SOVIET ATOMIC ENERGY VOL. 40, NO. 5

• Document Type: CREST
• Collection: General CIA Records
• Document Number (FOIA) /ESDN (CREST): CIA-RDP10-02196R000700070005-6
• Release Decision: RIFPUB
• Original Classification: K
• Document Page Count: 100
• Document Creation Date: December 23, 2016
• Document Release Date: April 10, 2013
• Sequence Number: 5
• Publication Date: November 1, 1976
• Content Type: REPORT

_ .. /,
.
. Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6 I
.._- --, -,,
,. v ? -
l
\
\
/ ,
? Rusian Original Vol..,40, No: 5,_11/laci, 1976 ,-
November, 1.976
. / ? ,
? < l?C
'
r
-r ,_. ,1
' . . ?
. -- ,
, . .
----\-- - ' ) AT010Fyty.1 ,3HEP1.14-fl - /
, ...?
IAT.0-MNAYNA ,NERGIYAY,- - '--,
,. ? f . ? , ? 4- - ,---??`'N
.? I ???, ?,(
.1 '
, -
/- tr-)
r
'IN
-\, I
?
?-? ? r
r ? .
'
_
)
italip':ULTANTS-1OREAli NEW YORK
N
'
? )
,
,? ,
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
: A
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
ATOMIC
,EKRGY-
SovietAtomk Energy is abstiacted or in7-
dexed in,Applied Mechanics Reviews,.Chem-
'kal Abstracts, Engineering Index) MISPEdL
Physics Abstracts and Electrical and Elec-
tronics Abstracti, Current Contents,- and
Nuclear Science Abstracts.
Soviet Atomic- E'rie'rgy is a cover-to-cover trans,latiOn of Atormiaya
Eriergiya, a publication of the Academy of Sciences of the USSR. \
,
Akagreement Witkllie Copyright/Agency of the USSR (VA),NP)
makes availablerboth? advance copies of the Russian journal and
,original glossy photographs and artwork. This serves`tO decrease
the necessary time lad between publication of the original and ,-
publication of the translation and heips,to improve The quality,.
of the latter. The ^translation 'began with the first isstle cot the ?
Russian journal. A
Editorial Beard of Atomnpya?Ellergiya:
Editor: M. D.,Milli,bn-shchikoy
/ ? Deputy Director '
" 11V: Kurchat6v Insfitute"of Atomic Energy
AtaclemY of Sciences Of the USSR
Moidow, USSR
1
,
Associate Editor (N. A. Vlasor
\
? A: A. Bochvar
N. A: bollezhar
V. S.' Fursov
GoleVin-
,
V. F. Kalinin
, . A. K: Krasip
'
. V. V.MetyeeV
4. _GyMeshaherifakov
I
? B.
Shevchenko
Smir\nov
A; P. Zefirov j
?
'Copyright ?1976 Plenum Publishing Corporation, 227 West 17tk Street, Neciv York,
'N.Y. 10011. All rights reterved. No"article contained herein ?nay be reproduced, -/
stbred in ELt:etrimial system, or transinitted, in any form, or by-any means, 'electronic,
mechanicalr photocOpying, Microfilming, recording orlOtherwise,..iivithoUI written
permission of the,publisher.
'
,
Consultanti:-Bureau jburnals' appear about six Months after the publication of)the
'original /Russian issue.- For biblioglaphic taLuracy, the, English, issue published1by
?
Consu!tants Bureu carries the same 'number and date'as the original Russtian .from
which it was tranila4d,. For example, a Rusiian issue published in 'December will
Appear in ,a_FonsUltants Bureau English translation about the following June, but the.
translation' issue will the Decernbr date. When ordering an'Y volume Or partitu-
lar/-issue .0f-a Consultants Bureau/journal, please-specify the date arid, Where a`ppli-,
Cable, 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 _ _ . ?
$'107.50 'Per vplurne (6 issues),
2 volumes per year ?
Prices somewilat higher outside the United States:
/-\ ^ Single I,ssue:r $50
Single Article:
. ?
CONSULTANTS - BUREAU, NEIN'YORK AND, LONDON;,
227/Weill 7th Street
New York; New YorK10,01,1 /
-
,
' Published Monthly. ,Second-class postage paid at Jamaica, New York,1 1431.-
, -J ?
_
'
Declassified and Approved\ For Release 2013/04/10 : CIA-RDP10-02196R000700070005-6
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
SOVIET ATOMIC ENERGY
A translation of Atomnaya Energiya
November, 1976
Volume 40, Number 5
May, 1976
From the Proceedings of the 25th Congress of the Communist Party of the Soviet
CONTENTS
Engl./Russ.
Union
445
363
20 Years of the Journal "Atomnaya Energiya"
448
367
ARTICLES
Steam-Superheating Fuel Elements of the Reactors in the I. V. Kurchatov
Beloyarsk Nuclear Power Station ?A. G. Samoilov, A. V. Pozdnyakova,
and V. S. Volkov
451
371
Some Physical Investigations in BFS-1 Fast Critical Assemblies
?V. A. Dulin, Yu. A. Kazanskii, V. F. Mamontov, and G. I. Sidorov
457
377
Some Results of Postreactor Testing of Six-Element Thermionic Units Operating
for 2670 h ? G. A. Batyrbekov, E. S. Bekmukhambetov, V. I. Berzhatyi,
S. E. Ermatov, Sh. Sh. Ibragimov, V. P. Kirienko, B. S. Kurmangaliev,
M. V. Mellnikov, V. V. Sinyavskii, Yu. A. Sobolev, and Yu. I. Sukhov
462
382
Effect of Heterogeneity on the Measurement of Integral Parameters
in Subcritical Systems ? L. N. Yurova, A. V. Bushuev, V. I. Naumov,
V. M. Duvanov, N. N. Khrennikov, and V. N. Zubarev
465
384
Investigation of the Liberation of Helium from Construction Materials
during Their Heating ?D. M. Skorov, N. P. Agapova, A. I. Dashkovskii,
Yu. N. Sokurskii, A. G. Zaluzhnyi, 0. M. Storozhuk, V. D. Onufriev,
and I. N. Afrikanov
468
387
The Evolution of Gas from Uranium Dioxide B. V. Samsonov, Yu. G. Spridonov,
V. Sh. Sulaberidze, and V. A. Tsykanov
472
390
Sputtering of Thin Films of Uranous?Uranic Oxide under the Influence of Fission
Fragments at Low Irradiation Doses ? V. A. Bessonov, G. P. Ivanov,
N. A. Grinevich, and E. A. Borisov
477
395
REVIEWS
Radiation Defects in Graphite ? T. N. Shurshakova, Yu. S. Virgil'ev,
and I. P. Kalyagina
481
399
DEPOSITED PAPERS
Asymptotic Solution of a y-Ray Transport Equation ? L. D. Pleshakov
493
411
LETTERS TO THE EDITOR
Correction for Time of Fall in Negative-Reactivity Determination by Rod-Drop
?0. A. Elovskii
500
418
An Investigation of the Parameters of a Critical Assembly ? E. Ya. Tomsons,
V. V. Bute, V. V. Gavar, A. S. Dindun, U. A. Kruze, and E. Ya. Platatsis
503
420
Ampule Devices in the VVR-M Reactor for Irradiating Carbon-Based
Materials ? G. Ya. Vasiltev, Yu. S. Virgiltev, V. G. Makarchenko,
and Yu. P. Semenov
506
423
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
Neutron-Spectrum Structure near a Resonant Absorption Line ? A. I. Dod',
CONTENTS
(continued)
Engl./Russ.
I. M. Kisil', and I. P. Markelov
509
425
Calorimeter Measurement of the Heat Released by Irradiated Nuclear Fuel
?A. P. Kirillovich, P. S. Gordienko, and V. P. Buntushkin
511
427
Structural Changes in Irradiated Dysprosium Titanate ? V. M. Kosenkov,
T. M. Guseva, S. A. Alekseeva, and V. K. Nevorotin
513
428
Separation of Isotopic Mixtures of Hydrogen in the Hydrogen?Palladium System
B. M. Andreev, A. S. Polevoi, and 0. V. Petrenik
516
431
Magnetic Susceptibilities of Beryllides ? V. P. Gladkov, V. I. Petrov, A. V. Svetlov,
D. M. Skorov, and V. I. Tenishev
519
433
CONFERENCES AND MEETINGS
Radiochemistry and Nuclear Technology on the 11th Mendeleev Congress
of General and Applied Chemistry ?E. V. Renard
521
435
Use of Radioisotope Techniques and Instruments in Machine Construction
?M. L. Golldin
525
438
The Annual Conference of the Plasma Physics Division of the American Physical
Society ? A. A. Kalmykov
526
439
IN INSTITUTES AND DESIGN OFFICES
Acceleration of 48Ca Ions and New Possibilities of Synthetizing Superheavy Elements
?A. A. Pleve
528
440
The Operator-Stereotelevision-Manipulator System in Nuclear Power Generation
? Yu. A. Gerasimov, V. P. Ivanov, and D. P. Malyuzhonok
529
441
EXHIBITIONS
New Exhibits in the Hall "Atomic Energy" the Exhibition of Achievements
of the National Economy of the USSR ? P. A. Sokolov
532
443
BOOK REVIEWS
A. A. Glazkov, I. F. Malyshev, and G. L. Saksaganskii. Vacuum Systems
for Electrophysical Apparatus ? Reviewed by A. P. Kukushkin and L. N. Rozanov.
534
445
V. I. Subbotin, M. Kho Ibragimov, P. A. Ushakov, V. P. Bobkov, A. V. Zhukov,
and Yu. S. Yur'ev. Hydrodynamics and Heat Exchange in Atomic-Power
Installations ? Reviewed by V. N. Smolin
535
445
B. V. Lysikov, V. K. Prozorov, V. V. Vasil'ev, D. N. Popov, L. F. Gromov,
and Yu. V. Rybakov. Temperature Measurements in Nuclear Reactors
? Reviewed by I. S. Kochenev
537
446
The Russian press date (podpisano k pechati) of this issue was 4/22/1976.
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 2013/04/10: CIA-RDP10-02196R000700070005-6
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
FROM THE PROCEEDINGS OF THE 25TH CONGRESS
OF THE COMMUNIST PARTY OF THE SOVIET UNION
With regard to energy technology, during the past five years we have begun the creation of technology
for atomic power stations. In years ahead atomic engineering will develop at an accelerated rate. Along with
reactors of 103 MW, we will exploit the integral equipping of atomic energy units with 1.5 ? 103 MW reactors.
The energy-producing industry will turn out turbines and generators of 500, 800, and 1000-1200 MW, and
steam-gas equipment of up to 250 MW.
From the address of the Chairman of the Council of Ministers of the USSR,
Comrade A. N. Kosygin, "The Main Directions of Development of the Na-
tional Economy of the USSR in 1976-1980."
We will ensure 1340-1380.103 kWh of electricity generated during 1980. Power-station capacity of the
order of 67-70.106 kW will be commissioned, including 13-15.106 kW at atomic power stations. We will con-
tinue to construct 4-6- 106-kW thermal power stations with energy units 500 and 800.103 kW each along with
atomic power stations with reactors 1-1.5. 106 kW each.
We envisage the rapid development of atomic energy in the European part of the USSR. We will acceler-
ate the construction and commissioning of fast-neutron reactors. We will instigate preparatory work in order
to use atomic energy for district heating.
We will organize mass production for atomic power stations of thermal-neutron reactors and associated
turbosets each of at least 1 .106 kW. We will exploit integral equipping for thermal-neutron units of capacities
up to 1.5 .106 kW.
In the European part of the RSFSR and in the Urals we will ensure the following.
The consolidation of electrical-energy resources through the construction and expansion of atomic and
thermal power stations. We will raise the capacity of the Leningrad Atomic Power Station to 4.106 kW, of the
Kostroma State Regional Power Station to 3.6.106 kW, and of the Reftinskii State Regional Power Station to
3.3.106 kW. We will commission into operation the Novovoronezh, Smolensk, and Kursk Atomic Power Sta-
tions, the Ryazan and Stavropol' State Regional Power Stations, and other power stations. We will expand the
construction of the perm' Thermal Power Station, and start the construction of two high-capacity atomic power
stations and Irganaisk State Power Station.
In the Ukrainian SSR we will ensure the commissioning into operation of the Chernobyl', Roventki, and
South Ukraine Atomic Power Stations and begin construction of two new atomic power stations.
In the Lithuanian SSR we will expand the construction of the Ignalina Atomic Power Station with reactors
1.5.106 kW each.
From the document "The Main Directions of Development of the National
Economy of the USSR in 1976-1980."
The large program of atomic-power-station construction will be fulfilled in regard to Soviet plans and
aid from the Soviet Union to countries that are members of the Council for Mutual Economic Aid, including the
construction of a large atomic power station in geographically remote, but ideologically close, revolutionary
Cuba. (Applause).
In the 10th Five-Year Plan we require the commissioning of at least 70 ? 106 kW of new generating capac-
ity, including at least 15.106 kW from atomic power stations, to ensure the output in 1980 of 1400.109 kWh.
It should be noted that recently it has been necessary, in the electrical and thermal energy industry, to corn-
Translated from Atomnaya Energiya, Vol. 40, No. 5, pp. 363-366, May, 1976.
This material is protected by copyright registered in the name of Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part
of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying,
microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $7.50.
445
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
mission generating capacity ahead of schedule; this reduces the reliability of the energy supply in the national
economy.
The 10th Five-Year Plan should serve as a beginning for further development of energy resources on a
qualitatively new technical basis. We will begin the creation of large atomic power stations with a sustained
capacity of 4-8.106 kW with thermal-neutron reactors of 1.106 kW, 1.5.106 kW, and 2.4.106 kW, with turbines
for these reactors each of 750.103, 1.106, and 1.2 .106 kW. In this period we will make the transition to fast-
neutron reactors and continue scientific and industrial development directed to the subsequent creation of ther-
monuclear power stations. (Applause).
Froin the speech of the Minister for Energy and Electrification of the USSR,
Comrade P. S. Neporozhnii.
Many specialists, engineers, and industrial workers sometimes wrongly assess the value of fundamental
research. Sometimes they have said: Fundamental science is science for academics, but applied science is
science for everyoneelse.
From a plethora of examples, I will cite only one that demonstrates the contribution of fundamental sci-
ence. In 30 years at the Leningrad Physical and Technical Institute, I. V. Kurchatov and other scientists insti-
gated studies of the physics of the atomic nucleus. A number of other institutes of the Academy began the de-
velopment of these studies, which seemingly had no relevance to practical matters. Academician N. N. Semenov,
in a completely different field, studying combustion and explosion phenomena, elucidated the mechanism of prop-
agation of chemical reactions that were termed autocatalytic, i.e., those that proceed with transfer of the re-
action from one atom to another.
