRAPS AND
VROUUM PUMP

GROUND

Fivtz FRITTED
GLADS DISK

_ (k ,
- EW9 RT ' J ‘ %
‘_Ti::T§9uu£TTl“_ ,.“w
Q\’)j)?fi VIGRE AUX
_ : %0 GOLUMN

PERIOD ENDING SEPTEMBER 10, 1951

OFFICIAL USE ONLY
WG, 12954

       

LIPS

 

 

@g
M H ’ § 25/4,
%E) 1 355 BALL dom

J LITERS

g 24{40

 

 

 

3 LITERS
- :
W”
E 0 o e
8
Fig, 12.5 - Scodium Hydroxide Purification Apparalus.
workers at the Carnegie Institute of has dissolved. The hot solutinon 1is
Technology(s) for use 1n corrosion passed through the fritted glass

studies and physical property deter-
minations. The purification steps
are effected 10 the pyrex apparatus
shown in Fig. 13.5. Ethyl aleonhol
{1.5 liters) is charged inte flask B
and refluxed over magnesium plus a
small guantity of 1odine to dehydrate
the aleohol. The dryv alcohol is
distilled over 1nto flask € which
contains 150 g of reagent grade
sodium hydroxide pellets and 1is
‘heated until most of the hvdroxide

 

(8)p. E. Sunyder and J. C. . Kelley, The
Low~-Temperature Heat Capacity of Sediua Hy-
droxide; The Entropy of Sodium Fydroxide at
298. 169K,, Carnegie Institube of Technology,
NP-1627 (Junp 1, 1950).

 

filter to remove
{ sodium

insoluble material
carbonate, sodium chloride,
etc.) and then heated to remove the
aleohol until a thick suspension of
the sodium hydroxide—~—alcoholate re-
in flask D. This flask is
removed to a dry beox, and the slurry
1s sucked fairly dry on a fritted
filter. The solid 1s then
washed with a small volume of cold
dry evhyl alecohal, and the solid
material 1s transferred to a suction
flask. The suction flask 1s heated
to about 130°C

mains

plass

under vacuum for at

least 24 bt until all the ale oholdte
1s decomposed aund 2 dry powder re-
Mains.

163
 

ANP PROJECT QUARTERLY PROGRESS REPORT

Obviously, the atmosphere in the
flasks and in the dry box must be
free of carbon dioxide and water.
This 1s accomplished in the case of
the flasks by introducing purified
nitrogen. This gas 1s used to force
the solution from flask C into flask
D and also 1s bled into flask D to
reduce bumping during the removal of
alcohol. The nitrogen atmosphere
prevents the formation of the brown
polymerization products which form
rapidly i1n the alcohol 1f air 1is
admitted at elevated temperatures.

The results of analysis of a
number of batches of purified sodium
hydroxide are given 1in Table 13.4
along with similar data for the
starting material. The first three
batches were prepared during the
early part of the program. Subsequent
modificationin equipment and technique
resulted in a purer product, as shown
by the results for Nos. 4 through 7.
At present, sodium hydroxide is being
routinely produced with a carbonate
content of 0.15% or less and a water

TABLE 13. 4

Purification of NaOH by Recrystal-
lization from Ethyl Alcohol

 

 

 

 

 

 

TOTAL
ALKALINITY (%) Na2C03 SiO2
MATERIAL (Calc. as NaOH) (%) (%)
Commercial NaCH 97.5 1.5 0.006
96.8 1.4 0.0086
98.0 2.0 0.005
Purified NaCH 1 99.73 0.31
Puri fied NaOi 2 99.81 0.36
Purified NaOH 3 99.83 0.31
Purified NaCil 4 99.90
Purified NaCH 5 99.98 0.13
Purified NalH 6 100.0 0.16 0.008
Purified NaQH 7 99.88 0.07 0.007

content of 0.1% or less. The silica
values show that there 1s virtually
no pickup of silica in the process
used. The yi1eld per batch 1s ap-
proximately 65 g, anda total of about
5 1b of purified sodium hydroxide
has been prepared by this method.

The results of several runs in-
dicate that pure sodium hydroxide
may also be obtained by removal of
carbonate from a concentrated aqueous
solution of the hydroxide, followed
by dehydration at 450°C under vacuum.
For this method a 50 wt % solution of
sodium hydroxide is made up in a
wax-lined vessel, the suspended
sodium carbonate 1s removed by
firltration, and the clear filtrate
1s dehydrated in a nickel vessel,
first at 200°C and finally at 450°C

under vacuum.

The results of the analysis of
sodium hydroxide purified by this
method are given in Table 13.5. A
comparison of these data with those
given previously indicates that the
hydroxide may be about as effectively
purified by this method as by the
ethyl alcohol process. This method
may be applied to quantity production
more easily than the ethyl alcohol
process.

TABLE 13.5

Purification of NaOH by Re-
crystajlization from H,0

 

TOTAL
ALKALINITY (%) [Na,CO, |Sio,
(Calc. as NaOH) § (%) (%)

 

 

164

Purified NaOQH 1 100.0 0.11 0.005
Purified NaOH 2 99.87 0.20 0.004
Purified NaQH 3 99.99 0.16 0.007
Purified NaQll 4 99.71

Purified NaOH 5 100.0

Purified NaQHl ¢ 00.78

Purified NaOH 7 99.94 0.15

 

 

 

 

 
 

FOR PERIOD ENDING SEPTEMBER 10,

Preparation of Other Hydroxides
(D. E. Nicholson, C. W. Harvill,
and D. R. Cuneo, Materials Chemistry
Division; F. A. Vingiello, Hesearch
Participant, Materials Chemistry
Division; and BR. P. Metcalf, Physics
Division). Preliminary studies
have shown that potassium hydroxide
cannot be purified by the ethyl
alcohol process used for sodium
hydroxide. Apparently the potassium
hydroxide~~alcoholate i1s extremely
soluble in ethyl alcohol and also
is relatively low melting and probably
more stable than the cerresponding
sodium compound. Limited success
has been obtained by using iscpropyl
alcoheol, and a study of the use of
this compound and of other alcohols 1s
underway.

Equipment 1s being set up for the
purification of strontium and barium
hydroxides. The method used entails
separation of the insoluble carbonate
in aqueous solution, recrystallization
as the octahydrate, and finally de-
hydration under carefully controlled
conditions. The secup, which is
constructed of nickel, should permit
purification of pound batches of the
hydroxides. - '

Purified lithium hydroxide mono-
hydrate, availlable commercialiy, was
debhydrated at 200°C under vacuum.
The dehydrated material so obtained
contains approximately 0.1% lithium
carbonate and less than 0.,1% water.
Rubidium hydroxide, reported to be
99.5% pure, was found to contain
approximately 5% carbonate and 15%
water. The dehydration of this
material is being investigated at
present.

Decomposition Pressures
droxides (E. O.

of Hy-~

Price, Research

 

1951

Participant, Materials Chemistry
Division). Experimental work has
been confined to a study of the
decompesition pressures of barium
hydroxide hydrates and anhydrous
barium hydroxide. The existence of
the monohydrate and octahydrate has
been established by numerocus ex-
perimenters, but there appears to be
some doubt regarding the existence of
the trihydrate. The decomposition
pressures for the following systems
have been reported:(%+19)

Ba(OH) , H ,0(s) fing, Ba(DH) ,(s) + H,0(g)

Ba(OH) ,* 81,0(s) ;::“.Ba(OH)£H20(s) + THO ()

These references also include com-
parable data for the strontium hy-
droxide systems. It mlght be noted
that for anhydrous barium hydroxide
the decomposition pressure reaches
1 atm at about 1000°C and the cor-
responding temperature for anhydrous
strontium hydroxide 1s 700°C. The
hydrates exert too high pressures at
elevated temperatures to be of very
much practical value.

Pressure measurements have been
made on barium hydroxide systems by
means of two different types of ap-
paratus. One type uses a nickel
chamber equipped with a diaphragm
for containing the sample, and the
pressure exerted against the diaphragnm
is measured by balancing against a
known pressure on the opposite side

 

£
(J)P M. G. Johnson, ‘“Der NDanpfdruck ven
%{ockr;em Salmiak, Z. phystk. Chemn. 61, 457
907 : :

(lo)b Tamaru and K. Siomi, *‘Dissoziation
von Sr((}H)z, Bafm)z und Hydraren, bestimmt mit

Hochtemperatur-Vakuumwaage," Z. physik. Chem.
A1T1, 221 (1935).

165
 

ANP PROJECT QUARTERLY PROGRESS REPORT

of the diaphragm. This instrument
was used for measuring the pressure
of a sample of barium hydroxide ap-
proximating Ba(OH),-(1.9)H,0 at
relatively low temperatures. Values
obtained are given in Table 13.6.
These values are 1n fair agreement
with those reported in the literature.
Measurements were not made at higher
temperatures because of difficulties
in temperature control.

TABLE 13.6

Decomposition Pressures of
Ba(QH) ," (1.9)H,0

 

 

 

TEMPERATURE | PRESSURE || TEMPERATURE | PRESSURE
(°C) (mm Hg) (°C) (om Hg)
44 6.6 69 19.7
45 6.8 73 21.2
51 8.1 76 24.5
52 9.3 78 25.4
54 10.6 82 27.8
66 16.3

 

 

 

 

 

The other type of apparatus
employed a tin manometer,using
nitrogen to balance the manometer.
Experiments with this apparatus
using barium hydroxide containing
about 2% water showed no measurable
pressure at 370°C. For material
corresponding to barium hydroxide
monohydrate, a pressure 1n excess
of 1 atm at 390°C was observed.
These measurements are all of a
preliminary nature and additional
work must be done before any con-
clusive results will be forthcoming.

Binary Hydroxide Systems (K. A.
Allen and W. C. Davis, Stable Isotopes
Research and Production Division;
C. J. Barton and J. P. Blakely,
Materials Chemistry Division)., The

166

system Ba(OH),-Sr(OH), has been
studied in some detail by thermal-
analysis methods because these
materials have been reported to be
somewhat less corrosive than the
alkali hydroxides. Figure 13.6
shows the completed phase diagram
for this system. There is evidence
for a solid transition in the com-
position range near the eutectic.
While these researches were made with
dehydrated commercial materials
which contained some carbonate, 1t
is safe to say that mixtures of
usefully low melting point are
available in this system., It 1s
likely, however, that the decom-
position pressures are relatively

high at 1500°F.

The system L10OH-NaOH has been
examined because certailn mixtures
of these materials dissolve ap-
preciable quantities of UO;. The
equilibrium diagram is shown in
Fig. 13.7. The low-melting eutectic
(219°C) contains 27 mole % LiOH;
there 1s evidence of a compound
NaOH*LiOH with incongruent melting
point. The thermal effects observed
at about 180°C are probably due to a
solid transition.

Phase diagrams for the systems
NaOH-KQOH, NaOH-RbOH, and RbOH-KOH
have been reported. (1) Phase
studies of the other possible com-
binations will be deferred until
more pressing problems are finished
or until it becomes clear that one of
the systems offers appreciable ad-
vantages from the corrosion stand-
point.

 

(11)G. v. Hevesy, “Uber Alkalihydroxyde. 1I,”
Z. physik. Chem. 73, 667 (1910).

