SOVIET ATOMIC ENERGY VOL. 47, NO. 1

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• Document Page Count: 96
• Document Creation Date: December 23, 2016
• Document Release Date: February 15, 2013
• Sequence Number: 1
• Publication Date: July 1, 1979
• Content Type: REPORT

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Russian Original Vol. 47, No. 1, July, 1979
ic
January, 1980
SATEAZ 47(1) 501-590 (1979)
Ls) Filz._
SOVIET
ATOMIC
ENERGY
ATOMHAFI 3HEP111/1
(ATOMNAYA iNERG1YA)
TRANSLATED FROM RUSSIAN
CONSULTANTS BUREAU, NEW YORK
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SOVIET oviet Atomic energy is a cover-tbzcoyer translation of Atomnaya
Energiya, a publication of the Academy of Sciences of the USSR.
ATOMIC
ENERGY
An' agreement with the Copyright Agency of the USSR (VAAP)
makes available both, advance copies of the Russian journal and
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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 begin with the ,first issue of the
Russian joyrnal.
Editorial Board of Atompaya Energiya:
Editor: 0. D. Kazachkovskii
Associate Editors: N. A. Vlasov and N. N. Ponomarev-Stepnoi
Secretary: A. I. Artemov
?
I. N. Golovin ' V. V., Matveev
V. I. ll'ichey I. D. Morokhov
V. E. Ivanov A: A. Naumov
V. F. Kalinin A. S. Nikiforov
P. L. Kirilloy ' A. S. Shtan'
Yu. 1. Koryakin B. A. Sidorenko
A. K. Krasin M. F. Troyanov
E. V. Kulov? E. I. Vorob'ev
B. N. Laskorin
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SOVIET ATOMIC ENERGY
A translation of Atomnaya Energiya
January, 1980
Volume 47, Number 1 July, 1979
CONTENTS
ARTICLES
Combination of Nuclear-Geophysical Methods and Apparatus for Increasing the Efficiency
of Prospecting, Extraction, and Reprocessing of Nonradiactive Mineral Raw
Materials (by the Example of Tin Ores) - S. A. Baldin, S. N. Voloshchuk,
B. G. Egiazarov, L. V. Zernov, I. A. Luchin, V. V. Matveev, L. Ch. Pukhal'skii,
Engl./Russ.
and N. I. Chesnokov
501
3
Algorithm for the Extremal Control of the Energy Distribution in a Power Reactor
-1. Ya. Emeltyanov, V. V. Postnikov, and G. V. Yurkin
506
8
Some Peculiarities of the Structure of the Flow in the Case of Critical Discharge
Conditions of Boiling Water through Cylindrical Channels - V. S. Aleshin
512
12
Dependence of the Ejection Coefficient of Uranium on the Flux of Thermal Neutrons
-G. P. Ivanov, V. A. Bessonov, N. A. Grinevich, V. A. Popovichev, and
E. A. Borisov
516
15
Effect of Irradiation on The Ultimate Fracture Strength of the Alloy Zr-2.5% Nb
- 0.A.Shat-skaya, E. Yu. Rivkin, A. M. Vasnin, V. V. Klyushin, A. V. Kozlov,
V. M. Nalesnik, and M. E. Rodin
519
18
Role of Impurities in Irradiation Embrittlement of Low-Alloy Steel - V. A. Nikolaev,
V. V. Rybin, and V. I. Badanin
523
21
Yields of Some Fragments from Fission of 235U, 238u, and 239Pu by Neutrons from
Spectrum of BR-1 Fast Reactor - L. N. Yurova, A. V. Bushuev, V. N. Ozerkov,
V. V. Chachin, A. V. Zvonarev, Yu. G. Liforov, Yu. V. Koleganov, V. V. Miller,
and 0. V. Gorbatyuk
528
26
Neutron Yield of (a, n) Reaction onOxygen -V. I. Bulanenko
531
28
Instrumental Neutron-Activation Analysis of Submilligram Amounts of Geochemical
Samples - V. I. Drynkin, E. V. Karns, V. D. Nartikoev, B. V. Belen'kii, and
A. L. Kerzin
534
31
Calculation of Photoneutron Yields from Thick Targets in Giant-Resonance Region
-V. I. Isaev and V. P. Kovalev
538
34
LETTERS TO THE EDITOR
An Accelerating Section for Reducing the Radiation Level in the Absorber Section of a Linac
- V. S. Balagura, V. M. Grizhko, I. A. Grishaev, L. K. Myakushko,
B. G. Safronov, and G. L. Fursov
543
39
y-Ray Recording Efficiency of a Spherical Detector - D. I. Konstantinov
546
41
Examination of Irradiated Metal Diborides by X-Ray Diffraction - Kh. Maile,
I. A. Naskidashvili, and T. Sh. Berdzenishvili
548
42
Neutron Spectra in the MeV Range in Fast Critical Assemblies - V. M. Lityaev,
V. A. Dulin, and Yu. A. Kazanskii
550?
44
Effects of Various Factors on the Absorbed-Dose Distribution in Thin Layers
-V. V. Krayushkin
552
46
A Calorimeter for Measuring Local Electron-Beam Absorbed Doses - V. A. Berlyand,
V. V. Generalova, and M. N. Gurskii
554
47
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Determination of Nuclear Constants as an Inverse Problem in Radiation Transport
-V. N. Dushin
CONTENTS
(continued)
Engl./Russ.
556 48
Effects of Bombardment by He+, Ni+, and Cr+ on Microhardness and Corrosion Cracking
of Stainless Steels - B. G. Vladimirov, V. M. Gusev, and V. S. Tsyplenkov
558
50
Nonstationary Prompt-Neutron Diffusion in a Fast Assembly - V. E. Kolesov,
0. I. Makarov, and I. P. Matveenko
560
51
Effects of Ion-Bombardment Dose and Previous Surface Treatment on the Erosion of
Molybdenum - B. A. Kadin, D. M. Skorov, and V. L. Yakushin
562
53
A Cadmium-Sulfide y-Ray Dosimeter with Elevated Stability under Irradiation
-V. K. Dubovoi
564
54
The Thermal-Neutron Fission Cross Section and the Fission-Resonance Integral for 243Cm
- K. D. Zhuravlev and N. I. Kroshkin
565
55
Absolute Measurement of the Branching Ratio for the 277.6-keV Line of 239Np
-V. K. Mozhaev, V. A. Dulin, and Yu. A. Kazanskii
566
55
Calculation of the True Volume Proportion of Steam in the Driving Section of a Natural-
Circulation Loop - L. N. Polyanin and A. L. Putov
567
56
A Local Approach to Determination of the Coordinates of an Interface - F. L. Gerchikov
and V. D. Kosarev
569
57
ANNIVERSARIES
Academician Pavel Alekseevich Cherenkov (On His 75th Birthday Anniversary)
- E. I. Tamm and B. B. Govorkov
572
59
The 50th Birthday Anniversary of Evgenii Vladimirovich Kulov
574
60
INFORMATION
The Accident at the Three Mile Island Nuclear Power Plant
575
61
CONFERENCES, MEETINGS, AND SEMINARS
Second All-Union Conference on the Chemistry of Uranium - g. A. Semenova
579
63
Symposium on the Scientific Foundations of Radioactive Waste Handling - V. I. Spitsyn
and A. S. Polyakov
580
64
Conference of Experts on the Effect of Nuclear Power on the Environment
- L. A. Win and V. I. Karpov
582
65
Scientific-Technical Conference "Energy and Environmental Protection"
-B. M. Stolyarov
583
66
International Seminar on the Practical Significance of the ICRP Recommendations
-Yu. V. Sivintsev
585
67
Conference on the Disruptive Instability in Closed Systems - V. G. Merezhkin
586 '
68
NEW BOOKS
V. L. Blinkin and V. M. Novikov. Liquid-Salt Nuclear Reactors - Reviewed by
Yu. I. Koryakin
588
69
V. V. Fisenko. Critical Two-Phase Flows -Reviewed by V. N. Sxnolin
588
69
Tritium Measurement Techniques - B. P. Maksimenko
590
70
The Russian press date (podpisano k pechati) of this issue was 6/27/1979.
Publication therefore did not occur prior to this date, but must be assumed
to have taken place reasonably soon thereafter.
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ARTICLES
COMBINATION OF NUCLEAR-GEOPHYSICAL METHODS
AND APPARATUS FOR INCREASING THE EFFICIENCY
OF PROSPECTING, EXTRACTION, AND REPROCESSING
OF NONRADIOACTIVE MINERAL RAW MATERIALS
(BY THE EXAMPLE OF TIN ORES)
S. A. Baldin, S. N. Voloshchuk,
V. G. Egiazarov, L. V. Zernov,
I. A. Luchin, V. V. Matveev,
L. Ch. Pukhal'skii, and N. I. Chesnokov
UDC 550.835.08
A promising trend for the utilization of ionizing radiations in the national economy is the development
of nuclear-physical methods and apparatus for solving the urgent problems of increasing the efficiency of
prospecting, extraction, and reprocessing of mineral raw materials. The greatest successes in this field
have been achieved with the high-speed analysis of uranium ores by a radiometric method, based on record-
ing the natural y radiation with reference to a simple apparatus [1]. When solving similar problems for non-
radioactive raw materials, use is made of the stimulated radiation induced in the substance by means of iso-
tope sources, particle accelerators of various types, nuclear reactors, etc.
One of the most effective methods of analysis of a substance in applied problems is the roentgenoradio-
metric method (RRM), consisting in the excitation in the substance of x-ray characteristic emission with
radioisotope sources and its subsequent recording by means of spectrometric equipment.
The energy spectrum of the induced radiation contains, together with the x-ray characteristic emission
of the required element, the characteristic emission of other elements occurring in the composition of the
ore, and also primary scattered radiation. Because of this, the necessary sensitivity and high measurement
speed in the majority of cases can be attained only by using spectrometric instruments which are relatively
complex in structure and composition, with devices for the automatic processing of the measurement results.
This makes the use of nuclear-physics methods extremely difficult for the complex analysis of nonradioactive
ores under production conditions, although several important problems at the present time have been success-
fully solved by these methods.
Nuclear-geophysical methods of logging, sampling uncrushed core samples and ores in the natural de-
posit, analysis of powdered samples, etc. have been developed and introduced, and in a number of cases
methods for monitoring sorted ore, concentration products, pulps and solutions are being used. At the same
time, automated methods and equipment for high-speed analysis and preliminary sorting in transportable
tanks, separation into finely divided batches and lumps have almost been uninvestigated and unused, although
on individual problems there are publications indicating the formulation of the problem, the results of labora-
tory investigations and certain positive experiments [2, 4].
In this present paper, the results of the development and introduction of a combination of nuclear-physi-
cal methods and equipment are recounted, which provide an effective solution for the production problems at
all the principal stages of monitoring the mining-metallurgical production of tin. Taking into consideration
the nuclear-physical characteristics of tin and the requirements for production, roentgenoradiometric [3] and
y-resonance methods of analysis [5] have been accepted mainly for the analytical monitoring of the complex.
The latter is used in those cases when it is necessary to monitor the tin in the form of cassiterite or to carry
out phase analysis on compounds of tin and iron.
Depending on the production problems to be solved, semiconductor detectors (SCD), scintillation and
proportional counters are used, and also isotope sources of 241AM, 147pm, and 119MSn. The constitution of
the equipment complex developed for monitoring in the different stages of the mining-metallurgical production
process is shown in Fig. 1.
Translated from Atomnaya Energiya, Vol. 47, No. 1, pp. 3-7, July, 1979. Original article submitted
January 30, 1979.
0038-531X/79/4701-0501$07.50 1980 Plenum Publishing Corporation 501
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Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020001-4
TABLE 1. Averaged Indices of Lump Separation
Techno-
logical
sample
Total
mass.
tons
Coarseness
grade, mm
Average
yield of
grade,%
Av. tin
content
(chem.
Tailings
Concentrate
Conc.
factor,
rel.
yield
from
yield
from orig.
tin
content,
yield
from
yield
fromorig
tin
content,
anal. ),0/0
grade,%
ore,010
010
grade. 010
ore,u/c,
010
units
1
69,6
63-100,
13,6
0,138
51,0
6,8
0,071
50,0
6,8
0,185
1,45
100-203
17,4
0,176.
44,0
. 7,7
0,060
56,0
9,7
0,267
1,52
2
48,0
63-100,
14,4.
0,096
85,0
12,2
0,055
15,0
2,2
0,328
3,4
100-200
18,0
0,130'
79,0
14,2
0,072
21,0
3,8
0,348
2,7
3
7,74
20-32,
8,5
0,174
63,0
5,4
0,100
37,0
3,1
0,3)1
1,72
32-63
15,6
0,220
81,0
12,6,
0,070
19,0
. 3,0
.0,860
3,9
Fig. 1. Structure of the complex and monitoring functions: A) monitoring for prospecting,
extraction, and sorting processes; B) control for treatment processes; 1) sampling; 2, 3)
logging of shot holes and bores; 4) high-speed analysis at the ore-monitoring station; 5)
sorting; 6) surface logging; 7) laboratory analysis of samples, core, and ores; 8, 9) batch.
and lump separation; 10) high-speed analysis of products; 11) tailings; 12) laboratory anal- '
ysis of samples; 13, 14) continuous monitoring for metallurgy (solution monitoring) and
concentration (pulp monitoring) processes, respectively.
In search and prospecting, the results of the roentgenoradiometric ore sampling at the site of the seam
and of the uncrushed core sample, logging of the drillholes, roentgenoradiometric and y-resonance analysis
of the powdered samples (see Fig. 1: 1, 6, and 7) are used for the selection of promising sections and for ob-
taining the initial data for calculating the reserves.
At the stage of extraction and preliminary concentration of ores, by means of roentgenoradiometric ore
sampling at the site of the seam and the uncrushed core sample, logging of shot holes and drillholes, high-
speed analysis and sorting of the ores into transportable containers, preliminary concentration of the ores into
finely divided batches and lumps, and analysis of the powdered samples by roentgenoradiometric and 'y-reso-
nance methods (see Fig. 1: 1-3, 6-9) are obtained starting data for calculating the reserves for detailed and
operational prospecting, and optimization of the extraction process; prerequisites are established for the se-
lective extraction and abandoning of substandard ores in the form of untouched blocks and rubbish, in order
to reduce losses and depletion, and the extraction of the part of the shattered rock massif which is substan-
dard in the content of tin.
In order to monitor and control the ore treatment process, the results of the roentgenoradiometric sam-
pling of the ore on a conveyor belt are used, as are monitoring of the tin content in the solid phase of pulp and solu-
tions in a continuous cycle, roentgenoradiometric and 'y-resonance analysis of samples for tin and its accom-
panying components (see Fig. 1: 10, 12-14). The data obtained from the measurements allows the content of
tin and accompanying elements to be determined in the ores and in the products of their treatment.
502
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Fig. 2 Fig. 3
Fig. 2. Irradiation and detection unit, with a semiconductor detector, at the ore-
monitoring station.
Fig. 3. Irradiation and detection unit of a facility for the separation of ores with a
semiconductor detector.
In the mining instruments for logging, sampling of the uncrushed core sample and ore at the site of the
seam, scintillation detection units are used, with a thin sodium iodide crystal. The sensitivity threshold in
the case of roentgenoradiometric sampling amounted to 0.05%, and in the case of logging - 0.07%. By compar-
ing the results of the roentgenoradiometric sampling with the conventional furrow sampling, the mean-square
error in determining the content of tin amounted to from *32 (for agradecontaininglesethan0.1%) to *13% (for
a grade containing more than 0.5%). ? The relative mean-square error in the data from a comparison of logging
with laboratory analyses of core samples is equal to ?30% for a grade containing less than 0.1%. It should be
noted that the error of conventional geological sampling is *40-60%. The sampling output amounts to not less
than 20 m per shift, and the logging output is not less than 100 m per shift on the instrument. At the present
time, using RRM, more than 30 km of the walls of mine workings can be sampled, and more than 10 km of
drillholes bored from underground mine Workings is logged.
In the facilities for high-speed analysis, sorting and separation of ores into transportable containers,
finely divided batches and lumps, detection units are used with silicon semiconductor detectors (area of sensi-
tive surface -,250 mm2). The time for the roentgenoradiometric analysis of a single truck is 15-20 sec, and
the output is more than 300 trucks per shift. More than 300,000 trucks have been analyzed in a production
cycle on this facility. The cost of the high-speed analysis of the ore mase in the truck by the roentgenoradio-
metric method is szi3 times less than the cost of handful sampling with a sharp increase of the measurement
output, providing automation of sorting.
An irradiation unit and a detection facility with a cooled semiconductor detector and 4 isotope sources
of 241Am in shielded equipments with collimators is shown in Fig. 2.
Investigations conducted on the facility for separation into finely divided batches and lumps under produc-
tion conditions have shown the feasibility in principle and technically of roentgenoradiometric concentration.
It has been established that the content of tin in the lumps or batches is determined with an error of 30-40%,
which is sufficient for the efficient separation of ores according to content, and the analytical sensitivity allows
sorting to be carried out under batch and lump conditions with an output of 5-15 ton/b or more, depending on
the coarseness of sorting of the material (Table 1).
The measurement section of the facility for ore separation is shown in Fig. 3, in which can be seen the
lower part of the detection unit with the semiconductor detector and 4 isotope sources, positioned in pairs on
503
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Cse,%
4
100
8
6
4
2
2 4 6 8 101 4 6 8 100
RRM, Sn %
Fig. 4. Comparison of results of high-speed analysis of
tin ores in trucks (roentgenoradiometric method) and
bulk sampling (Csn, chemical analysis); ?) skarn ores;
0) shales.
2
? .
both sides of the detector. Both facilities for the preliminaryroentgenoradiometric concentration operate-_.
in the automatic regime and are equipped with processor-analyzers for processing the databy a specified al-
gorithin and also monitoring and control devices with external slave mechanisms. It should'be noted that
facilities for the contactless automatic sorting and separation of ores; based on RRM, have been constructed
and tried out under industrial conditions for the first time. -
. Roentgenoradiometric and x-ray absorption instruments and facilities for the Continuous monitoring of '
concentration products (pulps and solutions) are based on the application Of saint-illation and semiconductor
detector units. The sensitivity threshold for the analysis of pulps and solutions with a Semiconductor detector
is 0.001-0.003 wt.% and 5-10 mg/liter, respectively, and foraoluticins with scintillatiOn detectors it is 70 mg/
liter. The instruments and facilities provide data extraction into external equipments (digital printout, pen
recorder) and to the production process monitoring and control system..
The laboratory instruments developed for the analysis of powdered bre samples and their treated prod-
ucts include 'instruments for determining the content of the total tin and tin oxide (in the form of tassiterite)
by roentgenoradiometric and y-resonance methods, instrtunents based on semiconductor spectrometers for
determining the content of tin and accompanying elements by the roentgenoradiometric method, and instru-
ments for the phase analysis of tin and iron in ores and processed products by the y-resonance method. The
accuracy of the analyses of powdered samples is not inferior to, and in some cases even exceeds, the accuracy
of chemical analysis. The sensitivity threshold is 0.005-0.02% of the total tin and tin in the form of cassiterite.
The mean-squire relative deviations of the results of analysis by the y=resonance method of the tin con-
tent in the form of cassiterite from the results of chemical analysis are given below:
504
Grade of tin No. of deter-
?Deviation ,%
content,% minations
0.05
454
? ?58.4
0.05-0.1
405
? 27.3
0.1-0.25
447
? 16.2
0.2571
229
?11.2
1 ?
88
?5.2
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TABLE 2. Distribution of a Mined-Ore Mass
According to Range of Content of Tin, Based
on High-Speed Analysis of Ores in Trucks by
the Roentgenoradiometric Method, %
Range of
tin content
Average I
content
Yield of I
ore
Yield of I
tin
Range of
tin content
Average I
content
lYield of I
ore
Yield of I
tin
0-0,10
0,04
33,5
7,6
0,30--0,40
0,34
8,8
17,0
0,10-0,15
0,12
18,3
12,9
0,40-0,60
0,51
6,8
19,7
0,15--0,30
0,21
37,4
39,3
0,60
0,89
0,9
4,5
The results of comparisons for other types of analyses of powdered samples on tin and incidental com-
ponents also have shown the high accuracy and sensitivity of roentgenoradiometric analyses. In all, ??=1 300,000
samples have been analyzed. The output of the analyses is 100-120 analyses per instrument per shift.
