SOVIET ATOMIC ENERGY - VOL. 34, NO. 3

• Document Type: CREST
• Collection: General CIA Records
• Document Number (FOIA) /ESDN (CREST): CIA-RDP10-02196R000400010003-7
• Release Decision: RIFPUB
• Original Classification: K
• Document Page Count: 117
• Document Creation Date: January 4, 2017
• Document Release Date: September 15, 2013
• Sequence Number: 3
• Publication Date: September 1, 1973
• Content Type: REPORT

Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
Russian Original Vol. 34, No. 3, March, 1973'
September, 1973
SATEAZ 34(3) 193-304 (1973)
SOVIET
ATOMIC
ENERGY
ATOMHAFI 3HEPitilfl
? (ATOMNAYA iNERGIYA)
TRANSLATED ,FROM RUSSIAN
CONSULTANTS BUREAU, NEW YORK
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
SOVIET
ATOMIC
ENERGY
?
Soviet Atomic -EnergY is .a cover-to-cover translatiOn of Atomnaya
Energiya,e publicatibn of Acader4 of Sciences of the USSR. ,
An; arrangement with, Mezhdunarodnaye Krfiga,. the Soviet book
export agency; makes available both advance copies of the Rus-
sian journal_ end origirial glossy ph,otogrephs and artwork.-.This
serves to decrease the necessary time leg befween' publication
of the original and publication of the translation and helps to im-
- prove the quality of the latter:The translation began with the first
issqe of the Russiari jqyrnal.
Editorial Board of Arciinnaya Energiya:
?
' Editor:, M. D.' Millionshchikov
DeputY Director. ,
I. V. Xurchatov Inititute of Atomic Energy
Academy of Sciences of the USSR
Moscow, USSR '
Associate Editors: N: A. Kolokol'tsov
N. A. Vlasov
A. Al?Bochvar
N. A: bolleihar
Fursov
, F. N. Golovin ?
V. F. Kalinin
A. K. Krasin
A. I. Leipuns'kii
;s A. R Zefirov
V. V. Matveev
M. G. Meshcheryakov
P. N. Palei
V. B. ShevC4nko
D. L. Simonenko,
V. I. Srnirnov
A. P. Vinogradov
Copyright?1073 Consultants Bureau, New York, a division of Plenum Publishing
Corporation, ?27 West 17th Street, New York, N.Y. 10011. All rights reserved.
No article contained herein may be reproduced for any purpose ,whatsoever
without permission of the publishers. A
-
Consultants Bureau Journals eppear' about six months after the publication of the
original Russian issue. For bibliographic acCuricy, the English iseue published by
Consultants Bureau carries the same number and date as the original Russian
from which it was translated. For example, a Russian issue published In pecern-
ber will appear,in a Consultants Bureau English translation about the ifollowing
June, but the tranelation issue will carry the December date. When ordering any
volume or particular issue of a Consultants Bureau journal, please specify the
date and, where applicable: the volume and issue numbers of the original Russian.
The material you will receive will be a translation of that Russian volume or issue.
Subscription,?
$80 per volume (6 Issues)' Single Issue: $30
2 volumes per year Single,Article: $15
(Add $5 for ordera-outaide the United States end Canada.) ,
CONSULTANTS BUREAU, NEW YORK AND LONDON -
227 West 17th Street
New York, New York 10011
' Davit House
S Scrubs Lane
,Harlesden, NW1016SE
England, ?
Publisfied monthly. 'Second-class postage paid, at Jamaica, New York 11431.
?
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
SOVIET ATOMIC ENERGY
A translation of Atomnaya Energiya
September, 1973
Volume 34, Number 3
March, 1973
CONTENTS
Engl./Russ.
On the Occasion Of the Sixtieth Birthday of Academician Georgii Nikolaevich Flerov . .
193
145
On the Occasion of the Sixtieth Birthday of Igor' Nikolaevich Golovin
194
146
ARTICLES
Development of Fuel Elements for Fast Power Reactors ? I. S. Golovnin,
Yu. K. Bibilashvili, and T. S. Menishikova
196
147
The SystemMo03?UO3 ? 0. A. Ustinov, M. A. Andrianov, N. T. Chebotarev,
and G. P. Novoselov
203
155
Buildup of Transuranium Elements in VK-50 Reactor Fuel ? V. Ya. Gabeskiriya,
V. S. Belokopytov, Yu. B. Novikov, V. G. Polyukhov, V. M. Sarychev,
G. A. Simakin, and A. P. Chetverikov
206
159
Regular System of Closely :Spaced Neutron Absorbers ? I. L. Chikhladze
and Ya. V. Shevelev
210
163
BIBLIOGRAPHY
New Books
217
169
BOOK REVIEWS
D. L. Broder et al. (editor). Manual on Radiation Shielding for Engineers, Vol. I
? Reviewed by U. Ya Margulis
220
170
ARTICLES
Test of Neutron Diffusion Theory in a Medium with Channels by the Pulsed Source
Method (Single Channel in a Moderating Block) ? I. F. Zhezherun
221
171
Production of Gamma-Active Isotopes in Soil by Neutrons with Energies up to 1 GeV
? A. A. Aleksandrov, E. K. Gel'land, B. V. Manyko, Yu. T. Mironov,
B. S. Sychev, and S. I. Ushakov
227
177
Acceleration of Electrons in the Slow-Wave Field of a Plasma Waveguide
? A. M. Egorov, Ya. B. Fainberg, V. I. Kurilko, A. F. Kivshik, L. I, Bolotin,
and A. F. Bats
230
181
Dose Fields of a Clinical Proton Beam Studied with a Radiation Track-Delineating Flaw
Detector ? M. F. Lomanov, G. G. S'himchuk, and R. M. Yakovlev
235
185
ABSTRACTS
Optimization of Reactor Reactivity Behavior by Burnable Poisons ? A. V. Voronkov
and V. A. Chuyanov
243
193
Solution of Neutron-Diffusion Problems in Heterogeneous Flat Reactors by the Direct
Variational Method ? N. V. Isaev and I. S. Slesarev
244
194
Effect of Space Charges in Insulator on Accuracy of Emission Detector Readings
? N. A. Aseev and B. V. Samsonov
245
194
Theory of the Transport of Nonstationary Gamma Radiation ? N. A. Seleznev
246
196
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
LETTERS TO THE EDITOR
Power Distribution in Fuel Element Meat - N. N. Ponomarev-Stepnoi, A-. M. Krutov,
CONTENTS
(continued)
Engl./Russ.
V. A Lobyntsev, and V. I. Nosov
248,
197
On the Use of an Electron Cyclotron for the Rapid Photon Activation Analysis of Ore
-- Samples for Gold = S. P. Kapitsa, Yu. T. Martynov, V. V. Sulin, ?
and Yu. M. Tsipenyuk
251
199
Features of Activation Analysis of Moving Matter Using a Fast Neutron Source
- V. V; Strelychenko and K. I. Yakubson
254
201
Some Characteristics of Electron-Emission Neutron Detectors with Ag, Ag109, Rh,
and Gd Emitters - I. Ya. Emeltyanov, Yu. I. Volodtko, 0. K. Egorov,
L. V. Konstantinov, and V. V. Postnikov
257
203
Radiation Chemical Conversions of Iodine in the System Tributyl Phosphate-Hexane -H20
-HNO3 - P. A. Zagorets, Z. I. Rasldna, G. P. Bulgakova, V. M. Makarov,
T. G. Sazhina, and T. N. Agafonova
261
205
Spiral Instability of a Plasma Filament of Elliptical Cross Section - L. S. Solov'ev
- L. S. Solov'ev
264
207
Synchronous Motion of Charged Particles in a Traveling-Wave Field - V. M. Mokhov
and V. V. Kushin
267
209
Production of Neutrons by Cosmic Rays at Various Depths Underground - G. V. Gorshkov
and V. A. Zyabkin
296
210
COMECON NEWS
XXIII Session of the COMECON Permanent Comission on Peaceful Uses of Atomic Energy
- V. A. Kiselev
272
215
Budapest Conference on Implementation of Radiation Processes and Radiation Facilities
- V. P. Averniaov
273
215
Collaboration Daybook
274
216
CONFERENCES
International Conference on Safety Engineering of Fast Reactors - Yu. E. Bagdasarov
276
217
Symposium on the Chemistry of the Transuranium Elements - N. N. Krot
and I. K. Shvetsov
280
219
September 1972 Symposium on Collective Methods of Acceleration - V. P. Sarantsev.
