SOVIET ATOMIC ENERGY VOL. 56, NO. 3

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' September, 1984
SATEAZ 56(3)133-206 (1984)
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SOVIET
ATOMIC
ENERGY
Soviet Atomic Energy is abstracted or in-
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SOVIET ATOMIC ENERGY
A translation of Atomnaya Energiya
September, 1984
Volume 56, Number 3 March, 1984
CONTENTS
Engl./Russ.
ARTICLES
Material Behavior Investigations of the VK-50 Reactor Fuel
Element Assemblies ? V. A. Tsykanov, V. K. Shamardin,
A. B. Andreeva, G. P. Kobylyanskii, G. I. Mbershina,
Yu. D. Goncharenko, R. E. Fedyakin, and V. P. Sadulin 133 131
Determining the Frictional Characteristics of Reactor Materials
? V. M. Shchavelin, A. V. Kostochka, A. M. Bolobolichev,
A. A. Kuznetsov, I. S. Golovnin, and Yu. K. Bibilashvili 138 134
Modeling the Gaseous Swelling of Fuel Elements
? Yu. G. Degal'tsev, V. F. Kuznetsov, and N. N. Ponomarev-
1
Stepnoi 41 137
Neutron Cross Sections for the Calculation of the Damaging Dose
in Reactor Materials ? V. I. Avramenko, Yu. V. Konobeev,
and A. M. Strokova 144 139.
In Situ Examination of Radiation-Induced Internal Stress
Relaxation in the Column of a High-Voltage Electron
Microscope ? I. I. Novikov, V. A. Ermishkin, V. G. Zharkov,
E. N. Samoilov, I. S. Lupakov, and B. S.'Rodchenkov 147 142
Occurrence of Gas Porosity of Annealing Nickel Containing Helium
? E. Ya. Mikhlin, V. F. Chkuaseli, Yu. N. Sokurskii,
and G. A. Arutyunova 150 144
Small Induction Motors for Nuclear Power Stations
? K. A. Alikhanyan 155 148
Automated System for Making Observations, Evaluations, and
Forecasts: a Foundation for Comprehensive Protection of the
Environment and of the Public Health ? E. I. Vorob'ev,
V. M. Prusakov, and V. A. Minchenko 157 149
Radioactive Contamination of the Sea Environment Near the
Leningrad Atomic Energy Plant in 1982 ? S. M. Vakulovskii
and A. I. Nikitin 162 153
Neutron Inspection of Moisture in Slightly Enriched UO2
? V. V. Frolov, V. I. Bulanenko, and V. V. Charychanskii 165 155
Average Characteristics of the Slowing Down of Low-Energy Electrons
in a Tissue-Equivalent Material ? V. A. Pitkevich 168 158
Measurement of the Ratio of the Fission Cross Sections of 238U and
235U for Neutron Energies in the Range 5.4-10.4 MeV
? A. A. Goverdovskii, B. D. Kuziminov, V. F. Mitrofanov,
A. I. Sergachev, S. M. Solov'ev, P. S. Soloshenkov,
and A. K. Gordyushin 173 162
Measurement of the Fission Cross Section of 238U and 235U Nuclei
by 14 MeV Neutrons ? A. A. Goverdovskii, A. K. Gordyushin,
B. D. Kuz'minov, A. I. Sergachev, S. M. Solov'ev,
and P. S. Soloshenkov 176 164
"Gas Target" in the Divertor of a Tokamak ? M. Z. Tokar' 173 165
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CONTENTS
LETTERS TO THE EDITOR
y Radiation Field Generated by Neutrons in an Unbounded Uniform Air
Medium ? A. V. Zhemerev
Determining the Density of Pyrocarbon Coatings on Micropins by
Gasification in a Glow Discharge ? A. A. Babad-Zakhryapin
(continued)
Engl./Russ.
188 173
and I. S. Alekseeva
192
175
Graphite Capsules for Accommodating Indicators of the Temperature
and the Neutron Flux in Irradiation Units
? T. N. Shurshakova, V. V. Gundorov, and K. V. Grigorteva
194
177
Radiation-Induced Changes in the Thermal Conductivity and the
Electrical Resistivity of Pyrolytic Graphite
? Yu. S. Virgiltev and I. A. Dmitriev
196
177
Preliminary Filtering of the Results of Measurements Performed
in Exploratory Gamma-Logging ? I. M.- Khaikovich and V. N. Popov.
198
179
Calibration of Individual Dosimeters by the Absorbed Dose of Photon
Radiation ? Yu. P. Bakulin, V. P. Bashmakov, E. A. Bogdanov,
T. I. Gimadova, N. I. Muratov, and A. V. Tultaev
201
180
Comparative Investigation of the Process of Defect Formation in Si02
under Gamma and Gamma-Neutron Irradiation ? I. Kh. Abdukadyrova.
