THE SOVIET JOURNAL OF ATOMIC ENERGY VOL. 8 NO. 5

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• Publication Date: June 1, 1961
• Content Type: REPORT

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Volume 8, No. 5
June, 1
THE SOVIET JOURNAL OF
TRANSLATED' FROM RUSSIAN
CONSULTANTS BUREAU
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Research by Soviet Experts
Translated by Western Scientists
Soviet Research'on the LANTHANIDE
AND ACTINIDE ELEMENTS, 1949-1957
' An important contr4ai n to the literature of nuclear chemistry, this collection of papers is a_
comprehensive presentation of Soviet re rch on the chemistry of lanthanides and actinides. The 106 reports
included in this collection appeared in the major Soviet chemical journals translated by Consultants Bureau,
as well as in the Soviet Journal of Atomic Energy, 1949-1957.
The five sections, totalling 657 pages, provide broad representation of contemporary Soviet
research in this important aspect of nuclear science. This collection, should . be accessible to all nuclear
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Each part may be purchased as follows:
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Complete collection $65.00
RADIATION CHEMISTRY,
PROCEEDINGS OF THE FIRST
ALL-UN1ON CONFERENCE MOSCOW, 1957
More than 700 of the Soviet Union's outstanding research scientists participated in this
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discussions which followed each report and reflected various points of view on the problem under analysis
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. .
Primary Acts in . Radiation Chemical Processes
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heavy paper covers 6 reports, plus discussion ` illustrated . $12.00
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EDITORIAL BOARD OF
ATOMNAYA ENERGIYA
A. I. Alikhanov
A. A. Bochvar
N. A. Dollezhal'
D. V. Efremov
V. S. Emel'yanov
V. S. Fursov
V. F. Kalinin
A. K. Krasin
A. V. Lebedinskii
A. I. Leipunskii
I. I. Novikov
E ditor-in-C hie f )
B. V. Semenov
V. I. Veksler
A. P. Vinogradov
N. A. Vlasov
(Assistant Editor)
A. P. Zefirov
THE SOVIET JOURNAL OF
ATOMIC ENERGY
A translation of ATOMNAYA ENERGIYA,
a publication of the Academy of Sciences of the USSR
(Russian Original bated May, 1960)
Vol. 8, No. 5 June, 1961
CONTENTS
Winners of Lenin Prizes
Determination of the Mean Number of Secondary Fission Neutrons from the Fragment
PAGE
341
RUSS.
PAGE
Mass Distribution. Yu. A. Zysin, A. A. Lbov? and L. I. Sel'chenkoy
343
409
An Investigation of the Properties of Metals and Some Steels after Irradiation
by Fast Neutrons. Sh. Sh. Ibragimoy, V. S. Lya.shenko, and A. I. Zay'yaloy
347
413
Vapor Pressure of T20. M. M. Popov and F. I. Tazetdinov
353
420
Radiometric Analysis of Ores on Conveyers. L. N. Posik, S. I. Babichenko,
358
425
and R. A. Grodko
Angle-Energy Distribution of y-Radiation Scattered in Water and Iron. Yu. A. Kazanskii
364
432
Universal Apparatus with a Coss y -Ray Source with an Activity of 60,000 g-eq of Ra
for Simulating Radiation-Chemical Apparatuses, and Investigations (The "K-60,000").
A. Kh. Breger, V. B. Osipoy, and V. A. Gol'din
371
441
LETTERS TO THE EDITOR
Investigation of the Spent Fuel Element of the First Atomic Power Station. A. P. Smirnov-
Averin, V. I. Galkoy, Yu. G. Seyastiyanoy, N. N. Krot, V. I. Ivanov, I. G. Sheinker,
375
446
L. A. Stabenoya, B. S. Kir'yanoy, and A. G. Kozloy
On Improving the Efficiency of Power Station Reactors with Gaseous Coolants.
T. Kh. Marguloya and L. S. Sterman
377
448
Measurement of the Fast Neutron Flux Distribution in the Core of the VVR-S Reactor
with Respect to Changes in the Electrical Conductivity of Germanium Specimens.
E. Aleksandroyich and M. Bartenbakh
381
451
Calculation of Thermal Shocks in Reactor Structural Parts. Yu. E. Bagdasaroy
383
452
600-key Proton Injector for a Linear Accelerator. Yu. N. Antonoy.L. P. Zinoy'ey,
386
454
and V. P. Rasheyskii
Mean Number of Prompt Neutrons Emitted in Photofission of Th2s2 and U238 by y -rays
Produced in the Fls (p, ay )016 Reaction. L. I. Prokhorova and G. N. Smirenkin . . .
390
457
Electron Acceleration in a Traveling-Wave Cyclical Wayeguide Accelerator.
A. A. Vorob'ey, A. N. Didenko, and E. S. Koyalenko
392
459
Use of Scintillation Counters in Gammascopy. V. E. Nesterov
394
461
NEWS OF SCIENCE AND TECHNOLOGY
Atomic Energy at the Soviet Exposition in Havana. L. Kimel', and V. Tsurkov
397
464
Atomic Energy of the All-China Exposition on Industry and Means of Communication.
Shen Chung-po
399
464
Annual subscription $75.00 ? 1961 Consultants Bureau Enterprises, Inc., 227 West 17th St., New York 11, N. Y.
Single issue 20.00 Note: The sale of photostatic copies of any portion of this copyright translation is expressly
Single article 12.50 prohibited by the copyright owners.
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CONTENTS (continued)
[Washington Conference of the American Nuclear Physics Society and Atomic Industrial
Forum. Sources; Nucleonics 17, No. 12. 17-23 (1959); Nuclear Power 5, No. 45,
111-116 (1960)
PAGE
RUSS.
PAGE
467]
[Development of Nuclear Power in the_Countries of South and Central America
467]
[Organic Moderated Reactors for Land-Based and Seagoing Facilities
470]
[Reactor as a Neutron Source
472]
Measurement of Magnetic Moment of Li8
400
473
[New Foreign Articles on Rolling of Uranium
474]
[On the Use of Statistical Analysis Techniques in Explorations for Uranium Deposits.
Source; R. Bates, Econ. Geol. 54, No. 3, 449 (1959)
476]
BIBLIOGRAPHY
New Literature
401
480
NOTE
The Table of Contents lists all material that appears in Atomnaya Energiya. Those items
that originated in the English language are not included in the translation and are shown en-
closed in brackets. Whenever possible. the English-language source containing the omitted
reports will be given.
Consultants Bureau Enterprises, Inc.
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. WINNERS OF LENIN PRIZES
Translated from Atomnaya gnergiya, Vol. 8, No. 5, pp. I-II,
May,1960
The Committee of the Council of Ministers of the.
USSR. for the Lenin prizes in the field of Science and,
-Technology has assigned the prizes for 1960 to the fol.-
lowing.scientists for their scientific research on the physics
of fast-neutron nuclear reactors: Academician (Academy
of Sciences of the UkrSSR) A. I. Leipunskii, .Drs. of Phys. :
and Math. Sci., 0. D. Kazachkovskii,and I. I,Bondarenko,
and Cand. of Phys., Math. Sci., L. N., Usachev. . ? .
Nuclear reactors working with fast neutrons occupy. ?
a special place in nuclear energetiC-s. Back 'in 1949;
A. I. Leipunskii calculated that in such reactors one could -
realize an extensive production of nuclear fuel. COnse-
quently,?the use of energy reactors working with fast
neutrons, together with the use of energy reactors working.
with thermal neutrons, would allow complete use of ex-
tracted uranium, which is equivalent.to increasing by a...
factor 100 the 'fuel resources of nuclear power engineering..
The main difficulty in the practical realization of
a chain reaction in a reactor aonscsted ip.the fact that
many problems of the 'physics of such reactors were still
unsolved. There appeared no publication concerning fast
reactors in the foreign literature until 1955. Therefore,
a:school of specialists was formed and developed indepen-
_dently in the Soviet Union in the physics of' fast-neutron
reactors. This school was headed by A. I. Leipunskii,
0, D. Kazachkovskii, I. I. BOndarenko, and L. N. Usachev.
Under their guidance the problems of the physics
of fast-neutron reactors were studied on a theoretical basis, ?
. ? .,
experiments for establishing the, constants necessary for
calculation were carried Out, the foundations of the theory
were elaborated, and practical work for building fast- '
neutron reactors was started.
? Subsequently, critical assemblies and fast-neutron.
reactors were constructed (BR-1, BR-2, BR-3, BR-4, and
BR-5). The construction of each new apparatus was a
logical continuation and generalization of experience
'accumulated, and a development, Of the theory and.. prac7
toe of reactors working with fast. neutrons.. -.
_
The last apparatus of /his seriesBR-5;.started in the
shintner of 1958?has all the main characteristics Of an'
atomic electric power plant, and is a prototype of the
future powerful atomic plants .with fast-neutron reactors.
The Lenin prize for the construction of the complex.
of research water- water reactors VVR-2, VVR:5, and ITR
was assigned to the scientists S. M. Feinberg, V. V.
Goncharov, G. A. Stolyarov, T. N. Zubarev, P. I. Khris-
tenko, V. F. Kozlov, and 0. I. Lyubimtsev.
The construttion of research reactors is the ha-sit' fOr
the. use of atomic energy. Perhaps there has been:no
development of its use that has not taken place as a direct
. or indirect result of work on research reactors. The obtain-
ment of radioactive isotopes, 'the radiation' treatment and
testing of materials, the study of the phYsical Problems
-Of high-power reactors; the testing of thermoemitting
elerneots for nuclear power plants and transport atomic
equipment under construction Or in project, the testing.
.of CoinPonenti,' joints and materials .for high power reactors,
A. I. Leipunskii 0. KazachkOvskii
I. I. Bondarenko
L. N. Usachev
341
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investigations in neutron physics, biology and medicine
investigations concerning the 'radiation of supply products,.
radiation polymerization, and many other problems are
solved with the help of research reactors.
Our industry produces at present in mass research
water-water reactors of the types VVR-2, VVR-S, and
ITR. Some scientific research institutes Of the Soviet
Union and of foreign countries are equipped with them.
The compactness, the low cost, thereliability of
operation,and the wide possibilities for experimental re-
search are the main advantages of the water-water re-
actors whose construction is recognized by the Lenin prize.
The first of these reactors?VVR-2?was.constructed
when the world literature included no publication con-
cerning such reactors. Nevertheless, the group of special-
ists solved successfully all the problerris?concerning the
construction and the physics of the reactor, its regulation,
its operational stability, etc.'
? The construction of the complex of research,water-
water reactors is a great achievement in the field of the
use of atomic energy for peaceful purposes. ? The vast
development of experimental work on such reactors that
exists?at present permits a wider use of atomic energy in
the national economies of the USSR and other ?countries.
