SOVIET ATOMIC ENERGY VOL. 42, NO. 5

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• Sequence Number: 5
• Publication Date: November 1, 1977
• Content Type: REPORT

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Russian Original Vol. 42, No. 5, May, 1977
? November, 1977
SATEAZ 42(5) 399-502 (1977)
SOVIET
ATOMIC
ENERGY
? ATOMHAR
(ATOMNAYA iNfRGIYA)
TRANSLATED FROM RUSSIAN
CONSULTANTS BUREAU, NEW YORK
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SOVIET
ATOMIC
ENERGY
Soviet Atomic Energy is abstracted or in-
dexed in Applied Mechanics Reviews, Chem-
ical Abstracts, Engineering Index, INSPEC?
Physics Abstracts and Electrical and Elec-
tronics Abstracts, Current Contents, and
Nuclear Science Abstracts.
Soviet Atomic Energy is a cover-to-cover translation of Atomnaya
Energiya, a publication of the Academy of Sciences of the USSR.
An agreement with the Copyright Agency of the USSR (VAAP)
makes available both advance copies of the Russian journal and
original glossy photographs and artwork. This serves to &crease
the necessary time lag between publication of the original and
publication of the translation 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:
Editor: 0. D. Kazachkovskii
Associate Editor: N. A. Vlasov
A. A. Bochvar
N. A. Dollezhal'
V. S. Fursov
I. N. Golovin
V. F. Kalinin
A. K. Krasin
V. V. Matveev
M. G. Meshcheryakov
V. B. Shevchenko
V. I. Smirnov
A. P. Zefirov
Copyright ? 1977 Plenum Publishing Corporation, 227 West 17th Street, New York,
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Consultants Bureau journals appear about six months after the publication of the
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SOVIET ATOMIC ENERGY
A translation of Atomnaya Energiya
November, 1977
Volume 42, Number 5 May, 1971
ARTICLES
Total 137Cs and 90Sr Contamination and External-Radiation Doses in the Territory
CONTENTS
Engl./Russ.
of the USSR ? L. I. Botneva, Yu. A. Izrael', V. A. Ionov, and I. M. Nazarov
399
355
Complex of Devices for Sampling and Measuring Tritium in Environmental Objects
? L. I. Gedeonov, V. A. Blinov, A. V. Stepanov, V. P. Tishkov, A. M. Maksimova,
and A. A. Antipov
405
361
Experimental Investigation of Algorithms for the Direct Digital Control of the Neutron
Field in an IRT-2000 Reactor ? E. V. Filipchuk, P. T. Potapenko,
A. P. Kryukov, A. P. Trofimov, V. G. Dunaev, N. A. Kuznetsov,
and V. V. Fedulov
409
365
Two-Zone System by Pulse Method ? B. P. Shishin, Yu. A. Platovskikh,
and T. S. Dideikin
414/
370
Total Reaction Cross Sections of Certain Metals and Gases for Very Cold Neutrons
? N. T. Kashukeev, G. A. Stanev, V. T. Surdzhiiski, and E. N. Stopyanova ... .. ?
?
417
373
Results of Testing Carbide Fuel Elements in the BOR-60 Reactor
? V. A. Tsykanov, V. M. Gryazev, E. F. Davydov, V. I. Kuz'min, A. A. Maershin,
V. N. Syuzev, I. S. Golovnin, T. S. Men' shikova, Yu. K. Bibilashvili,
R. B. Kotel'nikov, V. S. Mukhin, and G. V. Kalashnik
422
378
'Measurement of the Effects of the Reactivity of Materials in a Fast Reactor
? V. R. Nargundkar, T. K. Bazu, K. Chandramoleshvar, P. K. Dzhob,
and Rao K. Subba
428
383
The Effect of a High-Frequency Current in a Helical Winding on the Discharge in the
TO-1 Tokamak ? L. I. Artemenkov, N. V. Ivanov, A. M. Kakurin,
and A. N. Chudnovskii
432
387
X-Ray Detectors Based on Cadmium Telluride ? V. F. Kushniruk, L. V. Maslova,
0. A. Matveev, V. S. Ponomarev, S. M. Ryvkin, A. I. Terent'ev,
Yu. P. Kharitonov, and A. Kh. Khusainov
437
391
Use of Argon Ions for the Preparation of Nuclear Filters ? S. P. Tret'yakova,
G. N. Akap'ev, V. S. Barashenkov, L. I. Samoilova, and V. A. Shchegolev
441
395
DEPOSITED ARTICLES
"Some Reliability Aspects of Reactor Emergency Protection Systems
? A. I. Pereguda and A. A. Petrenko
445
398
Thermal Opening of Irradiated Uranium Oxide Fuel Elements of a BR-5 Reactor
with Fuel Separation by Dissolution ? G. P. Novoselov and S. E. Bibikov
446
398
Calculation of Differential Efficiency of Reactivity Controller by the Monte Carlo
Method ?Yu. P. Sukha.rev
447
400
Penetration of 'y Rays through Matter. Green's Function of Plane-Parallel Problem
with Azimuthal Symmetry ? L. D. Pleshakov
448
400
Neutron Transport in Half-Space with Sources ? V. P. Gorelov and V. I. Il'in
449
401
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CONTENTS
(continued)
Engl./Russ.
LETTERS TO THE EDITOR
Measurement of the Sensitivity of Neutron Detectors with Silver Emitter
during Long Service in Reactor - I. Ya. Emel'yanov, Yu. I. Volod'ko,
V. V. Postnikov, V. 0. Steklov, and V. I. Uvarov
451
403
Local Control of Profile and Magnitude of Energy Release of Loop Channels
- F. M. Arinld.n and G. A. Batyrbekov
453
404
Luminous Emittance of Neutron Beam in Air - A. V. Zhemerev, Yu. A. Medvedev,
and B. M. Stepanov
457
407
Effect of Pre-Irradiation of Oxidation of Alloy Zr + 2.5/1 Nb --- M. G. Golovachev,
V. I. Perekhozhev, V. E. Kalachikov, and 0. A. Golosov
459
409
Detection of Start of Boiling of Liquid Metal Coolant - K. A. Aleksandrov,
V. A. Afanas'ev, N. G. Gataullin, and V. V. Golushko
461
410
Direct Energy Conversion of Monoenergetic Ion Beams with Space-Charge Compensation
- 0. A. Vinogradova, S. K. Dimitrov, A. S. Luts'ko, V. M. Smirnov,
and V. G. Tel'kovskii
4-5_3_
411
Surface ,3-Activity of Soil and Vegetation Caused by Nuclear-Explosion, Products
and Its Dependence on the Vertical Migration of Isotopes - K. P. Makhon'ko
and A. S. Avramenko
465
413
Allowance for Fluctuations of Radiation Flux in Activation Analysis - Pham Zui Hien
467
414
Activation of Elements in (y , y') Reaction by "N y Rays - U. Akbarov, U. Uzakova,
and K. Umirbekov
468
415
Impact Toughness of Structural Graphite - Yu. S. Virgil'ev, V. V. Gundorov,
and V. G. Makarchenko
470
416
CsI(T1) Well-Detectors for Low-Background y Spectrometry - 0. P. Sobornov
472
418
COMECON CHRONICLES
**Third Symposium of COMECON Member-Nations on "Water Regimes, Water Treatment, and
Leak-Testing of Fuel Elements in Atomic Power Plants" - Yu. A. Egorov
and A. V. Nikolaev
477
422
INFORMATION
"Respiration" of the Sun - N. A. Vlasov
480
424
CONFERENCES AND MEETINGS
Franco-Soviet Seminar on "Conception of Atomic Power Plants, Technology,
and Operation of Water-Moderated-Water-Cooled Reactors" - V. A. Voznesenskii . .
481
424
Soviet-French Symposium on the Production and Application of Steel Piping in Industry
- G. V. Kiselev
484
426
Eleventh Conference on Energy Conversion and Research on Thermoelectronic Emission-
in the USA - V. A. Kuznetsov
485
427
International Meeting on Synthesis -of and Search for Transuranium Elements
- B. I. Pustyl'nik
488
429
Second IAEA Meeting on Large Tokamaks - L. G. Golubchikov
491
-431
Problems of Thermonuclear Equipment at the World Electrotechnical Congress
- V. F. Grishin and S. M. Sokolovskii
493
433
EXHIBITIONS
USSR Display at Electro-77 Exhibition - N. P. Longinova
495
433
BOOK REVIEWS
F. A. Makhlis. The Radiation Chemistry of Elastometers - Reviewed by E. D. Chistov
497
435
V. I. Vladimirov. Practical Problems of the Operation of Nuclear Reactors
- Reviewed by E. S. Glushkov
498
436
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CONTENTS
V. M. Gorbachev, Yu. S. Zamyatnin and A. A. Lbov. Interaction of Radiation
with Nuclei of heavy Elements and Nuclear Fission ? Reviewed by N. A. Vlasov
Yu. V. Kuznetsov. Ocean Radiochronology ? Reviewed by V. V. Gromov
The Russian press date (podpisano k pechati) of this issue was 4/25/1977.
Publication therefore did not occur prior to this date, but must be assumed
to have taken place reasonably soon thereafter.
(continued)
Engl./Russ.
499 436
500 436
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ARTICLES
TOTAL I92Cs AND 90Sr CONTAMINATION AND EXTERNAL-RADIATION
DOSES IN THE TERRITORY OF THE USSR
L. I. Boltneva, Yu. A. Izragl', UDC 614.876:632.15
V. A. Ionov, and I. M. Nazarov
Data on radiation background are basic to the establishment and application of criteria
for the safe use of atomic energy for peaceful purposes. The radiation background is deter-
mined by the radioactive contamination of a locality, the natural radioactivity of the en-
vironment, and cosmic rays.
In this article we consider the components of the radiation background for the territory
of the USSR. The actual data concerning the total 292Cs contamination and the natural radio-
activity were obtained from airborne y-spectral surveys [1, 2]. The errors of the airborne
measurements do not, for the most part, exceed 15-20%.
297Cs Contamination of the Soil and Vegetation Cover. The distribution of the surface
density (the concentration) of 137Cs for 1974 over the territory of the USSR is shown in Fig.
1. As is shown by the estimates of [2], the concentration of 192Cs varies only slightly with
time and will remain practically the same for the next few years. The average concentration
of 137Cs in the territory of the USSR amounts to 92 pCi/km2 (excluding the regions of high
mountains where no investigations have been conducted). The observed values of the concen-
tration amount in practice to 15-20 pCi/km2. The frequency distribution of the concentra-
tion values can be satisfactorily approximated by the normal law. The mean-square deviation
is 33 pCi/km2.
The distribution of I92Cs is characterized by marked differences between latitude zones.
Against the background of this overall rule, there is considerable spottiness in the distri-
bution of contamination levels, due to the local peculiarities of individual regions. The
maximum contamination levels are concentrated chiefly in the latitude belt from 50 to 60?
north. Here the 131Cs concentrations amount to 100-175 pCi/km2. North and south of this
belt the contamination levels decrease. The minimum value (25-50 pCi/km2) is found north of
70? and south of 45? north latitude.
In addition to varying with latitude, the contamination levels are higher for areas close
to the mountain systems: the Carpathians, the Crimean and Ural mountains, the Caucasian
ridge, the Tien-Shan, the Altai, and the Eastern Siberian and Transbaikal systems. East of
the Yenisei River, owing to the complex orography ? the presence of many mountain systems
occupying a large part of the territory ? the latitude-zone differentiation is less marked.
The latitude-zone distribution and the local deviations from it are closely related to the
observed patterns of precipitation distribution [2].
The distribution of the 37Cs concentration according to geographic and climatic zones
is shown in Table 1. These data indicate the connection between the contamination level and
the annual amount of atmospheric precipitation. A more detailed analysis has shown that the
relation between the 197Cs concentration and the annual average precipitation (P, mm) is ex-
pressed, for individual zones of the nonmountain areas of the country shown in Table 1, by
the following formulas:
Tundra and forest--tundra
Q =
0.25
+ 0.14P
Forest zone
Q =
0.3
+ 0.14P
Forest--steppe and steppe
Q =
0.45
+ 0.12P
Semidesert and desert
Q =
0.55
+ 0.35P
Translated from Atomnaya gnergiya, Vol. 42,
No. .5,
pp.
355-360, May, 1977. Original ar-
ticle submitted December 6, 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.
399
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400
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50?
TABLE 1. Surface Density (concentration) of 137Cs in Various Geo-
graphic Zone! ,
Zone
Soil, vegetation
Participa-
tion, mm/yr
Concen.,
pCi/km2
Tundra and forest?
Tundra gley and swampy soils
tundra
(mosses, lichens, bushes, low trees)
150-600
64
Coniferous forest
Podzol and swampy soils (coniferous
(taiga)
forests, swamps, meadows)
300-900
110
Mixed forest
Sod -podzol and gray forest soils
(coniferous and broadleaf forests,
meadows)
450-900
107
Forest? steppe zone
Chernozem, gray forest soils
(alternation of steppe massifs and
forest areas)
300-550
91
Steppe zone
Chernozem and chestnut-colored
soils (steppe vegetation)
300-400
87
Semidesert
Light chestnut and brown soils
(wormwood-type and halophytic
vegetation)
100-200
70
Desert
Gray-brown soils, serozems
(haloxylon)
75-150
59
60?
70?
80?
600 100? 140?
7.90?
700
60?
Isolines
80?
ryff: Mountainous regions
wr 120
Boundary of geographic zones
Fig. 1. Distribution of the concentration of 237Cs over the territory of the
USSR, 1.lCi/km2: I) tundra and forest--tundra; II) coniferous forest; III) mixed
-forest; IV) forest--steppe and steppe; V) semidesert; VI) desert.
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500 sr 70" 80" 60" 100' 140" 180"
60? 800 1000 120?
Fig. 2. 137Cs y-ray dose rate at a height of 1 m above the ground, pR/h (same
legend as in Fig. 1).
(here Q = q(P)/q is the ratio of the 137Cs concentration in a region with precipitation P to
the average concentration in the entire zone).
In the equations given above, the constant term is equal to the fraction of the concen-
tration due to dry fallout, and the regression coefficient characterizes the specific inten-
sity of washing out of radioactive aerosols by the atmospheric precipitation. The dry-fall-
out fraction increases from the northern to the southern regions (from 25 to 55%). In the
semidesert and desert zone the specific intensity of washing out of aerosols by the precipi-
tation increases sharply.
"Sr Contamination of the Soil and Vegetation Cover. An analysis of the ratio of "Sr
to 137Cs in the fallout [2] showed that the "Sr concentration can be found fromtheequation
qsr = 0.54
qcs with an error of no more than 10%.
We investigated specimens of different types of USSR soils. The ratio of "Sr to 137Cs
for gray forest soils, chernozems, and chestnut soils was 0.55 ? 0.12. For gray-brown soils,
serozems, and fixed sands it was 0.62 ? 0.15. The fairly large observed variations in the
ratio (25%) are due to the fact that the "Sr can be reliably determined in soil samples to
a depth of about 10 cm but is not taken into account when it migrates into deeper horizons.
The ratio in the soils is practically equal to the ratio in the fallout. Therefore the "Sr
concentration should be assumed to conform to the relation given for the fallout. The aver-
age "Sr concentration calculated in this way for the territory of the USSR is equal to 50
pCi/km2.
137Cs y-Ray Dose Rate. The distribution of the 137Cs y-ray dose rate at a height of
1 m for the territory of the USSR is shown in Fig. 2. In calculating the dose rate it was
assumed that in the virgin lands the penetration of 137Cs into the soil follows an exponen-
tial law. The penetration coefficient for 1972 was 0.6 ? 0.2 cm2/y. For this value of pen-
etration coefficient the dose rate at a height of 1 m is related to the 137Cs concentration
by the equation P (in pR/h) = 6.5.10-3*q (in pCi/km2). The variations in dose rate as a re-
sult of oscillations in the value of the penetration coefficient amount to 15%.
Between the 137Cs concentration map (see Fig. 1) and the dose-rate map (see Fig. 2)
there are certain differences, due essentially to the fact that in cultivated fields the
237CS has penetrated the topsoil. As a result, the dose rate drops by a factor of as much
401
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50? 60? 70? 80? 600 100? 140? 1800 70?
60 800 100 120?
Fig. 3. Dose rate of 'y rays from natural radioactive elements at a height of 1 m
above the soil, pR/h (same legend as in Fig. 1).
TABLE 2. Average y-Ray Dose Rate at a
Height of 1 m above the Soil
Soils
.
Dose
rate,
?RA
Coeff. of
variation
%
Character-
istic range
of variation,
?RA-
Tundra
3,5
43
2,0-5,0
Podzol
2,9
48
1,5-4,3
Sod-podzol

2,5-6,5
Gray forest soils
6,5
23
5,0-8,0
Chernozerns
7,1
21
5,5-8,6
Chestnut soils
7,6
20
6,1-9,1
Brown desert?
steppe soils
7,2
28
5,9-9,2
Gray-brown
desert soils
7,7
18
6,4-9,2
Serozems
8,6
18
7,0-10,2
as 2.5. Therefore, where there is intensive agriculture, owing to the large areas taken up
by plowed fields, the average dose rate is found to be considerably lower. The range of var-
iation of the dose rate is between 0.10 and 1.25 pR/h. The average value of the dose rate
when the plowed fields are taken into consideration is about 0.5 11R/h. As in the case of
the concentration, there is a distribution into latitude zones.
The frequencies of the dose-rate values have a logarithmic-normal distribution. The
marked asymmetry of the dose-rate distribution when there is a normal distribution of the
concentration is due to the large number of low dose-rate values for plowed areas.
Dose Rate of y-Rays from Natural Radioactive Elements. Natural radioactive elements
constitute a constantly present source of y rays. On the global and regional scales, the
external irradiation is affected, as a rule, by the radioactivity of the soils. Except for
mountainous regions, rock outcrops are limited in area, and therefore their y radiation is
local in character. The relative concentration of natural radioactive elements in soils is
402
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TABLE 3. Numerical Values of the TABLE 4. Distribution of Cosmic-Ray
Shielding Coefficient Dose over the Terriotry of the USSR
Geographic zone
i37cs
Natural
radioisotopes
Tundra and forest?tundra
0,40-0,80
0,60-0,90
Coniferous forest
0,50-0,85
0,70-0,95
Mixed forest
0,75-0,95
0,90-1,0
Forest?steppe and steppe
0,75-0,95
0,90-1,0
Semidesert
0,85--0,95
0,95-1,0
Desert
0,90-1,0
0,95-1,0
Dose, Mrd/yr
Exposure dose
rate,- ?R/h
Territorial
extent, ?Jo
28-30
3,65-3,9
60
30-40
3,9-5,2
33
40-50
5,2-6,5
5,5
50-100
6,5-13
2,0
100-150
13-19,5
0,5
found to lie mainly within the following limits: (0.5-3.5)4.10-4% for uranium, (1-14)?10-4%
for thorium, 0.3-3.0% for potassium. A characteristic of the spatial distribution of natural
radioactive elements in the soil is that on the regional level the relative concentrations
increase from north to south. The relative concentrations in rocks do not, for the most
part, exceed the maximum values of the relative concentrations in the soils. However, in
some varieties of rocks the uranium and thorium content may be greater by as much as one or-
der of magnitude than in the soils [1].
A map of the dose rate of y rays from natural radioactive elements in the territory of
the USSR, calculated on the basis of the uranium, thorium, and potassium content values nde-
termined by the aerial survey method [2], is shown in Fig. 3. The dose rate has a clearly
marked latitude-zone distribution. Genetic types of soils closely related to the geographic
and climatic zonality determine the spatial distribution of the dose field over the terri-
tory of the USSR (Table 2).
The frequency distribution of the dose rate within each genetic type of soil is satis-
factorily approximated by the normal law. The amount associated with the range of dose-rate
values (see Table 2) is 68%. This range was calculated on the basis of the normal distribu-
tion. There is a great deal of overlapping between the dose-rate values of different genetic
types of soils; this corresponds to the gradual transition from one type of soil to another
and is due to the effect of local factors, primarily the radioactivity of the parent rocks
and the mechanical composition of the soils.
Table 2 does not take account of soil radioactivity data for the regions east of the
Yenisei River. Here most of the territory is occupied by high plateaus and ridges which have
a high proportion of outcrops ('t,40%), so that it is difficult to distinguish the soil and
rock radioactivity values. The isoline contours in this region often correspond to the bound-
aries of individual rock complexes. The lowest radioactivity is found in intrusive rocks
of basic composition ? basalts, andesites, and their tuffs (1.5-2.5 pR/h). The maximum value
(9-14 pR/h) is found in areas of acid intrusive rocks (granites, granodiorites, quartz dio-
rites) and metamorphic rocks.
Attenuation of External y Radiation by Snow Cover. The snow cover can substantially re-
duce the external-radiation dose of field personnel and the general population. From the
data of 550 weather stations concerning the moisture content in the snow cover in the terri-
tory of the USSR, averaged over many years [3], maps have been prepared to show the attenua-
tion of the annual y-ray dose rate from 137Cs and natural radioactive elements [4].
The attenuation of the y rays is characterized by the dose-rate shielding coefficient
of the snow cover, K(x) = P(I, x)/P(I), where x is the moisture content of the snow cover;
x) and P(I) are the dose rates at a height of 1 m above the snow and above the soil,
respectively. In calculating the shielding coefficient, the nature of radioisotope penetra-
tion into the soil was taken into account. The limits of variation of K(x) over a 1-yr pe-
riod for the main geographic zones of the country are shown in Table 3. These changes char-
acterize the climatic nonuniformity of the zones considered. The greatest attenuation of the
annual dose values is found in the tundra and the forest--tundra. The annual dose value de-
creases by a factor of 1.5 for natural radioactive elements and 2.5 for 137Cs. In the south-
ern regions of the country the y-ray attenuation may be disregarded in practice.
Cosmic-Ray Dose. The cosmic-ray dose is caused mainly by the ionizing part of cosmic
radiation. The contribution made by neutrons for geomagnetic latitudes of 40-50? in the 0-4-
km altitude range is only 2-8% and is disregarded here.
