ORNL-2942.txt

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AEC RESEARCH AND DEVELOPMENT REPORT C-84 - Reactors-Special Features

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ORNL-2942

of Aircraft Reactors

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AIRCRAFT NUCLEAR PROPULSION PROJECT
SEMIANNUAL PROGRESS REPORT

FOR PERIOD ENDING APRIL 30, 1960

CENTRAL RESEARCH LIBRARY
DOCUMENT COLLECTION

LIBRARY LOAN COPY
DO NOT TRANSFER TO ANOTHER PERSON

If you wish someone else to see this
document, send in name with document
and the library will arrange a loan,

 

 

 

 

operated by

 

OAK RIDGE NATIONAL LABORATORY

UNION CARBIDE CORPORATION

for the =

f' X ';'

U.S. ATOMIC ENERGY COMMISSION 2. )
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LEGAL NOTICE

This report was preparsd as an sccount of Gevernment sponsored wark., Nelther the United Stares,

nor the Commission, nor any person acting on behalf of the Commission:

A. Makes any warranty or representation, expressed or implied, with respect to the gccuracy,
completeness, or usefulness of the information conteined in this report, or that the use of
any information, oppeoratus, methoed, or process disclosed in this report moy not infringe
privately owned rights; or

B. Assumes ony ligbilities with respect to the use of, or for domages resulting from the use of
any informotion, apparatus, method, or process disclesed in this rapart.

As used in the obove, "‘parson gcting on beholf of the Commission includes oany employee or

contractor of the Commission, or employse of such contractor, to the extent thot such employes

or contractor of the Commission, or employes of such controctor prepares, disseminates, or
provides access to, ony information pursuant 1o his employment or contract with the Commission,

of his employment with such contractar.

 

 

 

 
  
 

 

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C-84 — Reactors—Special Features
of Aircraft Reactors
M=-3679 (24th Ed.)

This document consists of 146 pages.

Copy )& of 225 copies. Series A.

ORNL-2942

Contract No. W-7405~eng=-26

ATRCRAFT NUCLEAR PROPULSION PROJECT
SEMIANNUAL PROGRESS REPORT
for Period Ending April 30, 1960

Staff

Oak Rildge National Laboratory

Date Issued

JUL 121960

Oak Ridge National Laboratory
Oak Ridge, Tennessee
operated by
UNION CARBIDE CORPORATION
for the
U. S. ATOMIC ENERGY COMMISSION

RTIN MARIETTA ENERGY SYSTEMS LIBRARIES

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3 4456 0251077 1

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 
 

 

    
 

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FOREWORD

The ORNL=-ANP program primarily provides research and development
support in reactor materials, shielding, and reactor engineering to
organizations engaged in the development of air-cooled and liquid~metal-
cooled reactors for aircraft propulsion. Most of the work described
here is basic to or in direct support of investigations under way at
General Electric Company, Aircraft Nuclear Propulsion Department, and

Pratt & Whitney Aircraft Division, United Aircraft Corporation.

 

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CONTENTS

SUWARY L A I N R N N N N N A A A N N N N N N E N Y ..

PART 1. MATERIALS RESEARCH AND FENGINEERING

Reactions With OX¥Een .. ieitiveeettneeneecneanosvonocones
Reactions with Air and Mixtures of Oxygen and Argon ......

Effects of Oxygen Contamination and Exposure to Lithium
on the Tensile Properties of Columbium .....ecevevuvnvroones

Mass-Transfer Effects in Systems Consisting of Dissimilar
Materials * 8 & & 0 & * 0 & & % & & 9 & s 0 e 2 ¢ 2 0 & P B 0SB RNSEE SRS Pe SR

Compatibility of Columbium-Zirconium Alloy and Haynes No. 25
Bl1OY v eeeererrnnncann. Gt et et s et nee tes s s e eesea vhen s

Compatibility of Boiling Potassium and Type 316 Stainless
Steel 0 0 8 P 8 S PSPPSR PRSP E BSOS PSP NSNS PN OSSO EE e

2. STUDIES OF THE AGING OF COLUMBIUM-BASE ALLOYS ...... Pt eeaan
Aging Studies on Wrought Sheet .......... ceserectsaaneerar s
Aging of Fusion Welded Material .....uivevoerecnceeeasnonoonnss

3. MECHANICAL PROPERTIES INVESTIGATIONS ...vevveeveervocnennenas

Effects of Alloying Additions on the Mechanical Properties
OfPure Columbiu-m " s 0 s ® 00 PENEPE PSRN S APt

Fatigue Studies of Inconel ....veiiereireroceneccescenuonnnes
4, ALLOY PREPARATION ........... et recae s et ee et e e s et
Electron-Beam Melting of Columbium Al11OYS ..vevreoeseo oo aen
5. CERAMICS RESEARCH . ..ttitttinresoreneneeensenonsennnnnenennens

Preparation of Beryllium Oxide ...cevieeeeerrrronnsoneonroonasns

Oxalate ProCesSS ..eiieieerosssossssosncssssons ceeo e tnans
Calcining BeC,0,*3H,0 to BeO ...... er e et et e e co
Sintering Studies ..iviiieiiiernnninnon S ha e
Evolution of Lithium During Sintering ....cieeeverecevenss
Analyses of Beryllium Oxide .....vivtiierneenenroeecorornones
Chemical MethodS t.iv.veiresesssssosssosssssnaocsacssannssns

opectrographic AnalysSes . u.eeevesveroecoonrsosconooososasas

 

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Compatibility of Beryllium Oxide and Other Metal Oxides .....
BeO=Y203 .i.iveeriniraresnnn ces e Ceer e Ceser e
BeO—Or0 ittt ittt sttt e e
BeO-LayCs and BeO-Mg0 .. viverrenerreorionnsrnceersonansess
BeO~Bal L.iuirererreerrsrsarssasresresenstacnraasenranenn

BeHaO L L A BN TR DN DN DAY N BN BN BN RN N I BN N B BN BN R L B I I DR R DN BN DK DN NN N NN BN NN N BN R NN BN DN NN I L N BN

Preparation of Refractory Oxides from Molten Salts ..........

ENGINEERING AND HEAT TRANSFER STUDIES ...... chedeess et e s
Molten Lithium Heat Transfer ......ceiveeervncrcnresanscnsans
Thermal Properties of Reactor Materials ..... ettt e e

Thermal Conductivity of Columbium~Zirconium Alloys .......

Enthalpy of Lithium Hydride ...... tercereesesnrra s cee e
Boiling-Potassium Heat Transfer Experiment ........0c00c00...
RADTATTION EFFECTS  tiieiiiernsotrssesosntossaroasssenssnanas ceos
Irradiation of Moderator Materials in the ETR ......ceevvvnss
Creep and Stress-Rupture Tests Under Irradiation ............
ADVANCED POWER PIANT STUDIES i isvuvsvvssnnsovssonsonnsocnses

Comparative Study of Power Plants for Space Vehicles ........

Space Radiator Designs ....civervrensns P e e e e st a e

PART 2. SHIELDING

SHEI'DINGTI‘}EORY e 8 08PN EEEAERNS RSNt Ad s

Monte Carloc Calculations of Response Functions of Gamma~Ray
SCintillationDetectorS s APPSR NS e * & & & B & 0 &

A Monte Carlo Code for Deep Penetrations of Gamma Rays ......

Prediction of Thermal-Neutron Fluxes in the Bulk Shielding
Facility from Lid Tank Shielding Facility Data .......... oo

Monte Carlo Calculations of Dose Rates Inside a Cylindrical
Crewcompartment 4 4 8 & ¢ 0 4 5 5 P B F S B S PP SRS PE T NSNS EE RS RN

Design of a Unit Shield ........... P aiseerras et e e s s sy
LID TANK SHIEIDING FACILITY et eveevrvocneoacsss Gt e e e e

Radiation Attenuation Measurement Behind Multilayer
Configurations ........cveeveene e e et r e se s senene oo

 

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BUIK SHIEIDING FACTLITY v tevervnnennnnnennoossonssosssennnnas
Stainless Steel-UO5 Reactor (BSR-II) ........ et

The Spectrum of Prompt Gamma Rays from Thermal-Neutron
Fis810n OF U237 i teteitiieteeeiieseansesnnesennssens

The Model IV Gamma~Ray Spectrometer ............ cese e ansas

Compilation of Data on Gamma-Ray Spectra Resulting from
Thermal-Neutron Capture .......c.vivvverecnoes cesserrssen s

TOWER SHIELDING FACILITY 4 ivvuuvsvrovsassvroscoenvonescananns
Tower Shielding Reactor II ....vevevereneeenenonansenss e e
Reactor Assembly ... ieiieiirireeennorenonsos P ess e st easae e
Critical Experiments ...eeeeriveeerersnsososcsoosnaresenoss
Nuclear CalculatiOnsS ..eeeesvvessnerosssrossnsssnsssossnns
Flow Distribution Studies ...vivvververvesss C it e
Experimental Program ......... Cee e s e i st e s cererar e
Detector Shield .......i0veve... erreererrnesenen o etoasas
Gamma=Ray Spectrometry ... iiererreenoorensas tereer e e ae e
Neutron SpectrometIy ...cvieieneereenens se s eseanrsncons o e

Data ProCesSSINE +ieitviverecrrvasesnsorooonorscorsosnsscens

 

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ANP PROJECT SEMIANNUAL PROGRESS REPORT

SUMMARY

Part 1. Materials Research and Engineering

1. Materials Compatibility Research

The effects of oxygen pressures 1in the range between 3 X 10~ and
5 x 10=3 mm Hg on the reaction rates of oxygen with columbium and the
correlation of the reaction rates with the formation of the oxides of
columbium were studied at 850, 1000, and 1200°C. As the oxygen pressure -
was increased from the lowest pressures, the following principal reactions
occurred: (1) solution of oxygen in columbium, (2) precipitation of CbO
and the formation of CbO, nodules, (3) CbO, film formation, and, finally,
at the highest pressures, (4) Cb,0s formation. Reactions that could be
attributed wholly -to the solution of oxygen in columbium were not en-
countered except at 1200°C and an oxygen pressure of 3 X 107° mm Hg.

The studies of the reactions of columbium with low-pressure alr were
extended to include tests at 1200°C. At an air concentration of 5 x 1074
mm Hg, a linear reaction rate constant of 6.3 X 1074 mg/cmz-min was ob-
served. In contrast, at an equivalent oxygen pressure, a reaction rate
constant of 4.5 X 10~3 mg/cem?.min was obtained in pure oxygen. Mixtures
of argon and oxygen also lowered the reaction rate in comparison with
the reaction rate in pure oxygen. Although the reaction rates at these
low pressures are several orders of magnitude less than in oxygen or air
at atmospheric pressure, the rates are sufficiently high to cause .
significant oxidation of columbium in a few hours.

Very small additions of oxygen to columbium were shown to increase
its room-temperature tensile strength. This increase in strength was
still evident after exposure to lithium at high temperature. However,
when the additions of oxygen were large enough to cause corrosion of
the columbium by high-temperature lithium, the addition of oxygen and

subsequent exposure to lithium resulted in decreased strength and ductility.

viii

 

 
 

EE—

Thin layers of carbide and nitride were formed on columbium~zirconium
alloy specimens after exposure to NaK in stainless steel containers at
elevated temperatures. The elevated~ and room-temperature mechanical
properties of the alloy were altered by this exposure.

Mass~transfer effects were observed in a NaK thermal-convection loop
constructed of Haynes No. 25 alloy which contained Cb—0.8% Zr specimens.

A very thin carbide layer was formed on the columbium-zirconium alloy, and
no significant alterations were found in its tensile strength or elongation.

Tests are being conducted to determine the compatibility of type 316
stainless steel with boiling potassium. Capsule and thermal-convection
loop tests conducted at 1600°F for several hundred hours showed no signifi-
cant attack on the steel. Tests of several thousand hours duration are

now in progress.

2. Studies of the Aging of Columbium-Base Alloys

It was shown that wrought sheet of Cb—l.15% Zr received from CANEL
responded to aging heat treatments. The aging heat treatments were carried
out at 1700°F for periods up to 500 hr on material which had been annealed
for 2 hr at either 2190, 2550, or 2910°F., The effect of the aging treat-
ment on the mechanical properties of the alloy annealed at 2910°F was
pronounced, whereas the effect on the properties of the alloy annealed
at 2190°F was minor. The microstructure of the alloy annealed at 2910°F
and then aged presented a classical aging pattern.

The aglng treatments were originally carried out 1in quartz capsules,
In order to determine whether the aging effects were due to silicon con-
tamination from the quartz, the aging experiments were carried out again
in columbium capsules on material annealed at 2910°F. The tensile strength
and elongation data were essentially the same as obtained with the quart:z
capsules.

Aging studies were conducted on Cb—1.15% 7Zr alloy weldments in the
temperature range from 1300 to 1800°F for time periods of 1 to 250 hr.
Ductile-~to=brittle transition curves were developed. It was shown that

at 1800°F, overaging occurred in 2 hr. Weldments made in an inert

R ix
atmosphere chamber were shown to age in a manner similar to that of weldments

made with the trailer shield, and weldments aged 1In columbium capsules

demonstrated the same properties as those aged in quartz capsules.

3. Mechanical Properties Investigations

Independent additions of nitrogen, oxygen, carbon, and hydrogen were
made to columbium. Room-temperature bend tests showed that nitrogen,
oxygen, and carbon increased the strength of the columbium. Large additions -
of nitrogen and oxygen reduced the ductility. Creep tests are in progress
to evaluate the effect of the impurities on the high-temperature mechanical
properties.

Measurements were made of the plastic strain which occurs during each
cycle of a conventional stress~fatigue test. The strain-fatigue curves
calculated from these data were compared with strain-fatigue data obtained

from tensile and cyclic creep tests.

4. Alloy Preparation

Techniques were evaluated for the preparation of columbium~zirconium
alloys in the electron-bombardment furnace. Electron-beam melting was

also carried out on other materials, such as Mo, U, ZrC and Al,0s.

5. Ceramics Research

One variation of the GEOM oxalate process was found to yield a -
beryllium oxalate monohydrate of fine particle size, and another led to
the production of fluffy BeO with a bulk density of 0.01 g/cm3.

The phase changes that occur during calcination of BeC,0,4+3H,0 to
BeC were further investigated as a function of calcining procedures, and
additional information was developed on the effect of calcining procedures
on the sinterability of the BeO. It was observed that lithium impurities
were volatilized rather effectively during standard sintering procedures.
Additional chemical and spectrographic methods were developed for analysis

of BeO,

 
 

 

Melting and quenching experiments were performed which yielded pre=-
liminary information as to the compatibility of BeO with Y03, SrQ, Li,0j,
MgO, BaO, and CaO,

A study was made of coprecipitation of UO, and BeQ by water vapor
from a mixture of UF,; in LiF-BeF,. In some instances, 20-u UO, particles

were coated on all sides with BeO.

6. Engineering and Heat Transfer Studies

Additional data covering an extended Peclet modulus range, were
obtained in the study of the forced-circulation heat transfer to molten
lithium flowing turbulently in a heated tube. A line through the data

mean has the equation

- 0.3
Ny = 1.251 N2 .

The thermal conductivities of two additional columbium-zirconium
alloys were measured. No apparent distinction on the basis of zirconium
content can be made between the several specimens that have been tested.

Over the temperature range 200 to 900°F,
k = 26.91 + 0.00857t

where k is in Btu/hr.ft-°F and t is in °F.

The enthalpy of lithium hydride was determined over the range 100 to
900°C.

Design and construction of stainless steel equipment is 1in progress

for studying heat transfer to flowing, boiling potassium in a heated tube.

7. Radiation Effects

 

The irradiation of yttrium hydride and beryllium oxide specimens in
the ETR was continued. Previously irradiated specimens were examined for
structural stability. In addition, the beryllium oxide specimens were
being checked to determine the amount of gas formed by (n,Q@) and (n,2n)

reactions in the beryllium.
Additional in=pile stress-rupture data were obtained on Inconel,

and in-pile equipment for testing columbium-zirconium alloys is under

construction.

8. Advanced Power Plant Studies

The study of reactor-turbine generator systems for auxiliary power
units in satellites was continued. A report is being prepared on
thermodynamic-cycle selection that is based on a critical review of radia-
tor, turbine, generator, reactor, and shield design considerations. Ad-
ditional work is in progress on space-radiator designs, with emphasis on

configurations for meteorite bumpers and coolant manifolds.

Part 2. Shielding

9. Shielding Theory

Three IBM=-704 codes have been completed that are designed to calculate,
by Monte Carlo methods, the response functions of NaI(Tl), CsI(Tl), and
xylene scintillators to gamma rays. Inclusion of secondary effects, such
as annihilation radiation and bermsstrahlung, plus an optional Gaussian
broadening to simulate an actual scintillator resclution, enables the
present codes to present an essentially complete response picture. The
codes permit calculations for either right circular cylinders or right
cylinders capped with truncated cones and inclusion of axial wells or
holes, and they are not limited as to dimensions. Gamma-ray energies
may range from 0.005 to 10,0 Mev.

A Monte Carlo code for the IBM~704 has been developed that is in=-
tended to calculate deep (1- to 20-mfp) penetrations of gamma rays. The
code employs the so-called 'conditional" Monte Carlo method, as well as
a "Russian Roulette" technique to assist in the calculation at the longer
distances. The relatively limited data available at present indicate
satisfactory agreement with calculations by the "moments"” method.

The final review of the calculation predicting thermal-neutron fluxes

in water around Bulk Shielding Reactor I from Lid Tank Shielding Facility

 
 

data showed reasonable agreement between calculated and measured results.
For distances of 65 cm or more from the reactor face, differences ranged
from O to 9%, whereas, for distances less than 65 cm, the results calcu-
lated from LTSF data were consistently higher than the experimental
measurements, with differences of from 6 to 22%.

A final report has been published covering the operation of the ABCD
code, an IBM=704 code for calculation of fast-neutron dose rates inside
a cylindrical crew compartment. The code calculates the dose from neutrons
uniformly distributed in space in the neighborhood of the compartment and
utilizes a so-called "similarity transform" feature to permit simultaneous
computation of dose for many different configurations. The Convair D=35
code, which computes the neutron distribution in air from a unit point~
monodirectional source will comprise the primary input for ABCD. A series
of problems run to check the performance of the code gave satisfactory
results.

A study has been initiated to develop calculational methods applicable
to the design of a unit shield, that is, an aircraft reactor shield con-
sisting solely of material surrounding the reactor, as opposed to shields
in which the shielding material is divided between reactor and crew
compartment. Preliminary studies have concentrated on an analysis of the
gamma~ray problem and the design and placement of heavy materials within
the shield. A simple model consisting of a point source of isotropic,
monoenergetic gamma rays imbedded at the center of a water sphere, with
a lead shadow shield of uniform thickness placed at a specified distance
from the source, was chosen for the first calculations. The first calcu-
lations have succeeded in giving a general knowledge of the magnitudes
of the factors involved and are being extended to more significant problems
involving variable~thickness shadow shields and considering the contributions

made by secondary gamma rays.

10. Lid Tank Shielding Facility

 

A large number of multilayer shielding configurations have recently

been investigated. The materials included borated and unborated
polyethylene, depleted uranium, lead, and steel. Measurements included

fast=neutron dose rates, gamma-ray dose rates, and thermal-neutron fluxes.

11. Bulk Shielding Facility

 

Extensive static and dynamic tests of the Bulk Shielding Reactor II,
the stainless steel, UOp-fueled, compact core for the BSF, have been
completed at the SPERT~I Facility of the National Reactor Testing Station,
and the core and control system are being returned to ORNL. Static tests
included uranium-foil flux mappings, vold~coefficient measurements, and
temperature=coefficient measurements. Dynamic tests, originally planned
to include tests of the self=limiting behavior of the core during damaging
power excursions, were terminated before core damage was appreciable, so
that the original core could be used in the BSF. It was observed that the
self=limiting behavior paralleled that of the APPR core P/8/l9, which was
previously tested at SPERT~=I in the range of periods down to 14 msec.
Since this behavior can be reliably applied to the BSR-=II and is well
understood on the basis of the energy model, the results obtained were
sufficient. Subsequent dynamic testing utilizing the full BSR=II control
system showed completely satisfactory performance of the control system.

Installation of the Model IV gamma-ray spectrometer at the BSF has
been virtually completed. Permanent reference points necessary for
calibration of the readout system are being established, and calibration
will shortly be completed. A test established that the weight of the
spectrometer and positioner causes only a small deflection of the beams
of the BSR building.

A crystal of CsI(Tl), 5 in. in diameter and 3-1/2 in. long, was
investigated for possible use in the Model IV spectrometer. Resolution
at 0.662 Mev was 9.2%, with a photofraction of 76%. Response was essentially
linear to gamma rays ranging from 0.5 to ~7.2 Mev., Decay time of the
light pulse, however, was about five times as long as in NaI(T1l). Com~-
parison of the photofraction given by the CsI(Tl) crystal with that of a
similar-sized NaI(T1) crystal showed about a 12% superiority in favor of

CsI(T1), a result, however, which may be due mainly to the greater density

 
 

of the CsI(T1l). A further comparison is being made by the use of the
Monte Carlo code mentioned above.

Data analysis is continuing for the experimental determinations of
the spectrum between 15 kev and 10 Mev of the prompt gamma rays from
thermal-neutron fission of U%3°, The remaining tasks are those of de-
termining the detailed absolute efficiency of the spectrometers for all
energies and translating the observed pulse-height spectra into the de-
sired absolute photon energy spectrum. Calibration of the spectrometers
has been handicapped by the very small number of monoenergetic radiocactive
sources which may be utilized as standards. Only yés8 (1.84 Mev), Na 24
(2.75 Mev), and Nté (6.1 Mev) have been available in the energy range of
interest. For these sources the shapes of the pulse~height spectral
response functions have been fitted using six parameters. An auxiliary
experiment utilizing the coincident 4.4~ and 12-Mev gamma rays from the
B (p,y)cr?* reaction has been initiated. The necessary 200~kev protons
are being supplied by the BSF 300-kv particle accelerator, and a thick
boron target is being used. It is hoped that a successful conclusion
to this experiment will preclude the necessity for further auxiliary ex-
periments, and data analysis may be completed. In connection with these
calibrations, the intensity calibrations of two of the high-pressure 4=
ionization chambers used for source strength determinations have recently
been checked by experiments.

A detailed and up=to~date listing of data on gamma-ray spectra re-
sulting from thermal-neutron capture in a total of 67 different nuclei
has recently been completed by NDA on an ORNL subcontract. The principal
sources of data were the work of a Canadian group led by Kinsey and
Bartholomew and a Russian group led by Groshev. Energies of from 0.3 to
12 Mev are covered by the survey. Tables of data giving the number of
photons per 100 captures for several energy intervals between O and >9
Mev, the highest-energy gamma ray, and the average number of photons per
capture have been compiled for 67 nuclei. Other tables give energies
and intensities of the discrete gamma rays resulting from thermal-neutron
capture by the 25 nuclei of greatest interest for shielding problems.

The data are also displayed in a collection of graphs.

AR a
--i

12. Tower Shielding Facility

The Tower Shielding Reactor II (TSR-II) has now been completely
assembled at the Tower Shielding Facility (TSF) and has been successfully
operated at low power. Initial assembly of the reactor was retarded by
the extensive fitting of parts found to be necessary to maintain the close
tolerances required to obtain the proper cooling water flow around the
fuel plates. In addition, a gradual distortion found to be occurring in
the fuel assemblies required machining and heat socaking of fuel segments
before proper assembly was possible, A delay was also suffered when
control system wiring was damaged in a small fire caused by a short circuit
in a temporary heater used to prevent freezing of water lines durlng
assembly.

The last major component of the TSR-II to be designed and assembled
was the control turret. The necessity, from flux symmetry considerations,
of minimizing the solid angle above the core occupled by materials other
than air and not spherically symmetric with the core complicated the de-
sign. The physical size of all members and mechanisms was held to a
minimum, resulting in a very compact assembly of control system components.

In order to assure safe initial operation of the TSR-II, a special
test setup was devised. The reactor was suspended in an elevated test
tank into which moderator water could be pumped from a separate reservoir.
A magnetically closed dump valve in the test tank could be rapidly released
to dump moderator at an appropriate signal from extra monitoring circuits
connected to the scram circuits, which were set at very low levels.

The calculations earlier proposed to be made of the TSR=II with final
geometry and materials have been completed with the use of the GNU-II code
with the IBM=704. The assumptions required by this code are fairly well
satisfied by the construction of the TSR~II. The calculations are being
used to predict the results of the initial critical experiments at the TSF.

Studies continue to optimize the flow distribution between the fuel
plates of the TSR~II., Some difficulty has been experienced in duplicating
with the actual fuel elements the results achieved with dummy fuel plates.
A part of the difficulty has been found to lie in the method of flow

 
 

S

testing. Careful tailoring of baffle plates, which proved successful with
dummy fuel elements, is expected to achieve satisfactory flow in the actual
fuel elements.