 
 

FOR PERIOD ENDING SEPTEMBER 10, 1951

OFFICIAL USE ONLY

 

 

 

 

 

- DWG. 12952
350 T | 1 I
500 |

o

W 450

=

4

v

& 400

=

i

b

350

& Q e . ®
Sr(OH), 10 20 %0 40 50 60 70 80 90  Ba(OH),

Ba{OH), (mole %)

Fig. 13.6 - The System Sr(fl“)z-na(OH)z.

Hydroxide-Fluoride Systems (W. C.
Davis and K. A. Allen, Stable Isotopes
Research and Production Division;
C. J. Barton and J. P. Blakely,
Materials Chemistry Division).
Scarpa®!?) has published studies of
the NaOH-NaF and KOH-KF systems.
According to his experiments the
former binary shows the formation of
mixed crystals with a solubility
break with the melting points of all
the mixtures intermediate between
those of the pure components. The
KOH-KF system 1s similar, with melting
points intermediate between those of
the pure components but with the

 

(12)g, Scarpa, ‘‘Thermal Analysis of Mixtures
of Alkali Hydroxides with the Correspondin
Halides,” Atti acced. Lincei 24, [, 738 an
955 (1915). '

 

compounds completely miscible in the
solid state.

These data have not been com-
pletely confirmed in this laboratory.
The incomplete study of the KOH-KF
system seems to show a eutectic at
about 40 mole % KF melting at 630°C.
Itis apparent, however, that materials
with a useful melting point may be
obtained only at high concentrations
of hydroxide; these mixtures would
probably be similar to pure hy-
droxides in so far as corrosion is
concerned.

Addition of NaOH to the termary
LiF-NaF-KF eutectic to the extent of
10 mole % can be accemplished without
great elevation of the melting point.
This system and the analogous system

167
'

ANP PROJECT QUARTERLY PROGRESS REPORT

OFFICIAL USE ONLY
DWG, 12953

 

500 I s r I

450

400

RATURE (°GC)

TEMPE
1o
o
&
q

250

] . | | |

 

 

200

 

 

 

 

 

NaOH 10 20 30 40

 

50 &0 70 80 20 L.iOH

LiOH (mote %)

Fig. 13.7 - The System NaOH-LiQH.

using KOH will be examined in more
deta1l.

lydroxide-Barate Systems (K. A.
Allen and W. C. Davis, Stable Tso-
topes Research and Production Divi-
sion; C. J. Barton and J. P. Blakely,
Materials Chemistry Division). These
materials are of 1nterest because of
their solvent action on UQ, and also
because of the possibility of re-
ducing considerably the corrosiveness
of the alkali by incorporation of the
borate. The data on the sodium
hydroxide~—sodium tetraborate system

168

indicate a eutectic at 95 wt % of
sodium hydroxide. The liquidus
cunrve climbs steeply from this
eutectic and shows a break at the
composition corresponding approxi-
mately to the orthoborate.

COOLANT DEVELOPMENT

Recently 1initiated studies leading
to development of nonmetallic coolants
have been largely confined to studies
of the alkali and alkaline earth
fluoride systems. While mixtures of

 
 

" FOR PERIOD ENDING SEPTEMBER 10, 1951

useful liquid range and low cor- TABLE 13.7
rosiveness are in hand, the heat-

. . : Low-Melting Non-Uranium Fluoride
transfer properties are not yet known

 

 

. .. .. . Eutectics

with precision sufficient to permit ~

evaluation of their usefulness. Table

13.7 lists the several fluoride MELTING

systems which are discussed in the COMPOSITION (mole %) POINT (°C)

literature or which have been studied : '

in the course of the fuel development 50% KF - 50% LiF . 492

program and which show melting poants

low enough to be of interest to the S6% KF - 447% AlF, 5?0

program, 60% KF - 40% PbF, 470
33% NaF - 67% PbF 505

It should be noted that these . 2

materials are largely by-products 30% LiF - 70% PbF 460

of the fuel research, and 1t 1is 58% NaF - 42% BeF, 342

likely that a concentrated effort 42% KF - 58% BeF 315

will serve to add to this list and 2 | |

to produce materials of lower melting 40% ROF - 607% BeF, 435

point and viscosity. A program of 48% LiF - 52% BeF, , 360

study of other fused salt systems 46.5% LiF - 42.0% KF - 11.5% NaF 454

which are likely to be better heat- _

transfer agents 1s underway. 30% NaF - 5% KF - 65% BeF, 345

 

 

169

 
 

ANP PROJECT QUARTERLY PROGRESS REPORT

14.

RADIATION DAMAGE

D. S. Billington, Physics of Solids Institute

A. J. Miller, Research Director’s Division

Radiation damage and radiation-
induced corrosion in the prime con-
stituents of the airplane reactor core
have been under examination during the
past few months. The X-10 graphite
pile, the Y-12 cyclotron, and the
Berkeley cyclotron have been utilized
as radiation sources., The limitations
in the significance of the information
being obtained are due mainly to the
low flux i1n the graphite pile and the
lack of methods for correlating effects
of accelerator particles with those of
pile radiations. In addition, the
Berkeley cyclotron has a low beam
current, and the Y-12 cyclotron needs
a more uniform beam and efficient
target~cooling mechanisms that will
allow it to be used for irradiating
with the high beam
While
using the presently available sources,
the experimental groups are expending

the specimens
currents that are available.

the maximum effort to design and con-
struct equipment suitable for use 1in
the LITR and MTR and to use higher
beam currents 1in the Y-12 cyclotron.
In the following description of the
experimental procedures and the re-
sults to date, the significant items
are the stability of the molten {fuel
in i1nconel capsules during graphite
pile and cyclotron irradiation, the
small changes 1n creep rate of 347
stainless steel during graphite pile
and the large decrease
thermal conductivity
in the

irradiation,
in inconel
detected
graphite pile.

in one experilment

170

CREEP UNDER IERRADIATION
J. C. Wilson J. C, Zukas
W, W, Davis

Physics of Solids Institute

A series of three fairly reproducible
cantilever creep tests conducted 1in
the ORNL graphite reactor exhibited
increased total creep strain,
to bench tests,

compared
after approximately
100 hr of operation. Up to this test
time there appears to be slightly less
creep under irradiation than on the
bench, The 1n-pile specimens show
about 10% greater creep strain after
200 hr and nearly 20% more strain than
the corresponding bench test after

about 250 hr, Type 347 stainless steel
was the metal tested, the maximum fiber
stress was 1500 psi, and the tempera-

ature was 1500°F.

The creep of 347 stainless steel
at 1500°F and a stress
of 1500 psi has been measursd by the
cantilever creep apparatus in the X-10
pile. The measurements previously
reported(!') have been extended, and to
date three fairly reproducible 250-hr
cantilever creep tests have been made.

maximum fiber

Up to 100 hr the creep under irradi-
ation is slightly less than that of

 

(). ¢. Wilson, J. C. Zukas, and ¥. W. Davis,
“Creep Under Irradiation,” Aircraft Nuclear
Propuflion Project Quarterly Progress Report for
Period Ending June 10, 1951, ANP-65, p. 200,
201 (Sept. 13, 1951}.

esp. p.

 
 

FOR PERIOD ENDING SEPTEMBER 10,

the unirradiated control specimens.
However, the in-pile specimens exhibit
about 10% greater elongation after
200 hr than the control specimen, and

nearly 20% greater elongataon after
250 hr.

A compesite plot of all the creep
data i1s given 1in Fig. 14.1. The
reproducibility 1s good except for the
test represented by cnrve O, which has
been disregarded in this analysis of
the data.  The diminished creep rate
observed during pile shutdowns (plotted
on a larger scale in the lower right-
hand corner of Fig., 14.1) substantiates
the conclusion drawn from the total
extensions, namely, that irradiation
increases the total creep strain after
the initial stages of the creep history,
1t was also the main reason for the
summary disregard of curve O in
analyzing the data., 1f the divergence
between the bench and in-pile curves
then the difference in

18 real, creep
strain should be guite large, per-
centagewise, after longer times.

a set of specimens will

for 500-hr periods and
the total strain measured after removal
from the reactor, In this way any
possible errors in strain measurement
because of radiation effects on the
microformer will be eliminated,. A

motre complete description of all the
2)

Accordingly,
be irradiated

above work has been given,

The observation of diminished creep
at shorter times is in agreement with
our earlier work.®) QOther workers{4®
have reported only decreased creep
under irradiation, but in the tests

 

(2)Phy3tC5 of Solids Institute Quartef!y
Progfess Report for Period Fndzng July 31, 1851,
ORNL- 1148 (in press).

(S)Ph sics of Solids Institute Quarterly
Progress geport for Period Ending April 30, 1951,
ORNL- 1095 (in press).

 

1951

reported here the duration of the test
is much longer and the total strains
much qmaller than 1n any other reported
creep tests under irradiation. This
could, in view of the ignorance of the
mechanisms accounting for creep defor-
mation at slower rates, be responsible
for the apparent difference in results.

These creep measurements under
irradiation will be supplemented by a
tensile creep apparatus which is being
designed for use with the LITR. This
tensile apparatus will provide amn
independent check of the cantilever
creep apparatus measurements of creep
under irrvradiation. Anocther creep
apparatus to be operated in the LITR
will provide stress-corrosion creep
data. Thisg apparatus, now in the
design stage, is basically a wmodified
sodium loop.

RADIATION EFFECTS ON THERMAL
CONDUCTIVITY

A. F., Cohen, Physics of
Solids Institute

Preliminary in-pile tests have been
made on one sample of inconel, (%2 ‘The
thermal conductivity at 820°C decreased
by a factor of 2 1in less than one week
of exposure in the X pile and then
maintained a fairly constant value.
The effect of an in-pile temperature
anneal (by increasing the temperature
of the specimen 70°C for 40 min and
then returning the specimen to 820°C)

 

(4)H P. Yookey et al,, Effect of Cyclotron
Irradiation.on Creep of Alumtnum Nnrth American
Aviation report NAA-SR-121 (June 8, 1951).

(S)Quarterly Progress Report for the Period
January 1.~ March 31, 19851, NEPA-1792.

(6)A. F. Cohen, "Radiation Effects on Thermal
Conductivity,” ANP-65, op. cit,, p. 205.

171
 

 

    

 

 

 

 

 

i~ SECRET =
™o PS|-B-180 o
DWG - 12450R!
| | % =
0.120— ] ] =
St
0 N M g
L l T ]
|\ ' g
O!GO"""— i i —_— ;
INDICATES PILE ” Q
- SHUTDOWN PERIOD /e _
£ s - =
< =
e
0.080}— ] _g
— =
< g
Z &
- W
©0.060— — w
—
i _.0.100+ =
0 L ;g ] ;g
z z 5
wl O
o = mad
0.040— 5 —
£0.0901
wl
_ ld —
=
<
0.020}— & _ —
_ 0.080 O/;LREFER.I‘ENCE SLOPE
7 B0 160 170 18O 190 _
0 | | |
50 100 150 200 250 300
TIME (hr)

Fig. 14.1 - Composite Plot of Three Bench (0) and Three In-Pile (0) Cantilever Creep Curves.
347 stainless steel at 1500°F; 1500 psi maximum fiber stress; irradiated in hole 15 of ORNL graph-
ite reactor, fast flux & 4 X 10'° neutrons/cm?/sec, On the ordinate, 0.001 deflection = 15 x 10-6
in./in. strain.
 

