SOVIET ATOMIC ENERGY - VOL. 39, NO. 4

• Document Type: CREST
• Collection: General CIA Records
• Document Number (FOIA) /ESDN (CREST): CIA-RDP10-02196R000400060003-2
• Release Decision: RIFPUB
• Original Classification: K
• Document Page Count: 96
• Document Creation Date: January 4, 2017
• Document Release Date: September 24, 2013
• Sequence Number: 3
• Publication Date: April 1, 1976
• Content Type: REPORT

Declassified and Approved For Release 2013/09/24: CiA-RDP16-02196R000406060003-2
? ?
5
'
- Russian Original Vol 39,Noi4, octob,er, 1975
1976
? \,, ,, ,,,,-- '? , ,
V. '. -`, ' ? ,
,.)
/
"
'
? ?
A'
( '- .'l
. .? .) . , --.
; 1, A% ' , \ ? " --- ,
.., ... ),
.. \ ? N ....,.
. .,... - ?--- - ... SATIAZ ,89t4).?863,952? (1975)- - / -
,,-,-: ?\ ' , ,
. --i,,, ,
/,? ,, .
...?, ,
) .
r- ? -?
-
-
L ATOMH;d1
(ATOPANAY A ,gNE,R9IY A)
)
\ 2 \-, ''' -.?, 0 [. ? H . N.
; /7 , y,
\ 7
' :/ ''?'' , , \. j)
--- ' '' , , ,
TRANSLATED FROM, RUSSIAN, -
? ; -?, ..
, ,. -!.,
, ? , . ,.... ?
. , / ? \--
-\ ), I , (--, . -?
? L / c
,'.; -
% (
,! ,
, . ' '-
, - , >-..,\,' _.. -\ :c
ic /
/
1 ) ' '"N k -
'N '4)
' - ) ? I -, ??
,-
\?-
,
,
-CONSUIO
PTS\ BUREAU NEW YORK'
\ ),
A
? -
/ ? ,
Declassified and Approved For Release 2013/09/24: bIA-RDP10-02196R000400060603-2
Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060003-2
1, , , ,, ; , , - :,7 , I.', --\ -,1,,,, / , - ' - - rz?,,,,,_ ?I i ._,---
,-- ----
,\_
,'-
`i i
, J j
, Soviet Atomic'Enerby is a co-ver-to-cover translati6n-of Atomnayq, .
,
Et-lrgiya, a?615;lic.ati,dn 8t,theYAcademy of,S0en.cee of the USSR. -----', ?
. - ,
r
?, .
-
, -
Soviet Atomic Energy is' abstracted of in-
dexed in Applied-Mechanics Reviews,' Chem-
ical Abstracts, Engineering Index,,INSPEC?
Physio's Ahsbacts and :Electrical' and ,Elec-
tronici (-Abstracts, ;Current Contents, and
-:-Nualear Science Abstracts,
....')
-- Ati, agieement with-thedopyright AgericY:oi'thi USSR (VAAP)
,
, , ,
, tnakevavallable both advance copies ofithe Russian fOurnal and ?.
? deiginal. glassy Photograrphs an4:.a,rtwork-.7This-erves to deCre-*
._ the -necessary time lag\between publication -ollthe original anel...`, .
publication of 'thetranstatiOn andhelps.loirnprovi theflyatity .
of, the latter. The "t'ran'slation .begari with, the first issue of the
- - tRussianjournal.
, . .., . , - --,--\-:,...;' :,, ? - -_-,
17
. r.-- 2 ?,' , , - - - - /
Editorial Board Of,Atomn\aya Energiya:-i -- -
( , _ 7- , ,. _ 1 ,
- . ,.,, .
. ,
--- i
? Editor: MI D. Millionahchikov L
- - `P? . +
P , -
DeputY 6Irectof
- ? -: I. ?V. Kurchatpv 'Instltute'of.,AfornieEner0
-
h-._' f ,;? Academy of Sciences of the USSR.,
\
? _ - Mopcow, USSR r ,?
) ' 1 -- i .
' (----, s 5 .c- .,
---', AisOciate Editor: N. A-VlasciV
k.,
, ? ,
1 I5. k: ? ' (5A
'
A. A. Bochvar ?, ?
,
- N. A,. Dollazhar
" X/..--S.:Fiirsov,
I. N bolOvirw -
? --"I', , , \IF. Kg1I0in, )
sTr-
)
A. I.,-Krasin ,.
*
,
, -
\ ,
.\
7- I-
V. V. Matveev;
')
,
m: MeshcheryakoV
? ? -
/ P. N. Palei
-
-
Shevchenko-
'
_
P. Vinogradov
r-
Ap:zefirv - '
?
---.. I
7 ../i ' . , ?
Copyright? 1976 Plenum ,P6
,F'ublshing Cor..poralion, 227 West-17th Street/ New', York,
' - W.Y. 10011. All rightS\ reserved. No article containecNherein 'may be,reprouced, 7 .,'''
stored in a retrievallsystem, or transmitted, in ,any form or by any meakielecfr'onic, -
--- mechanicel,'Phot-POopying; microfilming, rec.cird'ing or otnerWise,. vvithdut written ?
A - A- '-.--
?
/Permission' of the publisher. ' . ' / ? ,
.. )
7, .. \ , , . , ?,..? ' .._._? \ . _..,.. .
r /
Consultants Bureaujournalt appeai- about six-months after5the publication of the
' original ,Russian issue. For ;bibliographic accuracy, the Er'1611sh :issue publihed Vi'L
/ -)'Const.iltant Bureau cari'les.the-tame number anci`clate as the'o4inal.hussian from
7 i - ,/ _.
whi'ch t was translated. FOr-example, a Russian issue publisheein-December will '.
appear in a'.COns`ultants.Bureau' English translation a6Pul-the following ..1-Zine, but the ? ,(
tranda0on issue will carry the (5ecembe5date.'When ordering any volume or particu- ,--? /
_
lar issue/of .,e_ Coni"urtants Bureau' :journal, plea'se specify the date and, Where appli-
/cable.. the volume and !issuefew York, New-York 160,11
-
V,
,
\ .
/ - r, 7 !
7 , 1
?
Published monthly. 'Second-class-postage-paid at Jamaica ;? New York- 11431.
f
5' -7-` 7 \' ./ '\.' \ ' -7--'' 1 ' - '
)
? )- , . )
. . ._ , . ? , -%
. .
. ?.', ..,..._ ,.,..(..--,, ? ,- 1- ,
''-. ). /' Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060003:2 "
Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060003-2
SOVIET ATOMIC ENERGY
A translation of Atomnaya Energiya
April, 1976
Volume 39, Number 4 October, 1975
CONTENTS
Engl./Russ.
