SOVIET ATOMIC ENERGY NOL. 40, NO. 6

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• Document Page Count: 71
• Document Creation Date: December 23, 2016
• Document Release Date: April 10, 2013
• Sequence Number: 6
• Publication Date: December 1, 1976
• Content Type: REPORT

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Russian Original Vol. 40, No. 6, June, 1976
December, 1976
SATEAZ 40(6) 539-604 (1976)
rc
SOVIET
ATOMIC
ENERGY
ATOMHAH 3HENTIA
(ATOMNAYA iNERGIYA)
TRANSLATED FROM RUSSIAN
CONSULTANTS BUREAU, NEW YORK
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? SOVIET
ATOMIC
ENERGY
?
Soviet Atomie, Energy is abstracted or in-
dexed in Applied Mechanics Reviews,' Chem-
ical Abstracts,, Engineering Inde?c, INSPEC-7 '
Physles Abstracts 'and' Electrical and Elec-
tronics Abstracts, Current Contents, and
Nuclear Science Abstracts.
/ -
Soviet Atomic Energysis a'cover-to-cover translation of Atomnaya
Energiya, a,publicatiop of the Academy of Sciences of the USSR.
An agreement with the Copyrighit Agency of the USSR (VAAP)
? friakes,available.both- advance copies of the Russian journal and,
original glossy photographs and artwork. Tys serves to decrease
the necessary time lag between/publication -of the original and
publication of the trislation and helps to improve the quality'
' of the latter. The translation began 'with the first issUe of the
Russian journal.
Editorial Board of,Atomnaya energiya:
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Editor: M. D. Millionshchikov
Deputy Director ,
' I. V. Kurchatov Institute of Atomic Energy
'Academy of Sciences of the pssR
Moscow, USSR
'Associate Editor:' N. A. Vlasov
A. A. Bochvar
N. A. Dolle#al'
V. S. Fursov
I. N. Golcivin
! V. F..Kalihin
A. K. Krasin
Matveev
M. G. ?Meshcheryakov
V. B. Shevchenko
V.1. Sm(rnoV
A. P. Zefirov
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SOVIET ATOMIC ENERGY
A translation of Atomnaya Energiya
December, 1976
Volume 40, Number 6
June, 1976
CONTENTS
Engl./Russ.
The Kiev 240-cm Isochronous Cyclotron - A. F. Linev
539
451
Investigation of the Random Component of the Heat-Release Distribution in a Nuclear
Reactor - V. A. Karpov, V. G. Nazaryan, and V. V. Postnikov
546
456
Spectra of Fast Neutrons from a Pulsed Reactor - G. G. Doroshenko, S. N. Kraitor,
T. V. Kuznetsova, K. K. Kushnereva, E. S. Leonov, and G. A. Frolova
550
460
Numerical Investigation of the Optimum Conditions for the Power Reduction of a
Reactor - V. M. Desyatov, V. I. Pavlov, and-V. D. Simonov
555
464
Development of an Apparatus for Clarifying Solutions Prior to the Extraction
Reprocessing of VVER Fuel Elements - A. M. Rozen, K. A. Dolgova,
A. M. Nudel', I. M. Balakin, I. M. MaPtsev, V. I. Koblov, A. N. Levishchev,
and B. R. Borisov
558
Theory of a Mass-Diffusion Separative Unit- V. A. Chuzhinov, V. A. Kaminskii,
B. I. Nikolaev, 0. G. Sarishvili, G. A. Sulaberidze, and A. A. Tubin. . .
563
471
DE POSITED ARTICLES
Induced Activity of Building and Structural Materials in the 680-MeV Synchrocyclotron
Hall - V. F. Kas'yanov, M. M. Komochkov, Yu. G. Teterev, and V. V. Mal'kov . . .
570
478
Calculations of Some Characteristics of the y -Radiation Field Induced in Air by Fast
Neutrons - A. V. Zhemerev, Yu. A. Medvedev, and B. M. Stepanov
571
479
Electrochemical Behavior of Metals in the Radiation Field of a Nuclear Reactor
- G. Z. Gochaliev and S. I. Borisova
571
479
Production and Study of Corrosion Resistance in Zirconium Diboride and its Solid
Solutions with Titanium Diboride - V. V. Svistunov, A. R. Beketov, V. G. Vlasov,
and N. V. Obabkov
572
480
Thermalization of Neutrons in Solids - V. A. Baikulov
573
480
LETTERS
Using Diamond Detectors as Immersed a Counters - S. F. Kozlov, E. A. Konorova,
M. I. Krapivin, V. A. Nadein, and V. G. Yudina
574
482
Sensitivity of Emission Detectors toy Rays - G. V. Kulakov and B. V. Mukhachev
576
483
Migration of Radiogenic Lead in the Hydrothermal Metamorphism of Uranium Minerals
- V. M. Ershov
579
485
Quantitative Analysis of Various Factors Affecting the Intensity of the X-Ray Signal
Backscattered from a Semiinfinite Reflector - F. L. Gerchikov
581
487
Liberation of Helium in the Uniform Heating of Neutron-Irradiated OKh16N15M3B
Steel - N. P. Agapova, I. N. Afrikanov, A. I. Dashkovskii, A. G. Zaluzhnyi,
V. D. Onufriev, D. M. Skorov, Yu. N. Sokurskii, and 0. M. Storozhuk
585
490
Search for Fissile Isomers in the (n, 2n) Reaction - J. S. Browne R. E. Houve
587
491
Optimal Arrangement of Effective Absorber in a ReactOr -.A. M. Pavlovichev
and A. P. Rudik
589
493
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CONTENTS
Evaluation of Dose Rate from Radiation Heating of a Sample during Irradiation
(continued)
Engl./Russ.
? A. P. Balashov and A. M. Mamontov
591
494
The Crystal Structure of the Compounds Pu5Rh4 and Pu5Ir4 ? A. V. Beznosikova,
N. T. Chebotarev, A. S. Luk'yanov, M. P. Shapovalov, and L. F. Timofeeva
594
495
NEWS ITEMS FROM THE COUNCIL FOR MUTUAL ECONOMIC AID (CEMA)
Diary of Collaboration
597
499
INFORMATION
The I. V. Kurchatov Gold Medal Competition
598
500
Still No "Cosmion" ?N. A. Vlasov
599
500
Conference of the Fourth Committee of the International Commission on Radiological
Protection (ICRP) ? A. A. Moiseev
600
501
The Beta-Mikrometr-3 Double-Layer Coating Radioisotope Thickness Gauge
? I. I. Kreindlin, V. S. Novikov, A. A. Pravikov, and I. R. Rubashevskii
601
501
BOOK REVIEWS
A. M. Petroslyants. Nuclear Power Generation ? Reviewed by Yu. I. Koryakin .
? ?
602
503
The Russian press date (podpisano k pechati) of this issue was 5/21/1976.
Publication therefore did not occur prior to this date, but must be assumed
to have taken place reasonably soon thereafter.
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ARTICLES
THE KIEV 240-cm ISOCHRONOUS CYCLOTRON
A. F. Linev UDC 621.584.633.5
The development of the isochronous cyclotron project was started at the D. V. Efremov Scientific-Re-
search Institute of Electrophysical Equipment (NIIE FA) at the beginning of the 1960's. Approximately at the
same time, the basic concepts of isochronous cyclotrons were formulated: The theoretical principles of the
dynamics of acceleration processes and the system for withdrawal and monochromatization of external beams
were worked out, and possible causes for the loss of beam intensity, resonance phenomena and orbital selec-
tion were investigated.
The planning assignment provided for the construction of a versatile accelerator installation, satisfying
most completely the problems of the present-day nuclear physics of average energies. It was proposed that
the facility constructed should accelerate a large collection of particles with a wide range of maximum energy,
high monochromaticity and quality of the beam. Later, directly in the course of planning, a neutron spectro-
meter system was introduced additionally, which considerably complicated the design of the accelerator vacuum
chamber.
The important and most critical part of the project is the choice of form of the magnetic field and its
method of formation to ensure stability of the acceleration process. Initially, a three-sector design with a
controlled spirality was chosen. In order to regulate the spirality angle, the sectors consisted of two indepen-
dent parts, and the trough winding was connected in order to correct the magnitude of the flutter both for low
and high energy of the accelerated particles. However, as additional modeling of the magnetic field showed, the
trough winding and the nonuniformity of the sectors did not provide satisfactory quality of the magnetic field.
The new magnetic design does not require the trough windings to be connected during acceleration of protons
up to an energy of 80 MeV, and the shape of the sectors has been simplified considerably for the purpose of
improving the manufacturing technology (Fig. 1).
In order to increase the quality of the beam, an acceleration system with constant orbit geometry was
introduced, which required a current stability in the primary winding of 2 -10-5 and 10-4 in the correcting
windings. Stabilization and control circuits by currents in micromodule elements and thyristors have been
developed and manufactured, which have proved to be so compact that they have been arranged in a single stan-
dard rack of the Vishnya type.
The collection of plants and the various facilities permits one of three principal operating regimes of
the accelerator to be achieved: spectrometric, neutron-pulse, and heavy-ion acceleration regime after the
electrostatic recharging accelerator Ell. Each of these cycles requires the installation in the vacuum chamber
Fig. 1. Polar head with sector cover plates.
Translated from Atomnaya Energiya, Vol. 40, No. 6, pp. 451-456, June, 1976. Original article submitted
January 29, 1976.
rfThis material is protected by copyright registered in the name of Plenum Publishing Corporation, 227 West I 7th Street, New York, N.Y. 10011. No part
o this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying,
microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is availablc from the publisher for $7.50.
539
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TABLE 1. Principal Characteristics of Ac-
celerator Operating Cycles and Measurement-
Computational Systems
Energy of accelerated ions, MeV:
protons
deuterons
helium-3
heavy ions
Max. current, mA:
protons, deuterons, helium
heavy ions (by particles)
Beam emittance, mm ? mrad:
horizontal
vertical
Min. size of beam cross section, mm
Energy spread:
at the accelerator outlet
after the monochromator
Monochromator:
angular spread, deg
energy resolution
Neutron spectrometer cycle:
max. pulse duration, nsec
freq. repetition range, kHz
length of flight base, m
pulsed neutron flux, sec-1
ay. neutron flux, sec-1
Time structure of beam:
duration of cluster pulse, nsec
freq. of repetition of bunch, MHz
handling freq., Hz
handling pulse duration, msec
Beam extraction system:
Electrostatic deflector:
angular spread, deg
potential, kV
Current channel:
angular spread, deg
gradient, G/cm
Magnetic shield:
length, cm
gradient, G/ cm
Exptl. compartments:
working area, m2
No. of compartments
No. of beams
length of communication lines from
measurement center, m
10-80
20-70
20-180
140
50
10
20
20
8
2-10-3
1.10-4
270
1. 10-4
2.0
1-100
180
3.1018
1014
2-10
7-22
25-300
0.5-32
30
90
38
200
25
350
1200
6
16
50-200
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TABLE 1 (continued)
Measurement-computational center:
working area of computational
center, m2
working area of meas. center, m2
No. of Vishnya-type racks
No. of standard units
No. of computers:
M-600 ("Parameter")
AI-4096
M-4030
BESM-4
M-400
400
2x100
22
500
1
3
1
1
2
TABLE 2. Principal Parameters of Accelerator and Engineer-
ing Systems
Magnet
Max. induction, G
Diam. of pole pieces, cm
Final acceleration radius, cm
No. of sectors
Gap between poles, cm:
at humps
in troughs
Spirality near final radius, deg
Max. depth of variation, %
with trough winding
without trough winding
No. of pairs of concentric windings
No. of pairs of harmonic windings
Max. input power, MW:
primary winding
correcting windings
High freq. system
No. of dees
Aperture of dee, cm
Generator power, kW
Range of retuning, MHz
Freq. stability
Max. voltage on dee, kV
Stability of voltage on dee
Assumed time of retuning of operating cycles:
change of energy, type of particles, h
transfer of beam to another position, h
Correcting magnet:
length, cm
aperture, cm2
gradient, G/ cm
Beam transport system
SP-018 rotary magnets:
max. induction, kG
gap between poles, cm;
angle of rotation, deg
No. of magnets
SP-106 commutating magnets:
17.0
240
103
3
23.7
53.7
45
0.45
0.39
15
3 x3
0.9
0.8
1
5
450
7.5-22
2 .10-8
125
10-3
3.0
1.0
65.0
5x23
300
11.0
12.0
45
14
11.0
541
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TABLE 2 (continued)
max. induction, kG
gap between pciles, cm
No. of magnets
Quadrupolar magnetic lenses 15K-50A-1:
max. gradient, G/cm
aperture, cm
length, cm
No. of lenses
Quadrupolar magnetic lenses 20K-50A-1:
max. gradient, G/cm
aperture, cm
length, cm
No, of lenses
Sectional magnetic lenses 12-40Sh40:
max. field, G
aperture, cm2
length, cm
No, of lenses
Structure and engineering plant of accelerator
Working area, m2
Working volume, m3
Area of main hall, m2
Wall thickness of biological shield, cm
Installed power, MW
Consumable power, MW
Capacity of ventilation center, M3/h
Multiplicity of air exchange, min-1
Water consumption per day, in3
Circulating water:
external circuit, ni3/h
internal circuit (distillate), m3/h
12.0
8
500
1.5.0.
50.0
34
700
20
50.0
11
830
10 x35.4
40.0
4
12,000
126,000
1000
450
20
12.8
230,000
8
3900
1200
235
of the accelerator of a specific plant. A change of the operating cycle is associated with a considerable time
expenditure, especially as the disassembled plant becomes quite highly radioactive. -
Complication of the problems solved by means of the accelerator has taken place gradually, according
to the origination of new ideas, but the design of the vacuum chamber has remained as before and therefore it
has been found impossible to rearrange the equipment for all cycles of operation under these conditions. In the
future, if the versatility of the accelerator is to be justified, obviously a special chamber should be designed
in which the collection of equipment can be varied without disassembly. The principal parameters of the ac-
celerator are given in Tables 1 and 2.
In the spectrometric cycle, guaranteeing the production of a high-quality extracted beam, acceleration
is accomplished with a constant orbit geometry [2]. When changing the energy and, correspondingly, the mag-
netic-field, the amplitude of the accelerating voltage is changed, so that the number of revolutions for achiev-
ing the final energy remains unchanged. Seperation of the orbits and improvement of the quality of the extracted
beam are achieved because of internal selection and increase of stability of the current in the windings which
shape the magnetic field. A proton and deuteron energy of 80 MeV and 50 MeV, respectively, are achieved with-
out the use of the trough winding, which considerably improves the structure of the magnetic field and ensures
the production of betatron oscillations with minimum amplitudes.
The beam is extracted from the cyclotron by means of a short 30? deflector, the current channel, the magnetic
shield and the correcting magnet [3]. A special system compensates the effect of the current channel and the
magnetic shield on the cyclotron magnetic field.
The position of the deflector, the magnetic channel and the magnetic shield is remotely adjusted for
optimization of the coupling with the trajectories of the particles extracted from the accelerator. When chang-
ing the energy of the particles, because of the constant geometry of the orbits, it is not necessary to change
the position of the extraction elements. Only some correction of their position for optimization of the extrac-
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Fig. 2
Fig. 3
Fig. 2. Neutron pulse shaping system: 1) dee; 2) puller; 3) ion source; 4) internal deflector;
5) diaphragm; 6) stop frame; 7) external deflector; 8) target; 9) neutron beam; 10) orbits.
Fig. 3. Beam separation system: 1) cyclotron building; 2) cyclotron; 3) spectrometer; 4)mono-
chromator; 5) deflecting door; 6) neutron guide; 7) heavy-ion transmission line lens; 8) direction
of heavy-ion beam; 9) direction of neutron beam; 10) deflection magnet; 11) path length of neutron
spectrometer; 12) commutative magnet; 13) electrostatic recharging accelerator; 14) analyzer;
15) recharging accelerator building.
tion conditions is necessary.
The initial section of the ion conductor of the beam separation system has an aperture of 20 cm, and
after the analyzer on the rectilinear section it is reduced to 15 cm. The separation system of the corridor-
type standardizes the beam parameters at the various experimental positions. This makes it possible to dis-
pense with specialization of the compartments for specific experiments.
The separation system directs the beam in advance, analyzed with a resolution of 10-4, to any of 16 tar-
gets. In addition, the deflection of the beam from the corridor into the measurement compartments compen-
sates its nonmonochromaticity. Dispersionless deflection of the beam is effected by a system consisting of
two magnets (450) and lenses.