284
222
Second International Conference on Ion Sources - A. S. Pasyuk
287
223
Saclay October 1972 International Conference on Activation Analysis - B. S. Kudinov. . . .
290
225
On-Line-72 International Conference on Computerization Techniques - V. I. Prikhodiko
and A. N. Sinaev
293
227
Conference on X-Ray Spectral Analysis - S. V. Mamikonyan
296
228
Applications of Radioisotope Equipment in the Coal Industry - R. S. Morusan
298
229
Conferences and Seminars of the All-Union Isotope Association
300
230
NEW INSTRUMENTS
The Kvant-1 Direct-Reading Signal Dosimeter - I. E. Mukhin, G. A. Glinskii,
and V. S. Karasev
302
231
The Russian press date (podpisano k pechati) of this issue was 3/1/1973.
Publication therefore did not occur prior to this date, but must be assumed
to have taken place reasonably soon thereafter.
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
ON THE OCCASION OF THE SIXTIETH BIRTHDAY OF ?
ACADEMICIAN GEORGII NIKOLAEVICH FLEROV
The editorial staff of the periodical Atom.naya Energiya warmly greets Academician Georgii Niko-
laevich Flerov on the occasion of his sixtieth birthday, and wishes him excellent health, long life, and new
creative successes.
Translated from Atomnaya tnergiya, Vol.34, No. 3, p. 145, March, 1973.
0 1973 Consultants Bureau, a division of Plenum Publishing Corporation, 227 West 17th Street, New York,
N. Y. 10011. All rights reserved. This article cannot be reproduced for any purpose whatsoever without
permission of the publisher. A copy of this article is available from the publisher for $15.00.
193
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
ON THE OCCASION OF THE SIXTIETH BIRTHDAY OF
IGOR' NIKOLAEVICH GOLOVIN
I. N. Golovin was born in Moscow on March 12, 1913. He graduated from the Physics Department of
Moscow State University in 1936. His graduation thesis, The Present Status of the Theory of Nuclear
Forces, was recognized as outstanding. In the years 1936-1939, Golovin was a graduate student under the
supervision of I. E. Tamm. .For his work on the theory of vacuum polarization, he was awarded the learned
degree of candidate in physical and mathematical sciences. After completing his graduate work, he took a
teaching position at the, Moscow Aviation Institute.
During the first years of the war, Golovin was enrolled in the local home guard. He was later assigned
to Alma-Ata, where the Aviation Institute had been evacuated. There, in addition to his teaching duties, he
carried on scientific work at the Physics and Engineering Institute of the Ukrainian SSR Academy of Sciences,
which was also located at Alma-Ata during the period.
In 1944, I. V. Kurchatov invited Golovin to take part in work connected with the production of atomic
energy. For the next eight years he served as the first assistant director of the Institute of Atomic Energy.
Work on controlled thermonuclear fusion was in progress from the very outset at the Institute, and Golovin
became involved in that research, and soon became director of the OGRA thermonuclear division, which
was set up under I. V. Kurchatov's instructions.
Translated from Atomnaya Eneaitiy-a, Vol. 34, No. 3, ,p: 1467 March; 1973:
C 1973 Consultants' Bureau, a division of Plenum Publiahing CorporatiOn, 227 West 17th Street, New York,
N. Y. 10011. All 'rights reserved. This article ,cannot be reproduced for any purpose whatsoever without
! permission of the publisher. A copy of this article is available fr.om the publisher for $15.00.
194
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
Golovin was the initiator of a broad program of research involving storage and confinement of plasma
in open-ended magnetic traps, based on the concept put forth by Academician G. I. Budker. Under Golovin's
guidance, the large experimental machines OGRA-1, OGRA-2, and OGRA-3 were built, work went forward
on ultradeep vacuum in large volumes, and injectors of high-intensity streams of hydrogen ions and atoms
were devised.
The experimental and theoretical research carried out under Golovin's supervision played a major
role in the development of techniques for generating plasma and controlling high-temperature plasma, and
in the understanding of the processes at work, and exerted a substantial influence on the development of
thermonuclear reactor concepts and thermonuclear powder-generating station concepts.
High-output injector projects being worked out in Golovin's division are also of great importance in
the thermonuclear research program.
The administration of the Institute of Atomic Energy and the editorial staff of the periodical Atomnaya
6iergiya warmly greet this fellow-member of the editorial staff, Doctor of Physical and Mathematical
Sciences, Professor Igor' Nikolaevich Golovin, on the occasion of his sixtieth birth, and wish him excellent
health, long years of life, and new creative successes.
19.5
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
ARTICLES
DEVELOPMENT OF FUEL ELEMENTS FOR FAST
POWER REACTORS
I. S. Golovnin, Yu. K. Bibilashvili, UDC 621.039.54
and T. S. Men'shikova
The main stages in the development of sodium-cooled fast power reactors in the USSR are: the suc-
cessful operation of the BR-5 reactor, reconstructed now into the BR-10 reactor; the development and
start-up of the BOR-60 reactor, reaching nominal power in 1971; the completion of the construction of the
BN-350 experimental industrial reactor; the construction of the BN-600 reactor [1-4]. The experimental
data accumulated in the course of this program permit the development and construction of large-scale
industrial installations with fast reactors. Scientists in many countries have estimated that the optimum
electric power per unit lies in the 1000-2000 MW range. Practically all sodium-cooled fast reactor sub-
assemblies were difficult to complete because of technical innovations and lack of adequate experience in
related fields .of technology. This kind of installation requires heavy-duty sodium pumps and heat ex-
changers, steam generators, strong large-scale reactor vessels, turbogenerators, etc.
The development of fuel elements for fast reactors requires serious efforts. Studies of the nuclear
physics, chemical?metallurgical, and technological characteristics of a number of fuel materials and pos-
sible structural materials have sufficed to determine the direction of development of fuel elements for the
core and breeding blanket of sodium-cooled fast reactors for the next 10-15 years. Austenitic stainless
steel is the most suitable material for fuel-element cladding and will be the basic structural material during
the next decade.
Fast reactors use uranium oxide and uranium?plutonium fuel because of its good compatibility with
structural materials and the sodium coolant, its good radiation resistance, and the simplicity of its produc-
tion technology. It is appropriate to note that the development of oxide fuel elements for fast reactors
started in the Soviet Union on the basis of the experimental work on the BR-5 reactor, and has been taken
as basic by all European countries including France, England, and Italy, and at the present time the USA
also. The use of oxide fuel avoids a number of difficulties connected with the production of reliably
operating fuel elements, and accelerates the accumulation of fast reactor operating experience and data on
which the design of large-scale power systems can be based. This facilitates a possible subsequent shift
to carbide, nitride, or carbonitride fuel, and finally to the most alluring ? metallic fuel ? if favorable
scientific solutions are found.
The development of oxide fuel element designs permitting a burnup of up to 1n of the heavy atoms
for linear specific loadings up to 600 W/cm and cladding operating temperature of the order of 700?C is it-
self a difficult problem. The lack of experimental arrangements permitting the production of actual operat-
ing conditions of the fuel elements (irradiation by an integrated flux of more than 1023 fast neutrons/cm2,
dynamics of burnup, etc.) led to a somewhat belated discovery of such phenomena as the iodine?cesium
interaction of the core with the cladding, and the embrittlement and swelling of steel under high radiation
doses. These phenomena have still not been adequately investigated quantitatively and so far there is no
possibility of completely correcting earlier designs. However, there existence has not stopped the develop-
ment of oxide-fueled sodium-cooled reactors started earlier. Studies have enabled us to understand the
processes occurring in cores of oxide fuel elements at high burnups and huge temperature gradients, includ-
ing the mechanical interaction of the core and cladding, and to produce a dynamic model of these processes
serving as a basis for fuel-element calculations.