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TABLE 5. External Annual Radiation in
the Territory of the USSR
Geographic
zone
Natural
radioac-
tivity of
soils and
rocks
137Cs
Cos-
mic
rays
Atmo
spher-
ic
radon
Total
dose
Tundra and
26,9*
3,3
28,3
1,5
50,0
forest-tundra
20,5
2,2
28,3
1,5
52,5
39
4
54
3
100
Coniferous
27,7
5,5
28,5
4,2
62,9
forest
22,8
4,0
28,5
1,2
56,5
40
7
51
"2
100
Mixed forest
41,6
4,2
28,6
2,3
76,7
37,4
3,4
28,6
2,3
71,7
52
5
40
3
100 '
Forest-steppe
56,2
3,7
29,1
3;0
92,0
and steppe
50,5
3,0
29,1
3,0
85,6
60
3
34
3
100
Semidesert
56,2
4,0
29,3
4,5
94,0
53,3
3,6
29,3
4;5
90,7
59-
4
31
5
100
Desert
57,7
2,7
29,1
4,5
94,0
57,7
2,7
29,1
94,0
61
3
31
5
100
*The first line gives the average dose value in Mrd/
yr; the second, the dose value taking account of the
snow-cover shielding. Mrd/yr; the third line, the
contribution to the total dose (taking account of the
snow-cover effect), lo
The annual doses for points which have a geomagnetic latitude of co and an altitude of H
above sea level can be calculated by the approximate formula [5]
D(q), H)= a ((p)? b exp [Hie OM.
If D (yo, H) is expressed in megarads per year and H in kilometers, then b = 11,
1 15 for
a(p) 15+ 0.118 (q)- 25?) for 25 < q) < 42?;
? 17 for q) > 42?;
1.96 for q) < 10';
c( ) 1.96 exp [ - 0,0028 (q) --10?)]
q)
for
1,75 for 50?.
The formula-given above enables us to estimate the annual doses With an error of no
more than 5% in the 05-km altitude range.
The distribution of the cosmic-ray dose over the territory is characterized by the data
of Table 4. An isodose map for the territory of the USSR is given in [5]. For about 60% of
the territory of the USSR, including the European part of the country (except for the Car-
pathians and the Caucasus) and the Western Siberian plain, the dose is 28-30 Mrd/yr, while
the eastern mountainous part of the country has a dose of 30-50 Mrd/yr. The highest dose
-values Are found-in the area- of the Pamir-Tien-Shan mountain system and the Caucasus, where
the dose exceeds 100-150 Mrd/yr, with a maximum value of 200-350 Mrd/yr (u700 Mrd/yr on Lenin
Peak, Victory Peak, and Communism Peak). The dose rate, may differ from the average by 3-10%
because of variations in cosmic radiation.
Contribution Made by Various Sources to External Radiation in the Territory of the USSR.
The maps shown above enable us to estimate the contribution made by y-radiation from 137Cs,
natural radioactive elements, and cosmic rays to the external-radiation dose in the terri-
tory of the USSR.
In addition, y rays-are also emitted by radon decay products found in the atmosphere.
In order to take this factor into account,, we may assume that the average emanation coeffi-
cient is 10% and that the-concentration of decay products in the air decreases exponentially
with altitude and decreases by a factor of 2 at an altitude of 1 km [1].
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The annual doses of external radiation for the geographic zones of the USSR are indi-
cated in Table 5, from which it can be seen that the total dose of external radiation in-
creases from the tundra to the desert zone by a factor of almost 2: from 52 to 94 Mrd/yr.
The contribution made by the different components varies regularly for the different zones.
Thus, in the first two zones most of the annual dose is due to cosmic rays. In the other
zones the contribution of natural radioactivity is predominant. In the desert zone it is
one and a half times as high as in the tundra and the coniferous forests. This difference
is additional proof of the fact that the peaceful use of atomic energy has only a slight ef-
fect on the annual dose.
LITERATURE CITED
1. R. M. Kogan, I. M. Nazarov, and Sh. D. Fridman, Fundamentals of Gamma Spectrometry of
Natural Environments [in Russian], Atomizdat, Moscow (1976).
2. L. Boltneva et al., in: Collection of Reports of the First Radioecological Conference,
Star.); Smrkovec, Czechoslovakia [in Russian], Vol. 3 (1972), p. 95.
3. Handbook of the Climate of the USSR, Part IV, Nos. 1-34 [in Russian], Gidrometeoizdat,
Leningrad (1965-1970).
4. L. I. Boltneva, L. I. Kuznetsova, and I. M. Nazarov, in: Proceedings of the Institute
of Applied Geophysics [in Russian], Gidrometeoizdat, Leningrad, No. 34 (1977), p. 57.
5. L. I. Boltneva, I. M. Nazarov, and Sh. D. Fridman, Izv. Akad. Nauk SSSR, Fiz. Zemli,
No. 4, 66 (1974).
COMPLEX OF DEVICES FOR SAMPLING AND MEASURING TRITIUM
IN ENVIRONMENTAL OBJECTS
L. I. Gedeonov, V. A. Blinov,
A. V. Stepanov, V. P. Tishkov,
A. M. Maksimova, and A. A. Antipov
UDC 546.110.23.002.637:543.05
Among the contaminants of technical origin in water systems are many which remain in
the dissolved (ionic) state for a long time. Tritium may serve as an analog of their dis-
persal with bodies of water. At present, the tritium content in waters is sufficient for
studying migration. Such research makes it possible to predict the behavior of industrial
effluents in regions where new facilities are planned.
The use of nuclear energy has resulted in a marked global increase in the concentration
of tritium over the past decades in almost all water systems, in rainfall, and in surface
water. Thus, the mean monthly concentration of tritium reached 130 tritium units (t.u.) in
U.S. river basins in 1966 [1], from 60 to 360 t.u. in various U.S. rivers in the winter of
1971-1972 [2], an average of 150 t.u. in the rivers of the prefectures of Japan in 1971 [3],
and about 100 t.u. in the northern rivers of the USSR in 1972 [4].
Owing to dilution, the tritium concentration in large bodies of water is appreciably
lower. Highly sensitive techniques must be employed to determine it. Thehighestsensitivity
is ensured by methods employing scintillation and proportional counters. Analysis of work
on tritium determination in natural bodies of water shows that when enrichment is employed
the reproducibility of results is not good since there are no reliable methods of determin-
ing the degree to which the sample has been enriched with tritium if the initial concentra-
tion was low. An advantage of gas-filled proportional counters is that they ensure high sen-
sitivity (without enrichment) and make absolute measurements.
Translated from Atomnaya Pnergiya, Vol. 42, No. 5, pp. 361-364, May, 1977. Original
article submitted August 11, 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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Fig. 1. Unit for regen-
eration of silica gel: 1)
water collector; 2) con-
denser; 3) holder with
silica gel; 4) oven; 5)
thermocouples; 6) vacuum
cleaner.
Condematd
Apparatus for determining tritium in samples of water and atmospheric moisture were de-
veloped and described earlier [5]. Subsequently work was focused on improving the method.
The tritium concentration in hydrogen and water is determined in several stages: samp-
ling of water or trapping of atmospheric moisture; oxidation of atmospheric hydrogen; obtain-
ing hydrogen by decomposition of the sample; purification and introduction of the hydrogen
into the counter with propane or incorporation of the hydrogen into the counting gas by means
of synthesis; and measurement of the activity of the gas filling the counter.
The range of tritium concentration that can be measured is extremely wide. Accordingly,
the complex of apparatus consists of independent units of two classes: the first, for mea-
suring samples of "high" activity (with a tritium content exceeding 3?10s t.u.) and the sec-
ond, for low activities (from 3.10s to 30 t.u.).
The previous method used to prepare samples for measurement had the following short-
comings: small quantity of hydrogen obtained, long time required to obtain it (,2 h), and
a rather low degree of purity of the hydrogen which affects the process of synthesis and the
results of the measurement.
In the new variant, samples of water are taken in glass ampules of up to 250 ml in vol-
ume which are sealed at the site of the sampling. This ensures long storage of samples, ex-
cludes the possibility of samples becoming enriched with tritium as a result of exchange with
external sources, and if necessary permits measurement of further portions of water from the
same sample. As in the previous variant, atmospheric moisture is trapped in previously de-
hydrated silica gel. It consists of an aluminum holder with two apertures, containing 2 kg
of granulated silica gel. The air is drawn through an absorber by a vacuum-cleaner at a
rate such as not to allow water vapor to pass through. The principal parameters of the ab-
sorption unit are as follows: height of silica gel layer in absorption unit 10 cm; area of
absorber 300 cm2; air flow rate 0.5 ms/min; resistance of absorber layer 30 mm Hg.
The sampling time is determined by the air temperature and humidity and is chosen so
that the silica gel absorbs 250 ml of atmospheric moisture. The atmospheric moisture is ex-
tracted from the silica gel and the gel is regenerated by a stream of hot air circulating in
a closed system (Fig. 1). Then, in preparing the sample for measurement the required quan-
tity of water is mixed with calcium oxide and put into a stainless steel "reactor." When
samples of water of high specific activity (above 3000 t.u.) are to be measured to obtain
hydrogen a 0.2-ml sample of water is placed in a reactor consisting of a test tube of heat-
resistant glass. To measure samples of low specific activity 20 ml of water is placed in the
reactor.
When the reaction CaO + H20+ Ca(OH)2 is completed, powdered zinc is added to the reac-
tor. Completion of the reaction Ca(OH)2 + Zn + ZnO + CaO + H2 excludes any isotopic effects.
The time needed to obtain hydrogen is reduced to 5-10 min. Then the hydrogen is passed
through a filter (a layer of silica gel and activated carbon), cooled by liquid nitrogen.
Such purification enables us to obtain hydrogen free of impurities which poison the catalyst
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\
\
\
N
\ . ..
Stockholm
\ k
_ \
N
. 50
070
055 N, ' \
s.?N.
o"
*

;TM.lin
'
\
\
.4
.
. 30
00
om
30
046
90
?g
otx
\
\
'
Copenhagen
. of,5
060m,
41V.
55
Gdansk.
o60
altiisk..\,-?
Klaipeda
.4
Kaliningrad
\
\
Fig. 2. Tritium concentration (in t.u.) in samples of water
from the Baltic in 1975.
during synthesis of the filling gas for the counter and free of possible radioactive inert
gases. Purified hydrogen is used in a mixture with propane to fill counters in measuring
samples of high specific activity. When low concentrations of tritium are measured, the
hydrogen enters into the synthesis system where it becomes part of the filling gas, butane
The latter is synthesized with the aid of a palladium catalyst from a mixture of butadiene
and hydrogen obtained from the sample by the method described earlier [5]. The high purity
of the gas ensures good reproducibility of the measurements.
The arrangement for measuring high tritium concentrations uses three proportional coun-
ters, differing only as to volume or form of shielding against the radiation background. Two
of the counters have only passive shielding in the form of a 10-cm layer of lead and differ
only in length and, consequently, in volume (302 and 604 cm3). Such a combination is used
in order to exclude the fringe effect. Hydrogen obtained from a single sample is introduced
into both counters, up to the same partial pressure. Measurements in the counters yield the
counting rates n1 and n;, respectively, for volumes VI and V2 of hydrogen measured. The in-
strumental 0 spectrum recorded by a multichannel analyzer is used to introduce corrections
for incomplete counts of the low-energy part of the spectrum and corrected counting rates n1
and n2 are obtained instead of n1 and n;. Upon comparison with the absolute value of the ac-
tivity due to the fringe effect, both counting rates are reduced by the same quantity and
their difference is equal to the difference of the absolute values, whereby ao = (n2 ? n2)/
(V, ? V2), where ao is the volumetric specific activity of the hydrogen from tritium. The
third counter, of volume 302 cm3, is inside a passive shielding and is surrounded by a ring
of Geiger--Mueller counters operating in anticoincidence mode with the main counters. The
efficiency of the latter is evaluated with a sample of known activity and was found to be
'\,90% for standard measuring conditions.
More sensitive arrangements with proportional counters of volumes 3.1, 5, and 8 liters
are installed in underground premises with shielding with a water equivalent of 120 m. Under
these conditions there is no need of protection with the aid of an anticoincidence system.
The mean background of the 3.1-liter counter is 11.1 ? 0.4 counts/min (with a confidence co-
efficient of 95% the deviation of a single measurement of the background is ?1.2 counts/min).
A count that exceeds the background by 3 counts/min, which corresponds to q,30 t.u., is con-
sidered a reliable result. Signals from the counter anode are amplified, sorted according
to amplitude, and recorded. The entire apparatus is based on transistors and integrated mi-
cromodules. The schematic diagram of a similar arrangement was described earlier [5]. The
background of the 302-cm3 counter, with active shielding, is 3.4 counts/min.
The reproducibility of measurements on both arrangements was verified by repeated deter-
minations of the tritium content of the same sample. The high-sensitivity 3.1-liter arrange-
ment is characterized by a deviation of ?20% for a single measurement with a confidence
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600
Sevastopor
70.
Fig. 3. The tritium concentrations (in t.u.) in samples of
water from the lower course of the Danube and the northwestern
part of the Black Sea.
coefficient of 95%. The arrangement for determining higher activities (with a 302-cm9 coun-
ter) gives a deviation of ?8% with the same confidence coefficient.
The possibilities of ascertaining the spread of impurities in the environment with the
aid of the apparatus described above were evaluated during the course of work on the Baltic
Sea and in the lower courses of the Danube. Samples of water were taken at 20 points on the
Baltic in 1975. After purification by a single distillation without enrichment, the tritium
content of the samples was determined on a sensitive counter installed underground.
The sampling sites and the results obtained are shown in Fig. 2. The mean value of the
results of all determinations was 56 ? 20 t.u. with a confidence coefficient of 95%. This
differs little from the results of a 1972 analysis of five samples of water from the Baltic
[4]. The results range from 30 to 90 t.u.
With a confidence coefficient of 95%, the deviation of a single result from the mean is
65% which is more than 3 times the spread characterizing the reproducibility in the deter-
mination of the lowest tritium concentrations and, as mentioned above, is ?20% for the same
confidence coefficient. Only nine results obtained for Baltic samples deviate from the mean
value by less than 20%, whereas all the others deviate more. Thus, with a confidence coef-
ficient of more than 95% it can be said that these deviations reflect the variability of the
concentrations in the Baltic and not the random error of the method.
The points at which the tritium concentration deviates significantly from the mean were
not observed to be distributed systematically. This means that no effect of any local sources
of tritium was noted. This is probably due to the fact that water from the land drains into
the Baltic in various regions, freshening the seawater quite evenly. One may speak of some
tendency towards lower values near sounds.
The picture is different at the mouths of the large rivers falling into the Black Sea
(Fig. 3). The waters of the Danube bear a significant concentration of tritium, which is
roughly the same over its entire lower course, surveyed over the Kilya branch. These levels
are comparable to the tritium content in present-day atmospheric precipitation. In 1969-
1970 the mean tritium concentration in precipitation was, e.g., 244 t.u. in Tbilisi, 247 t.u.
in Odessa, and 133 t.u. in Rostov-on-Don [6]. The large mass of Danube waters is not diluted
immediately, but only at a considerable distance from the delta is the concentration observed
to drop to a minimum value which can be measured by our method without enrichment. Safe con-
centrations of tritium are involved here but the picture obtained here should also be char-
acteristic of other dissolved impurities borne by the waters of the Danube, some of which may
be toxic.
It is interesting that with a much smaller discharge from the Dnepr and the Bug the in-
crease in the tritium concentration in the waters of the Black Sea near their mouths is traced
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for a much smaller distance than at the mouth of the Danube. However, a halo of impurity
spread is traced in this case, too.
The values of the tritium concentration in the Black Sea and the Danube cover a wide
range. To take the mean in this case is to lose all the information about the differences
in various parts of the halo of tritium spread. As mentioned above, the reproducibility of
the results of measurements with the system described is characterized by an rms deviation of
?20%. This criterion can be used to evaluate the statistical reliability of the differences
in the results of tritium concentration measurements in the Danube and the Black Sea. Almost
all the differences observed are real.
Notwithstanding the increased tritium content in atmospheric precipitation and surface
waters, there are still many vitally important objects with a tritium content that is mea-
sured in several t.u. or even fractions of a tritium unit. A serious problem is posed by the
penetration of soluble poisonous impurities into the depths of the oceans and into subter-
ranean waters. The study of these objects and processes calls for extremely sensitive tech-
niques for tritium determination. It can be assumed that techniques employing proportional
counters will prove useful.
LITERATURE CITED
1. M. Chesnutt et al., Radiol. Health Data Reports, 7, 377 (1966).
2. Radiation Data Reports, 13, No. 5, 8 (1972).
3. Nat. Inst. Radiolog. Sci., Chiba, Japan, NIRS-RSD, 34, No. 34 (1972).
4. T. N. Zhigalovskaya et al., in: Meteorological Aspects of Radioactive Pollution of the
Atmosphere [in Russian], Gidrometeoizdat, Leningrad (1975), p. 223.
5. L. I. Gedeonov et al., in: Proceedings of the IAEA Symposium: "Environmental Surveil-
lance around Nuclear Installations," Vienna (1974), Vol. 1, p. 235.
6. V. N. Soifer et al., in: Polluted Natural Media [in Russian], No. 3(42), Gidrometeoiz-
dat, Moscow, p. 85.
EXPERIMENTAL INVESTIGATION OF ALGORITHMS FOR THE DIRECT
DIGITAL CONTROL OF THE NEUTRON FIELD IN AN IRT-2000 REACTOR
E. V. Filipchuk, P. T. Potapenko,
A. P. Kryukov, A. P. Trofimov,
V. G. Dunaev, N. A. Kuznetsov,
and V. V. Fedulov
UDC 621.039.562
With the development of nuclear energy, the problem of the use of a computer for control-
ling the neutron field of large nuclear reactors has become increasingly urgent. Significant
attention has recently been devoted to this problem. A review is given in [1] of methods of
utilizing computers for reactor control in a number of foreign atomic power plants, and the
problems of the synthesis of similar systems are discussed in [2]. It is being planned to
use computers on a research reactor for the control and recording of the spatial distribu-
tion of the neutron field [3].
This article is devoted to an experimental investigation of various algorithms for the
direct digital control of the neutron field of the Moscow Engineering Physics Institute IRT-
2000 reactor. The legitimacy of a study of the distinctive features of spatial regulation
on a research reactor has been shown in [4]. Research reactors can be successfully used as
a physical model in the conduct of such experiments, control and processing problems can be
Translated from Atomnaya gnergiya, Vol. 42, No. 5, pp. 365-369, May, 1977. Original
article submitted July 26, 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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1?TH 2
1 IK
Kt*
PPie
K2.
PI
16
Hi
3-14
17
18
19
5
21
20
Fig. 1. Structural layout of the IRT-2000 reactor experimen-
tal control system: 1, 16) operating mechanisms with the log-
ic unit; 2, 17) EMU-3A power amplifiers; 3, 18) TsAP-11/5
digital?analog converters; 4, 12) 1-37 measuring amplifiers;
5, 13) passive correction circuits; 6, 14) 1-102 normalizing
amplifiers; 7, 15) F-203 voltmeters; 8) UR-8 mismatch ampli-
fier; 9) integrating measuring amplifier; 10) Shch-1413 volt-
meter; 11) standard automatic regulator; 19) input?output
channel; 20) Nairi-S processor; and 21) "Consul" typewriter.
solved with their help, and experimental checks of the various algorithms and control system
structures are made on the basis of real equipment [5].
The present trend in the application of computers for reactor control consists of the
use of modular interconnected computer systems, which permits increasing the reliability and
flexibility of the digital control system. Safety requirements result in the necessity of
utilizing specialized computers which execute simple control functions but also possess high
reliability at the lowest level of the automated control system hierarchy ? direct digital
control of the reactor. The legitimacy of such an approach is explained again by the fact
that it is necessary to regulate the total power with very small cycles, which essentially
excludes the possibility of using complex control laws.
It has proved possible on the basis of these ideas to use a Nairi-S computer for the
experimental investigations. In order to couple similar input?output subsystems with the
processor, a special logic unit for input?output control was constructed on the basis of in-
tegrated circuits; this unit was connected to an integrator by means of a coincidence cir-
cuit built with transistors. Such construction of the channel for input?output of informa-
tion into the processor of the Nairi-S has permitted, without changing the command system of
the machine, realizing control by analog?digital and digital?analog conversions, entry into
memory, and the reading of information necessary for control. The indicated coupling unit
permits providing for the input?output of discrete information over three channels, and,
when necessary, over a. larger n1.10er of channels with small additions, which significantly
increases the computational power of the machine.
The complete structural.layout of the experimental system for direct digital control and
a diagram of the active zone of the IRT-2000 are given in Fig. 1. Type DPZ-11P emission de-
tectors, which were successfully used earlier in experiments with automatic reactor control
systems [4, 6, 7], arranged in expelled fuel elements were applied in the system as the neu-
tron flux detectors D1 and D2. The boron rods K1 and K2 were used as actuating members of
the control system, and the rods PP,. and PP2 were used to disturb the reactivity. In the
course of carrying out the experiment on direct digital control of the reactor, the total
power was stabilized by a standard automatic regulator (AR). Commercial measuring devices
and converters, incorporated into the corresponding analog input and output subsystems, were
used for connecting the reactor to the computer.
The analog input subsystem of the single control channel included a DPZ-11P detector,
an 1-37 current-measuring amplifier with a recording device, an 1-102 normalizing amplifier,
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1
2
4
6
cb
7
9
8
Fig. 2. Structural layout of the control program;
1) timer; 2) "Input" subprogram; 3) necessary in-
put of new settings; 4) subprogram for calculation
of the control actions; 5) check on violation of
constraints; 6) "Output" subprogram; 7) digital?
analog converter interrupt; 8) subprogram for in-
put of settings; and 9) forming of the control ac-
tion.
and an F-203 digital voltmeter, which was used as the analog?digital converter. A passive
correction circuit was included between the measuring and normalizing amplifiers [6].
The analog output subsystem consisted of a TsAP-11/5 three-channel digital?analog con-
verter, an EMU-3A power amplifier, a logic unit, which provides for different operating con-
ditions of the motor, and a standard servo.
The total power imbalance signal, picked off the UR-8 mismatch amplifier of the standard
AR system, was fed into the processor over a channel consisting of an integrating measuring
amplifier and an analog?digital converter on an Shch-1413 voltmeter base. Input of infor-
mation on the relative position of the AR rod was thereby provided.
The structure of the control program, which consists of several function subprograms,
is given in Fig. 2. The subprogram "Input" interrogates the appropriate analog?digital con-
verter, converts the double-decade output code of the voltmeter into the auxiliary code used
by the machine, and enters the measured value into memory. The subprogram "Output" provides
for control of the digital?analog converter ("zero" default, addressing) and converts the
information applicable for output to the form necessary for the normal functioning of the
converter.
A universal information-processing program calculates the control actions according to
various algorithms and systematically alters the structure of the control system and the
control law parameters. The functioning of the control program occurs in the following way.
Indexing of the subprogram "Input" occurs according to a signal from the counter, which simu-
lates the operation of a timer. After conclusion of input and the conversion operations con-
trol is transferred to the settings input program. If according to a requirement of the oper-
ator the input of new settings is necessary, an interrupt occurs in the digital?analog con-
verter and the machine proceeds to the readiness state.
