ORNL-1864.txt

 

   

g

 

 

3 " BN g Jree &~ l'_‘-: I
AFEA fpraer] :Ffitl a‘.“.fl Ewli*'.“ Tt ey
..':1} Ei}:.-:]:.m 15'-'1-.-[“}1 l.o'-t:i.} i:fl-..?i‘ ".m'nh:'.l T 1R
ORNL-1864
Progress
/44
FA
" fi.'k"-r.
¢ i
(@,

    
  
   
  
 

| ‘&'ea.
AIRCRAFT NUCLEAR PROPULSION PROJECT
QUARTERLY PROGRESS REPORT

-l 3 N FOR PERIOD ENDING MARCH 10, 1955

mgsmmnun Cru NGED T,-.

| i by - — g

= — | BY AUTHOR 'r ov... LEC f, /&‘éd N ot i
J_ =X iy B!_._.... -.,_----._---_-_G../?l..éé-—

-4'.‘)
—'.“

Sery
F

-‘ o =

   
 
 

N

3 445k 02509968 9

OAK RIDGE NATIONAL LABORATORY

OPERATED BY

CARBIDE AND CARBON CHEMICALS COMPANY 5"

A DIVISION OF UNION CARBIDE AND CARBON CORPORATION

(T3

POST OFFICE BOX P
OAK RIDGE. TENNESSEE
 

ORNL.-1864

This document consists of 205 pages.
Capy//? of 207 copies. Series A,

Contract No, W+7405-eng-26

AIRCRAFT NUCLEAR PROPULSION PROJECT Y
QUARTERLY PROGRESS REPORT

For Period Ending March 10, 1955

W. H. Jordan, Director
S. J. Cromer, Co-Director
R. 1. Strough, Associate Director
A. J. Miller, Assistant Director
A. W. Savolainen, Editor

DATE RECEIVED BY INFORMATION AND REPORTS DIVISION
(MARCH 25, 1955)

 

e (AR

Post Office Box P 3 quI:; DESDCIE!B g

Qak Ridge, Tennessee

 

 

 

 
PN GA LN~

0O

CMMEOOEANCPPIEILPOMMANTEIOOAOON-UPNZIOMMUMONDO

. Adamson
. Affel
. Baldock
Barton
. Bettis
Biltington
. Blankenship
. Blizard
. Boyd
. Bredig
. Bruce
. Callihan
. Cardweli
Cathcart
. Center (K-25)
. Chapman
. Charpie
. Clewett
. Clifford
. Cottrell
. Cowen
omer

Crouse
. Culler
. Emlet (K.25)
. Ferguson
. Fraas

Frye
. Furgerson
. Grimes
. Hoffman
. Hollaender
. S. Householder
7. Howe
. W. Johnson
H. Jordan
. W, Keilholtz
P. Keim
. T. Kelley
. Kertesz

. M. King

-

MOA LT TMErwOQooMIPAMSIETOPMOUMPLNE DO

. A, Lane

. E. Larson

 

"l‘|.‘

ORNL-1864

Progress

INTERNAL DISTRIBUTION

44,
45.
46,
47.
48,
49,
50.
51.
52.
53.
54.
53.
56.
57.
58.
59.
60.
61.
62.
63.
64.
65.
66.

67.

68,
69.
70.
71,

72.

73.
74.
75.

76.

77.

78.

79.
80.

81.

82.

83-92.
93-112,
113.
114-116.

4

 

. LaVerne

. Livingston

. Lyon

. Maienschein

. Manly

. Mann

. MeDonald

. McQuitken

. Miller

. Morgan

. Murphy

. Murray (Y-12)

. Nessle

. Oliver

. Patriarca

. F. Poppendiek

M. Reyling

W. Savage

W. Savelainen

. D. Shipley

Sisman

P. Smith

. H. Snell

. |. Strough

. Susano

. Swartout

. Taylor

Trice

. Yan Artsdalen

. YonderL.age

Warde

. Weinberg

White

. Whitman

. Wigner (consultant)
. Williams

Wilson

. Winters

. Zerby

X-10 Document Reference L ibrary (Y-12)
Laboratory Records Department
|_aboratory Records, ORNL R.C.
Central Research L ibrary

WUV NCLERP>»POOZYM

NOE-OMOE=PE-TMEMENADP>OOMP> LVITVTOAOSMAP>PTMIEICENDIZ

omOONwoOzTNnaPIrog

. e ‘;
X
S, s )
o "f‘:&r“:‘fi% o

ek
117.
118-119.
120.
121.
122.132.
133-137.
138.
139.
140.
141.144,
145,
146-149,
150-151.
152-153.
154.
155.
156-157.
158,
159.
160.
161.170,
171,
172.
173-178.
179-190.
191-205.
206.
207.

EXTERNAL DISTRIBUTION

Air Force Plant Representative, Burbank

Air Force Plant Representative, Seattle

Air Force Plant Representative, Wood-Ridge

ANP Project Office, Fort Worth

Argonne National Laboratory

Atomic Energy Commission, Washington (L.t. Col. T, A, Redfield)
Bureau of Aeronautics (Grant)

Chief of Naval Research

Convair, San Diego (C, H. Helms)

General Electric Company, ANPD

Glen L. Martin Company (T. F. Nagey)

Knolls Atomic Power Laboratory

Lockland Area Office

Los Alamos Scientific Laboratory

Materials Laboratory (WADC) (Col. P, L. Hill)

National Advisory Committee for Aeronautics, Cleveland
North American Aviation, Inc,

Nuclear Development Associates, Inc.

Patent Branch, Washington

Powerplant Laboratory (WADC) (A, M. Nelson)

Pratt & Whitney Aircraft Division (Fox Project)

USAF Project Rand

USAF Headquarters

Westinghouse Electric Corporation (Bettis Laboratories)
Wright Air Development Center (Lt. John F. Wett, Jr,, WCOSI-3)
Technical Information Service, Qak Ridge

Division of Research and Medicine, AEC, ORO

Atomic Energy Commission — East Hartford Area

s
Reports previously issued in this series are as follows:

ORNL-528
ORNL-629
ORNL-768
ORNL-858
ORNL-919
ANP-60
ANP-.65
ORNL-1154
ORNL-1170
ORNL-1227
ORNL.-1294
ORNL-1375
ORNL.1439
ORNL-1515
ORNL-1556
ORNL-1609
ORNL-1649
ORNL-1692
ORNL-1729
ORNL-1771
ORNL-1816

Period Ending November 30, 1949
Period Ending February 28, 1950
Period Ending May 31, 1950
Period Ending August 31, 1950
Period Ending December 10, 1950
Period Ending March 10, 1951
Period Ending June 10, 1951
Period Ending September 10, 1951
Period Ending December 10, 1951
Period Ending March 10, 1952
Period Ending June 10, 1952
Period Ending September 10, 1952
Period Ending December 10, 1952
Period Ending March 10, 1953
Period Ending June 10, 1953
Period Ending September 10, 1953
Period Ending December 10, 1953
Period Ending March 10, 1954
Period Ending June 10, 1954
Period Ending September 10, 1954
Period Ending December 10, 1954

 
FOREWORD

This quarterly progress report of the Aircraft Nuclear Propulsion Project at ORNL re-
cords the technical progress of the research on circulating-fuel reactors and all other
ANP research at the Laboratory under its Contract W-7405-eng-26. The report is divided
into three major parts: |. Reactor Theory, Component Development, and Construction,
il. Materials Research, and [ll. Shielding Research.

The ANP Project is comprised of about 400 technical and scientific personnel engaged
in many phases of research directed toward the achievement of nuclear propulsion of air-
craft. A considerable portion of this research is performed in support of the work of other
organizations participating in the national ANP effort. However, the buik of the ANP
research at ORNL is directed toward the development of a circulating-fuel type of reactor.

The effort on circulating-fuel reactors was, until recently, centered upon the Aircraft
Reactor Experiment. This experiment has now been completed, and the analyses of the
results of the operating experience are presented in Section 1 of Part |,

The design, construction, and operation of the Aircraft Reactor Test (ART), with the
cooperation of the Pratt & Whitney Aircraft Division, are now the specific long-range
objectives. The ART is to be a power plant system that will include a 60-Mw circulating-
fuel reflector-moderated reactor and adequate means for heat disposal. Operation of the
system will be for the purpose of determining the feasibility, and the problems associated
with the design, construction, and operation, of a high-power, circulating-fuel, reflector-
moderated aircraft reactor system. The design work, as well as the supporting research
on materials and problems peculiar to the ART (previously included in the subject
sections), is now reported as a subsection of Part |, Section 2, ‘“‘Reflector-Moderated
Reactor.”’
 

 
FOREWORD ..o

SUMMARY oo

CONTENTS

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

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

PART I. REACTOR THECRY, COMPONENT DEVELOPMENT, AND CONSTRUCTION

1. CIRCULATING-FUEL AIRCRAFT REACTOR EXPERIMENT ..o

Analyses of the Aircraft Reactor Experiment ...,

Dismantling of the ARE ......

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

Fission Product Investigaiions ..ottt e

Xenon Poisoning of the ARE ...
Neutron energy distribution ...
Xe137 cross section inm the ARE ...ttt

2. REFLECTOR-MODERATED REACTOR oo

Reactor Design .......ccccocee.

Reactor Physics ................

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

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

Activity of ART components after shutdown ...

Gamma-ray heating...........

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

Control FOd CONSIAOIIi OIS <o eeeee ettt ettt e et e s e e ettt ee s e e e e e e e et st temtmamaeenaeaeeanteananaen

3. EXPERIMENTAL REACTOR ENGINEERING ..o

In-Pile Loop Component Development ...

Instrumentation ..................
Melt-down hazard ..............
Fission-gas holdup .........
Flux-measuring loop ........

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

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

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

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

Hori zontal-shaft SUMP PUMP 1 oot

Heat exchanger.................

Pump Development ..............
ARE-type sump pumps......

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

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

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

Mechanical shakedown tests of ART pump (model 1) rotary elements............cooe

Design and Operation of Forced-Circulation Corrosion and Mass Transfer Tests ... v ——
Operation of fused-salt—Inconel 100ps ...

Sodium in MUIHMETAl TOOPS ..o ik

Heat Exchanger Tests ........

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

Intermediate heat exchanger test NO. 2 ...

Small heat exchanger test

Gas-fired heat source ......

Reactor Hazards Test..........

4. CRITICAL EXPERIMENTS

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

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

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

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

R eflector-Moderated R GCIOE ...ooo ettt et e e b e e

PART |l. MATERIALS RESEARCH

5. CHEMISTRY OF MOLTEN MATERIALS .o

Phase Equilibrium Studies

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

13
13
13
13

15
15
16

17
17

24
24
25
26

28

28
28
28
32
32
32
33

33
33

35
35
35

36
36
36
37

37

41
41

49
49

vii
viti

 

Solid phase studies in the NaF-UF, and NaF-ZrF ;-UF , systems. ..o,
Phase relationships and UF; solubility in BeF -bearing systems ...
Phase relationships in LaF ;- and UF ;-bearing systems ...
Preparation and Stability of UF ;-Bearing Melts ...
Reduction of UF, with uranium in alkali fluorides ...
Stability of UF; in alkali fluorides ...
Solubility of UF 4 in NGF R B L i F mMiXIUFES o oottt e e e e
Solubility of UF, in NaF-KF-ZrF ;o
Reduction of UF, in molten LijZrF_ and Li,ZrF oo
Preparation of UF, in NGF-ZEF s
Chemical Reactions in Molten Salts ... e
The equilibrium FeF, + Hz'—u—-—“A Fe® + 2HF in NaZrF, at B00°C i,
Reduction of UF, by structural metals ..., e
Stability of chromous and iron fluorides in molten fluorides ...
Production of Purified Molten Fluorides ...t e, _
Production-scale operations ... e
Pilot-scale preparation of BeF, mixtures ...
Pilot-scale preparation of UF ;-bearing mixtures ... et a ettt et enee
OPECHAl SEIVICES .. ittt et
Preparation of various fluorides .. ... e
Fundamental Chemistry of Fused Salts ...
Electrochemistry of fused salts ...
X-ray diffraction studies in the NaF-ZrF  system ...,
PRySical Chemistry ..o
CORROSION RESEARGCH oottt ettt ettt ettt et e bbbt sat s b
Thermal-Convection Loop Corrosion Studies ..o
Alkali-metal-base mixtures with UF, and UF ;.
Ceramic contamination in NaF-ZrF ,-UF , on Inconel ...
High uranium content in fluoride mixture ...
HAst@ IOy B HO0P S 1ottt etttk
Molybdenum JOOPS ..o e
Sodium in Hastelloy B 1oops ..o e
Sodium in Inconel loops with beryllium inserts ..o
Forced-Circulation Corrosion and Mass Transfer.........cooovoiiiiiiiic e
General Corrosion StUAies ..o et
High-temperature tests of molybdenum in contact with NGF—ZrF4-UF4 ............................................
Brazing alloys on Inconel and stainless steel in sodium and in fuel mixtures ...
Dissimilar metal mass transfer in the system zirconium—type 304 stainless steel—sodium ........
Diffusion of sodium into beryllium ...
Beryllium-Inconel spacer tests ... SSUUTUPRRURPPO
Cermets in NaF-ZrF ;~UF ;oo
Single-crystal specimens of magnesium oxide in lithium and inlead ...
Fundamental Corrosion Research ..ot
Mass transfer in Tiquid lead ... e
Mass transfer and corrosion in fused hydroxides .........c.coooiiiiiiiiiii e
Spectrophotometry of fused hydroxides ...,
Fused hydroxides as acid-base andalog systems ...

ey

49
50
52

53
53

56
56
57
57

57
57
58
61

63
63
63

66
66

67
67
68
68

70

70
70
71
73
73
73
73
74

735

77
77
/8
79
81
82

86

86
86
91
21
91
Chemical Studies of COTFOSION ..o oottt ce ettt ettt st bbbt 95
Effect of chromium additions on corrosion of Inconel by CrF in molten fluorides............coocon. 95
Protective action of small chromium metal additions ... 95
Effect of fisSion ProdUCES oo ittt es ettt 95
Time dependence of corrosion in tilting-furnace tests. ... 96
Effect of valence state of iron on corrosion of inconel by fluoride mixtures ... 96

. METALLURGY AND CERAMICS oottt ettt e s 97

Development of Nickel-Molybdenum Base Alloys ..o 97
FaBriC@tion STUAIES onoeeeeieeeee ettt ee ettt e et e et b ettt et s i e e b et b e et e 97
OXIAGHON STUATES .. e oottt ettt e ettt e e ae e s ea e e b s bbb e 104

Stress-Rupture Studies of Nickel-Molybdenum Base Alloys ... 104

Welding and Brazing Studies of Hastelloy B ... 109
R T GEOTS oo oee et ee ettt et ooz ee ettt e et a e ettt h e e eR bbb 109
Welding of thick SECHOMS ..ot 114
High-t@MPErGIUIE GQIMG o.euuveiieiree et 116

Stress-Rupture Design Curves for Inconel .. ... 120

Development of Brazing Aoy S ..o 120

Fabrication of Test Components . ..o it ittt ea ettt 128
High-conductivity=fin radiQhor ... 128
Intermediate heat exchanger No. 2 ..o 131
Radiator for Cornell Aeronautical Laboratory ... 132
Sodium-beryllium-Inconel compatibility festing apPAratus ..o 134
Heat eXChanGer BPGZES ..ottt e 134
Cermet=-10-Metal BIAZING oottt ettt e e 134

Special Materials FabriCation ..o 136
DIUPHEX FUBING 111ttt 136
GoE OOl FOAS oottt et et e b e e e 136
TUBULAE FUEL @E@MENTS oottt ettt e b b e bbbt 136
Boron Shield for AR T oot eee ettt b bbb 136
Al-UO, fuel plates for shielding experiment . ..o 137

P AIMI € RESEAICR .ottt et ettt ettt e ea e b bR e e 137
Oxidation reactions of UO, and of UO, in BeO . 137
Fabrication of rare earth oxide wafers for critical experiments ... 138

. HEAT TRANSFER AND PHYSICAL PROPERTIES ..o 139

Fused Salt HEat TranmsFor ..o oottt er s e bbb 139

Reactor Core HydrodymamiCs ...ooiu ittt 140

Electrical Heating and Flow in Tube Bends ..o 141

ART Fuel-to-NaK Heat EXChANGer ....c.ooiiiiiiiiiirite e 142

Reactor Core HEat TranSFer ..o iieiieee ettt bbb 142

Transient Boiling SHUGIES ... ittt 143

HEAE CAPACIEY 1.vvrreoieieeesereseearen e ema s s bR 143

VA SCOSTEY worrvvemieee et et s ceses s 143

Thermal CONUEHIVITY - vrrn ittt b 144

Electrical CondUEHiVITY oottt 144

Influences of the Physical Properties on Reactor Heat Transfer ..o 145

A x
10.

1.

12.

13.

14.

15.

 

RADIATION DAMAGE ... oottt e 146

MTR Stati € Corrosion TestS oottt ettt ettt e emae s et nn e sbe e e e e 146
MINTQEUFE I0-Pi1@ LOOP oottt ettt et etk ettt n e e b et et b e e 147
LITR Horizontal-Beam-Hole Fluoride Fuel Loop ..o e 150
Flux-Depression Experiments in MTR ...t 155
Creep and Stress-Corrosion TeStS ..ot st 155
ANALYTICAL STUDIES OF REACTOR MATERIALS e 157
Analytical Chemistry of Reactor Materials ... et 157

Determination of trivalent uranium in fluoride fuels ... 157

Determination of yranium metal in fluoride salt mixtures ... 158

Determination of oxygen in fluoride fuels ... 159

Determination of oxygen in metallic oxides by bromination ... 161

Differential spectrophotometric determination of beryllium ... 162

Determination of lithium in NaF-BeF,-LiF and NaF-ZrF ,-LiF base fuels ..., 162

Determination of potassium in fluoride fuels ... 162
X-Ray Spectrometer Investigations of Fluoride Fuel ... 163
ANP Service Laboratory . ...t e e 163
RECOVERY AND REPROCESSING OF REACTOR FUEL ..o 164
Pilot PLant DeSign oottt ettt b sttt et em et s ettt b 164
Process Development . ..ot et et 164

PART Itl. SHIELDING RESEARCH
SHIEL DEING ANA LY SIS oottt et ettt et eae et e et a e e s 173
Anisotropic Scattering of Neutrons in a Uniform Medium with Beam Sources ... 173
Energy Absorption Resulting from incident Gamma Radiation as a Function of Thickness

of Materials with SIab GEOMErY .....coioiiei et 173
Energy and Angular Distribution of Air-Scattered Neutrons from a Monocenergetic,

Monodirectional Point SOUFCE ...ttt et e et 173
LID TANK SHIELDING FACILITY oo et eb et e ettt 175
GE-ANP Helical Air Duct Experimentation ........... et eeeeteeeeoteeeseaeaeiteeeereeiettye e nre e e e e ne e eeeseeeas 175
Removal Cross S@CHONS ... oooiviiiiiiiiiiiieei et e et e 175
Reflector-Moderated Reactor and Shield Mockup Tests ..o e 176
TOWER SHIELDING FAC L TY oottt et e et e aets s ssessaasassseessesasemsesaessae s entebesbesbeanbesseenieseeaereaes 178
TSF Experiment with the Mockup of the GE-ANP R-1 Shield Design ....c...ocooiviiiiiciniiniiiinn 178
The Differential Experiments at the TSF: Phase 1 ... 178
Calibration of the Revalet, a Remotely Variable Lead-Transmission Gamma-Ray Dosimeter.......... 182
The Project ORANGE Primate Exposure at the TSF ... 182

PART IV. APPENDIX
LIST OF REPORTS ISSUED FROM SEPTEMBER 1954 TO MARCH 1955 ...ciiiii e, 191

 
ANP PROJECT QUARTERLY PROGRESS REPORT

SUMMARY

PART I. REACTOR THEORY, COMPONENT
DEVELOPMENT, AND CONSTRUCTION

1. Circulating-Fuel Aircraft Reactor Experiment

Analyses have been made of the experience ob-
tained through the operation of the Aircraft Reactor
Experiment and summary topical reports are being
prepared, The detailed analyses have revealed no
significant errors in the preliminary numbers re-
ported previously. The salvageable equipment in
the ARE Building is being removed and samples
are being obtained for metallurgical, physical,
radiological, and chemical analyses.

Attempts have been made to determine the fate of
several fission-product nuclides in the ARE. How-
ever, the results leave much to be desired because
no adequate plans were made before operation of
the reactor for the study of this problem. It is
clear that some ‘‘plating’’ of ruthenium onto the
walls of the fuel inlet line occurred and that some
ruthenium was volatilized, It also appeared that
the activity of the fuel in the ARE dump tank was,
in general, about what was to be expected from the
power history.

The energy distribution of neutrons slowing down
in an absorbing moderator of nonvanishing absolute
temperature was finally found to be obtainable by
Monte Carlo calculations on the ORACLE. As
applied to the xenon-poisoning problem of the ARE,
the neutron energy distribution is more closely
represented by a temperature of 2200°F than by
the moderator temperature of 1400°F. This reduces
the xenon absorption cross section, but, even with
this reduction, the observed xenon poisoning was
far below the poisoning which would have existed
if the xenon had not been removed in the off-gas.

2. Reflector-Moderated Reactor

A careful examination of the hazards associated
with the proposed operation of the Aircraft Reactor
Test was made, a summary report of the hazards
was prepared, and a presentation was made to the
Reactor Safeguards Committee. The hazards analy-
sis disclosed that no nuclear explosion could
occur that would damage the proposed sealed
reactor cell and that the reactor cell was more
than adequate to contain the worst conceivable

accident that might be experienced with the in-
stallation.

Work on the reactor layout is proceeding, and a
detailed stress analysis is under way. Much has
been accomplished on the detailed design and
fabrication of component test units, such as the
fuel pump, the core shells, the pump-~expansion
tank configuration, and the heat exchangers.

Comparison of earlier reactor calculations with
the results of the critical experiments on the re-
flector-moderated reactor has shown that the cal-
culations consistently gave critical masses which
exceeded the experimental values by about 2 kg of
U235, For the calculations, a quantity of Teflon
per cubic centimeter of fuel annulus was used that
was 50% of the quantity used in the experiments,
and it was found that this difference accounted for
the discrepancies in the critical masses, The
radioactivity of various parts of the reactor and
shield assembly at various times after shutdown
was computed, In the course of an investigation of
the gamma ray heating of the components of the
reactor assembly, curves were drawn of the gamma-
ray flux at various surface points of a cylinder
made up of gamma-emitting cylinders of fuel mo-
terial; the gamma absorption inside the fuel was
taken into account,

A brief study of the control rod problem showed
that the burnup of the rod will be rather serious
and thus indicated that it would be difficult to
use elements such as gadolinium and samarium
because they contain only a fraction of isotopes
of large absorption cross section. Europium is
being considered as a possible control rod material
because both isotopes of it have large absorption
cross sections; it appears to be obtainable, if
necessary; and it has, furthermore, the advantage
that the isotopes formed by absorption of a neutron
have large neutron absorption cross sections,

3. Experimental Reactor Engineering

All phases of work on the in-pile test loop for
irradiation in the MTR proceeded on schedule
during this quarter., The power density end tem-
perature differential specifications for the loop
were revised to make the first test more conserva-
tive, but the test conditions are still representative

S |
ANP PROJECT PROGRESS REPORT

of those expected in the ART. Loop layout draw-
ings have been completed, and detailed drawings
were started, A flux-measuring loop is being
fabricated for irradiation at the MTR in March.
This loop will provide much needed data on the
neutron flux to be expected in the actual in-pile
loop, since it closely mocks up the forward end of
the in-pile loop configuration. The horizontal-
shaft in-pile loop sump pump was operated at
temperature for 1000 hr, and performance was
satisfactory, Three heat exchangers for in-pile
use were tested, two with single walls and one
with a double wall. The double-walled exchanger
gave good heat transfer performance and should be
acceptable for in-pile use.

Performance tests on the ARE-type sump pump
were terminated following 3748 hr of nearly trouble-
free operation. The fuel mixture NaF-ZrF -UF,
(53.5-40-6.5 mole %) was used in these tests, and
the average fuel temperature at the pump was about
1350°F. The major portion of the operation was
at the design conditions for the pump, namely,
1500 rpm and 40 gpm. One of the desired uses of
this pump is for circulating NaK in intermediate
heat exchanger test loops where a pump output of
184 gpm and at a head of 300 ft is required. The
pump was tested with both water and hot NaK
(1400°F), and it was demonstrated that the required
performance could be obtained with a shaft speed
of about 3800 rpm.

Two ART pump (model 1) rotary assemblies were
given mechanical shakedown tests to determine
seal, -bearing, and vibratory performance character-
istics, Difficulty with excessive leakage of the
lower seal was experienced in both tests, and it
was found that the leakage was caused by dis-
tortions in the seals. The specifications for the
seal have been revised, and new seals have been
ordered.

Three electrical-resistance heated, forced-circu-
lation loops for testing corrosion and mass transfer
in fused-salt-Inconel systems were terminated
this quarter, These loops had accumulated 774,
625, and 521 hr, respectively, of operation before
leaks in the heater sections of two loops and a
pump seizure in the third loop necessitated the
shutdowns. The loop that operated 774 hr had «
Reynolds number of 10,000, @ maximum fluid tem-
perature of 1500°F, and @ temperature differential
of 200°F, The loop that operated 625 hr had
similar operating conditions except that the tem-
perature differential was 300°F, The loop that

*@‘Q o TR ey

A IPE!: ’

operated 521 hr had a Reynolds number of 15,000,
a maximum fluid temperature of 1500°F, and tem-
perature differential of 200°F, All three tests
were run with the fluoride mixture NaF -ZrF ,-UF
(53.5-40-6.5 mole %) in Inconel tubing.

The fourth of a series of tests of sodium in a
berylHium-Inconel loop was also terminated this
quarter after 1000 hr of trouble-free operation. It
had operated at @ maximum temperature of 1300°F
(at the beryllium insert), a temperature differential
of 300°F, and a maximum Reynolds number of
440,000 through the orifice of the beryllium insert,

Design of the test loop and fabrication and
assembly of the various components for the inter-
mediate heat exchanger test No. 2 are proceeding
satisfactorily., A small heat exchanger loop for
testing a bundle of 20 tubes was designed and is
being fabricated and assembled, This test has
been programmed to obtain heat transfer data for
use in the ART design by April 1.

Two tests were completed to ascertain the effect
of a rupture of the reactor pressure shell that would
spill fuel into the cell designed to contain the
ART., The problems of heat dissipation and re-
lease of radicactive material to the atmosphere
were investigated in these tests, The design of
the cell was modified as indicated by the tests so
that the over-all heat transfer rate will be adequate
to dissipate 3.5 Mw of continuous afterheat., It
was determined that the cell would prevent the
release of radioactive material to the atmosphere.

Other tests still in operation at the end of the
quarter included three resistance-heated forced-
circulation loops with fused salts in Inconel; one
gas-fired forced-circulation loop with fused salts
in Inconel; three thermal stress tests of Inconel in
fused salts and in sodium; three forced-circulation,
high-thermal-gradient loop tests of sodium in
Inconel; and one thermal-cycling test of sodium in
a beryllium-Inconel system,

Work is proceeding on facilities for dynamic
fluid tests of ART components, including the four
kinds of ART pumps, the heat exchangers, and
other components,

4. Critical Experiments

Three additional critical assemblies of the
reflector-moderated reactor were studied during
this quarter, An essentially spherical shell of
fuel surrounding a beryllium island and enclosed
in a beryllium reflector was used for two of these
assemblies, The fuel was separated from the

NPT
o BE S g
beryllium by Inconel shells. In the assembly with
]/] g-in.-thick shells the critical mass was 10.8 kg
of U233, Increasing the shell thickness to ! in,
increased the mass to 19.8 kg, with a density of
0.421 g of U235 per cubic centimeter of fuel region,
Various neutron-flux and fission-rate distributions
have been obtained. In the third experiment the
arrangement of materials was altered to include, on
opposite sides of the preceding assembly, a cy-
lindrical structure of beryllium surrounded by a
fuel annulus and a reflector to represent the
entrance and exit flow channels of the circulating-
fuel reactor. The first loading in this test was
24 kg of U235 at a density of 0.416 g of U?3° per
cubic centimeter of fuel region, which gave con-
siderable excess reactivity and an estimated
critical mass (without inserted control rods) of
about 19 kg. This assembly is to be reloaded at a
lower density to evaluate the mass more closely.
A comparison of the neutron-absorbing properties
of gadolinium and samarium, proposed as reactor
control rod materials, shows them to be equally
effective.

PART tl. MATERIALS RESEARCH

5. Chemistry of Molten Materials

Solid phase studies in the NaF-ZrF -UF, system
and its associated binary systems were continued,
and a revised phase diagram for the NaF-UF,
system has been prepared. Quenching studies of
the ternary system have served to define fairly
accurately the primary phase fields of most of the
phases that exist at liquidus temperatures and
have also given additional data on liquidus tem-
peratures and secondary phases for a number of
compositions.

Interest in obtaining fuels with improved physical
properties has refocused attention upon BeF,-
bearing systems, particularly alkali fluoride-BeF,
compositions with sufficiently low BeF, content
to have low viscosity. The potential importance
of such mixtures as fuel solvents is shown by the
low viscosity of a mixture containing 69 mole %
LiF and 31 mole % BeF, and by the high estimated
heat capacities of such mixtures. While very little
information on the solubility of UF, in such mix-
tures has been obtained, it appears that the solu-
bility at 600°C is at least as high as that observed
in NaF-ZrF, and in NaF-LiF-ZrF, mixtures con-
taining less than 50 mole % ZrF .

PERIOD ENDING MARCH 10, 1955

The usefuiness of LaF, as a “‘stand-in"’ for
UF, has been demonstrated. It was found that
there is a close correspondence in thermal effects
between LaF; and UF, systems with components
that form simple eutectics, for example, LiF and
UF,. This probably means that LaF, and UF,
have about the same melting point,

All measurements made to date on the stability
of UF, in molten salt systems and on the feasi-
bility of preparation of UF, by reduction of UF,
have shown poor reproducibility. The considerable
mass of experimental data suggests that this very
poar precision has a variety of causes.

The extreme susceptibility of the UF, mixtures
to oxidation by air or water and the difficulty in
sampling and analyzing the complex mixtures for
trivalent uranium are certainly contributing factors.
In some of the experiments, equilibrium is difficult
to establish because of coating of the reducing
agent by the sparingly soluble UF, product, and,
with the rather feeble stirring available, the heavy
metal reducing agents may not contact the charge
sufficiently. However, it appears that the most
significant reason for the lack of precision is the
as-yet-uncontrolled variation in activity of the
vranium metal in the vorious experiments., The
ability of the metal to alloy, over the temperature
range of interest, with zirconium, the container
metal, the filter, etc., appears to be primarily re-
sponsible for the lack of precision of the experi-
ments.

Accordingly, it appears that true equilibrium data
are available for few if any of the UFa-bearing
systems studied. Studies that are typical of a
larger number performed during the quarter are
reported, but it is obvious that techniques must
be altered and procedures adopted which will keep
the activity variable under control before data
which can be uniquely interpreted can be obtained.

6. Corrosion Research

Several additional Inconel and type 316 stainless
steel thermal-convection loops have been examined
which had circulated the alkali-metal base fluoride
mixture NaF-KF-LiF (11.5-42.46.5 mole %) with
various proportions of UF, and UF, added. The
fluoride mixtures were prepared and handled with
more precise procedures than those used previously,
and therefore results that could be more accurately
analyzed were obtained. Four Inconel and two
type 316 stainless steel loops showed visible and
ANP PROJECT PROGRESS REPORT

microscopic evidence of a uranium-rich layer on
the entire surface. Chemical analysis showed that
practically no Uf:3 was left in the Inconel loops
after operation; less disproportionation was evident
in the stainless steel loops.

The depths of attack have recently increased on
Inconel thermal-convection loops operated as
controls for 500 hr at o hot-leg temperature of
1500°F with NaF-ZrF ,.UF, (50-46-4 mole %), and,
in addition, more erratic results have been ob-
tained. Tests have shown that ceramic beads used
recently to prevent shorting of probes during the
filling operation may have contributed to the in-
creased attack, but it is evident that there are
other still-unknown variables that affect the cor-
rosion rate, Thermal-convection loop tests of a
NaF-ZrF ,-UF , mixture containing 25 mole % UF,
showed that the high uranium concentration re-
sulted in a small increase in the depth of attack
on Inconel. In tests of Hastelloy B thermal-con-
vection loops in both the as-fabricated and the
dry-hydrogen-cleaned condition with NaF-Zr F,-UF,
(50-46-4 mole %) as the circulated fluid, the ciean-
ing procedure appeared to have little or no effect.
No evidence of mass-transferred particles, sub-
surface void formation, or intergranular attack has
been found on type 310 stainless-steel-jacketed
moly bdenum thermal-convection loops which have
circulated N‘::F-ZrF"-UF4 (50-46-4 mole %). In
two Hastelloy B loops operated with sodium,
magnetic crystals were found in the lower cold leg
and in the hot horizontal section.

Preliminary corrosion results have been received
on the first three Inconel forced-circulation cor-
rosion and mass ftransfer testing loops operated
with NaF-ZrF ,-UF, (53.5-40-6.5 mole %) at high
temperature differentials and high velocities.
Despite the turbulent flow and the increase in the
number of cycles, the depths of attack found in
these loops are only two to three times the depths
found in thermal-convection loops operated for
comparable periods.

Static corrosion tests have indicated that brazing
alloys 10.8% P-9.2% Si-80% Ni and 11.6% P-
6.25% Mn—-82.2% Ni have fair corrosion resistance
to both sodium and fluoride fuel mixtures. Brazing
alloys of this type may be used as back-braze
material in heat exchanger fabrication.

Mass transfer tests of zirconium to type 304
stainless steel in sodium have been conducted in
seesaw apparatus in the temperature range 1000 to

1500°F. It appeared that no thermal-gradient mass
transfer occurred in any of the tests; there were no
zirconium deposits in the cold zone, However,
all zirconium samples gained weight during the
tests, and x-ray analysis revealed a layer of
zirconium oxide on each sample.

Tests of beryllium samples exposed to sodium
for 1000 hr in Inconel capsules have shown that
there is very little penetration of the beryllium by
the sodium at 1200°F. However, at 1500°F the
beryllium was very heavily attacked to a maximum
depth of 20 mils, and a 3- to 4-mil porous metallic
layer formed on the surface. The layer was aniso-
tropic and therefore was identified as either beryl-
lium metal or a beryllium-rich beryllium-nickel
solid solution. Tests are under way to determine
the optimum spacing between beryllium and Inconel
when exposed Yo sodium at various temperatures
and flow rates, Also, several cermets are being
tested as possible materials for use as valve
stems and valve seats.

A summary of the extensive studies of mass
transfer in liquid lead has been prepared, and the
study has been discontinued. The data obtained in
the study indicate that alloys containing an inter-
metallic compound will, in general, possess better
resistance to mass transfer than those in which
compound formation is not possible,

A high-temperature spectrophotometer has been
constructed, and approximate absorption spectra
have been measured for fused sodium hydroxide at
various temperatures up to 700°C in air. Also,
some quantitative results have been obtained in
the application of acid-base analog concepts in
fused hydroxide systems.

In chemical studies of corrosion, the addition of
chromium metal to NaF-ZrF, and NoF-ZrF4-UF4
melts exposed to Inconel in filting-furnace tests
was found to reduce attack on Inconel. The
metallic chromium reduces the corrosive chromic
ion to the noncorrosive chromous state.

7. Metallurgy and Ceramics

Investigations were continued in the study of the
properties of nickel-base alloys containing 15 to
32% molybdenum, ternary alloys with a nickel-
molybdenum base, and vacuum-meited Hastelloy B.
The experiments have shown that temperature is
the variable that affects the extrudability of
Hastelloy B. Sufficient tubing was obtained for
the fabrication of standard-size thermal-convection
loops of a 20% Mo—80% Ni alloy, a 24% Mo~767% Ni
alloy, and a 32% Mo-68% Ni alloy. Tensile and
bend test data were also obtained. The Cb-Mo-Ni
ternary alloy has been partially investigated
in melts containing 2, 5 or 10% columbium
and 20% molybdenum; the balance of each alloy
was nickel., The 2 and 5% columbium alloys
were readily fabricated into sheet, but the
10% columbium alloy fractured severely upon
rolling at 2100°F, Creep, oxidation, tensile, and
fabrication tests of several ternary alloys based on
molybdenum and nickel are planned. Extrusion
data have been obtained for several chromium-
molybdenum-nickel alloys, and the effect of the
chromium additions on corrosion resistance in
sodium is to be investigated. It has been found
that alloys containing over 5% chromium are more
resistant to oxidation at 1500°F than Hastelloy B.
Extensive stress-rupture studies of the nickel-
molybdenum base alloys are under way, and limited
data indicate that the corrosive action of the fused
fluorides does not adversely affect the creep-
rupture properties of Hastelloy B at 1500°F and at
1650°F.

Additional Hastelloy B radiator test components
have been fabricated for determining the feasi-
bility of using this material in applications in-
volving thermal shock and high-temperature oxida-
tion. Tests thus far have been encouraging, An
investigation is under way of the problems that
would be involved in welding the thick sections
of Hastelloy B that would be required in the fabri-
cation of a pressure shell.
study of the effects of high-temperature aging on
the microstructure and physical properties of
Hastelioy B has been initiated.

The program to obtain design data for Inconel at
1300°F and at 1500°F under reactor conditions is
nearly complete, and some data at 1650°F are
available. The results obtained, to date, in fused
salts, and, for comparison, in argon are presented
as a series of design curves.
results of tests of the oxidation resistance of dry-
hydrogen-brazed Inconel T-joints is also presented.
The tests made thus far indicate that most of the
nickel and nickel-chromium base alloys are suit-
able for service in an oxidizing atmosphere at
1500°F, and several are switable at 1700°F,

A dry powder method of brazing alloy preplace-
ment has been developed which provides a promis-
ing means for obtaining controlled amounts of

Also, an extensive

A summary of the

PERIOD ENDING MARCH 10, 1955

alloy on each tube-to-fin joint of high-conductivity-
fin sodium-to-air radiators. A procedure has also
been devised for aluminizing large numbers of
type 310 stainless-steel-clad copper fins to pro-
vide oxidation protection of the exposed copper on
the sheared edges of the fins. The fabrication of
the fluoride-to-sodium intermediate heat exchanger
No. 2, which requires the heliarc welding of 400
tube-to-header joints and the back-brazing of these
welds with a suitable corrosion-resistant alloy, is
under way, and about 200 joints have been com-
pleted.

The fabrication of a full-scale liquid-metal-to-air
radiator designed by the Cornell Aeronautical
Laboratory has been undertaken. The design in-
corporates integral-helical-finned tubing completely
machined from type 316 stainless steel bar stock.

The joining of cermets to Inconel is being studied
as part of the investigation of cermets for use as
valve seats. Special fabrication problems con-
sisted of further work on duplex tubing, tubular
fuel elements, boron shielding material, and AI-UO2
fuel plates for shielding experiments,

8. Heat Transfer and Physical Properties

Further heat troansfer experiments have been
performed with NaF-ZrF .UF, (53.5-40-6.5 mole
%) flowing in Inconel tubes. The heat transfer
coefficients continue to fall about 24% below the
general correlation for ordinary fluids, although no
corrosion deposits are observed on the tube walls.
Some quantitative velocity measurements in a
model of an 18-in. core for a reflector-moderated
reactor were obtained. The nature of fluid flow
ina simple separation region was studied by means
of the phosphorescent-particle technique. A study
of the nonuniform electrical heating and nonuniform
velocity distributions of fluids flowing in 180-deg
bends in high-temperature-differential, high-velocity
loops was conducted; an analysis indicated that
the wall temperature of the short-radius side of
the bend could be significantly higher than the
temperature of the long-radius side and thus,
perhaps, account for the large differences in cor-
rosion in the two regions,

In the current design of the ART, the flow on
the fuel side of the heot exchanger is in the middle
of the transition flow region where the Nusselt
number, in particular, can vary over a wide range.
Thus a study has been initiated to determine the
heat transfer and friction characteristics of the
ANP PROJECT PROGRESS REPORT

heat exchanger in this narrow Reynolds number
region, '

A preliminary value of the heat capacity of
NoF-ZrF4-UF_4 (56-39-5 mole %)} in the liquid
state was found to be 0.256 + 0.004 cal/g-°C over
the temperature range 570 to 890°C, The viscosi-
ties of three fluoride mixtures were determined.
The viscosity of NaF-KF-ZrF, (5-52-43 mole %)
varied from 7.9 cp at 550°C to about 3.5 cp at
750°C. The viscosity of NaF-BeF, (57-43 mole
%) varied from about 18.5 cp at 550°C to about
5.2 cp at 750°C. The viscosity of LiF-BeF,
(69-31 mole %) varied from about 10 cp at 550°C
to about 3.5 ¢cp at 800°C. A new radial thermal
conductivity device has been fabricated and found
to be satisfactory in checks with water. A study
was made of the influence on ART heat transfer
of the physical properties of three types of fluoride
fuels, namely, lithium-base, zirconium-base, and
beryllium-base fuels. Both the reactor core and
the fuel-to-NoK heat exchanger were examined.
From a heat and momentum transfer standpoint
the lithium-base fuel appears to be the more
desirable one.

9. Radiation Damage

Improvements have been made in both mechanical
design and temperature control of the facility for
irradiating Inconel caopsules containing fluoride
fuels in the MTR, and the capsule tests are con-
tinuing. A new method for welding thermocouples
to the surface of the capsules has been developed
which makes possible temperature measurements
that agree much more closely with optical pyrometer
readings than the measurements obtained with the
previously used spark-welded thermocouples. |In
the new method, the chromel and the alumel wires
are crossed and then resistance welded.

A miniature in-pile loop for circulating fluoride
fuel in a vertical position in the LITR is being
assembled.  This loop is to operate at a fuel
Reynolds number of 3000 with a temperature
gradient of 100°F along the length of the fuel tube.
The sump pump designed for this loop was tested
and found to be satisfactory. In an experiment to
measure the increase in excess reactivity that
would occur in the LITR if the loop ruptured, it
was found that the increase could not be more
than 0.2% Ak/k, which is within the limits of safe
operation.

The in-pile loop inserted in a horizontal-beam

hole of the LITR was operated 475 hr at full power
and is now being sectioned for examination. Flow
of the fuel mixture NaF-ZrF -UF, (62.5-12.5-25
mole %) was maintained at 8 to 10 fps (Reynolds
number, 5000 to 6000) during the entire run, and
the maximum fuel temperature was 1500°F. The
total power generation of the fuel under reactor
flux was determined by a series of heat balances
with the reactor off and at full power. Several
such measurements gave an average of 2800 w, or
about one-third the anticipated power. Experiments
are now under way to determine why the total
power was lower than expected.

Preliminary measurements of the thermal-neutron
flux to be expected in the fuel region of the first
in-pile loop scheduled for insertion in hole HB-3
of the MTR are being made by using the pneumatic
flux-measuring device presently in hole HB-3.
The preliminary results indicate that the power of
the presently conceived loop will be one-half to
one-third the desired power. Modifications are to
be made in the loop to increase the power.

Approval was obtained for irradiation of the
stress-corrosion apparatus in the LITR, and
examinations of the first irradiated Inconel speci-
mens are under way. Periodic checks of the re-
sistance between the stressing weight and the
weight probe throughout the 1120-hr test (~700 hr
at full power) indicated that no gross increase in
creep rate was caused by irradiation at a stress
of 1000 psi. Operation of the apparatus in the
LITR is satisfactory, but its performance in the
MTR is not assured because of gamma heating in
the relatively massive apparatus required to
achieve smooth temperature control. Therefore,
since stress-corrosion data at the power densities
available in the MTR are urgently needed, a
possible short-cut stress-corrosion apparatus is
being mocked up.

10. Analytical Studies of Reactor Materials

The coefficient of variation for the determination
of trivalent uranium in NaF-KF-LiF-UF -UF, by
the methylene-blue, one-step oxidation method was
calculated to be 2%, which is to be compared with
aprecision of 4% by the hydrogen-evolution method,
The addition of AICI; to samples of NaF-ZrF -
UF -UF, sufficiently accelerated dissolution of
these materials in solutions of methylene-blue in
1.5 M HCI for the methylene-blue procedure to be
applied.
A method was studied for the simultaneous deter-
mination of trivalent and total uranium in fluoride
fuels in which trivalent uranium is determined by
the methylene-blue procedure; the total uranium is
determined by direct titration of the solution from
the trivalent uranium titration with such standard
oxidants as K2Cr207, Ce(S0,),, KMnO4, and
Fe,(SO,);. The titration with each of the reagents
in the HC! medium was too slow for quantitative
application. There is some evidence that a stable
interaction species of pentavalent vuranium and
methylene-white is formed in this titration.

The hydrogen-evolution method for the deter-
mination of trivalent uranium has been modified so
that the hydrogen is converted to water which is
then titrated coulometrically with Karl Fischer
reagent. The primary advantage of this modifica-
tionis that low concentrations of trivalent uranium,
of the order of 4 mg, can be determined with a
precision of about 5%.

Further study of the determination of uranium
metal in fluoride salts by the method in which
the metal is converted to UH, and the volume of
hydrogen liberated upon thermal decomposition is
measured over KOH, showed that HCl and ammonia
are inferior to CO, as carrier gases, An apparatus
for a modification of this method in which the UH,
is ignited in oxygen to form water, which is then
measured volumetrically, was calibrated. It is
believed that the modification will extend the
range of the method to less than 1 pg of hydrogen,
which is equivalent to 80 pg of uranium.

The conductivity method for the determination of
the water produced by the reaction of metallic
oxides with KHF, was shown to be impractical
for the determination of microgram quantities of
oxygen because of the ‘‘carry-over’ of the KF in
the distillation of HF. An alternative procedure
was proposed in which the fused KHFz-MxOymelt
is electrolyzed to yield oxygen, which can be
separated and measured. Silver fluoride is added
to prevent formation of hydrogen at the cathode.
Preliminary results onNa,CO, showed quantitative
recovery of oxygen.

The bromination method for the determination of
oxygen in titanium was applied to the determina-
tion of oxygen in uranium and beryllium. In this
method the metal oxides are mixed with graphite
and brominated at high temperatures to form CO,
which is converted to CO, and measured. The
reaction for UO, is quantitative in 2.5 hr at 950°C.

PERIOD ENDING MARCH 10, 1955

In the case of beryllium, a flux of NaF and FeF,
is added to accelerate the rate of reaction. The
reaction was incomplete at 700°C after 2 hr.

A differential spectrophotometric method for the
determination of beryllium in NaF-BeF,-UF ,-UF,
was developed that is based on measurement of
the absorbance of the Be-p-nitrobenzeneazoorcinol
complex, The coefficient of variation is less than
1%. Direct application can be made in sulfate
solutions which contain a uranium-beryllium ratio
of 10 to 1.

By wusing an anion-exchange resin to retain
zirconium and sulfate ijons, the method of White
and Goldberg was adapted to the determination of
lithium in sulfate solutions of reactor fuels. |In
the presence of beryllium, the total of lithium and
beryllium is determined, and therefore the lithium
must be determined by difference.

A rapid, direct-precipitation method for the deter-
mination of potassium in sulfate solutions of
reactor fuels was developed. Potassium is precipi-
tated as the tetraphenyl boron salt, which is
filteredand weighed. The interference of zirconium
and beryllium is eliminated by complexing these
ions with fluoride. Uranium is held in solution
with a citrate buffer.

High-temperature x-ray spectrometer studies
were started to provide an additional aid in the
determination of phase diagrams, In the initial
work the compesition 2NaF-ZrF , was x-rayed at
temperatures from 400 to 600°C.

11. Recovery and Reprocessing of Reactor Fuels

Preliminary design is under way on a pilot plant
to recover, in seven batches, the 65 kg of U233 ijn
the 2500 |b of ARE fuel by a fused salt—fluoride
volatility process. The design is based on a flow
sheet that calls for a ninefold excess of fluorine
to be passed through the molten fuel at 650°C.
The UF6 and volatile fission-product fluorides
formed will pass from the fluorinator into a bed of
20- to 40-mesh sodium fluoride at 650°C, where
the wvolatile fission-product fluorides will be
absorbed. The UF, will collect in a series of
three cold traps ot +4, —40, and -60°C, which
will then be heated electrically under pressure so
that liquid UF, can be drawn off. An aqueous
caustic scrubbing system will be required for
disposing of excess fluorine. The preliminary cost
estimate for the piant, including @ 20% contingency

factor, is $285,000.
ANP PROJECT PROGRESS REPORT

Kinetic data obtained in further laboratory studies
on the process indicate that the fluorination step
will take place in three stages: (1) an induction
state in which the fluorine is completely consumed
by corrosion and formation of the relatively stable
NaF-UF . complex; (2) a period of constant UF
evolution while the fluorine is 100% utilized by
UF6 production and corrosion; and (3) a period in
which fluorine breaks through and serves as a
carrier gas to complete the volatilization of the
uranium from the molten salt.

The over-all corrosion rate of nickel for the
entire fluorination step was about 0.1 mph. The
data indicate that most of the corrosion took place
during the period that UF , was present.

PART HI. SHIELDING RESEARCH
12. Shielding Analysis

Multiple scattering in air is being calculated by
two different methods. In an analytical approach,
the Fourier transform is used. The method permits
any differential scattering law to be assumed,
although it is not permitted to change with energy
degradation. Also, the method is capable of han-
dling any arbitrary distribution in space and angle
of sources. Development of the method has just
been completed, and it has not yet been applied
to any calculations.

A stochastic (Monte Carlo) calculation for the
same problem permits energy degradation and the
special geometry of an aircraft shield to be taken
The calculations are to be carried
out on the Oracle.

The calculation of gamma heating in beryllium
slabs is being amplified to include higher energy
photons and to permit calculations for several
different-layered regions. This calculation will be
applied to gamma heating problems in the ART.

into account.

13. Lid Tank Shielding Facility

The GE-ANP helical air duct experimentation
has been completed, and an analysis of the data
has been made. An array of 35 ducts in a medium
of Raschig rings and borated water increased the
thermal-neutron flux a factor of 4000 and the
gamma-ray dose rate a factor of approximately 160.
The array introduced a 43.5% air void in the shield
in the region of the ducts; the reduced density
effect alone (without streaming) would increase
the thermal-neutron flux a factor of 770 and the

gamma-ray dose rate a factor of 175.

in an attempt to correlate the effective neutron-
removal cross section with other properties of the
atom, the ratio of the macroscopic removal cross
section to the density of the material has been
plotted as a function of atomic weight. These data
are compared with the total cross section at 8 Mev.

The static tests in the second series of experi-
ments with mockups of the circulating-fuel re-
flector-moderated reactor (RMR) and shield have
been started. In the dynamic tests that will fol-
low, the LTSF source plate will be replaced with
a continuous series of Al-U02 plates mounted on
a movable belt. The belt, which has been suc-
cessfully operated in a test rig at 1050 fpm, will
travel between the neutron window and the heat
exchanger region of the mockup.

14, Tower Shielding Facility

The first experiment with the mockup of the GE-
ANP R-1 reactor shield has been completed with
measurements having been made of the gamma-ray
dose rates along the x, y, and z axes of the de-
tector tank. The tank had 5 in. of lead instailed
1 ft from the rear face (reactor side). In ay trav-
erse (along reactor—detector tank axis) the gamma-
ray relaxation length (A) was 19.7 cm near the rear
of the lead. At the front of the tank, A was 9.6 cm.
At the sides of the tank, A was 10.3 cm.

The first series of differential experiments with
the reactor tank and the detector tank has been
started. Emphasis will be placed on obtaining the
dose rate distribution within the
detector tank as a function of the angle of beam
emission from the reactor tank.

In order to aid in the optimization of gamma-ray
shielding around the crew compartment of a nuclear-
powered aircraft, a method has been developed in
which small lead thicknesses that can be varied
with ease will simulate the gamma-ray crew shield.
For this purpose an anthracene scintillation counter
is enclosed in a thick tead shield with an aperture
in one side that can be covered, by remote con-
trol, with any of several lead disks that vary in
thickness from 0 to 0.7 in. The instrument has
been calibrated in an experiment with a known
geometry and source energy. The angle of inci-
dence of photons on the lead absorber disk was
varied, and the measured lead attenuation was in
good agreement with Monte Carlo calculations of
slant penetration.

fast-neuvtron
PERIOD ENDING MARCH 10, 1955

An experiment in which 25 rhesus monkeys were Air Force. The maximum dose administered was
exposed to high fast-neutron dose rates was per- 30,000 rep during an interval of approximately
formed as part of a program initiated by the U.S. 11/2 min.
Part |

REACTOR THEORY, COMPONENT DEVELOPMENT,

AND CONSTRUCTION
1. CIRCULATING-FUEL AIRCRAFT REACTOR EXPERIMENT

E. S. Bettis

J. L. Meem'

Aircraft Reactor Engineering Division

ANALYSES OF THE AIRCRAFT
REACTOR EXPERIMENT

W. B. Cottrell H. E. Hungerford
J. K. Leslie

Aircraft Reactor Engineering Division

The summary report of the ARE experience will
be issued as three separate ORNL reports: ORNL-
1844, ‘‘Design and Installation of the ARE'’;
ORNL.-1845, ‘‘Operation of the ARE’’; and ORNL-
1868, ‘'Post-Operative Examination of the ARE.”
The first of these reports is an orderly compilation
of the information which had previously been re-
ported piecemeal during the more than three years
which preceded the operation of the reactor. The
second is a comprehensive analysis of the infor-
mation that was obtained from the experiment
during the two-week operating period, starting with
the first uranium addition and ending with the final
scram. The final report will follow as soon as the
radioactive decay of the reactor and system will
permit examination of the various components.

The detailed analysis of the operational data is
virtually complete and the operating report, ORNL-
1845, will be issued soon. The detailed analysis
revealed no significant errors in the preliminary
numbers which were
quarterly report.

reported in the previous

DISMANTLING OF THE ARE

The equipment installed in Building 7503 for
operation of the ARE
salvaging and for obtaining samples for metallurgi-

is being dismantled for

cal, physical, radiclogical, and chemical testing.
Samples are being obtained from the main sodium
pump impelier; from the sodium piping, valves, and
bellows; and from the cold trap. In the fuel circuit
samples are to be obtained from various locations
in the piping, the main fuel pump impelier and
bowl, the fuel heat exchanger, and the fuel system
valves, bellows, and Rotameter. Samples will also
be taken of the V-belts and O-rings of the main
fuel pump; oil from the fuel pump, the sodium pump,

 

'Now with American Locometive Company, Schenec-

tady, N.Y,

ZE. S. Bettis and J, L. Meem, ANP Quar. Prog. Rep.
Dec. 10, 1954, ORNL-1814, p 11,

and the helium blower; electric and thermocouple
wires from the main fuel pump or heat exchangers;
and concrete from the cell wall near the off-gas
leak. Many of the instruments will be removed and
rechecked for calibration. In addition samples are
to be cut from the serpentine fuel bends in the
reactor core, from the pressure shell wall, and from
the beryllium oxide.

The samples are being submitted to the appropri-
ate groups for examination as they are obtained.
The results of the examinations will be reported
under the appropriate subject matter headings.

FISSION PRODUCT INVESTIGATIONS

M. T. Robinson
Solid State Division

S. A. Reynolds H. W. Wright
Analytical Chemistry Division

Attempts have been made to determine the fate
of several fission product nuclides in the ARE.
The results leave much to be desired since no
adequate plans were made hefore reactor operation
for the study of this problem. Nevertheless, several
interesting lines of research are suggested as a
consequence of the investigations.

During operation, copious amouynts of radicactive
gas evolved from the reactor and escaped into the
pit from a gas leak in the fuel pump. In an investi-
gation of this gas by P. R. Bell et al.,® the pres-
ence of Xels‘s, Xe]38, and Kr®® was revealed by
their own or their descendants’ gamma radiation.
It was also observed that poisoning of the reactor
was very much below the level expected if Xe!35
were efficiently retained.* After shutdown of the
ARE and dumping of the fuel, preliminary measure-
ments indicated that the radioactivity of the dump
tank was far below the expected level.® Therefore
several samples taken from the ARE after shut-
down were examined to determine what radicactive
nuclides they contained.

 

3p. R. Bell et al,, Measurement of Gamma Radiation
from Off-Gases of ARE, Dec. 22, 1954 (personal com-

munication).

4J. L. Meem and W. B, Cottrell, Preliminary Report —
Operation of the Aircraft Reactor Experiment, ORNL
CF-54-11-188 (Nov. 30, 1954).

5 o
W. K. Ergen, personal communication.

13
ANP PROJECT PROGRESS REPORT

A sample of pipe taken from the intake end of
the emergency off-gas line for draining the pit was
examined in @ gamma spectrometer. The only ac-
tivity clearly identified was due to Ru'®3, which
is characterized by a 0.50-Mev gamma ray and a
40-day half life. Chemical tests showed that this
material was probably on the outside of the pipe.

A similar study was made of three small samples
cut from the reactor fuel inlet line (line 120).
Gamma spectrometry showed Ru'®3, Ru'%6, and
Zr75-Nb%° as the only identifiable activities, The
following disintegration rates were observed at

62 days after shutdown of the ARE:
Ru'®® (1.3 +0.3) x 10° dis/min/em? ,

ZeP5NLS (1.3 +£0.2) x 108 dis/min/em?

The expected ratio of disintegration rates for
unsegregated fission products at this age is

Ru1%3/Z/95.NB%° = 0.8 ,

whereas a value 10 was found. It is clear that
some “‘plating’’ of ruthenium onto the walls of the
fuel line had occurred. Experiments are in progress
to determine the distribution of these activities
within the metal. A sample of the reactor fuel out-
let line (line 111) will also be examined as soon
as it becomes available in order to determine
whether or not the ruthenium activity is uniformly
distributed over the entire fuel circuit.

The decay of radioactivity of a sample of solid
ARE fuel taken as liquid from the dump tank has
been followed through the period from 31 to 81
days after ARE shutdown. The total activity of
the sample was observed with the large ion chamber
of the Radicisotopes Department. These results
were combined with gaomma spectra to yield both
total photon emission rates and differential decay
data. By observing gamma energy and half life,
the following nuclides have been identified:

Nuclide Half Life Gamma Energy (Mev)
Ba'¥%La'? 13days® 2.5, 1.60, 0.8, 0.5, 0.33
Ce“l 28 days 0.14
775 e % Very |cmg7 0.8

No Ru'% or 1'®! was detected. It seems likely
that the amount of the ruthenium present is rela-

 

6, . . . . .
This mixture of nuclides is at transient steady state

140

and decays with the Ba half life. Most of the gamma

rays are due to Lcn”o.

14

tively small. Continued study of the decay of the
sample will be necessary to determine just how
small the amount is. Because of the long delay
before decay measurements were started, the de-
tection of 1'3! was made difficult; however, the
differential decay data have been analyzed to
estimate the specific activity of the sample:

Time After Specific Averuge
ARE Shutdown Activity Gumma Energy
{days) (curies/kg} (Mev)

3 16 0.96
79 3.5 0.73

At 79 days, the dose rate from the sample (0.0074 g)
was measured with a calibrated ‘‘cutie pie.’’ |f
the dose rate is assumed to be given by r/hr at
1 #t = n CE, where C is the number of curies in the
sample and E is its energy, the results give n = 8,
Since a value of 6 or 7 is usually assumed for n,
it is apparent that the measurements of the specific
activity of the fuel are essentially in agreement
with the values obtained on small samples with a
““cutie pie.”’

Another sample of fuel was used for various
chemical and radiochemical analyses. The chemi-
cal results are given in Table 1.1. The uranium
analysis was lower after ARE operation because
of the use of barren fluorides to flush out the fuel
system. The iron presumably results primarily from
the drilling operation used to sample the salt.B
Aliquots of the sample from the dump tank have
been analyzed for Se8%, 7,95, Ruws, Cs'36, Ccs'37,
Ce'*', and La'*% all except lanthanum and the
two cesium isotopes were determined by conven-
tional radiochemical methods.

In order to estimate the efficiency of retention
of typical fission products, the radiochemical
analyses reported on ARE fuel were compared with
similar results obtained on a sample of solid fluo-
ride fuel (NaF—ZrF4-UF4, 8.5 wt % U) irradiated in
hole 12 of the ORNL Graphite Reactor. The irradi-
ation time approximately matched the high-power
operating time of the ARE. The following ratios
were then obtained between analyses reported for
the ARE fuel and those reported for the standard

 

"This mixture is still approaching equilibrium; the
apparent half life is too long to measure with limited
precision of the available equipment. Both isobars emit
gamma rays at about 0.8 Mev,

8W. E. Browning, Solid State Div. Semiann. Prog. Rep.
Feb. 28, 1955, ORNL-1851 (in preparation).
 

 

TABLE 1.1. RESULTS OF CHEMICAL ANALYSES
OF ARE FUEL
Fuel Before ARE High In
Component Power Operation* Dump Tank
U, wt? 13.59 5.97
Fe, ppm 25 140
Cr, ppm 420 250**
Ni, ppm 25 70**

 

*The U analysis was taken from the ARE Nuclear
Log Book; the other results were obtained from W, R.
Grimes, Jon. 4, 1955.

**Corrected to basis of undiluted fuel by multiplying
by 13.59/5.97.

sample (designated as MR-[}:

Nuclide Ratio (ARE)/(MR-f)
537 6.1

775 3.3

Ru'03 1.5 x 10~4

La 40 14

ce'! 15

The ratios were corrected to the ends of the re-
spective irradiations. Apparently Sr87, descendant
of 2.6-min Kr89, is reduced about a factor 2 below
the expected level in the ARE, presumably due to
partial escape of its noble gas ancestor. The low
value of Zr’% cannot presently be explained. It is
of interest to point out that the omount of Z¢%
reported radiochemically is in good agreement with
the amount deduced from the decay data on solid
fuel. It must also be mentioned that the amount of
Zr?5.Nb%3 found on the walls of the fuel circuit is
apparently negligible compared with the amount
found in the fuel (about 1 part in 2000 if the wall
activity is taken to be uniform over the entire
surface). The very low Ru'®® value in the ARE
sample may possible be due to a faulty analysis,
but it is presently felt to be real. The agreement
is fair between radiochemical results from the
ARE dump-tank material for Ce'?! and La'? and
the results deduced from decay of the solid sample
of ARE dump-tank material. The cerium and lantha-
num results in this experiment and the decay study
appear to be in reasonable agreement with what
would be expected from the ARE power history.

 

9J. L. Meem, The Xenon Problem in the ART, ORNL
CF-54-5-1 (May 3, 1954).

PERIOD ENDING MARCH 10, 1955

Gaomma spectrometry was used to determine the
ratio of amounts of Cs'3¢ and Cs'37 in the two
samples. The former isotope is ‘‘shielded’’; that
is, it must be formed directly in fission since
Xe'?¢ is stable and has a very low thermal-neutron
absorption cross section (0.15 barns). On the other
hand, Cs'37 is the daughter of 3.9-min Xe'37,
Thus, o difference between the ratios of the
amounts of these isotopes in the two samples is a
measure of the escape from the ARE fuel of Xe!37.
The results indicate that not over 20% of the
Xe'37 escaped from the fuel, and that, possibly,
none did.

It has become ws’romaryg‘]o to describe the
release of xenon from the flucride fuels in terms
of a quantity Ap' defined by

Rate of Xe escape = ?\p x amount of Xe in fuel.

From the ARE poisoning data it is roughly esti-
mated that for Xe'3>

)\p = 5 x 10-4 sec-l

This value is consistent with the observed be-
havior of Cs'37 also. From the $r®? data reported
above it appears that krypton isotopes have larger
valyues of A_ than do xenon isotopes.

Several lines of investigation are suggested by
the results reported here. In particular, the effects
of ruthenium on the physical properties of Inconel
and on corrosion by the fluoride fuels should be
studied. If all Ru'®® is removed from the fuel by
the walls, and if the ruthenium “‘plate’’ is uniform
over the entire reactor, the approximate rate of
deposition of Ru'®® is 0.7 (P/A) p/hr, where P is
the total reactor power in Mw and A is the surface
ared in cm2.

XENON POISONING OF THE ARE
W. K. Ergen

Aircraft Reactor Engineering Division

Neutron Energy Distribution

R. R. Coveyou R. K. Osborn
Aircraft Reactor Engineering Division
R. R. Bate

United States Air Force

If neutrons slow down in an infinite moderator of
small absorption cross sections, their equilibrium
velocity distribution will be nearly Maxweilian and
will correspond to the moderator temperature. |f

 

IOM. T. Robinson, Release of Xenon from Fluoride
Fuels: Proposal for an Experimental Program, ORNL
CF-54-6-4 {June 2, 1954).

