SOVIET ATOMIC ENERGY VOL. 31, NO. 2

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• Release Decision: RIFPUB
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• Document Page Count: 137
• Document Creation Date: December 27, 2016
• Document Release Date: February 26, 2013
• Sequence Number: 1
• Publication Date: August 1, 1971
• Content Type: REPORT

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Translation published March, 1972
i Russian Original Vol. 31, No. 2, August, 1971
SOVIET
ATOMIC
ENERGY
ATOMHAH 3HEP1"I4A
(ATOMNAYA ENERGIYA)
TRANSLATED FROM RUSSIAN
CONSULTANTS BUREAU, NEW YORK
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SOVIET
ATOMIC
ENERGY.
Soviet Atomic Energy is a cover-to-cover translation of Atomnaya
Energiya, a publication of the Academy of Sciences of the USSR.
An arrangement with Mezhdunarodnaya Kniga, the, Soviet book
,export agency, makes available both advance copies of the Rus-
sian journal and original glossy photographs and artwork. This
serves to decrease the necessary time lag, between publication
of the original and publication 'of the translation and helps to im-
prove the quality of the latter. The translation began with the first
issue of the Russian journal.
Editorial Board of Atomnaya Energiya:
Editor: M. D. Millionshchikov
Deputy Director
I. V. Kurchatov Institute of Atomic Energy
Academy of Sciences of the USSR
Moscow, USSR
Associate Editors: N. A. Kolokol'tsov
N. A. Vlasov
without permission of the publishers.
A. I. Alikhanov
A A. Bochvar
N. A. Dollezhal'
V. S. Fursov
1. N. Golovin
V. F. Kalinin
A. K.'Krasin
A. I. Leipunskii
V. V. Matveev
M. G. Meshcheryakov
P. N. Palei
V. B. Shevchenko
D. L. Simonenko
V. I. Smirnov
A. P. Vinogradov
A. P. Zefirov
Copyright ? 1972 Consultants Bureau, New York, a division of Plenum Publishing
Corporation, 227 West 17th Street, New York, N: Y. 10011. All rights reserved.
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SOVIET ATOMIC ENERGY
A translation of Atomnaya Energiya
Translation published March, 1972
Volume 31, Number 2 August, 1971
CONTENTS
Engl./Russ.
Optimization Studies of Fast Reactors -A. M. Kuz'min, Yu. V. Silaev, V. V. Orlov,
and V. V. Khromov . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 787 83
Some Characteristics of Selfoscillatory Regimes of a Boiling Water Reactor
- B. V. Kebadze and V. I. Plyutinskii . . 793 89
Dissociation of Uranium Oxides -I. S. Kulikov . . . . . . . . . . . . . . . . . . . . . 798 93
Radiochemical Determination of the Absolute Yields of Fragments. of the Fission of
Pu241 arid Pu239 by Slow Neutrons - A. V. Sorokina, N. V. Skovorodkin,
S. S. Bugorkov, A. S. Krivokhatskii, and K. A. Petrzhak . . . . . . . . . . . . . . 804 99
Measurement of Fission Cross Sections of U235 and Pu239 in a Neutron Spectrometer by
Means of the Moderation Time -A. E. Samsonov, Yu. Ya. Stavisskii,
V. A. Tolstikov, and V. B. Chelnokov . . . . . . . . . . . . . . . . . . . . . . . . 809 103
Neutron Radiative Capture Cross Sections in Silver, Au197,.Th232, and U238
-Yu. Ya. Stavisskii, V. A. Tolstikov, V. B. Chelnokov, A. E. Samsonov,
and A. A. Bergman . . . . . . . . . . . . . . . . . . . . . . . . . . 814 107
Experimental Investigation of the Feasibility of Increasing the Capacity of Indium
-Gallium Circuits - G. I. Kiknadze, V. S. Bedbenov, R. B. Lyudvigov,
and L. G. Sharimanova . . . . . . . . . . . . . . . . . . 820 113
Dosimetric Properties of Boron Nitride - G. A. Lubyanskii, V. V. Styrov,
and V. A. Sokolov . ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 826 119
A Variable-Energy Proton Linear Accelerator -V. A. Bomko, A. P. Klyucharev,
and B. I. Rudyak . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... . . . 831 123
Dynamics of Collisions of'Charged Clusters Associated with a Shock Acceleration
Mechanism - A. G. Bonch-Osmolovskii . . . . . . . . . . . . . . . . . . . . . . . 835 127
ABSTRACTS
Expanded Experimental Possibilities of Methods Based on Study of the Neutron Noise
of a Nuclear Reactor - V. V. Bulavin . . ... . . . . . . . . . . . . . . . . . . . . 841 133
Dynamics of Neutron Kinetics Processes in a Circulating-Fuel Reactor
- V. P. Zhukov and R. I. Kreer . . . . . . . . . . . . . . . . . . . . 842 134
Equivalent Systems of Kinetics Equations for a Circulating-Fuel Reactor
- V. P. Zhukov and R. I. Kreer . . . . . . . . . . . . . . . . . . . . . . . . . . . 843 134
The Transfer of Radioactive N13, N16, and F1$ along the Loop of the. VK-50 Boiling
Reactor - A. P. Veselkin, V. D. Kizin, I. G. Kobzar', V. Ya. Kucheryaev,
A. V. Nikitin, L. N. Rozhdestvenskaya, and Yu. V. Chechetkin. . . . . . . . . . . 843 135
Minimizing the Radial Temperature Drop in Cylindrical Disperse Fuel Elements
- Yu. V. Milovanov and R. I. Abramyan . . . . . . . . . . . . . . . . . . . . . . 844 135
Mechanism Underlying Release of Gaseous Fission Fragments from Ceramic Nuclear
Fuel - B. V. Samsonov and A. K. Frei . . . . . . . . . . . . . . . . . . . . . 845 136
Fluorination Kinetics of Nb205 - E. G. Rakov, D. S. Kopchikhin, B. N. Sudarikov,
and B. V. Gromov . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , 846 137
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CONTENTS
Engl./Buss.
Operation of the Facility for Deep Burial of Liquid Radioactive Wastes
- V. F. Bagretsov, S. I.. Zakharov, and S. V. Metal'nikov . . . . . . . . . . . . .
. . 847.
137
An Electrochemical Method of Determining the Radiation Dose Rate - G. Z. Gochaliev
. and S. I. Borisova . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
848
138
Determination of the Specific Activity of y-Emitting Isotopes in Extended Sources
without Sample Collection - V. I. Polyakov and Yu. V. Chechetkin. . . . . . . .
850
139
Portable Neutron-Irradiation Apparatus -V. K. Andreev, B. G. Egiazarov,
L. A.,Korytko, and Yu. P. Sel'dyakov . . . . . . . . . . . . . . . . . . . . . . . .
851
140
Measurement of Beam Parameters for a-Particles Extracted from the JINR
Heavy-Ion 2 Meter Isochronous Cyclotron - V. S. Alfeev, E. D. Vorob'ev,
G. N. Zorin, and Yu. I. Kharitonov . . . . . . . . . . . . . . . . . . ... . . . . .
851
140
Method of Measuring (p, n)-Thresholds for the Study of Accelerator Beam Analyzing
Systems - M. I. Afanas'ev, A. L. Bortnyanskii, and A. I. Graevskii . . . . .
853
141
LETTERS TO THE EDITOR
Expanded Capacity Radiation Loop at the IRT Nuclear Reactor in Tbilisi
- G. I. Kiknadze, E. L. Andronikashvili, V. S. Bedvenov, I. A. Gassiev,
G. V. Zakomornyi, D. M. Zakharov, B. I. Litvinov, R. B. Lyudvigov,
L. 0. Mkrtichyan, I. A. Natalenko, and L. I. Feldman . . . . . . . . . . . . . .
854
143
Fabrication of Liquid-Metal Working Materials for Radiation Loops - G. I. Kiknadze,
D. M. Zakharov, R. B. Lyudvigov, and L. I. Feldman . . . . . . . . . . . . . . .
858
146
Pneumatic Irradiation Channel Reloading System for the IRT-M Reactor
- T. S. Ambardanishvili, G. V. Zakomornyi, G. D. Kiasashvili, G. I. Kiknadze,
B. I. Litvinov, L. 0. Mkrtichyan, and A. M. Uvarov . . . . . . . . . . . . . . . .
860
147
Gas Evolution in the Primary Loop of a Pressurized-Water Reactor with Gas Volume
Compensators - Yu. F. Bodnar' . . . . . . . . . . . . . . . . . . . . . . . . . .
864
150
Effect of Reactor Radiation on Corrosion Cracking of Alloy AMg6M - Kh. B. Krast
and A. V. Byalobzheskii . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
867
151
Relation between Solutions of the Nonstationary and Quasicritical Transport Equations
- B. D. Abramov. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
869
153
Moments of the Neutron Density Distribution Function -T. E. Zima, A. A. Kostritsa,
and E. I. Neimotin. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
871
154
Spontaneously Fissioning Isomers of Uranium, Plutonium, and Americium from Neutron
Reactions -Yu. P. Gangrskii, T. Nad', I. Vinnai, and I. Kovach . . . . . . . . . .
874
156
Yields of Be 7 in the Irradiation of Lithium and Boron with Protons and Deuterons and
that of Beryllium with Protons, Deuterons, and a-Particles -P. P. Dmitriev,
N. N. Krasnov, G. A. Molin, and M. V. Panarin . . . . . . . . . . . . . . . . . .
876
157
Study of the Weak a-Activities of the Volatile Fractions of Lead-Zinc Ore by the
a-X Coincidence Method - V. Kush, V. I. Chepigin, G. M. Ter-Akop'yan,
and S. D. Bogdanov . . . . . . . . . . . . . . . . . . .
879
159
Neutron Resonance Apparatus with a Central Source Arrangement - B. S. Vakhtin
and E. M. Filippov . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
882
161
Longitudinal Stability of a Beam in a Linear Induction System - V. K. Grishin . . . . . .
884
163
Absolute Measurement of Particle-Beam Intensity by a Fluctuation Method
- Yu. P. Lyakhno and V. A. Nikitin . . . . . . . . . . . . . . . . . . . . . . .
887
164
Origin of Accelerated Atoms Accompanying a Plasma Blob - K. B. Kartashev,
V. I. Pistunovich, V. V. Platonov, and E. A. Filimonova . . . . . . . . . . . . . .
889
165
NEWS
XXIX Session of the Learned Council of the Joint Institute for Nuclear Research
-V. A. Biryukov . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
891
167
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CONTENTS
EngI./Russ.
The IV All-Union Conference on Heat Exchange and Hydraulic Resistance
- E. V. Firsova and B. L. Paskart . . . . . . . . . . . . . . . . ... . . . . . . . 897 170
II All-Union Conference on Charged-Particle Accelerators -V. S. Rybalko . . . . . . 898 171
Accelerators in the National Economy and in Medicine - L. G. Zolinova . . . . . . . . . 902 173
The Franco-Soviet Colloquium on Fast-Reactor Technology -Yu. E. Bagdasarov
and 0. D. Kazachkovskii . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 906 176
The Use of Nuclear Methods for Measurement and Control of Environmental Pollution
- L. V. Artemenkova . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 909 178
Soviet Nuclear Power Specialists' Tour of the Netherlands and Belgium
- L. V. Komissarov . . . . . . . . . . . . . . ... . . . . . . . . . . . . . . . . . 912 180
The Russian press date (podpisano k pechati) of this issue was 8/4/1971.
Publication therefore did not occur prior to this date, but must be assumed
to have taken place reasonably soon thereafter.
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A. M. Kuz'min, Yu. V. Silaev, UDC 621.039.526
V. V. Orlov, and V. V. Khromov
The selection of optimum characteristics of nuclear reactors requires involved considerations of
thermophysical, safety, hydraulic, and other questions and is a complex multiparameter problem. When
many parameters or equations are required for a fast reactor of the BN-350 type [1] a solution can be ob-
tained by using the optimization procedure for fast reactors given by Khromov et at. [2].
Description of the Algorithm of the Program. The optimization complex includes neutron physics,
safety, and thermotechnical reactor calculations and solves the following problem. Find the value of the
control vector u{ul; U2; . . . uq; , , , uk}, satisfying the relation
umin'- 11G u max (1)
for which some reactor characteristic Fo((p, u) has its optimum value (henceforth we will consider the
minimum of FO) and constraints on other quantities Fv (q , u) of the form
F ((p, u) GA? (v =1, 2, ..., p) (2)
are satisfied under the condition that the reactor state variables (P(r){w(1); (P (2); ... (P(m)l(2), such as neu-
tron flux, temperature distribution, etc., satisfy known equations, and the quantities umin, Umax, and A.
are assumed known.
The constant thermal power of the reactor W, the form and properties of the nuclear fuel, coolant,
and other reactor materials, the coolant temperature at the reactor outlet Tout, the fuel cycle parameters,
the operating characteristics of the reactor including the time of fuel reprocessing, the load factor of the
reactor, etc., are also assumed given.
The following parameters can be taken as components uq of the control vector u, where the subscript
i distinguishes those parameters which can vary from one reactor region to another: the core height H,
the height of the fuel element void for the collection of gaseous fission products Hg P., the thickness of in-
dividual reactor zones AR1, the outside diameter of an unclad fuel rod d1T), the fuel element cladding thick-
ness Ai 1, the relative pitch of the triangular fuel element lattice his the coolant velocity in a channel of the
i-th zone for maximum heat release rate v?, the fuel enrichment xi, the relative fraction of absorbers for
compensating reactivity ea, and the volume fraction of the fuel assembly wall in the reactor Cam.,
The neutron physics calculations are performed in the two-group diffusion approximation for a one-
dimensional cylindrical reactor with cross sections averaged over an 18-group neutron spectrum in each
reactor zone, and the axial neutron leakage is taken into account by the parameter x2 = ire/(H + 26eff)2?
The averaged microscopic cross sections and the effective and reflector savings 6eff are calculated with
the 18-4RZ-15 two-dimensional program. [3].
The heat engineering calculation is restricted to an estimate of the maximum fuel element cladding
temperatures, taking account of hot spots, and to a determination of the average temperature rise of the
coolant in the reactor ITcool? The calculation is performed under the assumption that there is. no axial
heat flow in the fuel elements. The calculation of the reliability of the fuel element cladding is carried out
under the assumption that the fuel does not exert pressure on the cladding in the swelling process [4]. The
maximum stress ua in a fuel assembly wall for a pressure drop Ap of the coolant over the height of the
reactor is determined under the assumption that the thickness of an assembly is very much less than its
dimension under the key.
Translated from Atomnaya Energiya, Vol. 31, No. 2, pp. 83-88, August, 1971. Original article sub-
mitted July 22, 1970.
? 1972 Consultants Bureau, a division of Plenum Publishing Corporation, 227 West 17th Street, New York,
N. Y. 10011. All rights reserved. This article cannot be reproduced for any purpose whatsoever without
permission of the publisher. A copy of this article is available from the publisher for $15.00.