P. h Khristenko
342
V. F. Kozlov ? O. h Lyul?imtsev
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DETERMINATION OF THE MEAN NUMBER OF
SECONDARY FISSION NEUTRONS FROM
THE FRAGMENT MASS DISTRIBUTION
Yu. A. Zysin, A. A. Lbov, and L. I. Serchenkov
Translated from Atomnaya gnergiya, Vol. 8, No. 5, pp. 409-412,
May, 1960
Original article submitted September 3, 1959
A method is presented for computing the mean number of secondary fission neutrons -r-) from the fission-fragment
mass distribution curves. The error of the method is estimated. It is shown that in those cases in which the frag-
ment mass distribution curves are carefully studied ,7-7 can be determined with satisfactory accuracy by this meth-
od.r The value of 7 is computed for fission in Th232, u233, u235, u238, pu239, Am241,and r252. The results are dis-
cussed and compared with results obtained by other methods. The partial values 7m are computed for thermal-
neutron fission of U233 and U235.
At the present time the mass distribution of fission
fragments has been studied in many cases of fission in
heavy nuclei [1-11]. In some cases the accuracy and
completeness of the experimental data are adequate for
computing the mean number of secondary fission neutrons
7 with satisfactory results. Although the quantity i7 has
been measured very precisely by indirect methods [12-
14], this calculation is of great interest since it allows us
to determine the values of F by an independent method.
Furthermore, this technique allows us to obtain the value
of for those cases in which it has not been determined
by other methods. Finally, in principle,using the fission-
fragment mass distribution curves it should be possible to
compute the partial values of 7m, i.e. the mean number
of fission neutrons corresponding to a given ratio for the
mass numbers of the heavy and light fragments (AH/AL)m.
Attempts to estimate r7 from the fission-fragment
mass distribution curves have been carried out earlier
[8-10, 15]; however, these estimates were extremely ap-
proximate. In the present paper, for the first time,the
possibilities of this method are investigated in detail, the
errors involved are studied, and calculations of 7 are
carried out for cases in which the error is shown to be
small.
In the general case
v =? 2A,
(1)
where Ao is the mass number of the fissioning nucleus
and 7i is the mean mass number of all fragments, where
ociA///a, (A1 is the mass number of the frag-
ment and cti is the fragment yield). It is well known that
Ea, = 2. In practice, however, it is not convenient to
use Eq. (1) to determine 77 because of the appreciable
errors involved. The error in the determination of i7 is
found to be much smaller if another expression is used:
(2)
which takes account of the fact that two fragments are
formed as a result of fission ? the light and heavy frag-
ments, which are characterized by the mass numbers
-AL and -A-H,respectively. This formulation of the problem
is valid in all cases of fission in which the original nucleus
is not highly excited since ternary fission and cases in
which protons and alpha particles are emitted can ob-
viously be neglected.
The quantities AL and -AH can be determined by
averaging the masses of the light and heavy fragments,
respectively, using the expressions
E,41a1 E Aka i,
?
AL =
and AH-
-
From the definition of yields it is obvious that cci and
cck summed separately over the light and heavy frag-
ments should be equal to unity. Because of the inaccuracy
in the experimental determination of the individual yields,
however, these quantities are found to be only approxi-
mately equal to unity; hence, in computing -73 from Eq.
(2) it is important to introduce two normalization condi-
tions: I a, = 1 and ak =1.
We now consider the mean-square-error AT and its
dependence on the uncertainty in the experimentally
determined quantities a. From the expression for the
determination of the error in 7 due to the ith term and
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Comparison of the Calculated and Measured Values of ir
emission
which ac-
companies
fission
Energy
Determination ofT/by the present
method
Determination of 1; by
other methods
Ao
AL
AH
reference
U233
U235
PU.238
Am241
Th232
1J235
U238
(j235
U238
U238
U238
U238
Cf 252
Neutrons
ft
ft
)' rays
10
Spontaneous
fission
Thermal
Reactor spec-
trum
Fission spec-
trum
The same
14 Mev
14 Mev
8-10 Mev
16 Mev
48 Mev
234
236
240
242
233
236
239
236
239
238
238
238
252
93,3+0,1
94,8+0,1
98,5+0,2
101+0,3
91, 4+0,1
95,6+0,2
97,3+0,1
96, 3?0 , 2
97,0+0,3
96,6+0,1
97,4+0,1
96,5+0, 2
106,3+0,3
138,2+0,1
138,8+0,1
139,0+0,1
138,3+0,3
139,8+0,3
138,4+0,3
138,9+0,1
135,5+0,2
137,0+0,4
138,7?0,3
137,5+0, 3
137,2+0,3
140,9+0,3
2,5+0,2
2, 4?0, 2
2,5+0,3
2,7+0 , 6
1 ,8?0, 4
2,0+0,5
2,8?0,2
4, 2+0 , 4
5,0+0, 6
2,7+0, 4
3,1+0,4
4,3+0,5
4,8+0,6
2,52+0,03
2,47+0,03
2,92+0,04
3,14+0,05
1,8***
2,50+0,06****
2;65?0,07*****
4,13+0,24
4,50+0,32
3,84+0,12
[12,131
[12,13]
[12,13]
[14]
[12]
[12,13]
[12,13]
[12,13]
[12,131
[13]
The uncertainty in iTis determined by the sum of the uncertainties for AL and A"H. The actual un-
certainty in Vis smaller.
" We give the experimental values of7/ obtained by other methods, which are averaged over many
measurements. Hence, the uncertainty is appreciably smaller than the experimental uncertainties in the
in the individual methods.
***This is obtained by direct extrapolation by the values of iTi=2.35?0.07 for En=3.5 Mev and TJ 2=
=4.64?0.20 for En=14.2 Mev.
****For an effective energy of 0.7-0,74 Mev.
For an energy of 1.5 Mev.
Vrn
3,0
2,0
1,0
I ' ob
0090
-
,
2,0 117=2'1B
1,0
'III Ili
m= R
Fig. 1. Partial values vm for thermal- Fig. 2. Partial values urn for thermal-
neutron fission of U. neutron fission of U.
344
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from the corresponding normalization conditions, it can technique it is possible to carry out the necessary calcula-
be shown that when Eq. (1) is used tions. It should be noted, however, that in this case it is
necessary to take account of the variation of the partial
value values vas a function of the type of fission which
g 171 at(i71--A)2;
while when Eq. (2) is used
(3)
L
Av = e 1/1 al (AL A02 isc4 (Alf? Ak)2 , (4)
where ? = A a/a is the relative uncertainty in the deter-
mination of a (for simplicity we assume that ? is the
same in all cases).
In the case of thermal-neutron fission of U235, com-
puting v from Eq. (1) we find 6,11 ? 1.3; comput-
ing this same quantity from (2) we find AT = ? 0.15
(in both cases we assume that I el 0.05). Thus, when
Eq. (2) is used the accuracy oi the calculation is an order
of magnitude better than the accuracy attained with Eq.
(1).
The accuracy depends strongly on the asymmetry in
the fission-fragment mass distribution curve, which is
determined by the depth of the valley. If the mass num-
ber AL (characterized by the yield a1). intermediatebe-
tween the numbers of light and heavy fragments, refers
to the heavy fragment rather than the light fragment, or
vice versa, an additional error Ai7=1(X-H?A-dall (with
an accuracy to second order) occurs. As an example
we estimate Ali for fission of U238 by thermal neutrons
(the asymmetry is well defined, aI 10-4) and
for fission by neutrons with energies of 14 Mev (in which
case the asymmetry is weaker, al =10-2). Taking (XH-
--XL =40, in the first case we have Aiit = 0.004; in the
second case AT/1=0.4. Thus for fission by 14-Mev neutrons
the quantity A71 can make a sizeable contribution in the
final uncertainty.
Since the error in the determination of 7 by the
method .described here depends on the uncertainty in the
yield determination, it is extremely important that the
yields (absolute or relative) ai and ak be known for the
greatest possible number of mass numbers Ai and Ak.
Interpolation of the fragment mass distribution curve for
A with unknown values of a increases the error. For this
reason the calculations reported here were limited to
those cases in which the fragment mass distribution curves
were known rather completely and characterized by well-
defined asymmetries. In this work we have used data
published up to January, 1959 [1-11]. It should be noted
that experimental refinement of the fission-fragment mass
distribution curves will make it possible to compute 1-)-
for many other cases of interest in the future. For the
cases which have been studied,refinement of the distri-
bution curves will make it possible to increase the ac-
curacy in the determination of 17 . In certain cases (for
example, for U238(y [2]) appropriate distributions have
not been studied with the necessary detail; however, by
using "reflected" points and a successive approximation
is involved.
The results of the calculations of ii by the method
reported here are given in the Table. For comparison
purposes, we also show values of V determined experi-
mentally by indirect methods. As is apparent from the
Table, in the majority of cases the values of v computed
by the present method are in satisfactory agreement with
the values determined by other methods.
It may be indicated that the data for 7 for fission of
U238 by y rays with energies of 8-10, 16,and 48 Mev are
new. In all cases the uncertainty in the determination of
'XL and AH are obtained from the spread in the results of
the calculations by different versions of the mass distribu-
tion curves; these curves in turn are obtained by different
methods and are affected by the spread in the experimen-
tal points and the individual uncertainties.
The fission-fragment mass distribution curves also
allow us to compute the partial values Um =A0?(AH +
+AL)m; the mass numbers AH and AT are coupled, i. e.
characterized by the same yield am (AH /AT =m). Rela-
tion between 77 and 7 is expressed by the formula
V --=
vm
711
E
711
(5)
The difference Ac,?(AH+ AL)m will determine the
partial numbers of secondary neutrons for a given pair of
mass numbers A and AL under the following conditions;
1) the mass distribution curve is measured very carefully,
2) am is a sensitive function of A, 3) there is no fine
structure in the mass distribution curve, 4) the curve it-
self is monotonic in the region considered, and 5)
and the distribution of probability for emission of one,
two, three, etc. neutrons are weak functions of A.
The calculation of 7m was carried out only for fission
of U233 and U238 by thermal neutrons, for which there are
experimental yield values for almost all the mass numbers
It will be apparent that this method of determining 7m
is not accurate in the region 1.3< m< 1.7. The curves
that have been obtained (Figs. 1 and 2) are similar in
shape to the analogous curves obtained for U233 by another
method [16]. The maximum value of v?m is obtained for
values m=1.5. In both cases there is a characteristic
sharp reduction in -Dm for fission into -fragments of ap-
proximately equal mass. If we consider the emission of
secondary neutrons to be the result of excitation of frag-
ments, which are not spherical in shape [17], it may be
assumed that the fragments in symmetric fission are more
spherical than the fragments due to asymmetric fission.
345
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LITERATURE CITED
1. S. Katcoff, Nucleonics 16, 4, 78 (1958).
2. R. Duffield, R. Schmitt, R. Sharp, Report No. 678
(USA) Second Int'l. Conf. an the Peaceful Uses of
Atomic Energy (Geneva, 1958).
3. A. N. Protopopov et al., Atomnaya fnergiya 5, 2,
130 (1958)!
4. M. P. Anikina et al., Second Int'l. Conf.on the Peace-
ful Uses of Atomic Energy (Geneva, 1958); Reports by
Soviet Scientists, Nuclear Physics [in Russian]
(Atomizdat, Moscow ,1959) Vol. I, p. 396.
5. L. Bunney etal,, Report No. 643 (USA) Second Int'l.
Conf. on the Peaceful Uses of Atomic Energy (Geneva,
1958).
6. K. Fritze, C. McMullen, and H. Thode, Report No.
187 (Canada) Second Int'l. Conf. on the Peaceful
Uses of Atomic Energy (Geneva, 1958).