15
ANP PROJECT PROGRESS REPORT

the moderator has an appreciable absorption, the
velocity distribution will be such as to favor higher
energies because some of the neutrons will be
absorbed before they reach thermal equilibrium
with the moderator. This phenomenon has been
investigated quantitatively by using the Monte
Carlo method (flrst used for a similar purpose by
G. F. von Dardel'") on the Oracle. The scattering
cross section was assumed to be constant and the
absorption cross section ~1/v. The higher ab-
sorption at low velocities further emphasizes the
shift of the neutron spectrum to higher energies.

It appears that the neutron velocity distribution
is well represented by a Maxwellian distribution,
corresponding to an ‘“‘effective temperature,’’ T,
if

22

1 = < 0.06 ,
1) “ .

 

where the macroscopic absorption cross section
E is measyred at the moderator temperature, T,
and % _ is the macroscopic scattering cross section.
The effective temperature is

(2) T, =T, (1 + aAK)
where a2 is a constant, approximately equal to 0.9,

and A is the atomic weight of the moderator. Corre-
lation of this work with the results obtained by

 

Y16, F. v. Dardel, Phys. Rev. 94, 1272 (1954).

16

Pratt & Whitney Aircraft'?

John'? is in progress.
Xe'33 Cross Section in the ARE
W. K. Ergen H. W. Bertini

Aircraft Reactor Engineering Division

and by Brown and St.

By using the results given above and computing
2 ond %_ by homogeneously distributing the
constituents of the ARE core, the ARE effective
neufron femperature was found to be 1.43 times the
moderator temperature, both temperatures being
measured on the absolute scale. Since the moder-
ator temperature was 1400°F or 1033°%K, the effec-
tive neutron temperature was 1477°K or 2200°F.
If the Maxwellion distribution corresponding to
this temperature is convoluted with the xenon
cross section, an effective xenon absorption cross
section of 1.37 x 10% barns is obtained, as com-
pared with 1.76 x 10° barns at 1400°F and about
twice that much at room temperature. Even by
using the reduced cross section, it is seen that
the xenon poisoning of the ARE would have been
much larger than the upper limit of the observed
value if the xenon had not been removed by the
off-gas system.

 

]2P ratt & Whitney Aircraft, Nuclear Propulsion Program
Engineering Progress Report No, 14, Oct. 1, 1954—-Dec.
3] 1954 PWAC-544.

H D. Brown and D. S. St. John, Neutron Energy
Spectrum in D 50, DP-33 (Feb. 1954).
PERIOD ENDING MARCH 10, 1955

2, REFLECTOR-MODERATED REACTOR

A. P, Fraas

W. K. Ergen

Aircraft Reactor Engineering Division

REACTOR DESIGN
A. P, Fraas

Aircraft Reactor Engineering Division

The preliminary design of the Aircraft Reactor
Test (ART) installation has been completed and
a careful examination of the hazards associated
with reactor operation has been made. The instal-
lation proposed makes use of a sealed reactor
cell installed in an extension of the present ARE
building, The hazards analysis disclosed that no
nuclear explosion could occur that would damage
the reactor cell and that the cell was more than
adequate to contain the worst conceivable accident
that might be experienced with the installation.
The cell was described in the previous quarterly
report, ! together with the other major components
of the proposed installation. The reactor hazards
report has been completed and a presentation made
to the Reactor Safeguards Committee. Work is now
beginning on the detailed design of the facility.

Work on the reactor layout is proceeding. A
half-scale model of the reactor has been completed
to disclose the major problems associated with
the assembly operation. The pump bearing and
seal layout has been modified to incorporate a
number of madifications to facilitate fabrication
and assembly., A number of possible heat ex-
changer header detail designs are being investi-
gated from both the fabricability and the stress
analysis standpoints. Preliminary load deflection
curves have been obtained on a model of the first
of these systems. A model of the most promising
of the others is under construction.

A detailed stress analysis for the pressure shell
is being carried out, with particulor attention
being given to the complex pump and header tank
region at the top. The problem is greatly com-
plicated by the need for evaluating the gamma
heating of the Inconel and the attendant unusual
temperature distribution,

Much has been accomplished in the past quarter
on the detailed design and fabrication of component

test units., A full-scale model of the fuel pump

 

]A. P. Fraas and F. R. McQuilkin, ANP Quar. Prog.
Rep., Dec, 10, 1954, ORNL-1816, p 29.

has been completed and is nearly ready for testing.
This unit will be used for determination of the
performance characteristics of the impeller, in-
cluding the cavitation limit, Four sets of core
shells are being fabricated by Pratt & Whitney
Aircraft for dimensional stability tests at temper-
ature and for fabrication investigations. Two test
rigs for investigating the flow characteristics of
full-scale cores have been fabricated and set-up
work is nearly complete. An intensive test pro-
gram is planned for these units. A pump—expansion
tank configuration designed to remove xenon has
been evolved which performs well hydrodynamically;
however, power required for its operation is higher
than is considered desirable. A careful re-
examination of each of the elements in this system
is being made so that a revised arrangement can
be designed and tested during the coming quarter.
The performance of this unit will be evaluated
analytically by using the data obtained on xenon
removal in the ARE and information from the
chemistry and physics programs,

Several heat exchanger test rigs are being
completed and should be ready for testing by
April. One of these will give information on the
heat transfer performance of this unusual con-
figuration with water, A second is designed to
yield performance data with NaK and the fluoride
mixture. The former rig is sufficiently flexible
so that various tube-spacing arrangements can be
employed (cf., sec. 8, ‘“‘Heat Transfer and Physical
Properties’’). This will, of course, not be pos«
sible with the welded-up test unit for operation
with fluoride mixture and NaK.

Design-study models have been built to show
the preliminary layouts for both the reactor
assembly and the facility, The model of the
reactor, pump, heat exchanger, and pressure shell
assembly is shown in Fig. 2.1. A 12.in. scale
was placed ot the bottom of the half-scale model
to give an indication of the key dimensions. (The
full-scale pressure shell will be approximately
54 in, in diameter.) The NaK outlet pipes project
axially from the lower part of the pressure shell,
and the NaK inlet pipes enter the upper part of
the pressure shell radially. The casings for the

17
ANP PROJECT PROGRESS REPORT

PHOTO 23368

SODIUM PUMPS

FUEL PUMPS

Nak INLET PIPES

) PRESSURE SHELL

* NaK QUTLET PIPES

 

Fig. 2.1. Preliminary Model of ART Reactor, Pump, Heat Exchanger, and Pressure Shell Assembly.

18
two fuel and two sodium pumps project vertically
upward from the ‘‘north head'’ of the reactor. The
fuel pump at the left is sectioned to show the
bearings and seal in the removable pump body in
the upper portion of the pump well. The pump and
expansion tank region is shown in more detail in

. SODIUM PUMP

 

 

 

 

 

 

 

 

 

FUEL PUMPS |

'SODIUM PUMP VOLUTE

 

PERIOD ENDING MARCH 10, 1955

Fig. 2.2, for which one of the sodium pump casings
was removed. The cylindrical fuel expansion
tank is at the center between the two fuel pumps,
while the sodium pump velute can be seen in the
foreground. A section cut through this region
just below the top plate, or ‘“‘upper pump deck,”’

 

 

 pHOTO THE70

 

Fig. 2.2. Detailed Yiew of ART Pump and Expansion Tank Region.

19
ANP PROJECT PROGRESS REPORT

is shown in Fig. 2.3, The fuel pump volutes are
arranged to discharge tangentially into the core
inlet at the center, The sodium-to-NaK heat ex-
changer tube bundles are along the upper and
lower quadrants at the periphery. Sodium returns
upward from the reflector through the circular
openings in the ‘“lower pump deck’’ between the
core inlet and the sodium-to-NaK heat exchangers.
The mss ¥ rising through the return opening in
the foreground flows to the leff, enters the sodium
heat exchanger near the fuel pump volute, passes
to the right through the heat exchanger, flows
back to the left around the right end of the baffle
enclosing the heat exchanger, and rises into the
sodium pump inlet (see Fig. 2.2). The NaK inlet
pipes to the sodium-to-NaK heat exchangers
project radially from the assembly in the lower
right and upper left in Fig. 2.3, The NaK outlet

  
  

 

NaK INLET PIPE o

FUEL PUMP VOLUTES

"NaK OUTLET PIPE

 

SODIUM = TO ~NaK HEAT EXCHANGER

pipes are at the lower left and upper right, A
section through the island, core, reflector, and
heat exchanger is shown in Fig. 2.4, The control
rod can be seen at the center of the island, The
rifle-drilled holes for the cooling passages through
the reflector and island can be seen in the
beryllium regions,

The preliminary l/iz-scule model of the reactor
test facility is shown in Figs., 2.5 and 2.6, The
full-scale cell will be 24 ft in diameter. The 10-
ft-dia reactor and shield assembly can be seen
to the right of the center of the cell in Fig. 2.5.
The NaK outlet pipes pass from the lower portion
of the shield, through a bulkhead in the cell wall,
and then upward to the radiator cores. The NaK
leaving the radiators rises to the four NaK pumps
at the top, from which it is returned to the upper
portion of the reactor. Four axial-flow blowers

..... N
PHOTO 23372 °

NakK QUTLET PIPE

 
  
   
 

e NoK INLET PIPE

 

 

Fig. 2.3. Section Through Fuel Pump Volutes and Sodium-to-NaK Heat Exchanger Region.

20
PERIOD ENDING MARCH 10, 1955

ey
PHOTO 23369

SODIUM PUMPS

FUELL PUMPS

EXPANSION TANK T
B SODIUM -TO~NaK HEAT EXCHANGER

§ BERYL.LIUM REFLECTOR- MODERATOR

PRESSURE SHELL NEEI R
& | O CORE FUEL ANNULUS

CONTROL ROD ==

e SODIUM COOLING PASSAGES

BERYLLIUM iSLAND

FUEL~TO-NgK HEAT EXCHANGER REGION
{ HEAT EXCHANGER NOT SHOWN )

 

Fig. 2.4. Section Through the Core, Reflector, and Island of the ART.

21
ANP PROJECT PROGRESS REPORT

         
  

 
 
 

i

GROUND -FLOCR LEVEL Nq_K PUMP DR?VE‘ MOTORS

REACTOR CELL

 

i

_ NaK EXPANSION TANK

    
   
 

; NaK __PUMP

COOLING BLOWERS %

B ok OUTLET PIPES

\ REACTOR AND SHIELD ASSEMBLY

Fig. 2.5. Preliminary Scale Model of ART Facility Showing Radiator Cores in Foreground.

22
£

 

MOTORS |

G S
MaK EXPANSION

 

 

Fig. 2.6. Preliminary Scale Model of ART Facility Showing Stack in Foreground.

 

BN

il
PHOTQ 23366

 

REACTOR CELL

 

SS61 ‘01 HOMYW ONIGONF a0l¥3d
ANP PROJECT PROGRESS REPORT

at the right force air through the radiators and
out the stack at the rear. The black line around
the upper part of the outer tank represents the
ground-floor level.

The NaK drain tanks can be seen in Fig. 2.6
to the left of the lower part of the reactor cell.
Note that four separate NaK systems are used,
each with its own pump, expansion tank, dump
tank, ete. The NaK expansion tanks are just
below and to the left of their respective pump
drive motors in Fig. 2.6.

REACTOR PHYSICS

W. K. Ergen
Aircraft Reactor Engineering Division
Activity of ART Components After Shutdown
H. W, Bertini

Aircrtaft Reactor Engineering Division

The octivities of the main components of the
ART as a function of time after shutdown are

 

109 ORNL_L,R-D,‘,N,G 5833
5
2
108
5
2
-
@
5 0o
&
g
E 5
c
=
= o
£, >
=
5 "\ 1000 hr
< 106 . . _
5
2 .
{00 hr
5
10
5
» ]  PREDOMINANT ACTMITY
ngs, NS ———
4
10
0 50 100 150 200 250 300

TIME AFTER SHUTDOWN (days)

Fig. 2.7. Activity of the ART Fuel After Opera-
tion at 60 Mw for 100 and 1000 hr.

24

given in Figs. 2.7 through 2,11. The activities
are given in gamma curies for 100 and 1000 hr of
operation at 60 Mw. Curtiss-Wright multigroup
2,3 were used as a basis for the
calculations, The detailed work? was done For
100 hr of operation, and the curves for ]OOOéhr
were obtained by multiplying the approximately
flat portions of the 100-hr curves by 10, This
method is valid for all curves except the fuel
curve, Fig. 2.7. Comparison with GE-ANP data®
indicates that the 1000-hr fuel-activity curve may
be high by a factor of 2 or 3,

calculations

 

2H. Reese, Jr., et al., Geomet?l Study for an ANP
Circulating- Fuel Reactor, WAD-1901 (Sept. 1, 1954).

Personal communication with S. Strauch, Curtiss-
Wright Corporation.

4H. W. Bertini, unpublished memorandum.

3. Motett, Miscellaneous Data for Shielding Calcu-
lations, p 66, APEX-176 (Dec. 1, 1954).

ORNL-LR-DWG 5834

INDICATES PREDOMINANT
CACTIVITY

ACTIVITY {gamma curies}

1000 hr

cB0cr® ]z

cef0

 

o 5¢ 100 150 200 250 300
TIME AFTER SHUTDOWN (days)

Fig. 2.8. Activity of the ART Sodium Coolant
After Operation at 60 Mw for 100 and 1000 hr.
T

ORNL—LR-DWG 5835

INDICATES PREDOMINANT ACTIVITY

ACTIVITY (gamma curies )

41000 hr

100 hr

 

0 50 100 4506 200 250 300 350
TIME AFTER SHUTDOWN (days)

Fig. 2.9. Activity of the ART NaK Coolant After
Operation at 60 Mw for 100 and 1000 hr.

For extrapolation purgo¥es, the following rules
may be considered to give approximate activities,
For times after shutdown greater than four days,
the height of the fuel curve can be considered to
be linear with operating time up to 1000 hr., The
1000-hr curve is approximately the curve for infinite
operating time, The heights of the approximately
flat portions of the other curves can be considered
to be linear with time for operating times up to
1 year. All portions of all the curves are directly
proportional to the reactor power,

The activities of the B'? layer around the re-
flector and of the lead shield were not examined
in detail, but estimates of their activities follow.
The activity of the B'® layer would be due mainly
to the sodium impurity.® This activity would be
about 5 x 10~3 of the activity due to the sodium

impurity in beryllium, The value for beryllium is

 

6Spectrographic report on sample 481(a) of H3E503
gives sodium present as 0.31 wt % in boron.

PERIOD ENDING MARCH 10, 1955

ORNL-LR-DWG 5836

INDICATES PREISOMINANT -AKCTIVITY

— _ f_fl\_lCONEL_CORE SHELLS {1000 hr)

ACTIVITY (gamma curies)

—_ —— —_— e — —

NEL CORE SHELLS
{100 fir) oS0

M MODERATOR {100 hr)

CoGO T

 

0 50 100 150 200 250 300
TIME AFTER SHUTDOWN {days)

Fig. 2.10. Activity of the Beryllium and the
Inconel Core Shells After Operation at 60 Mw for
100 and 1000 hr.

given in Fig. 2.10. The maximum activity of the
lead shield would be about 30 curies at shutdown,
At about two days, and beyond, the activity would
be a fraction of a curie.

The activities at time zero, that is, at shutdown,
are intended to be maximum possible activities,
The inclusion of this point in the curves has
introduced slight distortions in the interval from
one to five days after shutdown, I|f more accurate
activities are required in this interval, the values
are available in tabular form.4

Gamma-Ray Heating
L. T. Anderson, Consultant
Calculations have been made to get better
estimates of the gamma heating in the fuel pump
and expansion chamber regions of the ART., The

gemma enetgy flow at the axial-surface points of
a seties of circular cylinders of delayed-gamma-

25
ANP PROJECT PROGRESS REPORT

4”*-32’*‘

QRNL-LR-DWG 5837

10

 

 

 

 

 

 

10°
5
2
7
2
§ 10
o
o
E 5
£
o
e
>
£ 2
s
=
g o
I
NCONEL {00 hr)
> _ HEAT EXCHANGER TUBES 1000
2
e — INCONEL PRESSURE o
R
ESSuRe (G0 hr
5 : FEAT EXCHANGER TUBES (100 hr)
2
107!
0 50 100 {50 200 250 300 350
TIME AFTER SHUTDOWN (days)
Fig. 2.11. Activity of the Inconel Heat Ex-

changer Tubes and Pressure Shell After Operation
at 60 Mw for 100 and 1000 hr.

emitting, fpel was computed. A gamma interaction
coefficient of 0.13 em=! and a gamma activity of
10 watts/cm? were assigned to the fuel. Buildup
was not taken into account. Figure 2,12 presents
the results of these computations, The radidl
dependence of gamma energy flow was also com-
puted for a cylinder of 15 ecm height and diameter
and is shown in Fig. 2.13.

Control Rod Considerations
W. K. Ergen H. W. Bertini

Aircraft Reactor Engineering Division

The burnup in a control rod can be computed if
it is assumed that a rod, worth Ak, absorbs Ak
neutrons per fission. The number of grams burned
up during ¢ hours of operation at P megawatts is

[A%k (atoms destroyed/fission)

% 3.1 % 1019 (fissions/w-sec)
x P x 108 (w) x 3600 # {sec)
x M {g/g-atom)] /10.602 x 1024 (atoms/g-atom)]

= 1.85x 10=4 Ak PtM (g) .

For Ak = 5%, P = 60, ¢+ = 1000, and the atomic
weight M = 150 (corresponding to the rare earth
region), this amounts to 83g. For a density of 7,
representative of a rare earth metal, this would
correspond to 12 ecm3. A rod 7/16 in. in diometer

ORNL-LR-DWG 5838

A=15.24 cm -

A=12.70cem

h=1016cm-

v
-

 

ENERGY FLOW {w/cm?. sec)

 

 

 

CYLINDER RADIUS (cm)

Fig. 2.12.
Emitting Fuel vs Cylinder Radius and Height.

26

Gomma Energy Flow at Axial-Surface Points of Circular Cylinders of Delayed-Gamma-
ane—
ORNL-LR-DWG 5839

ENERGY FLOW (w/cm2-sec)

 

0 1 2 3 4 5 & 7 8
DISTANCE FROM AXIS (cm)

Fig. 2.13. Gamma Energy Flow at Plane-Surface
Points of 15-cm Square Cylinder of Delayed-
Gamma-Emitting Fuel vs Distance from Cylinder
Axis.

3 and a rod

and 12 in. long would have 30 cm
1 in. in diameter and 20 in. long would have
258 e¢cm3. Thus, if all the rod consisted of ‘‘burn-
able’’ material, the smaller rod would suffer con-
siderable burnup, but the larger one would survive

with little change. In considering the use of

PERIOD ENDING MARCH 10, 1955

samarium for the rod it was found that only 14%
of the element is made up of the large cross
section isotope; therefore 12/0.14, or 86 cm?,
would burn out, which is more than the total
volume of the small rod and aconsiderable fraction
of even the large rod. Furthermore, if the easily
obtainable oxide were used instead of the metal
and if a binder were used to give the rod me-
chanical rigidity, even the large rod would become
of marginal usefulness.

From the heating viewpoint, a power density of
20 w/cm® and a thermal conductivity of the rare
earth oxide of 0.003 cal/sec.cm:°C would yield
a temperature difference of 1200°F between the
center of a l-in. rod and the surface. Although
both the thermal conductivity and the power density
are considerably in doubt, the heating problem
appears to make impossible the use of a solid
oxide rod 1 in. in diameter. The use of a hollow
rod might present problems because of burnup
unless an element such as europium, which is
100% large absorption cross section isotopes,
were used,

27
ANP PROJECT PROGRESS REPORT

3. EXPERIMENTAL REACTOR-ENGINEERING

M. W. Savage

E. S. Bettis

Aircraft Reactor Engineering Division

Design work on the in-pile loop proceeded as
scheduled. A flux-measuring loop for obtaining
information on the flux to be expected is being
fabricated, and 1000 hr of trouble-free operation
of the horizontal-shaft sump pump was concluded.
A double-walled heat exchanger was found to be
acceptable for in-pile use.

Performance tests on the ARE-type sump pump
with fluoride fuel were completed, and tests with
water and with hot NaK (1400°F) indicated the
suitability of this type of pump for the heat ex-
changer test loop. Mechanical shokedown tests
of ART pump rotary elements are under way.

Three high-velocity, high-temperature-ditferential
loop tests of corrosion and mass transfer in fused-
salt-Inconel systems were completed, as well as
a fourth test of sodium in a beryllium-Inconel
system.

The test loop for the intermediate heat exchanger
test No. 2 is being fabricated, and work is under
way on a small heat exchanger loop for testing
a bundle of 20 tubes. Tests were made of the
integrity of the cell for containing the ART,

IN-PILE LOOP COMPONENT DEVELOPMENT
D. B, Trauger

Aircraft Reactor Engineering Division

The in-pile loop design specifications were
revised to make the first test a more conservative
one. The power density was reduced from
2.5 kw/cm?® to approximately 1 kw/em?® and the
temperature differential from 300 to 200°F; the
maximum temperature remains 1500°F. An extra
turn of tubing in the ‘‘nose’ section, approxi-
mately 2 ft of developed length, is necessary to
meet the temperature differential specification at
this power density. The total power is expected
to be about 22 kw. The fuel mixture NaF-ZrF -
UF4 (53.5-40-6.5 mole %) will be circulated at «
Reynolds number of 5000. These conditions, with
the exception of the temperature differential, are
at least as severe as those expected in the ART,
The design of the loop has progressed through
the layout stage and is now ready for detailing.
Figure 3.1 shows the present basic design. A
full-scale wooden model, shown in Fig. 3.2, was

28

prepared to aid in solving the difficult assembly
problems and for instructional purposes. The
numbers on the base indicate distance, in inches,
from the reactor lattice face. The calrod pre-
heaters and the fill tank had not been installed
when the photograph wos made. The water jacket,
which will completely encase the loop forward
of the rear header, is represented, in part, by the
plastic tube.

Instrumentation

R. A, Affel P. A. Gnadt

Aircraft Reactor Engineering Division

Instrumentation design for the in-pile loop is
nearly complete, and fabrication and testing of
most components are now under way, A simplified
diagram for the instrumentation and control system
is shown in Fig. 3.3. Since the power generated
in the loop is dependent on the reactor flux, it
is subject to sudden and unexpected changes.
These sudden and unexpected changes would come
about as a result of power failures, experimental
troubles, etc. The main control therefore is by
regulation of the cooling air in response to a
thermocouple signal at the exit end of the heat
exchanger. This signal will also energize the
preheat calrod circuits if danger of freezing exists.
The pump is driven by a Vickers hydraulic oil
motor, which, in turn, is activated by a pump unit.
Manual speed control will be provided by varying
the output of the drive unit. An electromagnetic
tachometer will operate from a toothed gear on the
pump shaft te indicate and record the pump speed.
Activity monitors on all fluid lines leaving the
loop will be connected to alarms or reactor scrams,
depending upon the potential hazard involved.

Melt-down Hazard
C. W. Cunningham

Aircraft Reactor Engineering Division

The principle hazard to the experiment would
be a pump failure or other sudden stoppage of
flow. In the event of a sudden flow stoppage,
the fuel temperature would rise rapidly, and the
fuel tube would be melted in a few seconds. The
6L

FACE OF LATTICE

. ¥ -
i
. ‘/F\LLE
u’
pil
L . - WATER OUT
T WATER IN
Tt~ —-THERMOCOUPLE LEADS
e .- OUTER WATER JACKET
e .~ INNER WATER JACKET
7
i ~-— — HYDRAULIC OIL OUT
//’ el
e -~ HYDRAULIC OIL IN
) 7 . -~ L HYDRAULIC MOTOR
/J‘/ - e
g . - -BEARING HOUSING
. P o
,,‘6‘"‘GRF‘P " ““-THERMOGOUPLE LEADS
il ek
r“"LLfiD . -~ PUMP
! ! .
“FILL LINE

““FUEL LINE
“VENT UNE

" FILL TANK

S WATER JACKET OUTLINE

T HEAT EXCHANGER

T SNIFFER LINE

- NOSE LOOP

Fig. 3.1. Loop for Circulating Fluoride Fuel in MTR Horizontal Beam Hole.

ORNL-LR-DWG 5663

o

-

20

  
  
    
     
    
     
  
  

T AR OUT

T
T AIR IN

SS61 ‘0L HOYVW ONIGNIT do1y¥3d
ANP PROJECT PROGRESS REPORT

 

 

 

 

 

 

 

 

BEARING HOUSING

R

 

 

 

 

 

 

 

. UNCLASSIFIED
ACKET | . PHOTO 23005

 

 

 

 

8-ft BARYTES CONCRETE SHIELD - 7
T i

STEEL SHOT SHIELD

Fig. 3.2, Model of In-Pile Loop.

ORNL Mathematics Panel has investigated this
problem with the use of the Oracle. A predicted
chronology of events follows: At 0.00 sec the
flow is assumed to stop and the salt temperature
is rising at 70°F per 0.10 sec. After 1.20 sec a
thermocouple on the outside of the tube will have
indicated a 100°F rise in temperature, which is
considered to be adequate to provide a reactor
scram signal. At 1.45 sec the salt at the tube
center is at its boiling point, 2430°F. The center
of the Inconel wall, however, will have reached
only 1650°F and will have sufficient strength to
resist considerable pressure if the salt should
superheat and suddenly boil,

If the assumption is made that the salt temper-
ature will continue to rise with no boiling, the
chronology continues, as follows: At 4.2 sec the
salt next to the inner wall will have reached
2500°F and the Inconel tube will begin to melt.
The entire tube would melt in 7.5 sec.

Calculations and circuit tests indicate that the
reactor can be shut down, on the basis of thermo-

30

couple or tachometer signals, in time to prevent
melting of the nose tube. Although it is unlikely
that a scram can be effected from a thermocouple
signal in time to keep the tube below temperatures
where crystal grain growth will become very rapid,
a signal from the tachometer should be fast enough
to avoid this type of damage to the loop. If the
scram fails, it seems likely that boiling will oceur,
aleng with subsequent displacement of the salt
into the pump sump and fill tank. This displace-
ment should result in additional time for secondary
protective circuits to function and shut down the
reactor.

If a melt-down or a sudden leak should occur,
it would be somewhat hazardous for molten salt
to reach the stainless steel water jacket wall.,
A molybdenum shield which will allow no line of
sight to the jacket has been provided as protection
in this event. Molybdenum has been demonstrated
to resist corrosion by the salt at boiling temper-
atures for 30 min or longer.
 

 

 

 

 

 

  

 

 

 

 

 

 

 

 

 

 

   
  

 

 

   
  

  

 
  
   
   

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

. * r T § €
UNCLASSIFIED
ORNL-LR-DWG 5670
VAB T v19,20,21,22,23,24
»
F Ve FILTER FILTE v-36
INSTRUMENT LTE ILTER PLANT
AIR SUPPLY AIR SUPPLY
NITROGEN
V-26
" Nl UNTREATED WATER DEMINERALIZED PROCESS
A FOR EMERGENCY USE WATER SUPPLY
ELECTRIC AR e
MOTOR - COMPRESSOR V.33
v-32 v-34
N e L
2000
3. L]
}_
W L e |
‘ o V=55
b=
- ‘l TO STACK :
' Y
“ 1 HELIUM
| 1500 | & —
— l AA DN A w TO STACK
L= | T Y % _ e
\ _l
=
e
' PUMP
BEARING\
‘ 1400 o«
| e U LIQUID NITROGEN W
< | T YTy 3 COOLED <
‘ HYDRAULIC «
N MOTOR
| 1300 OlL HOLD £ v-43 v-44 |4
- — BY-PASS OIL >
{ RETURN BLOWER N
<
| -
- | ’ 4
TO RETURN LINE
LEGEND : v-48 v-49
. —
A TEMPERATURE {°F) D PRESSURE (psig)
jlb-—J—— TO RETURN LINE
TO RETU
O FLOW (gpm) () FLOW (scth) / © LINE AN 70 Hy0
A30\
@ FLOW {scfm) 1/\ . l TO HG-5
l VoW VoW
I AR HEAT BEAM HOLE
A @ 125 EXCHANGER

— MAIN LOOP

——— PROCESS PIPING

L€

———— MECHANICAL
CONNECTION

Fig. 3.3. Flow Diagram of In-Pile Loop.

SS6L ‘0L HOYYW ONIONZ a0ld3d
ANP PROJECT PROGRESS REPORT

Fission-Gas Holdup

D, W. Magnuson
Aircraft Reactor Engineering Division

Fission gases will be purged from the pump
sump and from the pump bearing regions, with
purging from the bearing region serving to remove
leakage past the rotating face seal. These purge
gases will pass through liguid-nitrogen-cooled
charcoal traps and then through filters. The
design of the fission-gas holdup system is based
on adsorption data of the Linde Air Products
Company.!

The equations of Jury? were used to predict the
performance of adsorption traps in this gas holdup
system, and it was found that a 250-g activated-
carbon trap ot ~170°C wos five times more ef-
fective than required to give a reduction in
molecular density of more than 1 x 108, This
should constitute adequate decontamination. How-
ever, radioactive decay will further increase the
decontamination factor, and a second trap is
provided, in series, as an additional safety factor.
The use of an extremely small-pore metal filter
will prevent particles from reaching the MTR
stack.

Flux-Measuring Loop

D. M. Haines
Pratt & Whitney Aircraft

One of the greatest uncertainties that might
affect the successful operation of this in-pile loop
is in the estimation of the neutron flux, The
thermal flux will be “‘depressed’’ by the materials
of construction of the loop and by the fuel.
Several estimates have been made and an analog
computation was performed by Prott & Whitney
Aircraft. These estimates indicate that the ef-
fective flux will be about one half that of the
undisturbed beam hole. However, such estimates
are difficult to make, since the loop geometry is
quite complicated and the unperturbed flux pattern
is not well known. Accordingly, a flux-measuring
loop is being fabricated for early irradiation at the
MTR. This loop consists of the identical water
jacket and other components of the forward end
of the loop. The heat exchanger section will be

 

. N Burdick, Adsorption of Krypton and Xenon,
ORO-118 (Oct. 17, 1951).

25. H. Jury, Design of Percolators, ORNL CF-51-7-41
(July 9, 1951},

32

mocked up with approximately the correct ma-
terials. Cobalt foils placed inside and outside
the fuel tube and in a graphite bar along the heat
exchanger tubes will be counted after irradiation
to determine the flux. An exposure time of less
than 1 hr is planned.

Since this loop will not measure the flux de-
pression caused by the salt, a second and inde-
pendent experiment (cf., sec. 9, ‘‘Radigtion
Damage’’) is also being performed by the Solid
State Division. Small tubes, 0.269 in. ID by 7 in.
long, containing foil will be irradiated in the MTR
“rabbit’’ facility. The tubes will be filled with
various mixtures of material to match the effective
cross section of the fuel mixture NaF-ZrF -UF,
(53.5-40-6.5 mole %).

Horizontal-Shaft Sump Pump
J. A. Conlin

Aircraft Reactor Engineering Division

The first fused-salt test model of the in-pile
pump, which was described previously,® has com-
pleted a trouble-free 1000-hr endurance test. The
pump circulated 1400°F NaF-ZrF ,-UF, (53.5-
40-6.5 mole %) at 1 gpm in an isothermal loop
similar to the proposed in-pile loop.

Immediately after filling and priming, the pump
was operated at flows from 0.54 to 1.64 gpm and
found to be quite satisfactory. The pump, seal,
and drive motor are basically the same as those
proposed for in-pile use. However, in the final
in-pile pump design, the shaoft is cooled with
helium rather than oil, and the seals and bearings
are drop-lubricated.  These changes were in-
corporated to minimize the problem of oil disposal
in the loop assembly. A prototype pump is now
fully designed.

The priming difficulty previously reported® has
been solved by drilling a %z-in. vent hole between
the impeller cavity and sump in the vicinity of
the shaft. This priming difficulty was attributed
to a seal caused by the surface tension of the
water and the close radial clearance, 0.007 in.,
between the shaft and the pump casing. The seal
was strong enough to prevent venting of gas from
the pump and loop along the shaft and into the
sump.

A system has been proposed and tested with
water which will permit automatic filling and pump-

 

3). A, Conlin, ANP Quar. Prog. Rep., Dec. 10, 1954,
ORNL-1816, p 41.
sump:lewaly,control. It consists of a fill tank
connected to the pump sump by two tubes., One
tube, for filling, connects the bottom of the tank
to the bottom of the sump. A second tube, for
venting, extends from the top of the fill tank to
the pump. This latter tube enters the side of the
pump sump and is directed downward into the
sump, ending at the normal sump fluid operating
level. In operation, the pump and fill tank are
heated, with the fill line remaining frozen. The
fill line is then heated and fuel flows from the
tank into the sump and displaces gas to the fill
tank through the vent tube. When the sump level
reaches the vent tube, fuel is forced up the vent
until it balances the head in the fill tank, at which
time all flow stops. As the sump level lowers,
the bottom of the vent tube is uncovered, the
pressure balance is upset, and the sump again
fills.

In the test model it has been possible to control
sump level to within ]/] in. The operation of this
fill system is particularly sensitive to* vent-tube
design. The tube must have o gradual downward
slope from the fill tank to the sump, and, for best
results, have a sharp-edged flared end in the sump.
A unit is now being set up to test the system with
fused salts. In operation the fill line will be
frozen after transfer to eliminate the possibility
of additional fuel entering the sump.

Heat Exchanger

o L. P. Carpenter

Aircraft Reactor Engineering Division

The problems of sealing and shielding a reliable,
helium-recirculating cooling system for removal
of the heat generated in the in-pile loop are so
great that an air-cooling system appears to be the
least expensive in time and money.? Air will flow
from a compressor, through the loop heat ex-
changer, and discharge into the MTR pebble zone
through a second, little-used beam hole. A double-
walled heat exchanger is planned in which a tube
will be shrunk thermally around the salt-carrying
tube of the loop, and the tubes will be brazed
together at the ends. A third tube will form the
annular air passage, as in the original exchanger.
The double-walled tube is a safety feature in that
in the event of a break of the salt tube within the

 

4D. F. Salmon and L. P. Carpenter, ANP Quar. Prog.
Rep. Dec. 10, 1954, ORNL-1816, p 43.

PERIOD ENDING MARCH 10, 1955

heat exchanger, the radioactivity will be contained
within the second tube and will not enter the high-
A helical groove cut in the
inside wall of the second tube will direct any gas
leakage from the salt tube to an activity monitor.
Three tests of in-pile heat exchangers have been
run in which lengths equivalent to one half the
actual heat exchanger were used, Two exchangers
were of the single-tube-and-shell type that was
designed for helium cooling, and the third was a
double-walled tube type. The results of the tests
indicate that the required heat removal can be
accomplished with the double-walled tube, al-
though there is about a 20% loss in efficiency in
comparison with the single tube. The double-
walled heat exchanger removed 15 to 20 kw with
0.4 to 0.7 |Ib/sec of air and a temperature rise of
120 to 150°F. In the in-pile loop, the mass rate
of air flow will be about the same as that used
in the test, but the temperature rise at the exit
will be about twice that in the test, and therefore
heat-removal capacity will be about 30 to 40 kw.

velocity air stream.

PUMP DEVYELOPMENT
ARE-Type Sump Pumps
A. G. Grindell

Aircraft Rzactor Engineering Division

Performance tests of the ARE-type sump pump
were concluded. In these tests the pump charac-
teristics were obtained over the speed range 600
to 1400 rpm with the fluoride mixture NaF-ZrF -
UF4 (53.5-40-6.5 mole %) at 1300°F;> the critical
speed range was determined to be between 2850
and 3250 rpm; and the zirconium fluoride vapor
trap was demonstrated to be effective. The re-
liability test was terminated, despite continued
trouble-free operation, at a total operating time
of 3748 hr. The only interruptions in operation
were the shutdown for inspection at 2000 hr, and
a motor bearing failure at about 3000 hr (downtime,
21,5 hr). [t may be concluded from the tests that
the ARE-type sump pump, when used in con-
junction with a suitable vapor trap, may be
expected to operate with NaF-ZrF4-UF4 {53.5-
40-6.5 mole %) for about 4000 hr without trouble
at its design conditions: 1350°F, 1500 rpm, and
40 gpm.

 

5w, G. Cobb, A. G. Grindell, and W. R. Huntley, ANP
Quar. Prog. Rep. Dec. 10, 1954, ORNL-1816, Fig. 3.3,
p 44.

33
ANP PROJECT PROGRESS REPORT

Tests of the ARE-type sump pump were also
made to determine its suitability for delivering
NaK at 1400°F, 184 gpm, 300-ft head in the ART
intermediate heat exchanger tests. A preliminary
water test at 100°F indicated that the desired
conditions could be reached with the ARE-type
pump at a speed of about 3800 rpm (Fig. 3.4).
A hot test with NaK verified the preliminary test
data (Fig. 3.5). No life test of the pump at the
intermediate heat exchanger test conditions was
performed, but previous experiences indicate that
no serious trouble would be encountered during
the 1000-hr duration of the test. However, the
V-belt drive would be operating beyond its rated
speed limit, and it would probably have to be
replaced pericdically during the test.

Mechanical Shakedown Tests of ART Pump
(Model 1) Rotary Elements

A. G. Grindell

Aircraft Reactor Engineering Division

W. C. Snapp
Pratt & Whitney Aircraft

Included in the program for development and
testing of pumps for the ART circuits is a me-
chanical shakedown test of each rotary element
to be used. The objectives of the test are to
determine adequate assembly techniques, the ap-
propriate clearances, and the reliability of bearings
and seals and their lubricating and cooling equip-
ment. Particular emphasis is placed on assuring
that seal parts have no manufacturing defects and
that they have been properly assembled with
respect to the pump shaft and impeller. No load
is to be imposed on the impeller.

A mechanical shakedown test stand has been
erected, and two rotary elements have been tested.
Both elements were run at a constant 3800 rpm
with lubricating oil supplied at 2.0 to 2.5 gpm
at 150°F,

One rotary assembly (No. 1) was run for 96 hr
without vibration, bearing heating, or seal leakage,
and the test was shut down for disassembly and
inspection of the rotary element. No damage was
detected, The element was then reassembled and
the test restarted. A slight amount of oil leakage
past the lower seal was detected 22 hr after the
operation was resumed, and the test was termi-
nated at the end of 67 hr when the leakage rate
became excessive.

ORNL-LR-DWG 58Tt

 

400

    
    
     
 
   
 
      
 

3900 rpm

3600 rpm

 

 

300 e
S e
INTERMEDIATE HEAT
EXCHANGER TEST
DESIGN POINT

3300rpm

 

ON

T

; 3000rpm
o 200

<

w

T 2700rpm

2400rpm

 

100

 

 

 

 

 

o

100 200 300
FLOW RATE (gpm)

Fig. 3.4. Performance Characteristics of ARE-
Type Sump Pump in Tests with Water at 100°F,

ORNL-LR-DWG 5672
400 i

 

 
    
 
   

 

 

 

 

 

 

 

4000 rpm
36C0rpm
300 S
g
o
=4
£ 200
Q
o
oy
100 e —— - _— - - —_——— ]
0 ;, — - J__——L
o 50 {00 150 200 250
FLOW RATE {rpm)
Fig. 3.5. Performance Characteristics of ARE-

Type Sump Pump in Tests with NaK at 1400°F.
Inspection of the seal parts revealed that dis-
tortion of the bellows nosepiece had destroyed
the surface flatness of the seal face. A study
revealed that the material and hardness specifi-
cations for the bellows nosepiece precluded the
possibility of heat treatment for stress relief, and
the &'Portion was therefore attributed to stress
relief induced by test conditions. The bellows
nosepiece specification was rewritten, and new
seals were ordered.

The results of a test of a second rotary as-
sembly (No. 2) with seals manufactured to the
original specification verified those obtained in
the test of assembly No. 1. The total operating
time was 200 hr. Both these rotary elements will
be retested when seals manufactured to the revised
specifications are received.

DESIGN AND OPERATION OF FORCED-
CIRCULATION CORROSION AND
MASS TRANSFER TESTS

Operation of Fused-Salt«lnconel Loops
W. B. McDeoneld P. G. Smith

Aircraft Reactor Engineering Division

J. J. Milich R. A. Dreisbach
Pratt & Whitney Aircraft

Three fused-salt-Inconel forced-circulation
loops with large temperature differentials were
terminated after 774, 625, and 521 hr of operation,
respectively. In each case operation was termi-
nated short of the scheduled 1000 hr because a
leak developed; however, the operating periods
were long enough to provide valuable corrosion
and mass transfer data. The fuel used was
Nch-ZrF“-UF4 (53.5-40-6.5 mole %). The design
of the loop was illustrated in the previous report.®

The loop that operated 774 hr had a peak fuel
temperature of 1500°F, a maximum tube wall
temperature of 1740°F, a temperature differential
of 200°F, and a Reynolds number of 10,000. The
loop had to be terminated when a small leak
developed near one of the heater terminals. The
cause of this failure is not known. Because of
a plant power failure and a pump motor circuit
failure, the fuel had to be dumped twice during
this test.

The loop that operated 625 hr had a peak fuel
temperature of 1500°F, a temperature differential

 

5, A Mann, W. B. McDonald, and W. C. Tunnell,
ANP Quar. Prog. Rep. Dec, 10, 1954, ORNL-1816, Fig.
3.4, p 45.

PERIOD ENDING MARCH 10, 1955

of 300°F, and a Reynolds number of 10,000. The
maximum tube wall temperature in this test was
1610°F. The loop had to be terminated when the
pump (LFB model) seized because of a bearing
failure.

The third loop was terminated after 521 hr
because of a failure in the heater section. The
peck fuel temperature was 1500°F; the maximum
tube wall temperature was 1610°F; the Reynolds
number was about 15,000; and the temperature
differential was 200°F, During one period of the
test, the loop had to be operated isothermally for
69 hr because of a failure in the power supply
control. The results of examination of these
loops are reported in Sec. 6, ‘‘Corrosion Re-
search.”

The results obtained from these tests indicate
an undesirable temperature distribution in the
heater section. At the tube bends between the
heating elements, higher temperatures (as much
as 100°F) occurred on the inside of the bend than
on the outside. To determine the effect of the
uneven temperature distribution on corrosion, a
new loop in which the heating is accomplished
entirely in a straight section of tubing was put
into operation. A second loop incorporating the
old, coiled heater section was also started. A
zirconium-base fuel containing both UF4 and UF
is being circulated in these loops. The unit with
no bends in the heater is operating at a Reynolds
number of 15,000, a temperature differential of
200°F, and a maximum fuel temperature of 1500°F,
The other loop has a Reynolds number of 10,000,
a temperature differential of 300°F, and a maximum
fuel temperature of 1500°F.

In order to determine the effect of the method
of heating the loop on corrosion and mass transfer,
a gas-heated loop and an electrical-resistance
heated loop are being operated on a comparative
basis. They are both operating at a Reynolds
number of 1000, o temperature differential of
300°F, and maximum fuel temperature of 1500°F;
they are circulating the fluoride mixture NaF-

Zer-UI"'4 (53.5-40-6.5 mole %).

Sodium in Multimetal Loops

D. R. Ward W. B. McDonald
Aircraft Reactor Engineering Division

The fourth of a series of tests’ of sodium in
beryllium-inconel systems was completed. The

 

’D. R. Ward, L. A. Mann, and W. B. McDonald, ANP
Quar, Prog. Rep. Dec. 10, 1954, ORNL-1816, p 45.

35
ANP PROJECT PROGRESS REPORT

loop, which was made of Inconel and contained
a beryllium insert, was operated continuously for
1000 hr. The sodium was circulated at a maximum
temperature of 1300°F (at the beryllium insert)
and a temperature differential of 300°F. The
Reynolds number ot the minor diameter of the
beryllium insert was approximately 440,000.

A similar loop that is all Inconel is being
operated with sodium to provide a comparison with
loops containing other materials. Also, a loop
made of Inconel and type 316 stainless steel is
being operated. The steel is inserted in the cold
leg of the loop. The operating conditions for both
loops are a maximum temperature of 1500°F, a
minimum temperature on the cold leg of 1200°F,
and the Reynolds number in excess of 15,000.
The loops are scheduled to operate 1000 hr.

HEAT EXCHANGER TESTS
Intermediate Heat Exchanger Test No. 2

R. E. MacPherson
Aircraft Reactor Engineering Division

Design work is progressing on a heat exchanger
test loop to provide data on corrosion, mass
transfer, and reliability of a fuel-to-NaK-to-air
system operating under conditions comparable to
those postulated for the ART. The design will
incorporate two heat exchanger tube bundles con-
taining 100 tubes each in a regenerative type of
circuit. Component procurement and fabrication
are well under way, and it is intended that the
testing will begin in May.

The heat exchanger is basically similar to that
used in the first intermediate heat exchanger test.®
The NaK will flow from a 1-Mw gas-fired heater
through one tube bundle to a radiator. The stream
will then go to the liquid-metal pump and back
through the second tube bundle to the heater. The
fuel mixture NaF-ZrF4—UF4 (50-46-4 mole %) will
be circulated outside the tubes countercurrent to
the NaK flow and will be alternately heated and
cooled by the NaK stream.

A single tube bundle is composed of 100 Inconel
tubes 3/] in, in outside diameter, 0.017 in. in
wall thiciness, and approximately 6 ft in length.
These tubes are arranged in a 10 by 10 matrix
and contained in a square channel on 0.210-in.
centers with a 0.011-in. tube-to-wall clearance.

 

8'R. E. MacPherson and H. J. Stumpf, ANP Quar. Prog.
Rep. Dec. 10, 1953, ORNL-1649, Fig. 2.2, p 29.

36

The operating conditions are indicated on the
flow diagram, Fig. 3.6. The fuel mixture specified
for this test will restrict the attainable fluoride-
side Reynolds number to approximately 26G0.

o
ORNL-LR-DWG 5673

 

80" F 700° F

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

AIR IN — ) ——————— - AR OUT
154 psia - Mw 15 psio
NaK-TO- AR
1099° F RADIATOR 1279°F
“- 20 psic 30 psia A
ol &
e | o
o3 | =
Ztl ' E
FUEL
PUMP
w o - —_— -
» a 1 37 qpm wj{ u|o
| uie o 3 % | @
ol wla gl Slo
@fs Tl SR
\' | -
Lo | |
—
s 1] wl g8 |
| & l 'U < I ZZ
) 4 <X - |
[ 83 e mul ] of I
I ox 1 [Th el TRy
| W | w w| | aw |
g 1] = 2o
l I I I
e A e A
O g wie 08 a wio
Il BB ale o lg
1~ 2la -l 2B
2o n g
Y ai0tF 1590° F
74 psia 1~ Muw 67 psia
GAS-FIRED 80° F AIR IN
COMBUSTION 1000° E HEATER /,’ 30 psic 128 Ib/sec
PRODUCTS - <
ouT psia T~ 80°F GAS IN
30 psia  O.074 Ib/sec
Fig. 3.6. Flow Diagram of Loop for Testing

Intermediate Heat Exchanger No. 2 with NaF-ZrF .
UF, (50-46-4 mole %) on the Fuel Side at a
Reynolds Number of 2510.

Small Heat Exchanger Test
R. E. MacPherson

Aircraft Reactor Engineering Division

A small-scale heat exchanger test is now being
set up to provide a facility for investigating heat
transfer characteristics through the Reynolds
number range 0 to 5500 on the fluoride mixture
side of the fuel-to-NaK heat exchanger. The heat
exchanger to be used consists of 20 Inconel tubes,
3/16 in. in outside diameter, 0.017 in. in wall
thickness, and approximately 6 ft in length. These
tubes are arranged in a 4 by 5 matrix and contained
in a square channel on 0.22-in. centers with a
0.032-in. tube-to-wall clearance,

The operating conditions are indicated on the
flow diagram, Fig. 3.7. The fluoride mixture
NcF-ZrF4-UF4 (50-46-4 mole %) will be used. In
order to attain the Reynolds numbers desired in
this test, it has been necessary to increase the
tube center-to-center spacing and the wall clear-
ance in comparison with the spacing and clearance

QAN -LR-DWS 5674

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

° 200-kw ©
1441°F NokoTom AR 1344 ‘F
23 psia RADIATOR 20 psia
o
W
a
O
ol
NaK
PUMP
g
oy
@
[{s]
T
144 F 1386°F Y
, T35 oo
24 psia FUEL-TO-NaK 5 psia
'500° F HEAT EXCHANGER 1400 F
95 psia 24 psia
i
W
o
[
ol
Y
FUEL
PUMP
o
)
o
Q
1500° F 200-kw 1400° F \
, ELECTRIC _
96 psia HEATER 112 psia
Fig. 3.7. Flow Diagram of High-Velocity Loop

for Testing a Small Fuel-to-Nak Heat Exchanger
with NaF-ZrF -UF, (50-46-4 mole %) on the Fuel
Side at Reynolds Numbers of 0 to 5500.

PERIOD ENDING MARCH 10, 1955

specified for the intermediate heat exchanger
No. 2. In addition, the regenerative feafure is
not being employed so that the fluoride and NaK
circuit resistances can be kept within the head
range of available pumps (ARE type). The circuit
will consist of a 200-kw resistance heater (to heat
the fluoride stream), the heat exchanger, and a
NaK-to-air radiator for heat removal.

Gas-Fired Heat Source

R. E. MacPherson
Aircraft Reactor Engineering Division

R. Curry
Pratt & Whitney Aircraft

As a step toward development of a 1-Mw gas-
fired NaK heater for use in the intermediate heat
exchanger test No. 2, ¢ small-scale prototype
(100 kw) has been designed and is being as-
sembled. This unit was initially planned to test
the principles to be utilized in a larger heater,
but it has since been decided to fill the need for
high-capacity heat sources from outside vendors.
However, assembly and testing of the 100-kw
gas-fired heater are being continued to fill the
need for small utility heaters in many phases of
the development program.

The heater utilizes the Esso-type burner
a countercurrent NaK-to-flue-gas heat exchanger
with combustion air preheating provisions as
shown in Fig. 3.8. The unit, as designed, will
operate at an efficiency of 30 to 40%. No attempt
has been made to utilize the combustion gases
leaving the heat exchanger to preheat the com-
bustion air, since all the preheating that is desired
is accomplished in cooling the container wall of
the heat exchanger section. It is visvalized that
a full-scale unit could have considerably higher
efficiency, since the percentage of available heat
escaping through the container wall could be
appreciably less and a flue-gas economizer could

9 and

be used.

REACTOR HAZARDS TESTS
J. Y. Estabrook L. A. Mann

Aircraft Reactor Engineering Division

Tests were made to evaluate the integrity of
the cell designed to contain the Aircraft Reactor
Test and the effects of a reactor accident. An

 

9L. A. Mann and R. Curry, ANP Quar. Prog. Rep.
Dec. 10, 1954, ORNL-1816, Fig. 3.7, p 50.

37
ANP PROJECT PROGRESS REPORT

UNCLASSIFIED
CRNL.-L.R-DWG 5675

 

T
Al !

 

    
 
      

FROM GAS SUPPLY

 

 
 
 
 

ESSO-TYPE
NATURAL GAS
BURNER

7 Na INLET

 

il
£

 

 

 

 

 

 

 

COMBUSTION AIR AT
PREHEATER INLET “ T =

 

 

| TO sTACK

   

 

 

 

 

 

 

 

 

L

 

 

 

Fig. 3.8. Apparatus for Testing 100-kw Gas-Fired Heat Source.

accident simulated in which the reactor
pressure shell was ruptured during power operation
and the fuel was spilled into the cell. The appa-
ratus used in the test is shown in Fig. 3.9.

The test was performed in two parts. In the
first part, the spill was contained by the inner
wall of the double-walled cell (water-filled an-
nulus), and the afterheat was dissipated in boiling
of the water in the annulus.

To simulate the accident, 414 b of the fluoride
mixture NaF-KF-LiF (11.5-42-46.5 wole %) was
heated to 1500°F and dumped into the bottom of
the tank representing the inner wall of the cell;
the tank was surrounded by water at 185°F. The
salt mixture used has about twice the heat ca-
pacity of the proposed ART fuel, but the quantity
involved represented only about one-fourth the
volume of the ART fuel. Six thermocouples
attached to the outside bottom of the tank recorded
the temperature rise at the center and five other
points spaced at intervals of 6 in. from the center
outward. The water adjacent to the tank bottom

wdas

38

boiled briskly but not violently enough to shake
the tank visibly. Within 75 sec, the center thermo-
couple reading peaked at 940°F and fell off rapidly
as the salt mixture cooled (Fig. 3.10). No other
thermocouple read more than 300°F throughout the

test. The cooling rate shown by the recording
potentiometer indicated that the over-all heat
transfer rate would be adequate to dissipate

3.5 Mw of continuous afterheat if the center of
the bottom of the cell were modified to prevent
fuel from contacting that portion of the surface.

In the second part of the test, the reactor and
the inner wall of the cell were assumed to have
ruptured and permitted the fuel and the water in
the annulus to mix. To simulate the radioactive
reactor fuel, 2 kg of UF, was added to 505 b of
the fluoride mixture, and the mixture was heated
to 1150°F. An gluminum capsule containing 99 g
of N°2UF6 (59 g of normal uranium) that had been
irradiated in the ORNL Graphite Reactor was then
put into the melt, and the temperature was raised
to 1500°F. This simulated fuel was then injected
6¢

UNCLASSIFIED
ORNL-LR-DWG 5331

 
   
 
    
   
 
 
  
 
 
 
  
    
 

TRANSFER LINE

R
PRESSURE WATER LEVEL

TEST NO.2
*GASBLEED
PRESSURE

REGULATORS
HELIUM __
SUPPLY

PRESSURE GAGE (\
5 ft

WATER LEVEL

GAS SURGE TANK TEST NO.{

GROUND LEVEL

THERMOCOUPLES (WATER TIGHT}

Fig. 3.9. Facility for Testing Integrity of Reactor Cell in the Event of Rupture of the Reactor Pressure Shell.

S

SS61 ‘01 HOYVW ONIGN3 aol¥3d
ANP PROJECT PROGRESS REPORT

__
ORNL-LR-DWG 5677

 

  
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

   

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

1000 1 ’ \ { \
| ; ' !
‘ -COOLING RATE, ~2.6°F/sec
900 —- ‘ | b -
| /
800 ] S 1 -
w / ‘ ‘ 1 | :
e ‘ |
w i -
E.J 700 ‘ - | . e e
2
o i
Q
g \R
o 600 . ™~ - —
W ! |
I
- | AN
EJ 500 =~ - ’ - 1T - TN ‘
= .
Lt i 1
U '
X 400 - '
COOLING RATE,
& ~5.2°F/sec
2 ‘ !
Y o0 o — e - — — -
!I v
'%J | \&—_‘_-
W
200 } —
100 ] : - ; - e —
0l ; |
C 30 60 20 120 150 180 210 240 270 300 330 360

3%0

TIME AFTER START OF TRANSFER (sec)

Fig. 3.10.

Rate of Temperature Rise of Outer Surface ot Bottom Center of Reactor Cell When inner

Surface Was Flooded with a Fluoride Mixture at 1500 °F.

into the test tank (Fig. 3.9), which had been filled
with water to simulate the effect of a rupture of
the inner wall of the cell,

The time required for injection of the fuel was

about 50 sec. The physical agitation of the water,
tank, and transfer line (recorded by motion picture

40

camera) was violent during the injection, but no
explosion occurred. A negligible amount of steam
reached the surface of the water, and no entrain-
ment of water into the air was observed. No
measurable radicactivity was found outside the
steel tank. A water sample from the tank indicated
a total of 1 to 2 millicuries of activity.
PERIOD ENDING MARCH 10, 1955

4. CRITICAL EXPERIMENTS

A, D. Callihan
Physics Division

REFLECTOR-MODERATED REACTOR

D. Scott B. L. Greenstreet
Aircraft Reactor Engineering Division

R. M. Spencer
United States Air Force

J. J. Lynn D. V. P. Williams
Physics Division

J. S. Crudele E. V. Sandin
J. W. Noaks
Pratt & Whitney Aircraft

The current study of the nuclear characteristics
of the reflector-moderated aircraft-propulsion re-
actor was to include, as described in preceding re-
ports,! three critical assemblies of increasing
complexity approaching a structure similar to that
of the designed reactor. As the complexity has
increased, the assemblies have been successively
less amenable to calculation by using the presently
known multigroup methods and nuclear data. The
experiments have, therefore, provided results that
have wvalidated or extended reactor analysis
methods, and they are, more recently, providing
data for evaluating empirically the effects of certain
structural features of the reactor. In all these
assemblies the reflector is beryllium and the fuel
region contains alternate laminae of 93.2% enriched
uranium metal, 0.004 in. thick, and Teflon (CFz)n.
This critical experiment fuel satisfactorily repre-
sents the mixture of fluorides comprising the re-
actor fuel, and the thickness of the uranium and
of the Teflon sheets allows considerable flexibility
in the U23> concentration in the various loadings.
The first assembly consisted of an essentially
spherical fuel region surrounded by the reflector.
In the second, of which there have been three
variations, the fuel surrounded a core, or island,
of ‘berylium, also spherical, and was enclosed by
the reflector. The fuel was separated from the
island and from the reflector by metal core shells
to provide stability and, in the more recent experi-
ments, to represent the reactor structure. This
three-region assembly has been modified in the

 

A. D. Callihan et al.,, ANP Quarn Pro§' Re .,
ORNL-1692, p 45; ORNL-U?] p 44; ORNL-1816,

third experiment by the addition of a column of
beryllium to each side of the island and of cylindri-
cal shells of fuel to the former spherical fuel
shell to simulate the so-called ““end ducts.”” A
cutaway view of the assembly with the end ducts
is shown in Fig. 4.1. These end-duct additions
represent the inlet and outlet flow channels of the
reactor,

The data from the first experiment and some of
the critical parameters of the second experiment
have been presented.! The results from the second
experiment and preliminary data from the assembly
with the end ducts are presented here. The
dimensions and contents of all the assemblies are
summarized in Table 4.1, and a comparison is made
with the results of critical mass calculations by a
multigroup method.?2 The present fuel loading of
the assembly with the end ducts is excessive by
an amount too great to extrapolate with certainty
to a critical system with all poison rods removed.
The immediate program includes a reduction of the
loading so that an evaluation of the critical mass
can be made,

The distributions of neutrons detected by indium
activation, with and without cadmium covers, along
a vertical traverse in the midplane of the three-
region assembly with aluminum core shells (CA-20q)
is shown in Fig. 4.2, The cadmium fraction, that
is, the fraction of the neutrons having energies
below those absorbed by 0.02-in.-thick cadmium,
is also shown as a function of the radius. These
data have not been corrected for the absorption of
indium-resonance neutrons by cadmium. The calcu-
lated? bare and cadmium-covered indium activation
traverses are also shown. Figure 4.3 gives the
neutron distributions obtained from activated gold
toils also located on a vertical traverse in the
midplane of the assembly with the aluminum core
shells (CA-20a) and of the assembly with the
]/s-in.-thick Inconel shells (CA-20¢c). From these
data the fraction of neutrons with energies below
the cadmium cut-off in the center of the fuel region
of CA-20a is about 0.2, The corresponding value
for CA-20¢, with considerable uncertainty, is 0.1,

 

W. E. Kinney, private communication.

41
ANP PROJECT PROGRESS REPORT

   

  

ALUMINUM
SHEETS

ALUMINUM SHEETS

BERYLLIUM

CORE SHELLS

ALUMINUM
EXTRUSIONS

.

ORNL-LR-DWG 3689A

ALUMINUM
SHEETS

    
 

  

  
    
 

REGION

 
 
  

NN

REGION

    
   
 

  
  

   

TS
AT

\\ ) ».‘\“ N N N
\\\\ \\\\\\\ A ¥

  

“-PLANE OF TABLE SEPARATION

Fig. 4.1. Longitudinal Cross Section of Three-Region Critical AssemBIy with End Ducts.

The fission-rate distributions across the fuel in
the three modifications of the three-region assembly
are given in Fig. 4.4, The data were obtained
from the activity of fission products collected on
aluminum foils in contact with fuel sheets. Pairs
of points at the same abscissa value give the
results from opposite sides of a single 0.004-in.-
thick foil. From these and similar data from
cadmium-covered uranium-aluminum foil combi-
nations, it has been possible to plot the cadmium
fraction (that is, the fraction of all fissions pro-

42

duced by neutrons of energy below about 0.5 ev)
across the fuel for the assembiies with the aluminum
core shells and the ]/B-En.-thick Inconel shelis,
The lower values for the latter assembly (CA-20c)
are a consequence of the nuclear properties of the
Inconel and of the higher uranium density. This
variation in fission rate in the fuel sheets is also
shown in Fig. 4.5, where the results of self-
shielding measurements within the 0.004-in.-thick
uranium foils are plotted. A fuel sheet near the
beryllium island was replaced by four sheets, each
TABLE 4.1,

PERIOD ENDING MARCH 10, 1955

COMPOSITION OF REFLECTOR-MODERATED CRITICAL ASSEMBLIES

 

Three Region with Core Shells

 

Three Region
wi th l’B-in.-Thick

 

 

Two Region ]/lé-in.TThick ]/lé-in.-Thick ) «in.~Thick Core Shell's and
Aluminum Inconel Inconel
End Ducts
¥
Assembly number CA-19 CA-204q CA-20b CA-20c CA-21 %
Beryllium island *
Volume, ft3 None 0.37 0.37 0.37 1.27
Average radius, in. 5.18 3,18 5.18 4,20 (duct core
radius)
Mass, kg 19.4 19.4 19.4 67.0
Fuel region (excluding shells
and interface plates)
Volume, 3 1.05 1.78 1.78 1.72 2.06
liters 29,7 50.4 50.4 48.8 58.3
Average radius, in.
Inside 5.24 .5,24 5.31 4,33
Outside 7.48 9.51 9,51 9. 44 5.28}("“‘ radii)
Distance between fuel 0.173 0.639 0.284 0.142 0.142
sheets, in.
Mass of components, kg
Teflon 54,46 99.38 99.27 94.37 108.88
Uranium loading 11.88 5.00 11.74 22,07 26,02
U233 1oading 11.07 4.66 10,94 20.56 24,24
U233 density,* ¢/cm’ 0.372 0.092 0.217 0.421 0.416
Uranium coating 0.11 0.05 0.11 0.20 0.25
material
Scotch tape 0.09 0.1 0.11 0.1 0.15
Core shells and interface
plates
Mass of components, kg
Aluminum 0.92 5.85 1.10 1.10 1.10
Inconel 0 0 13.68 27.73 53.02
Reflector
Volume, f° 21.24 22,22 22,22 22,22 20,88
Minimum thickness, in, 12,9 11.5 11.5 1.5 11.5
Mass of components, kg
Beryllium 1102.5 1155.0 1155.0 1155.0 1094.1
Aluminum 16.8 29.2 29.2 29.2 29.2
Excess reactivity as loaded, % 0 0.9 0.3 0.4 ~3
Critical mass,** U235, kg
Experimental 11.Q7*** 4,35 10.8 19.8 19 + 2
Cal culated 9.5 to 4.5 11,3
10.3
*y235 s per unit volume of fuel region.

**Mass required for a critical system with poison rods removed.

***|t was necessary to increase the reflector thickness slightly to make this mass critical,

43
RELATIVE ACTIVITY (arbitrary units)

ORNL—LR—DWG 5778

 

 

 

28

 

 

 

 

 

 

 

 

 

22

 

N
Q

 

 

 

 

 

 

 

 

 

 

N

o

 

32
BERYLLIUM FUEL_ REGION BERYLLIUM
26 ——r e
EXPERIMENTAL)
24 e —
4}, :
i |
CALCULATED BARE INDIUM ACTIVATION
| I
|
18 | -
16 - - ‘
CADMIUM - COVERED INDIUM
4 ACTIVATION (EXPERIMENTAL) /
t.0
8 0.8
6 0.6
4 0.4
CADMIUM FRACTION
2 - 0.2
CALCULATED CADMIUM-COVERED INDIUM ACTIVA
0 0
C 1 2 3 4 5 6 T B 9 10 1 12 13 14 15 16 17 18 19 20 21 22

DISTANCE FROM REACTOR CENTER {in.)

Fig. 4.2, Radial Neutron Distribution in Midplane of Three-Region Assembly with Aluminum Shells (CA-20aq).