It is well known that the roentgenoradiometric method has a low penetration depth, not exceeding a few
millimeters, because of which the most important problem of the investigations was to estimate the represen-
tability and accuracy of the results of the analyses in the case of logging, sampling at the site of the deposit,
and especially in the high-speed analyses in the transportable containers. Comparison of the results of the
determination of the tin in a surface seam of ore loaded into a truck by the roentgenoradiometric method and
by total sampling with chemical analysis of selected samples (> 500 trucks with different tin contents) showed
that there is a direct correlation dependence between them. For a tin content of < 0.1 and > 0.1%, the corre-
lation coefficient amounted to 0.71 and 0.82 with a relative mean-square error for both content grades of
k's ?30% (Fig. 4). Similar data were obtained when comparing the results on representative bulk investigations
of roentgenoradiometric sampling at the site of the deposit and logging with furrow sampling and sampling of
drillhole core samples by conventional methods. Thus, according to the content of tin determined by the
roentgenoradiometric method in a surface seam, the tin content in the volume investigated can be judged quite
accurately, which confirms the calculations by the Materon?de Wies formula, and which showed the feasibility
of replacing bulk elements with linear or area equivalents of these samples with preservation of the statis-
tical representation of sampling [6].
The application of the combination of nuclear-geophysical methods and equipment provides an important
economic effect, due to the replacement of conventional methods of analysis and sampling by nuclear-geo-
physical methods and the change of technology of prospecting, extraction and processing of tin ores. Thus,
the cost of the nuclear-geophysical analysis of powdered samples is a factor of 6-8 less than that of chemical
analysis, and the cost of nuclear-geophysical sampling at the site of the deposit is a factor of 5-6 less than
furrow sampling, etc. Already at the present time the effect of using the instruments of the complex in pro-
duction has amounted to about 1 million rubles.
The use of logging of drillholes and shot holes and sampling at the site of the deposit allows the greater
part of drilling with sampling of the core to be replaced by more productive and cheaper coreless drilling,
the volume of the shattered rock mass to be reduced because of the part of the substandard ores remaining
in the depths in the form of untouched blocks, and prerequisites to be created for selective extraction and
reduction of depletion and losses. The introduction into the technology of extraction, of high-speed analysis
and sorting of ores in transport containers, preliminary concentration into finely divided batches and lumps
will allow 20-40% of the waste tailings with a tin content of 0.06-0.08% to be separated, which will make
it possible to reduce the volume and to increase the content of tin in the ore directed to further processing
by a factor of 1.2-1.5. For example, during the development of one of the experimental units, using nuclear-
geophysical methods, 5.6% of the ore with a tin content of ?0.08% remained in the depths in the form of un-
touched blocks; according to the data of high-speed analysis of the ores in trucks with a limiting content of
0.1%, 33% of the tailings was separated with a tin content of 0.04%; the concentration factor in this case
amounted to 1.37 (Table 2); and with a limiting content of 0.15%, 51.8% of the tailings was separated with a tin
content of 0.07% and a concentration factor of 1.61.
The preliminary concentration of ores into finely divided batches and lumps allows the tin content in ore
directed to further processing to be increased additionally, and its volume to be reduced. As investigations
have shown, in the case of lump sorting of ores with a size of 32-200 mm by the roentgenoradiometric method,
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15-25% of the tailings (from the original ore sent for concentration) is separated with a tin content of 0.05-
0.08%. Technicoeconomic calculations and the results of industrial operation of a complex show that the to-
tal effect of its introduction amounts to 2-3 rubles per ton of ore.
Conclusions. A combination of nuclear-geophysical methods and equipment has been developed, tested,
and introduced into commercial operation, for searching, prospecting, extracting, and processing lean tin
ores, encompassing the whole production cycle. The introduction of the complex significantly increases the
technicoeconomical production indices as a whole, allows the use of a more productive extraction technology,
ore processing, and the creation of automated production control systems. The use, inthe complex, of equip-
ment with semiconductor detectors in conjunction with analyzer-processors has allowed for the first time in
the practice of the mining and concentration industry the creation ,of automatic roentgenoradiometric facilities
for contactless high-speed analysis, sorting and preliminary concentration of lean tin ores into transport con-
tainers, finely divided batches and lumps. The design of the complex, the flow-sheet?technological principles
of equipment construction and the methods of measurement provide the capability for using it to solve similar
problems for other types of nonradioactive raw materials, such as molybdenum, tungsten, zinc, copper, and
other ores, including complex ores.
LITERATURE CITED
1. L. Ch. Pukhaliskii and M. V. Shumilin, Prospecting and Sampling of Uranium Deposits [in Russian],
Nedra, Moscow (1977).
2. A. P. Ochkur et al., Gamma Methods in Ore Geology [in Russian], Nedra, Leningrad (1976).
3. E. P. Leman et al., Nuclear Geophysics in Ore Geology [in Russian], NPO Geofizika, Leningrad (1977).
4. A. P. Tatarnikov, Nuclear-Physics Methods of Concentration of Minerals [in Russian], Atomizdat,
Moscow (1974).
5. V. I. Gol'danskii et al., Gamma-Resonance Methods and Instruments for Phase Analysis of Mineral
Raw Materials [in Russian], Atomizdat, Moscow (1974).
6. E. Karl'e, Procedure for the Quantitative Assessment of Uranium Deposits [in Russian], Atomizdat,
Moscow (1961).
ALGORITHM FOR THE EXTREMAL CONTROL OF THE
ENERGY DISTRIBUTION IN A POWER REACTOR
I. Ya. Emeliyanov, V. V. Postnikov, UDC 621.039.562
and G. V. Yurkin
It is necessary in the development and operation of large-capacity power reactors to equip them with
automatic systems for the control of the energy distribution (ED) in the active zone which are based on digital
computers.
For the extremal control of the ED, i.e., for the control of the ED with a search for the optimal values
of a goal function, programs were used which are based on implementation of different "classical" optimiza-
tion schemes [1-3]: Wiener's theory, dynamic programming, the maximum principle of Pontryagin, and
others. However, these methods are complicated with respect to instrumentation (they require a large mem-
ory and a lot of computer time), they are calculated for a single extremum determination, and in addition
they can lead to errors due to inadequacy of the process in the reactor and the model of it used in the control
algorithms [11.
Thus, e.g., in order to calculate a single control cycle in experiments on an HBWR, 5min of computer
time, 7K of operator memory, and 240K of disk memory (K represents 1024 machine words) are necessary.
The difference between the actual process and a linear model of it was eliminated by the introduction of a
large number of artificial variables [2].
Translated from Atomnaya Energiya, Vol. 47, No. 1, pp. 8-12, July, 1979. Original article submitted
February 2, 1978.
506 0038-531X/79/4701-0506$07.50 01980 Plenum Publishing Corporation
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Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020001-4
Deformation of the ED, %
0
1
-2
-
2 :b),
/h
_
3 15
-4
-
4 .=
-5
tt
-6
-
-6
.8
7
mz
-8
-
.g
8 t-.1
1,50
-10
-
1,46
-12-
-14
-
-16
-
i
1,38
-18
-
7,34
-20
I
i I I I
1,30
50 100
150
Time, sec
4%
22,25
21,95
21,65
21,35
21,05
i I I J I 20,75
40 44
0 4 8 12 16 20 24 28 32 36
Iterations
Fig. 1 Fig. 2
Fig. 1. Deformation of the ED in the 30-56 cell (-) at constant average power of the RBMK-1000
reactor of the LAgS, and the position of the CR in the 24-57 cell (---).
Fig. 2. Dependence of Kr (0) and a (0) on the number of iterations of CR control (p(k) =4).
The solution of the problem with the application of a reactor model improved by the method of dynamic
programming [3] required an operator memory of -40,000 words, 130,000 words on a magnetic drum, and
-500 sec of computer time (on a computer of the HITAC 5020E type); in the opinion of the authors, 104 times
more machine time would be required for the usual method of dynamic programming [3]. Therefore, the
question of choosing the most acceptable method of controlling the ED has not been conclusively resolved. On
the one hand practical testing and a sufficient number of experiments are necessary to accomplish this, and on
the other hand a comparison of the different methods in a specific class of problems is necessary.
The most important characteristics of the algorithms are their complexity, the convergence rate, sta-
bilityto noise, computation time, and so on. The difficulties of implementing general optimization schemes
associated with the processing of information of large volume and dimensionality have led to the appearance
of a different kind of iterative heuristic procedures based on a step-by-step (pulse) search for the optimum
and capable of finding the extremum of a goal function with simultaneous refinement of the algorithm param-
eters [4]. The heuristic side of step-by-step search methods has been investigated in [5]; however, it has
seldom been discussed with application to the problems of the optimal control of the ED.
It is advisable in this connection to discuss the extremal control of the ED with the help of a step-by-step
search of the extremum with a combination of test and working steps and in combination with a number of heu-
ristic methods of self-adjustment of the parameters. Heuristic and logical methods of self-adjustment, which
recall the procedure implementable by an operator behind a control panel, minimize the machine time costs
for solution of the problem.
STATEMENT OF THE PROBLEM AND A DESCRIPTION OF AN
ALGORITHM FOR ITS SOLUTION
The problem of extremal control as a method of automatic optimization [6] consists of the minimization
or maximization of some goal function whose value depends on controllable and uncontrollable parameters of
an object.
One can select as the goal function the nonuniformity coefficient of the ED with respect to the reactor
radius:
Kr = max [IV (r, P)/SW0 (r)1;
S W (r, p) du
S ?
W r) dv
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(1)
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where W(r, p) is the actual ED, which depends on the position of the control rods (CR), and Wo(r) is the spec-
ified ED.
Each component of the vector p is of the form
P./ = TIP1j7 (2)
(3)
and
0 a C. At the same time n(k) and nit Ix satisfy the relationships
nun
(4) (4) (OM
int
nmax?nmin-- ?
ctC / '
?ax (k) (4) 13(k)
nm+nmin=n ='T-c
1< p( ' ray incident on the detector near point A at an angle 0 to the
interior normal is 2R cos 0; the probability that the y ray will interact with the detector material is
5E, ft= 1 ?exp (-2?0R cos 0),
where tio is the linear attenuation coefficient for y rays in the material of the detector. We integrate LE ,R
with respect to a to get the number of y rays recorded by the detector in unit time that enter the detector
near A;
1
dl% r rec= (on dS/4p ) ?(a?1 dS/2p) [ (2410R)2 (1 -I- 210R)/(2P0R)2 exP(-2poR)]
We integrate (2) over the surface of the detector and divide the result by Ntot to get the efficiency for
y-ray detection for a homogeneous isotropic medium containing uniformly distributed activity:
(2)
e = (Nrec/Ntot )= I ? (1/2 (LoR)2) ?[(1+2PoR)/2 poR)21 exp (-211/4R). (3)
Translated from Atomnaya Energiya, Vol. 47, No. 1, pp. 41-42, July, 1979. Original article submitted
November 2, 1977.
546 0038-531X/79/4701-0546$07.50 CI 1980 Plenum Publishing Corporation
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Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020001-4
dv
Fig. 1 Fig. 2
Fig. 1. A spherical detector recording y rays in a homogeneous isotropic medium containing
uniformly distributed activity.
Fig. 2. Recording of y rays from a planar source with a spherical detector.
This expression differs from the corresponding formula of [4], and the difference is due to the inexplicit
assumption in the latter study that the angle of distribution of the radiation at the surface of the detector is
isotropic, which our argument shows to be incorrect.
We now consider the efficiency of this detector for y rays from a planar source having any angular
distribution for the radiation, which is however the same for any point on the source. Clearly, such a source
can be represented as the superposition of planar unidirectional sources. In many practical instances one can
neglect the attenuation of the radiation in the medium between the source and the detector, and then the spheri-
cal form of the detector makes the efficiency the same as that for any of the unidirectional sources constitut-
ing the planar source. It is therefore sufficient to consider the radiation from a planar unidirectional source.
The efficiency of a y-ray detector in an isotropic medium containing uniformly distributed activity can
also be derived from the efficiency in the field of a planar unidirectional source [2].
We introduce a cylindrical coordinate system with its origin at the center of the detector and with its z
axis parallel to the radiation direction of the unidirectional source (Fig. 2). The total number of y rays pas-
sing through the surface of the detector in unit time is
= aroaini (e)/cos 01,
where 0 is the angle between the z axis and the normal to the plane of the source, al is the specific activity,
and ni is the quantum yield.
The thickness of the detector m1(p, R) in the direction of arrival of a y ray at a distance p from the z
axis is 2, 1/72-7. Then the y-ray detection efficiency is
R
e (N reSI t'ot )= (1/nR2) S [i?exp ( ? 2?0 -1/ R2 ?p2)1 p dp
o o
1 ? 1-F2NR
=--- 1 2?0R),
2 (?0R)2 2 (p0R)2 exp (
which agrees with (3), as would be expected. The error in [2] is due to an error in, calculating the integral
in (4)
(4)
Numerical integration has been used [3] to determine the efficiency for such a detector in the y-ray field
set up by a planar source with a cosine distribution and arbitrary dimensions. The values for the efficiency
given in [3] for an infinite source agree with those from (3) within the error of the numerical integration.
This efficiency formula can be used in sensitivity calculations for radiometers, as well as in determin-
ing the photoelectric efficiency and the activity of bulk-distributed sources.
lam indebted to Professor V. V. Matveev for valuable comments.
LITERATURE CITED
1. C. Sybesma, Measurements of Continuous Energy Distribution y Rays in a Scattering Medium, Amster-
dam (1961).
2. Yu. A. Egorov, Scintillation Spectrometry of y Rays and Fast Neutrons [in Russian], Gosatomizdat,
Moscow (1963).
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3. D. I. Konstantinov et al., in: Dosimetry and Radiation Protection, Issue 6 [in Russian], Atomiz-
dat, Moscow (1967), p. 121.
4. Yu. A. Sapozhnikov, V. A. Lopatin, and V. P. Ovcharenko, At. Energ. , 40, No. 3, 246 (1976).
5. H. Hurwicz, G. Reau, and M. Storm, in: Physics of Intermediate Reactors [Russian translation], Gos-
atomizdat, Moscow (1961), p. 382.
EXAMINATION OF IRRADIATED METAL DIBORIDES
BY X-RAY DIFFRACTION
Kh. i. Maile, I. A. Naskidashvili, UDC 154-162:539.16.04
and T. Sh. Berdzenishvili
It has been shown [1-4] that metal diborides exposed to high thermal-neutron fluences (? 1020 neutrons/
cm2) undergo considerable changes, even extending to failure. We consider that these changes are due to the
production of helium atoms, which escape from the matrix and accumulate in the intergranular pores. How-
ever, the neutron fluence required to causefailure in metal diborides is reduced by up to a factor 100 if the
material is irradiated at 15?K [5]. This result cannot be explained in terms of the above mechanism, because
it is unlikely that the escape of helium from the matrix will be accelerated on reducing the temperature.
For this reason we examined the changes in lattice parameters of titanium and zirconium diborides and
the corresponding changes in unit-cell volume produced by low-temperature irradiation in a nuclear reactor.
The specimens of ZrB2 and TiB2 were plates of size 10 x 15 x 1.5 mm, which were examined before
and after irradiation by x-ray diffraction with a DRON-1 diffractometer at room temperature. Photographic
powder patterns were recorded for the specimens; the diffraction lines were broad rings composed of indi-
vidual spots. The distortion of the line shape caused by the large grain size was eliminated with the DRON-1
by oscillating the specimen in its own plane.
The specimens were irradiated in low-temperature channels of the reactor at 110 and 300?K, where the
temperatures did not fluctuate by more than ?5?K throughout the irradiation.
Figures 1-3 and Table 1 give the results for the ZrB2 and TiB2 specimens, in which the changes in lat-
tice parameters La/ain and 6.c/cin are given along with the changes in unit-cell volume ?V11. The reflec-
tions from the irradiated specimens are displaced towards smaller angles (the more so the higher the fluence),
which indicates an increase in the lattice parameters, which is usually caused by interstitial atoms or groups.
As the lattice parameters were measured at room temperature (although the specimens were irradiated at
110?K), only the helium atoms will largely persist as free interstitial atoms out of all those produced by the
irradiation (the helium is derived from the 10B). Therefore, the observed increase in the lattice parameters,
particularly in the c parameter, must be due to predominant accumulation of interstitial helium atoms in the
lattice.
Table 1 shows that (Lic/c)/(Atz/a) 1 fpr ZrB2 on low-temperature irradiation, which apparently means that the lattice of ZrB2 retains
more helium atoms than does TiB2. This is due on the one hand to the smaller size of the pores in TiB2 (be-
cause the lattice parameters are smaller than those of ZrB2) and on the other to the stronger bonds between
the atomic layers of metal and boron in TiB2 [1, 3, 6], and therefore the helium atoms are ejected from the
TiB2 matrix more readily. If we assume that the change in lattice parameters is due in the main to intersti-
tial helium, we have that the microcracks and failure are due to the internal stresses set up by these inter-
stitial atoms, not to the marked increase in the helium pressure in the pores. If on the other hand the cause
of the failure were such pressure rise, then reducing the radiation temperature, which greatly reduces the
mobility of the helium, should mean that the failure would occur at higher neutron fluences. However, the
failure in T1B2 in fact occurs at a much lower neutron fluence on reducing the irradiation temperature to 110?K
[3-5].
Translated from Atomnaya Energiya, Vol. 47, No. 1, pp,. 42-44, July, 1979. Original article sub-
mitted December 28, 1977.
548 0038-531X /79/4701-0548S07.50 ?1980 Plenum Publishing Corporation
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Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020001-4
TABLE 1
1110
20
15
10
Specimen
Low-temperature irradiation
Titr = "13?K
T ? 3= 00?K
ur
radiation dose,
neutrons/cm2
Aa
%
Ac %
AV
oz,
An 04
Ac
AV
gin
cin
Vin "' .
ain "
%
ein
%
viri
ZrB2
54016
2.1016
1.1017
--
0,08
0,19
0,04
0,10
0,42
0,04
0,26
0,80
0,01
0,04
0,21
0,01
0,06
0,35
0,03
0,14
0,77
TiB2
2.1016
1.1017
5-1017
0,04
0,10
0,46
0,01
0,08
0,25
0,09
0,28
1,17
--
0,08
0,34
--
0,06
0,22
--
0,22
0,90
High-temperature irradiation (4]
T 1.1T = 57 09X
ZrB.
TiB,
1070
1070
0,30
1,50
0,23
0,13
0,83
3,13
65
60
55
50
Zr 62
123
134 735 736 20- 153 154 155 2/9
Fig. 1 . Fig. 2 .
Fig. 1. The (023) diffraction peaks of TiB2 specimens: I) not irradiated; II) exposed to a neutron flu-
ence of 1.1017 neutrons/cm2 at 110?K; ILI) the same for 5.1017 neutrons/cm2.
Fig. 2. The (123) diffraction peaks of ZrB2 specimens: I) not irradiated; II) exposed to a neutron
fluence of 2.1016 neutrons/cm2 at 110?K; III) the same for 1.1017 neutrons/cm2.
No matter what the irradiation temperature, the maximum increase in unit-cell volume for ZrB2 before
the onset of failure is about 0.8%; we therefore conclude that there is a critical concentration of interstitial
atoms in ZrB2 above which the expansive force becomes greater than the binding forces between the alternat-
ing crystallographic basal planes. This leads to mechanical failure of the crystal. Prolonged irradiation
results in reduction in the grain size of the crystal without change in the increment in cell volume, whose
value remains around 0.8% for ZrB2.
Figure 3 shows that the increase Ac/c produced by the radiation in ZrB2 is almost completely elimina-
ted by annealing the specimen at 670?K, whereas Aa/a is then reduced by only 70%. This explains the differ-
ences in the data on the lattice-parameter changes obtained on low-temperature and high-temperature forms
of irradiation for ZrB2. At high temperatures (Tirr 600?K), the helium atoms escape into the intergranular
pores, and only atoms bound to lattice distortions remain in the diboride lattice, and the concentration of
these is much lower. Therefore, the diboride withstands a much higher neutron fluence at high irradiation
temperatures because most of the helium remains in the lattice at low temperatures.
When specimens irradiated at 15 or 120?K are heated to room temperature, there is fairly substantial
annealing of the radiation-induced defects [5], and it is therefore obvious that the stages of isochronous an-
nealing seen in the radiation-induced increase in the lattice parameters at 300-670?K (Fig. 3) cannot be due
to annealing of single defects. These stages in the annealing may be due to the flow of various defect com-
plexes to sinks (helium atoms plus vacancies, pairs formed by interstitial atoms with lithium, etc.), or else
the breakup of these into isolated defects, which at these temperatures also migrate to sinks.
549
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Fig. 3. Isochronous-annealing curves
for the radiation-induced increases in
lattice parameters for zirconium di-
boride: 0) Ac/Aco; *) 11a/ [Sao.
Therefore, these results and the available evidence [2-5] indicate that the failure in metal diborides
during reactor irradiation is due mainly to internal stresses in the matrix produced by interstitial helium
atoms.
Z. A. Titik assisted in the experiments, and we are also indebted to 0. I. Yurina for assistance with
the calculations and to our colleagues on the nuclear reactor at the Institute of Physics, Academy of Sciences
of the Georgian SSR, for general collaboration in performing the low-temperature irradiations.