FOR

was to cause annealing out of approxi-
mately 20% of the damage (i.e., the
decrease in thermal conductivity).
With further irradiation the thermal
conductivity returned to the former
equilibrium value. Experiments planned
to give information on the flux de-
pendence of the thermal conductivity
are in progress on types 310 and 316
stainless steel and on pure nickel.

IRRADIATION OF FLUDRIDE FUELS
G. W. Keilholtz, Materials

Chemistry Division

The effect of radiation on the
stability of the molten fluoride fuel
contained in inconel is currently under
investigation using the X-10 pile and
Y-12 cyclotron. Detailed analyses
of irradiated samples are not yet
available, Preliminary spectrographic
analysis of the first cyclotron-
irradiated capsule showed no detectable
radiation-induced corrosion., Capsules
containing fuel prepared with uranium
enriched 1in U%35 showed no evidence of
pressure build-up under pile irradi-
ation. ‘

Pile Irradiation of Fuel Capsules
(J. G. Morgan, C. C. Webster, P. R.
Klein, R. L. Cooper, D. F. Weekes, and
D. D. Davies, Physics of Solids In-
stitute). JInconel capsules containing
fuel prepared with normal uranium and
with uranium enriched in U?%33
tested for pressure build-up due to
fuel decomposition in ‘the X-10 pile.
The fuel composition was gtandard
NaF-KF-U¥F,, 46.5-27-26.5 mole %, The
containers were pressurized to 58 t 1
psi with helium at 1500°F in the pile.
During 200 hr of irradiation there was
no detectable increase 1in the pressure.
Figure 14.2 shows the construction of

were

 

PERIOD ENDING SEPTEMBER 10, 1951

S Y 4gad

 

 

14.2 -
tainer with Thermocouple Well and
Pressure Fitfings.

Fig. Liguid-Fuel Sample Con-

the inconel sample container wath
thermocouple well and pressure fittings.
Examination of the capsules and their
contents for radiation-induced cor-
rosion and for fuel decomposition 1s

in progress,

Cyclotron Irradiation of Fuel
Capsules (W. J. Sturm and M. J.
Feldman, Physics of Solids Institute;

R, J. Jones, J. 8. Luce, and C. L.
Viar, Electromagnetic Research Di-
vision). The cyclotron experiments

173
 

ANP PROJECT QUARTERLY PROGRESS REPORT

have been directed toward the determi-
nation of the effects of proton bom-
bardment¢’? upon the inconel capsules
containing the standard NaF-KF-UF,
ARE fuel,

Four capsules were bombarded with 2
to 3 pna of 20-Mev protons which pro-
vided an energy dissipation of approxi-
mately 20 watts per cubic centimeter

of fuel. Two l-hr and twa 4-hr runs
were made, Radiation cooling was
employed, The first capsule melted in

but 1n
temperatures were
maintained 1n the region 1300 to
1350°F. A partial analysis has been
made on one of the l-hr samples. No
significant changes in fuel composition,
trace element content, or corrosion
rate due to i1rradiation have been
detected., However, some grain growth
has been detected i1n the inconel from
both the irradiated and an unirradiated
control sample.

the region of intense beam,
subseguent runs

One capsule was mounted on a water-
cooled block and irradiated with a 20-
pa beam current for 1 hr at 1500°F,
This capsule has not yet been examined
owing to 1its high radioactivity.

CORROSION OF IRON BY LITHIUM UNDER
CYCLOTRON IRRABPIATION

North American Aviation, Inc,

W, W. Parkinson, Physics of
Solids Institute

Bombardments of irom 1mn contact
with molten lathium have been carried
out in the Berkeley 60-1in.
to provide information

cyclotron
about the

 

(")G. W. Keilholtz, J. C. Morgan, H. Robertson,
and C. C. Webster, “Irradiation of Fluoride
Fuels,” ANP-65, op. c¢it., p. 207.

174

behavior of structural materials under
irradiation while in contact with
liguid metals,

Corrosion data on the first three
capsules irradiated are given in
Table 14.1. There is no evidence of
accelerated corrosion due to 1irradi-
ation.

Micrographs were made of cross-
sections 1n the i1rradiated and un-
irradiated zones of the capsule from
run 1. These micrographs(8) show
that in some areas, apparently dis-
tributed at random through the irradi-
ated zone, an 1increase 1n graln size
has taken place. No conclusions have
been drawn yet as to the nature of
this change. The microhardness was
found to have an average value of 89.5
Knoop hardness numbers in the irradi-
ated zone and 89.3 in the unirradiated,
compared with an average of 106.5 over
the whole surface before bombardment.

Additional irradiations of molten
lithium 1n iron capsules have been
carried out, and the analytical re-
sults are being obtained. This will
complete the study of lithium 1in 1iron
for the present time. The next series
of irradiations will be made on molten
fluoride fuels in inconel capsules.

LIQUID METALS IN-PILE LOOP

C. D. PRaumann 0. Sisman
R. M. Carroll W. W. Parkinson
C. Ellis
Physics of Solids Institute

Lithium at 1000°F has been circu-
lated in the X-10 reactor for one week.

 

(8)

Fig. 14.5, ANP-65, op. ctit., p. 212.

 
 

FOR PERIOD ENDING SEPTEMBER 10, 1951

TABLE 14.1

Results of 31-Mev Alpha Irra@iatiun of Lithium in Iron Capsules

 

 

TEMPERATURE |  TIME AT EXPOSURE |Fe TN Li CHARGE®) |RADIOACTIVITY OF CHARGE(®)
AUN NO. (°C) TEMP. (hr) | (pa-hr) (%) counts/min)
1 625 9.0 25, 1 0.008 2
2 910 6.2 13.6 0.23(c) 2700¢¢)
3 675 17.3 3.8 9.004 - 0.006 2
Control 906 5.7 None 0.004 - 0.006 None

 

 

 

 

 

 

(®)Surface to volume ratio, 5.1 cm’/cm® with 1. 24 g of lithium in each capsule.

(b)Activity measured on step 1 of Berkeley Scaler, model Decimal 2000.

() eak developed in capsule. Jron oxide may have flaked into charge.

A detailed description of the experi-
mental equipment was presented in the
last guarterly report.(®) The princa -
pal components of the apparatus are a
Y-in.-1.d. 316 stainless steel loop
which extends to the center of the
pile connected to a 3/l6-in.-1.d. loop
external to the pile, an electro-
magnetic flowmeter, and three i1oni-
ration chambers tomeasure the activity
at various points on the external
circuit of the lithium loop. The
srimary purpose of this experiment was
te study the Bremsstrahlung activity

 

'(g)g, 0. Baumann, R. M. Carroll, and O. Sismén,
Liguid-Metals In-Pile Experiment,” ANP-65,
op. cit,, p. 211. , '

 

from the decay of 1i% created by the
{n,») reaction on Li’. The analysis
of this activity is reported in Sec.

9, "Nuclear Measurements,"

A cursory visual examination of the

tube from which the lithium had been
removed showed no signs of damage.
Also, there was no evidence found of

radicactive corrosion products in the
Yithium.  Forther investigation will
be made when the tubes hecome somewhat
less radicactive. Rather than pursue
the corrosion due to lithium any
further at this time, an inconel loop
containing sodium is to be operated
at temperatures up to 1500°F,

175
 

 
 
 

 
 

15.

Nuclear Development Associates,

As a result of an extensive analysis
of the supercritical water reactor
first proposed in WASH-24,(?) Nuclear
Development Associates has advanced
detailed specifications for a feasible
desi1gn, The reactor and shield com-
prise an ll-ft-diameter sphere of
water with a 2.5-ft square-cylinder
active core at its center., The reactor
will deliver 400 megawatts
maximum wall temperature of 1290°F and

with a

a fuel investment of approximately 25
1b. The proposed fuel elements are of
the sandwich design considered in the
last quarterly report,(?) and the core
design utilizes a flat neutron flux.
The pressure shell introduced some
problem of weight, but the overall
shield weight, 180,000 1b, remains well

within specifications,

OUTLINE OF A SPECIFIC DESIGN

Apart from some insertions {(pressure
shell, thermal shield for pressure
shell, gamma-ray shadow shielding cap,
flow headers, etc.) the reactor and
its shield may be approximately
visualized as an 1ll-ft-diameter sphere
of water, in the interior of which are
spaced some 200 parallel fuel assemblies

to form a 2.5-ft sgquare-cylinder

 

(D Abstracted from NDA Quarterly Report on ANP
Activities, June 1 to August 31, 1951, Y-F5-55.

(2 )Alrcraft Reactor Branch of USAEC, Appli-
cation of o Water Cooled and Moderated Reactor te
Aircraft Propulsion, WASH-24 (Aug. 18, 1950).

(S)Nuc]ear Developmwent :Assccisates, Inc.,
“Supercritical-Water Reactor,” Aircraft Nurlear
Propulsion Project Quarterly Progress Report .
Period Ending June 10 1951, ANP-65, p. 2 8
esp. p. 220. fgept 19515

 

SUPERCRITICAL

WATER REACTOR (D

Inc.

central core. Fach fuel assembly is =2

tube about an inch square, in 'the
itnterior of which 1s a fine-scale
structure of fuel-bearing plates and

water coolant passages, The fuel
assemblies occupy about one-fourth of
the core volume, the remaining space
between them being filled with moderat-
spacing between
at the axis of the
and increases radially cutward*
manner as to keep the thermal-

ing water, The
assemblies 1s least
core,
in such
flux approximately the same
for all the assemblies. Thus the heat
load is the same for all the tubes,
the stainless steel heat transfer
surface 1is

neutron

being used to maximum

effectiveness in all the tubes, and
the amounts of stainless steel and
fuel required 1n the reactor are

minimized.,

Two water flows may he dis
in this (1) the high-
speed flow throogh the fuel assemblies,
which picks up the fission recoil
(2} the slow flow of the
moderating water between the assemblies,
which picks up a smaller amount of
heat, principally from neuvtrons and
gamma rays. The relative speed of
these two flows can be adjusted by a
15.1.

shim

tinguirshed
arrangement

heat, and

valve as 1ndicated in Fig. This
has desirable features for con-
trol since it permits varying the
temperature rise {and thus the density)
of the moderating water without any

temperature or pressure changes in the

 

&

The cross-section of the assemblies may have
te be decreased, and their number increased, near
the edge of the core to avoid neutron lasses from
B too coarse moderator structurs. This WOJ]d
make the number of tubes larger than 200.

179
ANP PROJECT QUARTERLY PROGRESS REPORT

SECRET
DwWG. 12956

MODERATOR WATER INLET

|

 

  

PRESSURE SHELL

   
   

OUTLET
HEADER

 

 

EXIT TO
TURBINES

T =98C°F

  
 
  
  

 

 

 

 

 

T

FUEL ASSEMBLY

 

INLET FROM

PUMPS
7= 480°F j;aBERATDR WATER OUTLET
Fig. 15.1 - Supercritical Water

Reactor Flow Arrangement,.

Table
15.1 summarizes some of the numerical
values for the example selected,

external power plant system.