140d3Y SS§3YI0Yd LDIFr0y¥d dNV

CADMIUM FRACTION
RELATIVE ACTIVITY (arbitrary units)

RELATIVE POWER (arbitrary units)

PERIOD ENDING MARCH 10, 1955

ORNL—LR—-DWG 5779

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

   
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

] } ‘ 1
e —SERYLLIUM ISLAND —3=—|=€—FUEL ANNULUS ~meftt———— e BERYLLIUM REFLECTOR -
| :
— — —
Vg-in=THICK 1/ ~in-THICK
ALUMINUM INCONEL
SHELLS SHELLS
o0
BARE FOILS O °
CADMIUM-COVERED FOILS A A
® | : |
0.8 ‘ P -
%\
06 \
T
‘l\ * \\
~ y/4 e
\ . \
N X
°
° 2
0.2 A®, AL vl —O o
®. -/‘/ .\
—
-'-l—‘-/
0 2 4 6 8 10 12 14 16 18 20 22
DISTANCE FROM REACTOR CENTER (in.)
Fig. 4.3. Radial Neutron Distribution in Midplane of Assembly from Gold Foil Activation.
05 ‘ l | ‘ ORNL-LR-DWG 5780
TR
— BERTLILM FUEL REGION ir 8ERYLLIUM REFLECTOR | |
‘ |
| FISSION O 'ig-in-THICK ALUMINUM CORE SHELLS
‘ al RATE 0 Yg-in.-THICK INCONEL CORE SHELLS
A i DISTRIBUTION| A Yg-in.- THICK INCONEL CORE SHELLS
10— @ CADMIUM  |— 'g-in-THICK ALUMINUM CORE SHELLS
] | FRACTION  |—— Yg-in - THICK INGONEL CORE SHELLS
? NOTE: THE HALF SYMBOLS INDICATE THAT MEASUREMENTS
/ WERE MADE ON BOTH SIDES OF SOME URANIUM FOILS, |
\, THE OPEN SIDE OF THE SYMBOL BEING DIRECTED
xe‘;/‘/( TOWARD THE FOIL.
i | ]
0.5 A A | 1 _
3 /| i
\ : a_/fl/ \ 8
‘%\ v _ 2
o | ] 5
1 : 10 <
L - | ‘ ----- — 0.8 &
L N TTTH— 0/ — P . — 106 s
— ~ L | i } : - .4 g
\_\_ | ,_—/r ‘ ‘ ! ‘ ‘ — 0.2 A
0 T : . 0 g
0 1 2 3 4 5 6

DISTANCE FROM INNER EDGE OF FUEL ANNULUS (in.)

Fig. 4.4. Fission-Rate Distribution Across Fuel of Three-Region Assemblies.

45
0.001-iM%#htck, and the catcher-foil activity was

measured at each sheet. The experiment was

ORNL-LR-DWG 5784

 

 

 

 

 

 

 

 

| |

® DATA OBTAINED 043 in. FROM
ISLAND-CORE INTERFACE

i ——r-otp— -

4 DATA CBTAINED 171 in. FROM
ISLAND-CORE INTERFACE

| | .
1

t

. N
O 1 2 3 4 5 5 7
URANIUM THICKNESS {(mils)

 

 

 

 

RELATIVE ACTIVITY (arbitrary units)

 

 

 

 

 

 

 

 

Fig. 4.5. Self-Shielding in 0.004-in.-Thick Urani-

uvm Foil of the Aluminum Shell Assembly (CA-20q).
Wt 2o

46

repeated in a plane near the center of the fuel
region, and a more uniformly distributed fission
rate through the centrally located fuel was ob-
tained. These results, which agree at least quali-
tatively with the neutron-flux distribution measure-
ments in Fig. 4.3, are indicative of the spectral
differences in the neutrons in the fuel,

Some preliminary investigations have been made
of the suggested use of the rare-earth elements as
neutron poisons in reactor control rods, The
reactivity coefficients of several wafers® of Gd,0,
and Sm,0,, each about 0.01 in. thick and 4 em 2in
area, have been measured at two places in the
]/g-in.-ihick Inconel shell assembly (CA-20¢), and
little difference between the two materials has
been observed, At the center of the island, where
the cadmium fraction measured with gold foils is
0.5, the reactivity coefficients range from 0.10
¢/mg/em? for a 0.0l-in.-thick sample to 0.04
¢/mg/cm? for a 0.04-in.-thick sample. At the edge
of the island, where the cadmium fraction is 0,35,
the corresponding values are about 0.015 and
0.009 ¢/mg/cm?. The measurements show that a
sample of these materials a few thousandths of
an inch thick is 85% ‘‘black’’ to neutrons having
the spectrum they have in this experiment.

 

3F‘repc|red by J. R. Johnson of theMetallurgy Division.
Part |1

MATERIALS RESEARCH
PR 5. CHEMISTRY OF MOLTEN MATERIALS
W. R, Grimes

Materials Chemistry Division

Solid phase studies in the NQF-UF4 and NaF-
ZrF ,-UF, systems were continued, and phase
relationships and UF, solubility in the BeF,-
bearing systems were studied. The usefulness
of LaF, as a ‘‘stand-in” for UF, was demon-
strated. Extensive experimental studies of the
preparation and stability of UF, in alkali-metal-
and zirconium-base fluorides are reported. All
measurements made to date on the stability of UF,
in molten systems and on the feasibility of prepa-
ration of UF, by reduction of UF, have shown
poor reproducibility,

The 250-lb production facility, which was taken
out of service on January 1, was reactivated to
supply increased demands for zirconium-bearing
fluoride mixtures. The pilot-scale facility has
been used for developing a process for preparing
BeF ,-bearing mixtures and for the preparation of
various UF -bearing mixtures. The charge ma-
terial for an in-pile loop was also prepared.

PHASE EQUILIBRIUM STUDIES
C. J. Barton

Materials Chemistry Division

H. Insley
Consultant

Solid Phase Studies in the I"lch-UF4 and
NuF-ZrF4-UF4 Systems

R. E. Moore L. M. Bratcher
R. E. Thoma

Materials Chemistry Division

NoF-UF,. The gradient quenching technique
prevmusly descrsbedl was used to study phase
relationships in the NaF- UF4 system. Although
this system had been studied earlier in this
laboratory? and by others® by using thermal
analysis, visual observation, and quenching tech-
niques, it was never felt that o satisfactory under-

 

]C. J. Barton et al.,
1954, ORNL-18164, p 57.

2C. J. Barton et al., ANP Quar. Prog. Reps., ORNL-
858, p 14; ANP-60, p 128, ORNL-1692, p 52; ORNL-
1729, p 41; ORNL-1771, p 55,

3c. A, Kraus, Phase Diagrams of Some Complex Salts

of Uranium with Halides of the Alkali and Alkaline
Earth Metals, M-251 (July 1, 1943).

ANP Quar. Prog. Rep. Dec. 10,

standing of phase relationships in this system had
been achieved, particularly in the 25 to 50 mole %
UF, region. As a result of x-ray and petrographic
studies of quenched samples it was concluded
that the cubic phase that Zachariasen* postulated
to be a high-temperature form of Na,UF,, with a
homogeneity range extending to 40 mole %, is
actually a new compound Na U.F._ having «
rather narrow temperature range of stability. Below

630°C it decomposes to form (3,-Na,UF, and
Nao U F,,. At 672°C it melts incongruently to
give fi\la U F.,. and liquid. A revised diagram

that mc(udes the results of recent studies, as
well as some of the earlier thermal analyses and
visual observation results, is shown in Fig. 5.1.
Liquidus temperatures determined by the quenching
technique for mixtures having 27 to 40 mole % UF,
were in excellent agreement with the recently
presented partial phase diagram for this region.’
Two crystalline forms of Na,UF, reported by
Zachariasen? and observed in earfler quenching
work with this composition have not been observed
in recent studies; since their tfemperature range
of stability is not known, they were omitted from
the diagram.

NuF-ZrF4-UF4. The recent studies of the
NaF-UF, system have contributed materially to
a better understanding of phase relationships in
the ternary system. A number of gradient quenches
were carried out in an effort to define primary
phase fields more accurately. The results of
these experiments, presented in Table 5.1, seem
to indicate an invariant point near the composition
66 mole % NaF-9 mole % ZrF,-25 mole % UF,
where the primary phase fields of Na U F”,
N03U(Zr)F7 solid solution, and Na,U( r)6 11
solid solution meet. Invariant pomts are also
indicated near the composition 69 mole % NaF -4
mole % ZrF4—27 mole % UF4 where the N02UF6,
Na U(Zr) solid solution, and Na U,F . primary
phase fields meet and near 65 mole % NaF-28
mole % ZrF4—7 mole % UF4 where the Nazsz6,
Na,Zr(U)F_, and Na Zr(U)6 41 Primary phase

7!

 

4W. H. Zachariasen, J. Am. Chem Soc. 70, 2147
{1948).
5('.'. J. Barton et al.,

ANP Quar. Prog. Rep. Sept. 10
1954, ORNL-1771, p 55.

49
 

    

 

 
  
 

 

 

     

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

ANP PROJECT PROGRESS REPORT
ORNL-LR-DWG 5856
1100 l !
1000 — —
900 |- —_——— — —
, LBy=Na,UFg + LIQUID ; i
i / ! UF, + LIQUID
! - - L .
800 e X - 7 NagUzFyp+ B3-NayUFg ——
_ NaF + LIQUID | / ‘
o \ 3 f Nogll3F(, + LIQUID | | |
w i L Na,UgFyy + LIQUID j i |
E 00l % | /o o ]
g 700 : a-NagUF, + LIQUID | |
. [
; \ - {‘107U6F3'
+
‘ |
' : No U.F
i 5-3 17
600 | ——— ————fa-NagUF, 1 T | - | —— e __‘_/
NaF + a-Na,UF, \ + ‘ i ‘ w\
£3,-Na,UF ‘ ‘
3 2¥' 6 !
| - B5-Na,UF | |
+ ‘ i
-Na,UF
500 |[——— - - B-nazuFy L No,UgF s, — - MU o 4 UF —
NoF + 35-NazUF, + c7UgF 3 4
| | B5~Na, UF
| | |
400 | ———— - 1 s S e
Naf + 3,-Na,UfF g W W wo w :
:)m :)N :}m | Dh
° Q ] : o !
z = 2 z ;
300 l | J ‘ E l i
NaF 10 20 30 40 50 60 70 80 90 UF,
UF, (mole Yo

Fig. 5.1. Phase Diagram of the NaF-UF, System.

fields meet. The NaBZr4F19 primary phase field
boundaries have not been accurately located, as
yet. It is expected that a revised phase diagram
for this system will be completed in the near
future.

Phase Relationships and UF; Solubility in
BeF ,-Bearing Systems

L. M. Bratcher B. H. Clampitt

R. J. Sheil R. E. Thoma
Materials Chemistry Division
NaF-BeF .LiF. Data on the solubility of UF,

in NaF-BeF. mixtures obtained at another instal-
{ation® showed that it is necessary to go to low
concentrations of BeF. in this system before
usefully high concentrations of UF, dissolved in

50

the melt at 600°C can be obtained. It has also
been demonstrated that lower viscosities are
obtained as the BeF, concentration is reduced.
These developments encouroged a re-examination
of the ternary system, with emphasis upon the
LiF-LizBeF4-NazBeF4-NoF section of the system.
Phase diagrams for the LiF-BeF, and NaF-BeF,
binary systems have been published.7:8:% Un-

published thermal analysis data for the ternary

 

6Private communication, J. F. Eichelberger, Mound
Laboratories, to W. R, Grimes.

’D. M. Roy, R. Roy, and E, F, Osborn, . Am. Ceram.
Soc. 36, 185 (1953); op cit., 37, 300 (1954).

8A. V. Noveselova and M. E. Levina, J. Gen. Chem.
U.S.5.R. 14, 385 (1944).

7E. Thilo and H. A, Lehmann, Z. anorg, Chem. 258,
332 (1949).
PERIOD ENDING MARCH 10, 1955

TABLE 5.1, RESULTS OF QUENCHING EXPERIMENTS IN THE NGF-ZrF4-UF4 SYSTEM

 

 

 

Compos tion (mole %) Liquidus Secondary

Temperature Peimary Phase Temperature Secondary Phase
NaF ZeF, UF, °C) Q)
66.7 10.3 23.0 640 Na,U(Zr)F,*
66.7 6.3 27.0 646 Na UsF 640 Na ,U(Zr)F *
66.7 2.3 31.0 657 Na U,F .. 648 Na,UF
61.5 6.5 32.0 667 Na,U(Zr)F4,* 624 Na UsF (s
64.0 6.5 29.5 654 Na U.F, 613 Na U(Ze)F 4, *
64.0 11.0 25.0 640 Na,U(Zr) Fyy* 627 Na UsF
62.5 11.0 26.5 669 Na,U(Zr) F 4 p* 635 Na UsF o5
63.5 31.5 5.0 577 Na,ZrF, 567 Na, Ze(U)F 44 *
61.5 32.5 6.0 575 Na,Zr(U)F 4 * 547 Na, ZrF
59.5 33.5 7.0 580 Na_Zr(U) F 5 *
66.7 30.3 3.0 644 Na,Zr (U)F ,*
66.7 26.3 7.0 660 Na,Zr(U)F*
45.0 53.0 2.0 528 Na,Zr F o 513 Na,Zr(U)F 4 *
41.0 55.0 4.0 530 NayZr,F g
42.0 48.0 10.0 600 Zr(U)F > 553 Na,Zr(U)F
42.0 50.0 8.0 621 Ze(U)F
42.0 51.0 7.0 623 Zr(U)F *

 

*Solid solution.

system were obtained earlier in this laborctory.!?
These data showed low thermal effects, below
300°C, for a number of compositions, but eutectic
compositions could not be determined. Published
data on the binary systems show that compositions
in the ternary system near the LiF-Li BeF, eu-
tectic (3] mole % BeF2; melting point, 460 * 5°C)
offer the most promise of low melting points with
low BeF . concentrations. The compound Na_ Bef
melts at about 595°C, while the l\lql'—'-l\lc:,BeF4
eutectic (30 mole % BeF,) has a reported? melting
point of 570°C. Recently published datal! on
the ternary system show the existence of three
ternary compounds, NuLiBeF4, Noa(BeF4)2, and
NazLiBezFr The first compound apparently does

 

]OJ. P. Blakely, L. M, Bratcher, and C. J. Barton,
unpublished data.

”W. Jahn, Z. anorg. u. allgem. Chem. 276, 113
(1954); op cit., 276, 274 (1954).

not exist at liquidus temperatures, while the last
was reported to separate from the melt, together
with LiF, at a eutectic temperature of 320°C.
Preliminary thermal analysis results for ternary
compositions with 31 mole % BeF_, or less, seem
to confirm earlier indications that conventional
thermal analysis techniques do not give reliabie
liquidus temperatures in this system. Other tech-
niques, such as visual observation, filtration and
quenching, will be applied to the study of these
mixtures as soon as possible.

LiF-BeF -UF,. Attempts to prepare the LiF-
BeF,-UF, mixture, either by adding UF,; to
purified iiF-BeF2 compositions or by reduction
of UF, with excess uranium metal in similar
solvents, have yielded material with considerable,
and rather variable, concentrations of UF4. The
experimental evidence suggests that alloying of
uranium with the nickel equipment and the conse-
quent lack of control of the uranium activity may

31
ANP PROJECT PROGRESS REPORT

be responsible for the lack of reproducibility of
the data,

The considerable scatter in the data precludes
firm statements as to UF, solubility in this
medium. It appears, however, that the LiF-BeF
mixture containing 31 mole % BeF, will dissolve
at least 2 wt % of U3* at 600°C. Petrographic
examination of these materials shows the UF_ to
be present as large well-formed crystals, which
appear to have been deposited from solution. No
complex compounds of UF3 appear. lhermal data
on such mixtures suggest very low solubility of
UF, at the solidus temperatures.,

BeF -UF,. A 50-50 mole % mixture of BeF,
and UF, is the only mixture of these materials
that has been examined to date. No distinct
thermal effects were found on the cocoling curve,
the maximum temperature being about 900°C.
Petrographic examination of the preparation
showed UF3, crystalline colorless BeF2, and
BeF, with a yellowish color. Chemical analysis
of the material showed 95% of the uranium to be
in the trivalent form. The remaining tetravalent
uranium is presumed to be due to oxidation of a
part of the U3* by oxidizing impurities in the
BeF,, such as H20 and BeSO,. It appears from
this preliminary experiment that UF, and BeF,
do not form a compound and that solid solution,
if it occurs, is very limited,

PbFz-Ber. A phase diagram for the PbF,-
BeF, system, published recently,” showed that
these components form two compounds, 3PbF, -
BeF, and PbF,-BeF,. It was shown also that
the latter compound forms extensive solid so-
lutions with Ber. The statement was made that
mixtures in this system are quite fluid, even with
as much as 95 mole % BeF.,. This suggested
that the fluidity of alkali fluoride-BeF, mixtures
might be usefully increased by the addition of
PbF, if the resulting mixtures were compatible
with structural materials at high temperatures.
As o preliminary test of compatibility, mixtures
of PbF, and BeF, containing 50 and 75 mole %
BeF, were prepared by heating the compounds to
about 800°C in nickel crucibles equipped with
nickel stirrers. The breaks on the cooling curves
agreed fairly well with the values reported for
these compositions. The absence of metallic lead
in the resulting melts suggests that the activity
of the Pb*" jons in the melts was sufficiently
low that very little reduction of PbF, by the nickel

52

walls occurred. The compatibility of these ma-
terials with Inconel containers will be tested in
the near future,

NaF-Ber-UF . Three filtrations were carried
out to determine the solubility of U3* in the
Na,BeF ,-NaBeF, eutectic composition (43 mole %

BeF,). The U3* values obtained with samples
filtered at 600 = 10°C were 1.16, 1.22, and
0.82 wt %. These results are in agreement with

the findings at Mound Laboratory® that UF,
solubility in this system is very low except at
low Ber concentrations. This behavior might
be expected, since BeF, and UF, do not form a
compound, while NaF and UI:3 form a compound

believed to be NoUFd.

Phase Relationships in LaF ;- and UF -Bearing
Systems

L. M. Bratcher R. E. Thoma
Materials Chemistry Division

The usefulness of LaF, as a ‘‘stand-in”’ for
UF, was discussed in the previous quarterly
report.'2  These studies have been continued,
and it was found that there is o close corre-
spondence in thermal effects between L0F3 ond
UF, systems with components that form simple
eutectics (for example, LiF oand UF4). This
probably means that LaF, and UF_ have about
the same melting point.

LiF-LaF,. The LiF-LaF; system appears to
be a eutectic system with a eutectic temperature
of 770°C, the same as that reported for the
LiF-UF, system.'® The eutectic composition is
probably about the same also, that is, approxi-
mately 28 mole % LaFa.

UF,-LaF,. The components UF,-LaF, also
form a eutectic that melts at about 865°C. Insuf-
ficient data were obtained to locate the eutectic
composition, but, since UF, melts at a much lower
temperature than LGF3 (estimated melting point,
1425°C), the eutectic composition probably con-
tains much more than 50 mole % UF,. Cooling
curves with UFA--UI'—'3 mixtures have shown a
spread from about 835 to 875°C in thermal effects,
but the higher value is probably nearer to the
correct eutectic temperature.

 

12R. E. Moore and R. E. Thoma, ANP Quar. Prog.
Rep. Dec. 10, 1954, ORNL-1816, p 59; P. A. Agron and
M. A. Bredig, loc. cit., p 72.

13¢, J. Barton et al., ANP Quar. Prog. Rep. Sept. 10,
1954, ORNL-1771, p 59.
NaF-KF-Lqu. A mixture containing 25 mole %
NaF-25 mole % KF-50 mole % LaF, was found
to contain a single phase. This indicates solid
solution formation between NalaF, and l'(LnF4
and confirms the belief that NaUFfl and KUF4
form solid solutions.

RbF-LuFa. Dergunov’s thermal data'? for the
FbF-LaF, system were essentially confirmed.
The melting point of 585°C for the eutectic at
20 mole % Lcll:3 is the lowest melting point
in alkali fluoride-L.aF; systems, and
the corresponding RbF-UF3 eutectic would be of
interest if oxidation and disproportionation of UF3
in RbF mixtures could be avoided.

LaF, and UF, in ZrF4-Beuring Melts. A mixture
having 2 moles of ZrF, per mole of LaF; con-
tained some free Lan, while a mixture with a
3:1 molar ratio was essentially a single phase
that was believed to be a compound composition.
Therefore the compound previously designated
QZrFA-UF may actually have the composition
3ZrF ,-UF,. Free UF; was reported to be present
in the best preparation of the ZZrF4-UF3 compo-
sition; failure to find free UF_ in other prepa-
rations of this composition may have been due to
the existence of part of the uranium in the tetra-
valent state. Petrographic and x-ray studies of
a number of ZrF,-UF,-LaF, compositions were
made in on effort to identi?y more conclusively
the two compounds believed to exist in the ZrFA-
UF4-UF3 system, but phase relationships in this
system are not well understood and will receive
further study when time permits. A brief and
incomplete thermal analysis survey of the RbF-
LaFa-ZrF4 system failed to show usefully low

observed

melting points except at very low LaF, concen-
trations.

PREPARATION AND STABILITY OF
UF ,~BEARING MELTS

F. F. Blankenship C. J. Barton
Materials Chemistry Division

Reduction of UF , with Uranium in

Alkali Fiuorides
R. J. Sheil B. H. Clampitt

Materials Chemistry Division

Data on the reaction of metallic uranium with
Ul:4 dissolved in alkali fluorides have been

 

e p, Dergunov, Doklady Akad. Nauk S.5.5.R. 60,
1185 (1948).

PERIOD ENDING MARCH 10, 1955

obtained in the past mainly with preparations
ranging from 0.5 to 2.0 kg in weight.!3:16 |4
appeared desirable, however, to study the reaction
with smaller scale equipment in order to facilitate
investigation of the different variables involved
in the reaction. Therefore most of the recent
experiments have been performed with small nickel
reactors fitted with nickel filter sticks, and 20 ¢
of material has been used. The main advantages
of the small equipment are the larger number of
experiments that can be performed per man-hour,
smaller amount of materials required, and the
faster rates of heating and cooling that can be
achieved. The principal disadvantage of the
small-scale filtration apparatus is that it does
not lend itself readily to the use of the purification
procedures routinely practiced with larger prepa-
rations. In general, previously purified alkali
fluorides are ground and loaded into a reactor in
a dry box with the necessary amount of UF  and
freshly cleaned uranium metal. After equilibration
and filtration at the desired temperature, the
reactor is opened in a dry box and all the filtrate
and unfiltered residue are ground for analysis.
The data obtained to date are not sufficiently
complete or reproducible to accurately evaluate
the effect of all the variables.

Effect of Nickel Surface Area. Early efforts to
reduce UF, dissolved in NaF-LiF eutectic with
metallic uranium in small-scale filtration equip-
ment were uniformly unsuccessful, although the
reaction could be readily carried out in sealed
capsules in phase study apparatus and in large-
scale preparations. The filtration equipment used
in these experiments had the filter medium in
contact with the melt during the reaction period
(usually 2 hr or more). This apparatus was modi-
fied to permit the filter medium to be in contact
with the melt only during the actual filtration
time and for a short time thereafter while the
reactor was being cooled as rapidly as possible.
This change resulted in a large increase in the
degree of reduction of uranium in the filtrate.
Measurements of the surface area of the nickel
filter medium (0.0004-in. pore size) by the Carman

permeability method'” showed a surface area of

 

135G, M. Watson et al., ANP Quar. Prog. Rep. Sept. 10,
1954, ORNL-1771, o 77.

164 A, Friedman, ANP Quar. Prog. Rep. Dec. 19,
1954, ORNL-1816, p 61.

17p, C. Carman and J. C. Amell, Can. J. Res. 26,
Sec. A, 128 (1948).

53
ANP PROJECT PROGRESS REPORT

2.9 x 10° cm? per cubic centimeter of filter ma-
terial. For the standard Y%-in. sheet thickness,
this is equivalent to about 900 c¢m? per square
centimeter of apparent area. The reason for the
marked effect of this large surface area upon the
degree of reduction of uranium is not clearly under-
stood, but it can be explained by assuming that
disproportionation of UF, in alkali fluorides is
a heterogeneous reaction that occurs at metal
surfaces. Further studies of this variable are
under way.

Effect of Uranium Surface Area. Increasing the
surface area of the metallic uranium by converting
it to the hydride brought about a significant
increase in the reduction of UF, dissolved in
NoF-KF-LiF eutectic (11.5-42-46.5 mole %). The
effect of uranium surface area was larger at lower
temperatures, probably because the UF, produced
at the metal-salt interface is more soluble at the
higher temperatures. Conversion of the uranium
metal to the hydride was adopted as o standard
procedure.

Effect of Temperature. Varying the temperature
for the reaction between NaF-KF-LiF-UF, mix-
tures and metallic uranium from 550 to 750°C
produced no significant change in degree of re-
duction. The ratio of U3* to total uranium for
five experiments in this temperature range varied
from 0.57 to 0.70.

Effect of Time. Increasing the time of reaction
of the NaF-KF-LiF-UF, mixture with the metallic
uranium at 550°C from 15 min to 2 hr produced
an increase from 0.47 to 0.58 in the ratio of U3*
to total uranium in the filtrate. This difference
may not be significant in view of the poor repro-
ducibility of these experiments. The reaction
time for NaF-LiF-UF, mixtures with metallic
uranium at 750°C varied from 15 min to 6.7 hr.
Again, the variation of the ratio of U3 to total
uranium was of doubtful significance, but the
highest ratio (0.75) was obtained with the shortest
reaction time and the lowest ratic (0.57) with the
longest reaction time. The data obtained to date
show that the reaction of UF , dissolved in alkali
fluorides with finely divided uranium metal is
very rapid and thet it may possibly be followed
by disproportionation at a rather slow rate.

Effect of Excess Uranium. Parallel studies of
the reaction of NaF-KF-LiF-UF, mixtures with

 

8y, c. Whitley and R. J. Sheil, ANP Quar. Prog.
Rep. Dec. IO: 1954: ORNL']B]éJ P 60'

54

the theoretical amount of uranium required to
satisfy the equation

‘4U°+ 3/4UF4 = UF,
and 1.2 times the theoretical amount showed very
little difference in the ratio of U* to total uranium
in most cases. Where there was a ditference, the
preparation with the theoretical amount of uranium
gave the lower degree of reduction. It seems
likely that uranium can react with oxidizing
impurities in the fused salts which would other-
wise oxidize a part of the UF .

Stability of UF; in Alkali Fluorides

R, J. Sheil B. H. Clampitt
C. J. Barton

Materials Chemistry Division

The study of the stability of UF, in KF, which
was discussed in the previous repor’r,]8 was
continued and extended to other alkali fluorides.
The wvariables investigated were temperature,
solvent, container material, and concentration of
UF3. Some experiments were performed to in-
vestigote the effect of several variables simul-
taneously and the results are reported according
fo experiment type.

Sealed-Capsule Tests, A number of sealed
nickel, Inconel, and type 316 stainless steel
capsules containing KF-UF, and RbF-UF, mix-
tures were prepared. After t?'\e capsules had been
welded shut with a helium atmosphere over the
fluoride mixture, they were heated for 90 min at
the temperatures shown in Table 5.2,  After
cooling, the capsules were opened in a dry box.
A part of the fused salt was removed by drilling,
and as much as possible of the remainder of the
material was removed by chipping and was added
to the drillings for analysis. The capsule walls
were carefully cleaned and then drilled in air to
provide a sample of the wall material that had
been in contact with the fused salt for chemical
analysis. The drillings were weighed so that the
total amount of uranium in the walls could be
calculated. It was recognized that recovery of
metallic uranium alloyed with the walls was
probably not complete, and therefore the values
given in the last column of Table 5.2 are more
likely to be low than high. The percentages were
calculated on the basis of the amount of uranium
in the original mixture.
PERIOD ENDING MARCH 10, 1955

TABLE 5.2, SEALED-CAPSULE UF3 STABILITY TESTS WITH KF-UF3 AND RhF-UF3 MIXTURES

 

Fused Salt Analysis

 

o Test Ratio of U3+ Total Uranium
Composition Temperature Container Material Total y3t to in Walls
(mole %) °C) Uranium* (wt %) Total Uranium (%)
(wt %)
85 KF-15 UF3 780 Nickel 35.6 8.2 0.23 5.2
Inconel 34.8 8.8 0.25 4.6
Type 316 stainless 36.4 8.5 0.23 3.4
steel
800 Nickel 37.8 11.5 0.30 4.0
Inconel 35.6 11.5 0.32 4.0
Type 316 stainless 35.2 13.8 0.39
steel
850 Nickel 35.2 7.2 0.20 3.5
Inconel 36.0 12.6 0.35 3.8
Type 316 stainless 34.3 5.3
steel
200 Nickel 36.2 19.2 0.53 2.6
85 RbF~15 UI::3 800 Nickel 27.9 13.1 0.47 3.5
Incone! 28.9 14.9 0.52 4.4
Type 316 stainless 28.6 12.6 0.44 9.2
steel
200 Nickel 23.7 8.5 0.36 5.0
Inconel 22.8 7.4 0.32 6.2
Type 316 stainless 5.1

steel

 

*The theoretical total uranium for KF-UF3 mixtures was 38.1 wt %; for RbF-UF3 mixtures it was 27.6 wt %.

The resuits given in Table 5.2 show no sig-
nificant effect of temperature or container material
on the ratio of U3 to total uranium in the fused
salt, There was also no reproducible difference
between KF-UF, and RbF-UF3 mixtures.  In-
spection of the data indicates that the repro-
ducibility of results is not good enough to show
anything except large effects. There appears to
be a discrepancy between the ratio of the U™ to
total uranium for the KI':-UF3 mixtures in Table 5.2
and that previously reported'® for an Incone!
capsule heated to 1000°C, which showed 70% of
the uranium in trivalent form. In that case, the
material analyzed consisted only of the drillings
from the fused salt mixture. A fused salt layer
approximately 90 mils thick adjacent to the tube
walls was not sampled by this method, whereas
an attempt was made to get all the fused salt out
of the capsules in the later experiments. The

copsules were not agitated during the heating
period, and, therefore, if disproportionation of UF
occurred at the tube walls, it is possible that the
concentration of tetravalent uranium next to the
walls was higher than in the center of the
capsules. This possibility will be further in-
vestigated.

Open-Capsule Tests. The effect of varying the
concentration of UF | in NaF-KF-LiF eutectic was
tested in open Inconel and molybdenum capsules
heated in a helium atmosphere for 90 min at 750°C.
Molybdenum was used because it does not alloy
appreciably with uranium at this temperature. The
capsule walls were not analyzed for uranium in
these experiments, because it is difficult to effect
more or less complete removal of the fused sait
from the capsules without leaving the walls in a
poor shape for sampling, porticularly with the

35
ANP PROJECT PROGRESS REPORT

TABLE 5.3. OPEN-CAPSULE UF3 STABILITY TESTS WITH I(F-UF3 AND Nch-KF«LiF-UF’3 MIXTURES

 

 

. 3+ .
Ratio of U Th tical
Initial Composition Container Total Uranium ud* ate @ eore u:-q
le Material (w % (wt %) to Total Uranium
(mole %) atertd wt %) e Total Uranium (mole %)
85 KF-15 UF3 Inconel 34.8 17.5 0.50 38.1
Molybdenum 33.8 12.1 0.31 38.1
95 {NaF-KF-LiF*)-5 UF3 Inconel 19.7 12.2 0.62 22.1
Meolybdenum 19.9 10.6 0.53 221
80 (NaF-KF.LiF*)_20 UF3 Inconel 48.5 41.3 0.85 51.7
Malybdenum 49.5 44.0 0.89 51.7
60 (NaF-KF-LiF*)-40 UF;  Inconel 64.2 56.4 0.88 66.65
Molybdenum 66.0 60.8 0.92 66.65

 

*11.5-42-46.5 mole %.

brittle molybdenum capsules. The KF-UF; mix-
tures were included in these experiments for com-
parison with similar mixtures tested in sealed
capsules. The results of chemical analyses of
the fused salt mixtures are given in Table 5.3.

It appears from the results in Table 5.3 that
alloying of the metallic uranium formed by dispro-
portionation of UF3 hos little effect on the extent
of disproportionation. A part of the U3 results
are not in agreement with the values calculated
from total uranium values on the basis that one
gram-atom of uranium metal is produced when 4
gram-atoms of U3* disproportionate. The possi-
bility of inhomogeneity in the samples cannot be
ruled out, even though they were ground to —60
mesh in dry air.

Solubility of UF, in NaF-RbF.LiF Mixtures
R. J. Sheil

Materials Chemistry Division

Thermal analysis data for UF, dissolved in
NaF-RbF-LiF mixtures were reported earlier,!?
Since it had been demonstrated that thermal
analysis was not a reliable method for determining
solubility of UF, in such mixtures,?? it seemed
advisable to check the thermal analysis data for
this system by other techniques. Visual obser-
vation of liquidus temperatures and chemical
analysis of filtered samples gave the values shown
in Table 5.4 for the solubility of UF, in a mixture
near the ternary eutectic composition.

It appears from the data in Table 5.4 that UF
is significantly less soluble in the RbF-containing

56

TABLE 5.4. SOLUBILITY OF UF4 IN NaF-RbF-LiF
(10-50-40 mole %)

 

 

Temperature Uranium Content UF4 Content
(°C) (wt %) (mole %)
500 7.4 ~ 2.5
600 13.1 ~ 4.1
650 18.8 ~6.3

 

mixture than in the NaF-KF-LiF eutectic (11.5-
42-46.5 mole %) and that the thermal effect ot
425°C observed earlier with the NaF-RbF-LiF
mixture containing 2.5 mole % UF, represented
the solidus temperature for this composition rather
than the liquidus temperature.

Solubility of UF, in NaF-KF-ZrF,

B. H. Clampitt
Materials Chemistry Division

An attempt was made to determine the solubility
of UF; in a low melting NaF-KF-ZrF, mixture
(5-52-43 mole %; melting point, 410°C). This
mixture was recently reported to have low vis-
cosity.2!  After adding sufficient UF, to give

 

19y, P. Blakely, L. M. Bratcher, and C. J. Barton,
ANP Quar. Prog. Rep. Dec. 10, 1951, ORNL-1170, p 87.

20R, ). Sheil, ANP Quar. Prog. Rep. Dec. 10, 1954,
ORNL-1816, p 59.

215 |, Cohen and T. N. Jones, Preliminary Measure-
ments of the Viscosity of Compesition 20, ORNL CF-
55-2-20 (Feb. 2, 1955},
4.0 mole % concentration to a hydrofluorinated
preparation of the ternary:composition, the mixture
was heated to 600°C for 2 hr and then filtered at
that temperature. The filtrate analyzed 0.48 wt %
U3* and 1.55 wt % total uranium, while the corre-
sponding values for the unfiltered residue were
13.1 and 14.0% respectively. These results show
a yvery low ust solubiljty at 600°C in this solvent.

Reduction of UF, in Molten
Liasz7 and LiZZrF6

M. A. Friedman

Materials Chemistry Division

A mixture of LizZrF, (mp, 645°C) with sufficient
UF, to make the mixture 15 wt % uranium was
treated at 800°C with o 50% excess of zirconium
metal and stirred by bubbling hydrogen. The ap-
parent melting point of the mixture so obtained
was 630°C. A filtrate obtained at 645°C showed
3.9 and 4.0% U3 by two ditferent analytical
methods and 4.66% total uranium. Since UF, is
quite soluble in this melt the insoluble uranium
compound is certainly trivalent; about 95% of the
UF, charged was reduced to UF,; about 85% of
the uranium in the filtrate was trivalent.

When o similar experiment with Li,ZrF, was
performed the filtrate obtained at 600°C showed
only 1.1% U* and 2.89% total uranium.

Preparation of UF, in NaF-ZrF,

F. P. Boody
Materials Chemistry Division
Experiments on a 500-g scale demonstrated

that attempts to completely reduce UF, to UF, in
NaF-ZrF, produced mixtures containing both UF,
and UF,. Therefore a request for 3-kg quantities
reduced as completely as possible by ZrH, pro-
vided an opportunity to study the effect of batch
size on the UF,-to-UF, ratio. A preparation was
carried out in which 5 wt % uranium as UF, was
used in 3 kg of NaF-ZrF, (53-47 mole %) and
275% of the stoichiometric zirconium metal. The
metallic zirconium was added as smail chips
after a preliminary purification of the melt with
HF and hydrogen. After 24 hr of equilibration
while bubbling with hydrogen at 800°C, the melt
was filtered at 710°C. The product contained
5.17 wt % uranium, including 3.13 wt % U3*. In
view of the possibility that an incomplete reaction
had occurred because of precipitation of UF,; on

PERIOD ENDING MARCH 10, 1955

the zirconium metal, another trial was made in
which two 6- by ‘/2-in. zirconium bars were used
in addition to the chips. The bars were propped
up in the body of the melt to afford good surface
contact. After 5 hr of equilibration at 800°C, the
filtered product contained 5.06 wt % total uranium,
including 2.86 wt % U3*. Since it appeared that
the equilibrium amount of UF, had not been in-
creased, another method was tried in which 7 wt %
uranium metal was added to 3 kg of NaF-ZrF,
(53-47 mole %). Complete reaction between the
uranium metal and the ZrF, would change the
NaF-to-ZrF, ratio from 53:47 to 55:45. After 6 hr
of equilibration at 800°C and filtration at 780°C,
a product containing 4.68 wt % total uranium and
3.30 wt % U3" was obtained. The ratio of U 1o
total uranium in these trials was 0.605, 0.565, and
0.705, respectively. It was concluded that the
last and highest figure was representative of
equilibrium at 800°C, but that reproducibility
could not be expected because of uncontrolled
variation in the activity of solid uranium metal
and zirconium metal.

CHEMICAL REACTIONS IN MOLTEN SALTS

F. F. Blankenship L. G. Overholser
W. R. Grimes
Materials Chemistry Division

The Equilibrium
FeF, + H,=—Fe° + 2HF
in NaZrF, ot 800°C
C. M. Blood

Materials Chemistry Division

During the past year numerous attempts have
been made to determine the equilibrium HF con-
centrations resulting from the reduction of FeF,
solutions in NaZrF; by hydrogen. In addition to
the importance of this reaction in purification
procedures, there was interest in the possibility
of measuring the activity coefficient of FeF,.
Most of the attempts have been efforts to determine
the equilibrium HF concentration by extrapolation
to zero flow rate in a system with hydrogen bub-
bling through a melt containing FeF ..

During the past quarter several methods of
improving the contact between hydrogen bubbles
and the melt have been tried, and as the contact
has been improved the measured HF concentrations
associated with a given FeF, concentration have

57
ANP PROJECT PROGRESS REPORT

In each case, however,
the HF concentration at very low flow rates has
been about three times the value at flow rates of
200 ml/min in the same system. The shapes of
the concentration vs flow-rate curves have been
identical, and it has not been possible to obtain
a satisfactory extrapolation to zero flow rate be-
cause of the steep slope at low flow rates. By
using the best extrapolation that could be made
with each system and assuming the activity of the
reduced metallic iron to be unity, successive
improvements in contact of hydrogen with the melt
have increased the values of

consistently increased.

2
PouF

P
HZ

X

FeF2
from 2 to 8, where X is mole fraction and P is
pressure in atmospheres.

Since K, = 8 is about four times the K, expected
from the thermodynamic estimates, it is possible
that the assumption of unit activity for the metallic
iron is not valid. The reduced iron may alloy with
the nickel container. The largest HF values are
found in systems having the largest surface area
of nickel. It had been recognized that the nickel
surface was probably acting as a catalyst for the
heterogeneous reaction, and it was hoped that a

definite equilibrium could be reached in a system
containing nickel mesh baffles arranged so that a
bubble required 6 sec to rise through 10 in. of melt.
When a slow flow rate was resumed after the sys-
tem containing the baffles had been cllowed to
“‘rest’’ over night, an increase instead of the
expected decrease in HF concentration was noted
during the next few hours. This same apparatus
was also used in matching the HF concentration
in the inlet and outlet streams by adjusting the
inlet HF concentration. The results again agreed
with the values of K found by extrapolation to
zero flow rate, that is, K= 8 at 800°C. Typical
results for various systems are shown in Table 5.5.

Reduction of UF, by Structural Metals
J. D. Redman C. F. Weaver

Materials Chemistry Division

Apparatus and techniques for experimental de-
termination of equilibrium constants for the re-
actions ‘

Cr° + 2UF, ==CrF, + 2UF,

and

Fe® + 2UF, —=FeF, + 2UF,

were described in previous reports.??2 Equilibrium

 

22), D. Redman and C. F. Weaver, ANP Quar. Prog.
Reps,, ORNL-1692, p 56; ORNL-1729, p 50; ORNL-
1771, p 60; ORNL-1816, p 64.

TABLE 5.5. EFFECT OF BUBBLE PATH ON EQUILIBRIUM HF CONCENTRATION IN
THE REACTION H, + FeF, ——=Fe® + 2HF IN NoZrF_ AT 800°C

 

Calculated Activity

 

Conditions of Contact Between Gas and Melt K _* Coefficient**
for FeF2
Bubbles from 3/B-in.- tube rising through 5 in. of melt 2.1 1.3
Bubbles from 3/B-in. tube rising through 10 in. of melt 4.4 2.7
Bubbles from 3/8-in. tube rising through zigzag baffles 8 5

in 10 in. of melt

 

2
T
*Kx = , where P is pressure in atmospheres and X is mole fraction.
X
FeF2 H2
1 Ka
**K}/ = - — ;AF°=RT In K_; at 800°C, Ka = 1.6, if solid FeF2 in the standard state and an activity of metallic

iron of unity are assumed.

58
data were presented for these reactions in NaZrF,
with pure metallic chromium and metallic iron used
as the reducing agents. In addition, some experi-
ments with Inconel as the reducing agent and some
preliminary observations with chromium metal and
the NaF-KF-LiF eutectic as the reaction medium
have been presented.

The results of some recent studies on the re-
action of UF, with metallic iron in NaF-KF-LiF
(11.5-42-46.5 mole %) at 600 and 800°C are given
in Table 5.6. In these experiments, 2 g of hydro-
gen-fired iron wire was reacted with UF, (15 wt %)
in about 21 g of the NaF-KF-LiF eutectic contained
in nickel. The iron concentrations reported in
Table 5.6 are very nearly the same as those found
when NaF-ZrF, was used as the solvent. The iron
concentrations in the latter case were obtained
after reaction with 11.8 wt % UF, as compared
with 15 wt %, but this difference in UF, content
would result in only slightly higher values for iron
at the higher UF, concentration. It may be seen
from Table 5.6 that higher iron concentrations
result at 600°C than at 800°C, and this is also in
agreement with what was found in NaF-ZrF,. No
attempt has been made to calculate equilibrium
constants from these data, since the valence state
of the iron is not known with certainty.

Similar studies have been made of the reaction

of UF, with chromium in NaF-KF-LiF (11.5-42-

TABLE 5.6. EQUILIBRIUM DATA FOR THE
REACTION OF UF (15 wt %) WITH IRON
IN MOLTEN NaF-KF-LiF (11.5-42-46.5 mole %)
AT 600 AND 800°C

 

Conditions of Found in Filtrate

 

Equilibration

 

 

Total Total Total
Temperature Time Uranium lren* Nickel
(°C) (hr) {wt %) {ppm) (ppm)
600 3 12.7 655 100
3 12.6 750 125
5 11.7 685 85
5 10.8 680 95
800 3 11.9 470 80
3 12.2 425 90
5 11.2 525 130
5 11.9 485 110

 

*Blank of 80 ppm of iron at 800°C.

PERIOD ENDING MARCH 10, 1955

456.5 mole %) at 600 and 800°C. In these experi-
ments 2 g of hydrogen-fired metallic chromium,
21 g of the solvent, and the desired amount of UF,
were equilibrated in nickel apparatus. The total
uranium concentration was kept constant at 11.4
wt % but the ratio of UF, to UF; was varied. The
results are presented in Table 5.7.

The chromium concentrations given in Table 5.7
for 600°C are very much lower than the 2200 ppm
of chromium found when NaF-ZrF, was the solvent,
On the other hand, the values at 800°C when no
UF, was added are approximately equal to those
found in NaF-ZrF,. The larger differences in
chromium concentrations at 600 and 800°C in
molten NaF-KF-LiF than in NoF-ZrF, suggest that
mass transfer of chromium will take place much
more readily in NaF-KF-LiF. The addition of UF,
to the system decreased the equilibrium chromium
concentration markedly and indicated that corrosion
might be greatly reduced by using a mixture of UF,
and UF4. However, the total uranium concen-
trations found for those runs in which UF,; was
added were decreased significantly during the test,
probably because of disproportionation of the UF,.
Uncertainty in the valence state of the chromium
precludes any evaluation of equilibrium constants
at present.

Previous studies of the reaction between UF,
and Inconel in molten NaZrF,, as expected, gave
equilibrium chromium concentrations that were
much lower than those resulting from the reaction
of UF, with chromium. Similar studies with the
NaF-KF-LiF mixture as solvent for reacting hydro-
gen-fired Inconel turnings with 14.7 wt % UF , have
been made in apparatus of nickel. The data are
shown in Table 5.8. The values given in Table 5.8
are somewhat more than threefold lower than those
found when UF, reacts with metaltlic chromium.
However, there is some doubt as to whether equi-
librium was attained in the runs reported in Table
5.8, since the increase in chromium concentration
from 5 to 12 hr at both temperatures is small
enough that it may or may not be significant.

Some studies have also been made on the re-
action of metallic molybdenum with UF, in molten
NaF-ZrF, contained in nickel ot 600 and 800°C.
The results given in Table 5.9 were obtained at a
UF, concentration of 11.4 wt %. The results show
that UF4 is stable in contact with metallic molyb-
denum under the conditions used and confirm the
low molybdenum concentrations reported for similar

59
ANP PROJECT PROGRESS REPORT

TABLE 5.7. EQUILIBRIUM DATA FOR THE REACTION OF UF, WITH METALLIC CHROMIUM IN
MOLTEN NaF-KF-LiF (11.5-42-46.5 mole %) AT 600 AND 800°C

 

Conditions of Equilibration

Original Concentration

Found in Filtrate

 

 

 

 

 

 

 

Temperature Time UF4 UF3 Total Uranium Total Chromium* Total Nickel
(°C) (hr) (Wt %) (wt %) (wt %) (ppm) (ppm)
600 3 15.0 12.2 1190 30

3 15.0 11.8 1075 75
5 15.0 11.3 960 75
5 15.0 12.0 1130 40
800 3 15.0 1.2 2690 20
3 15.0 11.2 2600 50
5 15.0 11.9 2700 35
5 15.0 11.4 2810 15
5 10.0 4.7 11.3 55 95
5 10.0 4.7 10.6 30 70
5 10.0 4.7 10.7 25 45
5 7.5 7.1 10.1 120 70
5 7.5 7.1 10.4 130 80
5 7.5 7.1 10.2 175 75
5 7.5 7.1 10.4 135 135
*Blank of 500 ppm chromium at 800°C.
TABLE 5.8. REACTION OF !.JF4 WITH INCONEL IN MOLTEN NaF-KF.LiF
(11.5-42-46.5 mole %) AT 600 AND 800°C
Conditions of Equilibration Found in Filtrate

Temperature Time Total Uranium Total Chromium* Tetal lron Total Nickel

e {hr) (wt %) (ppm) (ppm) (ppm)
600 5 11.3 350 195 90
5 10.9 355 165 75
12 11.0 385 195 550
800 5 10.9 640 90 145
5 10.9 735 125 100
12 11.0 995 110 230
12 10.9 810 95 125

 

*Blank of 200 ppm of chromium at 800°C.

melts that had been circulated in thermal-con-
vection loops fabricated of molybdenum or of
Hastelloy B.

The data obtained for similar runs in NaF-KF-
LiF (11.5-42-46.5 mole %) at 600 and 800°C with
15 wt % UF, are given in Table 5.10. The results
indicate that the reaction of UF, with metallic

60

molybdenum proceeds in this solvent to a greater
extent than in NaF-ZrF,. They show that the
molybdenum concentration is lower at 800 than at
600°C and suggest that it decreases with time at
800°C. There is no apparent explanation for either
the high uranium values or for the poor precision
of the experimental results.
PERIOD ENDING MARCH 10, 1955

TABLE 5.9. REACTION OF UF, WITH METALLIC MOLYBDENUM IN MOLTEN NaF-ZrF, AT 600 AND 800°C

 

Conditions of Equilibration

 

Found in Filtrate

 

 

Temperature Time Tatal Uranium Total Melybdenum* Total Nickel
(°C) {(hr) (wt %) (ppm) (ppm)
600 3 8.4 7 155

3 8.6 7 135
5 8.5 7 230
5 8.6 9 215
800 3 8.5 8 90
3 8.4 8 85
5 8.5 9 30
5 8.6 11 85
5 8.6 9 80
5 8.6 9 10

 

*Blank of 20 ppm of molybdenum at 800°C,

TABLE 5.10. REACTION OF UF, WITH METALLIC MOLYBDENUM IN MOLTEN MNaF-KF-LiF
(11,5.42-46,5 mole %) AT 600 AND 800°C

 

Conditions of Equilibration

 

Found in Filtrate

 

 

Temperature Time Total Uranium Total Molybdenum* Total Nickel
(°Q) (hn) (wt %) (ppm) (ppm)
600 3 13.3 210 85

5 14.0 200 110
5 1.8 325 170
800 3 13.5 130 170
3 13.9 105 85
5 14,4 55 205
5 13.9 65 145

 

*Blank of 5 ppm of molybdenum at 800°C and 30 ppm at 600°C.

Stability of Chromous and lron Fluerides in
Molten Fluorides

J. D. Redman C. F. Weaver
Materials Chemistry Division

Previous studies?® had indicated that Fe*" was
relatively stable in NaF-KF-LiF and that the
solubility of FeF, was 12 and 19 wt % at 600 and
800°C, respectively; however, it was found that
Cr*? was not stable in this solvent and that it
apparently underwent disproportionation to metallic

 

23y, D. Redman and C. F. Weaver, ANP Quar. Prog.
Rep. Dec. 10, 1954, ORNL-1816, p 63.

chromium and Cr3*. Experiments have now been

performed in which ferrous, ferric, and chromous
fluorides were added to NaF-KF-LiF (11,5-42-46.5
mole %) to determine their stability. The results
of a number of runs in which FeF, was added to
the NaF-KF-LiF mixture and equilibrated ot 600
and 800°C
Table 5.11,

The nickel and ferrous concentrations found in
the filtrate suggest that the reaction

Ni + 2FeF,==NiF, + 2FeF,

in nickel equipment are given in

occurs under these conditions. However, the ratio
of Fe** present to that theoretically resulting from

61
ANP PROJECT PROGRESS REPORT

TABLE 5.11, STABILITY OF FeFy IN MOLTEN NaF-KF-LiF (11,5-42.46.5 mole %)

 

Conditions of Equilibration

 

Found in Filtrate

 

 

FeF; Added s
Temperature Time (% Fe) Fe Tetal Fe Total Ni
(°C) (hr) (wt %) (wt %) {wt %)
600 5 4.9 0.72 1.57 0.53
10.0 0.73 1.72 0.53
800 10.0 5.1 9.3 3.5
15 5.4 0.24 4.6 1.10
5.4 0.17 4.7 1.60
10.8 0.58 9.2 2.7
10.8 0.50 9.0 2.9

 

the Ni** content is only 0.75 at 600°C and about
0.1 at 800°C. This behavior indicates that al-
though Fe™™ is formed by the reaction of Ni with
Fe®* the resulting Fe** is not stable and probably
disproportionates as

3Fe** ==Fe° + 2Fe3*

The relatively large amounts of finely divided iron
observed in all the runs heated for 15 hr sub-
stantiate the belief that disproportionation occurs.

In an attempt to reconcile the results presented
in Table 5.11 with the earlier observation that
FeF, is stable in the NaF-KF-LiF mixture, some
runs were made in which FeF, was equilibrated
in the presence of NiF,, These experiments were
carried out in nickel, and the results given in
Table 5.12 show that Fe*" is not stable in the
presence of Ni**, and it must be assumed that
Ni** enters into some reaction which at 600°C
removes nearly all the Ni** from the melt. The
high nickel concentrations of the residues sug-
gested that metallic nickel was present. The
reactions involved may be the following:

3Fe*t ==2Fe3*t 4 Fe°

L+ N 44 )
Fe® + Ni —Fe + Ni°

Some studies with metallic iron and NiF, as re-
actants are to be made in an attempt to evaluate
the equilibrium concentrations for the reactions.
Previous studies?® have shown that CrFy is
stable in the molten NaF-KF-LiF mixture contained
in nickel at 600 and 800°C and data also were
presented which indicated that CrF, is not stable

62

TABLE 5.12. STABILITY OF FeF, IN MOLTEN
NaF-KF«LiF (11.5-42.46,5 mole %)
IN PRESENCE OF NiF,

FeF2 Odded: 6-0 wit % Fe
NiF2 added: 2.1 wt % Ni
Equilibrated 5 hr

 

Equilibratien Found in Filtrate

 

 

 

Temperature Fe'? Total Fe Total Ni
(°C) (wt %) (wt %) (wt %)
600 2.4 2.9 0.01
2.5 2,9 0.02
800 2.9 5.9 0.7
2.7 5.8 1.1
under these conditions but, rather, undergoes

disproportionation (3CrF2—> Cr® + 2CrF3).
Additional data that support this belief are given
in Table 5,13,

Analytical data obtained for the residues indicate
that the solubility of CrF, was not exceeded in
any of the runs and that the solubility of CrF,
was exceeded in all runs at 600°C but not at
800°C. Although the precision of these measure-
ments is not of a high order, the data have been
used to obtain an approximate value for the equi-
librium constant of the reaction

3CrF2:—f— 2CrF3 + Cr°

For concentrations expressed in mole fractions,
the value for K_ is of the order of 10% at 600°C.
TABLE 5.13. STABILITY OF CrF, IN MOL TEN
NaF-K F-LiF (11.5-42.46,5 mole %) AT 600°C

Equilibrated 5 hr at 600°C

 

Found in Filtrate

 

 

CrF, Added *+ 34
% Co (52 %) (E: %)
2.9 0.30 1.08
2.9 0,32 0.74
5.8 0.30 0.91
5.8 0.48 1.08

 

The value for K _ for this reaction at 600°C (calcu-
lated from the standard free energy of formation

of CtF, and CrF3) is about 103,

PRODUCTION OF PURIFIED MOLTEN
FLUORIDES

F. F. Blankenship G. J. Nessle
L. G. Overholser

Materials Chemistry Division

Production-Scale Operations

F. L. Daley J. P. Blakely

Materials Chemistry Division

A total of 2077.6 kg of processed fluorides was
produced in the 250-lb facility during the quarter.
The various compositions and amount of each
processed are listed below.

Amount Processed

 

Composition (kg)
NcF-ZrF4—U Fy (50-46-4 mole %) 496.8
NaF-ZrF 4-U Fy4 (53.5-40-6.5 mole %) 451.7
NaF-—ZrF4-U Fyu (56-39-5 mole %) 790.4
NaF-—Zer (54.1-45.9 mole %) 338.7

Totol 2077.6

The 250-Ib facility was taken out of service on
January 1, at which time a total of 2545 kg of
processed fiuoride was available for the various
testing programs. |t was anticipated that this
stockpile would be adequate for 4 to 6 months.
It now appears that increased demands resulting
from the general speed-up in the ANP program and
the increased interest in UF j-bearing materials

PERIOD ENDING MARCH 10, 1955

will require operation of the production facility
briefly in March and at frequent intervals there-
after. This facility is in standby condition and
can be returned to service on approximately a
two-day notice,

Pilot-Scale Preparation of BeF, Mixtures

J. P. Blakely C. R. Croft
4. Truitt
Materials Chemistry Division

A general process development program has been
instituted to determine the best method for proc-
essing fluoride melts containing BeF, in the pilot-
scale facility. A preliminary batch of NaF-BeF,
(57.43 mole %) was processed with no treatment
except melting and mixing under a helium atmos-
phere. This material could not be filtered. Visudl
observation of a melted sample revealed consider~
able amounts of an insoluble white powder, which
was assumed to be BeO. Later analyses of ¢
sample of this batch which had been used for
physical property studies showed approximately
5% BeO. This value may, perhaps, be high due to
imperfect handling of the batch, but it is cbvious
that more rigorous purification is required,

Accordingly, the standard procedure for the
preparation of zirconium~base fuels was applied
to mixtures containing BeF,. Briefly, the pro-
cedure consists of melting the batch under a HF
atmosphere, treating the melt with hydrogen for
1 hr at 800°C, hydrofluorinating it at 800°C for
90 min, and finally stripping it with hydrogen to a
value of 1.5 x 10=4 moles of HF per liter of
exit gas. It was immediately apparent that the
sulfur content of the BeF, was high; the odor of
H,S was readily detectable in the out-gas. Table
5.14 shows pertinent analyses for several samples
of the available BeF,.

The analyses indicate that both sulfur and
chromium are present in considerably higher con-
centration than previosuly found in ZrF,. Since
chromium is removed very slowly from the ZrF -
bearing mixtures with hydrogen, some trouble from
high CrF, concentrations in the product was ex-
pected. Chemical analyses are at present avail-
able for only three of the nine batches processed
by this method, and these results are presented
in Table 5,15,

The data are, as yet, insufficient for generali-
zation, but some differences in behavior between
BeF,- and ZrF,-bearing mixtures are apparent.

63
ANP PROJECT PROGRESS REPORT

TABLE 5,14 PURITY OF BeF, RAWMATERIAL

 

Major Constituents (wt %)

Container No.

Minor Constituents {ppm)

 

 

 

 

 

 

 

Be F Fe Cr Ni 5
1 19.3 76.7 0.032 230 150 30 2350
3 19.1 78.1 0.049 250 165 35 2205
5 19.1 76.7 0.029 250 135 35 2270
7 19.2 80.1 0.035 215 145 30 1810
8 19.3 80.4 0.038 245 135 35 1930
TABLE 5.15. FINAL PURITY OF NaF-BeF, MIXTURES
Major Constituents Minor Constituents
B atch Na. (wt %) {ppm}
Be F Fe Cr Ni S
EE330 8.70 59.9 235 12 1 19
EE331 8.87 59.4 225 50 120 55
EE332* 8.79 58.3 215 35 165 320

 

*This was batch EE330 broken up and reprocessed.

The CrF, content of the melt is appreciably low-
ered by using the zirconium-base fuel process, but
the iron concentration is not, There is consider-
able inconsistency in the sulfur analyses, but
there is no reason to doubt that the sulfur is readily
removable.

It was observed that the reactor dip-line tended
to plug as HF concentrations approached 1 x 10~4
moles of HF per liter of exit gas. Consequently
about 50% of the baiches processed were termi-
nated at readings of approximately 1.5 x 10—4
moles of HF per liter. It was also noted that
occasionally the HF level of a sample rapidly
decreased to a value near 1 x 10~3 moles HF per
liter, remained there for 2 to 4 hr, and then gradu-
ally rose to nearly 2 x 104 moles HF per liter
before decreasing again.

Successful filtrations of these preparations have
been accomplished with sintered nickel filters of
0.0024., 0.0015., and 0.004-in. pore diameter. The
finest of these filters is presently being used on a
routine basis.

Pilot-Scale Preparation of UF ;-Bearing Mixtures
G. J. Nessle J. P. Blakely

Materials Chemistry Division

A series of eight preparations was completed
that contained NaF-KF-LiF and a constant total
uranium concentration with wvarious U3+/UF4
ratios. The process used in the preparation of
these batches was identical with that used previ-
ously in which separate purification steps for the
main constituents were utilized. The NaF-KF-LiF
eutectic (11.5-42-46.5 mole %) is first heated and
stripped with hydrogen at approximately 800°C.
The melt is then cooled and the UF, is added.
The heating and stripping are again repeated; the
melt is cooled, and the uranium metal is added.
The final heating and stripping are carried out,
and the batch is transferred to a storage can,

In a previous study of the reaction

U + 3UF, —> 4UF,

in this environment, it was demonstrated that the
reaction does not proceed to completion. In the
previous series of preparations, sufficient uranium
metal was added to yield approximately 11 wt %
U3*, and fairly consistent values of around 5 wt %
U3* were obtained. In the present series of prepa-
rations, approximately 2.5 wt % U3* was desired,
so the uranium-to-UF , ratio was set to yield theo-
retical U3* values of 6 wt %. The results ob-
tained are presented in Table 5.16. Although the
level of U3* desired was not generally attained,
these batches have been released for corrosion
studies in thermal~-convection loops.

For another series of eight preparations con-
taining 50 mole % NaF, 46 mole % ZrF,, about
3 mole % UF,, and 1 mole % UF,, the proper
quantities of NaF and ZrF  containing the desired

PERIOD ENDING MARCH 10, 1955

amount of UF, were treated with HF and hydrogen
in the normal manner until less than 1 x 10~%
moles of HF per liter of exit gas was indicated.
The melt was then cooled to room temperature and
the UF; was added carefully under a helium atmos-
phere. The melt was again heated and agitated
with hydrogen until the temperature of the melt
reached 600°C, held at temperature and agitated
for 2 hr, and transferred through the filter to the
storage receiver in the normal fashion. The general
reproducibility of this type of preparation is indi-
cated in Table 5,17, 1t is believed that experience
with this procedure will make it generally useful
for preparation of several types of UFj-bearing
mixtures.

TABLE 5,16, CHEMICAL ANALYSES OF UF,-BEARING ALKALI FLUORIDE MIXTURES

 

 

 

 

 

 

 

 

Barch No. Einal U3+ Total Uranium Minor Con stituents (ppm)
(wt %) {wt %) Fe Cr Ni
EE323 2.0 1.5 50 40 70
EE325 1.2 111 80 30 35
EE326 3.4 13.3 100 30 105
EE327 1.6 11.3 115 45 30
EE328 1.6 11.0 105 45 30
EE410 1.4 1.9 70 35 30
EE411] 1.6 12,3 50 50 35
EE412 3.0 12,2 95 40 55
EE413 .7 10,6 65 S0 75
TABLE 5.17. CHEMICAL ANALYSIS OF UF;-BEARING NaF-ZrF ,-U Fy MIXTURES
5 atch No. Total Uranium Final U%" Minor Constituents (ppm)
(wt %) (wt %) Fe Cr Ni

EE420 11.8 0.98 260 125 40
EE421 7.4 1.87 50 70 295
EE422 8.5 1.60 65 55 105
EE423 8.7 2.19 95 175 120
EE424 9.4 1.66 40 45 30
EE426 2,00

EE427 1.08

 

65
ANP PROJECT PROGRESS REPORT

Special Services

J. E. Eorgan J. P. Blakely

Materials Chemistry Division

Hazards Experiments, Two 300-Ib batches of
molten NaF-KF-LiF eutectic were prepared for
hazards experimentation. The first test consisted
in dumping 500 Ib of the molten mixture at 1500°F
into a large steel tank whose hemispherical bottom
was externally cooled by immersion in water. The
second test involved injection of the 1500°F batch
of salt into a tank of water well below the water
line.. The transfers of 500 Ib of molten salt through
l1-in. lines required approximately 1 min in each
case; the transfers were effected by using 25 psig
of helium to force the salt from the storage reser-
voir into the test tank. The results of these tests
are described in sec. 3, ‘“Experimental Reactor
Engineering.’’

Charge Material for In-Pile Loop. A request for
1500 g of a mixture containing 63 mole % NaF,
25 mole % ZrF,, and 12 mole % UF, was received
from the radiation damage group. Since this ma-
terial was to be used for an in-pile loop, enriched
uranium was required, and the processing had to be
done in the facility in Bldg. 9212 at Y-12.

Assembly and repair of the processing unit in
Building 9212 were completed on January 9, 1955,
and the batch was finished on January 11, 1955,
Samples were submitted for analysis to both the
X-10 and Y-12 analytical laboratories. The ac-
countability transfer of the material was done on
the reported values of the Y-12 laboratories, with
the X-10 results used as checks.

A considerable amount of repair and parts re-

placement had to be done on this processing unit
before the material could be prepared. Complete
remodeling of the unit will be required before new
batches can be processed safely again.
Five thousand pounds of
low-hafnium ZrCl, was received for conversion
intfo ZrF, at Building 9211 in Y-12. It is esti-
mated that this stockpile of ZrF , will be sufficient
to meet all production demands of zirconium-base
fuels unless the present rate of consumption in-
creases or a new composition is needed.

ZrF, Processing.