Translated from Atomnaya Energiya, Vol. 34, No. 3, pp. 147-153, March, 1973. Original article
submitted September 14, 1972.
0 1973 Consultants Bureau, a division of Plenum Publishing Corporation, 227 West 17th Street, New York,
N. Y. 10011. All rights reserved. This article cannot be reproduced for any purpose whatsoever without
permission of the publisher. A copy of this article is available from the publisher for $15.00.
196
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
A-A
047,,gr2
Our basic ideas consist of the following [5-7]:
1. After a burnup of more than 3% of the heavy atoms
oxide fuel elements with loadings of 500-600 W/cm are
0,41-gos practically completely (more than 80-90%) free of gaseous
fission products. Thus swelling is a minimum as compared
with other forms of fuel. The volume of uranium dioxide
and mixed uranium and plutonium oxides increases 1% on
the average for 1% burnup, while for loadings of ? 200-
250 W/cm plutonium dioxide swells 1.5% for 1% burnup.
2. Oxide fuel softens at temperatures above 900?C,
its plasticity increasing sharply with temperature. Radia-
tion intensifies this process. In the compact oxide core of
a thin fuel rod operating at high linear loadings only the
outer 0.15-0.2 mm layer remains rigid and exerts the main
mechanical action on the cladding.
3. The high temperature gradients which arise during
a change in reactor power cause radial cracks in a compact
core. For loadings above ?350 W/cm these cracks are
"healed" during steady-state operation by an evaporation
?condensation mechanism with mass transfer into the
colder part of the core. As a result of radial mass transfer
the initial gap between the cladding and core is rather
rapidly eliminated under operating conditions until the
mechanism of "fragment" swelling is brought into play.
In the cold state the gap is determined by the difference in
thermal expansions of the materials.
4. During operation an oxide core undergoes structural changes leading to the formation of several
characteristic zones: an outer zone with the original structure, an equiaxed grain zone, and a columnar
grain zone. The zone boundaries correspond to the radial temperature distribution determining the radial
variation of mechanical properties of the core material. The structural changes in the core occur as a
result of the formation and migration of mostly large pores inward into the high-temperature zone, forming
a central hole or increasing the size of the existing hole during the initial period of irradiation if the fuel
element was constructed with a central void. The accumulation of solid fragments has a relatively small
effect on structural changes up to 10% burnup.
5. The outer rigid layer of an oxide core must have a uniformly distributed initial porosity to com-
pensate for the swelling of this layer during the accumulation of fission fragments. The mechanism of this
process can be explained by the production and diffusion of vacancies in the microscopic regions of the ther-
mal spikes produced in the slowing down of fission fragments. The minimum value of the initial porosity is
determined by the required burnup. It is assumed that the increase in volume due to the accumulation of
solid fission fragments does not exceed 0.4% for 1% burnup.
6. The mean effective fuel density in a cross section of a fuel element, computed by taking account of
the internal porosity of the pellets, the central hole in the core, and the gaps, must be limited to a value,
depending on the construction, which prevents melting of the inner portions of the core with subsequent
axial mass transfer.
7. The power density in a fuel element is limited to a value which does not cause melting of the central
part of the core during the operating period. In this case the contraction of the central hole in the core to-
ward the end of the operating period as swelling occurs under the restrictive action of the cladding is taken
into account, as is the lowering of the melting point of the dioxide with poisoning by fission products.
8. The swelling of steel in a neutron field significantly changes the pattern of stress and strain in the
fuel element cladding. Estimates based on a design model which assumes that the rate of swelling of the
core is independent of the extent of its mechanical interaction with the cladding shows that the swelling of
steel has a favorable effect on the efficiency of the central fuel elements of an assembly. The jackets of
these fuel elements "escape" from the core, as it were, and the mechanical loadings decrease. Because
A-A
Pitch of helix 100? 5
Fig. 1. Core fuel element design for load-
ing of BN-350 reactor: 1) lower cap; 2)
sleeve; 3) upper cap; 4) cladding; 5) wire.
197
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
? Fig. 2. Core fuel element design for the second load-
ing of the BN-350 reactor: 1) lower cap; 2) gas space;
3) can; 4) lower end shield briquet; 5) core briquet;
6) sleeve; 7) porous plug; 8) upper cap; 9) spacing
wire (tape); 10) cladding.
of the nonuniformity of the temperature around the perimeter of the cladding of a fuel element on the peri-
phery of an assembly further stresses arise as a result of the nonuniform swelling of steel. The magnitude
of these stresses depends on the temperature of the "rosette." Therefore it is desirable to take measures
to decrease the nonuniformity of the temperature around the perimeter of the peripheral fuel elements. In
Soviet reactors displacers [6] are introduced into the peripheral cells of an assembly to accomplish this [6].
The ideas presented above can be carried over completely to the operation of a fuel element with a
vibrocompacted core of powdered dioxide fuel. The only difference is in the initial period of irradiation
during which the powder is sintered into a compact rod with a central hole. For such fuel elements the
initial operation of the reactor at power must follow a special program to ensure the solidification of the
hollow core without melting.
The safety factor of the cladding was calculated by taking account of the thermal and mechanical stres-
ses from the gas pressure and the swelling of the core. The long-term strength and long-term plasticity
were also taken into account, as was the relaxation of stresses [6].
The design of fuel elements for the core of the BN-350 reactor, intended for the first loading, was
developed before the calculational and design methods for oxide fuel elements were in final form. How-
ever, the ability to operate up to 5% burnup was verified by direct experiment in a sodium loop of the
MIR-2 reactor and in the BR-5. The fuel element design is shown in Fig. 1. It consists of a stainless
steel tube 6.1 mm in diameter with a wall thickness of 0.55 mm filled with sleeves of sintered uranium
dioxide forming a core 1060 mm long. The average effective density in a cross section of the fuel element
is 8 g/c m3. The nominal initial diametral gap between the cladding and the core is 0.3 mm. The ends of
the cladding are closed by argon arc welds. There is practically no gas collector. The empty volume in
198
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
TABLE 1. Comparative Characteristics of
Fuel Element Designs
Characteristics
BN-350 fuel element
BN-600 fuel
element
basic
II variant
Diameter, mm
6,1
6,9
6,9
Length of active
1060
230+830
90+660=750
portion, mm
=1060
Maximum cladding
temperature, ?Co-
680
700
710
Maximum fuel tern-
perature,?C*
1800
2500
2500
Maximum heat
loading, W/cm
450
530
530
Maximum burnup,*
5
10
10
Gas pressure at end
of operating per-
iod, atm
140
100
40
Maximumt tangen-
tial strain of clad-
ding,0/0
0,25
1,8
1,6
Safety factor of
cladding (with re-
spect to tension)at
end of operating
period
1,4
1,37(1,05
1,55(1,06*)
*Temperatures indicated for hot spots.
tCalculations performed without taking account of
the effect of reactor radiation on material properties,
*Time to rupture decreased by a factor of 100 in
comparison with the properties of steel without taking
account of the effect of reactor radiation on them.
the fuel element is made up of the central hole of the core,
the gap between the cladding and the core, and the small
space in the upper part of the element (20-25 mm) which
compensates for the thermal expansion, taking account of
the allowance for the height of the core. The spacing of
the fuel elements in an assembly is maintained by a heli-
cally wound wire.
The design pressure of gases in the fuel element at
the end of the operating period is 140 atm, but the safety
factor of the cladding remains above one for a maximum
cladding temperature of about 680?C at the beginning of
the operating period and 650?C at the end. Hexagonal as-
semblies are formed of 169 fuel elements.
In the second loading of the BN-350 reactor it is
proposed to use the more refined core fuel element design
shown in Fig. 2. The diameter of the fuel element here
is increased to 6.9 mm, and 127 of them can be placed in
a hexagonal assembly of the same size, maintaining the
loading of the fissionable isotope. A gas space is provided
in the lower colder part of the fuel element, and the lower
end reflector is combined with the fuel core in a single
jacket. This permits a decrease in the pressure of fission
product gases inside the cladding to a maximum of 100 atm
for 10% burnup of the heavy nuclei. A certain increase in
the average temperature of the core decreases its mechan-
ical action on the cladding during swelling. Since this
was confirmed on experimental samples irradiated in the
SM-2 reactor, the fuel element design developed turned
out to be operable to a burnup of 100,000 MW days/ton of
UO2.