By comparing all this and assessing new scientific data, I. V. Kurchatov showed as early as 1940 the ne-
cessity for extending work on atomic technology in our country, but the war began. At the height of the war we
discovered that not only fascist Germany but also our own allies were clandestinely carrying out intensive work
on the creation of a nuclear weapon; and then in 1943 we began this research. Many institutes of the Academy
of Sciences, the biggest industrial organizations, and engineers were drawn to this type of work. Specialized
branches of scientific and industrial organizations were set up. Unimpeded contact and the concerted work of
Scientists of various specializations, of industrial workers, and constant supervision and assistance by the Cen-
tral Committee of the Party led within only five years to the solution of a problem of immense complexity.
Our country was saved from the nuclear threat. (Applause). This was fundamental work, but its practical re-
sults were needed for the very survival of our country.
And what is to be the future destiny of this fundamental line of research? Comrade delegates! You know
that on the basis of the resolutions of the 24th Congress of the Communist Party of the Soviet Union we have
developed atomic energy technology and in the Ninth Five-Year Plan we commissioned two more units at the
Novovoronezh Atomic Power Station, viz., the Kol'sk and Leningrad Atomic Power Stations, and assisted in the
construction of atomic power stations in the German Democratic Republic, Bulgaria, etc. The Leningrad Atom-
ic Power Station consists of units each of 1 .106- kW capacity, i.e., they are among the most powerful reac-
tors in the world. We have constructed a new, powerful atomic icebreaker, the "Arktika," and launched the
icebreaker "Sibir'." In the current five-year plan we intend to introduce atomic power stations of 12-15.106
kW with units each of at least 1 ? 106 kW.
Atomic technology is used today in thousands of plants to test manufactured goods by various methods;
during geological exploration for analyses of commercial minerals; and is widely used in medicine for the
diagnosis and treatment of a number of diseases...
Biology and agriculture use radiation to accelerate the selection of microorganisms and plants. For ex-
ample, the Institute of Cytology and Genetics of the Siberian Branch of the Academy of Sciences in conjunction
with the Institute of Husbandry of the V.I. Lenin All-Union Academy of Agricultural Sciences created by means
of radiation mutation a new variety of spring corn, "Novosibirskaya-67." The variety is zoned, and by 1978 it
is proposed to sow it over 2.106 ha of land; this will give an additional 600-106 kg of grain from the same land
since its cropping capacity is 1000-1500 kg above that attained up to this time.
Scientists continue to work successfully so that thermonuclear energy will not be used as bombs but in
controlled form for the production of energy. I am pleased to be able to inform the Congress that we have made
great advances here. (Applause).
446
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
We have the Tokamak-10 which Grigorii Vasil'evich Romanov neglected to mention although he was greatly
involved in its construction at the Leningrad plants. With this installation at the Institute of Atomic Energy we
already have an advanced thermonuclear reaction in the laboratory. (Applause).
Comrades! In the area of thermonuclear energy technology we have made good progress not only in the
laboratories of the Institute of Atomic Energy but also in Leningrad, at the Physics Institute of the Academy
of Sciences. We now are confident that these enormous thermonuclear resources will be devoted to the service
of the people. (Applause).
The generation of energy in lower forms, exploration for and production of natural energy resources for
this, and a system of converting the energy into a lower form for consumption demand more than 50% of the
country's total budget. The structure of the thermal-energy balance must gradually change. This will also be
affected by the need for economy in the use of petroleum and gas so that they are used more efficiently, the
broadening of the range of uses of atomic energy, and the inclusion in energy production by the close of the
century of thermonuclear sources, magnetohydrodynamic generators, and new methods of energy transmission,
as well as new trends in energy use. Enlargement of units at power stations to a few million kilowatts and the
needs of energy transmission and of thermonuclear energy technology require the use of superconductors, etc.,
in electrical engineering. The involvement of many areas of science is needed for this.
The developments will also be long-term, e.g., the transition from the laboratory carrying out the thermo-
nuclear reaction, which we visited today, to an operationally profitable power station will take place, just as in
atomic energy technology in the last 15 years, and thermonuclear power stations will probably have capacities
of not less than 10.106 kW, thus requiring study of the problems of distribution to consumers and transmission
of the energy in the optimum form.
Plainly we should not equate this long-term program with the concrete plan for the development of science
in the five-year plan ? it must contain long-term technicoeconomic forecasts. This state program has scien-
tific, technical, political, and economic aspects and must be constantly adjusted and refined. Economists, geo-
logists, energy technologists, atomic technologists, and specialists in systems analysis in the Academy of Scien-
ces are united in the solution of this problem.
From the speech of the President of the Academy of Sciences, Academician
A. P. Aleksandrov.
447
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
20 YEARS OF THE JOURNAL "ATOMNAYA E NERGIYA"
The first issue of the journal appeared in May 1956. The world's first atomic power station at Obninsk
was opened in its second year. Our physicists and engineers had already delivered lectures at the Sessions of
the Academy of Sciences of the USSR and at the 1st Geneva Conference. Atomic energy began to be used for
peaceful purposes and large reserves of information had accumulated at that time. To accelerate scientific
progress this information would have to be made available to a wide circle of researchers and engineers.
This task was assigned to our journal.
Naturally in the first issue of the journal an article was published on the world's first atomic power
station [1,2]. Its history and significance and the subsequent development of Soviet atomic energy technology
are not difficult to follow in the pages of the journal right up to the anniversary session, convened in Obninsk
in June 1974 [3]. At present, atomic energy technology, starting from a 5-MW reactor, has already commis-
sioned the 50th Anniversary of the USSR Novovoronezh Atomic Power Station (1455 MW) [4] and the V. I.
Lenin Leningrad Atomic Power Station (2000 MW) [5], not to mention several large stations in the construction
and planning stages [6]. High-capacity fast breeder reactors [7] are in industrial operation; these seem likely
to make an important contribution to the further development of energy technology. The 25th Congress of the
Communist Party of the Soviet Union took the decision to construct atomic power stations to a total capacity of
13-15.103 MW during the 10th Five-Year Plan.
In the first issue of the journal we published information on an exhibition where a model was demon-
strated of an atomic-powered icebreaker that was then under construction [8]. The two atomic icebreakers
? "Lenin" and "Arktika" are now operating successfully in northern waters; they have greatly expanded the pos-
sibility of polar navigation [9] and a third, "Sibir'," has been launched.
Studies of the controlled thermonuclear reaction were from the very start one of the main concerns of
contributors to the journal. In the third issue we published Academican I. V. Kurchatov's widely noted lecture,
"On the possibility of producing thermonuclear reactions in a gas discharge," which was delivered on April 25,
1956 at the British research center at Harwell [10]. This reported for the first time the researches of Soviet
physicists, who observed the appearance of neutrons as a result of pulsed pinching of a deuterium plasma.
After this, publicationof research into controlled thermonuclear synthesis began and the internation ex-
change of information started at conferences, seminars, and symposia. I. V. Kurchatov greatly expanded ther-
monuclear studies and attracted to them many physics laboratories in the country. The first advances in this
area were also reflected in the pages of "Atomnaya Energiya." Later, however, our journals proved inadequate
for the publication of all the material, and so the new journal "Fizika Plazmy" (Plasma Physics) was created.
Studies have now reached the stage where there is a definite possibility not only of the actual construction of
a thermonuclear reactor but also of elucidating the many important characteristics of its equipment and of prepar-
ing for trials and operation installations capable according to calculations of giving an actual yield from ther-
monuclear synthesis. Great progress in this was made at the I. V. Kurchatov Institute of Atomic Energy with
the creation of installations of the tokamak type [11]. The Tokamak-10, completed in autumn of 1975, is one
of the most perfect modern thermonuclear installations. The state of research now urgently requires studies
of ways of developing thermonuclear energy technology and its efficient combination with nuclear-fission energy
technology [12].
The peaceful use of nuclear explosions is beginning to acquire practical significance [13, 14, 15]. In the
large hydraulic-engineering plan to divert part of the flow of northern rivers to the south into the Volga basin,
nuclear explosions were considered a highly effective means of constructing channels [16].
The pages of the journal have reflected the progress of fundamental scientific researches. The first issue
carried an article [17] describing the then largest heavy-particle synchrotron, the 10-GeV machine at Dubna.
Later large electron accelerators were built at Khar'kov, Erevan, and Tomsk, a heavy-ion accelerator at Dubna,
Translated from Atomnaya Energiya, Vol. 40, No. 5, pp. 367-369, May, 1976.
This material is protected by copyright registered in the name of Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part
of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying,
microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $7.50.
448
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
a proton accelerator at Moscow, and one of the largest proton accelerators in the world, the 70-GeV machine at
Serpukhov [18]. It is equipped with large-scale research installations of Soviet ("Lyudmila") and foreign ("Mir-
abelle") manufacture [19-21]. The use of the accelerators promoted the discovery of new phenomena in the
interactions of particles and nuclei and manifestly expanded and consolidated international scientific collabora-
tions. Some methods of research developed in our laboratories have been used in the subsequently constructed
giant accelerator at Batavia (USA) by a group of Soviet physicists working there in collaboration with the Amer-
icans.
Together with the principal lines of energy research other areas of science and technology were also
developed, united under the concept of atomic energy and by the subject matter of our journal. The industrial
use of research methods developed in physical and chemical laboratories has grown and spread; radioactive
isotopes as well as technical devices and materials created in the atomic industry have been assimilated in
the practice of many areas of the national economy. Information on the main advances in these diverse fields
has constituted one of the aims of the journal, and its pages have reflected many achievements of a practical
nature, such as nuclear instrumentation [22], activation analysis [23], nuclear geophysics, radiation chemistry
[24], dosimetry and radiation protection [25, 26], radioecology [27], and the presowing irradiation of agricultural
crops [28].
The flow of scientific information sent to the journal is vast; consequently, in 1965 we introduced a method
for the deposition of papers that later was adopted by many other journals. The founding of new journals ("Yad-
ernaya Fizika," "Fizika Plazmy," "Khimiya Vysokikh Energii," etc.) and of the collection "Voprosy Atomnoi
Nauki i Tekhniki" has lightened the portfolio of the journal, but the deposition of papers has been retained and
remains highly advisable.
Atomic science and technology in the Soviet Union has made profound advances in the last 20 years, both
in important and subsidiary directions in energy research, and the editorial board and staff hope that the journal
"Atomnaya Energiya" has contributed to these achievements.
The quality and timeliness of the scientific information published in the journal depend on its numerous
authors and reviewers. The editorial board and staff acknowledge their services and share with them the satis-
faction of 20 years' activities on behalf of Soviet science and technology.
LITERATURE CITED
1. D. I. Blokhintsev, N. A. Dollezhal', and A. K. Krasin, "The reactor at the atomic power station of the
Academy of Sciences of the USSR," At. Energ., No.1, 10 (1956).
2. D. I. Blokhintsev, M. E. Minashin, and Yu. A. Sergeev, "Physical and thermal calculations for the reactor
at the atomic power station of the Academy of Sciences of the USSR," At. Energ., No. 1, 24 (1956).
3. D. I. Blokhintsev, N. A. Dollezhal', and A. K. Krasin, "Some conclusions from the operation of the world's
first atomic power station," At. Energ., 36, No. 6, 423 (1974).
4. A. N. Grigor'yants et al., "Ten years of operation of the 50th Anniversary of the USSR Novovoronezh
Atomic Power Station," At. Energ., 38, No. 1, 3 (1975).
5. N. A. Dollezhal' and I. Ya. Emel'yanov, "Experience in the creation of high-capacity reactors in the
USSR," At. Energ., 40, No. 2,117 (1976).
6. L. M. Voronin and E. Yu. Zharkovskii, "Atomic energy in the USSR in the Ninth Five-Year Plan," At.
Energ., 40, No. 2, 167 (1976).
7. 0. D. Kazachkovskii, "Development of work on fast reactors in the USSR," At. Energ., 36, No. 6, 444
(1974).
8. I. I. Novikov, "Exhibition on the peaceful use of atomic energy," At. Energ., No. 1, 102 (1956).
9. F. M. Mitenkov et al., "The icebreaker 'Arktika,' a new achievement of Soviet atomic shipbuilding,"
At. Energ., 39, No. 3, 163 (1975).
10. I. V. Kurchatov, "On the possibility of producing thermonuclear reactions in a gas discharge," At. Energ.,
No. 3, 65 (1956).
11. L. A. Artsimovich, "Studies of controlled thermonuclear synthesis in the USSR," At. Energ., 31, No. 4,
365 (1971).
12. I. N. Golovin, "On the place of hybrid reactors in the world energy system," At. Energ., 39, No. 6, 379
(1975).
13. I. D. Morokhov et al., "Third International Conference on the Peaceful Use of Underground Nuclear
Explosions," At. Energ., 34, No. 5, 407 (1973).
14. I. D. Morokhov et al., "Fourth International Conference on the Peaceful Use of Underground Nuclear
Explosions," At. Energ., 39, No, 2, 148 (1975).
449
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
15. I. D. Morokhov et al., "The peaceful use of atomic energy and the problem of restricting the spread of
nuclear weapons," At. Energ.-, 40, No. 2, 99 (1976).
16. A. K. Kruglov, "Atomic science and technology in the national economy of the USSR," At, Energ., 40,
2, 103 (1976).
17. V. I. Veksler, "Principles of charged-particle accelerators," At. Energ., No. 1, 75 (1956).
18. Yu, M. Ado et al., "Some results of bringing into operation the 70-GeV proton synchrotron at the Institute
of High-Energy Physics," At. Energ., 28, No, 2,132 (1970).
19. V. A. Vasillev, "The Lyudmila liquid hydrogen bubble chamber," At. Energ., 32, No, 3, 262 (1972).
20. "Inauguration of the French liquid hydrogen bubble chamber Mirabelle at the Institute of High-Energy
Physics," At, Energ., 31, No. 6, 641 (1971).
21. A. Berthelot and R. M. Sulyaev, "The Mirabelle bubble chamber for the Serpukhov accelerator," At.
Energ., 32, No, 5,371 (1972).
22. L. M. Isakov and V. V.. Matveev, "On the use of nuclear instrumentation in the control of environmental
pollution," At. Energ., 35, No. 6, 417 (1973).
23. A. S. Shtant, "Instrumental basis of activation analysis in the USSR," At, Energ., 33, No. 4, 858 (1972).
24. V. I. Goltdanskii, "Radiation photochemistry as the possible basis for efficient use of dual-purpose re-
actors," At. Energ., 39; No, 4, 243 (1975).