Preparation of Various Fluorides

B. J. Sturm E. E. Ketchen

Materials Chemistry Division

The preparation and purification of various

fluorides have continued. Increasing demands for

66

hydrofluorinated UF, and pure UF, made it neces-
sary to devote most of the effort to these materials,
although substantial amounts of the structural
metal fluorides and LaF, were also prepared.
Chemical analysis supported by x-ray and petro-
graphic examination has been used to establish
the identity and purity of the moterials,

Approximately 12 kg of UF, was purified by
hydrofluorination at 600°C to meet the requirements
for the preparation of UF, and various experi-
mental uses. Three batches of FeF, were pre-
pared by hydrofluorination of anhydrous FeCl,
at 450°C followed by a helium flush at this temper-
ature.  Approximately 8 Ib of (NH,),CrF, was
synthesized by heating an excess of NH, HF,
with CrF,3%H,0 at 200°C. A portion of the
(NH);CrF,  was thermally decomposed under
helivm at 700°C to yield CrF ;, and another portion,
after being decomposed to CrF, at 600°C, was
reduced to CrF, by heating under hydrogen at
800°C. Approximately 4 1b of LaF; was prepared
from La,0; by converting the oxide to an aqueous
solution of LaCl,, precipitating the fluoride by
decantation and centrifugation, and finally drying
at 150°C in a silver vessel.

The need for large quantities of pure UF, for
various studies has been met through a concerted
effort on the part of this group. The method being
used is essentially the same as that described
previously24 in which finely divided uranium metal
and UF, are reacted at 900°C while being rotated
in a steel capsule containing steel balls. Some
modifications were made in the apparatus which
permit the capsules to be [oaded and unloaded in
the vacuum dry-box, and, also, the heating period
has been extended from 24 to 32 hr. There is some
evidence that these changes have been instru-
mental in producing a product of higher purity than
previously obtained. The composition of nine
batches (740 g per batch), as determined by
chemical, and petrographic examination,
is given in Table 5.18, All runs were made with
1.3% excess of uranium metal. Runs 2 through 8
were heated for 24 hr and 9 and 10 for 32 hr.

An evaluation of the purity of the UF; is very
difficult. The x-ray method will not detect the
small amounts of UF, that are present in most
cases, and the petrographic examination, while
being more sensitive to the presence of UF,,
cannot give any quantitative values, The chemical

X=ray,

 

24y, C. Whitley and C. J. Barton, ANP Quar. Prog.
Rep. Sept. 10, 1951, ORNL-1154, p 159.
PERIOD ENDING MARCH 10, 1955

TABLE 5,18, COMPOSITION OF UF,

 

Chemical Analysis (wt %)

 

 

Run No. Petrographic Examination X-ray Data
Total U UF, F
2 81.1 95 18.8 Some UF, Found UF,, UO,
3 79.6 99 19.9 Trace of UF, Only UF,
4 80.5 99 19.5 Seme UF Found UF
5 80.2 93* 19.8 Only UF, Only UF,
6 79.8 97 20,7 Trace of UF, Only UF4
7 79.9 97 19.7 Trace of UF, Only UF,
8 80.1 5% 20.0 Some UF, Only UF4
9 80.6 100 19.6 Trace of UF,
10 79.1 98 19.1 Seme UF

 

*Found 97% for sample submitted six weeks later,

**Found 98% for sample submitted four weeks later.

method vused is probably not capable of giving
values that are more accurate than 2%, It should
be noted that there is rather poor agreement be-
tween the chemical results and the petrographic
results in the case of run No, 5, It is not known
why the reaction is practically complete in some
cases and not so in others. No UO, has been
detected in any of the batches except No. 2, and
this suggests that the cause of low UF, contents
is incomplete reaction rather than oxidation,

FUNDAMENTAL CHEMISTRY OF FUSED SALTS
Electrochemistry of Fused Salts
L. E. Topol

Materials Chemistry Division

Measurements are being made in an attempt to
find a reversible fluoride electrode for melts, It
is known that Ag/AgCl serves as a reproducible
reference electrode for molten chlorides, and
therefore the half-cell Ag/AgF is one possibility
for fluorides. However, preliminary experiments
with AgF indicate the salt to be very unstable in
the presence of traces of moisture., Even with
careful handling in a dry box the AgF heated at
650°C for 5 hr in helium appeared to have hydro-
lyzed almost completely to yield metallic silver
(silver oxide decomposes above 300°C), Further

 

25¢, Wagner and D. Balz, Z. Elektrochem. 56, 574-9
(1952).

work with AgF has therefore been abandoned for
the time being.

Another possibility for a fluoride electrode is
the half-cell Ni/NiF2 (saturated in other fluorides).
Wagner and Balz25 have studied the system
KF-NiF,, and have found that two compounds
exist = K,NiF, and KNiF,. According to their
data, solutions between 9.1 and 33.3 mole % NiF,
in KF in the temperature range of 797 to 930°C
would result in a solid phase containing K,NiF,
that would precipitate upon saturation. Thus two
half cells containing different concentrations of
NiF, in the above limits should produce an emf of
zero.

The emf apparatus consists of a fairly gas-tight
can of suitable size to fit in a dry box. Electrodes
of grade A nickel rod of 1/8--in. diameter are welded
to 40-mil nickel wire leads. These are insulated
from the can by Morganite recrystallized.alumina
thermocouple beads through which the wire fits
tightly. A gas inlet, an outlet, and a thermocouple
well complete the cell. The two half-cells con-
tained in nickel or Morganite crucibles sit on a
Morganite plate and are joined electrically by a
slightly porous ZrO, bridge previously impregnated
with an alkali fluoride (the same as that to be used
in the experiment as the solvent). All nickel
electrodes and vessels are annealed before each
run by heating in hydrogen at 770°C for 1 hr. A
helium atmosphere purified by a liquid-nitrogen,
activated-charcoal trap is used throughout.

67
ANP PROJECT PROGRESS REPORT

Measurements have been made, to date, in KF,
LiF-KF (50-50 mole % eutectic), and the eutectic
mixture NaF-KF-LiF (11.5-42-46.5 mole %)}. Half-
cells containing equal concentrations of NiF, in
the NaF-KF-LiF or LiF-KF eutectics give emf’s
of the order of 2 to 10 mv at 600 to 750°C, How-
ever, with differing concentrations of NiF,_ in the
two half-cells, emf's much higher than the above
were obtained (20 to 60 mv). Even though most
cells were run for two days, equilibrium conditions
may not have been obtained, especially in the
formation and solubility of the complex.

X-Ray Diffraction Studies in the NoF-ZrF , System

P. A. Agron M. A. Bredig
Chemistry Division

The use of high-temperature x-ray diffraction
techniques was continued in the study of the
polymorphous  transitions of the compounds
Nazsz6 and Noasz7 in the NoF-ZrF4 system.
Several runs were made up to 550°C with the
30 mole % ZrF4 composition to assist in locating
the various Na,ZrF, transitions and to determine
the extent of solid solution occurring in the
Na,ZrF, phase. A marked phase transition took
place at 520 to 525°C which did not reverse on
cooling and holding for 1 hr at 495°C. The ex-
tensive solid solution in the Na Zr[:7 phase
(Table 5.19) even at temperatures :Leiow 550°C
reduced the amount of Na,ZrF, phase and made
the identification of the latter rather difficult.

Table 5.19 lists the lattice parameters of the
body-centered tetragonal phase (isomorphous with
the body-centered tetragonal form of Na UF.)
at two temperatures and compositions. etro-
graphic examination?® of sample b-2 showed the
presence of two major phases, along with a minor
amount of a finely divided constituent that is
possibly an oxidation product. One phase is
index of refraction

uniaxial negative with an
slightly tower than 1.404, and the second is
biaxial positive with on index slightly higher

than 1.404.
the second phase are close to those?? belonging
to phase 4 of Na,ZrF , but these phases differ
in x-ray diffraction pattern. On the basis of the
resemblance of the x-ray pattern of this phase to
that of the lower temperature form2® of N°3UF7’
tentative orthorhombic cell dimensions are pro-
posed as belonging to a lower temperature NOBZrF
phase.

The petrographic characteristics of

7

Physical Chemistry
E. R. Yan Artsdalen
Chemistry Division

The density and electrical conductivity of the
pure fused salts (1) potassium bromide, (2) sodium

iodide, (3) cesium chloride, and (4) rubidium

 

264, 1n sley, Consultant.

27 ANP Quar. Prog. Rep. June 10, 1954, ORNL-1729,
p 40,

28)51e ported data.

TABLE 5.19. LATTICE PARAMETERS IN THE COMPOUND Na;ZrF, OF THE NaF-ZrF, SYSTEM

 

 

 

NagZrF,
Sample Composition Temperature Cell Dimensions
Designation (mole % ZrF ,) (°C) Crystal Symmetry (z)
a 25 25 Body-centered tetragonal a = 533
c = 10,53
b-1 30 524 Body-centered tetragenal a = 5,45
c = 10.90
b-2* 30 25 Body-centered tetragonal a = 537
c = 10.75
Orthorhombic = 8,36
= 573
c = 10,90

 

*Sample cooled in high-temperature furnace from sample b-1.

68
PERIOD ENDING MARCH 10, 1955

bromide were measured as functions of temperature. Specific conductances of these salts, respectively,
may be expressed by the equations:

(1) K = =3.226; + 1012, x 107% - 4.827, x 105/ (©

{(range 740 to 960°C)

(2) K = =0.820, + 5.9404 x 1073 ~ 1976, x 10-6,2 (¢ = 0.002,)
(range 675 to 915°C)

(3) K = —1.346, + 4,537, x 1073 - 1106, x 1078/ (0 = 0,002,)
(range 650 to 9205°C)

(4) K = —3.030 + 9.103, x 10~3: - 4.509, x 10~6:2 (@ = 0.002.)

5 9 6 2

(range 695 to 905°C)

where ¢t is in °C and o represents standard deviation. The densities of these salts, respectively, are

given by:
(1) p = 2733,
(2) p = 3.3683
(3) p = 3.482,
(4) p = 3.4464

The applicable temperature ranges

- 0.8252; x 103 (o = 0.0004,)
-3 -

~ 0.94905 x 10~3; (¢ = 0,0004))

- 1061 x 10~3; (0 = 0.0004)

— 1.0718 x 10~3; (@ = 0.0005)

are the same as those for «.

69
ANP PROJECT PROGRESS REPORT

6. CORROSION RESEARCH

W. D. Manly

G. M, Adamson

Metallurgy Division

W. R. Grimes

F. Kertesz

Materials Chemistry Division

Additional thermal-convection loop studies have
been made of Inconel and type 316 stainless steel
exposed to alkali-metal-base fluorides with various
proportions of UF; and UF , added, Inconel exposed
to NaF-ZrF ,-UF, with ceramic contamination,
Inconel exposed to the zirconium fluoride base
mixture with 25 mole % UF,, Hastelloy B and
molybdenum exposed to NaF-ZrF -UF,, Hastelloy B
exposed to sodium, and beryllium exposed to
sodium in Inconel. Preliminary examinations have
been made of three Inconel forced-circulation cor-
rosion and mass transfer tests, and the indications
are that the depths of attack are only two to three
times those found in thermal-convection loops
operated for comparable periods.

General corrosion studies were continued that
included high-temperature tests of molybdenum,
tests of brazing alloys on Inconel and stainless
steel, a study of dissimilar metal mass transfer in
the zirconium—type 304 stainless steel-sodium
system, tests of the diffusion of sodium into beryl-
lium in a beryllium-sodium-Inconel system, beryl-
lium-Inconel spacer tests, tests of the corrosion
resistance of cermets in NaF-ZrF -UF,, and a
study of the solid-phase bonding of cermets. The
investigation of mass transfer in liquid lead has
been completed and summarized, and additionadl
information on the fundamental properties of fused
hydroxides is presented.

In chemical studies of corrosion, the addition of
chromium metal to NaF-ZrF, and NaF-ZrF -UF
melts exposed to Inconel in tilting~furnace tests
was found to reduce attack on the Inconel. The
metallic chromium reduces the cotrosive chromic
ion to the noncerrosive chromous state.

THERMAL-CONVECTION LOOP
CORROSION STUDIES

G. M. Adamson, Metallurgy Division
V. P. Treciokas, Pratt & Whitney Aircraft

Alkali-Metal-Base Mixtures with UF; and UF,

Several additional Inconel and type 316 stainless
steel thermal-convection loops have circulated the

70

alkali-metal-base flyoride mixture NaF-KF-LiF
(11.5-42-46.5 mole %) with various proportions of
UF; and UF, added. The fluoride mixtures were
prepared and handled with more precise procedures
than those used previously, ! and therefore results
that can be more accurately analyzed were ob-
tained, The results obtained from these loops,
which were operated for 500 hr with a hot-leg
temperature of 1500°F, substantiate the previous
observation that wuranium fluoride can dispro-
portionate in these systems. Four Inconel and
two type 316 stainless steel loops operated with
fluoride mixtures containing more than 4.83 wt %
UF, showed visible and microscopic evidence of
a vranium-rich layer on the entire surface of the
loop. The layers were about 1 mil in thickness
and were thicker in the hot legs than in the cold
legs. No corrosive attack was noted in any of the
loops and no difficulty was encountered with
plugging of the stainless steel loops. The absence
of attack in these loops compared with the attack
previously reported is thought to be, primarily, the
result of the closer control of the UF, concen-
tration; however, the materials probably also con-
tained fewer impurities. The metallographic ex-
of the loops have not yet been
completed, but the data obtained thus far are
presented in Table 6.1. The results of chemical
analyses of the fluoride mixtures used are pre-
sented in Table 6.2. A typical hot leg of an
Inconel loop (loop 590) is shown in Fig. 6.1, and
Fig. 6.2 shows a typical hot leg of a type 316
stainless steel loop (loop 193).

The results of the chemical analyses {(Table 6.2)
of the batch and the fill samples should have been
identical since they were both taken from the

aminations

original material, but discrepancies existed and
That prac-
tically no UF; was left in the Inconel loops after
additional
portionation took place.

therefore both results were reported.

operation is evidence that dispro-
Less disproportionation

was evident in the stainiess steel loops than in

 

'G. M. Adamson and A. Taboada, ANP Quar. Prog.
Rep. Dec. 10, 1954, ORNL-1816, p 76.
PERIOD ENDING MARCH 10, 1955

 

 

 

 

 

 

 

 

TABLE 46.1. RESULTS OF METALLOGRAPHIC EXAMINATION OF THERMAL-CONVECTION LOOPS AFTER
CIRCULATING NaF-KF-LiF (11,5-42-46.5 mole %) CONTAINING UF3 AND UF,
Initial Uranium Maxi
Loop Loop Content (wt %) aximum
No. Material ° Atff:ck Metallographic Notes
U udt (mils)
589 Inconel 8.09 5.4 None Uranium metal layer to a thickness of 0.5 mil
throughout loop
590 Inconel 10.9 4.8 None Uranium metal layer to a thickness of 0.5 mil
throughout loop
595 Inconel 10.5 5.6 None Uranium metal layer to a thickness of 0.5 mil
in cold leg and 1.0 mil in hot leg
596 Inconel 11.5 5.0 None Uranium metal layer to a thickness of 0.5 mil
in cold leg and 1.0 mil in hot leg
193 Type 316 stain- 12.5 66 2 Thin uvranium metal layer throughout loop;
less steel attack in form of pits
194 Type 316 stain- 12.3 4.9 None Thin uranium metal layer throughout loop
less steel
TABLE 6.2. RESULTS OF CHEMICAL ANALYSES OF THE FLUORIDE MIXTURES CIRCULATED
Loo Uranium {wt %) U3+ (wt %) Nickel Chromium Iron
. P (ppm) (ppm)} (ppm)
Ce
Batch Fill Final Batch Fill Final Fill Final Fill Fina! Fill Final
589 8.1 7.9 7.7 5.4 4.7 0.7 30 45 40 55 95 125
590 10.9 10.8 10.1 4,8 7.7 1.1 75 39 55 55 125 210
595 10.5 11.0 10.1 5.6 5.3 1.3 30 75 35 20 75 140
596 11.5 10.7 10.7 5.0 5.8 1.5 110 140 40 17 130 140
193 12.5 11.7 11.6 6.6 6.4 4.4 65 70 85 55 130 335
194 12.3 10.8 11.4 4.9 6.0 3.3 35 55 45 18 75 160

 

the Inconel loops. The nickel and iron impurities
increased more in these loops than in loops which
circulated mixtures containing UF, and no UF,.
The chromium contents were comparable.

In addition to the loops described above, six
other Inconel loops have been operated, but the
results of examination have not yet been received.
Visually, all six loops showed evidence of metallic
deposits similar to those reported metallographi-
cally in Table 6.1. These loops inciuded one
which circulated a fluoride mixture containing as

little as 1.24 wt % UF,.

Ceramic Contamination in NaF-ZrF ,-UF,
on Inconel

The depths of attack have recently increased on
Inconel thermal-convection loops operated as
controls for 500 hr at a hot-leg temperature of
1500°F  with NaF-ZrF,-UF, (50-46-4 mole %),
and, in addition, more erratic results have been
obtained. Previously operated loops showed
attack to a depth of 8 mils, whereas the last
three control loops have shown attack to a depth
of 11 mils. One possible cause of the increased
attack has been thought to be the ceramic spacers

71
ANP PROJECT PROGRESS REPORT

that are occasionally used on the filling level
probes. These probes are about 18 in. long, and
the operators occasionally place a few ceramic
beads near the lower end to prevent shorting on
the side walls of the loops. The ceramic beads
are the type used on thermocouples, and they are
not completely resistant to attack by fluoride
mixtures, The fluoride mixture is always removed
from the probe as soon as possible, but even in a
short time some contamination could occur. The
loop design has now been changed so that «
shorter probe may be used.

Five Inconel loops have been operated for

500 hr at 1500°F to determine what effect the

UNCLASSIFIED
¥-7205

 

‘NI OO0 X082

Fig. 6.1. Hot Leg of Inconel Thermal-Convection
Loop After Circulating NaF.KF-LiF.-UF ,-UF, for
500 hr at 1500°F. Loop 590. 250X. Reduced 30%.

ceramic beads may have had on the rate of cor-
rosion. Loops 609 and 610 were operated without
ceramic beads as controls; loops 611 and 612
were operated with a string of ceramic beads on
the probes; and loop 613 was operated with a
fluoride mixture that had been contaminated with
ceramic beads placed in the fill pot. The results
of examination of these loops are presented in
Table 6.3,

Although essentially no differences in depth of
attack were noted for control loop 609 compared
with loop 613 and for control loop 610 compared
with loops 611 and 612, larger and more numerous
voids were observed on the loops operated with
fluoride

mixtures. The

UNCL ASSIFIED
T-7209

ceramic-contaminated

 
  

. ’\3

‘NI 0b0'0 X0§Z —#——-{

Fig. 6.2. Hot Leg of Type 316 Stainless Steel
Thermal-Convection Loop After Circulating NaF-
KF-LiF-UF;-UF, for 500 hr at 1500°F. Loop 193.
250X, Reduced 30%,

TABLE 6.3. RESULTS FROM INCONEL THERMAL-CONVECTION LOOPS OPERATED WITH Nch-ZrF4--UF4
(50-46-4 mole %) CONTAMINATED WITH CERAMIC BEADS

 

Loop

Metallographic Notes

 

 

N Variable
o Hot-Leg Appearance Cold+Leg Appearance

609 Control Subsurface voids to 7.5 mils No deposit

610 Control Subsurface voids to 8 to 11 mils No deposit

611 Ceramic beads on probe Subsurface voids to 11 to 13 mils; No deposit

surface rough
612 Ceramic beads on probe Subsurface voids to 11 to 12 mils No deposit
613 Ceramic beads in fill pot Subsurface voids to 7.5 to 8 mils Intermittent small metallic particles

 

72
surface of loop 613 was also much rougher than
the surfaces of the other loops.
the results for the two control loops indicates
that, while the ceramic may have some effect,
there are other unknown variables that affect
the corrosion rate.

A variation in

High Uranium Content in Fluoride Mixture

Determination of any unusual corrosive effects
of fluoride mixtures with higher uranium con-
centrations than those normally used was re-
quested by the Solid State Division for evaluation
of in-pile tests for which o high uranium con-
centration is required. Subsurface void formations
to depths of 10 and 15 mils were found in the hot
legs of Inconel thermal-convection loops after
circulating NaF-ZrF -UF, (62.5-12.5-25.0 mole %)
for 500 hr ot 1500°F. The maximum depth of
attack found in any loops in which the standard
fluoride mixture NaF-ZrF,-UF, (50-46-4 mole %)
has been circulated has been 11 mils. The higher
uranium concentration does therefore appear to

result in a small increase in depth of attack.

Hastelloy B Loops

Hastelloy B thermal-convection loops in both
the as-fabricated and the dry-hydrogen-cleaned
condition have been operated for 1000 hr af o
hot-leg temperature of 1500°F with NaF-ZrF,-UF,
(50-46-4 mole %) as the circulated fluid. The dry-
hydrogen cleaning was for the removal of surface
oxide layers. No appreciable reduction in cor-

rosive attack was noted in the dry-hydrogen-
cleaned loops as compared with the as-fabricated
loops. The metallographic data for these loops,
presented in Table 6.4, are in agreement with

those previously reported for similar Hastelloy 3

PERIOD ENDING MARCH 10, 1955

loops.! The chemical analyses of the mixtures
circulated in these loops have not yet been
received.

Hastelloy B loop 186 was operated with
NaF-ZrF,-UF, (50-46-4 mole %) for 1000 hr with
a hot-leg temperature of 1650°F.
constructed with the drain, which acts as a cold

This loop was

trap during operation, in direct line with the cold
leg. The lower loop joint was a sharp angle of
75 deg rather than the smooth bend used previously.
Visval examination showed a considerable deposit
of needle-like, magnetic crystals around the top of
the trap and on both sides of the lower joint. This
loop is to be compared with Hastelloy B loops
161 and 162, operated previously at 1650°F with
off-set traps, which showed only small scattered
deposits.

Molybdenum Loops

Three loops constructed of molybdenum jacketed
with type 310 stainless steel for oxidation re-
sistance have been operated at a hot-leg tempera-
ture of 1500°F for periods up to 1000 hr with
NaF-ZrF,-UF, (50-46-4 mole %) as the circulated
tluid.

subsurface void formation, or intergranular attack

No evidence of mass-transferred particles,

has been found. A rough surface, with depressions
to a maximum depth of 2 mils, was noted but is
believed to be associated with fabrication pro-
A thin, aitered surface or metallic-
appearing layer was also observed on the inner
surface in both the as-received condition and
following loop operation. The hot leg of loop 185,
which operated for 1000 hr, is shown in Fig. 6.3.

cedures.

Sodium in Hastelloy B Loops

Two Hastelloy B loops were cleaned with dry

TABLE 6.4. CORROSION FOUND IN HASTELLOY B THERMAL-CONVECTION LOOPS AFTER
CIRCULATING NoF-ZrF,-UF, (50-46-4 mole %) FOR 1000 hr AT 1500°F

 

Metallographic Notes

 

 

L.oocp No. Condition
Hot-Leg Appearance Cold-l.eg Appearance
163 Dry hydrogen cleaned  Pitting to a depth of 2 mils and subsurface voids No deposit
164 Dry hydrogen cleaned  Pitting to a depth of 2 mils and subsurface voids No depesit
179 As-fabricated Subsurface voids to o depth of 4 mils Small metallic deposit
181 As-fabricated Subsurface voids to a depth of 1 mil No deposit
182 As-fabricated Subsurface voids to a depth of 2 mils No deposit

 

73
ANP PROJECT PROGRESS REPORT

   
 
 

 

UNCLASSIFIED
T-8944

i
3
t
‘NE OO0 X0§2

 

Fig. 6.3, HotLeg of Motybdenum Loop After
Circulating NaF-ZrF ,«UF, for 1000 hr ot 1500°F.
Loop 185. 250X. Reduced 30%,

hydrogen and operated at a hot-leg temperature of
1500°F with sodium as the circulated fluid. Loop
188 was operated for 500 hr and loop 189 for
1000 hr. In both loops, deposits of magnetic
crystals were found in the lower coid leg and in
the hot horizontal section. The metallographic
examination of loop 188 showed a rough surface,
with pits and subsurface voids to a depth of
2 mils. These results confirm those presented in
the previous reporf.]

The susceptibility of Hastelloy B to the mass
transfer of nickel inthe circulated sodium prompted
the operation with sodium of three loops con-
structed from A-nickel. Two of the loops were
dry hydrogen cleaned prior to operation, while
the third was operated in the as-fabricated con-
dition. Dendritic, magnetic crystals were found
in the cold legs of all three loops after operation
for periods of 3500 and 1000 hr at a hot-leg
temperature of 1500°F.

Sodium in Inconel Loops with
Beryllium Inserts

G. M. Adamson A. Taboada
Metallurgy Division

Six Inconel thermal-convection loops with beryl-
lium inserts were operated with sodium as the
circulated fluid in a second series of sodium-
beryllium-lnconel compatibility tests similar to

74

those previously discussed. ! Loops were operated
for 500 hr at 1300 and 1500°F and for 1000 hr at
1000, 1200, and 1350°F. High-purity sodium from
the same batch was used in all the loops.
Metallographically, no attack and no deposits
could be found in any of the cold legs or in any
of the Inconel parts of hot legs that operated below
1350°F. The Inconel hot legs, including the
sleeves, of loops operated at 1350 and 1500°F
showed maximum intergranular attack to 2 mils.
The beryllium inserts showed subsurface-void
attack on the outer surface, as tabulated in

Table 6.5.

TABLE 6.5. DEPTH OF ATTACK ON QUTSIDE OF
BERYLLIUM INSERTS FROM INCONEL LOOPS
AFTER CIRCULATING SODIUM

 

 

Temperature of  Operating  Maximum

Loop No. Hot Leg Time Attack

(°F) (hr) (mils)
560 1000 1000 0
559 1200 1000 3
557* 1200 1000 4

555 1300 500 1.5
558* 1350 1000 9
556 1500 500 21

 

*These two loops may have been reversed in cutting.

Visual examination of the inserts reveaied no
general attack; however, large streaks or rough
areas were observed near the top or down one
side of all inserts. Both the inside and the
outside of the insert from loop 559 are shown in
Fig. 6.4. The spiral down the center of this
insert was found in all inserts except the one
from the loop operated at 1000°F. Microscopic
examination showed no change in the surface and
no attack under the spirals. The spirals were
accompanied by a dark nonmetallic deposit that
could not be identified by diffraction or seen under
the microscope. Neither of these areas of visible
attack seems to correspond to any expected flow
pattern. Loops are now being started to determine
whether the areas of attack correspond to machining
variables.
UNCLASSIFIED
T-6547

DIRECTION OF FLOW

T

 

 

 

e eesbrerarn L
3 OPNTHES

Fig. 6.4. Surface of Beryllium Insert from
Inconel Loop 559 After Circulating Sodium for
1000 hr ot 1200°F.

FORCED-CIRCULATION CORROSION
AND MASS TRANSFER

G. M. Adamson R. S. Crouse
Metallurgy Division

Preliminary corrosion results have been received
on the first three Inconel forced-circulation cor-
rosion and mass transfer testing loops operated
by Experimental Engineering (cf., sec. 3, “‘Experi-
All these loops
were terminated before their scheduled time, but

mental Reactor Engineering’’).

they operated long enough to provide some cor-
rosion data. Similar mixtures of NaF-ZrF,-UF,
(53.5-40-6.5 mole %) were circulated in all these
loops. The operating data provided by Experi-
mental Engineering and the corrosion data now
available are presented in Table 6.6.

Despite the turbulent flow and the increase in
the number of cycles, the depths of attack found
in these loops are only two to three times the
depths found in thermal-convection loops operated
for comparable periods. A typical hot-leg section
from loop 4690 is shown in Fig. 6.5. The data
are still too meager to permit any conclusions to
be drawn, but several results are worth noting.
A large difference was found between the depth of
attack on the inside of the bends and that on the
outside. Where a depth of attack on the inside of
a bend was 16 mils, directly opposite the attack
would be much lighter in intensity and to a depth
of only 5 mils. The inside and outside surfaces
of a bend from loop 4690 are shown in Fig. 6.6.
In loops 4696-C and 4694 differences in wall

PERIOD ENDING MARCH 10, 1955

  
 

UNCLASSIFIED
7-7130

X ¢ e
. e
® e i
* & 2 . o ;,".1 h i
» - .. / ¢ .//’V_, % \\"
. T -
> ‘\ g LA e
o v o . . ,, ‘ :~p &
.4 o » - . »
.. . - .O
& o
- -
s e 2 -
z

Fig. 6.5. Hot Leg of Forced-Circulation Inconel
Loop 4690 After Circulating NaF-ZrF ,-UF . 250X,
Reduced 31%.

thickness in the bends of from 42 to 47 mils were
found; but, as yet, no large differences have been
found in loop 4690. These loops were electrical-
resistance heated, and therefore more current and
higher temperatures existed where the walls were
thicker on the inside of the bends. |n loop 4694
differences were found thickness in
straight sections, and corresponding differences
in attack were noted. It has been shown (cf.,
sec., 8, '"Heat Transfer and Physical Properties'”)
that variations in flow also occur in the bends
that would cause higher temperatures on the inside.

in wall

A complete bend from loop 4694 was sectioned
longitudinally and examined. The depth of attack
was the same on both sides of the straight section
before the bend, but the difference increased very
rapidly as soon as any curvature was noted. The
difference remained approximately the same all
around the bend. At the exit end the difference
did not stop suddenly but continued into the
straight area beyond. The depth gradually de-
creased from the end of the curvature but some
difference was still present several inches beyond
the bend. This pattern of attack was predicted by
the flow studies. These differences seem to show
that mass-transfer is even more temperature sensi-
tive than it was thought to be. One other dis-
crepancy noted in the data was that in loops
4696-C and 4694 the deepest attack occurred in
the first leg of the heater rather than in the second
leg where the maximum temperature was supposed

75
ANP PROJECT PROGRESS REPORT

TABLE 6.6, DATA ON OPERATION AND CORROSION OF INCONEL FORCED-CIRCULATION
CORROSION AND MASS TRANSFER TESTING LOOPS

 

Loop Designation

 

4690 4696-C 4694

 

Operating Dota

Time, hr 774 625 519
Maximum fluoride temperature, °F 1500 1500 1500
Maximum wall temperature, °F 1740 1610 ' 1580
Temperature drop, °F 200 300 200
Approximate Reynolds number 10,000 10,000 15,000
Heater length

First section, ft 4 6 5

Second section, ft 4 8 7
Loop length, ft 48 54 52
Velocity, fps 6.67 5.94 9.19
Cause of termination Leak Pump bearing failure Leak

Corrosion Data

Maximum depth of attack, mils

First heater section

Straight 12 6 16*

Bend 12 15 20
Second heater section

Straight 15 7 8

Bend 17 11 11

 

*This straight section showed variations in wall thickness.

      

& * - ~
-;‘ i .- a8, '.,% ¢
. - . - . ™. .». o@*
- ‘ * * - #
s q - '
L
v o .- 1 . . b . .
e 5
* o
b4
. o
. o
' O
i x

. & el
."""‘0 \

- ¥

(@) ‘ . - (B)

4 -

Fig. 6.6. Inner (a) and Outer (b) Surfaces of a Bend from Forced-Circulation Inconel Loop 4690 After
Circulating NoF-ZrF ,-UF,. 250X. Reduced 23%.

76
to have occurred. However, the two heater legs
were found not to be of equal length, and therefore
the power input was higher in the first leg than
in the second.

Chemical analysis data have been received for
loop 4690. The batch analysis was reported to

contain 14.8 wt % U, 7 ppm Ni, 45 ppm Cr, and

60 ppm Fe. Two samples taken from the sump
tank after the loop was drained show the
following: 15.1 and 12.2 wt % U, 100 and

40 ppm Ni, 690 ppm Cr in both samples, and
115 ppm Fe in both samples.

The only cold-leg deposits found so far were in
loop 4696-C. Metallographically, a deposit, as yet
unidentified, was found in a single section near
the end of the cooling coil of this loop. In loop
4690 two deposits were found in the pump that
were not present in either of the other two loops.
There was a deposit of loose magnetic crystals
on the back of the impelier. These crystals ana-
lyzed 4.52 wt % Ni, 9.54 wt % Cr, and 1.18 wt % Fe
or (corrected to 100 % metal) 29.6 wt % Ni, 62.7
wt % Cr, and 7.7 wt % Fe. The second deposit
was at the liquid level on the sleeve around the
pump shaft. This deposit was not discrete crys-
tals, and, under the microscope, it was shown to
be a thin film of metal
crystals. A diffraction examination reported the
deposit to be a mixture of nickel and slightly
altered fluorides. Two different spectrographic
samples were submitted, and the resuits, which

surrounding fluoride

follow, gave the first indication of nickel mass

transfer,
Sample 1: Metal plus Fluoride Mixture
Cr 3 ppm
Fe 1 ppm
Ni Sppm
Co 0.1 ppm
Na Trace
U Trace
Zr Trace

Sample 1: Fluoride Mixture

Cr 2 ppm
Fe <0.02 ppm
Ni <0.05 ppm
Co <0.05 ppm
Na Trace
U Trace
Zr Trace

PERIOD ENDING MARCH 10, 1955

Sample 2: Magnetiec Phase

Ni >5 ppm
Cr <0.4 ppm
Fe >10 ppm

Sample 2: MNonmagnetic Phase

Ni <0.05 ppm
Cr >5 ppm
Fe <0.02 ppm

GENERAL CORROSION STUDIES
E. E. Hoffman

W. H. Cook C. F. Leitten, Jr.
Metallurgy Division

High-Temperature Tests of Molybdenum in
Contact with NoF-ZrF ,-UF

An attempt is being made to devise protection
for the cooling jacket of the fused salt—Inconel
in-pile forced-circulation loop experiment which
is scheduled for insertion in the MTR. A loop
failure would undoubtedly occur in the vicinity of
the nose of the loop if pumping of the fused salt
were interrupted for a short period. It is possible
that in the event of a pump failure, the fused salt
might reach a temperature as high as 2430°F,
which has been given as the approximate boiling
point of the fluoride fuel mixture to be used. It
has been proposed that a sheath of molybdenum
around the nose of the loop would afford sufficient
protection for the cooling jacket, and it was arbi-
trarily concluded that 30 min at 2430°F would be
more than adequate to determine the suitability of
A molybdenum
specimen was therefore placed in a molybdenum
container, and the container was then filled with
NaF-ZrF,-UF, (53.5-40-6.5 mole %). In one test
a molybdenum plug was welded into the top of the
capsule, while in a second test, the top was left
open.

molybdenum for this service.

These test containers were sealed first in
Hastelloy B and then in quartz to prevent oxi-
dation of the molybdenum. There was no weight
change of either specimen during the test, and
metaliographic examination showed no attack
(Fig. 6.7). However, the effects of alloying be-
tween the molybdenum and Hastelloy B capsules

may be seen in Fig. 6.8.

77
ANP PROJECT PROGRESS REPORT

Brazing Alloys on Inconel and Stainless Steel
in Sodium and in Fuel Mixtures

Static corrosion tests have been completed on
a series of T-joints which were submitted by the
Wall Colmonoy Corporation. The purpose of these
tests was to find a brazing alloy which had good
corrosion resistance to both sodium and the fuel

INCH

UNCLASSIFIED
¥-14263

 

 

0.003

 

 

 

 

 

Fig. 6.7. Surface of Molybdenum Specimen
After Exposure for 30 min at 2430°F to NaF.ZrF -
UF,. Etched with NH,OH + H,0,. 500X. Re-
duced 42.5%.

 

mixture NaF-ZrF4-UF4 (53.5-40-6.5 mole %).
Several of the brazing alloys listed in Table 6.7
as having been tested on Inconel have also been
tested on type 304 stainless steel. The results
Metallo-
graphic results indicate that only the brazing
ailoys 1-10 and B-11 of this series have fair cor-
rosion resistance to both sodium and to the fused
salt fuel mixture.

An Inconel T-joint brazed with alloy B-11
(10.8% P-9.2% Si—-80% Ni) is shown in Fig. 6.9 in
the as-received condition.
may be seen along the fillet surface of this brazing
alloy. Figure 6.10 shows brazing alloy B-11 after
exposure to NaF.ZrF -UF, (53.5-40-6.5 mole %)
for 100 hr at 1500°F; only nonuniform attack to a
depth of 1 mil is noticeable along the fillet surface.
Figure 6.11 shows the same brazing alloy after
exposure to sodium for 100 hr at 1500°F, The
very erratic surface attack to a depth of 4 mils
can be seen, along with several subsurface voids
to a depth of 9 mils. A brazing alloy such as
B-11 might be used in a heat exchange system as
a back-braze material where the surface of the
braze alloy would normally be exposed to the

of these tests were reported previously.2

Several small voids

 

2g, E, Hoffman, W. H, Cook, and C. F. Leitten, ANP
Quar, Prog. Rep. Dec. 10, 1954, ORNL-1816, p 80.

UNCL ASSIFIED
Y-14017

Fig. 6.8, Molybdenum-Hasteiloy B Capsule After Exposure for 30 min at 2430°F to NaF-ZrF -UF ,.

Bottom of capsule at left. Note alloying of inner molybdenum capsule with outer Hastelloy B container.

78
PERIOD ENDING MARCH 10, 1955

 

 

 

TABLE 6.7. RESULTS OF STATIC TESTS OF BRAZING ALLOYS IN SODIUM AND IN
NuF-ZrF4-UF4 (53.5-40-6.5 mole %) AT 1500°F FOR 100 hr
Alloy Alloy Weicht Ch
Designa- Composition Joint Material Bath eight Lhange Metallographic Notes
tion (wt %) (g) (%)
A-10 12 P—88 Ni Type 304 stain-  Sodium -0.0015 ~0.141 Attack along entire fillet surface
less steel to a depth of & mils
Flueride -0.0031 -0.30 No attack on surface of fillet
mixture
A-10 12 P88 Ni Inconel Sodium —0.0002 -0.028 Attack on entire fillet syrface to
a depth of 13 mils
Fluoride ~-0.0008 -0.108 Surface attack along fillet to a
mixture depth of 0.5 mil
B-11 10.8 P-9,2 Inconel Sodium ~0.0010 —-0.098 Erratic surface attack to a depth
S5i—80 Ni of 4 mils, subsurface voids to
a depth of 9 mils
Fluoride -0.0018 —-0.166 Fillet surface attacked to a
mixture depth of 1 mil
E-11 13 Si-87 Ni Type 304 stain~  Sodium -~0.0007 —-0.068 No attack present along fillet
less steel
Fluoride -0.0036 -0.358 Fillet completely attacked
mixture
F-11 9 Siw17.8 Cr— Type 304 stain-  Sodium 0.0 0.0 No attack on fillet
73.2 Ni less steel
Flueride -0.0052 -0.52 Attack on braze joint to o depth
mixture of & mils
H-10 10 P-4.3 Mo—  Inconel Soedium -0.0011 ~0.107 Uniform surface attack along
85.7 Ni fillet to a depth of 9 mils
Fluoride —0.0016 —-0.154  Attack along fillet surface to a
mixture depth of 0.5 mil
I-10 11.6 P=6.25 Inconel Sodium ~0.0014 -0.135 Erratic attack along fillet to a
Mn—-82.2 Ni depth of 4 mils
Flueride ~0.0016 —0.157 Attack along fillet surface to a
mixture depth of 0.5 mil

 

fluoride mixture and would be in contact with
sodium only in the event of a tube-to-header weld

failure.

Dissimilar Metal Mass Transfer in the System
Zirconium=Type 304 Stainless Steel-Sodium

The problem of dissimilar metal mass transfer
of zirconium to type 304 stainless steel in sodium
has been studied in seesaw tests in the tempera-
ture range 1000 to 1500°F. The test containers
were capsules made of type 304 stainless steel,
and the zirconium specimens were retained in the

hot zone by partially crimping the capsule wall.
Each specimen of zirconium was carefully cleaned
and weighed both before and after testing in order
to obtain weight-change data. Upon completion of
the test, each capsule was sectioned and samples
of each section were submitted for metallographic
and spectrographic analysis. Six samples were
cut from each container that taken together were
representative of all portions of the container.
Table 6.8 gives the conditions of the tests. In all
three tests a seesaw furnace with a speed of
1 cpm was used, and the duration of each test

was 100 hr.

79
ANP PROJECT PROGRESS REPORT

 

Fig. 6.9.

" UNCLASSIFIED
Y382

0.05

 

INCH

0.040

 

0.015

 

0.020

 

0.025

 

 

 

Inconel T-Joint Brazed with 10.8% P-9.2% Si=80% Ni in the As-Brazed Condition. Note

small voids at fillet surface. Etched with glyceria regia. 150X.

   

[ UNCLASSIFIED -
Y1360

 

0.015

0.020

  

 

 

A
g
L
fx? L £
L T et

 

 

Fig. 6,10, Inconel T-Joint Brazed with 10.8%
P-9.2% Si-80% Ni After Exposure to Static
NaF-ZsF ,-UF, ot 1500°F for 100 hr. Note slight
attack along fillet surface. Etched with aqua
regia, 150X. Reduced 39%.

80

Rk 2l 0
UNCLASSIFIED
Y8

L in.0s

 

NICKEL PLATE.

INCH

0.010

 

0.045

 

0.020

 

 

0.025

 

 

 

Fig. 6.11,
P -9.2% $i~80% Ni After Exposure to Static Sodium
for 100 hr at 1500°F. Note attack at fillet surface
and subsurface voids. Etched with glyceria regia.

150X. Reduced 37%.

Inconel T-Joint Brozed with 10.8%
TABLE 6.8. TEST CONDITIONS FOR STUDYING
DISSIMILAR METAL MASS TRANSFER IN THE
SYSTEM ZIRCONIUM-TYPE 304
STAINLESS STEEL-SODIUM

 

 

Hot-Zone Cold-Zone Temperature
Test No. Temperature Temperature Gradient
(°F) (°F) (°F)
1 1000 , 531 469
2 1200 700 500
3 1500 1030 470

 

It appeared that no thermal-gradient mass trans-
fer occurred in any of the tests, because there
was no deposit of zirconium found in the cold
zones. The cold-zone section of the type 304
stainless steel capsule used in test No. 3 is
shown in Fig. 6.12. Metallographic examination
showed a layer of fine particles in the hot-zone
section of each capsule.
nation of the surface of a specimen sectioned
from the hot zone (Fig. 6.13) of the capsule used
in test No. 3 showed no trace of zirconium or

However, x-ray exami-

compounds.  The precipitated layer
found in the hot zone reached a maximum thickness
of 2 mils, and it wos very similar to the layers
observed on type 304 stainless steel samples
carburized in the presence of sodium.
Spectrogrephic analysis revecled only a slight
trace of zirconium in the hot-zone section of the
capsule used in test No. 2. An increase in zir-
conium content was found in the capsule used in
test No. 3, but the amount could be considered to
be negligible, since it was of the order of 10~ 3%.
It was also noted in test No. 3 that a slightly
larger concentration of zirconium appeared in the
cold zone of the capsule than in the hot zone.
cases the zirconium samples gained
X-ray analysis revealed

zirconium

In all
weight during the tests.
a layer of zirconium oxide on each sampie. The

following tabulation gives the weight changes
found:
T N Zirconium Sample Weight
est Re. Change (g)
+0.0041
2 +0.0089
3 +0.0150

PERIOD ENDING MARCH 10, 1955

= UNGLASHFIED
- ¥-14370 :

NICKEL PLATE

INCH

 

7 o008

 

0.002

 

0.003

 

e 10.004

 

= 10.005

 

 

¢ 0.006

 

 

 

Fig. 6.12, Cold Zone of Type 304 Stainless
Steel Capsule After Exposure to Sodium in Seesaw
Apparatus for 100 hr at 1500°F, Etched with 10%
oxalic acid. 500X. Reduced 37%.

    

NOLASSIFIED
NICKEL. PLATE I e

  

INGH
|

 

 

L o ov . = ‘,“ C e ] ¥ . 0.005

 

0.005

 

 

Fig. 6.13. Hot Zone of Type 304 Stainless Steel
Capsule After Exposure to Sodium in Seesaw
Apparatus for 100 hr at 1500°F. Etched with 10%
oxalic acid. 500X, Reduced 42%,

The original weight of each sample was 17 g. |t
was apparent that under the conditions of these
tests, zirconium exhibits negligible amounts of
dissimilar metal transfer to type 304
stainless steel.

mass

Diffusion of Sodium into Beryllium

Tests were run to determine the extent to which
sodium penetrates beryllium metal in a beryllium-

81
ANP PROJECT PROGRESS REPORT

sodium-inconel static system. The beryliium
specimens and the sodium were sealed in Inconel
capsules and maintained at 1200 or 1500°F for
1000 hr. Five cuts 10 mils thick were then
machined from one surface of each specimen. The
sides of the specimen were machined off to a
depth of 50 mils to avoid sodiym contamination
from the edges. The turnings were carefully
collected and submitted for spectrographic sodium
analysis. The results of these analyses are
presented in Table 6.9.

TABLE 6.9. SODIUM CONCENTRATION IN
BERYLLIUM TURNINGS AFTER EXPOSURE OF
BERYLLIUM SPECIMEN TO MOLTEN SODIUM
FOR 1000 hr IN AN INCONEL CAPSULE

 

 

 

Depth of Sodium Concentration of Beryllium Layers

Be Layer (mg of Na/g of Be)
(mils)  After Test at 1200°F After Test ot 1500°F
0to 10 0.2 135

10 te 20 0.2 29.4

20 to 30 0.6

30 to 40 0.3 0.1

40 to 50 0.02 0.6*

 

*This opparent increase in sodium content is not con-
sidered to be significant.

It appears from the data in Table 6.9 that very
little penetration of beryllium by sodium will occur
at a temperature of 1200°F. As may be seen in
Fig. 6.14, the beryllium specimen in the test at a
temperature of 1200°F was attacked irregularly to a
maximum depth of 5 mils. In the test at 1500°F the
specimen was very heavily attacked to a maximum
depth of 20 mils (Fig. 6.15), and a 3- to 4-mil
porous metallic layer covered the surface of the
beryllium specimen. This layer is anisotropic and
therefore has been identified as either beryllium
metal or a beryllium-rich beryllium-nickel solid
solution. No surface layers could be found on the
walls of the Inconel capsules used in these tests;
however, there was quite a bit of fine precipitate
along the surface to a depth of 2 to 3 mils. This
precipitate may be either BeNi or Be,;Ni, particles.

Beryllium-Inconel Spacer Tests

Tests revealed previously3 that dissimilar metal
mass transfer of beryllium metal to Inconel across

82

LNCLASSIFIED
Y-14189

    

Fig. 6.14-
After Exposure to Static Sodium for 1000 hr at

Surface of a Beryllium Specimen

1200°F. Large voids are due to attack by sodium,
Unetched, 250X. Reduced 35%.

small sodium gaps is a serious problem at tempera-
tures in excess of 1200°F. Therefore a study is
under way to determine the effect of temperature
and spacer distance on the alloying of beryllium
with Inconel. Layers of the compounds BeNi and
Be, ,Ni,, both of which are very hard and brittle,
have been found on Inconel plumbing in past tests.
Thermal-convection loop tests® at a hot-zone
temperature of 1300°F for 1000 hr revealed a
Be,Ni; layer opproximately 20 mils thick where
an Inconel pipe and a beryllium insert were in
direct contact. In the same test in areas where a
6-mil clearance was present between the lnconel
and the beryilium, a 0.5-mil layer of the BeNi
compound was found on the surface of the Inconel.

The tests for determining the optimum spacing
between beryllium and Inconel in a sodium environ-
ment have been conducted with the sodium static,
because the maximum attack on beryllium speci-
mens and the only beryllium-nickel compound layers
on Inconel in thermal-convection loop tests have
been found in areas where the sodium was fairly
stagnant. In the tests completed to date, spaces
of 0, 5, and 20 mils were used between the Inconel
and the beryllium, and the specimens were exposed
to sodium for 1000 hr at 1200°F. The appearance

of the specimens after testing may be seen in

 

3G. M. Adamson et al., ANP Quar. Prog. Rep. Dec.
10, 1954, ORNL-1816, p 78.
UNCLASSIFIED

PERIOD ENDING MARCH 10, 1955

3 UNCLASSIFIED

 

Y-14190

   

Y-14191 |

 

 

 

 

0.007

 

0.008

 

0009

 

0.040

 

0.041

 

0.012

 

0.013

 

0.044

 

0.045

 

 

 

 

 

Fig. 6.15. Surface of a Beryllium Specimen After Exposure to Static Sodium for 1000 hr at 1500°F,

(z) Bright-field illumination.

(b) Polarized light — layer which formed on the surface is either pure

beryllium or a berytlium-rich beryllium-nickel solid solution. Unetched, 250X,

Figs. 6.16, 6.17, and 6.18.

alloying occurred where the specimens were in

As may be seen,
direct contact. Metallographic examination of the
surface of the Inconel specimen separated from the
beryllium by the 5-mil space revealed a maximum
of 0.2 mil of beryllium-nickel compound formation.
The surface of the Inconel specimen separated
from beryllium by the 20-mil space had no beryl-
lium-nickel compound layers; however, there was
an excessive amount of precipitate in the Inconel
grains to a depth of 1 mil. Beryllium and inconel
specimens which were held in direct contact during
testing are shown in Figs. 6.19 and 6.20, with the
6-mil layer of Be,,Ni, and BeNi which formed on
the surface of the Inconel during this 1000-hr test
being shown in Fig. 6.20. Spectrographic analyses
of drillings from the surface of the Inconel re-
vealed a beryllium concentration of 3.49 mg/cmz,
which is equivalent to a layer of pure beryllium
approximately 1 mil in thickness.

The beryllium insert and the Inconel sleeve
which surrounded it in a recent thermal-convection

 

BERYLLIUM
20~ MIL SPACE

INCONE L

5-MIL SPACE

BERYLLIUM

Inconel and Beryllium Specimens in

Fig. 6.16.
Positions Occupied During Exposure to Sodium for

1000 hr at 1200°F.

loop test with sodium as the circulated fluid are
shown in Fig. 6.21. This test operated for 1500 hr
with a hot-zone temperature of 1300°F, and ex-
cessive alloying occurred between the sleeve and
the beryllium specimen. These surfaces were
separated by a space of approximately 6 mils at
the beginning of the test.

83
ANP PROJECT PROGRESS REPORT

UNCLASSIFIED
-13863

-—— BERYLLIUM

~g——— INCONEL

 

 

Fig. 6.17. Surfaces of Beryllium and Inconel
Specimens Separated by the 20-mil Space During
Exposure to Sodium for 1000 hr ot 1200°F,

BUNCLASSIFIED
Y-12864

-«—— INCONEL

~——— BERYLLIUM

 

Fig. 6.18.
Specimens Separated by 5-mil Space During Ex-
posure to Sodium for 1000 hr at 1200°F, Gray and
black areas on beryllium indicate formation of
beryllium oxide, while the dark areas on the lnconel
indicate
Disregard areas of direct contact,

Surfaces of Beryllivm and Inconel

beryllium-nickel compound formation.

Cermets in NuF-ZrF4-U F4

The normally inert and refractory nature of cermets
makes them attractive for applications in which
solid-phase bonding (*‘self-bonding’’) is a problem.
Valve stems and valve seats are especially sensi-
tive areas where such bonding cannot be tolerated.
VYery clean surfaces promote solid-phase bonding,
and in fused fluoride salt systems at elevated

84

 

UNCLASSIFIED
8

Y-1395

—ag——— BERYLLIUM

-a—— |NCONEL

Fig. 6.19. Surfaces of Beryllium and Inconel
Specimens Held in Contact During Exposure to

Static Sodium for 1000 hr at 1200°F.

temperatures any surface contaminants are rapidly
reduced. Therefore, further evaluations have been
made of the corrosion resistance in seesaw appa-
ratys of Kennametal, Inc., cermet specimens K150A,
K151A, K152B, K162B, and D4675 in contact with
NaF-ZrF ,-UF, (53.5-40-6.5 mole %) for 100 hr at a
hot-zone temperature of 1500°F. These specimens
had previously shown promising corrosion resistance
in NaF-ZrF,-UF, (50-46-4 mole %), a somewhat
less severe corroding medium, but there were some
variations in duplicate tests.? The results of the
recent tests are given in Table 6.10, together with
the complete compositions of the specimens.

Comparisons of the as-received with the tested
specimens under a metallographic microscope did
not reveal attack on any of the specimens shown
in Table 6.10. In consideration of the medium
used and the temperature difference between the
hot and cold zones, the set of specimens from the
tests with the cold-zone temperature of 1230°F
should have had the most severe corrosion. The
discrepancies between past and present fused salt
corrosion attack on these Kennametal specimens
are believed to be due to the improved purity of
the fused fluoride salts used aond to improved
metallographic polishing techniques., The metallo-
graphie polishing of cermets to enable high magni-
fication examination of edges within accuracies
of 1 to 2 mils is very difficult. Special techniques
are being developed for polishing these very hard
materials.

 

4E. E. Hoffman et al., ANP Quar. Prog. Rep. June 10,
1954, ORNL-1729, p 68.
PERIOD ENDING MARCH 10, 1955

UMNCL ASSIFIED
Y-14365

INCH

 

0.001

 

0.002

 

 

0.003

0.004

 

0.005.

 

 

 

 

 

Fig. 6.20. The BeNi and Be,|Ni; Loyers Formed on Inconel Specimen Shown in Fig. 6.19 During Ex-
posure to Static Sodium for 1000 hr ot 1200°F. Unetched. 500X,

TABLE 6.10. RESULTS OF SEESAW TESTS OF TWO SAMPLES OF EACH OF SEVERAL CERMETS IN
CONTACT WITH NaF-ZrF4-UF4 (53.5-40.0-6.5 mole %) FOR 100 hr AT A HOT-ZONE TEMPERATURE OF 1500°F

 

Approximate

Sample Cha ngea (%)

 

 

Sample , Cold-Zone
Desi ) Composition (wt %) T
esignation emperature Dimensionclb Weight
(°F)
K150A 80 TiC-10 Ni-10 Nl:;Tc:TiC3 1385 +1.0 + 2.5
1230 +0.6 +0.05
K151A 70 TiC-20 Ni-10 NbTuTiC3 1385 +0.2 +0.1
1230 +0.2 +0.09
K152B 64 TiC-30 Ni~é Nch:TiC3 1385 +0.1 0.0
1230 +1.0 +0.09
K162B 64 TiC=25 Ni-5 Mo-6 NbTaTiC3 1385 +0.1 0.0
1230 +0.1 -0.02
D4675 97.5 WC-2.5 Co 1385 +0.2 +0.1
1230 +Q.1 +0.04

 

%Values based on figures to four decimal places; the dimensional measurements were made with a bench micrometer
and the weights were obtained with a Gram-atic balance. These percentu?e values should not be used to determine
relative corrosion resistance of different specimens; they are given here only to show their relative magnitudes.

Average percentage changes in thickness, width, and length,

85
ANP PROJECT PROGRESS REPORT

      
    

BERYLLIUM INSERT —————">

Fig. 6.21.

————— |[NCONEL SLEEVE

UNCLASSIFIED
Y-14407

Beryllium Insert and Inconel Sleeve from Thermal-Convection Loop in Which Sodium Was

Circulated for 1500 hr at a Hot-Zone Temperature of 1300°F. Clearance between Inconel and sleeve

was 6 mils. Note alloying of beryllium and Inconel,

The as-received and tested specimens of K151A,
K162B, and D4675 shown in Figs. 6.22, 6.23, and
6.24 illustrate typical specimen structures and the
absence of corrosion. In Figs. 6.22 and 6.23 the
TiC particles are the larger and darker colored
material and the lighter material is the binding
metal.

Single-Crystal Specimens of Magnesium Oxide

in Lithium and in Lead

Single-crystal specimens of magnesium oxide were
tested in static lithium and in static lead in Globe-

iron containers at 1500°F for 100 hr. These cor-
rosion tests supplement those previously reported
for single-crystal magnesium oxide tested in sodium

and in NaF-ZrF ;-UF (53.5-40-6.5 mole %).°

 

SE. E. Hoffman et al., ANP Quar. Prog. Rep. Sept. 10,
1954, ORNL-1771, p 94.

86

Two magnesium oxide specimens, nominally,
0.10 by 0.23 by 0.24 in., were cleaved from a piece
of synthetic magnesium oxide crystal. The speci-
men tested in lithium had a weight loss of 66.4%,
and it tock on the shape of the Globeiron container
in regions of contact. There was no attack on the
magnesium oxide specimen tested in lead.

FUNDAMENTAL CORROSION RESEARCH
G. P. Smith
Metallurgy Division
Mass Transfer in Liquid Lead

J. V. Catheart
Metallurgy Division

It was the purpose of this investigation to survey
the mass transfer properties in liquid lead of a
      
 
 

Y-13585

 
  

i

%

i

 

PERIOD ENDING MARCH 10, 1955

 

UNCL ASSIFIED |

INCH

0.002

UNCLASSIFIED
Y-13584

 
 

 

0.003

 

 

 

 

Fig. 6.22, Cemet K151A (70% LiC~20% Ni=10% NbTaTiC,) As-Received (a) and (b) After Exposure

to NaF-Z¢F,-UF, for 100 hr

1230°F in Seesaw Apparatus. Unetched.

    

{ UNCLASSIFIED [
TY13586

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

at a Hot-Zone Temperature of 1500°F and a Cold-Zone Temperature of
1000X. Reduced 5.5%.

 

INCH

0.002

 

| UNCLASSIFIED
yaser b

     

 

0.003]

 

1000 X

0.004

 

 

 

 

 

Fig. 6.23. Cemmet K162B (64% TiC~25% Ni-5% Mo=6% NbTaTiC,) As-Received (a) and (b) After
Exposure to NaF-ZrF -UF, for 100 hr at a HotZone Temperature of 1500°F and a Cold-Zone Tempera-
ture of 1230°F in Seesaw Apparatus. Unetched. 1000X. Reduced 6.5%.

87
ANP PROJECT PROGRESS REPORT

 

 UNCLASSIFIED |
Y-12679 f

   

INCH

0.002

UNCLASSIFIED
Y-12680

 

0.003

 

 

 

 

 

 

Fig. 6.24. Cermet D4675 (97.5% WC-2.5% Co) As-Received (a) and (b} After Exposure to NaF-ZrF .-
UF, for 100 hr at a Hot-Zone Temperature of 1500°F and o Cold-Zone Temperature of 1385°F in
Seesaw Apparatus. Etched with KOH-KsFe(CN)é. 1000X. Reduced 2%.

variety of metals and alioys with the aim of ob-
taining an insight into the roles of various important
alloying materials in fixing the resistance to mass
transfer of given alloys. Thus pure metals such
as chromium, columbium, and nickel were tested in
addition to alloys having practical structural im-
portance.  Work on this project has now been
terminated.

All tests were performed in small quartz thermal-
convection loops, the test specimens being fastened
in the hot and cold legs of the loops. In all cases
a temperature gradient of about 300°C was main-
tained across the loop, and the hot-leg temperature
was 800 to 810°C. Details of the construction and
operation of the loops, as well as of the experi-
mental results, may be found in earlier reporfs.6

The only new material tested during the past
quarter was type 310 stainless steel. A loop con-
taining this alloy plugged after 65 hr of operation
with hot- and cold-leg temperatures of 805 and
500°C, respectively. A transverse section of the

 

©J. V. Cathcart, ANP Quar. Prog. Reps. ORNL-1439,
p 148; ORNL-1771, p 100; ORNL-1816, p 88.

88

hot-leg specimen is shown in Fig, 6.25.

The results of this research project are summa-
rized in Fig. 6.26. The numbers at the ends of the
bars in the chart represent the time period of the
test, Except as noted on the chart, every loop
was operated until it plugged.

The pure metals tested, as a rule, showed rather
poor resistance to mass transfer, with the excep-
tions being molybdenum and columbium. These
{atter two metals were the only materials studied
which suffered virtually no mass transfer or cor-
rosion under the test conditions. At the other
extreme, nickel was found to be highly subject to
mass transfer; the nickel loops plugged in about
2 hr. Loops containing chromium and iron plugged
in about 100 and 250 hr, respectively, and there-
fore can be regarded as being intermediate between
molybdenum and columbium on the one hand and
nickel on the other with respect to resistance to
mass transfer.

The alloys studied could be conveniently grouped
on the basis of whether their resistance to mass
transfer was about the same as or greater than the
LEAD—STAINLESS STEE

- Pt b

¥

2.
e s
T -
& .

   

 

3T

‘*\ Fy‘ - - 5
s .
: CORRODED REG!ON
. “ i e .
S, ; : .
P LS
; Tl T~
;K . o fi’w #
- L #}"f
# o #
£ - W .
+ e
P ) A .
Yo - e S £
» . .
S f .‘\5‘
i
{ ‘

  

L INTERFACE —

  

+ vy

UNCLASSIFIED
Y-14253
T
. L
Z
r 0.001
T S 7 : 0.002
1 * - “Ey ) 3 . . "' "E."" 3
* \ " . F‘) < sw c 0003
4 \,\ =.>\ . . "_: .
' L S ~: L 3
.. £T ; ' . N i
.’ / x 3 vty me ¥
». # P h
* 0.004
‘f”’/ ) { 4 ! {’L
N ial 3 P
k'//{’\ ~‘Ab‘. . awfii s 2 &.S’;{' Lo
Lo oo "
_ o \ . 0.005
i - iw
* e :?’ ? \N“\: ur
e: :Xf* : "f:'.e N ) ‘
f. "?Vf - | L‘ ) * *fifl 0.006
b L x
# * Lt o
"; ; -~ O
S » F 0 -
i, ., ;; &\‘

PERIOD ENDING MARCH 10, 1955

 

 

 

 

 

 

 

 

 

 

Fig. 6.25. Transverse Section of Type 310 Stainless Steel Specimen Exposed to Liquid Lead in Hot

Leg (805°C) of Quartz Thermal-Convection Loop.
500X. Reduced 5%.

average of their pure metal components. Examples
of the former category are furnished by the 300
series stainless steels, Inconel, and Nichrome V.
On the other hand, loops containing the 400 series
stainless steels and alloys such as 45% Cr-55%
Co and Hastelloy B required much longer times for
plugging than would have been predicted on the
basis of the data for their pure metal components.
The results obtained with the 45% Cr~55% Co alloy
were particularly striking, more than 750 hr being
required for the plugging of a loop containing this
material. Plugging times of 100 and 80 hr were
recorded for comparable chromium and cobalt loops,
respectively.

A survey of the phase diagrams of the alloys
tested showed that those alloys having a greater
than expected resistance to mass transfer all con-
tain either a phase which is an intermetallic com-
pound or else the composition is relatively close
to that which would produce an intermetallic com-

Temperature gradient in the loop was about 300°C,

pound. Conversely, for the other alloys, either no
intermetallic compounds exist or the compositions
are far removed from those of possible compounds,
On the basis of the data available, it seems
reasonable to conclude that alloys containing an
intermetallic compound will, in general, possess a
higher resistance to mass transfer than those in
which compound formation is not possible. The
only exception to this rule among the alloys tested
was the 50% Cr-50% Fe alloy. This composition
corresponds closely to the theoretical composition
for the sigma-phase region inthe iron-chromiumdia-
gram, and it was thought that this alloy would pos-
sess arelatively great resistance to mass transfer.
However, the average plugging time for two such
loops was 3% hr. No completely satisfactory expla-
nation for the apparent anomaly has been found, but
it was thought that on account of the extreme fria-
bility of the specimens, the very short plugging
times might have been due to @ mechanical failure

89
06

 

COLUMBIUM
MOLYBDENUM
TYPE 446 STAINLESS STEEL
TYPE 410 STAINLESS STEEL
2% Si-14% Cr-84, Fe
HASTELLOY B (5% Fe ~28%, Mo -67 %, Ni)
25% Mo - 75%, Ni
45% Cr=55% Co
50% Mo -50%, Fe
16 /o Ni~37%, Cr-47%, Fe (AUSTENITE AND SIGMA PHASE) =] 380 hr
50 76 Cr-50% Fe (SIGMA PHASE ), 38 hr
BERYLLIUM
IRON
CHROMIUM, 100 hr
COBALT, 80 hr
TITANIUM, 5 hr
NICKEL, 2hr
30 % Ni— 70, Fe 275 hr
16 7o Ni-37 %5 Cr-47 Y Fe (AUSTENITE AND FERRITE)
TYPE 347 STAINLESS STEEL, {40 hr
TYPE 304 STAINLESS STEEL, 100 hr
INCONEL, 90 hr
TYPE 310 STAINLESS STEEL, 65 hr

NICHROME V, 12 hr

100 200 300 400 300
TIME (hr)

?