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1,7
1,6 14
1,3
1ti 1~h
/,3 1,35
1, 9
1, 5
1,6;
,
a
2,3 2,5
Liz 1P9
7
Fig. 1. Curves of Kr = const and gradient search for the minimum of
the coefficient of nonuniformity of heat release. a-b is the projection
of the crest of the surface Kr(ui, u2) on the ui-u2 plane.
The program includes algorithms for thermophysical and safety calculations which permit the solu-
tion of optimization problems taking account of the constraints:
Tmax = kfaf (E) V E (1 ~- S),
JB (t)
where v is the mean velocity of the neutrons at the time of moderation t, crf(E) is the fission cross section
for nuclei of the test substance, is the average over the neutron spectrum of N(E, t) in the mod-
erator at time t, S is a small correction term representing the replacement -- uf(E)fE and de-
pending on the width of the neutron energy spectrum N(E, t) and the energy dependence of the test cross
section uf(E), and kf is a normalizing factor.
The mean neutron energy E (keV) and the moderation time t (psec) are connected [2] by the relatior
183
(t+0.3)2 (2)
If we can neglect the correction term S, then the expression for the cross section in terms of the neutron
energy takes the form
of (E) = JB (t) i1
kf E
The energy dependence of the cross section can be normalized [2] from resolved resonances with
known parameters [3] or from the thermal cross section. However, the poor resolution of the spectrom-
eter does not always permit us to resolve resonances in the measured cross section with reliably measured
parameters. Furthermore, if the measured cross sections have low resonances with energies of the order
of a few tens of electron volts, then the statistical accuracy of measurements in the thermal region becomes
inadequate, partly owing to the marked fall in neutron density in the moderator [2],
JB (t) = const t-0-35e-t/T, (4)
where T is the mean lifetime of the neutrons (' 890 ?sec) in the lead prism, and partly owing to the long
times, relative to a neutron burst (2000, ?sec), to which in this case the thermal region corresponds (1/v
law).
Translated from Atomnaya Energiya, Vol. 31, No. 2, pp. 103-106, August, 1971. Original article
submitted August 31, 1970.
o 1972 Consultants Bureau, a division of Plenum Publishing Corporation, 227 West 17th Street, New York,
N. Y. 10011. All rights reserved. This article cannot be reproduced for any purpose whatsoever without
permission of the publisher. A copy of this article is available from the publisher for $15.00.
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---------- -------
TABLE 1. Thermal Cross Sections (in b)
Used for Normalization with the Aid of Mea-
surements in a Graphite Prism
ath
t
Fig. 1. Diagram of main prism of lead
moderator and graphite prism for nor-
malization of cross section curves by
means of thermal value. 1) Prism of
lead moderator; 2) position of zirconium
-tritium target; 3) channel in which fast
neutron bursts are registered; 4) main
measurement channel; 5) graphite prism;
6) measurement channel of graphite prism;
7) hole in graphite moderator.
An adequately intense thermal neutron spectrum can
be obtained in a graphite prism placed near the main prism
of the lead moderator (see Fig. 1). Since the thermal neu-
tron density is greater in the graphite prism than in the
lead moderator, for the same times relative to a neutron
burst we can reliably normalize the data from the thermal
cross section.
For normalization by the thermal cross section we
constructed a graphite prism, 120 x 60 x 60 cm in size.
By means of a boron detector we investigated the dependence of the neutron density Jfh(t) in the graphite
prism in terms of the time relative to a neutron burst. The results of JtB (t) measurements show that
In JB (t) has a linear graph which is established at time It > 1000 psec after the formation of an equilibrium
spectrum. In this time range the integral neutron density in the graphite prism is about 10 times greater
than in the measurement channel of the lead prism.
Jfh(t). Then by analogy with Eq. (1) we have
jt (1)
JB (t) =ks u E
of i. C
W
u0
0
w
Q)
U W C .
U
W
2 u S C
u
O
N b0 u
'E
C
y
X
u
tJ
I
One indium layer Xa thick at
78?3,9; 68?3,4;
68?3,4
1
0,37?0,02
1
the center of graphite assem-
65?3,3; 60?3,0;
bly. The indium layer con-
66?3,3; 64?3,2;
sists of eight indium foils
68?3,4; 75?3,7;
/8 thick each
II
Eight layers of indium Xa/8
136?6,8; 119?6,0;
105,9?5,3
1,56?0,16
0,58?0,03
1,57?0,16
thick each spaced symmetri-
110?5,5; 110?5,5;
cally with respect to the as-
99?5,0; 85?4,3;
sembly center at 20 mm in-
91?4,5; 97?4,8;
tervals
III
Eight layers of indium Xa/4
65?3,2; 48?2,4;
43,4?2,2
0,64?0,07
0,23?0,01
0,62?0,06
thick each spaced symmetri-
45?2,5; 44?2,2;
call with respect to the
Y
49?2,4; 39?2,0;
41?2,0; 40?2,0;
center at 20 mm intervals.
40?2,0; 45?2,3;
Each as/4 layer consists of
36?1,8; 40?2,0;
41?2,0; 42?2,1;
two as/8 indium foils
38?1,9; 42?2,1;
IV
One Xa/2 layer of indium at
137?7,0; 123?6,2;
125,3?6,3
1,84?0,18
0,66?0,03
1,78?0,18
the center of the graphite
115?5,8; 120?6,0;
128?6,4; 126?6,3;
assembly. The laye consists
Y?
12516,3; 128?6,4;
of eight Xa/16 indium foils
V
Eight Xa/16 indium layers
259?13,0; 248?12,5;
244,6?12,3
3,6?0,36
0,27?0,04
2,35?0,24
symmetrically spaced with
247?12,4; 244?12,2;
240?12,0; 236?11,8;
respect to the center at 20
241?12,0; 242?12,1;
mm intervals
As the average thermal neutron flux in activity generator models having a single y-carrier layer Xa and Xa
/2 thick is taken the neutron flux averaged over the measured thermal neutron distribution in the layer. Within
the limits of experimental accuracy, the average thermal neutron flux in models with Xa/4, Xa/8, and Xa/16
y -carrier layers is taken to be the neutron flux measured on the layer surface.
The experiments described below were an attempt to find a range of optimum parameters of various
dense indium-graphite lattices serving as activity generator models.
The model assemblies were cylinders consisting of alternating graphite discs and indium foils of
various thickness and spacing. The diameter and total height of all assemblies was 120 and 320 mm, re-
spectively.
The assembly construction, made up of individual graphite discs 20 mm thick and of thin indium foil
with a thickness Aa/8 = 125 mg/cm2 (0.0173 cm) and Aa/16 mg/cm2 (0.0086 cm), made it possible to vary
both the lattice spacing and the thickness of indium layers by varying the number of foils in a layer. We
have studied six different configurations of heterogeneous assemblies simulating six different activity gene-
rator models (see Table 1).
For plotting the neutron flux distributions in the assemblies and to find the specific activity (specific
radiation capacity) stored in the foils, the assemblies were irradiated in a vertical channel 180 mm in dia-
meter placed at the edge of the core of the IRT nuclear reactor in Tbilisi [8] (see Fig. 2a).
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9110
2 Q8
0,6
0,4
Q2
1.2
H1,01
46
0,4
12
a
M
8 12 16 20 14 28
b
0,2 ~~ 14 14089 H
4 8 12 16 20 24 28
C
1,2
1,0
0,8
0,6
8 12 16 20 24 29
d
I I I I I I I I ICI ~ i
LJ I 11-fTI I ITT
r_I- LtTT-TTTi-l-1-i 4T
Height, cm
8 12 16 18 20 28
f
Fig. 3. Thermal neutron flux distribution: a) in graphite block
without absorber; b) in single-layer model of activity generator
with Aa layer of y-carrier; c) in single layer generator model
with Aa/2; d) in eight-layer generator model with Aa/4; e) in
eight-layer generator model with Aa/8; f) in eight-layer model
with Aa/16.
To allow for reactor power fluctuations, monitors of 10 mg/cm2 metal gold foil were placed in all as-
semblies at a distance 60 cm from their top. All measurements were normalized for the counting rate of
this monitor. The distribution of thermal neutrons over the height of heterogeneous indium-graphite as-
semblies was measured by activating 20 mg/cm2 pure copper foils placed between graphite discs at 20 mm
intervals (see Fig. 2b).
The specific activity in the indium absorber layers was determined by measuring the radioactivity of
indium samples in the form of discs 10 mm in diameter cut from the center of irradiated Aa/8 and Aa/16
foils. The measurements were made with the aid of a pulse-height analyzer and a scintillation y-spectrorn-
eter. The measuring error was 4-5%. The thermal neutron flux distribution is shown in Fig. 3.
The distribution of thermal neutron flux in a graphite assembly without y-carrier (Fig. 3a) indicates
that the neutron production density in the assembly is constant over a length of 26 cm ('f2T for graphite)
as predicted by the Fermi age theory. The relative thermal neutron flux determined in this assembly and
normalized for the monitor count rate has been assumed as the average undisturbed flux Do in all subse-
quent measurements. In plotting the thermal neutron distributions in the investigated activity generator
models, the relative flux has been obtained by similar normalization. This makes it possible to compare
the results obtained in different models by relating it to a unit ordinate corresponding to the undisturbed
flux in the chosen relative units.
Figure 3b shows the thermal neutron distribution in a single-layer activity generator model with a
,y-layer Aa thick (model I in Table 1). The disturbance of the neutron flux by the absorber is high and the
average thermal neutron flux in the -y-carrier layer is 0.374)0.
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 Declassified and Approved For Release 2013/02/26: CIA-RDP10-02196R000300090001-2
A comparison of the curves in Fig. 3b and e shows that the average thermal neutron flux in the case
of a single y-carrier layer is 1.56 times lower than in the eight-layer model with the same total thickness
of y-carrier* (models I and II in Table 1). The specific radiation power of the eight-layer model is 1.6
times the radiation power in a single layer of the same total thickness.
The curve in Fig. 3d shows the thermal neutron flux distribution in a heterogeneous system of eight
indium layers (model III in Table 1) of a total thickness equal to 2Aa. The average specific activity of in-
dium foils as well as the relative average thermal neutron flux in this model is 35-40% lower than in a
single-layer system consisting of an indium block Aa thick (model I in Table 1). The total indium activity
is -4.25 times the total activity in model I.
If the indium foils were placed in the moderator not as a heterogeneous system, as in model III, but
as a single indium block of equivalent thickness, the total indium activity would amount to ' 0.85 and not
1.25 (relative units) (see Fig. 1).
The thermal neutron flux distribution in a single Aa/2 indium block (model IV) is shown in Fig. 3c.
The disturbed flux at the distribution minimum amounts to 0.640 and the average flux in the layer is 4)
0.66 (Do.
A comparison of the specific activity of single-layer models (I and IV in Table 1, Fig.3b and c) shows
that the specific activity of the y-carrier in a Aa/2 layer is 1.84 times than in a Aa layer.
The average thermal neutron flux, which determines the y-carrier specific activity, in a Aa/2 layer
is 1.78 times the flux in a Aa layer. Thus, the generation of radioactive nuclei in the -y-carrier layers of
such thickness takes place as a result of absorption of mainly thermal neutrons.
The specific activity of model V, stored in eight indium layers Aa/16 thick each and spaced in the
graphite at intervals of the order 0.7Atr, cannot be ascribed entirely to In116 generation by thermal neu-
trons. This is indicated by the data in Table 1 in which the ratio of average thermal neutron fluxes is 2.35
in the chosen system of relative units, while the ratio of specific activities is - 3.6. Such a discrepancy
can be explained only if the contribution of resonance neutrons into the generation of In116 nuclei is allowed
for. In fact, the resonance integral, calculated in accordance with [9], is - 600 b for thin indium layers
such as Aa/16 layers and barely 100 b for Aa layers. This fact can be of great importance in reactors in
which the fraction of fast and resonance neutrons in the leakage spectrum is considerable. In such reactors
it is possible to design activity generators in which the thickness of moderator layers between -y-carrier
layers with high self-shielding factors for resonance electrons is such that groups of fast neutrons can
achieve an age equivalent to indium resonance energies. In such a case it is possible to obtain considerable
gain in specific and total capacity of the radiation circuit.
The practically continuous distribution of thermal neutron sources in the investigated graphite struc-
ture (see Fig. 3a) makes it possible to compare the theoretical and experimental values of thermal neutron
fluxes in thin y-carrier layers. In particular, the distribution curve in Fig. lc indicates that the average
thermal neutron flux in a Aa/16 layer is -0.96340. Taking into account the influence function of such layers
spaced at 20 mm intervals in a graphite moderator we arrive at a figure of - 0.940 for the thermal neutron
flux in the assembly. The experimental value of thermal neutron flux is - 0.884)0.
LITERATURE CITED
1. A. P. Aleksandrov, At. Energ., 25, 356 (1968).
2. Yu. S. Ryabukhin et al., Industrial Use of Large Radiation Sources (Abstract of Reports of the USSR
Delegation), Vol. 2, IAEA, Vienna (1963), p. 175.
3. E. L. Andronikashvili et al., At. Energ., 13, 342 (1962).
4. G. I. Kiknadze et al., At. Energ., 19, 176 (1965).
5. G. I. Kiknadze et al., Abstracts of Reports of the Conference of Young Scientists on Radiation Chem-
istry and Radiobiology [in Russian], Obninsk (1969), p. 140.
6. Yu. S. Ryabukhin and A. Kh. Breger, At. Energ., 5, 533 (1958).
7. R. Murray, Nuclear Reactor Physics [Russian translation], Gosatomizdat, Moscow (1959).
*Neutron fluxes measured on the surface of indium foils are taken as the average thermal-neutron flux in
Aa/16, Aa/8, and Aa/4 layers. The error in the average flux in such layers is quite small and does not
exceed 5% because of low flux depression.
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 Declassified and Approved For Release 2013/02/26: CIA-RDP10-02196R000300090001-2
8. V. V. Goncharov et al., Nuclear Reactors and Nuclear Power Engineering (Reports of Soviet Scientists
at the Second Geneva Conference) [in Russian], Gosatomizdat, Moscow (1959), p. 243
9. V. V. Orlov, T. V. Golashvili, and A. I. Baskin, in: Neutron Physics [in Russian], Gosatomizdat, Mos-
cow (1961).
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 Declassified and Approved For Release 2013/02/26: CIA-RDP10-02196R000300090001-2
G. A. Lubyanskii, V. V. Styrov,
and V. A. Sokolov
The development of new and effective thermoluminescence dosimeters is closely associated with the
study of radiation processes in solids [1]. An investigation of the dosimetric properties of compounds of
the type of A(III) and B(V), which is now being paid considerable attention, is of interest. In this work we
investigated the thermoluminescence and dosimetric properties of one of the representatives of this type
of compound, boron nitride (BN).
Samples. Boron nitride, synthesized according to the method described in [2],* possesses a hexago-
nal lattice with parameters corresponding to the literature data [3]. Spectral analysis for impurities indi-
cated the presence (in wt. %) of: Al < 10-4, Mg < 10-4, Ca < 10, Ag < 10-4, Si < 10-3.