LITERATURE CITED
1. G. V. Samsonov, At. Energ. , 14, No. 6, 588 (1963).
2. V. V. Ogorodnikov et al., At. Energ., 23, No. 4, 341 (1967).
3. G. V. Samsonov et al. , At. Energ. , 24, No. 2, 191 (1968).
4. M. S. Kovalfchenko and V. V. Ogorodnikov, Fiz. Khim. Obrab. Mat., No. 4, 14 (1971).
5. L. S. Topchyan et al., At. Energ. , 42, No. 3, 226 (1977).
6. G. V. Samsonov, Dokl. Akad. Nauk SSSR, No. 83, 689 (1953).
NEUTRON SPECTRA IN THE MeV RANGE
IN FAST CRITICAL ASSEMBLIES
V. M. Lityaev, V. A. Dulin, UDC 621.039.51
and Yu. A. Kazanskii
Neutron spectra are extremely important reactor characteristics, because the quantitative aspect of
any process in a reactor may be represented by averaging the neutron interaction cross sections over the
neutron spectrum. A spectrum itself is dependent in a complicated fashion on the cross sections for inter-
action of the neutrons with the materials in the reactor. Further, the theoretical spectrum is also dependent
on the approximation used in the treatment. Therefore, neutron-spectrum measurements in reactors are of
some interest. Measured and theoretical values provide refinement of the inelastic-scattering matrix for
238k and estimation of the error of the group approximation arising near fission thresholds and thresholds for
(n, 2n) reactions.
A scintillation spectrometer with y-ray discrimination by decay time in a stilbene crystal of diameter
7 mm and thickness 8 mm was used to measure the neutron spectra at the centers of three critical assemblies;
this crystal was mounted in a Teflon jacket with a wall thickness of 2 mm. The working principle has been
described previously [1]. The energy scale was calibrated from the upper limits of the Compton distributions
given by y rays of known energy from 137Cs, 22Na 65Z n, -- RR
Y sources and the 4.42 MeV y-ray line from a Po-
a-Be source. The spectrometer was also calibrated with monoenergetic neutrons from a Van der Graaf ac-
celerator. The reference neutron source was a miniature isotropic 252Cf source. The apparatus recoil-pro-
ton spectra were converted to neutron energy spectra by differentiation [2] with respect to energy with a vari-
able step (-1-E) (Fig. 1). A correction was made for the edge effect [3], whose values were 1 and 13.5%, re-
spectively, at 0.8-1.4 and 6.5-10.5 MeV. In measurements on the 252Cf source, the temperature in the energy
range 0.4-10 MeV was 1.41 ?0.03 MeV, which agrees with published values.
Translated from Atomnaya Energiya, Vol. 47, No. 1, pp.44-45, July, 1979. Original article submitted
January 30, 1978.
550 0038-531X/79/4701-0550$07.50 1980 Plenum
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Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020001-4
14
12
10
4
2
2,0
4
1,5?
S
-4
2
45
0 2 4 6 8 E, MeV ?,51.0 1,4 1,8
Fig. 1 Fig. 2
Fig. 1. Neutron spectra: 1) at the center of the BFS-35-1 assembly; 2) the same for the
BFS-33 assembly; 3) for a 252Cf source.
2,2 E, MeV
?
Fig. 2. Ratio K(E) of the measured spectra to a 1/E spectrum for the following assemblies:
1) BFS-30; 2) BFS-35-1.
TABLE 1. Measured and Calculated Neutron
Fluxes
Assembly
Energy
range,
MeVMeV
g'exp
TleX p. ',Pith with
matrices given in
[7]
[8]
BFS-30
10,5--6,5
0,01047
0,878
0,914
6,5-4,0
0,0636
0,887
0,886
4,0--2,5
0,1640
0,965
0,927
2,5--1,4
0,3510
1,050
1,044
0,459
1,4-0,8
0,4107
0,996
1,018
0,0317
BFS-33
10,5-6,5
0,0114
0,943
1,029
6,5-4,0
0,0792
1,092
1,115
4,0--2,5
0,1675
0,991
0,950
2,5-1,4
0,3208
0,968
0,957
0,461
1,4-0,8
0,4212
1,015
1,037
0,0291'
BFS-35-1
10,5-6,5
0,01432
1,225
0,225
6,5--4,0
0,0755
1,282
1,083
4,0-2,5
0i1521
1,197
1,998
2,5--1,4
0,2820
1,161
1,046
0,457
1,4--0,8
0,4761
0,851
1,959
0,0274
1/E
2,5-1,4
0,467
neutron
spectrum
1,4-0,8
0,0338
The main errors in measuring the neutron spectra' arise from errors in converting the pulse-height dis-
tributions to energy spectra; if one assumes that the Terrell analytical representation applies for a fission-
neutron spectrum of 252Cf, the likely error in determining the neutron flux at 10 MeV relative to 1 MeV is
about 15%.
The BFS-30 assembly has a composition close to that in the high-enrichment zone of a fast power re-
actor with oxide fuel and sodium cooling, while the BFS-33 and -35-1 were constituted by media of uranium
oxide and metallic uranium whose enrichment was such as to produce Keo 1 [4, 5].
The neutron fluxes were calculated by means of the M-26 program [6] with the BNAB-70 constants [7]
(see Table 1). Table 1 also gives the ratios of the measured group neutron fluxes cpi to the theoretical ones
for the various inelastic-scattering matrices for 238U [7, 8]. The differences between the group fluxes
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calculated with the various a matrices increase with the ratio of 238U to 235U (this was 16.8 for the BFS-
35-1 assembly) and as the proportions of constructional and moderator materials decrease. The measured
group fluxes agreed best with those calculated from the matrix of [8]. Figure 2 shows how the measured
neutron spectra differ from a 1/E spectrum. Table 1 also gives the mean group fission cross sections for
2381J [9] as obtained by averaging for a 1/E spectrum and as derived from the measured spectra. The chan-
ges in the mean fission cross section for 238U are, respectively, ?1.2%,-1.1%, and-2%fortheBFS-30,-33,
and -35-1 when the measured neutron spectra are used instead of a 1/E spectrum in the averaging.
The value of the theoretical neutron flux above 6 MeV is important in correct allowance for the (n,2n)
reactions on 239PU (23Na, 238U); Table 1 shows that the calculations and measurements are in agreement in this
regi013.
LITERATURE CITED
1. F. Brooks, Nucl. Instrum. Meth., 4, 151 (1959).
2. C. Lanczos, Applied Analysis, Prentice-Hall (1956).
3. Fast-Neutron Physics [in Russian], Vol. 1, Gosatomizdat, Moscow (1963).
4. V. A. Dunn et al., At. Energ., 40, No. 5, 377 (1976).
5. E. N. Kuzin et al., FEI-698 Preprint, Obninsk (1976).
6. Sh. S. Nikolaishvili et al., in: Proceedings of the Trilateral Soviet?Belgian?Dutch Symposium on Fast-
Neutron Physics [in Russian], Vol. 1, Izd. TsNHatominforma, Moscow (1970), p. 37.
7. L. P. Abagyan et al., Group Constants for Calculations on Nuclear Reactors [in Russian], Atomizdat,
Moscow (1964).
8. A. S. Krivtsov and V. I. Popov, in: Neutron Physics [in Russian], Vol. 4, Izd. TsNHatominforma,
Moscow (1977), p.113.
9. Y. Stehn et al., BNL-325, Supp. No. 2 (1965).
EFFECTS OF VARIOUS FACTORS ON THE
ABSORBED-DOSE DISTRIBUTION IN THIN LAYERS
V. V. Krayushkin UDC 539.12.08
A specified distribution of absorbed radiation energy within the object must be provided in applied
radiation-treatment systems, including industrial ones; we have examined the distribution of the absorbed
dose D over the thickness of the material in relation to the angle of incidence of the electron beam, and also
the effects of back-scattering metal plates on the distribution of D in thin layers (thickness less than the
electron stopping length).
The experiments were performed with electrostatic accelerators at energies of 0.3-1.0 MeV. A thin
foil in the exit window resulted in a virtually monoenergetic electron beam. The dosimeters were cellulose
triacetate (CTA) films of thicknesses 70 and 140 Am (p = 1.25 g/cm3), which simulated thin layers of the target
material. The value of D was determined from the difference in the optical densities of the CTA films be-
fore and after irradiation as measured with an SFD-2 spectrophotometer by a standard method [1] giving a
coefficient of variation of ?8%. Stacks of CTA films were used to adjust the thickness.
The distribution of D in a homogeneous material exposed to a beam incident at right angles containing
electrons of energy 0.3-2.0 MeV can be put as
D (x) r=(Dmax/2) [I +sin (0.2 ? 0x)], (1)
where x is measured in the material in g/cm2,
Dmax is the maximum value of D within the material, and
is a parameter dependent on the electron energy [2].
Experiments with these CTA films were performed with the electrons incident at various angles a,
which led to the empirical formula
Da (x)=-- [(1 ? cos a)/41 Dmax (1+ sin [4.7 ? (0? ? 4.5) cos a]).
(2)
Translated from Atomnaya Energiya, Vol. 47, No. 1, pp. 46-47, July, 1979. Original article sub-
mitted February 13, 1979.
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44
1,0
48
-? 46
82 44
42
42 0,4
0,6 0,8 1,0 1,2
Thickness, mm
1,4 1,6
1,2
10
:74 48
gz.1 0,6
!?-
C
44
0,2
44 co 1,2 1,6 2,0 2,4
Thickness, mm
28
Fig. 1 Fig. 2
Fig. 1. Distribution of the absorbed dose over the thickness of CTA films for angles of inci-
dence of the electron beam of 0 (1), 20 (2), 40 (3), and 60? (4) (the points are from experiment).
Fig. 2. Absorbed-dose distributions for specimens of thickness 0.84 (1), 1.68 (2), and 2.1 mm
(3); (4) material irradiated without a substrate (semi-infinite geometry).
1,4
1,2
T;
40
48
42-
46
0 0,1 0,2 0,3 0,4
410 10 20 30 40 50
Thickness, mm
Fig. 3 Fig. 4
Fig. 3. Reflected-dose factor in relation to atomic number of metal substrate at energies
of 0.4 (1), 0.7 (2), and 1.0 MeV (3).
Fig. 4. Absorbed-dose distribution in CTA films of thickness 0.42 mm exposed to 0.5 MeV
electrons on substrates of: 1) lead; 2) nickel; 3) aluminum; 4)no substrate (semi-infinite ge-
ometry); the points are from experiment and the lines are from theory.
'ref
44
3
60 70 80 Z
45 46
If the beam is incident normally, i. e., if cos a = 1, then (2) becomes (1).
Figure 1 shows the distribution of D over the CTA films for various angles of incidence at 0.5 MeV;
there is agreement to within ?7%. It is also of interest to examine the distribution of D in a layer of material
irradiated on a metal substrate, since then there is back-scattering from the metal substrate if the thickness
of the material is less than electron range, and therefore there is an increase in D near the surface of the
substrate.
The metals used in the experiments were aluminum, iron, nickel, copper, tin, brass, and lead. These
were used as plates of thickness 1 mm (this thickness was greater than the electron range in the relevant ener-
gy), and these were fitted with several CTA films assembled in stacks.
Figure 2 shows the distribution of D in material on a tin substrate exposed to a 0.8-MeV beam. The
value of D in the material increases the more substantially the less the thickness for a given electron energy.
The distribution of D over the thickness on irradiation on a metal substrate can be put as the sum of two
functions:
D t(x)=--
r D (x) for x < 2a ?4,5/0;
1 D (x)? TfrefD (24 ?x) for x > 2a ?4,510,
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(3)
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where D(x) is the distribution of D when the material is irradiated without a substrate by a beam in normal
incidence, as (1) shows, and nlref is the dose factor representing backscattering from metal substrate i, and
a is the thickness of the irradiated material in g/cm2.
The backscattering correction factors were determined by experiment with the beam incident normally
on the target for various electron energies; D increased for CTA films on metal substrates of high atomic
number (under otherwise equal conditions). Graphs were drawn up relating the backscattering factor to the
atomic number for various electron energies (Fig. 3). The results for the effective reflection factor are in
good agreement with the measurements of [3].
Figure 4 shows the distribution of D for material on substrates of various metals for an electron energy
of 0.5 MeV; the following formula was devised to calculate D for various angles of incidence in the presence
of backscattering substrates:
Da (x) for x < 2a-4.5/0;
Dat (x) =
(x)? riref Da (2a ?x) for x>
> 2a ?4.5/0.
(4)
The observed and calculated values agreed to within ?10%, which indicates that one is justified in using
empirical formulas (2)-(4) to calculate the major parameters of radiation-treatment systems.
LITERATURE CITED
1. A. K. Pikaev, Dosimetry in Radiation Chemistry [in Russian], Nauka, Moscow (1975).
2. V. V. Krayushkin and V. P. Suminova, in: Proceedings of the Second All-Union Conference on Indus-
trial Uses of Charged-Particle Accelerators [in Russian], Vol. 1, Izd. NIgFA, Leningrad (1976),
p. 347.
3. W. Bothe, Ann. Phys., 6, 44 (1949).
A CALORIMETER FOR MEASURING LOCAL
ELECTRON-BEAM ABSORBED DOSES
V. A. Berlyand, V. V. Gen eralova, UDC 539.12.08
and M. N. Gurskii
Electron beams of energy between 0.2 and 10 MeV are widely used in radiobiology, industrial applica-
tions, and research in many areas; exact measurements on absorbed dose are performed by means of point
dosemeters or calorimeters. Many such calorimeters have been described [1] for electrons of energy over
1-2 MeV.
The instrument of [2] allows measurements to be made at electron energies above 150 keV; a single
component acts as absorber, temperature transducer, and heater, namely a conducting polyamide film [3],
which is made from polyamide lacquer with the addition of carbon black. These films are of high thermal
stability and can work at temperatures of 350-400?C for several months or briefly (a few hours) at 450-500?C;
the radiation resistance of such a film is also higher by several orders of magnitude than that of the most
radiation-resistant polymer, polystyrene. The thickness of the film can be varied over the range 1.5-6 mg/
cm2. The carbon content is 20-40 mass%. Measurements show that the thermal defect is not more than 0.3%.
The temperature coefficient of resistance of the material is negative, # = ?0.0004, and this is less by
about an order of magnitude than the values for copper and platinum, but the value is still quite sufficient for
the measurement of dose rates above 103 rad/sec.
The calorimeter proper is a strip of this film 10 x 20 mm; the inherent resistance is ?200-300 SZ, and
this is inserted in one arm of a de bridge. The bridge is balanced by adjusting the input current at some tem-
perature Tb, which is above the environmental temperature. The absorbed dose rate.P is defined by
p (I? ? 11) R Tb (1)
Translated from Atomnaya Energiya, Vol. 47, No. 1, pp. 47-48, July, 1979. Original article submitted
March 23, 1976; revision submitted February 6, 1979.
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1 2 3' 4 5 1/
\
r-c \ A

Fig. 1. The dose-rate meter: 1) calori-
meter; 2) screen of diameter 40 mm; 3)
nylon filaments; 4) radiator; 5) hole of
diameter 35 mm; 6) mounting ring; 7)
screening body of 60-mm diameter.
where I and 12 are the currents in the device for which the bridge is balanced before and during the irradiation
respectively, while R(Tb) is the resistance of the device at temperature Tb and M is the mass of the device.
Instruments have been devised in uhf engineering [4] for keeping such bridges constantly in balance automati-
cally; these can be used in conjunction with this ealorimeter in ionizing-radiation dosimetry.
In our dosemeter (Fig. 1), the calorimeter is surrounded by screens also made of the same material in
order to minimize effects from' environmental-temperature fluctuations and convection. The distance between
the calorimeter and the screen is 1 mm. .The design of the mounting ring allows radiators of various materials
to be mounted on both sides, and it is therefore possible to make measurements in any material at any depth.
The screens and the calorimeter are made of the same material, so the radiation itself sets up quasiadiabatic
conditions. If the measurement time is much less than the time constant of the calorimeter, which is about
30 sec, one can assume that
P = k (dR I dt),
(2)
where dR/dt is the rate of change of the resistance and k is a coefficient of proportionality. This means that
measurements can be made in dynamic mode, which is often preferable, because the electron-beam intensities
are high and the time stability is poor.
This instrument in principle provides a means of precision dose-rate measurement; the calorimeter
employing this material contains no foreign materials and therefore there are no errors due to energy deposi-
tion in foreign inclusions. The material combines the functions of absorber, heater, and transducer, and
therefore there is virtually complete identity in the substitution of electrical heating for radiation heating.
The systematic error in dose-rate measurement is 0.7% at the 0.95 confidence level, while the random error,
expressed as the coefficient of variation, is not more than 2%. The calorimeter provides for dose-rate mea-
surement on electron beams between 103 and 2.107 rad/sec for electron energies from 0.15 to 10 MeV or more,
while it substantially simplifies the methods used in calibrating film chemical dosimeters in which a total-ab-
sorption calorimeter is required [5]. Also, the film calorimeter can be used in the dosimetry of photon radia-
tion, including low-energy x rays, as well as the p radiation from large radiation sources.
,
LITERATURE CITED
1. S. Gunn, Nucl. Instrum. Meth., 85, No. 2, 285 (1970). .
2. V. A. Berlyand, V. V. Generalova, and M. N. Gurskii, Inventor's Certificate No. 593554, Byull. Izo-
bret. , No. 39, 212 (1978).
3. V. E. Gul', V. F. Blinov, and M. G. Golubeva, Plastmassy, No. 9, 45 (1976).
4. M. I. Bil'ko et al., UHF Power Measurement [in Russian], Soy. Radio, Moscow (1976).
5. V. A. Berlyand et al., At. Energ. , 42, No. 3, 199 (1977).
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DETERMINATION OF NUCLEAR CONSTANTS AS AN
INVERSE PROBLEM IN RADIATION TRANSPORT
V. N. Dushin UDC 539.125.52
The values of nuclear constants obtained by measurement are distorted by various effects arising from
the finite dimensions of the systems, including attenuation of radiation fluxes, multiple scattering, etc. The
requirements of reactor engineering impose closer tolerances on the determination of constants, and these
can no longer be met by correcting for such effects via average factors or terms. Therefore, one has to con-
sider the extraction of nuclear constants from the data of neutron experiments as an inverse problem in the
theory of radiation transport [1], for which we give a method applicable to the real geometry, and for which
we demonstrate the good performance in recovery of an inelastic-scattering cross section.
We write the transport equation for the system as
(x)--- K (x' x)(1)(x')dx' S (x),
(1)
where 4,(x) is the particle-collision density, x= (r, S2, E), S (x) is the collision density for the first collision,
and K (x' ?x) is the kernel of the transport equation. The unknown parameters (cross section, scattering
indicatrix) appear in K(x' ?x). The collision density 4,(x) is measured or is calculated from measurements
for various values of x and is therefore known approximately. This 4,(x) may not correspond to any physical
kernel K(x, ?x), and therefore one has to avoid solving the inverse problem in the ordinary sense of the word
and instead consider the definition of some general solution. Following [2], we take the functions a (x) as the
solution (o- (x) are the desired nuclear constants), which are such as to provide
IP 0= if (a),
o(x)EC
where
V (a) (x)-4)2 (x)j2 dx.
Here 4.e(x) is the experimental collision density, while 4) (x) is the collision density calculated by means of
o-(x),andD is the region of definition (measurement) of 4)e (x), 4, (x)? L2; the condition for physical reality of
the target parameters requires that
(2)
(3)
(4)
Therefore, the inverse problem in radiation transport amounts to defining the turning point in the function of
(3) subject to conditions (4) and (1). The Monte Carlo method is used to solve (1) on account of the complexity
of the geometrical conditions in real experiments; this makes the problems stochastic. Stochastic quasigra-
dients [3] are used in solving stochastic turning-point problems, where this gradient is a random quantity
whose mathematical expectation is similar to a gradient or generalized gradient in a certain sense. The turn-
ing point is defined by a sequence of random vectors
as, s= 0, 1, 2, ...; as+i = nx (as? psvs); s= 0, 1, 2, ..., (5)
where Irx is the operator for projection of the acceptable values of the parameters a on the region; uo is an
arbitrary initial point, as is the step size, and vs is the stochastic quasigradient vector, whose mathematical
expectation is
M (v'/a?, GI, GS) - slir (as)?b'.
(6)
Here as is a nonnegative random quantity, bs is a random vector, and ira(crs) is the vector for the generalized
gradient, i.e., a vector that for any z satisfies
Translated from Atomnaya Energiya, Vol. 47, No. 1, pp. 48-50, July, 1979. Original article submitted
May 22, 1978.
556 0038-531X/79/4701-0556807.50 0 1980 Plenum Publishing Corporation
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4
?
A
A
A ?
A
A A A A
A I
2 4 6 6 10
Iterations
12
/4
402-
0
0,913- b
0,015
4?1
4005
0
1
2 3 4 E, MeV
Fig. 1 Fig. 2
Fig. 1. Convergence of the minimization in a model experiment; specimen size about 0.5 mean
free path.
Fig. 2. Results from processing inelastic-scattering spectra for neutrons in lead; a) recovered
cross section; b) solid line ? observed spectrum, broken line ? spectrum calculated from recovered
cross section.
,11 (z)?T (as) Oa (a.), z ? a.).