Instruments and scanning devices for monitoring the beam parameters, and also collimators which shape
its requisite quality and reduce the radiation background in the measurement compartments, are arranged in
the transmission line. A system is used for regulating the scattering angles of the particles within the limits
of 50? without changing their direction [4]. Secondary particles, including also particles polarized by the scatter,
can be directed into any measurement compartment of the beam separation system. The analyzing magnet
reduces the energy spread to 1.0.10-4. At the inlet and outlet of the analyzer, hexagonal lenses are installed
for compensating aberration of the second order. A paired analyzer for secondary particles is located in one
of the measurement compartments, consisting of a precision spectrometer and a wide-range spectrograph with
a resolution of 10- 3.
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1 2 3 4 5 6 7 8 910 20 f, MHz
I pi
/I I I I I III I 1 1111111
3 4 6 8 10 20 30 40 60 80 100 .
E, MeV /nucleon
Fig. 4
100
80
60
40
a) 20
10
8
6
4
10
Fig. 5
100
Fig. 4. Energy of ions inthe cyclotron: 1) 40Ar8+; 2) 20Ne5+; 3) 14N4+; 12c4+; 5) 4He2+ ? 20.1-;
6) 311e2+; 7) 411+.
Fig. 5. Energy of accelerated heavy ions in the cyclotron (1) and in the TATsIT system (2):
- - -) Coulomb barrier at a uranium target.
A
In the neutron-pulsed cycle of operation, protons are accelerated to an energy of 100 MeV. In order to
ensure the necessary focussing at this energy, the trough coils are switched in to increase the amplitude of the
fundamental harmonic of the magnetic field up to the required magnitude.
The internal deflector shapes the pulse and consists of 25 - 100 microbunches. Subsequently, these
bunches are diverted in the axial direction by the external deflector, which occupies the region along the radius
from 86 to 100 cm, and are incident on a thick target, cooled by water, where a pulsed neutron flux is generated
with an intensity of 3.1018 sec-1 (Fig. 2). The pulse duration is 2 nsec at a pulse frequency of 1 - 100 kHz.
The beam of neutrons is directed into three neutron guides, pumped down to a pressure of 10-3mm Hg
in order to reduce losses during passage of the neutrons. The neutron guides are arranged in a special gallery
with three measurement pavilions, which serve simultaneously the compartment for conducting heavy ions.from
the electrostatic recharging accelerator to the cyclotron (Fig. 3). The central neutron guide has a length of
180 m.
The heavy-ion acceleration cycle, with a maximum central field of B = 16.7 kG and a final radius of 103
cm, enables particle energies of E = 140 Z2/A to be obtained, where Z and A are the charge and mass, respec-
tively. The radiofrequency system is tuned over the range of wavelengths from 13 to 43 m, which provides
acceleration for ions with a specific charge in the range Z/A = 0.2 to 1.0. Thus, by using multiple harmonics
q = 1, 3, and 5 of the accelerating field, the facility provides acceleration up to an energy in excess of the
Coulomb barrier at a uranium target for all heavy ions up to 40Ar. Heavier particles will have a lower energy
(Fig. 4).
In connection with the proposed design of the electrostatic recharging accelerator at an energy of 20 MeV,
the possibility was studied also of using the isochronous cyclotron for preacceleration of superheavy ions after
stripping? the TATsIT system [5]. In this system, the ions are accelerated first of all up to an energy of about
1 MeV/nucleon in the recharging accelerator, and then they are injected into the cyclotron using stripping
and they are additionally accelerated (Fig. 5) Curve 1 in Fig. 5 is the energy of heavy ions, accelerated in
the cyclotron with an intensity of 3.1014 sec-1, an energy resolution of 2 ? 10-3 and an emittance of 20 mm ? mrad;
curve 2 is in the TATsIT system at an intensity of 1012 sec-1 and with the same energy resolution and emittance.
Calculations were carried out of the limiting permissible energy for particles of different mass with
different preliminary acceleration conditions (position of source, conductor potential, intermediate stripping,
etc.). The intensity losses due to scattering by residual gas, the nonconformity of the time structures and the
phase volumes of the beam were also evaluated. The results of the calculations showed the feasibility of accel-
erating all particles, including ions of 200Hg up to an energy in excess of the Coulomb barrier in a uranium
target, with a beam intensity of ? 1012 sec-1. The heavy ions, at the outlet from this system, after passing
through the monochromator will possess spectrometric characteristics for carrying out qualitative experiments.
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Fig. 6. Accelerator building with laboratory structure.
In order to obtain ions of light particles, accelerated in the spectrometric and neutron-pulsed cycles, an
axial source will be used with a controllable position relative to the center of the accelrator. In the case of
acceleration of heavy ions, a radial source of the Penning type will be used, similar to that installed in the
300-cm cyclotron in Dubna.
At the present time, construction and commissioning of the cyclotron building with auxiliary units and
facilities has been completed (Fig. 6); plant in addition to the monochromator, which would be ready at the end
of 1976,has been installed and aligned. Tests of the vacuum and high-frequency systems, the cooling system,
and the power supplies have been carried out. The investigation of the magnetic field of the accelerator has
been completed. Assembly and alignment of the sectors have been effected with an error of up to 0.2 mm. The
bending of the I beam of the electromagnet with a maximum field does not exceed 0.4 mm, and when the load
is removed, the initial magnitude of the gap is restored. The magnetic field has been measured with a semi-
automatic equipment with an error of better than 1 G and a speed of response of 1.5 sec per point. In all, about
150 charts of the field and up to 6700 points on each have been measured. The results were processed on a
BESM-4 computer by a special program. The deviation of the corrected magnetic field from isochronous in
the working acceleration region does not exceed 10 G.
The Kiev isochronous cyclotron was started up on March 19, 1976. An internal beam of neutrons on the
final radius of acceleration was obtained with an energy of 50 MeV and an intensity of 10 pA. The beam in-
tensity was practically unchanged with change of radius. Protons were accelerated in the pulsed cycle with
a pulse frequency of 50 Hz, a pulse duration of 4 msec and an amplitude of the accelerating voltage of 70 kV.
The path length of the neutron spectrometer and the electrostatic recharging accelerator are being de-
signed. A polarized particle source with an axial injection system is being developed. The feasibility of auto-
matic control of the isochronous cyclotron using a computer is being studied also.
Extraction of the beam and its transport to the measurement compartments is expected at the end of
1976. The acceleration of heavy ions and the startup of the neutron-pulsed cycle will be effected somewhat
later, after acquiring operating experience with light charged particles.
LITERATURE CITED
1_ V. Kolotyi et al., in: Proceedings of the 6th International Cyclotron Conference, Vancouver (1972), p. 87.
2. A. F. Linev et al., in: Proceedings of the 4th All-Union Conference on Accelerators, Moscow, Session 5,
Report 5 (1974).
3. V. Belyakov, R. Litunovskii, and 0. Minyaev, in: Proceedings of the 7th International Cyclotron Confer-
ence, Zurich, Session D (1975).
4. Yu. G. Basargin, N. I. Zaika, and A. M. Yasnogorodskii, Izv. Akad. Nauk SSSR, 32, No. 2, 384 (1968).
5. A. F. Linev, Preprint KIYaI-73-9Ya [in Russian], Kiev (1973).
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INVESTIGATION OF THE RANDOM COMPONENT
OF THE HEAT-RELEASE DISTRIBUTION IN A NUCLEAR
REACTOR
V. A. Karpov, V. G. Nazaryan, UDC 621.039.512.45
and V. V. Postnikov
Increase of accuracy and identification of the sources of error in neutron-physical calculations of the
heat-release distribution (HRD) in nuclear reactors is of considerable interest for the planning and operation
of high-energy reactor facilities.
The discrepancy between the calculated and actual values of HRD [W(r) and W(r)] may by due to the
methodical inadequacy of. the numerical model and the actual unit, and the difference of the actual structural
and physical characteristics of the reactor elements from the characteristics used in the neutron-physical
calculations, which may be caused by inaccuracy in achieving the specified dimensions and nuclear concentra-
tions in the reactor elements, and also by errors in measurements and control of the various operating para-
meters: the positions of the control and safety rods, the power level, the fuel temperature and the coolant
temperature, etc.
The discrepancy W(r) = W(r)?W(r), due to the combined effect of these factors will be considered in the
future as the random component of HRD (RCHRD) in contrast to the determined component W(r).
The results of a neutron-physical calculation of HRD at present are most frequently used in the station
computers for monitoring the energy release field together with the readings of the intrareactor sensors, dis-
posed discretely in the active zone [1, 2].
In solving the problems arising during planning and operation of a reactor, first and foremost problems
of the discrete monitoring and control of HRD, it is essential to know the space?time characteristics of RCHRD
? the harmonic composition and correlation moments [1].
These characteristics may prove useful for the analysis of the causes of HRD misalignments, observed
sometimes during nominally symmetrical loading of the active zone of the reactor, and when estimating the
necessity for complication of the numerical model of the reactor for the purpose of increasing the accuracy
of the calculations, etc.
Previously, the space characteristics of the RCHRD usually were determined [1] by comparison of the
calculated and measured values of HRD on the reactor, which made difficult the solution of certain problems
in the reactor planning stages. In this case, the RCHRD included components due both to errors of the initial
calculated characteristics and to the inadequacy of the numerical model of the unit.
This method of determining the RCHRD characteristics required a large volume of laborious experi-
mental investigation.
In this present paper, a method is considered for determining the RCHRD space characteristics, based
on a statistical experiment. The results are given, obtained for the RBMK uranium?graphite reactor. In
order to study the RCHRD, the PINK program was used, which was developed for the BESM-6 computer, in
which a two-dimensional few-group diffusion approximation was used. The program calculates HRD in the
reactor, on the basis of an iteration solution of a system of finite-difference equations for a square reference
grid [3]. The applied algorithm permits the neutron flux and heat release in every fuel channel (FC) to be cal-
culated, taking into account the individual neutron-physical properties of each cell of the reactor.
Experiments on the uranium?graphite reactors of the Beloyarsk nuclear power station have enabled a
Translated from Atomnaya Energiya, Vol. 40, No. 6, pp. 456-460, June 1976. Original article submitted
June 23, 1975.
546
This material is protected by copyright registered-in the name of Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part
of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying,
microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $7.50.
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TABLE 1. Results of Calculations of the RCHRD Characteristics
of the Multiplication Factor of the RBMK Reactor
and the Random Component
Taking account of the characteristics,
RCHRD
mean-
Max. ampli
tude of HR
Variations for a single
"drawing"
subjected to "drawing"
square
deviation,
variation,
reactivy,%
AK/K
coeff. of varia-
tion of HRD, Kr
- K'
Spread of uranium dioxide charge
in FC's
1
3
0.0035
0,005
Spread in fuel enrichment of FC's
16
0.0170
0.060
Simultaneous and independent spread
in the uranium dioxide charge and the
fuel enrichment of FC's
16
0.0170
0,060
Simultaneous and independent spread in
the uranium dioxide charge in FC's,
in the fuel enrichment, and in the
absorbing properties of the control
rods and auxiliary absorbers
7
19
0.0300
0.080
method for computing the control rods located in intermediate positions to be developed, according to which
each partially inserted rod is replaced in the calculations by a completely inserted rod, equal in efficiency.
The PINK program has been used widely in the planning stages during the forming of the initial charge,
and also during operation of the RBMK reactor of the V. I. Lenin nuclear power station at Leningrad [4].
In the determination and analysis of the RCHRD, first of all a calculation was carried out of HRD by the
PINK program for some actual reactor charge [5]. Then, for one or independently for several parameters of
each cell of the active zone, the random deviations from the law of probability density were-calculated by the
Monte Carlo method, corresponding to a field of tolerances of the parameter and assumed to be uniform accord-
ing to the data of the design specifications. Subsequently, calculation of the HRD was carried out by the PINK
program for a reactor state which differs from the original state by the random deviations of the parameters.
Comparison of the first (initial) value of HRD [W(r)] and the second (perturbed) value of HRD [W(r)] enables
the RCHRD [W(r)Ito be obtained, which is the homogeneous centered random distribution. On the basis of the
RCHRD obtained, the space correlation moments Kir(r) were calculated, which are dependent on the distance
between two points of the active zone.
An analysis of the experimentally obtained values of RCHRD for the uranium-graphite reactors of the
Beloyarsk nuclear power station, carried out previously, showed that the RCHRD within the limits of practical
requirements [1] possess the property of ergodicity, and therefore its correlation function can be calculated
on the basis of a single application.
Analysis of the calculated RCHRD characteristics showed that the principal contribution to it for the
RBMK reactor is made by spreads in the enrichment, in the 235U charge of the fuel channels, in the absorbing
properties of the control rods and auxiliary absorbers (AA). Therefore, in future calculations random devia-
tions only of these parameters were introduced. As deviations in enrichment and the fuel channel charge have
almost no effect on the moderation cross section and the diffusion coefficient, these deviations were taken
into account only by means of changes of the macroscopic neutron fission and absorption cross setions.
The results of the RCHRD calculations and of the random component of the multiplication factor of the
RBMK reactor are shown in Table 1 for different versions of "drawing" of the characteristics of the active-
zone charge elements of the reactor.
The mean-square deviations of the RCHRD (o-*), shown in Table 1, are due only to the nonidenticity of
the neutron-physical characterstics assumed in the calculations and their actual values and therefore are almost
independent of the choice of methods for the neutron-physical calculation. The value of air = 7% agrees within
the limits of statistical accuracy with the similar quantity (5-6%) obtained experimentally during the physical
and power startup of the RBMK reactor.
Figure 1 shows the RCHRD for two mutually perpendicular directions through the center of the RBMK
core, obtained in the calculation of the charge for the physical startup. Figure 2 shows schematically the
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W(FYiti17)
;08
;06
1,02
1,00
;98
;96
;94
492
0,90
1
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1,04
1,02
700
0,98
496
0,94
;92
;90
;88
1
3 5
7 9
11
13
15
17
19
21 23 25 27 29 31
Channel No.
33 35 37 39 41 43 45 47 49
3 5 7 9 11 13 15 17 19 21 23 25 27 29 31 33 35 37 39 41 43 45 47 49
Channel No.
Fig. 1. Field distribution of the ratio W(r)A(r) for the central array of channels of the
RBMK core in the (a) latitudinal and (b) longitudinal directions for different spreads in the
charge component properties: , ---, and-) taking account of variations in the amount
of UO2 in the fuel rods, in the 235U fuel enrichment, and in the properties of the fuel rods, the
auxiliary absorbers and the control rods, respectively.
regions of positive and negative values of the RCHRD in the reactor core. Analysis of the results of the cal-
culations for the RBMK reactor showed that the main contribution to the amplitude of the RCHRD are the lowest
radial-azimuthal harmonics. In other words, the random spreads in the channel characteristcs give rise to the
appearance of smoothly varyingHRD misalignments. In this case, the sign of *(r) in each fuel channel of the
active zone is varying in a random manner from calculation to calculation.
The RCHRD correlation functions, used in the mathematical processing of discrete control for the RBMK
reactor, on a computer, and obtained by this method, are plotted in Fig. 3. For comparison, the correlation
function found experimentally by means of a detailed scanning of the active zone is plotted also. It follows
from Fig. 3 that the agreement of the results is completely satisfactory.
Also of interest is the investigation of the relation between the mean-square deviation of the nominal
values of the fuel channel macroscopic cross sections for the RBMK reactor and the mean-square deviations
of the RCHRD. Table 2 shows the values of a for different mean-square variations of the fission and neutron-
absorption cross sections for the fuel channels. Analysis of the data from Tables land 2 permits the conclu-
sion to be drawn that a close to linear dependence is observed of the mean-square deviation of HRD over the
reactor on the mean-square deviations of the cross-sections vf Ef and Ea. The effect of the deviations in the
fission and absorption cross sections on the RCHRD is approximately equal.
The results presented obviously permit the nature to be established of the significant unidentified mis-
alignments of HRD in the case of a symmetrical initial charge for a series of reactors, in particular the re-
actors of the Beloyarsk nuclear power station. These effects were explained previously as the consequence
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Fig. 2. Deformation of the heat release field as a
consequence of the spread in properties of the fuel
rods, the auxiliary absorbers and control rods,
within the limits of the physical bounds of the RBMK
reactor; 1) increase and 2) reduction of power.
TABLE 2. Dependence of Mean-Square Deviations of HRD on Devi-.
ations of the Fission and Absorption Cross Sections of the Fuel
Channels for the RBMK Reactor
Devia. of fission
and absorption
cross section of
FC's from '
nominal, %
Mean-square devia.
-of HRD over reactor
Corresponding variations
breeding properties
of reactor-, % /K
coeff. of variation
of HRD, Kr? IC,
0,25
2,70;
2,52
?0,0055;
0,008
0,011;
0,034
0,50
3,64;
4,44
?0,0023;
0,018
0,047;
0,063
1,00
6,74;
8,24
0,0003;
0,041
0,129;
0,144
2,00
12,66;
15,26
0,0188;
0,101
0,302;
0,307
3,00
17,82;
21,55
0,0552;
0,180
0,552;
0,474
Note. The first figure refers to the fission cross section and the second figure refers
to the absorption cross section.