The spherical lead=and-water shield to be used to collimate the radia~-
tion at the detector has been delivered to the TSF. Before it is placed
in use it will be slightly modified to accommodate a 9 by 9 in. NaI{T1l)
crystal which will be used for gamma-ray spectroscopy.

Photomultiplier tubes to be used for spectral measurements have been
tested for gain shifts, and several tubes showing less than a 2% change
over the range of interest have been selected for use. A 400=channel
pulse-height analyzer was received, and minor modification and redesign
of circuits has been performed to improve linearity and live time error.
Amplifier stability has also been improved by circuit modification. A DD2
amplifier equipped with 4~-usec pulse~shaping cables has been tested both
with pulses from a LiI crystal and pulses from a test pulse generator and
has been found to perform satisfactorily.

The use of surface-barrier silicon diodes in neutron spectroscopy is
being investigated. The configuration tested employs a pair of silicon
diodes ~1 cm square, with their contiguous faces coated successively with
gold and Li®F. The Lié(n,a)T reaction is used to produce simultaneous
charged-particle events that are detected as a collected charge in the
diodes, which feed an amplifier and analyzer system. The thermal-energy
peak appears at 4.78 Mev, the Q for the reaction, and the response is ex-
pected to be linear with energy. Efficiency, calculated from accepted
Li® cross section values, is of the order of 107°% for 1-Mev neutrons,
whereas the theoretical resolution is 5.3%. The spectrometer is expected
to be useful over the range from 750 kev to 5 to 10 Mev. Tests with high-
energy neutrons are planned at the ORNL High Voltage Laboratory.

An automatic data~acquisition and processing system for the TSF is
being developed. The system includes a punched~-paper-tape analyzer out-
put, automatic data plotting from the tape, and a processing system for
gamma-ray spectra, using first the IBM-704 and eventually the IBM=7090

computers. In the gamma-ray processing, response functions determined

 

xvii
 

I
by interpolation of single-energy data are least-squares fitted to the
unknown spectrum, without any assumptions concerning the unknown spectrum.
Nevertheless, artifical smoothing of the unknown spectrum required to in-
sure convergence of the method results in a net resolution somewhat better
then that of the spectrometer. Results obtained thus far indicete that

typical experimental response functions are well fittegd by a nonlinear

analytical function.

 
 

PART 1. MATERIALS RESEARCH AND ENGINEERING
                   

AT ST

     

RS S A A R

   

 
 

1. MATERTIALS COMPATIBILITY RESEARCH

Reactions of Columbium with Low=-Pressure Gases

 

Reactions with Oxygen

The effects of oxygen pressures in the range between 3 X 10~
and 5 X 1072 mm Hg on the reaction rates of oxygen with columbium and
the correlation of the reaction rates with the formation of the oxides
of columbium were studied at 850, 1000, and 1200°C. As the oxygen pressure
was increased from the lowest pressures, the following principal reactions
occurred: (1) solution of oxygen in columbium, (2) precipitation of CbO
and the formation of CbO, nodules, (3) CbO, film formation, and, finally,
at the highest pressures, (4) Cb,0s5 formation. Reactions which could be
attributed wholly to the solution of oxygen in columbium were not en-
countered except at 1200°C and an oxygen pressure of 3 X 1077 mm Hg. The
rate curve obtained under these conditions is shown in Fig. 1l.1l. The
specimen showed a weight increase equivalent to 0.22% 0,, which is well
within the solubility limit of oxygen in columbium at 1200°C, as reported
in the literature.l:? The shape of the curve reflects the diffusion
phenomenon in which the diffusing species is approaching its saturation
concentration. No physical evidence of a reaction was apparent; however,
x~ray diffraction patterns indicated the presence of weak CbO lines.
Since this specimen was slowly furnace~cooled to room temperature, it is
assumed that CbO precipitated during cooling. The reaction postulated

for the solution of oxygen is

Cb + 1/2 0, — Cb [0]

where [0] signifies oxygen in solid solution.

 

1A, U. Seybolt, "Solid Solubility of Oxygen and Columbium, " J. Metals
774 (June 1954),

°R. P. Elliott, "Niobium-Oxygen System,"” p. 5, Armour Research
Foundation Report ARF 2120-3 (April 27, 1959).
 

The first transition observed in the reaction rate coincides with
internal oxidation. The rates of internal oxidation (precipitation of
CbO) and the formation of CbO, blisters on the surface of the metal were
previously reported to be linear.? Further analyses of the data have
shown that the effects of the oxygen pressure on the reaction rate

constants can be described by the following equation:

Rate constant = a.(i—gpgg) s
where a and b are constants and p is the oxygen concentration at pressures
between 1 X 10~% mm Hg and that pressure which results in film formation.
Below 1 X 107% mm Hg, negative deviations from the above relationship
are observed, and the reaction rate approaches a linear relationship with
the pressure. Investigators of other gas-metal systems have Observed
similar effects and have attributed the deviation to the fractional cover-
age of the metal surface with a monomolecular layer of adsorbed gas.4
The reaction products have been identified as the oxides of columbium
by x~ray analysis, and it appears that the following two principal reactions

occur:
¢b [0] — CbO | (internal precipitation)
and
CbO (surface blisters) + 1/2 0, — CbO, (surface blisters) .
A second transition in the reaction rate from linear to parabolic

occurs with extended exposure time or increased oxygen pressure, as

shown in Fig. 1.2a and b. The shape of the rate curve is quite similar

 

3"ANP Semiann. Prog. Rep. Oct. 31, 1959," ORNL=-2840, pp. 6~9.

4C. J. Smithells, "Gases and Metals," pp. 77-136, John Wiley and
Sons, New York (1937).

 

 
 

to that of the curve for the simple solution of oxygen (Fig. 1.1);

however, in this case, a protective film of ChO, forms and the decrease

in the reaction rate is attributable to this film. The metal specimens
that produced the oxidation rate curve shown in Fig. 1.2c were embrittled
to the same extent as those specimens tested under cénditions which re-
sulted in a linear oxidation rate. Since film formation occurs immediately
at these pressures and temperatures, the embrittlement has been concluded

to occur in the following three successive steps:

Cb + 1/2 05 — Cb0O, (film)
CbO, + Cb — 2 CbO (subfilm)
CbO (subfilm) — Cb[O]

A third transition from the parabolic to a linear reaction rate was
observed at the highest pressures studied, as shown in Fig. 1l.2d. Rate
curves of this type are also characteristic of the oxidation of columbium
at 400°C and oxygen pressures of 1 atm. The formation of Cb,05 on the
specimens coincided with the transition from parabolic to linear. The

reaction which accounts for this transition is assumed to be
2 CbO2 + 1/2 0, — Cby0s5 (nonprotective)

The oxide formed in this case is black and adherent.

Reactions with Air and Mixtures of Oxygen and Argon

 

The studies of the reactions of columbium with low=pressure air have
been extended to include tests at 1200°C. At an alr concentration of
5 X 10™% mm Hg a linear reaction rate constant of 6.3 X 1074 mg/cmz-min
was Observed. In contrast, at an equivalent oxygen pressure a reaction
rate constant of 4.5 x 1073 mg/cmz-min was obtained in pure oxygen.
Mixtures of argon and oxygen also lowered the reaction rate in comparison

with the reaction rate in puare oxygen.
 

 

 

 

 

UNCLASSIFIED
ORNL-LR-DWG 46488

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

046 ///J
A
/.
—~ 042 7 *
o
£
g
o
E
z 008
3 S
(o] v
|-
= 004
$ ./’ 1200°C
o’ OXYGEN PRESSURE: 3x10°° mmHg
/
o l
o 50 100 150 200
TIME (min)
Fig. 1.1l. Rate of Solution of Oxygen in Columbium.
UNCLASSIFIED
ORNL-LR-DWG 49547
4 ’ 4 { T
| o
T & ‘ /»/. j
& 1 E L —
g 3 | ga—ue? L 3 } ,.’/.
L P o |
o - . E -8
E - = ./
-~ 2 ’/ N =z > o [ y
z .t = vy .
< // 0 ./ ‘
o -~ 1000°C T g
- - s
5 377 | OXYGEN PRESSURE: 1+0*mmHg —| 2 4|, 27— o 4000°C “-#
Lg .{‘4' = ./' w ® 1200°C
o i 3 ‘ \ Ky ' OXYGEN PRESSURE: 5% 10-4mm Hg
[ I i | l L | ! .
0
0 500 1000 1500 0 100 200 300
{a) TIME (min) (b) TIME {min}
I 2.8 ‘
i ) L I .,
. | 2.4 - A
./
—~ -
- e g 20 - 7
§ 3 o
g‘ --—"._J. -_CE:‘
= - ® z
zZ : —t —
g ‘ " g
~ .
E i ‘ .'/? . I’j—:
@ . | | o L .
* L)
= o 850°C = / 850°C
OXYGEN PRESSURE: 4x40™*mmHg 04t OXYGEN PRESSURE: 2x407 mmHg
/7 f
* | ‘ ‘ 0% J E
o 100 200 300 0 100 200 300
(c) TIME {min) (d) TIME (min)
Fig. 1.2. Effect of Time, Temperature, and Oxygen Pressure on the

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Rate of Oxidation of Columbium.

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 
 

 

Although several mechanisms for oxidation of columbium by oxygen
have been observed, as mentioned above, the pressure conditions under
which they occur are in a narrow range of only a few microns at the
temperatures investigated. Further, although the reaction rates at the
low pressures are several orders of magnitude less than in oxygen or
alr at atmospheric pressure, the rates are sufficiently high to cause
significant oxidation of columbium in a few hours.

It is significant that these studies indicate a lower than expected
rate of oxidation of columbium in low-pressure air. In this respect,
the effect of the nitrogen content of the air on the surface reactions

appears to be quite important, as suggested previously.3

Effects of Oxygen Contamination and Exposure to Lithium
on the Tensile Properties of Columbium

 

 

Columbium specimens to which various amounts of oxygen had been
added were tested, as discussed previously,5 to determine the effect of
exposure to lithium on the room-temperature tensile properties. The
specimens were exposed to lithium for 100 hr at 1500°F. In order to
determine the effect of the oxygen additions in the absence of lithium,
specimens were contaminated to approximately the same oxygen concen-
trations as those of the specimens tested previously and then heat treated
in argon for 100 hr at 1500°F. The curves presented in Fig. 1.3 summarize
the results of room-temperature tensile tests of these specimens. It may
be seen that prior to exposure to lithium, additions of oxygen to columbium
increase the tensile strength at approximately a linear rate (in the range
shown), with no apparent loss in ductility. When exposed to lithium,
however, both the strength and ductility of the oxygen~contaminated
specimens decrease drastically., Additions of less than the minimum oxygen
concentration required to support the corrosion process result in an in-
crease in the tensile strength upon exposure to lithium. Additions of

oxygen in excess of the amount required for the corrosion process cause

 

S"ANP Semiann. Prog. Rep. Oct. 31, 1959," ORNL=2840, p. 24.

e SN S RS P
 

3 UNCLASSIFIED
(x10°) ORNL-LR-DWG 49548
©0

 

50

i /]

 

 

 

 

 

2 HEAT TREATED IN ARGON
- FOR 100 hr AT 1500°F

}—

o

&

5 /\' ‘

= P EXPOSED TO LITHIUM

2 FOR 100 hr AT 1500°F

—

 

20 / \. i ’3

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

®
\-'
i
]
O 1
20 \
\ |
o HEAT TREATED IN ARGON
45 O‘-.--'——— / | oy
~ 7
2 \ |
2 10
E YEXF’OSED TO LITHIUM
S
& /"
—
wl \
° .\
S~e
0 \“
0 500 1000 t500

OXYGEN CONCENTRATION {ppm)

Fig. 1.3. Effect of Oxygen Contamination and Exposure to Lithium
on the Room~=Temperature Tensile Properties of Columbium.

 
i e e et R £ s o e e e e o e

progressively deeper corrosive attack and subsequent decreases in tensile
strength and duvuctility.
ey

Mass~Transfer Effects in Systems Consisting of
Dissimilar Materials

Studies of mass~-transfer effects in three-component systems were
continued with tests of a system consisting of columbium=-zirconium alloy
specimens in a NaK bath contained in type 316 stainless steel. A system
typical of that being used in these studies is shown in Fig. 1.4.
Columbium~zirconium alloy tensile specimens were used as test samples
in order to obtain a quantitative measure of any change in tensile
strength and ductility as a result of carbon and nitrogen pickup. Con=
trol specimens were heat treated in argon for 500 hr at the test tempera-
ture in order to obtain comparative data. The results of these tests and
previous tests of unalloyed columbium in sodium and NaK are summarized in
Table 1.1.

The columbium and columbium-zirconium alloy tensile specimens from
tests in which the columbium~to=-stainless steel surface area ratio was
greater than 1 showed an increase in carbon and nitrogen content. An
increase in the tensile strength and, in general, a decrease in the
elongation were found for both columbium and columbium~zirconium alloy
specimens., Some similar changes occurred in the tensile properties of
the control specimens, but they do not explain all the differences ob=
served. When type 318 stainless steel, which contains about 1%
columbium, was substituted for type 316 stainless steel, no decrease
in the amount of carbon and nitrogen pickup was noted. The tensile
properties of the stainless steel specimens that were included in these
tests were found to be unchanged. The stainless steel container presented
much more surface area than the stainless steel tensile specimens and was
the major source of the carbon and nitrogen. The oxygen content of both
the columbium and the columbium-zirconium alloy specimens decreased in

nearly all cases as a result of exposure to the sodium or the NaK baths.,
ARR—
ORNL-LR-DWG 46643R

    
    
  
 

 

 

 

TYPE 310
STAINLESS STEEL
1
TYPE 316
STAINLESS STEEL
flfl
24
1
5 /
10 4
Ch—Zr ALLOY
TENSILE SPECIMEN
r
g
TYPE 316
STAINLESS STEEL DIMENSIONS
SPACERS IN INCHES

Fig. 1l.4. Schematic Diagram of System for Testing the Compatibility
of Dissimilar Materials.

10

 

 
Table 1.1. Effect of Mass Transfer on Tensile Propertles of Columbium and Columbium=Zirconium Alloys®
Tested in Sodlum or NaK in Stainless Steel Containers

Specimen thicknesst 0.040 1in.
NaK composltiont 56% Na, 44% K

 

 

 

 

74 zeondum Change in Impurity Tensile Strength (psi) Elongation in 2.5 in. (%)
Content. Exposure Concentratlons of
Spe;i?en of Rath Teizrze?ggia— Surfizziofirea Time Specimens (ppm) At Room Temperature At Test Tempersture At Room Temperature At Test Temperature
S%zgigin (br) c N, 0, As-Received Control Test As-Recelved Control Test As-Annealed Control Test As=Recelved Control Test
Specimen Specimen Specimen Specimen Speclmen Specimen Specimen Specimen Specimen Specimen Specimen Specimen

1587°¢ 1 Na 1700 3 500 +430 +710 =311 53 500 68 000 47 000 47 000 74 000 19 8 7 1 1
1696°¢ 0.6 NaK 1700 6.5 500 +230 +690 =200 47 000 47 700 86 700 29 600 27 900 43 700 25 19.5 9.5 11.3 10 6.8
1713¢ 0.7 NaK 1800 20 500 4782 41040 =100 47 000 48 600 79 300 29 600 26 000 33 000 25 19.0 14.5 11,3 16.0 10.0
16974 0.7 NaK 1700 6.5 500 +642 +1340 =800 47 000 47 700 85 100 29 600 27 200 41 500 25 19.5 14.5 11.3 10 6
1592¢ 0 Na 1700 13.5 1000 4323 +1254 429
1561 ¢ 0 Na 1600 10 1000 #4452 415 +60
1562¢ 0 Na 1700 10 1000 4682 <+1320 =50
1563°¢ 0 Na 1700 0.1 1000 4112 =22 =60
1696¢ 0 NakK 1700 6.5 500 +238 4659 =70 22 800 34 400 48 200 6 550 : g 800 17.5 18 4.5 19.2 20.0
1713¢ 0 NaK 1800 20 500 +788 4859 =110 22 800 40 000 50 400 5 700 6 100 g 000 17.5 16 6.0 17.2 12.0 7.9
16974 0 NaK 1700 6.5 500 +18 41919 =800 22 800 34 400 46 200 6 550 g 770 17.5 18 11.5 19.2 19.0

 

aAll columbium and columbium-alloy materials utilized by the corrosion laboratory
in these tests were wrought material annealed at ORNL for 2 hr at 1600°C.

bRatio of surface area of stainless stesl to surface area of specimen.
cType 316 stainless steel container.

dType 318 stainless steel container.

11
 

 

 

 

Metallographic examination of the columbium and the columbium-
zirconium alloy specimens showed surface layers 0.2 to 1 mil in thickness
that were identified as CbC and CboN. These layers will be machined off
and the specimens will be reanalyzed to determine whether there was any

carbon and nitrogen pickup in the specimens.

Compatibility of Columbium-Zirconium Alloy
and Haynes No. 25 Alloy

A thermal=convection~loop test has been conducted to determine the
compatibility of Cb—0.8% Zr alloy with Haynes No. 25 alloy (nominal
composition: 50% Co—20% Cr—15% W—10% Ni—2% Fe) in a NaK environment.
A 25-in.=-long columbium-zirconium alloy liner was placed in the hot leg
of the loop, and five sheet tensile specimens (0.040 in. thick) were
suspended inside this liner. The presence of any mass transfer and its
effect on the tensile properties of the alloy could be determined with
this test system. The hot-leg temperature was held at 1600°F (871°C)
for 500 hr, while NeK (56-44 wt %) was being circulated in the loop.

Post-test examinations showed weight gains of 0.3 to 0.7 mg/in.2 on
the tensile specimens. The results of chemical analyses, presented in '
Table 1.2, indicated carbon transfer from the Haynes No. 25 loop material
to the Cb—0.8% Zr alloy specimens during test. An increase of 190 ppm
of carbon in the tensile specimens and a loss of 330 ppm of carbon from
the Haynes alloy were noted. No significant changes in nitrogen content
were detected in either the columbium-zirconium alloy or the Haynes
alloy. The brittle layer that is characteristically formed on columbium
and its alloys during testing in NaK or lithium was observed metallo~-
graphically, but it was only 0.1 mil in thickness. No significant
alterations in the tensile strength or elongation of the Cb—0.8% Zr alloy
were found. The ratio of the surface area of the Haynes alloy to the
surface area of the columbium-zirconium alloy specimens in the loop was

3:l, which is to be compared with a 10:1 ratio in the PWAC~11l reactor.

12
Table 1.2. Results of Chemical Analyses of Loop Material and
Specimens of Columbium=-Zirconium Alloy After Exposure to
NaK in a Thermal=Convection Loop Constructed of

Haynes No. 25 Alloy

Hot~leg temperature: 1600°F (871°C)
Test duration: 500 hr

 

Impurity Concentrations (ppm)

 

Item Analyzed Before Test After Test

 

C 0, N Hy C 0, N, H

 

Cb—0.8% Zr alloy o
tensile specimens 180 980 190 3 370 670 210 2

Haynes No. 25
loop material ge0 2000 190 27 530 130 140 18

 

a1Total cross sectilon analyzed.

b'I‘he before-test analyses were made on a total cross section;
the after-test analyses were made on 15-mil turnings from the in-
side surface of the loop wall.

Compatibility of Boiling Potassium and Type 316
Stainless Steel

In a recent study of various nuclear power systems designed to
generate electricity for space vehicles (see Chap. 8, this report),
it was concluded that the potassium and rubidium showed substantially
greater promise than any other potential working fluids. The study
indicated the need for information on the compgtibility of stainless
steel and potassium under boiling conditions to determine whether
corrosion, mass transfer, or related problems might limit the life of
such systems.

A test has been completed in which potassium was boiled in a
type 316 stainless steel capsule for 100 hr at 1600°F (871°C). The test
system 1s illustrated in Fig. 1.5. Type 316 stainless steel specimens
were exposed in the ligquid, vapor, and the liguid=vapor zones of the

test system.

13
x
M~
o
O
8% 5 - 8
b = - >
@ o
no w % =
44 g 8 3
0 w = w)
Z | Q w
o w o _AIH
z w & o
= @ i
o v

 

STEEL SPECIMEN
SEMICYLINDRICAL
HEATERS

 

 

 

POTASSIUM LIQUID
871°C (1600°F)
STEEL SPECIMEN
TANTALUM GETTER
SPECIMENS

 

 

 

e

T

INSULATION

 

 

6in.

 

 

1

18in.

 

System for Tests of Steel Specimens in Boiling Potassium.

1.5,

Fig.

14

 
 

Tantalum getter specimens that were located at the bottom of the
test capsule increased in nitrogen and oxygen content during the test,
but & net weight loss occurred as a result of transfer of tantalum to
the inner surface of the capsule in the bath zone., Tantalum was not
detected on the capsule surface in the vapor zone. Examinations made
after the test revealed a significant weight change, +2.8 mg/in. 2, of
the type 316 stainless steel specimen that was in the liquid-vapor zone,
whereas the weight changes of the specimens exposed in the vapor and
liquid zones were negligible., No attack was detected on any of the
specimens by metallographic examinations.

Similar tests of 500 hr duration have been conducted in types 310
and 316 stainless steel capsules, but examinations of the specimens have
not been completed.

Type 316 stainless steel specimens have also been tested in boiling
potassium in a thermal~-convection loop. The test system is shown in
Fig. 1.6. The duration of the test was 200 hr, the boiling-zone tempera-
ture was 1550°F (843°C), and the cold=leg liguid=-zone temperature was
1430°F (776°C). The total power input to the heaters was 7.5 kw, with
approximately 5.5-kw input to the boiling region.

Examinations of the stainless steel specimens (1 X 2 X 0.040 in.)
after the test showed the following slight weight changes of all the

specimens$

Weight Change

 

Specimen Locatlon
P (mg/cm?)

Boiling zone —1.7
Vapor-liquid interface

in boiling zone —2.3
Vapor region in hot leg 3.5
Vapor~liquid interface

in cold leg +0. 4
Ligquid zone in cold leg —0.3

The steel specimen exposed in the vapor region exhibited the greatest

weight loss. This is attributed to the effect of refluxing of potassium

15
 

UNCLASSIFIED
ORNL-LR-DWG 44927R

       
   
    
 
   
  
    
 

12-in. HEATER 3 L~ INSULATION

™~/ -in~SCHED. 40 PIPE

 

 

 

 

 

 

 

 

A 3 12-in. HEATERS
EXIT HOT AIR OF L |1
/é {
|| < tiouo Leve
L~ 12-in. HEATERS
COOLING FINS ~——_| [ E x
—1550°F e
COOLING 7
AIR IN 7

 

 

8-in. HEATER

1430°F 12-in. HEATERS

8-in. HEATER

 

8-in. HEATER
100°F
«— THERMOGOUPLE
LOCATIONS COLD TRAP
| — TYPE 316 STAINLESS
STEEL TABS

 

 

Fig. 1.6. Design of Thermal-Convection Loop for Testing Steel
opecimens in Boiling Potassium.

16

 

 
 

in this region during operation of the loop. No significant differences
could be detected in the specimens, and no attack was observed metallo-
graphically. Metallographic examination of various sections of the loop
revealed no attack of the type 316 stainless steel container. Chemical
analyses of the specimens and loop material showed no indications of
potassium pickup.

Based on the rate of heat extraction from the cold leg, calculated
from the air-flow velocity and temperature rise, the potassium-flow rate
in the system was approximately 64 grams per minute.

It is evident that loop tests of much longer duration will be re-
quired to evaluate type 316 stainless steel in boiling-potassium systems.
A second loop test is now under way with a higher boiling-zone tempera-
ture and a higher flow rate. The duration of this test will be several
thousand hours. The stainless steel inserts are in the form of tensile
test specimens so that the effects of the exposure on the mechanical

properties of the material can be determined.

17
 

2. STUDIES OF THE AGING OF COLUMBIUM~BASE ALIOYS

Aging Studies on Wrought Sheet

Further investigations of aging in columbium~zirconium alloys have
been completed. As reported previously,1 specimens of various columbium=
zirconium alloys were annealed 2 hr at 1600°C (29lO°F) in vacuum, sealed
in evacuated quartz capsules, and aged for various times at several
temperatures. These specimens were then tensile tested at room tempera-
ture or under vacuum at elevated temperatures. After completion of the
tests, the question arose as to what effect varying the annealing tempera-
ture would have on the aging phenomenon,

Several sheets of Cb~1.15% Zr (heat XM=339, PFYU) were received from
Pratt & Whitney Aircraft Division, CANEL, to be used in aging experiments.
Tensile specimens were stamped from these sheets and were annealed 2 hr
at 1200°C (2190°F), 1400°C (2550°F), and 1600°C (2910°F). The specimens
were then aged in evacuated quartz capsules O, 25, 50, 100, 200, and 500
hr at 1700°F and tensile tested under vacuum at 1700°F. The results of
these tests are shown in Fig., 2.1 The specimens annealed at 1600°C
(2910°F) showed a marked increase in ultimate tensile strength that
reached a peak of approximately 59 000 psi after approximately 75 hr at
1700°F. The alloy then overaged and the strength dropped to approximately
46 000 psi after 200 hr. The elongation of the alloy was initially 11%
but dropped to 5% after 500 hr. The specimens annealed 2 hr at 1400°C
(2550°F) and 1200°C (2190°F) showed some variations in tensile strength
and elongation that might be attributed to aging; however, the effects
were not as pronounced as for the specimens annealed at 1600°C.