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414
101
wo
99
98
15
20
25
1 2
3 4
DPZ-2
t,
H, cm
Fig. 3. Variation
DPZ-2 currents and
the control rods H
compensating for a
bance.
in
of the DPZ-1 and
the positions of
in the course of
reactivity distur-
An adjustment is set by the operator from a console (a "Konsul" typewriter), after which
control is transferred to the information-processing unit for calculation of the control ac-
tion. The possibility of changing the control law form and parameters is similarly provided
for in the program. Further, control is transferred to the constraint-checking unit, and
finally the action decided upon enters with the aid of the subprogram "Output" the input of
the appropriate digital?analog converter. Then control is transferred to the timer, and the
cycle is repeated during the cycling time. The memory code of the Nairi-S machine was used
as the master program.
One should emphasize that such hardware and software design permits successfully simu-
lating the operation of a real control computer in combination with a real configuration of
analog input and output subsystems.
The experiments conducted were concerned with investigations of system dynamics, the de-
termination of adjustments, and checking the efficiency of the system under different standard
'disturbances. Linear and relay systems with in-phase and out-of-phase channel operation were
similarly investigated with different control laws. As a result, the following basic quan-
titative characteristics have been obtained:
1. At a power of 2000 kW the current of each DPZ was approximately 2 pA.
2. The signal at the input of the F-203 analog?digital converter after correction and
the 1-102 normalizing amplifier was hundreds of millivolts, which made it possible to operate
stably at the voltage limit of the voltmeter with three significant figures. At the same
time stabilization of the DPZ currents with an error of 0.5% was ensured.
3. The analog output subsystem was adjusted so that based on safety ideas the rate of
reactivity introduction was restricted to a value of 0.03 0/sec for the maximum control sig-
nal. The cycling time was 1 sec.
4. The active zone of the reactor possesses well-expressed spatial effects. Thus, upon
a disturbance of the reactivity by the movement of the rod PPi upward by 5 cm, the current
of DPZ-1 increased by 1.5-2%, and the current of DPZ-2 decreased by approximately the same
amount. When rods K1 and K2 were simultaneously moved in different directions (the total
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/04
103
102
10!
WO
99 Fig. 4. Controlled alteration
98 of the neutron field profile.
97
96
95
power is constant), it proved possible to achieve variations of 10-15% in the DPZ current.
Since the action of the standard power regulator compensates for the fundamental har-
monic of the neutron field distribution, the dynamical properties of the direct digital con-
trol system are basically determined by the time constant of the emission detectors. Experi-
ments with analog correction according to the procedure of [61 and with digital correction
of the detectors (a correction based on the first difference was used, which is equivalent in
continuous systems to a correction based on the derivative) showed that with proper adjust-
ment the dynamical characteristics of the system are practically unchanged. However, when
digital correction is used, it is necessary to take into account the additional load of the
machine memory, which is especially significant when there is a large number of detectors.
The relay system with in-phase channel operation was selected as the base system after
comparative analysis of the different structures for direct digital control. This structure
possesses a number of advantages, among which one should note the simplicity of the analog
output system, the great flexibility, and the simplicity of adjustment, which permits execut-
ing practically any control algorithms.
One of the most important problems in reactor control is regulation of the total power.
It is possible that the necessity, for safety reasons, of an analog power regulator does not
diminish for a reactor equipped with computer control. In this case one of the functions of
the computer in the control process should be maintaining the AR rods in the most effective
position. The following algorithm is a possible way to solve this problem. An error signal
is formed in each channel from the sum of the local error, and it is possible to assume to
the accuracy of the conversion processes that this signal is proportional to the position
of the AR rods.
A transitional process affiliated with the compensation for the disturbance in the reac-
tivity caused by the movement of rod PP1 upward by 5 cm characterizes the operation of this
system (Fig. 3). First the redistribution of DPZ currents is determined by the rapid-acting
total power regulator. Further, a synchronous movement of rods K1 and K2 returns the AR rod
to its original position, and in conclusion a redistribution occurs in the positions of the
zone regulation rods in order to obtain the required field shape.
One should include high reliability and mobility in controlling the total power among
the merits of this algorithm; the necessity of exact agreement of the settings which deter-
mine the field shape with the total power setting is a shortcoming. An algorithm which in-
cludes the normalization of the local signal of the neutron field to the average level is
free of this disadvantage. In accordance with it a local error signal is determined by com-
paring the reference value of the field shape with the measured value. The main advantage of
the algorithm consists of the independence of the error signal from the power. It has been
experimentally confirmed that with appropriate adjustment of the channels a variation in the
total power does not result in the actuation of the rods K1 and K2.
One should emphasize that in contrast to the schemes discussed earlier the algorithm
which includes normalization of the error signal does not ensure regulation of the total pow-
er; thefefore, the presence of an automatic power regulator becomes a necessary operating
condition. In this case the introduction into the algorithm of the regulation of a signal
proportional to the integral of the imbalance signal of the total power permits synthesizing
a structure combining the advantages of the systems discussed earlier. Actually, this al-
gorithm essentially ensures stabilization of the total power based on a signal from the mis-
match amplifier of the standard regulator, maintaining at the same time a specified shape
of the neutron field according to signals of the appropriate intrazone detectors. The power
required for a specified field shape is determined by a unique setting.
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It is important to note that the presence of automatic power regulator rods is not com-
pulsory. An experimental check of the efficiency of a direct digital control system without
a regular AR has shown that in this case acceptable quality of the processes is provided both
under stabilization conditions (compensation of local disturbances) and in the case of a con-
trolled redistribution of the neutron field. As an example, the transitional processes in
the case of the controlled alteration of the field profile ? the specification of heteropolar
settings ? are presented in Fig. 4.
The experiments conducted are only the first step in direct digital control of the neu-
tron field of a reactor. However, the creation of new high-reliability installations with
a computer operating in real time and their increasingly widespread introduction into the
control systems of atomic electric power plants provide a basis for talking about the prac-
tical possibility of using a computer in a closed control loop. The results of the tests on
the IRT-2000 reactor of the different structures and algorithms of direct digital control
may be useful in the planning of real control systems for large energy reactors.
LITERATURE CITED
1. V. G. Dunaev and P. T. Potapenko, At. Tekh. Rubezhom, No. 12, 10 (1974).
2. E. V. Filipchuk et al., At. Energ., 39, No. 1, 12 (1975).
3. T. Apostolov, I. Uzunov, and A. Markov, in: Testing of the Operation and Use of Research
Reactors [in Russian], Symposium of the Council for Mutual Economic Aid, Predyal (1974),
p. 813.
4. E. V. Filipchuk et al., At. Energ., 39, No. 2, 90 (1975).
5. Ya. Bouzhik, A. Gadomski, and S. Lyatek, Kernenergie, 19, No. 6, 183 (1976).
6. L. G. Andreev et al., At. Energ., 40, No. 4, 335 (1976).
7. M. G. Miteliman et al., At. Energ., 39, No. 4, 272 (1975).
TWO-ZONE SYSTEM PULSE METHOD
B. P. Shishin, Yu. A. Platovskikh, UDC 621.039.51.12
and T. S. Dideikin
In theoretical and experimental investigations [1-3] of the behavior of neutron pulses
in multizone breeder systems, interest has centered on the principal eigenvalue Al, deter-
mining the neutron-pulse decay as t =. If the active-zone?reflector system is sufficiently
subcritical, Al depends entirely on the reflector properties and, if the system is only
slightly subcritical, Al depends predominantly on the active-zone properties [2-3]. In this
work it is shown that under certain conditions the main part of the function representing
the neutron-pulse decay with time is described by two time constants Al and A2, where A2 de-
pends mainly on the reflector properties and in particular, on its adsorption cross section.
Determining Al and A2 is intimately connected with finding the reactivity in the given un-
steady experiments.
The single-group diffusion approximation with the usual boundary conditions (equality
of neutron fluxes and currents at the boundaries of the media and zero neutron flux at the
boundary of the system) gives the following equations for the eigenvalues An
F1[wi(X)]=D2F2[4,2(X)].
where F2(w2) = wiJi(willi)/Jo(w2R2); F2(w2) = w2[I2(w2R1)Ko(w2R) I0(.02R2)1(1(w2R1)]/[10(w2R2)?
K0(w2112) ? I0(w2R2)1(0(w2R2)]for cylindrical geometry; F2(w2) = w1tanw2122; F2(w2) = w2cotw2h
for plane and spherical geometry. In the latter case the equation for A agrees with the
Translated from Atomnaya gnergiya, Vol. 42, No. 5, pp. 370-372, May, 1977. Original
article submitted July 13, 1976.
414
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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
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FI,F2
I
I
1
1
A 4 A;1
a2
6'113
A'21
A-10-3
sec_i
A'22
Fig. 1. Behavior of the
functions F1 (dashed line)
and F2 (continuous line).
corresponding equation of [2]. Here 111 is the radius (or semithickness) of the active zone;
R2 is the external radius of the reflector; h is the reflector thickness; w(A) = [A ? A. ?
(vD)04]/(vD)1; 4(A) = [(vEa)2 + (vD)214 ? X]/(vD)2; X., is the time constant corresponding
to an infinite medium with the properties of the active zone; 14 = (71/H)2 is the geometric
height parameter (for a finite cylinder of height H); the subscripts 1 and 2 on vEa, vD, and
w refer to the active zone and the reflector, respectively; Ea is the thermal cross section.
The value of vD determines the slope of the function X = X(B2) for each medium. To deter-
mine the limits on X2 and A2, the properties of the functions F1 and F2 are considered. F1
and F2 have the following properties: if X is equal to X1r, and An (the eigenvalues of media
1 and 2 with zero boundary conditions for the neutron flux at their boundaries), F1 and F2
have poles; if X is equal to 4n and XL (eigenvalues of each medium with zero boundary con-
ditions for the derivative of the neutron flux at r = R1), F1 and F2 have zeros between the
poles. It is simple to estimate X' and X" since they correspond to homogeneous media. F1 is
monotonic if X < Xli and F2 is monotonic if X < 41 (Fig. 1). If X;2 < A121 (as in the ex-
periments described below), the least (principal) eigenvalue X1 lies within the broad limits
41 < X, < 41 and depends on the height and breeder properties of the system. The second
eigenvalue A2 is close to the pole of F2 and is found within the limits 41 = (vEa)2 + (vD)2?
[(7/2h)2 +
< A2 < A21 = (VEa)2 (VD)2[(r11h)2 4- 14], i.e., between the eigenvalues of re-
flectors of thickness h and 2h. For a water reflector these limits are narrow.
Thus A2 ,=-1 (VEa)2 and depends very little on the height and other properties of the ac-
tive zone and also on the reflector dimensions. Therefore if the reflector height exceeds
that of the active zone (as in several of the experiments) A2 is largely unaffected. The
eigenvalue X1 corresponds to a monotonic eigenfunction of the neutron flux density with a
maximum at the center of the active zone and A2, to an eigenfunction which has a maximum at
the reflector and passes through zero close to the boundary of the active zone and the re-
flector. The formulas obtained for w(A) and LO(X) correspond to a linear dependence X(B2).
Taking into account the nonlinearity of these dependences, significant at singularities for
the active zone, leaves the conclusions qualitatively unchanged.
In the pulse experiments carried out on a two-zone reactor, the two neutron-density de-
cay constants were found and the relation of these values to the change in the breeder prop-
erties of the reactor was established.
The experiments were carried out on a subcritical uranium?water pile surrounded by a
supplementary water reflector. End reflection was practically absent. Asa protectionagainst
thermal-neutron scattering in the pulse experiments the pile was coated with cadmium sheet
below and at the sides. In addition, it was placed a distance of at least 1 m from the wall
and floor of the chamber. Sampling measurements showed that the neutron-density decay con-
stant in the water of the reflector for a dry active zone with porosity 80% was 4900 sec-1.
This value agrees with the constant in an isolated volume of the same order as the reflector
volume, and hence scattering of the thermal neutrons makes a negligible contribution to the
detector reading. Fast scattered neutrons were recorded in the 50-100-psec range (this in-
formation was not considered).
In the experiments a neutron gas with a neutron-pulse length of 1 psec was used. The
commutation scheme developed allowed information from four SNM-13 neutron counters to be re-
corded simultaneously on an AI-256 analyzer [4]. The resolution time of the counter--analyzer
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nWN
100
10
1
10 20
Channel number
TABLE 1. Results for Neutron Decay Con-
stant
30
Fig. 2. Time, dependence of
neutron flux density, re-
corded by counters (A) in
the active zone and reflector
in pulse experiments (the
figures on the curves cor-
respond to the counter num-
bers;--,S is the neutron source).
Ex pt.
State of
active
zone
'
State of
reflector
,
Boron
concn. in
soln.,
g /liter
Sub-
crit-
ical-
ity p /13
Xl.
sec -1
)4.
sec -1
1
Dry
with H20
_
?
?
4900
2
with 11,0
The same
_
3,8
760
5000
3
with H-20
"
?
4,9
950
5000
4
With bor-
ic acid
solution
"
0,93
4,3
1420
5100
5
The same
"
0,93
7,9
2400
5000
6
"
With bar-
ic acid
0,22
3,6
1250
7300
solution
.
7
"
The Sam
0,22
5,2
1670
7300
8
"
0,87
4,6-
1520
13600
9
.
0,87
15,5
4490
13050
system was 10-12 psec. The sites of the counters were at the center of the active zone, at
half the radius of the active zone, in the reflector close to the active-zone?reflector
boundary, and at the midpoint of the neutron reflector. The ratio of the active-zone or re-
flector volume to the volume of the SNM-13 counter was 104-105, and hence the counters had
no practically significant effect on the processes in the active zone and the reflector. In
the course of the experiments the active zone and the reflector were filled with water or
boric acid solutions. The reflector was completely filled; this level remained unchanged in
a series of experiments with constant contamination of the active zone with-boron. The level
of filling of the active zone was varied, thereby altering the degree to which the system
was subcritical.
The AI-256 pulse time analyzer recorded the information from the four counters on the
neutron-density decay in the reactor after the passage of a short pulse of fast neutrons.
Typical results obtained from one of the experiments are shown in.Fig. 2.
Analysis of the experiments shows that the neutron-density decay is described in the
active zone basically by the function exp (?Xlt) and in the reflector by the sum of exp
and exp e-X20. In the whole series of experiments states with A2 > Al were observed. The
value of A2 was constant with constant reflector composition and different degrees of sub-
criticality of the reactor, but depended significantly on contamination of the reflector by
boron; the value of Al depended on the level of filling and boric acid concentration in the
active zone and on the reflector properties (see Table 1).
The error in the measurements of Ai and A2 was -?3%. It follows from the experiments
that the asymptotic decay constant AI for the neutron density in the reactor indicates the
breeder properties of the reactor, while A2 is a physical characteristic of the reflector.
This puts in doubt the basis for the method of measuring the degree of subcriticality of two-
zone reactors [5], in which the whole of the decay curve for the neutron pulse is used to de-
termine the reactivity.
416
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LITERATURE CITED
1. R. Sher, Nucl. Sci. Eng., 29, 302 (1967).
2. E. A. Stumbur et al., in: Theoretical and Experimental Problems of Unsteady Neutron
Transfer [in Russian], Atomizdat, Moscow (1972), p. 282.
3. G. Wells, Trans. Am. Nucl. Soc., 9, 170 (1966).
4. S. P. Volkov and B. P. Shishin, Prib. Tekh. Eksp., No. 1, 95 (1975).
5. A. Walter and L. Ruby, Nukleonik, 8, 287 (1966).
TOTAL REACTION CROSS SECTIONS OF CERTAIN METALS AND
GASES FOR VERY COLD NEUTRONS
N. T. Kashukeev, G. A. Stanev, UDC 539.125.5.162.2
V. T. Surdzhiiski, and E. N. Stoyanova
Research on the storage of ultracold neutrons [1-4] has shown that there is a systematic
discrepancy between theory and experiment. To investigate this, the interactions of neutrons
with matter were studied at energies close to those of ultracold neutrons (UCN). The inter-
action of neutrons with certain materials was studied in [5] in the velocity range from 5 to
100 m/sec, and in [6, 71 in the range from 40 to 350 m/sec.
We describe the procedure and apparatus used to measure the interaction of very cold
neutrons with certain metals and gases, and list the measured values of the neutron total
reaction cross sections for velocities from 100 to 250 m/sec. The range of neutron veloci-
ties chosen is of interest since it is close to the range of ultracold neutrons, and also
opens up the possibility of studying the interaction of neutrons with the nuclei of samples
without the effects of coherent Bragg scattering and surface reflection from the samples.
Calculations and experiments showed that the latter is unimportant.
Results of the investigation of the law of interaction of very cold neutrons with cer-
tain gases in the velocity range from 60 to 250 m/sec are presented. These results may turn
out to be useful in investigating the propagation of: ultracold and very cold neutrons in
gases, and in the development of the theory of UCN gaseous converters.
Arrangement of Experiments to Study the Interaqltion of Very Cold Neutrons with Metals
and Gases. The measurements were performed at the 17-2000 reactor in Sofia on a 10-m hori-
zontal channel especially fitted out for this purpose. The total interaction cross sections
of samples for very cold neutrons were measured by he transmission method and the results
were processed by the least-squares method.
Fig. 1. Experimental channel
to study interaction cross
sections of metals and gases
for very cold neutrons: 1)
neutron guide; 2) chopper.
Institute of Nuclear Research and Nuclear Power, Bulgarian Academy of Sciences, Sofia.
Translated from Atomnaya inergiya, Vol. 42, No. 5, Op. 373-377, May, 1977. Original article
submitted April 26, 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.
417
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500 mm
2
5
6
Fig. 2. Mechanical chopper for very cold neu-
trons: 1) protective screen; 2) flat rotor with
two transmitting slits; 3, 4) system for receiv-
ing starting signal; 5) disk for attaching sec-
ond rotor; 6) detector of very cold neutrons.
A beam of very cold neutrons was obtained at the exit of a curved reflecting neutron
guide made ofan electropolished stainless-steel tube (Fig. 1) inserted to the bottom of the
channel. The neutron guide had a radius of curvature of 22 m, a diameter of 0.04 m, and an
overall length of 7.37 m. The neutron guide consists of two parts which can be evacuated in-
dependently or filled with the gas under study.
The spectrum of the very cold neutrons was analyzed with a special mechanical chopper
having a flat rotor with two transmitting slits and operating in a time-of-flight regime
(Fig. 2). lExperiments in the velocity range from 100 to 250 m/sec were performed with the
rotor turning 2500 rpm, a flight path of 0.5 m, and 4 or 11.5? slits. Between the rotor and
the detector there was a vacuum tube with a nonreflecting cadmium-plated surface which could
be filled With the gas being studied. The very cold neutrons were detected with a helium-
filled proRortional counter made at Dubna for recording ultracold neutrons [3]. The detec-
tor signals were time analyzed with a 400-channel analyzer. Both parts of the analyzer mem-
ory were used; each part contains 200 channels each 102.4 msec wide.
The total flux of very cold neutrons incident on a sample was 50 pulses/cm2ssec. Addi-
tional experiments to soften the spectrum or to increase the existing flux by using various
converters were unsuccessful since the converters (graphite, aluminum, flowing distilled
water) placed on the bottom of the neutron guide were not cooled.
Interaction Cross Sections of Metals. Experiments on the transmission of a beam of very
cold neutrons through metal samples were performed for two different positions of the metal
plates being studied.
1. A plate was rigidly fastened directly over the opening of the neutron guide in front
on the protective screen of the chopper. The beam was measured with and without the sample
and the results were compared.
2. A plate of suitable shape was fastened over one of the slits in the rotor and ro-
tated with it. Two peaks were recorded simultaneously in the analyzer (Fig. 3) corresponding
to the two slits? the spectra of the beam (1) without the sample and (2) with the sample.
The differences between the cross sections measured with the sample in the two positions
were within the limits of error as can be seen from the results for copper and nickel (Table
1). Consequently, the two methods are equally suitable under the present conditions. In
both cases the angle at which the detector is viewed from the sample is q,10?.
418
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800
25 50 75 100 125 150 175 200
Channel number
Fig. 3. Peaks corresponding
to the two slits in the ro-
tor of the mechanical chop-
per (silver, d = 0.128 mm,
t = 120 mm).
TABLE 1. Neutron Interaction Cross Sec-
tions of Metals for v = 100 m/sec. b
Metals
or
aa (BNL) t
Ag (1; 1)*
1356+120
1360
In (2; 1)
4080+370
4180
Cu (2; 1) technical
82+8
81
Cu (1; 1) )
84+8
81
Cu (2; 1) electropolished
technical
70+7
81
Ni (1: 1)
140+15
100
Ni (2; 1)
145+14
100
Fe (1; 1)
92+10
56
Mo (1; 2)
194+20
55
Mo (1; 3)
203+22
55
*Measured neutron total interaction cross section
of sample.
tNeutron absorption cross sections of metals cal-
culated from 1/y law.
The first number in parentheses denotes a station-
ary 1 or rotating 2 sample; the second is the number
of plates in the sample.
Recycled neutrons were not observed in any of the measurements, as can be seen from
Fig. 3, since there were no neutrons with velocities below 50 m/sec in the beam used.
The background was determined by averaging the number of pulses over all the channels
in which the pulse counts were distributed statistically uniformly (no fewer than a hundred
channels), and over the whole spectrum when measurements were made with both slits of the'
chopper rotor covered with cadmium screens. Both methods gave 0.004 pulses/sec per analyzer
channel.
The statistical errors quoted for the measurements of the total cross sections were cal-
culated for a neutron beam with an angular spread of 2? and an uncertainty of 0.05 m in the
flight path because of the thickness of the detector. These estimates are considerably lar-
ger than the statistical spread of the results, since evidently the effective depth of pene-
tration of neutrons into the detector is less than its thickness.
The range of neutron velocities for which the results are reported was chosen from a
broader range of neutron velocities, since for this range the statistical errors were small.
The results are less reliable for neutrons with somewhat lower velocities because of the very
low intensity, and for neutrons of higher velocity because of the decrease of the spectrom-
eter resolving power with increasing neutron velocity. A special channel was constructed
to investigate the interaction of lower velocity neutrons.
In order to investigate the role of surface coherent effects for the velocity range stud-
ied, measurements were performed with different numbers of plates of the same metal. If these
effects play any role in the present case, the results would depend on the number of plates,
i.e., on the number of reflecting surfaces. Table 1 shows that two and three plates of molyb-
denum give practically identical results.