1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17
Fig. 3. Normalized p(T) correlation functions, obtained for different
variations in the fuel charge (1), in the fuel charge and enrichment (2),
and the absorbing properties of the auxiliary absorbers (3); also the nor-
malized correlation function obtained experimentally (4) [T is the unit
pitch of the lattice].
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of filling up with water or contamination of the graphite stack, uncontrolled repositioning of rods, etc.
The data from the recent work show that the cause of the HRD misalignments might be entirely due to the
variations in the enrichment and charge of the fuel channels, the variations in dimensions of the control
rods, etc.
The method devised for calculating RCHRD and its characteristics, permits, at the planning stage, an
estimate of the accuracy of the calculated HRD, and also the provision of the necessary starting data for the
development of a system for the discrete monitoring of HRD with previously defined characteristics.
LITERATURE CITED
1. I. Ya. Emel'yanov et al., At. Energ., 34, No. 2, 75 (1973).
2. I. Ya. Emel'yanov et al., At. Energ., 29, No. 4, 298 (1970).
3. V. A. Karpov, in: Collection of Reports on the Programs and Methods of Calculating Nuclear Reactors
[in Russian], Izd. CMEA, Dimitrovgrad (1975) , p. 89.
4. I. Ya. Emel'yanov et al., in: Experience in the Operation of Nuclear Power Stations and Paths for the
Further Development of Nuclear Power Generation [in Russion], Vol. 1, Izd. FEL Obninsk (1974), p. 81.
5. I. Ya. Emel'yanov et al., Report on the Conference of Mutual Economic Aid Countries on Methods of
Monitoring and Controlling Nuclear Reactors and Nuclear Power Stations [in Russian], Warsaw (1973).
SPECTRA OF FAST NEUTRONS FROM A PULSED REACTOR
G. G. Doroshenko, S. N. Kraitor,
T. V. Kuznetsova, K. K. Kushnereva,
E. S. Leonov, and G. A. Frolova
UDC 621.039.526
The possibility of deriving the spectra of fast neutrons from the readings of threshold detectors by mini-
mizing the directional spread was considered in [1, 2]. It was shown in an experiment on paper that the method
is characterized by higher accuracy and renders more information than, say, the method of orthonormal ex-
pansion. But as far as we know, to date no neutron-spectrometrical measurements have been made in this
manner. Such measurements have been made in our work, in which the spectra of fast neutrons of a pulsed
HPRR-type reactor [3] were measured.
The core of the HPRR reactor consists of a cylinder with a diameter of 20 cm and a height of 23 cm,
which is made of an alloy of molybdenum (10%) and uranium enriched to 93% with the 235U isotope. The reactor
can be used in stationary or pulsed operation with about 1017 fissions per pulse.
A set of threshold detectors consisting of 237Np(n,f), 238U(n, f), 32S(n, p), "Alin, p), "Al(n, a), and supple-
mented by 235U(n,f) in a filter of 1 g/cm2 10B were used for the measurements. Both the calibration and the
determination of the detector readings were analogs to [4]. The detectors were irradiated in the experimental
hall of the reactor. The reactor core was at a height of 2 m in the center of a hall having the dimensions 11
x 10 x 10 m. The detectors were placed on that height at a distance of 3.3 m from the center of the core. The
measurements were made either without shielding or with shields consisting of 13-cm iron or 12-cm Plexiglas
inserted between the detectors and the core at a distance of 3 m from it.
The differential neutron spectrum yo(E) was derived from the readings Ni of a set of m detectors by
solving a system of integral equations
N1= (E)cri (E) dE,
(1)
where a(E) denotes the cross section of the corresponding reaction; i = 1, 2, 3 . ? m (m = 6). The informa-
Translated from Atomnaya Energiya, Vol. 40, No. 6, pp. 460-464, June, 1976. Original article submitted
July 27, 1975,
550
This material is protected by copyright registered in the name of Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part
of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying,
microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $7.50.
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070006-5
Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070006-5
fif,Nr
1,2
1,1
1,0
47
Np(n, n.
U(n,f)
_ , fl
S 01,0
_ 4
AL (n,p)
_
I
\\
,
.
1
-
!
a
000
_
U(n,f)
S (np)
,
.
b
c
3 (n,p) .
U(n,f)
17,p) .
AI (
i
r
A l(n,p)
?
I
.
1
I
I
E
Np (n, f)
AL (n,a-)
I
10 1001 10
Number of iterations
o
Fig. 1. Dependence of the ratio Ni/Nni of detector readings upon the number of iterations
when the neutron spectra were derived: a) without shielding; b) with 13-cm iron shielding;
and c) with 12-cm Plexiglas shielding.
100 1
10
100
tion measure
/08
45 ;0 2,0 5,0 /0 0,5 1,0 2,0 5,0 10 0,5 1,0 1,0 5,0 MeV
Fig. 2. Spectra of the fast neutrons of a pulsed
HPRR-type reactor: a) without shielding; b) with
13-cm iron shielding; and c) with 12-cm Plexiglas
shielding. Spectra derived after: 0) 3, A) 5, and
0) 25 iterations.
Ni
J IN , ?= Niln
(E) cri (E) dE
(2)
was used as a measure for the difference between the left and right sides of the above equations; the informa-
tion measure denotes the directional spread in aleatoric function space. In order to determine the minimum
of functional (2), an iteration process was developed. The algorithm of the iteration process was assumed in
the form of 12]:
\-1 oi(E) i=1
GLI
i=
(3)
(y9n(E) and con+i(E) denote the neutron spectra of the n and n+1 iteration, respectively; Ni denotes the measured
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10
101?
108
45 0 2,0 40 10 0,5 1,0 40
iii
4
hi
t
4
I
_
_
%
a
4
?
1
X
1 ?
4
i
i
i
b
\
1/4
AI
?1
1
i
?
a
/
/
,"`
a 11.
c
i
1/4
I
t
I I
//
I
i _1'
i
I
i
1
?
x
t %
4
.
1
1
I
i
1
i
40 10 45 1,0 2,0 5,0 MeV
Fig. 3. Spectra of the fast neutrons of a pulsed HPRR-type reactor: a) without shielding;
b) with 13-cm iron shielding; and c) with 12-cm Plexiglas shielding: El and A) minimiz-
ing the directional spread and orthonormal expansion, respectively; 0) calculations for
the HPRR reactor,
detector readings; and Ni denotes the detector readings which were calculated from the spectrum derived
after .n iterations). Calculations were made on a computer with a program written in ALGOL. The time
requii.ed for performing 100 iterations on a Minsk-22 computer amounted to about 2 min. q0(E) = 1 was as-
sumed as the zeroth approximation.
Figure 1 shows the ratio of the detector readings N?i. to Ni. As early as after 10-25 iterations, the detec-
tor readings calculated from the restored spectrum correspond to the experimental values with an error of
less than fractions of a per cent and are practically constant thereafter. Figure 2 proves that the form of the
spectra hardly changes. It follows from the spectra which are shown in Fig. 2 for various numbers of itera-
tions that the iteration process converges rapidly toward a stable solution and that no oscillations occur.
According to [5], this form of convergence indicates that the errors which are made in the determination
of the detector readings and of the energy characteristics used are insignificant.
Figure 3 shows the results of measurements for 25 iterations (the results are similar in the case of 100
iterations) and the calculated data for the HPRR reactor used with the above shieklings [6]. The structure of
the measured spectra results from the superposition of the spectrum of the fission neutrons with a total energy
of about 2 MeV and the spectrum of the inelastically scattered neutrons with the maximum at several hundred
keV. The contribution of the latter neutrons increases behind the iron shielding and dominates. The relative
contribution of the neutrons in that region increases also behind a water-containing shielding. But in that case
the increase is caused by the moderation of the incident neutrons in the shield and by the relative increase in
the contribution of neutrons reflected from the walls of the building.
It follows from a comparison of the calculated and experimental results that they are in good agreement
in the entire energy range from 0.5 to 10 MeV when no sheilding or when iron shielding is employed. The spec-
tra were measured with adequate accuracy even when only the set of 5 detectors was used. The agreement is
observed only above about 2 MeV in the case of Plexiglass shielding, because the neutrons which are scattered
in the building and have a softer spectrum influence the results. These neutrons were disregarded in the cal-
culations, because the HPRR reactor is practically in an open place.
Figure 3 includes the fast-neutron spectra which were derived from the detector readings by the ortho-
normal expansion of [7]. According to that method, the differential neutron spectrum is represented as a
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Eco(E). rel. units
a
0 I I I I I
0,5 1,0 2,0 5,0 10 0,5 1,0 2,0 40 10 0,5 40
I I 1 ? I I I
2,0 5,0 10 0,5 1,0 2,0 5,0 10 45 1,0 2,0 5,0 &MeV
Fig. 4. Spectra of fast fission-induced neutrons:
e) with polyethylene shielding; b) heavy-water reactor, and d) BR-1 reactor. The spectra
were derived with the full set of detectors (0) or without the 235U + 1-g/cm2-1?B detector
(x); -) given spectra.
a) without shielding, c) with
lead shielding,
linear combination of the efficiencies of the detectors used, i.e.,
111
(I) (E) =
h=1
where the coefficients elk are obtained from the condition that the difference between the unknown spectrum
(p(E) and its approximation to all the functions cri(E) must be orthogonal:
We therefore obtain with Eq. (1)
f (E) - I mho!, (E)] ni (E)dE= O.
h= 1
(4)
(1)(E) = E Ni E (E), (5)
1=1 /1=1
where pki = f 0- k(E) (E) dE denotes matrix elements which depend only upon the efficiency of the neutron
detectors.
The neutron spectra obtained in this fashion have the form of waves. In all cases maxima were noted in
the spectra at 1.2 MeV; the maxima do not depend upon the form of the spectrum, i.e., they are an effect
originating from the instruments. The neutron spectrum below 1 MeV is not reproducible by the method of
orthonormal expansion. The results confirm the relative merits of the two methods, which were assessed in
[8] on the basis of calculations. The results are important as far as applications are concerned, because the
results show that the method of minimizing the directional spread for deriving neutron spectra has advantages.
The advantages are essentially explained by the substantial differences between algorithms (3) and (5).
Algorithm (5) represents the spectrum as a linear combination of the efficiencies of detectors; the consequence
is the wave-like structure which originates from the efficiencies rather than from the energy distribution of
the neutrons. In algorithm (3), the dependence of the derived neutron spectrum upon the efficiency is less
pronounced and the wave-like structure of the spectrum is practically not noticeable. This algorithm precludes
the appearance of negative values of the spectrum, which are observed in certain cases when algorithm (5) is
used.
Since equation system (1) is poorly defined and since its solution in the form of Eq. (5) implies the deter-
mination of the inverse matrix (Pk(')' the resulting solution depends very strongly upon the errors of the
detector efficiencies and upon experimental reading errors. The error in the determination of the neutron
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spectrum can many times exceed the above errors. When the method of minimizing the directional spread is
employed, the inverse matrices need not be determined so that the influence of the errors in the initial data
is extremely small.
In addition to the threshold detectors, we used in the present measurements also a detectorAicji:gon7,.
tamed 235U and was surrounded by a boron filter with a thickness 1 g/cm2 10B. This detector records neutrons
with an energy in excess of ? 400 eV and has an efficiency which varies very little with energy. In order to
.determine whether the detector should be employed for measurements of fast neutrons, an initial experiment
was made "on paper" to derive spectra which sharply differ in their forms at energies of 0.5-10 MeV: spectra
of fission neutrons without shielding and with shielding by lead or polyethylene, the neutrons stemming from a
heavy-water reactor or the BR-1 reactor [2]. The spectra were derived by minimizing the directional spread
with algorithm (3) and by using the calculated readings of scaled threshold detectors inclusive (or exclusive)
the detector with 235U + 1 g/cm210B. The results (Fig. 4) show that the 235U + 1-g/ cm2-10B detector substantially
increases the accuracy at energies below 1.5 MeV and facilitates the reliable determination of the spectra with
0.5 MeV.
Obviously, the same effect can be obtained when one uses in place of the 235U + 1-g/cm2-10B detector some
other detector which responds to neutrons with energies below 0.5 MeV, e.g., a detector based on 63Cu (n, y),
6Li(n, a), etc. Though the spectrometric information which is obtained in that case is without interest in the
entire response range of the supplemental detector, the incorporation of it allows a better utilization of the
spectrometric possibilities of the neptunium detector near its energy threshold.
The authors thank I. B. Keirim-Markus for interest in the present work and useful discussions.
LITERATURE CITED
1. M. Z. Tarasko, E. A. Kramer-Ageev, and E. B. Tikhonov, in: Problems of Dosimetry and Radiation
Shielding [in Russian], No. 11, Atomizdat, Moscow (1970), p. 125.
2, G. G. Doroshenko et al., At. Energ., 35, No. 5,343 (1973).
3. I. B. Keirim-Markus, V. A. Knyazev, and S. N. Kraitor, in: Metrology of Neutron Radiation at Reactors
and Accelerators [in Russian], Vol. 2, Izd. VNIIFTRI (All-Union Scientific-Research Institute of Physico-
technical and Ratliotechnical Measurements), Moscow (1974), p. 121.
4. K. K. Koshaeva, S. N. Kraitor, and L. B. Pikel'ner, At. Energ., 32, No, 1, 68 (1972).
5. G. G. Doroshenko, S. N. Kraitor, and E. S. Leonov, in: Metrology of Neutron Radiation at Reactors and
?Accelerators [in Russian], Vol. 2, Izd. VNIIFTRI (All-Union Scientific-ResearchInstitute of Physico-
technical Measurements), Moscow (1974), p. 36.
6. J. Poston, J. Knight, and G. Whitesides, Health Physics, 26, 217 (1973).
7. I. B. Keirim-Markus and V. I. Popov, in: Problems of Dosimetry and Radiation Shielding fin Russian],
No. 10, Atomizdat, Moscow (1969) , p. 14.
8. G. G. Doroshenko et al., in: Metrology of Neutron Radiation at Reactors and Accelerators [in Russian],
Vol. 2, Izd. VNIIFTRI (All-Union Scientific-Research Insitute of Physicotechnical and Radiotechnical
Measurements), Moscow (1974), p. 45.
554
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NIVP-E?R,FCAL INVESTIGATION OF THE OPTIMUM
CONDITIONS FOR THE POWER REDUCTION OF A REACTOR
V. M. Desyatov, V. I. Pavlov UDC 621.039.516
and V. D. Simonov
The problem of the optimum cycle with respect to time for the power reduction of a reactor is considered
in [1, 2]. The optimum power reduction trajectory is shown in Fig. 1, where xi and x2 are the concentrations of
iodine and xenon, respectively, normalized to the fission cross section, and t is the curve of the reactor states,
stable with respect to iodine and xenon. In the general case, when limited to the maximum concentration of
xenon (x2= x2max), the trajectory has a section of minimum power (ab), a section where the power is varying
(bc), a i a section of maximum power (cdt). The section cdi is completed by the curve tangential to x2 = x2max,
and represents the change of the iodine and xenon concentrations during instantaneous reduction of the re-
actor power to the state defined by the point d on the curve
The special features of the individual sections of the optimum trajectory and the optimum time of reduc-
tion of power To depend on such reactor characteristics as the operative reserve of reactivity Ap, the maximum
x(3) and minimum x3 . permissible neutron fluxes, and also the values of the neutron fluxes corresponding to
max mm
the initial x(3) and final x(3) reactor power level and the rate of change of power a.
In this present paper a numerical investigation is given into the development of these relations which
affect significanly the feasibility of the practical achievement of the optimum reactor power reduction cycle.
The calculations were carried out by a specially devised program, which allows To, the time of operation of
the reactor with minimum (tab), variable (tbc), and maximum (tub.) power, to be found in terms of the known
values of Ap, X(g), xy), x(3) ;and x(3) . A reactor is considered, with physical characteristics which are typical
min max
for the water-cooled/water-moderated power reactor (VVER).
Figure 2 represents the dependence of the time of movement on the sections of the optimum trajectory,
on the operative reserve of reactivity, obtained on the assumption of a stepwise power change at the points
a, b, c, and di. When Ap< 1.5%, the main contribution to the total time of power reduction is the time of opera-
tion with variable power tbc. This time decreases rapidly with increase of Ap and when Ip = 1.9%, it becomes
equal to zero, so that over the range of values 1.9%.45_ 2.35%, the optimum transition is achieved by a two-
stage cycle: operation at the minimum and maximum permissible power. When Apa? 2.35, To = 0, i.e., it be-
comes equal to a "direct" transition at a power equal to 0.2 x(3), without a forced standstill in the "iodine well."