UNCLASSIFIED
ORNL-LR-DWG 5857

(/774 NO MASS TRANSFER

EE==] USUALLY LITTLE MASS TRANSFER

GROUP {

GROUP 2

6Rour 3 [ HEavy mass TRansFER

A—EXPERIMENT TERMINATED BEFORE COMPLETE
PLUGGING OF LOOPS

194 hr

 

600 700 800

Fig. 6.26. Summary of Tests of Mass Transfer of Various Materials in Liquid Lead.

LYOdIY SSTID0dd LIFrodd dNv
of the specimens.’ A topical report covering this
research project is being prepared.

Mass Transfer and Corrosion in Fused Hydroxides

M. E. Steidlitz W. H. Bridges
Metallurgy Division

Work that has been done during the past yec:lr8 on
the mass transfer of nickel in sodium hydroxide
has indicated a possibility of finding limiting top
temperatures and temperature gradients at which
mass transfer will not occur. In order to investi-
gate this more thoroughly and to extend the work
to other potential container materials, a new cor-
rosion test apparatus has been constructed.

The new system is essentially a multicontainer
duplicate of the single model used in the earlier
work. Five steel pots are connected through ap-
propriate valving to vacuum, purified helium, and
purified hydrogen lines and through a bubbler to
the exhaust line which extends up into a hood.

The sodium hydroxide is contained in a bucket
fabricated of the material being studied. The
bucket is hung from the cold finger, which is also
made of the test material, The pot, heated by an
external furnace, heats the bucket and the hy-
droxide, and an air jet, blown down the inside of
the tubular cold finger, provides the thermal
gradient. The hydroxide is circulated by thermal-
convection currents established between the bucket
and cold finger. Temperatures are measured on the
bucket and inside of the cold finger. A thermo-
couple outside the pot controls the furnace. A
few pilot runs have been made and scheduied test-
ing should begin shortly.

Spectrophotometry of Fused Hydroxides

C. R. Boston
Metallurgy Division

Construction of a high-temperature spectropho-
tometer for studying the fundamental nature of fused
salts has been completed. Approximate absorption
spectra have been measured for fused sodium hy-
droxide at various temperatures up to 700°C in
air with sodium hydroxide at 350°C as a reference.
Limitations of the instrument restricted measure-
ments to the wave length range 400 to 650 mpu.

 

7). V. Cathcart, ANP Quar. Prog. Rep. Dec. 10, 1954,
ORNL-1816, p 91. )

8w. H. Bridges, Met. Semiann. Prog. Rep. Apr. 10,
1954, ORNL-1727, p 52.

PERIOD ENDING MARCH 10, 1955

The experimental procedure consisted briefly of
placing a periclase {magnesium oxide crystal) cell
containing sodium hydroxide in the light path and
at a given temperature measuring the photocell
response at various wave lengths relative to the
response at 600 my, which is the wave length of
maximum response at 350°C. Repetition of this
procedure at various temperatures gave a series
of curves that intersected at 600 my and, by their
general form, gave some indication of how the
absorption was changing with temperature.

With increasing temperature a rapid increase in
absorption appears at the shorter wave lengths,
until, at 700°C, the absorption at 400 mu reverses
the trend and drops markedly; however, at 440 my,
the absorption again increases and results in a
sharp absorption peak at 440 mu. The long wave
length end of the spectrum does not appear to be
significantly influenced by temperature change.
These preliminary measurements are approximate
and will serve mainly as a guide in designing
future apparatus and experiments.

Improvements in experimental design are now in
progress. These include the fabrication of a deeper
cell to prevent the hydroxide from creeping out as
it frequently does with the present cell. In ad-
dition, auxiliary apparatus will be constructed to
permit maintenance of a controiled atmosphere over
the melt.

Fused Hydroxides as Acid-Base Analog Systems

G. P. Smith
Metallurgy Division

In the preceding quarterly’ an outline was given
of the application of acid-base analog concepts
in a qualitative way to reactions in fused hydroxide
systems. Some aspects of this application have
now been considered quantitatively. It was previ-
ously shown that acid-base analog reactions in
fused hydroxide mediums were controlled by the
following anionic self-decomposition reaction,

20H= = H,0 + 07~

As previously shown, the equilibrium constant for
this reaction is determined in part by the cations
which do not enter into the stoichiometry of the
reaction, that is, the alkali and alkaline earth
metal cations, because of the ability of these
cations to polarize the various species in the

 

%. p. Smith, ANP Quar. Prog. Rep. Dec. 10, 1954,

91
ANP PROJECT PROGRESS REPORT

reaction without significantly increasing the cation-
anion bond energy through resonance energy due
to covalent structures. Expressions have now
been derived which show quantitatively the way
in which these nonreactive cations control this
self-decomposition equilibrium. First, it should
be pointed out that fused alkali-metal carbonates
and sulfates when acting as reaction media form
acid-base systems which are closely analogous to
fused alkali-metal hydroxides. The corresponding
anionic self-decomposition equilibria are the fol-
lowing:

co,”" = COo, + 07~ ,

0,77 = 30, + 077
This correspondence will be useful inasmuch as
the literature contains very much more information
concerning reactions in fused carbonates and
sulfates thon in fused hydroxides.

For a reaction medium, whether hydroxide, car-
bonate, or sulfate, containing a single honreactive
cationic species, it has been shown thot there
exists a linear empirical relationship between the
ionic potential of the cation and the self-decompo-
sition constant K for the medium. Second, for a
reaction medium which contains two or more non-
reactive cationic species, ¢ thermodynamic formula
has been derived which relates the self-decompo-
sition constant K, of the medium to its cationic
composition. These two expressions can be com-
bined, in principle, to give a complete phenome-
nological description of the role of nonreactive
cations in acid-base analog reactions in the
mediums under consideration.

In 1951, Cart-
gave simple empirical formulas relating
the heats of oxysalt formation to the ionic potentials
of the cations based on their corrected univalent
radii. However, his relationships are not quite
those required. In 1952, Ramberg'' found that
there exist arbitrary parameters for each of the
alkali and atkaline earth metals which, when plotted
against the heats of oxysalt formation per two
equivalents, give a straight line for each kind of
oxysalt.

Empirical Relationships for K ;.
ledge'?

Ramberg’s work was therefore used as a
starting point to obtain the empirical relationships

shown in Fig. 6.27.

 

(1;20, H. Cartledge, J. Phys. & Colloid Chem. 55, 248
1}.

V14, Ramberg, J. Chem. Phys. 20, 1532 (1952).

92

UNCL ASSIFIED
ORNL-LR-DWG 5083

100 | I T | ! \ 1 |

 

 

Be

=
|

z (B
™o
o
\
4

 

 

 

 

 

 

20 40 [30] 80 100 120 140 160 180
AH (kcal)

Fige. 6.27. Heats of Self<Decomposition of
Hydroxides, Carbonates, and Sulfates as Functions
of the Cationic Potentials,

In Fig. 6.27 the AH values which are plotted are
for the following self-decomposition reactions
2MOH = H,0 + M0 ,
M,C0, = CO, + M0 ,
M50, = S0, + M,0 ,
where M is on alkali metal, and
M’(OH), = H,0 + MO ,
M’CO, = CO, + MO,
M’80, = 30, + M°O ,
where M’ is an alkaline earth metal. The reactants
and the products are in their standard states. Thus,
for example, in the case of the hydroxide curve,
the comparison is for the enthalpy change for re-
actions involving the same number of anions, and
therefore the only variable along the hydroxide
curve is the kind of cation. The numerical vaiuves
are taken from Rossini'2 and are for 25°C. Along
the vertical axis are plotted values of the icnic
potential (Z/7) on a logarithmic scale, where Z is
the charge on the cation and r is its ionic radius,
% The values for all the
metals, except cesium and beryllium, and mag-
nesium for the hydroxide curve, may be represented
remarkably well by straight lines.
Figure 6,28 is a similar plot for AF, rather than
AH. The values are from the same sources, except

as given by Pauling.’

 

]2F. D. Rossini et al., Selected Values of Chemical
Thermodynamic Properties, USNBS Cir. No. 500, 1952.

13, Pauling, Nature of the Chemical Bond, Cornell
Univ. Press, lthaca, 1940.
UNCLASSIFIED
ORNL-LR-DWG 5082

7.0

 

 

Mg

 

z/r (BTH

 

 

 

 

a 20 40 &0 80 100 120 140 160
AF {keal)

Fig. 6.28. Free Energies of Self~Decomposition
of Hydroxides, Carbonates, and Sulfates as
Functions of the Cationic Potentials,

for the sodium hydroxide value, which was computed
by Smith and Boston. 14 There are, of course,
fewer AF data than there are AH data. However,
it is evident that the straight-line relationships are
still preserved and the slopes of the AF vs Z/r
functions are the same as those of the AH vs Z/r
functions. Reliable data at elevated temperatures
are too scarce for presentation on plots such as
those given in Figs. 6.27 and 6.28., However, data
for the sulfates of caleium, barium, and potassium
were plotted for a series of high temperatures and
were found to follow straight lines as closely as
they do at 25°C.

Because of the relationship between the curves
in Figs. 6.27 and 6.28, it is reasonable to assume
that, over the straight-line portions of the curves,
the entropy terms are constants that are independent
of the cations. |f this result is combined with the
reaction isotherm, it can be stated that the straight-
line portions of the curves in Fig. 6.27 may be
represented by equations of the form

Z
|ong=a|og —| + &,
r

where K, is the self-decomposition equilibrium
constant, and a and b are constants which are
characteristic of the anions only:

The values for the beryllium compounds deviate
from the straight-line relationships in such a way
as to lead to greater stability than would be
expected. This effect is, perhaps, associated with
the extra resonance energy which is to be expected

 

146, P. Smith and C. R. Boston, ANP Quar. Prog.
Rep. Sept. 10, 1954, ORNL-1771, p 102,

PERIOD ENDING MARCH 10, 1955

from the very high polarizing ability of an ion of
such large ionic potential.

Dependence of K, on Cationic Composition. A
thermodynamic relationship has alse been derived
which gives K, as a function of composition for
mediums containing two or more kinds of cations.
The notation to be used to express concentrations
is presented first and then the derivation is out-
lined in terms of an isothermal-isobaric cycle of
four component processes.

The electrical equivalent fraction fii of Xl.+ ions
is by definition equal to the number of charges in
the melt due to Xi+ divided by the total number of
positive charges on all ions in the melt. There-
fore, by applying the requirement of electricadl
neutrality it is found that

and, for the Yl.++ ions,

 

In the first process,
2A0H—> 4,0 + H,0

A quantity of melt containing two gram-atoms of
OH is converted into one mole of water in its
standard state and solution of oxides. The free
energy change accompanying this process defines,
in terms of the reaction isotherm, an equilibrium
constant which is the self-decomposition constant
K, of the melt AOH:

AF] = —RT InK

In the second process,

d

A,0 + H,0—ZN,(X)),0
+ ENI.’(YI.)O + H20 ,
and thus the product of the preceding process is

separated into component oxides, each in its
standard state:

— NN o SN S o
AFg = XNAP ) o + ENJAFSy jg

- EN/F(y 5o

0 !
where AFO(X) o represents the standard free
"2

i

energy of formation of (X )20, and fi(x) o repre-
2

sents the partial molal free energy of {X.),0 in
the solution of oxides A,0. By combining terms

93
ANP PROJECT PROGRESS REPORT

in the above equation and applying the definition
of activity, there is obtained

+ 3N’lna

AF2 = RT [):Ni In a(Xf)2° ;

(Y)O J ’

where @, ) o is the activity of (X,),0 in the so-

lution of oxides A.,,O.
In the third process,

EN(X),0 + EN/(¥,)O + H,0 —> 32N (X,JOH
+ EN/(Y )OH), ,

and thus the products of the preceding process are
reacted individually with the water to yield the
corresponding hydroxides in their standard reference
states. For the free energy change for this process,
the following expression is obtained upon collecting
like terms:

AF, = SN, {QAFO(Xi)OH - APy o

- AFOHzo] + LN/ [AFO(YZ.)(OH)

- AFO(YI-)O - AF°H20
The quantities in brackets define, by means of the
reaction isotherm, two sets of equilibrium con-
stants, K, and Ki', which are the same as the self-
decomposition constants of the constituent hy-
droxides, (Xz.)OH and (Yz.)(OH)2, respectively, as
defined in an earlier progress report.” Thus,

AF, = RT(EN, InK, + 2N In K})
1 1 1 3
In the fourth process,

22N (X,)JOH + ENI(Y )(OH),— 240H ,

and thus the products of the preceding process are
mixed to form the starting solution. The free energy
change is

AF = E2Nz’F(Xi)0H + EN;F(Yi)(OH)z
- Z2’\71'A1E'@()<2.)OH - EN{AFO(YI.)(OH)Q :

Then, proceeding as for the second process,
3 = 2

+ ENZ.' In a(yi)(oH)z ,

94

where a(Xz.)OH is the activity of (X )OH in the

solution of hydroxides AOH.

The sum of the free energy changes for the steps
of the isothermal cycle must vanish. Therefore,
~AF = AF, + AF; + AF,. Writing these terms
out, changing the logarithm base to 10, dividing
through by RT, and collecting some of the like
terms, give

(1) logK, = EI_VFI. log K, + ):1-\71.' log K

“r o
+ 2N log —m8—
i
a(Yz.)(OH)z

%x,),0
+ 2N. log
! 2
“(x,)OH

where the activity terms are not the activities of
the various constituents in a hydroxide melt at
equilibrium with its self-decomposition products;
they are activities of oxides in a solution con-
taining nothing but oxides, and activities of hy-
droxides in a solution containing nothing but
hydroxides. Equation 1 is a thermodynamically
exact formula.

Next, it is assumed that the solutions consist
of jons and Temkin's rules'> are applied to the
activity terms. Equation 1 reduces to

(2) log K, = EN—I. log K; + EN; log Kz.’
Yy Yo
s i
+ ZN'T: iog———z—— ;
sz.YOH

2
Yx 7o
—_ 1
+ ENz‘ log
2 2
inVOH

where the y are ionic activity coefficients. In
particular, Yxu Yy and Yo are the activity coef-

7 7
ficients of Xz.+, Yz.++, and 077, respectively,
in a solution of oxides only, and yxz_, sz_, and ¥4
are the activity coefficients of Xz.+, YI.+)r
OH™ in a solution of hydroxides only,

The notation may be simplified by changing the
subscripts for the alkaline earth metals from
Yo Yo Ygooito¥ 00 Y ot Y s where
m is the number of kinds of alkali metals and 7 is
the number of kinds of alkaline earth metals. Thus,

, and

2/v
_ I yl. }/0
(3) |ong=EN.logK.+ZN.|og-——-—_ ,
1 1 1 2/7/
Y Y .
OH

 

15M.. Temkin, Acta Physicochim. U.R.S5.S. 20, 411
(1945).
where the summation is from unity to m + n and v
is the charge on the cation.

Two models of an ideal mixture of electrolytes
have been proposed as approximations to reality,
one model by Temkin'> and one by Herasymenko. '®
If Temkin’s model is a good approximation, then
the activity coefficients in Eq. 3 will approximate
unity to the same degree. |f Herasymenko's model
is a good approximation, then the above activity
coefficients will by no means approximate unity or
even constants. A considerable volume of compu-
tations published since 1945 not only overwhelm-
ingly supports Temkin's model as the better of the
two, but indicates that Temkin's model is a satis-
factory first approximation for a surprising variety
of fused electrolytes. Therefore, the equation

(4) log Kd = EN—z’ log Kz.

Furthermore,
Eq. 4 should be a better approximation than the
model of an ideal solution on which it is based,
inasmuch as the activity coefficients in Eq. 2
occur as ratios of similar terms which probably
tend to cancel.

is taken as a first approximation.

CHEMICAL STUDIES OF CORROSION

F. Kertesz
Materials Chemistry Division
Effect of Chromium Metal Additions on Corrosion
of Inconel by CrF, in Molten Fluorides
H. J. Buttram R. E. Meadows
Materials Chemistry Division
The addition of chromic fiuoride causes strong
attack of inconel exposed to NaF-ZrF, and
NaF-ZrF ,-UF, while chromous fluoride

apparently has little effect. Therefore experi-
ments have been carried out to determine whether

melts,

by addition of metailic chromium the chromic ion
can be reduced in the melt to the noncorrosive
chromous state and thus provide a protective
environment for the Inconel.

Inconel capsules containing the fluoride mixture
and CrF; additions with and without chromium
pellets were studied in customary tilting-furnace
type of test. The chromium concentration in the
melt after the test, when plotted as a function of
CtF, added before the test, was found to increase
at a faster rate when added metallic chromium was

 

Tép, Herasymenko, Trans. Faraday Soc. 34, 1245
(1938).

PERIOD ENDING MARCH 10, 1955

present than when the only chromium available as
the metal was that contained in the Inconel capsule
walls. Metallographic examination showed strong
attack (4 mils) when 3.3 wt % CrF; was odded to
NaF-ZrF, (50-50 mole %). When the same experi-
ment was run with the addition of chromium pellets,
the examination revealed only light to moderate
subsurface void formation to a depth of 0.5 mil.
Therefore the reaction

Cr + CrF3—> :.’:Crl'_'2

probably takes place with the chromium peilets
and protects the Inconel surface.

Protective Action of Small
Chromium Metal Additions

H. J. Buttram R. E. Meadows
Materials Chemistry Division

placed in
Inconel capsules containing 30-g quantities of
NaF-ZrF, (50-50 mole %) or NaF-ZrF,-UF, (53.5-
40-6.5 mole %) to determine the protective effect
of small chromium additions. The thicknesses
of plating utilized were 1, 2, and 3 miis, which,
on 6-in.-long, 40-mil, copper wire, introduced 32,
107, and 213 mg of chromium.

After tilting-furnace tests, analyses of the
capsule contents revealed very low concentrations
of copper (20 to 30 ppm) in all the mixtures.
Therefore, it is reasonable to assume that even
in the experiment in which the l-mil-coated wire
was used, all the chromium was not used. These
small additions were sufficient to limit the attack
on the Inconel to a depth of 0.5 mil.

Chromium-plated copper wire was

Effect of Fission Products
H. J. Buttram R. E. Meadows

Materials Chemistry Division

The effect of simulated fission products on the
corrosion of Inconel in fluoride melts was reported
previously.17 A repeat experiment in which
commercial-grade YF., was used as an oddition
to NoF-ZrF4-UF3 (53.5-40-6.5 mole %) tested in
a tilting furnace revealed that less than stoichio-
metric amounts of chromium were released from
the Inconel by the YF,. This is not surprising,
since consideration of the free energies of the
compounds involved indicates that no appreciable

 

174, J. Buttram and R. E. Meadows, ANP Quar. Prog.
Rep. Sept. 10, 1954, ORNL-1771, p 109.

935
ANP PROJECT PROGRESS REPORT

reaction of YF, (or other rare earth fluorides)

should occur.

Time Dependence of Corrosion in
Tilting-Furnace Tests

H. J. Buttram N. V. Smith

Materials Chemistry Division

Attempts to correlate corrosion data obtained
from tilting-furnace tests with those obtained from
thermal-convection loops have included some
experiments in which Inconel capsules filled with
NaF-ZrF,-UF, (53.5-40-6.5 mole %) were cycled
(4 cpm) at 800°C on the hot end and approximately
600°C on the cold end for 25, 100, 1000, and
2000 hr. Chemical analyses did not reveal any
substantial changes in amounts of chromium re-
moved from the capsule walls as a function of
time. Metallographic examination showed nearly
identical attack of 1 to 2 mils, with some evidence
that attack was more intergranular in the longer
exposures.

It seems apparent from these data that only the
reactions of chromium with the »
equilibration according to the reaction

“impurities’’ and

96

Cr + 2UF, = 2UF,; + CrF,

are important in this type of test. Mass transfer
in tilting-furnace tests with these materials is,
apparently, so slight as to be imperceptible.

Effect of Yalence State of Iron on Corrosion
of Inconel by Fluoride Mixtures

H. J. Buttram R. E. Meadows

Materials Chemistry Division

The effect of the valence state of iron on the
corrosion of Inconel by fluoride mixtures was
studied by adding FeF, and FeF; to Inconel
capsules  containing NaF-ZrF, -UF, (53.5-40-
6.5 mole %) and running the customary tilting-
furnace tests. The tests showed that the amount
of chromium removed from the Inconel was nearly
constant regardless of the quantity of FeF, added,
while it increased as a direct function of size of
the FeF, additions. Metallographic observations
agreed with the chemical analyses; the depth of
attack did not exceed 2.5 mils with any gquantity
of FeF, tested (up to 6 wt %), while the FeF,
was found to cause heavy subsurface void for-
mation to a depth of 8 mils.
PERIOD ENDING MARCH 10, 1955

7. METALLURGY AND CERAMICS

W. D. Manly

J. M. Warde

Metallurgy Division

Investigations were continued in the study of
the properties of nickel-base alloys containing
15 to 32% molybdenum, ternary alloys with a
nickel-molybdenum  base, and vacuum-melted
Hastelloy B. The studies include fabrication
experiments, tests of oxidation and oxidation
protection, stress-rupture examinations, and the
development of techniques for welding and brazing
the materials. An extensive study of the effects
of high-temperature aging on the microstructure
and physical properties of these materials is
also under way.

Stress-rupture design curves are presented for
Inconel tested in fused salts and in argon, for
comparison, at 1300°F and at 1500°F. Preliminary
data obtained at 1650°F are also given. A
summary of the results of tests of the oxidation
resistance of dry-hydrogen-brazed inconel T-joints
is presented. The tests made to date indicate
that most of the nickel and nickel-chromium-base
brazing alloys are suitable for service in an
oxidizing atmosphere at 1500°F, and several are
suitable at 1700°F,

Work also continues on the development of
fabricational techniques and their application to
the production of radiators, heat exchangers, and
other special items,

DEVELOPMENT OF NICKEL-MOLYBDENUM
BASE ALLQYS

H. Inouye J. H. Coobs
Metallurgy Division

M. R. D'Amore
Pratt & Whitney Aircraft

Fabrication Studies

Hastelloy B. The variables affecting the quality
of extrusions of billets of Hastelloy B were
studied. Hastelioy B melting-stock pellets were
used to prepare the forged billets supplied by the
Haynes Stellite Company. The Hastelloy B pellets
were vacuum melted and cast at pressures of
6 to 20 p Hg. The vacuum-melting technique was
used to eliminate volatile constituents in the

 

]H. Inouye and J. H. Coobs, ANP Quar. Prog. Rep.
Dec. 10, 1954, ORNL-1816, p 100,

commercial pellets in an effort to improve the hot
working characteristics of the alloy. Neutron-
activation, spectroscopic, and vacuum-fusion
methods of analysis of the alloy revealed that the
trace elements had not been effectively removed.

The alloys tested were vacuum-melted Hastelloy
B, vacuum-melted Hastelloy B plus 0.1% cerium,
airmelted and forged Hastelloy B, air-melted
Hastelloy B deoxidized with a titanium-nickel-
aluminum-manganese master alloy, and vacvum-
melted Hastelloy B decarburized with FeO and
deoxidized with calcium, In the tests of these
alloys the effects of homogeneizing the cast
billet, of various extrusion temperatures, of several
die designs, and of canning were studied.

The experiments have shown that temperature is
the variable that affects the extrudability of the
alloys. Consistently high recovery of sound metal
is obtained at an extrusion temperature of 2000°F,
Both tubing and rod have been made with fair
success, showed some
promise and will be continued, since, as shown in

Canning experiments

the previous repor'r,1 higher extrusion ratios are
attainable with the canned material because higher
temperatures are permissible. The importance of
increasing the deformation is indicated, since
during tube reducing ot room temperature the tube
blank made from a forged billet was reduced suc-
cessfully, while the tube blank extruded from a
cast billet fractured with no recovery of sound
metal.

Attempts to hot roll the extruded rods to sheet
were unsuccessful because of edge cracking and
center splitting during rolling. Center splitting is
characterized by the formation of two pieces of
approximately the same thickness while the hot
billet is being rolled. Rolling temperatures be-
tween 1950 and 2200°F were investigated at
reductions of 5 to 10% per pass. The alloy was
reheated in air between each pass,

Hastelloy B + 0.03% Ce.
experiments described above,
melts of Hastelloy B were being made. One ingot
was made with a 0.03% cerium addition in the
form of an 11% Ce—89% Al master alloy. Spectro-
graphic analysis revealed the residual cerium

Concurrent with the
smaller vacuum

97
ANP PROJECT PROGRESS REPORT

TABLE 7, 1.

HASTELLOY B PLUS 0.03% CERIUM (NOMINAL) IN AIR

Specimens annealed 1 hr at 2100°F in hydrogen
Strain rate — 0.05 in./min

ELEVATED TEMPERATURE TENSILE TESTS OF

 

 

 

) Test Yield Point, ] | )
Specimen Temperature 0.2% Offset Tensile Strength Elongation
Number ©F) (psi) {psi) (% in 2 in.)
vVT-9-9 1100 58,700 108,500 20
vT-9-8 1300 49,700 79,400 14
VT-9-7 1500 50,200 58,400 15
VT-9-6 1600 41,100 44,200 30
VYT-9-5 1650 41,800 43,000 38

 

TABLE 7.2, ROOM TEMPERATURE
BEND TESTS OF 0.065-in.-SHEET
HASTELLOY B PLUS 0.03% CERIUM

Specimen annealed 1 hr at 2100°F in hydrogen

Strainrate — 1 in./min

 

 

Time Aged
at 1500°F Bend Angle Results
(hr) (deg)
0 180 Broke, bent on itself
5 89 Did not break
50 103 Did not break
100 103 Did not break
200 85 Broke
500 106 Did not break

 

content to be 57 x 1073%.  The ingot was
successfully hot rolled at 2100°F, with a moderate
amount of edge cracking, and sufficient material
was obtained for some tensile and bend tests.
The results of these tests are given in Tables 7.1
and 7.2, The bend angles listed in Table 7.2
indicate that the alloy tested does not show
brittleness after aging at 1500°F,

The ductilities of vacuum-melted Hastelloy B
contdining cerium are compared in Table 7.3 with
the average values for commercial Hastelloy B,
as reported by Haynes Stellite Company. Impact
tests of commercial Hastelloy B were also made
at room temperature to determine the sensitivity

98

TABLE 7.3. COMPARISON OF THE DUCTILITIES OF
COMMERCIAL HASTELLOY B AND VACUUM-MELTED
HASTELLOY B CONTAINING 0.03% CERIUM

 

Elongation (%)

 

 

Test
Yacuum-Melted
Temperature Commercial
CF - llov B Hastelloy B
astelloy Containing Cerium

1100 22 20

1300 9 14

1500 13 15

1600 17 30

1650 18 38

 

of the test in showing the effects of aging. The
results are presented in Table 7.4. Impact values
generally decrease with aging time, and the em-
brittling effect of aging appears more pronounced

at 1300°F,

20% Mo-80% Ni. During this period, two 30-1b
vacuum melts and six extrusions of tube blanks
and rod were made of the 20% Mo-80% Ni alloy.
The tube blanks were drawn to 0.500.in.-dig,
0.055-in.-wall tubing and used for the fabrication
of three standard size thermal-convection loops.
The extrusion data were reported previously. !

The extruded rods were rolled to sheet at 2100°F,
and the reductions per pass were as hign as 30%.
The alloy showed no tendency toward edge
cracking, and no deleterious effects of aging were
evident from room-temperature tensile tests. The
tensile test data are tabulated in Table 7.5,

Elevated-temperature tensile tests were per-
formed with the alloy in the solution-annealed
condition and after aging for 500 hr at the testing
temperature. The ductilities of the specimens at
elevated temperature were disappointingly small
compared with the elongations obtained in room-
temperature tests. The elevated-temperature data
for this alloy are presented in Table 7.6. The
oxidation rate, weldability characteristics, creep-
rupture data, work hardening, and recrystallization
temperatures for this alloy were reported previ-
ously.]

24% Mo-76% Ni, Three 30-Ib vacuum melts of a
24% Mo-76% Ni alloy were prepared, and tubing
for three thermal-convection loops and a quantity

TABLE 7.4. STANDARD CHARPY IMPACT
TEST OF COMMERCIAL HASTELLOY B
(R-340, HEAT NO. 1235) AT ROOM TEMPERATURE

 

Impact Values (ft-lb)

 

 

 

PERIOD ENDING MARCH 10, 1955

of sheet were obtained from these melts, The
fabricability and oxidation rate of this alloy were
reported previously,! Room-temperature bend and
tensile tests and elevated-temperature tensile
and stress-rupture tests have now been completed.
Bend tests of specimens aged at 1500°F for periods
up to 500 hr showed no deleterious effects from
aging at this temperature, Since the alloy is a
single-phase at these temperatures, impairment
of the ductility would indicate impurity effects,

tests showed the
embrittling effect of the precipitation of beta
phase from the alpha-phase solid solution at
1300°F (Table 7.7). The elevated-temperature
tensile properties of this alloy have been de-
termined for both the annealed and aged con-

Room-temperature  tensile

ditions. Low ductilities were obtained between
1300 and 1600°F. The data are summarized in
Table 7.8,

32% Mo=-68% Ni. The extrusion properties and
the oxidation rate of the 32% Mo—68% Ni alloy
were described previously.! It is more difficult
to extrude this alloy than alloys containing less

 

 

 

 

Aging Time molybdenum, since it is tougher and lower ex-
(hr) Aged at 1300°F Aged at 1500°F trusion temperatures must be used because it is
- hot short, The alloyis readily hot rolled ot 2100°F
5}’2 112 and at room temperature, Sufficient tubing for one
25 38 49 thermal-convection loop for corrosion testing and
50 43 a quantity of sheet for mechanical property testing
were prepared.
100 43 Room-temperature bend tests show that aging is
150 32.5 rapid at 1500°F and that the alloy is character-
200 0.5 istically brittle after aging times greater than
' about 50 hr, as shown in Table 7.9. Plates of
500 17 35 beta plus gamma phase are found in this alloy
1000 29 after aging from the solution-annealed condition.
Aging of a 72% cold-rolled alloy at 1500°F shows
TABLE 7.5. ROOM TEMPERATURE TENSILE TESTS OF A 20% Mo—-80% Ni ALLOY
Specimen annealed 1 hr at 2100°F in hydrogen and aged 96 hr at temperature
Aging Yield Point,
Specimen Temperature 0.2% Offset Tensile Strength Elongation
Number (°F) (psi) {psi) (% in2in.)

DP1-26-1 Unaged 36,900 104,600 62

.2 1100 36,200 106,500 61

-3 1300 39,700 109,800 58

.4 1500 36,500 106,400 63

 

99
ANP PROJECT PROGRESS REPORT

TABLE 7.6. ELEVATED-TEMPERATURE TENSILE TESTS OF A 20% Mo-80% Ni ALLOY

Strain rate — 0.05 in./min

 

 

Speci Test Yield Point, T e S N El .
:Iec:;nEn Temperature 0.2% Offset en5|(e -;rengf (Vo'ngzuf'lor;
umber ©F) (psi) psi 2 in2 in.
Annealed 1 hr at 2100°F in hydrogen
DP1-26-29 1100 24,100 53,600 20
-30 1300 21,200 41,700 14
-31 1500 21,800 36,100 8.8
-32 1600 19,400 28,800 7.5
-33 1650 21,000 25,400 9
Aged 500 hr at testing temperature
-34 1100 22,500 45,100 1.5
-37 1500 21,500 34,800 6.5
-36 1600 20,900 29,000 7.5

 

TABLE 7.7. ROOM-TEMPERATURE TENSILE TESTS
OF A 24% Mo-76% Ni ALLOY (HEAT NO. VT-11)

Specimens annealed 1 hr at 2100°F in hydrogen

and aged 284 hr at temperature

Aging Yield Point,

 

Tensile El .
Temperature 0.2% Offset Strength ongs .ion
o _ ) (% in2 in.)
(" F) (psi) (psi)
Unaged 42,400 109,000 62
1300 88,500 123,000 4
1500 47,800 114,600 47
1650 42,000 110,000 47

 

equiaxed grains of beta plus gamma phase. The
properties of the alloy in this condition are not
known, except that the hardness remains high
(~500 VPN) up to aging times of 200 hr. Room-
temperature tensile tests of the alloy confirm the
brittle nature, as indicated in Table 7,10,

The elevated-temperature tensile tests of the
alloy show strengths comparable to those of
Hastelloy B; but the ductilities are low., Notch
sensitivity of the alloy was indicated by the

rupture of several specimens in the pin hole during

100

tensile tests. Plates spot welded onto the shoulder
to strengthen the area of the pin hole did not
help; the specimens failed in the heat-affected
zone. Adequate ductilities were obtained at
1100°F and at 1650°F (the alloy transforms from
beta plus gamma phase to alpha plus delta phase
at 1600°F). Data for this alloy in both the so-
lutionsannealed condition and after 500 hr of aging
at the test temperature are shown in Table 7.11,

Cb-Mo-Ni.  The columbium-molybdenum-nickel
ternary alloy system was partially investigated.
The alloys studied thus far contained 2, 5, or
10% columbium and 20% molybdenum; the balance
of each alloy was nickel, The 3-lb vacuum melts
of the 2 and 5% columbium alloys were readily
fabricated into sheet, but the 10% columbium alloy
fractured severely upon rolling at 2100°F.

The oxidation rate at 1500°F of alloys con-
taining 5 and 10% columbium was less than that
for the 20% Mo—80% Ni alloy but slightly more
than that for Hastelloy B. The scale that formed
in oxidation tests was nonadherent. The 2%
columbium alloy has not yet been tested.

The 2% columbium alloy is a single-phase be-
tween 1300 and 2100°F; however, Widmanstatten
structure appears in both the 5 and 10% columbium
alloys at 1300 and 1500°F. The 5% columbium

alloy shows no increase in hardness after 20 hr
PERIOD ENDING MARCH 10, 1955

TABLE 7.8. ELEVATED-TEMPERATURE TENSILE TESTS OF A 24% Mo-76% Ni ALLOY

Specimen annealed 1 hr at 2100°F in hydrogen
Strain rate — 0.05 in./min

 

 

Specimen Test Yield Point, Tensile Strength Elongation
. Tem;z)erature 0.2% O.ffse'r (oon G i
(" F) {psi)
DP1-25-19 1100 28,300 49,300 17.5
-20 1300 43,800 5
-21 1500 25,100 38,100 6.5
-22 1600 24,000 32,400 7.5
.23 1650 24,800 31,200 .3
-24* 1100 37,800 55,000 "
.27* 1500 24,200 38, 100 8
-26* 1600 25,200 35,000 10

 

*Aged 500 hr at test temperature.

TABLE 7.9. ROOM-TEMPERATURE BEND
TESTS OF A 32% Mo—68% Ni ALLOY (HEAT NO. YT-10)

AFTER AGING AT 1500°F

Specimens annealed for 1 hr at 2100°F in hydrogen

TABLE 7.10. ROOM-TEMPERATURE TENSILE TESTS
OF A 32% Mo-68% Ni ALLOY (HEAT NG, VT-10)

Specimen annealed for ]/é hr at 2100°F in hydrogen
and aged for 288 hr at temperature

 

 

Aging Time Bend Angle Hardness Resuire
(hr) (deg) (VPN)
Unaged 75 233 Did not break
5 27 265 Cracked

50 0 516 Broke
100 0 498 Broke
200 0 594 Broke
500 1.5 509 Broke

 

 

 

 

Aging Yield Point, Tensile El )
Temperature  0.2% Offset Strength (?O.ng;f.lor)]
(°F) (psi) (psi) st
Unaged 66,400 135,600 33
1300 122,500 1.0
1500 85,200 1.0
1650 121,800 2.5

 

 

of aging at either 1300 or 1500°F. The 10%
Cb-20% Mo-70% Ni alloy ages af both 1300 and
1500°F, as indicated in Table 7.12,

The room-temperature tensile data for the 2 and
5% columbium alloys were determined after aging
at temperatures of 1300, 1500, and 1650°F. The
results are shown in Tables 7.13 and 7.14. The
iow ductility of the 5% columbium specimen tested
after aging at 1650°F may be due to hydrogen
embrittiement,

Several other ternary alloys based on molybde-
num and nickel will be screened by creep, oxi-

dation, tensile, and fabrication tests. The stress-
rupture properties of the nickel-molybdenum and
the nickel-molybdenum-columbium alloys are
reported below.

Cr-Mo-Ni. The effect of chromium additions to
nickel-molybdenum base alloys is also being
investigated as a means of improving the corrosion
resistance of these alloys in sodium. [f small
chromium additions {up to 5%} are effective, these
alloys may also be useful in the fluoride salts,
since the small percentages of chromium involved
would not cause appreciable chromium mass trans-
fer. Also, the oxidation resistance, strength, and

101
ANP PROJECT PROGRESS REPORT

TABLE 7,11, ELEVATED-TEMPERATURE TENSILE
TESTS OF A 32% Mo~-68% Ni ALLOY (HEAT NO, DP1-24)

Specimens annealed 1 hr at 2100°F in hydrogen

 

 

Test Yield Point,  Tensile Elongation
Temperature  0.2% Offset Strength (% in2 in)
(°F {psi) (psi)
1100 44,800 113,400 32.5
1300 51,200 77,700*
1500 59,700*
1600 35,200 37,300* 1.0
1650 30,700 39,400* 11.3
1650** 30,400 41,800 17.0
1100 *** 53,700 85,500 18.8
150Q*** 82,700 1.3
1600*** 61,500 2.0

TABLE 7.13. ROOM-TEMPERATURE TENSILE TESTS
OF 2% Cb-20% Mo-78% Ni ALLOY (HEAT NO. VYT-12)

Specimens annealed 1 hr at 2100°F in hydrogen
and aged 284 hr at temperature

 

 

Aging Yield Point, Tensile £l '
Temperature 0.2% Offset Strength (70?9;?0';
°P (psi) (psi) o tm& n
Unaged 38,600 102,500 60.5
1300 39,400 104,600 65.5
1500 37,900 103,800 70.0

 

TABLE 7.14. ROOM-TEMPERATURE TENSILE TESTS
OF 5% Cb=20% Mo-75% Ni ALLOY (HEAT NO, YT.5)

Specimens annealed ]/2 hr at 2000°F in hydrogen
and aged 284 hr at temperature

 

 

*Specimens which failed in pin hole,
**Rerun of 1650°F specimen.

***Specimens aged 500 hr at testing temperature.

TABLE 7,12. AGING EFFECTS IN A 10% Cb-20%
Mo-70% Ni ALLOY (HEAT NO. AC.4)

Specimens annealed 1 hr at 2100°F in hydrogen

 

Hardness Mumber (VPN)

 

 

Aging Yield Point, Tensile '
Temperature  0.2% Offset Strength Elolngat.uon
°F) (psi) (ps)  (Pin2in)
Unaged 37,800 126,000 61.3
1300~ 128,500 2.5
1500 89,500 152,000 18.8
1650 65,200 102,500 6.5

 

 

Aging Time
(hr) Aged at 1300°F Aged at 1500°F
Unaged 193
5 332 216
25 353 2565
50 353 286
130 371 329
500 N 341

 

ductility of the nickel-molybdenum base alloys at
temperatures above 1500°F may be improved by
alloying with chromium,

Fourteen 30-lb vacuum-melted ingots of chro-
mium-molybdenum-nickel alloys have been ob-
tained, These alloys all have a nicke! base, and
they contain 20% molybdenum; the chromium con-
tent varies from 3 to 10%. All the ingots have
been machined into 3- by 3-in. extrusion billets.

102

*In vacuum.

Thirteen of the 14 alloy compositions have been
extruded into tube and/or rod. The extruded tubes
are being reduced to 0.50 in. in outside diameter
with 0.055-in. walls for use in thermal-convection
loops, The extruded rods are being rolled to
0.065-in.-thick sheet. The sheet is to be machined
into test specimens for determining strength prop-
erties of the alloy. The alloy compositions and
extrusion data are tabulated in Table 7,15,

The extruded rods showed severe edge cracking
during hot rolling at temperatures between 2000
and 2250°F. A chemical analysis of the electro-
lytic chromium used in the melting of these alloys
revealed an oxygen content of 0.45%. Therefore
to investigate the role of oxygen contamination
of the chromium on the fabricability of the alloys,
100-g arc-melt buttons containing high purity
TABLE 7.15.

PERIOD ENDING MARCH 10, 1955

ALLOY COMPOSITIONS AND EXTRUSION DATA FOR CHROMIUM-MOL YBDENUM-NICKEL ALLOYS

Atmosphere: Houghton's salt bath No. 1550
Heating Time: tube blanks, 30 min; rod, 45 min

Billets: 3 in. in diameter and 3 in. long

Die: 30- to 45-deg cone, alloy CHW

Mandrel:

Lubrication: glass wool in container, Fiberglas sleeving on mandrel stem

1-in. straight stem, alloy LPD

Dummy Block: Al + graphite for tube extrusions; 70-3Q brass + graphite for tube extrusions

Pressure Requirements {on 3-in. ram): 70 to 90 tons/in.2

 

Alloy Composition (wt %)

Extrusion Data

 

 

 

Soaking
Reduction E dabili
Cr Mo Ni Other Form No. Ratio Temperature xtrudability
(°F)
3 20 77 Tube 1 7.5:1 2200 Good
Rod 1 6:1 2200 Good
5 20 75 Tube 2 2300 Good
Rod 1 6:1 2200 Good
7 20 73 Tube 2 7.5:1 2250 Poor
Rod 1 6:1 2125 Good
10 20 70 Tube 2 7.5:1 2300 Poor
Rod 1 6:1 2200 Good
3 20 76.5 6.5 Cb Tube 2 5:1 2200 Goad
Rod 1 6:1 2125 Good
5 20 74.5 0.5 Cb Tube 1 7.5:1 2200 Shattered
Red 1 6:1 2200 Good
7 20 72.5 0.5 Cb Tube 2 7.5:1 2175 Good
Rod 1 6:1 2200 Good
10 20 695 0.5 Cb Tube 2 5:1 2175 One good; one shattered
Rod 1 6:1 2200 Goaod
3 20 76.9 0.1 Ce Tube 2 5:1 2200 Shattered
5 20 74.9 0.1 Ce Tube 2 5:1 2200 Shattered
7 20 72.9 0.1 Ce Tube 2 5:1 2250 Shattered
Red 1 6:1 2125 Good
10 20 69.9 0.1 Ce Tube 1 5:1 2125 Shattered
5 20 74.75 0.25 Ce Tube 2 7.5:1 2125 Good
Rod 1 6:1 2125 Goed
7 20 72.75 0.25 Ce Mot extruded

 

103
ANP PROJECT PROGRESS REPORT

chromium were prepared, The oxygen content of
the chromium was 0,011%. The buttons were hot
rolled at 2150°F from a thickness of 0.45 in, to a
thickness of 0.15 in, The alloy compositions and
the results of hotrolling are given inthe following:

Hot Rolling

Characteristics

Alloy Composition
{wt %)

3 Cr—-20 Mo-=77 Ni
5 Cr=20 Mo=75 Ni
7 Cr—20 Mo~73 Ni
10 Cr—20 Mo—=70 Ni

Slight edge cracking
Severe edge cracking
No edge cracking
No edge cracking

The edge cracking during rolling of the alloys
containing 3 and 5% Cr was attributed to melting
From the results obtained with the two
alloys of higher chromium content, it is apparent
that the difficulties encountered in hot rolling are
due to oxygen contamination of the chromium.
Therefore small amounts of cerium and columbium
were added to the melts for deoxidation purposes.
The columbium additions had no apparent effect,
but cerium added in a ratio of approximately 0.1%
cerium for each 3% chromium improves the hot
rolling properties of these alloys, It has not been
definitely determined whether the beneficial effect
of cerium on the hot rolling properties is due to a
deoxidizing action or to alloying. Arc-melts of
chromium-molybdenum-nickel alloys with various
percentages of aluminum or titanium added are
being prepared for studying the effect of known
deoxidizers on these alloys. Many of the alloys
tested thus far have shattered during extrusion in
the 2200°F temperature range. The shattering is
probably caused by too high an extrusion temper-
ature or by flaws within the extrusion billet.

practice.

Oxidation Studies

H. Inouye
Metallurgy Division

M. R. D’ Amore
Pratt & Whitney Aircraft

Oxidation tests at 1500°F of 168 hr duration
have been completed for the following alloys:
3% Cr-20% Mo-77% Ni, 5% Cr-20% Mo-75% Ni,
7% Cr-20% Mo—73% Ni, 10% Cr-20% Mo-70% Ni.
The data obtained are plotted in Fig. 7.1, and
oxidation curves for commercial and vacuum-
melted Hastelloy B are included for comparison.

The curve for vacuum-melted Hastelloy B is the

104

UNCLASSIFIED
ORNL-LR-DWG 5782

 

 

 

 

 

4.0
| i
35— - — — NOMINAL COMPOSITION (wt %)
O  30r-20Mo-T7 Ni
3.0 ® S5Cr-20Mo-75Ni  _ _ __ _ —
A 7 Cr-20 Mo-73 Ni
& 5 A 10 Cr-20 Mo-7C Ni
s S | O HASTELLOY B-VAGUUM MELTED
g B HASTELLOY B-COMMERGIAL
=
<
o
l._
x
@
o
=

 

 

 

 

 

 

 

 

 

0 50 100 150 200
ELAPSED TIME (hr)

Fig. 7.1.
Base Alloys at 1500°F in Static Air.

The Oxidation of Nickel-Molybdenum

The data reported
for the remaining alloys are the results of single
tests. Additional testing is planned to supple-
ment the present data. The oxide scale formed on
Hastelloy B and the 3% Cr-20% Mo-77% Ni ailoy

spalled completely during cooling to room temper-

average of two oxidation tests.

ature at completion of the test. Moderate spalling
of the oxide scale was observed on the alloys
containing 5, 7, and 10% chromium. From the
limited data it appears that nickel-molybdenum
base alloys containing over 5% chromium are more
resistant to oxidation than Hastelloy B at 1500°F.

STRESS-RUPTURE STUDIES OF
NICKEL-MOLYBDENUM BASE ALLOYS

R. B. Oliver D. A, Douglas
J. H. DeVan J. W. Woods
Metallurgy Division

An expansion of mechanical testing facilities
has recently been accomplished to provide ad-
ditional equipment for investigating the strengths
of Hastelloy B and related nickel-molybdenum
alloys under conditions encountered in a circu-
lating-fuel Six new lever-arm type of
machines (Fig. 7.2) have been
added for testing materials in fused salts, as well
as four direct-loading or dead-load type of units
(Fig. 7.3) for testing materials in argon, hydrogen,
Twelve new tube-burst units (Fig. 7.4)

reactor.
stress-rupture

or air.
PERIOD ENDING MARCH 10, 1955

| UNCLASSIFIED |
Y3952

 

Fig. 7.2. LeverrArm Stress-Rupture Machines.

105
ANP PROJECT PROGRESS REPORT

- A

UNCELASSIFIED

a

 

106

 

ines,

Rupture Mach

Dead-Load Stress-

3

ige 7

F
PERIOD ENDING MARCH 10, 1955

UNCLASSIFIED
Y-13946

 

 

Fig. 7.4, Tube Burst Test Unit,

107
ANP PROJECT PROGRESS REPORT

were also installed for the study of multiaxial
stress systems,

In order to provide space for this additional
equipment, it was necessary to outfit a new labo-
ratory, and thus there was some interruption of
the testing program during the past two quarters.
In addition to the machines mentioned above, the
new laboratory also includes six lever-arm units
and four tube-burst units used previously for
testing materials in fused salts and liquid metals.
Each testing machine and associated furnace has
its own temperature-recorder-controller, which is
mounted in the central control panel board shown
in Fig, 7.5.

The results obtained in the Hastelloy B testing
program are quite preliminary; however, a series
of creep-rupture tests on solution-annealed ma-
terial have been completed in an argon atmosphere
at 1500 and 1650°F, and a similar series of tests
is being run in fused salts. Based on the limited
data availeble, the corrosive action of the fused
fluorides has not adversely affected the creep-
rupture properties at 1500 and 1650°F. Rupture
times in the fluorides are compcrable to or greater
than those found in argon under similar stress
conditions. In all cases the results found for
Hastelloy B show marked improvement in strength

compared with Inconel for the same temperature
conditions. A comparison of the stress-rupture
properties found for Hastelloy B, type 316 stain-
less steel, and Inconel in argon at 1500°F is
presented in Fig. 7.6.

In conjunction with the testing of Hastelloy B,
tests are being conducted on the modified nickel-
molybdenum alloys described above.  Stress-
rupture results at 1500°F in argon have been ob-
tained for the alloys listed in Table 7.16. By
comparison with the data presented in Fig. 7.6,
it can be seen that these alloys are inferior in

TABLE 7.16. STRESS-RUPTURE PROPERTIES OF
NICKEL-MOLYBDENUM BASE ALLOYS
AT 1500°F IN ARGON

 

Time to
Stress

{psi)

Elongation

Rupture
°p (%)

(hr)

Composition (wt %)

 

78 Ni—20 Mo-2 Cb 8000 133 8
76 Ni-~24 Mo 8000 69 3

5000 358 2
68 Ni—32 Mo 8000 332 8

 

 

 

 

Fig. 7.5. Central Control Panel.

108
20,000

~

i
)
i

HASTELLOY B, ANNEALED AT 2100°F

10,000

5000

STRESS (psi)

2000

1000
10 20 50 100 200

Figo 7.6.
Inconel Tested in Argon at 1500 °F,

strength to Hastelloy B when tested under similar
conditions. The final elongations are also less
than those obtained for the commercial Hastelloy B.

Tube burst tests, which provide a study of
multiaxial stress systems, are presently under
way on %-in.-OD Inconel and Hastelloy B tubing
with 0.020- to 0.040-in. walls. A ratio of axial-
to-tangential stress of 1:2 and a test temperature
of 1500°F are being used. The tube is internally
pressurized with argon and is surrounded with
fluorides. Variable stress ratio tests, described
previously,? have been initiated to determine how
changes in tangential-to-axial stress ratios affect
tube failure. In conjunction with these tests, an
analysis will be made of the stress systems which
are present in pressurized tubes and of the rate
of creep deformation and failure which can be
expected to take place as a result of such stress
patterns.

 

2R. B. Oliver et al., ANP Quar. Prog. Rep, Sept. 10,
1954, ORNL-1771, p 112.

 

PERIOD ENDING MARCH 10, 1955

UNCLASSIFIED
ORNL—LR~-DWG 5783

TYPE 316 STAINLESS STEEL, AS RECEIVED

INCONEL., ANNEALED AT 2050°F

500 1000 2000 5000 10,000

TIME (hr)

Stress-Rupture Curves for Sheet Specimens of Hasteiloy B, Type 316 Stainless Steel, and

WEL DING AND BRAZING STUDIES OF
HASTELLOY B

P. Patriarca K. W. Reber
R. E. Clausing G. M. Slaughter
Metallurgy Division

J. M, Cisar

Aircraft Reactor Engineering Division

R. L. Heestand
Pratt & Whitney Aircraft

Radiators

Additional Hastelloy B radiator test components
have been fabricated, as described previously,®
to determine the feasibility of using this materidl
in applications involving thermal shock and high-

temperature oxidation. |t was thought that the

 

3p. Patriarca et al., ANP Quar. Prog. Rep. Dec. 10,
1954, ORNL-1816, p 103.

109
ANP PROJECT PROGRESS REPORT

characteristic aging of Hastelloy B with its ac-
companying loss in ductility might result in fis-
sures in the tubes or tube-to-header joints when
they were subjected to simulated cyclic service.
Table 7.17 presents a history of test radiators
built to date for evaluating these assemblies under
Typical components
of a radiator are shown in Fig. 7.7 before welding
of the split headers and after completion. Radiator
No. 1 (Table 7.17) differed from the other radiators
in that only a single row of five tubes was used.
The stress distribution onthe tubes in this radiator
was not comparable to that on a full-scale radiator,
and therefore this test was not run for the full
500-hr life test,

The tests of radiators No, 2 and No. 3 differed
in that radiator No. 2 was subjected to severdl
water quenches from 1500°F to simulate a more
severely stressed condition. Radiator No. 3 was

several service conditions.

 

carefully examined visually after 12 aqir cools
from 1500°F, since this was thought to be a real-
istic number of thermal cycles to which an actuadl
radiator might be subjected. As can be seen from
Fig. 7.8 the fin distortion was not severe enough
to cause a material increase in air-pressure drop
through the core. A testing temperature of 1200°F
was chosen for radiator No. 5 to obtain an experi-
mental check of the evidence obtained from micro-
structural studies that aging proceeded very rap-
idly at this temperature.

The test of radiator No, 4 was terminated after
69 hr at 1650°F because the severe, accelerated
oxidation caused leaks. (All specimens were leak
tested periodically by means of a helium leak
detector to determine soundness.,) Radiator No. 4
is shown in Fig. 7.9 after termination of the test.
Since the severe localized attack suggested the

S, UNCLASSIFIED |
E v.13913

Fig. 7.7. Hastelloy B Radiator Before Welding of Split Headers and Completed Test Radiator.

110
TABLE 7.17.

PERIOD ENDING MARCH 10, 1955

HISTORY OF HASTELLOY B TEST RADIATORS WITH STAINLESS-STEEL-
OR INCONEL-CLAD COPPER HIGH-CONDUCTIVITY FINS

Braze Alloy:

Coast Metals Alioy No. 52

Test Atmosphere: Air

 

Radiator Test Number

 

 

1 2 3 4 5
Number of tubes 5 10 10 10 10
Fin cladding material* Type 310 stainless  Type 310 stainless  Inconel  Inconel Type 310 stainless
steel stee! steel
Service temperature, °F 1500 1500 1500 1650 1200
Service time, hr 350 500 511 69 500
Number of eycles 8 27 265 8 191
Type of quench Water Water Air Air Air
Type of failure Nene Nene None Oxidized None

 

*Fifteen fins per inch.

INCHES

 

UNCLASSIFIED
Y-13930

 

Fig. 7.8. Hastelloy B Radiator No. 3 After 12 Air Cools from 1500°F, Note mild fin distortion.

111
ANP PROJECT PROGRESS REPORT

 

L INCHES

, UNCLASSIFIED
Y-13928

 

Fig. 7.9. Hastelloy B Radiator No. 4 After 6% hr at 1650°F in Air.

influence of a self-fluxing oxide, a series of ex-
periments was conducted to investigate the effect
of several oxides on the susceptibility of Hastelloy
B to this phenomenon., Each specimen was tested

for 24 hr at 1650°F and for 100 hr at 1500°F, The

results of these experiments are given in Ta-
ble 7.18.
Very serious accelerated oxidation occurred

both with Hastelloy B and Hastelloy C in contact
with copper oxide, A comparison of Hastelloy B
in contact with copper oxide for 24 hr at 1650°F
in air with an uncontaminated control specimen of
Hastelloy B tested under the same conditions is
shown in Fig. 7.10. Even though the copper oxide
was placed only in the area of pitting, which has
been brushed clean of scale, the entire surface of
the specimen shows a different type of scaling
than that on the control specimen. Since the
scaling was not observed on specimens exposed

112

to copper oxide at 1500°F, tests were carried out
at intermediate temperatures to determine the flow
point of the composite oxide, and it was found to
have a flow point of around 1525°F,

Since the Inconel-clad copper fins were also
adversely offected in radiator No, 4, it is sus-
pected that a volatile oxide, MoO,, condensed on
the fins near the headers with a consequent ac-
celeration in the rate of oxidation, As shown in
Figs. 7.9 and 7.11, the fins close to the headers
were completely oxidized, and the others were
only distorted. It is thought that the high velocity
air may have carried the volatile oxide away with
a consequent reduction in the deleterious effects
on the fins farthest from the headers.

These experiments illustrate that
protection of the copper exposed when the clad-
copper high-conductivity fins are punched out is of
extreme importance and that brazing techniques

complete
TABLE 7.18. EFFECTS OF VARIOUS OXIDES ON THE FLUXING OF

PERIOD ENDING MARCH 10, 1955

HASTELLOY B AND HASTELLOY C OXIDE

 

 

Type of Specimen Time (hr) Temperature (°F) Result
Hastelloy B + A1203 24 1650 Slight oxidation
100 1500 Slight oxidation
Hastelloy B + Cu0O 24 1650 Pitting to 40 mils
100 1500 Pitting to 15 mils
Hastelloy C + CuO 24 1650 Pitting to 40 mils
100 1500 Slight oxidation
Hastelloy B + SiO 24 1650 Slight oxidation
100 1500 Slight oxidation
Hastelloy B + FeO 24 1650 Slight oxidation
100 1500 Slight oxidation
Hastelloy B + type 316 stainless steel 24 1650 Slight oxidation
100 1500 Slight oxidation

 

 

UNCL ASSIFIED
Y-14038

 

Fig. 7.10. (a) Hastelloy B Oxidized at 1650°F. Note characteristic scale. {b) Hastelloy B + CuO
Oxidized at 1650°F. Note pitting and change in scale characteristics.

113
ANP PROJECT PROGRESS REPORT

 

UNCLASSIFIED
Y-140%94

Fig. 7.11. Finned-Tube Section of Hastelioy B Radiator No. 4.

should be developed to ensure complete coverage
by the brazing alloy. This problem is further
discussed below under the heading ‘‘High-Conduc-
tivity Fin Radiator,”’

Welding of Thick Sections

The fabrication of a pressure shell from Hast-
elloy B will present welding problems of an en-
tirely different nature from those encountered
in the tube-to-header welding of this material, As
a means of evaluating the weldability of the ma-
terial and to determine suitable procedures for
its fabrication, a heavy-welding investigation is
under way in which the metallic arc and the semi-
automatic heliarc welding processes are being
used,

The mefc”ic-orc welds were made on as-re-
ceived / by 3- by 4-in. pl afes by using a 90-deg
lncluded-ongle bevel, with a /] -in. land, and a
/16"” spacing between plates in all cases, except

114

plate No. 4 (Table 7.19). The welds were made
with both l/é- and 5/32-in. Hastelloy B coated elec-
frodes. All plates were tack-welded to a strong-
back to simulate a severely restrained condition.

Tests on the plates listed in Table 7.19 con-
sisted of 180-deg guided-root and face bends in
the as-welded and aged conditions. Tensile
specimens were also cut from plate No. 3 for
testing in the as-welded and aged condition. All
aged specimens were treated at 1500°F for 200 hr,

It may be noted that the bend-test results pre-
sented in Table 7.19 are somewhat inconsistent.
The plate No. 1 failures may be attributed to
incomplete fusion in the root pass., Plate No, 2
was welded in much the same manner as plate
No. 1, but particular care was used in the de-
position of the root pass. All bend tests on this
plate were made in the as-welded condition, and
no failures were observed., Plate No. 3 was tested
PERIOD ENDING MARCH 10, 1955

TABLE 7.19. RESULTS OF BEND TESTS OF WELDED HASTELLOY B PLATES (?'/8 BY 3 BY 6 in.)

 

 

 

i Bend Tests
b ter of
Plate No. Current {amp) El lcmeder ? N;mber of Bevel
ectrode (in.) asses Number of Tests Results
1 80 ]/8 Single 3 root Two welds failed
3 face None failed
2 20 ]/8 Single 2 root None failed
1 face None failed
3 100 5/32 Single 1 root None failed
1 face None failed
1 root, aged Failed
1 face, aged Failed
4 A 95 % 7 Double 2 None failed
100 562 7 2 One failed

 

to determine the effect of aging on the bend-test
resuits,

Several tensile and guided-bend test specimens
are shown in Fig. 7.12. It can be seen that speci-
mens 4 and 5 are typical as-welded face and root
bends, while 6 and 7 are the aged face and root
bends, respectively, from plate No. 3. An as-
welded tensile specimen and a welded-and-aged
tensile specimen were also prepared from plate
No. 3. The as-welded specimen (specimen 1 in
Fig. 7.12) exhibited good ductility in the parent
material but fracturing occurred in the weld. The
aged specimen (specimen 2 in Fig. 7.12) also
showed good ductility in the parent material, but,
in this case, the fracture occurred in the parent
metal.

One-half of plate No. 4 was prepared with a
1/S-in.-dio coated electrode, while the remaining
half was welded with a 5/32-in.-c|ic| coated elec-
trode. Radiographic inspection showed that all
welds were sound. Two bends from the fillet for
which the lé-in.-dia rod was used exhibited good
ductility, while one bend from the fillet made by
using 5/32-in.-<:|ic1 rod fractured during testing.
Inspection of the fracture showed no apparent
flaws,

The inconsistencies in the bend-test data could
possibly be attributed
yielding observed in the weid zone. The root-bend

to the inhomogeneous

specimen (specimen 5 in Fig. 7.12) shows the
extent to which such yielding was present. Speci-
men 3 in Fig. 7.12 was machined from a double-
beveled specimen to determine the effect of
bending axially with the weld. Some parting was
noticed along the fusion line, but there was no
fracture. It is evident that yielding in this longi-
tudinal test is more homogeneous than in the
transverse bend test.

Several weld beads have been
prepared with the semiautomatic Aircomatic ma-
chine available in the Welding Laboratory. These
welds were deposited on }/z—in.-thick Hastelloy B
plate with 0.060-in.-dia Hastelloy B wire as the
filler metal,

experimental

In this process, the filler wire is
continuously fed from a coiled spool. Conditions
have been established which produce good bead-
on-plate welds as determined from visual obser-
These conditions are listed in the follow-
ing to present an indication of the rapid travel
speed and deposition rate which can be obtained
by this process: arc current, 325 amp; welding
speed, 325 in./min; rate of wire feed,
200 in./min; argon consumption, 40 cfth. Metallo-
graphic examination of these welds indicated that

vation.

travel

porosity would not be a major problem when
welding with the semiautomatic equipment. How-
ever, further experiments will be conducted to
evaluate the physical properties of these welds in

115
ANP PROJECT PROGRESS REPORT

 

UNCLASSIFIED
©Y-14351

 

Fig. 7.12. Metallic-Arc Welded Hastelloy B Bend and Tensile Specimens. (1) As-welded — broken in
weld zone, (2) welded and aged — broken in parent material, (3) as-welded — bent along axis of welding,
(4) as-welded face bend, (5) as-welded root bend, (6) welded-and-aged face bend — broken along fusion
zone, {7) welded-and-aged root bend — broken through center of root.

order to compare them with the properties of welds
made by the conventional, manual, heliarc process.

High-Temperature Aging

It has been realized that at service temperatures
ranging upward from 1000°F copious quantities of
precipitated phases would be formed throughout
the microstructure of Hastelloy B. Preliminary
observations also indicated that the degree of
precipitation in weld deposits and in weld heat
affected zones might be noticeably different from
that found in wrought structures. An investigation
has therefore been initiated to determine the effect
of several variables on the type and quantity of
precipitate, as determined by metallographic

examination. An attempt is also being made to

116

correlate the physical properties of the material
with the observed microstructures,

The variables being investigated in this study
include aging temperature, time at this temper-
ature, prior thermal history, degree of residual
cold deformation, base metal composition, original
and aging environment, The
relatively homogeneous microstructure of a section
of 9’]6-in. Hastelloy B plate after a solution heat
treatment of 2 hr at 2150°F is shown in Fig. 7.13.
A similar sample after aging for 1500 hr at 1500°F
is shown in Fig. 7.14, which illustrates the typical
appearance of the precipitate formed at this
temperature after extended aging times. A photo-
micrograph of the same sample at a higher magni-
fication (Fig. 7.15) shows the diverse nature of

microstructure,
the precipitated phases. The grain boundaries are
clearly outlined by the more massive particles.
Aging at 1300°F for prolonged periods produces
the characteristic microstructure shown in Fig.
7.16. The marked difference in structures ob-

 

Fig. 7.13. Microstructure of 3/16-in. Hastelloy B
Plate After a Solution Heat Treatment of 2 hr at

2150°F. Etched with chrome regia. 150X, Re-

duced 30.5%.

PERIOD ENDING MARCH 10, 1955

tained at 1300°F and at 1500°F would be expected
to produce appreciable wvariations in physical
properties. Hot tensile specimens have been
subjected to long-time aging treatments and are
now being tested to confirm this and other pre-
dictions based upon this metallographic evaluation.

The quantity of precipitate formed was found to
increase proportionately with increasing time at
temperature within the limits of this investigation,
A 1950°F ‘“‘overaging’’ heat treatment, recom-
mended by the Hoynes Stellite Company, noticeably
reduced the rate of precipitation after short-time
aging, with the effect still somewhat evident up

to 1500 hr,

Residual cold work in Hastelloy B produces a
striking effect upon the rate of precipitation during
aging, as evidenced by the excessive quantities
of precipitate shown in Fig. 7.17, a specimen cold
worked (20% deformation) and held for 100 hr ot
1200°F. Evidence indicates that the induced
stresses produced during water quenching from
above 1950°F may increase this precipitation
rate. The influence of base metal composition is
now being studied, and aging treatments will be
performed on specimens of selected chemical
analysis. Extreme variations in microstructure

  

e RS NCLASSIFIED
SN Y40ss |
o e N \\”*4”‘ }*’4% x

\ ,‘? )';_'(.;f'& :'?‘“t

 

Fig. 7.14. Sample Similar to That Shown in Fig. 7.13 After Aging for 1500 hr at 1500°F, Etched with

chrome regia. 150X.