FUEL ELEMENTS AND ASSEMBLIES

Further consideration of the fuel-
element possibilities discussed 1in
the last quarterly report¢3’ has led
to the selection of an oxide type
element, The use of powder metallurgy
technigques to form a fuel sandwich
composed of a sintered UO, and stain-
less steel powder layer rolled between
two pieces of stainless steel seems
particularly attractive. The original
work at ORNL‘3***’ produced successful
sandwiches of this type with an over-
all thickness of 30 mils; however,
other research has indicated that this

 

(4L‘Powder Metallurgy — Fuel Element Fabri-
cation,” Metallurgy Division Quarterly Progress

1951,

Report for Period Ending Januery 31,
ORNL-987, p. 52 (June 7, 1951).

180

figure might be lowered considerably,
The powder metallurgy technique dis-
tributes the UO, in separate tiny
pockets in the interior of, in effect,
an otherwise continuous piece of
stainless steel. It appears that there
will be sufficient free volume to
accept the fission gases at moderate
pressure, Thus, pending in-pile tests
of this fuel-element type, we are
inclined toward some optimism regard-
ing 1ts chances for ultimate success-
ful development.

A stainless steel (347 or 316 Nb)
sandwich of 20 mils total thickness,
with the central layer about 10 mils
thick and containing, some 10% UO, by
volume, 1s suitable. There are many
conceivable ways of arranging such
plates into assemblies. Figure 15.2
illustrates one arrangement designed
to give stiffness, maintain spacings,
permit thermal expansions, etc. Here
alternate flat and corrugated fuel
plates are stacked to fill a square
unfueled tube. The contact of the two
fueled walls has been estimated from
NACA-TN-2257 to run 30°F hotter than

the average metal wall temperature,

REACTIVITY

It appears desirable to keep the
thermal-neutron flux radially flat in the
core to minimize the amount of stain-
less steel needed for heat-transfer
purposes, If the amount of fuel
needed was dependent only upon the
amount of steel used, then this flat-
flux design would obviously minimize
the critical mass as well, Furthermore
it 1s actually found, with appropriate
simplifying assumptions, that uniform
thermal flux in the core does accompany
minimum critical mass, Thus flattening
the neutron flux operates directly to.

 
 

FOR PERIOD ENDING SEPTEMBER 10, 1951

TABLE 15.1

Summary of Reactor Design-Point Values

Heat output

Core size

Water inlet temperature

Water outlet temperature

Maximum wall temperature

Fuel plate‘thickness '

Maximum heat flux from fuel plate
Longitudinal maximum/average power ratio
Average heat flux from fuel plate

Weight of stainless steel fuel plates
Percent of unfueled steel added in fabrication
Toral weight of stainless steel 1n core
Average density of water in core(®?
Critical mass in most favorable idealized case(®)
Pressure shell inside diameter (sphere)
Pressure shell thickness

Pressure shell weight

Cooling stream hydraulic diameter
Cooling stream inlet velocity

Cooling stream exit velocity

Cooling stream pressure drop

Water flow rate

Percentage of core volume occupied by:

Stainless steel fuel plates
Additional unfueled stainless steel
Water in high-speed channels .
Low-speed moderating water

400,000 kilowatts(?)
2.5-ft square chifider
480°F

980°F

1290°F (P

0,026 1in.

1,250,000 Btu/hr-ft?
1.4

900,000 Btu/hr- ft?
972 1b

22

700

0.73 g/cm?

10 - 12 kg of U35
4.5 ft

1.5 an.

400G 1b

0,065 in.

12 ft/sec

73 ft/sec

25 ps1

430 1b/sec

]
LA L N O
LN D W

(a)Ihe heat output required for level sgperéonic flight at altitude 1s approkimately 300,000 kilo-
watts in WASH-24 and NEPA-1843; WASH-24 indicdates a maximum power (sea-level) requirement of 398,000

kilowatts, which figure has not been checked by United Aireraft.

(b)lnc}uding allowance of 30°F for hot spots..

(C)When the moderator flow rate and density are as high as possible in Fig. 15.1.

d . . .
( )Unpolsoned water density at the above maximum value, no losses from coarseness of structure, ete.

 

181
 

ANP PROJECT QUARTERLY PROGRESS REPORT

SECRET
DWG. 12957

F

 

 

1.031n.

0.407in. |0,15?in.|

 

 

 

 

 

 

F;—*wwvv*%—wmuogam. ---------- v——fiw——fl

SCALE: ABOUT 245, TIMES ACTUAL SIZE

Fig. 15.2 - Cross-Section of Fuel-
Element Assembly.

minimize the fuel content, as well as
indirectly by minimizing the amocunt of
steel,

The problem solved was the follow-
ing: Given a region containing
moderator and given some fuel (pure
fissionable material, or a definite
mixture of fissionable material and
structure)}, what is the minimum amount
of the fuel and how should 1t be
distributed in the moderator in order
to obtain a ecritical assembly? By
assuming that (1) the reactor is
thermal, and (2) the volume occupied
by the fuel is negligible, 1t was
proved that there 1s a region in the
moderator in which fuel should be
placed, and in this region the thermal
flux should be constant. The fuel

182

distribution can be obtained for
suitably shaped systems.

The above restrictions are well
satisfied for pure U?3% immersed in
an infinite sea of water, heavy water,
beryllium, or carbon. Thus we are
now in a position to calculate the
minimum critical mass of a chain-
reacting assembly moderated by any of
the usual moderators. This calculation
has not yet been carried out. A
calculation for the same power output,
and the results of an illustrative
calculation using the same power
output and with the same assumptions
otherwise, comparing a reactor with
uniform fuel-tube loading with the
flat-flux reactor has been made. Both
reactors had cores 75 c¢m long and 67
cm in diameter and reflectors with
outer diameters of 100 cm. The flat-
flux reactor had 300 kg of stainless
steel. The uniform fuel reactor
required, for the same power output,
1.54 as much stainless steel and 1.41
as much urantum,

STABILITY

Calculations of the equivalent bare
reactor have been expanded to include
the detailed flux shape in the reflected
reactor and a fairly detailed repre-
sentation of the space variations of
flow and heat transfer. First results

have been obtained and will be reported
as soon as they have been analyzed.

The general procedure is to con-
sider first small oscillations from
equilibrium. The periods of the
system are then obtained as the
eigenvalues of the problem. A vari-
ational principle is used to convert
the system of partial differential
equations to a matrix problem, which

 

1y
 

FOR PERIOD ENDING SEPTEMBER 10, 1951

is then handled by methods discussed
in a recent report.(%) 1In this method
of approach the accuracy of the
results for the perioeods and time
behavior 15 limited only by one’s
diligence and ingenuity in applying
the variational principle. :

PRESSURE SHELL

the pressure shell aond the
iayer of heavy

Between
reactor core is a
material to cut down the gamma heat
release in the shell. Both this
thermal- shielding layer and the shell
1tself are of substantially constant
thickness all around the reactor and
represent 4 substantial fraction of the

Since
tend to

total gamma-shield weight,
divided-shield arrangements
concentrate the heavy material at the
side of the reactor facing the crew,
the pressure shell and 1ts thermal
shield reduce the efficiency of the
overall shield design by putting
substantial amounts of material in the
wyrong place.

To keep down the weight associated
with the pressure shell we may (1)
keep the shell nearly spherical and
its radius fairly small, (2) reduce
the thermal-shielding layer until heat
transfer from the shell begins to be
troublesome, {(3) make the shell of
high-strength alloys such as U. 8.
Steel stainless W or Armco 17-7 PH or
17-4 PH, (4) run the pressure-induced
stresses quite high in the shell, or
(5) permit thermal strains large

 

(S)NDA Quarterly Report on ANP Activities,
March 1 to May 31, 1951, OBNL, Y-12 site, Y-F5-47
(June 14, 1951}, :

 

eniough to cause yvielding in the outer
fibers.*

For application in the above-
mentioned gamma heating problem, methods
have been devised for valculating the
the gamma leakage from spherical
reactors with vadial wvarization in
component concentratrons., 1he gamma
build-up processes cccurring inside
the reactor have been included approxi-
mately, and calculations have also
been made of the angular distribution
of the emerging photons. Present un-
certainties in the energy distribution
of gammas produced as a result of
fission prevent accurate calculation
of the leakage; conservative assumptions
have been made, but they could be off
by 50% or more.

SHIELD

Preliminary estimates of divided-
shield weights were made for the super-
critical water cycle for a range of

reactor core slzes, reactor-crew
separation distances, and reactor
powers., The basic shield used was

essentially that described in ANP-
53,(%) modified to include several
inches of 1ron near the core because
of the pressure shell and its heat
shielding. This iren layer was
considered useful shielding material,
since the fast-neutron flux for this
design was considerably smaller at the
pressure shell than the primary gamma
flux. (The thermal-neutron flux
reaching the i1ron was considered
stopped by a boron laver.} Then an

 

*While it a pears unlikely that a ductile
material can faig by thermal stress, the allowable

thermal strains may be limited by practical
considerations, such as embrittlement, temperature
cycling,
gpenings,

strain concentration at hot spots,
or supporbs.

183
 

ANP PROJECT QUARTERLY PROGRESS REPORT

appropriate thickness of lead was
removed from the shadow disks and all
the lead was removed from the sides of
the compartment, * From the
results of this preliminary study,
assuming a 44-ft separationm distance

crew

 

*

Work hasz just started on estimating the
nzutron-indnced activity of the circulating
water. Preliminary estimates indicate that some
of the gsmma shielding wight have to be restored
at the sides of the crew compartmeni for designs
with condensers far out in the wings (as in a
modified B-47, with the engines in pods below the
wings).

184

(center of reactor to vear face of
crew compartment), the following shield
weights for the power and core diameter
of Table 15.1 were estimated:

Core shield weight 58,500 1b
Crew shield weight 21,500
Total shield weight 80,000 1b

The shield weight allowed for in the
airplane design (WASH-24 and NEPA-1843)
1s 91,600 lb.

 
 

FOR PERIOD ENDING SEPTEMBER 10, 1951

16. CIRCULATING-MODERATOR-COOLANT REACTOR: HKF

Atomic Energy Division, The H. XK. Ferguson Co., Inc.

A study of a subsonic anircraft
reactor using circulating sodium
hydroxide as both mederatoer and coolant
and a fixed fuel of UFJ3 dissolved 1in
alkalil and alkaline earth fluorides
has been completed by the Atomic Energy
Divisiocn of The H. K. Ferguson Ce,
under contract to OBNL. As a result
of this study it was concliunded thatr
the circulating-moderator-coeclant
reactor can provide ample power for a
subsonic airplane. Actually the
performance of the aircraft is inter-
mediate between the NaOH homogeneous
and the UBi circulating-fuel reactors
discussed in previous quarterly re-
ports. %37 Principal advantages of
the circulating moderator-ccclant
system over the other two include the
use of a moderator of known properties
and behavior and the fact that the
secondary coolant is not made radio-
active, which simplifies the shielding
problem. The main drawback, in addition
to the fact that a suitable container
for sodium hydroxide isnot yet assured,
is the multitude of metallic heat-
transfer tubes in the core through
which heat is transferred from fuel to
coolant, The present study was con-
fined to the reactor because it was

 

(I)Abstracted from the report by K. Cohen,
Circulating-Moderator-Coolant Reactor for Sub-
senic Aircraft, HKF-112 (Aug. 29, 1951). This
reactor i described in some detail in HKF-112.