ARTICLES
Radiophotochemistry as a Possible Basis for the Efficient Use of Two-Purpose Reactors
? V. I. Goltdanskii
863
243
Experimental Double-Fuel-Element Channel for the Reactor of the World's First
Nuclear Power Station? M. G. Bul'kanov, V. A. Kurov, V. D. Lazarevskaya,
V. G. Potolovskii, and V. S. Sever'yanov
867
247
Statistical Characteristics of the Temperature Fluctuations in a Direct-Flow
Sodium?Water Steam Generator ? B. V. Kebadze, V. S. Sroelov, B. V. Kul'min,
and A. I. Gavrilin
870
250
Analysis of the Conditions Required for Producing Highly Purified Carbide Fuel by
Reducing Oxides with Carbon, and Study of the Behavior of Oxygen and Carbon
Impurities in the Presence of Fission Products ? R. B. Kotel'nikov,
V. S. Belevantsev, S. N. Bashlykov, G. V. Titov, V. A. Zelyanin,
and A. M. Anuchkin
874
255./
Simulation of the Formation and Annealing of Defects in Regions of Damage Created by
Collision Cascades in Alpha Iron ? P. A. Platonov, V. F. Krasnoshtanov,
and Yu. R. Kevorkyan
879
260
Determining the Proportion of Plutonium Nuclei Undergoing Fission during the Burn-Up
of Slightly Enriched Fuel? K. I. Zykov and 0. A. Miller
884
265
Technological Sensitivity Factors in Atomic Electric Power Plant Optimization
?A. M. Kuz'min
888
269
Use of Internal Direct-Charge Detectors as Input Elements of Automatic Reactor Control
Systems ? M. G. Mitel'man, L. G. Andreev, I. V. Batenin, B. G. Dubovskii,
V. A. Zagadkin, V. F. Lyubchenko, K. N. Mokhnatkin, N. D. Rozenblyum,
V. S. Sever'yanov, V. B. Tregubov, Yu. M. Shipovskikh and A. I. Shtyfurko
892
272
An Investigation of the Transition Effect in Layered Absorbers
? V. I. Vittko, I. A. Grishaev, and G. D. Kovalenko
895
275
DEPOSITED PAPERS
Phase Extension of a Relativistic Electron Bunch Interacting with a Resonator in a
Storage Ring ? S. G. Kononenko, N. I. Mocheshnikov, and N. N. Naugol'nyi
899
279
Reliability of Electron Accelerators for Radiochemical Devices ? Yu. D. Kozlov
900
280.
Spectral Distribution of the Albedo of 137CS 'Y. Radiation for a Two-Layer Medium with
a Cylindrical Interface ? Yu. B. Davydov, A. A. Timonov, and A. V. Davydov
901
281
Optimization of the Health Protection Zone and Shield Parameters for Accelerators
? Yu. A. Volchek and A. Ya. Yakovlev
902
281
LETTERS
Elemental Analysis of Boron Carbide and Initial Components by Proton-Induced X Rays
? A. G.Strashinskii, G. K. Khomyakov, N. A. Skakun, N. V. Serykh,
and I. T. Ostapenko
904
283
Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060003-2
L
Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060003-2
CONTENTS
Cross Sections and Resonance Integrals for the Fission of 239Pu, Am, Cm, and 249Cf
(continued)
Engl./Russ.
? K. D. Zhuravlev, N. I. Kroshkin, and A. P. Chetverikov ? ? .
907
285
Instrumental Neutron Activation Analysis of Geological and Biological Objects Using
a Computer? V. B. Zlokazov, L. P. Kul'kina, and 0. D. Maslov
909
286
Determination of Manganese in Aluminum by Neutron Activation with a 252Cf Source
? K. Sailer, Sh. Darotsi, Sh. Nati', P. Raich, I. Chikai, and L. Gergei
911
288
Some Results of an Examination of a Dismantled RK-L Hot Loop
? D. M. Zakharov, V. V. Gavar, A. S. Dindun, and M. M. Kramer
913
290
Ratio of Radiative Capture and Fission Cross Sections for 239Pu at Neutron Energies of
0.2-30 keV ? A. A. Bergman, A. N. Medvedev, A. E. Samsonov, and V. A. Tolstikov
916
291
Evolution of Hydrogen from Proton-Irradiated Construction Materials
? Yu. P. Vasin, A. G. Zaluzhnyi, D. M. Skorov, and 0. M. Storozhuk
919
293
Cross Section for the Fission of 240Pu by Neutrons from a Nuclear Explosion
? E. F. Fomushkin, E. K. Gutnikova, G. F. Novoselov, and V. I. Panin
921
295
CONFERENCES AND CONGRESSES
Problems of Increasing the Service Reliability of the Metal in the Power Equipment of
Nuclear Power Stations ? L. M. Voronin and E. P. Karelin
923
297
Second All-Union Radiogeochemical Conference ? R. P. Rafarskii
925
298
All-Union Conference on the Chemistry of Neptunium and Plutonium? A. M. Rozen .
929
301
The Third All-Union Conference on Linear Charged-Particles Accelerators
? I. A. Grishaev
932
302
The 2nd International Symposium on Nuclear Electronics ? A. N. Sinaev
934
303
Principal Results of the April Session of the International Commission on Radiological
Protection (ICRP) ? A. A. Moiseev
937
305
The Third European Congress of the International Radiological Protection Association
(IRPA) ? V. N. Lystsov
939
306
Isotope Ratios as Indicators of Radionuclide Sources and Environmental Migration
Paths ? R. M. Aleksakhin
942
307
EXHIBITIONS
"Inventions and Branch Licensed Products" Exhibitions in the "Atomic Energy"
Pavilion at the Exhibition of Achievements of the National Economy of the USSR
? B. A. Sokolov and E. A. StrePnikov .
944
308
NEW INSTRUMENTS AND TECHNOLOGICAL PROCESSES
LBK fl Potassium Concentration Meter ? L. V. Matveev, 0. G. Mikhailov,
and E. A. Strel'chenko
946
309
A New Method of Regeneration of Traps for the Purification of Alkali-Metal Coolants
? L. G. Volchkov and F. A. Kozlov
948
310
BOOK REVIEWS
D. Bedenig. Gas-Cooled High-Temperature Reactors ? Reviewed by Yu. L Koryakin . .
949
311
Msdrgs6rit Pavelscu. Numerical Methods for the Calculation of Nuclear Reactors
? Reviewed by S. M. Zaritskii
951
311
The Russian press date (podpisano k pechati) of this issue was 9/25/1975.
Publication therefore did not occur prior to this date, but must be assumed
to have taken place reasonably soon thereafter.
Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060003-2
Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060003-2
ARTICLES
RADIOPHOTOCHEMISTRY AS A POSSIBLE BASIS
FOR THE EFFICIENT USE OF
TWO-PURPOSE REACTORS
V. I. Gol'danskii UDC 621.039.335
The recent extensive development of nuclear power has given rise to the extremely important problem
of creating two-purpose reactors producing both electrical power and chemical products. Reactors of this
kind may be considered with some justification as the basic installations of future large-scale radiochemical
production [1-4]. A number of fundamental ways in which two-purpose reactors may be used for chemical
production have been discussed in the literature, and these we shall now consider.
Chemonuclear Synthesis. This version is the most direct method of converting nuclear energy into
the energy of endothermic chemical processes. The five stages involved in the trivial use of a nuclear
reactor as an energy source for chemical production are replaced by a single stage (Fig. 1). The chief
part of the fission energy ? the kinetic energy of the fragments retarded in the reaction mixture (the pro-
portion of fission energy carried away by the fragments f = 0.85) ? is in this case directly harnessed for
chemical requirements. The mechanism underlying the conversion of the fission fragment energy into
chemical energy in the gas phase is purely radiochemical; in the condensed phase this effect may be sup-
plemented by the thermochemical action of the fragments [5, 6], associated with the formation of micro-
scopic regions of brief local heating along the tracks of the fragments. The radiation yield of the trans-
formations taking place under the influence of the fragments lies in the range G 1-10 for ordinary (not
chain) reactions, where G is the radiation yield (the number of fully reacting molecules per 100 eV energy
of the ionizing radiation), and differs little from that encountered in other forms of radiation [7].
The necessity of minimizing the primary radioactive contamination of the products of chemonuclear
synthesis (evenbefore subjecting these to specialpurification) compels us to use special highly dispersed
chemonuclear elements or cells ("chels") as nuclear fuel; these have a certain optimum pore size, en-
suring the liberation of a high proportion of the fission-fragment energy in the interior of the reagents,
the fission fragments then emerging from the latter and being retained in the walls of the chemonuclear
cell. This acts as a kind of self-filtration of the chemonuclear cell from radioactive contaminants [8-14].
The proportionp ofifission energy transferred to the reagents under optimum conditions is 0.3 [9].
Allowing for all these factors, the yield of chemonuclear-synthesis products for one act of fission of
the nuclear fuel (E ? 200 MeV) is approximately 2 ? 108(G/100)fp 5? 105 G. Here and subsequently we
shall take the fixation of nitrogen (M = 28) as a typical example; taking G = 1, we then obtain a yield of
Wchn 60 kg/g U for the chemonuclear synthesis.
However, remembering that after the elapse of a time t 10 sec from the instant of uranium or plu-
tonium fission the total activity of the fission fragments of 1 g of nuclear fuel equals 3.1011 t-1,2 Ci [15],
it is easy to see that in the absence of further purification the activity of the chemonuclear synthesis prod-
ucts will be 5.106 t-192 Ci/g, i.e., in four or five days Actin ? 1 Ci/g.
Requirements imposed upon the purification of chemonuclear-synthesis products are extremely strict
(purifications of the order of 106-1012 times are demanded); this makes the technology of chemonuclear pro-
cesses more expensive and (even more important) makes them partially or even entirely incompatible with
the elimination of pollution from the environment.
Translated from Atomnaya Energiya, Vol. 39, No. 4, pp. 243-246, October, 1975. Original article
submitted March 4, 1975.
?1976 Plenum Publishing Corporation, 227 West 17th Street, New York, NY. 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 $15.00.
863
Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060003-2
Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060003-2
8
Light
energy
Nuclear
ener
2
Thermal
gy
Alienergy
Chemical
energy
Thermal
energy
3
Mechanical
energy
Electrical
energy
Mechanical
energy
7
Fig. 1. Various ways of converting nuclear into chem-
ical energy: 1-2 ?3-4 ? 5-6) ordinary way; 1 ? 6)
chemonuclear synthesis; 1 ?2 ? 6) high-temperature
fuel element version; 1 ?8-6) radiophotochemical syn-
thesis.
Synthesis Using High-Temperature Reactors. The two-stage conversion of nuclear energy into chemi-
cal energy via a thermal stage (Fig. 1) was proposed and examined in relation to the oxidation of nitrogen
in [16]. Here it is essential to use high-temperature fuel elements; however, an extremely important ad-
vantage of this version is that there is no direct contact between the fission fragments and the reagents, so
that the radiochemical purity of the products of the chemical processes is assured. The yield of combined
nitrogen for fuel-element temperatures of 1900-2500?C is (according to the estimates of [16]) w
-temp ? 200-
360 kg/g U (Wtemp increases further in this case as the result of a chain reaction in the oxidation of nitro-
gen). A high yield of the desired products is achieved by the successful combination of high-temperature
kinetics and the rapid cooling of the reaction products, i.e., the quenching of the favorable high-tempera-
ture equilibrium. Unfortunately at the present time there are no fuel elements capable of prolonged opera-
tion without failure at can temperatures of 2000?C or even lower temperatures. Chemical synthesis in
high-temperature nuclear reactors thus remains for the time being a purely speculative version.
Synthesis under the Action of Mixed ny Radiation from Nuclear Reactors. Even in the case of special
reactor constructions ensuring the maximum use of mixed ny radiation for chemical purposes, the power
of these radiation sources is no greater than 1-6% of the thermal power of the reactor [1]. Hence the yield
of the desired product is only Wiry' (2.5-15) kg/g U, i.e., far lower than in chemonuclear synthesis.
The danger of radioactive contamination (activation) of the synthesized products of the radiochemical
processes nevertheless remains (although to a less serious extent than in the case of chemonuclear syn-
thesis); this contamination often includes radioactive isotopes of elements entering into the composition of
the products which are incapable of being separated chemically.
Averaging the data presented in [1] we obtain the following relationship (typical for the case of mixed
ny radiation), being the integrated dose of irradiation and the integrated thermal neutron flux: F/D ? (2-3) ?
1014 neutrons/(cm2?Mrad).
It is thus easy to see that the formation of every gram of the desired product in typical nitrogen- and
hydrogen-containing systems in a field of mixed ny radiation with a radiation yield of G 1 will be accom-
panied by the accumulation of ?10-1 Ci of I4C and ?2 .10-2 Ci of tritium distributed between the original
reagents and the radiochemical-synthesis products, in proportion to their contents of nitrogen and hydrogen
molecules.
Although these activities are many orders of magnitude smaller than those encountered in chemo-
nuclear synthesis, they are still quite unacceptable. It is clear that the large-scale manufacture of chemi-
cal products involves the creation of serious radioactive contaminants, the necessity of reliably separating
these from the chemical products, and necessarily a high degree of purification of the original reagents
from impurities liable to be severely activated by thermal neutrons (for example, halogens, sulfur, a large
number of metals, and so on). All this may make the technology of chemical synthesis under the action of
mixed ny radiation from nuclear reactors more expensive, so greatly limiting its practical value.