25. Yu, A. Izrael', "Peaceful use of atomic energy and the environment," At. Energ., 32, No. 4,273 (1972).
26. A. I. Burnazyan, "Radiation protection in the operation of atomic power stations," At. Energ., 39, No. 3,
167 (1975).
27. M. Zaduban, "Problems of radioecology associated with the development of nuclear energy technology,"
At. Energ., 34, No, 5,376 (1973).
28. N. M. Berezina et al., "On the introduction of presowing gamma irradiation of seeds in 'Kolos' equipment
into agricultural practice," At. Energ., 37, No, 1, 43 (1974).
450
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
ARTICLES
STEAM-SUPERHEATING FUEL ELEMENTS OF THE REACTORS
IN THE I. V. KURCHATOV BELOYARSK NUCLEAR POWER
STATION
A. G. Samoilov, A. V. Pozdnyakova, UDC 621.039.54:621.311.2:621.039
and V. S. Volkov
One of the most promising directions in nuclear power development is that of high-temperature thermal-
neutron reactors; the high efficiency of these reactors enables nuclear power stations to compete with thermal
power stations using organic fuels.
One of these is the high-pressure steam-superheating uranium?graphite reactor, which underwent its
initial development in the Soviet Union at the end of the fifties [1]. This reactor is one of the channel class,
in which the coolant pressure is taken up by individual, small-diameter channels. The reactor uses slightly
enriched uranium, the moderator is graphite, the number of fuel channels 998, of which 730 evaporating chan-
nels are provided for the preheating and partial evaporation of the water of the first circuit (P 150 atm,
T 340?C, mass content of steam at the outlet ? 30%) and 268 superheating channels are used for superheat-
ing the steam of the second circuit (P 90 atm, T 500? C) [1,2]. The superheated steam is passed directly, with-
out any intermediate heat exchanger, from the superheating channels to the turbine of the electrical generator. The
active zone of the reactor (7,2 m in diameter and 6 m high) is surrounded by a graphite reflector 0.8 m thick. The
thermal power of the reactor is 285 MW, the rated period of a campaign 730 days, and the amount of 235U consumed
during the campaign 243/198 kg (the denominator indicates the amount of fissile isotopes).
We see from the specification and parameters of the reactor that the fuel elements of the superheating
channels are expected to endure severe working conditions. Since this reactor constitutes a further develop-
ment of the reactor used in the world's first nuclear power station, it retains the basic structural concept of
a tubular fuel element, withstanding the coolant pressure during normal operation and preventing contamina-
tion of the circuit and turbine by fission activity should the fuel element lose its hermetic properties. The
designed operating conditions of the fuel elements are indicated in Table 1.
For the evaporating channels it is quite possible to use fuel elements similar to those of the first nuclear
power station, with the same fuel composition but slightly greater dimensions of the central and outer cans.
The question of using the same fuel elements in the superheating channels raised serious doubts and it was
considered essential to develop a more reliable fuel element. In creating a fuel element producing steam at
a temperature of ? 500?C and a pressure of ? 100 atm and working under fairly severe thermal loads lay one
TABLE 1. Operating Conditions of the Reactor Fuel Elements [I]
Characteristic
Channel of max. power
EC*
sct
Thermal flux, kcal/ (m2? h)
525-103
480 103
Max. coolant velocity, m/sec
0.0
57.0
Max. temperature, ?C:
inner fuel-element can
355
530
fuel composite
400
550
graphite
660
725
*Evaporating channel.
t Superheating channel.
Translated from Atomnaya Energiya, Vol. 40, No. 5, pp. 371-377, May, 1976. Original article submitted
July 9, 1975.
This material is protected by copyright registered in the name of Plenum Publishing Corporation, 227 West 17th Street; New York, N. Y. 10011. No part
of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying,
microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $7.50.
451
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
Declassified and Approved For Release 2013/04/10 : CIA-RDP10-02196R000700070005-6
TABLE 2, Design Characteristics and Working Conditions of the Superheating Channels [6]
Characteristic
Initial form for
reactor of first
unit
U-shaped channel with six fuel
elements for reabtor of second unit
down-going fuel
elements
up-coming fuel
'elements
Max. channel power kW
368
767
Min. channel power kW
202
548
Steam flow through max.-power
channel, kg/h
1900
3600
Steam flow through min.-power
channel, kg/h
1040
2570
Press, at channel entrance, atm
110
132
124
Press, at channel outlet, atm
100
125
110
Steam temp. at channel entrance, ?C
316
328
397
Steam temp. at channel outlet, ?C
510
399
508
Max. thermal loading, kcal/ (m2-11)
480.103
820 103
680.103
Max. steam velocity, m/sec
57
76
112
Max. temp., ?C:
inner can of fuel element
530
426
531
fuel composite
550
482
565
graphite
725
735
of the chief scientific and technical problems besetting the development of reactors with nuclear steam super-
heating.
The main requirements imposed upon the fuel elements of the superheating channls are: good radiation
resistance over long periods, compatibility between the fuel-element materials, and between the fuel-element
can, the graphite stack, and the superheated water vapor, a reasonably high thermal conductivity of the fuel
composite (so as to avoid large temperature drops in the fuel element, thermal stresses, and the possibility
of exceeding the temperature limit of compatibility of the fuel-element materials), good heat resistance and
hot strength in order to ensure the mechanical wholeness of the fuel element during the entire operating period,
and corrosion and erosion resistance of the fuel composite in superheated steam so as to avoid loss of material
on dehermetizing the fuel element.
In designing the fuel elements, certain mutually exclusive requirements were encountered, and these had
to be met with a compromise solution. Thus, e.g., the requirement of a high mechanical strength in order to
ensure the wholeness of the fuel element, insignificant radiation damage, containment of the fission products,
etc., contradictedthe requirement that rapid mechanical breakdown of the fuel element should occur after damage
to the central tube so as to ensure the rapid transfer of radioactive fission products into the reactor stack to-
gether with the stream of coolant, and hence prevent radioactive contamination in the circulation circuit and
the power equipment of the power-station machine room.
In the initial analysis of possible methods of developing the steam-superheating fuel element, a number
of solutions characteristic of other reactors were rejected. In particular it was found unacceptable to consider
such heat-resistant materials as graphite-based dispersion-type fuel composites, or composites of uranium
dioxide and stainless steel as used in certain foreign developments ("Pathfinder" superheated-steam reactor
[3], the high-temperature gas reactors HTGR [4], "Dragon" [5], and others). The basis of our rejection of these
materials was as follows. It was impossible to use graphite as a construction material in a tubular fuel element
sustaining a high water-vapor pressure. The UO2? stainless steel composite was unacceptable owing to its in-
adequate thermal conductivity, since for a heavy loading of the reactor with uranium a fairly thick active layer
would be required and this would involve extremely high temperature drops. In addition to this, in view of the
mechanical characteristics of this fuel composite, it would be impossible to use a high volumetric proportion
of uranium dioxide, so that a high proportion of stainless steel would have to be employed and highly enriched
uranium would be required.
The development of the steam-superheating fuel element proceeded in several directions. After prelim-
inary tests on the technology of making and checking the serviceability of fuel elements constructed in various
ways with various materials, a tubular fuel element with a stainless steel can and a uranium dioxide-base _ com-
posite fuel was chosen for more extensive development.
452
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
1.-j111 .1:
It 12
020x0,3
Fig. 1. Basic construction of the steam-superheating channel and the
fuel element. a) Channel: 1) release of steam from the channel; 2) entry
of steam into the channel; 3) upper head; 4) sealing rings; 5) fuel element;
6) linear-expansion compensator; 7) lower head. b) Channel cross section:
1) graphite collar; 2) fuel element; 3) steam tube. c) Fuel element: 1)
inner tube; 2) end cap; 3) outer can; 4) fuel composite.
In its initial form the fuel element was of tubular construction, formed by two coaxial cylindrical stain-
less-steel cans having dimensions of 4)9.4 x 0.6 and 4)20 x 0.3 mm, containing spaces distributed along the
annular gap and connected at the ends of welding on end caps. In the annular gap between the inner and outer
cans, a fuel composite consisting of UO2 fragments dispersed in a matrix alloy was placed.
The steam-heating channels containing fuel elements of this kind were similar in construction to the
evaporating channels, i.e., they consisted of six fuel elements arranged in a graphite collar around the central
steam-carrying tube. Passing along this tube the steam enters the central tubes of the fuel elements and is
superheated as it moves along them [61. The construction of the steam-superheating fuel element and channel
is illustrated in Fig. 1, and the designed working conditions are given in Table 2.
Later a U-shaped construction of the steam-superheating channels was developed; this was distinguished
by the fact that the central steam tube was eliminated and the steam was superheated by successively passing
first through three fuel elements in a downward direction and then through three fuel elements in an upward
453
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
Fig. 2. Cross section of the U-shaped channel and the fuel element:
Six-element channel (a); five-element channel (b); 1) inner tube of
the fuel element; 2) fuel composite; 3) outer can of the fuel element.
direction. In the place of the former central tube was a soft-regulation absorbing rod enabling the power of the
fuel channels to be to a certain extent equalized. This construction eases the operation of three of the fuel ele-
ments in the channel because of the lower temperature, and reduces the temperature of the graphite stack (-
100?C for a channel power of 360 kW), which favorably influences the working conditions of the graphite and the
physical characteristics of the reactor. The construction of the fuel elements was slighly altered in this kind
of superheating channel: instead of the central tube 9.4 x 0.6 mm in size we used a tube of 12 x 0.6 mm, al-
though leaving the can material and fuel composite as before. As a result of these constructional changes to
the fuel element and channel the hydraulic resistance of the channel was reduced,,while the reduction in the
thickness of the active layer led to a reduction in the nonproductive absorption of neutrons and to an improve-
ment in the physical characteristics of the reactor [2, 6]. Channels of such a construction were used in the re-
actors of the first and second units of the power station. The design characteristics of these channels and the
working conditions of the fuel elements are shown in Table 2 and the construction in Fig. 2.
Efforts to improve ,the physical and thermotechnical characteristics of the reactor led to the further
modernization of the channel and a change of fuel elements. One of the fuel elements in the channel was re-
moved; the steam was then superheated in the five fuel elements (three down-coming and two rising). The inner
tube of the fuel element was increased to a size of 4)16 x 0.7, the outer to 4)23 x 0.3 mm. Following this re-
construction the physical and thermotechnical characteristics improved sharply as a result of a reduction in
the volume of matrix material in the fuel element and an increase in the open cross section of the channel.
Superheating channels of this construction are being placed in the reactor of the second unit of the power sta-
tion whenever the existing six-element U-shaped channels are exhausted and require renewal [7]. The basic
construction of the channel and fuel element is presented in Fig. 2.
As can material of the steam-superheating fuel element we use stainless steel, which has a considerable
corrosion and erosion resistance in high-pressure superheated water vapor, excellent mechanical character-
istics, and an acceptable thermal conductivity, and is technologically convenient in the manufacture Of the fuel
elements (production of long, thin-walled tubes, welding, deformation, and so on).
As fuel composite for the steam-superheating fuel elements we use UO2 fragments dispersed in a matrix
alloy. Uranium dioxide has a number of favorable properties leading to its choice as fissile Component: It is
inert with respect to high-temperature water and steam, compatible with many construCtion materials (includ-
ing the stainless steel used for the cans), has a fairly high radiation resistance, and Contains a great deal of
uranium. A serious disadvantage of uranium dioxide is-its low thermal conductivity; which prevents its use in
pure form as a nuclear fuel in a tubular fuel element with unilateral heat take-off. The temperature at the
454
Declassified and Approved For Release 2013/04/10 : CIA-RDP10-02196R000700070005-6
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
TABLE 3. Average Characteristics of the First Unit of the Beloyarsk
Nuclear Power Station before and after Installing Standard Superheat-
ing Channels [8]
Characteristic
Electrical power of unit, MW
Steam press, in front of turbine, atm
Steam temp. in front of turbine, ?C
Press, of exhausted steam, atm
Water flow in 1st circuit, tons/h
Press, in separators, atm
Gross efficiency,%
Specific flow of steam, kg/ (kW. h)
Expenditure of electric power on
internal requirements, %
Before use of
steam-super-
heating ele-
ments
60-70
60-64
395-405
0.09-0.11
1400
95-100
29-32
4.6-4.9
10-12
After intro-
ducing su-
perheating
elements
100-105
80-85
490-505
0.035-0.04
2300-2400
120-130
35-36
3.8-4.0
7-9
surface of the fuel element should never exceed 630-650?C, the temperature of the steam being ? 500?C; hence
the total temperature drop in the fuel element, including the temperature drops in the outer and inner cans, in
the fuel layer, in the core-can contact resistance, and in the coolant layer close to the walls should not exceed
100-120?C. Thus uranium dioxide may only be used subject to its dispersion in a matrix with a high thermal
conductivity.
The matrix alloy developed for this purpose has a high thermal conductivity, so ensuring a high thermal
conductivity of the whole composite and satisfying the condition that the overall temperature drop in the fuel
element should lie below 100-120?C. The alloy is compatible with UO2 and stainless steel, its melting point is
well above the maximum fuel-element working temperature, it is corrosion and erosion resistant in water
vapor, its thermal-expansion coefficients ensure the absence of serious thermal stresses during the heating
and cooling of the fuel element, the mechanical characteristics satisfy the demands made regarding the ade-
quate strength of the fuel element, while providing for its breakdown by the pressure of the coolant if the inner
tube should rupture. A shortcoming of the matrix alloy is its relatively high neutron capture cross section.
However, this disadvantage is compensated by the fact that the number of superheating channels in the reactor
is only ?27% and the volume of the alloy in the composite is relatively small. Hence the use of steam-superheat-
ing fuel elements with the foregoing fuel composite has no serious effects on the physical characteristics of the
reactor.
A great complication in the fuel-element manufacturing technology was the size of the element (active
length 6 m) and the necessity of ensuring fault-free thermal contact between the fuel composite and the inner
can, as well as the requirement that the uranium distribution should be very uniform. Considerable difficulties
also arose in developing methods of testing the quality of the fuel elements. The technology eventually devel-
oped overcame these difficulties, was adopted by industry, and ensured the production of good-quality fuel
elements. Before placing the fuel elements in the reactor their characteristics were analyzed comprehen-
sively in various test systems and their efficiency was verified by tests in the loop of the First Nuclear Power
Station reactor.