(7)
The process of (5) may not give a monotonic reduction in the target function 4,(o-) at each step, and this
is a qualitative difference from the usual gradient method. Some theorems [3, 4] ensure that the process of
(5) converges in an infinite number of steps. Only a restricted number of tests can be performed in a practi-
cal case, which is equivalent to restricting the processing time, and this requires the introduction of some
statistical criterion that will provide some given error level in the functional, Such a criterion can be used
with known values for the variances of 4) (x) and ,I)e (x) to estimate whether a statistically significant deviation
of 4, (a) from zero occurs, and therefore to determine the points at which the search is stopped. If the target
function does not decrease during the minimization and also remains statistically significant, this means either
that the gradient describing the radiation transport is inadequate to describe the experimental conditions or that
there are systematic errors in the experimental data.
As an example, we give results on the inelastic-scattering cross section for neutrons of energy 4.7 MeV
[5]. The experiments were performed at the Khlopin Radium Institute with cylindrical geometry by the time-
of-flight method. A program was written for the BESM-6, which used a subroutine for solving the transport
equation by Monte Carlo methods. Local estimation of the neutron flux is employed [6] together with statisti-
cal weights that incorporate the loss of neutrons from the specimen and the absorption. Random termination
of the paths was also incorporated. The partial derivatives of the flux required to calculate the stochastic
quasigradient were estimated at the same time as the neutron flux itself [7]. The workability of the algorithm
was checked out on a model experiment. The spectrum of inelastically scattered neutrons calculated in this
way was used to recover the initial model cross section. Figure 1 illustrates the convergence of the minimi-
zation, where 6 is defined as
6= {s (Fexp(E)?F (E)121F (E) "2. (8)
Here Fexp (E) and F (E) are the measured and calculated neutron fluxes; as Fexp (E) and F (E) are calculated
with an error of about 1.6%, we can say that the minimization converges in 5-7 iterations.
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This method was used to recover inelastic-scattering cross sections from experiments with specimens
of size up to two mean free paths. Figure 2 shows a typical example.
This method can also be used to optimize the shape and size of the specimen, as well as the parameters
of the time-of-flight spectrometer.
LITERATURE CITED
1. A. V. Vantkpv, in: Nuclear Science and Technology, Nuclear Constants Series [in Russian], No. 16,
Atomizdat, Moscow (1974), p. 11.
2. A. N. Tikhonov and V. Ya. Arsenin, Methods of Solving Incorrectly Formulated Problems [in Russian],
Nauka, Moscow (1974)?
3. Yu. M. Ermol'ev, Stochastic Programming Methods [in Russian], Nauka, Moscow (1967).
4. Yu. M. Ermol'ev and Z. V. Nekrylova, Kibernetika, 6, 4 (1966).
5. V. N. Dushin et al. , [1], No. 24 (1977), p. 27.
6. V. G. Zolotuldiin and S. M. Ermakov, in:-Physics of Reactor Shielding [in Russian], Gosatomizdat,
Moscow (1963), p. 171.
7. E. V. Pletnikov and G. Ya. Trukhanov, [1] (1974), p. 106.
EFFECTS OF BOMBARDMENT BY He, Nit, AND Cr+
ON MICROHARDNESS AND CORROSION CRACKING OF
STAINLESS STEELS
B. G. Vladimirov, V. M. Gusev, UDC 620.197
and V. S. Tsyplenkov
One reason for the failure of components made of stainless steels and alloys working at elevated tem-
peratures in water or water?steam mixtures is intercrystallite corrosion followed by cracking under stress
[1]. The cracks arise under such conditions at the solid?liquid (or solid?gas) interface, where the state of
the metal surface largely determines the rate and type of corrosion failure, particularly in the initial stages.
Improved resistance to corrosion cracking is provided by mechanical treatment that produces compressive
stresses in the surface layer [2-4].
Also, bombardment with various ions alters the mechanical and physicochemical parameters of the sur-
face layers because the implanted ions produce chemical effects and various defects. In particular, the im-
plantation of He+, Ni+, and Cr+ has been used to improve the corrosion resistance in metals [5, 6]. It is
therefore of interest to examine the scope for improving the corrosion resistance of stainless steel by surface
treatment with He+, Ni+, and Cr+ ions at energies of rather more than 10 keV.
Methods
Rectangular plates (size 80 x 5 x 0.6 mm) of stainless steels type OKh16N15M3B of the austenite class
(after heating to 1050?C) and type 1Kh13S2M2 of the ferrite ?martensite class (after annealing at 850?C)
were used with grain sizes of 20-30 tim.
The steels were bombarded with He+, Ni+, and Cr+ ions at 40 keV in an ILU-3 accelerator [7] at a cur-
rent density of 15-20 ?A/cm2 to a total dose of 1018 ions/cm2. The specimens were set up in two different de-
vices. One of these provided for maintenance of a set temperature during irradiation (e.g., 500?C on treat-
ment with Cr+ ions). In the other, the specimens were mounted on a water-cooled table. However, the poor
thermal contact meant that the temperature indicated by a Chromel?Alumel thermocouple rose to about 100?C
during the bombardment.
The corrosion tests on the original and irradiated specimens were performed in an autoclave made of
OKh18N1OT steel of volume 0.35 liter with an aqueous solution of FeC13 having a chloride concentration of 100
Translated from Atomnaya Energiya, Vol. 47, No. 1, pp. 50-51, July, 1979. Original article submitted
June 26, 1978.
558 0038-531X/79/4701-0558807.50 1980 Plenum Publishing Corporation
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TABLE 1. Results of Tests on Corrosion Cracking
of Steels in FeC13 at 360?C and a Pressure of 190
mm Hg (chloride concentration 100 mg/liter)
DoOingele-
ment (ion)
Irradiation
temperature
?C
Time to failure, h
OKh16N15M13B
1Kh13622v12
He
Ni +
Cr +
Cr +
Original
specimen
?100
?100
?100
?500
350
350
230
30
130
90
30
60
70
40
210
190
170
150
. 130
...vaa 210
2
no
170
150
130
110
0 20 40 60 80 0 20 40 60 80
Distance from surface, um
Fig. 1. Change in the microhardness
of steels type OKh16N15M3B (open
circles) and 110113S2N2 (filled circles)
over the cross section after surface
implantation at 100?C of helium (a),
nickel (b), and chromium ions (c), and
also chromium at about 500?C (d).
mg/liter at 360?C and 190 mm Hg. The stresses in the specimens were produced by bending in a circular
mount [1]. The strain in the stretched outer surface layer after bending was 5%. Constant strain during the
test was ensured by contact welding of the ends of the specimen. The onset of failure (occurrence of crack-
ing) was observed by examination with an MBS-2 microscope every 10 h in tests lasting up to 100 h, 20 h for
tests lasting up to 200 h, and 90 h for tests lasting up to 500 h. The FeCl3 solution was replaced after each
cycle; the tests were continued until the first visible crack appeared. Also, a PMT-3 instrument was used
with a load of 100 g to measure the microhardness at various depths by oblique-section methods.
Experiment Results
The measurements showed that irradiation of Olch16N15M3B steel with nickel, helium, or chromium ions
at temperatures up to 100?C increased the microhardness down to depths of 40-60 pm; the ranges of He+, Cr,
and Ni + ions of energy 40 keV in steel are [8] substantially less than 1 pm, and the change in microhardness
to much greater depths may be due to radiation-stimulated diffusion of point defects and the formation of dis-
location loops, which distort the lattice and harden the material [9].
Figure la-c shows that the implantation of the metal ions (nickel and chromium) produces a greater in-
crease in the microhardness at the surface than does implantation of helium; however, helium produces a
layer of constant hardness (1-42kgf/mm2) throughout a thickness of about 30 pm, whereas the metals produced
a rapidly decreasing microhardness. A similar layer of thickness about 20 pm of constant but elevated micro-
hardness (about 210 kgf/mm2) was produced by bombarding 1Kh13S2M2 ferrite-martensite steel with helium
(Fig. la).
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Figure id shows that there is a complicated change in the microhardness for OKh16N15M3B steel ex-
posed to chromium ions at 500?C; this is clearly the joint effect of several processes, some of which produce
hardening and others softening. In particular, lattice distortion due to radiation defects and the formation
of finely divided phases tends to harden the implantation layer. However, the radiation may also accelerate
the stress relaxation. The peak (Fig. 1d) may reflect the joint effects of these various processes. Type
1Kh13S2M2 steel has a bcc lattice (Fig. lb-d), and stress relaxation appears to be the dominant effect on
bombardment with metal ions.
The corrosion results (Table 1) show that surface treatment with helium, nickel, and chromium ions
at about 100?C substantially increases the time to failure for OKh16N15M3B steel; by a factor 2.7 for nickel
and helium ions or by 1.7 for chromium. Table 1 shows that helium ions (and to a smaller extent chromium
ions) also increase the time to failure for ferrite- martensite steel.
Figure 1 and Table 1 show that there is a correlation between the change in microhardness and the cor-
rosion resistance in the implantation layer, apart from the case of 1Kh13S2M2 steel exposed to Cr + ions. The
improved corrosion resistance in the implantation layer may be due to a surface barrier of doping atoms,
which prevents the chloride ions from diffusing into the metal and also may prevent the diffusion of oxygen and
other corrosion stimulators. There was a certain increase in the corrosion resistance of 1Kh13S2M2 steel
after bombardment with Cr+ undoubtedly on account of the surface-barrier effect.
We are indebted to V. P. Voeikov for considerable assistance in performing the corrosion tests.
LITERATURE CITED
1. V. P. Pogodin, V. L. Bogoyavlenskii, and V. P. Sentyurev, Intercrystallite Corrosion and Corrosion
Cracking in Steels in Aqueous Solutions [in Russian], Atomizdat, Moscow (1970).
2. H. Suss, Corrosion, 18, No. 1, 17 (1962); 17, No. 2, 63 (1961).
3. X. Tozano, J. Jpn. Inst. Met., 24, No. 12, 786 (1960).
4. W. Ruttmann and T. Gunther, Werkst. Korros. , No. 2, 104 (1965).
5. Yu. M. Khirnyi and A. P. Solodovnikov, Dokl. Akad. Nauk SSSR, 214, No. 1, 82 (1974).
6. A. Autill et al., Corr. Sci., 16, No. 10, 729 (1976).
7. N. P. Busharov et al., Prib. Tekh. Eksp., 4, 19 (1967).
8. J. Lindhard, M. Scharff, and H. Schiott, Kgl. Danske Vid. Selskab, Mat.-Fys. Medd., 33, No. 14
(1963).
9. V. N. Bykov and V. A. Troyan, Phys. Status Solidi, 32, 53 (1975).
NONSTATIONARY PROMPT-NEUTRON
DIFFUSION IN A FAST ASSEMBLY
V. E. Kolesov, 0. I. Makarov, UDC 621.039.514.4
and I. P. Matveenko
The characteristic time dependence ipa (i-, V) exp (-a0t) occurs in experiments with pulsed prompt-neu-
tron sources after the transient period, where a0 describes the asymptotic decay of the neutron pulse and is
the same no matter what detector is used at any point in the system. If a0 is only slightly dependent on Fand
V/ (which is typical of fast reactors), one says that there is a quasiasymptotic neutron-flux distribution.
One can estimate the limits of an asymptotic or quasiasymptotic distribution by numerical calculation
on non stationary processes; we restrict consideration to the multigroup one-dimensional nonstationary diffu-
sion equation. The numerical method of solving this equation with two or three groups has been described [1].
Here we consider a multigroup treatment with over 20 groups. Consider the equation
I ? div grad .7+ 27p :?Tp
a n
(1)
Translated from Atomnaya Energiya, Vol. 47, No. 1, pp. 51-52, July, 1979. Original article submitted
July 17, 1978.
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Alc
30
20
10
10
8
7
0 0 10 20
0,6 47 0,13 4g R/Rcr Time, 10-6 sec
Fig. 1 Fig. 2
Fig. 1. Dependence of a/a er on radius for the BFS-32 assembly: solidline - fromtheory, points -
from experiment.
Fig. 2. Relative derivative of the count rate as a function of time for an assembly containing natural
uranium dioxide (symbols as in Fig. 1).
30
subject to boundary and initial conditions of the form
=?2: -P = --/q7; (r, It=o= (1)0 (r),
an y
where G is the number of energy groups, q (r, t) {(p (1) (r, t), (p(2) (r, t), , (pG (r, t)}; S is a lower trian-
gular matrix of order G with nonnegative coefficients, T is the complete matrix of that type, and v-1, n,
and 1 are diagonal matrices. There is always an asymptotic solution to (1), but the asymptotic distribution
may occur after excessive time following the pulse, whereas we are interested in time intervals such that the
neutron flux is still above the limit of detection of the instrument or above the background.
We solve (1) by a difference method. The stationary-state approximation has been examined in some
detail [2], so we pass at once to the differential-difference equation
(2)
ao7Pth -AniDh+ th(p-h,
where Ah, vh, Th are difference approximations for the operators div 15 grad- E + S, v-?, T; then -(Ph (r,
(pha (r) exp (-aot) for large times. It is necessary to follow the behavior of the flux from the start of the pulse
to the steady-state value, so the difference scheme must be asymptotically stable [2]. Only a purely inexplicit
scheme is unconditionally asymptotically stable amongst the various two-layer schemes for a parabolic equa-
tion, so we apply the operator Ah to the upper t layer. The T matrix is complete, and this is more convenient
to apply to the lower layer. The error in the asymptotic or quasiasymptotic state is reduced by introducing
corrections in terms of solutions exponential in t. As a result
1x Rh (r, --ihs (r, t ?T)JPE= AnTlyr (r, l'hiCPht (r, t?r.), (3)
where T is the time step and
x = ?ch/Il ? exp (ccr)];
_ ( II (Thc (r, t ? 2T)
cz??(Pits (r, t ? T)
(4)
The solution to (3) is found for each value of t by one-dimensional fitting for each group in sequence beginning
with the first. The scheme of (3) has unconditional asymptotic stability and convergence, and calculations
show that the corrections of (4) greatly improve the accuracy. It has been shown [3] for some simple cases
(e.g., for Th = 0) that a ?aa, (ph, exp (at) ? (pha const for t Q0 for any r. In practice these conditions are
met for any Th even in the more general case where the corrections of (4) are introduced separately for each
group.
561
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The scheme of (3) was implemented as the NESTOR program (nonstationary one-dimensional calcula-
tion) in Algol for the M-220; the experimental results were made comparable with the calculations by the use
of data for assemblies of two types. The first type was composed of a series of single-zone multiplication
systems whose composition was close to the core composition used in fast power reactors (the BFS-32 assem-
bly [4]). A characteristic of such a system is the presence of a quasiasymptotic neutron-flux distribution.
Assemblies of the second type corresponded to continuous variation of the neutron spectrum in time. We
used an assembly containing natural uranium oxide, which is the blanket material used in fast reactors. The
methods of measurement anti the parameters of the apparatus corresponded with those of RI (Figs. 1 and 2).
In the first case, a was calculated for detectors differing in spectral sensitivity e (E); the decrement a
remained virtually constant after a transient period of about 1.5 ?sec for assemblies of radius from Rer down
to 0.7 Rel.. Smaller assemblies gave a that varied appreciably with time and depended on the type of detector.
Figure 1 shows data for a detector having e (E) = Z15(E), and in that case the relation a (t) =RAE, (P)] TP)/at
after the transient period showed a part with very little slope (plateau). The maximum and minimum values of
a given in Fig. 1 correspond to the initial and final instants as taken symmetrically about the middle of the
plateau for falls in neutron flux by a factor 102. In the second case, the calculations and experiments were per-
formed for a detector having e(E) = Ef8(E) and placed at the center of the assembly. A 26-group system of
BNAB constants was used.
The experiments and calculations confirmed that these approximations are applicable on calculations on
the prompt-neutron kinetics in assemblies (down to Kef =0.7 or less).
LITERATURE CITED
1. B. I. Kolosov, FEI-275 preprint, Obninsk (1971)?
2. A. A. Sarnarskii and A. V. Gulin, Stability in Different Schemes [in Russian], Nauka, Moscow (1973).
3. V. E. Kolesov and 0. I. Makarov, FEI-822 preprint, Obninsk (1978).
4. I. P. Matveenko et al., At. Energ. , 44, No. 3, 262 (1978).
EFFECTS OF ION-BOMBARDMENT DOSE AND
PREVIOUS SURFACE TREATMENT ON THE
EROSION OF MOLYBDENUM
B. A. Kahn, D. M. Skorov, UDC 539.12.04:621.039.616
and V. L. Yakushin
Blistering under ion bombardment for some considerable time was considered [1-2] as the most impor-
tant form of erosion of the first wall in a thermonuclear reactor [3, 4]. However, subsequent study showed
that blistering in fact falls at high radiation doses [5, 6].
When neobium was bombarded by helium ions of energy 15 keV at 2 mA/cm2, the blistering vanished at
doses of (6.25-18.75)?1018 ions/cm2 [5]. It was suggested that this was due to alteration in the distribution
of the implanted gas when the tops of the blisters lift and that there is also an effect from cathode sputtering
of the surface layer [5, 6]. The final topography is very much dependent on the irradiation conditions, in-
cluding target temperature. It is therefore of interest to examine the effects of dose and previous treatment
on the irradiation erosion of molybdenum, which is one of the promising materials for these discharge cham-
bers.
Vacuum-melted molybdenum of MChVP grade and TsM10VD alloy were polished mechanically before
irradiation, with final finishing with diamond paste type ASM 1/0. Some specimens of the T5M10VD alloy
were additionally polished electrolytically. The state of the treated surface was monitored with a 201 PS
profilograph.
The specimens were bombarded with helium ions of energy 20 keV at a beam cukrent density of 660 ?A/
cm2; the MChVB molybdenum was irradiated at 650?C to doses of (5-500)?1017 ions/cm2, while the T5M10VD
Translated from Atomnaya Energiya, Vol. 47, No. 1, pp. 53-54, July, 1979. Original article submitted
December 28, 1978.
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41g
g 417
g 0,15
415
409
4 8 12 16 20
B?10-7 ions/cm'
24
Fig. 1. Erosion coefficient of TsM10VD
alloy in relation to dose.
Fig. 2. Surfaces of specimens of TsM10VD alloy after dose of 1 -1018 ions/cm2 (a) and of
vacuum melted MChVP molybdenum after a dose of 5-1018 ions/cm2 (b).
was irradiated at 700?C with doses of (4.8-20)?1017 ions/cm2. See [71 for more details of the bombardment
conditions. The surface topography was examined with an EMV-100L electron microscope by means of car-
bon replicas shadowed with chromium and also with a Kwikscan-50 scanning microscope. The erosion coeffi-
cient was deduced from the electron micrographs via the geometrical dimensions of the tops of the blisters,
for which the numbers of atoms were calculated.
Figure 1 shows the observed dependence of the erosion coefficient for TsM10VD in relation to dose; at
first, the coefficient increases, but there is a peak in the range 1.4-1018 to 2-1018 ions/cm2. Figure 2a shows
a typical electron micrograph of TsM10VD after a dose of 1-1018 ions/cm2. The surface has been substantially
eroded, and most of the blisters have burst, with many of the tops leaving the surface of the material.
It was found that mechanical polishing greatly reduced the tendency of the material to produce blisters
that fail, especially by comparison with the eldctropolished material, at all doses in the range from 5-1017 to
5.1019 ions/cm2. For example, a mechanically polished surface showed no broken blisters, whereas electro-
polished specimens under analogous bombardment conditions showed considerable erosion arising from bro-
ken blisters (Fig. 2a). The suppression of blistering appears to be related to microscopic roughness of height
0.04-0.18 pm produced by mechanical polishing. This roughness prevents the gas from forming coplanar
layers and thus hinders the union of helium bubbles into gas cavities, which represent the first stage of blis-
tering. Electropolishing also hydrogenates and embrittles the surface layers of molybdenum, and this appears
to favor erosion under irradiation.
Doses between 5-1017 and 2?1018 ions/cm2 to MChVP molybdenum produced blisters; the blisters vanished
on raising the dose to 5-1018 ions/cm2, but the surface of the material was etched, and cones of height about
0.7 pm were produced on further increase in the dose (Fig. 2b). Blistering is therefore a transient effect.
LITERATURE CITED
1. W. Primak, Y. Dayal, and E. Edwards, J. Appl. Phys., 34, No. 4, 827 (1963).
2. W. Primak and J. Luthraf, Ibid., 37, No. 6, 2287 (1966).
3. R. Behrisch,Nucl. Fusion, 12, No. 2, 695 (1972).
4. E. Kemp, ibid., 14, No. 2, 277 (1974).
5. J. Martel and S. Jacques, in: Proceedings of the Conference on Surface Effects in Controlled Thermo-
nuclear Fusion Devices and Reactors, Argonne, Jan. 10-15, 1974.
563
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6. J. Roth and R. Behrisch, J. Nucl. Mater., 57, No. 3, 365 (1975).