The microstructures of the specimens annealed 2 hr at 1600°C
(2910°F) and aged 50, 100, 200, and 500 hr at 1700°F are shown in Fig.
2.2. Specimen (a) shows some precipitation in the grain boundaries and
a generally clean grain structurej specimen (b) shows more gain-boundary

precipitation, with some precipitation in the matrix; specimen (¢) shows

 

1"ANP Semiann. Prog. Rep. Oct. 31, 1959," ORNL-2840, pp. 15-17.

18

 
 

e et A R R i

ORNL—LR-DWG 44801A~

-

 

 

 

 

 

 

 

(x10%) t l ;
® ANNEALED 2hr AT 1200°C R
A ANNEALED 2 hr AT 1400°C i

_1.15 Zr o
Cb O ANNEALED 2 hr AT 1600°C »
60 AGED AND TESTED AT {700°F
{6) { ) INDICATES % ELONGATION

. (617

= 55

o

T )]

|_

o

&

W50

|_

wy

lu (4)

2 3(5)

ul 45

-

i

k (6)

s

=

-

D

 

 

 

 

 

40
©) ",,/’r " (9)
% |
(10) ‘

 

 

 

 

 

 

 

 

 

-_'_'_-—-—.
35 ®) 8 \\\:9, :: i
(9)
30
0 25 50 100 200 500

AGING TIME (hr)

Fig. 2.1. Effect of Annealing Temperature on Aging Cb—1.15% Zr
Alloy Specimens.

a general coarsening of the precipitate, with some depletion of the matrix
around the grain boundaries; and specimen (d) shows some further coarsening
of the precipitate; however, it appears that precipitation is essentilally
complete.

The data indicate that the element or elements responsible for the
observed aging in the alloy are not put completely into solution by
annealing at 1200 or 1400°C. Annealing at 1600°C appears to be sufficient,
however, to dissolve the material in question., Upon subsequent aging,
the dissolved material precipitates and causes the observed aging phenomenon.

Tt was suggested that the aging observed in wrought sheet may have

been due to silicon contamination from the quartz capsules used during

19
 

 

B OB {dEddd{ad kT
5 Bl
= + \
& :
mn.__ T ¢ - .
I__ % ...1. g
o
vota ._c »_.‘ r
3 v - :
. | ) ..._ H . .
. 52 i ... 3
L\ S

 

o swm §H99998835899 oy

T-33234
h

m——

(a)

20

AGED 100 hr

AGED 50 hr

 

m_ mw_ I._m,_

 

 

 

AGED 500 br

AGED 200 hr

Effect of Aging Time at 1700°F on the Metallographic Appearance of Cb—l.15% Zr.

HNO5-HF-H,S04-H,0 (25-10-10-50, parts by volume).

Fige 2ol

Etchant:
2 As a check on this possibility, specimens of the

the aging treatments.
Cb—l.15% Zr alloy (heat XM-339 PFYU) were annealed 2 hr at 1600°C and
aged in argon-filled columbium capsules for 0, 25, 50, 100, and 200 hr.
Thefigfggffiens were then tensile tested at 1700°F in vacuum, and the test
results are compared in Fig. 2.3 with those for specimens aged for the
same times and at the same temperatures in quartz capsules. Although
some differences are noted in the ultimate tensile strengths and
elongations, it is still evident that the aging phenomenon 1s present
even when the silicon content of the alloy is very low. ©Silicon analyses
of representative specimens used in the aging studies are presented in
Table 2.1.

A comparison of the silicon concentrations and the tensile strengths
of samples 1 through 6 indicates that there is little or no correlation
of the silicon concentration with the observed changes in tensile strength
as a function of aging time.

Samples 7 through 10 show the variation of silicon concentration
with distance from the specimen surface. The as-received specimen showed
essentially uniform distribution of silicon. The specimen annealed at
1600°C and then aged 120 hr in quartz showed a high concentration of
silicon in the outer 0.003=in. layer and a slight decrease in the next
0.006-1in. layer.

The data for samples 11 through 13 compare the silicon contents and
tensile strengths of three Cb—0.75% Zr specimens from a previously re-
ported series of aging tests.? As for samples 1 through 10, there was
no apparent correlation of silicon content with tensile strength.

Three heats of columbium-zirconium alloys with nominal zirconium
content of 1% are presently on hand that were supplied by Pratt & Whitney
and are designated PFYU, PGVE, and PGTF. Tensile specimens have been
stamped from the alloy sheets in both 0.040-in. and 0.045-in. thicknesses.
After aging in quartz capsules for 120 hr, the 0.040-in.-thick specimens

 

2"symmary of Data Presented by Pratt & Whitney Aircraft at ANPO
Materials Meeting, Germantown, Md., Feb. 1, 1960, CNLM-2300.

3"ANP Semiann. Prog. Rep. Oct. 31, 1959," ORNL-2840, p. 15.

21
(103

ULTIMATE TENSILE STRENGTH {psi}

22

60

55

50

45

40

35

R
ORNL-LR-DWG 49549

 

 

© N

Cb—1.45 % Zr (HEAT XM—-339) ANNEALED
2hr AT1600°C AND AGED IN EVACUATED
QUARTZ CAPSULES

(6}

 

 

T

(5)

 

(9)

 

 

 

 

 

 

Ve

? NG
/// \\\“@) (4)

P
\ |
///<:\\Cb—hisZr(HEATXM-339) ¢

/7 ANNEALED 2 hr AT 1600°C

AND AGED IN ARGON-FILLED

(9)§ COLUMBIUM CAPSULES
z’/,(m

 

 

{5)

 

 

C
{41}

N

 

 

 

 

 

30

 

 

 

 

SPECIMENS AGED AND TESTED AT {7O0°F
{ )} INDICATES % ELONGATION IN 2%2 in.

L

 

 

 

 

 

 

Flg. 2.3.

20 50
AGING TIME (hr}

100 200

500

Effect of Container Material on the Aging Behavior of
Cb—1l.15% Zr Specimens.

 
Table 2.1. Results of Neutron Activation Analyses for Silicon in Columbium~Zirconium
Specimens Used in Aging Studies

 

 

‘o s Tensile ,
Samole Siliecon Strength Elongaticon
Sample Description P Test History Concentration g o In 2 in.
No. (ppm) at 1700°F (%)
PP (psi) ’
Cb—1% Zr, heat XM=339, 1 As received 20
0. =in.-thi

040~1n.~thick sheet Annealed at 1600°C 18, 31 36 340 11

3 Annealed at 1600°C; 19, 19 57 700 6

aged in S5i0, for
50 hr at 1700°F

L Annealed at 1600°C; 14, 24 56 400 6
aged in 810, for
100 hr at 1700°F

5 Annealed at 1600°C; 30, 40 47 100 4

aged in 5i0, for
200 hr at 1700°F

6 Annealed at 1600°C; 30, 32 46 400 5
aged 1in S1C; for
500 hr at 1700°F

Cb—1% Zr, heat XM-339, 7 As received 39, 56
0.003-in. cut from

surface
Same as above, but 8 66, 60

0.006~1in, cut taken
below 0.003=1in. cut

Cb—1% Zr, heat XM-339, 9 Annealed at 1600°C; 116, 78 52 600 5
0.003-in. cut from aged in Si0, for

surface 120 hr at 1700°F

Same as above, but 10 101, 45

0.006=1in. cut taken
below C.003-1in. cut

Cb—0.75% Zr, Fansteel 11 Annealed at 1600°C 26, 20 40 000 11

12 Annealed at 1600°C; 53 46 100 3
aged in Si0, for
500 hr at 1700°F

13 Armealed at 1600°C; 24 61 200 5

aged in Si0, for 750
hr at 1500°F

 

engile test conducted at 1500°F.

will be tensile tested at 1700°F. The 0,045-in.-thick specimens will
be machined after aging to 0.040 in. in thickness toc eliminate any
surface contamination received during aging and will then be tensile

tested at 1700°F.

23
Aging of Fusion Welded Material

Studies of the fusion welding of columbium-base alloy sheet material
have been continued. The primary emphasis in this work has been on the
determination of the bend characteristics of aged welds after they have
been exposed to various thermal treatments in vacuum.

The aging studies were carried out on welds made on 0.062-in.-thick
Cb—1% Zr alloy sheet specimens prepared from material (heat XM-339)
supplied by Pratt & Whitney. In order to assure a standard shielding-
gas purity, all the welds were made in the inert-atmosphere welding

chamber shown in Fig. 2.4. Use of this chamber minimized the possibility

*UNCLASSIFIED
PHOTO 43496

 

 

 

Fig. 2.4. Inert-Atmosphere Chamber for Columbium=Alloy Welding
Studies.

24

 
of inadvertent pickup of contamination, but it should be emphasized that
chemical analyses (see Table 2.2) have shown that the pickup of contami-
nation in trailer-shield welding4 is minor and approximately the same

as in controlled atmosphere chamber welding. Both welding processes
appear to be very satisfactory for the Jjoining of columbium and its alloys.
All the welds compared in Table 2.2 were made using the same bottle of
helium, which analyzed 99,998% He.

Table 2.2. Typical Chemical Analyses of
Cb—1% Zr Alloy Weld Metal Produced by
Different Welding Procedures

 

Impurity Concentrations

 

 

(ppm)
0, N H,p C
Base metal 130 170 1.0 230

Inert-atmosphere= 150 180 9.0 200
chamber weld
metal

Trailer~shield 140 190 4.0 200
weld metal

 

Three different encapsulating procedures were used to prepare the
weld specimens for aging. The initial procedure was to insert the
specimens in quartz tubes which had been evacuated to a pressure of
10"? mm Hg. Subsequently, however, the influence on the bend behavior
of slight amounts of silicon pickup during aging was considered, and,
in an attempt to guard against the possibility of such pickup, two
alternative encapsulating methods were investigated. The first of theée
involved wrapping the Cb—1% Zr specimen in tantalum foil before en-
capsulating it in an evacuated quartz tube. The other method was to

insert the specimen in a columbium tube which was filled with pure argon.

 

4Tbid., pp. 38%1.

25
The columbium tube was, in turn, encapsulated in an argon-filled
stainless steel tube,

Chemical analyses of the welds aged in evacuated quartz tubes are
shown in Table 2,3. The contamination with oxygen, nitrogen, and silicon
appears to be minor, and no correlation with ductility can be seen. The
encapsulating procedures and aging conditions utilized thus far in this
study and the results of the bend tests on the aged specimens are described
in Table 2.4. The data indicate that aging is most pronounced at 1500°F.
Embrittlement of the welds occurred after 25 hr at 1500°F regardless of
the encapsulating procedure. Some bend tests were made at 400°F in the
early experiments because Pratt & Whitney had reported improvements in
ductility at that temperature. The tests at 400°F were discontinued when
no significant differences were observed in the ORNL studies. Additional
specimens will be aged to determine the extent of aging at 1300 and 1400°F,
and duplicate specimens are being aged at 1700°F in argon~filled columbium

capsules.

Table 2.3. Results of Chemical Analyses of Aged Cb—1% Zr
Alloy Weld Metal

 

 

 

Impurity Concentrationsa Room-
{ ppm) Temperature
Sample Bend-Test
0, No Si Behavior
As~=received sheet 170 160 20, 39, 56
Weld metal aged 250
hr at 1300°F 150 160 38, 40 Ductile
Weld metal aged 250
hr at 1500°F 190 150 52, 45 Brittle
Weld metal aged 250
hr at 1600°F 220 170 53, 49 Ductile
Weld metal aged 250
hr at 1700°F 180 bty 43 Ductile
Weld metal aged 250
hr at 1800°F 260 150 33, 40 Ductile

 

 

 

——

aAnalyses were made of weld metal only; oxygen and
nitrogen were determined by the vacuum-fusion method, and
silicon was determined by activation analysis.

26

 
 

Table 2.4.

Description of Aging Capsule, Aging Conditions, and Results

of Bend Tests of Cb—1% Zr Weld Metal (Heat XM-339) Specimens

Welding conditionss

current, 100 amp

speed, 5 in./min
helium atmosphere

Bending conditions:

transverse guided bend

2T bend radius, where
material thickness

T

 

Results of Bend Tests

 

 

Aging Aging
Aging Capsule Temperature Time
(°F) (hr) At Room At 400°F
Temperature
Evacuated quartz tube 1300 100 Ductile Ductile
(specimen not wrapped) 1300 250 Ductile Ductile
1400 5 Ductile
Evacuated quartz tube 1400 25 Ductile
(specimen wrapped in
tantalum)
Evacuated quartz tube 1400 100 Ductile
(specimen not wrapped)
Evacuated quartz tube 1500 1 Ductile
(specimen wrapped in 1500 2 Ductile
tantalum) 1500 5 Ductile
1500 10 Ductile,
cracks
present
1500 25 Brittle
1500 50 Brittle
1500 100 Brittle Brittle
Evacuated quartz tube 1500 250 Brittle Brittle
(specimen not wrapped)
Evacuated quartz tube 1500 500 Ductile,
(specimen wrapped in cracks
tantalum) present
1600 1 Ductile
1600 2 Ductile
1600 5 Ductile
1600 10 Ductile,
cracks
present
1600 25 Brittle
1600 50 Brittle

27
Table 2.4 (Continued)

 

Results of Bend Tests

 

 

Aging Aging
Aging Capsule Temperature Time
(°F) (hr) At Room At 400°F
Temperature
Evacuated quartz tube 1600 100 Brittle Ductile,
(specimen not wrapped) cracks
present
1600 250 Ductile,
cracks
present
Evacuated quartz tube 1600 500 Ductile,
(specimen wrapped in cracks
tantalum) present
Argon=filled columbium 1700 1 Ductile
tube
Evacuated quartz tube 1700 2 Ductile,
cracks
present
Argon=filled columbium 1700 5 Ductile,
tube cracks
present
Evacuated quartz tube 1700 10 Brittle
1700 25 Brittle
Argon=-filled columbium 1700 50 Ductile,
tube cracks
present
Evacuated quartz tube 1700 100 Ductile
Argon~filled columbium 1700 250 Ductile
tube
Evacuated quartz tube 1700 500 Ductile
Evacuated quartz tube 1800 1 Ductile,
(specimen wrapped in cracks
tantalum) present
1800 2 Ductile
1800 5 Ductile
1800 10 Ductile
1800 25 Ductile
1800 50 Ductile

28

 

 
 

Table 2.4. (Continued)

 

Results of Bend Tests

 

 

Aging Aging
Aging Capsule Temperature Time
o At Room o
(°F) (hr) Temperature At 400°F
Evacuated quartz tube 1800 100 Ductile
(specimen not wrapped) 1800 250 Ductile
Evacuated quartz tube 1800 500 Ductile
{specimen wrapped in
tantalum)

 

Samples that illustrate the bend-test results reported in Table 2.4
are shown in Fig. 2.5. Figure 2.5a shows a weld that was completely
ductile; Fig. 2.5b shows a ductile weld with cracks present; and Fig. 2.5c
shows a brittle weld. The extent of the cracking in the ductile weld
is indicated in Figs. 2.6 and 2.7. The cracks are both transgranular
and intergranular. The microstructure of the crack in the brittle weld
shown in Fig. 2.5c is illustrated in Fig. 2.8. The fracture appears to
be primarily transgranular,

Examinations of the microstructures of these weld specimens have
shown that there is some change with aging. The change is illustrated
by comparison of Figs. 2.9 and 2.10. Figure 2.9 shows a weld in the
as-welded condition, and Fig. 2.10 shows a weld after aging for 100 hr
at 1500°F. The typical grain-boundary phase present in weld specimens
following aging is shown in Fig. 2.10. The relationship of this phase
to the brittle behavior of certain aged welds has not yet been determined.

Curves showing the time-dependent trend in weld ductility when
the welds are aged at different temperatures are presented in Fig. 2.11.
Although the ductility scale on the ordinate is purely qualitative, the

relative effects are evident.

29
= %
T

u; e
- W %
&

UNCLASSIFIED UNCL ASSIFIED
_;-,—_'_ B, Y-33346 e N
- b

      

UNCLASSIFIED
¥-33347

 

Fig. 2.5. Typical Cb—1% Zr Alloy Weld Bend-Test Samples. (a)
Ductile sample that was aged 250 hr at 1300°F. (b) Sample aged 250 hr
at 1600°F showing ductile behavior, with cracks present. (c) Brittle
sample that was aged 250 hr at 1500°F. 3.5%.

30

 
UNCLASSIFIED
Y-33953

 

0.4Q in./ DIV

Fig. 2.6. Cross Section of Ductile Cb—1% Zr Alloy Bend~Test
Sample with Cracks Present.

 

Lad
 

  
  

UNCL ASSIFIED
Y-33964

Fig. 2.7. Photomicrograph of a Crack Shown in Fig. 2.6. Both transgranular and inter-
granular cracks are evident. Etchant: HNOj3;, H,S0,, HF, H,0. 50X.
 

 

UNCL ASSIFIED
¥.33310

   

 

0.0

 

INCHES

 

 

 

 

 

 

 

 

» o

0,086

\ 4 0.08

2,

0.07

>

o

0

Fig. 2.8. hotomicrograph of a Crack in a Brittle Cb—1% Zr Alloy Weld. The cracking

appears to be primarily transgranular. tchant: HNO;, H,SO,, HF, H;0. 50X,

 
 

 

UNCLASSIFIED
Y-33073

&
INCHES

 

003 )

 

2 000X

o=

 

 

 

Fig. 2.9. Photomicrograph of As=Welded Cb—1% Zr Alloy. No grain-boundary phase is evident.
Etchant: HNO;, H»S04, HF, H;0. 1000X.
 

 

 

 

 

_ UNCLASSIFIED
. ' 0 Y-33042
. .’
t

. { u

i v w

: o ! T

¥ o
[ . A - E-.—

g ¢
3
!
: = — e o2
—_— —— e == — = =
. 003
- &

*

' o

o

. S

004

 

 

Fig. 2.10. Photomicrograph of Cb—1% Zr Alloy Weld After Aging for
100 hr at 1500°F. A definite grain-boundary precipitate is evident.
Etchant: HNO3, H»S50,;, HF, H»0. 1000X.

DRNL-LR-DWG 48980
BOOCF iEn

+ DUCTILE

~ DUCTILE

  
 

 
 
   
 

DUCTILE WITH CRACKING

  

DUCTILE WITH CRACKING

BRITTLE

BRITTLE

 

 
     
 

PO0OF BO0°F
| BUCTIL 1800
—_— e —— - # A
DUCTILE WITH / RLETHE Wi SREimin
CRACKING T el e K] S
BRITTLE BRITTLE
! 2 5 10 50 10K 250 500 | 2 25 50 10G 250 500
TIME {hrl TIME (hr)

Fig. 2.11. Influence of Aging Time at Various Temperatures on the
Bend-Test Behavior of Cb—1% Zr Alloy (Heat XM=339).

35

 
 

Microhardness data measurements were made on some of the specimens

in the as-aged condition, and the results are reported in Table 2.5. It

does not appear that there is any reproducible or systematic correlation

between the hardness data and the bend behavior.

 

 

 

Table 2.5. Results of Microhardness Measurements on Aged Welds
of Cb—1% Zr Alloy
Hardness, DPH Values
Aging Aging (500~-¢ 1load) Room-Temperature
Time Temperature Bend Test
(hr) (°F) Base Heat- Behavior
Affected Weld
Metal
Zone
250 1300 92 97 100 Ductile
250 1500 132 160 169 Brittle
250 1600 132 160 169 Ductile, cracks
present
250 1700 112 125 165 Ductile
250 1800 138 132 Ductile
1 1700 178 145 169 Ductile, cracks
present
2 1700 137 134 156 Ductile, cracks
present
10 1700 183 148 173 Brittle
2O swy 1700 178 200 Brittle
100 1700 156 173 Ductile
100 1500 211 Brittle
0 173 Ductile

 

36

 
3. MECHANICAL PROPERTIES INVESTIGATIONS

Effects of Alloying Additions on the Mechanical Properties
of Pure Columbium

 

 

Studies of the effects of alloying additions of nitrogen, oxygen,
hydrogen, and carbon on the mechanical properties of pure columfiium have
been continued. The results of creep tests at 1800 and 1850°F in nitrogen,
wet and dry argon, and wet and dry hydrogen were summarized previously.l
The current studies include tests for evaluating the properties of -
columbium to which small alloying additions are made prior to testing.

Bend and tensile tests are being used to evaluate the room-temperature
mechanical properties of these specimens. The effect of distribution of
the contaminant is being studied by testing samples as-contaminated, after
homogenization, and after homogenization and annealing at a lower tempera~=
ture to precipitate the contaminant as a second phase. Creep tests of
alloyed specimens in inert environments are being used for evaluating

the elevated-temperature mechanical properties of these alloys.

The following procedures are being used for adding carbon, oxygen,
nitrogen, and hydrogen to columbium. Different amounts of carbon are
added by annealing columbium for various times in an argon—10% methane
environment at 1100°C. An argon—benzene mixture has also been found to
be a satisfactory annealing atmosphere. Although both these environments
result in the addition of some hydrogen to the specimen, the hydrogen
content can be reduced to sufficiently low levels by annealing in vacuum
at 1300°C. Nitrogen is added by annealing at 1200°C in purified nitrogen
at a pressure of 1 atm. Oxygen is added by annealing at 400°C in oxygen
at a pressure of 1 atm. The oxide formed by heating under these conditions
for less than 50 hr is quite adherent and can be diffused into the sample
by a subsequent high-temperature anneal. Hydrogen is added by annealing
in purified hydrogen at a pressure of 1 atm. The analytical data presented
in Table 3,1 illustrate the suitability of these techniques for adding

one contaminant at a time.

1YANP Semiann. Prog. Rep. Oct. 31, 1959," ORNL-2840, pp. 44—47.

37
Table 3.1.

Contaminated by Various Treatments

Analytical Data on Columbium Specimens

 

Specimen
Treatment

Contaminant Analysis (wt %)

 

Oxygen

Nitrogen

Hydrogen

Carbon

 

As~received

Annealed 48 hr in

0.0340
0.0340

0.0048
0.440

<0.0001
0.0008

0,0140

nitrogen at
1200°C

Annealed 24 hr in
argon~methane
mixture at
1100°C

Annealed 24 hr in
argon-benzene
mixture at
1100°C

Annealed 8 hr in
oxygen at 400°C
and homogenized
by annealing 2
hr at 1300°C

0.0310 0.0027 0,0077 0.0450

0.0370 0.0024 0.0090 0.0520

0.0760 0.0033 <0.0001

 

Several of the specimens to which various amounts of nitrogen,
oxygen, and carbon have been added have been tested in bending. The
specimen was supported at two points 0.750 in. apart, and the load was
applied at a point half way between the two supports. In such tests the
specimen is subjected to shear and bending forces, and its behavior is
recorded as a load-deflection curve.

1 X 1.250 X 0,040 in.

The test-specimen dimensions were
A1l the specimens were annealed 2 hr at 1300°C
prior to alloying, and some of the specimens were given this same
annealing treatment to homogenize them after alloying. The data ob~
tained in the tests completed thus far are summarized in Tables 3.2,
3.3, and 3.4. Note that carbon, oxygen, and nitrogen increase the
strength of pure cclumbium. However, the data indicate that if nitrogen

and oxygen are present in large quantities they reduce the ductility.

38

 
 

 

 

Table 3.2. Effect of Nitrogen on Bend Properties of Columbium
Time in N> Cgifiggfegf Maximmum Deflection Specimen
at 1200°C ) Homogenized Load at Maximum p

(hr) Specimen® (1b) Load (in.) Behavior

(wt %) :

0 0.0048 Yes 81.2 0.25 Did not

crack

2 0.0420 No 127.2 0.25 Did not

crack

2 0.0483 Yes 155.2 0.25 Did not

crack

8 0.1274 No 113.2 b 0.029 Cracked

(93.2) (0.011)

8 0.1386 Yes 194.,7 0.145 Cracked
24 0.3010 No 85.2 0.010 Cracked
24 0. 2905 Yes 139.2 0.135 Cracked
48 0.5241 No 80.2 0.009 Cracked
48 0., 4447 Yes 163.2 0.036 Cracked

(102.2) (0.014)

 

®As determined by weight change.
Pannealed 2 hr at 1300°C.