42 0,6 0,8 1,0 701 cm-2
Fig. 1. Optimuni power _reduction trajectory
.(xgax = 43); xr = 0.1 )43); xinin = 0).
Translated from Atomnaya Energiya, No. 6, pp. 464-467, June, 1976. Original article submitted July 23,
1975.
This material is protected by copyright registered in the name of Plenum Publishing Corporation, 227 West 17th Street, New York. N.Y. 10011. No part
of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying,
microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $7.50.
555
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Fig. 2
F.
30
25
20
15
\ \
10 \\
0,5 1,0 1,5
Fig. 3
Fig. 2. Dependence of operating time of the VVER-440 reactor at various sections of the optimum
trajectory t on the operative reserve of reactivity Ap [xax
= 43)
; 3)r(
0.243); xgin = 0].
Fig. 3. Dependence of To, Ti.w., and Teff on the operative reserve of reactivity (xflax
x(3). = 0): 1) x(f3)= 0; 2) x(3) = 0.1x(3); 3)x(3)= 0.3 43); 4) x(3) = 0.543); ) To, ? ?) Teff.
min f o f
45 1,0 1,5 2,0 AA%
Fig. 4
30
20
10
1,0
Fig. 5
dp,
Fig. 4. Effect of the reactivity power effect on the magnitude of the optimum
time of power reduction fx(m3)ax = 43); xVin = 0 and 4; ) = 01: ? and - - -)
without and with account of the power effect, respectively.
Fig. 5. Relative increase of optimum time of power reduction E, for a cycle
with a limitation on the rate of change of power as a function of AP (xt)ax =
x (3)* X(3) = 0; a = 10% min).
0 ' min
The optimum reduction cycle of a power reactor has more significance from the point of practical achieve-
ment the lesser To is and the less the generation of energy which accompanies it. For this, it is obvious to
compare the optimum time with the time of stay of the reactor in the "iodine well" Ti .w., and to estimate the
power output in effective hours Teffr For each value of 343) values of Ip can be assigned for which the power
reduction cycle from the level x(3) to the level x(3) is achieved during the time To < Ti.w.. It can be seen in
Fig. 3, where the dependence to To, Ti.w., and Teff on the operative reserve of reactivity for a reactor of the
VVER-440 type are plotted, that, for example, for x(3) = 0, this region is determined by the inequality'Ap > 0.7%,
and for x(3) = 0.iX(3), >0.6%. Taking into account the negative power effect of reactivity, which is character-
istic for VVER reactors, the boundary of this region is displaced to the side of a lower value of Ap (Fig. 4).
556
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TABLE 1, Optimum Time of Power Reduc-
tion for Different Values of the Parameters
which Define the State of the VVER-440
Reactor in the Transition Process, h
.1.
*---='
2. E
H
-
P,,...,
H
(3) /x(3). .(3) (3)t
-Min 0 I Mid'wu
0
0,2
0,4
0
0,2
0,4
0
31,48
31,55
31,66
18,81
18,96
19.25
1,0
0,2
23,74
23,81
23,92
11,68
11,82
12,11
0,4
15,32
15,39
15,50
4,74
4,88
5.17
0
31,59
31,66
31,77
18,95
19,09
19,38
0,8
0,2
23,88
23,95
24,06
11,86
12,00
12,29
0,4
15,53
15,60
15,71
5.02
5,16
5.45
0
31,7831,84
31,96
19,17
19,31
19.60
0,6
0,2
24,15 24,22
24,33
12,19
12,34
12,62
0,4
16,09
16,15
16,27
5,74
5,89
6,18
*4=0,5%.
t,51)=1,0%.
The relations in Figs. 3 and 4 have been obtained on the assumption that the rate of change of power is not re-
stricted and at the points a, b, c, and di the power is changing stepwise.
The problem concerning the optimum power reduction cycle of a reactor was considered in [3] in the
case of a finite rate of change. This restriction is justified and is conditional on the capabilities of the control
system and the safety requirements.
The calculations show that a noticeable increase of the optimum time in comparison with the case of in-
finite rate of change of power is observed, in the case of the actual rate of change of power of power reactors
( a = 1%/min). The results of these calculations are shown in Fig. 5 in the form of the dependence of the rela-
tive increase of To when a = 1%/min on the operative reserve of reactivity and the power at which the reactor
changes over.
As a result, an increase of To leads to the appearance of sections of the trajectory which "smooth" the
power flash-up. Therefore, by reducing the size of the flash-up "step," the solutions of both problems can be
brought together. Moreover, a cycle with small fluctuations of the total power is simpler to use and is more
reliable with respect to safety conditions. The dependence of To on x(3) and x(3) is shown in Table 1. It can
min max
be seen that over a wide range of values of x(Ain and x(n)13 ax the optimum time varies only weakly. Obviously,
for every value of x(3), values of x(3) and x(3) can be calculated, for which the cycle, with an insignificant
min max
increase of To, will be suitable for practical achievement.
LITEREATURE CITED
1. A. P. Rudik, Nuclear Reactors and Pontryagin's Maximum Principle [in Russian], Atomizdat, Moscow
(1971).
2. A. P. Rudik, Xenon Transition Processes in Nuclear Reactors [in Russian], Atomizdat, Moscow (1974).
3. T. S. Zaritskaya and A. P. Rudik, At. Energ., 36, No. 2, 140 (1-974).
557
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DEVELOPMENT OF AN APPARATUS FOR CLARIFYING
SOLUTIONS PRIOR TO THE EXTRACTION REPROCESSING
OF VVER FUEL ELEMENTS
A. M. Rozen, K. A. Dolgova,
A. M. Nuclei', I. M. Balakin,
I. M. Malitsev, V. I. Koblov,
A. N. Levishchev, and B. R. Borisov
UDC 621.039.59.001.57
When reprocessing fuel elements from VVER (water-cooled/water-moderated power reactors) [1] by
extraction technology [2], the problem arises of the clarification of the solutions [2-4], as suspensions can clog
the control-and-measuring instrument and the analyzer, and also can stabilize emulsions, which leads to the
formation of interphase films and disruption of the normal flow of the process in the extractors [5]. Therefore,
it was decided to use filtration or centrifugation [3, 4j.
Preliminary investigations showed that when dissolving VVER fuel elements, highly dispersed, difficultly
separable suspensions arise, the clarification of which is a problem [4]: with settling (even during 24 h,
only 50-60% of the solid phase can be separated), the rate of filtration through a paper filter "blue ribbon" with
a drop in pressure of 0.5 kg/cm2 is negligibly small (curve 1, Fig. 1); with centrifugation, a fugate containing
< 5 mg/liter of solid suspension is obtained only with a separation factor of 20 thousand; in this case, the
decantate, filtrate, and fugate, just like the crude solution, on contact with 25% tributylphosphate (TBP) formed
interphase films. Because of the defined properties of suspensions, for complete separation of the suspen-
sion and in order to prevent the formation of films, it is necessary to prepare solutions ? use flocculants and
4025 0,050 4075 4100 4 ms / m2
Fig. 1. Dependence of the rate of filtration on the volume
of filtrate obtained: 1) actual (unprocessed) solution of
VVER fuel elements; 2) actual solution treated with gelatin;
3) simulated VVER fuel elements, treated with gelatin.
Pressure drop 0.5 kg/cm2.
Translated from Atomnaya Energiya, Vol. 40, No. 6, pp. 467-471, June, 1976. Original article submitted
July 9, 1975.
558
This material is protected by copyright registered in the name of Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part
of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying,
microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $7.50.
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Fig. 2. Diagram of experimental-industrial filter:
1) filtrate assemblage receiver; 2) casing; 3) filtra-
tion element; 4) conical slotted bottom.
subsidiary filtration methods.
In [3, 4], the results are discussed of investigations into the choice of conditions for clarifying solutions
on cartridge filters; these investigations were carried out on laboratory apparatus, using solutions of unirradi-
ated fuel elements; the results have been verified on irradiated solutions.
The present paper is devoted to the development of an industrial clarifying apparatus ? a filter and a
centrifuge. In order to test it, it was necessary to develop special simulant-solutions.
Simulant-Solution for Testing Semiindustrial and Experimental-Industrial Apparatus. In order to test
industrial filters, a large quantity of solution is required and the use of solutions of actual fuel elements would
be far too expensive. It was necessary to devise simulant-solutions.
Solutions of VVER fuel elements contain uranium (300-330 g/liter), 2-3 M HNO3, have a specific gravity
of 1.4-1.5, and a viscosity of 2.2-2.6 cp. The content of suspended matter amounts to up to 1 g/liter, the bulk
of which is finely dispersed graphite; also present are silicic acid and other impurities (including uranium
up to 5%), 80% of the particles have a size of > L'), the flow of the
vapor?gas mixture along the diaphragm in the space II can be expressed by the relation
U(y) =Q?u Py,
and, consequently, the expression for the gas flow in this plane is written as
G' (y)= (Q ? uP y) b (y) ?
(8)
(9)
(10)
The longitudinal variation of gas flow in the space II with Eq. (6) taken into consideration can be repre-
sented in the form
d
(T [kV uP Y) Vb (01 uP Va?qTb
0-1 ?
Integration of this equation under the boundary condition yb (y = 0) = 0 yields
Tb (y) =
a [1? (
u P \ /(4- 1)]
Y )
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(12)
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L.!
1,0
ln
5
t151--
1,5
2,0
3,0
4,0
5 10 q"
Fig. 3. Dependence of light-fraction flow on total
vapor flow for various values of Peclet number.
Considering that gas flow through the partition is much less than vapor flow, one can write
uPy=(1-0?) Q .
The expression for yb is then tranforrned to
(13)
(Y)= Ta {1 _[i ?(I ?fly)l-]h/(")}. (14)
Figure 2 shows the dependence on the longitudinal coordinate of the gas flux density through the partition in the
dimensionless form Tx/ en'a nDio for 9 = 0.5. The quantity Tx/ e/y a nDio is a function of Peclet number and ev
only and does not depend on the geometry of the unit or on the properties of the gas being separated.
The behavior of Tx/ e/yanDio shows that the gas flux density decreases with height and possibly changes
sign for a given relation between ev and in q. This is explained by the fact that reverse convective transport
try begins to predominate over diffusion gas transport through the diaphragm as the gas concentration rises in
the inner space. Having set Tx = 0 in Eq. (6), it is easy to obtain the inversion point for the flux density with
a given operating mode of the unit (given ev and ln q):
(q 110-i
H.
y* ? i?Ov
(15)
A change in the sign of Tx (y) means that a portion of gas which has diffused into the inner space of the unit is
retransported into the condenser space, i.e., the unit operates with an internal gas circulation. Assuming that
the coordinate of the inversion point is H, we obtain from Eq. (15) the condition under which internal circula-
tion is as yet absent in the unit:
(16)
For values ev < OZ'r, the unit operates with internal circulation.
To determine the flow of the light fraction emerging from the unit, we consider the gas balance in an
arbitrary cross section of the unit. Neglecting transport of matter by longitudinal diffUsion, we have
dG= ?P Tx 4, (17)
where G(y) is the flow in some cross section y of the outer space of the unit.
Considering Eqs. (8) and (15), and integrating Eq. (1'7) along the channel,
566
G
dG = ? P dy ,
(1-0).E
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(18)
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4 5 6 Ln q
Fig. 4, Dependence of enrichment factor e on In q
for various values of the vapor distribution factor
and 0 = 1/2.
we obtain the distribution of gas flow in the outer space:
G-(1-0)L-PH74-:1: ln q x
Considering that G(H) = L, we obtain from Eq. (19) tnhDelomagnitude
L'=OL=PH
10
0)y]/(_1)) .
of the light fraction,
(19)
(20)
(21)
v{1 11 (1
of the flow
1-10_`:1:2;
and of the flow into the unit,
reD
L --= P
ya0v q
1 ?0v1 /(")
lnq.
?01
Equation (21) shows that the feed flow L depends on the gas-flow partition factor 0, i.e., on the scheme
for combining units into a cascade. One should make note of the particular feature of a cascade of mass-dif-
fusion units operating in a given mode (for given Ov and In q) which is manifested by the fact that the flow L'
of the light fraction does not depend on 0.
Using the condition (13), Eq. (20) can be presented in the form of a criterial equation,
L*-Q* _J. !i) (. In q )1.1(-1)
(22)
which is suitable for explaining the physics of the process and for establishing a relation between total vapor
flow, Peclet diffusion number, and the flow of light fraction produced by the unit. In Eq. (22), L* = L'/e/PH ?
anDto and Q* = Q/e/PHnDio are dimensionless quantities respectively characterizing the flow of the light
fraction and the total vapor flow fed into the unit, Figure 3 gives the dependence of L* on Q* calculated from
Eq. (22) for various Peclet numbers. Setting y = H in Eq. (13), it is easy to obtain a relation between In q and
the total vapor flow:
(23)
It is clear from Fig. 3 that the flow of the light fraction depends on the total vapor flow and on the vapor dis-
tribution factor. When Qv = 0, which corresponds to passage of all the vapor through the partition, L* --- 0,
Therefore, each curve begins at the point Q* = In q when L* = 0. It is then also clear that the flow of the light
fraction increases as ev increases.
We turn to a calculation of the separation produced by a mass-diffusion unit. Equation (6) indicates that
Tx is independent of the x coordinate and therefore, for the assumed model of a mass-diffusion unit, one can
consider that transverse transport of the light component is only realized because of its concentration gradient
along the diffusion path /e. Then one can obtain from Eq. (3) the differential equation
567
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d (c' ?c) Tx (c, c) eoc (1 c) ddlnxy ? 0
dx nDy
(24)
for transverse transport of the light component, assuming TC where c' is its concentration at the begin-
ning of the diffusion path.
In separation of isotopes, their concentration changes little in the transverse direction of the unit, which
makes it possible to consider the quantity c(1? c) a constant in Eq. (24). Using this assumption, integration
Of Eq. (24) over the diffusion path length under the boundary conditions c(0) = c' and c(/ e) = c" leads to an ex-
pression for the concentration differential of the light component produced in the partition:
A = c"?c' =e c (1 c) q Ya -1". x [exp (tux/ einDio) ?11,
(Iva 7.?qvb
(25)
where = 1? (D10/D12).
Let the gas flow be G in some cross section y of the outer space and let the concentration of the light
component be Ca. In an interval dy, the flow through a channel boundary is dG at a concentration ca + A, with
the increment of concentration being given by Eq. (25). Denoting the small change of concentration along dy by
dca, we obtain from the balance of flows in the selected elementary segment
dGA =G dca (26)
neglecting smallness of second order and transport of matter by longitudinal diffusion. Using Eqs. (17) and
(25), introducing the notation
a?
Yb [eXP (tUrxleinDia) ? I]
f?tl'a Va TYa
and separating the variables, we rewrite Eq. (25) in the form
Integration of Eq. (28) along the unit,
dca ? CCPTTd
cc,(1?ca)? 8? G Y'
co
dca
c-
?01) 2?dY
yields a relation for the enrichment factor of the heavy fraction:
e- ln c-1(1?c-) cd(1?c?) ? sop c4GTx dy,
(27)
(28)
(29)
(30)
where Tx and G are defined by the appropriate equations.
Using the well-known relation between the enrichment factors for the light and heavy fractions (c+ and
c-) and the total enrichment factor c [4],
we obtain for the quantity c the expression
568
=-- = 08,
aTx
8= ?T P
(31)
(32)
The integral on the right side of Eq. (32) can only be evaluated numerically in the general case. In the
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particular case of very high vapor flow rates, where the value of the vapor distribution factor ev is close to
one, the value of gas concentration in the inner space of the unit can be assumed to be zero. Then the solution
of the equation system (2)-(3) is considerably simplified and yields for the enrichment factor and flow of the
light fraction expressions which agree with those given in [1, 2]:
qlnq 1
?
6=8? q-1
nDio
L' = In q
P .
q? 1
(33)
(34)
Figure 4 shows typical relationships between the enrichment factor and the diffusion Peclet number cal-
culated from Eq. (32) for various values of the vapor distribution factor and 0 = 1/2. The curve for O. = 1
corresponds to the limiting case yb = 0. The fact that the curves e = e(ev, ln q) practically coincide when In
q>> 2 indicates the insignificant effect of the gas in the space II on the diffusion of the components of the mix-
ture being separated. Therefore the approximate theory, which is based on the assumption of an absence of
gas in the inner space, can be used successfully to evaluate the enrichment factor. Equation (33) is particu-
larly convenient in that the dependence of e on the gas flow partition factor e is explicitly expressed in it.