(Z)H. K. Ferguson Co., “Homogeneous Circu-
lating-Fuel ' and Circulating-Moderator Reactory
Aircraft Nuclear Propulsion Project Quarterly
Praogress Report for Period Ending March 10, 1951,
ANP-60, p. 308 (June 19, 1951). :

(S)H. K. Ferguson Co., ““Circulating-Fuel
Reactor,” Aircraft Nucleer Propulsion Project
Quarterly Progress Report for Period Ending
June 10, 1951, ENP—GS, p- 224 (Sept. 13, 1951).

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

{1)
SEGRET
DWG. 129588
NOOH FLOWING INLET HEADER ‘
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W 777 7 # 7 d FPZ 77722 L
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NAOH OUT OUTLET HEADER

 

SECTIONAL ELEVATION

Fig. 16.1 - Schematic Drawing of
Reactor Core. :

known at the outset that the inter-
mediate heat exchanger and remainder
of the power plant would be almost
1dentical with those designed for the
homogeneous sodium hydroxide reactor
and that the power requirements are
similar, A schematic diagram of this
reactor core is shown in Fig, 16.1.

OPERATIONAL CHARACTERISTICS

The circulating-moderator-coolant
reactor is designed to provide power

185
-
3
-

 

ANP PROJECT QUARTERLY PROGRESS REPORT

to operate a modified B-52 type plane
crursing at 35,000 ft altitude and
Mach 0.8. The maximum metal tempera-
ture was set at 1500°F, The problems
of a solid moderator are eliminated
and a simpler structure 1s adegquate.
The liguid fuel has the

advantages that it can be drained upon

specified

landing, which permits a large saving
in fuel inventory for a fleet of nuclear
the fuel can expand out of
which provides a stabilizing

aircraft;
the core,
effect for reactor control.

The reactor delivers 140,000 kw at
the design point of 35,000 ft and Mach
0.8, with a coclant outlet temperature
of 1380°F, inlet 1180°F. The maximum
heat flux i1s 530,000 Bru/frt?2-hr, At
sea level 1t provides 230,000 kw.
This 1ncrease 1s accomplished by in-
creasing the temperature rise through
the reactor (inlet 1007°F, outlet
1335°F) while increasing the heat flux
{maximum 870,000), and i1t permits
climb from sea level at 720 ft/min at
a speed of 250 knots.

REACTOR CHARACTERISTICS

The present reactor design has the
characteristics of all fixed-{fuel
reactors: (1) the heat must be trans-
ferred first through the fuel and then
through the canning material; (2) the
coolant flow must be distributed in
accordance with the flux distribution;
and (3) the core must contain a compli-
cated structure subject to thermal and
mechanical stress and to radiation
damage. This reactor utilizes the
type of ligquid fuel planned for use in
the proposed ARE reactor at ORNL, the
essential difference being the use of
hydroxide as the combined
coolant and moderator as compared with
the OBNIL use of a fixed moderator of

sodium

186

BeO and a coolant of sodium. Sodium
hydroxide was specified as the moderator
coolant in the HKF proposal because 1t
is considered to have the best combi-
nation ol properties for this use:
adequate slowing-down power; does not
freeze, decompose, or have a high vapor
pressure 1n the operating temperature
range; and 1ts neutron absorption 1s
not excessive for this type of reactor.

An advantage of this design over
the homogeneous sodium hydroxide
reactor and the circulating-fuel
reactor 1s that the fuel does not
circulate ogutside the reactor during
operation. This tends to decrease the
reactor shield weight, and, more
1lmportant, avoids irradiation of the
heat exchanger and outside coolant
fluid with delayved neutrons. It also
eliminates loss of delayed neutrons
The large number of
common to all
noncirculating-fuel aircraft reactors,
1s a disadvantage compared to the
homogeneous reactor, Probably more
important than the problems and
advantages of the mechanical design
are the materials problems. 1{f ma-
terials resistant to corrosion by
sodium hydroxide with good high-
can be

outside the core.
heat-transfer tubes,

temperature characteristics
found, the circulating-moderator-
coolant reactor appears to be capable
of powering the subsonic aircraft,

CORE DESIGN

The reactor core is a square cylinder
32 in, 1in diameter. The sodium hy-
droxide moderator cooclant flows down-
ward through small, closely spaced
inconel tubes, Fuel occupies the
space between the tubes. A jacket of
sodium hydroxide around the core acts

 
 

FOR

as a reflector as does the sodium
hydroxide in the top and bottom headers.
l.Lead provides a gamma shield to pro-
tect the pressure shell against over-
heating. The sodium hydroxide flows
inside the tubes, thereby ensuring
against hot spots which could be
caused by interruptions in flow, which
would occur if the flow were outside
the tubes.’ This positive flow path
is believed necessary, as sodium
hydroxide has poorer heat-transfer
properties than sodium, for example.

The fuel was designed to be drain-
able, in order to decrease greatly the
uranium inventory for a group of planes.
This advantage, however, brings with
it a serious leakage problem, Since
the fuel 1s in a common container, it
could all escape through a large leak.
The effect of small leaks i1s minimized
by balancing pressures so that coolant

leaks into the fuel, rather than the
reverse.
REACTOR PHYSICS
The uranium weight is high, 187 1b

of U**® in the core. This is mostly

 

PERIOD ENDING SEPTEMBER 10, 1951

because of the large amount of inconel,
necessary for heat-transfer surface.
The reactor could not be much smaller
without exorbitant uraniunm content.
The uranium could be somewhat
by going to larger sizeg, at the
expense of increased shield weight,

reduced

The sodium hydroxide reflector
brings up the thermal-neutron flux,
and therefore power production, in the
The effect
toward the
center as would be desired, owing to
the large absorption in the reactor,
and the effect on leveling out the
neutron flux is only moderate.

core near the boundary.

does not extend as far

The amount of fuel 1n the middle of
the reactor is varied to give a total
shim control of 10% reactivity. For
inherent temperature stability the
fuel in the end zones, which is B0% of
the total, can expand to chambers out-
side the reactor. No provision is made
for xenon override, as draining and
replacing the fuel after shutdown makes
1t unnecessary.

187
 

ANP PROJECT QUARTERLY PROGRESS REPORT

17.

L. F. Hemphill

CIRCULATING-MODERATOR-COOLANT REACTOR:

ORNL

R. W. Schroeder

H. R. Wesson
ANP Division

A preliminary design of a hydroxide
cooled and moderated reactor has been
defined. A 2.5-ft spherical reactor
should be able to deliver 200 megawatts
with a maximum wall temperature of
1500°F and a maximum coolant outlet
temperature of 1450°F, Only 115
megawatts 1s believed adequate for
flight at Mach 0.75 at 35,000 ft. An
essential feature of the design 1s the
use of annular fuel elements to attain
a high ratio of heat-transfer surface
to fuel volume., These design studies
were based on the use of sodium hy-
droxide, as 1ts properties are better
known than those of the other possible
hydroxides.

FLUID-CIRCUIT SPECIFICATIONS

After exploratory investigations, it
was decided to attempt to design a 200-
megawatt reactor with a 2.5-ft spherical
core, with a maximum wall temperature
in contact with the coolant of 1500°F,
and with coolant inlet and outlet
temperatures of 1100 and 1450°F,
respectively., These temperatures
imply a permissible coolant film drop
of 400°F at the inlet end, and 50°F at
the outlet end. The thermal con-
ductivity of sodium hydroxide 1is poor
relative to the thermal conductivities
of liguid metals, causing heat-transfer
coefficients and film temperature drops
to suffer. Attainment of the 50°F
film drop at the outlet end therefore
indicated special treatment to be
necessary. With all other variables
fixed, film drop varies directly with

188

unit power and inversely with V928
where V 1s fluid velocity. Therefore
the design should be such as to permit
the coolant to exit from a region of
minimum power, and such as to permit
high coolant velocities at the dis-
charge end of the flow path. Con-
versely, lower velocities can be
employed at stations farther removed
from the discharge end,
mizing pressure drop.

thereby mini-

FUEL-ELEMENT DESIGN

Fuel-element studies indicated the
ratio of heat-transfer surface required
to fuel volume required to be high,
approximately 1000 ft? surface to
0.33 ft*® volume. (The fuel volume is
based on the use of fused fluorides
with a uranium density of 150 1b/ft3,
and an estimated critical mass of
50 1b.) The use of cylindrical fuel
elements with fuel filling the elements
would have involved a prohibitive
number of very small tubes. The use
of annular fuel elements, with fuel
in the annulus and coolant moderator
flowing inside and outside the element,
permitted acceptable fuel temperature
gradients, and a mechanical design of
enhanced simplicity and ruggedness.

REACTOR DESIEGN

The foregoing considerations led to
the configuration 1llustrated in Fig.
17.1. The entering fluid 1s passed
through the reflector region first

 
FREE FLOW 75 %
HEAT-- TRANSFER AREA 960 f12 .
NO. FUEL ELEMENTS 2446

 

FOR PERIOD ENDING SEPTEMBER 10, 1951

SECGRET
DWG. 12959

FUEL ANNULUS O(H2 =(L377 0D 1 0.OH WALL TUBING {(OUTER TUBE)

 
  
  

 

 

 

 

 

 

 

 

 

 
   
   
   
     
 
  
   

0323 0D 1 0015 WALL
“0 PER INGH HELICAL SPAGER THREAD
(INTEGRAL WITH INNER TUBE)

TYPICAL FUEL ELEMENT
DETAIL "A"

 

MATERIAL CONSTITUENGY
CODLANT-MODERATOR 5.9 17
STRUGTURE 16 £t
FUEL 0.37 112

_—MANIFOLD HEADERS
TOTAL  7.87i .

~GAS BLEEDER LINES
—~DUMP AND EMPTYING LINES

  
  
    
 

SRR

SEE DETAIL "A'3—2_

e eR T ey v Mo s

 

NaOH OUTLET PIPE 5-in.IPS
AND THERMAL SLEEVE
6 REQ'D EQUALLY SPACED

o
-

 

 

o

=N

 

Fe N2OH OUTLEY

PR

  

 

:}/‘; '
, o . . _ % NaOH INLET ROTATED
AND THERMAL CLERVE : \. © 3G° FROM TRUE POSITION
& REQ'D EQUALLY SPACED SLOQUTLET PLUG FOR GAS
- BLEED AND. EMPTYING LINES
ELEVATION
17.1 - Sodium Hydroxide Cooled ard Moderated Reactior, Dimensiocns are

inches unless otherwise specified.