7. L. Bunney et al., Report No. 644 (USA) Second Int'l.
Conf. on the Peaceful Uses of Atomic Energy (Geneva,
1958).
8. J. Cuningham, J. Inorg. and Nucl. Chem. 4, 1(1957).
9. J. Cuningham, J. Inorg. and Nucl. Chem. -5-, 1 (1957).
346
10. L. Glendenin and E. Steinberg, J. Inorg and Nucl.
Chem. 1, 45 (1955).
11. J. Cuningham, J. Inorg. and Nucl. Chem. 6, 181
(1958).
12. R. Leachman, Second Int'l. Conf. on the Peaceful
Uses of Atomic Energy (Geneva, 1958) Selected
Reports of Foreign Scientists, Neutron Physics (Atom-
izdat, Moscow, 1959) Vol. II, p. 342; Atomic Engi-
neering Abroad, 1, 11 (1959).
13. I. I. Bondarenko et al., Second Int'l. Conf. on the
Peaceful Uses of Atomic Energy (Geneva, 1958)
Report of Soviet Scientists, Nuclear Physics [in Rus-
sian](Atomizdat, MosCow, 1959) Vol. I, p. 438.
14. V. I. Lehedev and V. I. Kalashnikova, Atomnaya
gnergiya 5, 2, 176(1958).
15. R. Jensen and A. Fairhall, Phys. Rev. 109, 942(1958).
16. J. Fraser and C. Milton, Phys. Rev. 93, 818 c1954).
17. Ya. B. Zel'dovich and Yu. A Zysin, Zhur? Eksp.i
Teoret. Fiz. 10, 8, 851 (1940).
*Original Russian pagination. See C. B. translation.
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Declassified and Approved For Release 2013/02/19: CIA-RDP10-02196R000100050005-4
AN INVESTIGATION OF THE PROPERTIES OF METALS
AND SOME STEELS AFTER IRRADIATION BY FAST NEUTRONS
Sh. Sh. Ibragimov, V. S. Lyashenko, and A. I. Zav'yaloy
Translated from Atomnaya Energiya, Vol. 8, No. 5, pp. 413-419,
May, 1960
Original article submitted May 28, 1959
The article considers the effect of the irradiation by fast neutrons and of a subsequent heat-treatment on the
properties of some metallic materials. The change of the properties of materials upon irradiation is explained
in terms of the formation of various types of defects on the crystal lattice, which are annealed at appropriate
temperatures. The kinetics of the defect elimination processes leading to a strengthening of the material is
studied, and their activation energy is determined.
It is well known that under the action of nuclear
particles, especially of fast neutrons, significant changes
take place in the physical and mechanical properties of
various materials. The extent of these changesdepends
principally upon the particle energy, the integral dose,
and the temperature of irradiation.
The change of mechanical properties of metallic
materials under the action of heavy nuclear particles
has some similarity with the strengthening produced by
cold deformations. The cause of the property change in
both cases could be looked for in the appearance of
disturbances of the regularity of the crystal structure of
the materials and defects of the lattice. On the other
hand, some experimental data show an essential differ-
ende between the changes produced in a material by
nuclear particles and those produced by a cold plastic
deformations [1-3]. One can suppose that the disturbances
arising under the action of nuclear particles are a com-
plicated phenomenon: additional experimental and
theoretical investigations are required before they can
be explained.
The present paper reports an investigation of the
effect of fast neutrons on the structure and properties of
iron, nickel, molybdenum, and some steels,whose
chemical composition is given by Table 1.
Samples of the materials mentioned above were
irradiated in the active zone of an experimental BR-2
reactor [4] after an appropriate heat-treatment in
special hermetically welded tubes of 1Kh18N9T steel.
The average integral dose was 1.8.1020 fast neutrons per
1 cm2, and the irradiation temperature was 40-70?C.
The properties of the materials studied, before and
after irradiation, are presented in Table 2. The data of
the table show that as a result of irradiation by fast
neutrons, the strength, hardness, and electric resistance
increased, and the relative lengthening decreased. The
degree of property change depended upon the nature of
TABLE 1, Chemical composition of some steels
Contents of principal elements, 6/0
Steel type
Cr
Ni
Mo
Ti
?
1Kh18N9
0,14
16,0
9,5
?
?
1Kh18N9T
0,11
16,7
9,2
?
0,6
1Kh18N12
0,09
16,6
12,0
?
?
1Kh18N12M2T
0,10
17,2
12,4
2,1
0,5
1Kh18N17
1,12
16,1
16,9
?
?
the irradiated material: a more or less complex composi-
tion of a steel due to alloying with various elements did
not seem to have any important effect on the magnitude
of the change in question. The greatest change was ob-
served for molybdenum (its relative lengthening decreased
by a factor more than 10, its electric resistivity had a
33% increase); this may have depended upon the high
annealing temperature and upon the high elastic con.
stants of molybdenum.
With the purpose of studying the temperature sta-
bility of radiation defects and establishing the tempera-
ture of complete regression of the properties, the irradia-
ted samples were subjected to annealing at various temp-
eratures for 30 min. After each treatment the micro-
hardness and (in some instances) the electric resistance
of the samples were determined. The results of the micro-
hardness measurements are given in Figs. 1 and 2.
The data presented in Figs. 1 and 2 show that: 1) as
a result of heating at appropriate temperatures a com-
plete elimination of the increase in microhardrass took
place; 2) the increase in microhardness upon heat-
treatment took place in a temperature range higher for
materials with body-centered cubic lattices (iron, moly-
bdenum) than for materials with face-centered cubic
347 ?
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TABLE 2. Properties of metals and of some steels before and after irradiation
348
Specific electric
resistance, ? ohm?
?cm
UOptligA
00
uolnipvi
-al nip
,l'0,1
._
co co-
,-,
n ,caD
1--- I I I 1-: t?-:
1---N
uopv!pvi
-IT 910jaq
Cq cc.
- .
N
io c-
,-.
0 Cn
1 I I ' I 1---: CO-
t----r-
Microhardness,
kg/mm2
uopvinn
o 0
co-. 1,-
o co
000000
oo",n" c::)- c5 ?O. "-T.
cocc...-.0mo
-4-, .?....,
uopylpvio
-al iavg
0
Lo- c;
CO CD
..,4 ..,-.,
000000
00,i01.4. 20?, we find ree /50?
. it is reasonable to expect
that because of the increase in anisotropy of Compton
scattering with an increase in energy of the primary y
radiation, the angle eo will be reduced. However, the
function le (0), plotted from the data of [7] (a plane uni-
directional source of electron bremsstrahlung with elec-
tron energies of 10 Mev and a lead barrier 152.4 mm
thick) for 6 > 20? is proportional to e-0/18e; that is to
say it coincides with the le curve in lead for a Co60
source.
Energy spectra of scattered radiation. The energy
spectra Io plotted from the results of other measurements
of the angle-energy distributions measured under condi-
tions of semiinfinite geometry were compared with the
calculations [1] which were carried out by the method
of moments for a medium of infinite extent (cf. Figs. 6
and 8 and [5, 6]). When the Coe? is used as a radiation
source agreement within 5-10% is generally obtained
with the calculations for energies greater than 0.4-0.5
Mev. As a rule, for energies below 0.4 Mev the experi-
mental data lie below the calculated data; this finding
is explained by the difference in geometries in the cal-
culations and the experiments.
It is of great interest to compare the experimental
and calculated energy spectra for higher primary y -radia-
tion energies. In [7] a comparison has been made of the
energy spectra for scattered bremsstrahlung y radiation,
370
using the calculations in [1]. The discrepancy of about
25% is attributed by the authors to the difference in geo-
metry and the fact that no account has been taken of the
brehmsstrahlung due to secondary electrons in the cal-
culations. It should be noted that the discrepancy is actu-
ally more significant since the authors of [7] have com-
pared the energy spectrum of y radiation which intersects
a fixed unit area in all directions (flux) with calculations
[1] of the spectrum for scattered y radiation which inter-
sects a sphere of unit radius in all directions.
In conclusion the author wishes to thank I. I. Bondar-
enko and V. I. Kukhtevich and S. G. Tsypin for discussion
of the present work; the author is also indebted to A. N.
Voloshin and V. I. Popov for help in carrying out the
experiments.
LITERATURE CITED
1. H. Goldstein and J. Wilkins. Calculation of the Pene-
tration of Gamma Rays (New York, 1956, NYO- 3075).
2, L. Spenser and F. Stinson, Phys. Rev. 85. 662 (1952).
3. M. Berger, J. Appl. Phys. 26, 1504 (1955).
4. G. Whyte, Canad. J. Phys. 33, 96 (1955).
5. Yu. A. Kazanskii and S. P. Belov, Physics and Heat
Technology of Reactors. Suppl. No. 1 of Atomnaya
gnergiya, p, 123.**
6. Yu. A. Kazanskii, S. P. Belov, and E. S. Matusevich,
Atomnaya gnergiya 5, 2, 457 (1958).* *
7. J. Hubbell, E. Hayward. and W. Titus, Phys. Rev.
108, 1361 (1957).
8. W. Dixon, Canad. J. Phys. 36, 419 (1958).
9. Yu. A. Kazanskii, Pribor. i Tekh.gksp, 4,32(1959).
10. M. Weiss and M. Bernstein, Phys. Rev. 92, 1264
(1953).
11. V. I. Kulditevich, S. E. Tsypin, and B. P. Shemen-
tenko, Atomnaya gnergiya 5, 6, 638 (1948).
12. M. Berger and J. Doggett, J. Res. Nat.Bur. Standards,
56, 89 (1956).
t A comparison of the ratio of the energy build-up for
different geometries is possible because of the small
difference in the accumulation factors under conditions
of infinite geometry for a point source and a plane uni-
directional source with small values of ti or [1].
s'Original Russian pagination. See C. B. translation.
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Declassified and Approved For Release 2013/02/19: CIA-RDP10-02196R000100050005-4
UNIVERSAL APPARATUS WITH A Cow y -RAY SOURCE
WITH AN ACTIVITY OF 60,000 g-eq OF Ra
FOR SIMULATING RADIATION-CHEMICAL? APPARATUS,
AND.INVESTIGATIONS (THE "K-60,0009
A. Kh. Breger, V. B. Osipov, and V. A. GoPdin
Translated from Atomnaya Energiya, Vol. 8, No. 5, pp. 441-445,
May, 1960
Original article submitted August 20, 1959
The article describes a universal apparatus for radiation-chemical investigations with a Cos? y -ray source possess-
ing an activity of about 60,000 g-eq of Ra. The design of the apparatus makes it possible to ,simulate radiation-
chemical apparatus with powerful isotopic y -ray sources of various configurations and dimensions: a cylindrical
radiating element, a radiating element in the form of two plates, a radiating element in the form of a "heat ex-
changer," and a radiating element in the form of one or more rods. The dosage rate (without taking into account
the attenuation in the protective vessels) varies from-250 thee in a volume of 36 liters to? 3000 thee in a
volume of 0.1 liter. The apparatus is designed for carrying out radiation-chemical investigations under practically
any physicochemical parameters; it ensures the possibility of remote control and observations both of the experi-
mental conditions and the processes taking place in the investigated systems during irradiation.