7. D. M. Skorov et al., in: Proceedings of the All-Union Conference on Engineering Problems of Thermo-
nuclear Reactors [in Russian], Vol. 3, NIAFA, Leningrad (1977), P. 226.
A CADMIUM-SULFIDE y-RAY DOSIMETER WITH
ELEVATED STABILITY UNDER IRRADIATION
V. K. Dubovoi UDC 53.07/08+53.001.89
A major disadvantage of semiconductor y-ray dosimeters is that the irradiation resistance is poor,
which hinders their use in high-power irradiation systems, where the exposure dose may be many kR/sec.
In the present dosimeter, high radiation resistance is provided by the use of CdS crystals exposed to
fast reactor neutrons. The radiation defects make these crystals homogeneous in bulk properties, while
they have good sensitivity to y radiation and a specific resistance of 106-107 ohm ? cm, and the voltage-current
response is linear up to fields of 300 V/cm. Ohmic contacts were formed with indium. The dimensions of the
detectors were 0.5 x 0.1 x 1 cm, so they-ray distribution at the point of measurement is altered only slight-
ly. The current Ly is measured with a constant voltage applied to the detector. Shunting by the ionized air is
eliminated by sealing the transducer and fitting special microconnectors that eliminate current liakage. Fig-
ure 1 shows the dosimeter characteristic, where ly is proportional to the square root of the exposure dose.
There is an increase in Ly with temperature, and therefore in the sensitivity of the y-ray detector. Measure-
ments were made of the current at a fixed exposure-dose rate of 50 R/sec for various photon energies Ev; it
was found that the CdS dosimeters had only a slight hardness trend for y radiation. We used 60Co (Ey =1.17
and 1.33 MeV) and 137Cs (0.66 MeV) sources together with the y rays from a reactor, which are strongest at
Ey = 0.3 MeV [1]. The radiation resistance in a y-ray field was determined from the Ly curves in relation to
"Co y-ray dose at a dose rate of 6000 R/sec, the result being the value of -109 ,(Fig. 2).
3
2_,/ I
10 20 30 40 50 60 70 VP, sec /0 7 10 '
8 10 9 DA R
Fig. 1 Fig. 2
Fig. 1. Relation of ly to exposure dose for 60Co y rays at U = 5 V and T = 30?C.
Fig. 2. Dependence of 17 on "Co exposure dose at U = 5 V, T = 30?C, and Py =6000 R/sec.
LITERATURE CITED
1. V. M. Kolyada and V. S. Karasev, At. Energ. , 19, No. 6, 532 (1965).
Translated from Atomnaya Energiya, Vol. 47, No. 1, p. 54, July, 1979. Original article submitted
August 30, 1978.
564 0038-531X/79/4701-0564$07.50 1980 Plenum Publishing Corporation
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THE THERMAL-NEUTRON FISSION CROSS SECTION AND
THE FISSION-RESONANCE INTEGRAL FOR 2 4 3 CM
K. D. Zhuravlev and N. I. Kroshkin UDC 539.172.4
Thermal-neutron fission cross sections of and the resonant-fission integral If for 243Cm have been mea-
sured in determining nuclear constants for the actinides by the cadmium difference method; the standard was
a 235U target, for which the thermal-neutron fission cross section was taken as 582.2 ?1.3 barn, while the
resonant-fission integral was taken as 275 ?5 barn [1]. Measurements were made in a horizontal channel in
the SM-2 high-flux reactor. The cadmium ratio for 235U was 40 when the neutron beam was filtered with lmm
of cadmium.
The uranium and curium targets were deposited on aluminum substrates of thickness 0.1 mm. The
amount of substance in the target was about 1 ptg in each case, while the diameter of the target spot was about
4m
10 mm. The curium target had the following percent isotope composition; 242cm 0.la ? 0.07, 243cm 37.29 ?
1.34, 244Cm 61.61 ?1.34, 245Cm 0.94 ?0.10. The number of nuclei in the curium target was determined from
the spontaneous fission rates of 244Cm and 242CM. The spontaneous-fission half-life of 244CM was taken as
(1.270 ?0.007)407 yr, andthat for 242Cm was taken as (6.09 ?0.18)? 106 yr[2]. The number of nuclei in the
uranium target was determined by a counting in 2n geometry. The isotope composition of the uranium target
and the method of use have been described in detail [3, 41.
The measurements led to correction of of and If for 243Cm (K2014 = 0.998, K244 = 0.987 and K2,745 = 0.937,
K245 = 0.987), on account of the contents of 244Cm and 245CM, which relate respecTlively to the thermal-neutron
fission cross section and the resonant-fission integral.
The temperature of the neutrons at the reactor exit having a Maxwellian spectrum was 353?K [4], and
the g(T) correction incorporated the fact that the fission cross section of uranium does not follow VIM so the
value was 0.965 [5]. It was assumed for curium that the fission cross section follows 1/1/T and g(T) =1.
Table 1 gives the fission cross section and resonance integral, which are in agreement within the error
of measurement. The spread in the values is probably due to the lack of data on the behavior of the fission
cross section of 243Cm in the thermal range.
TABLE 1. Thermal-Neutron Fission Cross
Section and Resonant Fission Integral for 243Cm
in barn
Parameter Our results
Data of [1]
Data of [6]
cri
672+60
600+50
609,6?25,9
if
1480+150
1860+400
1575+136
LITERATURE CITED
1. Neutron Cross Sections, BNL-325, Third Edition, Vol. 1 (1973).
2. Yu. S. Zamyatnin, in; Nuclear Constants [in Russian], No. 14, Atomizdat, Moscow (1974), p. 22.
3. K. D. Zhuravlev, N. I. Kroshkin, and A. P. Chetverikov, At. Energ., 39, No. 4, 285 (1975).
4. K. D. Zhuravlev and N. I. Kroshkin, in: Atomic Science and Engineering, Nuclear Constants Series
[in Russian], No. 19, Atomizdat, Moscow (1975), p. 3.
5. P. E. igeltstaff, in: Handbook on Nuclear Physics [in Russian], Fizmatgiz, Moscow (1963), p. 268.
6. C. Bemis et al., Sci. Eng., 63, 413 (1977).
Translated from Atomnaya Energiya, Vol. 47, No. 1, p. 55, July, 1979. Original article submitted
August 7, 1978.
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ABSOLUTE MEASUREMENT OF THE BRANCHING
RATIO FOR THE 277.6-keV LINE OF 2 3 9Np
V. K. Mpzhaev, V. A. Dulin, UDC 661.039.51
and Yu. A. Kazanskii
It is common to measure the intensity of the 239Np y rays in order to determine the accumulation of
239Pu in breeder reactors; in some cases it is necessary to know the branching ratio of the 277.6-keV y-ray
line in the decay of 239Np. Also, it has been suggested that the set of standard calibrated radioactive prepa-
rations (from the IAEA) should contain 243AM instead of 293Hg. In that case it is also necessary to know the
absolute yield of the y line of 239Np, as this is formed by the a decay of 243Am.
We have measured the absolute intensity of the y line of 239Np (Ey = 277.6 keV)by means of an a particle
calibrated 243Am source and the 293Hg source from the OSGI set of standard radioactive sources. The 243Am
was calibrated in precisely known geometry from the a-particle counting rate provided by a silicon detector
[1]?
The y rays from the 243Ain and 293Hg preparations were recorded with a germanium detector having a
sensitive volume of about 35 cm3.
The absolute yield of the y line of 239Np (JyNp) was determined from
NyNpQHgEviig
YNP NvHgQAm6VND
where NyNp and Nytig are the count rates produced by y rays of energy 277.6 for 243AM and 279.1 keV for
293Hg; QAm and Qllg are activities of those isotopes; el,Np, eyHe are the efficiencies for y rays of energies
277.6 and 279.1 key, respectively, in the germanium defector; and JyHg is the branching ratio for the y line
(Ey = 291.1 keV) in the decay of 293Hg (0.8155 ? 0.0015 [21).
The ratio of the y-ray efficiencies (e.yrig/eyNp) is dependent on the slope of the efficiency curve for y
rays in the total-absorption peak. The trend with energy is governed by the size of the sensitive volume and
the counting geometry. The ratio of the efficiencies for y rays of energies of 279.1 and 277.6 keV was mea-
sured with the OSGI set as 0.995 ?0.004.
We give below the components of the error that determine the overall error in measuring the branching
ratio of the y line of 239Np having Ey = 277.6 keV, %;
Statistical error
Activity of 293Hg [3]
Activity of 243Am [1]
of 20311g
Ratio of efficiencies for y rays and 243Am.
Self-absorption of the y.rays in 243Am
Branching ratio for y line of 293Hg (E =279.1 keV)
Total error of measurement for the branching
ratio for the y line of 239Np (Ey = 277.6 keV)
0.40
1.50
0.33
0.40
0.001
0.20
1.65
The overall error of measurement is clearly governed by the error in the calibrated 293Hg source from
the OSGI set.
We give below the branching ratios for the y line of 239Np with Ey = 277.6 keV as measured here and
quoted elsewhere:
Translated from Atomnaya Energiya, Vol. 47, No. 1, pp. 55-56, July, 1979. Original article sub-
mitted September 1, 1978.
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f4]
0.145
?0.004
[2]
0.1450
?0.0015
[5]
0.141
?0.004
[6]
0.152
?1.005
Our result
0.1430
?0.0024
We are indebted to S. E. Lavrov and V. I. Ivanov for preparing the 243Am source and determining the
isotope composition.
LITERATURE CITED
1. V. A. Dulin and V. K. Mozhaev, At. Energ., 44, No. 6, 528 (1978).
2. C. Bowman, Reports US Nuclear Data Committee, USNDC-9 (1973), p. 225.
3. Certificate Nos. 227 and 228, OSGI (1975). ?
4. J. Ahmad and M. Walalgren, Nucl. Lnstrum. Methods, 99, No. 2, 333 (1972).
5. L. N. Yurova et al., At. Energ., 36, No. 1, 51 (1974).
6. C. Bowman, Reports US Nuclear Data Committee, USNDC-11 (1974), p. 274.
CALCULATION OF THE TRUE VOLUME PROPORTION
OF STEAM IN THE DRIVING SECTION OF A NATURAL-
CIRCULATION LOOP
L. N. Polyanin and A. L. Putov UDC 621.039.52,44
Here we calculate the true volume proportion of steam in the driving section of a natural-convection
loop operating at low pressures.
The true volume content of steam cp is related to the mass flow X and the steam breakthrough factor w
(p=1/(1 (,) ? p" ? 1 X -- X
p' )
(1)
where (3' and p" are the densities of water and steam, respectively. The coefficient co = Wv/Ww
appearing in (1) varies over the height of the driving section, where Ww and Wv are the true speeds of the wa-
ter and steam, on account of increase in the group rate of rise of the steam bubbles Wffs as the steam content
increases.
By definition
Wv war w gy
= + !is 1 ? (I) ris
Ww Ww ' 1?x wsls '
(2)
where Wods =G/piSds is the circulation rate, Scis is the area of cross section of the driving section, and G is
the mass flow rate upon the heat carrier in the driving section.
Measurements [1, 2] for (i9 :5_ 0.5 show that Wgrrs is related to the rate of rise of a single bubble W5 by
(3)
We substitute (3) into (2) to get
WgrW?. ?(p)
is ns
= (1+ Kw) (1? P"
)q
/[1?(1?Kw ?r)
P
(4)
where Kw = wksiwods.
Translated from Atomnaya Energiya, Vol. 47, No. 1, pp. 56-57, July, 1979. Original article submitted
September 1, 1978; revision submitted November 10, 1978.
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567
where
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co(40(z)
0,35
vo
425
0,2o
qi5
?
2
4
8
m
Fig. 1. Height variation of co (z)
calculated from (15) and the mean
value Tp (z) for pressures (kgf/cm2)
of: 1) 12; 2) 16; 3) 25; for X0 = 0.01.
The variation of X overthe height z of the driving section is as follows for quasiequilibrium flow conditions:
dX 1 di' 1 dp
dz ? r dz ? r dp dz'
dp
dz = ?g (1 --(p) +-rcp],
(5)
(6)
because the pressure loss due to4riction and acceleration of the flow can be neglected under these conditions.
Here it is the enthalpy of the water on the saturation line, p is pressure, r is the latent heat of vaporization,
and g is acceleration due to gravity. As dit/dp is comparatively small in the driving section, it can be taken
as constant.
where
From (1) and (4)-(6) we get an equation for the true volume steam content:
dz p'
Dr----rp"jg
(7).
(9)
The effect from boiling in the carrier (which causes co to rise in the driving section) is important only in
the region of comparatively low pressures, where p"63' ? 1, andthen (7) has the solution
=-1 17 1 -1- 2z
/ ? wo)2 D(4+ Kw)
The value cp at the inlet to the driving section (z = 0) is given by
o 1?X0 P"
p
To=1/[1+(l+Kw) ? ' X0 +Kw
P"
which is equivalent [1, 3] to the generally accepted formula
'Po = WO.I(WO+ W+ a),
(9)
(10)
where W0t and W0" are the reduced velocities of the water and steam respectively, while a is the character-
istic velocity of single-bubble rise \Vris appearing in (3). It is recommended [1, 4] that this speed should
be derived from
568
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a=
{(0,65-0,0039p)63-- , d ds 200 mm,
(12)
where dds is the diameter ofthe driving section in mm and p is pressure in kgf/cm2 (11 :s p Is 125 kgf/cm2).
It has been found by experiment [5] that the deformation of the steam-content curve at the inlet to the
driving section has virtually no effect on the mean value and therefore on the flow rate of the carrier in
this natural-circulation loop. Measurements on the VK-50 boiling-water reactor [6] provided an analogous
conclusion, namely that the total flow rate in the natural-circulation loop is independent of the energy genera-
tion in the individual cassettes in the core. Therefore, (p0 can be calculated from (10) with X0 taken as the
mean mass steam content over the section at the outlet from the core.
Figure 1 illustrates the solution by reference to theoretical cp (z, p) and p (z, p) = f co (zt, p)dzt for
the case X0 = 0.01 and Kw =2; clearly, the change in the true volume content of steam oveP the height aris-
ing from the boiling in the carrier is considerable at low pressures in an extended driving section, and this
should therefore be incorporated into the calculations on the driving head and the steam available in the loop.
LITERATURE CITED
1. A. I. Filimonov et al., Teploenergetika., No. 10, 22 (1957).
2. G. B. Wallis, One-Dimensional Two-Phase Flow, McGraw-Hill (1969).
3. A. Ya. Kramerov and Ya. V. Shevelev, Engineering Calculations on Nuclear Reactors [in Russian],
Atomizdat, Moscow (1964).
4. G. G. Bartolomei, V. A. Suvorov, and S. A. Tevlin, in: Water Preparation and Processes in Boilers
[in Russian], No. 1, Gosenergoizdat, Moscow (1963).
5. I. S. Dubrovskii, Teploenergetika, No. 2, 31 (1974).
6. A. P. Sarygin etal.,At. Energ. , 30, No. 4, 350 (1971).
A LOCAL APPROACH TO DETERMINATION
OF THE COORDINATES OF AN INTERFACE
F. L. Gerchikov and V. D. Kosarev UDC 539.171:539.12
X rays are used to considerable advantage at the present time in automatic-control systems [1]; this
requires an optimum apparatus design and also definition of the working conditions in the radiation fields,
so it is necessary to know the parameters of the direct beam and of the scattered radiation, including the
fields arising from backscattering at interfaces with air.
Much of the information on scattered-radiation characteristics for interfaces has been derived either by
Monte Carlo techniques or by integrating the distributions for point unidirectional sources over the surface of
the scatterer or over the volume of the air [2-4]. However, the methods and apparatus available for integral
evaluation of the scattered radiation and for determining the coordinates of the interface when one of the media
is air are subject to substantial limitations on account of the marked effect of the component scattered by the
air [2, 3]. Measurements show that the backscattered flux decreases away from the scatterer, whereas the
flux scattered by the air increases. These two fluxes become comparable at a certain distance. The point
where the fluxes are equal is the critical value licr as regards determination of the upper limit to the boundary
position.
It has been pointed out [2] that the y rays from 137Cs show a detectable effect in the air at a distance of
10 m, while Rcr = 22 m; our measurements have also shown that Rel. =12-15 m for x rays of effective quan-
tum energy 60 keV [4].
Translated from Atomnaya Energiya, Vol. 47, No. 1, pp. 57-58, July, 1979. Original article submitted
December 4, 1978.
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4i,
Interface
Fig. 1. Working geometry.
Therefore, on the whole, the integral approach to the identification of interfaces and the parameters
of the backscattering pattern in the presence of air show marked effects from the distance to the interface,
and these effects are extremely important in determining the upper limit to the coordinates of the interface.
Consider an air medium at a distance R from a scatterer exposed to point isotropic pulsed x-ray source
RS which is combined with a total-absorption detector in conditions of good geometry, in which an energy flux
E0 is emitted into an angle (Fig. 1). We estimate the air component of the scattering on the assumption
that the interface is a hemisphere at the center of which lie the detector and source, which are at a distance
R1 in vacuum from the hemisphere. The backscattered energy flux is [2] as follows in the air?vacuum geom-
etry:
Ea (E1)= E0Ae111012nRf,
where Ae is the energy albedo.
We increase R1 by a certain amount AR and then have by analogy with (1) that
Ea (R1-1- AR)= E0AeV,/2n (R14-AR)'.
The energy flux backscattered by the air layer at a distance AR from the detector is
Ea(AR)= Ea(Ri+ AR) ?4(E1)?
If a layer AR of air is placed at a distance R from the source, a survey [2] indicates that the backscattered
energy flux arising from the localized area of air as corrected for attenuation is
Ea (AR) = E 0AeY 0 exp (-2?0R)[1?exp (? 2ttoAR)] (2nR2)-i.
(1)
(2)
(3)
The density of the energy flux backscattered by the material is affected by the attenuation of the primary
beam tto and that of the scattered beam j., the result being [2]
Er= exp [ ? (tto ?I) (2nR2)-2 (4)
We use (3) and (4) with suitable steps to get
E a (A R) exp (? 2itoAR)
Er exp ? (pi ?go) El
As ?0AR ?1 and (iii?tio)R ?1 in (5) for soft x rays, we put
Ea (AR)2p0AR
Er
We determine Rer by equating (6) to 1 and allow AR to tend to zero (the localized air liolume is minimal).
Then we have
570
(5)
(6)
Rcr=(11?t1o)-1. (7)
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Comparison of (7) withthe analogous expressionRcr= (3p0+ ?1)-1 of [2] shows that local estimates of the
backscattered radiation and determination of the position of the interface when one of the media is air can
allow one to identify the backscattering from the air and the material over a range of coordinates of the in-
terface broaderthanthat for the integral method. 'Engineering implementation of the local-evaluation method
can be provided by dynamic strobing of the detector; a narrow strobe it (SR) is used with a synchronized
source to scan localized volumes of air in accordance with a set program, as well as the surface of the scat-
terer in order to identify the quanta scattered by the material and the air.
In experiments, At =10 nsec (AR =1.5 m), and the radiation sources were ones developed in ?col-
laboration with I. P. Karpinskii et al. [5] and Kalinin Polytechnical Institute in Leningrad, which have been
used in determining the coordinates of air-concrete interfaces for R ?80 m. The detector was a plastic
scintillator (polystyrene containing P-terphenyl and POPOP) working with an FEU-87 [6].
LITERATURE CITED
1. Apparatus and Methods for X-Ray Analysis [in Russian], Nos.19-21, Mashinostroenie, Leningrad (1978).
2. B. P. Bulatov and N. F. Andryushin, Backscattered Gamma Radiation in Radiation Engineering [in
Russian], Atomizdat, Moscow (1971), p. 201.
3. Radiation Technology, Issue 3, Papers from the All-Union Radiation Engineering Research Institute
fin Russian], Atomizdat, Moscow (1969), p. 117.
4. F. L. Gerchikov, At. Energ., 40, No. 6, 487 (1976); 41, No. 6, 414 (1976).
5. I. D. Ivanov et al., in: Nuclear Instrumentation [in Russian], Nos. 34-35, Atomizdat, Moscow (1977),
p. 126.
.6. -A. P. Karmanova and E. E. Stepanov, in: Papers from the All-Union Instrumentation Research Insti-
tute [in Russian],- Nos. 30-31 (1976), p. 69.
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ANNIVERSARIES
ACADEMICIAN PAVEL ALEKSEEVICH CHERENKOV
(ON HIS 75th BIRTHDAY'ANNIVERSARY)
E. I. Tamm and B. B. Govorkov
e
? 11,11.11.;,..
' 1 ?
On July 28, 1979,Pavel Alekseevich Cherenkov, the eminent physicist-experimenter, whose name is
associated with one of the most outstanding scientific discoveries of our time, was 75 years old. He greeted
his anniversary full of creative plans, and his energy and enthusiasm areto be enviedbymany young scientists.