?
?
?
203
182
The Russian press date (podpisano k pechati) of this issue was 2/24/1984.
Publication therefore did not occur prior to this date, but must be assumed
to have taken place reasonably soon thereafter.
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ARTIcL_Declassified and Approved For Release 2013/09/14: CIA-RDP10-02196R000300040003-5
MATERIAL BEHAVIOR INVESTIGATIONS OF THE VK-50 REACTOR
FUEL ELEMENT ASSEMBLIES
V. A. Tsykanov, V. K. Shamardin,
A. B. Andreeva, G. P. Kobylyanskii,
G. I. Maershina, Yu. D. Goncharenko,
R. E. Fedyakin, and V. P. Sadulin
UDC 621.039.548.3
In proportion with the increased duration of operation of fuel elements, the probability
of their failure due to the radiation environment and corrosion damage of the cans from the
coolant side also increases [1]. A study of these phenomena on fuel element cans of the alloy
Zr + 1% Nb, most widely used in power reactors, is of great interest. In the present paper,
the results of material behavior investigations of fuel element cans and the sheaths of fuel
element assemblies (FEA) of the VK-50 reactor after 6 (FEA 1) and 8 (FEA 2) years of operation
are considered. Each FEA contains 162 fuel elements and 6 breeder elements, the claddings of
which are made from tubes with a diameter of 9.15 x 0.65 mm (alloy Zr + 1% Nb), annealed at
580?C during 3 h. The hexagonal sheaths are made of Zr + 2.5% Nb alloy. The fuel consists
of pellets of 3%-enriched uranium dioxide. The length of the active section of a fuel element
is 1980 mm. The coolant flow rate (boiling water) at the fuel element assembly inlet is (0.9 ?
0.1) m/sec, the pressure is 6.9 MPa, and the temperature at the core outlet is 287?C.
The distribution of the density and bulk steam content of the coolant over the height of
the core (Fig. 1) is obtained on the basis of measurement of the power distribution and the
thermohydraulic characteristics of the reactor. During 5.2 yrs of operation of the fuel ele-
ments,thewater-chemical cycle of the VK-50 reactor was neutral uncorrected, then a neutral-
oxygen cycle was introduced [2], after which the water-chemical indexes of the coolant were
improved markedly (Table 1). The oxygen content in the coolant in the upper part of the core
is higher by approximately a factor of 100 than at the FEA inlet.
FEA 1 and 2,loaded simultaneously, were operated with variable reactor power (Fig. 2).
In this case, the average thermal flux density for an average reactor power value for all the
time of the investigations amounted to 240 kW/m2, and the maximum thermal flux density for a
reactor power of 200 MW attained 680 kW/m2. In all, for different reasons, 50 shutdowns oc-
curred during this time, including those due to operation of the scram system approximately
4 times per year; the time of power decrease in this case amounted to 50 sec. Cooling and
heating up with planned shutdowns were effected with a rate of 15?C/h. The coolant tempera-
ture in the case of prolonged shutdowns was 50?C and in the case of short-time shutdowns it
was -100?C. Both FEA in the initial period of time functioned in the peripheral and then in
the central regions of the core, and FEA 1 was located at the center -13% and FEA 2 -43% of
the total duration of their operation.
The fast neutron (E > 0.8 MeV) fluence distribution over the height of the fuel elements
is obtained by a numerical method [3], taking into account the distribution of the fuel burn-
up and the steam content of the coolant (see Fig. 1). Cooling of FEA 1 and 2 in the cooling
pond took place over 2.5 yrs and 5 months, respectively. Fuel elements of the outer row were
withdrawn for the investigations of the assemblies.
The absence of damage was established by external examination of the fuel elements of
FEA 1 and 2. The deposits on the surface of the fuel elements of FEA I are compact but those
of FEA 2 are loose, flaky, and easily peeled off (Fig. 3). During separation and transporta-
tion the deposits flaked off, and underneath were revealed white sections covered with zirco-
niumoxide. The most intense deposition peeled off at the section between -650 mm from the
bottom and -750 mm from the top (FEA 1), but for FEA 2 it was between 100-150 mm from the bot-
tom and 400-450 mm from the top of the fuel elements. The thickness of the deposits was de-
termined by measuring the diameters of the fuel elements with a micrometer with an error of
10 pm before and after chemical dissolution of the deposits. The results of these measurements
Translated from Atomnaya Energiya, Vol. 56, No. 3, pp, 131-134, March, 1984. Original
article submitted June 3, 1983.