As a result of the development of radiation-chemical
investigations [1, 2] and the imminent change-over from
laboratory work to processes on an enlarged scale it has be-
come necessary to develop new apparatus which meet
both ordinary requirements (for isotopic apparatus) [3]
and a number of new requirements.
On the one hand, the apparatus must meet require-
ments associated with the necessity of irradiating large
volumes of substances (liters, tens of liters)at high dosage
rates (102-103 r/sec) and adequate uniformity of the dos-
age field.* On the other hand, the necessity arises of
carrying out investigations of radiation processes in flow-
ing and circulating systems. In the latter case, uniform
distribution of the dosage field in a large volume is not
obligatory and it is more important to have high dosage
rates. Moreover, with the development of radiation-chem-
ical investigations the physicochemical conditions under
which the experiments are carried out become more com-
plicated. Finally, for a change-over from laboratory ex-
periments to industrial processes, experimental simulation
of radiation-chemical apparatus with powerful radiating
elements of various shapes and sizes is necessary. Reliable
scientifically-based methods of calculating pilot-plant
and industrial apparatus can evidently only be develop
ed by means of such experiments. This includes calcula-
tions of dosage fields of energy [4-6] that is absorbed
by the irradiated system,f the efficiency of the radiating
element and the apparatus as a whole [7, 8], the choice
of the optimum shape and size of the radiating element,
the heat conditions, etc.
When the K-60,000 apparatust was developed, to-
gether with ordinary requirements previously formulated
[3, 93 and included in the design of the apparatus
described in [10. 11],the above-listed requirements were
also taken into account.
It should be noted that in a number of works published
after the construction of series K apparatus, during the
development of apparatus for radiation-chemical invest-
igations,the authors took as their basis considerations and
requirements similar or closely similar to ours. This
applies particularly to those works whose authors
made efforts to analyze literature data and approach the
development of a new apparatus from the aspect of a
scientifically-based problem of general significance [12,
13], not a particular constructional problem applicable
only to the given conditions.
TheK-60,000 apparatus was also developed on the
basis of such considerations.
Principal parameters and design of the K-60,000
apparatus. The K-60,000 apparatus is designed for:
1) The simulation of radiation-chemical apparatus
with powerful y -ray sources of various configurations,
that is,
? The latter is important mainly during irradiation of
objects in the solid phase.
f The calculations of the absorbed energy in irradiated
systems were made by A. Kh. Breger, B. I. Vainshtein,
L. S. Gusei, N. P. Syrkus, and Yu. S. Ryabukhin.
t The development was commenced in July, 1957.
371
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a) with a radiating element in the form of a hollow
cylinder of height 32 cm, and an internal diameter of
2-38 cm; in this case, a mean dosage rate" of -3000
r/sec (in a volume of 0.1 liter) to -250 r/sec (in a volume
of 36 liters) is ensured;
b) with a radiating element in the form of two plates
of height 32 cm and length up to 32 cm, the distance be-
tween the plates being from 5 cm (in a volume of 4
liters the mean dosage rate** was ,-1300r/sec) to 25 cm
Fig. 1. Horizontal section of the K-60,000 apparatus;
1) transducers of the USID-1 dosimetry apparatus; 2)
transmission; 3) electrical drive; 4) working table; 5)
solenoid shutter,
dooroirtm-A;Effortzefieiroseresom
r
NMI liko,
1D111
1011111rAW
'ffr,
6i% r47441e0
Fig. 2. Vertical section of the K-60,000 apparatus; 1)
store; 2) working table; 3) radiating element tubes;
4) ropes for raising the cassettes with Co66; 5) mechan-
ism for moving the sources; 6) induction transducers;
7) switchboard connecting the electric cables with
the instruments of the physicochemical control-
board; 8) transmitting camera of the television
system.
372
(in a volume of about 26 liters the mean dosage rate"
was 450 r/sec);
C) with a radiating element of the "heat exchanger"
type with different arrangements of the sources in the ir-
radiated medium;
d) with a radiating element in the form of one or
more rods.
It is possible to vary (other conditions being equal)
the activity of the radiating element from 3000 to 60,000
g-eq of Ra.
2) For carrying out radiation-chemical processes in
various apparatus with the use of the radiating elements
indicated in section 1, and in practice under any physico-
chemical conditions.
The K-60,000 apparatus is located on two stages of
a specially equipped room. The labyrinth (Fig. 1) and
the working chamber (Fig. 2), in which irradiation is
carried out, are located on the lower stage. The working
chamber and the labyrinth are separated from the neigh-
boring rooms by concrete ceilings and walls (specific
density of the concrete 2.3 t/m3), which protect the per-
sonnel when the cassettes with the y -ray sources are in
the working position.
The following principal units of the apparatus are
located in the working chamber.
1) The y -ray source, consisting of 20 articulated
four-membered cassettes (Fig. 3). Each cassette contains
four standard Co6? preparations, each having an activity
of 700-750 g-eq of Ra (in its sheath, a preparation has a
length of 81 mm and a diameter of 11 mm).
2) The stgre of the y -ray source, consisting of 20
isolated, specially curved,tube-channels, the area be-
tween which is filled with iron shot of diameter 2-3 mm
(bulk density 4.5 g/cm3).tt During the intervals between
Irradiation of the objects the cassettes with the Co6? pre-
parations are kept in the lower parts of the channels of
the store. The layer of iron shot, of height 1.6 m, and the
curved channels attenuate the y radiation to the per-
missible norm.
3) The radiating element of the apparatus, consist-
ing of 20 individual tubes connected to the corresponding
channels of the store by means of flexible metal sleeves.**
4) The working table of the apparatus, on which the
objects to be irradiated are arranged, and which is used
?? The dosage rate is calculated without taking into ac-
count the attenuation of irradiation in the materials of
the cassette, the tubes of the radiating element, etc.
Such data on the calculation of the dosage fields and
their experimental investigation will be published later
ft The K-60,000 apparatus is made without the use of
lead.
**The channels of the store, the cassettes and the radia-
ting element tubes are made of mark Kh18N9T stainless
steel.
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to support the tubes of the radiating element, the con-
figuration of which is determined by the experimental
conditions (Figs. 2 and 4).
5) The ropes with electromagnets fixed on the ends
by means of which the required number of cassettes (from
1 to 20) is lifted to the radiating element of the appara-
tus.
6) The mechanism for moving the source, with trans-
mission gear and electric drive.
In addition to these components the working chamber
contains the outlets of the gas and liquid pipelines, the
switchboard for connecting the electric cables to the
secondary instruments on the control desk and for ob-
serving the physicochemical conditions of the experi-
ments: remote control and adjustment of the temperature,
pH, pressure, electrical resistance, dosage rate, etc. A
television system is installed in the working chamber for
visual observations on the irradiated objects. The control
desk of the apparatus has a light-signalling system for
indicating the position of the individual irradiation sources.
Working principle of the apparatus. Before the corn-
mencerhent of irradiation all the cassettes with the Co6?
preparations are located in the lower part of the channel;
this is signalled by the induction transducers installed
in the channels of the store. The induction transducers
Fig. 3. Cassette with Co60 y -ray source and electro-
magnet.
Fig. 4. Working table of the K-60,000 apparatus,
with radiating element in the form of two plates.
consist of double coils*** with an ac feed. As a result
of the presence of cores, which in the given system are
formed by the Co6? preparations, an emf is induced in
the secondary winding of each coil; this actuates a re-
lay which operates in the signal and locking circuit of
the inlet door of the labyrinth. Thus, entry into the laby-
rinth. Thus, entry into the labyrinth and working chamber
is only possible when all the cassettes with the Co6? pre-
parations are located in lower part of the channelsttfi.
To prepare the apparatus for radiation-chemical
experiments the operator installs and fixes both the tubes
of the irradiating element and the ropes with fitted elec-
tromagnets according to the selected configuration of the
radiating elementtn. When the objects to be irradiated
have been placed in the guard vessel, the cables can be
fed through the central tubel. The apparatus having been
made ready for work in this way, the operator in the work-
ing chamber switches on the locking device as a result
of which the core of the solenoid shutter of the door2 is
attracted and held for 20 sec. This period is sufficient
for the operator to leave the chamber and close the door.
The power supply of the electromagnets3is switched on
to raise the required number of cassettes with Co6? pre-
parations, by means of the toggle switches on the control
desk.
After the power supply has been switched on, the
cassettes are drawn toward the electromagnets by a plug-
yoke and are raised to the working position by means of
the displacement mechanism. The signal lamps serve as
The winding of the coils is made of mark Pg.TKSO-
0.35 wire, resistant to the action' of y rays, impregnated
with organosilicon lacquer.
1-1-1' In addition to the locking device by means of the
induction transducers, the USID-1 dosimetry device oper-
ates in the locking circuit of the inlet door. When the
door is opened the electromagnets are automatically dis-
connected by the key, and if cassettes with Co6? are pre-
sent in the radiating element they fall into the store via
the channels. In this way additional locking is obtained
during functioning of the apparatus.
M The change-over from one variant of radiating ele-
ment to another is carried out under safe conditions in
30 min.
'For any variants of the radiating element the objects to
be irradiated can be placed on the ends of the working
table and on the floor of the working chamber.
2When the core of the solenoid shutter is lowered the
door to the labyrinth can only be closed if the time re-
lay (the button of which is located in the chamber of
the apparatus) is switched on; this eliminates the possi-
bility of the presence of personnel in the chamber when
the source is raised.
3The supply for the electromagnets is obtained from an
ordinary buffer circuit with a VSA-6 rectifier and a 384
x 2 amp/lir storage battery.
373
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a means for indicating when each cassette has been lifted
into the tube of the radiating element.
When the experiment has been completed all the
operations are carried out in the reverse order; the irradia
ted objects can be removed from the chamber for further
investigation and the apparatus prepared for the next ex-
periment.
Method of assembling the powerful y -ray source of
the K-60,000 a2paratus. The y -ray source, consisting of
80 standard Co6? preparations, is assembled by means of
a special transport container, similar to that developed 5.
by us previously [8], by the method described in [10, 11].
The container is installed on a plate located in the room
above the working chamber of the apparatus. From the
center of the plate a pipeline is passed through the con-
crete ceiling; this pipeline is first connected in the work-
ing chamber to one of the channels of the store contain-
ing the empty cassette. Four Co60 preparations are passed
in succession into each cassette via the pipeline; the plug-
yoke is then screwed into the cassette by means of an
electromagnetic screw driver. When the plug is screwed
up (or unscrewed) each cassette is held securely in a hex-
agonal jack in the lower part of the channel.
The authors wish to express their thanks to V. I.
Vainshtein, M. A. Dembrovskii, and N. P. Syrkus, who
took part in the discussion of individual problems of the
design of the apparatus, and also to A. I. Dombrovskii,
V. M. Kasatkin, A. V. Tatov, D. V. Yasinskii, I. A.
Gromov, V. G. Ivanov, M. N. Demichev, V. V. Serebrya-
kobaya, N. V. Troinov, and others, who took part in the
construction and assembly of the apparatus.
2.
3.
4,
LITERATURE CITED
1. S. S. Medvedev, Proc. All-Union Scientific-Tech-
nical Conference on the Use of Radioactive and
Stable Isotopes and Radiations in the National Eco-
nomy and Science (April 4-12, 1957). Istopes and
374
6.