Samples activated by manganese, europium, and samarium were also produced. Boric acid served
as the flux. The charge was calcined at 750?C for 1 h and at 1400?C for 30 min in a stream of ammonia.
The samples were used in the form of powders; in individual cases tablets were prepared by pressing at
a pressure of 450 kg/cm2.
Experimental Method. The thermoluminescence of boron nitride was investigated after irradiation
of the samples at room temperature with -y-quanta on an RKh-y-30 apparatus and with thermal neutrons in
the vertical channel of the nuclear reactor of the Institute of Physics of the Academy of Sciences of the
Latvian SSR.
The apparatus for recording thermoluminescence curves, consisting of heating and recording units,
permitted the use of higher rates of heating (3.6-9.1 deg/sec) and ensured constancy of the selected system.
The sample was applied directly on a constantan plate, heated by current, with a depression for the lumino-
phore, in a thin dense layer in the form of a weighed sample (30 mg) or in the form of a tablet (60 mg). The
thermoluminescence curves were recorded with an FEU-39 with an SZS-14 filter, cutting out thermal radia-
tion. The time of recording of one curve was 40 sec. The emission spectra were measured on apparatuses
of P. Stuchka Latvian State University and A. F. Ioffe Physicotechnical Institute of the Academy of Sciences
of the USSR.'
Curves of Thermoluminescence and Spectra. The absorbed dose was varied from 10 to 3.108 rad.
The experiments indicated that without preliminary irradiation, both of inactivated and of activated BN, no
thermoluminescence is detected. However, at a dose of 10 rad, a thermoluminescence peak already appears
at a temperature of 125?C (rate of recording 9.1 deg/sec); Beginning with a dose of 5.102 rad, a second
peak appears at the temperature 425?C. When the dose is further increased, the second (high-temperature)
peak becomes dominant, and the light sum luminescing in it constitutes 95% of the total stored light sum.
*We should like to express our great gratitude to G. V. Samsonov and M. D. Lyutoi for providing boron
nitride.
tWe should like to thank V. Ya. Grabovskii and L. S. Druskina for providing these apparatuses and for
their practical aid in conducting the experiment.
Translated from Atomnaya Energiya, Vol. 31, No. 2, pp. 119-122, August, 1971. Original article
submitted July 30, 1970; revision submitted October 5, 1970.
? 1972 Consultants Bureau, a division of Plenum Publishing Corporation, 227 West 17th Street, New York,
N. Y. 10011. All rights reserved. This article cannot be reproduced for any purpose whatsoever without
permission of the publisher. A copy of this article is available from the publisher for $15.00.
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Fig. 1. Curves of the intensity J of thermoluminescence of hex-
agonal BN at various temperatures and absorbed doses: 1) 102
rad; 2) 103 rad; 3) 105 rad.
Fig. 2. Value of the stored light sum of hexagonal BN as a func-
tion of the dose D of y-radiation for the first (1) and second (2)
thermopeaks. The relative units of the light sum are different
for the two curves.
The dynamics of the curves of thermoluminescence of BN is shown as a function of the absorbed dose in
Fig. 1. For various samples of inactivated BN (samples of a different orikin, in addition to those indicated
in the experimental section, were also investigated), the first peak lies in the interval 110-150?C, the second
in the interval 425-490?C. From Fig. 2 it is evident that the stored light sum in the first peak increases
nonlinearly with the dose, and a deviation from linearity already occurs at relatively low doses (103 rad).
At the same time, the stored light sum in the second peak increases as a linear function of the dose all the
way up to 105 rad. In the same dose interval, the thermoluminescence intensity is approximately propor-
tional to the irradiation dose.
At a dose of 5.106 rad, the stored light sum reaches a maximum value and gradually drops when the
dose is further increased. Moreover, the maximum of this peak is shifted in the direction of high tempera-
tures with the dose. Thus, at a dose of 2.4 ?108 rad, it corresponds to a temperature of 510?C.
The emission spectrum of the thermoluminescence of inactivated BN lies in the blue-green region.
In both thermopeaks, radiation with wavelengths A = 380-520 nm is observed. (The position of the maxi-
mum of the spectrum for various samples varies in the region 388-429 nm.)
Boron nitride, activated by manganese (1 mole %), exhibits three thermopeaks. The first thermopeak
is situated at the same temperature (110-150?C) as the temperature of inactivated BN. The second is situ-
ated at the temperature 250-270?C, and the third at 375-440?C.
The activator thermopeak (250-270?C) begins to appear at a dose of 5.102 rad; the stored light sum
corresponding to it is approximately proportional to the dose of irradiation up to 106 rad. At very high
doses, 2.4 -108 rad, this thermopeak disappears. The activator thermopeak has somewhat lower intensity
than the high-temperature peak of inactivated BN at equal doses.
The emission spectrum of the thermoluminescence in the activator thermopeak takes the form of a
band with maximum at A = 497 nm to 518 nm.
In the case of activation of boron nitride by europium (0.1 mole %) and samarium (0.5 mole %o), no
activator peaks were detected, but a high-temperature thermopeak appears at a somewhat lower tempera-
ture (by 60-70?C) than for inactivated BN.
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From the results cited it follows that inactivated BN con-
tains at least two varieties of traps, the depth of which were
100 ~- ---F--~ determined from the thermoluminescence curves according to
the Urbach formula and the Lushchik equation [4] and are 0.7-
0.8 eV for the first thermopeak, and 1.4-1.6 eV for the second
fhnrmnnnak
We should mention the high reproducibility of the results
from sample to sample in a lot and from lot to lot for the same
501 1 i I method of synthesis of BN from raw material of the same qual-
10 20 30 r,h ity: ?5% for the position of the maximum of the thermopeaks,
Fig. 3. Loss of the stored light sum S ?1070 for their intensities.
of the high-temperature peak of hex- The method of synthesis and purity of the starting mate-
agonal BN as a function of the duration
of storage at various temperatures, rials appreciably influences the curve of thermoluminescence,
especially the light sum. Thus, BN synthesized from cp raw
material stored an order of magnitude smaller light sum than
BN synthesized from very pure xaw materials. We observed the same tendency for a shift in the thermo-
peaks of BN, synthesized according to the method of [2], after its supplementary calcination in air. Thus,
after 20 h of calcination at 1300?C, the maximum of the first peak was shifted from 120 to 140?C, and that
of the second from 480 to 400?C; there was a redistribution of the intensities of the peaks in favor of the
low temperature peak.
With respect to sensitivity to the.dose, BN is only two to three times inferior to LiF. Let us note
that the activator peak of BN-Mn, possessing linear dose characteristics and favorably situated in the
temperature scale, is also convenient for purposes of dosimetry.
Conservation of Stored Light Sum. We verified the ability of boron nitride to conserve the light sum
at different temperatures. The results are cited in Fig. 3. It is evident that in a period of 10 h at room
temperature, BN loses approximately 3% of the light sum (curve 1), while at 50 and 68?C it loses 8 and 17%,
respectively (curves 2 and 3). Dosimetric information with respect to the second thermopeak is preserved
(by approximately 30%) for 90 days (dose of irradiation 105 rad).
At a low dose rate of the irradiation of boron nitride for a long time, there are practically no losses
of light sum. Thus, in an experiment during which BN was irradiated for ten days at a dose rate of 0.6
rad/h, the stored light sum proved close to the calculated. This property permits the use of BN for dosi-
metric purposes, including low dose rates.
Dependence of the Sensitivity on the y-Radiation Energy. The storage of the light sum in the irradia-
tion of inactivated BN with -y-quanta of Co with energies equal to 1.17 and 1.33 MeV, and Tu170 with Ey
= 0.084 MeV, was compared. It was established that in these cases neither the shape of the thermolumines-
cence curve nor the dose characteristics depend on E 'Y.
Boron nitride, as a dosimetric material, possesses tissue equivalence: the effective atomic numbers
of soft biological tissue (7.42) and BN (6.8) almost coincide.
Possibility of Repeated Use. In the dosimetric respect it is important that a sample does not experience
irreversible changes after complete deexcitation of the stored light sum, and return to the initial state. The
reproducibility of the thermoluminescence curves was verified in the case of repeated (five times) irradia-
tion of the same sample with the same dose. After each irradiation, as a result of one temperature cycle,
the thermoluminescent emission entirely disappeared. When irradiation was repeated, the intensity of the
thermopeaks and their positions were almost unchanged. This shows the possibility of the repeated use of
BN as a dosimeter without special intermediate treatment.
Resistance to Interference. In thermoluminescence dosimetry, the source of interference, especially
at low doses, may be chemiluminescence and triboluminescence, which give a false signal. We did not de-
tect any emission of nonirradiated BN during heating in air. Triboluminescence also was not detected. To
verify this, the same sample was irradiated with the same dose twice, and the second time the powder was
mixed before heating. No differences in the thermoluminescence curves were detected.
We also verified whether the light sum is stored when BN is irradiated with sunlight. It was found
that in the case of 5 h exposure, the first thermopeak is detected for inactivated BN (at the temperature
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TABLE 1. Dosimetric Properties of BN in Neutron Irradiation
Sample
1010
1011
1012
1013
1014 - 1015
/l I tl I
J2I
t2
J1 I tl I
J2
12
Jl
tl J2 I 12
J1 tl J2 I t2
J11 tl I J2 I 121J11 III
J2I
12
I
BN,
4
120
-
-
6
150
-
-
10
160
-
-
12
163
-
-
16
165
-
-
25
165
-
-
cubic
BN
,
2
150
40
500
6I
150
70
510
11
150
155
520
22
151
170
522
8
150
180
522
3
155
7
52C
ll
xa-
gonal
BN -M ,n.
1m
e
102
851
126
260
133
107
132
282
25
1 150
165
365
1 22
1 150
42
315
9
165
47
387
1 2
164
14
39(
o
01.
B N Eu,
-
10
118
63
401
20
1
1120
74
405
130
137
1
129
412
133
152
183
362
80
155
43
387
13
147
180
37(
1
I
1
~
1
1
1
1
1
1
mole%
.
I
I
Note: J1 and J2 are the intensities of the first and second thermopeaks, rel. units; t1 and t2 are the temperatures
corresponding to them,?C; rate of heating 9.1 deg/sec.
150?C), equivalent to an absorbed dose of -y-radiation of 3 rad. In the case of 20 h exposure, the intensity
of this peak reached a level equivalent to 7 rad.
BN activated by manganese proved even most sensitive to daylight. In the case of a 5 h exposure, it
exhibited a thermopeak at the temperature 110?C, equivalent to a dose of 4.5 rad; in the case of 20 h expo-
sure, it exhibited two thermopeaks, and the high-temperature peak at the temperature 260?C corresponded
to a dose of 12 rad.
The experimental results are presented in Table 1. It is evident from it that cubic BN exhibits
thermoluminescence with one peak, the intensity of which increases nonlinearly with the thermal neutron
flux. In this case, the temperature corresponding to the maximum of the thermoluminescence curve is
shifted from 120 to 165?C.
Hexagonal BN, in the case of irradiation with thermal neutrons, just as in the case of irradiation
with y-quanta, possesses two peaks of thermoluminescence, lying at the temperatures 150-155?C and 500-
525?C, respectively. Probably the centers of capture responsible for these peaks are of the same nature
as in irradiation with photons., although in the case of neutron irradiation the peaks are somewhat shifted in
the high-temperature direction. These centers of capture existed before irradiation and did not appear as
a result of it. In addition, irradiation induces color centers in BN, which in the presence of large fluxes
(> 1014 neutrons/cm2) becomes evident, since the color of the sample changes from white to brownish. How-
ever, no new peak of thermoluminescence appears in this case, and the intensity of the old ones, on the con-
trary, decreases. At a flux of 1017 neutrons/cm2, thermoluminescence disappears entirely. The colora-
tion of BN may be explained by the formation of F-centers [5, 6].
Hexagonal BN, activated with manganese and europium, also possesses two thermopeaks, the positions
of which, as can be seen from Table 1, differ from the thermopeaks characteristic of inactivated BN. For
example, the high-temperature peak corresponds to considerably lower temperatures at the same doses.
The dependence of the temperature of the maximum on the neutron flux in this case is sharper.
Let us note also that the dosimetric information is well preserved during neutron irradiation. Thus,
16 days after irradiation, the light sum is still 80% of the original value. A comparison of the efficiency of
BN as a dosimeter in irradiation with y-quanta and neutrons shows that the intensity of the thermopeaks at
a -y-irradiation dose of 1 rad is approximately six times as great as at a dose of 1 rem of irradiation with
thermal neutrons. Thus, BN can serve as an effective dosimeter of y-irradiation.
In conclusion, let us express our sincere gratitude to Z. A. Grant and K. K Shvarts for their useful
advice and their aid in conducting this work.
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 Declassified and Approved For Release 2013/02/26: CIA-RDP10-02196R000300090001-2
LITERATURE CITED
1. K. K. Shvarts et al., Thermoluminescence Dosimetry [in Russian], Znanie, Riga (1968).
2. G. A. Meerson et al., Ogneupory, No. 2, 72 (1955).
3. G. V. Samsonov et al., Boron, Its Compounds and Alloys [in Russian], Izd-vo AN UkrSSR, Kiev (1960).
4. Ch. B. Lushchik, Dokl. Akad. Nauk SSSR, 101, 641 (1955).
5. M. B. Khusidman and B. N. Sharupin, Radiokhimiya, 9, 279 (1967).
6. M. B. Khusidman and V. S. Neshpor, Teor. i Eksperim. Khimiya, 2, 270 (1967).
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 Declassified and Approved For Release 2013/02/26: CIA-RDP10-02196R000300090001-2
V. A. Bomko, A. P. Klyucharev, UDC 621.384.643
and B. I. Rudyak
The most serious disadvantage of most linear accelerators for ions is that the particle energy cannot
be varied. An accelerating structure of Alvarez type [1] takes the form of a cavity loaded by drift tubes and
excited in the E010 mode; in principle, it does not allow the beam energy to be varied. Particles with inter-
mediate energies can be obtained only by dividing the accelerator into sections, which involves some diffi-
culties and which does not provide smooth energy variation.
There is a long history of attempts to obtain intermediate ion energies from linear accelerators. The
Berkeley heavy-ion accelerator uses the fact that a detuned linear-accelerator system can produce a beam
containing particles with intermediate energies as well as ones with the maximum energy. Some compo-
nents of intermediate energies can be isolated by deliberate deviation from normal operation (e.g., by
changing the accelerating field distribution and adjusting other elements). Such working conditions are un-.
stable, and the intensity is much less than that for a beam with the full energy. Various explanations have
been offered in terms of features of the particular accelerators: large apertures of the drift tubes and grid
focussing in the section before the extractor.
A new method of linear ion acceleration has been developed at Kharkov Physicotechnical Institute,
which provides smooth energy adjustment (in principle, to any value less than the maximum) without loss
of intensity or of monochromaticity.