To investigate the absorption cross section and the incoherent scattering cross section,
two groups of metals were studied: metals for which the thermal neutron absorption cross
section is many times larger than the incoherent scattering cross section (Ag, In, Cu), and
419
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6 b
200
160
120
80
142 24,3 344 445 A,
t I
4458 244 162 /22 98
m /sec
6t, b
4000
2000
41
16,2
24,3
344
40,5
488
244
162
122
98
A, A
V, M/sec
Fig. 4. Total interaction cross sections of certain metals for very cold
neutrons as functions of neutron wavelength.
metals for which the incoherent scattering cross section is comparable with the thermal neu-
tron absorption cross section (Ni, Fe, Mo).
Figure 4a and b shows the total interaction cross sections of certain metals for neutrons
in the velocity range 100-250 m/sec as functions of the neutron wavelength. In both groups
of metals this relation is linear; i.e., the interaction cross section obeys the 1/v law:
cr(0==adhktv. (1)
The total cross sections for weakly absorbing metals (Fig. 4a) become negative when ex-
trapolated to A = 0, but these values, like those for metals with large capture cross sec-
tions (Fig. 4b), lie within the limits of statistical errors. It is doubtful whether such an
extrapolation should be made, however, because of the multiplicity of different interaction
processes possible for neutrons with widely different wavelengths.
Table 1 lists values of the neutron total cross sections at 100 m/sec. The absorption
cross sections for this velocity were calculated from the thermal values by using the 1/v
law. For metals of the first group, the absorption cross sections are the same as the total
interaction cross sections, while for metals of the second group there are substantial dif-
ferences. Thus, the absorption cross sections of metals of the first group obey the 1/v
in the range from 100 to 2200 m/sec. If it is assumed that the absorption
metals of the second group obey the 1/v law in the 100-2200-m/sec range,
250-m/sec range there is another interaction, in addition to absorption,
obeys the 1/v law.
Table 1 shows that the differences between the measured total cross sections and the ab-
sorption cross sections calculated by the 1/v law are not the same for the metals investi-
gated: for Ag, In, and Cu there is no difference; for Ni and Fe the difference is the same
(40 b); for Mo the difference is appreciable (140 b).
The additional interaction of neutrons with nickel and iron in the indicated velocity
range is apparently related to th'e domain structure of the samples. For the molybdenum sam-
ple the reason for the observed difference is so far not clear. On the basis of further in-
vestigations it can be conjectured that the cause of the increase in the neutron total reac-
tion cross section of molybdenum is the small-angle scattering of neutrons by inhomogeneities
of the sample. However, this scattering should obey the 1/v law in the velocity range in-
vestigated, and is observed at appreciably larger angles, depending on the characteristics
of the experimental arrangement.
It should be noted that the measured values of the total cross sections of copper and
nickel are in good agreement with the values reported in [5-7] for this velocity range.
The data in Table 1 were obtained for metal samples of 99.95% purity. Some samples of
co,pper, nickel, and iron of technical purity were investigated also; the results obtained do
not differ significantly from those for the high-purity samples.
cross
law
sections of
then in the
whose cross
100-
section
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0
8,1
16,2
24,3 32,4
40,5 48,6
56,7
64,8
Fig. 5. Mean free path of very cold neutrons as a function of
neutron wavelength: 1) tube between chopper rotor and detec-
tor filled with nitrogen; 2) outer portion of neutron guide
filled with argon (gas pressure ? 760 mm Hg).
Total Interaction Cross Sections of Gases for Very Cold Neutrons. Two methods of lo-
cating the targets were used in the investigations with gases. The gas filled a straight
tube with nonreflecting walls located between the chopper rotor and the detector, or the out-
er portion of the neutron guide. In this case the geometry of the experiment is relatively
good, but the results might be affected by a possible change in the coefficient of reflection
of neutrons from the walls of the neutron guide when gas is present. Figure 5 shows the de-
pendence of the neutron mean free path in nitrogen and argon at a pressure of 760 mm Hg on
the neturon wavelength. The total interaction cross section varies with neutron velocity
according to a hyperbolic law:
cr(v)=(c?d/v)-1. (2)
The quantity u = 2d/NL, where N is the nuclear density of the gas and L is the length
of the tube, has the same value as the mean-square velocity of the gas molecules.
It should be noted that the neutron total interaction cross sections of gases appear to
deviate from (2) for lower neutron velocities.
Further studies will be made for different temperatures of the gas targets and for gases
with substantially different molecular weights.
The authors thank A. V. Antonov, V. M. Iobashov, I. M. Frank, and Khr. Ya. Kristov for
their stimulating interest in the work; V. V. Golikov, V. I. Lushchikov, A. V. Strelkov, and
Yu. V. Taran for helpful discussions of the results.
1.
2.
3.
F.
L.
L.
LITERATURE CITED
L. Shapiro, JINR Preprint R3-7135, Dubna (1973).
V. Groshev et al., JINR Preprint R3-5392, Dubna (1970).
V. Groshev et al., JINR Preprint R3-7282, Dubna (1973).
4.
I.
M. Frank, JINR Preprint R3-7810, Dubna (1974).
5.
A.
Steyerl and H. Vonach, Z. Phys., 250, 166 (1972).
6.
W.
Dilg and W. Mannhart, Z. Phys., 266, 157 (1974).
7.
R.
Lermer and A. Steyerl, Phys. Stat. Sol. (a), 33,
531
(1976).
421
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Declassified and Approved For Release 2013/04/01: CIA-RDP10-02196R000700090005-4
RESULTS OF TESTING CARBIDE FUEL ELEMENTS IN THE
BOR-60 REACTOR
V. A. Tsykanov, V. M.
E. F. Davydov, V. I.
A. A. Maershin, V. N.
I. S. Golovnin, T. S.
Yu. K. Bibilashvili,
V. S. Mukhin, and G.
Gryazev,
Kuz'min,
Syuzev,
Men'shikova,
R. B. Kotel'nikov,
V. Kalashnik
UDC 621.039.542.344
A great deal of attention has been given recently to the investigation of carbide fuel
for fast power reactors with higher thermal conductivity and a larger number of heavy atoms
than oxide fuels; this makes it possible to reduce the doubling time, increase the multipli-
cation factor, and thereby make the operation of fast reactors more efficient. In this paper
we present some results of a study of the behavior of carbide fuel and carbide fuel elements
under irradiation; these were used in the BOR-60 reactor to a burnup of 10% of the heavy
atoms (h.a.).
Characteristics of Bundles and Fuel Elements. Each of the four irradiated bundles con-
tained 19 fuel elements. In two bundles there were three fuel elements each with a sodium?
potassium layer. In the other fuel elements, the gap between the jacket and the core was
filled with helium. The construction of the bundles and the fuel elements has been described
earlier [1]. The density of the fuel pellets was 90-96% of the theoretical value, and the
carbon content was 4.7-5.1% by mass. If we take account of the oxygen and nitrogen content,
the mole ratio (C + 0 + N)/U ranged from 0.98 to 1.08. The cores were constructed by press-
ing and baking, and also by hot pressing of the original aarbide powders obtained by gas
TABLE 1. Parameters for the Testing of
Fuel Elements with Carbide Fuel
Maximal irradiation
regime
Bundle
II
III
IV
Burnup, ak h.a.
3,5
5,1
7,1
10,4
Fluence of neutrons ?
10-22, neutrons 7cms
for E > 0 MeV
2,1
3,5
51
7,6
for E > 0.1 MeV
1,7
2,8
4,1
6,1
Linear power,
550
550
560
700
W/cm
Temp. of jacket, ?C
600
600
650
680
Calculated temp. at
center of fuel element
with helium layer, ?C*
1120
1120
1160
1300
Operating time at
power, h
5760
8570
11500
13600
*The thermal conductivity of the fuel was taken to be
0.15 Wicm ? deg C; the conductivity of the contact be-
tween fuel and jacket was taken to be 1 W/cm2- deg C.
TABLE 2. Balance of Volume Variation in
Fuel Elements under Irradiation up to a
Burnup of 10.4% h.a., Expressed as % of
the Initial Volume within the Fuel-Ele-
ment Jacket along the Active Part
Part of fuel
element
Before
irradiation
After
irradiation
Change in
volume
Jacket
100/100
104,6/105,2
+4,6/5,2
Core,
including:
88,6/85,8
104,6/105,2
+16,0/19,4
fuel
73,8/79,7
75,2/81,1
+1,4/1,4
pores
4,8/5,1
17,9/23,9
+13,1/19,7
cracks
0/0
0,6/0,71-
+0,7/0,7
central
cavity
10,1/0
10,9/0
+0,8/0
Gaps,
including:
11,2/14,0
0/0
?11,2/14,0
radial
6,3/9,1
0/0
?6,3/9,1
axial
4,9/4,9
0/0
?4,9/4,9
*The data in the numerator refer to fuel elements with
a plug core, those in the denominator fo fuel elements
with pellet cores.
t Estimate.
Translated from Atomnaya inergiya, Vol. 42, No. 5, pp. 378-382, May, 1977. Original
article submitted October 4, 1976.
422
This material is protected by copyright registered in the name of Plenum Publishing Corporation, 227 West 17th Street, New York, IV. 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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Declassified and Approved For Release 2013/04/01: CIA-RDP10-02196R000700090005-4
700 g
600 .5,
4.)
500 -8
400
300
MO hV 300 RV 400
Height of fuel element, mm
Fig. 1. Distributions of neutron flux for E > 0
(1) and >0.1 MeV (2) and maximum fuel-element jack-
et temperature as functions of active-zone height.
Change in o.d. of jacket, 0/0
Fig.
element jacket as a function of burnup: 1, 2, 3,
4) gas carbidization: pellets, plugs, pellets
with covering, and with UC--PuC, respectively; 5,
6, 7, 8) carbothermal reduction: pellets, plugs,
hot-pressed pellets, and pellets with covering,
respectively.
3
0
2.
1 2 3 4 5 6 7 8 9 10 11
Burnup, 01011.a
Variation of outer diameter of the fuel-
carbidization of powdered uranium and carbothermal reduction of uranium dioxide [2-4]. Two
fuel elements had a mixed fuel containing '1,15% PuC by mass. The diametral gap between the
pellets and the jacket varied from 0.12 to 0.4 and from 0.2 to 0.6 mm, respectively, for the
fuel elements with a helium layer and those with a sodium--potassium layer. The 235U enrich-
ment was 90%. The jacket was made of OKh16N15M3B stainless steel, with an outer diameter of
6.9 mm and a wall thickness of 0.4 mm. In five fuel elements the fuel pellets had a chro-
mium--niobium-based covering.
Test Parameters. All the bundles were tested in the fifth row of the active zone of
the BOR-60 reactor (Table 1). The distributions of the neutron flux and of the maximum jack-
et temperature for different reactor power levels are shown as functions of active-zone
height in Fig. 1.
Condition of the Fuel Elements. The condition of the fuel elements and bundles after
the test was satisfactory. All the fuel elements remained airtight. The change in the outer
diameter of the jacket was observed after 5% h.a. burnup, and at the maximum burnup value of
10.4% h.a. it amounted to 0.8-2.5% (Fig. 2).
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100 ZOO 300 400 500 600
Distance from upper end of fuel element, mm
14
12'
10
4
0
4 6
Burnup, 070 h.a.
10
Fig. 3. Variation in outer
diameter (solid curve), plas-
tic deformation (dashed curve),
and diametral swelling of jack-
et (solid curve with circles)
AS functions of fuel-element
active-zone height.
Fig. 4. Variation of
relative flow rate of
gaseous fragments from
fuel as a function Of
burnup (same legend
as Fig. 2).
Estimates made on the basis of the change in the density of the steel jacket by the
method of hydrostatic suspension showed that for a fast-neutron fluence of 6?1022 neutrons/
cm2 CE
> 0.1 MeV) the increase in the diameter as a result of jacket swelling [for isotopic
swelling Ad/d = (1/3)AV/V] was (0.8-1.0) ? 0.2% (jacket temperature 500?C).. The remainder
of the change in the jacket diameter was apparently caused by a deformation of the steel un-
der the pressure of the swollen fuel.
The plastic deformation of the jacket, calculated as the difference between the total
change in diameter and the diameter increment due to the swelling of the jacket material at
a burnup value,of 10.4% h.a,, wasA-1.5%, i.e., constituted up to 60% of the total change in
the jacket diameter. The change in the outer diameter due to the swelling of the jacket ma-
terial and to plastic deformation of the jacket along the height of the active part of a fuel
element used to a burnup of 10.4% h.a. is shown in Fig. 3.
The relative flow rate of fission-fragment gases from the fuel element under the jacket
at a core temperature of less than 1300?C was 2-12% and 1-3% for fuel elements with helium
and sodium--potassium layers, respectively (Fig. 4), and increased somewhat with increased
burnup.
Balance of Volume Variation in Fuel Elements under Irradiation. Data on volume changes
for the irradiation of fuel elements with different carbide cores are given in Table 2.
Deformation of the Jacket. Experimental data on the deformation of the fuel-element
jackets were investigated by the method of regression analysis. We studied the variation of
the relative deformation with the different forms of free spaces in the fuel element, with
424
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2,0
2,5
//
5 10 15 20 25 30
Volume of pores and radial gap, %
Fig. 5. Variation of the outer diameter and plas-
tic deformation of fuel-element jackets as func-
tions of the porosity of the fuel and the radial
gap: 1, 2) UC, gas carbidization: pellets and
plugs; 3, 4, 5, 6) UC, carbothermal reduction: pel-
lets, plugs, hot-pressed pellets, and plugs, respec-
tively; +) confidence interval 0.9; ---) boundaries
of region of reliable values.
the presence of an axial cavity, and with the properties of the techniques by which the car-
bide and the core were obtained. The analysis was carried out for fuel elements irradiated
to burnup values of 7.1 and 10.4% h.a. To construct the regression models, we used the most
efficient step method [5]. For fuel elements irradiated to a burnup value of 10.4% h.a. we
found that the jacket deformation depended only weakly on the effective fuel density (effec-
tive density is the ratio of the fuel mass to the volume inside the jacket along the active
part of the fuel element) and did noti observe any effect produced by the central cavity and
the techniques by which the carbide and the cores were obtained. The best correlation was
found for the sum of the values for the fuel porosity and the radial gap (Fig. 5). It fol-
lows from Fig. 5 that in order to obtain zero plastic deformation (the dot?dash line) for
given irradiation conditions, the total volume of the pores in the fuel and the radial gap
must be '\,20%.
Variation in Fuel Structure. The typical structure of the irradiated fuel (Fig. 6) in-
dicates that there is no marked grain growth; this is in agreement with the calculated esti-
mate for a relatively low fuel temperature (1300 and 1150?C for pellet and plug cores, re-
spectively). In most of the investigated cross sections of the fuel elements, we found
cracking of the core after irradiation.
The main change in the carbide-fuel structure under irradiation consists in the forma-
tion of pores. The dependence of this process on the irradiation temperature, the burnup,
and, probably, the granularity and chemical composition of the fuel is very complicated. At
a temperature above 1000-1100?C the fuel forms pores with a diameter of more than 0.1 p,which
are visible with an optical microscope and are homogeneously distributed through the grain at
a burnup value of up to 5% h.a. An increase in temperature and burnup leads to the preferen-
tial formation of large pores along the boundaries of the grains, and in a number of cases
to a merging of the pores on the boundaries oriented perpendicular to the thermal gradient
(Fig. 6b). At a temperature of '1,1300?C and a burnup of 10% h.a. we found that the intercon-
nected porosity changed to isolated porosity (Fig. 6c). Metallographic investigation of the
irradiated fuel did not reveal any phases containing fission products, apparently as a re-
sult of their high degree of dispersion due to the relatively low irradiation temperature.
The uranium dicarbide present in the superstoichiometric fuel in the form of numerous
inclusions was dissolved under irradiation. This may be due to the fact that the oxygen
stabilizing the dicarbide is bound by the fragment elements [6]. The 0.4% decrease in the
monocarbide lattice parameter (Fig. 7) for 10% h.a. burnup indicates considerable solubility
in it of the fission products (zirconium, molybdenum, etc.).
425
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Fig. 6. Macrostructure (a) and micro-
structure (b, c) of fuel after itradia-
tion to a burnup value of 10.4% h.a.
5
Burnup, 16 ha.
10
Fig. 7. Variation of the uranium monocarbide lat-
tice parameter as a result of irradiation: 0) gas
carbidization; 0) carbothermal reduction; A) hot
pressing; x, +) simulators with 0.3 and >0.5% 0;
40 UC + Cr.
Swelling of the Fuel. The swelling of the fuel was determined on the basis of the change
in density and in the geometric dimensions of the fuel briquettes. When the core center tem-
perature was 1200 ? 50?C, the average rate of swelling was 1.5 ? 0.2 and 1.2 ? 0.4% per 1%
burnup on the basis of the density change and the geometric-dimension change, respectively.
426
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TABLE 3. Carbonization of the Jacket in
the Case of Fuel Elements with a Helium
Layer (burnup 10.4% h.a.)
Jacket characteristics
Carbon content of fuel,
% by mass
4,7
4,82
4,9
4,95
Max. depth of the zone
of interaction, ?
150
120
150
150
Microhardness kg/mm2:
interaction zone
350
500
360
350.
outside zone of
interaction
260
300
300
300
Av. carbon content of
jacket, % by mass
_
0,14
0,14
0,16
150
400 NW NW
Ternp.ofirmersurftice
ofjacket,?C
Fig. 8. Variation of depth of the zone
of interaction in fuel elements with a
helium layer as a function of tempera-
ture: 1) carbothermal reduction; 2) gas
carbidization.
700
On the basis of data on the change in the dimensions of the fuel column in the axial and di-
ametral directions, we estimated the degree of anisotropy of the swelling, which was found
to be 1.3. The average value of the anisotropy coefficient indicates that the core swells
somewhat more in the radial than in the axial direction.
Compatibility of the Fuel and Jacket. The carbonization of the jacket ? one of the prob-
lems involved in the use of carbide fuel ? is most dangerous when there is a heat-conducting
liquid metal layer between the fuel and the jacket. When a helium layer is used, the car-
bonization is slight; we found that the carbon content of the fuel in the range from 4.7 to
4.95% by mass did not substantially affect this process (Table 3).
In fuel elements with mixed uranium?plutonium fuel irradiated to burnup values of 3
and 5% h.a., the depth of the zone of interaction did not exceed 20 p. The variation of the
depth of the zone of interaction in fuel elements with a helium layer (burnup value 10.4%
h.a.) as a function of temperature is shown in Fig. 8.
In fuel elements with a sodium?potassium layer that were irradiated to burnup values
of 3.3 and 7.1% h.a., the maximum depth of the interaction zone was found to be 100-150 and
0,
where A is the wavelength of the radiation; x is the distance from the source to the point
of detection; E is the eigenvalue found by solving the problem about the penetration of y
rays to alarge distance*; b and cl are known constants; r(x) is the y function of yo =
and eo is the angle of y-ray emission from the source.
As an example of the use of Green's function, we found the distribution function 'of the
energy flux density for plane and point isotropic sources.
It was shown that the distribution functions of the energy flux density at large dis-
tances from a plane perpendicular and a point isotropic sourceagree to with in the factor (1)(z,
X0)/411-x2 (z is the nuclear charge of the matter through which the radiation passes).
*U. Fano, J. Res. Nat. Bur. Standards, 51, 95 (1953).
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It was shown that the factor (I)(Z, A0) can not only be found theoretically btt can also'
be expressed in terms of the ratio of energy build-up factors of the plane perpendicular and
point isotropic sources which can be taken from experiment or calculated by the momentum
method.
A comparison is made of the factor (I) calculated theoretically and in terms of the ratio
of the energy build-up factors.
NEUTRON TRANSPORT IN HALF-SPACE WITH SOURCES
V. P. Gorelov and V. I. Il'in UDC 621.039.51.12:539.125.52
This paper considers the distribution of neutrons in a half-space with sources. The
solution is found in the form of a series in the complete set of eigenfunctions of the trans-
port equation [1]. The approximations used for the coefficient of the expansion of the solu-
tion for the eigenfunctions of the continuous part of the spectrum lead to finite analytic
expressions that are convenient for calculations.
We considered the Milne problem with quadratic anisotropy. To solve the initial equa-
tion for original transport equation, the expansion coefficient of the solution for the
eigenfunctions of the continuous part of the spectrum is approximated by the expression
A (v , f 2) A (f 2) (1?v)
(where v is the eigenvalue of the continuous part of the spectrum and f2 is the anisotropy
parameter), satisfying the well-known properties of the vanishing of A(v, f2) at the point
v = 1 and the divergence of the derivative of the total neutron flux at the interface [1].
In the process, for the boundary-value problem we replace the exact boundary conditions by
Marshak conditions for the moments of the distribution function [2]. This allows simple
analytic expressions to be obtained for the extrapolated length of H(f2) and the angular dis-
tribution of the radiation emerging from the half-space.
Calculations show that the formulas derived give good accuracy. The results for H(f2)
confirm that the extrapolated length depends weakly on the anisotropy factor f2.
It was noted that the expressions obtained for the neutron spectrum and the extrapolated
length contain a nondiffusive term in explicit form and give the dependence on f2, while re-
maining simpler than other known results [1, 3, 4].
The paper solves the problem of a half-space with anisotropic neutron sources by the
method of generalized eigenfunctions [1]. The spectral coefficient in this case is approxi-
mated by
C v
A (v , -= A (11o) (1? v) {P (v) 6 (v To) 1
2 (v ?I10)
where A(v) = 1.? (Cv/2)1n[(1 + v)/(1 v)]; arc cos po is the angle of incidence of neutrons
from the source; d(v) is the Dirac delta function; A(P0) is an unknown coefficient found from
the boundary conditions; and the index p denotes that the relevant integrals are taken in
the sense of the Cauchy principal value [5].
The form chosen for the spectral coefficient A(v, po) conveys the salient features, known
from the exact solution, of its vanishing at the point v = 1 and the existence of a pole at
the value v = 112.
Replacement of the exact boundary conditions by the conditions for the moments of the
distribution function,
Rzh+i,j, (07 [0 dp,=40-1,
k 1 ,
enables the necessary equations to be obtained for determining the unknown constants.
The analytic form of the spectrum of radiation emerging from the half-space and the ex-
pression for the albedo of the half-space with neutron capture havebeen obtained. Calculations
449
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of the latter show that the results arrived at are in good agreement with the exact value
[6]. The computational formulas convey nondiffusive effects and are quite simple in form.
The quadratic anisotropy limitation is not fundamental and the consisleration can be
generalized to more complex scattering laws.