Radiations in Chemistry [in Russian] (Izd. AN 8SSR,
Moscow, 1958) p. 85.
?A. I. Topchiev, I. G. Alad'ev, and P. S. Savitskii,
Atomnaya &ergiya 5, 3, 321 (1958).4
A. Kh. Breger. Problemy Fiz. Khim, 1, 61 (1958).
Yu. S. Ryabuldiln and A. Kh. Breger, "Simulation
of isotopic radiation sources of possible industrial
radiation-chemical apparatuses," Proc. First All-
Union Conference on Radiation Chemistry [in Russian]
(Izd. AN SSSR, Moscow, 1958) p. 318.
A. Kh. Breger et al.,Problemy Fiz. Khim. No. 2,
132 (1959).
A. V. Bibergal', M. M. Korotkov, and T. G. Ratner,
Atomnaya inergiya 2, 3, 244 (1959).4
'7. B. I. Vainshtein, A. Kh. Breger, and N. P. Syrkus,
Zhur. Khim. Prom-ti, 7, 6(1959).
8. N. P. Syrkus, A. Kh. Breger, and B. I. Vainshtein,
Zhur. Khim, Prom-ti, 8, 1 (1959).
9. A. Kh. Breger et al. in: Symp. The Action of Ioniz-
ing Radiations on Inorganic and Organic Systems. .
Ed. by S. Ya. Pshezhetskii [in Russian] (Izd. AN SSSR,
Moscow, 1958) p. 379,
10. A. Kh. Breger et al., Proc. All-Union Conference
on the Use of Radioactive and Stable Isotopes and
Radiations in the National Economy and Science
(April 4-12, 1957). The Preparation of Isotopes.
Powerful y Ray. Appalatus. Radiometry and Dosi-
metry [in Russian] (Izd. AN SSSR, Moscow 1958)
p. 182.
A. Kh. Breger et al., Report No. 29, UNESCO Inter-
national Conference on the Use of Radioisotopes in
Scientific Investigations (Paris, September, 1957).
12, G. S. Murray, R. Roberts, and D. Dove, Report No. 19,
UNESCO International Conference on the Use of Radio-
isotopes in Scientific Investigations (Paris, September,
1957).
13. 0, Joklik, Atompraxis 10, 10, 355 (1958).
40riginal Russian pagination. See C. B. translation.
11.
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Declassified and Approved For Release 2013/02/19: CIA-RDP10-02196R000100050005-4 ?
Letters to the Editor
INVESTIGATION OF THE SPENT FUEL ELEMENT
OF THE FIRST ATOMIC POWER STATION*
A. P. Smirnov-Averin, V. I. Galkov, Yu. G.
Sevast'yanov, N. N. Krot, V. I. Ivanov, I. G.
Sheinker, L. A. StabenoVa, B. S. Kir'yanov,
and A. G. Kozlov
Translated from Atomnaya Energiya, Vol. 8, No. 5, pp. 446-447,
May, 1960
Original article submitted January 28, 1960
In designing new nuclear reactors for power stations
and in developing new operating parts, it is necessary to
study the changes occurring in fuel elements (FE) during
their operation. Thorough investigations of spent FE make
it possible to design power station reactors where the
nuclear fuel is utilized with maximum efficiency. One
of the stages in studying spent FE is the investigation of
the nuclear fuel isotope composition after operation in
the reactor, the burn-up variation along the FE length,
and the state of the outside and inside FE jackets.
We investigated the fuel element of the First Atomic
Power Station [1], which was in operation over an effec-
tive period of 104 dayst and then stored over a period of
1160 days after the end of the run. It was transferred from
the storeroom of the First Atomic Power Station to the
unloading room of the "hot" laboratory and then placed
into a protection cell by means of the remote-handling
unloading mechanism.
The fuel element was observed by means of special
binoculars through the lead glass of the cell. A thin oxide
film was detected on the outside jacket. This jacket was
not damaged. In measuring the outside diameter at differ-
ent spots along the FE length by means of a remote-meas-
urement micrometer, it was found that the jacket was
deformed. As a result of irradiation the average diameter
along the FE length increased from 14.11 ? 0.02 to 14.20?
? 0.02 mm.
In the course of further investigations, six specimens
1 cm long were cut from different spots along the FE
length.
An inspection of the specimen inside tube on the
side where it was in contact with the coolant (water) re-
vealed a brown deposit. By means of an MIM-6 metal
microscope, it was established that the deposit had a
thickness of 1 i. The metal structure at the metal?de-
posit interface was not disturbed, which indicated that
the layer consisted of scale and that it was not a product
of stainless steel corrosion.
The burn-up was determined with respect to the ac-
tivity of Cs', which evolved from the specimen. Cesium
Is the most suitable fragment for this purpose, since its
yield is well known [2] and its half life is long, due to
which the corrections used in calculating the burn-up are
small. The chromatographic method was used for separat-
ing cesium. The purity in chemically separating Cs137
was controlled by means of a scintillation y spectrometer
as well as by measuring the 8 spectrum according to the
absorption method. The absolute 8 activity of Cs137 was
measured by means of .a 4ir counter.
In burn-up calculations we took into account the
Cs137 activity due to Pu2" fission as well as the reduction
in U235 nuclei due to the radiation capture of neutrons.
Data on the burn-up at different FE spots are shown in
Fig. 1. The average burn-up is equal to 12.5%.
For the specimen taken from a spot 95 cm from the
lower FE end, the burn-up was determined by means of
a mass spectrometer. The uranium content in this sped-
20
5
20 40
60 80 100 120 140
Length, cm
Fig. 1. Burn-up along the FE length.
*First communication.
t The investigation of elements with deeper burn-up is
in progress.
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160
375
3
2
1
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Data
Pu238
Pu239
Pu2a0
P11211
AM241
,
Experimental
' 2,54.10-4
1,20
0,102
1,27.10-2
1,86.10-3
Theoretical
?
1,17
0,100
1,36-10-2
?
al)(1081
Bottom
Top
20
15
10
20 40 60 80 100 ' 120 ' 140 ' 150
Length, cm
Fig. 2. Variation of over-all a, 8, and y activities along
the FE length.
2,0
1,f
Bottom
40(x
yPu 238( 103).
Q,n24
Itx
Top
20, 40 60 80 100
Length, cm
120 140 150
Fig. 3. Distribution of the content transuranic element iso-
topes along the FE length.
men was 4.32%, which corresponded to a burn-up of
16.1%. This value was in good agreement with the value
obtained with respect to the Cs137 yield.
The over-all a activity was measured by means of
a standard Da-49 device, and the over-all 6 activity was
measured by means of a 4ir counter . The y -radiation
activity was measured by means of an ionization chamber,
which was calibrated with respect to the y activity of a
radium standard. The measurement results are shown in
Fig. 2. As was to be expected, the shape of the relative
376
family of the over-all 6-and y -activity curves was simi-
lar to the shape of the burn-up curve.
The content of transuranic element isotopes was
determined with respect to a-radiation spectra and with
respect to the number of spontaneous fission events. The
spectra were obtained by means of an ionization a spec-
trometer, which consisted of an ionization chamber with
a small collecting electrode (a ball 1.5 cm in diameter),
an amplifier, a discriminator, a second amplifier, a sta-
bilized rectifier, and a 50-channel pulse analyzer. The
a-spectrometer resolving power was 2%.
Complete data on the content of transuranic element
isotopes cannot be obtained directly from the spectrum
of the specimen a radiation, since the spectral peaks of
Pu239 and Pu24?,as well as Pu238 and Am241,coincide with
each other.
After we measured the a-radiation spectrum of
plutonium that evolved from the specimen, we succeeded
in separately determining the PU238 and Am241 content.
The Pu249 content was determined with respect to the
number of spontaneous fission events produced by the
layers prepared from the evolved plutonium.
Figure 3 provides data on the content of transuranic
element isotopes (in kilograms per 1 ton of uranium) at
different spots on the FE. The Pu241 content was deter-
mined theoretically by using data obtained from Am.
On the basis of experimental data, we determined
the amounts of various isotopes (in kilograms per 1 ton
of uranium), which were averaged with respect to the
FE length. The theoretical and experimental results ob-
tained in measuring the FE isotope composition are shown
in the table. A comparison of these results indicates that
? the theoretical and experimental data are in good agree-
ment.
The authors extend their thanks to G. M. Kukavadze
and R. N. Ivanov, who performed the mass-spectrometer
analysis of irradiated uranium, and also to V. N. Sharapov
who calculated the FE isotope composition.
LITERATURE CITED
1. D. I. Blokhintsev, N. A, Dollezhal', and A. K.
Krasin, Atomnaya energiya, 1, 10 (1956)4
2. S. Katcoff, Nucleonics 16, 4, 78 (1958).
tOriginal Russian pagination. See C. B. translation.
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Declassified and Approved For Release 2013/02/19: CIA-RDP10-02196R000100050005-4
ON IMPROVING THE EFFICIENCY OF POWER STATION
REACTORS WITH GASEOUS COOLANTS
T. Kh. Margulova and L. S. Sterman
Translated from Atornnaya ?Energiya, Vol. 8, No. 5, pp. 448-451,
May, 1960
Original article submitted September 3, 1959
An increase in the gaseous coolant temperature tc,1
at the reactor downstream end always leads to an increase
in the reactor efficiency. However, the choice of the
maximum temperature depends on the material of which
the fuel-element jacket is made. If the material cost is
high and is not justified by an increase in the cycle ef-
ficiency, it is advisable to use lower coolant temperatures.
The gas temperature te,2 at the reactor upstream end
cannot be determined without a careful analysis of the
power station thermodynamic layout. The power station
efficiency is usually the highest for a large gas tempera-
ture drop between the reactor upstream and downstream
ends (of the order of 150 to 200?C and over). If a two-
pressure operating cycle instead of a single-pressure cycle
is used in the power station layout, the resulting complica-
tions and the higher cost are entirely justified by the ad-
vantages offered by an increase in efficiency. The use of
a three-pressure operating cycle is hardly advisable, since
in this, the efficiency is very little improved in compari-
son with the case of the two-pressure cycle, and the power
station layout becomes much more complicated.
Figure 1 shows the dependences of the efficiency
(71e) on the vapor pressure Php in the high-pressure loop
for a power station with a two-pressure operating cycle.
The curves were plotted for the case without regenerative
?le
0,28
0,27
0,26
0,25
0,24
0,23
10 20 30 40 50 CO Php atm abs
Fig. 1. Variation of ne in dependence
on the vapor pressure Php for the follow-
ing temperatures tc,2 (?C): 1) 140;
2) 200;
preheating of the feed water, when the gas temperature
at the reactor downstream end was equal to 375?C. Each
value of t0,2 corresponds to a Ph value that secures
the optimum efficiency. The initial points of these curves
correspond to the Tie value for the single-pressure opera-
ting cycle.
In reactors with gaseous coolants, for constant tc,i
and te,2 and a chosen value of php, the vapor pressure in
the low-pressure loop drops with an increase in the feed
water temperature tfw, and the quantity of this vapor,
reduced to the same quantity of high-pressure vapor, in-
creases. This negatively affects the power station effi-
ciency. Therefore, regeneration exerts a positive influence
only in a certain definite feed water temperature inter-
val, which depends on the number of preheaters; the opti-
mum tfw value is lower than that for ordinary power sta-
tions.