Principle of Smooth Energy Adjustment
The device uses parts with uniformly distributed accelerating fields of varying length, so the cavity
cannot be excited in the E010 mode. If the cavity is so excited, such parts cannot be produced. The field
distribution can only be distorted (perhaps chaotically) by the perturbing tuned devices normally used to
produce acceptable uniformity in the field.
We concluded that excitation in the E011 mode (or higher E011 modes) can provide energy adjustment
when the frequencies of the individual sections of the accelerator are perturbed by tuned devices. We drew
this conclusion about this essentially parasitic wave type from the changes in the field distribution in the
cavity.
It is usual to employ a cosine distribution of the high-frequency electric field along the axis (curve 1
of Fig. 1) in order to excite the E011 mode in the cavity of a linear accelerator having drift tubes with all
the sections tuned to the same resonant frequency. Perturbations at one end of the cavity allow one to
shift the nodes in the electric field along the axis, and appropriate turning of the sections allows one to pro-
duce parts with uniform fields and sharp cutoff (curve 2 of Fig. 1).
Stepwise energy adjustment is provided on changing the length of a uniform part having ideally steep
cutoff, the steps being equal to the energy gain per stage in the accelerating structure. The practical cut-
off has a finite slope, which can be varied to provide smooth energy adjustment.
The beam must pass without loss through the part of the accelerating structure where there is no ac-
celeration, which requires a prearranged distribution of the magnetic fields in quadrupole lenses in the
drift tubes.
Translated from Atomnaya Energiya, Vol. 31, No. 2, pp. 123-126, August, 1971. Original article
submitted June 8, 1970; revision submitted January 4, 1971.
? 1972 Consultants Bureau, a division of Plenum Publishing Corporation, 227 West 17th Street, New York,
N. Y. 10011. All rights reserved. This article cannot be reproduced for any purpose whatsoever without
permission of the publisher. A copy of this article is available from the publisher for $15.00.
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 Declassified and Approved For Release 2013/02/26: CIA-RDP10-02196R000300090001-2
Fig. 1. Field distributions along the
1 accelerator corresponding to E011
2 excitation: 1) all sections tuned to the
same frequency; 2) frequencies of
sections adjusted to produce a part
with a uniform distribution in the ac-
2 celerating field.
Field-Shaping Devices
Fig. 2. Disposition of the tuned stub in
the cavity.
The cutoff must be fairly steep in the parts of uniform
distribution but vary in length in order to provide monoener-
getic beams accelerated to intermediate energies.
There is another reason why the beam becomes less
monoenergetic. If the uniform field distribution is produced
by deforming the left branch of the E011 wave (Fig. 1), there
remains a certain distribution of the electric field from the
right branch. This field is especially strong when short parts
with uniform fields are produced in order to accelerate par-
ticles to energies less than half the maximum energy. These
fields have no great effect on the mean energy, which is deter-,
mined by the left branch; in the right branch, the particles
deviate from synchronism, and their energies on average do
not alter. However, the residual right-branch E011 field in-
creases the energy spread, and it also causes useless dissi-
pation of high-frequency power in the cavity.
One needs very simple adjustment of the accelerator
from one output energy to another, and this should provide for
automatic control of the adjustment to optimum beam param- .j
eters at any intermediate energy as well for programmed
variation. I
These requirements impose severe demands on the tunec
devices. Some currently used tuned devices (cylinders, I
spheres, plates, etc.) attached to the cavity wall are ineffectivE
when the E011 mode is involved, since the fields are then much]
more stable, and hence they are less sensitive to perturbations
introduced into the cavity. Also, they do not allow one to eliminate completely the right-branch field of the
Eau mode.
All these requirements are met by a stub of variable length attached to the end wall of the cavity; thi,
is also simple and convenient. This new resonant system has effects somewhat similar to those of a half
drift tube attached to the end, whose length is also adjustable.
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 Declassified and Approved For Release 2013/02/26: CIA-RDP10-02196R000300090001-2
Fig. 4. Distributions of the
fields produced along a cav-
ity excited in the E011 mode.
I, arb.
units
1,0
5 - i i ++++ d
5 6 7 8 9
n W, MeV
5 6 7 8 9 Fig. 6. Elastic-scattering ex-
W,MeV
citation function for protons
Fig. 5. Energy spectra of accelerated produced with smooth energy
protons. adjustment.
A tuned stub was found to be most effective if placed near the side wall of the cavity (Fig. 2). As the
stub is inserted into the cavity, the E011 field node shifts along the cavity, and the right-branch field is ulti-
mately eliminated completely. This device is not only effective but also simple in design (Fig. 3); it is
convenient to use and provides for automatic smooth energy adjustment.
We examined the energy adjustment on a proton linear accelerator having a maximum energy of 9
MeV and an energy spread of 90 keV [3]. The working wavelength was A0 = 2.1 m, mean electric field along
the axis 18.4 kV/cm, cavity length 6 m, injection energy 500 keV, 34 drift tubes, and grid focussing. This
accelerator was commissioned in 1966 and since then had worked in the E010 mode at 0.2 ?A and a utiliza-
tion factor of 0.1%.
Early in 1969 it was converted to smooth energy adjustment. First we did development tests on
methods of producing parts with uniform accelerating fields and variable lengths in the Eo11 mode. We
examined the distribution of the loss of high-frequency energy in this method of field shaping, the effects
on the cavity Q from various methods of field shaping, and the stability and perturbation sensitivity of the
field. We also examined the working characteristics with smooth energy adjustment.
The resonant stub was attached to the end wall of the cavity 10 cm from the side wall. This distance
provided adequate performance while eliminating breakdown to the wall.
The resonant stub was used with tuned devices of ordinary type to shape the fields for producing
smooth energy adjustment. The stub shifted the E011 node through the necessary distance, while lateral
tuned devices provided field uniformity and steep cutoff. Figure 4 shows seven accelerating-field distri-
butions along the axis produced in this way.
Appropriate adjustment provided a uniform distribution for the E011 left-branch field throughout the
length of the accelerator. As we did not know the entire previous history of this distribution, we could not
distinguish it from the uniform E010 distribution usually obtained. The two uniform distributions all along
the cavity produced identical losses of the high-frequency power.
Figure 5 shows examples of the proton energy spectra for energies corresponding to the field distri-
butions of Fig. 4. The energy spread was 50 keV when the protons were accelerated to the maximum energy
by an E011 field uniform all along the accelerator, which is substantially less than the spread for ordinary
E010 acceleration. Figure 5 shows that the beam current and energy increase together, which is due to the
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 Declassified and Approved For Release 2013/02/26: CIA-RDP10-02196R000300090001-2
use of grid focussing, which was absent on the part where there was no accelerating field and where there
was some loss of the particles.
Figure 6 illustrates the scope for use of this method by reference to the excitation function for elastic
scattering of protons by Cr53 as measured at 60?. The points correspond to the energies used in the experi-
ment.
Since September 1969 the accelerator has been used in experiments involving smooth energy variation
from 3 to 9 MeV. The insertion depth of the tuned devices was varied in accordance with a set program
without breaking the vacuum and with the high-frequency power on. The power supplies were placed to in
ject the power near the entry section in order to provide the required levels for the short parts with uni-
form field distribution.
Prolonged use in the E011 mode has shown that this provides not only smooth energy adjustment but
also high stability in the beam characteristics. Even when maximum-energy operation is needed, E011
working is preferable to the ordinary E010 mode.
This method has the advantage that it can be operated on existing linear accelerators without major
design changes and hence without additional cost.
1. L. Alvarez et al., Rev. Sci. Inst., 26, 111 (1955).
2. A. Ghiorso et al., Proceedings of the Linear Accelerator Conference, Los Alamos, 1966.
3. B. I. Rudyak et al., Coll.: Nuclear Physics [in Russian], Kharkov, Izd. FTI AN UkrSSR (1967), p. 25
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 Declassified and Approved For Release 2013/02/26: CIA-RDP10-02196R000300090001-2
A. G. Bonch-Osmolovskii UDC 621.384.01
In 1956 Veksler [1] proposed a shock mechanism for accelerating clusters of charged particles, a
variant of the general coherent method which he considered there. At the 1959 Geneva Conference on Ac-
celerators, Veksler and Tsytovich [2] presented a brief analysis of shock acceleration, and included the
case of collision of charged clusters.
Recently interest has revived considerably in the collective acceleration method [3], on the one hand,
and in the question of obtaining large electron currents [4], on the other. There is reason to think that suc-
cess in this area could bring the accomplishment of shock acceleration considerably closer, since the real
problem has become the creation of a heavy cluster with the required parameters.
The present paper deals mainly with certain aspects of collision dynamics of charged clusters and
the establishment of the main characteristics of shock acceleration, including the basic parameters of the
charged clusters themselves. The generation and prior acceleration of these clusters are not considered
here.
Our notation will be: all quantities relating to a heavy cluster will have the numeral 1, and to a light
cluster, the numeral 2. For example, M1, vi, yi are the mass, velocity, and energy (in units of Me) before
a collision of a heavy cluster; and N2 , v2, Y12 are the number of particles (electrons), the "running" elec-
tron, and the transverse energy of the electrons in a light cluster, respectively. The running electron is
usually defined by: v = ron (ro = e2/mc2, n is the linear density of particle number, and is N/27rR for a thin
ring of radius R); Quantities before collision have subscript 0, and after collision, no subscript.
We assume that conditions for elastic collision of two clusters are met, and that the collision is head-
on (the target parameter is much less than the cluster size). The first assumption will be discussed below.
In the general case we assume that the heavy cluster has velocity vo and the light cluster vZ before
collision, the directions of the velocity vectors being the same, and vto > vZ. Then it is easy to show, using
relativistic velocity transformation formulas, that the energy of the light cluster after the collision (the
precise meaning of the word "after" is elaborated below) will be
1-2 L V2
C2
E2 = M2c272 z (1)
1-V/c2
where V is the velocity of the system center of gravity (the c-system):
V __ Mt'Pi -I-M2Y20VS
MjVi+M2Vz
MIT! >> M21's (3)
is satisfied, then V v1; and the c-system practically coincides with that in which the heavy cluster is at
rest. Then
Translated from Atomnaya Energiya, Vol. 31, No. 2, pp. 127-132, August, 1971. Original article
submitted July 10, 1970.
? 1972 Consultants Bureau, a division of Plenum Publishing Corporation, 227 West 17th Street, New York,
N. Y. 10011. All rights reserved. This article cannot be reproduced for any purpose whatsoever without
permission of the publisher. A copy of this article is available from the publisher for $15.00.
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 Declassified and Approved For Release 2013/02/26: CIA-RDP10-02196R000300090001-2
M2
V _, nr,r -2..0 YrelM1
1+2 MMI Yrel
-EM
1'rel = Yi70
2 - z 2 J ? (5)
The quantity yrel is the energy of the light cluster at infinity in the c-system with condition (3) satis-
fied. We derive a simpler formula for yrel in the important practical case when
Yi2 >> 1; Yee >> I; ',O2 >> Y82:
(6)
1 Yi
Vrel ~ 2 Ya
(7)
Now we can see that if a stricter condition than (3) is satisfied, viz.:
Mi >> 2M2Yrel
(8)
(here the heavy cluster has practically no energy change in the collision process), then:
E2 2M2c2Y?Yre1?
(9)
In the particular case when v2 = 0, yrel
= 'Yi and
E2 2Mzc2Ys1 2
(10)
The last formula was given by Veksler [1], and defines the well-known y2 effect in elastic collision of
relativistic particles.
Under the above conditions, an increase in energy of a light cluster by a factor of y2 in the laboratory
system during a collision indicates elastic reflection in the c-system of a light cluster from the force cen-
ter (the heavy cluster), with energy y at infinity.
The above collision kinematics assumes that charged clusters move and interact according to the law
applicable to the behavior of elastic spheres of mass Ml and M2. We should explain that the conditions for
this model are close to the actual picture of charged cluster collision.
Earlier papers on shock acceleration mainly considered neutral clusters with an intense azimuthal
current, and interacting by virtue of magnetic repulsion forces (for oppositely directed currents in the clus-
ters). But these currents cannot be maintained by a spatially uniform magnetic. field, and, additionally, the
magnetic interaction is relatively weak in this case, which is awkward from many points of view. There-
fore, the present paper gives attention mainly to interaction of charged clusters.
We consider a light cluster in the form of a thin electron ring with large radius R and small radius
a, and a small impurity of ions, the electrons rotating with energy -Y12 (before the collision) in the azi-
muthal direction. As regards the heavy cluster, we assume only that its characteristic geometrical dimen-
sion does not exceed It, and that the rotational energy of the electrons is y11. We also assume that it con-
sists entirely of electrons and that its mass. satisfies Eq. (8). This last point means that the heavy cluster
is at rest in the c-system, and its subsequent motion is not considered.
The large dimension of both clusters is maintained by a uniform magnetic field, so that the currents
in the clusters have the same direction.
The chief assumption made in considering shock acceleration is that the distance of closest approach
during the collision (in the c-system) is considerably larger than the characteristic dimension R of the
clusters. Then the interaction energy of the two clusters can be written approximately in the form
W . e2N1N2 e2N1N2 (R)2z 2z \z / (11)
In writing Eq. (11) it is assumed that the currents in the clusters are relativistic during the entire time of
collision. The number of ions in the light cluster is small, so that N2 NZ . We neglect the second term
in Eq. (11), considering that R > 1; then the term in Eq. (22) containing In can be
neglected, and
,_ 2zo VVre112_1
tc c (1-x)(1+Tre1';
for X = 0.1 Tc - 20 zo/c.
We determine one further characteristic of shock acceleration: the "interaction length" is the dis-
tance which the heavy cluster traverses during the collision time 'r.-Y1
(in the laboratory system):
L = Cie-Y0 ^ 2e2NjN270 1/4elx2_1 (24)
z M2c2(Yrel-1) 1-x
The quantity L is the path length over which the light cluster is accelerated by the electric field of the
heavy cluster.
We now estimate the fraction of energy which the charge of the light cluster loses in radiation. It is
known [5] that the energy loss in radiation by unit charge is
AE = 2e4 J ez (z) dt = 4e4N21 r dt (25)
3m2V12cs 3m2712cs J Z4 (t)'
0
The main part of the radiation spectrum is concentrated in the frequency region WTC - 1; the corresponding
wavelengths are considerably greater than the geometrical size R of a cluster, i.e., A 207rzo >> R. Over-
estimating the result somewhat, we consider that the radiation of the light cluster is fully coherent, and then
AE 4e"N22N12 dt
3m2Yl2es J 0 z4 (t)
It can be shown that the main part of the loss falls in the relativistic velocity region far from the stop point;
the losses near the latter are less by a factor of yrel than that in the relativistic region. Then we can sub-
stitute z = zo + ct in the interval, and calculations give
AE 4 N2 8R 7)+ m N?2
E g'!y1(Yrel-1)2 [1+Y z (Ina -4meYi ?NZJ (27)
12 1 2
Under ordinary conditions
reduces to the condition
VZ >> (?rel -1)2
V2/y12 bgr ;zz: 2 j - (4)
where 4, es are the relative photoefficiency of registration by a scintillation -y-spectrometer with colli-
mator of the radiation of volume and surface cylindrical sources, respectively; c , is the photoefficiency
of registration of a crystal for a narrowly collimated beam of -y-quanta; b is the distance from the source
to the crystals; 2a and L are the diameter and length of the collimator, respectively; 2R and t are the di-
ameter of the cylindrical source and thickness of the wall, respectively; k is a coefficient considering the
nonuniformity of the flux from the cylindrical source through the collimator, k1 '~ 1, k2 Fs 0.85 (experimen-
tal values); ?s, At, ?K are the coefficients of absorption of y-quanta in the source, shield, and walls of the,
collimator, respectively.