However, the approximation yb = 0 is unsuitable for calculation of the flow of the light fraction. The fact
is that there is a reverse convective transport of gas in actual separative units which depends on the magnitude
of Iv and neglect of gas concentration in the inner space is equivalent to neglect of this reverse transport.
Such a situation leads to an overestimate in the calculated values of the emerging flow L' in comparison with
the values achieved in practice.
The authors are grateful to I. G. Gverdtsiteli for discussions of the results, valuable advice, and continu-
ing consideration of the work.
LITERATURE CITED
1. L G. Gverdisiteli and V. K. Tskhakaya, in: Isotope Production [in Russian], Izd. Akad. Nauk SSSR, Moscow
(1958),p. 113.
2. I. G. Gverdtsiteli et al., in: Proceedings 2nd Geneva Conference [in Russian], Reports of Soviet Scien-
tists, Vol. 6, Atomizdat, Moscow (1959) , p. 69.
3. R. Ivans et al., J. Chem. Phys., 35, 2076 (1961).
4. N. A. Kolokol'tsov, At. Energ., 27, No. 1, 9(1969).
569
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Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070006-5
DEPOSITED ARTICLES
INDUCED ACTIVITY OF BUILDING AND STRUCTURAL
MATERIALS IN THE 680-MeV SYNCHROCYCLOTRON HALL
V. F. Kas'yanov, M. M. Komochkov,
Yu. G. Teterev, and V. V. Mal'kov
UDC 539.16.04:621.384.67
In this paper, we have investigated the principal mechanisms of activation of building and structural
materials in the scattered field of neutrons, which would permit the induced activity to be forecast at any point
of the accelerator hall. The problem is solved by determining the radioisotopic composition in irradiated
samples of materials and by measuring the radiation dose intensity from the induced activity. The activity of
the isotopes and the dose intensity from the induced activity were compared with the flux densities of neutrons
of different energy groups, measured by threshold detectors. The neutron flux density (neutrons! cm2 sec)
at different parts of the synchrocyclotron hall amounts to 105- 103, 105- 103 for thermal and intermediate
neutrons, 103- 103 for fast neutrons with E = 2 - 20 MeV and 102 - 103 for relativistic neutrons when E > 20
MeV.
Samples of iron, normal concrete, copper, and aluminum at different points of the accelerator hall served
as building and structural materials. The activity of the isotopes in the samples was measured with NaI(T1)
and Ge(Li) spectrometers.
Table 1 shows the specific activity of the isotopes in structural materials, which make the main contribu-
tion to the radiation hazard and the spread of its values. The samples of iron and concrete are taken at dif-
ferent parts of the accelerator hall. The presence in iron of 63Co is explained by impurities of cobalt 10"
g/g) and the isotope 56Mn with its origin is due mainly to impurities of 3.10-3 g/g of manganese.
Similar irradiation conditions permitted different materials to be compared with respect to their radia-
tion hazard from induced activity. The total of the specific activities of the products formed were compared
by their totoal y constants for each material. For Fe, Cu, Al, and concrete, the ratios of these values at in-
finite irradiation are as follows: 10:20:10:1, respectively, over 1 h, and 1:0.6:0.4:0.1 over 90 days after comple-
tion of irradiation.
A relation was established between the dose intensity generated by the y radiation from the induced
activity, and the flux density of neutrons with energies of 2 Me ?0 ? OH + OH'
Me ? 0+ 2e? H20 Me +20H'
OH+ H202+ OH' 0; +2H20
The corresponding scheme for acid solutions is stated.
(No. 849/8460. Paper submitted August 6, 1975. Complete text 0.65 author's folio,
5 Figs. and 25 Refs.)
PRODUCTION AND STUDY OF CORROSION RESISTANCE
IN ZIRCONIUM DIBORIDE AND ITS SOLID SOLUTIONS
WITH TITANIUM DIBORIDE
V. V. Svistunov, A. R. Beketov, UDC 546.831.271:620.193.2
V. G. Vlasov, and N. V. Obabkov
The practical application of zirconium boride is of special interest because of its high chemical durabil-
ity. In order to improve some of the physical and chemical properties of ZrB2, admixtures of TiB2 are added
to form solid solutions.
In these studies samples were used that were obtained by a combination method including high tempera-
ture self-ordering fusion and activated sintering. The samples had a residual porosity of 2.4 vol.% and were in
the shape of plates of size 2 x 6 x 6 mm. They were used for oxidation in a 3.10" liter/min flow of pure oxygen.
During oxidation of ZrB2 samples it was found that over the temperature interval 300-700?C the rate of
oxidation increases somewhat while remaining insignificant, then decreases (800?C) and increases sharply at
temperatures above 800?C. All the curves fit a paralinear oxidation law.
The dependence of the oxidation rate on the oxygen pressure is complicated. In the low pressure range
up to 50 torr the oxidation rate increases with a pressure rise. This is due to the high volatility of boron
oxides. A further increase in the pressure to 75 torr suppresses the volatility of the oxides, the amount of
B203 in the protective film increases, and the oxidation rate decreases and later increases again.
The kinetics of the oxidation process is complex and to a great extent depends on the amount of boron
oxides in the protective layer. During oxidation of mixed borides the oxide layer is enriched in titanium which
causes formation of ZrTiO4 in all samples, which sharply impairs the protective properties of the surface
scale [1].
The apparent activation energy for oxidation of ZrB2 is 20 kcal/mole. Small admixtures of TiB (about
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8 mol. %) reduce this to 14 kcal/mole, but depending on the increased aniount of TiB2 01 the solid soltition
the activation energy increases and for 30 mol. % TiB2, it is 30 kcal/mole.
The paralinear oxidation law, experiments with an inert tracer, and the appearance of the oxidized
samples [2] lead us to assume that oxidation of ZrB2 in solid solutions takes place in a diffusive regime with
predominant diffusion of oxygen to the reaction interface at the boride? oxide film.
LITERATURE CITED
1. I. N. Frantesevich, R. F. Voitovich, and V. A. Lavrenko, High-Temperature Oxidation of Metals and
Alloys [in Russian], Gostekhizdat UkrSSR, Kiev (1963).
2. J. Benard, Oxidation of Metals [Russian translation], Metallurgiya, Moscow (1968).
(No. 850/8465. Paper submitted August 12, 1975. Complete text 0.5 author's
folio, 3 Figs., 2 Tables, 19 Refs.)
THERMALIZATION OF NEUTRONS IN SOLIDS
V. A. Baikulov UDC 621.039.512.45
The variational method for calculating the distribution of slow neutrons in infinite homogeneous media
with uniformly distributed sources of fast neutrons described andtested in [1] for a gaseous moderator model
is extended to a solid medium.
No limitations of any kind are placed on the behavior of the neutron absorption cross section in the ther-
malization region. The incoherent Gaussian approximation is used for the differential scattering cross section.
A method for computing the dispersion is described which is based on use of experimental values of the
generalized frequency distribution for liquids and the characteristic normal oscillation mode functions for
crystals. A method is developed for calculating the matrix elements characterizing the scattering of neutrons
by solid systems.
The energy distribution of slow neutrons in water is computed at temperatures of 295 and 423?K with
19B, 113Cd, and 149Sm as absorbers. The effect of various regions of the generalized frequency distribution for
water on the slow neutron spectrum is studied. A comparison is made with known models and experimental
data. It is shown that a model using the low frequency distribution given in [2] in combination with the high
frequency distribution of Nelkin (two vibrational levels with energies of 0.205 and 0.481 eV with the "weights"
of the two levels in the ratio 1:2) with the ratio of the contributions from these distributions being 1:2.38 agrees
better with the experimental data of [3] than Nelkin's model (see Fig. 1).
(ME)
los
104
103 111111u I I 1 1 1 1111 1 1 111111 I I 1 11 1111
;001 401
0,1 10 E, eV
Fig. 1. Energy distribution of the slow neutrons,
4,(E), in water at a temperature of 295?K with the
absorption cross section of cadmium equal to 15.4
barns per atom of hydrogen at energy E
model in this work; -- - -) Nelkin's model;
- ?- ?-) gaseous model; 0) and x) experiment.
LITERATURE CITED
1. V. A. Baikulov, At. Energy., 31, No. 5, 507 (1971).
2. P. Egelstaff, B. Haywood, and J. Thorson, in: Inelastic Scattering of Neutrons in Solids and Liquids,
Vol. 1, IAEA, Vienna (1963), p. 343.
3. J. Beyster, Nucl. Sci. and Engng, 9, No. 2, (1961).
(No. 851/8483. Paper submitted Noverniber 16, 1975. Complete text 0.6 author's
folio, 7 Figs., 1 Table, 10 Refs.)
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LETTERS
USING DIAMOND DETECTORS AS IMMERSED a COUNTERS
S. F. Kozlov, E. A. Konorova,
M. I. Krapivin, V. A. Nadein,
and V. G. Yudina
UDC 621.376
The parameters of diamond detectors for ionizing radiations were studied in [1-3]. The fact that the
diamond detectors are not affected by chemically agressive media is one of the main advantages of these detec-
tors. It is interesting to explore the possibilities of using diamond detectors for measuring the a activity with-
out need for preparing thin, dry samples when the detector is in direct contact with the solution.
It is known from the previously published papers that when the sensitive surface of semiconductor detec-
tors, e.g., silicon detectors [4], is in contact with a liquid, there occur several phenomena which imply that the
characteristics of immersed detectors differ substantially from the characteristics of detectors with gold
coating.
We consider in the present work the characteristics of a diamond detector recording the a radiation from
solutions; the sensitive surface of the detector was in contact with the electrolyte. For the measurements on
the solution, the detector was hermetically glued to a lateral ground-in joint of a glass column (Fig. 1). The
a activity was measured in bimolar nitric acid solutions of 239PU in concentrations between 10 and 100 pg/cm3.
An "Amur" pulse amplifier, a 512-channel LP-4050 pulse analyzer, and a PP9-1 scaler were used to record
the a particles and to measure the amplitude spectra of the pulses. A detector in the form of a ? 100-p-thick
diamond plate was used for our measurements. One of the detector faces had a chemically stable contact which
had been obtained by ionic alloying of the diamond with boron; the opposite face of the detector carried a con-
tact of deposited gold.
Two series of measurements were made. In the first series, the detector was glued to the column so
that the diamond surface alloyed with the boron was in direct contact with the solution on which the measure-
ments were to be made. A bias voltage of +330 V was applied to the gold contact; the second contact was
grounded through the electrolyte with the aid of a stainless steel electrode. Figure 2a shows the amplitude
spectrum obtained in this case. In the second series, the gold contact was removed from the detector and the
diamond plate was glued to the ground-in column section with the diamond face from which the gold coating
had been removed. Thus, the a active solution was in direct contact with a surface of pure diamond. A bias
voltage of +450 V was applied to the boron contact; an external electrode was used for grounding via the elec-
trolyte (Fig. 2b). When the pure surface of the diamond crystal was in direct contact with the electrolyte
(gold contact removed), the maximum amplitude of the pulse spectrum was 4.5 times smaller than the maximum
amplitude of the spectrum obtained in measurements made with the detector provided with two contacts. This
result can obviously be explained by phenomena which occur at the contacts, i.e., at the diamond?electrolyte
interface.
The time-dependent counting characteristics of the diamond detector were checked with the surface of
the pure diamond in contact with the a active solution. Plutonium nitrate solutions with concentrations of
10.3, 20.7, 52.4, and 102.3 pg/cm3 were used for the measurements. The number of pulses counted per unit
time depended linearly upon the plutonium concentration of the solutions. No changes of the counting character-
istics of the diamond detector were observed during ten days when the diamond detector was continually in
contact with the plutonium nitrate solution.
The experiments have shown that, in principle, diamond detectors can be used as immersed a counters
Translated from Atomnaya Energiya, Vol. 40, No. 6, pp. 482-483, June, 1976. Original article submitted
December 30, 1974.
This material is protected by copyright registered in the name of Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part
of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying,
microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $7.50.
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Fig. 1
Number of pulses
50
Channel No.
Fig. 2
200
100
Fig. 1. Glass column with a diamond detector glued to it: 1) contact with the surface obtained by
ionic alloying of the diamond with boron; 2) diamond plate; 3) contact with the surface, sputtered
gold; 4) stainless steel electrode.
Fig. 2. Amplitude spectra of the pulses obtained by recording the a radiation of a solution: a)
when the electrolyte touched the boron contact; and b) when the electrolyte was in direct contact
with the diamond surface.
either when diamond is alloyed with boron or when pure diamond is in direct contact with the solution. However,
immersed diamond detectors with a chemically stable contact obtained by ionic alloying (doping) or carbidizing
have the best counting characteristics.
LITERATURE CITED
1. E. A. Konorova, S. F. Kozlov, and V. S. Vavilov, Fiz, Tverd. Tela, 8, 3 (1966).
2. E. A. Konorova and S. F. Kozlov, Usp. Fiz. Nauk, 98, No. 4, 735 (1969).
3. E. A. Konorova and S. F. Kozlov, Fiz. i Tekh. Poluprov., 4, No. 10, 1865 (1970).
4. L. Cathey and W. Jenkins, Trans. Nucl. Sci., 9, No. 3, 193 (1962).
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SENSITIVITY OF EMISSION DETECTORS TO y RAYS
G. V. Kulakov and B. V. Mukhachev UDC 539.1.08
An emission detector is a system of electrodes of different materials spatially separated in vacuum. The
current in the detector is caused by counter flow of electrons knocked out of the electrodes by y quantas.
The object of the present work is to construct analytical models and to investigate experimentally the
sensitivity of emission detectors with quartz tube between the electrodes in the fields of stationary high in-
tensity (up to 106 R/sec) y radiation. The main advantages of such detectors are the operation without external
supply sources, high thermal and radiation stability, small dimensions, and the absence of an upper limit of
the intensity of the exposure dose [1, 2].
The construction of a vacuum emission detector is shown in Fig. 1. Tantalum with 0.3 mm thickness,
zirconium or stainless steel with 0.7 mm thickness was used as the emitter. The outer diameter of the emit-
ters was 5.4 mm. A quartz tube of 3.2-mm diameter and 0.5-mm-thick wall was used as the insulator. The remain-
ing parts were made of stainless steel. The diameter of the collector was 1.8 mm. The length of the sensitive
part was 150 mm. A heat and radiation proof KTMS(S) cable with magnesium oxide insulation and stainless
steel -casing was used for signal transmission. The detector was assembled in 5.10-6 mm Hg vacuum.
When the detector is irradiated, electric current is produced not only in the sensitive part but also in
its connecting part and in the cable. In order to take into consideration the contribution of these currents to
the detector sensitivity measurements were carried out with a compensated detector having similar construc-
tion but without the sensitive part.
It is shown in [3, 4] that the basic characteristics of emission detectors are well described by an analyt-
4
5
K(c.C..c(KKK 10 m, the dependence of the signal on the base d becomes
insignificant (Fig. 3). For an energy of 0.06 MeV and H = 1-10 m, with d = 0.5 m and 211,0 = 30?, the values of
Ls for water, concrete, and iron are 10-5-10-7; 5-10-6-10-7; and 10-6-2 -10-8, respectively.
With increase in the radiation collimation angle, the value of the useful signal increases. For
Ey 0.06-0.2 MeV, d = 0.5-3 m, and 24,0 = 30-120?, Ls = 10-4-10-6 for a concrete refledtor (Fig. 4). With in-
crease in H, the dependence Ls = f(410) grows less significant.
For Ey = 0.06 MeV and d = 1 m, with 24,0 = 30? and H = 1-10 m, the values of Ls for water, concrete,
and iron are 10-7-6-10-6; 5 .10-8-2 ? 10-6; and 6-10-8-9-10-7, respectively.
Experiments on testing units and objects confirm that the theoretical assumptions and calculations are
correct (Fig. 5). In particular, the experimental measurements given in [1] show that the calculated signal
transmission coefficient Ls for energy quanta Ey = 0.661 MeV agrees closely with experiment. Both in theory
and practice, the maximum intensity of backscattered radiation is observed for reflection from water. The
measured values of the backscattered radiation for water and concrete [1], recorded systematically for H =
0-30 m, are in good agreement with the calculated signal transmission coefficient Ls in this range of H. Ex-
perimental measurements of backscattered x radiation for different values of the base d are in good agreement
with the calculated values. Both in theory and practice, it is clear that the effect of d on the value of the back-
scattered signal at the detector inlet becomes less pronounced for distances H a- 10 m.
Thus, the value of the signal at the detector inlet in x-ray control systems is determined by the energy
of the x-ray beam, the geometry of the receiver?transmitter channel, and the collimation angle of the emitter.
By varying these parameters for each type of reflector, a maximum value of the signal can be obtained at the
detector inlet for minimum size, weight, and energy characteristics of the apparatus.