117
ANP PROJECT PROGRESS REPORT

  

o Tt N ey o - Loy ML . . - :
o7 LAk o Mty s Ao - UNCLASSIFIED
A # i ¢ - o -~ i ¥ . o
Cog A LT e e AT T Y4058
-~ Ly 1 T P
0t . yfi B ¥ ! 5 ) \ i
(P, ‘ L = }1() :“% I‘ R A ‘
22 L }"{1 v i \ .y L\Eg Y *
\ oo L ~ .\"s‘ :
?‘x ©w & \ VT . o ‘."Xia/f
P % ey i 3 k.
e . i - Ar/(f N s
J”f't‘n ~ bk L = f; ofly . n7
F A % jff > ” 3 o o
Nl % 3"‘ e e G l) . 1 Y
*on " Ay e 3
PR . ::bmi‘ .7 : - ‘;7 ” '
f g o ‘i‘ %:‘”s Y A oY ok
Jab N AL
’ ;"gf‘\ ’ . (f{,-”s““
o . o
\ F \’? . idd;”“ ' / e,,"(/ W%
; 3 ) N4 e
’»}I.{ * 'f “?“, ) :-"""P 3 E i.‘ “\_hrg“%fia*—fi‘w?“
S Lo * L e T Te
/ Bt L e ’&?\ ;
f I oo ? o,
ek - R W - \«
}},,{ # o s Ji -

  

: G L
- \ R it .. . W\fii .«-«""”‘f‘ ‘
je i . e

Fig. 7.15. Higher Magnification (750X) of Fig. 7.14. Note the several types of precipitate present and
especially, the more massive particles outlining the grain boundaries. Etched with chrome regia.

r

 

Fig. 7.16. Hastelloy B Aged at 1300°F for Prolonged Periods. Note characteristic precipitate,
Etched with chrome regic. 150X,

118
~ e smpons

. UNCLASSIFIED

 

Fig. 7.17. Hastelloy B Cold Reduced 20% and
Subsequently Aged 100 hr ot 1200°F. Note ex-

cessive precipitation. Etched electrolytically.

150X. Reduced 30.5%.

are prevalent in welded joints, as would be ex-
pected since they are susceptible to both segre-
gation in the fusion zone and fo stresses from the
welding process. The original microstructure, as
developed by fabrication techniques or thermal
treatments, has a decided effect upon the aging
process, but each individual case should be
evaluated separately. Except under corrosive or
oxidizing conditions, the environment does not
appear to be important,

Early experiments indicated that it would be
desirable to develop a heat treatment which would
result in improved ductility ot room temperature
and at high temperatures. Since a spheroidized
structure should be beneficial, a heat treatment
has been developed to produce it. This structure,
which is shown in Fig. 7.18 at a magnification of
150 diameters and in Fig. 7.19 at a magnification
of 2000 diameters, can be obtained by a heat

treatment4  which is not deemed practical for
structural components. However, physical test
specimens containing the spheroidized micro-

structure are being prepared to evaluate the effect
of spheroidization onductility and other properties.
If the results are promising, further work will be
aimed at developing a more practical heat treat-

ment,

 

4Aging of a 20% cold reduced specimen at 1200°F for
200 hr followed by spheroidization at 1540°F for 150 hr.

PERIOD ENDING MARCH 10, 1955

UNCLASSIFIED
¥.i3se8

 

 

 

 

 

 

 

 

 

 

 

 

 

Fig. 7.18. Microstructure of Hastelloy B Ob-
tained by the Experimental Spheroidizing Heat
Treatment.  Etched with chrome regia., 150X,
Reduced 30.5%.

       
 

UNCLASSIFIED

;‘*21 *
i‘d i *

&
Y
L

i
.

aanu
:.'.\'gs

P
:%i

&

&

G

* |

»

o

oo ® il
”~
*
o
35 s
¥

G

€
¥
2, i,
e
&s'f&i.g

Fig. 7.19.
Fig. 7.18.
Reduced 31%.

Higher Magnification (2000X) of
Etched with chrome regia. 2000X.

As a part of the correlation of physical proper-
ties with microstructure, hot tensile tests have
been completed on wrought specimens aged for
100 hr at five different temperatures. The yield
strengths and elongations were determined for
specimens
aged’’ conditions and are shown in Fig. 7.20. |t
can be seen that a minimum elongation exists at
1300°F, while the yield strength does not appear
to be as seriously affected.

in the solution annealed and ‘‘over-

119
ANP PROJECT PROGRESS REPORT

UNCLASSIFIED
ORNL-LR-DWG 5784
80

L LT o]
T

70 24 SPECIMENS SOLUTION-ANNEALED FOR 2 hr AT ——- 4 7Q
2150°F AND OVERAGED FOR 70 hr AT 1950°F |
— BEFORE 100-hr AGING AND SUBSEQUENT TESTING —]

 

80

 

 

 

 

 

f | | \ ! i I |
60 [-#0 SPECIMENS ANNEALED FOR 2 hr AT 2450°F BEFORE # 60
100~ hr AGING AND SUBSEQUENT TESTING

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

_ o
! Q
. 0
—~ 50 < ] 50 =
& ] | | E
< - ) | 2
— T
z | | A =
2 a0 — YIELD STRENGTH — 1= “I f— 40 ©
E | T / i
o r" } o
Z S / .
S ! ; s 0
w 30 - 4 30
1 1| &
T ,/ >

| —— g

20 . : C—t™ — 74 20
§ -
s\\ J_,// ]
1} \..‘\ //,‘
40 ‘ T 3 o S = 1 ! 10

 

 

 

 

 

 

 

 

 

 

 

 

 

 

] DUCTILITY 4{
- | ‘. _
. BEREEENEEE .

1000 1400 4200 1300 1400 {1500 4800 4700
TEMPERATURE (°F)

 

Fig. 7.20. Strength and Elongation of Hastelloy B
as a Function of Testing Temperature After Aging
for 100 hr at the Test Temperature. All physical
testing performed at the aging temperature,

Long-time aging of hot tensile test specimens
is now in progress to permit a more complete
analysis of the precipitation effects which would
occur in Hastelloy B during service at these
elevated temperatures. An apparatus for the hot
bend testing of wrought and weld metal specimens
is also under construction. The determination of
aging rates by hardness measurements has also
been proposed as a means of supplementing the
data accumulated by other methods.

STRESS-RUPTURE CESIGN CURVES
FOR INCONEL

R. B. Oliver D. A. Douglas
J. H. DeVan J. W, Woods
Metallurgy Division

The program to obtain design data for lnconel
at 1300 and at 1500°F under reactor conditions is
and some data at 1650°F are
A series of tests at an extremely low
range of stresses has begun that will provide data
on rupture times in the 5,000 to 10,000-hr range,
which is longer than the present period assumed
for aircraft reactor operation.

nearly complete,
available,

The low stresses

120

involved in these tests are, however, representa-
tive of the principal stresses with which the
reactor design engineers are concerned. The
to date, in fused
salts and, for comparison, in argon are presented
as a series of design curves shown in Figs. 7.21
through 7.30. The times to 0.5, 1, 2, 5, and 10%

elongations and to rupture are plotted against
stress for both annealed and as-received sheet

results obtained for Inconel,

specimens.

In the case of Inconel specimens annealed at
2050°F and tested in argon at 1500°F, it should
be pointed out that considerable scatter exists
in the data as a result of a precipitation hardening
phenomenon which sometimes occurs. The nature
of the precipitate and the causes of its occurrence
are presently uncertain; since it may or may not
appear in entirely similar material under identical
stress and temperature conditions. The precipi-
tation tokes place at from 300 to 4000 hr and
significantly
strength of the alloy.
earlier

the creep and rupture
The higher the stress, the
in the test the precipitation effect is
observed. Because of the unpredictable behavior
of this precipitate, the design curves which are
presented represent those tests in which no pre-
cipitation hardening was observed.

improves

DEVELOPMENT OF BRAZIMG ALLOYS

P. Patriarca K. W, Reber
R. E. Clausing G. M. Slaughter
Metallurgy Division

J. M, Cisar

Aircraft Reactor Engineering Division

R. L. Heestand
Pratt & Whitney Aircraft

The resistance of brazing alloys to high-temper-
ature oxidation is a primary factor to consider in
the choice of these materials for use in the fabri-
cation of liquid-metal-to-air radiators.
ation program is therefore being conducted to
determine the suitability of 28 potential high-
temperature alloys for this application. Inconel
T-joints were prepared with these materials and
small smaples of these joints were subjected to
static air at 1500 and at 1700°F for various times.
The results of these tests, as determined by me-
tallographic examination in the as-polished con-
dition, are presented in Table 7.20. Some of the
information found in this table has appeared in a

An evalu-
 

 

 

 

PERIOD ENDING MARCH 10, 1955
TABLE 7,20, OXIDATION RESISTANCE OF DRY-HYDROGEN-BRAZED INCONEL T-JOINTS
Brazing Oxidation in Static Air*
Brazing Alloy Composition (wt %) Temperature At 1500°F At 1700°F
°F) For 200 hr For 500 hr  For 1300 hr  For 200 hr For 500 hr
Commercial Alloys
Nicrobraz 70 Ni~14 Cr—6 Fe—5 B~4 Si—1C 2150 Slight Slight Slight Slight Slight
Low-melting Nicrobraz 80 Ni-5 Cr—6 Fe=3 B-5 Si-1 C 1950 Slight Slight Slight Slight Slight
Coast Metals No. 50 93 Ni—-3.5 §i—-2.5 B-1 Fe 2050 Slight Slight Slight Slight Slight
51 92 Ni-4.5 5i-3 B-0.5 Fe 2050 Slight Slight Slight Slight Slight
52 89 Ni-5 5i~4 B~2 Fe 1840 Slight Slight Slight Slight Slight
53 81 Ni-4 Si—-4 B-8B Cr-3 Fe 1950 Slight Slight Slight Slight Slight
NP 50 Ni-12 $i—28 Fe—4 Mo—-4.5 2050 Slight Slight Slight Slight Moderate
P-1 Mn-0.5 Cr
Mond Ni Co. Alloy 64 Ag—33 Pd-3 Mn 2150 Severe Severe Severe Complete
Copper 100 Cu 2050 Complete Complete
Experimental Nickel-Base Alloys
G-E No. 62 69 Ni=20 Cr-11 Si 2150 Slight Slight Slight Slight Moderate
81 66 Ni=19 Cr—10 Si—4 Fe—1 Mn 2150 Slight Slight Slight Slight Moderate
Mi-Cr-5i 73.5 Ni-16.5 Cr—10 Si 2150 Slight Slight Slight Slight Moderate
Ni-Si 88 Ni—12 §i 2200 Slight Slight Slight Slight Slight
Ni-Ge 75 Ni-25 Ge 2150 Slight Slight Slight Moderate  Severe
Ni-Ge-Cr 65 Ni—-25 Ge—~10 Cr 2130 Slight Slight Slight Slight Moderate
Electroless Ni-P 88 Ni-12 P 1740 Slight Slight Slight Above melting point
of alloy
Ni-P-Cr 80 Ni—=10 P-10 Cr 1830 Slight Slight Slight Above sclidus of alloy
Ni-Mo-Ge 50 Ni-=25 Mo~25 Ge 2150 Slight Stight Slight Moderate  Severe
MNi-5n 68 Ni-32 5n 2150 Slight Moderate Severe Severe Compiete
Ni-Mn 40 Ni—-60 Mn 1950 Complete Complete
Ni-Mn-Cr 35 Ni-55 Mn—10 Cr 2050 Severe Severe Complete  Severe Complete
Experimental Precious-Metal Base Alloys
Pd-Ni 60 Pd—40 Ni 2300 Very Slight Moderate Very Slight
slight slight
Pd-Ni-5i 60 Pd—37 Ni-3 Si 2150 Very Slight Moderate Slight Moderate
slight
Pd-Al 92 Pd-8 Al 2020 Very Very Very Very Slight
slight slight slight slight
Pd-Ge 90 Pd—10 Ge 2050 Very Slight Severe Complete
slight
Au-Ni 82 Au-18 Ni 1830 Very Very Slight Moderate  Moderate
slight slight
Au-Co 90 Au—10 Co 1830 Very Very Moderote Slight Severe
slight slight
Au-Cu 80 Au-20 Cu 1740 Moderate  Complete Complete

 

*Very slight, less than 1 mil of penetration; slight, 1 to 2 mils of penetration; moderate, 2 to 5 mils of penetration; severe, greater than §
mils of penetration; complete, fillet completely destroyed.

121
ANP PROJECT PROGRESS REPORT

UNCLASSIFIED
ORNL-LR-DWG 5785

 

20,000

 

 

 

 

10,000

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

5000

 

 

 

 

 

 

 

STRESS (psi)

 

 

 

 

 

 

 

 

2000

 

 

 

 

 

 

1000 - : :
100 200 1000 2000 5000 10,000
TIME Lhr}

Fig. 7.21. Stress-Rupture Characteristics of As-Received Inconel Tested in Argon at 1300 °F.

UNCLASSIFIED
ORNL—1.R-DWG 5786

20,000

\

10,000

5000

STRESS (psi)

2000

1000
1 2 5 10 20 50 100 200 500 1000 2000 5000 10,000
TIME. (k)

Fig. 7.22. Stress-Rupture Characteristics of As-Received Inconel Tested in Fused Salt at 1300°F,

122
PERIOD ENDING MARCH 10, 1955

UNCLASSIFIED
ORNL-LR-DWG 5787

 

 

30,000

20,000

RUPTURE
10
,
10,000

5000

STRESS {psi)

2000

1000 2000 5000 10,000

5 10 20 50 100 zZ00 500
TIME {hr)

1000 .

Fig. 7.23. Stress-Rupture Characteristics of Annealed Inconel Tested in Argon at 1300°F.

UNCLASSIFIED
CRNL-LR—-DWG 5788

20,000

 

10,000
'E
£ 5000
o
wn
L
o
f
2000
1000
\ 2 5 10 20 50 100 200 500 1000 2000 5000 10,000
TIME (hr)

Fig. 7.24. Stress-Rupture Characteristics of As-Received Inconel Tested in Argon at 1500°F.

123
ANP PROJECT PROGRESS REPORT

UNCLASSIFIED
ORNL-LR-OWG 6789
T

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

|o,ooor T
5000 |-
_ |
“w
e
w
w 3
ul
o
'_
n
2000 Lo .
N \
il
1000 i _ |
f 2 5 10 20 100 200 500 100G 2000 5000 10,000

   

TIME (hr)

Fig. 7.25. Stress-Rupture Characteristics of As-Received Inconel Tested in Fused Salt at 1500°F.

10,000

5000 |

STRESS {psi)

 

UNCLASSIFIED
ORNL-LR-DWG 5790

UPTURE

100 200 500 1000 2000 5000
TIME (hr)

10,000

Fig. 7.26. Stress-Rupture Characteristics of Annealed Inconel Tested in Argon at 1500 °F.

previous quarterly repmt,5 but it is presented
again in a more complete form to permit a compre-
hensive and detailed evaluation.

The void formation along the braze metal-Inconel
interface, which was noted on the samples brazed
with G-E No. 62, G-E No. 81, and the 73.5%
Ni-16.5% Cr-10.0% Si alloys, and which was

previously thought to result from internal oxi-

 

5P. Patriarca et al.,, ANP Quar. Prog. Rep. June 10,
1954, ORNL-1729, p 94.

124

dation, is now believed to be associated with a
diffusion phenomenon. Check samples, which
were fested in vacuum under identical conditions
of time and temperature, contained voids at similar
locations along the interface. Although the nature
of the void formation is not yet completely under-
stood, it has been noted that the quantity of one
constituent appears to increase with increasing
time at temperature and with increasing temper-
ature of test. The identification of this con-
stituent, which is probably a complex intermetallic
10,000

5000 [~ - -

STRESS (psi)

2000

1000

 

PERIOD ENDING MARCH 10, 1955

UNCLASSIFIED
ORNL-LR-DWG 5794

! i

100 200 500 1000 2000 5000 10,000
TIME {hr}

Fig. 7.27. Stress-Rupture Characteristics of Annealed Inconel Tested in Fused Salt at 1500°F.

10,000

5000

STRESS {psi)

2000

|
|
2 5 10 [4e] 50

10001

UNCLASSIFIED
ORNL-LR--DWG 5792

  

100 200 500 000 2000 53000 10,000
TIME (hr}

Fig. 7.28. Stress-Rupture Characteristics of As-Received Inconel Tested in Argon at 1650°F.

compound, is to be attempted by x-ray techniques,
since knowledge of its composition is required for
the determination of the diffusion mechanism.

This investigation of brazing alloys has shown
that a majority of the nickel and nickel-chromium
base alloys are suitable for service in an oxi~
dizing atmosphere at 1500°F, and several are
suitable ot 1700°F. Oxidized Inconel joints
brazed with two of the alloys of prime interest for
liquid-metal-to-air radiators are shown inFigs.7.31

and 7.32. The excellent resistance to attack of
alloy No. 81 after 500 hr at 1500°F in static air
is illustrated in Fig. 7.31. This alloy has been
used for the fabrication of units containing austen-
itic stainless steel or Inconel fin materials, while
Coast Metals alloy No. 52, shown in Fig. 7.32,
is of interest in the fabrication of high-conduc-
tivity-fin assemblies. Only minor oxidation is
evident on the joint brazed with Coast Metals
alloy No. 52 after being subjected to the oxidizing

125
ANP PROJECT PROGRESS REPORT

10,000

5000

0.5% 1% 5%

STRESS (psi)

2000

1000

ra
o
)

20 50

 

UNCLASSIFIED
ORNL-LR-DWG 5793

RUPTURE

100 200 500 1000 2000 5000 10,000
TIME {hr)

Fig. 7.29. Stress-Rupture Characteristics of As<Received Inconel Tested in Fused Salts at 1650°F.

UNCLASSIFIED
ORNL-LR-DOWG 57949

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

10,000 1 _
.. . i . —
5000 ~4]7j’ B
- | f
g ¢ ——‘ ]
5 ' RUPTURE . \ Lo '
o - T — ——— - ol _:'*AH
£ j \ \ |
2000
| ar
. |
1000 \R || L

  
    

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

1000 2000 5000 10,000

TIME {hr)

Fig. 7.30. Stress-Rupture Characteristics of Annealed Inconel Tested in Fused Salts at 1650 °F.

atmosphere for the same time and temperature as
the joint brazed with G-E alloy No. 81, The
resistance of the precious-metal-base alloys to
attack is also very good, but their application
to liquid-metal-to-air radiator fabrication is not
considered to be promising. Their poor compati-
bility with liquid metal environments makes their
use extremely risky, since severe tube-wall di-
lution during brazing or solid-state diffusion in
service may produce a high concentration of the
precious metal near the circulating liquid.

126

Cyclic oxidation tests are now being conducted
to determine the effect of thermal fluctuations on
the adherence of protective oxide films. Pre-
liminary results indicate that the extent of attack
on most of the alloys does not increase appreci-
ably in the cyclic tests in comparison with the
static tests, but a more complete analysis is
required before definite can be
reached.

Corrosion tests have
alloys of the

conclusions

indicated that brazing
nickel-chromium-germanium and
 

PERIOD ENDING MARCH 10, 1955

 

 

 

 

UNCLASSIFIED
Y-13320
: 0.0t
I
O
=
0.02
0.03
i fl"i'; é' 3 :? “ T
j( W A ,,>) V(’ ‘:W X
o S X
S e

 

 

Fig. 7.31. Inconel T-Joint Brazed with G-E Brazing Alloy No. 81 and Tested in Static Air for 500 hr
at 1500 °F, Note negligible attack. Etched with electrolytic oxalic acid. 100X,

nickel-chromium-germanium-silicon types may be
of interest in fluoride-to~sodium heat exchangers.
In order to conduct further research on these alloys
and to provide enough material for tests on heat
exchanger components, l-lb quantities of each
alloy and of the 75% Ni-25% Ge binary alloy were
arc-melted. Several techniques have been investi-
gated for reducing these arc-melted ingots to the
more desirable powder form, but no suitable labo-
ratory method has been obtained. Grinding with a
mortar and pestle or with a ball mill has not been
satisfactory because the alloy possesses suf-
ficient inherent ductility to prevent easy powdering.
Experiments in which the alloy was induction
melted in a quartz tube and the molten metal then
permitted to drop from a small hole into water
have not been promising. However, 50-g samples
have been submitted to a commercial manufacturer
of brazing alloy powders for experimentation in
Other possibilities for
obtaining powder lie in the fabrication and instal-

powder-making equipment.

lation of a small atomizing machine or the use
of larger atomizers available in industry.

Methods for the production of brazing alloy
powders from of elemental
powders are now being developed, and preliminary
have indicated that this technique
may be very promising. The elemental powders
are carefully mixed and sintered in dry hydrogen
for several hours at a temperature very near that
required to fuse the lowest melting point eutectic
in the system. This sintered material may then be
crushed easily to provide a fine powder suitable
for preplacement on heat exchanger or radiator
components. Experiments for determining the
optimum time and temperature required to obtain
sufficient diffusion without seriously impairing
the friability of the compact are being conducted,
and the possibilities of applying this presintering
method to other alloy systems of interest are being
studied.

sintered mixtures

experiments

127
ANP PROJECT PROGRESS REPORT

 

| UNCLASSIF 1ED .

 

INCH

0.02

 

0.03

 

 

 

Fig. 7.32. lInconel T-Joint Brazed with Coast Metals Brazing Alloy No. 52 and Tested in Static Air for
500 hr at 1500°F. Only very slight oxidation can be seen. Etched with electrolytic oxalicacid. 100X,

FABRICATION OF TEST COMPONENTS

P. Patriarca K. W. Reber
R. E. Clausing G. M. Slaughter
Metallurgy Division

J. M. Cisar

Aircraft Reactor Engineering Division

R. L. Heestand
Pratt & Whitney Aircraft

High-Conductivity-Fin Radiator

The fabrication of a sodium-to-air radiator with
6 in. of type 430 stainless-steel-clad copper high-
conductivity fins was described in a previous
report.® The tube-to-fin joints were brazed with
Coast Metals alloy No. 52 and the tube-to-header
welds were made by the semiautomatic heliarc
welding technique. The welded tube-to-header
joints were back brazed as a precaution against
the formation of leaks during service.

During the fabrication of this radiator it was

 

b, Patriarca ez al., ANP Quar. Prog. Rep. Sept. 10,
1954, ORNL-1771, p 120.

128

found that it was necessary to maintain critical
control over the quantity of brazing alloy placed
on each tube-to-fin joint. A sufficient amount of
alloy was required for the oxidation protection of
the exposed copper at the punched holes of the
high-conductivity fins. The presence of braze
metal in excess of that required for the protection
of the copper might result in “‘puddling,”” that is,
concentration of the excess on the bottom fins,
which would be undesirable because the air pas-
sages between the fins might be sealed and lo-
calized solution of the base metal might occur.

The use of extruded brazing alloy wires as a
means of obtaining controlled quantities has been
investigated extensively, but the lack of a suitable
binder makes their use somewhat unatiractive at
the present time. An acrylic binder material
produces wires which are relatively easy to handle
when freshly extruded, but they become embrittled
upon aging for a few hours at room temperature.
Thus the assembly of large complicated radiators
is seriously hampered by crumbling and subse-
quent movement of the brazing alloy from the
desired tube-to-fin location. Another binder ma-
terial, Castolite, produces wires that are extremely
ductile and weak at room temperature and therefore
require meticulous care during handling.

A dry-powder method of brazing alloy preplace-
ment has been developed which provides a prom-
ising means for obtaining controlled amounts of
alloy on each tube-to-fin joint. A stainless steel
template containing holes that have been precision
drilled is placed over a sheared fin so that each
hole in the template is centered over a punched
hole in the fins. Since the drilled hole is larger
than the punched hole, the template fits securely
against the flat portions of the fin. The dry
powder is then applied, and the excess powder is
After careful removal of the template,
the powder is secured to the fin with a methyl-
acrylate cement and allowed to dry. A 36-hole
fin with brazing alloy preplaced by this technique
is shown in Fig. 7.33.

The oxidation protection of the exposed copper
on the sheared edges of the fin is also required in
order to minimize oxidation and thus overcome the
severe fin distortion that would result from the
volume changes that would occur during formation
of the oxide. An aluminizing process has been
developed in which edge protection is obtained by
the formation of a highly oxidation-resistant
copper-aluminum alloy. A procedure has also been

removed,

PERIOD ENDING MARCH 10, 1955

devised to permit the aluminizing of large numbers
of type 310 stainless-steel-clad copper sheet fins.
This procedure, which requires that all fins be
precision sheared to an exact size, is described
in the following:
chlorethylene or another suitable organic solvent,
the fins are stacked in groups of approximately
200 and securely clamped together with ]/4-in.-
thick stainless steel end plates. The exposed
copper edges are then sprayed with three coats
of Kestron acrylic spray to seal the cracks be-
tween fins and prevent the flow of the aluminum-
bearing slurry onto the stainless steel cladding.
A coat of slurry, consisting of 100 cm?® of acrylic
resin to 40 g of atomized aluminum powder (~325
mesh), is then applied evenly to the fin edges,
After sufficient drying the clamped fins are heated
in helium at 750°F for 2‘/2 hr to accelerate the
formation of the oxidation-resistant aluminum
bronze. A high helium flow rate (80 cfh) should
be maintained for the first hour, but it can then
be reduced to 40 cth for the remainder of the heat
treatment.  After cooling, the excess aluminum
can be easily removed from the fin edges with the
aid of a brass wire brush. These fins can then be
punched by using the conventional techniques.
Test specimens of fins with this type of edge
protection have been metallographically examined

after 100, 200, 500, and 1000 hr in air at 1500°F,

After degreasing in poly-

UNCLASSIFIED
Y-14294

 

Fig. 7.33. High-Conductivity Fin for Radiator with Brazing Alloy Preplaced by Dry-Powder Technique.

129
ANP PROJECT PROGRESS REPORT

A protected specimen tested in air for 1000 hr
at 1500°F is compared in Fig. 7.34 with an un-
protected fin tested under the same conditions.
The small voids near the aluminized edge of the
protected fin are atiributed to the migration of
copper into the aluminumerich surface metal.
Cyclic tests are also being conducted on alumi-
nized fins to evaluate their stability under simu-
lated aircraft service conditions., After 200 hr at
1500°F and 70 air-cools to room temperature, no
adverse effects were evident,

In preparation for the subsequent assembly,
welding, and brazing of two 500-kw NaK-to-air
radiators for heat exchange experiments, 2200
stainless-steel-clad copper high-conductivity fins
have been sheared to the desired 2.in. by 8-in.
final dimensions. They were degreased, inspected
for surface imperfections, and aluminum-bronze

 

edge-protected by the procedure described above.
They will be punched and ossembled with the
tubes by utilizing the dry-powder preplacement
technique. The use of a l/lé-in.-thick template
containing holes 0.246 in. in diameter will provide
the minimum quantity of brazing alloy required per
joint to permit good coverage of the exposed
copper on the punched lips. To further reduce the
tendency toward the undesirable accumulation of
excess alloy as a result of normal variations in
the quantity deposited, two thin sheets of Inconel
are to be placed tightly together at 4-in. intervals
along the 12 in. of stacked fins. The capillary
joint between these two Inconel sheets will act
as a sump to remove any excess brazing alloy that
may be present. This technique is applicable to
the fabrication of the 500-kw radiators because
l/Ié-in.-thick Inconel sheets are required at 4-in.
intervals for structural support.

UNCLASSIFIED
Y¥-14334 .

 

 

Fig. 7.34. Type 310 Stainless-Steel-Clad Copper High-Conductivity Fin After Oxidation for 1000 hr at
1500°F. The top fin illustrates the excellent edge protection afforded by the aluminizing treatment. The
bottom fin illustrates the severe attack of the unprotected fin.

130
Difficulties have been encountered in the use
of the 6% aluminum-bronze high-conductivity fin
material when brazing the fins to Inconel tubes.
The aluminum oxide film present on the fins
prevents wetting by the conventional brazing
alloys. An electroplate of nickel on the fin did
not act as an adequate diffusion barrier to permit
wetting by the alloys. A chromium electroplate
served as a diffusion barrier, but chromium is
extremely difficult to wet in the hydrogen dew-
points readily available in large-scale brazing
operations. However, with an electroplate of iron
on the aluminum bronze, tube-to-fin joints brazed
with Coast Metals alloy No. 52 have been obtained
that exhibit excellent wettability, Fig. 7.35.
Joints brazed in this manner have also shown
excellent resistance to oxidation after 100 hr at
1500°F. The iron pickup during brazing does not
appreciably reduce the oxidation resistance of the
nickel-silicon-boron brazing alloy. The excess
iron can be removed from the fins after brazing by

 
    

Fi g. 70350

PERIOD ENDING MARCH 10, 1955

a suitable pickling operation, or it can be re-
moved early in service upon the formation of a
nontenacious iron oxide. In view of these prom-
ising developments, a quantity of commercially
available 5% aluminum bronze has been punched
and is being iron-plated for the subsequent fabri-
cation of a large-scale radiator.

Intermediate Heat Exchanger No. 2

The fabrication of the fluoride-to-sodium inter-
mediate heat exchanger No. 2 requires the heliarc
welding of 400 tube-to-header joints and the back-
brazing of these welds with a suitable corrosion-
resistant alloy, such as Nicrobraz or nickel-
chromium-germanium. The welding of approximately
200 of these joints has been satisfactorily com-
pleted by the semiautomatic rotating-arc method,
but prior to initiation of the actual fabrication, a
set of experiments was conducted to determine the
optimum combination of welding conditions.
Several of the sample joints were examined under

 UNCLASSIFIED
U NT14098

 

lron-Plated Aluminum Bronze High-Conductivity Fins After Brazing with Coast Metals
Brazing Alloy No. 52. Good wettability of the fins was obtained. 60X.

131
ANP PROJECT PROGRESS REPORT

high magnification to determine the presence of
weld microfissuring, but this type of cracking was
not found to constitute a problem.

In these experiments, two Inconel test headers,
of the same size and physical shape as those on
the actual unit, were machined, and sample tube-
to-header welds were made under various con-
trolled conditions. It was immediately evident
that the preparation of the header surface after
insertion of the tubes, but prior to welding, should
definitely not be done by abrasive grinding. En-
trapped abrasive in the joint caused severe arc
instability and inconsistent welds. A more reliable
method consists of preshaping the tubes to
conform to the curvature of the header before
assembly, The tube can then be heliarc tack-
welded to the header and expanded with a special
tool before final welding. With the method of
header preparation prescribed, welding variables
were evaluated to ascertain those which gave the
most consistent weld penetration and which were
least likely to result in excessive hole constric-
tion or in undersirable preferential melting of the
tube wall,

Prior experience had shown that an arc distance
of 0.050 in. and a weld time of approximately 6 sec
produced consistently good welds when joining
3/] ¢ in-0D, 0.017-in. wall tubing to relatively
thick headers. These values were therefore used
for the determination of the optimum diameter of
electrode rotation and the proper welding current.
A rotation diameter of between 0.21 and 0.22 in.
was found to be desirable, since diameters of less
than this often resulted in preferential melting of
the tube wall and greater diameters produced the
maximum weld penetration in the header plate
rather than at the joint where it is of most impor-
tance. An arc current of 60 amp at an arc voltage
of 10 produced consistently satisfactory welds with
penetrations of approximately twice the tube wall
thickness, A typical weld produced under these
optimum conditions is shown in Fig. 7.36. No
weld porosity or cracks are evident and excellent
penetration was achieved. The two-pass effect
evident in the nugget occurs because a weld over-
lap of one-fourth revolution is used after the com-
plete peripheral weld has been made. As the weld
overlap is being made the weld current is gradually
decreased to prevent the formation of undersirable
arc craters. The uniformity of welds made under
these conditions can be seen in Fig. 7.37, which

132

shows a 100-weld header section of the partially

completed intermediate heat exchanger No. 2.

The fabrication of approximately 50 ‘‘comb"
spacers for this unit has also been completed.
The necessary jigs were prepared and cone-arc
plug welding conditions were determined for the
heliarc welding of the 0.020-in. by 0.040-in. wire
spacers into the 0.010-in. Inconel strip headers.
After these spacers are attached to the assembly
and the tube-to-header joints are back-brazed, the
unit will be ready for insertion in the heavy Inconel
pressure shell,

Radiator for Cornell Aeronautical Laboratory

The fabrication of a full-scale liquid-metal-to-
air radiator designed by the Cornell Aeronautical
Laboratory has been undertaken.  The design
incorporates integral-helical-finned tubing com-
pletely machined from type 316 stainless steel bar
stock. The tubing is machined to 0.200 in. in
outside diameter with a 0.037-in. wall, and the
header plates and other structural components of
the radiator are machined from 1!/1 - and Y-in.
type 316 stainless steel plate. The partially
completed unit is shown in Fig. 7.38.

After assembly of the tube-to-header section of
the radiator, the 312 tube welds were manually
heliarc welded by qualified operators using pre-
scribed procedures. Longitudinal grooves were
machined in the header plate before welding to
simulate trepanning. The grooves in the header
aid in minimizing microfissures because they
substantially reduce the strain restraint near the
weld, A ‘‘skip sequence’’ was also employed
during welding to equalize the heat distribution
in the header plate,

The filler plates that can be seen in Fig. 7.38
were welded to the side plates to occupy most of
the space between these plates and the outer rows
of tubes. To prevent severe distortion when
welding the side plates to the headers, the side
plates were reinforced with 1-in.-thick stainless
steel strong-backs, as shown in Fig. 7.38, and
then welded to the unit, Figure 7.38 shows the
unit after completion of the root passes. The re-
maining heavy welding and back-brazing with an
alloy such as Coast Metals alloy No. 50 should
be completed in the next few weeks.

The choice of an alloy for use in back-brazing
the welded tube-to-header joints in this radiator
depends to a large extent upon the ability of the
 

PERIOD ENDING MARCH 10, 1955

 

 

 

 

 

UNCLASSIFIED
Y-14185
Q.01
X
O
2
0.02
0.03
Q.04
0.05

 

 

 

Fig. 7.36. Inconel Tube-to-Header Weld Prepared Under Optimum Welding Conditions. Excellent pene-
tration and weld quality obtained. Etched with electrolytic oxalic acid, 75X.

UNCLASSIFIED
¥-14355

 

A 100.Weld Header Section of the

Fig. 7.37.
Intermediate Heat Exchanger No. 2 Welded with
the Semiautomatic Heliarc Welding Equipment,
Note weld uniformity.

alloy to withstand the severe strain imposed upon
cooling as a result of the heavy sections involved.
Although several high-temperature brazing alloys
possess adequate resistance to high-temperature
oxidation, many of them, such as Nicrobraz, G-E
alloy No. 81, and Coast Metals alloy No. 52,
crack upon furnace cooling from the brazing
temperature. However, Coast Metals alloy No. 50
and the nickel-chromium-germanium alloy show no
evidence of cracking under these circumstances.
Since Coast Metals alloy No. 50 can be purchased
in a powder form and it possesses good resistance
to both oxidation and to liquid sodium, it will be
seriously considered for this application. MHow-
ever, the extent and effects of boron diffusion from
the braze into the stainless steel base material
will be determined before it is used. The large
mass of the unit suggests that a 12-hr-heating and
12-hr-cooling cycle should be used during brazing,

133
ANP PROJECT PROGRESS REPORT

UNCLASSIFIED
¥-14282

 
   

Fig. 7.38.

Cotnell Aeronautical Laboratory
Radiator After Welding of the Tube-to-Header
Joints and Deposition of the Heliarc Root Passes
in the Side Plate-to-Header Joints, The heavy
stainless steel strongbacks were used to prevent
severe distortion of the side plates during welding.

and metallographic test specimens have been
prepared to test this cycle under simulated condi-
tions, The extent of diffusion will be determined
in the as-brazed condition and after heating for
25, 50, 100, and 500 hr at 1650°F. Also, nine
tensile-test specimens have been coated with a
slurry and brazed under conditions approximating
those to be used on the actual radiator. These
specimens were submitted to Cornell Aeronautical
Laboratory for longitudinal drilling and testing to
simulate tests on tubing. Two uncoated tubes
wete subjected to the some thermal cycle as the
brazed samples to serve as controls. If the re-
sults of these metallographic and physical tests
are promising, Coast Metals alloy No. 50 will be
used for the brazing operation. Otherwise, an

134

alloy of the nickel-chromium-germanium type will
probably be used.

Sodium-Betryllium-Inconel Compatibility
Testing Apparatus

A second sodium-beryllium-Inconel compatibility
testing apparatus was fabricated; the components
are shown in Fig. 7.39 in an exploded view and
in Fig. 7.40 after completion, The unit comprises
a cylindrical beryllium insert inside an lnconel
housing. The Inconel housing, the sodium pots,
and the thermocouple housings were manually
heliarc welded into position, and nine thermocouple
assemblies were brazed with G-E alloy No. 62.
All welded and brazed joints were leak-tight, as
determined by testing with a helium leak-detecting
apparatus.

Heot Exchanger Brazes

It was reported previously? that the 82% Au-18%
Ni brazing alloy should be considered for use in
the fabrication of fluoride-to-air heat exchangers,
and it has now been established that the solid-
state migration of gold from this alloy into Inconel
tubing is minor after an extended peried at an
elevated temperature. Thus corrosion tests of the
tube interior after service would not be obscured
by the presence of localized gold-rich areas.
Recent tests with copper as a brazing alloy for
fluoride-to-helium heat exchangers have shown
similar promising results. A typical joint was
held at 1500°F for 900 hr and then examined for
copper diffusion, Microspark spectrographic
techniques coupled with a metallographic exami-
nation revealed that a maximum of 0.003 in. of
copper diffusion had occurred. |

Dynamic corrosion loops are also tobe fabricated
for studying the corrosion of joints brazed with
several alloys and tested in intimate contact with
circulating fluoride fuel, These loops will in-
corporate several sleeve-type sections in the hot
leg to assist in the evaluation of brazing alloys
for in-pile loop applications.

Cermet-to-Metal Brazing

The joining of cermets to Inconel will be required
if cermets are to be used for applications such as
valve seats, and, in evaluation studies of cermets,

 

7P. Patriarca et al., ANP Quar. Prog. Rep. Dec. 10,
1954, ORNL-1816, p 109.
 

Fi Ge 7-39¢

PERIOD ENDING MARCH 10, 1955

UNCLASSIFIED
- Y-14142

Unassembled Sodium-Beryllium-Inconel Compatibility Testing Apparatus No. 2 Showing

Beryllium insert, Incone! Housing, Sodium Pots, and Thermocouple Assemblies.

joins between these two materials are now re-
quired for the preparation of samples for cermet
self-welding tests. A study of cermet-to-Inconel
brazing is therefore being conducted in an attempt
to develop suitable techniques for the production
of these joins.

Preliminary tests have consisted of determining
the wettability of certain cermets with fluoride-
resistant alloys such as copper, gold-nickel, and
nickel-phosphorus, Although the nickel binder is
easily bonded, it appears that wetting of the non-
metallic portion of the cermet may be much more
difficult. Also of prime importance in the success-
ful production of these joins will be the solution
of the problem posed by the unequal thermal ex-
pansions of the two materials, The cermets of
immediate interest have thermal expansion coeffi-
cients ranging from 4.0 x 10=% to 5,7 x 10-6
in./in..°F as compared with 10,2 x 10~¢ in./in..°F

for Inconel in the range 1000 to 1400°F. The
best solution to this problem seems to lie in the
use of a combination of a ductile brazing alloy
and a ductile pad of metal between the cermet and
the Inconel.

A technique now being investigated consists of
the deposition on the cermet of a 0.0001-in. layer
of ‘'electroless’’ nickel-phosphorys, which wets
both the metallic and nonmetailic portions of the
cermet, When this layer is covered with a 0.005-
in-thick acid copper plate and then joined to
Inconel by heating to above the melting point of
both the nickel-phosphorus layer and the copper,
a join that possesses fair mechanical properties
is produced. The phosphorus apparently aids in
wetting the cermet and then diffuses into the sur-
rounding metal sufficiently topermit some ductility.
Physical tests on this type of join will be con-
ducted and screening tests on other alloys will be

135
ANP PROJECT PROGRESS REPORT

UNCLASSIFIED
Y-14179

 

Fig. 7.40, Assembled Sodium-Beryllium-Inconel
Compatibility Testing Apparatus No. 2,

performed in an attempt to obtain more satisfactory
brazing material for this opplication. Methods for
bonding ceramics, such as the manganese-molyb-
denum and titanium hydride techniques, will also
be investigated.

SPECIAL MATERIALS FABRICATION

H. Inocuye J. H. Coobs
Metallurgy Division

M. R. D'Amore
Pratt & Whitney Aircraft

Duplex Tubing

The study of flow patterns in impact-extruded
duplex and three-ply tubing was continued, Four
extrusions of composite stainless steel—carbon
steel tubing have been completed. An examination
of the layers of the extruded tubes revealed that
fairly close control of the final layer thickness
was obtained by adjusting the layer thickness of

the extrusion billet, The extruded tubes were

136

subject to poor material recovery because of metal
loss at both ends of the tube. An attempt to im-
prove the material recovery by varying the ex-
trusion billet design is planned,

G-E Control Rods
The 34 control rods ordered for the GE-ANP

Project have been completed and delivered for
reactivity testing. They were prepared by filling
tubes with a mixture of 50% aluminum and 50%
boron carbide by tamping with a pneumatic hammer.
The tubes were then evacuated while being heated
at 300°C to remove the paraffin binder and cold
swaged to the final diameter.

The tubes, as loaded, had a boron density of
about 0.82 g/cm3 that was increased during the
final swaging operating to more than 0.85 g/cm?,
which was considerably more than the 0.70 g/cm?
minimum specified,

Tubular Fuel Elements

Twelve more plate-type fuel elements were
prepared by using either elemental or prealloyed
type 304 stainless steel in combination with
either high-fired or steam oxidized UO, in the
cores. These elements have cores 2 in. wide and
5 in. long, and they will be formed into single,
seamed tubes for hot and cold drawing experi-
ments. Several more tubular assemblies are being
prepared by using seamless materials for reduction
by hot swaging or hot rod rolling techniques.

Boron Shield for ART

Samples of an 85% B C-15% SiC composition
with a boron density of 1,26 g/cm3 were received
from the Carborundum Company, This composition
is reputed to be easy to fabricate, with very close
dimensional control, by cold pressing and firing.
Nicholson of the Carborundum Company stated that
the boron content could be increased somewhat
by the addition of boron or boron nitride to the
composition.

Metal lographic examination of samples from a
recent test designed for comparing the reactions
of boron and of boron carbide with Inconel has
been completed, In the test, three samples of
boron-containing materials were in contact with
Inconel in a helium atmosphere for 100 hr at
1500°F.  The boron-containing materials were
commercial, hot-pressed B ,C (Norton Company’s
““Norbide’’), hot-pressed boron, and cold-pressed
Total depth of penetra-
tion in Inconel of the reaction with these materials
was 0,004, 0.006, and 0.010 in., respectively.

The diffusion of boron from these materials into
Inconel seems to proceed along grain bound-
aries, and @ second phase which is boron-rich
is gradually formed. In the case of severe reac-
tion, as from the amorphous boron powder, the
second phase forms a continuous layer at the

amorphous boron powder.

interface; the layer separates slightly from the
inconel during cooling. Thermal cycling of a
system in which considerable reactionhad occurred
could thus result in spalling of the layer and an
increase in the rate of attack.

Another compatibility test has been completed
that was designed to compare the reactions of hot-
pressed B,C and the B ,C-SiC composition with
Inconel. In this test the samples were exposed in
helium at 1600°F for 100 hr. In an attempt to in-
hibit the reaction, one surface of each of two
pieces of the B C-SiC was sprayed with a sus-
pension of Norton alumina (Grade 38-900) in
acetone before assembly for testing., Visual exami-
nation showed some reaction, even on those sur-
faces protected with alumina. Samples are now
being prepared for metallographic examination.

Al-UO, Fuel Plates for Shielding Experiment

A series of fuel elements was prepared with both
525 and 24S aluminum alioys for cladding in an
attempt to make aluminum-clad sandwiches con-
taining UO, mixed with aluminum that will be
strong enough to withstand the high stresses
expected in the proposed shielding experiments
that mock up the reflector-moderated reactor (cf.,
sec. 13, ““Lid Tank Shielding Facility’). In the
test the plates will be cycled at speeds ranging
up to 20 fps on a link chain running on 3.5-in.-dia
sprocket wheels. No difficulties were encountered
in fabricating the cores for these plates or in
maintaining uniform fuel distribution. The cores
were prepared by mixing 57 wt % steam-oxidized
UO, in -140 mesh atomized aluminum and then
cold pressing at 33 tsi. The sandwiches were
reduced about 92% in 10 roll passes with no
evidence of edge cracking or separation in the
cores.

In preparing the 525 aluminum-clad plates, much
difficulty was encountered in obtaining bonding

PERIOD ENDING MARCH 10, 1955

between the cladding and the frame during hot
rolling. Therefore a second set was prepared with
2S5 aluminum frames and 525 aluminum cladding.
These plates bonded satisfactorily. They were
strain-hardened by cold rolling to the required
thickness and finally bent slightly on a 5]/5-in.
radius to straighten them and add ridigity, How-
ever, attempts to bend them to a smaller radius,
to achieve greater ridigity, resulted in cracking of
the cladding and exposure of the UO -bearing core,
and therefore these plates also were unsatisfactory.

Next, a series of plates was prepared with 245
These plates
were rolled successfully, with good bonding and
fuel distribution, and they could be solution
annealed and bent cold to a cross sectional radius
as small as 1]/2 in. without cracking the cladding.
After age hardening, these plates were tested and
found to be very close to the required strength.

A set of 17 such plates has been prepared for an
endurance test. Each plate contains the required
91 g of UO, with 245 aluminum cladding. They
were bent to a ]%—in. radius and age-hardened to
the T-81 condition. Fabrication of enriched plates
will probably begin as soon as results of the test
are available,

aluminum as the cladding material,

CERAMIC RESEARCH

C. E. Curtis J. A, Griffin
J. R. Johnson
Metallurgy Division

Oxidation Reactions of U02 and of UO2 in BeO

Uranium oxide in the form of a powder or bars
previously sintered at 2900°F in hydrogen gained
up to 2%in weight when fired for 66]/2 hr at 2500°F
in still air. Under the same conditions, up to 14%
was lost from beryllium oxide tubes containing
16.2 wt % of UO, previously sintered at 2900°F.
The accelerated volatilization may have been due
to the formation of a compound between uranium
oxide, beryllium oxide, and possibly water vapor,
The losses in UO, were much lower than those
obtained recently by H, C, Brassfield with similar
tubes heated in moving air. However, all results
point to the necessity of stabilizing UO, against
oxidation under these conditions. For this purpose
either the formation of a stable compound of UO
or use of a protective glaze may be found to be
feasible. Both approaches are being investigated.

137
ANP PROJECT PROGRESS REPORT

Fabrication of Rare Earth Oxide Wafers
for Critical Experiments

Five wafers each of gadolinium and samarium
oxides (?93 in. in diameter and about 10 mils thick)
were supplied for testing at the Critical Experi-
ment Facility (cf., sec. 4, *"Critical Experi-

138

ments’’). About 32 slugs of a mixture of Sm,0,
and Gd,0, are being fabricated for subsequent
canning in an Inconel rod, The slugs will be 0.45
in. in diameter and 0.75 in. in length. These
slugs will also be used at the Critical Experiment
Facility, Some physical property measurements
will be made before delivery.
PERIOD ENDING MARCH 10, 1955

8. HEAT TRANSFER AND PHYSICAL PROPERTIES

H. F. Poppendiek

Reactor Experimental Engineering Division

Additional heat transfer experiments have been
performed with NaF-ZrF ,-UF, flowing in Inconel
tubes. Some quantitative velocity measurements
were obtained for the 18-in. ART core, and the
nature of fluid flow in a simple separation region
was studied. Analytical studies of heat and ve-
locity distribution in the high-temperature-differ-
ential, high-velocity loops were made,

A preliminary value was obtained for the heat
capacity of NaF-ZrF ,-UF, in the liquid state, and
the viscosities of NaF-KF.ZrF,, NaF-BeF,, and
LiF-BeF, were determined, A study was made of
the influence on ART heat transfer of the physical
properties of three types of fluoride fuels,

FUSED SALT HEAT TRANSFER
H. W. Hoffman

Reactor Experimental Engineering Division

Further heat transfer experiments have been
performed with the fuel mixture NaF-ZrF,-UF,
(53.5-40-6.5 mole %) flowing in an Inceonel tube.
The data obtained were for the period of operation
from O to 2 hr and were in agreement with the
previously reported results for 24 to 115 hr of
operation. Failure of the apparatus caused termi-
nation of this sequence of experiments before data
for longer operational times could be obtained.
The results of the heat transfer measurements in
the system NaF-ZrF ,-UF ;-Inconel are summarized
in Fig. 8.1, and a comparison is made with the
general correlation for ordinary fluids, as well as
with the results for the NaF-KF-LiF (11.5-42-46.5
mole %)-Inconel system in which corrosion deposits
were found on the tube walls.

Surface deposits did not occur with NaF-ZrF ,-UF ,
in |lnconel, but vapor blanketing on the inside tube
surface could, perhaps, cause the 24% difference
between the general correlation and the current
data. Therefore, a system is being readied for
studying the pressureedrop characteristics of
NaF-ZrF ,-UF ; in circular tubes with electrical ly
heated wall s,

Photomicrographs (Figs. 8.2 and 8.3) of the two
Inconel tubes used in the NaF-ZrF,-UF, heat
transfer experiments were made by R, Crouse of
the Metallurgy Division. The conditions prevailing

during exposure are presented in Table 8.1, In
tube 1, extensive subsurface void formation oc-
curred to a depth of 5 mils. However, the total
void volume was not large enough to affect the
thermal conductivity of this 5-mil thick region.
This is further indicated by the agreement between
the heat transfer results of tube 1 and tube 2; no

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

B
ORNL—LR—DWG 4
0.010 %0
&
a
= 0005
?"’ " GENERAL CORRELATION, /=0023 Wp02
| 1 ;
T “-” -0.‘._l T
= //’{ 3 :
e o 5 &
o s -~ T
o ko |
= o NoF - ZrF, —Ur, (53.5~40 ~6.5 mole %)
& go0z s \ IN INCONEL | f
m o
2 dinys o
NaF — KF—LiF (11.5-42—46.5mole %) IN INCONEL
L |
Q.00 ! -
2000 5000 10,000 50,000
NR:
Fig. 8.1. Comparison of Current Heat Transfer

Measurements on

NaF.ZrF UF, (53.5-40.6.5
mole %) in Inconel with the General Correlation
for Ordinary Fluids and the Data for NaF-KF.LiF
(11.5-42.46.5 mole %) in lnconel.

 

Fig. 8.2.
NoF-ZrF,-UF, for 115 hr.
1300°F. Average velocity: 10.5 fps.

Inconel Tube Exposed to Flowing
Average temperature:

139
ANP PROJECT PROGRESS REPORT

TABLE 8.1. CONDITIONS PREVAILING DURING HEAT TRANSFER EXPERIMENTS ON NaF-ZrF ,-UF

 

 

Tub Exposure Temperature Reynolds Velocity
uhe (hr) (°F) Number {fp s)
1 24 1200 5,500 to 6,500 8.5 to 10,0
85 1300 ~ 8,500 10.5
6 1400 9,000 to 10,000 10.8
2 3 1420 ~ 8,500 10.5

 

 

Fig. 8.3.
NaF.ZeF,-UF, for 3 hr.
1420°F. Average velocity: 10.5 fps.

Inconel Tube Exposed to Flowing
Average temperature:

apparent void formation occurred in tube 2. Ad-
ditional heat transfer studies with NaF-ZrF -UF,
are planned in a system modified so that the cause
of previous apparatus failures is avoided.

Since the ART fuel-to-NaK heat exchanger has
been designed to operate in the transition flow
regiomy-experiments to define more precisely fused
salt heat transfer in both the transition and the
laminar zones are now under way. Experiments
with NaF-KF-LiF eutectic in stainless steel and
a zirconium-base fuel in Inconel will be conducted.

REACTOR CORE HYDRODYNAMICS

J. O, Bradfute L. D. Palmer
F. E. Lynch

Reactor Experimental Engineering Division
G. L. Muller
Pratt & Whimey Aircraft

Preliminary, quantitative velocity data for a
model of an 18-in. core for a reflector-moderated
reactor were obtained by the photographic tech-

140

nique. This core has straight-through flow and
no entrance vanes. Qualitative velocity infor-
mation was obtained by using the visualization
technique in which a phosphorescent material is
employed.‘ This technique revealed a very large
region of flow separation along the outer core wall
starting at a point close to the entrance and con-
tinving past the midplane, Figure 8.4 shows an
estimate of several velocity profiles in the flow
channel. Additional smaller separation regions
were observed next to the island wall, but they
were much less distinct. The results of these
experiments are described in detail in a separate
repori'.2

The design of a one-quarter scale model of a
proposed 2l-in, core has been undertaken. It is
planned to utilize the flow facilities and the lighting
and photographic equipment assembled for the
smaller core model to describe the velocity pro-
files. The phosphorescentparticle and the photo-
graphic techniques will again be used.

A vaned section to impart a rotational component
to the flow immediately upstream from the core
model has been fabricated (Fig., 8.5). The effect
of rotation on the large separation region will be
studied qualitatively by using the visuvalization
technique.

A variable-angle divergent channel system was
designed, fabricated, and installed in the flow-
visualization system so that the nature of flow
in separation regions could be studied. Thick,
laminar-like flow layers were noted next to the wall
where flow separation occurred under turbulent

 

L. D. Palmer and G. M. Winn, A Feasibility Study
of Flow Visualization Using a Phosphorescent Particle

Method, ORNL CF-54-4-205 (Apr. 30, 1954),

2J. 0. Bradfute, Qualitative Velocity Information
Regarding the ART Core: Status Report No. 4, ORNL
CF-54-12-116 (Dec. 14, 1954).
SECRET
ORNL-LR-DWG 5619

Ut
l VERY LOW VELOCITY,
\ PERHAPS NEGATIVE
\
|

VERY LOW VELOCITY,

PERHAPS NEGATIVE
VAN

A
L———VERY LOW VELCCITY,

/T PERHAPS NEGATIVE

N1y

SEPARATION REGION,
LARGE-SCALE TURBULENCE

 

 

 

 

LIMIT OF SEPARATION REGION
(OBSCURE DUE TO INTENSE TURBULENCE)

| v/

 

 

Fig. 8.4. Qualitative Velocity Profiles in a
Model of an 18-in. Core of a Reflector-Moderated
Reactor for a Reynolds Number of 3000.

 

PERIOD ENDING MARCH 10, 1955

(high Reynolds number) conditions. Also, regions
of partial flow stagnation were observed next to
this laminar-like layer, Experiments dre now in
progress in which these separation regions are
being wiped out by the insertion of very fine screens
| ocated upstream.

ELECTRICAL HEATING AND FLOW
IN TUBE BENDS

H. W. Hoffman L. D. Palmer
N. D. Greene

Reactor E xperimental Engineering Division

It has been observed that peripheral corrosion
on the inside surface of the 180-deg bends of a
high-temperature-differential, high-velocity loop is
not uniform. The preliminary resuvlts of the experi-
ments (cf., sec. 6, ‘‘Corrosion Research’’) indi-
cated the attack on the short-radius side of the
bend to be 2 to 3 times that on the long-radius
side. If it is presumed that the corrosive attack
is related to the temperature of the surface and
that, probably, it is greater at a high surface
temperature, two possible mechanisms by which the
short-radius side temperotures could be greater
than the long-radius side temperatures can be
hypothesized: (1) nonuniform heat generation in
the wall and (2} nonuniform fluid velocity distri-
butions.

UNGLASSIFIED
PHOTO - 23481

Fig. 8.5. Vaned Entrance Section for Obtaining Flow with a Rotational Component.

141
ANP PROJECT PROGRESS REPORT

It was shown that for a tube bend with an elec-
trical current passing through the tube walls, the
heat generation in the wall on the shorteradius
side of the bend is greater than that on the long-
radius side. For the specific dimensions of the
bends in the loops used in these experiments
(tube OD, 0.5 in.; wall thickness, 0.045 in.; radius
of curvature, 3.5 in.), the short-radius side heat
generation was 30% greater than that of the long-
radius side. Therefore the surface temperature on
the short-radius side was higher than that on the
long-radius side. Experimental verification of
this conclusion was obtained by instrumenting a
typical bend with thermocouples and observing the
temperatures as the power generation was varied
while the outer surface was being cooled by uniform
natural convection and there was no flow on the
inside of the tube.

Velocity profile observations were made of the
flow through a 180-deg glass bend having the same
ratio of radius of curvature to tube radius as that
in the experimental loop. The phosphorescent
flow visualization method was used for these
observations.! |t was found that the maximum
fluid velocity occurred close to the long radius
of the bend and that there was a large difference
in the amount of wall cooling between the short-
and long-radius sides because of the large differ-
ences in fluid velocity, Thus the difference
between the wall temperature on the short-radius
side of the tube and that on the long-radius side
was further increased by the difference in wall
cooling. Indeed, it is probable that the effect of
nonuniform fluid flow distribution is more important
in yielding high wall temperatures than is the
effect of the nonuniform heat flux in the tube wall.

Estimotes of the combination of the heating by
the two mechanisms have yielded differences in
temperature between the short- and long-radius
sides of the order of 100°F. Some measurements
have been made that tend to confirm this rough
estimate (cf.,, sec. 3, ‘‘Experimental Reactor
Engineering’’).

ART FUEL-TO«NaK HEAT EXCHANGER
J. L. Wantland H. W. Hoffman

Reactor Experimental Engineering Division

In the current design of the ART, the flow on the
fuel side of the heat exchanger is in the middle
of the transition flow region (Reynolds number

ranges from 2000 to about 5000). The fuel will

142

flow parallel to a bundle of tubes through which
NaK will be flowing, and, for a zirconium-base
fuel in this system, the Nusselt number may vory
from about 4.4 to 35 through this narrow Reynolds
number range. Thus, a study has been initiated
to determine the heat transfer and friction charac-
teristics of the heat exchanger in the transition
flow region,

An experiment has been designed which utilizes
a full-scale heat exchanger tube bundle containing
100 tubes. Water is to be used as the heat trans.
fer fluid at a temperature level that will give
Prandtl numbers and kinematic viscosities similar
to those of the fuel. The heat transfer charac-
teristics will be determined by two different
methods. First, measurements will be made with
the system operating as a water-to-water heat
exchanger with high fluid flow rates through the
tubes to yield low and calculable thermal re-
sistances in the flowing water inside the tubes,
In the second experiment, the tube bundle will be
resistance heated by passing an electric current
through it, and it will be cooled by water flow
between the tubes. Pressure-drop measurements
will also be made for the system. These experi-
mental studies will yield the variations of Nusselt
number and the friction characteristics throughout
the transition flow region.

REACTOR CORE HEAT TRANSFER
N. D, Greene J. A. Russell

Reactor Experimental Engineering Division

A study of the feasibility and desired objectives
of o volume heat source experiment for investi-
gating uniform volumetric heat generation within
divergent and convergent channels has been com-
pleted. The design of several components for this
experiment, for example, the heat exchanger and
the test channel, has been established, and con-
struction of the divergent-flow test channel .is
nearing completion. Construction of the associated
heat exchanger is awaitingreceipt of the necessary
metal.

The 300-kw, 460-v power supply, which is
required both for the volume heat source experi-
ment and the ART heat exchanger experiment, has
been ordered, and all necessary power instru-
mentation has been designed to adapt this power
source to both experiments. Means for recording
rapid temperature changes in the fluid, which will
be at an electrical potential of 440-v (alternating
current), are being studied. A saturable reactor is
being tested to determine its ability to control
large amounts of alternating current. The charac-
teristics of the reactor will be evaluated as soon
as a source of direct current (for control purposes)
is available,?

TRANSIENT BOILING STUDIES
M. W. Rosenthal

Reactor Experimental Engineering Division

The phase of the transient boiling program which
involved a series of delay-time and superheat
measurements was completed. A number of runs
with exponential increases in power were under-
taken, and several experiments were performed
with step increases and with [inear increases in
power. Reduction of the data and analysis of the
results are continuing.

HEAT CAPACITY
W. D. Powers

Reactor Experimental Engineering Division

Two copper calorimeters have been installed and
are now in operation. Because of the much greater
precision that can be obtained, these two calo-
rimeters are expected to furnish enthalpies and
heat copacities at least as fast as did the five
ice calorimeters previously in use. -

Preliminary results for the heat capacity of
liquid NaF-ZrF -UF, (56-39-5 mole %) were ob-
tained. The heat capacity was found to be
0.256 + 0.004 cal/g.°C over the temperature
range of 570 to 890°C.

YISCOSITY
S. I. Cohen

Reactor Experimental Engineering Division

The program of refining viscometry techniques
has been substantially completed. The specific
changes made during the course of this program
were described previously.* Results of the
program suggest that subsequent measurements
will yield viscosities somewhat lower than those
given by early measurements on the same compo-
sitions,

 

3Most of the power supply development and instru-
mentation work is being done by J. A. Russell of the
Instrumentation and Control Group of the Reactor Ex-
perimental Engineering Division,

4s. I. Cohen and T. N, Jones, ANP Quar. Prog. Rep.
Dec. 10, 1954, ORNL-1816, p 117,

PERIOD ENDING MARCH 10, 1955

Measurements have been made with the refined
techniques on three fluoride mixtures. Data were
taken on NaF-KF-Z¢F , (5-52-43 mole %) with both
the Brookfield and capillary viscometers, and the
results yielded by the two devices were in good
agreement (Fig. 8.6); the viscosity varied from
about 7.9 cp at 550°C to about 3.5 cp at 750°C
and may be represented throughout this range by

po= 0.1233 3427/T
where T is in °K,

UNCLASSIFIED
ORNL-LR-DWG 5605

TEMPERATURE (°K)
700 800 200 1000 100 1200

3427/T(°K)

 ——=p=01233 e
© BROOKFIELD VISCOMETER
® CAPILLARY VISCOMETER

 

8
}_
=
oy
)
Q
0
>
i

400 500 800 700 800 900 1000

TEMPERATURE (°C)
Fig. 8.6. Viscosity of NoF.KF.ZrF, (5-52.43

mole %) Obtained by Both the Brookfield and the

Capillary Viscometers.

Measurements were mode on two beryllium-
bearing mixtures in a small, disposable drybox
fabricated expecially for beryllium work from «
20-gal drum. The data were taken with capillary
viscometers, The viscosity of NaF-BeF, (57-43
mole %) varied from about 18,5 cp at 550°C to
about 5.2 cp at 750°C (Fig. 8.7) and may be
represented throughout this range by

g = 0.0308 5240/T |

143
ANP PROJECT PROGRESS REPORT

UNCLASSIFIED
ORNL~LR—DWG 5606

TEMPERATURE {°K)

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

700 800 200 1000 1100 1200 1300
] . T T T TTTTITTIT T
T T T T
- |
_ 1ol ~ 1
NeF —Bef,(57—43 mole %) 4\
N 2=00308 £ 5240/T (oK)~ : \
o
=10 —— l \g —L '
roF NN H
I | - N
g [ [ —1 NN Bl
= 5 A ‘ | . _
< | i 1
i L|F BeF2(69 31 mole "Fo)
—‘ p= —0.148¢3650/T(°K) ¥
5 , \ L
0 CAP\LLARY VISCOMETER NO 27 \
® CAPILLARY VISCOMETER NOQ. 28
\ L] N
300 400 500 600 TOO 800 90Q 1000

TEMPERATURE (°C)

Fig. 8.7. Viscosities of NciF-BeF;2 (57-43
mole %) and LiF.BeF, (69-31 mole %).

where T is in °K. The viscosity of LiF-BeF,
(69-31 mole %) varied from about 10 cp at 550°C
to about 3.5 ¢p at 800°C (Fig. 8.7) and may be
represented throughout this range by

= 0.118 6’3650/T

where T is in °K. It was expected that the vis-
cosity of the LiF-BeF, mixture would be lower
than that of the NaF.BeF, mixture because the
former has a lower density. However, the vis-
cosities of both are higher than would be predicted
from the general trend for other fluorides, which
indicates a proportionality between the viscosities
and densities. The beryllium-bearing mixtures
are apparently glass-like in character.