T4 c:rrTT-
t
LITERATURE CITED
1. K. M. Case and P. F. Zweifel, Linear Transport Theory, Addison-Wesley (1967).
2. G. I. Marchuk, Theory and Methods for Nuclear Reactor Calculations, Plenum Publ. (1964).
3. L. N. Romanova, in: Some Mathematical Problems of Neutron Physics [in Russian], Mosk.
Gos. Univ. (1960), p. 8.
4. N. V. Ptitsyna, ibid., p. 28.
5. F. D. Gakhov, Boundary Value Problems, Pergamon (1966).
6. C. Crosjean, in: Proceedings of the Second International Conference, Vol. 16, Geneva
(1958), p. 431.
450
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LETTERS TO THE EDITOR
MEASUREMENT OF THE SENSITIVITY OF NEUTRON DETECTORS WITH
SILVER EMITTER DURING LONG SERVICE IN REACTOR
I. Ya. Emel'yanov, Yu. I. Volodtko,
V. V. Postnikov, V. O. Steklov,
and V. I. Uvarov
UDC 621.039.517
Beta-emission neutron detectors (BEND), especially BEND with silver emitters, are used
extensively to monitor the neutron flux density in reactors [1, 2]. The technology used in
the manufacture of cables with magnesian insulation is often employed in the fabrication of
BEND. This technology makes it possible to organize the series production of BEND with prac-
tically no limitations as to the length of the sensitive part of the detector.
Tests have been
cable (TUMI 098-69),
sulat ion. The cable
of analytically pure
sistant steel 0.5 mm
made with BEND with silver emitter, constituting a segment of a KDMS(S)
manufactured by the technology used in making cables with magnesian in-
has an Sr-999 silver core (GOST 7222-54) of 0.55-mm diameter, insulation
magnesium oxide (GOST 4526-67), and a sheath of Kh18N1OT corrosion-re-
thick. The outside diameter of the cable is 3.0 mm.
On-site radiation testing of BEND was carried out in an IVV-2 reactor up to a thermal-
neutron fluence of 2.4.1021 neutrons/cm2 at 650?C, as well as in the reactors of the first
and second blocks of the Beloyarsk atomic power plant up to a fluence of 1.1021 neutrons/cm2
at '1,300?C. The tests showed that the mean BEND current is proportional to the reactor power
and the mean neutron flux density, up to a power close to the nominal power of the first and
second Beloyarsk reactors and the IVV-2, i.e., at a thermal-neutron flux of up to 1.1013,
2.1013, and 1.1014 neutrons/cm2.sec, respectively. The proportionality is maintained during
long use up to the given values of the thermal-neutron fluence. The BEND specimens (seven
in the IVV-2 reactor, four in the second Beloyarsk reactor, and more than 30 in the first
Beloyarsk reactor) retained a sufficiently high insulation resistance during the tests. Thus,
at nominal reactor power it was no worse than 106 Qom.
Along with the specimens of BEND consisting of a straight segment of cable-detector,
five-section BEND were also tested to monitor the energy distribution over the height of the
first Beloyarsk reactor. The BEND sections, at 80?C, consisted of segments of cable-detec-
tor twisted into a cylindrical spiral. The signal from each section was transmitted to re-
cording equipment by a KNMS(S) cable [3]. The five-section BEND retained signal linearity
and a sufficiently high resistance of insulation during use in a reactor for more than 2years.
The tests confirmed the long service life of cable-type BEND with silver emitter. However,
precise data about the variations in the sensitivity of BEND sensitivity as a function of the
thermal-neutron fluence are also required in order to monitor the energy distribution in the
reactors. To obtain such data the IVV-2 reactor was provided with two experimental ports
holding, respectively, 7 and 9 BEND specimens. The cable-detector length irradiated was 0.7 m.
At nominal reaCtor power the temperature of the BEND specimens was80?C, the mean densityof
the thermal-neutron flux was 4.701013 neutrons/cm2.sec, the mean density of the fast-neutron
flux (E > 1 MeV) was 1.201013 neutrons/cm2.sec, and the mean y-ray dose rate was 1.00106 rd/
sec
In addition to determining the sensitivity, we periodically checked the linearity of the
BEND and measured the resistance of their insulation under various intensities of irradiation.
For precision monitoring of the neutron flux density, over a period of 1.5 yr we measured the
activity of a 2.0-mm-diameter wire made of Sr-999 silver. The wire was irradiated in an exper-
imental port along with a number of BEND specimens for 10 min at a reactor power close to the
Translated from Atomnaya Lergiya, Vol. 42, No. 5, pp. 403-404, May, 1977. Original
article submitted February 13, 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.
451
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0,8
......
? 3
1
la on
46
I
0,2 44 0,6 = 41 t,o 42 44
FT- 10-21, neutrons/cm2
Fig. 1. Sensitivity of BEND vs thermal neutron
fluence FT: 1) curve corresponding to formula
given in text; 2, 2') confidence limits with a
confidence coefficient of 0.98; 3) calculated
curve; first (0) and second (m) experimental
ports.
TABLE 1. Resistance of BEND Insulation at Various Values of Flux Density and
Thermal-Neutron Fluence
cpT, neu-
trons /cm2. ,
sec
FDIuftaromica0
1012
0,03.1020
1,0.1020
2,0.1020
0,4.1020
4,0.1020
4,0.1020
0,3.1020
0,1.1022
0,2.1020
0,0.1020
1,2.1021
4,7.1012
_
6,4.108
?
?
?
5,3.108
?
?
?
?
.5,8408
9,6.1012
2,3-108
?
?
?
?
?
4,2.108
. ?
?
?
?
?
1,4.1013
1,6.108
?
?
?
?
?
3,8.108
_
?
?
?
4,2.108
2,2.1013
1,2.108
1,4.108
?
?
?
2,7.108
?
1,6.102
?
?
?
2,6.1013
?
?
1,0408
?
3,8.108
?
2,2.108
?
2,0408
_
?
3,0.103
3,1.1013
?
?
2,5.108
?
?
?
3,1.108
?
2.0.108
?
?
?
4,2.1018
n,1-108
9,5-107
7,8.107
--
3,6.108
4,8.108
1,9.108
2,3.108
2,1-107
8,4.108
1,7.108
1,7-108
nominal value. After exposure in this port for 3-5 min activity with half-lives of 24.4 and
2.4 min decayed practically completely in one hour above the active zone. The residual ac-
tivity with a half-life of 259 days is not high (the y-ray dose rate at a distance of 0.1 m
is less than 100 pR/sec) and practically does not change during the measurements, thus making
for convenient work with a tracer and more accurate results of the measurements.
All the specimens retained their efficiency at a thermal-neutron fluence of 1.2.1021 and
8.8.1020 neutrons/cm2 in the first and second experimental ports. The tests showed that the
BEND readings remain linear in the neutron flux to within ?2% at an indicated radiation in-
tensity up to the values of the thermal-neutron fluence attained.
The results of the measurements, given in Fig. 1, were processed by the least-squares
method. Finally, we obtained an empirical formula chaiacterizing the relative variations in
the sensitivity of the a-emission neutron detectors with silver emitter as a function of the
thermal-neutron fluence FT:
Figure 1 gives the curve corresponding to this formula and the confidence limits for a con-
fidence coefficient of 0.98. It also gives the calculated curve of the variations in the
BEND sensitivity as a result of burn-up of the emitter material and an increase in the ther-
mal-neutron fluence. As seen from Fig. 1, the calculated curve is close to the experimental.
At a fluence of 1021 neutrons/cm2 the calculated variations in the sensitivity are 6%,
whereas the formula above yields 8%.
The resistance of the BEND insulation depends little on the irradiation intensity at
constant temperature [2]. A slight drop in the insulation resistance was observed in the
given experiment as the irradiation intensity increased; this is evidently due to a relatively
small rise in temperature which occurs in the process, regardless of well-organized heat re-
moval from the specimens. In the course of the data processing, the geometric mean of the
resistance of the BEND insulation was found separately from specimens placed in each port.
These values for various intensities of irradiation and neutron fluence are listed in Table
1. It can be seen that at a constant irradiation intensity the BEND insulation resistance
does not display any tendency to change with an increase in the neutron fluence, up to 1.2?
10" neutrons/cm2, and is q,108 Q?cm.
452
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Knowing the neutron fluence for each detector, we can make allowance for variations in
its sensitivity by means of the given relation. The detector sensitivity variations due not
to burn-up of the emitter material but to changes in the insulation resistance [2] can evi-
dently be disregarded, at least for a fluence of up to 1.2?1021 neutrons/cm2 since there are
no significant changes in the resistance.
The results of the given studies confirm the usefulness of cable-type 0-emission neutron
detectors with silver emitter for precision monitoring of energy distribution during long
periods of use in the active zone of a reactor with a power load.
1.
I.
Ya.
Emel'yanov et
al.,
At.
2.
I.
Ya.
Emel'yanov et
al.,
At.
3.
I.
Ya.
Emel'yanov et
al.,
in:
LITERATURE CITED
Energ., 30, No. 3, 275 (1971).
Energ., 37, No. 1, 72 (1974).
Problems of Atomic Science and Engineering. "Reactor
Construction" Series [in Russian], No. 4 (11), Izd. TsNIIatominform, Moscow (1974),
p. 51.
LOCAL CONTROL OF PROFILE AND MAGNITUDE OF ENERGY RELEASE'
OF LOOP CHANNELS
F. M. Arinkin and G. A. Batyrbekov UDC 621.039.51.519
This paper gives the results of studies on the possibilities of controlling the neutron
field and energy release over the height and diameter of an experimental channel of a water-
moderated?water-cooled reactor by means of an annular chamber with gaseous neutron absorber
3He.
Chamber 1 (absorbing shield) consists of two stainless-steel coaxial tubes 2, welded
into the common end plate and flange. Along its height the chamber is divided into seven
isolated sections 3 with an autonomous gas supply system (Fig. 1). The number and dimensions
of the sections were chosen in accordance with the conditions of nuclear safety of the ex-
periment (the reactivity of one section did not exceed 0.7 fieff, the fraction of delayed neu-
trons) and acceptable control of the energy-release profile along the channel height. The
height of the active part of the sHe-filled chamber was 580 mm, which is slightly less than
that of the active zone (600 mm). The chamber is connected to cut-off valves and manometers
via a collector with a tank for 3He by a system of tubes 4. The chamber was installed in the
central cavity of a critical assembly measuring 140 mm in diameter [1], formed by a shaped
plunger 5 made of CAV alloy. The chamber was provided with 3-mm inside and outside gaps for
coolant ducts. In the middle of the chamber was an experimental channel of 96-mm diameter
into which a model fuel assembly was inserted. The assembly consisted of uranium dioxide
tablets enriched to 90%, measuring 12 mm in diameter and 4-6 mm in height, and packed in a
tungsten can with a wall thickness of 1 mm. The assembly contained seven model fuel elements
arranged along the channel axis and had a height of 400 mm with a clearance of 60 mm between
tablets.
For nuclear safety the chamber and all the tubing were tested for a long time at a pres-
sure of 1.5 times the working pressure. Upon being filled with gas, the sections were cut
off from the collector by the cut-off valves; this eliminated the possibility of 3He leaking
from all the sections simultaneously if the main tube burst. When the assembly was in an
automatic control mode, the automatic control rod (AC) was inserted almost completely (450 mm)
so that in the event of leakage of sHe from the system complete insertion of AC would activate
the scram system.
Translated from Atomnaya fnergiya, Vol. 42, No. 5, pp. 404-407, May, 1977. Original ar-
ticle submitted March 9, 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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454
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Fig. 1. Diagram of chamber and recorder chart of
active-zone load (all dimensions in mm).
Rcd
2
7
1,2
1,0
g8
0,4
0,2
Sections
_
_
',-.4..........___
1
.
3
_
7
6
5
4
3
2 ?
/
b
_
-
7
2
3
/
-------'N
ZOO 100 0 100 200 1-155, mm
Fig. 2. Distribution of (a) Rcd and (b) thermal-
neutron flux 0 along the height of the experimental
channel: a) cadmium ratio for (ad), where m is the
number of the curve; n is the number of the section:
1.163; (Ead)r5=0.831; (Ead)'5= (Ead) = 0.665; (Ead) = 0,416;
(Ead4 = 0.582; (Ead)i--7 = 1.828: the other sections for all
curves Ead = 0 for Ead = 1.828, and P = 1114.6 kPa;
b) thermal-neutron flux: (2ad)i-6 = (Ed)i = 1.163; (Ead)g-6 = 0,831;
(Ead)1-6 = 0.665; (zad) =- 0.249; (Ead) = (Ead)? = 0.419; (Ead)j = 0.582;
(ad)_5 = 1.828; (ad) 3 = 1.496; in the other sections, for all
curves Ead = 0.
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Sections
400
350
300
a 250
?;
200 -
150
100
SO
7
a
300 200 100 0 100 200 300 0 3 6
H n
ss, R, mm
un
Fig. 3. a) Distribution of energy along height of fuel assembly and b)
uranium fission density (11235) along radius of fuel-element model: a)
(madl-5 = 1.163; (I.d)1-5 = 0,831; (Z.61)4,6 = (1,f)a ,..._ 0.665; (Z.(1) = 0.249; (ad) g = (ad) = 0.582;
(Za1:)63_6 = 1.828; in the other sections, for all curves Ead = 0;b) Ead = 0 (1);
0.996 (2), and 1.828 (3).
46
45
44
Fig. 4. Negative reactivity of
individual sections of chamber
43
42
as function of 3He pressure: 1,
2) sections 4 and 5 (in the other
sections Ead = 0); 3, 4) sections
5 and 4 (in the other sections
0,1
Ead = 1.328, P = 911.9 kPa).
2 4 6
P, kPa
8
10
In studying the properties of the chamber, we employed the concept of shield "blackness,"
determined by Ead (where Ea is the macroscopic cross section for the absorption of thermal
neutrons by a working substance of thickness d). The essence of the method of control pro-
posed consists in measuring the shield blackness over height by varying the pressure of 3He
in separate sections.
The purpose of the experiment was twofold: to study the possibility of controlling the
distributions of the neutron flux and energy release along the radius and the height of the
experimental channel; and to study the effect of the chamber on the physical characteristics
of the active zone and to analyze the safety aspects of work with such devices.
The distributions of the energy release in the fuel assembly, of the thermal-neutron
flux, and the' cadmium ratio Rcd in the experimental channel were measured as a function of
the 3He pressure, as were the reactivity introduced into the active zone by the chamber as a
whole and by the individual sections as well as the interference between the sections.
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The energy distribution along the height of the fuel assembly was determined by calori-
metric methods [2]. The energy microdistribution inside the fuel elements was also measured;
upon integration, this gave the total energy release in each element [3]. The thermal-neu-
tron flux and Rcd were measured with 197Au and "Cu detectors. In regions with a large neu-
tron-flux gradient we resorted to autoradiography of the detecting material (in the given
case, strips of "Cu) onto x-ray film which was subsequently analyzed on a microphotometer
[4].
Figure 2 shows the distribution of the thermal-neutron flux and Rcd on the empty exper-
imental channel as a function of (ad)n, where n is the number of the section counted from
the top. Curve 1 corresponds to the case when the 3He pressure in all sections is zero.*
The coefficient of nonuniformity is 1.24 in this case. The asymmetry in curves 1 and 2 ("sag"
on right-hand side of plot) is caused by control elements inserted to a depth of 200-220 mm.
The energy Q along the height of the channel with a seven-element assembly (Fig. 3) is
of a similar nature, from which it is seen that the attenuation of the energy release at these
values of Ead is less than the attenuation of (1). The energy distribution along the height
of the assembly (see Figs. 2 and 3) and the various fuel elements can be controlled contin-
uously within wide limits; in particular, it is possible to shape an almost even profile of
energy distribution over the height of the assembly (curves 5 and 6). With the chamber a
maximum change of 2.4 times was obtained in the energy released in section 5, the mean energy
release in the assembly decreasing by a factor of two in the process.
The variation of the uranium fission density (p233) in the cross section of the central
fuel element is shown in Fig. 3b. The solid curves were obtained by calculation in the Sy
approximation of the Carlson method [5], whereas the empty circles denote experimental points.
The large experimental errors at the ends of the curves are explained by the difficulty en-
countered in determining the detailed shape of the microdistribution of the energy at the
boundary of the fuel element.
Calculations of the distributions of the thermal-neutron flux along the radius of the
reactor in the Sy approximation for various values of Ead showed that there is good agreement
with experiment within the limits of the active zone. In the empty experimental channel the
calculation underestimates the value of the neutron flux by 5-6%; this can be obviously con-
sidered a drawback of the Sy approximation in calculations of large air cavities. The cal-
culations were carried out by a one-dimensional kinetic program [6] with a system of 26-group
constants. The program permitted us to take account of a neutron leakage from the end planes
of the active zone. The leakage from the experimental channel was taken to be zero.
The reactivity measurements showed that the chamber without sHe introduces a negative
reactivity (-0.3%), whereas the experimental channel introduces a positive reactivity (+1.7%).
The reactivity of some sections of the chamber is plotted in Fig. 4 as a function of the 3He
pressure. The difference between curves 1, 2 and 3, 4 is due to interference between the
sections. The total contribution to the reactivity from all sections for P = 1114.6 kPa is
?2.4%.
In conclusion, the authors would like to thank the personnel of the critical test stand
at the Institute of High-Energy Physics, Academy of Sciences of the Kazakh SSR, as well as
V. P. Kiselev and Sh. Kh. Gizatulin for their assistance in performing the experiments.
LITERATURE CITED
1. Zh. S. Takibaev et al., 'FITE, Preprint PP-3 [in Russian],
Nauk Kaz. SSR, Alma-Ata (1973).
2. Yu. L. Tsoglin et al., in: High-Dosage Dosimetry [in Russian],
p. 136.
3. V. B. Klimentov, G. A. Kopchinskii, and V. G. Bobkov, At. Energ
4. A. Ertaud and P. Zaleske, J. Phys. Radium, 14, 191 (1953).
5. B. Carlson and J. Bell, in: Proceedings of the Second International
Peaceful Uses of Atomic Energy. Reactor Physics [Russian translation]
dat, Moscow (1959), p. 408.
6. L. N. Yaroslavtseva, NIIAR Preprint
Inst.
Fiz.
Fan,
Vys. inerg., Akad.
Tashkent (1966)9
, 29, No. 4, 283 (1970).
Conference on
Vol. 3, Atomiz-
P-9, Melekess (1968).
?
*For the other curves, Ead is given only for those sections which contain helium.
456
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LUMINOUS EMITTANCE OF NEUTRON BEAM IN AIR
A. V. Zhemerev, Yu. A. Medvedev, UDC 551.594.5
and B. M. Stepanov
The paper considers a method of visualizing neutron fields in air in the optical region
of the spectrum. Luminescence stimulated in air by neutrons with an energy of 'k,10 MeV or
less, in principle, occurs under the action of charged particles (electrons, protons, and a
particles) with an energy of 'k,1 MeV formed during the propagation of neutrons in air. It
has been experimentally shown [1-3] that the character of the excitation of luminescence in
air by protons and a particles with an energy of '\,1 MeV is roughly the same as in excitation
with electrons and the luminous efficiency n (ratio of luminous energy in the optical
to the absorbed energy of fast charged particles) does not depend
the ionizing particle. Thus, the intensity of neutron-stimulated
termined by the absorbed neutron energy.
Production of heavy charged particles (protons, a particles)
right during the interaction of neutrons with the nuclei of atoms
pheric air, the ratio of y-ray yield
dependence of the nuclear reactions.
protons and a particles are absorbed
rays runs into hundreds of meters so
on the type and
luminescence
in
range
energy of
air is ,de-
and y rays is possible
in the make-up of atmos-
to charged-particle yield being determined by the energy
Since their paths in air are ''l cm, we can assume that
right where they are produced. However, the path of y
that the absorbed y-ray energy is distributed over some
volume. The absorbed energy of a neutron consists of the absorbed energies of short-range
radiation (protons, a particles) Ip(r, 0 and long-range radiation (y rays) Iy(r, t). These
quantities are related to the neutron field N(r, c, t) by
Ip(r, t)==1 devN (r, e, t) cyi (e) E (e);
(r, t)==1 dcv521 cleN (e. 8, 0 ai (8) x 'a [E1(0, r--e, t--e=r7c,
(1)
(2)
where v and e are, respectively, the neutron velocity and energy; cri(E) is the macroscopic
cross section of the i-th reaction leading to the production of a charged particle or a y
ray with an energy of Ei(c); Ia(Ei, r, t) = Ia(Ei, 08(0; Ia(Ei, r) = A exp (-110[1 + Cur,'
exp (Dpr)]/4nr2 is the absorbed energy from a stationary, isotropic point source of y rays
with an energy of Ei [4]; and t' is the time at which the y ray is produced. The delay of
y rays is taken into account in Eq. (2). The summation over i in Eqs. (1) and (2) is per-
formed for all possible reactions. The neutron field N(r, c, 0 can be calculated, e.g.,
by the Monte Carlo method.
Note that luminescence in air can also be stimulated by recoil nuclei which arise in
the elastic scattering of neutrons with an energy of '?,2 MeV or more.
Neutrons with E4'1.; 0.45 MeV may be absorbed as a result of the radiative capture reac-
tion 14N(n, y)15N and the 14N(n, p) 14C reaction with the emission of a proton [5]. The en-
ergy release of these reactions are
E pr zoP pr iv); Env= nvEvE4/Eg.,
a
where E? y, E? t and E? are, respectively, the coefficients of radiative capture, total cap-
ture, and neutron capture with emission of a proton [6] (Epr/41y = 25 [5]); Qp = 0.62 MeV is
the energyof thereactionwith emissionof a proton [5]; ny = 2.25 is the average number of y
Translated from Atomnaya inergiya, Vol. 42, No. 5, pp. 407-408, May, 1977. Original
article submitted May 14, 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.
457
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Declassified and Approved For Release 2013/04/01: CIA-RDP10-02196R000700090005-4
rays and Ey= 4.96 MeV is the mean energy of y rays produced in the radiative capture of a
neutron [7]. On substitution of the numerical values, we get Epr = 0.6 MeV and Eny = 0.43
MeV.
We determine the luminous intensity of an elementary volume of air at the outlet of. a
thermal-neutron reactor. Taking f = 1012 cm-20sec-1 as a typical value of the thermal-neu-
tron flux, we obtain
I riEpr 10-9 W/ cm',
where n 10" [1-3]; and Za 100 m is the path of thermal neutrons to absorption. Such an
intensity can be recorded easily.