864
Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060003-2
Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060003-2
Synthesis in Radiation Circuits. According to the results presented in [4], the power drawn from re-
actors in the form of the activity of the radiation circuits may be estimated as -1% of the electrical power
of the reactor, which amounts, for example, to -108 g-eq ?Ra for an electrical power of 100 MW or a
thermal power of 100/K MW, where K is the thermoelectric energy-conversion efficiency. From this we
may estimate the yield of end product as Weirc GK ? 1.7 kg/g U, which, although well below even the
mixed n7-radiation output, nevertheless has a considerable qualitative advantage (the products contain no
radioactive contaminants). Hence the use of radiation circuits is at the present time the most promising
version of the radiochemical use of nuclear reactors [4].
Radiophotochemical Synthesis. One further possibility of two-stage transformation remains promis-
ing, that of nuclear ? light ? chemical energy (Fig. 1). In this version the fuel elements act as photochem-
ical elements or cells ("fels"); the fission fragments are retarded in specially selected (solid or gaseous)
scintillators, while the light so created is focused by a set of reflectors into the ends of glass-fiber or
other light guides, which bring this light into the zone of photochemical transformation. The yield of the
end product Wphch - 18,000 4:11a/E kg/g U, where (13 is the fragment/light energy conversion coefficient
(the so-called light yield), E is the energy of the light quanta in eV, a is the quantum yield of the photo-
chemical reaction, allowing for the incomplete focusing of the light into the photochemical zone (in the
present case there is no need to concern oneself with stopping the fission fragments in the interior of the
scintillator, so that the factor p encountered in Wchn is no longer required). There are certain scintilla-
tors for which 41 reaches -0.1, for example, c13. 0.08 in NaI(T1) and (1) 0.15 in CaI2(Eu), the value of .13
not depending on the ionizing capacity of the particles but being the same for both p particles and fragments
[17]. For E 3 eV and a - 0.1 we then have Wphch^- 60 kg/g U. Of particular interest for radiophotochem-
ical synthesis are gaseous (rather than solid) scintillators of the xenon type, possibly with the addition of
light converters for displacing the center of the scintillations from the ultraviolet to the visible region at
pressures amounting to tens of atmospheres.
As in the case of synthesis with the aid of radiation circuits, the products of radio photochemical syn-
thesis contain no radioactivity. The yield of these products may, in fact, be much greater than that of the
circuit version, and (a factor of no mean importance) the perilous necessity of removing radioactive ma-
terials (activity carriers in the circuits) from the active zone of the reactor is avoided (this procedure is
replaced by the collection and transmission of light).
Naturally it is desirable to ensure selectivity of the chemical transformation, and the preferential
yield of a prespecified product. This may be done more easily in a photochemical transformation than in
a radiochemical process, for example, by choosing the appropriate light converters so as to optimize the
light of the radiophotochemical sources as regards intensity and spectral composition. We should take
note of some important initial successes in the conversion of fission fragment energy into not simply
monochromatic but also coherent light, involving the use of nuclear reactors to pump lasers [18, 19], the
efficiency of the nuclear/laser energy conversion reaching several percent.*
Of course there are a number of problems in which photochemical interaction cannot replace radio-
chemical processes (as in chemical transformations, or the modification of materials opaque to visible light
or other radiations close to the optical region); however, there may also be processes in which radiopho-
tochemical synthesis may prove to be the best method of using nuclear energy for chemical purposes and
may provide a firm basis for the operation of the most economic two-purpose (power and chemical) nuclear
reactors.
Investigations into the following problems of radiophotochemical synthesis are thus extremely vital:
1. The choice of systems ensuring the maximum conversion factor for the transformation of kinetic
fission-fragment energy into light energy (it is desirable to obtain a monochromatic or bright-line light
spectrum of photoradiation origin, the radiative pumping of lasers being of particular interest).
2. Provision for the most efficient method of collecting and transporting light from the active zone
of the reactor (this requires, in particular, a high radiation resistance of the scintillator, which should
clearly be of the gas type), and also the transportation of the whole system of light reflectors in the active
zone and the initial section of the light guide from the active to the photochemical zone.
*We are not here considering other aspects of interaction between laser and radiation-based processes,
such as the possibility of increasing the radiation resistance of chemical compounds or improving the se-
lectivity of radiolysis by using laser beams as a background to the irradiated substances [20, 21].
865
Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060003-2
Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060003-2
3. Choice of the best photochemical processes, with the allowance for the requirements of the chem-
ical industry, the emission spectra of scintillators severely irradiated with fission fragments, and also the
possibility of using various sensitizers and spectral converters.
LITERATURE CITED
1. A. Kh. Breger et al., Fundaments of Radiochemical Apparatus Construction [in Russian], Atomizdat,
Moscow (1967), p. 500.
2. V. I. Golidanskii, Vestn. Akad. Nauk SSSR, No. 9, 62 (1970).
3. Ya. M. Kolotyrkin and V. L. Karpov, Zh. Vses. Khim. Obshch. im. D. I. Mendeleeva, 18, 242
(1973).
4. A. Kh. Breger et al., ibid. , p. 312.
5. V. [Goldanskin and Yu. Kagan, Intern. J. Appl. Rad. and Isotopes, 11, 1 (1961).
6. E. A. Borisov, L. A. Bulanov, and E. V. Starodubtseva, Khim. Vys. Energ., 4, 550 (1970).
7. V. I. Gordanskii et al., At. Energ., 30, No. 3, 262 (1971).
8. P. Harteck and S. Dondes, Nucleonics, 14, 22 (1956).
9. M. Steinberg, Chem. Engng. Progr. , 62, 105 (1966).
10. M. Beller, L. Epel, and M. Steinberg, Chem. Engng. Progr, Symp. Ser., 63, 71 (1967).
11. V. I. Gol'danskii et al., Izotopy v SSSR, No. 12, 7 (1968).
12. B. G. Dzantiev et al., At. Energ., 29, No. 2, 71 (1970).
13. B. G. Dzantiev, A. K. Krasin, and V. T. Kazazyan, At. Energ., 33, No. 6, 803 (1972).
14. E. A. Borisov and V. D. Timofeev, Zh. Vses. Khim. Obshch. im. D. I. Mendeleeva, 18, 323 (1973).
15. S. Glasstone and M. Edlund, Fundamentals of Nuclear Reactor Theory [Russian translation], IL,
Moscow (1954), p. 460.
16. Ya. B. Zel'dovich, V. I. Gol'danskii, and B. G. Dzantiev, in: Fixation of Atmospheric Nitrogen [in
Russian], GIAP, Moscow (1963), pp. 35-43.