189

 
 

ANP PROJECT QUARTERLY PROGRESS REPORT

 

 

 

 

 

  
     
  
    
 

 

 

SECRET
DWG. 12960
] i T T T
2101 MAX
{n
\\\\\////, FUEL TEMP.
026 MAX < (PEAK; AT CENTER ”1
1600 Ry e . /1543 mMaAX 1500 MAX CF FUEL SLAB) .
\\\ //” \\\\ ///” > \\\ ’/’,—G—_ ) -9_ _J
1500 \ " \i;///// \\\\i////f:w%;::“, T FUEL SIDE
1;‘:] - . \/ ‘\,/ m—lifip MAX CUEL B
P 1400 T T CO .LANjT WAt ]
x ; ,/’”//A COOLANT TEMP.
a.
= 1300 bt ‘?EQT_MAX }
k- 7 e __:“’ O g ~—=-0
1200].. e o 1280 MAX fi\l250 MAX
SOLID LINE, 200,000 kw
100!l BROKEN LINE, 115,000 kw -J
TWO TOP CURVES DRAWN ON TEMP
SCALE (2); AlLl. OTHERS ON SCALE ()
1000 ) | { L { | | | | i
A 8 CD E FG H IJ K LM N 0
REACTOR STATIONS (REFER TO FiG. I7H
Fig. 17.2 - Temperature Throughout Sodium fiydroxide Core.

(the fluid comprising the reflector)
to enable the pressure shell to operate
at the lowest temperature in the cycle.
It 1s then passed through the core via
five parallel passes, starting at the
core centerline and discharging from
the peripheral pass.

flow veloci-
and heat-transfer coefficients
are plotted against flow path length
in Figs., 17.2 and 17.3. Data are
plotted corresponding to two design
conditions in Table 17.1. The 115-
megawatt conditions correspond to the

requirements for Mach 0.75 at 35,000

Reactor temperatures,
ties,

190

TABLE 17.1

Design Coalant Condition for Maximum
and Cruise Power

 

 

COOLANT COOLANT
INLET OUTLET
TEMP. (°F) TEME. (°F)
Maximum power 1100 1450
200 megawatts
Cruise power 1050 1250
115 megawatts

 

 

 

 
 

FOR PERIOD ENDING SEPTEMBER 190,

1951

 

 

 

 

 

SECRET
e DWG. 12961
< I T ( I
B
s
Ca 200 -
oE INSIDE_TUBE
m \(\) i : e e \\‘-.‘
S%ac | / A -
5 é e ________.___'“‘""_'""'—Tfll‘f: ----------------------------------------- /
“oh \OUT SIDE TUBE
3 .
% 40k |
= CINSIDE TUBE '
£ - ,
>_ 50”’ \\\ e -
> _ -
S 200 | e
3 l Oh“ ';LWZL:::::::::::M / )
'5‘;’ . QUTSIDE TUBE
| [ N b, J C | N [
O3 B o E G H 1 K M N 0

REACTOR STATIONS (REFER TO FIG.{7.1)

Fig. 17.3 - Sodium Hydrexade Veloclty and Film Coefficient’ ’Fhruughout fore.

Velocity and A for 200,000 kw; same for 125,000 kw.

ft, based on NEPA studies for their Design features currently under
type IL-14 power plant. The 200- investigation include a fuel header
megawatt conditions were selected and fission product gas bleed-off
initially for design purposes and were system, and a control system involving

intended to include calculated require- displacement of reflector

sodium

ments with a generous contingency Thydroxide by inert gas. Both systems
margin., ' : include negative acceleration features.

 

191
 

ANP PROJECT QUARTERLY PROGRESS REPORT

18.

North American Aviation,

At the request of Oak Ridge National
Laboratory, North American Aviation
has conducted an investigation of
high-temperature (above 1800°F) helium
and sodium liquid-vapor power cycles
with regard to their application to
the Phase II (Mach 1.5, 45,000 ft)
arrcraft, This work has been com-
pleted, and detailed reports are
being prepared for publication.
A brief summary of the conclusions
reached as a result of these studies
1s presented here,

Aerodynamic studies here indicate
that L/D ratios of the order of 4.0 to
5.0 represent a maximum attainable
range for large supersonic aircraft
at Mach 1.5 and 45,000 ft. Values at
the higher end of the range can be
achieved only with an exceptionally
clean airframe and high-impulse
engines., (This 1s in sharp contrast
to L/D values of the order of 5.7 to
6.5 quoted generally in the literature
for systemswithrelatively low-impulse
engines.) A consequence of reduced
aerodynamic performance i1s i1ncreased
power plant impulse, and this in turn
requires engineering a higher tempera-
1s presently
As a

ture power plant than
under consideration elsewhere.
reference, the necessity of high-
temperature power plants for the
supersonic propulsion of nuclear air-
craft has previously been i1ndicated
by NAA.(?)

 

(I)Reprint of an official letter from North
American Aviation to Oak Ridge National Laboratory
outlining the results of their power plant studies
{gglfhc ANP program, NAA-AFR-51A-1653 (Sept. 10,

192

HIGH- TEMPERATURE POWER PLANT STUDIES

Inc.

North American has investigated
compressor Jets as a means for the
attainment of the necessary high
temperatures. In the binary com-
pressor-jet cycles, the working medium
in contact with the high-temperature
components need not necessarily be an
oxidizing fluid. If the operating
temperature of the radiator is main-
tained below 1800°F, the problems of
obtaining suitable high-temperature
materials may be significantly reduced,
since all high-temperature components
are not subject to oxidation.

On the basis of experiments con-
ducted at North American and elsewhere,
there appear toexist materials capable
of withstanding high temperatures
(up to 3500°F) in nonoxidizing atmos-
pheres. DBefore these materials can be
considered seriously for engineering
application, a major developmental
effort 1s required to establish tech-
niques by which they can be fabricated
into useful components. Furthermore,
much has to be learned about the cor-
rosion resistance of these materials
to the various coolants and the work-
ing medium proposed. Some pre-
liminary experimental work along these
lines has been undertaken at this
laboratory. To date, results are
encouraging but not yet conclusive.
For example, static corrosion rates of
3250°F sodium vapor on molybdenum are
apparently well within the required
Iimits for the aircraft power plants.

 

(z)J. A. Malone, A. S. Thompson, and H.
Schwartz, Remarks on the Necessity of High-
Temperature Power Plants for the Attainment of
Supersonic Nuclear-Powered Aircraft, North
American, NAA-SR-Memo-33 (March 29, 1951).

 
ey

 

FOR

North American has conducted an
analytical investigation of high-
temperature compressor jets to de-
termine the plausibility of such
systems and to provide a guide for any
future experimental work that might be
undertaken, The aim in this work has
not been to establish the feasibility
of systems that might be constructed
in the near future. The aim has been
rather to determine whether an extensive
matertals and component development
program for high-temperature com-
pressor-Jjet application would be
justified. |

An integrated reactor, power plant,
and airframe study has been conducted
for several proposed systems, carried
out 1n sufficient detail to determine
if a plausible supersonic airplane is
achievable., Some systems guickly
appeared severely limited in per-
formance and onlvy a small amount of
analytical detail was reguired. This
was true for the gas-cooled reactors.
The sodium liquid-vapor compressor jet
was developed in considerable detail
and appears to be one plausible
system, In all cases the analysis was
dependent on extrapolaticens of rela-
tively meager experimental data on
properties of materials and working
fluids. The numerical results that
have come out of these analyses are
not intended to be final and fixed
and should be considered in the light
of their intended exploratory purpose.

SODIUM LIQUID-VAPOR COMPRESSOR JET

A liquid-metal-cooled reactor
operating in conjunction with a sodium
liquid-vapor compressor-jet system has
been investigated. On the basis of
this study the following specifications
for a supersonic aircraft have been
established: |

 

PERIOD ENDING SEPTEMBER 16, 1951

General:

Mach no. {cruise) - 1.5
Alrvitade {(cruise) 45,000 f¢
L/D (cruise) 5.0

Takeoff speed
Landing speed

70, knots
TG knots

Weight:
Gross weight 400,000 1b
Power plaat weight 145,600 1b

feactor and shield (divided) 120,000 1b

Structure weight and miscel Janeous 120,000 b

Pay load weight 20,000 1b
Reactor:
Diameter 4.0 ft
Leagth 4.0 ft
Moderator BeC + €
Fael ‘ U homoge-
neously
impreg-
nated in
moderator
Conlant Tin
Feactor wall temperature ~A000°F
Fuel inventory 73 1b
Reactor power 535,000 kw
Power plant: ,
Sodium eveporation temperature 260005'{di¥
saturate
vapor)
Sodium evaporation pressure 290 psia
Sodium condensate temperature 1700°F
Maximum air temperature ~1450°F
Air flow rate 1518 1b/sec
Air specific impulse 528 1b
thrast/1hb
air/sec

Several airframe configurations and
power plant arrangements were con-
sidered, each being optimized to
arrive at the best overall performance.
In the final configuration the entire
power plant is housed within the
fuselage, and side inlet scoops are
provided for air intake. A shielded
crew compartment 1s in the nose of
the ship some 50 ft forward of the
reactor and capable of housing five

193
 

ANP PROJECT QUARTERLY PROGRESS REPORT

crew members and the necessary control
equipment. The power plant is aft of
the reactor; the fuselage diameter is
approximately 13 ft at this section.
The power plant consists of a sodium-
vapor generator heated by molten tin,
located immediately aft of the shielded
reactor. Sodium vapor passes to five
turbines mounted circumferentially
around and entirely within the fuselage.
Each turbine is directly coupled to a
303-1b/sec air compressor which drives
atr aft and across a common radiator,
the sodium condenser, which occupies
the entire fuselage diameter. Air
continues aft through a nozzle pro-
ducing the propulsive thrust; the
sodium condensate 1s pumped forward
to the vapor generator,

The task of developing such a power
plant as described above would require
a very extensive effort. However, 1f
the appraisal of supersonic aerodynamic
performance is correct, and 1f amanned,
supersonic nuclear vehicle 1s required,
a program of this sort should be under-
taken, The sodium cycle is suggested
as one possible approach, not neces-
sarily aunique one, which on the basis
of preliminary studies appears to have
no fundamental limitations that pre-
clude 1ts development,

HELTUM-~COOLED REACTORS

in addition to compressor-jet
systems in which helium serves both as
a reactor coolant and working medium,
several turbo-jet systems were investi-
gated in combination with helium-
cooled reactors.

These investigations have indicated

that a major development effort uti-
for

lizing helium-cooled reactors

194

supersonic propulsion of aircraft
cannot be justified. In no case does
it appear that a supersonic vehicle
can be achieved, even when considering
reactor temperatures as high as 3300°F
and helium pressures in the range of
1000 to 2000 psi. There are fundamental
limitations involved having to do with
the comparatively poor heat-transfer
characteristics of the gaseous coolants.

A basic requirement for a high-
performance aircraft is light, compact
heat-exchange equipment with minimum
auxlliary power requirement. Jt 1s
well known that at moderate pressures
and temperatures gaseous coolants are
several orders of magnitude poorer in
these regards than the liquid coolants.
Increasing gas pressure tends to reduce
this difference, as does 1ncreasing
temperature differentials between hot
and cold fluids. At high pressures
equipment becomes more compact and
pumping power 1is reduced, but little
or no weight saving 1s accomplished
since heavier walled equipment 1is
required to accommodate the increased
pressure. The requirement that temper-
ature differentials in heat-exchange
equipment be large forces the maximum
power plant temperatures to 2 high
level without any realizable thermo-
dynamic benefit, While temperatures
in this analysis as high as 3300°F
were considered, air specific impulses
remained relatively low (approximately
30 1b thrust/lb air/sec). In all
specific designs considered for the
helium system, no plausible supersonic
vehicle was found. Specific power
plant weights in excess of 2.6 1b of
power plant per pound thrust were
obtained in all instances. Even
assuming an L/D ratio of 5.0, which 1s
hardly realizable with these low-
impulse engines, an aircraft of a gross
weight well in excess of 450,000 1b 1s

 
 

FOR PERIOD ENDING SEPTEMBER 10, 1951

required. ' On this basis it was recom- devoted to the helium, or other gas,
mended that no additional effort be cycle. '

195

 
 

 

 

 
 

19.