Let us consider the luminescence stimulated in air by a point isotropic source of neu-
trons with an initial energy of co -"s 0.45 MeV. Let us calculate the luminous intensity (we
call it the intensity of luminescence over a disk) recorded by a radiation detector whose
angle of collimation is formed by two conic surfaces with slightly different apex angles;
the line joining the center of the source and the detector is the axis of symmetry. Suppose
that the light detector is at a large distance (Ro) p-1) from the neutron source; then the
luminous intensity over a disk is determined by a light quanta emitted by a volume formed by
two infinite cylindrical surfaces of radii p and p + Ap with the source-detector line as the
axis of symmetry. The intensity of the luminescence over a disk, stimulated by protons, is
given by
Dp (p, [Epr NopApEopr/8aR8r exp { -Eat- r2/4T (t));
the intensity of luminescence over a disk, stimulated by y rays, is given by
Dy (p, (p, t) Dp (p, t),
where
(3)
(4)
00 00
f 42., 0 = dy exp (?y214) r dz ,
exp ( - f3z -x2/4) [1-+ Cwaxp pluA sh 72T 1/224- 02y2;
lin Jo 1/ x2-1-024-0.
a = 2nyEflyA/Eprqr; p 1; a = 11)/7; x pp; T(t) is the size of the delayed neutrons [6]; No
is the total number of neutrons emitted by the source. In writing Eqs. (3) and (4) we used
an analytic expression for the captured-neutron field which is proportional to exp -
r2/4T(t)//[47T(t)]3/2.
Next, we study Eq. (4). Let C = 0, i.e., we take no account of y-ray scattering. For
small x (x41(20) we expand the hyperbolic sine of Eq. (4) in a series, retaining the first
two terms. Upon integrating, we have
/(p, t)--= [71 +202 (I-1)]/12,
where
= exp (02) [1 - erf (M];
erf(B) is the error function. When 04 1, f(p, t) ilT8/2.
For large values of x 20, the hyperbolic sine of Eq. (4) can be replaced by exp(x)/
2. Using the saddle-point method to evaluate the integrals with respect to z [saddle point
z = (1/z2 02y2 _ /B 202) , J and with respect to y [saddle point y = 01, we have
f (p, Vrtx/2 [02/x (x-2132)1 exp f(x2/4[32)?02- xl.
Thus, visualization of the luminous field yields information aboutthe absorbed energy
and, for small x, also gives us an idea of the field of captured neutrons.
Let us evaluate the luminous flux by Eq. (3): (Er)1 0.06 sec [5] and No 1023 [8].
Let p = 300 m, Ap = 10 m, T = 20104 M2, Ro = 30 p, and t = 0.06 sec; then Dp 10-8 W/cm2,
which is perfectly measurable by present-day optical means.
LITERATURE CITED
1. C. Fan, Phys. Rev., 103, No. 6, 1740 (1956).
2. W. Borst and E. Zipf, Phys. Rev., 1A, No. 3, 834 (1970).
3. M. Hirsh, E. Poss, and P. Eisner, Phys. Rev., 1A, No. 6, 1615 (1970).
458
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4. M. N. Wrobel (Vrubel'), S. N. Sidneva, and A. S. Strelkov, At. Energ., 34, No. 1, 47
(1973).
5. V. M. Kuvshinnikov et al., in: Nuclear Constants [in Russian], No. 16, Atomizdat, Mos-
cow (1974), P. 53.
6. A. V. Zhemerev et al., At. Energ., 38, No. 3, 174 (1975).
7. Nuclear Data [in Russian], Atomizdat, Moscow (1969), p. 391.
8. H. Sandmeier, S. Dupree, and G. Hansen, Nucl. Sci. Eng., 48, 343.
EFFECT OF PRE-IRRADIATION ON OXIDATION OF ALLOY Zr + 2.5% Nb
M. G. Golovachev, V. I. Perekhozhev, UDC 621.039.531:669.296'293
V. E. Kalachikov, and 0. A. Golosov
Oxidation of the Zr + 2.5% Nb alloy is usually intensified by the action of radiation
from a nuclear reactor [1]. To understand the nature of this effect it is desirable to con-
sider the influence of pre-irradiation on subsequent oxidation of Zr + 2.5% Nb.
The studies were carried out with specimens which had undergone quenching from 840?C
with subsequent 40% deformation and annealing at 550?C for 5 h in a vaccum no worse than 5?
10-5 mm Hg. Before oxidation, some of the specimens were irradiated in the active zone of
a reactor at 80?C with a fluence of 2.6.1015 neutrons/cm2 (E 1.1 MeV). The effect of the
irradiation on the properties of the alloy was evaluated by the changes in the electrical
resistivity, measured at room temperature on specimens with dimensions of 50 x 2 x 0.5 mm.
Isochronal annealing of the irradiated and unirradiated specimens was carried out in a vacuum
furnace over a range of 50?C. The holding time at each temperature was 2 h since it was es-
tablished in preliminary experiments that even at a low temperature (170?C), annealing for
1.5 h was sufficient for practically complete restoration of the initial resistivity.
The variations in the electrical resistivity of the irradiated and unirradiated alloy
during the process of isochronal annealing are plotted in Fig. 1. The error in determining
the resistivity of the alloy was 0.5%. Each experimental point corresponds to the mean of
measurements of two specimens.
On the basis of data concerning the variations in the electrical resistivity under iso-
chronal annealing of the irradiated alloy, we chose three oxidation temperatures at which the
resistivity of the irradiated alloy is greater than (250?C), almost equal to (330?C), and
less than (400?C) that of the unirradiated alloy. Specimens measuring 30 x 5 x 1 mm were
oxidized in a stream of moist (5-7 vol.% vapor) of commercially pure nitrogen with an overall
gas flow rate of 5-10 liters/h. The specimens were weighed on VLAO-100 scales with a sen-
sitivity of 5010-5 g. The increases given in Table 1 for each temperature are averages
for 3 or 4 specimens.
TABLE 1. Increase in Weight of Zr +
2.5% Nb Alloy after 100 h of Oxidation,
m8/dm2
Testing temp., ?C
State of alloy
250
330
400
Unirradiated
2,0+0,9
7,2+1,5
16,1+1,3
Irradiated with fluence
of 2.6 .1012 neutrons/
cm2 (E 1.1 MeV)
4,5-T1,3
8,1-T1,3
10,7+1,2
'
Translated from Atomnaya Energiya, Vol. 42, No. 5, p. 409, May, 1977. Original article
submitted July 13, 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.
459
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10

*
?
0
100 200 300 400 ?C
Fig. 1. Relative variation
of resistivity of Zr + 2.5%
Nb alloy during process of
isochronal annealing: A)
irradiated witha fluence
of 2.6.1019 neutrons/cm2;
0) unirradiated.
When the temperature is raised from 250 to 400?C the process of oxidation is intensified
for both unirradiated and irradiated specimens. However, the effect of pre-irradiation on
the oxidation of the alloy is not unique at each temperature investigated. At 250?C the ir-
radiated specimens display a larger increase in weight than do the unirradiated. At 330?C
this difference practically does not exist and at 400?C the pre-irradiated specimens are more
resistant to oxidation than are the unirradiated.
A correlation can thus be observed between the variations in the resistivity and the
resistance to oxidation for irradiated specimens of the Zr + 2.5% Nb alloy; a lower resistance
of oxidation corresponds to a higher electrical resistivity.
A multitude of radiation-induced defects are formed in the crystal lattice of the alloy
in the process of irradiation. During annealing, the defects either form complexes or mi-
grate to dislocations, grain boundaries, etc. However, some radiation-induced defects are
preserved up to a temperature of 300-350?C, as evidenced by the increased electrical resis-
tivity. This increased concentration of defects (in comparison with the unirradiated mate-
rials) evidently leads to accelerated oxidation of the alloy.
Annealing of irradiated specimens at 400?C probably gives rise to a further stabiliza-
tion of the alloy structure. This, as is known [2], enhances the oxidation stability, which
proves to be higher than in unirradiated specimens.
LITERATURE CITED
1. Sueo Nomura, Nippon Genshiryoku Gakkaishi, 11, No. 6, 353 (1969).
2. J. Le Surf, in: Proc. Symp. ASTM (STP) Applications-Related Phenomena for Zirconium
and Its Alloys, Philadelphia (1969), p. 286.
460
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Declassified and Approved For Release 2013/04/01: CIA-RDP10-02196R000700090005-4
DETECTION OF START OF BOILING OF LIQUID METAL COOLANT
K. A. Aleksandrov, V. A. Afanas'ev, UDC 621.039.534.6
N. G. Gataullin, and V. V. Golushko
One method of detecting when the coolant starts boiling in the core of fast reactors is
an acoustic method [1], based on the recording of noise from forming and "collapsing" vapor
bubbles. The principal difficulty in using this method in power reactors is attendant upon
the presence of an intense noise background generated by the work of the pumps and the flow
of the coolant. It is, therefore, necessary to make the proper choice of the frequency range
of detection and the best signal-to-noise ratio while preserving an adequate intensity of the
useful signal.
Detection of the start of boiling of sodium with various sensors was studied in a BOR-
60 reactor. The assembly used incorporated a boiler and acoustical transducers: of thewave-
guide type (steel rod 10 mm in diameter and 7 m in length with a piezoelectric element of
TsTS-19 ceramic attached to the upper end) and of the immersible type (with a piezoelectric
element made of lithium niobate [2] and TsTS-19 ceramic). The diameter and thickness of the
piezoelectric element of the first immersible transducer are 10 and 3 mm and those of the
second are 12 and 1.5 mm. The assembly was installed above the core: the transducers were
at the level of the packet heads at a distance of 250 mm from the boiler.
The operation of the boiler is based on the passage of an alternating current through
a tube containing sodium; the volume of the sodium being heated is 3 cm3. During boiling
the vapor bubbles leaving the tube "collapse" in the volume of underheated sodium. The in-
stant at which boiling begins is monitored by an abrupt drop in the supply current. Boiling
was started 25 times at various coolant flow rates through the reactor.
Studies of the intrinsic noise of the BOR-60 reactor showed that the effective value of
the noise amplitude is related only to the variations in the coolant flow rate; the shape of
the spectrum remains practically unchanged. Temperature and power variations over a broad
range do not appreciably affect either the amplitude or the spectrum of the noise, which is
in agreement with the data of [1]. There is practically no reactor noise at frequencies above
100 kHz.
The experiments were carried out at zero reactor power and a temperature of 210?C. This
made it possible to simplify the design of the assembly and to use a piezoceramic immersible
transducer with a high sensitivity. In addition to the acoustic transducers mentioned above,
in the experiments we also used a clamping transducer with a piezoelectric element of TsTS-
19 ceramic; this transducer, measuring 10 mm in diameter and 1.5 mm in thickness, was set up
on the reactor lid. This transducer recorded a higher level of noise background at all fre-
quencies than any of the other transducers did.
The signals from the piezoelectric transducers were transmitted to a wideband preampli-
fier from which they went to the main amplifier via a high-frequency filter with a "cutoff"
frequency of 50 kHz. Then the signals were fed into a random-process analyzer [2] for ampli-
tude, time, and parallel frequency analysis in the frequency range 20-500 kHz.
In the case of all measured spectra of boiling noise, in order to prevent signals being
recorded in the absence of boiling the discrimination level of the instrument was set slightly
above the level of background noise recorded by the transducer on the lid at the maximum flow
rate. Figure 1 shows the boiling noise spectra recorded by the various transducers at zero
and maximum rate of coolant flow through the reactor. The coordinates of the individual
Translated from Atomnaya fnergiya, Vol. 42, No. 5, pp. 410-411, May, 1977. Original
article submitted July 13, 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.
461
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30
25

0
50 100 1 5 0 200 250 300 350
Frequency. kHz
Fig. 1. Boiling noise spectra of sodium: 0)
clamping transducer of lithium niobate (zero flow
rate); 0) waveguide-type transducer (flow rate
960 m3/h); A) clamping transducer on reactor lid
(zero flow rate); A) piezoceramic clamping trans-
ducer (flow rate 960 m3/h).
00 4 5 0
points are determined by the mean frequency of the resonance filters of the analyzer and the
time-averaged measurements of the signal amplitude, referred to the discrimination threshold
which was the same in all experiments.
Signals of boiling are recorded most successfully by clamping transducers (with piezo-
electric element of TsTS-19 ceramic). It should be noted that boiling can be recorded even
by transducers set up on the reactor lid. For the immersible transducer the ratio of the
effective value of the signal during boiling to the background noise in the frequency range
above 50 kHz is no less than 30 dB at maximum coolant flow rate. This ratio grows as the
frequency increases.
The measurements showed that to detect when sodium begins to boil in a reactor the most
acceptable frequency range is 80-300 kHz, where the maximum signal-to-noise ratio is observed.
The use of piezoelectric transducers with resonance at frequencies in this range makes it
possible to isolate the useful signal more reliably. The boiler of simple design used in the
experiments can be a convenient means of adjusting and monitoring acoustical systems for the
diagnosis of the onset of boiling by sodium coolant in a fast reactor.
LITERATURE CITED
1. V. M. Baranov, Ultrasonic Measurements in Atomic Engineering [in Russian], Atomizdat,
Moscow (1975).
2. K. A. Aleksandrov et al., NIIAR Preprint P-9(275), Dimitrovgrad, Bulgaria (1976).
462
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DIRECT ENERGY CONVERSION OF MONOENERGETIC ION BEAMS WITH
SPACE-CHARGE COMPENSATION
O. A. Vinogradova, S. K. Dimitrov,
A. S. Luts'ko, V. M. Smirnov,
and V. G. Telikovskii
UDC 621.039.6
The injection of fast atoms into a plasma can serve as one of the methods for plasma
heating in tokamaks [1]. Estimates show [2] that a sufficiently high energy efficiency
0.7) can be obtained if direct conversion of the energy of ions that are not neutralized in
the neutralizer is employed.
This paper discusses a converter using compensation of the ion space charge by an elec-
tron flux in crossed magnetic and electric fields (E x H) which makes it possible to convert
the energy of comparatively dense ion beams. The device is more compact than those previously
proposed. At the present time, several alternative locations of the converter in the injec-
tion system are under consideration. It was suggested [2] that the beam of unneutralized
ions be deflected by about 90? by means of a bending magnet and the energy then converted.
A beam can be deflected by means of the magnetic field of a tokamak. A system for conversion
of ion energy along the flow channel was also proposed, i.e., the so-called in-flight con-
version.
It is expedient to use the converter with E x H fields discussed here in an arrangement
with a bending magnet. The instrument is placed immediately behind the magnet and makes it
possible to convert the energy of a beam at the same density as that with which it merged
from the magnet. In this case, there is no necessity for expansion of the beam to a low den-
sity since the space charge is compensated by electron fluxes in the converter with E x H
fields. The transverse dimensions of the flow are relatively small, and a lens, rather than
a grid, can be used to remove electrons from the beam, which makes it possible to increase
the energy flux density considerably.
The scheme for conversion with E x H fields is shown in Fig. 1. The ion flux is encom-
passed by loops, one side of which is an electron emitter (E) and the other side a collector
(C). The potentials (pi, (P2, ..., coN are applied to the loops; the potentials increase lin-
early with distance in proportion to the approach to the ion collector, which consists of
jalousie louvres at the stopping potential cp.r.
Power losses during stopping of a dense beam of high-energy ions in such a system (ni =
109-1011 cm-3 is the density of the ion beam and Wi = 100-200 keV is its energy) were esti-
mated in this work. These values of density and energy are typical of ion fluxes in the
N
v
____
ig
-
vdr
E
Fig
1 '
\
Adi
1w- -
Ad 11.'
=Ili11
L
N
?_
collector
Fig. 1. Diagram of
converter with E x H
fields.
Translated from Atomnaya gnergiya, Vol. 42, No. 5, pp. 411-412, May, 1977. Original
article submitted July 30, 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.
463
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1,6
1,2
0,8
-0,4
2 U. kV 0 ZCO 400 600 800 1000 EI, Oe
7
2
Fig. 2 Fig. 3
Fig. 2. Dependence of collector current on collector potential for H = 0
(1) and H = 250 Oe (2).
Fig. 3. Dependence of collector current on magnetic field intensity for
Uc = 0 (1) and Uc = 2.85 kV (2).
injector of thermonuclear reactors based on a tokamak or open trap. In the dependence on the
energy and density of the ion beam, limitations are imposed on the parameters of the conver-
ter (H, external magnetic field; L, length of stopping space; S and d, tfansverse dimensions
of the ion beam along the drift velocity vdr and along H respectively; Z, distance from the
emitter loop to the ion beam) which are determined by the following conditions [3]: small-
ness of the stopping space in comparison with the ion Larmor radius, al = rL/rLi 0.35 MeV). The level of the natural background is indicated
by a dashed line. The migration of nuclear-explosion products was measured. This effect in
the surface layer of the soil vegetation cover can be calculated from the difference between
the sum of the cumulative atmospheric fallout of the various isotopes formed during a nuclear
explosion and the surface 0-activity of the soil actually observed. The calculations (Table
1) were based on experimental data [1, 2] and the ratio of the activity of short-lived iso-
topes over a period of 1 year [6]. Table 2 gives the numerical values of the fraction Adet/A
Translated from Atomnaya Energiya, Vol. 42, No. 5, pp. 413-414, May, 1977. Original ar-
ticle submitted November 29, 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.
465
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Declassified and Approved For Release 2013/04/01: CIA-RDP10-02196R000700090005-4
ILADLEJ L. Auci?v?soLupic uomposJ.cion 01
Soil Pollution in Summer of 1963
Radio-
nuclide
A, pCi/ Adegi
km2 A. ?10
Radio-
nuclide
A. ?Ci/Adet/
km2 A, 0/0
841kIn
45
0
125Sb
107
4,2
89Sr
36
51,5
127Te
30
12,5
90Sr
33
4,1
137Cs
56
6,3
oy
33
67,4
141ce
7
3,0
Ny
80
54,0
"'lee
200
0
95zr
250
1,1
144Pr
200
74,0
95Nb
250
0
147pm
185
0
io3Ru
25
0,4
15iSin
3
0
i?8Ru
465
0
i55Eu
25
0
106RU
465
77,0
200
100
1063 1964 1955 1965 1957 Yr
100
50
TABLE 2. Weighted Mean Depth of Penetra-
tion of Nuclear-Explosion 0 Products in-
to Soil,.d mm
Yr
Calc.
Expt.
Yr
Calc.
Expt.
1963
3
4
1966
7
10
1964
5
7-8
1967
12
10-14
1965
4
9-10
Fig. 1. Hard 0 radiation
from soil surface: 1)
variation with time; 2)
attenuation A0/0 due to
vertical migration of
isotopes.
of surface activity, where Adet is the 0 radiation detected and A is the excess of the iso-
tope in the soil. The calculation of the accumulation of hard 0-emitters on the soil sur-
face for the middle of each year enables us to find the attenuation of the radiation owing
to the penetration of isotopes into the soil with time. A large proportion of 0-activity
with Emax > 0.35 MeV is absorbed by the soil and the radiation drops by almost an order of
magnitude in 5 years as a result of vertical migration (curve 2 in Fig. 1).
Let us find the depth to which nuclear-explosion products penetrate into the soil. The
profile of their concentration is described by an incomplete y function [4], but in a number
of cases or in the surface layer of the soil it is quite close to the exponential function
[4, 5] with exponent a. It is not difficult to show that in this case the weighted mean
depth of penetration of nuclear-explosion products into the soil ("excess center") is given
by
A-13
d= 1/a=
(1)
where A and 0 are, respectively, the calculated and observed excess of nuclear-explosion
products in the soil, the observed value being found from the R radiation; p = 6.10-3 cm2/mg
is the coefficient of 0-particle absorption by the soil, equivalent to the 0-particle absorp-
tion by the real mixture of isotopes. Table 2 gives the values of d calculated by Eq. (1)
and the experimental values [5]. They are seen to be in satisfactory agreement. Thus, the -
weighted mean depth of the penetration of 0-emitters into the soil increases at the rate of
several millimeters per year (without corrections for the decay of isotopes in the process
of migration.
LITERATURE CITED
1. N. V. Vasil'eva et al., Tr. Inst. tlektromekh., No. 25, 166 (1972).
2. K. P. Makhon'ko et al., Tr. Inst. Electromekh., No. 6(64), 64 (1976).
3. F. I. Navlotskaya, Migration of Radioactive Products of Global Fallout in Soils [in Rus-
sian], Atomizdat, Moscow (1974).
466
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Declassified and Approved For Release 2013/04/01: CIA-RDP10-02196R000700090005-4
4. K. P. Makhontko, in: Radioactive Isotopes in Soils and Plant 8 [in Russian], kblbs,
Moscow?Leningrad (1969), p. 48.
5. V. F. Brendakov et al., Tr. Inst. Elektromekh., No. 5, 143 (1970).
6. M. P. Grechushkina, Tables of Composition of Products of Instantaneous Fission of 2351J,
238U, and 233PU [in Russian], Atomizdat, Moscow (1964).
ALLOWANCE FOR FLUCTUATIONS OF RADIATION FLUX IN
ACTIVATION ANALYSIS .
Pham Zui Hien UDC 543.53
In activation analysis experiments the effect of radiation flux is usually eliminated
by simultaneous irradiation of a specimen and a standard or by the use of an external stan-
dard [1]. Although these methods, in principle, ensure good accuracy, they do have short-
comings which limit the efficiency of the method of analysis and cannot always be used suc-
cessfully, especially in cases when short-lived isotopes are formed. This paper gives a
simple method of determining the induced activity with allowance for the fluctuations of the
radiation flux during irradiation. The source of radiation used was a neutron generator
with an integrated yield of the order of 1010 particles/sec. The neutron monitor consisted
of a long counter with an SNM-11 detector. With the aid of scalers controlled by a timer
(or time analyzer) the monitor also recorded the time distribution of the intensity of the
neutron generator during irradiation. The activity was calculated by summing the activities
induced by neutron irradiation in every channel of this time distribution. It is easily
shown that the solution of the activation equation can be written as
(1)
The quantity AI is the part of the activity associated with the mean value of the neutron
flux:
i1r=aA711--expOb (2)
where A is an isotope decay constant; t is the irradiation time; N is the number of atoms of
the element being determined; f is the average number of neutrons per channel; and a is a
coefficient. The quantity 6 in Eq. (1) is a correction associated with the neutron flux
fluctuations during irradiation:
6
e
-"
[(51,1(extin_f)+6/2 (e2Wri_ektin+ +6in ?
( --e )1.
f e-
Here, n is the number of channels; fi is the number of neutrons detected in the i-th channel;
and6f1=fi--f=f1.?(1/r)Ef.(1] 6f1 = 0). If neutron flux fluctuations are neglected
(6f1 = 0) and if n = 1, we have 6 = 0. In the general case, the neutron flux fluctuations
are taken into account by calculating the correction 6 and using Eq. (1) and the experimental
value of A to find Al which, according to Eq. (2), is proportional to the number of atoms of
the element being determined. Clearly, the more channels, the more accurate the calculation
of 6 by Eq. (3). However, in practice it is always possible to calculate the correction by
Eq. (3) with sufficient accuracy with a small number of channels.