It should be noted that the lower 230 mm of the fuel
core directly adjoining the end reflector (Fig. 2) is made
of solid rather than hollow briquets. The increase in effective fuel density in the cross section of the
fuel element achieved in this way leads to an increase in the surface temperature of the briquets and to a
decrease in the mechanical action of the core on the cladding in this part. Figure 3 shows the longitudinal
distribution of the tangential strain of the fuel element cladding for the second loading. The increase in
effective fuel density in the lower part of the fuel element from 75 to 86% of theoretical decreases the strain
of the cladding toward the end of the operating period from 2.3 to 1.8%, which significantly increases its
operating reserve [7, 8].
An alternative fuel element design provided for the incorporation of both the lower and upper end
reflectors into a single jacket. This design was not successful in the BN-350 reactor, however, because
of a significant increase in the hydraulic resistance of the assembly. This idea has been employed in the
design of a BN-600 fuel element now being developed. One version of this fuel element is shown in Fig. 4.
The fuel element was designed for a 10% burnup of heavy nuclei, and can use a fuel core of both uranium
dioxide and a (UPu)02 mixture. The jacket has a large gas space (800 mm) and a "heating" region of the
fuel core (90 mm) for decreasing the tangential strain of the cladding. Samples of fuel elements close to
the design described are being tested in the BOR-60 reactor at the present time. Table 1 lists the com-
parative characteristics of the fuel element designs described.
The quality of manufacture of the fuel core has an appreciable effect on the efficiency and operating
characteristics of fuel elements. Pelletization is an accepted production process in fuel core manufacture
in the USSR and other countries. Rather efficient automatic presses ensuring low production losses have
been developed [5, 6, 9]. Although the charging material intended for processing by an automatic press
requires more careful preparation to ensure constancy of the bulk density and the duplication of sizes and
properties of the individual pellets, the amount of plasticizer acceptable in it is significantly less than in
199
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
0 250 500 750
Length of active part, mm
1060
Fig. 3. Tangential strain of cladding along the length
of a core fuel element of the BN-350 reactor in which
the effective density along a length of 230 mm from
the lower end of the active part of the fuel element is:
) 75% of theoretical; ) 86% of theoretical.
powders used in other methods of forming oxide cores. This ensures higher quality of the product after
sintering: uniform density, correct geometric shape, smaller deviations of dimensions from nominal. The
latter permits the omission of the grinding process except to correct pellets rejected, for example, as the
result of wear of the pressing device.
A point infavor of the pelletization process is the possibility of using automatic presses with a lubri-
cated pressing device to form "damp" pellets of powder practically without the addition of a plasticizer.
This further improves the accuracy of the pellet manufacture and their quality, and permits an increase in
the fuel charge in the fuel elements. The optimum technology uses a starting powder with a minimum
amount of materials which are eliminated in the sintering process.
Some remarks on the purity of the starting material are in order. In choosing the condition of the
uranium dioxide the developer and user generally start from the possibilities of the supplier but try to use
the purest product. In principle the supplier can produce a product of any degree of purity, and a high degree
of purity may turn out to be economically advantageous to him. An increased contamination of the initial
uranium dioxide, particularly by highly volatile admixtures, can affect the quality of the pellets produced,
the capacity of the fuel with respect to heavy atoms, and the efficiency of the fuel elements. Not all impuri-
ties worsen the working capacity of fuel element cores, however, and the presence of impurities below a
certain level has a negligible effect on the fuel element properties. There has been little research on the
dependence of the technological properties and radiation stability of fuel elements on the purity of the start-
ing material, yet it is one way of reducing the cost of the fuel cycle.
Another aspect of the problem of quality and economy of the production of ceramic cores is the choice
of plasticizer. In our practice the most widely used binders are aqueous solutions of high-molecular alco-
hols. Plasticizers of this type permit the use of simple technological equipment. However, they are not
optimum in at least two respects: first, they require the selection of rather narrow pressure limits in
forming pellets; second, they prolong the sintering process because of the difficulty of eliminating moisture.
Anhydrous plasticizers are more suitable: high-molecular fatty acids, their salts (stearates and
behanates), particularly if there is a problem of obtaining a given uniform initial porosity in the sintered
material [10]. The binding properties of these substances are manifested even for small additions to the
charge, and can improve the quality of the production.
The sintering process is the most important technological operation for obtaining products of compact
uranium dioxide. It can be performed in a vacuum or in various atmospheres. The most common is sinter-
ing in a hydrogeneous atmosphere. This ensures adequate stability of pellet size and properties (density,
stoichiometric composition) and guarantees a low carbon content in the product.
200
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
A-A
Central 0,41004
-56;gf13134
Zg5?a08,
Fig. 4. Alternative design of a fuel element for the
core of the BN-600 reactor: 1) lower Cap; 2) gas
space; 3) can; 4) lower and shield briquet; 5) core
briquet; 6) sleeve; 7) upper end shield briquet; 8)
porous plug; 9) upper cap; 10) spacing wire (tape);
11) cladding.
Uranium dioxide will be used as core fuel in the first loadings of the BN-350 and BN-600 fast power
reactors. After reactors of this type have been mastered the fuel will be mixed oxides (UO2 + 15-20% Pu02).
It is proposed to make fuel elements of this type by pressing pellets.
The technological process of making fuel elements with cores of mixed-oxide fuel involves certain
special features. The first feature has to do with the method of making the original powder. This powder
can be obtained either by mechanical mixing of powdered uranium and plutonium oxides, or by coprecipitat-
ing them from solutions [11]. Choosing one or the other method requires taking account of the necessity of
a uniform distribution of plutonium in the fuel core. A solid solution of plutonium dioxide in uranium dioxide
formed by sintering pellets pressed from mechanically mixed powders may have significant nonuniformities
in the distribution of components in small volumes, which leads to alarge Doppler coefficient of reactivity
of the system. Further technological operations to equalize the distribution of concentrations in such fuel
may incr ease the cost of the process. The coprecipitation process and the subsequent firing lead to the forma-
tion of crystals of a solid solution of oxides, and the distribution of the fissionable component in such pow-
ders is very uniform. It should be kept in mind that the use of coprecipitated mixtures does not require the
complete separation of uranium and plutonium in the reprocessing of spent fuel elements, and this may
reduce the cost of the external fuel cycle. Of course in the initial phase of production the choice of the
method of manufacture of mixed-oxide fuel may be determined by the current possibilities of the manufac-
turing plants in the country, but for a stabilized process of multiple reprocessing of fuel the method of
chemical coprecipitation appears to be preferable.
201
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
A second feature of the process of manufacturing pellets of mixed-oxide fuel is the difference in
affinities of uranium and plutonium for oxygen and hydrogen. This leads to a significant effect of the atmo-
sphere onthe sintering process. In sintering in a reducing atmosphere a two-phase structure of solid solu-
tions is formed in the core with the phase concentrations depending on the sintering regime. A single-phase
structure can be obtained by sintering in an oxidizing atmosphere, but the necessity of such a process to
improve the radiation stability of the fuel must be confirmed experimentally.
A third feature of the production of mixed oxides is the high toxicity of the product. The presently
existing areas for the manufacture of fuel elements with cores containing plutonium dioxide [10] are equipped
with glove boxes and provide for carrying out the process manually. However, for large-scale industrial
production of plutonium fuel the safety requirements may demand significant corrections and necessitate
partial or complete remote control, particularly for multiple reprocessing of fuel. Remote control ordinar-
ily involves an increase in production costs. From our point of view a reasonable mechanization and auto-
mation of technological processes will be advantageous in the large scale production of plutonium fuel and
for sufficiently developed equipment will permit the elimination of hand work except for brief manual opera-
tions to correct faults or replace equipment. This decreases the requirements for high reliability, dura-
bility, and dependability of operation which can be demanded of a remote control system reprocessing ma-
terials with a biologically dangerous radiation level.