During recent years, by the initiative of Pavel Alekseevich, many new, interesting, and promising investiga-
tions have been initiated, covering broad problems of contemporary physics. These include questions of the
structure of elementary particles and the investigation of new nuclear structures (hypernuclei) and, at first
sight far removed from these questions, problems of the generation and utilization in different fields of sci-
ence and technology of new forms of electromagnetic radiation, like the so-called synchrotron and undulatory
radiation. This is far from a complete list of the problems, on the solution of which P. A. Cherenkov is now
working, as Director of one of the most prominent laboratories of the P. N. Lebedev Institute of Physics of
the Academy of Sciences of the USSR. The creative life of Pavel Alekseevich is associated with this Institute.
In 1932 he became a graduate of the Institute of Physics of the Academy of Sciences (FIAN) and, on the initi-
ative of S. I. Vavilov, he started to investigate the luminescence of solutions of uranium salts under the action
of y radiation. These investigations also led Pavel Alekseevich to the discovery of a new, remarkably beauti-
ful physical phenomenon. He discovered that y radiation creates a weak luminescence of the solution, which
differs sharply from normal luminescence. In extremely time-consuming experiments, in which the method
of photometry at the threshold of vision was used, Pavel Alekseevich studied certain characteristic features
of the emission discovered by him. In the investigations the main traits of P. A. Cherenkov's nature were
displayed ? enthusiasm, unusual tenacity in achieving the designated goal, ability to find the simplest solution
to the problems arising, and the knack of paying attention to the "trifles" of the experiment.
After 2 years it already had become clear that the fluorescence discovered had no relation with lumines-
cence, its polarization was measured, and an increase of energy in the emission spectrum with a reduction of
Translated from Atomnaya Energiya., Vol. 47, No. 1, pp. 59-60, July, 1979.
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the wavelength of the primary Y quanta was discovered. This allowed S. I. Vavilov in 1934 to arrive at the
conclusion that the new form of luminescence was related with the electrons formed in the solutions by the
Compton scattering of,y radiation. For the next few years, Pavel Alekseevich studied experimentally all
the fundamental characteristics of the new physical phenomenon, the nature of which was realized in 1936,
after its theory was developed by I. E. Tamm and I. M. Frank. They showed that the luminescence discovered
by P. A. Cherenkov is the emission of a charged particle, moving with a velocity exceeding the velocity of
light in the substance. In 1936-1937, P. A. Cherenkov verified quantitatively the Tamm?Frank theory by
measuring the characteristic angle of emission and its dependence on the refractive index of the medium,
established the energy distribution in the emission spectrum, and the absolute brightness of the luminescence.
These investigations brought P. A. Cherenkov world recognition. G. S. Landsberg called them the "decoration
of Soviet physics." The new form of radiation is known now as "Cherenkov radiation."
In 1946, after this work, P. A. Cherenkov, 5.1. Vavilov, E. I. Tamm, and I. M. Frank were honored with
the State Prize of the Soviet Union.
The discovery by P. A. Cherenkov, in addition to the enormous scientific interest, has a great practical
importance. Thus, in high-energy physics the most outstanding experimental research conducted over the
period of the last decades has proved to be possible only because of the use of methods for recording par-
ticles based on the application of Cherenkov radiation, or as it is now customary to say, Cherenkov counters.
Threshold and differential (gas) Cherenkov counters, shower Cherenkov spectrometers, various Cherenkov
chambers ? all these are the instruments without which it is impossible now to describe the experimental phys-
ics of elementary particles. The use of Cherenkov detectors in physics, astronomy, and other fresh fields of
science has reached such scales that, without fear of being mistaken, it may be asserted that P. A. Cherenkov
is now one of the world's most well-known physicists.
In 1958 P. A. Cherenkov, L E. Tamm, and I. M. Frank were awarded the Nobel Prize for Physics "for the
discovery and explanation of the Cherenkov effect."
In the years of the Great Patriotic War, Pavel Alekseevich was occupied with the development of an instru-
ment of military designation, based on the use of certain methods of nuclear physics. Since 1946, Pavel Alek-
seevich participated in the development and construction of the first electron accelerators in the laboratory
directed by V. I; Veksler. After his participation in work on the construction of a 250-MeV electron synchro-
tron, Pavel Alekseevich, together with the team of authors, was awarded the State Prize of the Soviet Union.
From 1959 Pavel Alekseevich directed the Laboratory of Photomeson Processes of FIAN, investigating the
electromagnetic interactions of elementary particles. For a start, Pavel Alekseevich carried out fundamen-
tal research of photon ? nucleoli interactions, in particular the photoscattering of the lightest nuclei at an
energy of 250 MeV. After this work, -in 1977 P. A. Cherenkov with co-authors was honored for the third time
with the State Prize of the Soviet Union.
Pavel Alekseevich headed the work on the creation of the FIAN scientific group for the investigation of
electromagnetic interactions, including the high-intensity "Pakhra" accelerator, operating at 1.3 GeV, and
the modern measurement?recording center. Without waiting for the completion of construction of this com-
plex, P. A. Cherenkov in the laboratory started to work on a study of high-energy electromagnetic processes,
creating jointlywith the Institute of High-Energy Physics (IFVE) and the Erevan Institute of Physics (ErFI) the
electron beam on the Serpukhov proton accelerator.
P. A. Cherenkov was one of the first physicists who paid attention to the abundant possibilities opened up
by the use of synchrotron radiation. Even in the 1950s synchrotron radiation was the principal proposed and appliectt
method, -proposed by him together with Yu. M. Ado, for the storage of electrons and for the production of
colliding electron?positron beams in synchrotrons. Subsequently, Pavel Alekseevich and authors addressed
a paper in which the possibilities were considered of using synchrotron radiation in science and technology.
Now, synchrotron radiation is being widely used everywhere where there are electron synchrotrons and
storage devices, and in some countries sources of synchrotron radiation have been specially constructed.
P. A. Cherenkov, with young colleagues, has done much for the theoretical and experimental study of
undulatory radiation on the tmdulator, built by them for the first time and operating in the rectilinear gap
of the "Pakhra" synchrotron.
Pavel Alekseevich Cherenkov is widely known not only to the world scientific community. He devotes much
effort and energy to the struggle for peace and to the expansion of scientific relations between the scientists
of different countries. For many years he has been a Member of the Presidium of the Soviet Committee for
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World Protection, a Member of the Soviet Committee for European Safety and Collaboration, and a partici-
pant of the Pugwash movement of scientists.
Pavel Alekseevich Cherenkov has been rewarded with two Orders of Lenin. two Orders of the Red Ban-
ner, the Order of the "Badge of Honor," and medals of the Soviet Union. The high merits of P. A. Cherenkov
have been marked also with orders of foreign countries.
The friends and colleagues of Pavel Alekseevich warmly wish him good health, success, and new crea-
tive happiness.
THE 50th BIRTHDAY ANNIVERSARY OF
EVGENII VLADIMIROVICH KULOV
The editorial staff of the Journal heartily congratulate Evgenii Vladimirovich Kulov, Member of the
Editorial Board of the Head of the Central Administration of the State Committee for the Utilization of Atomic
Energy, on his 50th birthday, and wish him good health and new creative successes.
Translated from Atomnaya Energiya, Vol. 47, No. 1, p. 60, July, 1979.
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INFORMATION
THE ACCIDENT AT THE THREE MILE
ISLAND NUCLEAR POWER PLANT
As is known, on March 28, 1979, there was an accident at the Three Mile Island nuclear power plant
which attracted the close attention of the world for several days. At that power station there is a Babcock
and Wilcox pressurized water reactor with a power of 956 MW (gross). The reactor was first brought to
criticality on March 9, 1978, and produced its first electricity on March 15, and was in commercial opera-
tion since December 30.
There were many sensational and contradictory reports about the accident. In some of them the danger
of the accident was unjustifiably exaggerated. Recently publications have appeared which are used as the basis
of this report.* It makes it possible to follow the causes, development, and resolution of the accident and to
evaluate its consequences.
Chronology of the Development and Resolution of the Accident
0 sec (4:00:0 local time). At a power of 98% of nominal, all three feed pumps shut off. The cause of
the shutting off was said to be either a cutting off of the supply of condensate to the condensate cleaning sys-
tem due to a fault in a valve on a connecting pipe or an obstruction in the ion exchange filters. Similar shut-
offs of the feed pumps had been noticed previously (at least eight times since October, 1978). When the main
feed water system fails the emergency supply system is supposed to be turned on automatically. However,
this time, although the emergency feed pumps were turned on, water did not come into the loop since the gates
in this system had been left closed after earlier teats. As a result, the steam generators ceased to be supplied
with feed water which caused the turbines to be shut off. According to a statement by the NRC, operation of a
nuclear power plant with closed gates on the emergency feed water supply system is forbidden.
3-6 sec. Because heat removal had ceased the temperature and pressure in the reactor began to rise.
At a pressure of 158 kg/ cm2 in the primary loop the relief valve 7 (see Fig. 1) on the volume compensator 8
operated and part of the coolant went into the overflow tank 3.
9-12 sec. When the pressure in the loop reached 166 kg/ cm2 the reactor was automatically stopped.
15 sec. The pressure in the loop fell to 155 kg/ cm2 and reached roughly an atmosphere below the value
at which valve 7 should close; however, this did not occur. Evidently, the operators did not know that the
valve had not closed since at that time the apparatus which indicated the water level in the volume compensator
was out of order and gave high readings (the apparatus went off scale).
30 sec. Sensors indicated that the emergency feed water supply pumps were working but fruitlessly
since, as noted, the gates in this system were closed.
60 sec. The water level in the volume compensator 8 rose sharply and reached a minimum in both
steam generators.
2 min. When the pressure in the loop fell to 112 kg/ cm2 the emergency core cooling system (ECCS)
turned on automatically, sending water directly into the primary loop. It is intended to operate as well if the
water level in the volume compensator is reduced or the pressure in the containment vessel rises.
3 min. The pressure in the overflow tank 3 reached 7 kg/ cm'.
4 min 30 sec. Since the ECCS supplied more water than had flowed out of the volume compensator, the
operator manually turned off one of the pumps supplying water to this system in order to avoid completely fil-
ling the volume compensator, at which time it would become impossible (under normal conditions) to regulate
the pressure in the primary loop.
*This report is compiled from material published in: Nuclear Engineering International, 24, No. 285, p. 10,
(1979); RUN-Actualites, No. 2, p. 190 (1979); Nuclear News, Special Report, April 6, 1979; Atomwirtschaft,
24, No. 5, p. 222 (1979); and SVA-Bulletin, No. 7, p. 11 (1979).
Translated from Atomnaya Energiya, Vol. 47, No. 1, pp. 61-63, July, 1979.
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121
Fig. 1. Schematic longitudinal cross section of the Three
Mile Island nuclear power station: 1) reactor; 2) waste wa-
ter (sump) pump; 3) overflow tank; 4) relief diaphragm; 5)
shutoff valve; 6) containment vessel; 7) relief valve; 8) vol-
ume compensator (surge chamber); 9) steam generator; 10)
valve for release of steam to the atmosphere; 11) main
steam pipe; 12) auxiliary building; 13) waste (drainage) wa-
ter; 14) emergency feed (makeup) water system pump; 15)
feed water pipe; 16) machine room; 17) turbine; 18) conden-
ser; 19) ejector; 20) feed pump; 21) pipe from emergency
feed water supply.
6 min. The pressure at the output of the reactor fell to 94 kg/cm2, that is, to saturation at a temper-
ature of 307?C.
, 7 min 30 sec. When the pressure in the overflow tank rose to 10.5 kg/cm2, in accordance with the
design a system was automatically turned on to pump water from this tank to a reservoir in the auxiliary
building 12.
8 min. The gates in the emergency feed water supply system were opened manually. The thermal
shock of the cold water entering the almost dry steam generators resulted in the rupture of the seal on one
of them. Next the damaged steam generator was shut off. Starting the steam generators sped up the drop
in pressure in the primary loop.
10 min. The second pump which supplies water to the ECCS Was turned off manually.
11-12 min. For the first time since 7 minutes after the accident began, the readings of the water level
in the volume compensator indicated a level in the operating range, evidently due to turning off the last ECCS
pump.
15 min. Despite the fact that water was pumped out of the overflow tank, the pressure in it rose to
13.3 kg/cm2, the relief diaphragm on the tank ruptured, and about 40 tons of water poured out onto the floor
of the reactor building.
20-60 min. Attempts were made to stabilize the reactor at a pressure of 71 kg/cm2 and a temperature
of 288?C. A steam bubble began to form in the upper part of the reactor.
1 h 15 min. Both circulation pumps in one loop of the reactor were turned off because noise and vibra-
tion in the pumps created a fear that cavitation would set in.
1 h. 40 min. Both circulation pumps in the other loop were turned off.
1 h 45 min-2 h. The water temperature at the inlet to the reactor was 65?C and at the outlet 327?C.
This jump indicated the absence of natural circulation of water and heat removal through the steam generators.
The steam bubble expanded and covered the top of the core. The fuel elements were thus damaged and part
of the fission products fell into the coolant. Because of the reaction of the zirconium in the jackets on the
fuel elements with steam, hydrogen formed which collected in the upper part of the reactor vessel. However,
because there was no operational method of analysis, the operators did not suspect its existence in the steam
bubble at this time.
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2 h 18 min. The volume compensator was closed by the shut off valve 5 (evidently, up to this time the oper-
ators did not know that valve 7 had been jammed in an open position). However, the pressure in the reactor
continued to fall and the steam-hydrogen bubble continued to expand.
2 h 30 min. The pressure in the reactor fell to 49 kg/cm2.
3 h 00 min. The pressure in the reactor rose to 148 kg/cm2.
3 h 15 min. A pressure rise of 0.35 kg/cm2 was noticed in the overflow tank 3.
3 h 48 min. A pressure rise of 0.77 kg/cm2 was noticed in the overflow tank. Both these rises could
have been caused by escape of steam since the pressure in the reactor again fell to 123 kg/cm2 (evidently, in
the preceding 1-2 h the operators, trying to stabilize the reactor, opened and closed the volume compensator
and turned the ECCS off and on). The excess pressure in the containment vessel rose from 0.07 to 0.21 kg/
C2
111 .
5 h 00 min. The pressure in the containment vessel reached 0.32 kg/cm2. It was isolated at the design
pressure. of 0.28 kg/ cm2 with the waste water pipe being closed. Thus, direct escape of radioactivity from
the vessel was stopped. Nevertheless, it took place from the auxiliary building to which some of the water
from the overflow tank had been pumped previously when it had fallen on the floor because of overflow or
breakage of the seal in that loop. Then krypton and xenon entered the atmosphere but iodine was mainly re-
tained in the ventilation system. Somewhat earlier there was an escape of radioactivity during a release of
steam, which had been slightly contaminated during leaks in the steam generators, through valve 10.
7 h 30 min. In more than 2 h the pressure changed from 87 to 147 kg/cm2 with a large jump in the
temperature between the inlet and outlet of the reactor. In order to reduce this jump, as is necessary for
cooling the reactor with a low-pressure system, the operator opened valve 5. As a result there was a further
discharge of coolant.
8-9 h. The pressure in the reactor fell to 35 kg/cm2. At 42 kg/cm2 a system for sending borated water
into the core operated.
10 h. In the containment vessel there was an abrupt rise in the pressure to 2 kg/cm2, probably due to an
explosion of detonating gas. The system went into operation and 19 m3 of sodium hydroxide solution was scat-
tered into the vessel. (The containment vessel had been designed and tested for a pressure of 4.2 kg/cm2.)
13 h 30 min. Valve 5 on the volume compensator was closed. The circulation pump in the loop with the
undamaged steam generator was turned on. Over the next 3 h the pressure in the reactor again rose from 46
to 162 kg/cm2. The reactor was cooled by the undamaged steam generator at an inlet temperature of 200?C
and an outlet temperature of 293?C. The rapid cooling prevented a gas bubble under the reactor cover. If the
pressure were reduced it could expand and cover the core which would in turn cause additional damage to the
fuel elements. The size of the bubble was calculated using actual data on the pressure, temperature, and
location of the water level in the volume compensator.
The hydrogen bubble was eliminated after several days with the aid of the coolant cleanup system as
well as by removal of water at the inlet to the reactor and injection into the steam cavity of the volume com-
pensator. Thus, by using the variable solubility of hydrogen at different pressures, it was possible to trans-
fer the hydrogen to the volume compensator and then remove it to the inner volume of the containment vessel.
Radiation Situation
The radioactive releases from the nuclear power station consisted mainly of noble gases which were
separated from the overflow water in the auxiliary building and then entered the atmosphere. They were too
small to have any effect on the population. True, there is a report that the maximum dose rate measured
outside the power plant must have been (apparently at the moment the cloud passed by) 20-25 mrem/h, while in
the Harrisburg region (roughly 16 km from the plant) it was 0.3 mrem/h. However, on April 7, observers
from the NRC, determined that the maximum dose 1.5 km from the plant was 0.05 mrem.h. The natural irradia-
tion dose to a resident of Harrisburg is estimated to be 88 mrem/ yr. According to data from the NRC and
?the Environmental Protection Agency, the dose due to the accident is roughly the same. For comparison, it
is noted that the natural irradiation dose to residents of Denver, Colorado, is about 147 mrem/ yr because
of the high altitude of the city.
In the neighborhood of the power plant various organizations made analyses of many samples of milk,
air, water, soil, and plants. The content of radioiodine in those samples where it was observed corresponded
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to a specific activity of 10-20 pCi/liter (in one sample 31 pCi/ liter), which is much less than the amount,
12,000 pCV liter, allowed by the public health authorities when domestic cattle have to be transferred to
stored feed. Radioiodine was observed in only 8 out of 152 air samples taken outside the plant. The highest
activity was 2.440-14 Ci/liter. The concentration of radioactive inert gases which was observed in the air a
distance of up to 65 km from the plant was less than a quarter of the allowed value. Radioactive iodine was
not observed in a single one of 130 water samples. Based on the amount of radioiodine which went beyond the
confines of the plant, the calculated irradiation dose to the thyroid gland of any user of the water was esti-
mated to be less than 0.2 mrem. Radioiodine was not observed in a single one of 147 soil samples or of 171
plant samples.
The radiation situation at the power plant is worse. At the site the dose rate on April 7 was 85 mrem/h
and increased somewhat in later days. In the state of March 31, three employees of the power plant received
a dose of from 3.1 to 3.8 rem during an attempt to drain and close the auxiliary building, that is, a dose larger
than the allowed quarterly dose (3 rem). Twelve other employees received enhanced doses; however, the
doses they received were much less than 3 rem. Access to the containment vessel is closed.
MAIN CHARACTERISTICS OF THE PWR REACTOR
Power, MW:
electrical
959 (gross)
thermal
2770
Uranium load, tons
82
Number of fuel assemblies
177
Number of fuel elements per assembly
208
Specific energy density in the core, kW/liter
90
Uranium enrichment, %
2.57
Pellet diameter, cm
1.09
Maximum fuel temperature, ?C
2299
Fuel-element shell:
material
Zircalloy-4
thiclmess, mm
0.67
Parameters of the water in the primary loop:
pressure, kg/cm2
154.7
inlet temperature, ?C
292
outlet temperature, ?C
320
Parameters of the water in the secondary loop:
pressure, kg/cm2
63
temperature, ?C
297
The reactor has two loops, each with two main circulation pumps with a capacity of 16,000 tons/h; the
steam generator is a vertical type with a capacity of 2760 tons/h.
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CONFERENCES, MEETINGS, AND SEMINARS
SECOND ALL-UNION CONFERENCE ON
THE CHEMISTRY OF URANIUM
E. A. Semenova
The conference took place in October, 1978, in Moscow. Work on increasing the efficiency of all stages
of the nuclear power fuel cycle was discussed.
More than 500 experts took part in the conference. In conveying greetings to the participants from the
President of the USSR Academy of Sciences, Academician A. P. Aleksandrov, Academician I. V. Tanaev
noted the achievements of Soviet science in the field of uranium chemistry and the need to expand their work
in view of the planned expansion of nuclear power.
One hundred six talks were heard in two plenary sessions and five sections. Review talks by B. N.
Laskorin, A. G. Samoilov, N. P. Galkin, and G. B. Naumov dealt with the mechanism for formation of urani-
um deposits, new natural uranium compounds, and the physical and chemical basis for extracting it from
mineral slurries, underground leachate solutions, and natural waters. The following were mentioned: the -
efficient use of autoclave processes for processing difficult and complex ores, the extensive use of sorption
and extraction methods in the fuel complex, the creation of efficient extractants and sorbents of new types
(sorbents impregnated with organic solvents, organophosphorus superextractants, amphoteric cation exchange-
solvating complexing agents, cyclic ethers, etc.), and the use of laser techniques in different stages of the
fuel cycle. Chemical and physical--chemical methods of separating uranium isotopes were discussed and in-
formation was presented on a large class of uranium fluoride compounds which are important in the develop-
ment of the scientific and technical foundations for dealing with nuclear-pure materials and for the regenera-
tion of spent fuel. Considerable attention was devoted to materials problems which arise in the development
of designs and technology for manufacturing fuel elements.