For an experimental verification of the possibility of using expressions (1)-(4), we used volume cy-
lindrical sources filled with an aqueous solution of a radioactive isotope, and surface cylindrical sources
with self-absorption in water and without it, crystals of NaI(Tl) with dimensions 70 x 70 and 40 x 40 mm,
and collimators 150 and 250 mm long, 5, 10, 20, and 50 mm in diameter. For the entire range of y-quan-
tum energies considered (0.1-1.5 MeV), distances (25-170 cm), dimensions of sources (diameter 16-255
mm), and collimators, the deviation of the experimental points from the calculated values of the relative
photoefficiency of recording lies within the limits of the experimental errors and comprises 10-15%; only
at b bgr does the error increase to 40%.
The value of the minimum measurable specific activity of the isotopes is 10-7 Ci/liter for volume
cylindrical sources and 10-9 Ci/cm2 for surface cylindrical sources. The method of measurement was
used to determine the activity of a coolant and to analyze deviations in the circuit of an atomic electric
power plant.
tThe relative photoefficiency of recording of the radiation of a detector with a collimator is the rate of
count of pulses at the peak of total absorption from the source of a unit volume (surface, linear) specific
activity.
Translated from Atomnaya Energiya, Vol. 31, No. 2, p. 139, August, 1971. Original article sub-
mitted October 12, 1970.
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 Declassified and Approved For Release 2013/02/26: CIA-RDP10-02196R000300090001-2
PORTABLE NEUTRON-IRRADIATION APPARATUSt
V. K. Andreev, B. G. Egiazarov, UDC 542.1:541.28
L. A. Korytko, and Yu. P. Sel'dyakov
A portable pulsed neutron-irradiation apparatus is described which can carry out the following tasks
in the case of natural rock formations: 1) spectrometry of the natural y-radiation, 2) spectrometry of the
y-radiation from inelastic scattering of fast neutrons (ISFN), 3) spectrometry of the y-radiation from the
radiative capture of neutrons (GRCN), 4) measurement of the temporal distribution of captured y-radiation,
and 5) spectrometry of the induced-activity y-radiation (IAG) in two situations - during and after irradia-
tion of the medium by means of a pulsed neutron source (for relatively short-lived isotopes). Figure 1 il-
lustrates the operation of the apparatus.
The apparatus consists of a remote unit containing a pulsed neutron source and a scintillation detec-
tor, a control and synchronization unit, and a recording device (a pulse analyzer).
TISFN
TGRCN
+LAG
W/A M/
The pulse neutron source provides a neutron output
of up to 107 neutrons/sec at a pulse repetition frequency of
500-10,000 Hz, and a pulse length of 5-10 psec. The neutron
source may operate continuously.
The control and synchronization unit detects y-radiation
during specified time intervals and synchronizes the operation
of the apparatus at the various stages of the measurement.
Amplitude-to-time pulse analyzers may be used as recorders.
The apparatus weighs about 12 kg (without the recorder).
It draws no more than 20 W. The remote unit is 110 mm in
diameter and 1300 mm long.
The complete arH.ilo contains experimental results
Measurement of natural and induced activity Lt illustrating the operation of the apparatus and its capabili-
? Gate closed
p Gate open
ties for directly determining the coal content (with a bore
model) and the content of the basic rock-forming elements.
The results of measurements with a semiconductor y-detec-
Fig. 1. Operation of the apparatus. tor are also reported.
MEASUREMENT OF BEAM PARAMETERS FOR a-PARTICLES
EXTRACTED FROM THE JINR* HEAVY-ION 2 METER
ISOCHRONOUS CYCLOTRONS
V. S. Alfeev, E. D. Vorob'ev, UDC 621.384.02
G. N. Zorin, and Yu. I. Kharitonov
In this study the energy of a-particles was measured using Rutherford scattering on a thin gold foil
(0.25 ?) at an angle of 20? l.s. (laboratory system), since in this region of angles for the Au197 (a, a) reaction
tTranslated from Atomnaya Energiya, Vol. 31, No. 2, p. 140, August, 1971. Original article sub-
mitted November 9, 1970.
tTranslated from Atomnaya Energiya, Vol. 31, No. 2, pp. 140-141, August, 1971. Original article
submitted October 12, 1970.
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 Declassified and Approved For Release 2013/02/26: CIA-RDP10-02196R000300090001-2
0=20?keV
MeV He4
1
Fig. 1. The energy spectra of
the scattered a-particle beam
extracted from the JINR U-200
heavy ion 2 meter isochronous
cyclotron and the a-calibration
lines.
It 4 if
190 220 800 840
The number of the analyzer channel
channel, the cross section of elastic scattering of a-particles in the energy range of 30-40 MeV exceeds
the cross sections of inelastic processes by several orders of magnitude. The beam, with a geometric
radius of 860 mm, was extracted from the JINR U-200 cyclotron using a charge-exchange technique [1].
To detect the a-particles a semiconductor surface barrier detector with a 27 mm2 cross section was
used, which is made from high resistivity silicon and which has certain advantages over a Li-drifted detec-
tor [2]. The detector was situated at a 45? angle l.s. relative to the scattered beam, and at a bias of 300 V
provided a sensitive layer of -800 p, corresponding to the mean free path of a-particles with an energy of
-45 MeV. At 200 V bias on the detector the resolution was equal to 20 keV for a known a-line from Am241
(5.482 MeV).
After the removal of the spectrum of scattered a-particles, the scale of the 1024 channel analyzer
was determined by a precision pulse generator which was calibrated at the known a-lines from Po212 (8.776
MeV) and Po 211 (6.774 MeV).
The measured energy of a-particles extracted from the U-200 is equal to (36.5 f 1.0) MeV, and the
degree of monochromatism of the a-beam, allowing for the inherent resolution of this procedure, is equal
to (350 ? 50) keV, which is 1% of the beam energy (see Fig. 1).
It has been established that one of the basic contributions to resolution is the instability of the high-
frequency voltage on the dees of the cyclotron.
As a check on the measurements, a stack of aluminum foils with a common density of (100 ? 4) mg
/cm2 was introduced. The thickness of each foil was 6 p. The energy of the a-particles after the absorber
was equal to 17.5 MeV, which corresponds to an energy of primary a-particles of (36.0 ? 1.4) MeV [3].
The energy measurement by means of absorbers agrees well within the limits of error with the mea-
surements by the precision pulse generator.
LITERATURE CITED
1. I. A. Shelaev et al., Pribory i Tekh. Eksperim., No. 3, 53 (1970).
2. G. Andersson-Lindstroem, Nucl. Instrum. and Methods, 56, 309 (1967).
3. L. Northcliffe, Ann. Rev. Nucl. Sci., 13, 67 (1963).
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 Declassified and Approved For Release 2013/02/26: CIA-RDP10-02196R000300090001-2
METHOD OF MEASURING (p, n)-THRESHOLDS FOR THE
STUDY OF ACCELERATOR BEAM ANALYZING SYSTEMS
M. I. Afanas'ev, A. L. Bortnyanskii, UDC 621.384.6
and A. I. Graevskii
A system is proposed for automatically recording the yield of nuclear reactions. The energy of the
incident particles is changed by modulating the potential on the target. A sawtooth voltage on the target is
obtained by means of its charge through capacitive coupling to the beam current and its rapid discharging
using a high-voltage relay. As the potential increases, the channels of a pulse-height analyzer are shifted
by a current integrator. The yield of the nuclear reaction is registered in the memory of the analyzer
synchronously with the increase of the target potential. The analyzer operates in a multi-channel accumu-
lator mode.
A cassette of boron counters in a paraffin moderator was used as a neutron detector. The construc-
tion of the chamber permits the detection of the total neutron yield in a cone with an opening angle of 90?;
at the same time, the generating cone does not touch the metallic components, which makes it possible to
detect all neutrons in the reaction A127(p, n)Si27 in the range 8 keV above the threshold.
Small leakages in the chamber insulation (5.10-9 A) permit operation with beam currents up to 10-3A
without significant distortions in the form of the excitation function.
A curve is cited to illustrate the relative neutron yield in the A127(p, n)Si27 reaction in the vicinity of
the threshold. The scanning range of the target potential is 25 kV.
With the same target chamber insulation, measurements may be taken in an energy range two times
greater, by charging the target negatively beforehand. In this case the target potential will increase to
zero, and then to the previously given maximum positive potential. The apparatus, described permits mea-
surements in the bipolar scanning mode of the target potential. At the same time, before the beginning of
each cycle the high-voltage relay connects the target to a rectifier, at the output of which the specific volt-
age is established.
The excitation function of the C13(p, n)N13 reaction near the threshold is presented, which was obtained
in the bipolar scanning mode. The scanning amplitude is 25 kV.
Translated from Atomnaya Energiya, Vol. 31, No. 2, p. 141, August, 1971. Original article sub-
mitted November 9, 1970.
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 Declassified and Approved For Release 2013/02/26: CIA-RDP10-02196R000300090001-2
EXPANDED CAPACITY RADIATION LOOP AT THE
IRT.NUCLEAR REACTOR IN TBILISI
G. I. Kiknadze, E. L. Androriikashvili,
V. S Bedvenov, I. A. Gassiev, G. V. Zakomornyi,
D. M. Zakharov, B. I. Litvinov, R. B. Lyudvigov,
L. O. Mkrtichyan, I. A. Natalenko, and L. I'. Feldman
The first immersion type RK-P indium-gallium radiation loop was commissioned in 1963 at the nu-
clear reactor of the Institute of Physics of the Academy of Sciences of the Georgian SSR [1]. In - 30,000 h
of steady operation, this facility demonstrated excellent reliability and ease of operation [2].
Fig. 1. Basic layout of the RK-PM radia-
tion loop facility: 1) activity generator;
2) overflow tank; 3,10) float level gages;
4, 9) globe valves; 5) electromagnetic in-
duction pump; 6) electromagnetic flow-
meter; 7, 8) large irradiator and small
irradiator; 11) gas holder.
The radiation loop was dismantled in 1967 while the re
actor was being rebuilt. The entire dismantling operation
took 5-6 h. The radiation environment presented no serious
hazard to the personnel engaged in the dismantling operation.
After the reactor had been rebuilt and its power output
had been brought up to 4-5 MW [3-5], construction work was
begun on a modernized RK-PM radiation loop, installation of
which was based on the same design principles.
The new radiation loop was built by the end of 1968 and
was brought up to rated irradiation conditions. in January,
1969. The -y-radiation carrier in this renovated facility was
again a binary indium -gallium alloy of eutectic concentration
[6]. Contact between the alloy and oxygen present in the at-
mosphere was prevented by a protective blanket of inert gas.
The layout of the facility is shown in Fig. 1, and a gen-
eral view of the facility appears in Fig. 2. The components
of the facility are combined in the following four principal
structural units.
1. The activity generator, comprising a four-layer sys-
tem of branching slotted channels 3 mm thick and with trans-
verse dimensions 410 ? 500 mm, joined by common collectors
(Fig. 3). The layers are separated by graphite slugs 40 mm
thick. The casing of the generator and the slotted channels
are made of titanium.
The choice of that design reflected the possibility of
more complete utilization of leakage neutrons from the re-
actor, and the possibility of increasing the neutron self-shield-
ing factor in multilayered systems [7, 8].
The total volume of the alloy present in the activity
generator is 2.6 liters. Two resistance thermometers were
installed in the top and central parts of the activity generator,
in titanium wells sunk in the graphite slugs.
Translated from Atomnaya Energiya, Vol. 31, No. 2, pp. 143-145, August, 1971. Original article
submitted August 6, 1970.
o 1972 Consultants Bureau, a division of Plenum Publishing Corporation, 227 West 17th Street, New York,
N. Y. 10011. All rights reserved. This article cannot be reproduced for any purpose whatsoever without
permission of the publisher. A copy of this article is available from the publisher for $15.00.
Declassified and Approved For Release 2013/02/26: CIA-RDP10-02196R000300090001-2
 Declassified and Approved For Release 2013/02/26: CIA-RDP10-02196R000300090001-2
6,660
Fig. 2. General view of the RK-PM radiation loop: 1) activity
generator; 2) tray under activity generator; 3) underwater
chamber; 4) tray with irradiator; 5) tubular girder; 6) central
channel of large irradiator; 7) central channel of small irra-
diator; 8, 9) process channels; 10) channel for electrical cable
and drain valve control switch; 11) channel for irradiator globe
valves control switch; 12) topside support area; A) water level;
B) irradiator positioning level.
A protective pan made of 1Kh18N9T steel is placed under the activity generator and under the feed
piping, to help cope with any alloy leakage emergency.
2. The underwater chamber, made of 1Kh18N9T steel, is shaped as a cylinder 520 mm in diameter
and 800 mm in height. An electromagnetic linear induction pump and a magnetic flowmeter completely .
identical to the ones used earlier in the RK-P radiation loop are installed inside the chamber. Also located
there are the overflow tank of 15 liters capacity, also made of 1Kh18N9T steel, and provided with a float
level gage with a differential-transformer type sensor. The working recorded travel of the sensor float
(50 mm) monitors the top half of the tank capacity, and aids in checking on how completely the y-ray carrier
is drained from the loop piping. A stop valve, also located in the underwater chamber, is used to control
overflow.
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 Declassified and Approved For Release 2013/02/26: CIA-RDP10-02196R000300090001-2
3. The system of two irradiators, the stop valves serv-
ing them, and the float level gage are mounted on a rigid tray
which is coupled to the bottom of the facility by four standpipes.
These standpipes function simultaneously as process lines
through which loop piping, electrical cableware, and the remote
2 control switch for the globe valves pass. The irradiators are
3 fabricated in the form of hollow cylinders with the alloy flow-
Fig. 3. Activity generator for RK-PM
radiation loop: 1) top collector; 2) graph-
ite slug; 3) slotted channel; 4) mounting
bracket for generator; 5) reference ad-
justing screw; 6) tray.
ing along a helical track between them. The globe valves make
it possible to switch either of the irradiators, or both simul-
taneously, into the circulation system. The irradiators are
equipped with central and peripheral channels to facilitate ir-
radiation exposures in a field of pure y-radiation and dosi-
metric monitoring of the y-emission dose rate of the. facility.