LITERATURE CITED
1. A. I. Leipunskii et al., (USSR) SMEA Symposium on the State and Prospects of Construction of Fast
Reactor Power Plants [in Russian], Vol. 1, Obninsk (1967), p. 249.
2. A. I. Leipunskii et al., ibid. ,p. 123.
3. A. I. Leipunskii et al., Atomnaya Energiya, 30, No. 2, 165 (1971).
4. A. I. Leipunskii et al., Atomnaya Energiya, 25, No. 5, 380 (1968).
5. I. S. Golovnin et al., Paper at the Franco-Soviet Symposium on the Development of Fuel Elements for
the BOR-60 Reactor [in Russian], Kadarash (1970).
6. A. I. Leipunskiiet a1., Paper 49/P/460 at the Fourth Geneva Conference [in Russian] (1971).
7. I. S. Golovnin et al., Atomnaya Energiya, 30, No. 2, 216 (1971).
8. R. Klipot and A. Smolders, Powder Metallurgy, 12, 24, 305 (1969).
9. M. Batler et al., Paper M 88/33 at a Symposium on the Use of Plutonium as a Reactor Fuel [in Rus-
sian], Brussels (1967).
10. E. A. Evans et al, Paper P/236 (USA) at the Third Geneva Conference (1964).
11: C. Sory et al., J. Nuclear Mat., 35, 267 (1970).
202
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
THE SYSTEM Mo03 ? UO3
0. A. Ustinov, M. A. Andrianov, UDC 548.736
N. T. Chebotarev, and G. P. Novoselov
An x-ray analysis [1] of the system UO2?UO3?Mo03, performed on specimens obtained by sintering
stoichiometric amounts of uranous?uranic oxide and molybdenum trioxide in vacuum at 750?C, revealed
that the reaction of these oxides leads to the formation of UMo08 and U2Mo08. Sintering of uranous?uranic
oxide with molybdenum trioxide in air at 750?C leads to the formation of uranyl molybdate UO2Mo04 [2].
This compound has a monoclinic structure [3] with the following unit cell constants: a = 7.200 A b = 5.480
A, c = 13.59 A; [3 = 104?36'; the space group is p21/c; the unit cell of this compound contains four formula
units.
We have investigated the reaction of Mo03 and U308 in air in the range from room temperature to
1000?C in order to construct the phase diagram Mo03?UO3. We used x-ray and thermographic analysis for
this purpose. Thermographic analysis was performed in an NTR-64 thermorecorder. The standard was
MgO. Weighed amounts (0.5 g) of the mixtures were placed in quartz crucibles and heated at a rate of 20?C
/min. No reaction of the crucible material with the melt was observed. The temperature was measured by
Pt?Pt/Rh thermocouples to within ?5?C. The specimens for x-ray analysis were obtained by heating
mixtures of the powdered initial oxides for 4 h at 700?C, after which the sintered mixtures were ground and
then reheated under the same conditions. The temperature was measured to within ?10?C by a Chromel
?Alumel thermocouple. The x-ray diffraction patterns of the powders thus obtained were recorded in RKU-
86 cameras in Co- and Cr-radiation.
We used uranous?uranic oxide (13308), obtained by heating uranium dioxide in air at 600?C for 10h,
and molybdenum trioxide (Mo03) of cp grade. The x-ray diffraction patterns of the initial U308 and Mo03
powders exhibit only lines of the corresponding phases.
TABLE 1. Temperature of Thermal Effects
Recorded on Differential Heating Curves of
Mixtures of 1.1308 and Mo03
U308 content of
Mo03 ? U308 mixture
I mole oh (in
wt.% 'terms of
Temperature of effects, ?C
II
III
. IV
0
10
17
20
30
40
45
60
70
80
90
100
0
5,4
9,5
11,4
18,0
25,4
29,5
43,4
54,4
67,2
82,1
100
600
610
610
610
610
600
610
600
600
610
740
740
740
740
740
740
740
800
780
760
760
780
830
860
930
980
980
980
TABLE 2. Phase Composition of Mixtures
of Mo03 and 13208 after Heating to 700?C*
0303 content of Mo03 ?
U308 mixture
wt. To
role oh (in
terms of
02. sr)
Phase composition (from
x-ray analysis data)
10
20
30
40
50
60
70
80
90
" 100
5,4
11,4
18,0
25,4
33,8
43,4
54,4
67,2
82,1
100
*Residence time 8 h.
Mo03
Mo03
M003>> UO2Mo04
Mo03> UO2Mo04
UO2Mo04> Mo08
UO2Mo04+ traces of Mo03
UO2Mo04
UO2MoO4d-U308
U308-1-UO2Mo04
U308
Translated from Atomnay.a Energiya, Vol. 34, No. 3, pp. 155-157, March, 1973. Original article sub-
mitted May 15, 1972.
09 1973 Consultants Bureau, a division of Plenum Publishing Corporation, 227 West 17th Street, New York,
N. Y. 10011. All rights reserved. This article cannot be reproduced for any purpose whatsoever without
permission of the publisher. A copy of this article is available from the publisher for $15.00.
203
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
TABLE 3. Comparison of Interplanar Spacings in the Structures of the Molybdate Investi-
gated by the Authors and UO2Mo04 [3]
Indices Ilk/
Interplanar spacings
Indices Ilk/
Interplanar spacings
authors' data
[3]
authors data
PI
'vie
a, A
'reit
d, A
Ivis.
d, A
'reit
d,A
102
110, 111
012
111
200
202
112
113
004
210
212 .
114
014
202
204
,020
104
121, 120
214
122
w
m
s
m
1 m
1 m
m
1 m
1 v.w.
v.w.
w
v. w.
m
1 ni. la.
5,53
4,30
4,20
3,921
3,471
3,351
3,288
2,927
2,827
2,790
2,753
2,715
2,544
2,453
14
19
100
24
20
18
11
12
31
11
5
4
5
7
2
5
10
3
6
5,51
4,30
4,20
3,915
3,480
3,457
3,367
3,341
3,286
2,933
2,833
2,820 ,
2,801
2,762
2,739
2,721
2,548
2,465
2,455
300
122
512
504
513
123
311
302, 016
214
216
222
312
' 025
206, 116
223
031
-4-02
017
320
130, 313, 131
1 ni. w.
1 m,w,
1 m
1 m
w
1rri.w.
w
w
m
w
1 m
1
} av.(d)
.1
2,308
2,178
2,118
2,030
1,986
1,962
1,912
1,897
1,838
1,821
1,793
1,763
5
7
4
1
10
8
4
4
2
4
4
6
4
4
6
1
5
5
2,321
2,305
2,188
2,172
2,122
2,036
1,992
1,969
1,958
1,909
1,899
1,841
1,823
1,809
1,800
1,777
1,771
1,767
*In this column the following abbreviations are used: w weak, av average, s strong, v. w very weak, and
d diffuse.
1-Juenke and Bartram [3] determined the reflection intensity I from diffractometric data.
RESULTS
When mixtures of U308 and Mo03 are heated in air to 600-1000?C, bright yellow powders or cakes are
formed. The differential heating curves of mixtures of the initial oxides exhibit an exoeffect I at approxi-
mately 600?C in all cases (Table 1). This effect is not observed on the thermograms after reheating. For
sintered mixtures containing up to 60 wt. % U308 we observe a constant endoeffect II (740?C) and a variable
endoeffect III. Effects II and III are also observed when the thermograms of these mixtures are recorded
again. Between the compositions of mixtures corresponding to 60 and 70 wt. % U308 the effect at 740?C
disappears and effect IV appears at 980?C.
X-ray analysis revealed formation of a molybdate (Table 2) with a complex structure in this system.
X-ray analysis showed that this compound is uranyl
molybdate UO2Mo04, the crystal structure of which was
mmo
established by Juenke and Bartram [3]. Table 3 gives the
interplanar spacings in the molybdate structure, calculated
from several leading lines of the x-ray diffraction pattern
obtained in Cr-radiation, together with the interplanar
spacings in the structure of UO2Mo04, taken from [3], for
comparison.