“Numbers in parentheses indicate load at which first crack formed;
other numbers indicate loads at which complete failure occurred.

Table 3.3.

Effect of Oxygen on Bend Properties of Columbium
Homogenized by Annealing for 2 hr at 1300°C After Alloying

 

Time in O,

Oxygen Content

Maximum

Deflection

 

at 400°C of Specimen Load at Maximum Spec1@en
(hr) (vt %) (1b)  Load (in.) Behavior
0 0.034 81.2 0.25 Did not crack
4 0.050 1.6 0.25 Did not crack
8 0.078 114,2 0.25 Did not crack
65 0.300 01.2 0.012 Cracked

 

39
Table 3.4. Effect of Carbon on Bend Properties of Columbium

 

Carbon Content Maximum Deflection

 

. . Specimen
History of Specimen Load at Maximum Behavior
(wt %) (1b) Load (in.)
As-received 0.0140 81,2 0.25 Did not crack
Annealed for 24 hr 0. 0450 84,2 0. 25 Did not crack
at 1100°C in
AI'_CH4
Annealed for 24 hr 0.0520 91.9 0. 25 Did not crack
at 1000°C in
Ar—CgHg

 

The data for the series of specimens to which nitrogen had been added
illustrate the relative effects of equivalent amounts of nitrogen when
present as a surface layer and when distributed throughout the specimen

by homogenization.
Fatigue Studies of Inconel

The fatigue properties of Inconel were studied under a subcontract
at Battelle Memorial Institute,? and the results of that work have been
correlated with the results of tests at ORNL. The objective of the
Battelle investigation was to measure the plastic strain which oceurs
during each cycle of a conventional stress-fatigue test and to use the
strain data to obtain strain-fatigue curves. These curves are shown
in Fig. 3.1 and are compared with strain-fatigue data obtained from

tensile and cyclic creep tests performed at ORNL, >

 

’R. G. Carlson, "Fatigue Studies of Inconel," BMI-1355 (June 26,
1959).

3C. R. Kennedy and D. A. Douglas, "Plastic Strain Absorption as a
Criterion for High Temperature Design," ORNL-2360 (April 17, 1958).

40

 
 

UNCLASSIFIED

ORNL-LR-DWG 43290

1000 A
F hr

1500| 70 P
1 360 |CONVENTIONAL FATIGUE
1 3600 |CONVENTIONAL FATIGUE
1 70 CYCLIC CREEP
13 3600 |CONVENTIONAL FATIGUE
150 TENSILE DUCTILITY
1300 TENSILE DUCTILITY

- 0
~ o S

5, TRUE PLASTIC STRAIN RANGE (%)
e

 

0.01 L
oX 1 10 100 1000 10000 100,000

N ,CYCLES TO FAILURE

Fig. 3.1l. Strain-Fatigue Properties of Inconel Rods at 1300 and
1500°F.

Conventional fatigue data for tests at 360 cycles per hour agree
with cyclic creep data. The higher frequency fatigue curves are displaced

to the right. A similar frequency effect was reported previously.4

 

4R, W. Swindeman and D. A. Douglas, "The Failure of Structural
Metals Subjected to Strain=Cycling Conditions," Trans. ASME 81, 203-12
(June 1959).

41
4. ALLOY PREPARATION

Electron-Beam Melting of Columbium Alloys

Experimental melting studies have been initiated with the electron-
bombardment furnace shown in Fig. 4.1 from the operator's station. Part
of the control panel is visible at the left. The power supply that
delivers 60-kw d-c power (20 kv, 3 amps) to the electron gun is located
at the left of the furnace. The gun is mounted on top of the drift-
tube, which is the cylindrical water~cooled section at the top center
of the picture. A water-cooled mold for casting 2.9-in.~diam ingots 1is
shown clamped to the bottom support plate of the furnace. Molds are
also available for 2.3- and 3.8-in.-diam ingots; however, most of the
melting to date has been done with a button plate.

The electron-beam melting furnace is being used for the preparation
of columbium alloys with carefully controlled compositions. Initially,
the material of most interest is a Cb—l% Zr alloy with wvarious amounts
of carbon added.

A series of melts was made to determine to what extent the final
composition of the alloy could be controlled by adjusting the starting
composition and the melting time. The columbium starting material used
was a 1 1/8-in.-diam bar cut into 5/8-in. lengths (approximately 100 g
each). Todide zirconium was added as strips. The buttons were held
molten on one side for a predetermined time and then turned over and
held molten for an equal time on the other side. The results of this
melting study are presented in Table 4.1. The results indicate the
need for additional melting experience before alloys can be prepared
with assurance of achieving a preselected composition. However, it is
interesting to note that alloys with very low oxygen and nitrogen
contents have been prepared that retained 1% or more zirconium.

There appears to be anomalies in the carbon analyses, based on the

work of Smith,l and therefore the analyses are being repeated. ©Smith

 

1H, R. Smith, Jr., Electron Beam Melting Techniques, Chap. 14 of
"Vacuum Metallurgy,” R. F. Bunshah (ed.), Reinhold, New York, 1958.

42

 
UMCL ASSIFIED
PHOTO 35332

 

 

 

Fig. 4.1. Electron Bombardment Furnace,

 
Table 4.1, Effect of Melting Time and Initial Composition on the
Final Composition of Selected Columbium=~Zirconium Alloys

 

Zirconium and Impurity Concen-

 

 

Melting L 0tvial trations of Melt
\ Zirconium
Melt No. Time@
. Content
(min) (vt %) Zr 0 N C
(wt %)  (ppm) (ppm)  (ppm)
Columbium stock <0.05 320 60 120
Zirconium stock 1.00 62 10 80
34 4 1.5 1.06 70 37 90
35 8 1.5 1.08 82 26 100
36 12 1.5 0.67 46 18 100
37 16 1.5 0.79 26 10 110
38 20 1.5 1.04 60 22 120
39 12 0 <0, 05 47 13 130
40 12 2.0 1.21 34 13 120
41 12 1.0 0.52 28 16 140

 

aTotal time: one=half on each side of button.

reports removal of oxygen from columbium preferentially by the volatili~-

zation of CbO or Zr0O with some competition from the carbon deoxidation

(CO) reaction. It is expected, therefore, that some decrease in carbon

should have been achieved; in no case should there have been even the

minor increases indicated.

An electron-beam melted columbium-zirconium alloy button containing

4% zirconium was readily cold rolled from approximately 3/8 in. to
0.040 in. Previous work with inert-atmosphere arc-melted buttons in-
dicated a susceptibility to cracking during cold rolling of columbium
alloys containing as little as 1.5% zirconium.

Other materials which were successfully electron=beam melted

include Mo, W, Ta, V, U, Y, Fe, Ni, Al,03, CbC, and ZrC. Uranium

monocarbide was melted, but it showed a tendency to change composition

from stoichiometric UC to UC + UC,. High vaporization rates which
precluded successful melting were encountered with Cr, Zr0,, and UO,.

No systematic evaluation has yet been made of these electron~beam

melted materials.

 

 
5. CERAMICS RESEARCH

Preparation of Beryllium Oxide

Oxalate Process

 

The preparation of sinterable beryllium oxide by the oxalate process
was studied further. The sintering characteristics of beryllium oxide
seem to be influenced by particle size and degree of hydration of the
beryllium oxalate from which it is calcined. Additional microcrystalline
beryllium oxalate trihydrate was prepared by rapid cooling in a silver
container, as described previously.l

When a concentrated solution of beryllium oxalate was boiled, beryl=-
lium oxalate precipitated that was of much finer particle size than that
obtained by rapid cooling of a solution. Also, the material so obtained
was the monohydrate instead of the usual trihydrate.

An attempt was made to prepare anhydrous beryllium oxalate for
comparison with the hydrates. Oxalyl chloride was refluxed with beryl-
lium hydroxide, but the product was merely beryllium hydroxide containing
some chloride.

An attempt was made to prepare improved spectroscopic standards for
use 1n the analysis of beryllium oxide. To separate beryllium from im-
purities, the perfluorobutyrate was extracted with diethyl ether from an
aqueous solution containing ethylenediamine-tetraacetic acid. Phosphorous
was absent from the beryllium oxide made from the ether layer, otherwise
there was no significant improvement over previous standards. Spectro-
scopic methods are not very good for the determination of phosphorous
in beryllium oxide.

A very fluffy beryllium oxide has been prepared. When hot, concen-
trated, aquecus oxalic acid was reacted with double the molar quantity
of beryllium hydroxide, a clear, viscous solution was formed. The

solution was colloidal in nature; that is, it displayed the Tyndall

 

L"ANP Semiann. Prog. Rep. Oct. 31, 1959," ORNL-2840, pp. 64—65.

45
effect. Oven drying of the solution usually formed a glass. Ignition
of the glass at 700°C formed a highly expanded white mass with a bulk
density approximeting 0.01 g/cm3. (Bulk density of commerical beryllium
oxlde 1s about 0.25 g/cm3.) The material is & poorly crystallized

beryllium oxide, the usefulness of which has not yet been evalusted.

Calcining BeC,04¢3H,0 to BeO

 

Further studies were made of the calcination process in which beryl-
lium oxide is obtained by decomposing BeC;0,4+3H,0 in order to define the
variables that may affect the sinterability of BeO powder obtained by
this process. Samples of BeC,0,°3H,0 were celcined, in air, to BeO
under dynamic conditions. Phase changes occurring during the process
were indicated by differential thermal analysis, and weight changes were
recorded continuously with respect to temperature. Attempts were then
made to define each step in the calcination process by gquenching samples
at selected temperatures and analyzing the samples by x=ray diffraction
and optical techniques.

The differential-thermal=-analysis and weight=-change curves recorded
simultaneously with BeC,0,4+3H,0 was calcined at the rate of 7°C/min to
beryllium oxide are shown in Fig., 5.1. The first indication of decom~
position of the trihydrate is at approximately 80°C. The reaction proceeds
rapidly to where a "shoulder" (quench point No. 1) appears on the
differential~thermal~analysis curve at approximately 130°C. The sig-
nificance of the "shoulder" was not determined. At this temperature
and heating rate the weight change corresponds very closely to the
theoretical weight loss for a dihydrate (~12%); however, if such a
compound exists at this temperature, it is too unstable to be quenched-
in and is not detectable by high~-temperature x=ray diffraction. Only
two phases, BeC,0,°3H,0 and BeC,0,+H,0, were shown to exist from the
beginning of decomposition until decomposition was completed at approxi-
mately 150°C, where the monohydrate was the only phase detected. The
monohydrate phase is stable up to approximately 250°C, where a second

decomposition reaction begins that is completed, as indicated by the

46

 
 

differential~thermal-analysis curve, at approximately 330°C, and a
noncrystalline phase appears. This phase has not yet been identified,
although a characteristic carbonate reaction is noted when it is treated
with HCl. Beryllium oxide first appears at approximately 370°C as a

very poorly crystallized material, according to the x-ray diffraction
pattern. The calcination process is completed as the temperature is
raised to approximately 400°C, and the material develops rapidly into

a well-crystallized beryllium oxide powder. The completion of the process
is marked by the exotherm on the differential-thermal-asnalysis curve at
this point.

Thermograms of the calcination process at different heating rates
are presented in Fig. 5.2. X=ray analyses show the phases present on
the lower temperature weight plateau to be BeC,0,-H,0 and to BeO on the
final weight plateau. These data show the calcination process to be
rate dependent with respect to temperature, that is, the slower the
heating rate the lower the temperature at which reactions are initlated
and completed. Reaction rates are now being determined under static

conditions.
Sintering Studies

Studies were initiated in an effort to relate calcining history to
fired densities of beryllium oxide powders obtained by decomposing
BeC50,¢3H,0 to BeO. The first approach was to compare densities ob-
tained by calecining the trihydrate directly to BeO at 900°C with densities
obtained by slow conversion to the monochydrate before final calcination.
To eliminate the variables of purity and chemical preparation, each ex-
periment was conducted on the same batch of starting materialj each of
these batches was divided into two parts, one-half being first converted
to the monchydrate and then calcined and the other being calcined from
the trihydrate under the same conditions. In order to eliminate any
possible variation in sintering conditions, specimens from both halves
of each batch were fired at the same time. The results of preliminary

experiments are presented in Table 5.1. With batches 1 and 2 from which

47
200

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

UNCLASSIFIED
ORNL-LR—DWG 43194A
T T T T
DIFFERENTIAL - THERMAL —ANALYSIS CURVE
------ WEIGHT-CHANGE CURVE |
O QUENCH POINTS ‘
100 ’ l ‘
| 1
. HEATING RATE;: 7°C/min
0 bmmema |
\T\
¢ \\
_'IOO : \\ e
7
= ‘\ 3 |
£ >e— |
w -200 S ;
2 \ka 6
2 \
O A
= \
I 300 7 35
2 \
ut
= U \\
2 Ty
-400 1
\
¥
i A
: \
-500 L
\
\
\
v |
-600 ;
\s.--—-.-..i._
|
-700 i
0 100 200 300 400 500 600
TEMPERATURE (°C)

Fig. 5.1.

Differential~Thermal-Analysis and Weight-=-Change Curves
Obtained During Calcination of BeC,04;°3H,0 to BeO.

100

UNCLASSIFIED
ORNL-LR—DWG 43435A
HEATING RATE: 3°C/min

=== = HEATING RATE: 7°C/min

80

60

WEIGHT LOSS (%)

20

 

100 200 300 400 500 600
TEMPERATURE (°C)
Fig. 5.2.

Thermograms of the Calcination Process for Decomposing
BeC,0,4°3H,0 to BeO.

48

 
 

Table 5.1. Effects on Fired Densities of BeO Powders of
Variations in the Heat Treatments Prior to Calcination
for Converting BeC,04¢3H,0 to BeO

 

 

Batch Heat Treatment of Phase Present Fired Densitya
No BeC,0,4* 3H50 Before (% Theoretical)
’ Before Calcining Calcining
1 None BeC,0,+3H,0 78
Held at 110°C for
9 hr BeC,0,°H,0 88
2 None BeC 50,4+ 3H,0 76
Held at 150°C for
96 hr BeCQO4-H20 79
3 None BeC50,°3H50 95
Held at 110°C for
96 hr BeC50,°H,0 96
4 None BeC,0,4+ 3H,0 96
Held at 175°C for
96 hr BeC 50,4+ Hp0 %

 

aSpecimens fired at 1650°C for 1 hr in a carbon=tube in-
duction furnace containing a helium atmosphere,

poorly sinterable BeO powder was obtained by calcining directly from the
trihydrate, sinterability was improved by first converting slowly to the
monohydrate; this effect on sinterability was more marked when conversion
to the monohydrate was accomplished at 110°C than at 150°C. With batches
3 and 4 from which an easily sinterable powder was obtained by calcining
directly from the trihydrate, sinterability was not appreciably improved

by prior conversion to the monohydrate.

Evolution of Lithium During Sintering

In the interest of specifying limitations of the lithium content in
beryllium oxide powder to be fabricated for reactor application, ex-
periments were conducted to determine to what extent lithium would be
retained in a beryllium oxide body under various conditions of heat

treatment. The results of the experiments are presented in Table 5.2.

49
Table 5.2. Retention of Lithium in
Beryllium Oxide as a Function of
Heat Treatment

 

Temperature

 

* Heat Duration of Lithium
© ca Heating Content,
Treatment
4 (nr) (pm)
Starting
material 0 100
1000 24 35 -
1200 1 25
1400 1 15 -
1650 1 10

 

The data indicate that 65% of the lithium present in the starting
material is removed during the removal of volatile constituents before
sintering. During sintering, more lithium would be removed as the
sintering temperature was increased until only 10% would remain if final

densification were obtained by heating at 1650°C for 1 hr.

Analyses of Beryllium Oxide

Chemical Methods

Development work on chemical methods of analysis for the determination
of silicon, lithium, and magnesium in beryllium oxide has been satisfacto-
rily completed. Brief outlines of the methods selected are presented
belcw,

Silicon, Silicon is distilled as fluosilicic acid, H,5iFg, and
determined spectrophotometrically as the silico-molybdic acid complex?
in sulfuric acid solution. The distillation is carried out in a gold
still, since hydrofluoric acid is necessary in order to dissolve the

difficultly soluble BeO,. The distillate containing the H,SiF¢ is

 

°G. N. Cade, Anal. Chem. 17, 3723 (1945).

50

 

 
 

collected in a solution of sodium chloride in a polystyrene cylinder.
After evaporation to dryness in a platinum dish, the residue is dissolved
and the silico~molybdic complex is formed. For 10 to 100 y of Si, the
coefficient of variation is 3%. Samples as large as 1 g of BeO have been
used, although smaller samples are preferable because of time involved

in dissolving the sample; 0.5 g of BeO can be dissolved in HF in about

30 min under reflux conditions. The results obtained with this method
have been compared with those obtained by the spectrographic technique
for the determination of silicon, and the over=~all agreement for 10 to
800 ppm of silicon in BeO is 9%.

3 has Dbeen

Magnesium. The spectrophotometric method of Mann and Yoe
tested for the determination of magnesium. Magnesium forms a colored
complex with Magon [sodium—l—azo-2-hydroxy—3-(2,4-dimethylcarboxanilide)-
napthalene~1~( 2-hydroxybenzene-5-sulfonate) ] at pH 9. The method is
extremely sensitive; for example, 1 pg of magnesium in 25 ml of solution
has an absorbance of 0.05 in a l-cm cell. Beryllium interferes, however,
and must be removed. This has been accomplished by extracting beryllium
with 1 M sodium perfluorobutyrate, C3F7;COONa, at pH 3.5 with ethyl ether.
Magnesium is not extracted under these conditions. Traces of beryllium
remaining in the aqueous layer are rendered harmless by adding a solution
of polyvinyl alcohol, a dispersing agent, prior to the development of the
magnesium-Magon complex. The effect of other impurities in BeO on the
complex has not yet been ascertained.

Lithium. Lithium is determined by extracting it with 0.1 M dipivaloyl-
methane in ether from an alkaline scolution. Beryllium is kept in solution
by complexing with fluoride. The measurement of lithium is completed
colorimetrically with the use of the thoron method.# The method can be
applied for the determination of less than 1 ppm of lithium in BeO with

a coefficient of variation of 5%.

 

3¢. K. Mann and J. H. Yoe, Anal. Chem. 28, 202 (1956).
4P, F. Thomason, Anal, Chem. 58, 1527 (1956).

51
Development work is continuing on the determination of other impuri-

ties in BeO, such as fluoride, phosphorus, and iodine.

Spectrographic Analyses

Preliminary curves have been established for determinations of Al
(100~1000 ppm), Fe {100—2000), Mg (50—500), Ca {50-1000), and Si (75—2000)
in a Be0 matrix with the photoelectronic spectrometer. Elements such as
Bi, C4, Cr, Co, Cu, Pb, Li, Mn, Mo, Ni, K, He, Sr, Ti, Zn, Na, and Sn can be
determined to ppm levels by photographic methods.

Compatibility of Beryllium Oxide and Other Metal Oxides

A preliminary investigation of the compatibility of beryllium oxide
with several metal oxides of the groups II and III elements was conducted.
As a general approach, the various compositions were reacted on a platinum-
strip resistance heater and air-quenched from the melt. The samples
thus obtained were subjected to x-ray and microscopic examination. In
some instances, heating and cooling information was obtained by comparing
various compositions with a reference sample of beryllium oxide, the

thermal effects being noted with a differential couple.

BeO—Y,03

A seriss of Be0O—Y,03 samples was prepared by taking compositions in
10 mole % increments from 50% BeO—50% Y,03 to 90% BeO—10% Y,03, heating
them on a platinum strip until molten at temperatures less than 1600°C,
and then air~-quenching them. X-ray diffraction and microscopic ex-
amination indicated the 50/50 composition to be predominantly single-

phase birefringent material having an index of refraction of 1.75 %+ 0.05.

BeO—Sr0

Acicular, single crystals, some as large as 1.5 mm in length, were
grown from a melt of 60 mole % BeO—40 mole % Sr0. The spectrochemical
analysis of the crystals showed barium {of the order of 500 ppm) to be

52

 
 

the most predominant impurity. The index of refraction of the crystals
is approximately 1.77. Lattice parameter values of approximately 9.04
and 7.16 A, respectively, for the a and b axes have been determined from
a zero=-layer Weissenberg film. The ¢ dimension is still in question and

further studies are in progress.

Be0O—La,03 and BeO—MgO

 

Both the BeO—La,03 and BeOMgO systems have been investigated and
found by x-ray diffraction to exhibit compound formation. Beryllium oxide-
rich compositions of BeO—La,0; were melted on the platinum strip at tempera-
tures lower than those required for BeO—Y,03 mixtures. The BeO-MgO
reaction was found to take place at a relatively high temperature of ap=
proximately 1900°C. Resolution of information pertaining to this system
is complicated by the high volatility of MgO at temperatures in excess of
1700°C.

Be0—Ba0

Mixtures of BeO and Ba0 were made and melted on the platinum strip.
Although melts were readily obtained, no evidence of compound formation
was found. This may be dus in part to a reaction between the BaO and the

platinum strip.

BeO—Cal

Compound formation in the system BeO—CaO was noted previously.5 The
chemical formula (CaBe30s5) of the phase has been somewhat verified by
a chemical analysis that indicated 41.7 wt % Ca and 14.5 wt % Be. In an
additional investigation, compositions of 90 mole % BeO—10 mole % Cal to
10 mole % Be0O—90 mole % Cal were heat treated at 1000°C for two weeks and
air-quenched. Although not at equilibrium, a three-phase region of CaQ,
BeO, and an xCa0-yBeO compound was attained that eliminates the possibility

of a simple eutectic system.

 

SUANP Semiann., Prog. Rep. Oct. 31, 1959," ORNL~2840, pp. 75-76.

53
Two compositions, 60 mole % BeO—40 mole % Ca0 and 80 mole % Be0O—20
mole % Ca0, were subjected to heating and cooling and, at the same time,
were compared with a reference sample of BeQ. With bare Pt-Pt—10% Rh
thermocouples inserted directly into the centers of samples contained in
platinum cups, it was possible to note the temperature of the sample, the
temperature of the reference, and the measured differential temperature.
The preliminary data obtained with this arrangement have shown melting
upon heating at approximately 1335°C and 1370°C, respectively, for the two

compositions. .

Preparation of Refractory Oxides from Molten 3Salts

In experiments designed primarily for studying fuel-reprocessing
schemes for the molten-salt reactor, water vapor was found to react with
UF, dissolved in molten fluorides to precipitate uranium. ® The stoichi-
ometry of the evolved HF with the quantity of uranium removed from solution
indicated the formation of UO,. Similar experiments further indicated
that water vapor would react with BeF,; dissolved in molten fluorides to

& The possible

produce BeQ’ and would also form ThO, by reaction with ThF,.
application of these reactions for the preparation of refractory oxides
prompted further investigation.

Initially, helium saturated with water vapor at room temperature was
passed over a mixture of UF,; dissolved in molten LiF-NaF (60-40 mole %).
The relative reaction rate was controlled by a gas flow rate of approximately
500 cm3/min over a melt surface area of 1l.4 cm® maintained without stirring
at 800°C., During a reaction period of three weeks, a constant evolution
of HF was observed that corresponded to the conversion of approximately 2 g

of UO,; per day. Upon cooling, the reactiocn mixture was found to contain

UO, crystals that had the optical properties shown in Table 5. 3.

 

6J. H. Shaffer, "MSR Quar. Prog. Rep. Oct. 31, 1958," ORNL=2626, p. 92.

7J. H. Shaffer, "Chemical Reprocessing Status Meeting of September 28,
1959, " ORNL CF-59~10-97, Oct. 26, 1959, p. 5.

83. H. Shaffer, unpublished data.

54

 
Table 5.3. Optical Properties of Oxides Precipitated
From Fluoride Melts

 

 

Maximum
Oxide Character Refractive Index Color Particle
Size {(p)
U0, Isotropic 2,36 £ 0,02 Red 100
ThO, Isotropic 2,11 * 0,02% Colorless 120
BeO Uniaxial b
positive 1.715 = 0.005 (Nw) Colorless 100

 

%A value of 2.20 has been reported in the literature (p. 137,
Bulletin of the National Research Council, Number 118, published
by the National Research Council National Academy of Sciences,
Washington, D. C.); however, that measurement was for mineral
thorianite which is reported to contain as much as 33 wt % of UO,,
U053, and U30g as impurities. Consequently, the difference between
the literature value and the value in this table should not be
construed to imply that there is anything "unusual" about the ThO,
precipitated from the fluoride melts.

bThis value has been reported in the literature as 1.719
(p. 16 of reference in footnote a).

The experiment was repeated using ThF, and BeF, separately in
solutions of LiF-NaF (60-40 mole %). Examinations of the reaction pro-
ducts of these expsriments also yielded the anticipated results. The
optical properties of the ThO, and BeQ obtained are also shown 1n
Table 5, 3.