The float level gage located at the uppermost point of the
circulating system makes it possible to measure the amount of
y-ray carrier present in the system, and to reliably record
cases of accidental leakage of alloy with a sensitivity of 20 cm3.
1) 4. The top support area with the pipe-suspension system,
forming a rigid girder truss connecting all the units enumer-
ated above in a single structure is shown in Fig. 2. The chan-
nels of the irradiators in the irradiation loop, and the channels
for the remote control switches actuating the globe valves, are
mounted on the top support platform. A system of control adjustment screws on the support platform makes
it possible to adjust the activity generator with precision relative to the boundary of the IRT reactor core
[91.
The loop control panel .is in the form of a memory panel, giving the operating personnel up-to-the-
minute information on the performance of the facility.
The instruments mounted on the control panel take care of monitoring the temperature in the activity
generator and in the channel for the induction pump, the level of the indium-gallium alloy in the circulating
system and in the overflow tank, currents flowing through the windings of the induction pump, the flowrate
of y-ray carrier as it circulates through the system, and the dose rate in the two irradiators.
The emergency and alarm annunciation system (which is transistorized) rings an audible alarm signal
and switches on the appropriate alarm bulb on the memory panel of the loop in the event the temperature
rises, the flowrate declines, there is leakage of inert gas, the pump malfunctions or shuts off, or the fill
level of the alloy declines.
The dose rate of y-radiation at the center of the loop's large irradiator is -400 r/sec; that at the
center of the loop's small irradiator is - 500 r/sec, for each MW of reactor power output when the two ir-
radiators are operating simultaneously.
When the large irradiator is shut off, the dose rate in the small irradiator is 850 r/sec for each
MW of reactor power output. The y-equivalent of the arrangement is 60,000 g-eq Ra per megawatt. The
operating flowrate of the y-ray carrier is - 7 cm3/sec, the duration of a single alloy circulation period is
30 min. The temperature of the y-ray carrier attains its peak in the activity generator, specifically 63?C
when the reactor output is at 1 MW, and 155?C when the reactor output is at 4 MW. The total volume of
y-ray carrier is 13.1 liters, of which 2.6 liters are present in the activity generator, as indicated earlier,
7.5 liters are present in the large irradiator, 2.5 liters in the small irradiator, and 0.5 liters in the piping.
The peak radiation power output of the RK-PM facility, attained when the thermal power output of the
IRT reactor is 5000 MW, was 300,000 g-eq Ra.
The RK-PM radiation loop differs from its precursor not only in the significant increase in radiation
dose rate, but also in terms of compactness and ease of assembly. A five-man team carried out the rigging
and assembly work for the whole facility in the reactor vessel within eight hours.
A large number of exposures have already been performed in the channels of the RK-PM radiation
loop in pilot production and research work. The commissioning of this radiation loop has expanded the ex-
perimental capabilities of the reactor facility considerably. Radiation loops of this type.can be constructed
with ease at any existing pool type reactor facility.
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LITERATURE CITED
1. G. I. Kiknadze et at., At. Energ., 19, 176 (1965).
2. E. D. Chistov et at., Nauchnye Raboty Inst. Okhrany Truda VTsSPS, 48, 42 (1967).
Safety Institute of the All-Union Council of Trade Unions], 48, 42 (1967).
3. Sh. P. Abramidze et al., Proceedings of the Jubilee Session of the Institute of Physics of the Academy
of Sciences of the Georgian SSR devoted to the 50th Anniversary of the October Socialist Revolution,
Tbilisi, 1968 [in Russian], Inst. Fiziki Akad. Nauk GruzSSR (1969).
4. Sh. P. Abramidze et al., At. Energ., 27, 547 (1959).
5. Sh. P. Abramidze et at., Report to the Vth International Conference on the Physics and Engineering
of Research Reactors, Warsaw, 1968 [in Russian].
6. G. I. Kiknadze et al., At. Energ., 19, 178 (1965).
7. G. I. Kiknadze, Doctoral Dissertation (in Russian), Tbilisi (1968).
8. G. I. Kiknadze et al., Report to the IVth Working Conference on the Physics and Engineering of Re-
search Reactors, Budapest, 1965 [in Russian].
9. V. N. Chernyshevich et al., Report to the Vth International Conference on the Physics and Engineer-
ing of Research 'Reactors, Warsaw, 1968 [in Russian]..
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FABRICATION OF LIQUID-METAL WORKING
MATERIALS FOR RADIATION LOOPS
G. I. Kiknadze, D. M. Zakharov, UDC 621.039.573.669.87
R. B. Lyudvigov, and L. I. Feldman
The fabrication of liquid-metal y-ray carriers, for example, indium -gallium or indium -gallium -tin
alloys, is one of the principal problems in the design of radiation loops. In each specific case, the proce-
dure followed in the fabrication of these materials may differ in certain technological features of the pro-
cess, but the following general rules do have to be observed consistently in each instance.
1. Structural materials of high stability to attack by the liquid-metal environment must be used in
the melting facility, and most particularly in the piping and parts coming into direct contact with the liquid
-y-ray carrier.
This requirement pursues the aim of eliminating any possibility of the y-ray carrier being contami-
nated by extraneous elements (as a result of possible dissolution of the structural material in the alloy) or
by interaction products forming on the interface between the structural material and the melt. Quartz or
titanium single-phase and two-phase alloys (such as VT1-1, VT-6, etc.) can be recommended as suitable
structural materials in these applications. The use of titanium is preferred for technological reasons. It
would also be desirable to carry out a preliminary passivation of the titanium in air at temperatures 550-
570?C for 100 to 150 h. This treatment leads to the formation of a TiO2 oxide film on the surface of the
titanium, thereby preventing any interaction between the titanium and the indium-gallium alloy [1, 2].
2. A special degassing of the metals comprising the charge to the -y-ray carrier is required, during
the meltdown and in heating to the specified temperature.
This requirement pursues the aim of lowering the oxygen content and lowering the content of other
gases dissolved in the original metals, and also of removing (at least partially) any volatile gallium oxides
or indium oxides that may be present. The degassing process must be carried out gradually, by pumping
out the vapor phase from the melting vessel into which the charge has been introduced previously, down to
a residual pressure of 10-3 to 10-4 mm Hg over the entire period that the metals are being melted down and
heated to the specified temperature. The heating rate is 1.5-2 deg/min.
3. Special attention must be given to isothermal holding of the melt for several hours with bubbling of
the inert gas through the melt. The isothermal holding temperature must not be higher than 500?C, and in
case titanium is used as the structural material the maximum heating temperature must be lowered to
300?C. Bubbling with inert gas (helium best of all) makes it possible to: a) carefully mix the components
of the melt by mechanical means; b) bring the oxide phase present in the bulk of the liquid y-carrier in the
suspended state to the surface of the melt.
Isothermal holding of the liquid phase promotes an equalization of the concentration of the components
of the y-ray carrier throughout the bulk of the melt through intensification of the diffusion processes at
work. It should be stressed that bubbling of inert gas through the melt may result in the formation of a
finely dispersed phase of black coloration (this means that gallium oxides and indium oxides, or gallium
alone, may turn up in the melt [3-5]), and this phase would be present in the suspended state throughout the
free space of the melting vessel, gradually settling on the walls and on the surface of the vessel as a "cap"
of oxides covering the melt.
Translated from Atomnaya Energiya, Vol. 31, No. 2, p. 146, August, 1971. Original article sub-
mitted August 6, 1970.
? 1972 Consultants Bureau, a division of Plenum Publishing Corporation, 227 West 17th Street, New York,
N. Y. 10011. All rights reserved. This article cannot be reproduced for any purpose whatsoever without
permission of the publisher. A copy of this article is available from the publisher for $15.00.
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 Declassified and Approved For Release 2013/02/26: CIA-RDP10-02196R000300090001-2
4. The next stage is cooling of the liquid-metal -y-ray carrier to temperatures close to room tem-
perature, and careful filtration of the carrier.
The cooling rate is 2-3 deg/min. The filtering system consists of an array of filters featuring a grad-
ual transition from rough clean-up to fine filtering. Stainless steel mesh combined with Petryanov fabric,
caprone fabric, and quartz shavings is suitable for filter aids; Schott filters and porous graphite can be
used for the fine filtration step.
5. Periodic analysis of samples of the y-ray carriers for their content of basic components is ad-
vised.
Analysis of the y-ray carrier during the isothermal holding process and during the bubbling process
is particularly crucial. A constant level of the concentration of components of the y-ray carrier with time
is an objective indicator of the quality with which the alloy has been prepared, and can serve as a signal to
terminate the isothermic holding and bubbling steps. Analysis of the y-ray carrier for the content of the
basic components is also required before the radiation loop is filled with alloy (after the y-ray carrier has
been through the filtering system). This makes it possible to secure reliable information on the composi-
tion of the y-ray carrier fulfilling the functions of a working fluid for the radiation circulation loop.
6. The entire complex of operations involved in preparing the y-ray carrier and in filling the radia-
tion loop with it must be carried out in the strict absence of any contact between the alloy and oxygen in the
air, or moisture.
This stresses the necessity of monitoring the oxygen content and moisture content in an inert gas,
and where the need arises taking measures to purify the inert gas. Traces of moisture must be removed
from the piping and from the various units involved in the preparation of the alloy.
The principles formulated here were arrived at in the course of work done at the Institute of Physics
of the Academy of Sciences of the Georgian SSR [6], and can be considered a basis for the design of equip-
ment intended to prepare large batches of alloy for industrial-scale radiation loops.
The authors express their sincere thanks to B. I. Litvinov for his highly appreciated participation in
the design and operation of the facilities for preparing the liquid-metal y-ray carrier.
LITERATURE CITED
1. L. S. Moroz et al., Titanium and Titanium Alloys [in Russian], Vol. 1, Sudpromgiz, Leningrad (1960).
2. G. I. Kiknadze, D. M. Zakharov, and L. V. Mel'nikova, At. Energ., 19, 177 (1965).
3. M. Robert, Compt. Rend., 5, 51 (1964).
4. A. G. Godzhello, Trudy Mosk. Energ. Inst., No. 5, 354 (1964).
5. S. P. Yatsenko, G. N. Perel'shtein, and D. V. Lokshin, Trudy Inst. Khimii, Ural'skii Filial Akad. Nauk
SSSR, No. 18, Sverdlovsk (1968).
6. I. G. Kiknadze et al., At. Energ., 19, 176 (1965).
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 Declassified and Approved For Release 2013/02/26: CIA-RDP10-02196R000300090001-2
T. S. Ambardanishvili, G. V. Zakomornyi,
G. D. Kiasashvili, G. I. Kiknadze,
B. I. Litvinov, L. O. Mkrtichyan,
and A. M. Uvarov
The pneumatic reloading system for the vertical channels of the IRT-M reactor in Tbilisi incorporates
the following principal components (Fig. 2).
1. The specimen irradiation channel located in the core or in the reflector is an array of two coaxial
tubes of 1Kh18N9T steel, 38 mm and 57 mm in diameters, wall thickness 2 mm and 2.5 mm, respectively.
At the blind end of the inner tube in the channel is mounted a special valve normally closed when the channel
is being loaded and open under the pressure of the working gas when the channel is being unloaded. This
Fig. 1. Irradiation channel: 1) bottom;
2, 3) inserts;
4) channel inner tube;
5)
channel body;
6) distance spacer;
7)
nut; 8) flange;
9) bolt; 10, 13, 17,
19)
washers; 11) pipe connection; 12) union
nut; 14) pipe connector; 15) pipe; 16)
travel stop; 18) valve.
Translated from Atomnaya Energiya, Vol. 31, No. 2, pp. 147-150, August, 1971. Original article
submitted July 31, 1970.
? 1972 Consultants Bureau, a division of Plenum Publishing Corporation, 227 West 17th Street, New York,
N. Y. 10011. All rights reserved. This article cannot be reproduced for any purpose whatsoever without
permission of the publisher. A copy of this article is available from the publisher for $15.00.
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 Declassified and Approved For Release 2013/02/26: CIA-RDP10-02196R000300090001-2
I
1 0 U
Fig. 2. Pneumatic reloading system of vertical channels of IRT-M reactor:
1) specimen irradiation channel; 2) main tube; 3) air tube; 4) distributor; 5)
sensor; 6) bypass valve; 7) motor; 8) charging device; 9) pneumatic valve;
10, 22) globe valves; 11) recess in reactor shielding; 12) reactor hot cave; 13)
receiving chamber in reactor hot cave; 14) hot cave for radiochemical labora-
tory; 15) receiver in hot cave of radiochemical laboratory; 16) receiver tank;
17, 18) pressure gages; 19) high pressure emergency controls; 20) compres-
sor; 21) decontamination unit.
valve also acts as a stop valve converting the channel into a pulsation-dampening volume to allow smooth
descent of the specimen during the loading operation. The layout of the channel is shown in Fig. 1.
When a specimen is being loaded into the channel, the excess amount of working gas is dumped
through the bypass valve located between the feed line for the working gas and the pneumatic shuttle con-
veying channel.
An electrical sensor records the entry of the specimen into the zone where it is expedited to the
vertical portion of the channel. This sensor then shuts off the system pumping the working gas which
transports the loaded specimen through the channel, and records the passage of the specimen in the appro-
priate cell of the memory panel.
2. The distributor (Fig. 3) is a device providing a combination of any of five channels reserved for
the pneumatic shuttle system at the IRT-M reactor in Tbilisi, and one of the three channels reserved for
loading of specimens, the path to the radiochemical laboratory, or the path to the operational hot cave in
the reactor building.
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 Declassified and Approved For Release 2013/02/26: CIA-RDP10-02196R000300090001-2
Fig. 3. Distributor: 1) container duct to channels; 2) gasket; 3) hood; 4) elec-
trical microswitches; 5) combination channel; 6) five-groove Maltese cross; 7,
9, 17) gears; 8, 11, 18) driven gears; 10) insert; 12) loading line; 13) radio-
chemical laboratory line; 14) operational hot cave line; 15) half-body of container
duct; 16) electric power drive; 19) channel half-body.
The distributor thus consists of two parts kinematically linked with each other by a system of trans-
missions.
The two parts of the distributor are brought into action by an electric power drive, with the aid of a
five-groove Maltese cross. The combination of the pneumatic shuttle channels and their mutual fixed posi-
tions is brought about by using a special sector of a driving disk. Teflon gasketing is used to provide seal-
ing between the five-channel and three-channel parts of the distributor.
The sensors for the combination of channels in the three-channel and five-channel parts of the dis-
tributor are microswitches included in the control circuitry of the electric drive. The electric power
drive, and with it the combination of channels on the two parts of the distributor, is stopped by a pro-
grammed matching of protrusions on the rotating parts with microswitches mounted on the frame of the
distributor.
The accuracy with which the reactor channel is matched to one of the three specimen-conveying chan-
nels is.achieved by using the microswitch to shut off the power drive motor at the instant the Maltese cross
is engaged by the sector on the drive disk. The system transmitting motion from the five-channel part of
the distributor to the three-channel part is designed such that displacement of the five-channel part by one
channel corresponds to displacement of the three-channel part by one channel also, so that the switchover
can be accomplished through the action of the same electric power drive.