1 WOO
900
800
? 700
UO2M004-FU0,
UO2M004 + L
M003+ UO2 M004
I ? I
POO
BOO
700
0 10 20 JO 40 50 50 70 80 SO UO3,mole%
0 10 ZO JO 40 50 50 70 80 SO UO3,1qt,ob
Fig. 1. Phase diagram of the system Mo0
-UO3.
204
DISCUSSION OF RESULTS
According to the x-ray data and visual observations
on the change in color of the mixtures during heating,
exoeffect I at 600-610?C cdrresponds to the onset of the
reaction of U308 and Mo03. When mixtures containing up
to 65 wt. % U308 were heated to 740?C or above, they partly
or completely melted. Therefore endoeffect II at 740?C is
due to eutectic melting of the mixtures. Endoeffect Ill is
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
observed when the mixtures pass completely into the liquid state, i.e., it characterizes the liquidus line of
the system. Endoeffect IV at a constant temperature of 980?C is due to peritectic decomposition of UO2Mo04,
formation of which was established by x-ray analysis. When UO2Mo04 is heated in vacuum to 1000?C it
decomposes with formation of U308 and Mo03. Since the composition 1J02Mo04 (UO3 ?Mo03) corresponds in
the ternary system U?Mo?O to the cross section UO3?Mo03, not the cross section U308?Mo03, we may
assume that during the reaction of U308 and Mo03 (taken as the initial materials) in air, U308 is oxidized
to UO3. Therefore our data must be regarded as the result of the reaction of Mo03 with UO3.
The phase diagram of the system Mo03?UO3 (Fig. 1) was constructed from the thermographic and
x-ray data. The diagram exhibits a compound UO2Mo04 which melts incongruently at ?980?C; together
with Mo03, this compound gives a eutectic containing ?14.6 wt. % UO3, the melting point being 740?C. After
our experimental work was completed, Serezhkin et al. [4] published their results; these confirmed the
data in [3] on the structure of UO2Mo04 and agreed with our results.
LITERATURE CITED
1. V. K. Trunov et al., Zh. Neorgan. Khimii, 10, No. 11, 2576 (1965).
2. V. K. Trunov et al., Dokl. Akad. Nauk SSSR, 141, No. 1, 114 (1961).
3. E. Juenke and S. Bartram, Acta Crystal., 17, 618 (1964).
4. V. N. Serezhkin et al., Radiokhimiya, 13, No. 4, 659 (1971).
205
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
BUILDUP OF TRANSURANIUM ELEMENTS IN VK-50
REACTOR FUEL
V.
Ya.
Gabeskiriya, V.
S. Belokopytov,
UDC 621.039.524.4-97:
Yu.
V.
and
B.
M.
A.
Novikov, V. G.
Sarychev, G. A.
P. Chetverikov
Polyukhov,
Simakin,
621.311.25:621.039
In order to find the isotopic composition of irradiated fuel of a water-moderated water-cooled boiling
reactor we have experimentally investigated the isotopic composition of irradiated fuel in samples cut from
VK-50 reactor fuel elements. We have sought to determine: a) the isotopic composition of uranium, pluto-
nium, and americium after irradiation; b) the amount of plutonium, americium, and curium isotopes pro-
duced by irradiation; c) the dependence of buildup of, transuranium elements on burnup.
Preparation and Dissolution of Samples
The assembly out of which the fuel element was extracted has been irradiated for 657 effective days
and held up for more than a year. The fuel was 2%-enriched uranium dioxide.
Irradiation did not cause any significant structural changes in fuel. The fact that no temperature
domains (fusing, formation of acicular or equiaxial grains) have been observed indicates that the tempera-
ture of uranium dioxide at the center of the pellet did not exceed 1600?C (Fig. 1). Four samples were cut
at different length of the fuel element (Fig. 2) in order to obtain information about the variation of isotopic
compositionandbuildup of transuranium elements with burnup. The weight of the samples was 0.6 to lg.
The samples were then dissolved in a mixture of concentrated perchloric and nitric acids (heated). The
solution was brought up to 100 ml by adding 8 N nitric acid.
EXPERIMENTAL METHOD
Radiometric Analysis. Without chemically separating the transuranium elements, aliquots were
taken from the investigated solution and made into targets from which the alpha spectra and total alpha
activity were determined. The content of Cm242 and cm244 was calculated from the results of alpha spectro-
metric analysis taking into account the data of absolute alpha count.
The measurements were carried out with the aid of an alpha spectrometer with a silicon semiconduc-
tor detector. The Am241 line (5486 key) resolution of the spectrometer was ?40 keV. The absolute activity
of targets with alpha emitters was measured with a proportional flow counter of 47r geometry. The accuracy
of alpha activity measurements was ?2%. The accuracy of the final results of the determination of CM242
and Cm244 content was ?20%.
The fission product content was determined from gamma activity of Cs137 and CO" isotopes using a
gamma spectrometer with a coaxial Ge(Li) detector having a volume of 21.7 cm3 and an active surface area
of 6.76 cm2.
The Ba137 line (662 keV) resolution of the gamma spectrometer was ?6.5 keV. The accuracy of
determination of Cs137 and Ce144 content in the samples was not less than 15%.
Determination of Isotopic Content of Uranium, Plutonium, and Americium. To find the isotopic com-
position of uranium a small amount of the starting solution was deposited on the evaporator of the three-
strand ion source of the mass spectrometer.
Translated from Atomnaya Energiya, Vol. 34, No. 3, pp. 159-162, March, 1973. Original article
submitted April 20, 1972.
,0 1973 Consultants Bureau, a division of Plenum Publishing Corporation, 227 West 17th Street, New York,
N. Y. 10011. All rights reserved. This article cannot be reproduced for any purpose whatsoever without
permission of the publisher. A copy of this article is available from the publisher. for $15.00.
206
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
Fig. 1. Macro- and microstructure of fuel.
For isotopic analysis of plutonium the latter was
separated out from the solution and decontaminated of fis-
sion elements. Separation and decontamination was done
by extraction using a 0.5M solution of D-2 EHPA [11.
Plutonium was first stabilized in a tetravalent state by a
30% solution of hydrogen peroxide in the presence of 8 N
nitric acid by heating in a water bath. Extraction was
carried out at a ratio Vwater :Vorg = 1: 1 for 2-3 min
followed by phase separation.
The water phase containing americium, curium,
and fission products was poured off, concentrated to two-
three drops by evaporation, and prepared for subsequent
americium analysis.
The remaining organic phase was rinsed with 7 N
hydrochloric acid to remove any trace of iron. For re-
extraction we used a 10% solution of ammonium carbonate.
The reextract was rinsed with decane and calcinated to
remove ammonium salts. The dry residue was dissolved
in 1 ml of 8N nitric acid while heating.
The isotopic composition of uranium, plutonium,
and americium was determined using an MI-1311 and a
modified MI-1305 mass spectrometers. A three-strand ion source was added to the latter to produce the
ion beam. A type SI-01 ion counter was used to record the ion current. A rhenium foil served as an ion-
izer. The analytic technique was similar to that described in [2].
Determination of the Content of Uranium, Plutonium, and Americium Isotopes. The amount of ura-
nium inthe analyzed sample was determined by the Sakhorov method [3], and the content of plutonium and
americium isotopes was determined by the method of isotopic dilution. Reference tracers of Pu242 and AM243
were added to aliquot parts of the analyzed solution. Plutonium and americium were separated chemically
after careful mixing. The technique described in [5] was used in further analysis. The content of plutonium
and americium was determined in two or more parallel analyses.
RESULTS
Isotope Composition of Uranium, Plutonium, and Americium. The content of uranium, plutonium,
and americium in the investigated samples is listed in Table 1, together with the data of radiometric anal-
ysis of the content of CM242, cm244, CS137, and Ce144. As seen in Table 1, heavy plutonium isotopes con-
stitute 39% of sample 2.