0038-531X/84/5603-0133$08.50 ?1984 Plenum Publishing Corporation 133
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TABLE 1. Water-Chemical Indexes of the
Coolant of the VK-50 Reactor
Operating
period of
FEA, yrs
Concn. of impurities,
t1ardness,
1-il
gequiv./ke
ng/kg
Coolant
pH
02
Fe
Cu
Zn
up to 5,2
Feed water
25
10-30
5-10
5-20
6,0
0,5
Reactor
water
160
20-50
5-20
5-20
6,2
2,5
above
5,2
Feed water
200
5
5
2
6,0
0,5
up
Reactor
water
200?
250
10
20
3
6,2
to 3
Fluence, 1021 nicm2
?
- 05
4 1i
U
2 -'0,2
0 c I I 0
0 0,5 1,0 1,5 0
Distance from bottom of core, m
Fig. 1. Variation of the bulk steam content of coolant (3), the
fast neutron (E > 0.8 MeV) fluence (1, 2) and the fuel burnup (4,
5) in the fuel elements of FEA 1 and 2 respectively over the height
of the VK-50 reactor core.
60,
le
20?2
200
150
a/00
0
50C.
Unloading
of FEA No. 1
nloading
of FEA No. 2
2 .1 5, 5 5 7
Reactor operating time, yr
Fig. 2. Graph of the power variation of
the VK-50 reactor in the operating period
of FEA 1 and 2.
for sections with unpeeled deposits are presented in Table 2. The thickness of the deposits
above on more energy-stressed sections of the fuel elements also increases with increase of
the residence time of the fuel elements in the reactor.
Radiographic analysis of the deposits was conducted on the DARD remote-controlled dif-
fractometer using the characteristic CuKa,emission. It was established that the deposits
consist mainly of oxides of iron (Fe203) and copper (Cu0). The presence was detected also of
manganese oxides (Mn304). The oxide film is zirconium dioxide with monoclinic modification,
134
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Fig. 3. External view of the fuel
elements of FEA 1 (a) and 2 (b).
TABLE 2. Thickness of Deposits on Differ-
ent Sections of Fuel Elements, pm
Distance from bottom
of active zone. mm
FEA 1
FEA 2
100
10
20
600
50-60
75-100
1600
25
50
2000
10
15
Fig. 4. Macrostructure of.transverse thin sections of fuel
elements of FEA I (a) and 2 (b) in the upper part of the ac?
tive zone ? 1920 mm from the bottom of the active zone (x9).
The differences, in the values of the diameter. of the fuel elements in the original state. -
and after residence in the reactor are small and are due mainly to oxidation of the external
surface of the fuel element cans. The internal,and external surfaces of the sheath tubes of
.both FEA were covered with a nonuniform thickness of oxide film, somewhat more.for FEA 2, .the'
? maximum thickness of which amounts to 200 pm at,the fins. The formationof the thickest ox-
ide films on these sections obviously'is'caused by the presence of residual etiesses,
stimu-
lating corrosion.
135
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Z,
200
150
100
50
10 20 JO 40 JO
t, thousand h
Fig. 5. Kinetics of thevariation of the max-
imum thickness of the oxide films Z on fuel
elements of the VK-50 reactor.
Fig. 6. Microstructure of transverse thin sections of the
cans of fuel element assemblies I (a) and 2 (b) in the upper
part of the core (1920 mm from the bottom of the core).
No appreciable deformation of the cladding, impairment of its intactness and squeezing
of the fuel was observed during metallographic examinations of FEA 1 and 2. A strong pitting
of the outer surface of the claddings was observed in the cladding material in the zone of the
fuel column and had a focal nature (Fig. 4). The foci, together with the continuous oxide
film, are distributed over the whole of the outer surface of the cans, and corrosion damage
of the FEA 2 cans is expressed to a large degree. It should be noted that for both FEA the
intensity of oxidation of the outer surface varies as a function of the height of the fuel
elements. On sections distant from the active zone by 1500-1800 mm, an increased amount of
foci with a length of less than 800 pm is observed, having moreover maximum depth. It should
be noted that in the zone of compensating volume, where there is almost no energy release dur-
ing operation of the fuel elements, the corrosion of the outer surface of the claddings has a
uniform nature and proceeds with the formation of thin (-10 pm) oxide films. On the power-
stressed sections of the fuel element claddings, the origination of more intense boiling is
possible, which can lead to increased corrosion. Using the published data [5], a curve was
constructed of the maximum thickness of the oxide films on the fuel element claddings vs their
residence time (Fig. 5). The nature and intensity of the corrosion vary in proportion to the
increase of operating time. In the initial stage, a predominantly continuous corrosion is
observed, accompanied by the formation of oxide films with a thickness of up to 50 pm; in the
next stage, an acceleration of thecorrosionfoci develops, after which a stage of gradual in-
crease of thickness of the foci starts with approximately the same speed as for the cOntinuous
corrosion. A marked difference in the state of the outer surface of the fuel element claddings
is detected in the region of the spacer grids of stainless steel and far from them. Thus, for
FEA 2, the thickness of the layered oxide films below the spacer grids attained -500 pm, and
between them -250 pm.