The distributor is housed in a recess in the biological shield of the reactor, with pressuretight access
doors. This recess is hooked up to the special ventilation system (to evacuate the radioactive gas liber-
ated, when specimens are discharged, through possible clearances between the five-channel and three-chan-
nel parts).
3. The loading device, acting as a locking mechanism, is accommodated in the same recess in the
shield as the distributor. The working gas is fed during the loading operation through the action of electro-
magnetically actuated valves. The working gas exists from a receiver tank with tank pressure maintained
at set level by a compressor. The compressor motor is controlled by using an electrical contact pressure
gage which sets the upper and lower pressure limits in the receiver tank. The working pressure in the re-
ceiver tank is 4 atm. In case of emergency, there is a reserve supply of working gas available in pressure
cylinders, which can be connected up to the receiver tank by opening globe valves.
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A container for conveying the specimen is made up of two thread-engaged parts made of aluminum or
plastic, and forms a cylinder with spherical ends. The diameter of the container is 32 mm, the length 65
The pneumatic shuttle control system makes it possible to carry out the following operations:
1) combination of either of five channels in the core and the loading device, either with the discharge
lines leading to the radiochemical laboratory or with those leading to the operational hot cave;
2) loading of containers with the test specimens in the five channels;
3) discharge of containers with test specimens to the radiochemical laboratory or to the operational
hot cave;
4) interlock against loading of more than one container into any one channel;
5) interlock of loading and unloading until the combining operation has been completed;
6) interlock of distributor motor starting until the end of the loading and unloading operation;
7) air pressure in receiver tank maintained within set range;
8) loading and unloading interlocked when pressure in receiver tank falls below or rises above set
range;
9) annunciation of combining of channels for container reloading operations, lowering or raising pres-
sure in receiver tank, and also on movement of container through one of the three conveying lines, by
switching on appropriate light bulbs on the front panel of the control desk.
In addition, the pneumatic shuttle control system takes care of restoring the basic signals recorded
in the overall system memory, when the electrical power supplies are switched on again after being dis-
connected or deenergized accidentally.
Measures were taken to decrease the number of contacts and to decrease the amount of current flow-
ing through them, with the object of enhancing system reliability, and this was achieved through the use of
transistors and semiconductor diodes.
An indication of container loading into the irradiator channel is achieved through the use of a memory
cell in the pneumatic shuttle control system. A twin-coil polarized relay acts as the memory element.
When the container passes by the corresponding load sensor, the circuit supplying one of the relay
coils is broken, and the current flowing through the other coil then throws the yoke into the "loaded" posi-
tion. A green light bulb then goes out and a red bulb goes on on the control panel, and the corresponding
transistor is driven to cutoff, thus interlocking any loading of a second container while the channel is "busy."
Electrical sensors feeding information on the progress of the container are located every ten meters
along the travel path. Signals from these sensors light up bulbs on the memory panel of the pneumatic
shuttle control system, making it possible to pinpoint the position of the container in the event of an acci-
dental shutdown.
It has been proposed to work out a completely noncontacting system using controlled valves, in the
future.
The control panel is made in the form of a memory block for the pneumatic shuttle system.
The receiver device placed in the hot receiving cave for reception and forwarding of specimens to the
radiochemical laboratory consists of two coaxial tubes (1Kh18N9T steel) 38 mm and 57 mm in diameter,
with respective wall thicknesses 2 mm and 2.5 mm. The valve for venting the conveying gas discharges
spent gas into the tube space as the container passes by, and the tube space communicates to the hot cave,
where the spent gas is drawn into the special ventilation system of the radiochemical laboratory. After the
valve has done its work, the vertical portion of the receiving channel becomes converted into a damping
volume which lowers the container down smoothly into the hot cave. The container falls into a spherical
trap with a swing-out flap cover, and is withdrawn from there by a remote-controlled manipulator.
The basic parameters of the pneumatic shuttle are: operating pressure when container is loaded in
0.5 atm, gas flowrate 20 liters, pressures of 1.0, 1.3, and 1.5 atm for discharge at respective speeds 2, 4,
and 6 m/sec; average flowrate of conveying gas over a 100 meter stretch of piping: 310 liters.
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 Declassified and Approved For Release 2013/02/26: CIA-RDP10-02196R000300090001-2
GAS EVOLUTION IN THE PRIMARY LOOP OF A
PRESSURIZED-WATER REACTOR WITH GAS
VOLUME COMPENSATORS
The conditions that guarantee that there will be no gas evolution in the primary loop of a pressurized-
water (water-cooled, water-moderated) reactor with gas volume compensators (VC) were considered in [1,
2]. The presented temperature dependence of the solubility of a gas in water for a constant total pressure
P0 taking account of the vapor pressure Pv of the water is correct for a heterogeneous gas -water system
[2].
In the primary loop, the only such system is the VC. In the articles indicated above, an estimate of
the solubility of the gas in the primary coolant, where there are no gas pockets, was also carried out taking
the vapor pressure of water into account. However, there is a basic difference between the concepts of gas
solubility in heterogeneous and in homogeneous systems. Therefore, the conclusions drawn concerning the
conditions for gas evolution in the primary loop are incorrect.
According to Henry's law, the solubility of a gas in water for given P0 and To depends on the partial
pressure of the gas Pg in the vapor-gas mixture above the level of the water, where, with increasing par-
tial pressure of the gas, its solubility in the water increases [3]. We thus expect that the maximal solu-
bility, for given parameters, will occur for Pg = P0, i.e., when there is no water vapor in the gas phase.
Such conditions are possible, e.g., because of the constant exchange of gas above the level of the water.
To confirm the above, we carried out the following experiment. Completely degasified water was
poured into an open container, and the temperature was kept at 100?C during the entire experiment - pre-
venting boiling. Measurement of the concentration of nitrogen in water over several hours showed that the
nitrogen content is equal to the solubility at 100?C and nitrogen partial pressure of 1 kgf/cm2. The experi-
ment was conducted at atmospheric pressure. The solubility of nitrogen at 100?C and atmospheric pres-
sure, determined taking into account the partial pressure of the water vapor above the water level, was
equal to zero. No evolution of nitrogen within the volume of water was observed.
Fig. 1. Temperature dependencies of
the partial pressure Pg and also of Pg
+ Pv for C = const [all concentrations
are in liters per kilogram at normal
temperature and pressure (NTP)]: 1)
C=0; 2)C=1; 3)C=2; 4)C3; 5)
C5; 6)C=7.
Translated from Atomnaya Energiya, Vol. 31, No. 2, pp. 150-151, August, 1971. Original article
submitted August 3, 1970; revision submitted November 11, 1970.
? 1972 Consultants Bureau, a division of Plenum Publishing Corporation, 227 West 17th Street, New York,
N. Y. 10011. All rights reserved. This article cannot be reproduced for any purpose whatsoever without
permission of the publisher. A copy of this article is available from the publisher for $15.00.
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 Declassified and Approved For Release 2013/02/26: CIA-RDP10-02196R000300090001-2
Hence it follows that the solubility of a gas in a homogeneous system Chom equals the solubility of
the gas in a heterogeneous system Chet determined for the condition that the partial pressure of the gas
above the water level equals the total pressure.
Thus, there should be no evolution of gas bubbles in a homogeneous system for concentration of the
gas in water less than Chom. Especially because the formation of centers in the gas phase is limited, for
a radius of curvature of the bubble surface less,than 10-5 cm, the pressure inside them noticeably increases
because of the surface-tension forces [4].
The gas concentration in the primary coolant has a definite steady-state value for constant pressure
and constant temperature in the VC [2]. Therefore, in order to determine the conditions for gas evolution
in the primary loop, it is more convenient to use the temperature dependence of the gas partial pressure
for constant gas concentration in water C.
The temperature dependence of the nitrogen partial pressure calculated according to the data of [5,
6] is shown by the continuous lines in Fig. 1. The partial pressure of nitrogen decreases with increasing
temperature.
The dot-dash lines show the temperature dependence of the total pressure of the vapor-gas mixture
for an equilibrium distribution of gas in the heterogeneous gas -water system at a constant nitrogen con-
centration in water. The curves have a minimum, which, with increasing nitrogen concentration in water,
shifts toward higher temperatures.
The condition for the formation of a gas phase within the water volume is Pg > P0. For a hetero-
geneous gas -water system, gas will evolve when Pg + Pv > PO.
For a known temperature TVC in the VC and pressure Po, the nitrogen concentration Ci in the water
found in the VC, and hence also in the primary coolant, is determined from curve A, since for a hetero-
geneous gas-water system, the condition Pg + Pv = Po is satisfied. The dependence of the nitrogen partial
pressure Pg on the temperature, for a nitrogen concentration in water Ci, is represented by curve B.
From Fig. 1 it follows that evolution of nitrogen inside the volume of water is possible only for T11
< Tmin, where T11 is the temperature of the coolant in the first (primary) loop. Note that Tmin < TVC.
For artificial introduction of gas into a homogeneous volume or local boiling of the coolant, a heterogene-
ous system forms. The conditions for evolution of nitrogen from the water are T1l < TVC and Til > Tm,
where Tm is the maximal temperature characterizing the equilibrium state of the heterogeneous system,
for a nitrogen concentration in water Ci and specified parameters. In the temperature range TVC < Til
< Tm, the gas in the gas pocket that was formed earlier will go into solution.
It follows that when there is dissolved gas in the water, the boiling point decreases somewhat. When
the water boils, the gas that was dissolved in it produces a definite partial pressure Pig in the volume of
vapor that forms. We also have that the pressure of the water vapor is Piv = PO-Pil. Hence, the water
will boil at the temperature Tis that corresponds to the saturation state for pressure Piv.
The decrease in boiling point AT = Ts-Tis, where Ts is the boiling point of pure water at pressure
Pp, was calculated for various nitrogen concentrations in water at pressures 100 and 170 kgf/cm2. The
results can be represented by
OTS00=0.090C, ?C;
AT17o=O.133C, ?C,
where C is the concentration of nitrogen in water ,(liters per kilogram at NTP).
Thus, the condition for the absence of gas evolution in the primary loop of a pressurized-water re-
actor with gas volume compensators is the maintenance of a temperature in the VC lower than the maximal
temperature that is feasible in the primary coolant; also, a temperature regime must be observed that en-
sures that there will be no local boiling of the water. In order to dissolve the vapor-gas pockets that were
formed earlier, it is necessary to reduce the temperature in the loop for a short time down to a value slightly
less than Tm.
1. V. S. Sysoev, At. Energ., 26, 461 (1969).
2. N. V. Bychkov and A. I. Kasperovich, At. Energ., 28, 145 (1970).
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 Declassified and Approved For Release 2013/02/26: CIA-RDP10-02196R000300090001-2
3. V. A. Kirillin and A. E. Sheindlin, Thermodynamics of Solutions [in Russian], Gosenergoizdat, Mos-
cow-Leningrad (1956).
4. A. I. Rusanov, Phase Equilibria and Surface Phenomena [in Russian], Khimiya, Leningrad (1967),
p. 165.
5. H. Pray et al., Ind. Engng. Chem., 44, 1146 (1952).
6. T. Andersen, Trans. Amer. Nucl. Soc., 10, 507 (1967).
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 Declassified and Approved For Release 2013/02/26: CIA-RDP10-02196R000300090001-2
Kh. B. Krast and A. V. Byalobzheskii UDC 621.039
It is well known that the hardness and brittleness of metals increase under the influence of reactor
radiation as a result of increased internal stresses [1]. At the same time it was shown [2] that a relaxa-
tion of the microstresses occurs, for example, in uranium under the effect of the neutron flux. These
phenomena, which are observed also for nickel and zirconium, are explained by the diffusion of the defects
under the influence of the stress field.
Based on this an assumption was made that the reactor radiation must also affect corrosion under
stress. Due to the lack of published information the experimental procedure had to be worked out anew.
The experiments were carried out on the reactor IRT-2000 [3]. In view of the impossibility of carry-
ing out experiments at constant load the method of constant deformation was used, for which samples in the
form of loops [4] made from a 1.5 mm thick sheet of alloy AMg6M was. used. The blank was 160 x 15 mm
in dimensions [5].
Before the experiment the samples were subjected to artificial aging for 24 h at a temperature of
170?C. The position of the sample in the experimental unit is shown in Fig. 1. After bending the sample
into the loop a slit of 0.2-0.3 mm remained between the ends; this would not lead to any noticeable change
in the stress. An electric current from a 6 V source was applied to insulated contacts built-in in the sam-
ple. When the sample disintegrated, the contacts were closed due to the presence of a spring and a signal-
ing device for the current was switched on (a lamp, a bell, or such things).
After checking several solutions recommended for rapid determination of the susceptibility of alloys
to corrosion cracking [5] a solution consisting of 0.25 N NaCl + 0.2 N CH3COONa + 0.05 N CH3COOH (pH
= 5.15) was chosen, which did not cause ordinary corrosion. The volume of the solution was 150 ml. The
results of the experiments are presented in Table 1. As evident from the table, the time taken before the
Fig. 1. Experimental unit for experi-
ments on corrosion cracking in reactor:
1) potentiometer; 2) plug; 3) thermo-
couple in glass tube; 4) quartz test tube;
5) polyethylene container; 6) sample
loop; 7) solution; 8) insulated contacts;
9) clamp; 10) spring; 11) signaling de-
vice for current.
Translated from Zhurnal Atomnaya Energiya, Vol. 31, No. 2, pp. 151-153, August, 1971. Original
article submitted September 14, 1970; revision submitted February 1, 1971.
? 1972 Consultants Bureau, a division of Plenum Publishing Corporation, 227 West 17th Street, New York,
N. Y. 10011. All rights reserved. This article cannot be reproduced for any purpose whatsoever without
permission of the publisher. A copy of this article is available from the publisher for $15.00.
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 Declassified and Approved For Release 2013/02/26: CIA-RDP10-02196R000300090001-2
TABLE 1. Effect of Neutron Irradiation on Corrosion Cracking (Intensity of radiation
1012 neutron/cm2 ? sec)
Conditions of irradiation
Duration of the
experiment under
irradiation, h
Tempera-
ture,?C
pH at the
end of the
experiment
Time taken before-
disintegration
(average forthree
samples),h
Without irradiation
-
53
5,2
5-7,5
11,12 *t 11
50-55
6,4
120-140
Irradiation of the sample in solution
11
50-55
6,4
120-140
36
50-55
6,4
450
di
I
i
f
l
i
h
l
lit
50-55
6,4
6-10
rra
at
t
e so
on o
ut
on a
one
lit 53
7,1
5-6
Irradiation of the sample alone
11
53
5,2
150
' Without irradiation with subsequent irradiation for 11 h.
fi Thermal power of the reactor 1500 kW.
$ Thermal power of the reactor 2200 kW.
disintegration increases sharply on irradiation of the corroding system in the reactor. The pH of the sys-
tem also changes simultaneously. It was established from experiments that without irradiation the change
in pH was 0.1-0.2 units in 300 h.