In conclusion, the author would like to thank N. F. Andryushin and V. N. Barkovskii for critical comments
made on reading the manuscript.
LITERATURE CITED
1. B. P. Bulatov and N. F. Andryushin, Backscattered y Radiation in Radiation Engineering [in Russian],
Atomizdat, Moscow (1971).
2. V. I. Glagolev et al., in: Nuclear Instrument Design, Proceedings of the Soviet Scientific-Research
Institute of Instrument Design [in Russian], No, 10, Atomizdat, Moscow (1969), p. 85.
3. I. Ya. Serebrennikov etal., in: Radiation Engineering [in Russian], No. 3, Atomizdat, Moscow (1969),
p. 114.
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LIBERATION OF HELIUM IN THE UNIFORM HEATING
OF NEUTRON-IRRADIATED OKh16N1 5M3B STEEL
N. P. Agapova, I. N. Afrikanov,
A. I. Dashkovskii, A. G. Zaluzhnyi,
V. D. Onufriev, D. M. Skorov,
Yu. N. Sokurshii, and 0. M. Storozhu.k
UDC 621.039,546;541,1845:546.291
It is known that, under the action of neutron irradiation, (n, a) and (n, p) nuclear reactions in construc-
tional materials lead to the formation of helium and hydrogen, accumulation of which in the material may cause
undesirable changes in its mechanical and physical properties. Therefore, it is important to investigate the
accumulation and behavior of gaseous nuclear-reaction products in materials and, in particular, the processes
by which they are liberated from the material.
In [1] an account was given of the method used and the results obtained in a study of the kinetics
of helium liberation from samples of nickel and OKh16N15M3B steel (in its initial state, a foil of thick-
ness ,-, 100 p with recrystallized grains of size 30-50 p) , in which helium was produced by bombardment with
a particles in a cyclotron. In the course of uniform heating of the given samples at a rate of 7 degC/min,
6 peaks in the rate of helium liberation were observed.
The present paper reports an investigation of the liberation of helium from samples of steel OKh16N15 ?
M3B of analogous initial state, irradiated in a reactor at a temperature not exceeding 500?C by a neutron flux
of 4.0 ? 1022 neutrons/cm2 (E > 0.1 Mev). The kinetics of gas liberation in the course of uniform heating at a
rate of 7 deg C/min was investigated by the method described in [1]. A typical curve for the liberation of
helium in these conditions, obtained for a constant rate of drawing off the helium, is shown in Fig. 1. At tem-
peratures above 500?C on the kinetic curve, five high-temperature peaks of helium-liberation rate can be ob-
served (II'-VI'). Comparing this curve with that for the cyclotron-irradiated samples, we see that they are
identical at temperatures above 500?C (the peaks of gas-liberation rate are labelled in accordance with [11).
At temperatures above 500?C, the liberation of helium from samples of steel in which helium is formed
200 400 600 BOO 1000 T,T
Fig. 1. Kinetics of helium liberation from irradiated
samples of OKh16N15M3B steel (4.0 ? 1022neutrons/cm2)
in the course of uniform heating at a rate of 7 deg C/min.
Translated from Atomnaya Energiya, Vol. 40, No. 6, pp. 490-491, June, 1976. Original article submitted
May 26, 1975; revision submitted December 15, 1975.
This material is protected by copyright registered in the name of Plenum Publishing Corporation, .227 West 17th Street, New York, N.Y. 10011. No part
of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying,
microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $7.50..
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in the (n, a) reaction on irradiation by fast neutrons in a reactor is characrerized by the same stages as the
helium liberation from samples subjected to irradiation in a cyclotron. Hence, the liberation of helium from
the steel at temperatures of 600-700?C corresponds to the diffusion of gas atoms by a vacancy mechanism,
while the liberation of helium at temperatures of 900?C and above corresponds to the migration of gas bubbles
to the sample surface [1].
Investigation of the irradiated samples of steel by transmission electron spectroscopy showed the pres-
ence of pores with distinct boundaries. The pore size lay in the range 50-250 A (mean value 150 A), while the
pore concentration was ? 2 .1015 cm-3. Calculation showed that the creation of an equilibrium pressure of
helium in the pores would require a helium content larger by a factor of 35 than that observed experimentally
using the apparatus and the procedure described in [2]. This means that the pores have the character of vacan-
cies, i.e., they are not filled with helium atoms. The vacancy pores, naturally, provide a channel for the helium
atoms present in the lattice of the material, but evidently they do not absorb all of the helium atoms diffusing
by a vacancy mechanism, and some of them appear at the surface, causing the peak in the gas-liberation rate
in the range 600-700?C. When the temperature is increased, the helium atoms evidently collect in the pores.
At high temperatures, helium liberation from irradiated samples of OKh16N15M3B steel occurs through the
migration of pores to the sample surface and the release of their helium content. In the absence of neutron
irradiation, the solution in the lattice of helium present in the pores is unlikely [3].
The curve of helium liberation (Fig. 1) lacks the low-temperature peak in the helium-liberation rate
(300-400?C) observed on the corresponding curve for samples of the same steel saturated with helium in a
cyclotron. The explanation is that, on irradiation in the reactor, the maximum temperature of the samples
reaches 500?C, and the process of gas liberation at these temperatures goes to completion in the course of
irradiation. Hence, in the subsequent uniform heating of the samples of steel irradiated in the reactor, there
is scarcely any further liberation of helium at temperatures below 500?C (Fig. 1).
LITERATURE CITED
1. D. M. Skorov et al., At. Energ., 40. No. 5, 387 (1976).
2. A. I. Dashkovskii et al., At. Energ., 40, No. 3,251 (1976).
3. D. Rimmer and A. Cottrell, Philos. Mag., 2, 1345 (1957).
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SEARCH FOR FISSILE ISOMERS IN THE n, 2n) REACTION*
J. S. Browne and R. E. Houve UDC 539.172.4
The formation of spontaneously fissioning isomers of various actinide nuclei in the (n, 2n) reaction
produced by 14-MeV neutrons has been reported [1, 2]. A ratio between the isomer formation cross section
and the prompt fission cross section in the range (1-2)-10-4 was obtained for the reactions 240Pu(n, 2n)239mPu
(T112 = 8.5 psec) and 242Pu(n, 2n) 2411nPu (T1/2= 27 psec).
It was suggested [3] that one should expect to find an as yet unknown isomer 243MPU with a half-life of the
order of 10 psec according to Metag's systematics [4] for isomeric lifetimes. The only known isomer of 243Pu
has a lifetime of 33 nsec, which is not in agreement with the systematics based on calculations of the fission
barrier. Since the (n, 2n) reaction appears to be a good mechanism for the formation of fissile isomers, we
decided to look for an isomer of 243PU in the reaction 244(1, 2n)243Pu.
A device at the Lawrence Livermore Laboratory having a high-intensity beam of 14-MeV neutrons was
used for this purpose. A beam of 500-keV deuterons created in the device was incident on a rotating titanium
?tritium target at a current of 10 mA. With a continuous beam, the neutron yield was of the ordei of 3-1012
n/sec. The beam was pulsed at frequencies of 10 and 100 KHz, and 2.5 MHz for isomer searches between
pulses in the time intervals 100 and 10 psec and 400 nsec, respectively. The pulse widths were 1.5 psec, 130
and 1.5 nsec, respectively.
A diagram of the device is shown in Fig. 1. Samples of 238U, 242?.r,
u and 244Pu were set up in the ioniza-
tion chamber. The mass and isotopic composition of the samples are given in Table 1. Fission events were
recorded with a fast current-sensitive preamplifier (rise time < 10 nsec) having a time resolution of 2 nsec.
Analysis of the time between pulses from neutrons and from fission was made with a fast discriminator and a
TABLE 1. Sample Composition
' Sample
Mass; mg
. ,
Ptirity, ?Jo
Density,
lig/drn2
238ti ?
242Pu
244Pu
1,0
1,0
0,59
99,9
99,9
99,1
?500 '
?500
?500
Fig. 1. Diagram of arrangement of ionization cham-
ber and 14-MeV neutron source: 1) deuterons; 2)
collimator; 3 ionization chamber (96% Ar and 4%
CO2); 4) rotating tritium target.
*Work supported by the Energy Research and Development Administration (ERDA).
University of California Lawrence Livermore Laboratory. Translated from Atomnaya Energiya, Vol.
40, No. 6, pp. 491-492, June, 1976. Original article submitted July 7, 1975.
This material is protected by copyright registered in the name of Plenum Publishing Corporatick 227 West 17th Street. New York, N.Y. 10011. No part
of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying.
microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for$7.5.0.
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io7 107
a
Number of fissions
/0 ? Lb ?sec
ID 105--
104
10 103
102 ? 102_
/01 I I I I 1 I IQ
60 50 40 30 20 10 0 70 60 50 40 30 20 10 0
Time, ?sec S Time, ?sec
Fig. 2. Time spectra corresponding to fission events in samples of 242Pu and 244Pu: a) 242pu.
(n, 2n) 241 pu; b) 244pu(n, 2n) 243PU (background level given by the solid line outside the bound of
the prompt fission peak).
1 I I I I I I
time-to-pulse-height converter (TPC), the output signal of which was fed to a multichannel analyzer. Pulse-
height spectra from the TPC for 242PU and 244PU at 10 KHz and 1.5 ?sec are shown in Fig. 2. None of the spec-
tra showed the presence of a decaying isomer. A 27-?sec activity was reported [2] for 242PU.
The total number of counts in two successive half-lives is compared in order to determine the sensitivity
of the measurements. If an isomer with a half-life of 27 ?sec is present during an experiment, there should
be a difference between the number of counts in the intervals 0-27 ?sec (first half-life) and 27-54 ?sec (second
half-life). Denoting these sums by Si and S2, we find that Si = 5700?75 and S2 = 5715? 75; consequently, there
is no activity with a half-life of 27 ?sec observed within the limits of the statistics. If one takes a deviation of
three standard errors from the mean (Poisson distribution) as the minimum observable activity in order to
record some kind of activity, one can then determine the ratio of isomeric decay to prompt decay (only 0.13%
of the cases were outside the 30- limit in the Poisson distribution in our measurements; consequently, the
probability limit of the present work is better than 99%). For 242Pu, 30- = 3.75 = 225. Since only half the
decays fall in the first half-life, the required minimum is 450. The number of prompt counts is 1.0 ?107and con-
sequently the ratio is 450/107 = 4.5.10-5, i.e., four times less than that observed in [2]. If the ratio given in
[2] were correct, 700 counts above background would have been observed in first half-life. Note that the 244back-
ground is produced by spontaneous fission in measurements of 242PU and Pu. Our conclusion that the ratio
of isomeric decay to prompt decay for the reaction 242PU(n, 2n)241MPU (Ti = 27 ?sec) must be less than 4.5 ?
10-5 disagrees with the data of [2].
We also failed to observe an activity in the 10-?sec range, the appearance of which was expected in the
244Pu(n, 2n)243PU reaction according to [3, 4]. The upper limit of the ratio established by the same method is
2.10-5 for the same 30- criterion.
As should be expected, a fissile isomer was not observed for 238U either since an isomer of 237U has not
been observed thus far.
These results for the (n, 2n) reaction can be compared with other results in which the compound nucleus
evaporates two neutrons, the (p, 2n) reaction for example. The (p, 2n) reaction was used [5-7] for the produc-
tion of isomers of various actinides. The maximum ratios of isomeric to prompt fission fall within limits
from 5 ? 10-5 to 5 ? 10-6 for E = 13 MeV. This is in agreement with the limit obtained in our experiment.
Therefore, if isomers are formed in the (n, 2n) reaction produced by 14-MeV neutrons, their formation cross
sections for 24IPu and 243PU are at least four times less than previously reported.
The authors thank Dr. G. Otampo for furnishing the 244PU sample and Dr. M. S. Coops for preparation of
high-purity 242pu an ?
o 238U samples.
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LITERATURE CITED
1. Yu. P. Gangrskii et al., At. Energ., 31, No. 2, 156 (1971).
2. A. Belov et al., JINR Preprint E15-6807 (1972).
3. J. Browne and C. Bowman, Phys, Rev., 9, 1177 (1974).
4. V. Metag et al., in: Proc. IAEA, Physics and Chemistry of Fission, Vol. 1, Vienna (1969), p. 325.
5. J. Borggreen et al., Phys. Letters, 25, 402 (1967).
6. N. Lark et al., Nucl. Phys., A139, 481 (1969).
7. S. Bjornholm et al., Nucl. Phys., A95, 513 (1967).
OPTIMAL ARRANGEMENT OF EFFECTIVE ABSORBER
IN A REACTOR
A. M. Pavlovichev and A. P. Rudik UDC 621.039.51
Among the many reactor optimization problems associated with the arrangement of fuel and absorber
within the volume of a reactor [1], the problem of optimal arrangement of effective absorber has not yet been
considered. Strictly speaking, one should introduce in its formulation two control factors ? the concentration
of effective absorber and the concentration of fissile material ? and one should also take into account the
burnup of nuclides during an operating period. To simplify matters, we assume that the distribution of fissile
material is constant over the reactor volume and changes only because of displacement during the arrangement
of effective absorber. It is required to find an arrangement for which the reactor is critical and for which the
amount of neutrons captured by the effective absorber at fixed reactor power is a maximum. Questions asso-
ciated with the burnup of nuclides are not discussed.
We discuss this problem for a thermal reactor in a two-group diffusion approximation where the initial
equations for the one-dimensional problem have the form
with the boundary conditions
Di AcPi =Dicq.:1)t ? (E t ?FE I-2) ?--
Lf ?1?vEf202=
D2 6,02?D2Gq02?E202+Ii_201
V(Pi. (0) --v(P.2 (0) 0;
cPt (R)=gi VOi (R) gz V 02 (R);
(R) g3 V 4) i (11) gi V 02 (R). f
(2)
(3)
Here, 01 and 02 are, respectively, the fluxes of fast and thermal neutrons; R is the external radius of the re-
actor region in which redistribution of absorber is allowed; the other notation is standard. The boundary con-
ditions (3) make it possible to take into account the existance of a region on the periphery of the reactor in
which redistribution of absorber is not allowed, e.g., a reflector.
The concentration of useful absorber per unit volume of the reactor was taken as the control factor u(r)
which was subject to the limitation umin u(r) umax. The maximizable functional was selected either in
the form
J f u (1) (r)
(4)
Translated from Atomnaya Energiya, Vol. 40, No. 6, pp. 493-494, June, 1976. Original article submitted
June 10, 1975.
This material is protected by copyright registered in the name of Plenum Publishing Corporation, 227 IVest 17th Street, New York, N.Y. 10011. No part
of this publication may be reproduced, stored in a retrieval system, or -transmitted, in any form or by any means, electronic, mechanical, photocopying,
microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $7.50.
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if the effective absorber captured only fast neutrons, or in the form
= (r) cl) 2(r) dr.
V
if the effective absorber captured only thermal neutrons. The total reactor power was assumed given:
W= "(Ift t Ef202) dlr.
V
(5)
(6)
If necessary, a limitation can be placed on the energy density ? the power removed from a unit volume of the
reactor.
A linear dependence of cross sections on the control factor was considered: (?Et? E1-2 + PEfd /Di = al
4- bill; 042/ = a 2 + b2u; E1_2/ D2 =, a 3 + b311; E2/ D2 = a 4 + = a 5 + k5U; T.12 = a 6 + b6U; I i I , > ,
i = 1, 2, . , 6; b4 0, the remaining bi 5_ 0; a1 < 0, a4 < 0, the remaining a i 0. The diffusion coefficients
were assumed independent of the control factor. The numerical values of the coefficients a i and bi selected
for the calculations were typical of a heavy-water reactor.
Using the methods of the mathematical theory of optimal processes [2, 3], one can obtain some informa-
tion about the optimal distribution of effective absorber prior to numerical calculations. It is well known that
in the absence of limitations on phase variables in linear control problems, the optimal arrangement can consist
of three types of regions: u = u min, u = umax, and a region of special control u = usp (in [1], the last region is
called the region of classical control). An analysis of the switching function in a "bare" reactor (g1 = g2 = g3
g4 = 0) shows that the region u = umax should be located on the periphery of the reactor. In addition, by using
the Kelly condition [3], or the zero value of the switching function in the region of special control, one can
determine in certain particular cases whether a special control region is included in the optimal arrangement.
For example, when 131 = b3 = b5 = 0, the special control factor satisfies the Kelly condition for J = J1 and is
included in the optimal arrangement but the special control factor is not included in the optimal arrangement
for J = J2.
The complete solution of the optimization problem was carried out by using an optimization program
based on the gradient method. An optimal arrangment was sought for various values of 4, R, and reflector
thickness. The quantitative solutions obtained had the following qualitative features. The optimal value of the
functional J increases with decrease in al, and with increase in R and reflector thickness.