17. V. V. Matveev and B. I. Khazanov, Instruments for Measuring Ionizing Radiations [in Russian],
Atomizdat, Moscow (1967), p. 697.
18. L. I. Gudzenko, I. S. Slesarev, and S. I. Yakovlenko, Preprint 109/1974 of the P. N. Lebedev
Physical Institute, Academy of Sciences of the USSR.
19. T. Wimett, H. Helmick, and R. Schneider, Science News, 106, 229 (1974).
20. V. I. Gol'danskii, Khim. Vys. Energ., 9, 78 (1975).
21. V. Golidanskii, Intern. J. Radiat. Phys. Chem., 7, 339 (1975).
866
Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060003-2
Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060003-2
? EXPERIMENTAL DOUBLE-FUEL-ELEMENT CHANNEL
FOR THE REACTOR OF THE WORLD'S FIRST
NUCLEAR POWER STATION
M. G. Bultkanov, V. A. Kurov,
V. D. Lazarevskaya, V. G. Potolovskii,
and V. S. Severtyanov
UDC 621.039.5
The construction of new types of fuel elements requires that reactor tests be carried out on both the
fuel components and on experimental half-scale fuel elements. The channel being described is intended
for verifying, under the conditions of the reactor of the world's first nuclear power station, new technical
solutions for the fuel elements of the Beloyarsk nuclear power station and the Bilibinsk nuclear heat and
electric power plant. The positive experience accumulated on the basis of tests of unique experimental
fuel elements has enabled an experimental channel to be constructed which, when loaded with fissile ma-
terial, is identical with the regular-fuel-element channel and can be used in place of it.
The channel is designed for a regular cell of the reactor of the world's first nuclear power station
and has the same subcoupling dimensions. In the construction of the channel, certain improvements are
taken into account which were adopted in the channels of the Beloyarsk nuclear power station and the Bili-
binsk nuclear heat and electric power station, and directed at increasing their efficiency.
The experimental channel (Fig. 1) consists of two fuel elements, descending and ascending tubes,
installed in metal and graphite sleeves, which form a cylinder with diameter 64 mm for the graphite sleeve
and 63 mm for the steel components. The length of the channel is 6.6 m.
The coolant from the conduit reaches the dispensing collector through a connecting tube and an inlet
central connecting pipe, and enters the pressure chamber located in the lower part of the channel through
two descending tubes with diameter 12 x 0.6 mm. The coolant passes upwards, washing the inside surface
of the fuel element tubes, from the chamber through spiral tubular compensators, which take up the tem-
perature expansions between the descending and ascending tubes. Later, the coolant enters the collecting
vessel and leaves the channel through a side connecting pipe. The dimensions of the fuel elements are:
TABLE 1. Results of Fuel Element Tests in the Experimental Channel
Cell
Cell
designation
Date of in-
sertion (1970)
Date ofremoval
Bumup, kg/t
Av . heat flow,
,
106 kcal/m-
Temp.of fuel
lement clad-
ding, ac
05-14
DTM-87
22.V
12.1.74
20,3
1,31
400
05-08
DTM-88
10.V1
12.1.74
18,5
1,13
380
05-18
DTM-90
6.IX
12.1.74
17
1,36
420
09-16
DTM-92
29.VI
12.1.74
17,1
1,56
430
12-19
DTM- 93
10.VIII
12.1.74
16,4
1,16
390
08-15
DTM- 95
10.VIII
12.1.74
18,1
1,7
440
11-12
DTM- 96
3.X
12.1.74
17,2
1,3
400
09-18
DTM-86
21.V
15.111.72
6,5
1,5
430
04-11
DTM-85
21.V
7.X.73
17
1,23
390
01-12
DTM-84
21.V
7.IV.72
5,1
1,19
390
06-13
DTM- 94
13.VII
6.1V.72
5,3
1,5
430
13-10
?DTM-91
28.VI
7.X.73
13,1
1,1
380
08-19
DTM-89
18.VII
7.X.73
16,5
1,0
360
02-11
DTM-99
11.1X
Being tested
15
1,21
. 390
10-05
DTM400
8.IX
Being tested
14,5
1,16
390
03-20
DTM-97
8.1X
6.1V.72
4,95
1,35
410
05-04
DTM-98
11.IX
Being tested
16,7
1,17
390
Translated from Atomnaya Energiya, Vol. 39, No. 4, pp. 247-249, October, 1975. Original article
submitted October 7, 1974.
?1976 Plenum Publishing Corporation, 227 West 17th Street, New York, NY. 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 $15.00.
867
Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060003-2
Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060003-2
A-A
(A%
diameter of the surrounding cladding 20 x 0.2 mm and inside tube
12 x 6.6 mm.* The compensators are made of tubes with diameter
9.4 x 0.6 mm and have a spiral pitch of 30 mm.
Below the distributing and collector vessels are assembled the
metal plugs of the biological shield, with openings 12x 0.6 mm
diameter below the tube. The lower plug is braced to the outside
jacket, made from 63 x 1.5 mm tubing. The jacket is connected
with the channel head and is a load-carrying structure for the plugs
of the biological shield. The lower part of the descending tubes
and the fuel element are enclosed in graphite sleeves. The opening
in the center of the sleeve, with diameter 10 mm, is designed for
emergency withdrawal of the channel from the reactor cell. In the
zone of the compensators, graphite sleeves are arranged for cen-
tering them.
In the experimental dual-fuel-element channel, with ther-
mometry of the outside cladding of the fuel element, additional an-
nular recesses are provided in the graphite sleeves for positioning
the bracing units of the thermocouples and openings for their lead-
out. A special packing gland is provided at the channel head for the
thermocouple lead-out. There is also a fitting for the pressure
bleeder.
? Chromel? Alumel cable-thermocouples, type KTMS, were
used, of size 2 x 0.06 mm, and which had shown excellent efficiency.
Measurement of the temperature of the thin-walled cladding of the
fuel elements is one of the most complicated types of reactor ther-
mometry. Most frequently, a bracing is used for the hot junctions
of the thermocouples by means of clamping or bimetallic rings ?_
10 j this is a unique method of avoiding damage to the fuel-element
cladding (Fig. 2). A considerable drawback of this method is the
loss, in the course of time, of the clamping properties of the ring
which leads to breakdown of the contact between the cladding and
the thermocouple hot junction, and to the appearance of an error in
the thermocouple readings.
The channel components are made from grade 12Kh18N1OT
stainless steel. For avoiding cracks, thin-walled tubes with diame-
ter 12 x 0.6 and 9.4 x 0.6 mm were welded only with butt seams.
In order to form the necessary space for the radial expansion of
the fuel elements during swelling and for reliable monitoring of the
leak-tightness of the fuel elements, the gaps between them and the
graphite sleeves are 0.5 to 0.7 mm (at the side). Heat from the graphite stack is removed largely through
descending tubes, disposed in the graphite sleeves and with a minimum gap of about 0.15 mm. Part of the
heat from the graphite is also transferred to the fuel elements.