In a study of the coprecipitation of U0, and BeO from a mixture of
UF,; in LiF-BeF, (63-37 mole %), the experimental techniques were altered
by sparging the melt with helium to provide agitation. This modification
was used to suspend the U0, crystals in soluticn in an attempt to coat
these particles with BeO. Petrographic examination of the crystalline
products again verified the formation of U0, and BeO, and confirmed that
in some instances UO, particles (~20 microns in size) were coated on all
sides with BeO.

The successful preparation of pure refractory oxides by this method

will be contingent upon the absence of ineclusions in the oxide crystal

55
and the ease of separation of the product from the solvent material.
Consequently, experiments are now in progress to study the properties

of crystalline oxides produced from other molten-salt solvents.

56

 
e 3400 e b e A e e e 7 0 i e A L e L b AR s b 0 L i 5 ke St it B e o e 8 P . Vil e b —

6. ENGINEERING AND HEAT TRANSFER STUDIES

Molten Lithium Heat Transfer

Additional data, covering an extended Peclet modulus range, have
been obtained in the study of forced-circulation heat transfer to molten
lithium flowing turbulently in an electrical-resistance-heated tube. The
experimental system, as previously described,l utilized as the test section
a thick=-walled type 347 stainless steel tube (3/16 in. i.d., 7/8 in. o.d.)
which was partitioned by the power leads into a hydrodynamic entrance
region of 25 x/d and two heated regions each of 40 x/d. Tube=-wall tempera=-
tures were measured by thermocouples located both on the outer tube surface
anl within wells extending to within 0.010 in. of the inner surface.

The experimental data and calculated results based on the thermowell-
couple data are summarized in Table 6.1. The discrepancy still remains?
between the inner tube-wall temperatures calculated on the basis of the
thermowell~couple data and those calculated from the outer wall thermo-
couple measurements. There is some indication, however, that this dif-
ference may be associated primarily with the uncertainty in the value for
the thermal conductivity of the tube wall., The determination of the thermal
conductivities of several type 347 stainless steel specimens is in progress.

The heat transfer results presented in Table 6.1 are plotted in
Fig. 6.1. The experimental circumstances that dictated the division of
the data into the five series indicated in Table 6.1 and Fig. 6.1 are
given in Table 6.2, which shows that between the runs of series I and II
tha thermowzll couples were replaced and that between the runs of series
TT and IITI the lithium was partially changed and the thermowell couples
were again replaced. During the runs of series III, a partial plug In the
lithium circuit limited the Reynolds modulus to values below 40 000. Prior

to the start of the runs of series IV, the fluid was circulated isothermally

 

L"ANP Semiann. Prog. Rep. March 31, 1959," ORNL-2711, p. 77.
RU"ANP Semiann. Prog. Rep. Oct. 31, 1959," ORNL-2840, p. 55.

57
8%

 

 

Table 6.1, Lithium Heat Transfer Data and Results of Forced~Circulation Tests
Flow Heat
Series Egn qf/A Rate tm,i tm,o w,1 - (Btu/hi-ftz °F) Balance NRe NPe NNu
(Btu/hr.£t?) (1o/hr) (°F)  (°F) (°F) BRI CIVERD
I 10 81 530 369 703.1 717.8 4,75 17 160 1.013 30 050 1270 11.37
11 81 450 369 703.5 1718.0 4 .60 17 700 1.070 30 050 1270 11.70
12 76 880 246 706.8 727.3 5.40 14 240 0.999 19 850 834 9.42
13 76 290 492 708.9 719.3 4,35 17 540 1.020 39 830 1680 11.59
iI 14 76 190 244 700.6 721.0 4,00 19 050 0.992 19 960 844 12.62
15 75 370 367 702.8 715.8 3.45 21 850 0.965 30 050 1270 14.40
16 75 090 488 703.2 713.0 3.25 23 100 0.964 39 830 1680 15.00
17 75 480 603 706.3 714.2 2.60 29 030 0.964 49 400 2090 19.24
18 75 040 428 703.6 714.5 3.20 23 450 0.949 34 990 1480 15.54
I1T 1-60 58 590 61.6 696.5 754.8 7.55 7 760 0.933 5 070 215 5.19
2-60 68 940 88.8 692.1 742.2 8.07 8 540 0.892 7 290 306 5.66
3-60 88 890 213.5 692.4 718.0 7.35 12 090 0.936 17 400 736 8.00
4=60 92 650 337.0 694,1 711.7 6.85 13 520 0.974 27 460 1161 8.94
5-60 94 420 439.4 6©96.0 709.6 6.30 14 990 0.963 35 810 1514 9.90
6=60 100 800 486.7 690.6 705.4 7.70 13 100 1.087 39 580 1670 8.63
IV Al-60 79 040 66.6 680.1 752.0 8.40 9 410 0.922 5 460 230 6.24
A2=-60 80 260 123.4 689.1 731.5 7.20 11 150 0.922 10 080 426 7.39
A3-60 93 680 227.1 690.2 716.9 6.30 14 870 0.985 18 510 783 9.82
A4 =60 114 700 332.7 674.9 697.9 7.85 14 620 1.015 26 850 1140 9.56
A5=-60 114 700 434.0 678.2 695.6 7.15 16 050 0.996 34 910 1482 10.50
A6-60 121 100 540.5 673.9 688.9 6.25 19 380 1.018 43 310 1839 12.64
A'7-60 125 800 651.0 671.1 684.4 6.10 20 620 1.047 52 010 2210 13.42
A8-60 130 100 773.5 669.3 680.8 6.65 19 570 1.040 61 800 2626 12.74
Vv A9=-60 35 450 36.8 713.2 775.5 11.45 3 0% 0.985 3 080 130 2.08
A10-60 52 410 164.7 726.6 746.3 5.0 10 480 0.942 13 700 578 7.06

 
1 i e e 5 103 s £ 6 L e e 5t i e U el e e e P s meFna 7 e e ot et g i ) e - e bt e o G 1+ e e el

Table 6.2. Operational Summary of Lithium Heat Transfer Tests

 

 

Date Description of Loop Operation
8~4=59 Lithium circulating; run 3
8-5-59 Runs 4 through 6
Em7=59 Runs 7 through 9; lithium dumped to sump and frozen
9=3=59 Tube~wall thermowell couples installed
9-15-59 Series I; runs 10 through 13; lithium dumped to sump and
frozen
9=29-59 Thermowell couples replaced
10-1-59 Loop filled from sump with lithium used in test runs of

series I; series II, runs 14 through 18; lithium dumped
to sump and frozen

1-25=60 Thermowell couples replaced

2=12-60 Pump repaired; oxide noted on pump bowl and impeller;
lithium circulated at 800°F to dissolve oxide and dis-
carded; fresh low-oxygen=-content lithium charged to loop
through sump

2=-23~60 Series III, runs 1-60 through 6-60; system operating with-
out external circult guard heating; partial plugging
noted; lithium circulated with line heaters on and heat
sink heaters off; full flow regained

2-24-60 Seriles IV, runs Al~60 through A8-60; lithium dumped to
sump and frozen

3-4=60 Loop refilled from sump with lithium used in runs of series
IIT and IV} series V, runs A9-60 and A10-60; lithium
dumped to sump and frozen

 

for 16 hr while using the heat sink as a cold trap for reducing the oxygen
content of the lithium. The lithium was returned to the sump between the
runs of series IV and V. The data of each series were analyzed by least-
squares methods.

Two characteristics of the data are immediately apparent from Fig. 6.1:
(1) on the whole, the data are resclved into individual groupings with an
average deviation of ilO%, and (2) the slopes of the curves associated with

data obtained in each series of runs are nearly equal (0.30 * 5%)., If 0.3
is taken as the correct slope (a regression analysis in which all data

are taken as a unit gives the slope as 0.38), a line through the data

mean will have the equation

_ 0.3
Ny, = 1.251 N2 (1)

The scatter band is iBO%, corresponding approximately to three standard
deviations. Equation (1) is compared in Fig. 6.2 with the empirical

correlation of Lubarsky and Kaufman,~

Ny = 0.625 Ngé4 , (2)

the theoretical equation of Martinelli and Lyon,4

Ny, = 7 +0.025 NgéB , (3)

and the equation of Kutateladze et al.,”

— 0.73 yO+ 081
Ny, = 6.8 + 0.0765 Np.'2 Np. (4)

Thermal Properties of Reactor Materials

 

Thermal Conductivity of Columbium~Zirconium Alloys

The thermal conductivities of two additional columbium~zirconium

alloy specimens have been measured using the longitudinal-heat-flow

 

’B. Lubarsky and S. J. Kaufman, "Ligquid-Metal Heat Transfer," NACA
Report 1270 {1956).

“R. N. Lyon (ed.), "Liquid Metals Handibook," 2nd ed., NAVEXOS P-~733
(Rev.), June 1952,

°S. 8. Kutateladze et al., "Liquid-Metal Heat Transfer Media,"
Supplement No. 2, Atominaia Energiia, Translation by Consultants Bureau,
Inc., New York (1959).

60

 
UNCLASSIFIED

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

20 ORNL-LR-DWG 49548
®
2~
-
- A .~
‘,--.‘ A::'/'. A
.—/0’.
1) T
Q - - /(-\, A
CE) 8 - ant -"P-.
- ‘__a"A _.l-"”
— L -
5 =T
/‘
Z Y -
5
£ Q= SERIES I
4 ——e—— SERIES II
— =~ — SERIES III
—-A-— SERIES IV
u SERIES V
2 B
100 200 500 1000 2000 5000
Npe =NRe* Mpy, PECLET MODULUS
Fig. 6.1. Lithium Heat~-Transfer Data.
ORNUNCLL\SSIFIED
40 L-LR-DWG 49519
KUTATELADZE (REF 5) 4"
. 20 7
2 _~"MARTINELLI-
3 LYON (REF 4)
E — ///
w / //
3 ___—/
=z pe
< 8 g
ORNL-LITHIUM _ ==
(THIS REPORT) ==
6 -
==~ LUBARSKY —KAUFMAN (REF 3)
4 |
100 200 500 1000 2000 5000

Npg= N, Np, , PECLET MODULUS

Fig., 6.2. Comparison of ORNL Lithium Heat Transfer Data with Other
Liquid-Metal Heat Transfer Correlations.

6l
comparison-type apparatus described previously.6 The amounts of zirconium
in these two samples were somewhat higher (0.78 and 0.94 wt %) than in the
earlier specimens (0.48, 0.62, and 0.69 wt %).’ The results obtained

thus far are compared with the data of Tottle® and Fieldhouse et al.? for
pure columbium in Fig. 6.3.

The data of Fig. 6.3 are represented to within *7% by the equation
k = 26,91 + 0.00857 t

over the temperature range 200 to 900°F, where k is in Btu/hr°ft-°F and
t is in °F. No apparent distincticn can be made between the several

specimens on the basis of zirconium content.

Enthalpy of Lithium Hvdride

 

The enthalpy of a sample of lithium hydride has been determined using
a copper-block calorimeter. The results are presented in Fig. 6.4. For

the solid (100 to 650°C), the enthalpy above 30°C is expressed as
Hy — Hzo = —32.17 + 0.914 t + (71.30 x 10™°) t?
In the liguid phase, between 700 and 900°C, the enthalpy is given by

Hy = Hio = 291.47 + 1.988 t — (6.35 x 107°) t2

The heat of fusion at 680°C was found to be 694.4 cal/g.

 

6"ANP Semiann. Prog. Rep. March 31, 1959," ORNL-2711, p. 78.
7"ANP Semiann. Prog. Rep. Oct. 31, 1959," ORNL-2840, p. 59.
8. R. Tottle, J. Inst. Metals 85, 375 (April 1957).

°L., B. Fieldhouse, J. C. Hedge, and J. I. Lang, "Measurements of
Thermal Properties,' WADC-TR-58-274 (Nov. 1958).

02

 
 

ORNL-I_LR-DWG 49520

UNCLASSIFIED

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

40
"
L
+ 0 -
3 /d)_——__—— = -
3 / .. — —-‘——A
: TOTTLE (ref 8) / o A ’_.“——— 4! - A A
=
> 30 — "
— —-—'___—-
Q
2 B 0.94 wt % Zr
= e A 0.78 wt% Zr
3 S FIELDHOUSE ef a/ (ret 9) O 0.48 wt % Zr
r A 0,62 wt T Zr —
g ® 0.69 wt% Zr
o
L
I
—
20
0 100 200 300 400 500 600 700 800 900
TEMPERATURE (°F)
Fig. 6.3. Thermal Conductivities of Five Columbium~-Zirconium Alloy

Specimens Compared with the Data of Other Investigators for Pure Columbium.

2500

2000

1500

1000

Hy -Hzq, ENTHALPY {cal/ g)

500

UNCLASSIFIED

ORNL-LR-DWG 49524

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

.
o
o
g |
=
& e?
W |
a
=y
=1
ol
Z1
=
w
=y
’I
.’
/“/
®
—
/“/
/u
®
l}/
_/
100 200 300 400 500 600 700 800
TEMPERATURE (°C)
Fig. 6.4. Enthalpy of Lithium Hydride.

63

 

 

900
 

Boiling-Potassium Heat Transfer Experiment

The characteristics of a nuclear auxiliary power unit for satellite
application that utilizes a rubidium- or potassium=-vapor cycle have been
discussed in reports in this series. Because of its availability and low
nuclear activation, potassium now appears to be the better fluid.l® Lack
of information on two-phase heat transfer to ligquid metals, in general,
and to potassium, in particular, prompted the design of an experimental
system for measuring boiling and condensing coefficients for potassium -
under conditions of interest in reactor design.

In order to minimize pumping power, space, and weight, it is necessary
to vaporize as large a fraction of the liquid as possible per pass.
Furthermore, weight considerations impose a limitation on tube wall thick-
ness and, hence, on operating pressure. The problem, then, is one of
absorbing heat at a relatively high thermal flux level by the boiling
liquid under conditions of very high exit void fraction (resulting from
the combination of the weight per cent of liquid vaporized and the high
specific volume of the vapor phase). A study by Dengler and Addomsll of
heat transfer to water in saturated, forced-convection boiling at 20
psia in a vertical tube has shown that the heat-transfer coefficient de-
creases rapidly for values of the quality (weight fraction vaporized) in
excess of 0.7. This reduction results from a "dry-wall" condition
assoclated with the change from annular to spray flow of the liguid.

Since analogous behavior might be expected with boiling potassium, .
the delineation of the region of potential burnout becomes important.

This will be studied by determining experimentally the variations in the
heat-transfer coefficient as a function of the quality for a range of
saturation pressures and mass velocities., Another problem which arises
in the vaporization of a liquid to a high void fraction is the high

outlet vapor velocity attained at low operating pressures. This results

 

10"ANP Semiann. Prog. Rep. Oct. 31, 1959," ORNL-2840, p. &7.

11C. E. Dengler and J. N. Addoms, "AIChE Heat Transfer Symposium'
at Louisville, p. 95 (1956).

 
in large two-phase pressure drops. Since existing two-phase flow corre-

lations have not been verified for liquid metals, it is planned to measure
such pressure drops in the boiling-potassium system. Further, limited
data suggest that heat-transfer coefficients in condensation of liquid
metals are lower by an order of magnitude than would be predicted from

theory.12

Thus, the apparatus is being designed so that condensing
coefficients for potassium may alsc be determined.

A flow diagram of the system proposed for these studies is presented
in Fig. 6.5. Liquid potassium is pumped through a throttle valve (re-
quired for flow stabilization) into a vertical, electrically heated boiler
at the saturation temperature corresponding to the inlet pressure, The
effluent liquid-vapor mixture then passes to a centrifugal separator from
which the liquid fraction is returned to the liquid reservoir at the pump
inlet and the vapor fraction is passed to the condenser. TFlow rates of
liquid and condensate streams are measured by collection in hold tanks
provided with liquid~level indicators or by electromagnetic flowmeters;
thus, the exit quality is determined directly. Pressure-measuring devices
are included for establishing the pressure drop across the boiler,
separator, and condenser, and for control purposes. A vacuum vent through
a small vapor condenser above the surge tank is provided for purging non-
condensable gases. Two electromagnetic pumps in series are required to
provide sufficient head for circulation at high exit guality. The heat
generated in the pump cell by the electric current is removed by means
of relatively large bypass flow through the air-cooled liquid reservoir.

A circulating cold trap is provided across the pumps for oxide removal.

It is expected that the liquid potassium flow rate will range from
0.1 to 1.0 gpm, corresponding to an inlet liquid velocity to the boiler
of 0.4 to 4.0 ft/sec. Initial operation will be such that approximately
atmospheric pressure will exist at the pump inlet, and the corresponding

saturation temperature of potassium will be 1400°F. Pressures and

 

128, Misra and C. F. Bonilla, "AIChE Heat-Transfer Symposium" at
Louisville, p. 7 (1936)

65
saturation temperatures at other points around the loop will, of course,
be higher.,

The experimental boiler for the system is shown schematically in
Fig. 6.6. The boiler consists of a 0.375-in.-o.d., 0.028-in.-wall, type
347 stainless steel tube surrounded by a stack of twenty-one 2-in.-thick,
5=-in.=o.d. copper disks furnace brazed to type 347 stainless steel tube
on the inside and to a type 310 stainless steel contaimment jacket on the
outside. Thermal isolation in the axial direction is achieved by spacer
sandwiches of type 321 stainless steel foil and screen. Each copper disk
is provided with an individually controlled clamshell heater of about
o-kw capacity. Sheath-type Chromel-Alumel thermocouples (0.045 in. o.d.)
are positioned within each copper disk to measure the temperature at
three radial locations.

The copper disks function as heat meters so that the local heat fl1
can be determined from the measured temperature gradient. The temperature
gradient can also be extrapolated to give the boiler=-tube wall temperature.
Sinze the temperature of the vapor will be known, the local heat-transfer
coefficient can be determined. Because of the large ratio of outside-
to-inside area of the disks, internal heat fluxes up to 500 000 Btu/hreft?
may be attained by utilizing easily replaceable clamshell heaters. The
copper disks also serve as 'thermal capacitors" to prevent a rapid in-
crease in tube wall temperature as the burnout point is approached. It
has been calculated that if complete loss of internal cooling were to
occur at a given position in the tube, the time required for the tube
wall temperature to increase from ~1500 to ~1900°F at full power input
is 3 min., This gives adequate time to reduce the power input to that
particular section of the boller. Automatic controllers are provided so
that continuous operation at potential burnout coanditions will be possible.

The most difficult problem associated with the fabrication of the
boiler is that of brazing the inner stainless steel tube and the outer
jacket to the copper disks. The method being tested involves (1) nickel
plating the inner and outer surfaces of the copper and (2) Joining the

disks to the stainless steel with Coast Metals No. 52 alloy by the

66

 
UNCLASSIFIED

ORNL-LR-DWG 46124A

 

SEPARATOR

' PRESSURE
A ELEMENT

 

VAPOR

 

 

 

VACUUM

 
   
 

— VENT
CONDENSER

SURGE
TANK

 

BOILER

 

 

 

 

CONDENSER

 

 

 
  
  
 
   
 
 
   
   
 

 

MIXING CHAMBER T/C

HOL D
TANKS
T/C

T/C
FLOWMETER

 

LIQUID
RESERVOIR

EM PUMPS

 
 

FREEZE
VALVE

 

 

1
=1
SUMP TANK

COLD TRAP

L9

T/C

N MIXING
CHAMBER

~—"0" TUBE
LEVEL
INDICATOR

Fig. 6.5. Schematic Flow Diagram of System
for Liquid-Metal-Bolling Experiments.

UNCL ASSIFIED
ORNL-LR-DWG 45166A

 
 
 
 
 
 
 
  
             
   

0.375-in.-0D, 0.028-in.
WALL, TYPE 347 STAINLESS
STEEL BOILER TUBE

TWENTY-ONE 5-in.-0D, 2-in.-THICK
COPPER DISKS BRAZED TO BOILER
TUBE AND CONTAINMENT JACKET

INSULATING SPACER SANDWICHES
OF TYPE 321 STAINLESS
STEEL FOIL AND SCREEN

    

    
 
 
 
  

g}\ PSR L7

&
.:
5
L

K
::
::
S

N2y
S~

ST
.0‘0 >
2R

258

55

 

  

   

A“

2-in.-LONG CLAMSHELL
HEATERS, ONE FER DISK
INDIVIDUALLY CONTROLLED

 

 

       
  

KB
SR
4l
SRR
SRS E5d

  

 

 

‘-\___
0.040-in.-OD CHROMEL-ALUMEL
SHEATH-TYPE THERMOCOUPLES,
THREE PER DISK

SECTION B-B

    
    
  
  

  

&
SN

2

     

R,
R
X0
P

S
&S

ELH
XX

< >
558
5%

 
 
   

o
&

    

TYPE 310 STAINLESS STEEL SHELL

SECTION A-A

Fig. 6.6. Experimental Boiller for Liquid-

Metal-Boiling Experiments.
furnace-brazing technique. This technique requires heavy plating
thicknesses, close tolerances, and careful control of amount and pre-
placement location of the brazing alloy. Three brazing tests have been
performed, none of which has been completely satisfactory, although the
latest test showed definite improvement.

The condenser (Fig. 6.7) for the system is similar to the boiler
in that 1t consists of a stack of copper disks around a central tube; in
this case, however, the heaters are replaced by a water-cooled radiation
heat sink, Resistance heating of the outer wall is provided to control
the rate of condensation. Thermocouples (not shown in Fig, 6.7) are
provided in the copper to measure the heat flux and tube-wall tempera-
ture so that local condensing coefficients can be calculated. It is
anticipated that, for initial operation, the experimental condenser will
be replaced by a bare coil of 3/4-in.-diam stainless steel tubing cooled
by a forced alr draft. The rapid response of this condenser to variations
in flow of the cooling air will enable close control of condensation
rate and, hence, of the loop pressure during shakedown operations.

The liquid-vapor separator, illustrated in Fig. 6.8, consists of a
4 l/2-in.-diam vorcex tube with a single 5/8—in.-i.d. feed nozzle entering
at a radius of 1.75 in., a 5/8-in.-i.d. liquid line, and a 5/8-in.-1i.d.
vapor line whose opening is about 3 in. above the level of the inlet.
The circular baffle on the top of the vapor line is necessary to eliminate
entrainment of drops of liquid which creep up the vapor line. The
separator 1s shown in Fig., 6.9 in operation with an air-water mixture
generated by atomization at the base of a 0.31-in.=-i,d. vertical tube
simulating the experimental boiler. Good dispersion of the water by the
air stream was obtained. Separation was excellent for flow rates (based
on equivalent liquid) of up to 0.5 gpm and qualities of up to 90%.
Extrapolation to conditions for potassium boiling at 14.7 psia indicate
that, at the corresponding liquid flow of 0.25 gpm and 90% quality, the
pressure drop across the separator will be about 1.5 psi.

The possibility of incorporating a gemma~ or x-ray densitometer

between the boiler and separator in order to obtain a continuous and

68

 
69

UNCLASSIFIED
ORNL-LR-DWG 452094

     

5o —e WATER QUT
AIR JACKET
_ <P

 

 

 

LAVA INSULATOR

Fig. 6.7. Experimental Condenser for
Liguid-Metal~Boiling Experiments.

UNCLASSIFIED
ORNL—LR-DWG 46123A

 

 

 

 

 

 

 

 

 

 

 

 

 

MIXTURE INLET [
%
7
{ !
A !
fi :::Ln
10in.
|
LANY
W 7 _+
N
A
q Y
LIQUID OUTLET \

 

VAPOR QUTLET

o 2 a4 6
1 L L L i

SCALE N INCHES

Fig. 6.8. Liquid-Vapor Separator for Liquid-
Metal-Boiling Experiments.

 
  
  

UNCLASSIFIED
PHOTO 35688

 

Fig. 6.9. Liquid-Vapor Separator in Operation with an Air-Water
Mixture.

70

 
et s ek s il e o 1 14 £ i 1 4w b i e 2 o e e e P e g e B e

essentially instantaneous measurement of vapor void fraction is being
considered., Adequate precision of measurement at void fractions cor-
responding to 50% quality or higher will, however, require a prohibitively
long vapor peth if a stainless steel wall is employed in the gamma-~ray
path. Use of a beryllium window would shorten the vapor path required

but would introduce the need for a reliable mechanical seal to operate

in contact with potassium vapor at 1400°F. It is unlikely, therefore,

that a densitometer will be available for initial operation of the system.

71
7. RADIATION EFFECTS

Irradiation of Moderator Materials in the ETR

A series of irradiations of yttrium hydride and beryllium oxide
specimens is under way in the Engineering Test Reactor in order to determine
the extent to which these materials maintain their integrity under reactor
conditions. The specimens are examined after irradiation for indications
of radiation damage, such as deformation and crystal growth, and for com-
patibility with containing materials. 1In addition, the beryllium oxide
specimens are checked to determine the amount of gas formed by the (n,a)
and (n,2n) reactions in beryllium. The relationship between the irradia-
tion temperature and th= rates of diffusion of the gas while the beryllium
oxide is being irradiated is also being investigated. For this study,
specimens are irradiated in fluxes that produce temperatures from about
100 to approximately 1000°C.