Figure 2 shows the efficiency vs. feed water tempera-
ture curves for different numbers of regenerative preheaters
for single-pressure (a) and two-pressure (b) cycles.
Figure 3 shows the dependences of the maximum ef-
ficiency value on the coolant temperature to for single-
pressure (curves 1 and 3) and two-pressure cycles (curves
2 and 4) in the presence (4eT solid curves) and absence
(nrx, dotted curves) of regenerative preheating. It can
be seen that regenerative preheating for the single-pressure
operating cycle is beneficial only if the tc.2 value in the
reactor exceeds ?170?C; for the two-pressure cycle, re-
generative preheating leads to an increase in efficiency
even for lower t 2 values. Obviously, in deciding on the
C,
power station layout, it is necessary to determine to what
extent the advantages resulting from the use of regenera-
tion with a given number of preheaters exceed the addi-
tional costs involved.
For higher tc.2 values, the efficiency as well as the
coolant circulation losses increase. The efficiency cal-
culated by taking into account these losses is determined
by the equation
H= _Ne?KNe Ne?KNe
1)
Qvap?Kriblmotil bl mechPle . (
where Q. is the reactor heat output, Ne is the reactor
electrical output, l< is the portion of the electrical energy
377
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Declassified and Approved For Release 2013/02/19: CIA-RDP10-02196R000100050005-4
ite,n
0,30
rle,n
0,28
0,27
0,25
0;25o .
23
0,29
0,28
0,270
50 100 t fw, ?41
a
It
50 100 150 tfw
Fig. 2. Variation or /leo.' in dependence of the feed water temperature tfw and the number
of regenerative preheaters z (mixing-type preheaters): 1) z=1; 2) z=3; 3) z=5; 4) z=10;
tc,1=375"C; te,2=200?C.
max
,
qe,n
0,31
0,30
0,29
0,28
427
0,26
0,25
0,24
0,23
3
1
140 160 180 200 220 240 te,2,?C
Fig. 3. Dependence of the,maxi-
mum values n Tax and n emnax on
the coolant temperature tc,2 for
the constant value of to,1= 375?C.
Ne expended on coolant circulation, Qvap is the heat
transferred to the vapor generator, nbl mech is the blower
?
mechanical efficiency, and 71b1 mot is the blower motor
efficiency.
Equation (1) can be reduced to the following form:
(1?K)ne
e 1-- Knell bl motnbl mech
(2)
The maximum efficiency values, (nell, n), calculated
by taking into account circulation losses in the presence
378
of feed water regenerative preheating for a single-pressure
(a) and a two-pressure cycles, are given in Fig. 4. The
variation of rig n in dependence on to,2 for constant loop
hydrodynamic characteristics and for a constant reactor
heat output is shown by dotted lines. Under these condi-
tions, the following relation holds:
(3)
K"?ltc, 1?.1C, 2 ) ) p, ) Tle,n
where K' is the fraction power expended on coolant cir-
culation, y is the average coolant specific weight, Cp,1
is the average coolant specific heat, and n' is the elec-
trical efficiency without taking into account the piping
losses for t" ? 2 K" V2' Cp,2,arld ne",n are the values of
C'
the same qu,antities for *,2.
If we know the fraction power K' for the given opera-
ting conditions, we can determine its value K" for the
same reactor output, but for a different value of tc,2.
As can be seen from Fig. 4, there are optimum tc,2
values which correspond to the maximum efficiency value
where coolant circulation losses are taken into account.
An increase in the coolant temperature at the re-
actor upstream end initially leads to such an increase in
efficiency that Ile% also increases (regardless of higher
circulation losses). Subsequently, increasing circulation
losses lead to the fact that lie Hn begins to decrease with
an increase int It is obvious from Fig. 4 that the opti-
mum H values for the single-pressure operating cycle
e,n
correspond to te,2 values higher than the tc,2 value for
the two-pressure cycle. The fraction power expended on
coolant circulation in the single-pressure cycle is also
higher than that in the two-pressure cycle under condi-
tions securing the optimum efficiency, and the efficiency
value where circulation losses are taken into account is
always lower.
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i?g,n
428
427
0,26
0,25
0,24
0,23
0,22
0,21
140 160
0
rS
/1
/I
AFJ
V
/
r /
r/
?yr/
//r
/
.
180 .2011
a
i?e8,n
0,29
0,28
0,27
0,26
0,25
0 24
220 2 ,?C 140 160
%
o
o
%
\N
- o? ,
MY
-
iP
Wirg-FriP'
AWAPPArv?1>
////_70?7
IP
\
. /
180 200 ty,T
Fig. 4. Dependence of n on the coolant temperature for various values
of the fraction power expended on coolant circulation (tc,f-375?C).
e,n
0,27
0,26
0,24
0,23
Fror,
torm
issler%,
mow
v?,
rharir\
EMT
aria
a
An
will.
150 180 200 to ,2,?C
a
ne,n
0,30
0,29
0,28
0,27
0,26
0,25
pill
lUll, AA
Illp
A
MON
MEMON
MOM
A
Wraggr
graggin
MOMMEMA
WIRMitil
MEM
?
ra7Jill
150 180 200 220t
Fig. 5. Dependence of ne,n on the coolant temperature tc,2 for various
relative values of the fraction power expended on coolant circulation
(two-pressure cycle): a) te.1=340?C, b) te.1=400?C.
379
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The above analysis pertains to conditions where
=375?C. In order o determine the effect oft
tc,1 t ffvalues
c,1
on the power station efficiency, we also performed cal-
culations for different tc.1 values.
Figure 5 shows the neHn vs. tc,2 curves. From Figs.
4b, 5a, and 5b, it is obvious that, for equal relative cir-
culation losses, eFin increases with an increase in the gas
temperature at the reactor downstream end, and the maxi-
mum efficiency values are obtained for higher tc.2values.
We are justified in increasing the reactor heat out-
put if the increase in the fuel cost in producing electrical
380
energy due to a lower power station efficiency is com-
pensated by lower capital expenditures. For a given re-
actor heat output, it is necessary to strive for the maxi-
mum power station efficiency while taking into account
circulation losses. It is necessary to bear, in mind that
the maximum efficiency value decreases with an increase
in the heat output of a given reactor, i.e., as can be seen
from Figs. 4 and 5, the lower dotted curves correspond
to large reactor heat output values.
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Declassified and Approved For Release 2013/02/19: CIA-RDP10-02196R000100050005-4
MEASUREMENT OF THE FAST NEUTRON FLUX
DISTRIBUTION IN THE CORE OF THE VVR-S REACTOR
WITH RESPECT TO CHANGES IN THE ELECTRICAL
CONDUCTIVITY OF GERMANIUM SPECIMENS
E. Aleksandrovich and M. Bartenbakh
Institute of Nuclear Research, Polish Academy of Sciences, Warsaw
Translated from Atomnaya Energiya, Vol. 8, No. 5, pp. 451-452,
May, 1960
Original article submitted December 28, 1959
In connection with the investigations of the behavior
of semiconductors under neutron irradiation that are
in progress at the Institute of Nuclear Research in Warsaw,
it was necessary to determine the fast.neutron flux distri-
bution in the reactor core.
Under the action of fast-neutron irradiation, the
electrical conductivity of a germanium monocrystal of
the n type gradually decreases, and, in the final stage,
the germanium monocrystal is converted to the p type
[1-7]. The typical curve representing the variation of
germanium electrical conductivity in time during irradia-
tion by fast neutrons is shown in Fig. 1. The initial portion
of this curve can be considered as linear, which is notice-
able only in the case where the neutron energy exceeds
?300 ev.
Nine germanium specimens cut from the mono-
crystal were used for measurements. The specimens were
in the shape of parallelepipeds whose dimensions were
1.5x 1.5x10 mm. They were selected in such a manner
that their electrical conductivity was at the maximum.
In order to eliminate the influence of thermal neu-
trons, each specimen was enclosed in a cadmium shell
Time, hr
Fig. 1. Variation of the electrical conductivity (a) of
a germanium specimen (monocrystal) in time during
irradiation by fast neutrons.
0.25 mm thick. The specimens were placed at equal
distances from each other in a special aluminum probe
and then introduced into the core of the VVR-S reactor
between the fuel elements. Thus, seven specimens were
distributed along the reactor core height and two speci-
mens were placed outside the core: one was above and
the other below it.
The electrical conductivity of the specimens in the
reactor was measured during irradiation with respect to
the current intensity through the specimens; the current
was supplied by a dc voltage source. The current intensity
was such that the specimens were not heated during meas-
urements. The reactor power, equal to 1.5 kw was kept
-constant.
1,8
1,5
1,4
1,2
1,0
0 0,8
0,5
0,4
0,2
0,0
X
t''`'.41111
14Nh---
into reactor I
Probe
moved /
Probe /
moved.,
Probe inserted
0
2 Time, hr
Fig. 2. Variation of the electrical conductivity (a) of
germanium specimens lowered to various depths into
the core of the VVR-S reactor during irradiation by
fast neutrons.
381
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Declassified and Approved For Release 2013/02/19: CIA-RDP10-02196R000100050005-4
The variation of the electrical conductivities of all
specimens during irradiation is shown in Fig. 2. A sharp
change in the magnitude of da/dt at the points t=1 hr and
t=2 hr was caused by changing the probe position in the
reactor.
The slope of the a (t) curve is the measure of the
intensity of the fast neutron flux at the point where the
specimen is located. Figure 3 shows the fast neutron flux
distribution along the core axis. In plotting this curve,
we used the time interval enclosed between the points
t=1 hr and t=2 hr in Fig. 2.
This method can also be used for absolute measure-
ments.
Effect of
control
rods
Bottom
Core height =
= 50 cm
9
Top
Fig. 3. Fast-neutron flux distribution along the VVR-S
reactor core axis.
382
LITERATURE CITED
I. R. Davis et al., Phys. Rev. 74, 1255 (1948).
2. W. Johnson and K. Lark-Horovitz. Phys. Rev. 76,
442 (1949).
3. J. Crawford, jr., K. Lark-Horovitz, Phys. Rev. 18,
815 (1950); Phys. Rev. 79, 889 (1950).
4. J. Cleland et al. Phys. Rev. 83, 312 (1951). Phys.
Rev. 84, 861 (1951).
5. H. James and K. Lark-Horovitz. Z. phys. Chem.
198, 107 (1951).
6. J. Cleland, J. Crawford and J. Pigg. Phys. Rev. 98,
1742 (1955).
7. B. Buras et al., Report No. 101/I-B. Institute of
Nuclear Research, (Warsaw, 1959).
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Declassified and Approved For Release 2013/02/19: CIA-RDP10-02196R000100050005-4
CALCULATION OF THERMAL SHOCKS IN REACTOR
STRUCTURAL PARTS
Yu. E. Bagdasarov
Translated from Atomnaya Energiya, Vol. 8, No. 5, pp. 452-454,
May, 1960
Original article submitted July 13, 1959
Thermal shocks in the walls of reactor structural parts
arise due to a sudden drop in the coolant temperature
at the core downstream end after reactor emergency
shutdown. Since sharp temperature changes lower the
strength of such important and irreplaceable parts as
tanks, collectors, and pipes, it is necessary to know how
to calculate the thermal shock in reactor structural parts
if power station devices are to be designed correctly.