In order to reduce the corrosive action of the reactor medium on the fuel-element cladding and tubes
of the channel water circuit, provision is made for purging the channel with nitrogen. In order to ensure a
reliable purging of the entire channel cavity, the compensation zone is enclosed on the outside with a tube
of diameter 63 x 1.5 mm, which has three openings 20 mm in diameter for outlet of the gas. Gas gaps are
used also for monitoring the leak-tightness of the outer cladding of the fuel elements and the channel tubes,
and this is effected by pumping out gas through the gaps and measuring its radioactivity. A connecting pipe,
located in the head of the channel, is used for feeding in the gas. In channels with temperature measure-
ment of the outer fuel-element cladding, this connecting pipe is located at the thermocouple outlet joint.
The principal technical data and operating conditions of the standard and experimental fuel channels
are given below [1, 2]:
Fig. 1. Experimental double fuel
element channel: 1, 2) inlet and
outlet pipes for coolant of the first
loop; 3) channel head; 4) steel
sleeves; 5, 6) descending and as-
cending tubes; 7) graphite sleeves;
8) fuel element; 9) compensator;
10) rotary chamber; 11) thermo-
couple outlet assembly; 12) pres-
sure bleeder pipe.
*Channels of this design allow fuel elements with maximum external diameter of up to 25-26 mm to be
tested in the regular reactor cells (diameter 65 mm).
868
Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060003-2
Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060003-2
1 3 1 2 4
Weld clamp
Fig. 2. Methods of securing the regulus of a thermocouple by means of a
shoe (a) and a strip of foil (b): 1) thermocouple; 2) clamping ring; 3) shoe;
4) strip of foil.
Working pressure, kg/cm2
. ?
100,
100
Water temperature, ?C
at inlet
at outlet
180-190,
270-309,
180-190
270-309
Coolant flow rate, t/h
1.9-2.4;
1.2-1.5
Average thermal flow from fuel element to water,
kcal/m2?h
(0.8-L8) ? 106,
(1-1.7) ?
106
Maximum temperature of outer cladding of fuel
element, ?C
450,
450
Uranium charge in channel, kg
4,
4
Number of fuel elements, pieces
4,
2
Active length of fuel element, mm
1700;
1700
Outside diameter of fuel element, mm
14 x 0.2,
20 x 0.2
Diameter of central tube of fuel element, mm
9 x 0.4,
12 X 0.6
Quantity of steel per channel, cm3/cm
1.05,
1.2
The design of the channel permits the fuel elements to be tested under cavity boiling conditions, which
is achieved when the channel is installed in the PV-2 experimental loop [2].
The operating parameters of the experimental channel under boiling conditions are shown below:
Pressure, kg/cm2 130
Water temperature at inlet, ?C 250
Coolant flow rate, t/h 1.2
Channel power, kW 170
Steam content on exit from channel, To 7
Maximum temperature of outer fuel element cladding, ?C 470
Tests of the channels under cavity boiling conditions were conducted with a steam content at the out-
let of up to 7% by weight, which corresponded to the target of the experiment.
The main part of the experimental dual-fuel-element channels (see Table 1) was loaded into the reac-
tor of the world's first nuclear power station in 1970. It can be seen from Table 1 that the fuel burnup in
the fuel elements, the greater heat flows and the higher temperatures of the fuel element claddings confirm
the operation of the experimental channels in more rigorous conditions by comparison with the standard
fuel channels. The dual-fuel-element channels have operated in the reactor for about four years and are
continuing to be used up to now. At present, fuel elements with heat exchange intensifiers are being tested
in the experimental channels, with increased diameters and with a greater fuel charge.
LITERATURE CITED
1. G. N. Ushakov, The First Nuclear Power Station [in Russian], Gosenergoizdat, Moscow ? Leningrad
(1959).
2. Ten Years of the World's First Nuclear Power Station, SSSR [in Russian], Atomizdat, Moscow (1964).
869
Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060003-2
Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060003-2
STATISTICAL CHARACTERISTICS OF THE TEMPERATURE
FLUCTUATIONS IN A DIRECT-FLOW SODIUM? WATER
STEAM GENERATOR
B. V. Kebadze, V. S. Sroelov, UDC 621.039.534.63:621.039.534
B. V. Kul'pin,* and A. I, Gavrilin
The temperature fluctuations in heat-transfer systems produce thermal stresses and fatigue damage
on prolonged operation. Considerable temperature fluctuations can arise in steam boilers at the start of
the zone of deteriorating heat transfer on account of the alternating contact with liquid and steam [1]. The
turbulent flow of the heat carrier in the presence of a considerable temperature gradient also produces
temperature fluctuations which penetrate into the wall [2]. Finally, temperature fluctuations can also arise
to some extent on account of the stochastic nature of the boiling, with random production and detachment of
bubbles at the heated surface. Although there have been several studies of temperature fluctuations, it is
not at present possible to estimate in advance the magnitude of the fluctuations for a particular apparatus,
mainly on account of the difficulty of incorporating thermal and hydraulic characteristics of any particular
design.
The basic purpose of statistical experiments is to determine the sources of the pulsations, the am-
plitude characteristics, and the characteristic times. The measurements were made with a model of a
steam boiler using two standard modules made in Czechoslovakia. The body was 159 x 7 m and length 6 m,
which contained 19 tubes of diameter 8 x 3, which were fixed in double tubular assemblies. The thermal
power of the testbed was 600 kW, but this did not allow us to produce normal mass flow speeds for the water
in the 19-tube module. We therefore left only six working tubes, which lay at the edge of the tube bundle.
The other tubes were blanked off and served for monitoring purposes. This enabled us to make measure-
ments at flow rates in the range 400-850 kg/see-m2, but the sodium speed in the generator was below the
nominal value by about a factor of three. The tubes were made of NT8Kh6 steel. We mounted 50 microther-
mocouples in stainless-steel jackets of outside diameter 0.5 mm on one working tube at intervals of 190 mm in the
evaporator module near the outer surface of the body. The ther-
mocouples were contained in longitudinal slots of depth 1 ram
i
and introduced through the body. In the space between the
134 138 145 150 112 ? Steam tubes of the evaporator module, along an internal generator,
there were 17 thermocouples in stainless-steel jackets of
67
outside diameter 0.8 mm. Figure 1 shows the thermocouples
used in most of the statistical experiments. The sodium
flow rate was measured by an electromagnetic flowmeter
calibrated to 2%. The water flow rate was determined from
the pressure difference with a standard throttle using a dif-
ferential manometer DM of class 1.5.