In ETR position C33H10, which is the position being used for these
irraditions, the gamma flux is 25 w/g. Typical neutron flux values that
were measured in this position during ETR cycle 18 with cobalt and nickel
wires as monitors indicate a maximum thermal-neutron flux of 6.2 X 104
neutrons/cm?-sec and a fast-neutron flux (>1 Mev) of 4.3 x 10%4
neutrons/cm?- sec.

The specimens irradiated in the initial experiment, as described
previously,l are being examined in hot cells at ORNL. The minute radial
cracks (mentioned previously!) around the thermocouple well in the
yttrium hydride sample are shown in Fig. 7.1. Examinations of unirradiated
pieces with thermocouple wells have shown similar, less pronounced cracks.
Measurements of the irradiated specimens have shown no changes in physical
dimensions.

Visual inspections of the irradiated yttrium hydride specimens have
revealed a possible composition change in the material at the interface

of the first and second specimens, as shown in Fig. 7.2. A spot of

 

1"ANP Semiann. Prog. Rep., Oct. 31, 1959," ORNL-2840, p. 79.

72

 
. UNCLASSIFIED
. PHUTO 50359

y—

 

Fig. 7.1. Irradiated Yttrium Hydride Sample Showing Minute Cracks
Around Thermocouple Well.

UNCLASSIFIED
PHOTO 50360

 

Fig. 7.2. Top Surface of Second Specimen of Irradiated Yttrium
Hydride.

]

 
metallic crystalline material formed on the face of the secoad specimen
and developed into a deposit extending 2 to 5 mils above the normal surface
plane. The deposit fits a corresponding depression in the face of the
first specimen. The formation of the metallic material may be attributed
to either of two mechanisms: migration of hydrcgen to cooler surface
regions of the capsule or actual diffusion of hydrogen through the metal
capsule wall. The location of the deposit near a high-temperature thermo-
well wall tends to add weight to the diffusion theory.

The total radiation which the specimens received was determinad by
counting the Co®% in the Inconel sheath of the thermocouple wires. The
dose profile obtained for the first experiment is presented in Fig. 7.3.

As may be seen, the thermal-neutron flux was about 20% less at the top

of the BeO samples than at the lower end of the yttrium hydride samples.
The fast-neutron flux (>1 Mev) was estimated from the flux data of
Phillips Petroleum Company that were obtained from cobalt and nickel wires
in the irradiation facility. The fast-neutron dose estimate is also pre-
sented in Fig. 7.3.

Irradiation of a second set of specimens has bzen completed and a
third set is bzing irradiated. Examinations of the second set of specimens

are under way.

Creep and Stress-Rupture Tests Under Irradiation

 

In-pile stress rupture testing of Inconel at 1500°F in air is being
continued in the poolside facility of the ORR. Details of the design,
construction, and operation of the experimental system have been described

3

previously.z’ The results of experiments completed since the previous

report in this series* are tabulated in Table 7.1.

 

2J. C. Wilson et al., "Solid State Div. Ann. Prog. Rep. Aug. 31, 1958,"
ORNL=2614, p. 106.

’N. E. Hinkle et al., "Solid State Div. Ann. Prog. Rep. Aug. 31, 1959,"
ORNL=2829, p. 214.

“"ANP Semiann. Prog. Rep. Oct. 31, 1959," ORNL-2840, p. 79.

74

 
UNCLASSIFIED
ORNL~-LR-~-DWG 49597

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

(X 4020) \kl\
THERM
Al -
NEUTRONE\
\\

2.0 N\\~;‘~\~s-
& ~
£
U ——
~ e
s -

-
g 1.5 YH TR BeO
- 2 "'"--.___ €
=2
= o
F4 -~
L Sr. ~~
© 1o ON Do~
~+— CENTER LINE Sg
OF REACTOR
0.5
0 1 2 3 4 5 o T 8 9 10 11 12

DISTANCE ALONG

Fig. 7.3. Graph of nvt vs.
of reactor at 4.5 in.

EXPERIMENTAL ASSEMBLY (in.)

Distance Along Experiment.

Center line

75
Table 7.1. Effect of Neutron Bombardment on the
Time to Rupture of Inconel In Air

 

 

 

 

 

Irradiation Time to
Specimen Stress Dose at Ru
. } pture
No. (psi) Rupture (br)
(Mwhr) r
First Heat
11-5 5000 2391 1252
11-6 5000 1671 73a
11-7 5000 1733 93
11-8 5000 1711 21
Second Heat
14-1 5000 1726 66
14=5 5000 1803 70
14-6 4000 3536 181
14-9 4000 2476 108
14-10 4000 1106 35
14=4 3000 9075 443
14=7 3000 8835 431
14-8 3000 6209 299

 

e ——— o —— e —— e et e e

aSpecimens were stressed at the operating
temperature for approximately 42 hr prior to
reactor startup; the time to rupture is in hours
after startup.

In order to investigate the hypothesis that the presence of boron
and the associated gas production resulting from the B0 (n,a) reaction
may be responsible for the reduced time to rupture of Inconel stressed
in a radlation field, the additional tests were conducted on two different
heats of material. Two specimens of the first heat were stressed to
5000 psi at 1500°F for about 42 hr prior to the reactor startup. Despite
this preirradiation treatment, the observed times to rupture after neutron
bombardment began (93 and 125 hr) are not appreciably different from those
obtained for specimens stressed shortly after reactor startup.

The results of the tests on the second heat indicate s shorter in-
pile time to rupture life than for the first heat. The relative effect

of the neutron bombardment on the second heat will not be known until

76

 
UMNCL ASSIFIED
¥-31rm

 

Fig. 7.4. Parts of Tantalum Furnace for High-Temperature In-Pile

UNCL ASSIFIED
1z

 

T

Fig. 7.5. Completed Tantalum Furnace.

17

 
the control tests are completed. ©Specimens of special heats of Inconel
with controlled amounts of boron are now being prepared for testing.

As reported previously,4 a program of study of the effects of neutron
bombardment on the creep and stress-rupture properties of columbium alloys
has been initiated. Designs for in-pile and out-of-pile tube-burst test
asgsemblies are essentially complete, and fabrication of parts is under
way. A Cb—1% Zr alloy will be tested in the first experiments. The
specimens will be similar to the Inconel specimens in size and geometry.
The in~pile furnaces will be of the type shown in Figs. 7.4 and 7.5, which
has undergone considerable development and testing out of pile. 1In the
fabrication of the furnace, a tantalum tube is first sand-blasted and
sprayed with alumina. Tantalum windings are wrapped over this form and
the furnace is then sprayed with more alumina. The completed furnace is
shown in Fig. 7.5.

Successful completion of the in-pile columbium test program will
depend on achieving a high-purity helium atmosphere. To accomplish this,
the supply helium will first be passed through a gas-purification train
and then into the expcsure can in which additional zirconium "getter"
material will be used. Further, the experimental system has been designed
to minimize the use of materials which would create atmospheric contami-

nation.

78

 

 
 

8. ADVANCED POWER PLANT STUDIES

Comparative Study of Power Plants for Space Vehicles

The power plant requirements for space vehicles and the performance
characteristics of various types of plants, including solar cells, radio-
isotope energy converters, etc., have been studied. Based on the available
information, it is concluded that by 1970 there will be need for power
plants capable of producing as much as 2000 kw of electrical output, and
that, in the range above 20 kw(e), a fission reactor coupled to a turbine
generator shows the greatest promise for development.

In view of the wide variety of systems that have been proposed for
this application, the principal objective of this study became the
development and application of criteria for Judging the relative merits
of typical systems and their components. A detailed report on this study
is being prepared.

Two major precepts guided the study. Experience with various types
of power plant has indicated that the reliability requirement for space
power units will be extremely difficult to meet, and hence the most
promising type of system is considered to be one that employs a single
working fluid to cool the reactor, carry out the thermodynamic cycle, and
serve as the lubricant and coolant for the turbine and generator. Further,
experience in other reactor programs has shown the importance of ruling
out systems requiring difficult new materials discoveries or inventions
to achieve a short-~term objective.

Consistent with these guide lines, a wide varlety of possible working
fluids and thermodynamic cycles was considered, and ten working flulds in
one or another of three thermodynamic cycles were chosen for detailed
analysis. The weight of the radiator alone was found to be excessive for
gas-turbine cycles and rather high for steam and mercury-vapor cycles.

A critical review of radiator, turbine, generator, reactor, and shield
design considerations led to the conclusion that the most promising

working fluids are potassium and rubidium, with potassium having the

79
advantage that it gives a much less serious radiation dose from the
radiator. On the basis of this analysis, experimental work was under-
taken on heat transfer in boiling potassium (Chap. 6) and corrosion in
boiling-potassium stainless steel systems (Chap. 2).

Some of the more important implications of the study can be deduced
from Table 8.1, which summarizes the effects of the working fluid on the
estimated weights of major components of the power plant other than the
reactor and shield for a set of 100-kw(e) power plants. All these data,
except those for a system using lithium as the working fluid, were ob-
tained for a consistent set of design conditions, including the use of
stainless steel (or a similar iron~chrome-nickel alloy) as the structural
material, Although a more refractory structural material and higher
temperatures were assumed for the lithium cycle, it may be seen that the
turbine and radiator manifold would be very much larger and heavier than

for any of the other working fluids.

Table &.1., Effects of Working Fluid on Major Design Parameters and Specific Weights
for a Series of 100~kw(e) Power Plants

 

Working Fluid

 

Fluid
H, He Air  AlCls H,0 He Rb K Na Li

 

Turbine conditions

Inlet temperature, °R 2000 2000 2000 2000 1360 1710 2000 2000 2000 2500

 

Outlet temperature, °R. 1375 1380 1357 1568 830 1200 1500 1500 1500 1800
Pressure, psia 81 81 81 60 1330 500 53 29.3 9.7 2.5
Rotor diameter, ft C.332 0.308 0.350 1.09 0.0808 0.287 0.644 0,799 1.84 5.86
Specific weights, 1b/kw(e)
Turbine 0.35 0.11 0.03 0.63 0.13 0.21 0.098 (0.161 1.69 99.0
Compressor 0.35 0.11 .03 C.38 0.13 0.01 0.005 C.008 0.0%9 5.0
Radiator 33 33 33 18 31 7.5 3.0 3.0 3.0 1.2
Manifold 0.02 0.03 0.2 3.8 0.0002 0.009 0.10 0.20 1.7 19.3
Working fluid 0.9 7.8 0.8 0.5 0.5 0.1
Total 33.72 33.25 33.26 22,71 32.16 15.53 4,00 3.87 6.98 124.6

 

Space Radiator Designs

Design studies of radiator systems for auxiliary power units in

space vehicles have been continued. The program has been expanded to

80

 
encompass studies of the vulnerability of space radiators to penetration
by meteoroids and of methods of reducing this vulnerability.

It has been proposed that the placing of space radiator systems inside
plastic "balloons” might reduce the vulnerability to meteoroid penetration,
and a preliminary study of the effectiveness and design of such plastic
bumpers has been com.pleted.l A detailed study of the thermal performance
of bumpers and bumper materials has been initiated, and preliminary work
has been completed in both the analytical phase and the design of an ex=
perimental program for the determinaetion of the thermal performance of
bumper materials.?’2 The results of these studies indicate that plastic
bumpers composed of "Teslar" or "Mylar' approximately 0.001 in. thick
will drastically reduce the hazard of meteoroid penetration while reducing
the thermal performance of the radiator by as little as 2%. Such bumpers
will result in an over-all increase of the radiator system weight of about
15 to 20%.

Comparative studies have been conducted to determine the most desir-
able configuration of radiator tubes and manifolds to be employed in the
design of a radiator system for a potassium Rankine cycle for a 100=kw(e)
reactor-turbine generator auxiliary power supply. The weilght of the
radiator system was considered the most important variable to be minimized,
but an effort was made to minimize the void volume and surface areas which
would be vulnerable to meteoroid penetration, since these quantities in-
directly affect the weight of the power unit by requiring greater quantities
of working fluid and providing more material for protection from meteoroids.
The results of this investigation have indicated that a promising radiator
design consists of a single vapor feed manifold with radiator tubes placed

on two sides of the manifold in a plane and perpendicular to the manifold.

 

1R, J. Hefner and P. G. Lafyatis, "Protection of Space Vehicles from
Meteorite Penetration,” ORNL CF-60-1-67 (January 20, 1960).

°R. J. Hefner, "Thermal Radiative Heat Transfer to Space from a Body
Enclosed by a Semitransparent Body," ORNL CF-60-5-2 (May 6, 1960).

3R, J. Hefner, "Space Radiator Test Program,” ORNL CF-60-4-43 (April 8,
1960).

81
 

Table 8.2. Comparisons Showing Effect of Radiator Configuration on
Specific Weight, Void Volume, and Vulnerable Area

 

Relative Relative

 

 

 

 

Radiator Number of Rows of Fins Specific Average Rela?lve
. . . Tubes Per Per . Void
Configuration®* Manifolds . Weight  Vulnerable
Manifold Tube . Volume
Radiator Area

Radiator Tube Taper Ratio: 3
A 1 2 2 1 1 1
B 2 2 2 0. 97 0.96 1.01
C 3 2 2 1.08 0.97 1.02
D 4 2 2 1.02 0.99 1.02
E (Inv) 1 4 4 1.17 0.90 0.77
F (Inv) 1 4 0 2.36 1.79 1.86
G (Inv) 1 1 4 1.61 1.52 1.00
H (Inv) 4 2 4 1.38 1.03 1,00
Radiator Tube Taper Ratio: 1
Ay 1.18 1.23 1.74
Radiator Tube Taper Ratio: 10
Ajp 0.92 0. 89 0.81

 

*¥See Figs. 8.1 and 8.2.

UNCLASSIFIED
ORNL—LR- DWG 45148

 

VAPOR INLET

LIQUID OUTLET

/

Fig 8.1. Radiator Configuration for Space Vehicle Power Plant.

82

 
 

 

The radiator tube assembly consists of round tapered tubes with two
longitudinal fins placed on each tube in a plane coinciding with the
plane of the radiator tubes on the manifold.

A study is presently in progress for determining the most desirable
materials and the sizes of all component parts of such a radiator system
in order to minimize the weight, void volume, and vulnerability to
meteoroid penetration. Preliminary results of this investigation Indicate
that the optimum radiator system for a 100~kw(e) power plant will weight
approximately 225 1b, exclusive of the meteoroid bumper; it will have a
void volume of 20 ft3, of which about 10% will be filled with liquid
working fluid; and it will have an area vulnerable to meteoroid penetration
of approximately 150 ft4, The radiator system will be about 13 ft long,
16 £t wide, and 1 ft thick, exclusive of the bumper. The bumper will

weigh approximately 25 1b and will be an ellipsoid with its major axis

UNCLASSIFIED
ORNL-LR-OWG 49394

 

Fig. 8.2. Radiator Configurations for Space Vehicle Power Plants
Showing Various Manifold Designs.

g3
perpendicular to the flat side of the radiator. The specific weight of
this radlator system willl be approxinately 2.5 lb/kw of electrical output,
Including the radiator and the bumper, but excluding the working fluid.

Teble 8.2 gives a comparison of the weights of some of the radiator
configurations that have been investigated. The table lists the specific
weight, void volume, and vulnerable area of each configuration as compared
with the same quantity for the configuration shown in Fig. 8.1 by consider-
ing each of these quantities as being unity for the reference system.

The configuration shown in Fig. 8.1 corresponds to configuration A .
in Table 8.2. Configuration B, C, and D of Table 8.2 are basically the
same as A, with the difference being in the number of manifolds and con~
sequently the number of tubes per manifold. The radiator tube designs
for configurations E (Inv), G (Inv), and H (Inv) are shown in Fig. 8.2b
and the radiator tube design for configuration F (Inv) is shown in
Fig. 8.2a. The manifold designs for configurations A, B, C, D, E and F,

G, and H are shown in Fig. 8.2c, 4, e, f, g, h, and i, respectively.

84

 

 
 

PART 2. SHIELDING

R G b, AR A 5 R RN S 20w R R Tk L S
 

 
 

9. SHIELDING THEORY

Monte Carlo Calceculations of Response Functions of
Gamma=Ray Scintillation Detectors

The development of IBM-704 codes to calculate, by Monte Carlo methods,
the response functions of organic and inorganic scintillators to gamma
rays has been reported.l At the present time three essentially perfected
codes are available., The first gives the response to monoenergetic gamms
rays of thallium=activated sodium lodide crystals, and, through proper
choice of a key word within the code, the results may include only primary
responses, may consider the results of annihilation radiation as well, or
may include the effects of both annihilation radiatlon and bremsstrahlung
radiation. When, as is possible, the output histograms are "smeared" by
a Gaussian broadening to simulate the resolution of an actual crystal,
the final result is an essentially complete picture of the behavior of
a particular scintillator.

The second code 1s designed to calculate the response functions of
xylene. For this material only primary responses and annihilation radiation
are considered.

The third code, which is similar to the first, calculates the behavior
of activated cesium iodide crystals. At the present time the sodium iodide
bremsstrahlung data are being used with this code, since little difference
is expected to exist.

The codes permit calculation of the response of crystals in the
form of either a true right cylinder or a right cylinder capped with
a truncated cone. (The latter shape is the normal growth form of some
crystals.) Wells or axial holes in the crystals are permitted as a
calculational varilant. Unlimited dimensional variations are possible,
while input photon energies may range from 0.005 to 10.0 Mev. Since,
as noted above, the most recent refinement of the code gives an essen-
tially complete description of crystal behavior, lacking only the ex-

perimental background and collimator effects, 1t 1s expected that this

 

1"ANP Semiann. Prog. Rep. Oct. 31, 1959," ORNL-~2840, p. 97.

87
code will be valuable, not only in the "unfolding" process necessary

to understand experimental spectra, but also in evaluating the magnitude
of background and collimator effects, and thus will assist in their
minimization. Running time for a problem varies over broad ranges,
depending upon choice of input energy, number of photon histories run,

crystal geometry, desired statistics, etc.
A Monte Carlo Code for Deep Penetrations of Gamme Rays

The development of a Monte Carlo code to calculate the energy and
angular distributions, at a point detector, of gamma rays emitted from
monoenergetic point isotropic sources in an infinite, homogeneous, uniform
medium has been reported.? Further work has been done on the application
of the "econditional” Monte Carlo method to this problem, with the result
that this method, which previously had seemed to require further mathe-
matical development, now appears to be suitable.

At the present time the available data are not sufficient to formulate
a definitive analysis of the code. Some results have been obtained, however,
for gamma-ray penetrations of from 1 to 20 mean free paths in media of
iron and water. A "Russian Roulette" technique was used to assist in the
calculation of the longer distances. Initial energies have ranged from
1 to 10 Mev. The differential energy spectrum obtained appears to be in
reasonable agreement with calculations by the "moments" method.? A detailed
analysis of the code, together with its mathematical basis, 1s to be
published when sufficient experience with the code has been accumulated to

verify it.

Prediction of Thermal-=Neutron Fluxes in the Bulk Shielding
Facility from Lid Tenk Shielding Facility Data

 

 

A calculation of the thermal-neutron flux in water around the Bulk

Shielding Facility (BSF) Reactor I with loading No. 33, based on

 

2Tbid., p. 105.

°H. Goldstein and J. E. Wilkins, Jr., "Calculations of the Penetrations
of Gamma Rays, Final Report,” NDA-15C-41 or NYO~3~75 (1954).

88

 

 
transformation by standard methods of Lid Tank Shielding Facility (LTSF)
data, has been reported.4’5 The purpose of the calculation was to ex~
amine the agreement between the independently measured powers of each
source and to validate the geometrical transformations employed. The
calculation has now been completed in final form, and the detailed re~
sults have been published.6

A final review of the data showed reasonable agreement between pre-
dicted and measured BSF reactor fluxes for distances of 65 cm or greater
from the reactor face, with the differences ranging between 0 and 9%;
while for distances less than 65 cm the calculated value of the flux is
uniformly greater than the measured value, with the differences ranging

from 6 to 22%. The comparison is shown in Fig. 9.1.

Monte Carlo Calculations of Dose Rates Inside
a Cylindrical Crew Compartment '

The ABCD code, an IBM~704 code designed to calculate, by Monte Carlo
methods, the fast-neutron dose rates inside a cylindrical crew compartment
of finite length, has been described.® The code computes the dose from
neutrons uniformly distributed in space in the neighborhocd of the cylinder.
A so-called "simllarity transform”" feature, originally described by
Steinberg,9 is utilized to permit the simultanecus computation of doses
for many different confilgurations.

The code in its final form permits a choice of incident neutron

distributions in angle and energy: (1) constant angle with the axial

 

4M"ANP Semiann. Prog. Rep. Mar. 31, 1959," ORNL-2711, p. 116.
S"ANP Semiann. Prog. Rep. Oct. 31, 1959," ORNL=-2840, p. 106.

SA. D. MacKellar, "Prediction of Thermal-Neutron Fluxes in the Bulk
Shielding Facility from Lid Tank Shielding Facility Data,” ORNL CF=59~1-24
(Jan. 12, 1960).

“The results reported here are based on work performed for ORNL under
Subcontract 1393 by Technlcal Research Group, New York.

8"2AWP Semlann. Prog. Rep. Oct. 31, 1959," ORNL-2840, p. 104.

°H. Steinberg, "Monte Carlo Calculation of Gamma-Ray Penetration
II," LNP-NR-47, Vol. II, Aug. 1958.

89
Pool.

20

UNCLASSIFIED
2—01-058—-0—-479A

CALCULATED

MEASURED

THERMAL—NEUTRON FLUX (v, /watt)

2

 

1o~
25 35 45 55 65 75 85 95 105 {115 125
DISTANCE FROM BSR (cm)

Fig. 9.1. Calculated and Measured Thermal-Neutron Fluxes in BSR

 
 

direction at each point, independent of azimuth, and (2) an angle-energy
distribution given by the output of the Convair D-35 code for neutron air
scattering, where the angle variable is again the angle with the axial
direction. The shield material may be water or polyethylene.

A series of problems has been run with the D-35 input to check the
performance of the code in the calculation of the neutron penetration
through a typical crew compartment. The dimensions of the compartment

for which the runs were made were:

Rear wall (essentially

black) 91.44 cm thick
Front wall 50.80 cm thick
Cavity 182.88 cm long
Cylinder outer radius 66,24 cn

Side wall 5.08-20.32 cm thick

The neutron source was considered to be monoenergetic, either 2.7 or 1.1
Mev, and to be located 64 ft from the detector. The results appear to be
satisfactory in all cases run. A complete report that covers details of
operation of the ABCD code and includes an operating manual for users

has been published.lO

Design of a Unit Shield

The shields designed for nuclear-powered aircraft are of two general
types. The first is the unit shield in which the attenuating material
is placed around the reactory the second 1s the divided shield in which
the shielding materisl is divided between the source and the recelver.
Although the unit shield can be expected, in general, to be "shaped"

less severely than the divided shield, there is still much to be gained

 

104, Steinberg, "Monte Carlo Code for Penetration of Crew Compartment-
II," TRG=211=3~FR, Dec. 31, 1959; see also H Steinberg, J. Heltner, and
R. Aronson, "Monte Carlo Code for Penetration of Crew Compartment,” Final
Report on ORNL Subcontract 931 under W~7405-FEng~26 (Dec. 1958).

91
in weight savings through efficient design. Accordingly, a study has
been initiated of some of the problems in the specification of unit
systems and of calculational methods for obtaining optimum configurations.
The systems considered utilize two basic materials: a light, hydrogenous
material, primarily to attenuate neutrons, and a dense material of high
atomic weight, primarily to attenuate gamma rays.

Preliminary studies have been directed toward a better understanding
of the gamma-ray problems and to the design and placement of heavy
materials within the light neutron-shielding material surrounding the
reactor core. The gamma-ray dose at the detector position is essentially
composed of gamma rays born in the core plus gamma rays resulting from
neutron interactions in the shield. Most of the work thus far has been
a study of the gamma rays born in the core, which were subdivided into
photons that penetrate all the layers of heavy material, photons that
scatter "favorably'" in the light shield material so as to bypass at
least part of the heavy material, and photons that escape the reactor
without penetrating all the layers of heavy material, scatter favorably
in the surrounding air, and reach the detector.

The simple model (Fig. 9.2) for the first sample computations consists

of a point source of isotropic, monoenergetic gamma rays imbedded at the

UNCLASSIFIED
ORNL-LR~DWG 4979t

 

 

Fig. 9.2. Model Used for Calculating Gamma-Ray Doses.