By way of illustration we give the results of making allowance for neutron flux fluc-
tuations during activation of silicon, aluminum, and iron in bauxites in the reactions "Si(n,
p)28A1 (T1/2= 2.31 min), 25A1(n, p)27Mg (T1/2 = 10 min), and "Fe(n, p)55Mn (T1/2 = 2.58 h)
[2]. Each specimen and standard was irradiated for 8 min, the counter recording the number
Translated from Atomnaya Energiya, Vol. 42, No. 5, pp. 414-415, May, 1976. Original
article submitted June 11, 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.
467
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Declassified and Approved For Release 2013/04/01: CIA-RDP10-02196R000700090005-4
TABLE 1. Neutron Flux Fluctuations in
an Irradiation
Channe
fi
(X10-1)
Of ?If, %
i
Chan-
nel
ri,
(x10-1)
VW, %
/
0-1
10 935
2,88
4-5
10 715
0,08
1-2
10 980
3,28
5-6
10 428
?1,9
2-3
10 701
0,65
6-7
10 379
?2,35
3-4
10 525
?0,98
7-3
10 378
?2,35
TABLE 2. Calculated Corrections d
n
ssAl
27Mg
56111T1
1
0
0
0
2
?0,0079
?0,0022
0
4
--0,011
--0,0031
' 0
8
?0,0115
?0,0035
0
of neutrons per minute. This permitted the correction to be calculated with n = 1, 2, 4,
and 8. Table 1 gives the experimental data on neutron flux fluctuations in an irradiation.
The "fluctuation amplitude" (dfi/f)max is 3%. The results of calculations of d are given
in Table 2. As the half-life of the isotope decreases, the correction increases and becomes
significant in comparison with the other experimental errors (1.2% for 26A1). It is also
seen from Table 2 that to obtain an accuracy of better than 0.1% for the isotopes "Al, "Mg,
and 66Mn it is sufficient to calculate the correction from Eq. (3) with n = 4-8.
LITERATURE CITED
1. D. De Soete, R. Gijbels, and J. Hoste, in: Chemical Analysis, Vol. 34, Wiley-Inter-
science, New York (1972), Chap. 10, p. 449.
2. B. S. Dzhelepov and L. K. Peker, Decay Schemes of Radioactive Nuclei. A < 100 [in Rus-
sian], Nauka, Moscow (1966).
ACTIVATION OF ELEMENTS IN (y, yl) REACTION BY I6N y RAYS
U. Akbarov, U. Uzakova, UDC 639.144.7
and K. Umirbekov
Activation determination of elements by isomers formed in the (y, y') reaction has been
the subject of a number of papers. In most of the papers the y sources are electron acceler-
ators [1, 2], which are expensive installations. Some of the papers discussed work in which
,
y quanta from 6oco [3j, I16mIn [4], and 24Na [5], with a maximum energy of 1.33, 2.17, and
2.75 MeV, respectively, were used for the (y, y') reaction. However, because of the low
values of these energies for obtaining isomers with a measurable activity, it is necessary
to use high-intensity y sources and large-volume specimens. Taking this circumstance into
allowance, we used radioisotopes with more energetic y rays to obtain isomers in the (y, y')
reaction. Among artificial radioisotopes the highest y-ray energy is possessed by I6N which
has a half-life of 7.4 sec and emits y rays of 6.14 and 7.11 MeV (69 and 4.9%). This isotope
can be formed from nitrogen, oxygen, and fluorine under the action of neutrons in I5N(n,
16N, 160(n, p)I6N, and "F(n, a)I6N reactions. The "F(n, a)I6N reaction was chosen as the
most convenient for using reactor neutrons.
To do this, a fluorine compound (containing 100 g fluorine) was packed in a cylindrical
vessel with a well and was irradiated for 20 sec in the channel of a VVR-SM water-moderated ?
water-cooled reactor. The I9F(n, a)I6N reaction was observed under the action of the fast
neutrons of the reactor. After irradiation the vessel containing the fluorine was removed
from the neutron field of the reactor within 1-2 sec and placed on the element under analysis
from which an isomer is formed in the (y, yl) reaction. In view of the short half-life of
I6N the study was carried out mainly on elements whose isomers have a half-life of up to
1 min.
Translated from Atomnaya t'nergiya, Vol. 42, No. 5, pp. 415-416, May, 1977. Original
article submitted September 8, 1976.
468
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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Declassified and Approved For Release 2013/04/01: CIA-RDP10-02196R000700090005-4
TABLE 1. Nuclear Characteristics and
Sensitivity of Determination of Ele-
ments
Element
Irradia- I
don
time, sec
Isomer
c.)
Measur-
ingtime,
se
1
Sensitiv-
ity. 8
Selenium
16
77771Se
19
160
40
2.10-2
Yttrium
15
nonY
16
915
30
10-i
Bromine
8
79mBr
5
210
10
:40-2
Silver
22
107mAg
45
93
90
2.10--2
109mAg
40
88
Hafnium
16
172m11 f
19
215
40
2?10-
Tungsten
9
183nINV
5,5
105
10
5.10-2
Gold
11
197m
7,4
279
15
10-2
After irradiation with y rays, the target element was delivered within 5 min to the
measuring apparatus which consituted a scintillation y spectrometer with NaI(T1) crystal with
well. The activity-measuring time was double the half-life of the isomer of the element un-
der analysis. The nuclear characteristics of the isomers studied and the sensitivity ob-
tained for each element, i.e., the minimum quantity of element for which a measurable iso-
mer activity is produced, are listed in Table 1. The test of a measurable activity is the
activity which corresponds to the number of counts in the photopeak of the analytical line
of the isomer and is equal to 3/RIT), where Nb is the number of background counts under the
photopeak.
The studies point to the potentialities for achieving a high sensitivity in the analy-
sis of elements by means of the (y, yl) reaction by using the y rays of I6N formed from
fluorine under the action of reactor neutrons.
The sensitivity values given in Table 1 were obtained at a neutron flux density of 1.8.
1013 neutrons/cm2osec, where the fraction of neutrons energetic enough to initiate the 13F(n,
a)16N reaction does not exceed 10%. Consequently, the sensitivity of obtaining isomers
rises significantly at a flux density with a high content of energetic neutrons. The sen-
sitivity can be raised by using fluorine in large quantities by cycling irradiation of fluor-
ine and the target element with subsequent cyclical measurement of the activity of the iso-
mer formed, and by obtaining I6N by irradiating fluorine circulating in a loop.
LITERATURE CITED
1. J. Otvos et al., Nucl. Inst. Methods, 11, 187 (1961).
2. S. Kodiri and L. P. Starchik, Zavod. Lab., No. 2, 191 (1970).
3. A. Veres, Acta Phys. Sci. Hung., 16, No. 3, 261 (1963).
4. I. A. Abrams and ?L. L. Pelekis, in: Methods and Application of Neutron-Activation Analy-
sis [in Russian], Zinatne, Riga (1969), p. 71.
5. I. A. Abrams and L. Lakoshi,Izv. Akad. Nauk Latv. SSR, Ser. Fiz. Tekh. Nauk, No. 6, 3
(1969). .
469
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Declassified and Approved For Release 2013/04/01: CIA-RDP10-02196R000700090005-4
IMPACT TOUGHNESS OF STRUCTURAL GRAPHITE
Yu. S. Virgil'ev, V. V. Gundorov, UDC 621.039:532.21
and V. G. Makarchenko
Tensile strength cannot serve as a full-valued strength criterion in the design of struc-
tures.in which cracks may form [1]. Therefore, to evaluate the work capacity of graphite
prone to brittle fracture, it is advisable to use the characteristics of viscoelastic prop-
erties, especially impact toughness whose determination is one of the principal forms of
mechanical testing [2]. This paper considers the impact toughness of domestic graphite struc-
tural materials and its relation with compressive strength.
The law of similarity is not satisfied in tests of brittle materials for impact tough-
ness; therefore, the shape and absolute dimensions of specimens should remain unchanged [2,
3]. Moreover, the significant scatter of results necessitates a quite large number of tests.
Some data [4] from tests of prismatic specimens indicated that the ratio an/ac is lower in
stronger graphite than in graphite of lower strength. Irradiation of graphite resulted in an
increase in the impact strength [5].
Variations in the impact toughness are related [6] to variations in other strength char-
acteristics, the modulus of elasticity and bending strength, by
Oa ? OEIE +2 acrb/ab.
In this study impact toughness was determined on an MK-0.2 pendulum impact testing ma-
chine with specimens measuring 8 mm in diameter and 10 mm in height. The specimens were
mounted on the base of the machine with a holder which ensured that the pendulum would hit
the specimen in the middle. The impact toughness was found from the formula
(Q//0.785d2) cos a,
"
where Q is the pendulum mass, equal to 1.136 kg; 7 is the pendulum length, 20 cm; d is the
specimen diameter, 8 ? 0.1 mm; and a is the angle of deflection of the pendulum, in deg.
The average values of 15 determinations was taken for the results of measurements. The
error of measurement did not exceed 10%. In the case of the strong materials KPG and VPG,
an. kgf ? cm/cm2
20
10
200
400 600 800 1000
ac, kgficm2
Fig. 1. Relation between impact
toughness and compressive strength
(a) before and (b) after irradiation
in the range 70-300?C (the value
given for ER specimens is the average
for impact toughness for parallel and
perpendicular specimens): 0) GMZ;010
GMZ after irradiation;N7, A) VPG; A)
ZOPG; 0) ER; and 0) MPG-6.
Translated from Atomnaya fnergiya, Vol. 42, No. 5, pp. 416-418, May, 1977. Original
article submitted September 3, 1976.
470
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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Declassified and Approved For Release 2013/04/01: CIA-RDP10-02196R000700090005-4
TABLE 1. Properties of Materials Tested
30
bZG
Grade
Matrix
Method of
shaping
No. of
soak-
ings
with
pitch
Temp.
of
treat-
ment,
?C
Density,
g/cm3
Co.mpressive
strength, kgf/
cm2 ?
ER and
variants
GMZT
GMZ*
VPG
KPG
MPG
Natural
graphite
Calcinated
petroleum
coke
The same
* *
Uncalcinated
petroleum
coke
The same
Pressing
Extrusion
The same
D
Pressing
The same
Up to 3
None
None
2 .
None
None
2300
2300
2303
2800
2400
2600
1,56-1,70
1,65-1,70
1,65-1,69
1,80-1,85
1,75 --1,85
1,80
70-370
150-350
350-450
300-350
300-420
600-700
400-500
600-800 *.
1000 **
? The numerator is the value for a specimen cut parallel to axis of blank and the de-
nominator, perpendicular to the axis.
-(anTilnit
o 0
_
0
1 o 0
0 00
0
10 ---=i--r
1018 log MW
F, neutrons/cm2
Fig. 2
an, kgf ? cm/cm2
WV 0 200 400 600
ob,kgf/cm2
Fig. 3
800 1000
Fig. 2. The dependence of the impact toughness of graphite on the fluence at
temperatures of 70-300?C: 0) VPG (perpendicular cut of specimens); 0) KPG.
Fig. 3. Relation between impact toughness and bending strength of VPG graphite
up to (v, A) and after (I, A) irradiation: A,y and A, A) parallel and perpen-
dicular axes of shaping of blank for cutting specimens.
we determined the bending strength instead of the compressive strength. Graphite materials
that differed substantially as to strength were tested (see Table 1).
To determine the effect of neutron irradiation on impact toughness, specimens of GMZ,
VPG, and KPG graphite 8 mm in diameter and 50 mm in length were irradiated in the tempera-
ture range from 70 to 500-550?C with a fluence of 1019 to 1021 neutrons/cm2 with E 0.18
MeV. The irradiated specimens were tested together with unirradiated control specimens. The
bending strength was first determined and then specimens of the dimensions given above were
prepared from the fractured specimens. Half were tested under compression while the other
half were tested for impact toughness. In view of the considerable radiation hardening only
the bending limit was found for graphite VPG and KPG.
A tentative comparison was made of the resillts of parallel determination of the impact
toughness of prismatic specimens on an MK-0.5 tester and cylindrical specimens on an MK-0.2
tester. It was found that with an error not exceeding ?40% there is a direct proportionality
(with a coefficient of 0.2) between the values obtained for the impact toughness by the two
methods. The difference was due to the fact that when more massive specimens were used, a
large proportion of the pendulum energy was carried off by flying fragments.
The impact toughness is anisotropic, being higher in the direction of the pendulum blow,
which is perpendicular to the preferentially oriented (001) plane. This can be seen from
the results of tests of anisotropic ER graphite whose strength was changed by impregnations
with pitch alternating with annealings.
471
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Declassified and Approved For Release 2013/04/01: CIA-RDP10-02196R000700090005-4
For all specimens tested the impact toughness and the compressive strength were found
to be directly proportional, with a proportionality factor of q,3?10-2. However, as the
strength grows above roughly 600 kgf/cm2 the direct proportionality is not conserved and
the increase in the impact toughness ceases (Fig. la). Some data give reason to suppose
that a deterioration in the perfection of the crystal structure leads to a decrease in the
factor of proportionality between an and an. ,
,k.
Under the effect of neutron irradiation, the impact toughness increases, beginning from
a fluence of '1,1019 neutrons/cm2. After a fluence of '1,2.1019 neutrons/cm2 the change in the
impact toughness because of the effect of radiation, which is more pronounced in high-strength
graphite (Fig. 2), practically levels off. Since other strength characteristics, including
compressive strength [5], also increase under neutron irradiation, the ratio between the im-
pact toughness and these strengths is not affected (see Fig. lb) up to an kgf/cm2.
A direct proportionality also exists between the impact toughness and the bending
strength, although in a limited interval (Fig. 3). Comparison of the straight lines in Figs.
1 and 3 reveals that the slope tends to decrease for irradiated graphite.
LITERATURE CITED
1. J. Amesz and G. Volta, in: Proceedings of the Eleventh Biennial Conference on Carbon,
Galtinburg, June 4-8 (1973), MP-5.
2. V. N. Barabanov, S. E. Vyatkin, and N. A. Lobastov, in: Structural Materials Based on
Graphite [in Russian], No. II, Metallurgiya (1966), p. 135.
3. N. A. Shaposhnikov, Mechanical Testing of Metals [in Russian], Mashgiz, Moscow--Lenin-
grad (1954).
4. V. P. Sosedov (editor), Properties of Structural Materials Based on Carbon [in Russian],
Metallurgiya, Moscow (1975).
5. Yu. S. Virgil'ev, At. Energ., 36, No. 6, 479 (1974).
6. R. E. Nightingale (editor), Nuclear Graphite, Academic Press, New York (1962).
CsI(T1) WELL-DETECTORS FOR LOW-BACKGROUND y SPECTROMETRY
O. P. Sobornov UDC 539.1.074
Well-detectors are of interest in the analysis of small amounts of radionuclides in
samples of small volume (5-30 cm3).
This paper compares spectrometric parameters, intrinsic background, and photoefficiency
of CsI(T1) and NaI(T1) detectors 100 x 100, 80 x 80, and 60 x 60 mm in size (diameter x
hef.ght) with wells 33 (35) mm in diameter and 60, 50, and 47 mm in depth, respectively. Mea-
surements were made under identical conditions using a single FEU-1l0 photomultiplier and
the equipment described in [1]. The detectors 100 x 100 mm in size had conical quartz light-
pipe--window combinations 20 mm thick and the other windows were of quartz or sodium glass.
The outer diameter of samples in the wells of all detectors was 30 mm. A point source (135Cs)
and distributed y sources [40K (crystalline KC1), 137Cs, 232T n, and 238U (with decay prod-
ucts)] were used. The latter were distributed in a Na3PO4 medium (analytic grade) and had
a density of 1.3-1.4 g/cm3 for a source size of 27 x 50 mm. Resolution was determined with
distributed y sources of 137Cs and 40K (Rcs, RK, %) and also with a point source of 137Cs
(R, %) at a distance equal to the diameter of the detector using standard techniques. In
this paper, "resolution" is understood to mean the resolution of the spectrometer, which is
determined by the equation
Translated from Atomnaya Energiya, Vol. 42, No. 5, pp. 418-421, May, 1977. Original
article submitted September 14, 1976.
472
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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Declassified and Approved For Release 2013/04/01: CIA-RDP10-02196R000700090005-4
'TABLE 1. Parameters of CsI(I1) and NaI(T1) Well-Detectors
Detector
Resolution,
%
Background count. rate (counts/
min) in interval A E y. MeV
ep, %
& and M(rel. units)in interval AE y, Me
0,55-0,751,35-1,55
1,70-2,00
2,50-3,40
e
M
a
M
C
M
C
M
Size, mm
Type
R
Rcs
Ric
0,40-
3, 40
0,55-
0,75
1,35-
1, 55
1,70-?
2, 00
2,50-
3,40
100 x 100;
well
33 x 60
Csl (T1)
NaI (T1)
11,0
9,4
10,7
9,4
8,1
7,1
304,0
(155,8)*
108,2
163,2
(45,6)
25,5
8,00
(7,30)
5,76
5,35
(5,52)
4,44
5,64
(5,76)
4,80
19,14
14,20
4,62
3,18
0,95
(1,79)
1,66
6,17
4,94
2,66
2,51
7,80
5,52
3,40
2,63
7,38
5,60
3,89
3,20
80 x 80;
well
35 x 50
CsI (T1)
NaI (T1)
10,6
8,8
10,1
8,4
7,8
6,5
89,3
71,4
24,9
16,6
4,42
3,78
3,28
2,64
3,61
2,88
10,00
7,62
2,60
2,08
1,37
1,34
3,33
2,54
1,94
1,60
3,67
2,65
2,04
1,64
3,80
2,82
2,50
2,08
60 x 60;
well
35 x 47
Csl (T1)
NaI (T1)
16,-
7,7
10,6
8,0
8,2
6,0
39,3
29,5
10,0
7,04
2,28
1,50
1,27
1,01
1.80
1,56
4,96 1,60
3,001,0
1,34
1,0
1,65
1,0
1,33
1,0
1,70
1,0
1,51
1,0
1,74
1,0
1,63
1,0
*Second detector of similar size.
Counting rate, rel. units
40
30
20
10
1-
Back-
ground
40K
1461
1620
A
0 0 0
0
00
1
2615-511
2615
/1
I
I
II '1 ' I I
I k I II
/
?
/ ? I
? / /
\ /
?/
0 20 40
60 60
Channel No.
1100
?
120
208T1
3198
140
Fig. 1. Spectra from background and from distributed y emitters
23 -
2Th (with decay products) and 40K (KC1). Measurement geometry
A and B: ---) 76 x 76 mm CsI(T1) scintillation unit, 80 x 80
mm CsI(T1) detector with 35 x 50 mm well; 1) detector; 2) sample;
3) photomultiplier.
473
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Declassified and Approved For Release 2013/04/01: CIA-RDP10-02196R000700090005-4
40
10
Back-
ground
609
4?K 1461
769
787
806
214Bi
1764
214BL
2128i
1620
2615
t
2071
3198
2204
?
?
?
0 ZO 40 50 80
Channel No.
100
120
140
Fig. 2. Spectra from background and from the natural y emit-
ters 4?K, U--Ra, and Th (with decay products). 100 x 100 mm
detectors with 33 x 60 mm well: ---) NaI(T1); CsI(T1).
where rS, rp, and ra are, respectively, the intrinsic resolution of the scintillator, photo-
multiplier, and analyzer. Since rp and ra are constants, R is arbitrarily considered to be
the resolution of the detector. Furthermore, the correct resolution is considered to be
that which is determined under the actual measurement conditions, i.e., in a detector well
with a distributed source. It was found that Rca = 1.32 RK in such a system.
A comparison of photoefficiency and intrinsic background was made in energy intervals
(AE, MeV) which contained the total absorption peaks (TAP) of the y sources as well as the
TAP and sum peaks for the cascades 1.120 + 0.609 in 214Bi and 2.616 + 0.583 in 208T1,
which
were selected for application to the problem of individual determination of these radionu-
clides in mixtures. The photoefficiencyP of the detectors was determined from the TAP of
40K (1.461 MeV) assuming that 1 g of potassium emits 3.25 y quanta per second [2]. For de-
tectors 100 x 100 mm in size, the efficiency for 40K was 58.9 and 65.6% for NaI(T1) and
CsI(T1) for well measurements; the photofractions were 23.3 and 29.2%, respectively. Dif-
ferences in the spectral shape and in the efficiency of detectors of equal sensitive volume
but different shape are shown in Fig. 1, where y spectra from identical sources are presented
which were obtained with a scintillation unit having a 76 x 76 mm CsI(T1) crystal and with
an 80 x 80 mm CsI(T1) detector having a 35 x 50 mm well. Sources of 4?K (51.5 g KC1) and Th
474
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Counting rate, disintegrations/min
15
10
5
I
1
1 I
I
I
)1
.C-
; 3
46.K 11461
1 ?Th 1 1
0 20 40 60 30 100 120 140
Channel No.
Fig. 3. Spectra from background and from rock (40 g of granite) in
measurement geometry A: ---) NaI(T1); ...) CsI(T1), ----) CsI(T1);
1) detector; 2) sample; 4) photomultiplier; 3) light-pipe--window.
(standard 55 g SVT-16A sample) were measured for the same time in a counting geometry optimal
for the specified crystals (on the face and in the well, respectively). Measured results for
the parameters of the detectors compared are given in Table 1. Because of the difficulty of
producing distributed y sources of optimal size for each detector, the experimental data have
an error of '0_0%.
Figure 2 shows spectra from natural y emitters and Fig. 3 shows the background spectra
for one NaI(T1) and two CsI(T1) detectors 100 x 100 mm in size with 33 x 60 mm wells along
with spectra from rock (granite) obtained with both types of detectors. The intrinsic back-
ground Nb of a y spectrometer can be considered to be made up of two components recorded by
the detector: environmental radiation no and the intrinsic radiation ns of the scintillator,
i.e., Nb = no + ns. Since the first component (for specific shielding) depends on the re-
lation of the cross sections for y-ray interactions with the scintillator material and the
second, in addition, on the content of 137Cs, 40_,
87Rb, and other radionuclides in the de-
tector, the data presented in Table 1 for the background is in terms of relative values. The
content of the 135Cs fragment in CsI(T1) detectors [3] limits the possibility of their use at
energies below 0.8 MeV. In some detectors, however, the 137Cs content is beyond the limits
of detection. This situation gives rise to the need for special purification of raw material
for CsI(T1) crystals.