In each working state (Table 1) we measured the tem-
perature distribution in the sodium and at the tube wall along
the length of the boiler, which went with the input and output
parameters of the sodium and water (steam) to provide a
Fig. 1. The steam generator with trans- basis for calculating the heat fluxes and heat-transfer coef-
ficients, for comparison with theoretical values.
=-44111111
111111
170 169 166 167 166
I 11110
1111" Water
113 114
1Na
ducers.
*Deceased.
Translated from Atomnaya Energiya, Vol. 39, No. 4, pp. 250-254, October, 1975. Original article
submitted October 3, 1974.
?1976 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 $15.00.
870
Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060003-2
Declassified and Approved For Release 2013/09/24 : CIA-RDP10-02196R000400060003-2
1
k
L464:-
0,6
4 6 710 2 10-1 100 f, Hz
Fig. 2
Fig. 3
Fig. 2. Autocorrelation functions for temperature fluctuations: 1) T138, a = 4?C; 2) T134, g =
1.4?C; 3) T134 ? T145; 4) T167, 0- = 2.4?C; 5) T166 ? T167.
Fig. 3. Spectral densities corresponding to the correlation functions of Fig. 2: 1, 2, and
4) as in Fig. 2; 3) modulus of the cross spectral density.
In the dynamic experiments, we measured primarily statistical characteristics of the thermocouple
signals near the crisis region. We used electronic potentiometer recorders with 2-mV scales to record
the temperature fluctuations.
The signals were amplified for statistical processing and converted to a pulse-frequency code for
magnetic tape recording. This recording was performed simultaneously on 4 channels. Specialized com-
puters (correlograph) calculatedthe mathematical expectation, dispersion, and autocorrelation functions,
while a digital computer calculated the spectral densities. Pulsations of elevated amplitude occurred near
the crisis region, which was first restricted to a length of 200 mm or so. The modes of operation were
chosen so that the deviations of the thermocouple signal from the crisis zone to both sides of the mean were
approximately equally probable. The pulsations remained stationary for 10-15 min. The precise mode of
pulsation away from the crisis zone was less dependent on the working parameters.
In all we performed over 30 runs, in which we varied the flow rates, water temperature, pressure,
and other parameters. We now consider some typical statistical characteristics. The normalized auto-
correlation functions of Figs. 2 and 3 (except for curves 4 and 5) relate to the first mode (Table 1). The
mode corresponds to a somewhat larger heat flux in the sections containing thermocouples 166 and 167.
Thermocouples 134, 138, 145, and 150 were mounted in the tube, while 166-170 were in the sodium. The
crisis zone lay near thermocouple 138. The fluctuations in the crisis zone had a standard deviation of
about 40?C and contained considerable damp component with a period of about 2 sec, which was almost ab-
sent at adjacent thermocouples. The periodic form of pulsation was observed in many runs. Preliminary
analysis showed that the reason may be some features of the water part of the system. For technical
reasons, only part of the steam passed through a condenser and cooler, while the rest was mixed with the
condensate directly ahead of the pump, which can lead to flow-rate fluctuations and temperature variations
in the crisis zone. To test this, the fluctuations in water flow rate were written along with the other sig-
nals to the magnetic tape. We analyzed the normalized autocorrelation and cross-correlation functions
for the flow rate and temperature in the crisis zone (Fig. 4, second mode, Table 1), and found a high
(about 0.5) correlation coefficient, which was determined as
R xy Mmaxi
X ?
V R xx (0) Rim (0)
TABLE 1. Working Parameters of Steam Generators
GI120,
tons/h
W0, kg ini2 .
sec
t 1 14, ?C
t 113, ?C
GN a, M3 ill
tin, ?C
t67, cC
qi.
kcal/m ? h
q+
1.
kcal/m .h
1,08
1,25
445
512
252
287
341
333
12,2
9,4
462
508
304
306
41,7
51,4
147
208
Note. Steam pressure 100 atm; qt and qi linear heat fluxes before and after heat-transfer crisis zone.
871
Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060003-2
Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060003-2
R
48
46
0,4
012
0
-42
-44
0
/
1 1 A i
1 1 ? '`.....,
% ?
?.2
1 1 1 1
2 3 4 5
Fig. 4 Fig. 5
Fig. 4. Autocorrelation functions for temperature fluctuations and water flow rate: 1)
T138, a = 6.3?C; 2) GB, a = 0.6%; 3) T138 ? GB.
Fig. 5. Distributions for the temperature fluctuations: 0) T138; e) T134; normal
distribution.
as well as agreement between the periods and time shifts in the peak for the cross-correlation function,
which indicates delay in the temperature with respect to the flow-rate variation. One can say that the lat-
ter variation is the external factor determining the character of the temperature pulsations. The measured
relative standard deviations in the flow rate were small (maximal value about 1.2% in all runs), but the tem-
perature fluctuations in the crisis zone became considerable (a = 9?C), which corresponds to a doubled
fluctuation amplitude of about 25?C. The amplitude of the temperature fluctuations increased as the fre-
quency was reduced and as the amplitude of the flow-rate fluctuations increased. The temperature fluc-
tuations in the crisis zone were anharmonic, as is clear from the content of higher harmonics in the spec-
tral density (curve 1 of Fig. 3). The distributions (thermocouple 138, Fig. 5) differ substantially from
normal, whereas the distribution for thermocouples outside the crisis zone was closely normal (thermo-
couple 134, Fig. 5).
The fluctuations in the thermocouple signals away from the crisis zone contained low-frequency and
high-frequency components (curve 2 of Fig. 2). The periodic component was weak even when the water
flow-rate fluctuations were present, for quite obvious reasons. The coefficient of heat transfer from the
steam to the tube was low in the region of deteriorating heat transfer, and the thermocouple in the wall
perceived in the main the fluctuations on the sodium side, whereas the thermocouple monitoring the steam
at the output produced a periodic component correlated with the flow rate. In the section before the crisis
region, the temperature in the tube was only slightly dependent on the flow-rate fluctuations.
We also used cross-correlation methods to detect the source of the temperature fluctuations in the
tube away from the crisis zone. Figure 2 (curve 3) shows the cross-correlation function for thermocouples
134 and 145, which lie one on each side of the crisis zone; this contains almost solely a time-of-flight
component, whose time shift agrees satisfactorily with the calculated transit time for sodium in that sec-
tion. The same applies to the cross-correlation function for series of thermocouples in sodium (curve 5 of
Fig. 2). Here the spectrum of the fluctuations was wider than that for the thermocouples in the tube (curve
4 of Fig. 2 and 3).