The reactor in the first unit of the Beloyarsk Nuclear Power Station started operations in April 1964,
and evaporating channels were initially used for superheating the steam. At this time the steam temperature
was no greater than 400-410?C and the electrical power 60-70 MW. In November 1966, 20 experimental
superheating channels were installed in the reactor, and in July-August 1967 all the heating channels used for
superheating the steam were replaced by standard superheating channels. The use of the superheating fuel
elements greatly improved the thermotechnical and economic indices of the NPS: The reactor was raised to
its rated power, the efficiency of the unit and the power production were increased, and the expenses were
lowered [8]. The average service characteristics of the first unit before and after installing the standard su-
perheating channels are shown in Table 3.
The reactor of the second unit was furnished with standard superheating elements from the very begin-
ning . On initial operation in 1967 the charging of the fuel channels was incomplete: 452 evaporating and 138
455
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
superheating channels. Complete charging (732 items) was effected in 1968 and complete charging with super-
heating channels (266 items) in 1969. The thermal power of the reactor was twice that of the first unit and
amounted to 560 MAT, Hence the specific power of the fuel and the mean thermal loading in the fuel elements
of this reactor were also twice those of the first unit. The reactor of the second unit also differs from that of
the first unit in having a one-circuit cooling system; the water in the evaporating channels is preheated and
partly evaporated, while the steam separated from the water in drum separators passes into the superheating
channels and is superheated in the fuel elements of these.
At the present time the fuel elements of the evaporating and superheating channels have operated for
quite a long period in the reactors of the first and second units, and it has become possible to assess their
true efficiency under full-scale conditions, as well as the technicoeconomic indices of the NPS, and also future
prospects. Details were presented in earlier papers [7, 9]; in the present article we present only information
relating directly to the steam-superheating fuel elements.
Since 1967, 300 superheating elements have been installed in the reactor of the first unit and about 30
have been taken out for various reasons (failure due to errors of the service personnel, damage in the coolant
tract, and manufacturing faults, as well as for test inspections and monitoring). The remaining superheating
channels continue satisfactory operation, none having failed as a result of radiation damage to the fuel elements
or the incompatibility of their materials; premature extractions of the channels have been reduced to individual
cases in each year. In view of the low burn-up intensity, the fuel elements have not yet reached the rated energy
development of 720 MW. days per channel, and in order to exhaust the rated reserve they will have to operate
in the reactor for another three years, when their total operation in the reactor will amount to about 10 years.
Over 450 superheating elements have been installed in the reactor of the second unit since 1967. In this
time about 200 channels have been extracted for partial recharging and replaced by fresh six- and five-element
superheating channels. The mean energy development of the discharged channels is 600-850 MW? days per
channel, or 18-26 MW days/kg U; their service life in the reactor has been five to six years, with over 200
complete cooling and heating cycles. Some of the discharged (incompletely used) channels have been installed
in the reactor of the first unit for further burn-up. Over the period indicated only eight channels were pre-
maturely extracted from the reactor of the second unit because of the breakdown of service conditions or for
test inspections and monitoring.
The maximum energy development of the superheating channels working in the reactor of the second
unit is 950 MW. days per channel.* Since the frequency of the premature extractions of the channels is
apparently not increasing with time, it has been decided to increase the energy development of a con-
siderable group of channels to 1200-1300 M1AT? days per channel (37-40 MW ? days/kg U). Failure of the
five-element channels has never yet been encountered; the energy development achieved is 275 MW? days
per channel (the rated energy development is 980 MW ? days).
Accounting for channels failing for constructional-technological reasons (experimental channels and
channels failing because of a breakdown in service conditions not being taken into consideration), the proba-
bility of the fault-free operation of the superheating channels up to the rated energy development of 720 MW.
days per channel is over 0.96, which is a very high value. Analysis of the operation of the superheating chan-
nels indicates that these have still not reached their limiting energy evolution.
During the operation of the superheating channels, considerable adjustments have been made to their
temperature conditions. Initially the maximum temperature of the steam at the outlet of the superheating
channel was limited to 510?C. However, the good results achieved in the use of steam-superheating fuel ele-
ments enabled this to be raised to 535?C in 1967 and 545?C in 1969. Prolonged service (more than four years)
under these conditions has in no way reduced the efficiency of the fuel elements, so that in 1973 it was decided
to raise the temperature at the outlet from individual channels to 560-565?C.
The high reliability of the steam-superheating fuel elements has ensured stable operation of the reactor in
the nuclear-superheating mode of operation, and has constituted one of the main factors in ensuring a high
efficiency and relatively low net cost in electrical power production. In the period 1970-1973 the mean net
cost of electrical power in the Beloyarsk NPS was 1.15-1.16 for the first and 0.92-0.93 kopecks/(kW ? h) for
the second unit. The net cost of electrical power in the second unit equals the mean net cost of electrical
*As at the end of 1975 the energy development of the superheating channels was 1150 MW. days ? Editor.
456
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
power in thermal power stations in the Ural region; this makes the second unit competitive with ordinary
thermal power stations [7,9].
LITERATURE CITED
1. N. A. Dollezhal' et al., in: Transactions of the 2nd Geneva Conference 1958, Papers of Soviet Scientists
[in Russian], Vol. 2 Atomizdat, Moscow (1959), p. 36.
2. P. I. Aleshchenkov et al., At. Energ., 16, 489 (1964).
3. M. Novick et al., in: Proc. 3rd Intern. Conf. Geneva, Vol. 6 (1965), p. 225.
4. F. Hoffmann and C. Rickard, ibid., Vol. 5, p. 101.
5. C. A. Rennie et al., ibid., Vol. 1, p. 318.
6. N. A. Dollezhal' et al., in: 3rd Geneva Conference, Paper No. 309 [in Russian] (1964).
7. N. A. Dollezhal' et al., in: Experience in the Use of Nuclear Power Stations and Ways of Developing
Nuclear Power Further [in Russian], Vol. 1, Obninsk (1974), p. 149.
8. N. A. Dollezhal' et al., At. Energ., 27, No. 5, 379 (1969).
9. N. A. Dollezhal' et al., At. Energ., 36, No. 6, 432 (1974).
SOME PHYSICAL INVESTIGATIONS IN BFS-1 FAST
CRITICAL ASSEMBLIES
V. A. Dulin, Yu. A. Kazanskii, UDC 621.039.526:621.039.519.4
V. F. Mamontov, and G. I. Sidorov
The accuracy with which the fundamental physical characteristics of prospective fast reactors may be
predicted is as yet far from adequate. Systems of nuclear-physical constants and computing methods are there-
fore usually checked by reference to integrated (macroscopic) experiments in fast critical assemblies [1-4].
TABLE 1. Tablets of the BFS-1 Critical
Assembly
Material Index
Thickness. mm
235U
2
5
UO'
3
9,5
239p11,T
4
2,4
Nat
5
10
C
6
10
Steel
7
8
10
5
Al
9
0
0.3 mm thick.
0.3 mm thick.
10
5
*In an aluminum can
1-In a stainless steel can
Translated from Atomnaya Energiya, Vol. 40, No. 5, pp. 377-381, May, 1976. Original article submitted
November 27, 1974.
This material is protected by copyright registered in the name of Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part
of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying,
microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $7.50.
457
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
TABLE 2. Composition of the Unit Cell
Assembly
ell
Indices of materials*
BFS -22
BFS -23
/BFS -26
BFS -27 {
BFS -28{
BFS ? -30 {
A
A
A
A
A
A
35253553535253532535235
535435353453534535354353
6967692676967
161018161018161018160181610118161018
608608608111111111111111111608608608
5656526565
115116115116115116115116115
565651111111111111111116565
9392393
111911131119111311191113
5352353
111511131115111311151113
555111111111111111111333
* Table 1, column 2.
TABLE 3. Homogeneous Nuclear Concen-
trations of the Central Insertions ? 1021,
nuclei/cm3
Assembly71-1
e
4
c,
E-,
cr>
,o
(-)
2
2
2
2
2
zg
-'7 '
(13...
BFS -22
A
1,44
7,76
-
7,29
-
10,5
2,4814,5
BFS -23
A
-
6,47
1,07
7,29
-
13,0
1,9316,9
BFS -26
B
1,21
0,14
-
-
28,8
-
12,5 20.9
BFS -27
B
1,57
0,18
--
7,67
24
--
6,57 3,95
BFS -28
B
2,32
6,34
--
--
--
12,3
29,4 --
BFS -30
B
2,30
6,30
--
6,63
--
12,12
9,3 3,14
Note. The critical assembles contained traces of hydro-
gen owing to the presence of glue in the sheaths of the
sodium and 5 mm uranium tablets (Table 1, index 2).
The maximum possible H concentration, referred to the
concentration of 235U , was 3% in BFS-22, 1% in BFS-23
(referred to 239P U ), 0% in BFS-26, 1% in BFS-27, 050in
BFS-28, and 1% in BFS-30.
The accuracy of calculations relating to a number of characteristics has been studied on various occa-
sions using the latest critical assemblies (BFS-22; 23; 26; 27; 28; 30) [5-7]. Measurements have included the
central reactivity coefficients of the main fissile elements (235u, 238u, 239pu) ? ,
as well as elements with well-
known cross sections, including those of the absorbing (10B, 137Au, 6Li) and scattering (111,12C,23Na, Pb) classes.
The ratio of the cost of the generated neutrons to the cost of the absorbed neutrons and the ratio of the
reaction rates of a number of fissile and absorbing elements have been measured. The neutron spectra have
been determined in the critical assemblies BFS-26, 27, 28, and 30 [8, 9]. The results presented in this paper
constitute a refinement of the data presented in [7].
The results of the experiments were compared with calculations using the 26-group BNAB-70 system of
constants [2, 3]. Corrections chiefly associated with the finite dimensions of the samples and the heterogeneous
structure of the critical assemblies had to be introduced into the experimental data. The values of these cor-
rections were determined experimentally and also by calculation. An analysis of the estimated experimental
data revealed the principal reasons for the differences between experiment and calculations. The character
of the discrepancies demanded some specific changes in the macroscopic constants.
In this paper we shall briefly describe the critical BFS assemblies and the experimental techniques,
verify the methods of introducing the corrections, and present the principal experimental results together
with the results of homogeneous calculations.
A detailed description of the BFS-1 installation was presented earlier [10]. The critical assemblies
were made up of tubes containing tablets of construction and fissile materials 46 mm in diameter. The thick-
ness of the tablets was varied from fractions of a millimeter to 10 mm (Table 1). The existence of thin 235U
tablets enabled us to create critical assemblies with different degrees of heterogeneity in the center of the
active zone ("homogeneous" insertion pieces). The homogeneous insertion pieces were situated in the center
of the active zone and comprised 57 cells (in 19 central tubes with three cells in each).
Table 2 shows the composition of the elementary (unit) cells of the critical assemblies. A represents
a cell of the active zone and B a cell of homogeneous insertion. The concentrations of 235U and 238U in the A
and B cells are well known (error no worse than 0.5%) and differ from one another by no more than 3%. Table
3 shows the homogeneous nuclear concentrations of the central parts of the active zones. The remaining zones
and the reflector of the critical assemblies were described in more detail in [11].
The reactivity introduced by the sample was determined by measuring two asymptotic periods of the re-
actor (with and without the sample). The mean square error of the measurement was (1-3) ? 10-7 AK/K for a
mean scatter of 5 ?10-7 AK/K in an individual measurement.
For measuring the spectral ratios of the indices of the fissile elements we used fission chambers con-
taining 235U and 238PU and chambers containing natural uranium. The captures in 137Au were recorded by ref-
erence to the residual y activity. We used the method of calibration in a thermal column.
The (crb97)/ (o-i36) ratio was also measured by an absolute method. The absolute rate of captures in gold
was determined by the y coincidence method and also by determining the total number of interactions in an
458
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
TABLE 4, Central Reactivity Coefficients
ssembly
Cell
2350
236pu
61A
loB
197Au
1.11
12c
23 N a
2081'%
BFS -22
A
4,48+-0,03
6,50+0,06 ? ?4,93+0,10
?1,22+0,018
?0,0046+0,0020
BFS -23
A
6,05+-0,06
8,1+0,13 ? ?6,83+0,3
?1,8+0,02
? ?
?
BFS-26
Bhet
8,47+-0,14
11,8?0,25-10,6?0,25 ?
?11,3
1,5+0,03 0,156+0,003
?
?
Sham
7,66+0,08
10,35?0,16 ?8,45+0,2 ?28,0+0,4
?8,88?0,3
2,1+0,07 0,174+0,006
?
?
Bhet
6,27+-0,10
? ?6,45?0,15-18,5?0,6
?6,22+0,16
2,22+-0,06 ?
?
?
BFS -27
Bhom
6,2?0,1
9,30+0,16 ?5,77+0,15 ?16,8+0,6
?5,54+0,14
2,24+0,06 0,168+0,003
?
BFS -28Bhom
8,15+-0,04
12,9?0,1 ?3,75+0,06 ?7,90+0,19
?1,645?0,018
1,73+0,03 0,047+0,001
?
?0,056+0,005
Bhet
7,3+-0,09
? ?3,76+0,08 ?7,59+0,08
?
1,83+0,03 0,0703+0,004
?
?0,05+0,01
BFS ,-30
Bhom
7,24+0,08
11,6+0,16 ?3,69+0,09 ?7,54+0,08
1,70+0,04
1,89?0,030,071?0,0015
0,0472?0,002
?0,056+-0,005
TABLE 5. Ratio of the Average Cross Sections to the Fission Cross Section of 239PU and
the (43x+) /Kitc+) Ratio
Assembly
Cell
235U
197Au
23813
BFS-22
BFS-23-
BFS-26
BFS-26
BFS-27
BFS-27
BFS -28
BFS-30
BFS-30
A
A
Bhet
Bhom
Bhet
Bhom
Bhom
Etiet
Bhom
1,00+0,02
1,03+0,02
1,03+0,02
0,99+0,02
0,98+0,02
0,915?0,020
0,91?0,02
0,92+0,02
0,30?-0,01
0,37+0,01
0,69+0,02
0,60+0,02
0,51+0,01
0,49+0,01
(0,50+-0,01)*
0,207?0,005
(0,212?0,003)*
0,220?0,005
0,218+0,005
(0,222+-0,003)*
0,0240+0,0005
0,0227+0,0005
0,0300+0,0006
0,0322+0,0006
0,0341+0,0008
0,0365+0,0008
0,0369+0,0007
0,97+0,03
0,765+0,025
0,80+0,02
0,80+0,02
1,01+0,019
0,88+0,015
The ratio was measured by determining the absolute capture rates in gold and the fission rates in plutonium.
NaI(T1) crystal. The absolute number of fissions was measured with a chamber containing a known number of
239Pu nuclei and an experimentally determined fission-recording efficiency [12].