THERMAL CONDUCTIVITY

W. D. Powers
Reactor Experimental Engineering Division

A radial thermal conductivity apparatus has been
fobricated, and a drybox has been designed and
fabricated to be used with this apparatus. The
sample to be studied in this apparatus will be
contained between concentric cylinders. A known
amount of heat will be generated in the inner
cylinder, and the temperature difference across
the sample will be measured. The thermal conduc-
tivity may then be calculated from these measure-

144

ments and the physical dimensions of the system.

Measurements were made on water for a number
of different heat flows. For low heat flows the
calculated thermal conductivity is constant, but
when the heat flow reaches a critical value the
conductivity increases with increasing heat flows.
Above the critical value, heat is transferred not
only by conduction, but also by free convection.
A plot of the observed conductivity of water vs the
heat flow measured in watts per inch of sample
is presented in Fig. 8.8; the results are in good
agreement with the known values. A radial thermal
conductivity apparatus is being designed for de-

termining the thermal conductivity of lithium

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

hydride.
UNCLASSIFIED
QRNL-LR-OWG 5807
0.6 l : ‘ T ‘ ’ ‘j
R R R N
~ ’ o *
'
c.
= . .
L.
@ N
r S LITERATURE VALUE | i
E e —— l_—f,‘_—,‘
5 \ \ ‘
>
a ‘ ‘ _
Z | |
8 . ‘ : ‘ !
Q2 ——— — - —_— . - _
— '
2 ‘
s
> |
< | ® EXPERIMENTAL DETERMINATION
ot F— ‘ - '
O S R R ——
0 2 a 6 8 10 12
HEAT FLOW (w/in.)
Fig. 8.8. Thermal Conductivities of Water

Measured by the Radial Device and Compared
with the Literature Yalue.

ELECTRICAL CONDUCTIVITY
N. D. Greene

Reactor Experimental Engineering Division

The platinum electrical conductivity cell has
been completed and standardized at room temper-
ature. An attempt to standardize the cell at high
temperatures was forestalled by electrode ex-
pansion and binding of adjacent components. The
machine work necessary to alleviate this problem
is now being undertaken.
INFLUENCES OF THE PHYSICAL PROPERTIES
ON REACTOR HEAT TRANSFER

H. F. Poppendiek

Reactor Experimental Engineering Division

It is of interest to compare the heat and momentum
transfer characteristics of the ART for the three
different types of fluoride fuels now being con-
sidered, namely, lithium-base, zirconium-base, and
beryllium-base fuels. The core, as well as the
fuel-to-NaK heat exchanger, must be considered.

In the case of the core, the wall-to-fluid temper-
ature difference which must be reduced by cooling
to prevent excessive wall temperatures is of prime
importance. For a given core geometry, power
density, and system pressure drop the wall-to-fluid
temperature difference can be related analytically
to the physical properties. Calculations for the
three different types of fuels indicate that the
wall-to-fluid temperature difference for the zir-
conium~base fuel will be about two times as great
as that for the lithium-base fuel and that the
wall-to-fluid temperature difference for the be-
ryllium-base fuel (based on somewhat limited
physical property data, at present) will be about
1.3 times as great as that for the lithium-base fuel.

PERIOD ENDING MARCH 10, 71955

In the case of the heat exchanger, a comparison
has been made on the basis that the sum of the
radial temperature difference plus one half the
axial fluid temperature difference is equal to a
constant; if all other parameters remained the
same, this condition would always yield the same
air outlet temperature in the radiator. An analysis
was made for a typical ART heat exchanger by
the method described previously.® The results
indicate that the zirconium-base fuel has a
Reynolds number of about 3000 and a corresponding
pressure drop about 1.6 times as great as that for
the lithium-base fuel and that the beryllium-base
fuel has a Reynolds number of about 2300 and a
corresponding pressure drop at least 1.4 times as
great as that for the lithium-bose fuel. The
Reynolds number for the lithium-base fuel is about
4000. The Reynolds number of the beryllium-base
fuel is low because its viscosity is relatively high
and its density relatively low (high kinematic
viscosity).

 

5M. W. Rosenthal, H. F. Poppendiek, and M. R.
Burnett, A Method for Evaluating the Heat Transfer
Effectiveness of Reactor Coolants, ORNL CF-534-11-63
(Nov, 4, 1954).

145
ANP PROJECT PROGRESS REPORT

9. RADIATION DAMAGE

D. S. Billington

J. B, Trice

Solid State Division

The program of irradiating Inconel capsules con-
taining fluoride fuels in the MTR is continuing.
A new method for resistance welding thermo-
couples to the capsule surface coupled with a
new control system has resulted in much improved
temperature control during the irradiations.

Design work has been completed on the miniature
in-pile loop, and many of the parts have been
fabricated, The miniature sump pump for the loop
has been tested and found to be satisfactory. The
horizontal-beam hole in-pile loop was operated
in the LITR and is now being disassembled for
metallographic examination. A pneumatic flux-
measuring device is being used for preliminary
measurements of the thermal-neutron flux to be
expected in the fuel region of the MTR in-pile
loop. Approval was obtained for irradiation of
the stress-corrosion apparatus in the LITR, and
examinations of the first irradiated specimen are

under way. The MTR creep apparatus has been
shipped to NRTS,

MTR STATIC CORROSION TESTS

W. E. Browning G. W, Keilholtz
Solid State Division
H. L. Hemphitl
Analytical Chemistry Division
irradiations in the MTR of Inconel capsules con-
taining fluoride fuels are continuing but have been
somewhat retarded because of interruptions in the

MTR operating schedule.

ments

Three recent improve-
in the MTR capsule irradiation facility,
which is being used to study static corrosion and
chemical stability in ANP fuel-container systems,
have consisted of a method for assembly of the
air annulus onto the capsule prior to insertion,
a controller with faster temperature and air flow
response, and an improved method for fabricating
thermocouples and for attaching them to the fuel
container walls.

The revised capsule—qir annulus assembly,
which is already being used in the MTR, was
shown schematically in the previous report.! The
arrangement has the advantages that the annulus
alignment can be performed and visually inspected
outside the reactor and that all the thermocouples

146

can be attached to the capsule assembly before
insertion into the reactor. Also, the problem of
maintaining thermocouples in the permanently
installed part of the facility to indicate capsule
alignment has been eliminated.
sleeve is easily removable in the hot cells in the
post-irradiation examination of the capsules.

The new system of control for maintaining a
steady capsule temperature through the proper
metering of cooling air to the capsule has de-
rivative action plus proportional control and fast
reset. This combination of controls has been
shown in bench tests to quench thermal oscil-
lations caused by inherent instabilities in the
system and to request an increase in cooling air
from the controller with sufficient speed to handle
any foreseeable sudden increases or decreases in
reactor power. It is now operating successfully
in the MTR capsule facility.

Two improved methods for welding thermocouples
to the Inconel capsules for the present series of
tests were made by R. J. Fox of the Central
Machine Shops. One method (illustrated in Fig.
9.1) involves crossing the chromel and alumel
wires, pressing them against the capsule with a
copper electrode, and passing current through them
to weld them to the capsule by resistance heating.
The other method uses the same welding tech-
nique, but the wires are laid parallel along the
surface before the current is applied. In each
case spot-welding parameters were optimized for
producing a thermocouple bead which appeared,
under the microscope, to have the best shape for
good heat transfer and mechanical strength. These
methods are better than the previous method, in
which the thermocouple was fused to the Inconel
surface with an arc discharge, in that a thermo-
couple junction is obtained which lies closer to
the Inconel capsule surface and therefore yields
a better measure of the capsule wall temperature.
A comparison of the crossed-wire type of re-
sistance-welded thermocouple and the previous
spark-welded type of thermocouple is presented

The annulus

 

w. E. Browning, G. W. Keitholtz, and H. L. Hemphill,
ANP Quar. Prog. Rep. Dec. 10, 1954, ORNL-1816,
p 120.
UNCLASSIFIED
PHOTO 13864

 

TRIMMED THERMOCOUPLE
-« DIRECTION OF AIR FLOW

 

AS-WELDED THERMOCOUPLE

|<g—*/4 in.

    

Fig. 9.1. Resistance-Welded Thermocouples.

in Fig. 9.2, in which the differences between the
temperatures measured with them and with an
optical pyrometer ot points adjacent to the re-
spective thermocouple beads are plotted against
air flow rate. It is apparent that the resistance-

welded thermocouples are subject not only to

PERIOD ENDING MARCH 10, 1955

smaller errors in temperature measurement but
they are also less sensitive to changes in cooling-
air flow rates. The new thermocouple assemblies
have been given thermal and mechanical service
tests far more severe than they will experience
in any of the MTR tests presently planned, and
they have survived these tests well.

MINIATURE IN-PILE LOOP

W. R, Willis M. F. Osborne
H. E. Robertson G. W. Keilholtz
Solid State Division

The components of an in-pile dynamic-corrosion-
testing loop for insertion in position C-48 in the

UNCLASSIFIED
ORNL-LR-DWG 5858
T

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

450 T T T T
Q SPARK -WELDED THERMOCOUPLE R-{3
A SPARK-WELDED THERMOCOUPLE R-15
#  SPARK-WELDED THERMOCOQUPLE R-16
400 |- ®  RESISTANCE-WELDED CROSSED-WIRE-TYPE ]
THERMOCOUPLES F-4, F-9 F-{0, F-il, AND F-12
—-m == DATA OBTAINED PREVICUSLY WITH SPARK-WELDED
THERMOCOQUPLES
w 350 |—~— AREA OF ACCEPTABLE READINGS WITH RESISTANCE - —
g WELDED THERMOCOUPLES
g TEST TEMPERATURE, 800°C
w AIR FLOW ANNULUS FORMED BY 0.2-in.-0D CAPSULE
w 300 AND 0,37in-ID TUBE —— —p— i
a :
Q
L]
o
z /
u 250 f
F ; / |
a :
g i ;
G 200 ° / :
¥ .
o o ;
3 |
i ;
z
/] ;
Z {50 ;
N o2
F
2 | |
A"
w i
& 100 1 *
g 58 2l B
E A’_j—v‘ ;
L %;u/lz: _ :
o - L~ !
ot - ;
50 = = 7
o —
- ,
e Wil « —— b e e
ey T, L
¥ ° e eas
5 e ® ® o -___._______'. -
®
-50
0 1 2 3 4 5 6 7
AIR FLOW RATE (scfm)
Fig. 9.2. Comparison of Spark-Welded and
Resistance-Welded Crossed-Wire-Type Thermo-
couples.

147
ANP PROJECT PROGRESS REPORT

LITR have been designed and partially con-
structed. When assembled, the unit will consist
of a miniature sump pump with a fuel loop and an
air heat exchanger, ond it will be encased in a
stainless steel jacket. The jacket will then be
connected to the exit facility on the northwest
face of the LITR by means of a flexible steel
tube which will communicate thermocouple wires
and other leads to the loop controls. It is
expected that the loop will operate at a fuel
Reynolds number of 3000 with a difference in
temperature of 100°F along the length of the fuel
tube. The loop, as installed, is shown sche-
matically in Fig. 9.3. After irradiation, the loop
will be transferred, under water, into the removal
cask, which will, in turn, be moved to a position
outside the reactor building. The stainless-steel-
jacketed loop will then be separated from the rest
of the unit and taken to the Solid State Division
hot cells where the fuel-container system will be
prepared for both metallographic and chemical
examination.

Performance tests have been made on the
miniature sump pump. A characteristic curve of
flow vs head, obtained with water at a pump speed
of 5000 rpm, is shown in Fig. 9.4. Similar
measurements of flow rate vs head for the fused
salts at elevated temperatures are not planned
because such tests would require the use of a
valve. However, flow rate vs pump speed has
been measured by using a metered volume of fused
salt and by using a venturi type of flowmeter
calibrated with water. A compoarison of the
measurements by the three methods is shown in
Fig. 9.5. It is apparent from the curves in Fig. 9.5
that the relationship between pump speed and
head can be obtained within the experimental
error from water tests.

The thermocouple corrections for the loop
geometry of the first in-pile model were obtained
with a special 30-in. test section inserted in a
mockup of the in-pile loop. The relationships
between the temperatures observed with an optical
pyrometer and those measured by the thermo-
couples were found and recorded as a function
of air flow rate along the annulus and power
generation in the fuel,

During a presentation of the safety features of
the miniature in-pile loop before the ORNL Ex-
perimental Review Committee, a question arose
regarding the adequacy of the LITR control rod
system to override the sudden increase in Ak/k

148

UNCLASSIFIED
55D-A-1123
ORNL-LR-DWG 53444

 

 

    

o

| I

T
Al

[—— FLEXIBLE METAL HOSE

 

 

 

 

 

   

 

 

 

X
L

 

7]

 

  

 

 

 

STAINLESS-STEEL-
JACKETED LOOP

 

 

 

 

—
T I

 

PRl

 

Sk

>,

 

 

 

 

 

 

 

 

 

¢ REACTOR

HL_’_’ JLA—IZJA‘

5
L

Fig. 9.3. Installation of Vertical In-Pile Loop
in Position C.48 of the LITR.

in the event the in-pile loop wall ruptured and
allowed all the fuel to spill into the bottom of
the stainless steel jacket of the loop. The bottom
of the loop will be in a very high thermal-neutron
flux region and, hence, in a very sensitive region
with respect to influence on reactivity of the
LITR. Since flux calculations indicated that an
excess reactivity increase of 2 to 4% might occur,
it was decided to mock up the situation experi-
mentally to obtain a more accurate prediction of

the increase. For the experiment the fuel mixture
was contained in an annulus around a long rod,
as shown in Fig. 9.6. This assembly was placed
in an outer aluminum jacket, which formed an air
annulus, and then lowered into the C-48 fuel
element position in the LITR. The fuel mixture
used contained 120 g of U233, which is within
12 g of the amount which will be used in the first
test. The annular geometry was chosen to repre-
sent the conditions which would produce the
highest increase in Ak/k.

UNCLASSIFIED
55D0~B-10444
ORNL-LR-~DWG 3222A
T

 

 

 

 

 

 

.l ol PUMP SPEED, 5000 rpm
TN t% -in--dia STRAIGHT -VANE
46 | e e SN IMPELLER

 

 

 

HEAD (ft OF H,0)
o

 

14 R

 

 

l
Il
I
k
|
I
i
|
!
I
|
|
:
[
[
i
|
|
I
!
|

e -

 

 

 

 

 

 

 

 

 

6 p— - —
o
G fommemm e e
f——DESIGN POINT: 1480 cm¥/min = \
2 : 4 fps (0.200-in.-1D TUBE) \
|
0 |
0 1000 2000 3000 4000 5000
FLOW (cm3/min)
Fig. 9.4. Head vs Flow Characteristics of

Miniature Sump Pump for In-Pile Loop.

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

58D-A-1040a
104 ORNL-LR-DWG 3221A
T T T T T T T T T
- @ EXPERIMENTAL RESULTS IN METERED VOLUME OF FUEL —
™ 0 EXPERIMENTAL RESULTS IN FUEL WITH -
—  VENTURI FLOWMETER o —
- EXPERIMENTAL RESULTS IN WATER 1~ ="
: -~ -~
. 5 _ . 9’4 D .
€ 0e /.‘///
& ”* e -
o + Ta f//
A ol
l - T N
o -
2 - 547
= /\fi’///
S 2f - e |
T o EXPECTED LIMITS OF ACTUAL HEAD
7 (CENTRIFUGAL PUMP) |
-
//
10° | I
1 2 5 10 20 30
CUTOFF HEAD (f)
Fig. 9.5. Head vs Pump Speed of Miniature

Sump Pump for In-Pile Loop.

PERIOD ENDING MARCH 10, 1955

The excess reactivity resulting from this simu-
lated accident was determined by measuring the
UNCLASSIFIED

SSD-C-11t44
ORNL-LR-DWG 50704

 

   
 

FILL TUBES PINCHED OFF
AND WELDED AFTER

ANNULUS IS FILLED\

 

 

 

 

 

 

 

 

ANNULUS\

 

 

 

0.203in —*

 

 

 

 

 

 

 

Fig. 2.6. Excess Reactivity Apparatus.

149
ANP PROJECT PROGRESS REPORT

LITR control rod position with the mockup in and
with the mockup out of the lattice. The LITR rod
calibrations were checked by following the known
xenon buildup in the reactor. After a correction
was made for buildup of xenon during the experi-
ment, it was found that the simulated accident
would odd less than 0.2% excess reactivity, which
can be handled easily by the LITR control rods.

LITR HORIZONTAL-BEAM-HOLE FLUORIDE
FUEL LOOP

J4. G. Morgan
M. T. Morgan

0. Sisman

W. E. Brundage

C. D. Baumann A. S, Olson

R. M. Carroll W. W. Parkinson
Solid State Division

The circulating-fluoride-fuel experiment has been
successfully conducted in hole HB-2 of the LITR,
and the loop is being disassembled for metallo-
graphic examination in the hot cell of Building
3025. The general description of the apparatus
and facility, as presented previously,Z applies
to this test, with a few modifications.

All loop welding was supervised by the Metal-
lurgy Division, and the recommended specifi-
cations were met. The schematic drawing shown
in Fig. 9.7 gives the details of the construction
of the loop. By x-raying the cairod heaters to
determine their effective heating lengths, it was
possible to position them in such a way as to
avoid hot spots on the fuel tubes. After the loop
was assembled, it was leak tested at 1500°F with
a helium leak detector, and then the loop and the
pump were brought to 1500°F under a helium
atmosphere. The molten salt NuF-Zer (50-50
mole %) was then charged into the loop to the
level of the lower probe in the pump bowl. After
further leak checks the loop was operated for
7]/2 hr at 1500°F with flow rates of 5 to 15 fps
based on the 0.225-in.-ID nose piece section.

The loop was then drained and cooled under g
helium atmosphere.  The filling and draining
methods used are shown schematically in Fig. 9.8.
In filling the locp all salt lines were heated to
1500°F under a purified helium atmosphere. With
the pump idling, the salt tank was pressurized
and the salt forced into the loop until it made

With

contact with the probe in the pump bowl.

 

2W. E. Brundage et al., Solid State Semiann. Prog.
Rep. Aug. 30, 1954, ORNL-1762, p 21,

150

the pump turned off, all pressures were equalized
and the fill line was frozen at point B, In draining
the loop, flow was stopped, and a portion of the
loop on the inlet side of the pump was frozen off.
By using the draining-pressure system, the salt
was then forced back into the salt tank until the
fill line was flushed empty,

The loop was inserted into LITR hole HB-2 on
December 7, 1954, with the reactor shut down,
and electrical and piping connections were made
before the shielding was completed. After bringing
the loop to temperature it was again filled with
the barren salt mixture, operated for 8% hr, and
drained. The enriched fuel NaF-ZrF  -UF, (62.5-
12.5-25 mole %) was then charged into the loop
and the fill line was sealed off. When 4.86 kg of
the enriched mixture had been added, the loop
was filled to the bottom proke of the pump bowl.
With the fuel circulating at temperature and the
reactor at zero power, the remainder of the external
shielding was added. A cutaway sketch of the
external shielding required is shown in Fig. 9.9,
Because of the difficulty in placing the paraffin
close to the hot pump enclosure, more gamma
shielding was required around the periphery than
was originally planned. Additional concrete block
walls were placed in front of the instrument panel
for protection of the operating personnel.

Full reactor power was reached ot 4:30 PM on
December 11, and the electric heaters were ad-
justed to maintain the desired operating tempera-
ture. Flow during the entire run was maintained
at 8 to 10 fps (Reynolds number of 5000 te 6500).
A composite plot of reactor power, loop tempera-
tufe, and salt Reynolds number vs time is shown in
Fig. 9.10.

The total power generation of the fuel under
reactor flux was determined by a series of heat
balances with the reactor off and at full power.
Equilibrium conditions were established with the
reactor at full power, and then upon reactor
shutdown the electric heaters were increased to
match the same temperature conditions along the
loop. This increase in electric power to duplicate
thermal conditions under flux was taken as the
power added to the system by the fissioning fuel.
Several such measurements gave an average of
2800 w, or about one-third the anticipated power.
An experiment to duplicate the perturbed flux
pattern is being conducted to determine the
highest wattage densities obtained in the loop
LGt

 

0L IN THROUGH SHAFT oren s s06s

   
     
   
    
  

HELIUM EXHAUST-

~JACKET ASSEMBLY _--~SPARK PLUG PROBE

  
  
 
  

ClL INLET FOR
RING COOLING —~{1

4

RADIATOR

 

_.—CENTRIFUGAL PUMP MODEL LFA

I

_—-——CONTROL GAS

-~ FUEL TUBING KOVAR SEAL ..
) FITTINGS —* - PUMP ENCLOSURE
_ _NOSE COVER

DRAIN TUBE SEAL-.
e ALSMAG 202 "} ~HELIUM ATMOSPHERE

zd HEATER CORE EXPANSION BELLOWS -,

i T NICHROME V .

CORE HEATER e [

‘*—f% . !

SECTION A-A

 

 

WATER OUTLET
EXPANSION BELLOWS-~_
N

JACKET ASSEMBLY-. HEAT EXCHANGER W ' LINER SEAL
CADMIUM SHEILD~ _Vs-in. DIA SUPPORT ROD N [ {AND FLANGE

HELIUM PURGE LINE - N / _FUEL TUBING 3 ~CALRORS 1oy e
. 0" RING

  
    

\ ) CFUELTUBING | ‘.
\ /o (OUTS | \ \33 P

 

AR INLET

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 
  
  
  

        

 

v \ - 5
RADIATOR - y | ANNULLS FOR / \ VENTURI- i/ ' |
BORON SHIELD~ /. WATER FLOW-~ “TRANSITION N3, “-FUEL TUBING {IN) “KOVAR SEAL
,_/ PLATE /a-in_DIA x FITTINGS
NOSE COVER SUPPORT ROD FUEL LEADS *PRESSURE N
TO DIAPHRAGM” TRANSMITTER CELLS %~ ~WATER INLET
HELIUM LEADS TO
20 in. REF — ———————— : DIAPHRAGM

 

 

“———12ft 5in. REF ~———————— e

 

 

Fig. 9.7. Schematic Diagram of LITR Horizontal-Beam-Hole Fluoride-Fuel Loop.

§S61 ‘0L HOY¥YW ONION3 doiy3d
ZGi

UNCLASSIFIED
SSD~-B— 11134
ORNL-LR—-DWG 5065A
CHECK VALVE

 

 

  
 

 

 

 

 

 

FREEZE-OFF LINE

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

=" B

- LOOP

OIL-FILLED BURBLER - /
< —3 ¢
FILL LINE — \ . >>
A/

(N

{

| '[

| \ CHARGE TANK

|
I | MELT-0UT FURNACE — =i
HIGH-PURITY HELIUM DRAINING HIGH - PURITY HELIUM

PRESSURE SYSTEM

Fig. 9.8. Fill-and-Drain System for LITR Horizontal-Beam-Hole Fluoride-Fuel Loop.

L30d3Y §534908d LID3F0dd ANV
test and to see why the total power was lower
than expected.

On January 7 the drive belt to the pump motor
became inoperable, and, with the reactor shut
down, a portion of the external shielding was
removed and the belt replaced. It was found to
be impossible to get circulation started again
because of a cold region in the loop, and therefore
the test was terminated. The loop was withdrawn
from the reactor and taken to the hot cells,
Building 3025, for disassembly. The tubing will
be sectioned for metallographic examination. The
loop had operated for a total of 645 hr, with 475 hr

at full reactor power.

CONCRETE BLOCK SHIELD
6 ft HIGH

PERIOD ENDING MARCH 10, 1955

The flow-measuring device described previously?
proved to be very successful and enabled con-
tinuous monitoring of the salt velocity to be
recorded. Design changes were made (Fig. 9.11)
that permitted the transmitter to withstand a
pressure as high as 10 psi across the diaphragm
without causing a pressure shift. The transmitters
were subjected to pressures as high as 40 psig
during operation with the first filling of npon-
uranium-bearing salt. Each cell has a sensitivity
of about 0.1 psi. k

 

3W. E. Brundage et al., Solid State Semiann. Prog.
Rep. Aug. 31, 1953, ORNL-1606, p 29.

PR
SSD-B-1121A
ORNL-LR-DWG 53424

HB-2

LITR SHIELD

 

 

 

400 mr/hr

 

 

7 mr/hr

  

 

 

 

 

 

HB-3 EQUIPMENT
SHIELD

 

 

 

L PUMP

 

 

 

 

 

 

 

%

16in. —m=t

   

3S5Smr/hr

 

 

 

 

 

 

&0 mr/hr

3-in. PARAFFIN \H1GH~DENS|TY BLOCK walLL

gft HIGH

/___ CONTROL PANEL

Fig. 9.9. Horizontal Section Through Shielding Required During Operation of LITR Horizontal-Beam-

Hole Fluoride-Fuel Loop. Gamma fluxes indicated.

153
¥Sl

1500

 

1450

 

 

1400

NOSE TEMPERATURE (°F)

6500

 

 

6000

5500

5000

43500

REYNOLDS NUMBER

4000

 

 

 

 

REACTOR POWER ( Mw)

TIME {days)

Fig. 9.10. Reactor Power, Flyoride Fuel Reynolds Number,
Operating Time.

 

ORNL—LR—DWG 5343 A

and Temperature of Nose of Loop vs

LY0dIY SSIYI03d LDIrodd dNV
UNCLASSIFIED
550-A—-1t09A4
ORNL-LR-DWG 5065A

HELIUM GAS LINE
TO CONTROL PANEL

MODIFIED CHAMPION V-1 SPARK PLUG
WITH PLATINUM TIP

     
 

 

 

 

 

 

 

PLATINUM CONTACT

 

 

WELD

 

LIQUID LINE
TO VENTUR|

Fig. 9.11. Pressure-Transmitter Cell.

FLUX-DEPRESSION EXPERIMENTS IN MTR

J. B. Trice H. V. Klaus
Solid State Division
R. H. Lewis

Phillips Petroleum Company

Preliminary measurements of the thermal-neutron
flux to be expected in the fuel region of the first
in-pile loop scheduled to go into hole HB-3 of
the MTR are being made by using the pneumatic
flux-measuring device which is presently in hole
HB-3. The decrease in the flux of the MTR that
will be caused by the fuel and the fuel container,
as well as by the auxiliary equipment of the
in-pile loop, is being determined. The loop, which
will be made of lnconel, was mocked up by using
a straight Inconel tube filled with a simulated
fuel, which consisted of a mixture of cadmium and
magnesium for one series of tests and a mixture
of aluminum and boron carbide for another series.

The preliminary results of the flux-depression
experiment indicate that for the fuel which is to
be used in the first in-pile loop tests, the de-
pression will be on the order of 70%, which means,
in terms of loop power, that the power of the
presently conceived loop design may be expected
to be 5 to 10 kw rather than the desired power of
15 to 30 kw. The nose of the loop is therefore
being modified to increase the expected power.
Before the loop is inserted in the MTR, it is

PERIOD ENDING MARCH 10, 1955

planned that a full-scale mockup of the loop will
be placed in the MTR to determine the expected
power (cf., sec. 3, ‘‘Experimental Reactor Engi-
neering’').

CREEP AND STRESS-CORROSION TESTS

W. W. Davis J. C, Wilsen
N. E. Hinkle J. C. Zukas
Solid State Division

Approval by the ORNL Experiment Review Com-
mittee for irradiation of the stress-corrosion appa-
ratus previously described? was received shortly
after the completion of bench tests involving
compatibility of component parts in case of sodium
leakage. Alarm circuits to signal sodium leakage,
water in-leakage, and excessive temperatures were
installed in the apparatus and it was inserted in
HB-3 of the LITR. The Inconel fuel chamber,
which contained 0.52 g of Nc:l"'—ZrF“-UF4 (63-
25-12 mole %), was surrounded by approximately
25 g of sodium. Because of the rapid power
changes in the furnace required to maintain the
specimen control temperature at 1500°F during
reactor startup or shutdown periods, a Leeds &
Northrup Speedomax and air-controller combination
was used. Thermocouples in wells in the sodium
chamber recorded fluctuations that did not exceed
10°F at the outset of the test; the fluctuations
gradually decreased to one-half this value after
135 hr of test, and the temperature remained steady
thereafter, The chamber was ot control tempera-
ture for 1120 hr, during which time the reactor was
up to power for approximately 700 hr.

Periodic checks of the resistance between the
stressing weight and the weight probe throughout
the test indicated that no gross increase in creep
raote was caused by irradiation at a stress of
1000 psi. The rig was removed to a shield to
decay sufficiently to permit handling and dis-
secting. The transverse cross section of the
specimen below the fuel level will be polished
and etched for metallographic examination. A
companion bench test is now in operation. Several
attempts to change or eliminate the baffle ar-
rangement now in use to simplify both assembly
and sectioning of the apparatus reintroduced the
temperature excursions which were so troublesome
at the outset of design of the rig.

 

4J. C. Wilson er al.,, ANP Quar. Prog. Rep. Sept. 10,
1954, ORNL-1771, p 142.

155
ANP PROJECT PROGRESS REPORT

Operation of the present apparatus in the LITR
is satisfactory, but its performance in the MTR is
not assured because of gamma heating in the
relatively massive apparatus required to achieve
smooth temperature control, Stress-corrosion data
are vurgently needed at fuel power densities of
1000 w/cm® in the MTR, and therefore a possible
short-cut  stress-corrosion apparatus is being
mocked up. The specimen tube consists of a
cylinder stressed by gas pressure on the inside
and in contact with fused salts on the outside.
The salt is in an annular space around the
specimen tube; the outside of the salt annulus
is the inside of a container tube that is finned
on the outside to transfer heat to helium in a
water-jacketed con. To preserve a low surface-
to-volume ratio it was suggested (by R. G.
Berggren) that the salt be allowed to remain solid
at the container tube wall. This cuts the surface-
to-volume ratio approximately in half if any inter-
action between the solid salt and the container
wall is ignored. Whether good heat transfer can
be obtained between the solid salt and the wall
remains to be determined. The experiment is
unique in that the hottest part of the molten salt
is in contact with the specimen tube (at 1500°F).
In capsule tests the inside of the salt volume has
frequently been several hundred degrees hotter
than the salt-metal interface.

Since rupture of one of the metal members is a
possibility, cooling in a gas stream is probably
unsafe from a reactor operations standpoint. Con-
vection cooling by fins to helium in a water jacket
or conduction by fins to a water-cooled heat sink
appears to be attractive. Experiments on con-
vective heat transfer from fins have shown that
there is a good chance of success. [f this is not
successful, conductive heat transfer will achieve
the results. Empirical heat transfer data are being
obtained for longitudinally finned, vertical cyl-
inders, and this work will be extended to cover
conducting fins. Generation of sufficient heat in
small cylinders to simulate the fission heat has

156

proved to be difficult, Passage of an electric
current through the finned tube works well for
low-conductivity fins, and carbon-arc and platinum-
radiation heaters are being tried. As soon as.
dota can be reliably extrapolated a test will be
put in the LITR.

The MTR creep apparatus and accompanying
instrumentation have been shipped to the MTR.
Details to be supplied for approval of the stress-
corrosion rig have been assembled but cannot be
completed until the first LITR test has been
analyzed.

Tests have begun on an electromagnetic trans-
ducer that is advertised to be operable at 1300°F.
Similar to a microformer in principle, it is
‘‘canned’’ in stainless steel and has ceramic
insulation on all windings. The device will be
tested under irradiation upon completion of bench
tests. If operable in-pile, the transducer will be
suitable for strain measurements in the bending-
type stress-corrosion apparatus. An extensometer
for the internally pressurized ‘‘tube burst’’ speci-
men is possible, but no development work is
being carried out. For strain data, the internal
diameters of the specimen tube will be measured
with o pneumatic goge after irradiation, and
multiple specimens will be irradiated, if neces-
sary, to obtain a strain-time curve.

Two bench tests of Inconel tubes in bending in
a helium atmosphere with NaF-ZrF -UF, (53.5-
40-6.5 mole %) were completed. After 432 hr at
1500°F a few subsurface voids were visible that
were distributed about equally around the periphery
of the tube. Another test was operated for 866 hr,
and the number of subsurface voids per unit length
of periphery was greater by a factor of 2 at the
tension and compression sides of the tube than
at the neutral axis. This tentatively confirms the
hypothesis that stress-corrosion in the usual sense
does not take place in these tests and that the
phenomenon should be called

“*strain-rate’’ or

**strain’’ corrosion.
PERIOD ENDING MARCH 10, 1955

10. ANALYTICAL STUDIES OF REACTOR MATERIALS

C. D. Susano
Analytical Chemistry Division
J. M. Warde
Metallurgy Division

Research on the determination of frivalent
uranium and uranium metal in fluoride-base reactor
fuels was continued. Further studies were made
of the methods for determining oxygen as oxides
in fluoride salts, A bromination procedure was
applied to the determination of oxygen in uranium
and beryllium compounds.  Modifications were
developed of methods for determining beryllium,
potassium, and lithium in fluoride fuels. High-
temperature x-ray spectrometer studies of fluoride
mixtures were initiated as an aid in the determi-
nation of phase diagrams.

ANALYTICAL CHEMISTRY OF
REACTOR MATERIALS

J. C. White
Analytical Chemistry Division

Determination of Trivalent Uranium in

Fluoride Fuels

A. S. Meyer, Jr. D. L. Manning
W. J. Ross
Analytical Chemistry Division

Oxidation of Trivalent Uranium by Methylene
Blue. Evaluation of the data obtained for the
determination of trivalent uranium in fluoride fuels
by the methylene-blue, one-step oxidation method’
has shown that the coefficient of variation for
NaF-KF-LiF-base materials is 2%. The coeffi-
cient of variation of the hydrogen-evolution method?
for similar materials, including NaF-ZrF -base
eutectics, is 4%. In LiF, NaF, and a mixture
which contains NaF-KF-LiF, the agreement be-
tween the results for trivalent uranium obtained
by both methods is, in general, satisfactory. The
results of the determination of trivalent uranium
by the use of methylene blue in compositions of
NaF-ZrF ,-UF ,-UF ; and KF-UF ;-UF, are difficult

to reproduce and are significantly lower than those

 

]A. 5. Mevyer, Jr., et al., ANP Quar. Prog. Rep.
Dec, 10, 1954, ORNL-1816, p 130.

2D. L. Manning, W. K. Miller, and R. Rowan, Jr.,
Methods of Determination of Uranium Trifluoride,

ORNL-1279 (Apr. 25, 1952).

obtained by the hydrogen-evolution method.. Since
the major portion of the samples received for
analysis are from the NaF-ZrF, system, an effort
has been made to adapt the methylene-blue method
to these samples. The low values obtained by
this method may result from incomplete dissolution
of the samples or from partial oxidation of the
trivalent uranium by ionic hydrogen. It has been
found that these samples can be dissolved if
stirred for 2 hr in a methylene-blue solution that
is almost completely saturated with AICI,. Al-
though more reproducible results are obtained
when complete dissolution is attained, the in-
stability and heterogeneity of the samples make
impractical a direct comparison of the results of
the modified methylene-blue procedure with those
of the hydrogen-evolution method., Accordingly,
dissolutionis now being carried out in an apparatus
in which any hydrogen that is formed by the
reaction

UF, +H" — U4" +3F- + L H,
can be measured.,

Preliminary experiments on the dissolution of
samptes of UCl, have indicated that when the
samples are dissolved in solutions of methylene
blue that are 3 M in HCl and saturated with AICI,
a small quantity of hydrogen is liberated but that
the volume of hydrogen is negligible when the con-
centration of HCl in the reagent is 1.5 M. In
solutions of lower acidity the oxidation of trivalent
uranium is carried beyond the tetravalent state.
Dissolution tests in solutions of intermediate acid
concentration are now being performed.

Simultoneous Determination of Trivalent and
Total Uranium.  Further efforts were made to
develop a procedure for the determination of the
total uranium content of these samples by direct
titration of the tetravalent uranium in the solution
after the determination of the trivalent uranium by
the methylene-blue method. Potentiometric titra-
tions of the reduced solutions have been carried
out with K2Cr207, Ce(504)2, KMnO4,Gnd Fe2(504)3
as oxidants. The titration with each of the above
reagents is too slow for application to analytical

157
ANP PROJECT PROGRESS REPORT

procedures.  Titration was not improved by re-
ducing the chloride ion concentration or by adding
H,S0, or H,PO,. The solutions could not be
titrated at elevated temperatures because the
methylene blue decomposed rapidly when heated.
When the solutions were treated with an excess of
oxidant and back-titrated with a standard solution
of ferrous sulfate, a portion of the excess reagent
was reduced, either by chloride ion or by methylene
blue.

During the course of the above investigations,
solutions of UO,‘_,SO4 in an excess of methylene
blue were titrated with a solution of CrS0,.
Definite potentiometric end points were obtained
that were similar to those reported ! for the titration
of solutions of tetravalent uranium in methylene
white with K,Cr,0,. The volume of titrant con-
sumed at the first end point, which coincided with
the decoloration of the methylene blue, corres-
ponded to the reduction of the methylene blue to
methylene white plus a one-electron reduction of
the hexavalent uranium. An additional equivalent
of Cr50, per mole of uranium was required to
titrate to the second end point. When solutions
which contained a molar excess of U02504 over
methylene blue were titrated, a similar titration
curve was obtained in which two equivalents of
titrant per equivalent of methylene blue were re-
quired to titrate to the first end point. In order to
eliminote the green color of trivalent chromium the
titrations were repeated with solutions of trivalent
titanium. At the first end point a colorless solu-
tion was obtained, and on the addition of an excess
of titrant the green color of tetravalent wuranium
was developed. The absorption spectra of the
solutions that had been titrated to the first end
point did not correspond to those of solutions of
any combination of tetra- and hexavalent uranium.
These results are consistent with the postulation
that an interaction species of pentavalent uvranium
and methylene white is stable in aqueous solu-
tions. A more detailed study will be necessary
before the existence of such a complex can be
established. The only procedure which appears
to be practical for the determination of total
uranium in these solutions consists of destroying
the methylene blue and removing the chloride by
a wet-oxidation procedure; the uranium content
of the resulting solution can then be determined
by a conventional titration method. Methylene
blue was found to be rapidly oxidized by fuming

158

with HNO, and HCIO,,,

Determination of Trivalent Uranium by the Karl
Fischer Modification of the Hydrogen-Evolution
Method. A procedure is being tested in which the
hydrogen that is evolved by the reaction between
UF, and an acid solution is converted to water,
and the water is determined by a modified Karl
Fischer titration. The hydrogen-evolution method?
is limited to samples which contain enough tri-
valent uranium to liberate a volume of hydrogen
that is sufficient for precise volumetric measure-
ment (1 mg of U liberates ~0.04 ml of H,, STP).
An additional limitation of this method is that all
gases which are insoluble in solutions of KOH
are measured as hydrogen. In the analysis of
samples which yield only a few tenths of a milli-
liter of hydrogen, adsorbed gases and gaseous con-
taminants of the reagents and the sweep gas may
introduce serious error.,

In the procedure being investigated the sample
is dissolved in an 80% solution of HCI. The
hydrogen liberated is passed through concentrated
H,50, and two drying towers of MgClO, by a
stream of purified helium. The dried gases are
then passed over CuO at a temperature of 500°C
to convert the hydrogen to water, which is then
absorbed in a solution of Karl Fischer reagent in
ethylene glycol and finally titrated coulometrically
with iodine. When samples of UF, that contained
from 4 to 20 mg of trivalent uranium were analyzed
by this procedure, the titrations corresponded to
96% of the theoretical trivalent uranium content,
with a coefficient of variation of 5%. The precision
of the method appears to be limited by the large
and variable blank titrations which are probably a
result of a mixing between the solutions in the
cathode and ancde compartments of the coulometric
titration cell. The cell is being modified in an
attempt to reduce this source of error.

Determination of Uranium Metal in
Fluoride Salt Mixtures

A. S. Meyer, Jr. B. L. McDowell
Analytical Chemistry Division

Further studies were carried out to improve the
method for the determination of uranium metal in
fluoride fuels by converting the metal to the
hydride and measuring the quantity of hydrogen
evolved by the thermal decomposition of UH,.
Although results of satisfactory precision and
accuracy can be obtained by the procedured in
which the decomposition of the hydride is carried
out under an atmosphere of CO,, it was found that
for some samples of UF ., periods in excess of 8
hr were required for the complete evolution of
hydrogen. In addition, no completely dependable
method has been found for removing the CO formed
when the CO, is reduced by uranium metal and
by trivalent uronium., Although the CO can be
quantitatively oxidized to CO, by passing the
gaseous reaction products over a mixture of 1,0,
reagent and powdered pumice stone at a tempera-
ture of 150 °C, the 120 reagent has been found to
become inactivated after limited, but varied,
periods of service., The reagent is, therefore, of
doubtful value when extended ignition periods are
required to complete the decomposition of the
hydride. In order to eliminate the possibility of
interference by CO and to reduce the ignition time,
modifications investigated in which the
decomposition of the hydride was carried out under
atmospheres of HCl and NH.,. When NH, was
used, the hydrogen was measured over a solution
of H2504 instead of KOH. In the presence of each
of the gases the volumes of hydrogen obtained
from the decomposition of UH, were not reproduc-
ible and were less than those predicted by the
postulated reactions:

UH, + 3HCl —> UCI, +3H,
2UH, + 4NH, —> 2UN, + 9H,

As an additional complication the decomposition
of NH, is catalyzed by the residue of uranium
nitrides to produce a slow evolution of insoluble
gases which continves for several hours after the
initial, rapid evolution of hydrogen from the
reaction between NH, and UH.,.

A modification is now being investigated in
which UH, is ignited in a stream of oxygen, and
the effluent gases are passed over heated CuO to
ensure the conversion of hydrogen to water, which
is then measured volumetricaily at reduced pres-
sure. The reaction between UH; and oxygen has
been reported? to be rapid and quantitative when
applied to the determination of hydrogen in large
samples of UH,. A modification of the method of

were

 

3A. S. Meyer, Jr,, and B. L. McDowell, ANP Quar.
Prog, Rep. Dec. 10, 1954, ORNL-1816, p 129,

4. C. Warf, The Composition of Uranium Hydride and
Its Decomposition at 250°C, CC-105% (Oct. 9, 1943).

PERIOD ENDING MARCH 10, 1955

Naughton and Frodyma® for the microdetermination
of carbon and hydrogen will be used for the deter-
mination of microgram quantities of hydrogen as
UH,. In this method the water produced by the
reaction between UH, and O, is first isclated in
an evacuated freezeout trap. It is then allowed to
volatilize into an evacuated vessel of known
volume, and the equivalent quantity of UH, is
calculated from the pressure which is measured
on a mercury-sealed, oil manometer. The apparatus
has been constructed and is now being calibrated
by using samples of BaCi,.2H,0 as a standard for
hydrogen in the form of water. On the basis of
design calculations, it would appear that less
than 1 ug of hydrogen can be determined by this
technique.

Determination of Oxygen in Fluoride Fuels

A. S. Meyer, Jr. J. M. Peele
Analytical Chemistry Division

Further tests of the procedure® for the determi-
nation of oxygen as oxides were carried out with
new components which were incorporated in the
apparatus to prevent the carryover of nonvolatile
electrolytes during the transfer of HF to the con-
ductivity cell. The new components include a
larger reactor and a splash trap in the transfer
line. The experiments indicate that, although the
conductivity method for the determination of the
water produced by the reaction of metallic oxides
with  KHF, is theoretically applicable, the
method is not practical for the determination of
microgram quantities of oxygen. It has been found
that repeated distillations with HF are required to
transfer the H,0 from the KHF, solution of the
sample to the conductivity cell. In the course of
these distillations, a sufficient quantity of KF is,
apparently, carried into the cell to mask the in-
crease in conductivity that would be produced by
small quantities of water.

An alternate procedure has been proposed for
the determination of the H,O that is formed on
dissolution of the oxides in KHF ,. Tests of this
procedure are now being carried out in parallel
with some further studies of the original method.

 

5J. J. Naughton and M. M. Frodyma, Anal. Chem, 22,
711 (1950).

6A. S. Meyer, Jr., and J. M. Peele, ANP Quar. Prog.
Rep. Sept. 10, 1954, ORNL-1771, p 148.

159
ANP PROJECT PROGRESS REPORT

The new method is based on the electrolysis of
the water in the fused bifluoride melt to yield
oxygen which ¢an then be separated from the other
reaction products and measured,

Melts of KHF, have been reported to be readily
dehydrated by carrying out the electrolysis until
rapid evolution of fluorine occurs.” During this
drying period, hydrogen is liberated at the cathode,
while O,, OF,, and F, are liberated at the anode.
Analysis of the product obtained during the initial
electrolysis of a commercial fluorine generator®
indicates that the H,O is removed during the early
stages of the electrolysis and that less than 10%
of the oxygen is evolved as OF,. The OF, is
generated only from relatively wet melts, and
negligible quantities are formed after the concen-
tration of H,O in the electrolyte is reduced to a
few tenths of a per cent.

Tests of the method are now being carried out
by using samples which contain milligram quantities
of oxygen. For these samples the oxygen can be
measured volumetrically by sweeping it into an
azotometer with CO,. Hydrogen is eliminated
from the electrolysis products by adding AgF to
the melts so that metallic silver rather than
hydrogen is deposited at the cathode. Two addi-
tional advantages from the Ag* ions in the elec-
trolyte may be anticipated. The standard oxidation-
reduction potentials? in acid solution of pertinent
couples are tabulated below:

Reaction E° (v)

U4+ 2H,0 — U0, ™ + 4H" + 2¢~ ~0.41
Ag —> A'g+ +e” —-0.80
2H,0 —> 0, + 4H" + 4~ 1.298

Agt — Agtt i e- 1.98

2F~ — F, + 2¢™ 2.85

The potential of the first reaction was obtained
from the Handbook of Chemistry and Physics, 10

7'J. H. Simons (ed.), Fluorine Chemistry, |, 227,
Academic Press, New York, 1950.

BR. C. Dawning etal., Ind. Eng. Chem. 39, 259 (1947).

9W. M. Latimer, The Qxidation States of the Elements
and Their Potentials in Aqueocus Solutions, p 296,
Prentice-Hall, New York, 1938.

IOC. D. Hodgman {Editer-in-Chief), Handbook o(
Chemistry and Physics, 34th ed., p 1552, Chemica
Rubber Publishing Co., Cleveland, 1952.

160

If the potentials of the above couples in the
KF-HF system are of similar relative order,
tetravalent uranium would be oxidized to the
hexavalent state by Ag* ions. In the absence of
the silver salt it would be necessary to oxidize
the wuranium electrolytically before the oxygen
could be quantitatively evolved. Since the method
is to be applied to samples which contain tri- or
tetravalent uranium as major constituents and only
traces of oxygen, the time required for the elec-
trolytic oxidation of the uranium would be prohibi-
tive. Furthermore, if Ag** forms a stable solution
in the fused electrolyte, it would be expected to
act as a fluorine carrier and thereby increase the
efficiency of the generation of O,. Under ideal
conditions Ag** could serve as a coulometrically
generated reagent for hydrogen in the fused salts.

A silver-lined, nickel cell has been assembled
for the purpose of studying this reaction, The
cell is charged with sufficient KF, AgF, and HF,
purified by distillation, to produce approximately
40 g of a composition of KF-2HF containing 5 g
of AgF.

The electrolyte is fused by heating the cell to
100°C, Air is removed from the cell by bubbling
CO, through the melt, The effluent gas from the
cell is freed of HF and fluorine by passing it
through NaF and mercury, and the insoluble
gaseous components are then collected and
measured over KOH in an azotometer. The charge
is electrolyzed until no more insoluble gas is
obtained. The sample is then added to the cell
and the gas that is obtained on electrolysis is
measured as oxygen. A recovery of 96% of the
oxygen present in the H,O produced when a
sample of Na,CO, was dissolved in the electrolyte
was attained,

Difficulties in electrolysis have been experienced
because of corrosion of the electrodes. When a
platinum anode is used, only a small quantity of
fluorine is liberated before the evolution of oxygen
is completed. The anode is, however, corroded
rapidly, particularly in the final stages of the
electrolysis. When nickel ancdes are used, corro-
sion is less severe, but the evolution of fluorine
is so rapid that it cannot be conveniently removed
by a mercury scrubber, An additional problem is
presented in that after extensive periods of elec-
trolysis, dendritic deposits of silver accumulate
and grow to such lengths that they provide a
metallic electrical circuit between the electrodes,
it is believed that this problem will not be of
major importance when samples which contain only
microgram quantities of oxygen are analyzed.

Determination of Oxygen in Metallic
Oxides by Bromination

J. C. White G. Goldberg
J. P. Young
Analytical Chemistry Division

A simple, precise method for the determination
of oxygen present as oxides in metals has leng
been desired. Recently, Codell and Norwitz!!
reported on the successful application of a bromi-
nation procedure for the determination of oxygen
in fitanium and titanium alloys. These investi-
gators passed bromine at 815°C over a sample of
titanium (which had been intimately mixed with
graphite) to form CO, which was, in turn, con-
verted to CO,, absorbed, and weighed. The
reaction time was of the order of 2 hr. Experi-
ments are now being made in an attempt to apply
this technique to the determination of oxygen in
metals of interest to the ANP program. The
progress of experiments on uranium ond beryflium
is presented here,

Uranium. The apparatus for the determination of
oxygen in uranium is shown in Fig. 10.1. Bromine
vapor is carried by helium over a boat (maintained

 

YIM. Codell and G. Norwitz, '‘New Method for De-
termining Oxygen in Titanium and Alloys,”’ Chem. Eng.
News 32, 4564 (1954).

TYGON TUBING

HELIUM

 

 
  
 
   

DEHYDRITE

ASCARITE

X1
TUBE FURNACE /?

 

 

 

300-mi FLASK

SCRUBBER BROMINE BUBBLER

DEWAR FLASK

ALCOHOL AND DRY ICE

QUARTZ SAMPLE BOAT
QUARTZ TUBE (14in. BY 1in.}
g YQ
|/‘_\"/lx:x:z::x

PERIOD ENDING MARCH 10, 1955

at about 950°C) that contains a mixture of a
uranium compound and spectrographically pure
graphite in a weight ratio of about 1 to 10, The
volatile reaction products are condensed in two
dry-ice—alcohol baths and the COis passed through
CuO, where it is converted to CO, and then
absorbed in a standard solution of Ba(OH),. The
reactions involved when UO, is the standard
material are

UO2 + 2Br, + 2C —> UBr, + 2C0O
CO + Cu0 — CO;2 + Cu
CO, + Ba(OH), —> BaCO, + H,0

The data indicate that for quantities of UO, of
the order of 100 mg, the first reaction is quantita-
tive in 2.5 hr at 950°C. A blank of 0.4 mg of
oxygen per hour must, however, be subtracted from
the total recovered, At lower temperatures, the
rate of reaction is considerably decreased. A
precision of approximately 1% is indicated. Future
work will involve a study of essentially micro
quantities of oxygen with direct measurement of
CO by 1,0, or by other methods.

Beryllium. The bromination procedure has also
been applied to the determination of oxygen in
beryllium. Preliminary experiments showed that
the rate of reaction between BeO, graphite, and
bromine at 950°C is far too slow for analytical
application. For example the reaction was only
5% complete after 1.5 hr. If a flux of Na,FeF is
added to the mixture, however, the rate of reaction

]
ORNL-LR-DWG 5608

TUBE FURNACE
TUBING {Smm 1D) xfl)
SN

=

250-mi FLASK

   
  
   
       
    
 

   

CuD TUBE
{20mm [D AND
14in. LONG)

Ba(OH)o
BUBBLER

 

 

™~ DEWAR FLASK

Fig. 10.1. Apparatus for the Determination of Oxygen in Uranium Compounds.

161
ANP PROJECT PROGRESS REPORT

is increased substantially, The probable reactions
are

3BeO + 6NaF + 2FeF3 —_ Fe203 + 3NazBeF4
Fe203 + 3Br2 +3C —> 2FeBr3 + 3CO

The reagents are mixed in a platinum boat which
is placed in a quartz tube encased with a platinum
shield to prevent attack by fluoride on the quartz.
The experiments made to date indicate that at
temperatures of the order of 700°C the reaction is
80% complete within 2 hr, It is anticipated that
higher temperatures will accelerate the reaction
rate. The determination of CO is completed as
previously described. Future work will involve
the application of this procedure to the determi-
nation of oxygen in NaF-BeF2 mixtures,

Differential Spectrophotometric Determination
of Beryllium

A. S. Meyer, Jr. D. L. Manning
Analytical Chemistry Division

The determination of beryllium in NaF-LiF-BeF,-
UF ,-UF, samples is required. The method of
Vinci'? in which the absorption of the colored
lake that is developed when alkaline solutions of
beryllium are treated with p-nitrobenzeneazoorcinol
is essentially specific for beryllium, but it is not
sufficiently precisg to permit application to the
determination of beryllium in reactor fuels, When
the technique of differential spectrophotometry was
applied to this determination a precision com-
parable to that of the titrimetric method!3 was
found..

In the differential procedure the absorbance of
the solution at 510 mu is measured against a blank
which contains 10 pg/ml of beryllium. On the
basis of replicate determinations of standards and
of a limited number of fuel samples, the coefficient
of variation of the method is less than 1% for
solutions which contain between 10 and 16 pg/ml
of beryllium.

Citrate ion is added to the solutions to prevent
the precipitation of interfering ions, principally
U02++. The presence of U02++ introduces a small
positive error which is approximately a linear

When the

function of the uranium concentration.

 

26 A, Vinci, Anal. Chem. 25, 1581 (1953),

13}, C. White, ORNL Master Analytical Manual, Method
No. 9 00711070, ORNL CE-53-1-235, Vol. 1.

162

weight ratio of uranium to beryllium is 10 to 1, the
error is only 3%. When the ratio is increased to 25
to 1, uranium is precipitated.

Determination of Lithium in NaF-Ber-LiF
and NuF-ZrFA-LiF Base Fuels

J. C. White G. Goldberg
Analytical Chemistry Division

The method of White and Goldberg'4 was applied
to the determination of lithium in NaF-BeF,-LiF
and NaF-ZrF -LiF base fuels, In this procedure
the sulfate solution of the alkali metals is con-
verted to a chloride solution by passage through
an anion-exchange resin in the chloride form. The
LiCl is then extracted with 2-ethyl-1-hexanol, and
the chloride ion is titrated in the nonaqueous
medium. If the acidity of the sulfate solution is
adjusted to approximately 1 N, the zirconium, like
the wuranium,
resin as the anionic sulfate complex and is thereby
separated from lithium.

Beryllium does not form an anionic complex with
sulfate ion and will pass through the column with
lithium and the other alkali metals. Moreover,
BeCl, is soluble in 2-ethyl-1-hexanol and ac-
companies the LiCl through the procedure. The
extraction of BeCl, was shown to be quantitative.
The described method for lithium is thus used to
determine the sum of lithium and beryllium. The
beryllium concentration is then determined by the
fluoride titration method, '3 while the lithium con-
tent is obtained by difference,

is retained on an anion-exchange

Determination of Potassium in Fluoride Fuels

C. R. Williams
Analytical Chemistry Division

A rapid, direct method for the determination of
potassium in fluoride fuels has been developed.
This procedure is based on the work of Wittig!S
and his co-workers, who found that the potassium
salt of tetraphenyl boron is very slightly soluble
in acidic solution, in contrast to the sodium and
lithium salts, Investigations have shown that of
the cations of concern in the ANP fuel program,

 

M, C. White and G. Goldberg, Application of the
Volbard Titration to the 2-Ethyl-1-Hexanol Separation
Method for the Determination of Lithium, ORNL-1827
{to be published).

13G. Wittig et al., Ann. 563, 11026 (1949).
only nickel forms a slightly soluble precipitate
with the reagent, Those elements that hydrolyze
in the acid solution, such as zirconium, beryllium,
and uranium, are held in solution by complexing
with fluoride or citrate ions (for hexavalent
uranium). The method can also be applied in
sulfate solutions and thus offers a distinct advan-
tage over the perchlorate method'¢ for deter-
mining potassium. The standard deviation of the
method is approximately 0.5%.

X-RAY SPECTROMETER INVESTIGATIONS
OF FLUORIDE FUEL

G. D. White, Metallurgy Division
T. N. McVay, Consuyltant

In addition to routine petrographic examination
of fuel samples, several samples were x-rayed on
the high-temperature x-ray spectrometer.  This
work was inaugurated as an additional aid in the
determination of phase diagrams, To date, the
composition 2NaF-ZrF, has been x-rayed at
temperatures from 400 to 600°C. The samples
were held by a nickel holder and were heated in
a vacuum of from 1 to 2 microns or in a purified
helium atmosphere. The atmospheres achieved so
far cause slight oxidation which becomes excessive
only after heating for over 4 hr. High-temperature
x-ray patterns have been obtained on two of the
five polymorphs of Na,ZrF .

 

wT. R. Phillips, ORNL Master Analytical Manual,
Method No. 9 00716450, ORNL CF-53-1-235, Vol. |,

PERIOD ENDING MARCH 10, 1955

ANP SERVICE LABORATORY

J. G, White W. F. Yaughan
C. R. Williams
Analytical Chemistry Division

The determination of beryllium in fluoride fuel
mixtures was resumed during this quarter. The
mixture of fluoride salts is dissolved in H,50,
and the solution is heated until copious fumes of
30, are evident. This process of fuming is re-
peated in order to remove the fluotide completely
from the solution. Uranium as UQ,50, is re-
moved from the solution by sorption on an anion-
exchange resin. Beryllium is then determined by
the modified volumetric method!3® of McClure and
Banks., The major portion of the work continues to
be analyses of fluoride salts, mixtures of fluoride
salts, and alkali metal hydroxides.

The total of 1110 samples analyzed involved
7905 determinations. The backlog at the end of the
quarter consisted of 335 samples. A breakdown
of the work load is given in Table 10.1.

TABLE 10.1. SUMMARY OF SERVICE
ANALYSES REPORTED

 

 

 

Number Number
of of
Samples Determinations
Reactor Chemistry 793 5642
Experimental Engineering 304 2172
Miscellaneous 13 91
Total 1110 7905

 

163
ANP PROJECT PROGRESS REPORT

11. RECOVERY AND REPROCESSING OF REACTOR FUEL

D. E. Ferguson

M. R. Bennett
G. I. Cathers

J. T. Long
R. P. Milford

S. H., Stainker
Chemical Technology Division

PILOT PLANT DESIGN

Design of the pilot plant to recover, in seven
batches, the 65 kg of U?3% in the 2500 Ib of ARE
fuel by a fused salt—fluoride volatility process
is in progress. Completion of construction by
December 31, 1955, is planned, The January
1955 flow sheet calls for a ninefold excess of
fluotine to be passed through the molten fuel at
650°C. The UF, and volatile fission-product
fluorides formed will pass from the fluorinator
into a bed of 20- to 40-mesh NaF at 650°C, where
the volatile fission-product fluorides will be
absorbed, The UF, will collect in a series of
three cold traps ot +4, —40, and ~60°C, respec-
tively, |f sufficient decontamination has occurred,
the traps will be isolated and heated electrically
in order to liquefy the UF,, which can then be
drawn off into receivers, [f further decontami-
nation is required, the UF, will be volatilized
out of the cold traps into another system con-
sisting of an NaF absortber and three cold traps
identical with those mentioned. An aqueous KOH
scrubbing system will be required for disposing
of excess fluorine. A schematic flow sheet for
the plant is shown in Fig. 11.1. The preliminary
cost estimate for the plant, including a 20%
contingency factor, is $285,000,

PROCESS DEVELOPMENT

Laboratory-scale studies were made on the
efficiency of fluorine usage and the corrosion of
the reaction vessel in the fluorination step. During
the first few minutes of the reaction, fluorine
was absorbed completely but no UF ; was evolved.
Fluorine continued to be absorbed completely
until 80% of the UF6 had been volatilized off;
after this, fluorine was detectable in the exit gas.
The data indicated that the UF, was essentially
all evolved by the time a fivefold excess of
fluorine had been used. Therefore the ninefold
excess shown in the flow sheet is believed to
provide a sufficiently large safety factor. The

164

average cortosion rate of nickel during the whole
fluorination step was about 0.1 mph, but the rate
appeared to be considerably higher than the
average during the time that UF6 was present.

Two methods were used in studying the reaction
kinetics of the direct fluorination of ARE-type
fuel. The first consisted in passing the exit gas
from the fluorination vessel through two absorbent
traps filled, respectively, with NaF and NaCl.
These wete weighed before and after each ex-
periment., The NaF trap, held at about 100°C,
was capable of trapping an equal weight of UF
by absorption. The NaCl frap lost weight as the
NaCl was converted to NaF by the fluorine.
Weighing these traps at 5-min intervals provided
a means of following the progress of the fluo-
rination, The second method, Fig. 11.2, con-
sisted in using three flowmeters! to measure the
fluorine input flow rate, the combined UF, and
fluorine output flow rate, and the fluorine output
flow rate after all UF, in the gas stream had been
absorbed. Kinetic data obtained by this method
were more precise than those cbtained by the first
method, since a Brown 5-mv recorder was used to
obtain flow readings at 20-sec intervals,

By means of the NaF-NaCl absorptionstrap
method, calibration curves for UF,, N,, and F,
were prepared which show the flow in cubic
minute corresponding to the

voltage readings obtained in the flowmeter method.
The procedure usually used with both techniques

centimeters per

was to fluorinate, in a l-in,-dia nickel reactor,
70 g of ARE-type fuel (NaF-ZrF ,-UF,, 52-44-4
mole %) at 650°C with a fluorine input rate of
about 60 ml/min,

The variation in flow rate with timein one run in
which thermal flowmeters were used is shown in

 

'K. 0. Johnson, W. F. Peed, and G. H. Clewett, A
Thermal Type Flow Meter jor Low Flow Rates of An-
bydrous Hydrofluoric Acid, C-5.355.8 (June 24, 1946).
PERIOD ENDING MARCH 10, 1955

5001R

     

QOOo00000DO000CO0QO0O0QQ 0

 

 

 

 

 

 

 

 

 

 

 

   

 

         

B B

 

 

   
   

=
P2
ot
a
i_L o
= T
© 2
3 =
3 o
5 oy
3 >
g oy
2 a
i <1
g
T
9
Lt
5 3
T
&
o]
=
-
<I
o
-
-
e
e
o
[+
o °
Sz -
85 Q.
y &ao >
|'1TLan B
O |}
2 >
ity o
S v
fo 0
2 23 o
8y : T
g8 ] © 2
= TR Iz w
oo
. 223 [
% W W W%M_ <
mv_ ANH %N.u n\un_umm -
o S @ a5 N L
4w > oY o w
oo w <~ S = e
CF b, £ o
of - STy @
- T T ey ye? =
B
2
2
|19
u
-
o
£
@
£
v
v
L]
—
—
—
-
2
i,

 

 

T
5 .
o [ad]
=
ML 03.‘,60%
ol O < 0=
> =zfy e " o
@mm f///////////////////m w
EHE |
<1
cgs [ A T
5% W
1 /] .
1 2 pEaaa W
7 TR
z
=9
=C
—_ 5
<4
< o3
£z
=
\/ -
£\

 

 

 

 

165
ANP PROJECT PROGRESS REPORT

 

 

 

 

 

 

 

 

 

 

 

v
ORNL-LR-DWG 5609

 

 

 

 

 

 

 

 

 

 

 

F, F, F, + UFg Rrufg [ ] R F 5
——=| FLOWMETER »|  REACTOR = FLOWMETER w{ NGF ABSORBER [—pm| FLOWMETER |——m OUTUET
Fig. 11.2. Schematic Diagram of Flowmeter Method of Measuring Gas Flow Rates in Fused Salt-

Fluoride Volatility Process.

Fig. 11.3a. |n this run three curves were obtained
from the three flowmeters. The first curve shows
the fluorine input rate, which remained practically
constant during the entire run, The second curve
gives the flow of gas from the reactor to the NaF
trap; the UF, calibration curve was used to in-
terpret this flow. However, the first part of this
curve is somewhat in error, since the system
originally contained nitrogen, and nitrogen or a
mixture of nitrogen and UF_ initially passed
through the second flowmeter. The third curve
gives the output flow from the NaF trap. The gas
passing through this meter was initially nitrogen;
it was fluorine after 31 min, at which time nitrogen
displacement was complete and a chemical test
for fluorine was first obtained. The hump in the
nitrogen part of the third flowmeter curve results
from displacement of nitrogen gas by UF, from
the apparatus between the reactor and flowmeter.
The initial flow of about 5 ml/min was probably
the result of some inert impurity in the fluorine
supply.