In measurements on the statistical characteristics of the sodium thermocouples it was found that the
dispersion decreased away from the sodium inlet, and hence it was much lower in the lower part of the
module. On the other hand, there was a marked rise in the dispersion on passing through the crisis re-
gion from the zone of deteriorating heat transfer to the region of dispersed annular flow, with the heat flux
increasing by about a factor 4. The standard deviations for sodium thermocouples 166, 167, 169, and 170
were 3.3, 3.0, 1.9, and 4.2?C, respectively,when the crisis zone lay between thermocouples 169 and 170.
The relatively high-frequency fluctuations recorded by the thermocouples in the sodium and in the
tube walls are due to motion of hydrodynamic nonuniformities of eddy type in the transverse temperature
gradient. The strength of the eddies is particularly high where there is a destabilized flow near perturba-
tion source (here the sodium inlet). The heat flux in the section between thermocouples 166 and 169 alters
872
Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060003-2
Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060003-2
-t.
2' 3 4 5
q ? 10-5, kcal/(m3.h)
Fig. 6 Fig. 7
Fig. 6. Standard deviation of temperature fluctuations in sodium as a function of
heat flux for Re =const=25,000: X) T167; 0) T169; 0) T170; A) Tim; A) T180.
Fig. 7. Correlation functions for the thermocouples in one cross section: T166
? T150.
only slightly, and the decreased amplitude of the temperature fluctuations is due to damping of the hydro-
dynamic perturbation away from the sodium inlet. The increased pulsation near thermocouple 170 is due
to the discontinuity in the heat flux. The sodium flow rate varied only slightly under the various conditions.
The standard deviation of the temperature in the sodium as a function of heat flux is roughly of straight-
line type (Fig. 6) for several thermocouples, and the fluctuations are larger for couples closer to the so-
dium inlet.
Figure 7 shows the cross-correlation function for thermocouples in the same cross-section at diamet-
rically opposite edges of the body (in the sodium and tube). The sharp fall to negative values for small time
shifts indicates that there are eddies comparable with the geometrical size of the steam-generator body
(Dgeom)? The large eddy size also explains the high correlation coefficient (0.4-0.5) for the thermocouples
separated by a considerable distance along the axis (L/Dgeom > 10).
Then the temperature fluctuations in the wall in the crisis zone are largely determined by the flow-
rate fluctuations; fluctuations in flow rate very minor from the operating viewpoint (about 1%) can result
in temperature fluctuations of doubled amplitude up to 20-30?C. More detailed studies could be made of the
relation between the flow-rate fluctuation and temperature fluctuation on a one-tube model.
Another appreciable source of temperature fluctuations lies in the large-scale eddies, which migrate
in parts of destabilized flow due to the transverse temperature gradient. It is difficult to predict the exact
extent of the fluctuations on a full-scale model on account of the differences in the temperature gradients
and the hydrodynamic perturbations, but one should bear in mind that the level of the latter increases ra-
pidly with the sodium flow rate. To reduce the importance of such fluctuations, one should avoid elevated
temperature differences at parts where the flow is destabilized. In the present design, the zone of transi-
tion from deteriorating heat transfer to dispersed annular flow should be as far as possible from the so-
dium inlet.
LITERATURE CITED
1. V. A. Vorob'ev et al., in: Exchange of Accumulated Experience on the Design and Commissioning
of Fast-Reactor Systems Based on the BOR-60 Reactor [in Russian], Izd. NIIAR, Dmitrovgrad
(1973), p. 266.
2. M. Kh. Ibragimov, V. I. Merkulov, and V. I. Subbotin, in: Liquid Metals [in Russian], Atomizdat,
Moscow (1967), p. 71.
873
Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060003-2
Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060003-2
ANALYSIS OF THE CONDITIONS REQUIRED FOR
PRODUCING HIGHLY PURIFIED CARBIDE FUEL BY
REDUCING OXIDES WITH CARBON, AND STUDY OF
THE BEHAVIOR OF OXYGEN AND CARBON IMPURITIES
IN THE PRESENCE OF FISSION PRODUCTS
R. B. Kotel'nikov, V. S. Belevantsev,
S. N. Bashlykov, G. V. Titov,
V. A. Zelyanin, and A. M. Anuchkin
UDC 621.039.542.394
The use of carbide fuel in fast reactors is one of several possible ways of achieving high economic
efficiency [1, 2]. One of the most promising techniques for producing carbide fuel on an industrial scale
is by reducing the oxides with carbon [3-5]. Certain questions associated with the production of carbide
fuel by this method and with its subsequent use are still under discussion and require further investigation.
This applies, in particular, to the problem of determining the optimum conditions for the production of
carbide fuel with the minimum oxygen and carbon contamination and also the behavior of these impurities
under conditions of irradiation.
A theoretical and experimental determination of the temperature corresponding to the onset of the
reaction between uranium and plutonium oxides and carbon [6, 7] failed to produce the necessary agree-
ment. We therefore calculated the temperature dependence of the oxygen potential in the U? C-0 system
by the Pourbaix method [8] using the latest thermodynamic data (Table 1) and plotted the corresponding
diagram (Fig. 1). The free energy of formation of UCi_x0x was calculated from the equation
LF =(1 ?x) AFuc ? 4.576T (1 ?x) lg (1 ?x)d- xAFuo + 4.5767'x lg x
on the assumption that ideal solid solutions of the UC?UO system were formed. The limiting solubility of
oxygen in UC was taken as 17.5 at. % [12].
The minimum reaction temperatures obtained from the Pourbaix diagram for various pressures are
given in Table 2, from which we see that the reaction underlying the production of uranium monocarbide by
TABLE 1. Thermodynamic Characteristics
of Compounds Belonging to the U? C ?0 Sys-
tem
Compound
Equation for AF. Temp-
cal/mole range, *K
Ref.
CO(g)
UO2 (0
1.10(g)
(s)
UC2(s)
U2C3(s)
UCO .6500 .35(0
?26 700 ? 20,95 T
?258 000+40,00 T
?258 300+40,50 T
?121 200-h 22,46 T
?24 260-1,12 T
?25 500
?21 300-3,00T
?22 240-1,96T
?24 250-1,20 T
?58 150+5,83T
?58 950+6,56 T
298-2500
298-1405
1405-2000
298-2500
298-1200
1000-2200
298-2200
1100-1400
1400-2000
298-1200
1000-2200
JZO io-4atm CO-.
wozwo 1 atm
\ wai;(ici)
6 keV where the influence of
resonance screening is small, the ratio of the count rates of the y counter should be constant; this was
confirmed by the experiment, and so by using the data from the "thick" specimen we could reduce the sta-
tistical scatter in this energy range.
Normalization of the data with respect to a was effected with thermalized neutrons in a graphite prism
with a gold specimen with n = 1.8.1021 nuclei/cm2 and with a "thin" plutonium specimen for which screening
in the thermal flux was small.
The measurements with the "thick" specimen were related to the measurements with the "thin" speci-
men in the energy range 8 keV < En