The thermal shock effect has been studied by a number
of investigators [1-3],
An effective method for protecting the walls of
important reactor structural parts from thermal shock is
the provision of a thermal screen between the wall and
the coolant.
The wall thickness of tanks, collectors, and pipes
used in reactor loops is considerably smaller than their
diameters. From the point of view of changes in tempera-
ture fields and thermal stresses, these walls can be consi-
dered to be flat. Therefore, we shall hereafter consider
only the two-dimensional geometry. For technical reasons,
in the majority of practical cases, the carrying walls
and the thermal screen are made of the same material.
This makes it possible to obtain analytical equations for
the calculation of thermal stresses if we assume that
ideal thermal contact exists between the thermal screen
layers and between the thermal screen and the carrying
wall and that liquid layers are absent. The neglect of
the thermal resistance and the heat capacity of the cool-
ant layers lead to somewhat higher thermal stress values
in the carrying wall, which provides a safety margin in
calculations.
The temperature changes in the plane of the wall
cross section are found from the following system of
equations [4]* :
a2t at
a --- - ?
aX2 aT'
tit= 00 = const;
at
BxIxO
at
-ax--1 x=o =a [t/x=6? 0(v)].
The solution of this system is given by the equation
t (x, r) ?0 (T)-1- 11, A (x) cos (8
n 6
where
bE2 T beAt
A, (T).--- ?Bne -e
'
4 sin En
Bn-
2en+ sin 2e?
The proper values of a n are found from the transcendental
equation tg ?n=Bik n, and, for Bi > 50 we can assume
that ? n=(2n- 1)(7r/ 2),where n=1, 2. . ..The thermal stress
on the outside surface of the kth layer is determined
from the relations [5]
E.., ?
CI (T)= a1"--; [th (T) (Ch, i)];
ck
ch
eh 1 vt, t (x, or) dx.
"Vh
Each form of the 0(r) function yields a separate
form of the expression for ock(T). The table provides the
final formulas for the cases of linear and exponential
changes in 0(r). As an example, the figure shows the
results obtained in calculating thermal shocks in the carry-
ing wall and the thermal screen of a fast reactor vessel
with sodium cooling [6].
The following initial data were used in calculations:
the material was Steel 1Kh18N9T; 0(T)=00t, 00e-mr ;
6, 00=140?C; m=0.1 sec-1; 6 =0.2m; Y10; c1=0.05 m; 7N
=0.19 m; cm=6 =0.2 m.
The following notation was used in these calculations:
t(x,T) is the carrying wall and the thermal screen tempera-
ture (?C); 00, 0(T), and 0.0 are the coolant temperatures at
the initial instant of time, at the time r, and the asymp-
? The notation used is explained at the end of the article.
383
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0, kg/cm2
4000
1000
2000
1000
0 10 20 30 rbsec
a
6, kg / cm2
200
100
0 500 1000 1500 r, sec
6, kg/cm2
2000
1000
0 10 20 30 t", .sec
Thermal shock calculation. a) Carrying wall without thermal screen; b) carrying wall with thermal screen;
C) end layer (in contact with sodium) of the 10-mm thermal screen.
Calculation Formulas for Determining Thermal Stresses
Coolant
temperature
Equation for calculating thermal stresses
For exponential
changes
0 (t)=-0+6,00e-r"
a (t)._\I I 1 n"1
-r" -- e-b841.]
-_-A6,00 x
ch L.IbEA?m
n=i
ek
sin(en?.5 ) --sin (t e, Yit )
6
cos (en 75-ch
eh? Vk
i5
For linear
changes.
0 CO= 00- pv
* Linear temperature
-it
changes
. sin (eri -Let,(e, it)
1 6
B
ch I
eePT0-11 e-be4v
_bov
[1 j
qin ch ? yk
6? en
for v < To;
.
a A 12-
(t). E --B X
ch b
n=1
[sin en ck )? sin (c -Y1)
11 6
15 ch
cos (en
eh? ?it a
6 en
for "r ->- TO
occur during the time r0, when 0 = 0,0.
totic value of the coolant temperature for 00 (?C),
respectively; 00=00- 0 00 is the maximum change in
the coolant temperature (?C);Tk(T) is the average temp-
erature of the kth layer (?C); T is the time (sec); T0 is
the time during which the coolant temperature drops
linearly (sec); x is the coordinate measured from the ves-
384
sel outside surface toward the interior (m); S is the over-
all thickness of the carrying wall and of the thermal screen
sheets (m); ck is the coordinate of the kth layer outside
surface (m); y k is the coordinate of the kth layer inside
surface (m). Here, the carrying wall represents the first
layer.
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The structural material parameters are the following;
a is the diffusivity coefficient (m2/sec); X is the thermal
conductivity coefficient (kcal/mxsecx?C);aw is the line-
ar thermal expansion coefficient (1/?C); Ew is the elasticity
modulus (kg/cm2); v is the Poisson coefficient; cc is the
coefficient of heat transfer from the wall to the coolant
( kcal/m2x sec x ?C); b=a/8 2; Bi=a8 A ; A= otwEw/ 1?u; a ek(T
is the thermal stress at the kth layer outside surface
(kg/cm2); 2 is a parameter (deg/sec); m is a parameter
(sec 1)
LITERATURE CI TED
1. R. Tidball and M. Shrut, Transactions of the ASME
76, 4, 639 (1954).
2. A. A. Klypin, Teploenergetika, 1, 33 (1957).
3. M. Heisler, J. Appl. Mech. 20, 2, 261 (1953).
4.
5.
6.
A. V. Lykov, Theory of Thermal Conductivity [in
Russian] (Gostekhizdat, Moscow, 1952).
S. P. Timoshenko, Strength of Materials [in Russian]
(Gostekhizdat, Moscow, 1946) Vol. 2.
A. I. Leipunskii, et al., Transactions of the Second
International Conference on the Peaceful Uses of
Atomic Energy (Geneva, 1958). Reports by Soviet
scientists; Nuclear Reactors and Nuclear Pdwer
Engineering [in Russian] (Atomizdat, Moscow, 1959)
Vol. 4, p. 212.
385
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600?key PROTON INJECTOR FOR A LINEAR ACCELERATOR
Yu. N. Antonov, L. P. Zinov'ev, and V. P. Rashevskii
Translated from Atomnaya Energiya, Vol. 8, No. 5, pp. 454-457,
May, 1960
Original article submitted December 10, 1958
The necessity for increasing the injection current
at the 10-Bey proton synchrotron at the Joint Institute
for Nuclear Studies prompted the authors of the pre-
sent paper to undertake the following problem: using
an ion source capable of furnishing a large proton cur-
rent, extract and focus this current for acceleration to
an energy of 600 key. In doing this it is necessary that
the geometric dimensions and angular convergence of
the beam at the output of the accelerator tube satisfy
the injection conditions in the linear accelerator.
Ion source. This is a gas discharge with a dou-
bly contracted pinch [1] in which' electron oscillations
are used. The source was developed in NII-5 and then
modified by the authors and V. M. Blagoveshchenskii,
T. I. Gutkin, and Yu. V. Kursanov of the High-Energy
Laboratory of the Joint Institute for Nuclear Studies.
A schematic diagram of the source is shown in Fig. 1.
In this source we used cathodel of a TGI-90/8 thyraton.
The cathode leads 2 are cooled by circulating water and
isolated from the cathode flange 5 by porcelain insulators
4. The hydrogen is introduced into the discharge chamber
through a palladium filter and tube 3.
The internal cavity of the intermediate anode 6
forms the cathode region of the discharge; the presence
in the intermediate anode of the channel 15, 9 mm in
diameter and 10 mm long, causes the formation of a
double layer of hemispherical shape. The spherical part
points toward the cathode and serves to contract the pinch.
The intermediate anode is also cooled by water which
flows in the cavity '7. Directly on the intermediate anode
there is a coil 8 that produces an axial magnetic field
that contracts the plasma in the space between the
channel 15 and the emission aperture 17. The intermediate
anode 6, the wall of the source, 16 and the output flange
12 form an open magnetic conductor. The copper anode
10 has an aperture 9 mm in diameter which is concentric
with the channel 15 and the emission aperture; the elec-
tron oscillation takes place in the space between these
two electrodes. The output flange of the source (anti-
cathode) has an insert 13 made from nonmagnetic mate-
rial (tungsten or stainless steel) with an emission aperture .
2 mm in diameter and approximately 0.5 mm long.*
Since the source is at a potential of 600 kv with respect
to ground the entire power supply circuit, which is moun-
ted on porcelain insulators 2.6 m in height, has its own
386
generator (220 v, 50 cps, 3.5 kw) and a GSM-1 generator
(220 v,500 cps, 0.6 kw). The drive that connects the
generator and the motor, which is at ground potential, is
a shaft made from 6 wood which is approximately 2 m
in length.
The power supply for the source is shown in Fig. 2.
The filament supply for the cathode of the source comes
from transformer Tr The filament requires 7.8 v at 6.4
amp. The hydrogen pressure in the discharge chamber is
maintained at about 1.5'10-2 mm Hg; the hydrogen is
admitted through the palladium filter Pa and the flow
rate is controlled by heating the filter.
The magnetic field in the source is produced by a
coil with 320 turns. The magnetic induction in the space
between the intermediate anode and the anticathode is
1000 gauss for a coil current of approximately 2 amp.
When the artificial line is discharged through the
thyratron a negative pulse approximately 500 ?sec long
appears at the cathode; the amplitude of this pulse is con-
trolled by the voltage from the power supply B2 which is
used to charge the line. The thyratron is fired by a pulse
14
5
16 14 14
011602I
4,1!
2
4
A F.ircwirACT
irg?0
za
6 8
Cop- rma
per vzzajSteel
eitkl
10 11 s=.
Porce- ma Rubber
lain Nom gasket
17
13
12
Fig. 1. Schematic diagram of the source; a) copper,
b) steel, c) porcelain, d) rubber gasket.
? If the emission aperture is a channel the loss of ion
current to the wall increases appreciably; hence, the
condition bcd must be satisfied for the aperture.
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which is obtained from a photomultiplier which is moun-
ted on an insulator; this tube is triggered by the flash from
an MN-3 neon lamp which is located at the base of the
insulator.
The pulse from the line (amplitude of 400-500 v) is
applied to the cathode of the source and ignites an arc
between the cathode and the intermediate anode. The
arc current produces a potential difference across the
resistance R1=200 ohm; this potential difference causes
the ignition of an arc in the channel in the intermediate
anode. Any disturbance of the uniformity of the discharge
gap in the channel causes electron drift in the region of
the anode plasma; the hemispherical double layer accele-
rates and focuses the electrons, providing the required
electron current density in the region of the anode plasma.
The potential jump in the double layer automatically
reaches a level such that the generation of ions in the
region of the anode plasma is sufficient for satisfactory
stability of the space-charge limited bipolar current;
ip ( 771
--
le
)1 / 2
where I is the directed current of positive ions; Ie is
the directed electron current; me and mp are the mass
of the electron and positive ion,respectively.