In talks in the section entitled "Natural uranium compounds," data were presented on structural studies,
the composition and properties of natural and synthetic uranium oxides, uranyl phosphates, and mourite, and
on the formation and dissolving of uranium oxides in natural processes and during underground leaching. It
was noted that oxidation?reduction and acid?base conditions in the water and the class of the geochemical
barrier determine the existence of the many geochemical types of uranium concentrations. A classification
was proposed in the form of a table and matrix with which it is possible to predict new geochemical types of
hydrogeneous deposits.
The talks in the section "Physical?chemical foundations of uranium extraction" reflected deeper study?
of the extraction and sorption processes employing different types of ionites and extractants (research on the
kinetics and chemistry of sorption and extraction, the effects of anions, the determination of the thermodyna-
mical characteristics of the processes), the extraction of uranium in the form of a uranyl halogen acetate as
one promising approach, and the wider use of modern instrumental techniques in research on uranium extrac-
tion processes and the use of computers for analyzing the results.
A large fraction of the talks in the section "Coordination compounds of uranium in solutions" was de-
voted to the thermal stability of uranium salts (uranium and sodium fluorides, uranium trioxide dihydride,
ammonium-uranyl tricarbonate, hexahydrauranyl nitrate, the carboxylates, etc.) with the use of high resolv-
ing power devices (the MTV-10-8 microthermal balances and others) and to studies of the structure and prop-
erties of synthetic uranyl sulphate complexes with bivalent metals, with uranyl molybdates and tungstates of
the alkali metals, and with the uranyl perbromates.
Many of the talks given in the section "Behavior of uranium in melts and solid phase reactions" touched
on more detailed studies of the chemical processes which occur in halide melts of uranium and the alkali met-
als. Complexing and the diffusion of uranium in halide melts were examined; the physical?chemical condi-
tions for formation of solid phases based on the phosphates of uranium and ruthenium in melts of the alkali
Translated from Atomnaya Energiya, Vol. 47, No. 1, pp. 63-64, July, 1979.
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metal chlorides and the kinetics of its removal by gaseous fluorination were investigated; and results were
presented on the volatility of UC14 and ThCI4 from melts with alkali metal chlorides. The precipitation of
uranium from fluoride melts by the oxides of Ca or Al was considered. The theoretical requirements for
electrorefining of uranium in molten halides with liquid metal electrodes were presented. The results of a
study of thermal dissociation and synthesis of the uranium oxides were presented. New results on the pyro-
hydrolysis of uranium hexafluoride, on the effect of an external force field on dissociation reactions, and on
the recovery and conversion of uranium fluorides were discussed. Reports were given on studies of the com-
plex ternary suicides of uranium with transition metals, of mixed chalcogenides of uranium, and of the lantha-
noids which are new magnetic and semiconducting materials.
At the conference talks were also heard on methods of determining uranium, impurities in it, and the
quality of nuclear fuel. V. K. Malcarov noted the possibility of, limits of application of, and errors of gravi-
and titrimetric, electrochemical, chemical?spectral, atomic absorption, mass and y spectroscopic, neu-
tron measurement, and x-ray fluorescence anniytic techniques. Volume, mass spectrometric, spectropho-
tometric, and other methods were proposed for determining the uranium in spent fuel, which permit a mea-
surement error of about 0.3%. In other talks remote automatic or nondestructive methods of determining
uranium were shown as possibilities with less accuracy (roughly 5-10%) but greater convenience for monitor-
ing the production of pure uranium compounds. An automatic rapid method for determining the nonstoichiom-
etry of nuclear fuel was proposed that was based on the emf of a high-temperature solid-electrolyte galvanic
cell which makes it possible to determine the ratio of oxygen to uranium to a high accuracy not attainable with
other methods.
The problem of environmental protection was examined in connection with the possible disposal of urani-
um in the form of solid and liquid wastes in the nuclear power fuel cycle. In this regard, the talk by D. I.
Gusev was of interest. In it he presented data on the uranium content in various environmental objects, in-
cluding human tissues, and demonstrated the effect of uranium on aqueous organisms when it is in a reservoir
for a long time. Ways in which uranium enters the human body were examined. In one of the talks the ecolog-
ical parameters for monitoring the uranium level in the hydrosphere and the biochemical parameters for
natural water reservoirs were cited.
At the conference attention was turned to research on quantum chemistry that is needed for a scientific
basis for the synthesis of uranium compounds with an unusual valence for this element and for a choice of li-
gands which stabilize the coordination compounds of uranium with different degrees of oxidation of its atoms.
The participants in the conference noted its high scientific level.
SYMPOSIUM ON THE SCIENTIFIC FOUNDATIONS
OF RADIOACTIVE WASTE HANDLING
V. I. Spitsyn and A. S. Polyakov
The symposium took place in November-December 1978, in Boston (USA). Seventy-seven talks were
presented there. Experts working on the treatment of highly active liquid wastes hold to the view that before
final burial the wastes must be embedded in stable solid materials. In Great Britain, Belgium, France, and
West Germany the technology of solidifying wastes to obtain borosilicate glasses is most developed. In France
a two-stage process of vitrification at the AVM installation in Marcoule was developed and first carried out in
1978. In West Germany and Belgium they are planning to build similar installations. In these countries a pro-
cess for embedding glass granules in metallic matrices has been developed. In Great Britain work is contin-
uing on the development of calcination with subsequent vitrification in a single crucible device.
The idea of creating "barriers" which are meant to provide a better guarantee of protection of the bio-
sphere than glass alone was discussed in many talks. The first barrier is to be a solid chemically and ther-
mally stable composition (core) which includes the highly active wastes. The material in the inner core is
expected to be a baked ceramic supercalcinate and glass granules. The second barrier is a layer of pyrolytic
carbon and aluminum oxide which coats the core. The third barrier is a metallic matrix (Pb+Sn; Cu; Ti, etc.)
Translated from Atomnaya Energiya, Vol. 47, No. 1, pp. 64-65, July, 1979.
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which is highly thermally conducting and resistant to corrosion and mechanical action. The container and,
finally, an appropriate geological medium can serve as subsequent barriers.
The embedding of highly active wastes in minerals known to be stable over millions of years was dis-
cussed. Fixation of radionuclides by titanate (inorganic) ionites followed by hot isostatic pressing was re-
ported (Sweden, USA, and others). Embedding of wastes in a glass ceramic based on fresnoite Ba2TiSi208
(West Germany), in synthetic rock ("synrock") containing silicate products and titanates such as BaAl2Ti8018,
Z r02?C a, Ti, CaTiO3, BaAl2Si208, KAISiO4, KA1Si206 (Australia), and in polludite (USA and others) was studied.
In many talks the properties of the hardened materials were introduced. Thus, in the Hahn-Meitner
Institute (West Berlin) the behavior of He formed in glasses due to alpha decay of actinides in it (the nuclear
reaction 19B(n, ce)7Li was used to simulate alpha decay) has been studied.
H. Healey and others (USA) used a scanning electron microscope to study the microstructure of solidi-
fied wastes, a pycnometer to determine the fraction of pores in the ceramic that were connected to the sur-
face, and an ion beam spectrochemical analyzer to study leaching and determine the content of Si, Al, alkali,
alkali-metal, and transition elements Fe, Zr, and Mc (W. Houser and others, USA).
The possibility of obtaining more certain data on the kinetics of the transfer of radioactivity into the
environment from solidified wastes by means of relatively short-term experiments was analyzed in detail
(USA). The solidification of the longest lived and one of the most volatile radioisotopes 1291 is being worked
out. It is proposed to embed up to 18% iodine in the form of barium iodate in concrete. It is considered that
constant isolation for the immense length of time required for 1291 cannot be guaranteed but the combination
of isolation and dilution will prevent danger to the present and future generations. Natural analogs are being
sought for determining the possible changes in solidified wastes over a prolonged retention period with limited-
duration laboratory experiments, the stability of natural glasses (the oldest volcanic glasses are about 40
million years old but most of them are much younger) is being studied, and the rates of hydration of volcanic
glasses are being determined as functions of the composition, temperature, and climate.
In the talk by G. MacCarthy (USA) the feasibility of hydrothermal contact of solidified wastes with rock
as a result of which some radionuclides (and possibly the majority of them) will interact with alumosilicate
rocks (but not with salt formations) was examined. Such contact may lead to formation of new thermodynami-
cally stable and low-solubility mineral-like phases.
The interaction of vitrified simulated wastes with rocks is being studied in the USA. A system was kept
in a heated state for amonth, after which x-ray structural and electron micrographical analyses were carried
out on 17 elements. Salt formations were discussed in detail as a medium for burial of solidified wastes. It
was noted that although these formations contain less than 1% water on the average, locally the water content
may be higher by a factor of five or more. Under the action of various forces the liquid migrates in salt with
a speed of from several millimeters to 1 m/yr. In addition a solution of chlorides may accumulate near waste
containers and cause them to corrode (USA).
The action of radiation on natural and artificial rock salt and the hydrothermal action of salt brine on
glasses at 350?C were discussed at the symposium. Tufa (a material with high sorption properties because
of its high (up to 70%) zeolite content; clinoptilolite is the main zeolite mineral in deep laying tufas and is
stable up to 700?C) and CaSO4 (low solubility, high thermal conductivity, and insignificant liquid content char-
acterize this material as a medium more suitable for wastes than rock salt) have been proposed as geological
media for permanent burial of solidified wastes. The importance of organizing a "bank" of experimental re-
sults for models which would make it possible to predict the behavior, including migration, of radionuclides
over a long period of time, as well as the extent of extrapolation of laboratory investigation, was noted. In
this regard the observations made at the site of the natural nuclear reactor in Oklahoma may yield interesting
results.
Talks by experts from the USA, Sweden, and West Germany were devoted to an examination of the ther-
modynamics of radionuclide migration in the geosphere and to a study of the sorption properties of clay and
rock. Pumping bentonite into crushed rocks is an effective means of inhibiting migration of radionuclides
from a waste reservoir into the environment (Sweden).
At the symposium it was reported that the chemical interaction between wastes and the nearby geo-
sphere may play a controlling role in their migration. Thus, corrosion of a metal container may lead to en-
try of polyvalent metal ions into the ground water in an amount sufficient to inhibit ion exchange retention of
radionuclides in nearby minerals, and sorption on silicate materials may lead to separation and creation of
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zones of enhanced concentration of certain radionuclides within the depository. The required lifetime for the
canisters (one of the barriers against entry of radionuclides into the biosphere) will be determined either by
a systematic analysis of the degree of risk or by legislated requirements (USA). Solidified waste canisters
made of titanium (Sweden) and canisters for spent fuel elements made of copper or a -A1203 (Sweden) or of
glass-ceramic p-spodumene (USA) are planned. Presentations on the corrosion of eleven samples of mater-
ial for bottles for long-term containment of 85Kr, on concrete and other low-temperature binders for solidify-
ing radioactive wastes, and on the development of mathematical models for evaluating the main aspects of
treating highly active wastes were of interest.
CONFERENCE OF EXPERTS ON THE EFFECT OF
NUCLEAR POWER ON THE ENVIRONMENT
L. A. Il'in and V. I. Karpov
Specialists from Australia, Austria, Great Britain, Argentina, Spain, Canada, China, Poland, the
USSR, Turkey, France, Switzerland, Sweden, and Japan, as well as representatives of international organi-
zations (UNEP, IAEA, ICRP, SCRD, WHO, EEC, and the World Council of Churches), participated in the
conference which took place in November, 1978, in Geneva, Switzerland. The chairman of the meeting was
the prominent Swedish scientist and director of the Institute of Radiation Protection, B. Lindell.
A draft of the talk, "The effect of nuclear power on the environment," prepared by a group of specialists
headed by El-Hinnawi, chairman of the energy program and director of the council on the state of the environ-
ment at UNEP, was discussed at the conference. This meeting was one of a series on evaluating the effect of
various energy sources (organic, nuclear, etc.) on man and the environment.
In this draft talk, data were presented on the present state and predicted estimates of the development
of traditional and nuclear energy in the world through the year 2000, on dosage due to various forms of radia-
tion on people in the USA, Great Britain, and Canada, and on the types of nuclear reactors and their use
through the year 2000. Special attention was devoted to the sources of and ways that radioactive and chemical
substances and heat enter the environment, the possible effects on man and the environment, and the human
dose levels due to all operations in the nuclear fuel cycle, from the mining of uranium ore to reprocessing of
spent nuclear fuel. The vital operations in the nuclear fuel cycle of burial of highly radioactive wastes and .
disassembly of nuclear power stations at the end of their operating life were discussed.
According to IAEA estimates, which agree with the data of other organizations, nuclear power will pro-
vide 11-13% of the total electrical power needs of the world in 1985 and about 17-20% in 2000. The installed
power of nuclear power plants is 350-400 GW(e) in 1985 and will be 1500-1800 GW(e) in 2000. The world pro-
duction of uranium in 1972-1975 was 19,000-20,000 tons/yr, in 1977 it was 30,000 tons/yr, in 1990 it is esti-
mated that it will be about 0.8-1 million tons/yr, and toward the year 2000 it will be 2-3 million tons/yr. The
water cooled?water-moderated reactor will be the main commercial reactor for electrical-power generation.
According to the data presented in the UNEP talk the main factors in the effect of the nuclear fuel cycle
on man and the environment and the possible criteria for comparing the fuel cycles for electrical-power gener-
ation could be the following:
the amount of fatal accidents and illnesses among workers involved in power production;
the specific (per GW(e)/yr) yields of harmful substances and heat into the environment and their effect
on man and the biosphere;
the cost of reducing the release of harmful substances from dumps and tailings accumulations to estab-
lished norms;
destroyed and nonreclaimable land;
the unit values of the natural resources used (land, water, organic fuel, air). ?
Translated from Atomnaya Energiya, Vol. 47, No. 1, pp. 65-66, July, 1979.
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We now consider some quantitative data according to these factors on the effect of the nuclear fuel cycle
on man and the environment.
Among uranium miners the number of fatal accidents is 0.1-0.5 per GW(e)/yr and the number of lung
cancer cases is 1.5?10-i man/yr. Several experts stated that similar data could be presented for other stages
of the nuclear fuel cycle as well.
In the talk the characteristic waste products and numerical estimates of their amounts were given for
operations in the nuclear fuel cycle. Thus, in mining the main environmental effects are the destruction of
land, dumps, mine drainage water, and contamination of the air with dust and radionuclides. The annual re-
lease of radon from the surface of a mine which provides for the operation of a 1 GW(e) nuclear power station
is 100 Ci. In processing, about 70% of the activity of the ore remains undissolved in tailings. Storage of the
tailings requires an area of 120 ha for 1 GW(e). Tailings accumulations can be a source of atmospheric pollu-
tion and pollution of ground water by radionuclides. Thus, special effort must be made to reduce their entry
into the environment to established norms. Similar data are given for other operations in the nuclear fuel
cycle.
An estimate was given in the talk of the population irradiation dose when the nuclear fuel cycle is opera-
ting. It appears that at present the irradiation dose to the population due to nuclear power is small (0.003-0.25
mrem/yr with an overall dose to individuals of 160-180 mrem/yr due to natural radiation, medical procedures,
etc.). The development of nuclear power through the year 2000 will not cause significant change in these num-
bers.
The talk is a summary and reference on the sources and amount of wastes produced and entering the
environment in all phases of the nuclear fuel cycle and on the possible dose burdens on man through the year
2000. Because of the comments by these experts the talk will be further improved and used for comparing
the effects of various energy sources on man and the environment. It is presumed that the concluding confer-
ence on comparing the effect of different forms of energy production will take place in October-November, 1979.
SCIENTIFIC?TECHNICAL CONFERENCE
"ENERGY AND ENVIRONMENTAL PROTECTION"
B. M. Stolyarov
? This conference, which took place at the Exhibition of Achievements of the National Economy (VDNKh)
of the USSR in November, 1978, dealt with the problems of protecting the air and water supply from the wastes
of thermal and nuclear power stations. Three sections were working at the conference, including one on the
protection of the environment during energy extraction at nuclear power stations, in the work of which repre-
sentatives of scientific-research, construction, and design organizations and nuclear power stations partic-
ipated.
Considerable attention was devoted to the hygiene aspects of protecting the environment from the wastes
of nuclear power stationa. In evaluating the level of possible radioactive contamination of objects in the en-
vironment, the main criteria are the industrial hygiene standards which regulate the limit of the annual dose
to a critical population group due to external and internal exposure and the operational limits on the amount
of individual radionuclides and their mixtures in the atmosphere and water. The talk by N. G. Gusev examined
the basic assumptions of the new draft regulation "Hygiene rules for the design and operation of nuclear power
stations" (SP-AES-78). These rules will replace "Hygiene rules for the design of nuclear power stations,"
No. 38/3-68. The main topic of the talk was the physical basis of the dosage quotas, accidental irradiation
dosages, gaseous aerosol wastes, etc. Stricter standards are proposed for the allowable normalized emis-
sion (ANE) per 1000 MW(e) of nominal power and the maximum allowable emission (MAE) at a nuclear power
station as a whole (Table 1). There RNG denotes any mixture of radioactive noble gases of artificial origin.
The long-lived nuclides are aerosols deposited in a filter during 1 (24 h) day and measured 1 day after the
sample has been taken. They include aerosols of all radionuclides remaining on the filter two days after
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TABLE 1. Norms for Gaseous Emissions, Cu/ TABLE 2. Normalized Emissions, Ci/day
day
Nuclide
ANE
MAE
Nuclide
ANE
MAE
RNG
500
3000
Long-lived
0,015
0,090
131I (gaseous
0,01
0,06
and aerosol
phases)
Short-lived
0,1
0,6
Nuclear power
station
Year
RNG
1311
Kolskaya
1976-1977
5
, 0
Armyanskaya
1977
4
1,7-10-4
Kurskaya
1977
180
? 0
Chernobylskaya
X1.77-1V.78
220
,-, 0
,
deposition began. Short-lived nuclides refer to radioactive aerosols deposited on a filter over a period of 2 h
and measured 1 h after the sample has been removed. They include aerosols of radionuclides remaining on
the filter 3 h after deposition began. A single (or daily) release of radionuclides 5 times in excess of the daily
mean value is allowed provided the total emission for the week does not exceed the corresponding calculated
value.
In the new rules, ANE and MAE for 90Sr, 89Sr, 137Cs, 69Co, 54Mn, 51 Cr are introduced. Experience in the
operation of nuclear power stations has shown that the releases of radionuclides are considerably less than
these provisions (Table 2). 'The same may be said`of the long-lived aerosols.
Scrubbing of air in recirculating ventilation systems at nuclear power stations to remove radioactive
aerosols and iodine by means of sprinkler systems was the topic of E. M. Klement'eva's talk (F. E. Dzher-
zhinskii All-Union Heat Engineering Institute). The essence of the method is that the flow of recirculating air
is washed in special chambers by water that is sprayed from a jet. Water is a solvent for iodine and its com-
pounds. Studies of the properties of aerosols at a nuclear power station with a water-cooled?water-moder-
ated power reactor (VVER) reported by S. S. Chernyi and V. P. Grigorov (F. g. Dzherzhinskii All-Union Heat
Engineering Institute) showed that the concentration of radioactive aerosols in rooms and air ducts at a nu-
clear power station is usually small and does not exceed 5-10-13 Ci/liter. During reloading of fuel it may reach
10-11 Ci/liter in isolated places. It was found that the bulk of the radioactive aerosols have much larger size
than previously assumed. Thus, the parameters of the logarithmically normal distributions of particle activity
averaged in terms of their aerodynamic dimensions in different ventilation systems vary and during normal
use the aerodynamic median diameter for the long-lived nuclides is 0.9-4.5 with'a standard deviation of
1.9-2.6. During reloading and repair work the median diameter of the aerosols in certain rooms and,ventila-
tion systems increases and may reach 8-10 Am. These results suggest the use of a filter material that is
coarser than FPP-15 cloth and therefore more retentive of dust.
G. V. Matskevich (F. E. Dzherzhinskii All-Union Heat Engineering Institute) examined the prevention of -
contamination of the environment by liquid waste from nuclear power stations. It was noted that most water
and chemical systems in a nuclear power station have no outlet. Water that is contaminated by radionuclides
is cleaned in special reprocessing systems and returned to the cyCle. The reprocessing wastes stay on the
reactor site under constant dosimetric monitoring. Only balanced amounts of water are released to the envi-
ronment after cleanup, storage in monitored tanks, and dilution to an extent that the concentration of radio-
nuclides meets the standard for drinking water.
Part of the talks dealt with the experience in designing equipment for ventilation systems and cleanup
systems for special gases and water.
Representatives of the NVAES and Kurskaya nuclear power stations spoke at the conference and shared
their experience in adjusting and operating gas cleanup and ventilation systems.
Recommendations were adopted for further improving the means and methods of preventing environ-
mental effects of nuclear power stations.