92

 
 

center of a sphere of water and a single lead shadow shield of uniform
thickness placed at a specified distance from the source. For a first
approximation, the "direct" contribution of photons penetrating the full
thickness of heavy material was computed by assuming that the shadow
shield formed a complete spherical annulus. Scattering effects were
accounted for by the assumption that a suitable approximation is given
by the use of the Goldstein buildup factors for watertl over a distance
equal to the total number of lead and water mean free paths along a
radius of the sphere. The calculation of the "direct” contribution is
expected to be least realistlc in the case of shadow shields of small
half-angle.

The contribution to the dose at the detector from photons that
scatter around the shadow shield was calculated by the use of a first-
collision model, taking into account the photon energy after scattering
through a favorable angle and assuming line~of=-sight attenuation with
buildup from point of scatter to the shield periphery. Only photons
scattering outside the shadow shield radius and not intercepted by the
shadow shield were consldered; that is, the shadow shield was assumed
to be "black." This assumption is most valid for relatively thick
shadow shields.

The air=scattered contribution was calculated using differential
energy spectra reported by Goldstein and Wilkins,ll assuming radial
emission at the shield surface, and applying alr-scattering transfer

1,12 Again, the shadow shield was

functions reported by Lynch et a
assumed to be "black.'
On the basis of the model discussed above, the gamma-ray dose rate

at the detector can be written as

gla,t) =g (t) +g (@) +¢g (o) ,

 

11y, QGoldstein and J. E. Wilkins, Jr., "Calculations of the Penetration
of Gamma Rays,” NY0=3075 (June 30, 1954).

12R, E. Lynch et al., "A Monte Carlo Calculation of Alr-Scattered
Gamma Rays,” ORNL-2292, Vol. IV (Jan. 28, 1958).

93
where gd(t) is the "direct" contribution in the direction of the detector,
gc(ofl is the first=collision contribution, ga(OO 1s the alr=scattered
contribution, t 1is the shadow shield thickness, and @ is the half-angle
subtended by the shadow shleld.,

The conditions for a minimum-weight design, consistent with a

specifled dose rate Do’ are

g%mng-—.o ) (1)
%H\%flj , (2)

where W = W(,t) is the weight of the shield, D = D (a,t) — D, is a

cale
dose rate functilon (the dose-rate constraint for this problem is given
by the condition D = 0), and A 1s a constant, the so-~called Lagrange
multiplier.

Solving (1) and (2) for )\ yields

 

 

OW 5 _ _
o St _ 2flro(ph pl)(l cosq) ()
oD dg, ’
ot —_—
dt
and.
oW 20 _ :
L 56 2 ro(ph 0,)t sinm ()
- oD deg dg ’
da L&, 2
ao ac

where ro is the inner radius of the shadow shield, ph and pl are the
densities of the heavy and light materials, respectively, and the other
quantities are as previously defined. In these equations a linear
dependence of shadow shield weight on thickness is assumed. Equating

the right~hand members of (3) and (4) give-

%

 
 

& dgd ~ dgc + dga 1l — cost (5)
dt ~\ a do sinc ’

In (5) the left side 1s a function of t only, whereas the right side is
a function of . Hence, 1if each side 1is plotted against 1ts argument,
equal ordinate values from the two plots will yield pairs of values
(t,a) which define the shadow shield thickness as a function of the half
angle. Each distinct palr corresponds to a specific value of A and the
calculated dose rate.

Because of the assumptions involved, the results should be most
realistic for thicknesses of the order of several centimeters and half-
angles greater than, say, 30 deg. The work thus far has served principally
to give some insight into the relative magnitudes of the quantities in-
volved and to test a simple optimization scheme. In future work more
realistic problems will be considered that involve variable=thickness

shadow shields and includes secondary gamma-ray contributions.

95
10. LID TANK SHIELDING FACILITY

Radiation Attenuation Measurement Behind
Multllayer Configurations

 

 

During this period, experiments were designed and slab materials were
procured for a series of tests in connection with the design of the shield
for the Pratt & Whitney PWAR~11C test reactor. The Lid Tank Facility was
used for several experiments on the general problem of secondary gemme-ray
production and shield optimization in connection with several applications.

Studies of the vulnerability of armored vehicles to attack by tactical
nuclear weapons have indicated that, for an enclosed vehicle such as a tank,
the personnel hazard for which the effective range of the weapon is the
greatest 1s the nuclear radiation produced by the explosion. Although tests
have indicated that the armor of present~day tanks offers significéht
radiation protection, it has been shown that armor steel does not provide
the best shielding per unit shield weight in a field of mixed neutron and
gamma radiation. Tests performed during Operation pPlumbbob suggested that
slightly better radiation protection than that afforded by armor steel
could be obtained, with a weight saving of 33-1/3%, by using a shield
combining steel with a hydrogenocus material. The tests also demonstrated
the importance of the secondary gamma=ray dose from the incident neutron
field. It was apparent that additional data were needed before the design
of a satisfactory vehicle could be undertaken.

The required information has been obtained at the LTSF by an in-
vestigation of the radiation~attenuation characteristics of materials
satisfying both the structural design requirements for the vehicle and the
shielding criteria. The purpose of the experiments was to measure the
neutron~ and gamma~ray-attenuation properties of various combinations and
configurations of materials and to obtain sufficient data to calculate the
secondary gamma-ray dose component directly behind the shield from a unit
fast-neutron dose incident on the shield.

The fast-neutron dose-rate attenuations were readily obtained from

dose~rate measurements made in front of a sample configuration and behind

926

 
 

each configuration. These dose rates were measured in air to eliminate

the possible error involved in extrapolating the usual water measurements

to a zero water thickness. The gamma-ray information, which included both
the dose from incident gamma rays and the secondary gamma-ray dose resulting
from neutron capture and neutrons being inelastically scattered within the
shield, was obtalned in borated water. Thermal-neutron flux measurements,
using gold foils, were made in the liquid nitrogen preceding the shield
configuration and also through a sample shield conflguration of steel for
calculation of capture gamma rays. Thermal-neutron flux mappings in borated
water behind the shield configurations were included for calculation of the
boron capture gamma rays.

The materials used in this study consisted of borated polyethylene
(containing either 3 or 8 wt % B;C), depleted (0.28% U?3°) uranium, lead,
and steel. Liquid nitrogen was used to modify the incident neutron spectrum.
All configurations consisted of ~5=~ by 5-ft slabs of various thicknesses.
Table 10.1 lists the configurations examined. The data obtained in these

measurements are being analyzed.

Table 10.1. Configurations Used in Attenuation Measurements

 

Radiation Measured

 

 

Configuration . . e s
Number Configuration Description Qa Fast Thermal
Rays Neutrons Neutrons
731X°> Ny, 7 in. steel, 3 in. X
polyethylene, 1 in, steel
434X No, 4 in. steel, 3 1in. x
polyethylene, 4 in. steel
494X Ns, 4 in. steel, 9 in. X
polyethylene, 4 in. steel
791X No, 7 in. steel, 9 in. X
polyethylene, 1 in. steel
A134X N,, 1 in. steel, 3 in. borated X
polyethylene, 4 in. steel
A4 31X No, 4 in. steel, 3 in. borated X

polyethylene, 1 in. steel

 

8The character "X" at the end of the configuration number indicates
a measurement made in air,

97
Table 10.1.

(Continued)

 

Confilguration
Number

Conflguration Description

Radlation Measured

 

Gamma,
Rays

Fast
Neutrons

Thermal
Neutrons

 

A4 61X

A164%

A25-3-2.5%X

A2.5=12~2.5%

A4=12-1X

Al=12=-4X

A491X

A194X

A2.5-9=2.5X

AU'1=-1-8=-2X

AU'1-1-10-, 5=
.SUX

98

Ny, 4 in. steel, 6 1n.

borated polyethylene, 1 in.

steel

No, 1 in. steel, 6 in.

borated polyethylene, 4 in.

steel

Np, 2.5 in, steel, 3 in.
borated polyethylene,
2.5 1n. steel

Nz) 2.5 iIl. Steel} 12 ino
borated polyethylene,
2.5 1n. steel

No, 4 In., steel, 12 in,
borated polyethylene,
1 in. steel

No, 1 in. steel, 12 in.
borated polyethylene,
4 in. steel

N,, 4 in. steel, 9 in.
borated polyethylene,
1 in. steel

Ny, 1 in. steel, 2 in.
borated polyethylene,
4 1in, steel

Noy, 2.5 in. steel, 9 in.
borated polyethylene,
2.5 1In. steel

Np, 1 in. steel, 1 in,
uranium, & in, borated
polyethylene, 2 in. steel

Ny, 1 in. steel, 1 in.
uranium, 10 in, borated
polyethylene, 0.5 in.
steel, 0.5 in. uranium

X

 
 

Table 10.1 (Continued)

 

Radiation Measured

 

 

Configuration . . s
Number Configuration Description Gamms, Fast Thermal
Rays Neutrons Neutrons
AU'1-,5=10~.5~ N,, 1 in. steel, 0.5 in. X
1UX uranium, 10 in. borated
polyethylene, 0.5 in,
steel, 1 in. uranium
AU'1-1-12-1.5X N,, 1 in. steel, 1 in. X
uranium, 12 iIn. borated
polyethylene, 1.5 in,
steel
Al-12-1, 50X N,, 1 in. steel, 12=in. X
borated polyethylene,
1.5 in. uranium
B'A1~12~1.5= N,, 0.5 in. boral, 1 in. X X
1UB steel, 12 in. borated
polyethylene, 1.5 in.
steel, 1 in., uranium,
borated H0
BTAU'1~,5=10=~ Np, 0.5 1n. boral, 1 in. X X
.5=1UB steel, 0.5 in, uranium,
10 in. borated polyethylene,
0.5 in. steel, 1 in.
uranium, borated H,0
BIAU!'1~1~-12~ No, 0.5 in, boral, 1 in. X x
1 1 3 steel, 1 in. uranium,
2 12 in. borated polyethylene,
1.5 in. steel, BHZ0
BTAU'-1-10~ N,, 0.5 in. boral, 1 in. X X
.5=.5UB steel, 1 in. uranium, 10
in. borated polyethylene,
0.5 in. steel, 0.5 in.
uranium, BH,O
B'AU'1~-1-8=2B Np, 0.5 in. boral, 1 in. X X
steel, 1 in. uranium, 8
in. borated polyethylene,
2 1In. steel, BH,O
464X No, 4 in. steel, 6 in. x

polyethylene 4 in. steel

99
Table 10.1 (Continued)

 

Configuration
Number

Radiatlion Measured

 

Configuration Description Cama. Fast Thermal

Rays Neutrons Neutrons

 

167X

197X

137X

1-12-6X

1-12=7X

1-12-4X

0=-15=0X

B'A2.5=3=2.5B

B'A134B

B'A431B

B'8 or B

100

Np, 1 in. steel, 6 in. X
polyethylene, 7 in.
steel

Ny, 1 in. steel, 9 in, X
polyethylene, 7 in.
steel

Np, 1 in. steel, 3 in. X
polyethylene, 7 in.
steel

Ny, 1 In. steel, 9 in. X
polyethylene + 3 in.

borated polyethylene,

6 in. steel

No, 1 in., steel, 9 in. X
+ 3 in. borated
polyethlene, 7 in. steel

No, 1 in. steel, 9 in. X
polyethylene + 3 in.

borated polyethylene,

4 1in, steel

Ny, 9 in. polyethylene + 6 X
in. borated polyethylene

Ny, 1/2 in. boral, 2.5 in. x x
steel, 3 in. borated
polyethylene, 2.5 in. steel,
BH,0

No, l/2 in. boral, 1 in. X X
steel, 3 in. borated

polyethylene, 4 in., steel,

BH,0

No, 1/2 in. boral, 4 in. x X
steel, 3 in. borated

polyethylene, 1 in. steel,

BH,O

Ns, 1/2 in, boral, 8 in, Foils
steel, BH,0 (shutter open)

 
 

Table 10.1 (Continued)

 

Radiation Measured

 

 

 

Configuration . . .
Number Configuration Description Camma. Fast Thermal
Rays Neutrons Neutrons
B'A-12-4B  N,, 1/2 in. boral, 1 in. x x
steel, 12 in. borated
polyethylene, 4 in. steel
BH,0
B'A4~12-1B N,, 1/2 in. boral, 4 in. x x
steel, 12 in. borated
polyethylene, 1 in. steel,
BH,0
8 or B N,, 8 in. steel, BH0 Foils
(shutter open)
B1A2,5-12- N,, 1/2 in. boral, 2.5 in. x x
2.5B steel, 12 in. borated
polyethylene, 2.5 in. steel,
BH,0
791B No, 7 in. steel, 9 in. X X
polyethylene, 1 in. steel,
BH,0
Wood placed at side of polyethylene slabs from here on
0=12-0X No, 9 in. polyethylene -+ 3 1in. X
borated polyethylene
0=6~-0X Ny, 6 in. polyethylene
0=15-0X Ny, 9 in., polyethylene + 6 in.
borated polyethylene
A464K Ny, 4 in. steel, 6 in. borated x
polyethylene, 4 in. steel
B'A461B N, 0.5 in. boral, 4 in. X X
steel, 6 in. borated
polyethylene, 1 in. steel,
BH,0
B'A164B N, 0.5 in. boral, 1 in. X x

steel, 6 in. borated
polyethylene, 4 in,
steel, BH,0

101
Table 10.1 (Continued)

 

Configuration
Number

Configuration Description

Radlation Measured

 

Thermal
Neutrons

Fast
Neutrons

Gammsa,
Rays

 

B'A2.5~6=2.5B

434B

731B

B761B

761B

B'A401B

B'A194B

B'A2,5-9-2.5B

N,, 2.5 1In., steel, 6 in,
borated polyethylene,
2.5 in. steel, borated
H,0

No, 4 in. steel, 3 in.
polyethylene, 4 in.
steel, BH,O0

Ny, 7 in. steel, 3 in.
polyethylene, 1 in. steel,
BH,0

No, 7 in. steel, 0.5 in,
boral, 6 in. polyethylene,
0.5 in, boral, 1 in. steel,
BH,0

N,, 7 in., steel, 6 in.
polyethylene, 1 in. steel,
BH,0

N,, 1/2 in. boral, 4 in.
steel, 9 in. borated
polyethylene, 1 in. steel,
BH,0

N2, 1/2 in. boral., 1 in.
steel, 9 in. borated
polyethylene, 4 in. steel,
BH,0

No, 1/2 in. boral, 2.5 in.
steel, 9 1in. borated
polyethylene, 2.5-in.
steel, BH,O

X X

 

102

 
 

11. BULK SHIELDING FACILITY
Stainless Steel~UO, Reactor (BSR-II)

The design and fabrication of a stainless steel, UO,~fueled, compact
core (BSR~II) for the Bulk Shielding Facility reactor has been described,?!
Extensive static and dynamic tests of the BSR~II, performed at the SPERT-I
Facility of the National Reactor Testing Station, have now been completed,
and the core elements and control system are being returned to ORNL.
Although the data from the tests are still being analyzed, some general
results may be reported.

The static tests, which included uranium-foil flux mappings, void
coefficient measurements, and temperature coefficient measurements, were
completed as programmed., The dynamic tests of the BSR-IIL were originally
planned to include tests of the self-limiting behavior of the core during
power excursions of sufficient severity to invite core damage. Subsequent
decisions regarding replaceability of this core in the event of damage
led, however, to a final decision to prevent excessive core damage and
return the original core to the BSF pool for use. It was observed during
early excursion tests that the self=shutdown behavior of this reactor
essentially duplicated the behavior of the similar APPR core P/8/19, which
had been tested previously at SPERT~I in the range of periods down to
14 msec. At a period of 14 msec the peak power of the BSR~II was 226 Mw,
and a very slight amount of plate deformation was observed in the inner-
most fuel plate of an element adjacent to the central fuel element. Since
the APPR behavior can be applied relilably to the BSR=II and is well
understood on the basis of the energy model, no further self=limiting
tests with periods of 14 msec or shorter were run.

Subsequent dynamic testing was therefore done with the full reactor
safety system installed. Peak powers were held to magnitudes of less than

the 226 Mw of the self-limiting tests, and a representative fuel plate

 

1"ANP Semiann. Prog. Rep. Oct. 31, 1959," ORNL-2840, p. 112.

103
was examlined after each test to insure against excesslve cumulative plate
damage. Two=, three-, and occasionally four-rod scrams were utillzed.
The performance of the control system was completely satisfactory and

was 1n accordance with design expectatlons in all respects.

The Spectrum of Prompt Gamma Rays from Thermal=
Neutron Fission of U<°~

 

Data obtained in measurements of the spectrum between 15 kev and 10
Mev of the prompt gamma rays from thermal-neutron fission of U?3° are
being analyzed. All the necessary filssion data have been obtained,2 and
the analysis'has been completed except for the major tasks of determining
the detailed absolute efficiency of the spectrometers at all energies
and translating the observed scintillation~counter pulse~height spectra
into the desired absolute photon energy spectrum. Efforts are being
made to prepare an analytical representation of the shape of the
spectrometer response to monoenergetic gamma rays and to determine the
absolute strength of the sources used to calibrate the spectrometers.
Since the response of the spectrometers must be interpolated through-
out the range of important energies, calibration of the scintillation
palr spectrometer used 1n the energy region above 1.5 Mev is handicapped
by the very small number of monenergetic radiocactive sources which may
be utilized as standards. Only Y%% (1.84 Mev), Na®* (2.75 Mev), and N'©
(6.1 Mev) have been available up to the present in the energy region
of interest. For these sources the shapes of the pulse-height spectral
response functions have been fitted satisfactorily using six parameters.
Shape analyses of the data taken with these sources at various times
throughout the experiment show consistency within the expected statistical
and analytical errors. The pauclty of calibration data available in the
high-energy region has led to the auxiliary experiment described below

using the B'Y(p,y)Cl?* reaction.

 

2"ANP Semiann. Prog. Rep. Oct. 31, 1959," ORNL-2840, p. 110.

104

 
 

Determinatlon of absolute spectrometer efficiencles requires a
knowledge of the disintegration rates of the calibration sources utilized.
In only a few cases could absolute coincidence experiments be performed
on sources measured Iin the spectrometers. Durlng the course of the ex-
perimental work, source strengths were "determined" by iInserting the
sources into two of the high-pressure 47 ionization chambers avallable
at the Laboratory. Final analysis of these data requlres detalled
information about the output current from these 1lonization chambers as
a function of source strength and photon energy. The intensity calibration
of each ion chamber used, fitted to an analytical function covering all
instrument ranges, has now been compared with the accurately known decay

198, The time dependence (on a scale of years) of

of a foil containing Au
the readings from these "standard" ion chambers is also being evaluated.
The absolute calibration of the ion chambers as a function of photon
energy will rest on a calculated energy dependence and on a serles of
absolute gamma-gamma coincidence experiments which have recently been
performed with a precision of somewhat Dbetter than 1% on sources of Sc46,
Coso, Na24, and Y88, Absolute calibrations performed by other workers
will also be employed whenever possible.

In order to provide an absolute calibration of the spectrometer at
higher energles, preparations are belng made for an auxiliary experiment
using the coincident 4.4= and 12-Mev gamma rays from the B(p,y)cl?*
reaction. The BSF 300~kv particle accelerator will be used to produce
the necessary 200=kev protons. A thick boron target will be placed in
the spectrometer at the usual source position, and an auxiliary
scintillation detector placed below the target should permit a coincldence
determination of the absolute source strength simultaneously with the
observation of the pulse~height spectrum. While the beam Intensity will
be inadequate for this experiment, the data obtained should improve
knowledge of the response of the spectrometer at high energies. The
only high=energy calibration data previously available were from an
even more difficult absolute measurement made with the use of the 6.1-

Mev gamma rays 1in the decay of 7-sec N16, Tt 1s hoped that after

105
completion of the present experiment further auxiliary experiments will
be unnecessary and that all efforts may be concentrated on the analysis

of the existing data.

The Model IV Gamma~Ray Spectrometer

Work has continued on the installation of the Model IV gamma-ray

3 All the controls and safety devices have now been in-

spectrometer.
stalled and checked. Before a complete calibration of the readout system
could be made, it was necessary to measure the effect of the weight of

the spectrometer and positioner on the beams of the building. A check

was made by measuring the elevations of points on the beams and then
allowing the beams to support the weight of the spectrometer for 24 hr.

At the conclusion of the test the elevations were again measured. The
maximum deviation observed was 0.03 in. Permanent reference points are

now being established 1n Building 3010, and the calibration of the

readout system willl be completed soon.

In continuation of the investigation of suitable detectors for the
Model IV spectrometer, a crystal of thallium-activated cesium iodide,
¢sI(TL), has been under study. The crystal is a right-circular cylinder
5 in. in diameter and 3 1/2 in. long. The pulse-height distribution
obtained for the 0.662~Mev gamma ray from Csi37 is shown in Fig. 11.1.
The resolution at thils energy was 9.2%, and the photofraction was 76%.
The crystal responded linearly within 0.50% of full scale to gamma-ray
energies between 0.5 and ~/.2 Mev, the maximum energy available for
these measurements. Thils linearity of response is shown in Fig. 11.2.

It was observed that the decay time of the light pulse in CsI(Tl) was
approximately five times as long as that in NaTI(T1).

The photofraction, or the ratio of the area under the total adsorption
peak to the total area of the distribution, is shown for sources of
various energies in Table 11l.1. In this respect the CsI(Tl) crystal
appears to be some 12% better than a NaI(Tl) crystal of similar size, a

 

3"ANP Semiann. Prog. Rep. Oct. 31, 1959," ORNL-2840, p. 115.

106

 
 

UNCLASSIFIED
ORNL—LR—-DWG 46598

 

A

 

 

 

 

 

 

 

 

 

COUNT RATE {arbitrary units)

 

 

¢ ®

 

 

 

 

 

 

 

T
.-—.____.

 

'.......

®

 

 

 

 

 

 

 

.’. ¢
= —— = g *Toeranese®*”® o7 e0-000gg00q 000000’ .."0'.-"0.0"

 

 

 

 

|
\.\

 

0 10 20 30 40 50 60
CHANNEL NUMBER

70 80 90 100

Fig. 11.1. Response of a 5-in.-diem, 3 1/2-in.-Long CsI(T1)

Crystal to the 0,662~Mev Gamma Ray from cst37.
collimated along the axis of the crystal.

The gamma rays were
UNCLASSIFIED
ORNL-LR-DWG 46599

/./
)
e

 

8.0

 

 

-
/
/

/

o 20 40 60 80 100 120 140 160 180 200 220
ANALYZER CHANNEL NUMBER

GAMMA-RAY ENERGY (Mev)

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Flg. 11.2. Measured Linearity of a 5-in.-diam, 3 1/2-in.-Long
CsI(T1) Crystal.
result which may be due mainly to the greater density (4.51 g/cm3 Vs
3.67 g/em®) of the CsI(TL).% A further comparison is being made by
means Of a Monte Carlo calculation, the results of which are not yet

complete.

 

4Other reported characteristics of CsI(Tl), which were not investi-
gated at this time but may be of possible future value, include a varia=
tion in the decay times of light pulses produced by differing particles
that suggests the possibility of classification of emitted particles by
type and the reported peaking of its emissions spectrum in the red that
invites its use wlth a red-sensitive photomultiplier tube.

108

 
Table 11.1. Experimental Response of a 5~in.=-diam,
3 1/2~in.~long CsI(T1l) Crystal to Various
Gamma~Ray Energies

 

Crystal Response

 

 

Source Gamma~-Ray
= Energy (Mev) Photofraction Resolution
(%) (%)
Hg <03 0. 279 75 13
Spti3 0.393 83 11
Na?? 0.511 g2 10.9
cs+37 0.662 76 9.2
y&sé 0. 889 71 8.4
7n6? 1.114 61 7.9
Na?? 1. 277 57 7.6
Na 4 1.368 54, 8.5
788 1. 840 45 6.8
Na 24 2.75% 36 6.8
N6 6.12 oty

 

Compilation of Data on Gamma=Ray Spectra Resulting
from Thermal-Neutron Capture

 

A detailed and up~to-date listing of data on gamma-ray spectra
resulting from thermal-neutron capture in a total of 67 different nuclel
has recently been completed.5 The compilation makes use of two principal
sources of data, the first being the work of Kinsey and Bartholomew at
Chalk River over the energy range from ~3 to ~12 Mev and the second being
the results of a Russian group, headed by Groshev, whose data, although
it extends to 12 Mev, is most detalled in the lower regions of a 0O,3-
to 12-Mev range. In the considerable range of overlap, the Chalk River
data have arbitrarily been given precedence in cases of disagreement.
Where other sources of data have existed, use has been made of such
results.

The results of the compilation are presented as two tables of data

and two sets of graphs. The first table gives the number of photons

 

5The work reported here was performed for ORNL under Subcontract
1216 by Nuclear Development Corporation of America.