ANALYTICAL CHEMISTRY"

C. D. Susano, Analytical Chemistry Division

Satisfactory chemical methods have
been developed or suitably modified
for the determination of metallic cor-
rosion products -~ principally iron,
nickel, chromium, molybdenum, and
platinum — in reactor fuels composed
of eutectic mixtures of fluoride
salts., Two colorimetric methods for
low concentrations of cobalt in fluoride
salt mixtures and alkali hydroxides
were found to be too i1nsensitive
(sensitivity, 100 ppm) for the purpose.
Work is continuing on this problem.

A crystalline material, which was
deposited in the lithxum at a hot-
cold junction in a coolant test systen
during an extended shutdown period, was
identi1fied by X-ray-daiffraction
and chemical analysis as a
mixture of slightly contaminated
lithium hydride and hydroxide. A
method for the determination of carbon
in lithium, both carbide and carbonate
carbon, has been developed.

analvysis

A review of the analytical work for
the various phases of the ANP Program
shows that approximately 50% of the
work is concerned with the analysis of
reactor fuels of the mixed fluoride
salt types, 30% with the analysis of
sodium and lithium, and the remainder
with miscellaneous samples quite
in nature. A total of 1781
(75% chemical, 25%

were: made on 310

varied
determinations
spectrographic)
samples.

 

(I)Abstracted from the report by M. T. Kelley
and C. D. Susano, Analytical Chemistry — ANP
Program, Quarterly Progress Report for Period
Ending August’ 31, 1951, ORBNL, Y-12 site, report
Y-B31-283 (Aug. 30, 1951).

.fi

 

The n-butyl bromide method, which
is now being used regularly for the
determination of oxygen in sodium, is
found to yield excellent results for
oxygen concentrations of 0.015% or
greater. Studies are being made of
refinements of this method so that it
may be applied with reasonable accuracy
at even lower concentrations. A search
is being made to find a liquid medium
suitable for use in melting sodium
samples from metal tube containers;
certain hydrocarbons which are under
test appear to be promising for this
application. Preliminary tests
directed toward the adaptaticn of
the method of Pepkowitz and Judd®?)
(for oxygen in sodium) to the determi-
nation of oxygen and nitrogen in
lithium do not appear promising.
However, an apparatus 1s under con-
struction for use in determining the
oxygen content of metallic lead.
The method to be used incorporates
modifications and refinements of the
technique used by Funston and Reed(??
for the determination of oxygen in
bismuth.

The oxygen content of inert gases
{(principally helium) is now being
determined by a colorimetric method
which gives excellent precision below
25 ppm; at higher concentrations of
oxygen the precision drops off sharply.
An alternate method will be tested as

 

SZ}L. P. Pepkowitz and W. C. Judd, “Determi-
nation of Sodium Monoxide in Sodium,’ A4nal., Chea.

22, 1283 (1950).
(3)g. s, Funston and S. A. Reed, “Determining

Traces of Oxygen in Bismuth Metal,’ Anal. Chen.
23, 190 (1951).

199
 

ANP PROJECT QUARTERLY PROGRESS REPORT

soon as
apparatus

fabrication of the required
1s complete.

ANALYSIS OF REACTOR FUELS

K. Bowan, Jr. J. C. White
W. J. Ross
Analytical Chemistry Division

In the course of the development of
methods for the determination of me-
tallic corrosion products in NaF-KF-UF,
and NaF—Ber—UF4 euntectic fuel mix-
tures, a considerable amount of infor-
mation has been accumulated as to the
solubilities of these mixtures 1n
various reagents. This information is
being compiled and will be 1issued as
a separate report.

Methods and techniques have been
developed for the determination
of metallic corrosion products -—
chiefly iron, nickel, chromium, and
molybdenum — 1in two reactor fuels,
the ternary eutectics NaF-KF-UF, and
NaF-BeF, -UF,. At present only the
NaF-KF-UF, mixture is of interest from
the standpoint of corrosion studies.
Platinum and silver are found in the
eutectic since thermocouples of these
The technigue of
from metal

metals are used.
removing the eutectic
sample tubes after heat treatment,
which was described in previous re-
ports, *+3) has also been applied
successfully to samples contained 1in
tubes of unusually small size (1/8 1in.
i.d.), which have to be used in cor-
rosion studied in the X pile,

 

(4)M. T. Kelley and C. D, Susano, Analytical
Chemistry -~ ANP Program Quarterly Progress Report,
ORNL, Y-12 site, report Y-B31-260 (May 31, 1951).
See also ANP-65, p. 234.

(S)M. T. Kelley and C. D. Susano, Analytical

Chemistry Division (Y-12 Branch) Quarterly
Progress Report, URNL, Y-12 site, report Y-B31-270
{June 9, 1951).

200

Spectrographic Results. The eu-
tectic mixture was prepared for spec-
trographic analysis of corrosion
products as described in previous
reports. (**5) The agreement of spec-
trographic and chemical results was
satisfactory for iron, nickel, and
chromium. Molybdenum was not de-
termined spectrographically. Since
the spectrographic preparation is time-
consuming and the results essentially
substantiate the chemical results,
spectographic tests have been dis-
continued.

Development of Colorimetric Methods,
Two reagents for the determination of
traces of cobalt in fluoride eutectics
and fused alkali metal hydroxides,
tetraphenyl arsonium chloride and
sodium diethyl thiocarbamate, were
investigated. Neither showed sufficient
sensitivity under the conditions
imposed in this case, 1.e., the limited
amount of sample available and the
number of other determinations required
on the same sample. The lower limit of
sensitivity for these reagents 1is of
the order of 100 ppm. Other reagents,
nitroso R salt in particular, will be
tried, and the procedures examined for
possible means of inmcreasing sensi-
tivity.

The procedure for the colorimetric
determination of chromium has been
standardized to give 5% accuracy 1in
the range 100 to 20,000 ppm by adopting
the procedure of G. F, Smith(%) which
assures holding all the chromium in
the same valence state.

Excellent precision has been obtained
in determining molybdenum in quantities

 

(G)Mixed Perchloric, Sulfuric, and Phosphoric
Acids and Their Applications in Analysis, G. F.
Smith Chemical Co., Columbus, Ohio (1942).

 
 

FOR PERICGD ENDING SEPTEMBER 10,

less than 100 ppm by using the thio-
cyanate-—stannous chloride complex and
extracting with butyl acetate,

Determination of Platinum. Platinum
is currently being used as a thermeo-
couple material in studies of the
physical properties of the fluoride
eatectic fuel mixtures, and 1t 1s
therefore necessary to analyze these
mixtures for platinum. Samples
analyzed for platinum have generally
been found to contain less than 100 ppw.

Platinum®’’ may be determined
colorimetrically as the yellew chlore-
platinpus ion, PtClfiéf, the reduction
frem Pt{IV)} to Pt(II) being ac-
complished by stannous chloride in
acidic solution. The color is stakble
at acid concentrations of about 0,25 N,
Palladium interferes in the procedure.
The interference of yvellow U(VY) ion
1is eliminated bv extraction of the
uranium with tribautyl phosphate.

OXYGEN IN SODIUM

Jr., and J. C. White
Analytical Chemistry Division

B. Bowan,

The n-butyl bromide method for the
determination of oxygen in sodium is
being used on a routine basis
samples of oxvygen content of about
0.015% or wmore. Studies are now pro-
gressing to determine the accuracy of
this and other methods (Pepkowitz and

Judd in particular) in the range 0.001
to 0.005%.

for

 

(?)E. B. Bandell, Celorimetric Determination
of Traces of Metals, p. 494, Interscience, New
York, 1950.

 

1951

GXYGEN IN LEAD

D, L. Mananing and W. K. Miller
Analytical Chemistry Division

The need for s method for the
determination of small amounts of
oxygen in lead has led to the in-

vestigation of the method of Funston
and Reed.{(®} This method invelves the
measurement of the reducticn in volume
of hydrogen when heated in centact
with the sample; however, the total
volume of the apparatus is so large in
comparison with the swall change in
volume during the reaction that the
method is highly inaccurate. A new
apparatus is being designed to correct
this failing in which a small known
volume of hydrogen is introduced 1into
a combustion chamber and allowed to
burn with the oxygen from the lead,
When the reaction is complete, the
remainder of the hydrogen 1s swept out
of the chamber with C0,, the CO,
absorbed in reaction with KOH, and the
volume of the rewaining hydrogen
determined,

OXYGEN IN HELIUM

J. €. White and W, J. Ross
Analytical Chemistry Division

The method of Brady(®’ discussed in
the last quarterly progress report(9®>
has been modified and used with good
results. The new procedure 1s to
measure the velume of test gas re-
quired to produce a 5% change 1in
transmission of the sodium anthraquinone
sulfonate reagent, which changes

 

(B)L. J. Brady, “Determination of Small Amounts
of Oxvgen in Gases,” Anal. Chea, 20, 1033 (1948).

(QJJ, C. White and W. J. Ross, “Oxy
felium and Argon,”

gen an

Aircraft Nuclear Propulsion

Project Quarierly Progress Report for Period
%gg{?g June 10, 1351, ANP-65, p. 238 (Sept. 13,

201
 

ANP PROJECT QUARTERLY PROGRESS REPORT

directly proportional to the oxygen
content of the gas. The precision of
the results is good at oxygen concen-
trations of less than 25 ppm, but
drops off sharply at higher concen-
trations because of the smaller test
gas volumes used. This 1s 1llustrated
by the following typical results of 26
12 helium

determinations made on

cylinders:

 

 

0, (ppm)

 

TEST VOLUME OF HELIUM (m})

A 3.96 12
2.26 10

11 (avg.)
B 0.45 81
0.17 : 130
0.23 300

170 (avg.)

 

 

 

A modification of the Winkler method
for the determination of oxygen as
developed by Pepkowitz(19) will be
tried for comparison with the method
now 1n use, The necessary apparatus
is being fabricated.

OXYGEN AND NITROGEN IN LITHIUM

R. Rowan, Jr. W. K. Miller
D. L. Manning
Analytical Chemistry Division

An attempt was made to apply the
method of Pepkowitz and Judd(?2)
oxygen in sodium to this problem with-
out success.

for

The reaction does not
proceed until the temperature of the
mixture of mercury and lithium 1s
raised almost to the melting point of

 

(10),

.. P. Pepkowitz, private communication.

202

lithium,
much violence. Adaptation of this
reaction to a controllable procedure

and then takes place with

does not seem imminent.

CARBRON IN LITHIUM
R. Rowan, Jr. W, K.