Solutions of different compositions were irradiated in order to find out the reasons for the change in
the pH of the mixture. These experiments showed that the pH of sodium acetate solution changed most ap-
preciably. Thus the pH of 0.2 N solution before irradiation was 8.1 and after irradiation it was 9.45. The
increase in the alkalinity of the sodium acetate solution obviously determines the increase in the pH of the
solution as a whole. The reason for the increase in the pH of the acetate solution is not definitely known.
Perhaps it is related to the intensification of the process of hydrolysis of the acetate under the action of
the isotope Na24.
The slowing down of cracking (see Table 1) is caused not by the change in the composition of the cor-
rosion solution under the action of the reactor radiation, but by the action of the radiation on the metal lead-
ing to a relaxation of the existing stresses. Actually, if the unirradiated metallic loop is placed in a pre-
irradiated solution, a slowing down of the process of its cracking is not observed. On the other hand a pre-
irradiation of the metallic sample in air leads to a slowing down of its cracking even in an unirradiated solu-
tion. The slowing down of cracking is caused by neutron irradiation, since test experiments with y-radia-
tion of 600 rad/sec did not show any noticeable effect on corrosion cracking.
It must be mentioned that the inferences made on the basis of the present work pertain only to the in-
vestigated form of mechanical stresses. In the presence of stresses, less liable to relaxation, the results
may be different.
In conclusion the authors express their sincere thanks to A. A. Tsvetaev and E. M. Zaretskom for
help in setting up the experiments and for taking part in the discussion of the obtained results.
LITERATURE CITED
1. D. M. Skorov (editor), Metallography of Reactor Materials [in Russian], Part 2, Gosatomizdat, Mos-
cow (1962).
2. S. T. Konobeevskii, Effect of Radiation on Materials, Atomizdat, Moscow (1967).
3. Kh. B. Krast et al., At. Energ., 27, 286 (1969).
4. I. L. Rozenfel'd and K. A. Zhigalova, Fast Methods of Corrosion Tests of Metals [in Russian], Metal-
lurgiya, Moscow (1966), p. 283.
5. T. M. Sigalovskaya and E. M. Zaretskii, Zashchita Metal., 3, 730 (1967).
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 Declassified and Approved For Release 2013/02/26: CIA-RDP10-02196R000300090001-2
RELATION BETWEEN SOLUTIONS OF THE NONSTATIONARY
AND QUASICRITICAL TRANSPORT EQUATIONS
There are two types of equations describing noncritical reactors in the one-velocity approximation:
1) a nonstationary transport equation taking account of delayed neutrons
anti '--XiNi+ 4rz bivEf d12'ip; i=1, 2, ..., m
where 0 (r, Sl, t) is the neutron flux and Ni(r, t) is the concentration of the precursors of delayed neutrons
of group i; 2) a quasicritical equation of the form
E9=-52Vt+4- rdgp,
v at =-QV*-E*+4n (Es+bovEf) d4't-- XiNi;
i+1
m
where 11 is some parameter related to keff?
In general the relation between the solutions of these equations is rather complicated. From several
papers [1, 2] devoted to this problem it follows that a simple relation exists only for reactors that are
slightly noncritical.
Our procedure leads to a very simple relation in the one-velocity approximation even for a homo-
geneous bare reactor that is far from critical.
We seek a solution of Eq. (1) in the form 0(r, 9, t) = eatcp(r, n). Then we obtain
// a+1
vE +1l Ecp=-12VW-{ 4n Ug(a) d52'cp, (3)
(a)= ( +1/
1_~ aNi
a+?
i=1
Es+vE f . bivXf
c=
E L Es+y`,f.
Introducing the notation r' = r(1 + ct/vZ) and co(r) = cp(r'/[1 +(cx/vE)]) .= ip(r'), we transform Eq. (3) to the
form
EcpQVT 4yc (a) dQ'(-
The solutions of Eqs. (2) and (5) agree if 71 = t(a) and at some boundary L = R(1 + cx/vf) with normal
n both solutions satisfy the boundary condition
~)=W(L, ()=0, (On) -vs.
Thus the true neutron distribution in a nonstationary reactor, i.e., a solution of Eq. (3), can be deter-
mined.from a solution of the quasicritical Eq. (2) written for a reactor of a different size. The inverse
procedure is possible also.
1. L. N. Usachev, Proceedings of the First International Conference on the Peaceful Uses of Atomic
Energy, Vol. 5, Geneva (1955), p. 503.
2. A. Henry, Nucl. Sci. and Engng., 20, 338 (1964).
3. J. Mika, Nukleonik, 9, 46 (1967).
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 Declassified and Approved For Release 2013/02/26: CIA-RDP10-02196R000300090001-2
MOMENTS OF THE NEUTRON DENSITY DISTRIBUTION FUNCTION
T. E. Zima, A. A. Kostritsa, UDC 621.039
and E. I. Neimotin
Finding the moments of the neutron density distribution function N by Marshak's method [1] does not
require a knowledge of the function N or its Fourier transform N. It is sufficient to know the values of the
Fourier transform and its derivatives at the origin. This method is ordinarily used to determine moments
of the P1-approximation to the. distribution function. We interest ourselves in the moments of the exact
function N.
Let us consider first the one-velocity kinetic equation for the distribution function N(x, i, t) in plane
one-dimensional geometry with a localized pulsed neutron source
aN
?v? Bz +vEN= 2 1~i Pi (?) Pi (pV) N(x, [1', t) d?'+ 2 b (x) b (t-to). (1)
In writing Eq. (1) it was assumed that the scattering function ?g(4)) can be represented by a finite sum
of Legendre polynomials:
ftg (CD) _ ?iPi (cos (D).
Here ' is-the angle between the neutron velocity vectors before and after scattering; ? is the cosine of the
angle between the neutron velocity vector and the x axis; E is the total cross section; and Es is the scat-
tering cross section. We.assume also that ?k > ?k+i?
We introduce the Fourier transform of the function:
N (9, ?, t) N (x, It, t) eigx dx. (3)
By using the equation for the Fourier transform obtained from Eq. (1) it can be shown that the total number
of neutrons in an unbounded medium varies according to the well-known law
+-4-1 J J N(x, tt, t)dxd?=r S N(4, W, t) d?] _ = No(9=0)=S*(t-to)e-0Za(t-to)' (4)
where
flt-to) = 1 for t > to;
{0 for t < to.
Simple calculations give the value of the second derivative of No = N(q, ?, t)d? at the origin in the
plane of the complex variable q and then the second spatial moment of the neutron distribution function x2
as a function of time:
2v 1 -vzs ( 1- ?t ) (t-to)
t-to- [1-e
In principle there is no difficulty in finding higher moments.
By integrating No(q = 0) and (22ND/8q2
)q=o with respect to to from -- to t we
tionary source: .
can find xst for a sta-
Translated from Atomnaya Energiya, Vol. 31, No. 2, pp. 154-156, August, 1971. Original article
submitted August 3, .1970.
0 1972 Consultants Bureau, a division of Plenum Publishing Corporation, 227 West 17th Street, New York,
N. Y. 10011. All rights reserved. This article cannot be reproduced for any purpose whatsoever without
permission of the publisher. A copy of this article is available from the publisher for $15.00.
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 Declassified and Approved For Release 2013/02/26: CIA-RDP10-02196R000300090001-2
2
2
xst = .
EEaI1- 1 , 'z -
It is worth noting that the second moment of the exact function and its value in the Pi-approximation
agree if by the P1-approximation we understand the full form of the differential equation obtained. Thus,
for example, in the telegraphic equation it is necessary to take account of the time derivative of the source.
The diffusion equation with the first time derivative gives the familiar relation x2 = 2D(t -to).
Higher moments are obtained differently. Earlier we found the moments of the distribution function
for certain problems in the theory of entrainment of neutrons by a moving medium. In a similar way it is
possible to study the effect of an acceleration of the system on neutron transport.
Bucci [2] noted a possible effect of large accelerations of the system on the critical state of a reac-
tor. Kostritsa and Neimotin [3] considered the effect of a sudden acceleration of the system on the density
of a neutron gas in a medium with a uniform distribution of sources.
Let the acceleration w be along the x axis. Then the term in the kinetic equation taking account of
the effect of inertial forces can be written in the form
wV?N=w? 8v ON + w u (1-?2) ON
a? .
For simplicity let us assume that the scattering is isotropic. Then the kinetic equation for a prob-
lem with a stationary source has the form
+i
ON ON
v? ax +w? a 1 v?2 ? ON +vEN= v2 sS N d?+ S6 (Z 2 v0) 6 (x), (9)
?
where N = N(x, p, v). In spite of the monoenergetic nature of the source v has a spectrum because of the
inertial forces. It is difficult to determine N or even its Fourier transform N in Eq. (9). Let us take Ea
= 1/vT and Is = 1/vT, where T and T are constants, and find the moments of the distribution function
X= 5 dx 1 v2 dv 1 d?xN (x, ?., v) wT?t (10)
5dx J v2 dv J d?N (x, ?, v) i '
T
dx ~ v2dv 1 d?x2N (x, ?, v) 31. 2 '+T
T 1
x2 - lI
5 dx f V2 dv 5 d?N (x, ?, v) = xn=o 1 f U2 1
? 1+T i
V2
2 v0 Pr
_o- 3 i (12)
1 T
Thus the neutron distribution becomes asymmetric with respect to the plane x = 0 when there is an
acceleration. We note that x and x2 can be obtained from the solution of Eq. (9) in the Pi-approximation.
The value of x agrees with Eq. (10), and x2 differs from (11) by a small quantity. The drift velocity vdr
can be found from the relation x = vdrT.
The same result can be obtained from the solution of Eq. (9) for monoenergetic sources uniformly
distributed in an infinite medium if N is written in the form
N= ~\1 fn (v) I'n (?)
n=0
and we limit ourselves to the Pi-approximation in considering the infinite system of differential equations
k+1 dfk+i k dfk-i w (k+1)(k+2) w k(k-1)
2k+3 du +w 2k-1 do + n 2k-f-3 fk+i- v 2k-1 fk-i
+vEfk= CvE.,fo+2 22 6(v-v0), (2k+1)6k0? (14)
Having determined the neutron density and current we obtain the expression
vdr=
~\1+T1
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 Declassified and Approved For Release 2013/02/26: CIA-RDP10-02196R000300090001-2
where t takes on various values for different types of dependence of the cross sections on the neutron ve-
locity:
1 for E 1 1
s and Ea
o o
3 for Es=const, E?const; (16)
o 2.4 for E8= const, la 1
V
Let us consider how an acceleration might manifest itself in actual neutron experiments. The quantities
vdr, x, and the corrections to x2 are small for thermal neutrons and accelerations of the order of the ac-
celeration due to gravity g. For w ? 100 g one might expect changes in the operation of a thermal reactor
[3]. Since vdr ^- wls/vo, where is is the scattering mean free path, the drift velocity may be a few centi-
meters per second for ultracold neutrons* diffusing in a tube a few centimeters in diameter. If the lifetime
of the ultracold neutrons is taken into account x and the correction to x2 will be significant.
The authors thank F. L. Shapiro and E. P. Shabalin for valuable comments..
LITERATURE CITED
1. R. Meghreblian and D. Holmes, Reactor Analysis, McGraw-Hill, New York (1960).
2. P. Bucci, Atomkerenergie, Vol. 11/12, 435 (1966).
3. A. A. Kostritsa and E. I. Neimotin, Izv. AN KazSSR, Ser. Fiz.-Mat., No. 2, 39 (1969).
*The effect of a gravitational field on ultracold neutrons was first studied by F. L. Shapiro and co-workers
[OIYaI Preprint RZ-5392 (1970)].
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 Declassified and Approved For Release 2013/02/26: CIA-RDP10-02196R000300090001-2
SPONTANEOUSLY FISSIONING ISOMERS OF URANIUM,
PLUTONIUM, AND AMERICIUM FROM NEUTRON REACTIONS
Yu. P. Gangrskii, T. Nadi, UDC 539.173.7
I. Vinnai,* and I. Kovach*
At the present time more than 20 spontaneously fissioning isomers from uranium to berkelium are
known. Most of these isomers have lifetimes in the nanosecond range and can be studied by the time-of-
flight method [1]. We use this method to investigate spontaneously fissioning isomers formed in fast neu-
tron reactions. A schematic diagram of the experimental arrangement is shown in Fig. 1. A beam of
neutrons falls on a target surrounded by a mica-muscovite dielectric ring detector. The target is 1 cm in
diameter and the outside diameter of the detector is 6 cm. The relative positions of the target and mica
are such that prompt fission fragments cannot strike the mica. Fragments are recorded only when recoil
nuclei undergo spontaneous fission more than 1 mm from the target. From the angle and the coordinate of
the track left in the mica by the fragment it is possible to determine the distance the recoil nucleus trav-
eled before decay. This distance determines the lifetime of the fissioning nucleus since the velocity of the
recoil nucleus is known from the kinematics of the reaction.
This procedure gets rid of the background of fragments arising from the spontaneous fission of the
target material or from fissions induced by thermal neutrons after the beam of the bombarding particles is
shut off. Therefore the targets can be made of isotopes with a short spontaneous fission half-life (even iso-
topes of plutonium and curium) or of materials having a large cross section for fission by thermal neutrons
(odd isotopes of uranium and plutonium). At the same time the use of neutrons to obtain spontaneously fis-
sioning isomers imposes special requirements on the cleanliness of the target surfaces since the momen-
tum of a recoil nucleus is small and any contamination of the surface significantly reduces the reaction
yield. In addition the dielectric detectors for recording fragments must have a low uranium and thorium
*Members of the staff of the Central Institute of Physical Studies, Budapest.
units
4
It 5
Fig. 2
Fig. 1. Schematic diagram of experimental arrangement. 1) T or Be target; 2) alu-
minum foil; 3) thorium, uranium, plutonium, or americium target; 4) mica ring de-
tector; 5) vacuum chamber.
Fig. 2. Calculated spectrum of neutrons from the irradiation of a thick beryllium tar-
get by 3 MeV deuterons.
Translated from Atomnaya Energiya, Vol. 31, No. 2, pp. 156-157, August, 1971. Original article
submitted February 16, 1971.
C 1972 Consultants Bureau, a division of Plenum Publishing Corporation, 227 West 17th Street, New York,
N. Y. 10011. All rights reserved. This article cannot be reproduced for any purpose whatsoever without
permission of the publisher. A copy of this article is available from the publisher for $15.00.
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 Declassified and Approved For Release 2013/02/26: CIA-RDP10-02196R000300090001-2
TABLE 1. Cross Sections for the Forma- content since these detectors are irradiated by practically
tion of Spontaneously Fissioning Isomers the same neutron flux as the target.
Reaction
vi O
cm2
Q Xlo-24 a'
9cm2 ag 1Q-4
Th232(n, 2n)Th231
c0,09
1,56?0,16
c0,06
U235(n, 2n)U234