+
The ratio of the cost of the fission neutrons to the cost of the neutrons absorbed in the gold (4)x 4'c)
was determined by measuring the reactivity pc, introduced by the absorbing sample (gold) and the absolute
rate of absorption in gold Nc, as well as the pseudoreactivity p-L-f of a 252Cf spontaneous-fission source of
known absolute activity [13]. The activity of the 252Cf was determined by several methods, the error being
0.6%.
One of the aims of these experiments was to study the influence of the heterogeneity of the critical
assemblies on the measured functionals. For this purpose we introduced homogeneous insertion pieces
intothe BFS-26, 27, 28, and 30 critical assemblies, which enabled us, on the one hand, greatly to reduce the
influence of heterogeneity, and on the other to measure effects due to the heterogenization of the insertion
pieces (Table 2). Tables 4 and 5 give the results of measurements in homogeneous and heterogeneous inser-
tions. We see from Tables 4 and 5 that in certain cases homogeneous insertions change the measured func-
tionals very substantially. Estimates regarding the influence of the finite dimensions of the homogeneous
insertions and the magnitudes of the heterogeneous effects in the BFS-22 and 23 critical assemblies will be
given later. Another aim of the experiments was to allow for the finite dimensions of the samples.
There are several ways of describing the calculated dependences of the effects under consideration on
the sample dimensions; these are based on the fact that the neutron spectrum in the sample differs from the
spectrum obtained before introducing the sample. In order to find the average flux pertubation in the sample
we must allow for the resonance and nonresonance neutron absorption, the slowing down of the neutrons in
the sample, and also their multiplication (breeding) in fissile samples. The resonance absorption was
459
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
1,0
49
46
47
10
FS-22
BFS -26
a
1,0
48
46
0,6 0,4
0 401 402 0,03 0,04 0,05 0 goo 0,010 4015 4020
BFS - 22
BE.S.-26,Bhom
0
BS -24,5het
08 1,0
c' 1,02
1,000
1,05
49
0,8
47
402 404 406 408 460 40015 40030 goos
BFS 28
1,00
BFS16
445
0 41
;0
1,5
42
143
fti
------ - A e
405 410 45 49 0 405
LA hr102,4 nucl/ cm2
0,10 415
Fig. 1. Specific reactivity (per unit mass) and
activation of the samples as functions of their
dimensions: a) -,0) calculated and experi-
mental reactivity; b) -, 0) calculated and ex-
perimental activation -- - ,A) calculated and
experimental reactivity; c) -, 0) calculated
and experimental reactivity; d) 0) cal-
culated and experimental activation---),
calculation of activation with an increased so-
dium concentration in the critical assembly;
e, f) -, 0) calculated and experimental re-
activity, ---) calculation of the reactivity
based on the altered system of constants.
TABLE 6. Comparison between Certain
Computed and Experimental Results
Material
Assembly. cell
BFS -22IBFS-2 3 13M-26113n-270S -281E1B-30
A Bhom
2391311
197All
1.0B
61A
1H
izc
zospb
'Na
238U
(131/035) exp /(Pi/P235)aik
1,03
1,27
1,27
0,33
1,035
1,37
1,38
1,08
1,015 1,00
1,25 1,43
1,20 1,20
1,16 1,12
1,03 1.12
1,11 1,08
1,05
1,14
1,21
1,00
1,64
4,7
((cr1)/Irei
9
0,899+0,018
0,971?0,007
25 ?
1,004?0,020
0,989?0,008
49
.0,989+0,015
0,989?0,010
- 81
1,000?0,012
1,000+0,016
vestigations of the physical parameters of uranium-graphite systems with different numbers of cells were
carried out. For the theoretical analysis of the effects of heterogeneity, the Galanin- Feinberg method was
used [3].
The height of the system was assumed to be large so that, considering neutrons lying in planes perpendi-
cular to the axes of the block, their distribution far away from the source is exponential, with decay constant
13; the space-energy distribution of the neutrons can then be written in the form
(I)(x, y, z, E) (Di (x , y, E) exp(-13z),
(2)
where the energy and space variables characterizing the distribution of the neutrons in the transverse section
cannot be separated. If Eq. (2) is substituted into the heterogeneous equation, the problem is then to determine
the parameters depending on the space-energy distribution of the neutrons in the transverse section. A simi-
lar problem was solved for the homogeneous system with equivalent properties, permitting complete separa-
tion of variables. In the calculation, the aging model of deceleration was used, neglecting anisotropy of the
neutron diffusion.
The basic parameters chosen to characterize the heterogeneity of the propagating system with lattice
spacing a were the parameter abrr- and the thermal constant y, the ratio of the number of neutrons absorbed
by the block in unit time to the density of neutrons at its surface.
Calculations were carried out for the uranium-graphite system with different numbers of cells for dif-
ferent values of the characteristic parameters. In Fig. 1 results are shown for the calculation of the correc-
tion Al32. pn0in - fl2het caused by the heterogeneity of the system with abrr = 1.11 and y /2/rD = 0.249 for a
ratio of volumes of graphite and uranium Vc/Vu = 38. As seen from Fig. 1, the correction A,32 rises mark-
edly as the number of cells decreases. In particular, for a lattice with natural uranium having a physical
parameter .K2 100.10-6cm-2 in a system of 20 cells, 6.82 R$ 15.10-6 CM.-2, i.e., the relative shift in the deter-
mined value of the physical parameter is about 15%. If x.2 increases, the critical size predicted on the basis
of the data of the exponential experiment becomes too low. The results obtained are in qualitative agreement
with the known calculations of the value of the critical load for a system with a small number of blocks
[4].
In order to understand how the heterogeneity affects the neutron spectrum in the system, the ratio of the
fluxes of resonance and thermal neutrons 4.r/43t was calculated for systems with different lattice spacings as
a function of the number of cells. The thermal constants were chosen in such a way that the square of the mi-
gration length of the lattice, calculated from the relation
M2=-= P(t-0)+T,
(3)
was retained unchanged in all cases.
In Table 1, calculated values of (4)1,./(1,)rel are shown for central and peripheral blocks. From these
data it is evident that (11./4.t depends on the number of cells and is not constant for all blocks of the system.
With increase in the number of cells, .1)r/41t tends for all blocks to an asymptotic value; the more strongly
466
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
expressed the heterogeneity of the block, the more slowly does (1,1./4).t tend to the asymptotic value.
If we take, e.g., 2% as an acceptable deviation of (131./(1,t from the asymptotic value, then for the uranium?
graphite lattice in the range a bri- = 0.8-1.4 (which is sufficiently characteristic for the system with natural
and slightly enriched uranium) the minium size of the subcritical system that ensures an acceptable divergence
of the neutron spectrum is found to be 20-40 cells, depending on the degree of heterogeneity. Analogous results
are obtained for other systems.
Experiments on the effect of the number of cells on the measurable parameters were conducted on sub-
critical uranium?graphite systems with natural uranium using an IRT reactor as the source of the neutron
beam. The number of cells in the investigated lattice with a/r-r = 1011 for Vc/Vu = 38 was changed from 9 to
81. The height of the system was not changed: in all experiments it was 270 cm. Measurements were made
in the central channel of the system at a distance of 135 cm from the lower end, known to be in the region of
the asymptotic neutron spectrum. The cadmium ratio R2c81 and the ratio (018)/(a9) were measured. The mea-
surements were made by the activation method, using a Ge(Li) spectrometer. The rate of the reaction 239U(n, y) was
determined from the intensity of 239Np y radiation of energy 277 keV, and the rate of fission from the intensity
of I43Ce y radiation at 283 keV. The procedures for the measurements and the treatment of the data were de-
scribed in [1]. The results for the parameters of the system with n = 81 cells are shown in Table 2.
From Table 2 it is evident that the dependence of both Rd and (0-2f8)/(0-18) on the number of cells is
significant only in passing from 25 to 9 cells, when R2&I reduces by 10% and (018)/(0-18) by 3%.
Thus in the given case, Rd is the parameter most sensitive to change in size of the system. For 25
cells and more, the physical parameters are unchanged within the limits of experimental error. If we take into
account that the value of (528 in these lattices scarcely depends on the size of the system, we may conclude that,
when the number of cells is varied, the value of Rd serves as a sufficiently objective criterion for the choice
C
of a system size ensuring that the set of integral physical parameters characterizing the neutron spectrum in
the lattice can be measured. Here the height of the system should be such as to ensure that the characteristic
spectrum is established and that the flux can be represented in the form given in Eq. (2).
The results of relative measurements of Rd are in satisfactory agreement with heterogeneous calculations
C
of (Pr/cIt. It may be assumed that for a heterogeneous system of thermal neutrons the acceptable size satisfy-
ing the required level of accuracy may be determined on the basis of calculations using the Galanin? Feinberg
method.
LITERATURE CITED
1. L. N. Yurova et al., At. Energ., 38, No. 4, 245 (1975).
2. A. M. Weinberg and E. P. Wigner, Physical Theory of Neutron Chain Reactors, University of Chicago
Press, Chicago (1958).
3. S. M. Feinberg, in: Reactor Construction and the Theory of Reactors [in Russian], Izd. Akad. Nauk SSSR,
Moscow (1956), p. 152.
4. A. D. Galanin, ibid. p. 191.
467
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
INVESTIGATION OF THE LIBERATION OF HELIUM
FROM CONSTRUCTION MATERIALS DURING THEIR HEATING
D. M. Sko.rov, N. P. Agapova,
A. I. Dashkovskii, Yu. N. Sokurskii,
A. G. Zaluzhnyi, O. M. Storozhuk,
V. D. Onufriev, and I. N. Afrikanov
UDC 621,039.546:541.183.5:546.291
It is a well-known fact that neutron irradiation generates helium and hydrogen in construction materials
during the (n, a) and (n, p) reactions where the' materials are composed of certain isotopes. The formation
of these gases implies changes in the mechanical and physical properties of the materials, particularly in
their radiation-dependent high-temperature embrittlement and swelling.
The present paper is concerned with investigations of the behavior of helium in nickel and OKh16N15M3B
steel during tempering.
The nickel (99.99%) and OKh16N15M3B steel (,-,100-p-thick foil with 30-50 p recrystallized grains) sam-
ples were saturated with helium on the ILU-100 magnetic mass separator of the Physics Power Institute by
bombarding the samples of the materials under inspection with 70-keV a particles to a dose of 3.10" parti-
cles/cm2 (flux density 1?A/cm2). The sample temperature did not exceed 100?C during the irradiation process.
The helium distribution over the thickness of samples treated on the magnetic mass separator corresponded
to a Gaussian distribution [1]. According to calculations, the layer saturated with helium extends from the sur-
face to a distance which is much smaller than the grain size in the samples of the materials under inspection.
Samples were saturated with helium by bombarding foils of OKh16N15M3B steel with a particles having
energies of up to 40 MeV in the cyclotron of the I. V. Kurchatov Institute of Atomic Energy. In order to obtain
samples of the alloy which were uniformly saturated with helium, a rotating disk with appropriate filters was
inserted between the ion source and the foils to be saturated. The sample temperature did not exceed 70?C in
the course of the irradiation. The calculated helium concentration in the irradiated samples was 10-2 at. %.
The liberation of helium from nickel and OKh16N15M3B steel was studied with the aid of a special setup
[2] whose operation is based on the mass spectrometric determination of the partial pressure of an inert gas
which has been liberated at some instant of time from the sample annealed in vacuum. However, in contrast
to the previously described method of investigating the liberation of gases during isothermal annealing [21,
the present work was based on investigations of the kinetics of helium liberation from construction materials
heated at a constant rate. The uniform heating was ensured by changing the voltage applied to the heater in
accordance with a preset program. In order to maintain the required vacuum in the annealing chamber, a
getter-ion pump and an electrical discharge pump were simultaneously used during the experiments. The helium
Fig. 1. Kinetics of helium liberation from
a sample of OKh16N15M3B steel.
Translated from Atomnaya Energiya, Vol. 40, No. 5, pp. 387-390, May, 1976. Original article Sttinnitted
November 14, 1974.
This material is protected by copyright registered in the name of Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011..No part
of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying,
microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $7.50.
468
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
25
ZO
0 15
czt1:
10
it ILE
500
200 .400 600 800 1000M(
Fig. 2. Kinetics of helium liberation from
a nickel sample.
Fig. 3. Kinetics of the liberation of helium
from an 01(1116N15M3B steel sample saturated
with gas in cyclotron (? 0.01 at. % He).
liberated in the pressure interval of the experiments was therefore pumped off at a practically constant rate;
the kinetic curves of gas liberation which were obtained logically characterize only the rate of helium libera-
tion from the material: An ascending part of the curve indicates that the rate of liberation exceeds the rate
Of pumping; a descending part means that the rate of pumping exceeds the rate of gas liberation. Three exper-
iments were made with each material heated at a rate of 7 and15 deg C/min to establish the reproducibility of
the results.
Figure 1 shows the liberation of helium from an OKh16N15M3B steel sample which had been saturated
with the inert gas on the ILU-100 apparatus; the sample was heated at a rate of 7 deg C/min.
One can distinguish five peaks in the rate of helium liberation. The characteristic temperatures of the
peaks differ in the experiments by less than 20-25?C. Figure 1 shows that the first peak ( I ) of the rate of gas
liberation appears in the low-temperature region between 300 and 400?C. The other peaks (II-V) are observed
at temperatures in excess of 600?C.
The curves representing the liberation of helium were obtained when nickel samples saturated with the
inert gas on the ILU-100 apparatus were uniformly heated (at a rate of 7 deg C/min). The curves representing
the rate of helium liberation from OKh16N15M3B steel and nickel are identical. The only difference is that the
second peak is stronger in nickel than in steel and slightly shifted on the temperature scale: In the case of steel,
the peak is observed at temperatures between 580 and 620?C, whereas the peak appears at temperatures
between 600 and 700?C in the case of nickel.
Figure 2 shows the liberation of helium from a nickel sample which had been saturated with the inert gas
on the ILU-100 magnetic mass separator and which was heated at a rate of 15 deg C/min. A similar curve was
obtained from an OKh16N15M3B steel sample which had been saturated with the inert gas on the above appara-
tus and which was heated at the same rate. By contrast to the curves representing the separation of helium
469
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
from steel and nickel samples at a heating rate of 7 deg C/min, the latter curves have only three peaks. As
far as the form is concerned, the curves coincide with the curves representing the helium liberation from the
corresponding materials heated at a rate of 20 deg C/min [3]; Three peaks of the rate of gas liberation are
observed (one in the low-temperature region of ? 300?C, and two peaks at temperatures in excess of 600?C).