For maximization of the functional (4), the optimal arrangement of the reactor, both with and without a
reflector, had the following form (from center to periphery): u = umin, u = usp, u = umax. For maximization
of the functional (5), the optimal arrangement consisted of two regions: u = umin and u = umax. For a cylindri-
cal geometry, the maximum gain which is achieved as the result of redistribution of absorber in comparison
with its uniform distribution is obtained in a "bare" reactor where R = 85 cm, which is the minimum
of the values considered. This gain was ? 30% and ,?,40% for Ji and J2, respectively. The gain from absorber
redistribution was reduced as R and reflector thickness were increased. It was 8% and 12% for J1 and J2 when
R = 140 cm and the reflector thickness was 25 cm.
It should be noted that with optimum absorber distribution from the viewpoint of maximum J, the maximum
energy density is considerably greater than for a uniform absorber distribution. If it is necessary that the
maximum energy density in an optimized device not exceed the maximum energy density in a reactor with uni-
form absorber distribution, the gain in the value of the functional J is decreased by approximately a factor of
two. In this case, a region of heat-engineering limitation appears in the center of the reactor.
LITERATURE CITED
1. A. P. Rudik, Optimal Arrangement of Nuclear Fuel in a Reactor [in Russian], Ser., Physics of Nuclear
Reactors, No. 2, Atomizdat, Moscow (1974).
2. L. S. Pontryagin et al., Mathematical Theory of Optimal Processes [in Russian], Nauka, Moscow (1969).
3. R. Gabasov and F. M. Kirillova, Special Optimal Control Factors [in Russian], Nauka, Moscow (1973).
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EVALUATION OF DOSE RATE FROM RADIATION HEATING
OF A SAMPLE DURING IRRADIATION
A. P. B'alashov and A. M. Mamontov UDC 539.12.08
Nonuniformity of energy deposition and of the temperature field can be neglected for a sample of small
size irradiated by high-energy particles or quanta. Under conditions of convective heat transfer, heat losses
are proportional to the temperature difference between the body and environment to a power close to one [1].
The temperature variation of a body during irradiation can be described by the equation
dT XS r D
cPc cpV n c '
where T is the temperature difference between the body and its environment (T = 0 at T = ; T is the irradia-
tion time; D is the dose rate, which is constant in time; c is the specific heat capacity of the sample; p, 5, and
V are the density, surface area, and volume of the sample; A is the coefficient of heat transfer.
We introduce the coefficient k = XS/pV which takes into account the dimensionality of the factor T.
In the initial stage of irradiation, heat loss is negligibly small, and the rise in temperature is linear,
(1)
dT D (2)
cis = c '
which makes it possible to evaluate the absorbed radiation dose rate from an experimentally determined rate
of temperature rise [2].
With further irradiation, the heat losses gradually increase, the rate of temperature rise tends to zero,
and its maximum value Is found from Eq. (1) when dT/dT = D. The dose rate can be determined from the stable
temperature difference Tmax,
D= kTn .
max
For an arbitrary value of the parameter n, we write the solution of Eq. (1) in the implicit form
dT
c D?kTn
,(3)
(4)
which defines the inverse relation T = T(T); the value of the dose rate can be found from a comparison of cal-
culated and experimental relations T(T). In practice, the problem reduces to minimization of a function of the
form
m2./ Ti )
Ti
E ft .1=1
1=1
(5)
where Ti and Ri are experimental and calculated values of the time in which the temperature difference Ti is
attained; m is the number of experimental points; the fi are weighting factors which take into account the
"density" of the points on the experimental curve.
Translated from Atomnaya Energiya, Vol. 40, No. 6, pp. 494-495, June, 1976. Original article submitted
July 18, 1975.
This material is protected by copyright registered in the name of Plenum Publishing Corporatioll," 227 West 17th Street, New York, N.Y. 10011. No part
of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form Or by any means, electronic, mechanical, photocopying,
microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $7.50.
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80
60
40
20
0 ZOO 400 600 800 1000 , sec
Fig. 1. Time variation of sample temperature:
45) during irradiation; 0) after end of irradiation.
TABLE 1. Results of Analysis of Radiation
Heating of Sample for Irradiation by 8.5-MeV
Electrons
i, AA /cm 2
0(104
erg/sec ? aeg
Dixie',
rad/sec
Dox los,
rad/sec
0,033
1,92
0,64
0,68
0,067
1,75
1,42
1,43
0,083
1,99
1,67
1,90
0,100
1,72
1,68
1,96
0,117
1,51
2,52
2,42
0,130
1,91
2,23
2,57
0,133
1,30
2,75
2,65
The values of Ri(Ti) are found from Eq. (4) by numerical integration for known parameters n and k.
An a priori estimate of the latter is difficult because of the dependence of the coefficient of heat transfer on
many factors [1]. It is simple to determine n and k by analyzing the curve for the drop in sample temperature
after the end of irradiation. The cooling process is described by Eq. (1) with D = 0 and the initial conditions
T = To when T = 0. The solution has the form
T = To exp (_-r) , n=i;
(1?en)k TT/1-3, n 1.
(6)
The value of n can be found by fitting a function similar to (5) by means of well-known methods of mini-
mization [3]. Furthermore, the value of k is determined by the least-squares method through linearization of
the experimental relation T(T) in appropriate coordinates.
In order to evaluate the effect of experimental error on the parameters found, test problems were solved
in which cooling curves T(T) (20-30 points) with an introduced random error of 1-5% normally distributed were
constructed for given values of n, k, and To. The relations obtained were subject to the analysis described above
in order to determine the parameters n and k. The results of the calculations indicate a noticeable effect of
the introduced errors on the magnitude of the heat-transfer parameters, which were calculated from a small
number of "experimental" points. However, it was established that this error has a very slight effect on the
value of the dose rate determined by means of Eq. (3), or by means of the minimization procedure (5), which
is explained by the dependence of the dose rate on both parameters.
The proposed method for evaluation of the dose rate was tested in a series of experiments. Aluminum
samples 10 x 15 x 30 mm in size wereirradiated with 8.5-MeV electrons incident on the broad face under
conditions of natural convective heat transfer. A platinum resistance thermometer glued to one of the faces
of the sample and a digital voltmeter were used for temperature measurement during irradiation. The error
in the determination of temperature was no more than 0.3?C. The electron current density j was determined
with a Faraday cup and had a relative error of 3-7%. The stability of the beam current was monitored with an
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induction sensor.
Figure 1 shows typical heating and cooling curves for a test sample corresponding to a current density
of 0.1 pA/cm2. The calculated heating curve (solid line) agrees with experimental temperature values with an
error of < 4%; for cooling, the calculated values practically coincide with the experimental values (calculated
values omitted in the figure).
Some results of analysis of temperature curves for different current densities are given in Table 1. It
was found experimentally that for a choice of the exponent n = 1.085, the calculated cooling curves mAtched
the measured curves with an error of no more than 4%.
The selection of a fixed value of n ensures a relatively slight spread in the parameter k, which depends
on the coefficient of heat transfer. The dose rates D1 were obtained from the initial portion of the heating
curves in accordance with Eq. (2); the dose rates D2 were obtained from analysis of the entire set of experi-
mental points by the proposed method. The dependence of D2 on electron current density is characterized by
a small deviation from linearity (maximum deviation, < 12%) while the analogous relation Di(j) is character-
ized by a relatively large spread in values, particularly at high current densities.
The proposed dosimetry technique, which is based on analysis of radiation heating of a body during ir-
radiation, makes it possible to obtain an independent evaluation of the dose rate from radiation directly absorbed
in the irradiated object. It can be used for studies in the fields of radiation physics and materials testing. For
a known dose rate, the results obtained make it possible to predict the kinetics of the temperature variation
of an object during irradiation. The required heat-transfer parameters can be determined in advance from heat
experiments. It should be noted that the possibility of using the proposed method is limited to conditions of
small radiative heat loss. If radiative heat transfer predominates, one can use the results in [4], in which the
characteristics of radiative heating of solids in a vacuum were studied.
LITERATURE CITED
1. H. Carslaw and J. Jaeger, Conduction of Heat in Solids, Clarendon (1959).
2. V. D. Anan'ev and I. M. Matora, At. Energ., 29, No. 4, 285 (1970).
3. D. J. Wilde, Optimum Seeking Methods, Prentice-Hall (1964).
4. Sh. Sh. Ibragimov, V. F. Reutov, and I. Yu, Abrashitov, Izv. Kazakh SSR, Ser. Fiz. Mat., No. 6, 84
(1974).
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THE CRYSTAL STRUCTURE OF THE COMPOUNDS
Pu5Rh4 and Pu5Ir4
A. V. Beznosikova, N. T. Chebotarev,
A. S. Luktyanov, M. P. Shapovalov,
and L. F. Timofeeva
UDC 548.73
Monocrystals of the componds Pu5Rh4 and Pu5Ir4 were studied by Laue's method, vibrations and photo-
graphy-of the reciprocal lattice (molybdenum emission). The intensities of the reflections were estimated
visually by the quality of blackening, taking account of the Lorentz and polarization factors. For the compounds
Pu5Rh4 and Pu5Ir4, the 354 and 173 reflections respectively, were used. The dimensions of the elementary
lattice were determined by oscillating crystal x-ray photography and they were refinedby powder x-radiography.
Calculations to determine the structure were carried out on a Minsk-32 computer according to a program
devised by the authors.
It was established that the compounds Pu5Rh4 and Pu5Ir4 are isostructural, related to a rhombic system
of the Pnma space group, Z ---- 4.
In the first stage of determining the structure, the coordinates of the atoms were found by the minimiza-
tion of the R factor as a function of the coordinates. In order to find the extremum, the complex-method of
Box was used. By achieving a value of R of the order 0.25-0.30, the distribution of the electron density in the
cell was found. The maxima of the electron densities were compared with the provisional values of the atom
coordinates. After quite good agreement, the final values of the coordinates were found by the least-squares
method, using a linearization procedure, which permitted the dispersion of the coordinates to be estimated.
In order to reduce the effect of factors which are dependent on the ratio of the sine of the angle of re-
flection and wavelength of the emission, the theoretical and experimental values of the intensity were reduced
to a single scale according to groups, with near values of this ratio. There were 12-22 reflections in each
group. The final values of the R factors for the compounds Pu5Rh4 and Pu5Ir4 amounted to 0.17 and 0.18, respec-
tively.
TABLE 1. Structure of Pu5Rh4 and Pu5 Ir4
Compounds
Compound
lattice
period
Positions
of atoms
Atomic parameter
Pu55h4
Pu51r4
Pu5Rh4
Pill- 8 (d)
0,158+1
0,155?1
a =7,263
11
0,122+1
0,123?1
b = 14,48
0,338+1
0,338?1
c = 7,464
Purr- 8 (d)
0,001+1
-0,002?1
9
0,090+1
0,091?1
-0,180+1
-0,183?1
Pu5Ir4
Pum- 4(c)
0,322?1
0,319?2
a = 7,245
-0,011?1
-0,007?1
b=14,60
Xi- 4 (c)
-0,041+2
-0,047?2
c=7,455
0,107?2
0,114?2
XII- 4 (c)
0,190?2
0,193?2
-0,356+2
-0,368+1
XIII - 8 (d)
0,194?2
0,183?1
-0,034?1
-0,035?2
-0,451?1
-0,460?1
'Density 13.59 and 16.54 g/cm3, respectively.
Translated from Atomnaya Energiya, Vol. 40, No. 6, pp. 495-498, June, 1976. Original article submitted
September 23, 1975.
594
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of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying,
microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $7.50.
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Interatomic Distances and Coordination of
Atoms in Pu5 Rh4 and Pu5 Ir4 Compounds
- Distance 1
PusRh i
PusIrd
Distance
Pu5Rh4
Pu5Ird
X,---Rh
X=Ir
X=Rh
X=Ir
Pili-Ptli
3,70 A
3,71 A
Pun-KR
3,24A
3,22 A
-2Pu1
3,85
3,86
-Xll/
2,91
2,97
-Pun
3,49
3,51
-.X///
3,03
3,12
=Pun
3,59
3,60
-XIll
3,06
3,08
-Pull
3,80
3,78
-Xll/
3,20
3.05
-Pum
3,32
3,31
Pu111-2Pu1
3,32
3,31
-Pum
3,41
3,39
-2Pu1
3,41
3,39
-XI
2,89
2,87
-2Pu1/
3,51
3,52
-XI
2,92
2,89
-2Pu11
3,52
3,54
-Xn
2,95
2,88
-XI
2,78
2,80
-XIII
2,73
2,82
-X/
3,18
3,08
-Xm
2,77
2,75
-XII
2,75
2,84
-XIII
2,98
2,90
-X/r
2,85
2,87
Pull-Pux
3,49
3,51
-2X1/
3,15
3,15
-NI
3,59
3,60
XI-2Pu1
2,89
2,87
-Pui
3,80
3,78
-2Pui
2,92
2,89
-Pau
3,75
3,81
-2Pu/1
3,16
3,22
-2Pun
3,78
3,76
-Pull'
2,78
2,80
-Pull/
3,51
3,52
-Pum
3,18
3,08
-Pum
3,52
3,54
-.XII
2,71
2,63
-X/
3,16
3,22
X/I-2Pu1
2,95
2,88
XII
2,99
3,05
-2Pu11
2,99
3,05
X11-2Pum
3,24
3,22
X11I-Pu1
2,98
2,90
-Pun/
2,75
2,84
-Pun
2,91
2,97
-Puni
2,85
2,87
-Pull
3,03
3,12
-XI
2,71
2,63
--Pun
3,06
3,08
XIII-Pur
2,73
2,82
-Pull
3,20
3,05
Pill
2,77
2,75
--Pull'
3,15
3,15
0 R'
A,B, c1111 c2D2 A282
Fig. 1. Projection of the structure of the Compounds Pu5Rh4 on the plane (001).
Table 1 shows the dimensions of the elementary cell, position of the atoms, atomic parameters, and the
theoretical density of the compounds.
Table 2 shows the jnteratomic.distances and coordinations of the atoms. The maximum error in deter-
mining the interatomic distances is ? 0.03 A.
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Fig. 2. Type AB layers in the structure of
the compound Pu5Rh4.
Fig. 3. Type CD layers in the structure of
the compound Pu5Rh4.
Figure 1 shows the projection of the structure of the compound Pu5Rh4 on the density (001). The struc-
ture consists of an alternation of plane layers of the type AiBI and A2B2, passing through the axis b perpendic-
ularly at a distance of V/2 from each other. Between these layers, two zigzag layers of type CiDi and C2D2
are disposed. Figure 2 shows the position of the atoms in the layers of type AB. Each atom of plutonium is
surrounded by a tetragon of rhodium atoms. The distances Pu ?Rh are 2.75, 2.78, 2.85, and 3.18 A. The tetra-
gons are linked between themselves by means of a common atom of rhodium and a short bond Rh?Rh = 2,71 A.
Figure 3 shows the arrangement of the atoms in the zigzag layer of type CD. Here, on the contrary, each
rhodium atom is surrounded by a tetragon of plutonium atoms. The distances Pu?Rh are 2.73, 2.91, 2.98, and
3.20 A. The tegragons are linked by means of a common atom of plutonium and the Pu?Pu bond is equal to 3.59
A. Bonds are achieved between adjacent layers of type CD because of the short distances, Pu?Rh = 2.77 A.
There are short bonds Pu?Rh = 2.89, 2.92, and 2.95 A between layers of types AB and CD. Thus, bonds within
the layers of types AB and CD, and also between the layers, are formed mainly because of the short distances
between atoms of plutonium and rhodium. The structure of the compounds Pu5Rh4 and Pu5Ir4 is close to the
structure of the compounds Sm5Ge4 [1] and Gd5Si4 [2], which have very much in common, although they belong
to different structural types [2].
One of the differences of these structures consists in that in the compound Gd5Si4, there is an additional
bond between atoms of Gdi and Gdip which there is not in the compound Sm5Ge4. This bond, however, is not
strong as the distance Gdi?Gdii is greater by 8% than twice the radius of gadolinium. In the structure of the
compounds Pu5Rh4 and Pu5Ir4 this distance Pur?Pull amounts to 3.95 and 4.02 A respectively, which is approxi-
mately 25% greater than twice the plutonium radius. Another difference between the structures of Sm5Ge4 and
Gd5Si4 consists in the presence near the atom of silicon (in Gd5Si4), and located in the type CD layer, of a short
bond with another atom of silicon from the adjacent layer. This distance is equal to 2.47 A. There is no simi-
lar bond in the Sm5Ge4 structure. This bond is absent also in the plutonium compounds. However, in the plu-
tonium compounds near the atoms of rhodium and iridium, and located in the type CD layer, there is a bond
with atoms of plutonium (Pun) from the adjacent layer, which is not present in the compounds with the rare-
earth elements. These distances Pull?RhIII = 3.03 A.and
LITERATURE CITED
3,12 A.