To compare the "quality" of the detectors, the criterion M = eb/RT, was used where e is
the counting efficiency in a given interval. Table 1 gives values of c and M for all detec-
tors relative to the 60 x 60 mm NaI(T1) detector with a 35 x 47 mm well. These data indicate
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that the use of CsI(T1) with a well makes it possible to increase the productivity of analyses
significantly or (for the same exposures) to increase the sensitivity and reduce the error.
This is facilitated by the great stability of spectrometric parameters in time which reduces
the number of measurements made for monitoring metrologic parameters and-for calibration of
gamma spectrometers. Notice also that the 60 x 60 mm CsI(T1) detector with a well 4,of,bet-
ter "quality" than the 76 x 76 mm CsI(T1) detector (without a well), the sensitive volume of
which is about 2.8 times greater.
The author is grateful to L. L. Nagorna and Ya. A. Zakharin for the preparation of high-
quality CsI(T1) detectors 100 x 100 mm in size.
LITERATURE CITED
1. Yu. A. Surkov and 0. P. Sobornov, At. Energ., 34, No. 2, 125 (1973).
2. J. Adams and P. Gasparini, Gamma-Ray Spectrometry of Rocks, Elsevier Publ. Co., Amster-
dam--London--New York (1970).
3. 0. P. Sobornov and 0. P. Shcheglov, At. Energ., 39, No. 1, 63 (1975).
476
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Declassified and Approved For Release 2013/04/01: CIA-RDP10-02196R000700090005-4
COMECON CHRONICLES
THIRD SYMPOSIUM OF COMECON MEMBER-NATIONS ON "WATER REGIMES,
WATER TREATMENT, AND LEAK-TESTING OF FUEL ELEMENTS IN
ATOMIC POWER PLANTS"
Yu. A. Egorov and A. V. Nikolaev
In accordance with the plan of operation of the COMECON Standing Committee on Atomic
Energy, a regular scheduled symposium was held in Neubrandenburg (German Democratic Republic)
to sum up work done on water regimes, water treatment, and leak-testing of the jackets of
fuel elements in atomic power plants with water-moderated?water-cooled reactors in COMECON
member-nations. Participating in the symposium were specialists from Bulgaria, Czechoslova-
kia, the GDR, Hungary, Poland, and the USSR as well as representatives of the COMECON Secre-
tariat and the IAEA. A total of 63 papers were read in five sections at the symposium.
The section on "Water Regimes, Corrosion, and Water Treatment in the Primary Circuit"
discussed papers on experience gained in operating water purification systems, ways of moni-
toring corrosion processes, and the behavior of active and inactive corrosion products in
coolants. A number of papers in this section were devoted to the study of water-purification
technology, especially the capabilities of electromagnetic filters, purification of coolants
by ion-exchange filters, the study of the sorption kinetics of some corrosion and fission
products, the effect of continuous purification on the concentration of active products in a
coolant, as well as the results of studies on the processes in a coolant with boron control
and the properties of ion-exchange materials, including their radiation stability.
The papers discussed in this section showed that in recent years the COMECON member-
nations have developed water-treatment systems and apparatus for atomic power plants with
various sources of water supply and that they have also developed and tested designs of elec-
tromagnetic filters whose application significantly improves purification of water from cor-
rosion products and, consequently, reduces the density of active and inactive deposits on
the surfaces in technological circuits. Much work has been done on developing technological
regimes for boron control with the aid of ion-exchange materials and on developing systems
for the regeneration of boric acid from drain water. Studies of the properties of ion-ex-
change resins made it possible to select requirements for nuclear-grade ion-exchange mate-
rials and to propose methods for testing them. Recommendations have been formulated for op-
timization of the water regime in atomic power plants with VVER-1000 (water-moderated?water-
cooled) reactors.
The section noted some major areas in which investigations should be pursued in coming
years. These include analysis and generalization of experience from the operation of water-
treatment installations, the study and evaluation of the corrosive state of equipment, de-
velopment and testing of high-temperature systems for coolant purification, including sys-
tems with electromagnetic filters, improved ion-exchange materials, and the technology of
regeneration of boric acid from effluents employing membrane processes, etc.
Papers presented to the section on "The Second-Circuit Water Regime" generalized the
experience from the use of equipment at the Novovoronezhskaya, Koslodui, and Bruno Loishner
atomic power plants with uncorrected water regime (e.g., the experience from the use of
OKh18N1OT steel evaporation surfaces in steam generators) discussed problems related to the
organization of the water regime in some atomic power plants. Interesting material was con-
sidered in papers devoted to the study of purification by condensation, electromagnetic and
ion-exchange filters, especially the use of electromagnetic filters in the high-temperature
part of the secondary circuit (after the deaerators), and design studies on the hydrodynamics
and lifetime of ion-exchange filters. The section took note of the need to standardize the
Translated from Atomnaya gnergiya, Vol. 42, No. 5, pp. 422-423, May, 1977.
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.
477
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secondary-circuit water regime, to study processes of corrosion of structural materials for
secondary circuits under various water regimes, technological schemes of the circuit, and
structural materials and systems for coolant purification, and to study the capabilities of
electromagnetic filters for purifying secondary-circuit water as well as drain water.
The papers heard by the section on "Radioactive Deposits in the Primary Circuit" con-
sidered the results of studies on the effect of transient processes on the behavior of some
radionuclides produced by corrosion in the Rheinsburg Atomic Power Plant and observations of
the transport of radionuclides in a repeated-circulation circuit in the course of the opera-
tion of the Leningrad Atomic Power Plant. The section also discussed the results of labora-
tory research on the sorption and solubility of corrosion products, and particularly on the
effect of the pH on the sorption of "Co by austenitic steel surfaces. An interesting paper
among the materials of this section was one which considered a mathematical model of the
sorption of active products on the surfaces of primary-circuit equipment, based on micro-
scopic constants. The section noted the necessity of extending the research on the forma-
tion and transport of active corrosion products in the primary circuit, and especially the
need to extend such studies in operating atomic power plants. Moreover, the section felt it
would be useful to study the behavior of corrosion radionuclides on reactor and in-pile loops.
The section on "Leak-Testing Fuel-Element Jackets" discussed a number of papers on vari-
ous aspects of leak-testing of fuel-element jackets in atomic power plants with water-moder-
ated?water-cooled reactors, the method of appraising the state of the active zone according
to the results of radiation monitoring of the primary-loop coolant, as well as various meth-
ods of detecting nonhermetic elements (by iodine radionuclides, by delayed neutrons, etc.), a
method of detecting defective fuel assemblies by overcompensation of the field of energy
liberation in the active zone. The section heard two reports on an experimental loop in the
Eva reactor and the possibilities of using it to simulate leaks and to detect fission prod-
ucts in the coolant.
At the present time a working approach has been developed to the monitoring of leaks of
fuel elements in atomic power plants with water-moderated?water-cooled reactors. This ap-
proach has three successive stages: continuous monitoring of the concentration of fission
fragments in the coolant, periodic monitoring of the nuclide composition of fission frag-
ments in samples of the coolant, and local monitoring (of each fuel-element assembly) in the
shut-down reactor. Demagnetization is detected in the first stage and its degree is esti-
mated in the second, whereas the third stage permits the defective assemblies to be rejected.
Just as the other sections did, this section also noted that it will be necessary to
focus on the study of the possibilities of determining the direct contact of the coolant with
the fuel by measuring the 239Np concentration in the coolant, on the study of the mechanisms
of leakage of radionuclides from fuel elements, particularly on fuel elements with artificial
defects, on the development of a method of localizing defective fuel-element assemblies by
overcompensation of the neutron field in the active zone, as well as on the unification of
the criteria for rejection of defective fuel-element assemblies, and the methods and instru-
ments of leak-testing of fuel elements.
The section on "Chemical-Analytic Control Monitoring" had papers on the development and
use of automatic systems for the chemical monitoring and control of the water regime of the
primary and secondary circuits of the atomic power plant. It was noted that such a system
is in successful use in the Beloyarsk Atomic Power Plant. A considerable number of papers
presented in this section were devoted to the development of techniques and instruments for
the chemical monitoring of various coolant properties: to determine the concentrations of
chlorine, boric acid, various corrosion-product and fission-fragment radionuclides, etc. At
the present time, borometers are being developed and tested at the Kozlodui, Novovoronezh-
skaya, and Rheinsburg atomic power plants. A number of COMECON member-nations are commer-
cially manufacturing instruments required for the water-chemical regime, e.g., for the auto-
matic measurement of the pH of the coolant.
Chemical and radiochemical monitoring constitute the basis of reliable and safe opera-
tion of atomic power plants and problems should be solved comprehensively. Therefore, in
coming years it is necessary to carry out work on the unification of ways and means of mon-
itoring the water-chemical regime in atomic power plants and on putting together unified
monitoring schemes, this being preceded by comparative tests of existing and newly developed
instruments and methods in existing atomic power plants.
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The discussion at the symposium on the problem under consideration was extremely useful
and showed that problems of the water-chemical regime of atomic power plants are being suc-
cessfully resolved. The symposium was held in a warm, businesslike setting.
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INFORMATION
"RESPIRATION" OF THE SUN
N. A. Vlasov
Academician A. B. Severnyi and his co-workers of the Crimean Astrophysical Observatory
have made an extremely interesting discovery: the sun's surface oscillates with a period of
160 ? 0.5 min [1]. A wide spot observed in the central region of the solar disk rises and
falls. The surface moves at speeds of up to 2 m/sec and the amplitude of the displacement
is about 10 km. It was possible to detect and measure such small oscillations by sensitive
methods developed for measuring very small shifts of spectral lines. A large portion of the
solar surface (perhaps even the entire surface) undergoes periodic radial oscillation, com-
pressing and expanding. The sun "breathes." Elastic waves propagate inside the sun and one
of them causes the respiration with a period of 160 min. For solar physics this discovery
gives approximately the same possibilities as seismometry does for terrestrial physics.
The elastic waves may "show" the interior of the sun. Up to the present time this was
expected only of the neutrino but tracking neutrinos has hitherto given an incomprehensible,
almost negative result [2].
The period of 160 min refers to the most pronounced oscillation of the surface, but os-
cillations with other periods are also possible, although with a smaller amplitude. The stan-
dard theoretical models of the sun predict radial oscillations with a period of less than
1 h. The observed period is in agreement with the theoretical period of quadrupole oscilla-
tions. Quadrupole oscillations of large amplitude would be quite surprising. But if the ob-
servations confirm the global-radial character of the oscillations, then the discrepancy be-
tween the period and the theoretical value is all the more surprising. The study of "sun-
quakes" today is one of the most interesting, new areas of solar physics and has caught the
attention of many astrophysicists [3].
LITERATURE CITED
1. A. Severny et al., Nature, 259, 87 (1976).
2. N. A. Vlasov, At. Energ., 41, No. 4, 251 (1976).
3. J. Christensen-Dalsgaard and D. Gongh, Nature, 259, 89 (1976).
480
Translated from Atomnaya tnergiya, Vol. 42, No. 5, p. 424, May, 1977.
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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Declassified and Approved For Release 2013/04/01: CIA-RDP10-02196R000700090005-4
CONFERENCES AND MEETINGS
FRANCO?SOVIET SEMINAR ON "CONCEPTION OF ATOMIC POWER PLANTS,
TECHNOLOGY, AND OPERATION OF WATER-MODERATED?WATER-COOLED
REACTORS"
V. A. Voznesenskii
The seminar was held at the Nuclear Research Center in Saclay from Dec. 8 to 10, 1976.
The 15 papers read at the seminar dealt with general conceptions and prospects for the de-
velopment of water-moderated?water-cooled power reactors, the experience from the operation
of such reactors in the USSR, the fabrication of equipment for atomic power plants, standards
and techniques for strength calculations, monitoring the state of the equipment prior to
start-up and during operation, and purification of primary-circuit coolant from corrosion
products.
The French program of construction of atomic power plants with water-moderated?water-
cooled reactors and the characteristics of these power plants were the subject of a paper
by M. Komez.
Notwithstanding the favorable experience from the operation of gas-cooled-graphite-mod-
erated reactors, it has been found economically inadvisable to build any more such reactors
for the production of electricity. The Soviet program provides for the construction by 1984
of 30 water-moderated?water-cooled reactor facilities with a power of up to 900 MW(E) and
four with a power of up to 1300 MW(E). As a result, by the beginning of the 1980s a large
proportion of electricity will be produced by atomic power plants, as shown by the data in
Table 1.
The primary circuit of the 900-MW atomic power plant has three circulating loops without
valves, as well as main circulating pumps, and vertical steam generators. To ensure the
safety of the population and to protect the environment, the equipment is installed in a
cylindrical protective shell (height 56 m, diameter 37 m), designed to withstand a pressure
of 5 bars, which could arise during maximum design failure related to the loss of coolant.
For emergency flooding of the active zone, there are three water tanks connected to the cold
branches of the loop, three high-pressure pumps for emergency introduction of boric acid, and
two low-pressure pumps, connected to the cold and hot branches of the loops. The active
safety system and reliable power supply system have a double back-up.
Atomic power plants are protected against flooding, earthquakes, and other natural dis-
asters and premises containing radioactive substances are protected from flying objects,
both external (falling aircraft, parts of turbines) and internal, arising during leaks from
the primary-circuit tubing. Special measures are undertaken against radioactive and chemical
substances getting into the groundwater.
A paper by L. Gendry and B. Moreau reported on construction plans, technology, and char-
acteristics of primary-circuit equipment for atomic power plants with water-moderated?water-
cooled reactors. The first reactor vessel for the atomic power plant in Chooze was built in
1964. Nineteen reactor vessels had been fabricated for France, the USA, Switzerland, and
Belgium by the end of 1976. The Framatom works (10,000 employees) can turn out eight reac-
tor vessels and volume compensators as well as 18-24 steam generators per year. Water-mod-
erated?water-cooled reactors can be constructed with a power of up to 2000 MW. The main
stages in reactor-vessel fabrication are: fabrication of the individual parts of the vessel
and cover (flanges, cowlings, bottoms, branch pipes), welding of the stainless facing onto
the various parts, and subsequently welding the parts together. The bottom is pressed and
Translated from Atomnaya Energiya, Vol. 42, No. 5, pp. 424-426, May, 1977.
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.
481
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TABLE 1. Generation of Electricity in
France, %
Year
Thermal
Hydro
Atomic
1975 '
56,5
33,5
10
1985
12,0
18,0
70
2000
10,0
9,0
81
the other parts are forged. The vessel material is carbon steel, alloyed with molybdenum,
manganese, and nickel. The facing is applied in two layers, each 4 mm thick; the first is
Kh24N12 steel and the second is Kh2ON10 steel. Much attention is paid to quality control.
The problems of making provision for seismic activity in designs for atomic power plants
were discussed in a paper by M. Livoland. Atomic power plants are constructed in the country
in regions with a seismicity of up to seven on the Richter scale. The increased force of
the underground shock is taken into account when designing the systems and premises contain-
ing radioactive substances. The spectrum of earth tremors is placed in correspondence with
the intensity of the earthquake. Various methods of calculation were considered for finding
the displacements and stresses in the structures of atomic power plants with a known soil
spectrum. In cases when the seismic-resistance of a structure is difficult to determine,
studies are carried out on a vibrations stand. Of some interest is a way of reducing the
seismic effects on the entire complex of premises of the reactor section by setting their
common foundation on special sliding plates of bronze or lead, resting on a layer of syn-
thetic rubber on top of the foundation. In this case, equipment designed for ordinary atomic
power plants can be used in regions of high seismicity.
Special regulations worked out in France cover the design, fabrication, and use of pres-
sure vessels for atomic power plants (paper by R. Roche). The main emphasis here was on the
definition of the equipment to which the regulations apply; the responsibility of the fabri-
cator and the operating organization; the choice of miterials; and the quality program as
well as documents to be presented to inspection agencies. The strength is evaluated with
reference to excessive strain, plastic and elastoplastic instabilities, progressive deforma-
tion, and low-cycle fatigue.
Normal and emergency operating conditions of atomic power plants are divided into four
categories with a different safety factor for each and the strength based on different limit-
ing states. The strength is analyzed by various methods; in particular, the method of finite
elements is used to design one of the most complex parts of the primary circuit as far as
configuration is concerned, i.e., the spiral of the main circulating pump (paper by J. Bey-
lac).
Various aspects of the quality control of the fabrication and the state of the primary-
circuit equipment during assembly, start-up, and operation of the atomic power plant were
considered in papers by R. Saglio and P. Bernard. Thus, to detect defects and to keep track
of their development the French Atomic Energy Commission developed new nondestructive meth-
ods: inspection of the reactor-vessel metal by focused ultrasound during periodic checks,
recording of acoustic emissions during hydraulic tests, and inspection of the connecting
pipes of the steam generator by means of Foucault or eddy currents. A special machine has
been built for setting up on the flange of the vessel for ultrasound and TV inspection of
the vessel from the inside. In 1976 the machine was used successfully on the reactors of the
Fressenheim and Chooze atomic power plants. During inspection of the steam-generator tubing
with Foucault currents, a pickup travels over the entire length of the tubing. The charac-
teristics of the signals corresponding to various types of damage are determined with the aid
of standards with predetermined defects.
As a result of theoretical investigations, it is considered possible to construct ef-
ficient systems for monitoring the state of equipment on the basis of analysis of the noise
in various parameters. For example, it is proposed to determine the displacements of devices
inside the vessel by analyzing the signals from ionization chambers.
High-temperature filters have been developed for purifying the primary coolant from ra-
dioactive corrosion products (paper by L. Dolle). Two prototypes of such filters (electro-
magnetic and graphite) have been tested in use in loops. In the long term it is planned to
construct filters for a coolant flow rate of up to 100 tons/h.
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As a result of visits to the nuclear research centers at Grenoble and Cadarache, the
Soviet specialists became acquainted with equipment and programs for research on the safety
of water-moderated-water-cooled reactors.
Experimental thermotechnical test beds in Grenoble make it possible to study both steady-
state heat exchange and the hydrodynamics in bundles of electrically heated rods as well as
the processes associated with the formation of large leaks. The program of work with the
Omega loop (pressure 170 bar, power up to 9 MW) envisages studies on the heat-exchange crisis
under steady-state conditions for bundles of 25 rods with a diameter of 9.5 mm, with a pitch
of 12.6 mm in a square array, and lengths of 2.12, 3.65, and 4.2 m (with uniform and cosinu-
soidal power distribution), as well as experiments on coolant loss. In the latter case, the
tubing used in the first stage had an internal diameter of 12 mm and a heated segment 3.65 m
long. An assembly of 36 rods of 3.65 m length is planned. In other loops (water and Freon),
in addition to the heat-exchange crisis, a study is made of the distribution of the flow rate
and the mixing between particular cells of bundles and there are plans to study the heat trans-
fer in the post-crisis region. Special facilities have been provided for studying the heat
transfer when the active zone is refilled following a big leak, as well as for studying crit-
ical flow rates, vapor condensation under the protective shell, and reactive forces during
leaks.
In Cadarache, the Soviet specialists saw the Aquitaine-2 test beds and the Phebus loop,
which were in the final stages of assembly. The test bed (pressure 170 bar, temperature
340?C) consists of two tanks connected by tubing of various diameters, lengths, configuration,
support designs, etc. It is intended for studying the stresses and strains that arise in
tubing elements during various types of rupture and the effects of the action of jets on con-
crete and other parts of the installation. The test bed is to be reconstructed later for
studying the stresses in devices inside the vessel.
At the center of the active zone of a 50-MW(E) reactor is a high-pressure loop for study-
ing the behavior of fuel elements under the conditions of large leaks. The leaks are organ-
ized in a special tank. Among the parameters whose role is to be elucidated are: the size
and location of the leak; the linear power and the internal pressure of fuel elements; and
the pressure, temperature, and the inlet of the injected coolant. Two types of assemblies
will be used: one consisting of a single fuel element and one consisting of 25 fuel elements
with an active-length diameter of 800 mm and pitch corresponding to the fuel assemblies used
in atomic power plants. Present plans call for about ten experiments with one fuel element
and 40 experiments with 25 fuel elements. To check the applicability of the results obtained
with fresh fuel elements, some tests will be carried out with irradiated fuel elements. Pro-
vision has been made for on-site inspection of assemblies by optical means and by n and y
scanning, after which they will be sent to the hot laboratory.
The Franco--Soviet seminar permitted a quite broad exchange of information on many as-
pects of the construction and operation of water-moderated reactors and promoted the further
strengthening of the scientific--technical ties between the USSR and France.
483
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SOVIET--FRENCH SYMPOSIUM ON THE PRODUCTION AND APPLICATION
OF STEEL PIPING IN INDUSTRY
G. V. Kiselev
The symposium was held in Moscow from Jan. 11 to 13, 1977. Papers on behalf of the
French side were presented by the Vallurec company which specializes in the manufacture of a
wide range of piping for various purposes, including for thermal and atomic power plants.
The company has divisions which produce piping designed for work with corroding agents at
low and high temperatures, under irradiation, etc. The materials used for the piping are
stainless and pearlitic steels and nickel, titanium, and zirconium alloys.
It should be noted that it is possible to make long seamless tubes (up to 36 m) for the
heat-exchange equipment of atomic power plants. For this purpose, use is made of two presses,
of 1500 and 3000 tons, with vitreous lubricant. Cold-drawn tubes of small and medium sizes
are produced of Soviet-made two-high and three-high mills. Weld seams on the tubes are made
by plasma-arc welding.
The Vallurec company produces a wide range of tubing for "nuclear" applications, such
as for fuel-element cans, heat exchangers, and steam generator of atomic power plants (see
Table 1).
The characteristic features of the production are of interest. Thus, a shop in the town
of Montbard making tubing for steam generators of atomic power plants measures 340 m long by
25 m wide, has been fitted with equipment for supplying cleaned and conditioned air so that
the temperature and humidity can be controlled, and has specially insulated walls and a spe-
cial floor, thus ensuring exceptional cleanliness during the process of fabrication. The
shop has an annual capacity of '1,1 million m of tubing, which can be increased to 1.5 million.
It makes straight and bent tubes with diameter of 12-25 mm, wall thickness 0.8-3 mm, and a
bend radius of 2 tube diameters to 1.5 m. The tolerance of the wall thickness of the cold-
drawn tubes is ?8%.
The tube move without turning on a special-purpose automatic line and then pass three
measuring heads in succession. These devices permit the tubes to be inspected with the aid
TABLE 1. Chemical Composition of Steels
Grade
Cr
Ni
Mo
Ti
Nb
AFNOR
AISI
DIN
Z6CN18.10
304
1.4301