The flowmeter readings are interpreted in Fig.
11.35 to show the UF, and fluorine evolution as a
function of time. During the first 14 min, fluorine
was completely absorbed. Material balance calcu-
lations indicated that only about one half this
amount of fluorine could be attributed to the
reaction

UF, + F,—> UF, .

The remainder is assumed to have been utilized
in corrosion. At the end of 14 min, UF6 evolution
began and rapidly increased to 35 to 40 ml/min
after about 20 min had elapsed. The UF ; evolution
remained essentially constant over the next 10min
and began to drop off simultaneously with the
breakthrough of flucrine, which occurred after
31 min,

The fluorine absorption was essentially com-
plete until 85% of the UF, had been evolved.
Based on the amount of fluctine absorbed during

166

this period, an average corresion rate of 4 mph
was calculated,

Qualitative work reported previously? on the
solubility of UF, in a NaF-ZrF, mixture had
suggested the existence of a stable NaF-UF6
complex that is soluble in molten NaF-ZrF,.
The induction period and the subsequent plateau
in UF  production shown in Fig. 11.3a2 and b are
also indicative of this, The reaction mechanism
is believed to be:

UF, (soln in NaF-ZrF,) + F, (g)—>
chlI:-UF6 (soln in NaF-ZrF ) —> UF6(g) .

The induction period is the result of both fluorine
consumption in corrosion and of a buildup in the
concentration of the NaF-UF . complex until the
saturation solubility is reached, Then UF, begins
to evolve as fast as fluorine is supplied in excess
of the rate of utilization in corrosion, The pos-
sibility that oxides dissolved in the fused salt
could account for the delay in UF, generation
wos discounted by the results of a run in which
the ARE fuel was sparged with HF for 45 min
and then with nitrogen for 10 min. Any oxides
present would have been eliminated by this treat-
ment, but the induction period was the same as
that in the run plotted in Fig. 11,32 and 5,

The sodium-to-zirconium atom ratioc in ARE-
type fuel (4 mole % UF,) is about 5:4. Addition
of more ZrF, to change this ratic to 4:5 had
practically no effect on the curves shown in
Fig. 11.32 and 4. This suggests that the NaF-ZrF,
complex is relatively weak in comparison to the
product of interaction between NaF and UF,, The
main effect of a change in the sodium-to-zirconium
ratio is on the melting point of the salt mixture,

The corrosion work to date has consisted mainly
of gravimetric tests on metal coupons and deter-
minations of nickel in the fused salt residues

 

2p. E. Ferguson et al,, ANP Quar. Prog. Rep. Dec.
10, 1954, ORNL-1816, p 134.
70

&80

W b wm
o O o

FLOW RATE {ml/min, STP)

n
O

70

60

50

40

30

FLOW RATE {ml/min, STP)

20

PERIOD ENDING MARCH 10, 1955

ORNL-LR—DWG 5610

 

s~

- FLOWMETER (F2 FLOW TO REACTOR)i

 

A A—A_A

 

 

 

VESSEL . NICKEL
——— TEMPERATURE: 650°C

|
|

FUEL: 70 g of NaF-ZrF,—UF, (52-44-4 mole %)

FLOWMETER 2 (UF6 FLOW FROM REACTOR)

 

 

 

 

 

 

FLOWMETER 3 —__
(N2 DISPLACEMENT BY
UFg AND F, CUTPUT )

.

®
./
o’o

 

o0

®
/

/
°

 

 

 

50
TIME {min)

 

 

 

 

 

 

 

/

 

 

 

 

 

 

 

 

20 30

TIME {min)

Fig. 11.3. (o) Gos Flow Characteristics in ARE-Type Fuel Fluorination Process. (b) Kinetics of
Fuel Fluorination. Dotted lines indicate uncertainty resulting from mixing of gas with nitrogen.

167
ANP PROJECT PROGRESS REPORT

from fluorinatiem studies (Table 1L1). In the
gravimetric tests the corrosion rate was generally
less than 0.1 mph for nickel and somewhat higher
for Inconel and Monel.

The cortosion rate was higher in the runs in
which UF, was present than in the others. In
these five runs there was a UF ,-te-UF, con-
version period of less than 1 hr, followed by a
period of several hours during which fluorine was
passed through the salt. That the average cor-
rosion rate was, in general, greater in the short
runs is believed to be the result of the very high
corrosion during the time that UF, was present.

The lower average rate obtained in long runs is
due to less corrosion occurring after the vranium
has all been evolved.

The corrosion rates determined from salt anal-
yses were obtained in five fluorination runs
made with simulated ARE fuel. The test periods
were probably greater than specified, since the
time spent in starting or stopping each experiment
is not accurately known., The internal surface
area of the nickel reactor used in each run was
about 66 cm?  The greatest corrosion rate,
0.66 mph, was again obtained in the run of
shorte st duration,

TABLE 11.1. CORROSION IN ARE-TYPE FUEL FUSED SALT-FLUORIDE VOLATILITY PROCESS AT 650°C

Gravimetric studies: corrosion coupons were cut from round rod, weighed, immersed for time shown in fused salt

through which fluorine was being passed, and reweighed

Salt analyses studies: fused salt residues from fluorination studies carried out in nickel reactors were analyzed

for nickel

 

 

 

Sale* Test Period Corrosion Rate {mph)
(hr) Ni Inconel Mone!
Gravimetric Method
Nch--ZrF4 (56-44 mole %) 14** 0.031 0.021 0.016
14** 0.36 0.40 0.76
Na F-ZrF4-UF4 (52-44-4 mole %) grxx 0.056 0.063 0.12
5,5%** 0.081 0.017 0.11
6, 5xx* 0.066 0.1 0.15
gxx* 0.077 0.064 0.48
| Rkl 0.37 0.56 0.44
chF-ZrF’4 {56~44 mole %) Jrxx 0.058 0.10 0.10
Salt Analyses
NuF-ZrF4-UF4 (52-44-4 mole %) 1.5 0.66
3 0.33
44 0.06
5 0.25
S 0.21

 

*Two runs were made with the first batch of salt,

run.

In all other experiments a fresh batch of salt was used in each

* % . . ;
Same corrosion samples used in the two successive runs,

%k . . ) .
Same corrosion samples used in the six successive runs.

168
Some information on corrosion can also be ob-
tained from the kinetic data in Fig. 11.3a and 6.
The discrepancy between the maximum UF, flow

rate of 40 ml/min and the flucrine flow rate of -

62 ml/min represents a nickel cortosion rate of
about 1.8 mph. This is perhaps effective only
during the period of existence of the NaF-UF,
complex. However, an even more rapid average
corrosion rate of 4 mph during the period UF,
was present was indicated by the fluorine material
balance. The low corrosion rates obtained in the
gravimetric test and fused salt analyses are
possibly obtained only over the process as a

PERIOD ENDING MARCH 10, 1955

whole, that is, when the UF ,-to-UF, conversion
period is accompanied by several hours of fluorine
sparging fo remove the last traces of vuranium.

In all laboratory fluorination tuns, 50% or more
of the fluorine was consumed in cotrosion of the
nickel reactor. This need not be true in scaling-
up the process to a pilot plant level, It is esti-
mated that the consumption of fluorine in cor-
rosion for a 10-kg uranium batch will be less
than one-tenth that in a batch of 6 g. This is a
consequence of the greater volume-to-surface
ratio in the 10-kg case.

169
Part |l

SHIELDING RESEARCH
12. SHIELDING ANALYSIS
E. P. Blizard

F. H. Murray

C. D. Zerby

Physics Division
S. Auslender
Pratt & Whitney Aircraft

H. E. Stern
Consolidated Vultee Aircraft Corporation

Air-scattering measurements made at the Tower
Shielding Facility indicate that for optimized divided
shields the multiple scattering of radiation in air
is quite important — more so than in many shield
designs considered heretofore. Consequently a
large fraction of the analysis effort has been
pointed toward more complete calculations of air
scattering. Two approaches are used. In one,
certain concessions are made in the physical hy-
potheses, such as assuming constant neutron
velocity, to enable analytical treatment of the
problem. In the other, a stochastic {Monte Carlo)
method is adopted, with as nearly exact data being
used as are available., Neither method has yet
been completely developed or applied.

An offshoot of a Monte Carlo calculation of slant
penetration of gamma rays in crew-shield side walls
has been the development of a similar calculation
of gamma heating in a multilayered region, This
method will be applied to the presently conceived
reactor designs to find the gamma heating to be
expected in regions near the core.

ANISOTROPIC SCATTERING OF NEUTRONS
IN A UNIFORM MEDIUM WITH BEAM SOURCES

F. H. Murray

Formulas useful in the analysis of problems of
multiple scattering of neutrons in an unbounded
medium have been derived. These formulas are
being applied in some problems of scattering in
stratified mediums. The particle velocity is as-
sumed to be constant in the analysis, and the
scattering law is assumed to depend only on the
angle between the initial and final directions of
the scattered particle. _

This analysis is based directly on the Fourier
transform, and the spatial distribution of sources
and the intensity os a function of angle from any
point may be arbitrary. The formulas and their
derivations are presented in a separate report.1

ENERGY ABSORPTION RESULTING FROM
INCIDENT GAMMA RADIATION AS A FUNCTION
OF THICKNESS OF MATERIALS
WITH SLAB GEOMETRY

C. D. Zerby  S. Auslender

Initial calculations of the heat generation in
berytllium slabs as a result of the transport of
slant incident gamma radiation were described in
a previous report,2 The method used for these
initial calculations is being extended to include
photons with energy up to approximately 10 Mev,

The coded caiculation for the Oracle, from which
the initial results were obtained, is also being
revised to make it more versatile. It will be pos-
sible in the future to study the heat generation
resulting from the transport of gamma radiation
through homogeneous slabs of materials containing
any number of elements. The code for the calcu-
lation is being designed for simplicity of operation
so that cases of interest can be calculated with a
minimum of effort and time. In particular it is
anticipated that the method will be applicable to
caleculations of heat deposition in the laminations
around the reflector-moderated reactor.

ENERGY AND ANGULAR DISTRIBUTION OF
AIR-SCATTERED NEUTRONS FROM A
MONOENERGETIC, MONODIRECTIONAL
POINT SOURCE

C. D. Zerby

A calculation of the energy and angular distribu-
tion of air-scattered neutrons from a monoenergetic,
monodirectional point source is being made to pro-
vide data from which the energy and direction of
the neutron flux at an aircraft crew compartment

 

1F. H. Murray, Anisotropic Scattering of Neutrons in a
Uniform Medium with Beam Sources, ORNL CF-54-11-83
{to be published).

2c. b. Zerby, ANP Quar, Prog. Rep. Dec. 10, 1954,
ORNL-1816, p 144,

173
ANP PROJECT PROGRESS REPORT

shield resuvlting from air-scattered neutrons from an
arbitrary source can be determined at any altitude.
The Monte Carlo method is being used for this
calculation, and it is presently being coded for the
Oracle,

The cross sections being used were taken from a
previous report® and include the complete resonance

 

3Neutron Cross Sections, AECU-2040 {(May 15, 1952).

174

structure, as reported. The scattering will initially
be taken as isotropic in the center-of-mass system;
however, later, it is intended that the anisotropy,
as determined from experiments or calculations,
will be included. For comparison with approximate
solutions to this problem, the energy and angular
distribution of single, double, triple, and all other
multiply scattered neutrons will be
separately.

recorded
PERIOD ENDING MARCH 10, 1955

13. LID TANK SHIELDING FACILITY
G. T. Chapman

J. M. Miller

D. K. Trubey'

Physics Division

J. B. Dee
W, J, McCool

H. C. Woodsum

J. Smolen

Pratt & Whitney Aircraft

The GE-ANP helical air duct experimentation
has been completed, and an analysis of the data is
presented, A new attempt to correlate the removal
cross section data with other properties of the atom
has resulted in a plot of the ratio of the macro-
scopic removal cross section to the density of the
material against the atomic weight., These data
are compared with the total cross section at 8
Mev. Final preparations for the second series of
tests on the reflector-moderated reactor and shield
mockup have been completed.

GE-ANP HELICAL
AIR DUCT EXPERIMENTATION

J. M. Miller

All experimental work has been completed at the
Lid Tank Shielding Facility (LTSF) on the GE-
ANP helical air ducts and a final report on the
work is being written, As described previously,?
the ducts were fabricated by shaping 3-in.-ID
flexible tubing around a 9-in. core. After removal
of the core, the ducts were stiffened with Fiberglas
wrapping. The projected length of each duct along
the z axis was 46.5 in., and the duct arrays were
arranged so that there was 5 in. between the duct
center lines, '

Radiation measurements were made beyond a
single duct, a three-duct array, and a 35-duct array
in plain water. Measurements were also mode
beyond the 35-duct array in a medium of Raschig
rings (hollow steel cylinders, ]/4 in. in diamefer
and ]/2 in. long) and borated water, The Raschig
rings were packed so that the medium around the
ducts was 33% steel and 67% borated water. Various
thicknesses of lead were also placed alongside
the 35-duct array in the Raschig ring—borated water
medium to mock up a side shield for the ducts,
Some of the results of the experimental work are
given in Table 13.1.

 

1On leave to attend ORSORT.

2J. M. Miller, ANP Quar. Prog. Rep., ORNL-1771, p
166; ORNL-1816, p 153.

REMOVAL CROSS SECTIONS
G. T. Chapman

Previous attempts to correlate the effective
neutron-removal cross section data obtained at the
LTSF with more rigorously defined properties of
the atoms, such as atomic weight and total cross
section, have been limited because of the scarcity
of data. However, a plot of the data has been made
that may prove useful in shield calculations. The
plot is presented in Fig. 13.1. The ratic of the
macroscopic removal cross section to the density
of the material (/p, cm?/g), a number proportional
to the removal cross section per nucleon, is plotted
against the atomic weight. These data are com-
pared with the total cross section at 8 Mev, as

2-01-057-0-93

 

———— T T ]
TOTAL CROSS-SECTION VALUE AT 8 Mev  _
{TAKEN FROM REF, 3) Lo
REMOVAL CROSS-SECTION VALUE DERIVED

 

  
 
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

— FROM MEASUREMENTS BEHIND COMPOUND 1T}
2| OF THE ELEMENT SRR
o FEMOVAL CROSS-SECTION VALUE FROM
PE MEASUREMENTS BEHIND ELEMENT
" £ = REMOVAL CROSS SECTION OR TOTAL
~ 051 CROSS SECTION AS DESIGNATED
S p = DENSITY OF MATERIAL
g ‘ I |
=~ 02 , P ot
o : | i
P N\ @ 1 |
‘ \ o ;] P
————3% i &
vos | T e | | .i
' 00852 | 4 54 o |
N Vi o :‘?j? : . +
A 3 T - !O T
002 |—— e S~k Co .
I 'Q-ng\‘“ S
ool | 1 L Mo T
; | S| I 1
i - 1 1
0.005 = -
12 5 40 20 50 400 200 500 4000
A, ATOMIC WEIGHT
Fig. 13.1. Neutron Shielding Ability per Unit

Weight of a Material as o Function of Atomic

Weight.

175
ANP PROJECT PROGRESS REPORT

TABLE 13.1, INTENSITY INCREASES RESULTING FROM THE REDUCED
DENSITY EFFECT AND THE PRESENCE OF G.E HELICAL AIR DUCTS

 

Calculated Factor Total Factor

Increase in lncrease in

 

Duct Non-Shield
'UC Shield* Quantity Measured en-snte Intensity Resulting Intensity Resulting
Configuration Component*
from Reduced from Presence
Density Effect of Ducts
Single duct Plain water Thermal-neutron flux ~2
Gamma-ray dose rate ~1
Fast-neutron dose rate o Qrk
Three ducts Plain water Thermal-neutron flux ~6
Gamma-ray dose rate ~ 1.4
Fast-neutron dose rate ~4
35 duets Plain water Thermal-neutron flux 43.5% air 210 ~ 480
Gamma-ray dose rate 49% ducts 2, 1%%* ~2.4
Fast-neutron dose rate ~ 200 ~ 400
Raschig rings in  Thermal-neutron flux 43.5% air 770 4000
borated water Gamma-ray dose rate 49% ducts 175 ~160

 

*Does not include water thickness between end of duct array and center of detection.

**Poor statistics.

***Corrected for iron in ducts.

obtained by N. G. Nereson et al.® at the Los Alamos
Scientific Laboratory,

All removal cross section measurements at the
LTSF have been compiled and will be published in
a separate repor’r.4

REFLECTOR-MODERATED
REACTOR AND SHIELD MOCKUP TESTS

J. B. Dee

The second series of experiments® with mockups
of the circulating-fuel reflector-moderated reactor
(RMR) and shield has been started ot the LTSF,
The final preparations involved some changes from
the plans previously published.® The 1I/B-in.-i'hiczl-c
Inconel window originally built into the source side

 

N G. Nereson et al, Survey of Hverage Neutron
Total Cross Sections from 3 to 13 Mev, LA-1655
(July 13, 1954).

‘. T. Chapman and C. L.. Storrs, Effective Neutron
Removal Cross Sections for Shielding, ORNL-1843 (to
be published).

sFor first series see C. L. Storrs et al., ANP Quar
Prog. Rep. Sept, 10, 1953, ORNL-1609, p 128.

6). B. Dee et al., ANP Quar. Prog. Rep., ORNL-1771,
p 164; ORNL-1816, p 155,

176

of the large tank for holding all the dry components
of the configurations has been replaced with a
% _in.-thick aluminum window. With the removal of
the lIncone!l and, consequently, the high-energy
nickel capture gamma ray, a more accurate siud),r
can be made of the smaller effects resulting from in-
elastic scattering and capture gammas in the other
components of the shield mockup. As originally
planned, Inconel will be used to simulate the RMR
core shell, but it will be in the form of a removable
slab placed immediately behind the aluminum
window of the dry tank., An advontage of this ar-
rangement is that it will be possible to place the
[nconel inside the fuel belt during the dynamic
tests and thus provide a more realistic mockup for
this series of tests,

The solid Al-UO, plates, which will simulate the
fuel in the dynamic fission-source tests, were
mounted in @ continuous series on a sprocket-driven
chain-link belt and tested under operating condi-
tions in a special rig outside the LTSF (Fig. 13.2).
The belt, which will move continuously from the
neutron window to the heat exchanger region and
back to the neutron window, was successfully oper-
 

 

PERIOD ENDING MARCH 10, 1955

UNCLASSIFIED
PHOTO 13853

Fig. 13.2. Apparatus for Testing the Chain-Driven Fuel Elements for the RMR-Shield Mockup.

ated at speeds up to 1050 fpm, which corresponds
to a transit time of 0.73 sec in an actual mockup;
the maximum design speed for the mockup is 900
fpm or a transit time of (.83 sec.

As mentioned previously,® this second series
of RMR experiments will consist of two sets of
tests: the static-source tests and the dynamic-
source tests. In the static tests, the first measure-
ments of which are now being made, the primary
and secondary sources of radiation that reach the
outside of the shield will be determined. This
will be accomplished by analysis of differences
in dose curves obtained by varying the thicknesses
of the mockup regions, as follows: (1) an Inconel
core shell % in. thick, (2) a beryllium reflector 8
to 16 in. thick, (3) a boron curtain consisting of
2!{‘ in. of boral, (4) a heat exchanger consisting of
one to four ]1/2-in. slabs of NaF contained in nickel,
(5) a second boron curtain, (6) a pressure shell
consisting of 0 to 2 in. of nickel, and (7) a shield

* termined,

made up of lead, 0 to 10]/2 in. thick, and water,
usually borated. In some tests, bismuth and copper
will be substituted for part of the beryllium re-
flector for determining the effect of placing a
gamma shield closer to the reactor. These tests
will not include radiations resulting from fuel
circulation, but the sodium activation within the
heat exchanger region will be studied.

For the dynamic tests a typical mockup will be
used, primarily, and the LTSF source place will
be replaced with the continuous series of Al-UO,
plates mounted on a movable belt. Several transit
times will be used — the minimum will be 0.66
sec — and the sodium activation in the heat ex-
changer caused by delayed neutrons will be de-
The fuel belt will also give a source
of fresh fission products. The attenuation of the
gamma rays resulting from these fission products
by various thicknesses of lead will be studied.

177
ANP PROJECT PROGRESS REPORT

14. TOWER SHIELDING FACILITY

C. E. Clifford
T. V. Blosser J. L. Hull
L. B. Holland F. N, Watson

Physics Division

D. L. Gilliland, General Electric Company
M. F. Valerino, NACA, Cleveland

J. Yan Hoomissen, Boeing Airplane Company

The first experiment with the mockup of the
GE-ANP R-1 reactor shield has been completed,
and the first series of differential experiments has
been started. A specially designed gamma-ray
dosimeter with which it is possible to simulate
the addition of small thicknesses of lead to an
aircraft crew compartment has been calibrated.
Some results of the exposure of primates to high
fast-neutron dose rates in a program initiated by
the U.S. Air Force are reported.

TSF EXPERIMENT WITH THE MOCKUP OF
THE GE-ANP R-1 SHIELD DESIGN

T.V. Blosser M. F. Yalerino
D. L. Gilliland J. Van Hoomissen
F. N. Watson

The first experiment with the mockup of the
GE-ANP R-1 reactor shield has been completed
at the TSF with measurements of the gamma-ray
dose rates along the x, y, and z axes of the
detector tank. The experimental arrangement for
these measurements was the same as that for the
earlier thermal-neutron flux measurements.! The
reactor-detector altitude (b) was 195 ft and the
G-E shield was a horizontal distance (d) of 64 ft
from the detector tank. Five 1-in.-thick lead slabs
were installed 1 ft from the rear face (reactor side)
of the detector tank to simulate the shielding in
the crew compartment.

In order to gain a knowledge of the dose rates
from gamma rays which will be incident on the
various faces of the crew shield mockup and of
the gamma-ray attenuation characteristics of the
hydrogenous shielding of the mockup, the gamma-
ray dose rates were measured with both plain and
borated water in the detector tank. From the
borated water traverses the reduction of the

 

]C. E. Clifford es al., ANP Quar, Prog. Rep. Dec. 10,
1954, ORNL-1816, p 158.

178

gamma-ray dose rate caused by the suppression
of the gamma rays resulting from the capture of
thermal neutrons in the water could be determined.

The y traverse (Fig. 14.1) measured in plain
water indicates an average relaxation length of
19.7 e¢m to the rear of the lead. The shape of the
curve in front of the lead indicates that the gamma
dose is due to the gamma rays which enter the
tank from the sides rather than those which
penetrate the [ead. Therefore, an effective relax-
ation length for the lead attenuation cannot be
obtained from these data. The reduction of the
gamma dose in the center of the tank which
resulted from the boration (0.4 wt %) of the water
indicates the presence of substantial quantities
of secondary gammas from captures in hydrogen.
The effective relaxation length at the front of the
tank along the y axis (reactor—detector tank axis)
was found to be 9.6 c¢m and was obtained by
subtracting from the curve the gamma dose pene-
trating the sides of the tank. The relaxation
length obtained at the sides of the tank (Figs.
14.2 and 14.3) in a similar manner was 10.3 cm
and was apparently a single exponential for pene-
tration of at least 25 cm of water. An error in the
calibration of the gamma-ray detector caused by
the neglect of the decay of the cobalt source
introduces a correction factor of 0.94 for all
gamma-ray dose rates measured in the detector
tank during this experiment.

THE DIFFERENTIAL EXPERIMENTS AT THE

TSF: PHASE |
T.V. Blosser M. F. Valerino
L. B, Holland J. Van Hoomissen
J. L. Hull F. N, Watson

The experimental program at the TSF is now
being arranged so that the mockup experiments
are interspersed with differential experiments
(those in which the 12-ft-dia reactor tank and the
2-01-056-3—~18+24-82

h =195 ft
d =64 ft

———— i -

CENTER OF SCINTILLATION CRYSTAL AT:

x£=0cm _T ' ' ‘
y=VARIABLE | |

Z=0cm

PLAIN WATER

GAMMA —RAY DOSE RATE (mr/hr/watt)
MULTIPLY ORDINATE BY 0.94

.
BORATED WATER

 

0 12 24 36 48 60 72 84 96 108 120 132 144
¥, DISTANCE FROM REAR FACE OF DETECTOR TANK TO DETECTOR CENTER {(cm)

Fig. 14.1. Gamma-Ray Dose Rate Along the y Axis of the Detector Tank.

641

 

10
156

104

-7

GAMMA —RAY DOSE RATE (mr/hr/watt)

S561 ‘0L HOYVW ONIONZ QOId3d
081

GAMMA—RAY DOSE RATE (mr/hr/watt)

<
0
O
)—
m
[T}
-
<l
=
O
o
c
3
o
5
2
=

—78

2-01-056—-3—-194+25-83

A =195 ft
d =64t

CENTER OF SCINTILLATION CRYSTAL AT:

X = VARIABLE
¥=70cm
Z2=0cm

PLAIN WATER

BORATED WATER

—66 —54 —42 —30 —18 -6 QO 6 18 30 42 54
x, HORIZONTAL DISTANCE FROM ¢ AXIS TO DETECTOR CENTER {cm)

Fig. 14.2. Gammu-R'uy Dose Rate Along the x Axis of the Detector Tdnk,

.

66

 

78

LId0dIY SSIAD08d LDIr0¥d dNV
10=4

195 ft
= &4 f¢
CENTER OF SCINTILLATION DETECTOR AT:
2 0.0cm --
7O cm

= VARIABLE
10-°

PLAIN WATER

<
”
o
ol
0
L
b=
<
=
o
o
o
5> 1
—
o
~
—J
D
=

GAMMA DOSE RATE (mr/hr/watt)

1077
-78 -66 -54 —42 -30 -8 -6 0 6 18 30 42 54

z, VERTICAL DISTANCE FROM & AXIS TO DETECTOR CENTER (cm)

Fig. 14.3. Gamma-Ray Dose Rate Along the z Axis of the Detector Tank.

18t

2-01—-056-3-21-88

66

 

78

$S61 ‘01 HOYVYW 9NIAON3I @old3d
ANP PROJECT PROGRESS REPORT

detector tank are used). A series? of differential
experiments has now been started with emphasis
on obtaining the fast-neutron dose rate distribution
within the detector tank as a function of 6, the
angle between the axis of symmetry of the beam
from the reactor tank and the source-detector
axis.  The thickness (p) of the water layer
shielding the reactor, as measured from the edge
of the reactor tank, is being held constant at
45 em. This thickness was chosen because it
is in the region of interest of the side shielding
on an gircraft reactor, and it also allows sufficient
intensity in the detector tank for accurate measure-
ments. Further, the prediction of the dose within
the detector tank for a variation of p is felt to be
less uncertain than for the variation of 4.

The quantities to be determined are the magni-
tude of the doses impinging on the front, side and
rear of the detector tank and the rate at which
these doses are attenuated. It is hoped that the
rate can be expressed in terms of the relaxation
length of a simple exponential function. The
results of these measurements will be compared
with single-scatter calculations which are now
being coded for solution on the Oracle. The
calculations are being devised so that the doses
arriving at the various faces of the tank can be
calculated separately, and, for the 0- and 90-deg
case, the angular distribution of the dose arriving
at the faces will also be calculated. The calcu-
lations will be sufficiently general that any
angular distribution at the source may be used.

The measurements which have been made to
date have been planned as a quick survey to
indicate the most interesting regions for the more
detailed measurements to follow.

CALIBRATION OF THE REVALET, A
REMOTELY VARIABLE LEAD-TRANSMISSION
GAMMA-RAY DOSIMETER

D. L. Gilliland

The optimization of gamma-ray shielding around
the crew compartment is one of the major problems
in the design of nuclear aircraft shields. To
permit optimization the attenuation of lead for
the gammg-ray dose penetrating the crew com-
partment must be determined as a function of the
lead disposition within the crew compartment. A

 

250me preliminary differential experiments were re-
ported previously; C. E. Clifford e al., ANP Quar. Prog.
Rep. Sept. 10, 1954, ORNL-1771, p 175.

182

full-scale experiment to obtain this information
by varying the lead thickness inside an actual
crew shield mockup would be both ‘difficult and
time consuming; therefore, a method has been
developed in which it is hoped the crew com-
partment gamma shield can be simulated with lead
thicknesses that can be varied with ease. This
has been accomplished by enclosing an gnthracene
scintillation counter in a thick lead shield which
has an aperture in one side that can be covered
with lead disks of varying thicknesses from 0 to
0.7 in. Some of the details of the instrument,
known as the Revalet, are shown in Fig. 14.4.

In order to ensure the validity of the measure-
ments to be made at the TSF with the Revalet,
an experiment with a known geometry and a source
of known energy (Co®%) was devised in which the
angle of incidence of photons on the lead absorber
disk could be varied. The measured lead attenu-
ation for the various angles of incidence (0 to
60 deg) was in good agreement with Monte Carlo
calculations of slant penetration.® A comparison
of the experimental measurements with the Monte
Carlo calculation is shown in Fig. 14.5. The
sensitivity of the counter as a function of angle
of incidence is shown in Fig. 14.6 for the open-
aperture case. The details of the instrument are
given in a separate repori’.4

THE PROJECT ORANGE PRIMATE EXPOSURE

AT THE TSF
T.V. Blosser L. B. Holland
D. L. Gilliland J. L. Hull

F. N. Watson

A preliminary experiment® in which primates
were exposed to mgssive doses of radiation has
been completed at the TSF as part of a program
initiated by the U.S, Air Force. Prior to these
measurements the only experimental data available
had been obtained with massive gamma-ray doses
from barium-lanthanum sources. A more realistic
approximation of the doses that would be received
from an atomic warhead would, of course, include
neutron as well as gamma-ray doses. It would

 

3c. b Zerby, Preliminary Report on the Penetration
of Composite Slabs by Slant Incident Gamma Radiation,

ORNL CF-54-9-120 (Sept. 21, 1954).

AD. L. Gilliland, Calibration of the Revalet, a Re-
motely Variable Lead-Transmission Gamma-Ray Dosime-
ter, ORNL CF-55-2-111 (to be published).

SA future experiment involving several hundred animals
is planned.
Lt 7T s

   

FRONT VIEW

== LEAD DISK IN ABSORPTION WHEEL
{0.1-in. GRADATION IN THICKNESS) ¥

Y, X Y4-in. ANTHRACENE CRYSTAL

PERIOD ENDING MARCH 10, 1955

2-01{-056-8-D1C4

GENEVA-LOC ACTUATOR TO
DRIVE LEAD-ABSORPTION WHEEL\\

 

\ )
X oo W

‘ L

| by

—COUNTERBALANCE

 

N ~POTENTIOMETER
. ! PREAMPLIFIER
Ll

| ~

 

PN Y, % Y, ~in. ANTHRACENE
N - 12 4 :

 

 

 

 

 

 

 

 

 

 

 

P CRYSTAL
& LEAD SHIELD
|
:..,‘ _ N“ LRI - .
; = A
DUMONT 6292 PHOTOMULTIPLIER
TUBE

LUCITE LIGHT PIPE

SIDE VIEW

Fig. 14.4. The Revalet, a Remotely Variable Lead-Transmission Gamma-Ray Dosimeter.

2~01-056-8-3~105

| !
" ¢ CALCULATED VALUE, WITH STANDARD
CEVIATION INDICATED

FRACTION OF INCIDENT ENERGY PENETRATING SLAB

EXPERIMENTAL ANGLE OF INCIDENCE
| 0 Odeg
30 deg

a
A 60 deg
A 75 deg
0 eXi 0.2 0.3 0.4 0.5 0.6 0.7
LEAD THICKNESS (in.}

 

Fig. 14.5. Lead-Absorption Measurements with
the Revalet,

also include a more diverse spectrum of the
gamma-ray energies. Since high radiation levels
exist outside the shielding of the TSF reactor,
it was chosen as a convenient source for further
experiments. This preliminary experiment, desig-
nated Project ORANGE by the USAF, involved

the exposure of 25 rhesus monkeys.

Since little was known of the effect of fast-
neutron doses, emphasis was placed on maxi-
mizing the fast-neutron dose rate and minimizing
the gamma-ray dose contamination. The maximum
neutron dose achieved was 30,000 rep for an
interval of approximately 1Y% min, the gamma-ray
contamination being approximately 1 r/7.5 rep.

The experimental arrangement for the exposures
is shown in Figs. 14.7 and 14.8. Two monkeys
were exposed simultaneously in the two cy-
lindrical Lucite cages shown in Fig. 14.7. The
cages were placed on a motorized turntable which
was rotated at 1 rpm, and they were separated
from the reactor tank by a 3-in.-thick lead shield
and @ ¥ -in.-thick boron-impregnated plastic sheet.
The horizontal midplane of the reactor coincided
with the bottom edge of the white stripe around
the reactor tank, shown in Fig. 14.7. The front
face of the reactor was 5.5 cm from the tank wall.
(Before the exposures were actually made the pine
two-by-four running across the top half of the

183
ANP PROJECT PROGRESS REPORT

g ST

S
2-01-056~8-5-106

ANGLE OF INOC!DENCE (deg)

30

OPEN APERTURE CASE

B

90
3 2 4

 

 

60

90

INTENSITY {(arbitrary units)

Fig. 14.6. Solid Angle of Detection of the Revalet.

cages, Fig. 14.7, was removed. The three boron
carbide stabs in front of, behind, and under the
turntable were also removed.)

Measurements of the fast-neutron doses were
made both with chemical dosimeters and with a
Hurst-type fast-neutron dosimeter. The chemical
dosimeters were attached to the outside of the
cages and on the monkeys. In some instances
cylindrical polyethylene containers filled with a
solution of sugar, water, and urea to simulate the
body composition of a monkey were used as
phantoms in the cages and chemical dosimeters

184

were attached to the phantoms. The Hurst do-
simeter measurements were made inside the cages
with and without the phantoms.,

The gamma-ray dose measurements were also
made with chemical dosimeters placed on the
cages, monkeys, and phanfoms. Measurements
with the 900-cm® ion chamber were made inside
the cages without the phantoms.

The total radiation doses to which the animals
were exposed are given in Table 14.1. The
results of these preliminary exposure studies will

be reported by the USAF,
G8l

Fig.

ST
1-01-056~-20

 

 

 

 

 

 

 

14.7. TSF Experimental Arrangement for Monkey lrradiation Showing Lucite Cages in Position Near Reactor Tank.

§S61 ‘01 HOMVW ONIGNI dol¥3d
ANP PROJECT PROGRESS REPORT

2-01-056-4~-D107

 

 

 

 

AR R

INSIDE SURFACE COVERED
WITH BLOTTING PAPER \

    
  
 

PLASTIC CAGE

’ TSR
|

CAGE HOLDER

 

 

 

 

 

 

 

 

(- 924 in.

 

CAGE ROTATJO_N//
MECHANISM ;

 

WOODEN STAND —— |

 

 

Ya-in- THICK PLYWOOD

 

 

 

 

 

 

 

 

Yo-in-THICK LEAD

  
   
 

SUPPORT CHANNEL

 

1-in.~THICK LEAD

 

1" SUPPCRT FRAME REACTOR TANK

 

 

 

 

7 1 3 L‘* =~—rooL EDcE

A IR NOTE: ALL DIMENSIONS ARE APPROXIMATE

 

 

   

 

 

Fig. 14.8. TSF Experimental Arrangement for Monkey Irradiation (Side View).

186
PERIOD ENDING MARCH 10, 1955

TABLE 14.1. TOTAL RADIATION DOSES TO WHICH MONKEYS WERE EXPOSED

 

 

 

Total Neutron Dose (rep) Tota! Gamma Dose (r)
Exposure NJ;T: of FTIil"’: ot Chemical Chemical
Number ye vl Fower Hurst Dosimeters lon Dosimeters
Exposed (min) Dosimeter _— Chamber B
Cage A Cage B Cage A Cage B
1 2 1.5 31,900 28,846 28,461 4280 7763 7894
2 2 1.5 31,100 , 26,154 27,692 4190 6578 6842
3 2 1.5 31,300 23,077 24,615 4210 7105 7026
5 2 1.5 2,180 2,300 2,153 292 150 263
6 2 1.5 2,240 2,305 2,164 299 180 296
7 2 1.5 12,300 10,000 10,770 1650 2368 2237
8 2 1.5 4,500 5,153 4,615 602 o921 789
9 2 1.5 11,600 10,000 11,540 1600 1973 2000
10 2 1.5 4,400 4,386 4,769 589 723 723
11 2 1.5 11,500 10,000 13,077 1550 1842 1973
12 2 1.5 4,420 3,946 3,846 591 592 592
13 1* 1.5 2,200 2,154 2,154 295 263 263
14 2 2.0 14,700 15,381 16,153 1960 2194 2389

 

*Plus 1 phantom,

187
Part IV

APPENDIX
REPORT NO.

CF-54-9-160
CF-54-11-188

CF-54-12-120

CF-54-12-209

CF-54-9-111

CF-54-10-106

CF-54-11-69

CF-54-12-154

CF:55-2-16

ORNL-1721

ORNL-1835

CF-54-9-24

CF-54-10-97

CF-54-10-119

CF-54-10-168

CF-54-11-33

CF-54-11-150

CF-54-12-21

15.

TITLE OF REPORT

I. Aircraft Reactor Experiment

ARE Instrumentation List

Preliminary Report — Operation of the Aircraft Reactor

Experiment

The Amount of Na22 in the Ne Coolant of the ARE
After Operation at 2 Mw for Two Days

Examination of the ARE

II. Reflector-Moderated Reactor

On Gamma Ray Heating in the Reflector-Moderated

Reactor

Thermal Stresses in Beryllium — Test No, 1

Temperature-Time History and Tube Stress Study of the

Intermediote Heat Exchanger Test

Shield Weights for the CFRE
Preliminary Evaluation of Possible Poisons for Use in
the ART Control Rod

ORNL Aircraft Nuclear Power Plant Designs

Aircraft Reactor Test Hazards Summary Report

), Experimental Engineering

Program for Continuing Study of Cavitation Phenomena

in Sodium

The Design of @ Small Forced Circulation Corrasion

Loop
IV. Critical Experiments

Reflector Moderated Critical Assembly Experimental
Program — Part [I

Experimental Program for Reflector Moderated Critical

Assemblies

The First Assembly of the Three-Region Reflector

Moderated Reactor

The Second Assembly of the Three-Region Reflector

Moderated Reactor

Preliminary Critical Assembly for Supercritical Water
Reactor, Part I

LIST OF REPORTS ISSUED FROMSEPTEMBER 1954 TO MARCH 1955

AUTHOR

Ro Go Aff3|

Jl L- Meem
W. B. Cottrell
H. W. Bertini

W- Dl Monly

P. H. Pitkanen
R- W- BUSSOI’d

R. E. MacPherson
R. In Gl’oy

J. B. Dee, Jr.

H. C. Woodsum

J. W. Noaks

A. P. Fraas

A. W. Savolainen

W. B. Cottrell ez al,

Je Mv Trummel

W. K. Stair

B. L. Greenstreet

B. L. Greenstreet

R. M. Spencer

R. M. Spencer

B. L. Greenstreet

J. W. Noaks
Jo Sc Crude|e

DATE ISSUED

9-20-54
11-30-54

12-23-54

12-30-54

9-14-54

10-25-54

11-30-54

12-20-54

2-2-55

12-3-54

1-19-55

9-3-54

10-11-54

10-19-54

10-29-54

11-5-54

11-24-54

12-1-54

191
ANP PROJECT PROGRESS REPORT

‘REPORT NO.

CF-55-1-123

ORNL-1770

CF-54-9-98

CF-55-1-54

CF-54-10-139

CF-54-10-140

CF-54-11-37

CF-54-11-63

CF-54-12-110

CF-55-2-20

ORNL-1769
ORNL-1777

CF-54-9-36

CF-54-12-65

CF-54-9-119
CF-54-9-120

CF-54-11-3

CF-54-11-113

192

TITLE OF REPORT

Three-Region Reflector Moderated Critical Assembly R.
with ]/“5 ins Inconel Core Shells

Preliminary Critical Assemblies of the Reflector R.
Moderated Reactar

V. Metallurgy

Examination of Sodium, Beryllium, Inconel Pump Loops G.
Numbers 1 and 2 E.
Materials Handbook L

¥I. Heat Transfer ond Physical Properties

Measurement of the Thermal Conductivity of Molten

Fluoride Mixture No. 44

Heat Capacity of Composition No. 40 W.
G.

A Laminar Forced-Convection Solution for Pipes Ducting H.

Liquids Having Volume Heat Sources and Large Radial

Differences in Viscosity

A Method for Evaluating the Heat Transfer Effectiveness M.
H. F.
M.

of Reactor Coolants

Qualitative Velocity Information Regarding the ART
Core: Status Report No. 4

Preliminary Measurements of the Viscosity of Composi-

tion 20

Free-Convection in Fluids Having a Volume Heat Source

AUTHOR

M. Spencer

M. Spencer

M. Adamson
Long

D. Manly

W. D. Powers
S' Ju C|aiborne

D. Powers

Blalock

Peoppendiek

Rosenthal
Poppendiek
Burnett

J. O. Bradfute

S. 1. Cohen
T. N. Jones

Du Cu Hdmilfon el al-

Fused Salt Heat Transfer — Part |I: Forced Convection He W. Hoffman
Heat Transfer in Circular Tubes Containing NaF-KF- J. Lones
LiF Eutectic
V1. Radiation Damage
Removal of Xenon from Fluoride Fuels: Preliminary M. Robinson
Design of In-Pile Equipment
Volatilization of Fission Products from Fluoride Fuels M. Robinson
VIIl. Shielding
The Time Variation for Injury from Radiation E. Blizard
Preliminary Report on the Penetration of Composite Ce Zerby

Slabs by Slant Incident Gamma Radiation

Measurement of an Effective Neutron Removal Cross G.

Section of Lithium at the Lid Tank Shielding Facility

Fraction of Biological Dose Due to Thermal Neutrons in E.

Aircraft Reactor Shields

Chapman et al.

Blizard

DATE ISSUED

1-21-55

11-22-54

9-13-54

1-5-55

10-26-54

10-26-54

11-5-54

11-4-54

12-14-54

2-2-55

11-15-54
2-1-55

9-3-54

12-10-54

9-21-54
9-21-54

11-2-54

11-19-54
REPORT NO.

ORNL-1682

CF-54-10-20
CF-54-10-48
CF-54-10-49
CF-54-10-138
ORNL-1771

ORNL-1816

TITLE OF REPORT

Reactivity Measurements with the Bulk Shielding Reactor
IX. Miscellaneous

ANP Information Meeting of August 18, 1954
ANP Research Conference of September 28, 1954
ANP Research Conference of September 7, 1954
ANP Research Conference of Qctober 26, 1954

Aircraft Nuclear Propulsion Project Quarterly Progress
Report for Period Ending September 10, 1954

Aircraft Nuclear Propulsion Project Quarterly Progress
Report for Period Ending December 10, 1954

PERIOD ENDING MARCH 10, 1955

A.
A,
A.
A.
A.

A.

AUTHOR

G. Cochran et al.

W. Savolainen
W. Savolainen
W. Savolainen
W. Savolainen

W. Savolainen (ed.)

W. Savolainen (ed.)

DATE ISSUED

11-19-54

10-6-54

10-11-54
10-11-54
10-26-54
10-21-54

1-20-55

193
 

THE AIRCRAFT NUCLEAR PROPULSION PROJECT
THE OAK RIDGE NATIONAL LABORATORY

MARCH 1, 1955

 

ANP PROJECT DIRECTOR W. H
CO.DIRECTOR 5. 4. CROMER RD
ASSOCIATE DIRECTOR R. I

Al

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

ASSISTANT DIRECTOR . J. MILLER RD
D. HILYER, SEC. RD
PRATT & WHITHEY AIRCRAET REPORTS
R. I. STROUGH, PROJECT ENGINEER  PWA A. W, SAVOLAINEN ARE
A, GIANGREGORID, ADM. ASS'T.  PWA P. HARMAN, SEC. ARE
0. A. BRADEN, $EC. ARE LITERATURE SEARCHES
H. C. GRAY, DESIGN INEER Pwa
. ENGIKE A. L. DAVIS ARE
ARCRAFT REACTOR ENGINE ERIMG DLYISION
5. J. CROMER, DIRECTOR RD
H. MCFATRIDGE, SEC. ARE SUPPORTING RESEARCH
W. H. JORDAN
ASSISTANT 70 DIRECTOR A. J. MILLER
R. 5. CARLSMITH ARE
J. P. LANE ARE
REACTOR PHYSICS ENGINEERING DESIGN REACTOR CONSTRUCTION POWER PLANT ENGINEERING EXPERIMENTAL ENGINEERING STAFF ASSISTANT STAFF ASSISTANT STAFF ASSISTANT STAFF ASSISTANT STAFF ASSISTANT STAFF ASSISTANT
W. K. ERGEN ARE H, C, GRAY PWA E. 5. BETTIS ARE A. P. FRAAS ARE H. W, SAVAGE ARE W, R. GRIMES W W. D, MANLY M E. P. BLIZARD P D. S, BILLINGTON 5§ H. F. POPPENDIEK REE w, K, ERGEN ARE
P. HARMAN, SEC. ARE 1. ZASLER ARE J. PARKER, SEC. ARE A. 5. THOMPSON PMC D. ALEXANDER, SEC, ARE
L. E. FERGUSON, SEC. ARE P, HARMAN, SEC. ARE
H. W, BERTINI ARE s, J. FOSTER, CLERK ARE PUMPS PLANNING AND CONTROL.
w, E. KINNEY ARE
A.G. . ADMINISTRATION LR, DY
R, K. OSBORN ARE A. A, ABBATIELLO ARE . R ae wLscored ARE B B T PSON . CHEMISTRY WETALLURGY SHIELDING RADLATION DAMAGE HEAT TRAMSFER AND PHYSICAL REACTOR CALCULATIONS
A, H. ANDERSON ARE W, R. GRIMES MC ¥. D, MANLY M o 5. 5. BILLIMGTON 5 PROPERTIES RESEARCH R, A. CHARPIE RO
MATHEMATICIANS D, €. BORDEN PHA CORE HYDRODYNAMICS AND THERMODYHAMICS ART COMPONENT TEST FACILITIES b, R. CUNEQ MC G. M. ADAMSON® M HIELDING RESEARCH 18, TRICE 55 H. F. POPPENDIEK REE i
A. FORBES ARE ¥. G, COBB ARE G. . WHITMAN ARE . T. FURGERSON ARE E.STORTO ARE D. E. CALDWELL, SEC. He #. H. BRIDGES* M E. P. BLIZARD F T. K. CARLSMITH, SEC. REE R, R, COVEYOU* ARE
M, TSAGARIS ARE J, ¥, ESTABROOK ARE R. L. BREWSTER ARE M. E. LACKEY ARE J. CROWLEY ARE M. J. WHITMAN USAF R, RICKMAN, SEC. P . D. BAUMANN 5% . D' GIVEN* WP
::. ::;ms P¥A WEAT EXCHANGER G, SAMUELS ARE A. G. GRINDELL* ARE SPECIAL, PROBLEMS 1. THOMAS, SEC. M L. 5. ABBOTT " W, E. BROWNING $5 HEAT TRANSFER W. C. SANGREN* WP
CONSUL TANTS > 3 PWA W, ). STELZMAN ARE R, E. MACPHERSDN* ARE R. F. NEWTON RD S, ALSLENDER PWA W. E, BRUNDAGE 88 J. 0. BRADFUTE REE
L R. E. HELMS ARE R. E. MACPHERSON® ARE M. W. HINES, COMPUTER  ARE L. A, MANN" ARE BENERAL CORROSION F. H. MURRAY P R, W, CARROLL. 88 B. C. HAMILTON REE
A m”"“hwfim A TECH, T. L. HUDSON ARE J. €, AMOS ARE 1. 4, TUDOR, DRAFTSMAN  ARE C. H, WELLS P& PHASE EQUILIBRIUM STUDIES E. E. HOFFMAN M H. E. STERN oy W.W. DAYIS 5§ H W, HOFFMAN REE
A, D, JARRATT PWA L, H, DEVLIN PHA €. J. BARTON e W, H. COOK o C.D. ZERBY P M. J. FELDMAN 55 F. £, LYNCH REE
s ':zERLRE erioer ;;;E R. D, PEAK PWA APPLIED MECHANICS AND PHYSICS IN-PILE LOOP TESTS L. M. BRATCHER M C. F. LEITTEN M D. E. GUSS USAF G L MULLER PHA
VE A 1. W, KINGSLEY AR D. 8. TRAUGER LID TANK SHIELDING FAC JE. oL
Fo Ly MAGLEY ARE € M.E. LAYERNE e ARE B. H. CLAMPITT e TECHNICIANS Ty M. E. HIBKLE 55 L.D. PALMER REE CRITICAL EXPERIMENTS
et DRIVES AND AUKILIARIES 7. J. BALLES PWA S. EUBANKS, STENO. ARE R. E. MOORE WC 6. T. CHAPMAN P H. L. HEWPHILL AC 1. L. WANTLAND REE
¥ NELs'BN m: C.5. BURTNETTE USAF R. G. AFFEL ARE R. J. SHEIL WC 3. €. POPE M ). B. DEE PWA G. W, KEILHOLTZ 55 - A 0. CALLIHAN® P
b P 5. M. DECAMP ARE F. A. FIELD USAF L. P. CARPENTER ARE D. L. STOCKTON ME L.R. TROTTER M ¥, J. MCCOTL YA H. V. KLAUS I TECHNICIANS . L. RUEFF,* SEC, P
N C. B, THOMPSON MH J. A CONLIN ARE J. M, MILLER P R, E. MAYES PwA NES REE
:. :. :x PWA CONTROLS W, I. GALYON, DRAFTSMAN ARE %, CUNRINGHAM i FUEL PREPARATION RESEARCH DYNAMIC CORRCSION 1 SMOLEN PHA J. G, MORGAN 55 w CLer REE J. 5. CRUDELE PWA
s FyA J- EASTHAN . HEAT TRANSFER P, A, GNADT ARE AND PHASE BEHAVIOR G. M, ADAMSON" M D. K. TRUBEY P M, T. MORGAN 8 G, M. WINN REE 3. FLELLIS® v
DRAFTSMEN J. W, ALLEN ARE D. M, HAINES PHA F. F. BLANKENSHIP We T, PRICE FWA H. €. WOCOSUM PHA A5, OLSON 5§ B. L. GREENSTREET ARE
E. R. MANN 1 R. B, SCHUL THEISS ARE D, W, HAGHUSON ARE ) WC A, TABOADA " M. F. CSBORNE 3 PHYSICAL PROPERTIES ¥, G, HARNESS* P
¥, C. GEORGE ARE S, C. SHUFORD PWA M. M. YAROSH ARE o ¥. P. TRECIOKAS PWA TECHNICIANS W, W, PARKINSON 88 ‘e
F. P. B0ODY MC 5. |, COHEN REE %, G. KENNEDY PWA
G. R. HICKS ARE A. L. SOUTHERN ARE TECHNICIANS R E CLEARY PWa E. BECKHAM p H. E. ROBERTSON 55 GREENE REE LT v
G, L. HOLLIS AR SPECIAL PROBLEMS | - E. TECHNICIANS . M T, ROBINSON - N. D. GRE: L J.LYN
. L E C. J. GREEN ARE H, A, FRIEDMAN MC H. JARVIS 1c "ot W. D, POWERS REE J. H. MARABLE* P
FILL AND DRAIN AND BR L H.
J. R. LARRABEE PWA ENRICFMENT SYSTEM . 8. COTTRELL ARE F. M, GRIZZELL ARE 5. LANGER MC G, O, BRADY M 4. R, TAYLOR P E. 5. SCHRARTZ 55 1. W, NDAKS PWA
g- G. MICHELSON ARE Ca . ARE H. E. HUNGERFORD ARE C. C. NANCE ARE R. P. METCALF MC R. M. EVANS " J. W. WAMPLER p f-é'Sfir‘-;:PLETON g: TECHNICIANS E. R. ROHRER® P
- A MILLS ARE R CMR"R“! J. K, LESLIE ARE B, L. JOHNSON ARE F. W, MILES MC PHYSICAL CHEMISTRY OF LIQUID METALS BULK SHIELDING FACILITY s G.C. BLALOCK REE E. V. SANDIN PWA
C, F, SALES AHE . Cu PWA SPECIAL PROBLENS I F. J. SCHAFER ARE R. E. THOMA Me C.C. WEBSTER 5 R M. BURNETT REE D, SCOTT, JR. ARE
C. E. STEVENSON ARE G. P. SMITH " F. C. MAIENSCHEIN P R. A, WEEKS 58 L 5, SNYDER PWA
SPECIAL PROBLEMS 5. J. CLAIBORNE REE 3
R. W. BUSSARD ARE CORROSION STUDIES C. R. BOSTON ] G, M. ESTABRODK e W R.WILLIS 5 REE R. M, SPENCER USAF
W. €. TUNNELL ARE R. 1. GRAY PWA PROCUREMENT E. KERTESZ MC J, ¥, CATHCART M 1B, FLYNN & 1. C, WILSON $5 T. . JONES D. V. P. WILLIAMS* P
H, €. HOPKINS PWA W. F. BOUDREAU® K.25 " K. 1. BUTTRAM MC G. F. PETERSEN M WP HAYDON® P +. C. ZUKAS S E. L. ZIMMERMAN® P
R, CLARKE® K-25 F. A. KNOX MC M. £. STEIDLITZ " K. 4. HENRY P TECHNICIANS s
ART FACILITY CONSUL TANTS 4. R, EVANS ARE R. E. MEADOWS M TECHNICIANS E. B, JOHMSON P
A. H. FOX, UMION COLLEGE W, R. O$BORN ARE N, V. SMITH MC K. M. JOHNSON PWA J. R. DUCKWORTH ARE
¥. G. PIPER* K25 W. LOWEN, UMION COLLEGE R. . ANDERSON I W, ZOBEL P C.ELLIS $5
R. L. MAXWELL, CLERKS TECHN!CIAN L.l HALL M ' ¢. BOUNDS, SEC. P
T, £, CRABTREE ARE UNIVERSITY OF TENNESSEE G. C, JENKINS ARE J. M. DIDLAKE MC WECHANICAL PROPERTIES ’ . CONSULTANT
¥. J. KELLEGHAN ARE C. F. WISLICENUS, F. LEWIS ARE TECHNICIANS D. F. WEEKES, TEXAS A & M COLLEGE
£. R. MCQUILKIN ARE PENNSYLVANIA STATE UNIVERSITY E. 8. PERRIN ARE PRODUCTIGN OF PURIFIED FUELS R.B. DLIVER M D. J. KIRBY P
G. C. ROBINSON ARE D. STOREY ARE 2. H. DEVAN " 1. SELLERS c
G. J. NESSLE We D. A. DOUGLAS M - SELL. !
C. F. VEST ARE 1. F. BLAKELY MC 1. ¥, WOODS W R, M, SIMMONS P
TEST SERYICES C. R. CROFT e . S, SMIDDIE Ic
W. B. MACDONALD ARE F. L. DALEY W TECHNICIANS G, G. 5TouT Ic
D. HARRIS, STENO. ARE 1. €. EDRGAN Mec J.T. EAST W H. WEAVER P FUEL REPROCESSING
A. MONTGOMERY, CL. TYP.  ARE 3 TRUITT HC G. M. GONZALEZ M —
. N
C. P COUGHLEN ARE AN S O HUDSON . TOWER SHIELDING FACILITY F. R, BRUCE cT
R. A. DREISBACK PwA V. G, LANE M €. E. CLIFFORD p
R. HEL TON ARE F. A. DOSS ARE B. MCNABE W .. BLOSSER P CHEMICAL DEVELOPMENT
E. M. LEES ARE B, F., HITCH MC C. K. THOMAS M L, B, HOLLAND B D, E. FERGUSON* T
J. 1. MILICH PWA W, JENNINGS MC C, W, WALKER M J L. HULL P G. I. CATHERS T
W, J. MUENZER ARE G, A, PALMER MC w. F. VALERING NACA M. R. BENNETT cT
¢ E. MURPHY ARE B. C. THOMAS MC FABRICATION J. VAN HOOMISSEN BAC TECHNICIAN
P, G. SMITH ARE R. G. WILEY ARE 1. H, CODBS " F. N, WATSON P CHRIC
L DY . M. DUFF cr
D. R. WARD ARE CHEMICAL EQUILIBRIA n :?NODU¢E‘DRE SWA E. MCBEE, SEC. " Rt
TECHNICIANS L. G. OVERHOLSER WC TECHNICIANS UNIT OPERATIONS
NOTE: THIS CHART SHOWS ONLY THE LINES OF TECHNICAL COORDINATION OF THE ANP PROJECT. THE ), ADKINS ARE E. E. KETCHEM MC TECHNICEAN E. D. CARROLL IC W, K, EISTER* cT
VARIOUS INDIVIDUALS AND GROUPS OF PEOPLE LISTED ARE ENGAGED EITHER WHOLLY OR PART TIME ON .5, ADDISON ARE . g- ;;g“f“ MC R. W, JOHNSON " J. N, MONEY P N R cr
RESEARCH AND DESIGH WHICH IS COORDINATED FOR THE BENEFIT OF THE ANP PROJECT IN THE MANNER R wELL e PN s WELDING AND BRAZING o o e - <
JNDICATED GN THE CHART, EACH GROUP, HOWEYER, IS ALSC RESPONSIBLE TO ITS DIVISION DIRECTOR G. 5. CHILTON ARE L. E. TOPOL MC P. PATRIARCA ') - DESIGH
FOR THE DETAILED PROGRESS OF ITS RESEARCH AND FOR ADMINISTRATIVE MATTERS, 1. M. COBURN ARE C. F. WEAVER MC :- E fié’é??k':fo *;w A CONSUL TANT H. E. GOELLER" cT
J. R. CROLEY ARE L L. R. P. MILFORD cT
AL HE, R ERSIT
THE KEY TO THE ABBREVIATIONS USED IS GIVEN BELOW. 7. M. CUNNINGHAM ARE SUPPORTING STUDIES K. W, REBER W H. A, BETHE, CORNELL UNIVERSITY
R.E. DIAL ARE P. A. AGRON c G. M. SLAUGHTER " CONTRACTORS
AC ANALYTICAL CHEMISTRY DIVISION — ORNL W. H. DUCKWORTH ARE J- E. SUTHERLAND < TECHNICIANS YORIDES, INC
ALC  AMERICAN LOCOMOTIVE COMPANY 1. P. EUBANKS ARE Tt DoV bhe)
W. K. R. FINNELL ARE CONSULTANTS 0. E. CONNOR M NUCLEAR DEVELOPMENT
ARE  AIRCRAFT REACTOR ENGINEERING DIVISION = ORNL H. FOUST ARE ). 4. CARTER, CARTER LABORATORIES B. MCDOWEL L M ASSOCIATES, INC.
BAC  BOEING AIRPLANE COMPANY W. D. GHORMLEY ARE B G FILL. DUKE UNIVERSITY g. €., SHUBERT M
P BENDIX PRODUCTS, DIVISION OF BENDIX AVIATION CORPORATION il o H. INSLEY - . SHOOSTER M
c CHEMISTRY DIVISION - ORNL C: G: HENLEY ARE T. N. MCYAY, UNIVERSITY OF ALABAMA COMSULTANTS AHP STEERING COMMITTEE
cr CHEMICAL TECHNOLOGY DIVISION — ORNL D. E, MCCARTY ARE N. CABRERA, UNIVERSITY OF VIRGINIA
cv CONSOLIDATED YULTEE AIRCRAFT CORPORATION G.E. MILLS ARE CoMTRACTORS M. J. GRANT, MASSACHUSETTS METALLOGRAPHY W. H. JORDAN, CHAIRMAN
B. H. MONTGOMERY ARE AMES LABORATORY INSTITUTE OF TECHNOLOGY 2. ). GRAY™ " e :
GE GENERAL ELECTRIC COMPANY 1. ). PARSONS ARE BATTELLE MEMORIAL INSTITUTE J, L, GREGG, CORNELL UMIVERSITY - E. 5, BETTIS
Ic INSTRUMENTATION AND CONTROLS DIVISION = ORNL M. A. REDDEN ARE CARTER LABORATOR(ES W, D, JORDAN, UNIVERSITY OF ALABAMA R, 5. CROLSE M D. 5. BHLLINGTON
" METALLURGY DIVISION  ORNL ®. REID ARE METAL HYDRIDES, INC. E. F, NIPPES$, RENSSELAER T M KEGLEY M E. P, 8LIZARD
A G. TOWNS ARE MOUND LABORATORY POLYTECHMIC INSTITUTE E L LONG M 80D
e MATERIALS CHEMISTRY DIVISION - QRNL B. C. WILLIAMS ARE A. R, NICHOLS, W. F. SAVAGE, RENSSELAER £ 3 MANTHOS W §. J. CROMER
ME MERCK & COMPANY, INC, CRAFT COORDINATION SAN DIEGO STATE COLLEGE POLYTECHNIC INSTITUTE s :. l;- ::fii:
. SISTARE, HANDY & HARMON TECHNICIANS - o
Mid
WINNEAPOL [$-HONE YWELL REGULATOR COMPANY . L. MATTHEWS Y12 P. €. SHARRAH, UNIVERSITY DF ARKANSAS €. R BOYD " . R. GRIMES
WP MATHEMATICS PANEL ~ ORNL D. L. CLARK ARE F. G, TATNALL, B, F, DAY M G e
HACA  NATIDNAL ADVISORY COMM TTEE FOR AERONAUTICS D. L. GRAY ARE cb Sug:::;f"“" CHEMISTRY AC BALDWIN-LIMA-HAMIL TON CORPORATION R. L. FITZGERALD M : ?’ ,::?_T'E:
P ARYSICS DIVISION — ORNL || O N e s A E. C. WRIGHT, UNIVERSITY OF ALABAMA B C. LESLIE M b b R RENDIEK
PG PACKARD MOTOR CAR COMPANY CONSULTANTS J.C. WHITE AC CONTRACTORS H. W. SAVAGE
PWA  PRATT & WHITNEY AIRCRAFT, DIVISION OF LUAC J. F. BAILEY, UNIVERSITY OF TENNESSEE E. M. ZARZECKI,* SEC AC BATTELLE MEWORIAL INSTITUTE :' ?.Ssr;’gbg
RD RESEARCH DIRECTOR'S DEPARTMENT ~ ORNL :’ :_‘- S:M"‘E‘:ERS RESCARCH 22;’%?:2,:;"'”” COMPANY CERAMIC RESEARCH 2. A, SHARTOUT
REE  REACTOR EXPERIMENTAL ENGINEERING DIVISION — ORNL ¥. K. STAIR, UNIVERSITY OF TENNESSEE A 5. MEYER, JR, AC GLENN L. MARTIN COMPANY D s+ SEC : A. M, REINBERS
8 STABLE ISGTOPE RESEARCH AND PRODUCTION DIVISION =~ ORNL D, L, MANNING AC METAL HYgRID:i'Tlng P : HOBBS, -
B, L, McDOWELL AC NEW ENGLAND
5 $OLID STATE DIVISION — ORNL 1M PEELE AC TESTING LABORATORY €. E. CURTIS: M
USAF  UNITED STATES AIR FORCE W, J. ROSS AC RENSSELAER POLYTECHNIC INSTITUTE L. M. DONEY W
s N 5. D, FULKERSOR® M
“PART TIME DEVELOPMENT UPERIOR TUBE COMPANY 5. . FULKERSONY M
1 P YOUNG ac UNIVERSITY OF TENNESSEE Lo AGH A
G. GOLDBERG AC M, P, HAYDON® P
SPECTROGRAPHIC ANALYSIS M. A MARLER AC J. R, JOHNSON* »
LR eRALLY + H SERVICE &, J. TAYLOR* "
- Rohcl ) W, F. VAUGHAN ac G. D. WHITE® »
J. A, NORRIS* AND OTHERS 5t €. R WILLIAMS ac TECHNICIAN
R. F. APPLE AC . "
MASS SPECTROMETRY D. E. CARPENTER AC . A4 GRIFFIN
C. R, BALDOCK" st TECHNICIANS CONSULTANTS
1. R.SITES* si R, C. BRYANT AC ™. T. CORY, GRAPHITE SPECIALTIES
L. R, HALL ac CORPORATION
L. E. IDOM AC H. INSLEY
A, H. MATTHEWS AC T. N, MCVAY, UNIVERSITY OF ALABAMA
C. E. PRATHER AC T, 5. SHEYL IN, OHIO STATE UMIVERSITY
A. D. WILSON AC H. THURNAUER, AMERICAN LAYA
C. M. WILSON AC CORPORATION

 

 

 

 

 

 

 

 

 
e
~7

sm=

    

the Atomic