The error in determination of uranium by the Sakhorov method was ?5%. The accuracy of determina-
tion of plutonium by the isotopic dilution method was ?3% and that of americium better than ?5%. The
amount of americium in sample No. 4 was estimated from the results of radiometric analysis. The content
of transuranium and fission elements was referred to the
20 time the fuel assembly was unloaded from the reactor.
0
bo
-
6.
a
"8
a 0
1100
5100
1000 1500
Length of element
100 5 0
1100
1900
2000
Fig. 2. Burnup distribution along fuel ele-
ment.
Dependence of Isotopic Composition of Uranium,
Plutonium, and Americium on Burnup. The amount of
fission products of U235, PU233, and P241 was calculated
from the experimental results listed in Table 1. Burnup
X due to U235 fission was calculated from
X-1000 [?(11+71)11z0)]kg/tonU, (1)
where y,!,' and is the relative content of the i-th isotope
before and after irradiation (i being the last digit of the
mass number of the given isotope), and z is the measured
ratio of plutonium to uranium content in the sample.
207
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
0.
,4
r.
PI
208
8
CO ?44 OD CO
Co CO CO CV
CV t-? VID CO
Co CO C) CO
CV .4, CV
CO CV
I I
CoCo
CO
CV c0
-c-r CO- e-1
^ CO CC
^ CO
c;c7c7?
-H+1+1
cn`'D
Co cv-
cv
`lualuoo
turrpTioui
CC ?-.7, CC
CV CO CO
CCIC $
CV tO CO
Content of plutonium isotopes ,%
? CoPCCV
000
d'Oc7c7
+1-H +1-H
000o o)
CCC CO CO
CO; C4.r:O'
PC
PC
CV C-? CV CV
CV
c7c>c7c;
+1+1+1+i
C- C')'4'
CO CD CO
Co 00)
CO CO CO 4:73
Stu 'inaluoo
turquotnaki
t?-?
CO
CC C7)
Content of uranium isotopes,%
CV CV C4 CV
0000
0.0.0.0;
CO 114 CO
? Co CV
CO C) CO Co
cn cn CC
Stu '1u21u03
tunTuvin
CO 00
CO a-I r- co
COC- CO
?oN atcltuvs
-s-t CV CO NI.
104
v . IU
/04
a
a.
101 5 10 15?
Pu2i9
PU24?
w
p11241
?
pu242
20
Burnup, kg/ton u
Fig. 3. Isotopic composition of plutonium as
a function of burnup.
The factor co in (1) is found from
(0=77+ -tocco
(2)
where a ? is the ratio of the effective radiative capture cross section to the
fission cross section of the i-th isotope.
The values of a 9 and al for PU239 and Pu241 isotopes were adopted from
[61.
The contribution of plutonium isotopes in total burnup Y was calculated
from
Y? z(I000? X)(co ?1)
kg /tonU.
1+ va
(3)
The calculated results of burnup in the samples and data on the content of
plutonium, americium, and curium per ton of starting uranium are listed in
Table 2 and represented by the curves in Figs. 2-4.
The error in the determination of U235 burnup is due first of all to the
error in isotopic analysis and decreases with increasing burnup. The error
in burnup in samples with low and high burnup was *8% and ?3% respectively.
The contribution of plutonium into total burnup was determined to
within ?(12-15)% since a9 and a1 are known to within ?(5-10)% and the error
in z is ?6%.
0-1
?
a 10-2
.4
0 f0-.7
/0'
1 5 10 15
Burnup, kg/ton U
Fig. 4. Content of uranium, plutonium, amer-
icium, and curium isotopes as a function of
burnup.
u 2X5
Plia-'---4----?
9--
i
,?241i
................?
' "
p11242
i?
r
f Am243
iem244
i
20
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
TABLE 2. Content of Uranium, Plutonium, Americium,
and Curium Isotopes (kg/ton U)
Isotopes
Sample No.
2
3
4
U238
8,34+0,10
6,16+0,10
9,14?0,19
16,73?0,20
U238"
1,75+0,09
2,12+0,11
1,76+0,09
0,68+0,03
U238
971,70?0,97
966,26?0,97
974,15+0,97
979,05+0,98
P11238
3,02+0,18
3,72?0,22
2,34+0,14
0,892?0,054
pu240
1,02?0,06
1,62?0,10
0,70?0,04
0,0657?0,0039
pu241
pu242
0,31?0,02
0,07+0,01
0,60?0,04
0,21+0,01
0,26?0,02
0,06?0,01
0,0087?0,0005,
0 we have
N, z. a.
(z ? nza)k E n! L dzn
,n=1 m=1 n=0
and analogously for m < 0
(z ma)]z=
_ 2 2 znni (k+ n ?1)! 1
(k? 1)! (ma)k+n
m=1 n=0 '
00.3 CO
*Ci Zn r dn zn (k+ n ?1)1
- -
(z ma)h r! E E T (z ma)-1 = ? n
n! ) (k ?1)1 (ma)h+ ?
m=1 nz=1 n=0 m=4. n=0
(12)
211
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
Fig. 1. Homogeneous regular lattice.
Using Eqs. (10) to (12) we write Eq. (2) in the form
00 OD
n
z = 1401n nalzi +Re [E +Ao (-1)n 22n-1B
n1 n (2n)!
=" n=i
co
X (az
?)2n + (a)n Ahdhk+ Bhz.].
a a
n=.0 h=1 n=0
Here Ao, Bhp Bo, and dnk are real numbers, B1 is an imag-
inary number, and the Ak are complex numbers,
00
=
[( ?1)n (a )fl (k n ? 1)1 1
dot
m1I (k ?1)! (ma)kl-n '
(13)
(14)
It follows from Eq. (14) that dnk 0 only if n and k have the
same parity.
We write the solution (13) in the form
. .
Itr 1 (-1)n 2271-1-B, i 3tr \ 2n
(1) (r, (p) = Ao ln ?a + 2 (cos np ? Re An + sin n(p?Im An) ? + An 2
7. - n (20 \ a 1
n=i n=--1
x cos 2ncp + 2: (..'' )n [cos rap E dm, Re Ak ?sin /up E dhh im Ad + 2 doh Re Ak+ Bo? Im Bi ? r sin (p.
a
n=1 k=-1 k=1 k=1
Then, using Eq. (9), we obtain the equations
Ao (In 3:13
a
Im /3161n (yn2q, ?p) = n7nXtr + 1)
CO
YoXtr E
P? /
14=2, 4, .,.
00
in=1
2gh
?Re AA+ 130=0;
ah
? h' ? n=0;
2m
00
Im An + (1 nYnA,t nn
gk+n
k=1, 3, 5, ...
Re ilk =0,
(k+n-1)!
nu 14,
(15)
where 15 in is the Kronecker delta. In Eqs. (17) and (18) n and k are odd. In the following two equations n
and k are even;
00
(1 TnX?trn \ ROA, nynXtr ) (k n ?1)! .2gh+
pflRe Ak+ A013;,,
pj
/ Pn '
k=2, 4, ... (k? n)! n! an+nm
x (? 1)-r 2n nYnAir) 0;
n(n)! P
Ak = 0 .
(19)
(20)
The system of Eqs. (7), (8), and (16) to (20) give the solution of the problem in closed form for an
arbitrary approximation. One must remark that Eqs. (17) and (19) contain either only ImAk, or only Re Ak,
of the same parity. This makes the solution of the system of equations much easier. After truncation
(that is, in the n-th approximation), the system contains 2n + 5 equations with 2n + 3 unknowns. After elim-
ination of these unknowns, two boundary conditions are obtained instead of the two usual equations of con-
tinuity. We look at different approximations.
Zeroth Approximation-: Re Ak = Im Ak = 0; k = n = 1, 2,... co. One can easily obtain from the system
of equations the equations:
212
cp+o ?43-o;
(10+o deLo rn
dx = dx = ay') +0 -o);
v o =
2np
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
(21)
2,0
4,0
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
The system (21) coincides with the system obtained in {11.
The approximation we are considering is valid when the absorbers
are at large distances from each other (a/d >> 1).
First Approximation. Re Ak = Im Ak = 0; k = n = 2, 3,..., co.