D. L. Manning
Analytical Chewmistry Division

Miller

Most of the carbon in lithium 1is
present as lithium carbide, Li,C,,
which decomposes on contact with water
C,H,; however, some
carbon mavy exist as carbonate and
graphite., Preliminary investigation
is being made of a method for simul-

to form acetylene,

taneously determining carbide and
carbonate which involves dissolving
in dilute acid a sample containing
lithium carbonate and carbide. The
evolved carbon dioxide is absorbed in
ascarite and weighed; the acetylene 1is
bubbled through an alkaline solution
of potassium iodomercurate, K, Hgl,,
and titrated as the acetylide (HC=C) Hg;
and the graphitic carbon is determined
by combustion of the wet residue.

URANIUM TREFLUOBIDE IN URANIUM
TETRAFLUORIDE

W, K. Miller and D. [, Manning
Analytical Chemistry Division

Uranium trifluoride has recently
been prepared in anticipation of 1ts
use in esutectic fuel mixtures for
nuclear reactors. Since 1ts preparation
generally includes reduction of UF,,
analytical procedures used to determine
UF, purity must be able to separate
the two fluorides. Tt was hoped to
find an oxidizing agent which would
oxidize U(ITT) to U(IV) with no effect

on the tetravalent 1on already present.

 
 

FOR PERIOD ENDING SEPTEMBER 10,

Of those mild oxidizing agents which
were tried, including water, ammoniacal
silver nitrate, and ferric sulfate—
hydrofluoric acid, only water appeared
promising, although ammoniacal silver
nitrate is still under investigation.

The amount of UF, may also be
determined by measuring evolved hydrogen
by the reaction

+ 3UF, + 2H,

4UF, + 4HC] ——> uCl1, ,

The apparatus used for this is similar
to that for the Dumas nitrogen determi-
nation except that a reaction flask
fitted with a dropping funnel replaces
the Dumas combustion tube., The evolved
hydrogen is swept into the gas buret
by a stream of CO, Farly results are
not too consistent, but it is believed
that by reducing the volume of the
apparatus better precision can be
attained. | '

The most promising method so far
considered 'is based on the total
reducing power of the sample. An
excess of standard ceric sulfate 1s
added to a mixture of the sample and
aluminum sul fate, and the excess is
titrated with standard ferrous ammonium
sulfate, Results obtained from this
method are compared with those from
the hydrogen evolution method on a
single sample in Table 19.1.

IDENTIFICATION OF RESIDUE IN
LITHIUM-METAL COOLANT SYSTEM

W. K. Miller and D, L. Manning
Analytical Chemistry Division

Puring a two-week period of repair
one

on a lithiuwm-metal coolant system,

 

 

1951

TABLE 19.1

Determination of Uranium Trifluoride
by Two Methods

 

 

 

UF-3 (%)
CERIC SULFATE HYDROGEN EVOLUTION
TEST METHOD METHOD
1 91.1 97.5
2 91,2 92.5
3 © 89.9 86.2
4 - 89.2 91.5
5 93.5 89.7
6 90.7 98. 1
Avg. . 90.9 92.9

 

 

 

portion of the system remained at the
operating temperature of about 500°F
while an adjacent section was cooled
to room temperature., The entire
system was then allowed to cool.
Clusters of pink transparent crystals
were found embedded in the lithium
matrix in the zone of contact between
the cold and hot sections. The
crystals dissolved completely in water
with a vigorous evolution of gas,
indicating that they were probably
lithium carbide or lithium hydride.
A few crystals were mechanically
separated from most of the matrix and
submitted to the Isotopes Physics
Department for X-ray-diffraction
analysis. The sample was identified
as a mixture of lithium hydride
and lithium hydroxide and p0331b1y_
small amounts of lithium nitride. : It
is probable that the LiOH was formed
by hydrolysis at the surface of the
sample and that freshly separated
crystals are almost pure LiH. A
hydrogen determination by combustion
was unsuccessful, probably because%of
the presence of some adhering lithium;

however, it is planned to determine

203
 

ANP PROJECT QUARTERLY PROGRESS BREPGRT

hydrogen by a gas evolution procedure
when the apparatus described above
under "Oxygen in l.ead" becomes availa-

ble.

ANALYTICAL SERVICES

L. J. Brady and J. W. Rbbinsan
Arnalytical Chemistry Division

J. A. Norris, Isotope Hesearch and
Production Division

Fifty percent of the service analyses
during the past quarter involved
reactor fuels, 1.e., the two eutectic
mixtures UFy -Nal'-KF and UF,-BeF, -NaF;
30% involved sodiumand lithium metals;
and the remainder involved beryllium

204

comhnounds, rtesidues, metals, etc.
There were 1781 determinations made on
310 samples, 75% chemically and 25%
spectrographically. A suwwmary of
service analyses performed 1s shown

in Table 19.2.

TABLE 19.2

Backlog Summary

Samples on hand Masy 26, 1951 58
No. of samples reczived 403

Total number of samples 461
No. of sawples reported 310
Backlog as of August 10, 1951 151

 
 

FOR PERTIOD ENDING SEPTEMBER 10, 1951

20. LIST OF REPORTS ISSUED

DATE

REPCRT NO. TITLE OF REPORT AUTHOR( S} ISSUED
Design of the ARE

Y-¥27-8 ARE Operation with Especial Regard to the Coolant W. M. Breazeale 7-9-51
Circuit

Y-F20-15 - Detection of Leaks in the Fuel Elements by Means W. K. Ergen 7-10-51
of Radicactive Tracers

Y-F10-57 Activity of Fission Products and Heavy Elements in Glen Putnam 7-12-51
ARE Fuel ' ' '

Reactor Physics

Y-¥10-56 = Simple Correction on Multiplication Constant for N. M. Smith, Jr. 5-22-51
Difference Between Assumed and Resulting Fission
Distribution in Multigroup Calculations

Y-F10-55 = The Contribution of the (s,2n) Reaction to the C. B. Mills 6-5-51

~ Beryllium Moderated Reactor N. M. Smith, Jr.

Y-F10-58 ~ A Discussion of Normalization in IBM Adjoint M. J. Nielsen 6-26-51
Calculations E ‘ :

Y-F10-61 = 1IBM Procedure for B4C Layer Between Core and Re- C. B.;Mi]ls 6-28-51
fiector by the Coveyou Method '

Y-F10-60 Recommendation on Alternative Loading N. M. Smith, Jr. 6-29-51

Y-F10-59  The Spherical Reactor with a B,C Layer Between C. B. Mills 7-6-51
Core and Reflector

Y-F10-62  The Transmission Coefficient of the B,C Curtain C. B. Mills 7-19-51

- in the ANP Reactor '

Y-F10-65  NeOH Cooled and Moderated Reactor N. M. Smith, Jr. 7-24-51

Y-F10-68  The Multiregion Reactor Problem as Applied to the C. B. Mills 8-16-51
Multigroup Method

Y-Fl0-566 ; Numerical Technigque for Criticality Ca]cu]ations J. W. Webster 3-20-51
on Hydrogen Moderated Reactors : :

Y-F10-64 = Heating in the B,C Cartain Due to Neutron : C. B. Mills 8-16-51
Absorption and the Blo(n,a)Li7 Reaction ‘

Y-F10-67 Effect on Radicactivity of Fleoding Coolant J. W. Webster 8-14-51
Channels with Borated Water - :

Y-F10-70 © Reduction of ?eak Temperatures in Fuel Tubes R. . J. Beeley §-23-51

1205

 
 

ANP PROJECT QUARTERLY PROGRESS REPORT

REPORT NO. TITLE OF REPORT AUTHOR(8)

Y-F10-72 The Calculation of Eigenvalues of Differential R. R.

Systems by Numerical Integration

Shielding Research

CF-51-5-74 Calibration of the Fast Neutron Dosimeter Used at R. G.

the Bulk Shielding Facility

CF-51-8-7 Suggested Program for Divided Shield Measurements J. L.
and Calculation of Air Scattering R. H.
ORNL~1027 Determination of the Power of the Shield Testing J. L.
Reactor. I. Neutron Flux Measuremesnts in the E. B.

Water-HBeflected Reactor

ORNL- 1046 A Nuclear Plate Camera for Fast Neutron Spectroscopy J. L.
at the Bulk Shielding Facility E. B.

CF-5-8-253 Preliminary Gamma-Ray Spectrel Messurements at the F. C.
Bulk Shielding Facility R. H.

CF-51-8-252 Experiment 5 at the Bulk Shielding Facility — The H. E.
Shadow Shield

Y-F5-55 NDA Querterly Report of ANP Activities from June 1 NDA
to August 31, 1951 — Divided Shield Studies

Y-F5-57 The Divided Shield L. A,

CF-51-6-53 The Shielding of Mobile Heactors E. P.

T. A

Heat-Transfer Research

CF-51-8-32 Status Memorsadum on the Analysis of Heat-Transfer H.
Charecteristics of the Lithium Figure-Eight System

3

Coveyou

Cochran

Mezem
Ritchie

Meem
Johnson

Meem

Johnsen

Maienschein
Hitchie

Hungerford

Wills

Blizard

. ¥elton

DATE
ISSUED

8-30-51

5-11-51
8-1-51

8-13-51

9-14-51
8-27-51
8-8-51

8-31-51

9-17-51

To be

published

in Begctor

Science -

and

- Technology

Claiborne

OBRNL- 1040 The Design and Construction of an Ice Calorimeter R. F¥. Bedmond
J. Lones

Radiation Damage

ORNL-928 Physical Properties of Irradiated Plastics 0. Sisman
C. D. Bopp
ANP-67 Radiation Damage snd the ANP Reactor L. P. Smith

206

 

8-6-51

8-27-51

6-29-51

7-25-51
REPORT NO.

Y-F31.2
Y-B4-16

Y-F31-3

CF-51-8-256

Y-F23-5

Y-¥30-1

Y-F8~22

Y-F31-4

NDA Docu-~
ment

HKF-112

Y-B31~260

 

FOR PERIOD ENDING SEPTEMBER 10, 1951

TITLE OF REPORT

Metallurgy
Cleanliness of Sodium Circuits
Literature Search on Metal-Ceramic Materials

Techniques for Molybdenum Plating from Carbonyl
Vapor of Solution

Literature Survey on Columbium

Experimeutal Engineering

Tests Required for Beactor Components

Report on Test of Extinguishing Agents for Lithium
Metal Fires

Preliminary Engineering Study of NaQOH Ccoled and
Moderated Reactor

Testing and Examination on Thermal-Convection
Loops Operated with Lithium and Lead

Alternative Systems

Estimated Divided Shield Weights for Supercritical
Water Cycle '

Circulating Moderator Reactor for Subsonic Aircraft

Miscellaneous

Analytical Chemistry—ANP Program Quarterly Progress
Report for Period Ending May 31, 1951

 

AUTHOR(S)

E. . Miller
E. P. Carter

E. C. Miller

W. C. Hagel
G. A. Cristy
W. C. Tunnel]
D. F. Salmon
E. 5. Wilsen
P. O

Nadler

B. W. Schroeder

. Day
. Brasunas

>
o m

NDA

H. K. Ferguson Co.

DATE
1SSUED

7-4-51
7-11-51

7-27-51

8-21-51

6-13-51

6-18-51
8-9-51

8~20-51

6-8-51

9-1-51

Analytical Chemistry 5-31-51

Division

207