109
per 100 captures for the seven energy intervals (1, l~2,-2—3, 35, 57,
7—9, and >9 Mev, the highest-energy gamma ray, and the average number of
photons per capture for 67 muclel. The second table gives the energiles
and Intensities of the dlscrete gamma rays resulting from thermal-neutron
capture by the 25 nuclel of most Interest in shielding problems. The
two sets of graphs show elther continuous capture spectra or bar-graph
line spectra according to whether the capture spectrum of the particular
nucleus 1s essentlally continuous or essentially a line spectrum. The
tables are accompanied by complete references, as well as a detailed
bibliography of pertinent works.

The general subject of capture gamma rays is also discussedj some
exlsting gaps in the present information are notedj and the user of the
compilation is cautioned that the capture spectra resulting from neutrons
of energles greater than thermal may differ importantly from the thermal
results presented. The detailed compilation, now 1s press, is being
published as ORNL~2904, 6

 

E. Troubetzkoy and H. Goldstein, "A Compilation of Information on
%amma;Ray Spectra Resulting from Thermal-Neutron Capture,’" ORNL=2904
1960).

110

 
12. TOWER SHIELDING FACILITY

Tower Shlelding Reactor II

 

Reactor Assembly

The design and function of the Tower Shielding Reactor II (TSR-IT)
have been described in detail.'s2s2 The TSR=II has now been completely
assembled at the Tower Shielding Facility and successfully operated at
low power.

During inltlal assembly of the TSR~II core in the pressure vessel,
extensive fitting of parts was found to be necessary, and many parts were
reworked after exact measurements were obtained of the place they were
to be located in the assembly. This was largely the result of the close
tolerances required in the wildths of water passages 1n and around the
core in order to obtaln the water flow distribution required for adequate
cooling of the fuel plates. In addition, it was discovered that for some
time after fabrication the fuel assemblles suffered gradual distortion
that increased the included angle of each segment and prohibited their
assembly into a full 360-deg arrangement., Side plates of the fuel
assemblies were machined, and the elements were heat soaked for a total
of 86 hr at 280°F to insure against further distortion.

In addition to the difficultles encountered during assembly of the
core, a further delay was indirectly caused by the unseasonably cold
weather which prevailed during the assembly period. Although the water
cooling system for the reactor is normally protected from freezing by
built=in, thermostatically controlled heaters, the heating system was
not connected for the preliminary experiments, and considerable dif=
ficulty was encountered from freezing of lines and components. A

delay of about one week was experienced when a short circuit in a

 

10, E. clifford and L. B. Holland, "ANP Quar. Prog. Rep. Dec. 31,
1956, " ORNL~2221, p. 352.

21,, B, Holland et al., "Neutron Phys. Amnn. Prog. Rep. Sept. 1, 1959,"
ORNT~2842, p. 39.

3"ANP Semiann. Prog. Rep. Oct. 31, 1959," ORNL-2840, p. 146.

111
temporary heater used to prevent freezing ilgnited a tarpaulin covering

the control turret. The fire caused minor damasge to control system
wiring.

The last major component of the TSR=II to be designed and assembled
was the control turret, which is mounted on the top plate of the
ionizetion chamber guide assembly. The control turret contains the shim
and control rod drive motors, speed reducers, and position transmitters}
the seat switch pump, manifolds, and differential pressure switchesj; and
the shim rod clutch manifold, magnet valves, and pressure transducers;
as well as electrical and water system quick-disconnects, the fission
chamber water accumulator, and the fission chamber drive rack. Since it
was necessary, 1n order to preserve the symmetry of flux desired, to
minimize the solid angle above the core occupied by materials other than
air and not spherically symmetric with the core, the design of the
control turret was complicated. The physical size of all members and
mechanisms was held to a minimum, and the resulting assembly of control
system components was very compact.

From the control turret, the five water tubes for the shim rod
clutches extend to O-ringed unions in the top head of the control
support tube, which runs down through the ionization chamber guide
assembly to the control mechanisms in the internal reflector region.
Flve seat-switch tubes make a union in the same area. In addition, two
5/8-in. tubes carrying shim bypass water, one l/2~in. tube for shim pump
supply, one 3/4=in. tube for central-sphere coolant, one 1/4~in. tube
furnishing reference pressure supply for seat-switch transducers, and
one 3/8-in. tube for the fission chamber water accumulator, all pass

from the control turret into the ionization chamber guide assembly.

Critical Experiments

To assure safe initilal operation of the TSR-II, a special test
setup, shown in Fig., 12.1, was constructed. The reactor was suspended
in an elevated test tank, and water for moderation was pumped from the

reservolr shown in the figure. To allow water to cover the fuel, a

112

 
£TT

RESERVOIR TANK

 

 

 

UNCLASSIFIED
ORNL—-LR-DWG 49395

 

REACTOR :
SUPPORT
COLUMNS/'

CONTROL TURRET

  
  
     

REACTOR ASSEMBLY

4-ft-dia ALUMINUM

I SCRAM-TRIPPED
: SOLENOID DUMP
VALVE

 

 

 

 

 

 

 

 

22 ft

 

 

 

 

 

 

SOLENOID-OPERATED [ -
J DRAIN VALVE coE?ggsfingg i
- VAL ) 3in—-dia HOLE IN

— - P REACTOR c
— - < PRESSURE ®
) - VESSEL T
_ ) Y L—M— y 0
I = &

e SHIM PUMP \k\‘CONCRETEPAD

{(POSITIVE DISPLACEMENT TYPE)

Fig. 12.1.

{TURBINE TYPE)

Test Setup for TSR-II Preliminary Experiments at the Tower Shielding Facility.
3=in.=-diam hole was left in the bottom of the reactor tank. Extra

monitoring circuits were connected to the scram circuits, and the scram
levels were set at very low levels. If the reactor power reached the

trip level, or 1f the reactor period became less than 1 sec, two magnets
connected to the reactor scram circuits were de-energlzed. This permitted
the dump valve in the test tank to open and draln the water from the

tank.

Nuclear Calculatlons

 

The originally proposed4r5 miltigroup, miltiregion calculations of
the TSR=II with final geometry and materials have been completed utllizing
the GNU=II® code for the IBM=704. The calculations required the assumption
of a uniform distribution of the constituents through a glven region, with
each reglon assumed to be a continuous spherical shell. This assumption
is falrly well satisfied by the construction of the TSR~IIL. The results
of some of the computations and the corresponding reactor geometry are
shown in Fig. 12.2. These results are being used to predict the results

of the Initial ecritical experiments at the TGSF.

Flow Distribution Studles

The modification of the hydraulic flow test system for studying
the flow distribution through the fuel elements in the central cylinder
was described previously.7 With only minor changes the flow through
elther the lower or upper elements can now be measured., Flow measure-
ments in the fuel element channels are made by injecting a "slug" of
salt solution into the water flowing through the central cylinder and
measuring the time required for the salt solution to travel between two

probes which sense the change 1In electrical conductivity of the water.

 

4"ANP Quar. Prog. Rep. Dee. 31, 1957," ORNL=2440, p. 279.
5"Appl. Nuclear Phys. Ann. Prog. Rep. Sept. 1, 1957," ORNI~2389, p. 6l.

éC. L. Davis, J. M. Bookston, and B. E. Smith, "GNU-II, A Multigroup
One-Dimension Diffusion Program,"” GMR~101 (1957).

7"ANP Semiann. Prog. Rep. Mar. 31, 1959," ORNL-2711, p. 145.

114

 
 

 

Keft
VALUES (%)
RELATIVE TO

CASED

+0.98%
-5.59%
-1.96%
~0.62%,
-0.459%
+0.65 %
+14,57%
+1.10%
-1.79%
-0.005%
-0.97 %
-0.26 %
+1.739%
+047 %

J
K
L
M
N
P
Q
R
S
T
U
v
W
X

REGION ¢

 

UNCLASSIFIED
ORNL—LR-DWG 49396

REGION 7

 

>

l

+2.38% L

L] 1

 

-7.80% | B |

11 1

 

-6.10 % | B+CD |

 

-2.18% | B—CD]

1L |

 

+0.93% lD+cnl

L [

 

+3.04%, ] D-CD [
0

K

FOR

STANDARD
D
COMPARISON

W x T C A YO0 VD ZT IR -

B+ CD
B -CD
D+ CD
D-CD

Fig. 12.2.

Computed Values of TOR-IT k

sl=f=l=l=l=

[1
[]

I [T 1
[]
]
[T

T [T , |

30 40 50 60 70
REACTOR RADIUS {(cm)

1
10

n
o

EY:
Hp_O+AI B4C @Pb AU+ +al [ ]H0 ONLY

DESCRIPTION OF CASES
CONTROL PLATES (REGION D3) WITHDRAWN TO NEAR CRITICAL (k¢ = 0.9954)

REGION J2 CONTAINS AIR AND Al INSTEAD OF H,O AND Al

REGION K4 CONTAINS AIR AND Al INSTEAD OF H,0 AND Al

REGION L6 CONTAINS AIR AND Al INSTEAD OF H,0 AND Aj

REGION M5 CONTAINS ONLY 50% AS MUCH U AS BEFORE

REGION N7 WATER IS AT 140°F INSTEAD OF 66°F TO SIMULATE BUBBLES

REGION P12 CONTAINS 50% Pb AND 50% H,0 INSTEAD OF 100% H,0

REGION Q12 CONTAINS 50% Pt AND 50% AIR INSTEAD OF 100% H,0

REGION R12 CONTAINS 100% LiH (75% OF Li IS Li®) INSTEAD OF 100% H,0

H,O IN ALL REGIONS IS AT 183°F INSTEAD OF 66°F

INNER 50% OF REGION T6 CONTAINS H,0 ONLY; QUTER 50% 5 H,0 + Al AS BEFORE
INNER 50% OF REGION US CONTAINS AIR +10% H,0; OUTER 50% IS H,0 + Al AS BEFORE
REGION V12 CONTAINS 100% AIR INSTEAD OF H20

REGIONS W9 AND W10 CONTAIN 100% H,0 INSTEAD OF Pb AND BORAL

REGION X5 IS SHIFTED 0.5 cm INWARD

CONTROL PLATES (REG!ON A3) COMPLETELY WITHDRAWN

CONTROL PLATES (REGION B3} SCRAMMED

LIKE CASE B EXCEPT WITH 0.5 cm CADMIUM INSTEAD OF B,C IN REGION 3

LIKE CASE B EXCEPT WITH 0.00577 cm CADMIUM INSTEAD OF B,C IN REGION 3
LIKE CASE D EXCEPT WITH 0.5 cm CADMIUM INSTEAD OF B,C IN REGION 3

LIKE CASE D EXCEPT WITH 0.00577 cm CADMIUM INSTEAD OF B,C IN REGION 3

off Resulting from Assumed

Deviations from Normal Core Arrangement.

115
Times of the order of milliseconds are recorded by two pens on a strip
chart. A number of readings are taken for each poilnt, and the readings
are averaged.

The baffle plate intended to distribute the flow through the lower
elements 1s hemispherlcal, and is spaced 1/8 in. above the lower central
elements. It is perforated with a pattern of 0,106~in.~diam holes. The
pressure drop measured across the control mechanlsm housing, the baffle
plate, the lower fuel elements, and the end of the cylinder is approximately
18 psi. Some difficulty is still being encountered, however, in optimizing .
flow distribution in the lower central fuel elements. Baffle plates that
have been tallored to give quite satlsfactory results with the dummy fuel
element do not appear to yleld the same results with the actual fuel ele~
ments. Typlcally, flow 1n two or three of the 40 channels of the actual
fuel elements appears to be swirling or to be otherwise unsatisfactory.

Re=-examination of the salt-conductivity measuring technique has
shown that the indication of swirling was in most cases the result of
poor measuring geometry. The technlque depends for its accuracy upon
the inserted slug of solution maintaining its spatial distribution as it
passes the two conductivity probes. The salt slug tends to develop
filaments or streamers on the wave front and gives a false indication
of random swirling. An Indication of good flow, however, has never been
found to result from bad conditions.

Since near-optimum flow distribution through the dummy elements was
obtained by careful tailoring, an effort is now being made to obtain .
similar results in the actual fuel elements by the same careful tailoring
of the baffle plates. Measurements have also begun on the flow distribution
through the upper elements, but no conclusive results have been obtained

at present.

Experimental Program

Detector Shield

 

The spherical lead~and-water shield® to be used to collimate the

radiation at the detector has been delivered to the TSF. It is shown

116

 
suspended from its ground support platform in Fig. 12.3. Before the shield
can be suspended from the TSF towers and completely checked out, it will
be modified slightly to accommodate a 9 by 9-in. NaI(T1l) ecrystal, which

will be used for gamma-ray spectroscopy.

Gamma-Ray Spectrometry

 

Available photomultiplier tubes to be used for spectral measurements
have been tested for gain shifts as a function of count rate over a range
of from 1 000 to 100 000 counts/sec. Several tubes were found which have
a gain change of <2% and are satisfactory for use.

A 400-channel pulse-height analyzer was received from Radiation
Instruments Development Iaboratory in September. Tests showed nonlinearity
of ~5% from channel & to channel 390. Very minor modifications were made,
and the tests now show a nonlinearity of <0.5%. A major fault was noted
in the live timer. The error was found to be a function of the count rate,
and under the best conditions varied from 5% at an output of 3 000
counts/sec to >20% at 30 000 counts/sec. A redesign of the output circuit
of the live timer reduced the error to ~0.6% and made the error independent
of count rate.

The input channels of the pulse-height analyzer have been modified
to permit the trigger circuits to be triggered by a pulse from the
amplifier PHS circuit ~0.5 pusec before the arrival of the pulse to be
analyzed. This modification improves amplifier stability and contributes
to the improved linearity of response,.

A DD2 amplifier equipped with 4-psec pulse shaping cables was tested
using both pulses from a LiIl crystal and pulses from a test-pulse generator
simulating those from the crystal. Both the amplifier and analyzer perform

well with the shape and length of pulses produced in this manner.

Neutron Spectrometry

 

A new approach to neutron spectroscopy is being investigated at the

TSF. The configuration of a detector consisting of two surface-barrier

 

8"ANP Semiann. Prog. Rep. Oct. 31, 1959," ORNL-2840, p. 138.

117
UNCL ASSIFIED
PHOTOD 49739

"y

ettt

] Y

L Sl W =

..."I,-

|
|

s

i e G v

s B X SN WY | T

 

4

Shield.

Detector

TSF Collimating

Fig.

118
 

silicorn diodes? 10 coated successively with gold and I1i®F and having surface
areas of ~1 cm? is shown in Fig. 12.4. Typical thicknesses may be 50 to
100 pg/em?® for gold and 100 to 150 ug/cm? for LiSF. The circuit in which
the signal 1s developed is also shown in Fig. 12.4. The well-known
1i(n,Q)T reaction is utilized to introduce the charged particles into the
diodes, which are electrically connected in parallel. The resultant
collected charge is developed across Ry and fed into an amplifier and
analyzer system for analysis. Since the Q for the Li6(n,a)T reaction is
.78 Mev, the thermal energy peak appears at this energy. The response

to energies above thermal should be linear with neutron energy. Tests of
the response of the spectrometer to high~energy neutronsg will be made at
the ORNL High Voltage Laboratory.

 

9. . Halbert and J. L. Blankenship, "Response of Semiconductor
curface-Barrier Counters to Nitrogen Ions and Alpha Particles," March 1960,
unpublished.

107. 1. Blankenship and C. J. Borkowski, "silicon Surface-Barrier
Nuclear Particle Spectrometer," presented at the Seventh Scintillation
Counter Symposium, Washingtonm, D. C. (Feb. 1960) .

UNCLASSIFIED
ORNL-LR-0OWG 49397

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Cy
PREAMP

 

Ry (1 Meg)

0-30v

o

Fig. 12.4. Silicon Diode Neutron Spectrometer.

119
Calculations, based on L1% cross sections given in BNL-325,11 predict
an efficiency of the order of 10”® for 1-Mev neutrons. The theoretical
resolution for thermal neutrons is 5.3% and should be improved somewhat
at higher energies. Such a spectrometer will be useful in the energy
range from ~750 kev to ~5 to 10 Mev.

A typical curve obtained from a silicon-diode detector at room tempera-
ture using thermal neutrons from a Po-Be source imbedded in paraffin is
presented in Fig., 12.5. The peak at the lower part of the curve is due
to amplifier noise, while the small peak is attributed to events in which
elther the alpha particle or the triton escaped because 4T geometry was
not realized. This peak can be reduced by placing the diodes closer to-

gether and reducing the area on which the Li®F is coated.

Data Processing

 

An automatic data acquisition and processing system for the Tower
Shielding Facility is under development that consists of three parts.
The first part is a punched-paper=tape output which is being installed
on the 400-channel RIDL analyzer, and the second part is a Moseley plotter
that will accept the analyzer tape for immediate plotting and inspection
of data. Both these units are scheduled for delivery and installation by
June 15, 1960. The third part is a processing system for gamma-ray
spectra that is being developed using the IBM~704 (and eventually the
IBM~7090) computer. The approach taken in the spectra analysis assumes
nothing about the shape of the spectra. Assuming nothing sbout the
spectra facilitates analysis when bremsstrahlung is a sizeable portion of
the input or close peaks are not resolved by the analyzer. The analysis
considers an input gamma ray to be possible in every one of the channels.
Response functions, determined by interpolation of single-energy data,
are least-squares fitted to the unknown spectrum. In order to insure

convergence, artificlal smoothing of the unknown spectrum is used. The

 

1p, J. Hughes and R. B. Schwartz, "Neutron Cross Sections,” BNL~325,
2nd. ed. (1958).

120

 
 

 

 

 

 

 

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UNCLASSIFIED
PHOTO 50296

i M

12.6. Nonlinear Function Fit to a
Experimental Gamma-Ray Response Function.
result of the smoothing is to change a sharp line input to one of finite
width; nevertheless, the net resolution is somewhat better than that of
the spectrometer. Results obtained thus far indicate that a typical ex-
perimental response function can be very well fitted by a nonlinear

function of the form

R(E’,E7) = Ay exp — [(x - XOi)/in]z

photo peaks
escape peaks
backscatter peaks

=[G =) | em =[G - <)%,
1/2

1 + K? exp [(x - xoo)/xjj]

 

-+

exp [(x - xoo)/xjj] + {l — K exp [(x - xoo)/x.J] :

J

A typical fit is shown in Fig. 12.6, in which the natural logarithm
of the total counts is plotted as a function of channel number. The
larger points are data points, the smaller are the fitted curve, and
the points enclosed by triangles represent the differences between

fitted curve and data.

122

 
 

 

 

ORNL-2942

ors—opecial Features
Aircraft Reactors
[1-3679 (24th ed.)

   
    
   

DISTRIBUT]S

    
   

1. J. W. Allen 37.48R. N. Lyon
2. D. S. Billington 3887, S. Luce
3. F. F. Blankenship 3§f 'H. G. MacPherson
4. E. P. Blizard A F. C. Maienschein
5, A, L. Boch ¥ V. D. Manly
6. G. E. Boyd A, J. Miller
7. R. B. Briggs K. Z. Morgan
8. A. D. Callihan F. J. Muckenthaler
9. C. E. Center (K-25) E. J. Murphy
10. R. A. Charpie J. P. Murray (Y-12)
11. W. B, Cottrell M. L. Nelson
12. F. L. Culler P. Patriarca
13. D. A. Douglas S. K. Penny
14. L. B. Emlet (X-25) P. M. Reyling
15. A, P. Frass H., W. Savage
16. J. H. Frye A. W. Savolainen
17. R. J. Gray E. D. Shipley
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2. N. E. Hinkle D. K. Trubey
25, M. R. Hill G. M. Watson
26. E. E. Hoffman A. M. Weinberg
27 H. W. Hoffman J. C. White
28, A, Hollaender E. P. Wigner (consultant)
29. L. B. Holland C. E. Winters
30. H. Inouye Laboratory Records Department
31. W. H. Jordan Laboratory Records, ORNL R. C.
32. G. W. Keilholtz t ORNL — Y~12 Technical Library
33. C. P. Keim Document Reference Section
34. J. J. Keyes f__Central Research Library
35. P. G. Lafyatis P
36. R. S. Livingston

123

 
 

EXTERNAL DISTRIBUTION

g1, AiResearch Manufacturing Company
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9293, AFPR, Lockhed Marietta
94. AFPR, North Arfrican, Downey
9596, Air Force Spec¥Rl Weapons Ce
97—98. Air Research arf Developmeng
99, Air Technical Ifjelligence
100—-102. ANP Project Offif, Convaij
103. Albuquerque Oper"f‘ons of
104. Argonne National % g borato
105—106., Army Ballistic Ml bile 7
107. Army Rocket and Gg~ped 51le Agency
108. Assistant Secretari@of i
109-114. Atomic Energy Comm¥ig#y Washington
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126. Bureau of Yards §
127—128. Chicago Operati
129, Chicago Patent
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131. Director of N
132. duPont Compan
133. Engineer Rese
134—141. General Elec
142—144, General Ele

   
    
    
    
     
    
    
     
    
      
    
    
  
    
  
  
  
  
    
  
 
 
  
   
  
   
  
 
  
  
  
   
    
  
 
    
 
 

l’

    

  

  

 
 

ter

 
 

  

  

  

  

  

Intelllg ace
Aiken '
h and Develglment Laboratories
ic Company (ANSD)
ic Company, Rid@land
145. General Nuck#ar Engineering C@poration
146. Hartford Aifferaft Reactors Ard@ Office
147. Idaho TestyDivision (LAROO) @g
148-149, Knolls Atg@liic Power Laboratory
150. Lockland craft Reactors Ope atlon Office
151, g Scientific Laborat@?y
152. Marquard Aircraft Company
153, Martin COmpany
154, National Aeronautics and Space Administration, Cleveland
155, National Aercnautics and Space Administration, Washington

'j,“

 

 

124

 
156,
157.
158.
159.
160,
16l.
162.
163.
164.
165.
166,
167-168.
169—172.
173.
174.
175—176.
177.
178.
179,
180.
181.
182.
183184,
185—187.
188-199.
200—224.
225.

 

 

National Bureau of Standards
Naval Air Development @enter
Naval AW Material Ceyjs
Naval Aif Turbine TegiStation
Naval ResQgrch Laboyory
New York ORgrationsirfice

Nuclear Met¥hs, Ingl

Oak Ridge Opdgatigs Office

Office of Navay Riearch

Office of the Wil of Naval Operations
Patent Branch, Whington

Phillips Petroliilh Company (NRTS)

Pratt & Whitneyj craft Division
Public Health J@¥FWce

Sandia Corporafn
School of Aviggon @edicine
Sylvania-Cornjis NuSlear Corporation
Technical Recf@rch (Roup

USAF Headqua @ers
USAF Project JEAND
U. S. Naval Estgraduayge School

U. S. Naval @ediologicaf Defense Laboratory
University off Californid@ Livermore
WestinghouscfBettis Atom¥ Power Laboratory
Wright Air Ovelopment DiW@sion

Technical Information ServMe Extension
Division of Research and Dev§lopment, AEC, ORO

  
 
       
   
  
   
 
   
  
  
  
 
   
 
  
 
 
    
   
 

  

 

125
Reports previously issued

126

ORNL=-528
ORNL=629
ORNL-768
ORNL-858
ORNL-919
ANP~60
AlNP=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-17"71
ORNL~-1816
ORNL~1864
ORNL~1896
ORNL=1947
ORNL=-2012
ORNL~-2061
ORNL=-2106
ORNL~2157
ORNL=2221
ORNL=2274
ORNL=~2340
ORNL=2387
ORNL~-2440
ORNL-2517
ORNL=-2599
ORNL=~2711
ORNL=~2840

Period
Period
Period
Pericd
Period
Period
Period
Period
Period
Period
Period
Period
Period
Period
Period
Period
Period
Period
Period
Period
Period
Period
Period
Period
Period
Period
Period
Period
Period
Period
Period
Period
Period
Period
Period
Period
Period

in this series are as follows:

Ending
Ending
Ending
Ending
Ending
Ending
Ending
Ending
Ending
Ending
Ending
Ending
Ending
Ending
Ending
Ending
Ending
Ending
Ending
Ending
Ending
Ending
Ending
Ending
Ending
Ending
Ending
Ending
Ending
Ending
Ending
Ending
Ending
Ending
Ending
EBnding
Ending

 

November 30, 1949
February 28, 1950
May 31, 1950
August 31, 1950
December 10, 1950
March 10, 1951
June 10, 1951
September 10, 1851
December 10, 1951
March 10, 1952
June 10, 1952
September 10, 1952
December 10, 1952
March 10, 1953
June 10, 1953
September 10, 1953
December 10, 1953
March 10, 1954
June 10, 1954
September 10, 1954
December 10, 1954
March 10, 1955
June 10, 1955
September 10, 1955
December 10, 1955
March 10, 1956
June 10, 1956
September 10, 1956
December 31, 1956
March 31, 1957
June 30, 1957
September 30, 1957
December 31, 1957
March 31, 1958
September 30, 1958
March 31, 1959
October 31, 1959