This could be expected because in both the present work and in [3] the materials under consideration were uni-
formly heated at almost the same rate. Selection of the optimum heating rate of the samples under inspection
is a rather difficult experimental problem, because in the case of a low rate, the peaks are smeared on the
temperature scale, whereas in the case of a high rate, a superposition of neighboring peaks is observed.
The liberation of helium from OKh16N15M3B steel samples which had been uniformly saturated with the
inert gas on a cyclotron was also studied. Figure 3 depicts the liberation of helium from such a steel sample
during its uniform heating at a rate of 7 deg C/min. The liberation of helium from that sample took place in
the same fashion as in the uniform heating of similar steel samples which had been saturated by the gas upon
bombardment with low-energy helium ions in the ILU-100 apparatus: One peak (I) of the rate of helium liberation
is clearly visible in the low-temperature range (280-380?C), whereas the other peaks appear at high tempera-
tures (II-VI). In the case under consideration, one more peak is observed in the high-temperature range. The
appearance of peak VI can be explained by the.influence of boundaries upon the distribution of the gas in the
material, because when steel is irradiated in the cyclotron, the entire sample volume rather than a narrow
layer near the surface was saturated with helium ions (provided that the magnetic mass separator was used
for irradiation).
An analysis of the results of [3-7] leads to the conclusion that the liberation of helium from the materials
under inspection during their uniform heating takes place in several stages.
Investigations which were made with an EM-300 electron microscope on similar nickel and OKh16N15M3B
steel samples annealed at 300, 400, and 500?C after irradiation with low-energy helium ions have shown that
the helium does not cause porosity in the materials under these conditions. According to the results of elec-
tron microscopy, in nickel and steel saturated with helium ions, the black spots and small dislocation loops
which are observed in the initial stage are converted into clearly distinguishable large dislocation loops when
the temperature is increased to 500?C; the large dislocation loops are loops of interstitial atoms, as put into
evidence by the work of [8]. The size of the loops increases at increasing annealing temperature and their con-
centration decreases (Fig. 4). One must assume that, after irradiation, most of the helium induced in the sam-
ple is in a metastable solid solution or forms complexes with a size of ? 10 A, as indicated by the result's of
electron microscopy. It is possible that some of the helium atoms remain in the form of the atoms introduced,
which are characterized by increased mobility and which may leave the sample while they generate the first
peak of gas liberation (see Figs. 1 and 3). It is also possible that a certain fraction of the induced helium atoms
form a Cottrell atmosphere near the dislocation loops. This conclusion is corroborated by the fact that the
first bubbles usually appear on dislocation lines. A sharp growth of the loops and the reduction of their concen-
tration imply a reduction of the total length of loop dislocations and a decrease in the concentration of points
at which helium atoms are adsorbed. The helium atoms which are liberated in this process leave the sample
and cause the first peak of gas liberation in the corresponding temperature interval (according to the results
shown in Fig. 4).
The results which the electron microscope helped to obtain in regard to the onset of helium-induced
porosity in nickel at 600?C and in steel at 700?C indirectly confirm that the helium atoms in these materials
have a significant diffusion-induced mobility. The second peaks of the rate of helium liberation in this tem-
perature interval (600-700?C; see Figs. 1 and 3) indicate that vacancies are involved in the diffusion mechanism
of the gas atoms.
If the liberation of a gas from a material which is heated at a constant rate is given by the volume diffu-
sion of the gas atoms, one can determine the activation energy of the diffusion of the gas atoms in the material
from the temperature Tm of the corresponding maximum [4, 5, 9]; the relation
-9? p.e. 69,5+4.6 lg - +4.6 lg 1015,
T Tr, p2f3
is used, in which p denotes the number of atomic layers between the layer saturated with gas and the sample
surface; and Ko denotes the oscillation frequency of the diffusing atoms; a value of 1015 sec-1 can be assumed
for the oscillation frequency of many metals [6]. In this manner one can obtain only tentative values of the
activation energy, because the p and Ko values which are employed are results of calculations. Calculations
have shown that the activation energies which were calculated for the diffusion of helium in the materials under
470
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
180
Fig. 4. Dependence of the average diameter
cT, of the concentration p of loops, and of the
number N of displaced atoms in dislocation
loops upon the temperature of isochronous
annealing cycles with a duration of 0.5 h each;
0) nickel; 0) OKh16N15M3B steel.
inspection on the basis of the temperature of the second peak on the inert-gas liberation curves are approxi-
mately the same for nickel and OKh16N15M3B steel; The activation energies amount to 45-50 kcal/(g ? atom)
because the temperatures of the peaks are practically identical (the difference is within the spread of the
experimental data). The activation energies of the diffusion of helium in nickel and steel coincide with the re-
sults of [3]. The activation energy of the diffusion of helium in nickel corresponds to the activation energy of
the self-diffusion of nickel [10]. This attests to a helium-diffusion mechanism in nickel involving vacancies
at temperatures between 600 and 700?C.
Electron microscopical investigations of the gas bubbles which develop in the materials at the tempera-
ture of the third stage of gas liberation (800-900?C) have shown that almost all the introduced helium is enclosed
in the bubbles. Therefore at temperatures of 900?C or more helium can be liberated from the material only
by nondirectional migration of gas bubbles to the sample surface. It is unlikely that redissolved helium ap-
pears in the pores without neutron irradiation, because the solubility of this gas in metals and alloys is ex-
tremely low.
LITERATURE CITED
1. H. Glyde and K. Mayne, Phil. Mag., 12, 919 (1965).
2. D. M. Skorov et al., At. Energ., 35, No. 4, 269 (1973).
3. D. Whitmell and R. Nelson, Radiation Effects, 14, 249 (1972).
4. R. Kelly and H. Matzke, J. Nucl. Mater., 20, 171 (1966).
5. E. Rued l and R. Kelly, J. Nucl. Mater., 16, 89 (1965).
6. R. Kelly and F. Brown, Acta Met., 13, 169 (1965).
7. K. Willertz and P. Shewmon, Met. Trans.,!, 2217 (1970).
8. N. P. Agapova et al., Izv. Akad. Nauk SSSR, Ser, Fiz., 38, No. 11,23 (1974).
9. J. Farrell et al., Vacuum, 16, No. 6,295 (1966).
10. R. Rantanen et al., J. Vacuum Sci. Technol., 7, No, 1, 18 (1969).
471
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
THE EVOLUTION OF GAS FROM URANIUM DIOXIDE
B. V. Samsonov, Yu. G. Spiridonov, UDC 621.039.343
V. Sh. Sulaberidze, and V. A. Tsykanov
Characteristic features of the macrostructure cross section of highly stressed container-type fuel ele-
ments with uranium dioxide cores (Fig. 1) are the zones of equiaxial grains and columnar crystals separated
by ring-shaped cracks (in the majority of cases) and sometimes by radial cracks, which arise after irradiation
due to a number of causes [1].
Fig. 1. Cross section of GD-122-type fuel
element (after irradiation).
Translated from Atomnaya Energiya, Vol. 40, No. 5, pp. 390-395, May, 1976. Original article submitted
January 15, 1975; revision submitted November 25, 1975.
This material is protected by copyright registered in the name of Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part
of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying,
microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $7.50.
472
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
-0
O 7It
t 4 ri_l
o c L
I
->-..?. o K '
O ? ,
CY. to
Operating conditions of reactor
TABLE 1. Parameters of Fuel Elements
Fuel
ele-
ment
External
and in-
ternal
diams.
of can,
Fuel- Fuel-
col-
quet umn
diam. ht.,
snm mm
Urani-
um
diox-
ide
wt.
Ab-
sorp-
tion
by
135LF/0
Initial
filling of
fuel ele-
ment,
atm.
D-6
D-7
GD-122
GD-125
GD-115
6; 5,4
6; 5,4
9,1; 8,2
9,1; 8,2
9,1; 8,2
5,2
5,2
8,15
8,15
8,15
201
200
350
349
351
36 21
36 21
182 21
182 21
182 36
He; 4
He;4
Air;
1
He;
3,5
Air;
1
*Fuel density for all fuel elements 10.2 g/cm3
Fig. 2. Nature of gas evolution from the vari-
ous structural zones of a fuel element: I) the
equiaxial grain zone only; II) the columnar
crystal zone only; III) the equiaxial grain and
columnar crystal zones.
TABLE 2. Thermal and Physical Param-
eters of Fuel Elements Tested
Fuel
ele-
ment
Fuel-
ele-
mert
power,
kW
Ther-
mai
power
per
unit
length,
W/ cm
Surface
temp.
of fuel,
?
C
Max.
temp.
of fuel,
,?.-.
?
Radius
of col-
umnar
crystal
zone,
cm
Rel.: ra-
dius of
col-
=nu
crystal
zone, %
D-6
D-6
GD-115
GD-122
CM-125
2,6
3,7
22
26
27
128
185
630
740
770
1300
1300
1000
1000
1000
1600
1700
1800
2400
2460
?
?
0,1
0,3
0,32
_
_
25
73
76
In fuel which is run with a large temperature gradient, there is a transport of mass connected with the
displacement of pores into hotter regions. The cause of this process during the first rise in fuel-element
power is the initial porosity of the fuel, while in the following rises it may be due to the porosity of the fuel
in the boundary region between the columnar crystal zone and the zone of equiaxial grains or the ring crack
separating these two zones. The mechanism of pore migration is connected with the vaporization of uranium
dioxide at the hotter wall of a pore and its condensation on the colder wall. The time needed to form the struc-
ture depends on thermal and physical conditions of operation of the fuel element and does not exceed some tens
of minutes. The growing columnar crystals are formed under temperature gradient conditions; therefore a
temperature stress does not arise in them during operation. In the case of the crystals within the equiaxial
granular zone, the appearance of a temperature gradient does give rise to a temperature stress. When the
fuel element ceases to operate, the stress is relieved in the equiaxial grain zone whereas it appears in the
columnar crystal zone. The zones tend to break apart under the effbct of these stresses. As the fuel element
is run again, the process of forming these structures is repeated. The boundary of the zones lies along the
1700?C isotherm.
Consequently, above 1700?C the dioxide forms its original structure consisting of elongated columnar
crystals lying adjacent to each other and having an almost theoretical density. The gaseous fission products
from such crystalline structures are not released into the free volume in practice, but are fixed at their
boundaries. It is as though all the gaseous fission products accumulated during the operation of the fuel
element are cleared out of the columnar crystal zone at the same time as the structure of the transient pores
is reformed.
Bearing in mind the variation in structure we have described, the kinetics of gas evolution from uranium
dioxide can be represented in the following manner (Fig. 2). The gaseous fission products leave the equiaxial
grain zone by thermal activation diffusion, complicated by the presence within the uranium dioxide crystal
structure of pore-trap vacancies, which capture atoms of the gaseous fission products and so prevent them from
leaving the fuel. The mechanism of this process is investigated in [2, 3]. If there is no columnar crystal zone
in the fuel element, then gas evolution must be constant (curve 1). Gas evolution in fuel elements having only
a columnar crystal zone must change suddenly at the instant the reactor comes on load and thereafter remains
constant (curve II). If both characteristic zones are present in the fuel element, then gas evolution will show
the features of both the extreme cases (curve III).
In order to test the proposed model experimentally, a method of direct measurement of the quantity of
473
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
2 16 20 24 28
Burn-up, GW days/ton UO2
Fig, 3. Characteristics of gas evolution in fuel elements type GD: ?
1-3) pressure in can, measured at rated power; 0, +, A) ditto, at no
load for fuel elements types GD-122, 125, and 115, respectively.
32
36
40
Gas evolution
Surge weak intensive
? 4 1 4 10 12 14 16 18 20 26
Burn-up, MW days/ton ti02
Fig.4
10 20 30
Time, hours
5 10 20 30
Burn-up x 10, MW days/ ton u02
Fig.5
Fig. 4. Characteristics of gas evolution in fuel elements type D: 1, 2) pressure in can measured
at rated load for fuel elements types D-7 and 6 respectively; 3,4) ditto, at no load; 5) evolution of
gaseous fission products in dioxide; - - - -) quantity of gas sorbed in fuel elements types
D-7 and 6, respectively.
Fig. 5. Variation of the quantity of gaseous fission products in a fuel element can during irradi-
ation: 1) with the fuel element in operation; 2) with the reactor shut down.
gaseous fission products during irradiation of the fuel elements was developed. This enables us not only to
verify in general terms the picture of the gas evolution mechanism that is at work, but also to detect some
new features of the process, such as an initial surge of gas evolution, absorption of gaseous fission products
in uranium dioxide, and sorption of the gaseous fission products when the reactor is shut down.
Arrangement of Experiments. The kinetics of gas evolution were studied on two series of fuel elements:
one having a columnar crystal zone (series GD) and one without it (series D).
The fuel elements of series GD were irradiated in a water loop of a type SM-2 reactor; the series D ele-
ments were irradiated in an ampoule channel [4] (Table 1). The thermal and physical parameters of the fuel
elements did not remain constant from run to run; therefore, Table 2 gives the characteristics of the most
typical conditions.
All the fuel elements were equipped with means for measuring the quantity of gaseous fission products
evolved from the fuel, the results of the measurements being little affected by the temperature of the fuel
in the element [4, 5]. The summation error in determining the quantity of gas did not exceed ? 8% over all the
experiments.
474
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
P maxl P 0
0.)
a
a
dP
(r.
10 15
Time, hours
Fig. 6. Nature of variation oi pressure
within a fuel element during a surge: 1)
leading front; 2) trailing front; 3) reactor
power.
20
Experimental Results. Figures 3 and 4 show the variation in the amount of gaseous fission products in
fuel elements series GD and D for various durations of irradiation. A fuel-element type GD-122 was removed
from the reactor after 165 effective days so that the structure of its cross section could be examined (see Fig.
1). Figure 3 shows the pressure measured with the reactor shut down. It was considerably lower than when
the reactor was operating, a fact which cannot be fully explained by the lower gas temperature. A reduction
occurred in the quantity of gas inside the can, linked, in our opinion, with absorption by cooled uranium dioxide.
A sudden increase in the pressure within the can during increases in reactor load after a shut-down
is characteristic of series-GD fuel elements having a columnar crystal zone (see fig. 3). In the case of series-D
fuel elements, which do not have a columnar crystal zone, a uniform pressure in the can prior to shut-down and
after start-up of the reactor is characteristic (see Fig. 4).
Discussion of Results. Analysis of the evolution of gaseous fission products from uranium dioxide during
irradiation can best be carried out on the basis of a generalized curve of the kinetics of gas evolution (Fig. 5),
in which it is possible to identify the sections of initial surge evolution, weak evolution, and intensive evolution.