The use of some simple transformations results in:
15
25 3 35 4 4'5p
Fig. 2. Dependence of 'y0 and yi
on the relative radius p.
c10-32 0(0+0+0-0;
dx dx avo
(RDA -o .
+o ? &0 = 17;" k az dx I
Vi ? - )?
Tt&tr 6 2g2;
P P
1?
Yiktr
oo
xi 1 112
g2 LI m2 = 6
m=1
Second Approximation. Re Ak = Im Ak = 0; k = n = 3, 4,
..., co. In an analogous fashion we write
Tta d0+0 da:Lo
(I)" CD-0 =
dc1)+0 dcb_o
dx dx a (vo ? A) (1113+0 (1)-0);
2g2 1?
A=
\ 4 g
P 2Y2kt r
) 42;tr
(22)
(23)
7E4
g4.-- 2j ?.=) ?
Trt=
From these results it is easy to induce a general law: in each succeeding approximation one of the
boundary conditions remains unchanged (this is a result of the separation of the equations into equations
with coefficients with even indices and with odd indices), while the second boundary condition contains a
correction of order (p/a)4 relative to the preceeding approximation. This guarantees the fast convergence
of the approximation.
We now look at a lattice in which the absorbers are located in an absorbing medium and are close to
each other. In this case the thermal neutron flux is described by the equation
? V20 (r) y,2 (r) = Lq2z ?
(24)
The solution of Eq. (24) for the geometry of Fig. 1 (not as yet assuming that a/L and p/a are small)
must be sought in the form
tt (r) = oreg(r) + 2 .An ebum Kn(Ir ?Lma ) .
The regular solution corresponding to Eq. (24) must have the form
X X
reg(") reg
(X) = C .071: C _e-17
Consequently the asymptotic flux averaged over a lattice spacing is determined according to Eq. (25) as
(25)
(26)
a
Cri (X) = (I) reg(X) E An E? eing)moivcn (17 x2+ (7 ma)2 ) dy. (27)
11=s? oo in= ? c0
One can obtain, after lengthy calculation from Eq. (27),
EC?) -1"1+k
IX! I (-1)k (1;?1)r(1.11-1 k)21n142-1-k 'xi ) 2 K
a 71.- 1
?1) (x) = Oreg (x) E A
2
n=2v=- co h=0
213
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
25
20
15
10
0
/
/
/----
1
1
-
.---
...-----
,
_I
t-10
-.:8
t=7
--------
I
-----
?..../4"9
c=.16
E4
g=4
E-3
?
r.
05 1 1,5 2 25 3 3,5 4 4 5 jo
co In' - 1
E ( 2 )
1,2.3)
1,2,3)
12,3)
1,2,3)
0
12,3)
0)
1,2,31
0
0,1,2,3)
Fig. 3. The dependence of the effective extra-
polation distance 6 on p and for the reactor
IGR in the case yo (p) and in the case yi = yi
? (p) (the order of the approximation is presented
in parenthesis).
? 2 An x sign (n)E (_i) h (2)en+li
n=2v+1=-- co k=0
Inl ._,,
ml
---r-
2
I In] _k\ 22-h I lx11
K 04 04.1) , (28)
\ 2 / \ L I --1-14 \ 1,
2
where F is the gamma function of Euler, E is the integral part, (In1/2k) is the number of combinations of
ml things taken 2k at a time. After using representations of these functions, and of the McDonald function
K, and after summing, one can obtain the relatively simple expression
1.1 lxi
ixi
L A4e L
TEL '-- + ? ? ? ?
111:1 (X) "?= (1)reg(X) AO- L + 2 Ai e 2 A ?2 A2e 3 L + 2 51! a
All harmonics enter into Eq. (29), in contrast to Eq. (6), which contains only Ao and A1 (for the case
L = 00). However, as calculation demonstrates, An decreases as the power (a/L)n, and in the limit L 00
only the two terms ?A0 and ?A1 remain in the equation, and the equation itself takes the form of Eq. (6).
On the basis of Eq.' (29) and their derivatives one can easily find four equations for the fluxes and their
derivatives on the two different sides of the band.
In order to obtain the boundary conditions on the surface of the absorber, it is necessary to represent
the solution (25) in such a way as to display its dependence on the azimuthal angle measured, for example,
from the central absorber. We make use of the theory of complex Bessel functions for the right and left
half-planes and rewrite expression (25) as
(29)
(1) (r) = Oreg(r) E An EIk (-i-) [( ( ?1)h] 2 if_n+7,('?lei-po+ 2 AK (.f-) (30)
n=-0o k=- co m=1
It follows from Eq. (30) that the middle term differs from zero only when the numbers n and k are of
the same parity. An analogous situation took place for the expansion of the function 2 (1/(z? ma)k) in
711= ?Oa
the case L = cc (the coefficients dnk). As one would expect, in the limit L ?00 the quantities
[(-1)-+(-1)hi E K-n+1, (7)
m=1
go over (with an accuracy of up to an insignificant multiplier) to the coefficients dnk. In order to use the
boundary condition (9) it is necessary to make a transformation of the regular solution (26). We substitute
x = r sin coo in this solution and use the generating function for Bessel functions [2]
214
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
Declassified and Approved For Release 2013/09/15: CIA-RDP10-02196R000400010003-7
TABLE 1. Dependence of the Effective Ex-
trapolation Distance 6 on the Absorber Radius
p for = 1 and for Different Choices of yn
and Different Approximations
Vo
= Vo (P)
Vi (p) V2 (P)i
= V2 (P) = Vi (P)
as in [4]
Vi (p) = V2 (P)
= V2 (p) = Vo (P)
as in [5]
ii
Nri g
approximation
approximation
II
III
II
III
0,01
8,860
8,857
8,857
8,857
8,869
8,868
8,868
0,05
8,804
8,791
8,790
8,790
8,837
8,835
8,835
0,10
8,747
8,711
8,708
8,709
8,795
8,792
8,792
0,15
8,691
8,603
8,599
8,601
8,745
8,742
8,742
0,20
8,627
8,611
8,605
8,624
8,683
8,678
8,678
0,30
8,523
8,696
8,689
8,693
8,577
8,570
8,570
0,40
8,440
8,525
8,515
8,516
8,489
8,479
8,479
0,50
8,361
8,416
8,404
8,404
8,401
8,390
8,389
0,60
8,284
8,321
8,307
8,307
8,315
8,301
8,301
0,70
8,210
8,232
8,216
8,216
8,231
8,215
8,215
0,80
8,138
8,146
8,129
8,129
8,147
8,130
8,130
0,90
8,067
8,062
8,044
8,044
8,065
8,047
8,047
1,00
7,998
7,981
7,962
7,962
7,985
7,966
7,966
1,25
7,822
7,772
7,755
7,755
7,777
7,759
7,759
1,50
7,651
7,565
7,559
7,559
7,571
7,560
7,560
1,75
7,483
7,360
7,393
7,393
7,366
7,380
7,380
2,00
7,315
7,152
7,425
7,425
7,158
7,277
7,277
2,50
6,980
6,728
6,484
6,485
6,735
6,424
6,424
3,75
6,141
5,611
5,402
5,406
5,620
5,404
5,404
5,00
5,296
4,400
4,139
5,149
4,413
4,147
4,147
'xP
E ik(z) tk
We put t = eicPb, and then after some little transformations
we obtain:
result
00
(1)reg(r) =E(4,-)
kr=
c_ - irk) ow? cosho]
k Pkik (-iT) (7o8h5 eihw?;
(31)
(32)
After substituting Eq. (31) into (30) we obtain the
D (,)== I {Dh/h +cook?
11=Co-
?
2 /h (-E-) [Or+ -1)h] An
n=---co
x ic_ii+? (-7) +At& (i_)}
(33)
By substituting Eq. (33) into the boundary conditions (9) we obtain equations for each of the 2n1-1- 1 harmonics:
00
Firh_i ( P ) ik L (--)] {Dh+fl -00
ic-ir+c-i)hiii.
L
rna 2L p 2L
X K -n-Fk (-27)} Ah [Kk-I (77) + 1,04, K z) K GT) yhxt, C 06 k0 = 0;
-n