1. G. Smith, Q. Johnson, and A. Jharp, Acta Crystallogr., 22, 269 (1967).
2. J. Iglesias and H. Steinfink, J. Less-Common Met., 26, 45 (1972).
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NEWS ITEMS FROM THE COUNCIL FOR MUTUAL
ECONOMIC AID (CEMA)
DIARY OF COLLABORATION
Conference of Specialists on Problems of the Transportation of Spent Fuel Elements and the determina-
tion in them of the fissile material content by a nondestructive method took place on December 1-4, 1975 in
Moscow. Reports were heard from specialists from the German Democratic Republic, Poland, Romania, the
Soviet Union, and Czechoslovakia on thermal and stability calculations of the maximum constitution of gas
mixtures and an estimate of their explosion hazard, model studies of the service life of containers, the trans-
portation of spent fast-reactor fuel, experience in the transportation of spent fuel from the VVER-440 and KS-
150 reactors, and the determination of the extent of burnup and the content of fissile materials in them. Also
reported was the development of documentation for the transportation of spent fuel and the taking of special
measures for the prevention and elimination of possible consequential hazards. The timeliness, and high technical
and scientific level of these papers is noteworthy.
The specialists drew attention to the advisability of:
1. Further development and unification of transportation methods for the conveyance of spent fuel assem-
blies from the unified reactors of the member countries of the CEMA.
2. The development of recommendations for the conditions of transportation of spent fuel elements as a
function of the extent of burnup, time of cooling, and leak tighteness.
3. The improvement of procedures for thermophysical and stability calculations of transport containers
with their verification under test bench conditions.
4. The investigation of new structural materials for transportation containers.
5. The improvement of the method of y-scanning and methods based on the recording of natural and
induced neutron radiation for determining the depth of burnup and the content of uranium and transuranic
elements in spent fuel.
The conference.of.specialists considered and approved plans for technical assignments on the stages of
a working plan on the theme "The transportation of spent fuel elements and nondestructive methods of deter-
mining in them the content of fissile meterials" by 1976 to 1980, presented by the delegation from the German
Democratic Republic, the USSR, and CzSSR in the Council for Scientific and Technical Collaboration on the
processing of irradiated fuel from nuclear power stations.
The Conference of Secretaries of Delegations of the Member Countries of CEMA in the Permanent Com-
mission of CEMA on the utilization of nuclear energy for peaceful purposes took place on February 17-19, 1976
in Moscow.
The conference considered and provisionally ratified the report of the Permanent Commission on the work
carried out in 1975 and its future activity, and a plan of resolution of the Commission on the report. These
documents were referred by the delegations of the countries in the Commission for their agreement in the
working agenda.
The participants of the conference heard and discussed information from the division on the utilization
of atomic energy for peaceful purposes of the Secretariat of CEMA, concerning the path for accomplishing the
work plan of the Commission in 1976-1977.
An exchange of opinions also took place at the conference, concerning the problem of the manufacture of
materials and proposals in the departments of nuclear power generation toward a plan for a long-term purpose-
ful program of cooperation for guaranteeing the economics of the basic requirements of the member countries
of CEMA on the principal forms of energy, fuel and raw materials in the period up to 1990.
Translated from Atomnaya Energiya, Vol. 40, No. 6, p. 499, June, 1976.
This material is protected by copyright registered in the name of Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part
of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying,
microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $7.50.
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INFORMATION
THE I. V. KURCHATOV GOLD MEDAL COMPETITION
The Division of Nuclear Physics of the Academy of Sciences of the USSR reports on conducting the com-
petition for the I. V. Kurchatov gold medal of 1977 with financial prize. The gold medal is conferred for out-
standing work in the field of nuclear physics.
In the competition for the I. V. Kurchatov medal, papers or series of papers may be presented, on a
single thematic, as a rule, of individual authors. In presenting the collective papers, only the leading authors,
but not more than three, are to be mentioned. Papers, honored by the Lenin and State Prizes of the Soviet
Union, and also prizes of the Academies of Science of the Union Republics and Branch Academies, are not
accepted in the medal competition.
The right to put forward candidates for the gold medal competition is vested in the academicians and
corresponding members of the Academy of Sciences of the USSR and the Academies of Science of the Union
Republics; the scientific institutions; universities; scientific and engineering-technological societies; scientific-
technical councils of the state committees, ministries and departments; the technical councils of industrial
concerns; the constructional bureau; the Scientific Councils of the Academy of Sciences of the USSR, and other
departments concerned with the most important problems of science.
Organizations and individuals putting forward a candidate for the medal competition must present to
the Presidium of the Academy of Sciences of the USSR (Moscow, V-71, Lenin Prospekt, 14) with the inscrip-
tion "The I. V. Kurchatov Gold Medal Competion," the following information (each in triplicate):
1. Motivating presentation, including the scientific feature of the paper, its value for the development of
science and the national economy.
2. The published scientific paper (or series of papers), and data of scientific discovery or invention.
3. Information about the author (surname, Christian name, patronym, year of birth, academic degree
and title; list of principal scientific papers, discoveries and inventions; place of work and occupational function,
official and domestic address and telephone number).
Date of presentation of papers to the competition, up to October 12, 1976.
Division of Nuclear Physics, Academy of Sciences of the USSR. Translated from Atomnaya Energiya,
Vol. 40, No. 6, p. 500, June, 1976.
598
This material is protected by copyright registered in the?name of Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part
of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying,
microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $7.50.
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STILL NO "COSMION"
N. A. Vlasov
Recent investigations of the annihilation of antiprotons with a neutron have detected a certain unexpected
excess of y quanta [1]. In order to explain it, the authors have suggested that before annihilation a bound state
of the antiproton with one of the nucleons is formed, which they have also designated a "cosmion." The surplus
y quanta, according to the author's supposition, were emitted by the cosmion during transition from one state
to another.
The resounding title attracted attention to the paper, and reports on it were published in many journals
(e.g., [2]). Meanwhile, for conclusiveness, data were lacking about the spectrum of the surplus y quanta [3].
Recently, special investigations of the spectrum have been published [4]. Spectral lines were sought, which
should correspond to transitions between states with a quite long lifetime and, consequently, with a small
energy width. The result was negative ? the expected spectral lines were not detected. The discovery of the
cosmion was found to be premature. The existence of bound states of an antinucleon with a nucleon is not
surprising, as their interaction is stronger than pairs of nucleons. Even from the time of E. Fermi, attempts
have been well known to interpret pions and other mesons as bound states of a nucleon with an antinucleon.
The supposedly detected radiative transition proved to be surprising. It is well known in nuclear physics, that
in those cases when the excited nucleus can emit a nucleon (not very slow), radiative transition is of low prob-
ability. A nucleon pair plus an antinucleon in any state, has many channels of decay (annihilation) into adrons
by the action of a strong interaction. Therefore, the detection of a radiative transition should be very interest-
ing, if it should occur. So far this has not happened.
LITERATURE CITED
1. T. Kalogeroponlos et al., Phys. Rev. Lett., 33, 1635 (1974).
2. I. S. Shapiro, Priroda, No. 12, 68 (1975).
3. N. A. Vlasov, At. Energ., 39, No. 1, 73 (1975).
4. T. Kalogeroponlos, Phys. Rev. Lett., 35, 824 (1975).
Translated from Atomnaya Energiya, Vol. 40, No. 6, p. 500, June, 1976.
This material is protected by copyright registered in the name of Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part
of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying,
microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $7.50.
599
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CONFERENCE OF THE FOURTH COMMITTEE
OF THE INTERNATIONAL COMMISSION ON RADIOLOGICAL
PROTECTION (ICRP)
A. A. Moiseev
The regular session of the Fourth Committee of the International Commission on Radiological Protection
(ICRP) took place from December 1-6,1975 in Moscow, in the work of which nine of the 13 members partici-
pated. The only problem discussed at this conference was the project, Publication No, 29. It provides an ac-
count of the approach of the Commission to the normalization of the level of irradiation of professional workers
and the population. It is proposed that the new ICRP publication will be approved by the Principal Commission
in 1976 and issued in 1977 by Pergamon Press in English.
The Joint Session of the National Commission on Radiological Protection with Minzdrave USSR and the
Fourth Committee of the ICRP, held on December 6, 1975, took place in a businesslike and friendly environment.
In Moscow the foreign participants of the conference visited the Radiological Center of the Central Insti-
tute for the Advancement of Medicine. A large group of scientists departed for Leningrad, where the group
acquainted themselves with the principal trends of the work on radiation safety at the Leningrad Scientific-
Research Institute of Radiation Hygiene.
During the stay of the foreign scientists in the Soviet Union, meetings and discussions were held with
leading Soviet scientists, with the Chairman of the Fourth Committee of the ICRP,A.Jammet, the Scientific
Secretary of the Scientific Committee of the United Nations on the effects of nuclear radiation, D. Benninson,
and the chief of the division of radiation medicine of the World Health Protection Organization, V. Seelentag.
During these meetings and talks, plans were discussed for the further development of scientific research in
the fields of radiological protection, the participation of Soviet specialists in the work of the International
Association on Radiological Protection and the Fourth International Congress on Radiological Protection,
which will be conducted by the International Association in April 1977 in France (Paris).
Translated from Atomnaya t nergiya, Vol. 40, No. 6, p. 501, June, 1976.
This material is protected by copyright registered in the name of Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part
of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying,
microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $7.50.
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THE BETA-MIKROMETR- 3 DOUBLE -LAYER COATING ?
RADIOISOTOPE THICKNESS GAUGE
I. I. Kreindlin, V. S. Novikov,
A. A. Pravikov, and I. R. Rubashevskii
In the All-Union Scientific-Research Institute of Radiation Technology (VNIIRTe) production models of
the Beta-micrometr-3 radioisotope instrument have been manufactured (see Fig. 1), which enables noncontact,
nondestructive measurements of the thickness of double-layer metallic and nonmetallic coatings of complex
configuration to be carried out (as in Fig, 1) on small-scale fragile metallic and nonmetallic manufactured
articles, with a minimum monitoring area of 0.25 mm2. Beta-mikrometr-3 is designed on the basis of previ-
ously developed instruments, Beta-mikrometr and Beta-mikrometr-2 [1, 2]. In contrast from the first two
modifications, the instrument is provided with three interchangeable 13-emitting sources with a wide range of
boundary energies. In order to select the optimum source for monitoring a specified combination of materials
of coatings and backings, a procedure was developed for calculating the effect of variation of the layer thick-
ness, located above or below the layer being monitored.
The circuit of the DOT-3 primary transducer is achieved by integrated microcircuits, which increases
significantly the reliability of the instrument. The measurement time is recorded by small-scale duplexers
and can be set up from 1-1000 sec depending on the required monitoring accuracy. The Beta-mikrometr-3
instrument has an output at the BZ-15 pen recorder. In combined operation with the BZ-15, the measurement
result is printed out on a paper strip: in the first column? the serial number of the measurement and in the
second column ? the readings of the instrument signal panel. The instrument is powered from 220-V, 50-Hz
mains; power requirement is not more than 100 VA. Working range 10-35?C.
LITERATURE CITED
1. I. I. Kreindlin, V. S. Novikov, and A. A. Pravikov, At. Energ., 30, No. 5, 47 (1971).
2. T. G. Zvereva et al., At. Energ., 36, No, 5, 413.(1974).
Fig. 1. Beta-mikrometr-3 double-layer coating thickness gauge: 1) PU-2
scaler; 2) DOT-3 primary transducer; 3) BZ-15 pen recorder.
Translated from Atomnaya Energiya, Vol. 40, No. 6, p. 501, June, 1976.
This material LS protected by copyright registered in the name of Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part
of this publication may be reproduced, stored in a retrieval system, or transmitted, in any forth or by any means, electronic, mecheinical, photocopying,
microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $7.50.
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BOOK REVIEWS
A. M. Petros'yants
NUCLEAR POWER GENERATION*
Reviewed by Yu. I. Koryakin
To write books about nuclear power generation aimed at analyzing and reviewing the power-economic
and nuclear power generating situation in the countries of the world at present has become incomparably more
difficult than a few years ago. A nuclear power station was small then, the specific features and structural
characteristics were not so diverse as at present, and the problems solved by means of them appeared to be
easily examinable for the future. The quantitative and qualitative factors responsible for the connection be-
tween nuclear power generation and the surrounding medium simply did not enter into the order of the day, in
view of the negligibly small role of nuclear power stations in the energy supplies and power-economy of indus-
trially developed countries.
The position now has changed, but not so much in the qualitative aspect as in the quantitative aspect
which, on the whole, is characterized by nuclear power generation acquiring the qualities of a complex system
and strict complexity. Therefore, an attempt is hopeless at present and the method is unacceptable of the
simple listing and descriptive account of the engineering-design decisions of nuclear power stations which was
possible a few years ago.
The author of the book being reviwed, clearly understanding these difficulties, has undertaken the solu-
tion of the difficult problem of interpreting the complex picture of the state of nuclear power generation in the
world and, it would appear, has dealt with it successfully, and although the description of nuclear power sta-
tions and their structural features predominates in the book, the main consideration even within these limita-
tions is given to the typical properties of nuclear power stations, the development of tendencies in structural
and circuit decisions, constraints and origins of these decisions arising from the systematic aspects and
specific characteristics of the development of nuclear power generation in every country.
It is important that the author has not confined himself to consideration of nuclear power stations as
electric power generating facilities. The conditions, which shape the national demand on them and the neces-
sary scientific and production basis of nuclear power generation (nuclear research centers, nuclear mechani-
cal engineering factories, etc.) in the book have been an important element to demonstrate the systemic qual-
ities and complexity of the nuclear engineering industry. A description is referred here of the expected role
and position of nuclear power generation in solving the problem of the power demands of countries in the long-
term and also the participation of nuclear power stations and their special operating features in national
power systems.
Certainly, the systemic aspects of nuclear power generation recounted by the author are the results of
an analytical approach to the numerous publications in Soviet and foreign nuclear power literature. An analysis
of this literature, even in itself, is not a simple task and a discussion of the results of this analysis and gener-
alization of the data in the thematic plan of the book at the level, and in the key, which corresponds to the pur-
poseful task of the author is no less difficult.
On the whole, this book broadly and systematically informs the Soviet reader about the outside situation
and the state of nuclear power generation in the Soviet Union and abroad. Naturally, Soviet nuclear power
generation is afforded special consideration and among the types of nuclear power stations are nuclear power
stations with channel-type reactors. The latter, as is well-known, made an overwhelming contribution to the
introduction of the nuclear power capacities of the Soviet Union in the Ninth Five-Year Plan, and this situation
in the 10th Five-Year Plan will be maintained. Moreover, nuclear power stations with these reactors have the
*Nauka, Moscow (1976).
Translated from Atomnaya Energiya, Vol. 40, No.6, p. 503, June, 1976.
This material is protected by copyright registered in the name of Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part
of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying,
microfihning recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $7.50.
602
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greatest capabilities for the enlargement and concentration of individual outputs, which is particularly importanf
for the Soviet Union, whose power system is capable of accepting the extremely large outputs of the nuclear
power-generating units. Obviously, the large unified electric power of the units is an important factor for
accelerating the growth of the nuclear power station outputs, economy of financial means and labor resources
in the field of power reactor construction and nuclear engineering. These qualities of channel-type reactors
are emphasized by the author. It is only a pity that the author did not succeed in introducing into the book a
description of the intermediate modification of the RBMK-1500 channel-type reactor, recently defined and
brought on stream (intermediate between the RBMK-1000 and the RBMK-2000), achieved by the first nuclear
power station to be built with theee reactors in accordance with the current five-year plan ? the Ignalinsk
nuclear power station (Lithuanian SSR).
The scope of the review does not permit the information in the book concerning foreign nuclear power
generation to be dealt with extensively. Nevertheless, it should be noted that it leaves at the disposal of the
reader, rich factual and substantial reference data. The numerous photographs of nuclear power stations and
other nuclear facilities, both Soviet and foreign, are immensely valuable for illustrating and brightening up
the book.
Unfortunately, it has not been managed without individual slight errors, inaccuracies, or even obsolete
data. The reason for the majority of these, is that the rate of scientific-technological progress in nuclear
power generation has outstripped the period of writing of the manuscript of the book and its editing?publish-
ing progress.
On the whole, the issue of the book "Nuclear Power Generation" can be evaluated as an undoubtedly
positive event in nuclear scientific-technological literature.
603
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,
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