The evolution of gaseous fission products averaged over the series of rims is used in the case of fuel elements
of series GD.
It is characteristic of all fuel elements that a sharp surge in pressure accompanies the first increase in
power (Fig. 6). The duration of the reduction in pressure depends on a number of factors, but the surge does
not usually appear on the scale of the abscissas of Figs. 3 and 4, as it lasts for some hours.
The difference between the pressures before and after the surge(AP = ? Pfin) is basically deter-
minedPinit
by the initial gas medium in the fuel element. If the fuel element was filled with air at atmospheric
pressure, then a vacuum of the order of 10 to 20 mm Hg is formed at the end of the surge. If the fuel element
is filled with helium at 1 to 3 atm, AP will be less than it would be if the filling was air. If the fuel element is
filled with argon, AP will be less, but the duration of the pressure surge will be greater.
When the fuel element is filled with an inert gas at a pressure of not less than 5 atm, the pressures before
and after the surge will be practically the same.
The variation in the quantity of the gases within the can due to the pressure surge is caused by desorption
at the surfaces of the heated uranium dioxide of gases that have accumulated during the processing and storage
of the fuel, This is confirmed by the fact that the curve of pressure growth during the surge can be satisfacto-
rily described by the relationship P Pinit 'max
exp(-E/RT) (E = 5000 to 8000 cal/mole). The subsequent
=
absorption of these gases apparently follows from chemical interactions between the evolved gases and the
incandescent uranium dioxide.
The boundary of the section of weak gas evolution for fuel elements which do not have a columnar crystal
zone (see Fig. 5) corresponds to a burn-up of 10 GW ? days/ton UO2. For fuel elements having columnar crystal
zones, gas evolution commences after the first period of continuous operation of the reactor at shut-down and
during subsequent load shedding. However, in these fuel elements, as calculations show,the gasses arenot
evolved from the equiaxial grain zone over a period of time. Gas evolution only commences in this zone at a
particular time.
475
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
We have calculated the accumulation of gaseous fission products in the equiaxial zone on the basis of
experimental data, up to the moment when intensive evolution commences. In the case of fuel elements without
a columnar crystal zone, the accumulation comprises ? 0.3 N? cm3/g UO2.
It is difficult to calculate the accumulation of gaseous fission products in fuel elements with columnar
crystal zones as there is an appreciable nonuniformity of specific energy evolution with respect to the radius
of the fuel element, which varies with enrichment, as this has an effect on the temperature distribution through-
out the fuel, on the magnitude of the equiaxial grain zone, and on the quantity of gaseous fission products evolved.
However, by estimating the accumulation of gaseous fission products in fuel elements with columnar crystal
zones, we find that this will again be in the vicinity of 0.3 N. cm3/g UO2. Consequently, the evolution of gases
from the equiaxial zone remains significant after the accumulation of gaseous fission p7ducts to a level of
,??? 0.3 N. cm3/g UO2. We should note that over the section of curve with low gas evolution, the quantity of gases
within the can is determined basically by quantity of gases that remain after the end of surge. Therefore,
rarification can occur within the fuel element in a number of cases.
Intensive gas evolution in fuel elements without columnar crystal zones is linked with evolution of gaseous
fission products from the equiaxial grain zone, while for fuel elements that do have columnar crystal zones, it
is linked with the structural reformation process (provided the accumulation of gaseous fission products in
the equiaxial grain zone is less than ???? 0.3 N. cm3/g 102) or with the combined action of reformation and evolu-
tion of gaseous fission products from the equiaxial grain zone (if the accumulation in this zone has reached a
level of ??? 0.3 N. cm3/g UO2). Data about the evolution of gaseous fission products from the equiaxial grain
zone is in satisfactory accord with calculations carried out according to the following scheme. Every zone is
divided by radius into an annular volume of thickness 2a, where a is the radius of an equivalent sphere which
depends upon the initial density [6]. The fuel temperature and specific density of energy evolution are taken
to be constant for each annular volume. The quantity of stable gaseous fission products evolved from the fuel
during a period of irradiation T can be calculated by solving the diffusion equation for the sphere.
ON IOT AN + fY, (1)
where N is the number of atoms of gaseous fission products in the solid body, atoms/cm3; D is the diffusion
coefficient, cm2/ sec; f is the rate of fission of 235U, nuclei/(cm3 ? sec); Y is the summation evolution of stable
gaseous fission products for each fission event.
The solution of this equation for initial condition [N(r, 0) = 0] and boundary condition [N(a, =-- 0] for
the number of gaseous fission product atoms found in the solid body (N') and volume V takes the form
N' fYTITF,
where F characterizes that portion of the final concentration of atoms evolved from the sphere:
00
6a2 6a2
F=1? ' NI 1
n4 exp(
90DT MDT
a-- 1
The coefficient can be determined from the expression
7_70R07?0)
D= 1.65 -10-6 exp
n2n2DT
a2
(2)
(3)
(4)
The referred diffusion coefficient D' = D/a2 increases with increase in burn-up, due to the reduction in
a. The relative radius of the equivalent sphere (a/ainit) is 1.0, 0.45, 0.31, and 0.22 at burn-ups of 10, 20, 30,
and 40 GW? days/ton UO2, respectively.
This variation in a can be calculated from the increase in the surface of the fuel accompanying the increase
in sorption of gaseous fission products due to the increase in burn-up, this being observed in experiments when
the reactor is shut down (see Figs. 3 and 4). The quantity of sorbed gases can be determined from the differ-
ence between the quantity of gaseous fission products inside the can, measured when the reactor is operating
and when it is shut down. It is clear from Figs. 3 and 4 that the quantity of sorbed gases varies little at first,
and then rises considerably. This indicates an inciease in the sorption surface of the fuel in the equiaxial
grain and columnar crystal zones when the reactor is shut down.
CONCLUSIONS
We have examined the basic laws governing the character of gas evolution and the behavior of uranium
dioxide during the operation of a fuel element prior to thorough burn-up. The behavior of the uranium dioxide
476
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
and the evolution of gases depend directly on the distributions of energy evolution and temperature with re-
spect to radius in the fuel element, and on the process of structure formation and burn-up in the different
zones of the fuel. The evolution of gases from uranium dioxide during irradiation can be estimated on the
assumption that all the gaseous fission products evolve from the columnar crystal zone and that evolution of
gases from the equiaxial grain zone takes place by a mechanism of thermally activated diffusion.
Sorption of gaseous fission products at the surface of the fuel can lead to errors in determining the
quantity of gaseous fission products evolved from the uranium dioxide during postreactor determination by
the fuel-can puncture method.
LITERATURE CITED
1. B. V. Samsonov et al., preprint NIIAR 11-177, Dimitrovgrad (1972).
2. B. V. Samsonov and A. K. Frei, At. Energ., 30, No. 4, 358 (1971).
3. B. V. Samsonov and A. K. Frei, At. Energe, 31, No. 2, 136 (1971).
4. V. A. Tsykanov and B. V. Samsonov, The Technique of Irradiating Materials in Reactors at High Neutron
Fluxes [in Russian], Atomizdat, Moscow (1973).
5. B. V. Samsonov, G. P. Lobanov, and V. V. Sidorov, Byull. Izobret. i Toy. Zn., No. 41, 48 (1974).
6. B. Lastman, Radiation Phenomena in Uranium Dioxide [in Russian], Atomizdat, Moscow (1964).
SPUTTERING OF THIN FILMS OF URANOUS - URANIC
OXIDE UNDER THE INFLUENCE OF FISSION FRAGMENTS
AT LOW IRRADIATION DOSES
V. A. Bessonov, G. P. Ivanov, UDC 539.211:546.79
N. A. Grinevich, and E. A. Borisov
It is generally accepted that thin uranium-containing layers are sputtered in a neutron flux by tearing
pieces of the material from a surface [1-4]. At a temperature of 104?C on the track and a density of 11 g/cm3
of the material, the torn-off conglomerates can reach a length of 40 A [5].
Observations made with an electron microscope have shown that by sputtering irradiated thin, 40-400-
X-thick UO2 films and samples of metallic uranium in neutron fluxes exceeding 1014 neutrons/cm2, particles
in the form of "islands" or spots with linear dimensions of up to 400 A can be precipitated on a collector.
These "islands" of material contain about 104 uranium atoms [1, 3].
When accumulations of sputtered material were analyzed with the track detector technique [4], pieces of
material appearing on the track detector as "stars" were detected. The authors assumed that the "stars" were
formed by secondary irradiation of the pieces of the material which were observed under the electron micro-
scope [1]. Sputtered thin films of uranium oxides had the form of conglomerates of large size (up to 109 atoms
in a piece) [6].
Kaminsky and Das [7] have observed the sputtering of a niobium surface; large conglomerates contain-
ing as many as 1011 atoms were formed under the influence of 14-MeV neutrons. But other authors [8] could
not detect a sputtering of the surface of certain metals under the influence of fission fragments nor the forma-
tion of conglomerates with a size greater than 1000 A. It is possible that the formation of the conglomerates in
the experiments made by Rogers results from the irradiation-induced redistribution of the sputtered material
on the collector surface [9]. Furthermore, it is unclear how the sputtering of conglomerates depends upon the
surface under inspection (metal-nonmetal).
The possibility of sputtering uranium-containing materials and pure metals by fission fragments (large
Translated from Atomnaya Energiya, Vol. 40, No. 5, pp. 395-398, May, 1976. Original article submitted
July 1, 1975.
This material is protected by copyright registered in the name of Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part
of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying,
microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for S7.50.
477
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
/o5
log
Sputtering coeff.
103
102
10'
Off
7012
/015 /014 1015 /0/5
Neutron flux (neutrons / cm2)
1011
Fig. 1. Dependence of the sputtering coefficient
(235U atoms per fragment) upon the neutron flux.
conglomerates torn from the surface) was not clearly established. Investigations of the effect are certainly of
interest.
The present article reports on an investigation of surface sputtering under the influence of fission frag-
ments; solid-state track detectors were employed. A thin film of uranous?uranic oxide enriched with 235U was
inserted into the setup consisting of an emitter (thin uranous?uranic oxide film with a thickness of 4 mg/cm2
on aluminum foil) and a collector (aluminum disk with a diameter of 10 mm); the emitter was separated from
the collector by a 0.5-mm-thick diaphragm with an internal diameter of 3.5 mm. The setup with the quartz
tube was evacuated to high vacuum and irradiated in a VVRTs reactor. After irradiation, the collector was
withdrawn from the setup, brought into contact with a glass plate, and irradiated again in the reactor with a
neutron flux of 1012-1015 neutrons/cm2. After etching in 1.4% hydrofluoric acid solution, the glass detectors
were inspected under an optical microscope and photographed. Before the secondary irradiation, the concen-
tration of uranium on the collectors was determined from the alpha activity with the aid of a semiconductor
alpha detector and the "Amur" instrument set.
The sensitivity of the solid-state track detector technique amounted to 1012 g for uniformly distributed
uranium atoms; the "Amur" instrument set had a sensitivity of 10-9 g uranium.
In order to determine the dependence of the sputtering coefficient upon the neutron flux, the emitter?
collector assemblies were irradiated in a reactor with various integral-doses of thermal neutrons. After that,
the uranium concentration on the collectors was determined with the above methods and the coefficient of
ejection was calculated with the well-known equations of [10].
At low irradiation doses, the sputtering coefficient can reach several thousand uranium atoms per frag-
ment (Fig. 1). An analogous overall dependence of the sputtering coefficient upon the dose has been described
in [11, 121, though the neutron flux exceeded 1014 neutrons/cm2.
Since particularly at low irradiation doses the highest sputtering coefficients were observed, it was
logical to assume that the increase is related to the initial state of the surface of the layer subjected to sput-
tering [13]. In the initial moments of irradiation, substantial changes can occur in the surface layer, because
when the irradiation with neutrons and the bombardment of the layer with fission fragments leaving the uranous?
uranic oxide sample begin, stresses can be generated on the surface sections through which the fission frag-
ments pass; protruding pieces, "islands," can be torn off if they are located in the zone in which the track of
a fission fragment is effective; and sections which are not very well joined with the base layer can be destroyed.
A surface which had not been initially prepared was subjected to sputtering, i.e., a surface which had not
been ground, polished, or modified with some other operation for leveling and smoothing the surface. When a
cross section profile of such a uranous?uranic oxide layer was inspected with an optical microscope, many
coarse irregularities, pits, and protrusions could be recognized on the sample surface. Therefore, the fact
that pieces containing a large number of uranium atoms are torn from the emitter at the beginning of emitter
irradiation can be one of the reasons for the noticeable increase in the coefficient of ejection. The high sensi-
tivity of the track detectors and observations of the distribution of the material precipitated on the collector
made it possible to follow the changes in the uranium precipitation process while the flux of the irradiating
neutrons was increased.
Some characteristic formations which could be recognized on the collectors with solid-state track detec-
tors are shown in Fig. 2. At a neutron flux of 1012_1014 neutrons /cm2, pieces which on the track detector are visible
as dense "stars" with a size of 30-50 ? (Fig. 2a and b) are present on the collector surface.
478
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
o,imm.
Fig. 2. "Stars" on the detector surface in the
case of primary (emitter?collector) and second-
ary (collector?detector) irradiation (neutrons/
cm2): a) 1011, 1014; b) 1013, 6 ? 1012; c) ?,? 1013, 1014;
d) 1013, 1014 (large "dissected star"; distance
between the collector and the detector 0.5 mm);
e) 1016, 6.1013; 1) 1016, 6.1012, respectively.
Both the size of a "star" on a detector and the track density in the detector depend upon the dose of the
secondary irradiation and the amount of material in the piece precipitated on the collector. In the case of a
secondary irradiation with a neutron flux of 1015 neutrons/cm2, pieces, each of which contains less than 106
uranium atoms, give one or two fragments leaving on the side of the detector. When such a detector is ana-
lyzed, one cannot detect an accumulation of uranium atoms on the collector, because the entire field of the
detector is uniformly covered with individual tracks. "Stars" can be obtained only from pieces containing at
least 107_108 uranium atoms.
When the size of "stars" was estimated, pieces which on the collector had linear dimensions of less than the
range of the fission fragments in the particular material (several microns) were considered as point sources.