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INTERNATIONAL SEMINAR ON THE PRACTICAL
SIGNIFICANCE OF THE ICRP RECOMMENDATIONS
Yu. V. Sivintsev
The seminar, which took place in Vienna in March, 1979, was organized by the ICRP, the IAEA, and
international public health and labor organizations. More than 250 experts discussed 42 talks. The reports
of the prominent scientists, the members of the ICRP, in which the idea of the ICRP was described and its
difference from previous recommendations was explained in detail, were grouped separately. Previously
the ICRP supported numerical values of the maximum allowable radiation dosage of 5 rem/yr for professional
workers and 0.5 rem/yr for individuals in the general population. The radiation environment was regarded
as acceptable if the irradiation dose did not exceed the appropriate maximum allowable values. In Publication
No. 26 this principle and the numerical values of the basic dose limits (this new term has replaced the pre-
vious maximum allowable dose) are retained but only without stochastic effects. The dose limit for nonsto-
chastic effects is set at 50 rem/ yr. In order to prevent nonstochastic effects of radiation, the ICRP has
accepted a lower limit (30 rem/yr) for the lens of the eye.
Another task of radiation protection was declared to be to limit stochastic effects to an acceptable level.
The ICRP relates the introduction of the concept of acceptable risk to the assumption of a nonthresliold linear
dependence of radiobiological effect on dose and to the principle of "justification of practice." The use of
sources of ionizing radiation is justified only when the benefit gained by using them exceeds the harm to per-
sonnel and limited groups of the population due to chronic low-intensity irradiation. B. Lindell noted in his
talk that the justification procedure was identical to a "risk-benefit" analysis.
The resulting new principle, the optimization of protection, represents, in the opinion of the ICRP, a
practical means of applying the recommendation to keep the dose at as low a level as is reasonably acceptable
(the ALARA principle). In the same talk by Lindell, it was shown that optimization is identical to applying
estimates of the "cost-effectiveness of protective measures." In such calculations the dominant factor is the
estimate of the monetary value per unit collective dose (man-rem). As noted in the talks by D. Beninson, C.
Jamme, and E. Pochin, the limits must be regarded as marking the region of unacceptable irradiation. In
arguing this position, E. Pochin noted that according to present estimates a radiation dose of 5 rem/yr cor-
responds to a risk level of about i-! yr (or 100 fatal cases per year per million radiation workers). This is
equivalent to the mean frequency of fatal accidents in industry in the developed countries. Nowadays people
consider work conditions with a risk of about 10-6/yr as "safe" and for the general population a risk of 10-6/
yr.. Based on these initial data, the ICRP considers that the average radiation dose for personnel and individ-
uals in the population must not exceed 1/ 10 the corresponding dose limit.
There was considerable interest in a talk reviewing previously published data on lung-cancer fatalities
among uranium miners. The seminar participants took into consideration that a group of experts from the
IAEA is completing a numerical analysis of dosimetric data, which were possibly higher in previous work,
and noted information On the dust content in air from other hazardous substances.
The ICRP recommendation that the mean annual dose to personnel not ekceed 0.5 rem/yr signifies a
need to greatly strengthen radiation protection at almost all stages of the nuclear fuel cycle. Thus, accord-
ing to data in a talk by a group of staff members from the Trombay nuclear center (India), the mean annual
radiation dose to personnel (rem/yr) is: 1 for uranium extraction in France (lung irradiation 1.5-2); 0.4-0.5
in Great Britain and 0.47-0.59 in the USA for nuclear fuel manufacture; 0.74-1.2 in the USA (1969-1975), 0.16-
0.55 in France (1964-1974), 0.89-1.15 in West Germany, and 0.98 (2.07 for personnel involved in repairs) in
Canada for nuclear power plant operation; and 1.25 in Great Britain and 3.15 in the USA for reprocessing of
spent fuel at a radiochemical plant. As French trade union representatives noted, the sharp increase in the
price of nuclear fuel in recent years has not been accompanied by greater expenditure on protection. As a
result the irradiation dose to personnel has not been reduced, but has increased to 1.2-2.8 rem/yr (0.46 at
Translated from Atomnaya Energiya, Vol. 47, No. 1, pp. 67-68, July, 1979.
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uranium mines; 0.32-0.85 at nuclear power plants; and 0.4-1.5 at radiochemical plants). Thus, the benefits
and costs associated with the use of nuclear energy go to different professional groups and this is an additional
circumstance which complicates the practical application of the principle of optimizing radiation protection.
Part of the talks were devoted to optimization of protection in the medical use of ionizing radiation.
The experience with applying this principle in Danish radiotherapy centers is of interest. In that country,
with its population of 5 million, there is a relatively high incidence of cancer. It is the cause of 23% of deaths:
18,500 per year; 390 and 350 per 100,000 men and women, respectively. Nevertheless, radiotherapy has a
positive effect. Thus, there were 1250 cases of uterine cancer (7%); as a result of treatment of 1100 cases,
the fatality rate was reduced to 500 women/yr (4%). According to the estimates of K. Ennou and K. Jensen,
for several years the irradiation dose to personnel, due mainly to hand-operated irradiation of patients, has
been considerably greater than 0.5 rem/yr and in some cases exceeded 5 rem/yr. Application of the prin-
ciple of optimization, however, has led to the conclusion that the use of the technically complex method of
gradual introduction, which is distinguished by a sharp reduction in the radiation danger, is economically
unjustified.
Of the talks devoted to evaluating the value of 1 man-rem, considerable attention was drawn to work
according to which the estimates published previously by the ICRP (100-250 dollars/man-rem) are too low.
Thus, in the just-mentioned Danish talk, the value was taken to be 2000 dollars/man-rem. Among the data
cited by Beninson, a set of numbers was published which characterize the growth in the value of the coeffi-
cient a as a function of the extent of protection of the population of the region next to a reactor with gaseous
radioactive material releases. An analysis was made for the B3000 reactor of the ASEA-atom firm assum-
ing 0.1% defective fuel elements in the core for four different methods of cleaning up the refuse: a gas holder
in the form of a tank filled with sand of the type used at the Oskarshamn nuclear power station; the same type
of gas holder supplemented with an apparatus for recombining radioactive gases of the type used at the Bar-
sebeck nuclear power station; a gas holder with a recombiner with an adsorbent based on activated carbon of
the type used at the Forsmarkt nuclear power station; and three stage cleanup with a system for pumping
the gases into hermetically sealed vessels. On going from the first to the last version, the coefficient a
rises from 79,000 to 1,165,000 dollars/ man-rem. A group of investigators from West Germany showed that
for collecting long-lived gaseous-phase nuclides at radiochemical plants with a production of 1500 tons/yr at
an a of 100-200 dollars/man-remthe cost for clean up of emissions of 3H is 1-2.5; 14C, 1-2; 85, 25; 1291,
2.5-4.5; and aerosols, 1.5-2.5 million dollars.
At round-table discussions the members of the ICRP and representatives of the leading nuclear centers
discussed the problems of justified risk and the practical significance of the ICRP recommendations. The
IAEA plans to publish the proceedings of the symposium in 1979.
CONFERENCE ON THE DISRUPTIVE INSTABILITY
IN CLOSED SYSTEMS
V. G. Merezhkin
About 50 experts from Great Britain, the USSR, the USA, France, and Japan participated in the con-
ference which was organized by the IAEA in February, 1979, in Garching (West Germany).
As is known, the disruptive instability determines the maximum attainable current and plasma density
in tokamaks. The problem was the topic of this conference, where the results of theoretical and experimental
research from the past few years were discussed. Experiments on stelarators and tokamaks were discussed,
the possibility of stabilizing the disruptive instability was analyzed, and the rates of loss of energy and cur-
rent drop as disruptions develop in planned large machines were evaluated.
Up to the present time four types of disruptive instabilities in tokamaks have been identified and are
being studied: internal, during current rise, minor, and major. The internal disruption, which causes saw-
tooth oscillations in the temperature in the center of the plasma column, was observed in 1974 in the ST toka-
maks with a peaked electron temperature and current density profile. The first investigations showed that the
Translated from Atomnaya Energiya, Vol. 47, No. 1, pp. 68-69, July, 1979.
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disruption is initiated by the development of a helical mode (m = 1, n = 1). B. B. Kadomtsev proposed a
model of the internal instability which explained the rapid drop in the temperature and current density in the
plasma region near the axis by recrossing of magnetic field lines near the resonance surface at q = 1. Recent
experiments on the PLT (USA) have shown that this model needs refinement since in some cases the field line
recrossing effect is manifested weakly or is absent. In the theoretical work of A. Ware presented at the con-
ference, an explanation for the internal disruption was proposed based on an instability in the poloidal rota-
tion of the plasma.
Disruptive instabilities during the current rise have been studied on the TFR (France), PULSATOR
(West Germany), and ALCATOR-A (USA) machines. Experiments on ALCATOR-A showed that in the disrup-
tions which occur during the current rise the number of periods of the perturbations in the poloidal field, m,
is related to the parameter q (calculated for a plasma boundary at the limiter radius) by the roughly constant
relation q- 1.6 m. Analyzing these results, the experimenters concluded that the peculiarities in the beha-
vior of the disruptive instability during the current rise are related to the skin effect distribution of the cur-
rent during this stage of the discharge and can be avoided by increasing the plasma density and reducing the
rate of rise of the current in the initial stage.
In the stationary stage of a tokamak discharge a minor disruption, in which one of the harmonics m = 4,
m = 3, or m = 2 remains isolated throughout the entire disruption, may occur or a major disruption may occur,
developing in two stages. In a typical major disruption a harmonic with m =2 appears at first and a harmonic
with m = 3 or m = 4 occurs in the middle. During the transition between the first arid second stages and at the
- end of a major disruption, harmonics with m = 1 and m = 0 occur. The structure and dynamics of development
of the instability in minor disruptions have been studiedin many experiments in terms of external perturbations
of the magnetic field and perturbations of the x-ray emission from the central region of the plasma column.
Experiments have established that in a minor disruption with an isolated m = 2 harmonic the harmonic with
m = 1 also appears, rigidly coupled to the m = 2 mode. Coupled perturbations appear in the initial stage of a
major disruption as well. As experiments on the T-3, T-4, ORMAK (USA) and DITE (Great Britain) have
shown, a major disruption occurs due to a narrowing of the current channel in the plasma because its periph-
ery has been cooled by impurities. This explains the appearance of the m = 2 mode at the beginning of a
disruption in the case of a fairly large value of 0, - 4-5 when a disruption usually occurs with a large amount
of impurities in the plasma. Recently it has been demonstrated on several tokamaks that the amount of light
impurities can be sharply reduced and that highly stable discharges are observed with a low value of = 2-
2.5. In experiments on DITE, JIPP, JFT-2, and DIVA (Japan) it was noted that the perturbations in the poloi-
dal field in this stability window are reduced to about 5.104.
The report by K. Odayama on the possibility of obtaining stable regimes in a tokamak at low values of
- 1.3-1.7 without a disruptive instability attracted great interest at the conference. Such regimes were
obtained in the DIVA machine with a shell placed close to the plasma. Despite the regular instability in regimes with
small and significant losses associated with this instability, the energy lifetime of the plasma was TE 2.1-5.7
msec, about twice that predicted by the dependence observed on ALCATOR, TE nAT? (where n is the mean
plasma density).
In studies of the instability of the plasma in the W WI-A (West Germany) and Uragan stellarators in
Ohmic heating regimes a close correspondence was observed in the behavior of disruptions and helical per-
turbations of the plasma in stellarators and tokamaks. The perturbation in the W VH-A occurred primarily
near the q = 1 and q = 2 rational surfaces. For small values of the rotational transform angle, about 0.5,
coupled m/n = 1/1 and m/n=2/1 modes were observed in this stellarator as well as isolated min =1/1 and
m/n = 3/2 perturbations.
H. Furth proposed an "economical" theory of disruptive instabilities which starts from the assumption
of "competition" among tearing modes. B. Careras presented the results of a numerical simulation of the
tearing instability and noted the possibility of a sharp increase in the amplitude of the m/n = 3/2 mode during
the nonlinear stage of development of the m = 2 mode. The presence of magnetic islands near the q = 3 reso-
nant surface in a minor disruption, which agrees with the tearing instability model, was demonstrated in ex-
periments on the TOSCA machine (Great Britain).
587
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NEW BOOKS
V. L. Blinkin and V. M. Novikov
LIQUID-SALT NUCLEAR REACTORS*
Reviewed by Yu. I. Koryakin
Current developments in nuclear power, the success of which is well known, often overshadow the
essence of the assumptions behind its development. In the meantime there is sufficient justification for
formation of opinions or doubts about the correctness of the operating principles in practice in the nuclear
power economy. This is the sort of critical evaluation which is made by the authors of the book reviewed here.
Despite the fact that the book is mainly based on foreign materials and is thus a review, the authors' position
appears rather clearly and definitely. The authors see and prove that liquid salt nuclear reactors, which
make possible a compact and promising fuel cycle, can be regarded as reactors of the future and as an alter-
native to the liquid metal fast breeder reactors presently in existence and under construction.
The authors corredly focus attention on the problem of the external fuel cycle rather than on the prob-
lem of producing energy. It can be said directly that the scale and effectiveness of nuclear power generation
even in the distant future depends ultimately on the solution of this problem.
The approximately ninety references enabled the authors both to make a scrupulous examination of the
main features of liquid salt reactors and their external fuel cycle and to assemble the material in the book
in monograph form.
Of course, the authors' bias becomes noticeable upon acquaintance with the book. It appears, as is
usual and as a rule, in a veiling of the negative sides of the reactor concepts being described and in an empha-
sis on the positive sides. But, even understanding this, it is impossible to refuse the authors their due: Their
great merit is that they have systematically and in detail illuminated a little-known but possibly very promis-
ing approach to the development of nuclear power. Of no less value is the fact that the book leads one to re-
flection and, thus, stimulates new research.
V. V. Fisenko
CRITICAL TWO-PHASE FLOWSt
Reviewed byV. N. Smolin
Interest in two-phase flows has been dictated by the practical needs on new branches of industry such
as nuclear power, cryogenics, and rocket technology. Over the last decade a huge amount of experimental
and theoretical material has been accumulated on the structure and flow regimes of two-phase systems, on
heat exchange and hydraulic resistance in two-phase flows, on the thermodynamic and acoustic properties
of tow-phase systems, etc. However, most of the material is unrelated and has not been systematized, which
makes it difficult to use for practical purposes. Thus, the appearance of a monograph which analyzes and
generalizes research on two-phase flows is extremely necessary for engineering and technical workers at
research and design institutes as well as at construction offices specializing in power production, particularly
nuclear.
In the first chapter the physical features of the critical flow of a two-phase system are described on the
basis of the general assumptions of molecular kinetic theory and continuum mechanics. The conditions for
* Atomizdat, Moscow (1978), 112 pp., 1.10 rubles.
t Atomizdat, Moscow (1978), 160 pp., 1.50 rubles.
Translated from Atomnaya inergiya, Vol. 47, No. 1, pp. 69-70, July, 1979.
588 0038-531X/79/4701-0588$07.50 1980 Plenum Publishing Corporation
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formation of a flow crisis in single-phase compressible flows are examined and the theoretical models for
determining the critical flow rate of a two-phase flow are analyzed. Later, the propagation speeds of per-
turbation waves in a two-phase medium and the adiabatic index of two-phase mixtures are examined together
with the validity of using this index for establishing the coupling between various parameters of two-phase
media in an isentropic process in the range of gas contents from 0 to 1.
In the second chapter a dependence is proposed for the adiabatic index on the critical flow parameters
of two-phase single- and multicomponent mixtures of chemically nonreacting substances. This dependence
makes it possible to determine the critical outflow velocity and critical flow rate of a mixture for known pa-
rameters of a retarded flow, assuming that the medium is homogeneous at the critical cross section of the
channel, as well as to determine the propagation of small perturbations in a homogeneous two-phase medium.
In the third chapter the peculiarities of formation of a critical flow of a two-phase medium in the case
of boiling water flowing out through channels with various geometries are discussed. The phenomena accom-
panying formation of a critical flow in various self evaporating fluids are analyzed. Conclusions are drawn
about general effects for materials with different physical properties. An expression is proposed for the
isentrope index in a dimensionless form for calculating the critical flow rate and the critical outflow velocity
of various materials. The method for calculating critical two-phase flows is based on the assumption that
irreversible losses do not occur as the heat transfer fluid moves toward the critical cross section.
The fourth chapter is devoted to solving the problem of safety in nuclear power plants when the circula-
tion loop seal is broken, leading ultimately to ensuring core cooling under conditions of falling pressure and
disruption of the coolant circulation regime, as well as to taking the necessary steps to localize the conse-
quences of an accident. The interrelations between a change in pressure in the first loop and the rate it is
emptied due to critical outflow of a single-component two-phase mixture are analyzed. It is concluded that
the compressibility of the coolant during critical outflow of the two-phase mixture has a controlling effect on
the nature of this process. For this case the hydrodynamics of the flow of the medium is determined by the
propagation speed of sound, which is close to the thermodynamical equilibrium speed of the flow. Based on
an analysis of experimental data, a quantitative estimate is made of the effect of the Mach criterion on the
dissipative component of the pressure losses. As an example an algorithm is given for calculating the change
in the coolant parameters when the loop is depressurized.
In the fifth chapter flows of highly humid two-phase systems in variable cross section channels are de-
scribed. Such flows must be studied in order to solve problems associated with the efficient use of the energy
in the gaseous component to increase the kinetic energy of the condensed phase in special apparatus and com-
ponents of modern technology.
In the concluding sixth chapter the appearance of nonstationarity in two-phase critical flows is examined.
The theoretical study of this problem is still in a developmental stage. Various methods of solving nonlinear
differential equations for the behavior of the medium under nonstationary conditions are used in this case. In
particular, the author examines the solution of the nonstationary problem by the method of characteristics
taking several peculiarities of two-phase flows into account and by a numerical difference method based on
the equations for the directions and conditions for compatibility of the characteristic curves. In the latter
case some approximations are determined for the parameters of state at the vertices of a grid.
One may argue with individual conclusions and assumptions, such as those about the reasons for the
hydrodynamic instability and the heat transfer crisis in steam generator channels. It is possible, naturally,
to propose other methods and computational algorithms; however, there is no doubt that this book will be use-
ful to design engineers and scientific workers in institutes involved in the design of reactors as well as to the
engineering and technical staff at nuclear power stations.
589
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TRITIUM MEASUREMENT TECHNIQUES *
Reviewed by B. P. Maksimenko
Tritium is presently used on a rapidly expanding scale both in various areas of science and technology
and in industry. Progress in thermonuclear research in recent years is opening the way to production of
thermonuclear power based on the D?T reaction, whichwill require working with a substantial amount of tri-
tium. In this regard, the appearance of a book which examines, discusses, and systematizes the modern tech-
niques for measuring tritium in various media, including atmospheric air, water, soil, and biological objects,
is most topical and useful.
Considerable attention is devoted to selection of representative samples for measuring tritium and to
their preparation for quantitative analysis. In the case of liquid and solid samples the problem is complicated
by the fact that the beta particles from the tritium are mostly absorbed in the sample material because of
their small energy and it is thus difficult to use nondestructive monitoring techniques. It becomes necessary
to initially prepare the samples chemically or physically.
In the discussion of measurement techniques the criteria for choosing the most appropriate of them are
examined: sensitivity, accuracy, and duration of the measurements, the particular way in which the tritium
is released or its specific chemical or physical form, the commercial availability of equipment and apparatus,
and the need to process isolated samples or to carry out continuous detection. In the section devoted to the
measurement of surface activity, the methods of using gas discharge and scintillation counters and photograph-
ic emulsion for this purpose are described, along with a method for taking samples by wiping followed by
a measurement of activity. Less used methods which are discussed include mass spectrometry, gas chroma-
tography, avalanche semiconductor detectors, bremsstrahlung detection, electron multipliers, photographic
film dosimetry, calorimetry, and thermal stimulation of exoelectron emission. The book also describes the
calibration of standard emitters (tritiated water, tritiated toluene, etc.) by microcalorimetry and activity
radiometry with the aid of proportional counters with internal filling, as well as the use of standard emitters
for calibrating liquid scintillation counters (with the aid of internal or external standard emitters) and gaseous
counters. Underground water which has been in low ground levels for a long time compared to the half life of
tritium is recommended as a material which is free of tritium for determining the background counting rate of
measurement systems. Another possible such substance might be deep sea water which is not in contact with
subsurface currents. Two techniques are analyzed from the standpoint of estimating the errors in the corre-
sponding calibration of the measurement system.
Among the definite merits of this book are its great informativeness and the large number of practical
recommendations based on a deep understanding of the advantages and disadvantages of the methods discussed
there and which can be called upon to help the researcher in choosing the appropriate method for solving a
given problem. The good translation and high level of scientific editing of the material should also be noted.
There is no doubt that this book will be useful as a systematic handbook for measurements of tritium.
* Atomizdat, Moscow (1978), 96 pp., 0.90 ruble.
Translated from Atomnaya Energiya, Vol. 47, No. 1, p. 70, July, 1979.
590 0038-531X/79/ 4701-0590$07.50 ? 1980 Plenum Publishing Corporation
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L from
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