The strong inhomogeneous magnetic field in the
region of the anode plasma causes a still greater contrac-
tion of the discharge, leading to an additional increase
in the density of carriers. The double contraction of the
discharge and the use of electron oscillation in the space
between the intermediate anode and the anticathode
causes an appreciable degree of ionization in the anode
region of the discharge. The current extracted from the
source can reach values of 270 ma. Knowing the pass
spectrum of the beam and the gas consumption in the
source (source parameters are given below) we can show
that the ionization percentage in the discharge is close
Tr2
220v. 31
50 cps
220v
50 cps
Tri
TGI-1-90/8
FEU
82
Li
MN1)-4,
Fig. 2. Power supply for the source (R1=200 ohm; R2=
=R3=12 ohm; R4=5.10 3 ohm; C=10 f; L=25mh).
to 100/0. Under the effect of the electric field the ions
generated in the anode plasma are directed toward the
emission aperture of the source and extracted by the ex-
tractiom system.
The geometry of the discharge chamber of the source
is chosen experimentally for optimum conditions for ignit-
ing the pulsed arc. The opening angle of the cone of the
intermediate anode in the working version if 120?. If this
angle is varied the current extracted from the source falls
off sharply since the axial magnetic field has a strong
effect on the electron kinetics and parameters of a low-
pressure plasma [2].
When the diameter of the emission aperture is in-
creased from 0.8 to 2.0 mm the current increases approxi-
mately as the square of the diameter (Fig. 3). However,
when the diameter of the emission aperture is increased
to 3.0 mm it is difficult to ignite the arc in the source
because the field of the extraction electrode "drops" in
the discharge chamber. A very important factor is the
presence of the nonmagnetic insert 13 (cf. Fig. 1). The
absence of this insert causes a noticeable reduction in
the extracted current and deteriorates the focusing. In
the working model the diameter of the insert is 10 mm.
Some increase in current is achieved if the diameter of
the insert is increased to 20 mm.
The current at the output of the injector, as a func-
tion of discharge current, is shown in Fig. 4. For any
magnetic field configuration the contraction falls off as
the discharge current is increased, i.e. with an increase
In charge concentration. (In a discussion of the relaxa-
tion length of a Maxwellian distribution, I. Langmuir [3]
50
10
0 1,0 20
Diameter of the emission aperture, mm
Fig. 3. The current extracted from the
source as a function of the diameter of
the emission aperture.
387
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has shown that with an increase in electron concentration
causes a reduction in mean free path. Since the magnetic
field is effective over the mean free path length, the
effect of the field is reduced as the concentration is in-
creased.) In Fig. 5 we show the source mounted on the
accelerator.
The parameters which characterize the operation of
the source are as follows:
Filament voltage, v
7.8
Filament voltage, amp
6.4
Arc voltage, v
90-120
Arc current, amp
30
Magnetic field in the gap, gauss
1000
Gas pressure in the discharge
chamber, mm Hg
1.540-2
Gas consumption
150
Proton component of the extracted
current, percent.
75
cvi
20
20
10
0 10 20 30 0 jam, amp
Fig. 4. The extracted current as a function of
arc current. (pH2=1.2.10-5 mm Hg; Uext=45
imag=1.1 amp).
Fig. 5. The source mounted on the accelerator
tube.
388
The percentage proton composition of the beam
which passes through the linear accelerator in the absence
of resonator excitation is measured by means of a sectored
magnet (deflection angle, 750).
Accelerator, extraction and focusing of the beam.
The useful diameter of the 600-kv accelerator is 350 mm
while the length is 1670 mm. The tube has 70 irises and
every other iris has an anticorona ring which is connec-
ted to an appropriate tap on a voltage divider. The divider
is a vinyl plastic tube 8 mm in diameter with uniformly
distributed taps, along which distilled water flows (this
water is also used for cooling the source).
The source is mounted directly on the input flange
of the accelerator. The extraction and focusing electrodes
are located inside the accelerator tube and the power
supply leads are introduced through the first anticorona
rings.
The arrangement of the extraction and focusing elec-
trodes in the source and accelerator and the power supply
system are shown on Fig. 6. A TRVV pulse transformer
is used to apply a pulse to the accelerator tube. The
amplitude of this pulse is 600 kv and its length is ap-
proximately 300 usec. A special system is used to main-
tain the voltage of the flat part of the pulse with an ac-
curacy of better than 0.5%. The transformer load is the
divider D (R=430 kiloohm) part of which is used to de-
rive the 50-kv extraction voltage. The resistance of the
accelerator tube divider is approximately 5.106 ohm.
The voltage to the focusing electrode is taken from a
high-voltage power supply PF which is located on the
source insulating column. In order tofocus the beam, at
the output of the tube (1 m from the output iris) we apply
a focusing voltage of 3 kv with respect to the source. This
voltage determines the energy spread of the beam at
the output due to ion generation in the extraction region.
The energy spread in the beam at the output is less than
0.5%.
The beam is monitored visually at the output of the
tube by means of the radiation from a quartz shield. The
Fig. 6. Diagram showing the arrangement and power
supplies for the extraction electrode (1) and the
focusing electrode (2).
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ion current is measured with a Faraday cylinder; a coil
which provides a field of approximately 500 oe at the
axis of the cylinder is mounted on the Faraday cylinder
in order to protect it from secondary electrons.
The angular divergence of the beam at the input to
the linear accelerator is estimated to be.3.10-3. The dia-
meter of the spot at the input to the linear accelerator
(a distance of approximately 6 m from the output of the
tube) is 8-10 mm. With a beam focused, at a distance
of approximately 1 m from the output of the tube the
spot diameter is approximately 2 mm.
In conclusion the authors wish to take this opportunity
to thank M. S. Vasil'ev and V. V. Slesarev for taking
part in this work.
1.
LITERATURE CITED
?M. von Ardenne, Tabellen der Elektionon-physik,
Ionenphysik und Ubermikroskipie. Deutscher Verlad
der Wissenschaften (1957).
2. 0. N. Repkova and G. V. Spivak, Scientific Records:
of a Plasma in a Magnetic Field, Moscow State
University, (Izd. MGU, Moscow, 1945).
. I. Langmuir, Proc. Nat. Acad, Sci. USA, 14, 627
(1928).
389
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MEAN NUMBER OF PROMPT NEUTRONS EMITTED
IN PHOTOFISSION OF Th232 AND U238 y RAYS PRODUCED
IN THE F19 (p, ay) 016 REACTION
L. I. Prokhorova and G. N. Smirenkin
Translated from Atomnaya Energiya, Vol. 8, No. 5, pp. 457- 459,
May, 1960
Original article submitted January 3, 1960
In [1] attention has been directed to the fact that
the data available in the literature concerning the mean
number of neutrons 7 in spontaneous fission of U2" (2.1?
0.1 [1], 2.26? 0.16 [2] ) and for photofission of U238by y
rays from bremsstrahlung with a maximum energy Emax
=5.5 Mev ( 1.65? 0.5 [3]) are not in agreement with a
linear dependence of 7 on the excitation energy of the
fissioning nucleus Ex [4]. The values given for if in photo-
fission of U238 indicate that all neutrons are released in
one fission event:
v-j n) -1 2 a (y, 2n) _L.
(v, f) a (17, 1) -1
(1)
since the energy Emax7_5.5 Mev is smaller than the re-
action threshold for U'38(y .n). A similar value of 17 for
Th232, 3.15?0.5 (Emax=6.8 Mev), can be obtained from
[3]. Other information on the number of prompt neutrons
in photofission is not available in the published literature.
In the present note we report on measurements of
the mean number of prompt neutrons emitted in one
photofission event U2" and Th"2 by y rays from the re-
action F19 (p, ay) 016. The reaction is realized by irradia-
ting a CaF2 crystal with 2.6-Mev protons. They -ray
390
Values of 7; in photofission of U238 and Th292
spectrum for the F19 (p, ay )016 reaction [5] consists of
three lines: 6.13, 6.9,and 7.1 Mev. The proton energy
for the last two lines is 3.2 times greater than the com-
ponent at 6.13 Mev. The admixture of y rays with an
energy of 12 Mev for the F19 (p.7) Ne" reaction is less
0.2%. The mean energy of the y rays causing fission of
U238 and Th"2 is approximately 6.7 Mev. The measure-
ments were carried out with a van de Graaff generator.
The accelerator target and the fission chamber are
surrounded by 12 B16F3 counters in paraffin which serve
as neutron detectors. The pulses due to fission neutrons
selected by means of a coincidence circuit with a resolu-
tion time of TF=52.10 -4sec. The method of making these
measurements is well known and has been described in
detail in [6]. Accidental coincidences, which amount to
approximately 70% of the recorded coincidences, are
eliminated by measurement of the resolution time T.
The spread 7 is less than 1% for the entire measurement
time.
The number of neutrons recorded for one photofission
event in U238 and Th'' is compared with the correspond-
ing quantity for spontaneous fission of Rum? for which the
value of 70 (accuracy of 2%) is 2.26?0.05 [7]. Thus the
Ref.
Emax
Mev
kx,
Mev
U238
Th232
a(V,f)
a(v,i)
131
5,5
5,3
1,65+0,5
0
1,65+0,5
6,8
5,9
4,3 +0,5
1,15
3,15+0,7
3,15+0,5
0
3,15+0,5
8,6
6 , 6
4,5+0,5
3,65
0,85+1,0
10,3
7,7
5,3+0,5
3,8
1,50+1,6
8,0+0,5
6,25
1,75+1,0
[9]
8
6,3
7,07
1,8
5,3?0,75
4,65
2,00
2,65:170,5
9
6,8
6,92
2,5
4,4+1,0
6,35
4,35
2,0 :7E1,1
10
7,5
7,25
3,25
4,0?1,3
8,65
5,7
2,95+1,4
11
8,2
7,47
4
3,5+1,6
11,4
7,0
4,4 +1,8
*Error ?40% ?
**Error +25%.
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ratio 3/iTo is measured directly in the experiment. The
measured results are modified by certain corrections?
2%)for variation in detection efficiency for prompt
neutrons in fission of the materials being studied and
neutrons from spontaneous fission of Pu249. This difference
arises as a consequency of the angular anisotropy of the
fragments in photofission and the correlation between
the prompt neutrons and the fragments. After corrections
are introduced, the experimental values of the ratio IV
//70 are 1.86?0.09 for uranium and 1.42?0.09 for thorium.
Using the value of 70 given above we find vu=4.2? 0.2
andTh? =3 2?? 0 2
?
Starting with the data on v for fission in the neigh-
boring nuclei U239 and Th233, produced with the same
excitation energy in neutron irradiation, and the small
variation in v in going from one isotope to another [8],
one would expect substantially smaller values of 17 and
approximately 2.7 for U238 and approximately 2.2 for
Th232. Large values are obtained for V if we exclude the
fraction of (y ,n) neutrons from the total neutron yield
measured in [9]. The earlier data reported in [3] are not
in agreement with the results reported in [9]. In the table
we show values of 7 computed from the neutron yield n
[3, 9] and the energy dependence o(y ,n)/a(y .1) [10],
according to Eq. (1) (for Emax