The critical angle of incidence was assumed as 35? for glass [14]; the track diameter after etching was assumed
to be 7-9 p. It is possible to show for these conditions that pieces of the material on the collectors must result
in "stars" with a diameter between 30 and 40 p on the track detector. When the distance between the collector
and the detector is increased to several dozen or more microns, the tracks in the glass of the detector appear
as "porous (dissected) stars." When the collector with a detector spaced 0.5 mm from it was irradiated for a
second time, images of the uranium pieces were obtained in the form of "dissected stars" (see Figs. 2c and
d). "Stars" of large diameter and with dissected structure can also result from a collector or detector surface
which is contaminated by occasional aerosol particles containing uranium (Fig. 3) or can be found after irradi-
479
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
Fig. 3. "Dissected" background "stars" obtained
upon irradiating a track detector of glass which
was in contact with a collector of pure aluminum;
thermal neutron flux 1014 nentrons/cm2.
ating collectors containing large pieces of the material (106 or more 235U atoms) with small doses (6,6
10,7
3,0?0,5
4
Vacancy
7,3?1
5,45?0,05
2,4--4,2
Double vacancy
7
Interstitial C
atom
6,75?0,4
9,76
0,45?0,05
0,140
9,90
2,8?0,2
0,016
Vacancy + inter-
stitial atom
11,5?2
neutrons
36,5
neutrons
C?C
0,9?0,1
Remark. The activation energy on the base plane is indi-
cated in the numerator; the activation energy in per-
pendicular direction is indicated in the denominator.
Translated from Atomnaya Energiya, Vol. 40, No. 5, pp. 399-408, May, 1976. Original article submitted
January 24,-1975,
This material is protected by copyright registered in the name of Plenum Publishing Corporation, 227 TVest 17th Street, New York, N.Y. 10011. No part
of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying,
microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $7.50.
481
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
5 1,0 1,5 2,
cp? lo22, neutrons / c m2
Fig. 1. Dependence of the relative change in the
parameters of the crystal lattice of graphite [2]
upon the accumulated dose at various tempera-
tures (?C) :0) 300; A) 350-440; A) 450-600; ?)
650; 0) 900-1350.
a high degree of perfection of the crystal lattice. When radiation defects are studied in polycrystalline graphite,
the structure-dependent properties are restored by thermal annealing and radiative annealing.
We discuss below papers describing research on radiation defects in synthetic graphite materials, among
them materials which were intensively irradiated. When the relation between the defects and the changes in the
properties of graphite during its irradiation is to be analyzed, a systematic order must be established in the
publications in regard to the influence of the radiation defects upon the various properties of the graphite in the
light of the present concepts concerning the mechanism of radiation-induced damage to materials.
At the present time, most of the authors assume that neutron irradiation creates point defects, vacancies,
and interstitial atoms in the structure of graphite. The equilibrium concentration of these defects depends upon
the intensity of the damaging flux, the time and the temperature of irradiation, etc. The resulting defects do
not exist in isolated form. Under certain conditions the defects interact and reduce the internal energy of the
crystal. It was theoretically calculated that the energy of migration of isolated atoms amounts to 0.016 eV [4]
and, according to recent data [3], to 0.14 eV. The corresponding value for vacancies is 4 eV (see Table 1).
The high mobility of the interstitial atoms causes them to associate into C2 and C3 pseudomolecules at 78?K
[4]. At temperatures in excess of 110?K, the molecules form (C2)n accumulations. Since the (C2)2 molecules
disintegrate into the simpler components C2 and since the latter migrate, a peak appears at a temperature of
?200?C on the curve representing the liberation of the stored energy during annealing.
Accumulations of 2-4 interstitial atoms [5] were detected (by scattering of slow neutrons) in graphite
which had been irradiated at ? 30?C. The size of the accumulations increases at this irradiation temperature
with increasing dose and, finally, saturation is reached. The size of the accumulations and the distances be-
tween them increase in the interval 30-600?C with the temperature of irradiation [6, 7].
After irradiation with (0.6-6.0) ? 102? neutrons/ cm2 at 350?C or higher temperatures , the accumulations have
a size of 30-40 A and reach 200 A upon irradiation at ?-? 650?C. It was assumed [4] that when the dimensions of the
molecular accumulations increase, the accumulations can transform into hexagonal structures. These are basically
new atomic planes formed by condensation of molecules. The binding energy in such a complex is obviously
equal to the sublimation energy of 7.4 eV [3]. At temperatures in excess of 1000?C, the atoms of the new planes
interact with free vacancies and, hence, the new planes disappear. Vacancies which are formed during irradia-
tion usually remain isolated and are hardly mobile at temperatures of up to 350?C; but when the temperature
is increased further, the vacancies become mobile [7] and condense into stable accumulations corresponding
to the removal of a part of an atomic plane [8]. These structural disturbances can be destroyed only by self-
diffusion processes having an activation energy of 7-8 eV [3].
It is also assumed that the form of the defects depends upon the degree of perfection of the crystal struc-
ture in carbonic materials [2]. However, there exist no methods of directly observing the defects in materials
having a low degree of perfection of the crystal structure. In materials having a high degree of perfection of
the crystal structure, the type of admixtures and their amounts noticeably influence the damage to the structure
[9]. The degree of radiation-induced damage is a function of the following three variables: intensity of the
damaging flux, and time and temperature of irradiation [10]. A 100-fold increase in the intensity of irradiation
is equivalent to a noticeable reduction of the irradiation temperature.
482
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
Irradiation of graphite results in the accumulation of radiation-induced defects and, hence, in a distor-
tion of the crystal lattice. At low irradiation temperatures, the spacing c of the planes in axial direction in-
creases (Fig. 1) owing to isolated point defects and small groups of them [11, 12]; at the same time a compres-
sion takes place on the base plane.
An empirical formula [14]
r=1.034+-7-,
(1)
can be used to describe the dependence which was observed in some investigations [13] between the order n of
binding (number of electron pairs participating in the chemical bond) and the length of the C?C bond.
The authors of [14] have considered the possible positions of the shifted atoms and vacancies in the crys-
tal lattice of graphite and have analyzed the variation of n in the above formula; the conclusion was that the
expansion of the lattice parallel to the c axis and the compression on the base plane are caused by vacancies.
However, since the calculations are not in a satisfactory quantitative agreement with the experimental results,
defects in the form of interstitial atoms must noticeably contribute to the expansion of the lattice in the direc-
tion of the c axis.
The reduced increase in the parameter c in graphite upon increasing irradiation doses and temperatures
is associated with the coagulation of the interstitial atoms into larger compounds [9]. The broadening of the
profile of the diffraction lines upon increasing the irradiation dose means particularly that the size of the
regions of coherent scattering decreases because the crystallographic planes are distorted. The crystal
structure can become completely amorphous at very high irradiation doses.
The net result of changes in the crystal structure is that the physical properties of the graphite samples
are modified.
The changes in the lattice parameters correspond to the changes in the macrodimensions only at low ir-
radiation doses. At neutron fluxes exceeding 1020 neutrons/ cm2, the expansion of a recrystallized pyrolytic
graphite sample in the direction of the c axis considerably exceeds the increase in the parameter c. This is
explained by the development of the aforementioned molecular accumulations whose dimensions are comparable
with the size of the crystallites [15]. Other authors [7] believe that the lack of correspondence results from
defects on the boundaries of the crystallites.
A compression of the sample occurs on the base plane. The dependence of the changes in the sample
dimensions upon the radiation dose is complicated: The high-temperature setting of any graphite is replaced
by a "swelling" which may reach 100% at high irradiation doses. In [16] the changes which occur in the sample
dimensions in broad temperature and irradiation dose intervals were explained by the dynamics of both the
formation and the transformation of complexes and accumulations. The conclusion was that the secondary
"swelling" of graphite is caused by the oriented formation of a vacancy-induced porosity in the lattice.
The mechanical properties are basically determined by point defects. Polycrystalline graphite has in-
creased strength and brittleness after irradiation. Its hardness increases and reaches saturation at a neutron
flux density of 7.2 ? 1018 neutrons/ cm2 (E a?1 MeV); thereafter the hardness decreases [9]. When graphite is ir-
radiated at temperatures of 56-140?C with a neutron flux density of up to 1 .1020 neutrons/ cm2, the relative in-
crease in the hardness depends linearly upon the absolute value of the neutron flux density; the observed effect
decreases exponentially with increasing irradiation temperature [9]. Electron-microscopical investigations
of the radiation defects [7] have shown that the hardness reaches its maximum when the average spacing of the
defects amounts to 20-50 interatomic distances. When the size of the defects increases further by coagulation,
the spacing of the accumulations increases, which results in a decrease in hardness.
In the case of synthetic graphite, the modulus of elasticity of the first kind increases during irradiation.
At low temperatures (up to 300?C), the change in the modulus of elasticity is nonmonotonic: A sharp increase
at neutron fluxes of 1020 neutrons/ cm2 is followed by a decrease, whereafter the absolute value of the modu-
lus increases again. At temperatures in excess of 300?C, the modulus of elasticity increases gradually; its
absolute value stabilizes and is constant at neutron fluxes of up to 1.5.1022 neutrons/cm2. The increase in the
modulus of elasticity is related to dislocation locking [2]. This effect can also be explained by the influence of
defects upon the elastic deformation of the lattice. The presence of interstitial atoms can reduce the effective
spacing of the layers. But the dislocation mechanism is generally accepted. A decrease in the absolute value
of the modulus is explained by changes in the aggregate state of the interstitial atoms during the irradiation
process.
483
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
The radiatiOn defects in graphite are per se electron traps and create an excess concentration of hole-
type charge carriers by reducing the Fermi level in the valence band of graphite. Therefore, when the irradia-
tion dose is increased, the Hall coefficient changes its sign from negative to positive, reaches its maximum,
and decreases gradually. Layer defects are per se scattering centers which reduce the mobility of the charge
carriers. Since the effect of the reduced mobility is superimposed on the effect of the increased charge car-
rier concentration, the resistivity of an irradiated sample exceeds the initial resistivity. The diamagnetic
susceptibility decreases, because the Fermi level in the valence band is reduced.
Electroparamagnetic resonance studies made it possible to identify the paramagnetic centers with simple
types of defects, namely with interstitial atoms and vacancies [17]. Measurements of the electroparamagnetic
resonance in pyrolytic carbon [18] have confirmed the assumption that trapped paramagnetic centers are related
to interstitial atoms and their simplest complexes of the type (C2)n [19]. The relatively low activation energy
(Emin 0.5 eV) of the movement of complexes of interstitial atoms [20] facilitates their annealing at tempera-
tures of up to 1000?C; this is in good agreement with a reduction of the concentration of trapped paramagnetic
centers. Nonlocalized paramagnetic centers are associated with vacancies and radiation-induced dislocations
which have developed in the layers.
Calculations of the concentration of defects accumulated in pyrolytic graphite (which is identified with a
single crystal) are made under the assumption that interestitial atoms and vacancies exist in the form of indi-
vidual defects, small groups, and, in some number, in the form of loops [21]. These defects change the linear
dimensions Xn and Xa because the lattice parameters are modified. The changes can be considered the reason
for the anisotropic internal pressure which is proportional to the vacancy concentration Cv and the concentra-
tionC1 . of interstitial atoms. In this case the formulas for the concentration of defects [22] assume the form
cz AC/C-I- 1.434a/a ,
2.8
?21.4Aa/a? AC/C
2.8
(2)
(3)
These dependencies are valid only at low temperatures and for low irradiation doses when only point de-
fects and small compounds thereof exist. When both the dose and the temperature of irradiation are increased,
the probability of forming large defect accumulations increases and, hence, the modification of the lattice para-
meters is retarded, whereas the concentration of the defects decreases. The total number of displaced atoms,
which was theoretically calculated with the cascade model of Kinchin and Peace, substantially exceeds the defect
concentration calculated with formulas (2) and (3) [23]. This means that, during irradiation, an intense anneal-
ing of the defects takes place.
It was attempted [14] to explain the observed changes in the parameters and to estimate the defect con-
centration in irradiated graphite with the aid of the dependence of the carbon atom spacing upon the order of
the bond, i.e., r = f(n). It was found with the well-known relations of [22] that in the case of graphite irradiated
at temperatures of up to 100?C, Aa/a =-0.4% and Ac/c = 4.6; the concentrations of vacancies and interstitial
atoms are the same and amount to 1.4%. Estimates of the vacancy concentrations from the above dependence
render r = f(n) 1.8% (from Aa/a) and 3.2% from Ac/c. The value of 3.2% is evidently too high, because a
part of the expansion in c direction can be explained by the influence of complexes of interstital atoms.
The type of defects and their migration energy are usually determined when changes in the physical prop-
erties of materials are studied in the course of thermal annealing and radiation-induced annealing.
Defect annealing is defined as the disappearance of defects from a crystal oversaturated with defects,
when the crystal is heated above the irradiation temperature at which the defects accumulated.
Radiation-induced annealing takes place under the influence of a second irradiation at an increased tem-
perature. Such a secondary irradiation of graphite containing radiation defects reduces the total damage to the
samples to a greater extent than conventional heating to the same new temperature, but without irradiation [24].
In order to determine the defects which were formed in a crystalline material after its irradiation, ther-
mal annealing is usually employed. The extent of annealing and its rate depend upon the mobility of the defects
proper. Excess defects can be removed from the crystal basically by redistributing the defects to sinks and
by recombination.
The theory of thermal annealing has been well developed [11, 25] but an analysis of-the determination of
the activation energy with one of the methods indicated [26] is not possible if one of the properties changes non-
monotonically. For example, the microhardness of irradiated pyrolytic graphite first increases upon isothermal
484
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070005-6
Fig. 2. Dependence of the relative change in the
electrical resistivity upon the temperature of
annealing: 1-6) Gor' Mi Metallurgical Plant
graphite irradiated under various conditions
with the following doses: 7.1021, 7.1021, 7.102?,
? 1019, and ~1018 [32] neutrons/cm2,
respectively; 7) graphite for nuclear applica-
tions, 5.42 ? 1021 [30]; 8) WSF graphite, 8.4 ? 1018
[12]; 9) pyrolytic graphite, additionally treated
at 2400?C, 9 .1020. The samples were initially
irradiated at 50-250?C.
annealing and decreases only thereafter to the initial microhardness values [27].
The defects which were accumulated during irradiation become mobile owing to their activation by ther-
mal annealing and undergo further transformations.
Small groups of interstitial atoms were found in electron-microscopical investigations of the annealing
processes in thin films of natural graphite crystals which had been irradiated at 50?C [28]. When the temper-
ature of annealing is increased, the groups become larger, but disappear at temperatures in excess of 1700?C.
Owing to the condensation of vacancies at 1700?C, loops with a radius of 150 A appear; the loops grow at 750 A
at 2600?C, but the area of all the loops is constant thereafter. The number of condensed vacancies, which
amounted to 0.5% of the total number of atoms, was calculated.
The electron-microscopical investigations of [3] have shown that dislocation loops with a diameter of