The orientation of hydrides over the perimeter of the claddings is predominantly annular,
and only on sections adjacent to points of local oxidation are radially oriented hydrides ob-
served (Fig. 6). The hydrogen content in the fuel element claddings was estimated by the
136
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rroperuies
neLaiduicai
of the Material of the FEA Fuel Element
Claddings
FEA
Section
of ele-
ment
Fluence;
rieutronSt
cm2
Tte = 20 'C
st
Ttest .---- 3?? ?C
'
-.?
'
"
.01'
tp `.
,d9'
95% of the theoretical density) pellets of small-grain uranium dioxide in the shell,
on the universal loop apparatus of a water-water reactor of the type considered in [10]. The
142
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12
0 0.2 0,4 go gg P/P
Fig. 3. Distribution of the local gas swelling S(p) (1) and
the relative energy distribution Q(p) (2) on the radius of a
cylindrical UO2 sample: the curves correspond to calculation
and the points td experiment (r, sample radius).
experimental conditions were close to those in [11] (T = 1750?C, burnup -1020 fissions/cm3),
the only difference being that the temperature gradient for the given samples was -200?C,
cm-1. The theoretical curve and experimental value of the mean bulk swelling of UO2 (initial
grain size 10 pm) after irradiation to -0.1% burnup are shown in Fig. 2. The swelling was
determined by the method of hydrostatic weighing with an error of no more than 10%.
The spatial distribution of local gas swelling over the radius of the UO2 pellets was
studied metallographically with analysis of the results on the Kvantimet-720 apparatus. Ex-
perimental and theoretical radial distributions of the gaseous swelling of the fuel over the
cross section of a cylindrical sample irradiated to a burnup of -3.101 fission/cm3 are shown
in Fig. 3, together with the relative energy distribution [7].
As follows from the given comparison, the space-time distribution of the gaseous swelling
of UO2 under irradiation obtained by calculation on the basis of the given method is in satis-
factory agreement, within the limits of experimental error (-10-15%), with the results of
measurements; this is acceptable for practical purposes.
LITERATURE CITED
1. J. Turnbull, J. Nucl. Mater., 38, 212 (1972).
2. W. Chubb, V. Storhok, and D. Keller, J. Nucl. Mater., 44, 136 (1972).
3. J. Turnbull and C. Friskney, J. Nucl. Mater., 71, 238 (1978).
4. A. Booth, GRDC-721 (September, 1957).
5. N. N. Ponomarev-Stepnoy and A. A. Khroulev, in: Proc. of TEPG Conference, JUlich, Ses-
sion E129 (1972).
6. Yu. I. Likhachev and V. Ya. Pupko, Strength of Heat-Liberating Elements of Nuclear Re-
actors [in Russian],-Atomizdat, Moscow (1975).
7. N. N. Ponomarev-Stepnoi et al., At. Energ., 34, No. 3, 197 (1973).
8. V. F. Kuznetsov, At. Energ., 47, No. 6, 410 (1979).
9. V. F.Kuznetsov, Preprint of the Institute of Atomic Energy, IAE-3275/5 [in Russian],
Moscow (1980):
10. Yu. G.NikolaeV et al., First Geneva Conference, USSR Paper No. R/621 (1955).
11. J. Turnbull, J. Nucl. Mater., 50, 62 (1974).
143
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NEUTRON CROSS SECTIONS FOR THE CALCULATION OF THE
DAMAGING DOSE IN REACTOR MATERIALS
V. I. Avramenko, Yu. V. Konobeev, UDC 621.039.53
and A. M. Strokova
In comparing the results of irradiating reactor materials and various neutron spectra
with data obtained in the irradiation of materials by charged particles, the damaging dose
is taken to be characterized by the number of displacements per atom. Calculation of the
number of displacements per atom is based on the formula
CO
D (1 crd (E) F (E) dE (1), (1)
where adis(E) is the neutron cross section of displacement of the atoms of the material at a
neutron energy E; F(E) is the energy spectrum of the neutron flux normalized to unity; t is
the flux of neutrons of all energies (E > 0). The cross section adis(E) is determined by the
sum of the cross sections of elastic and inelastic neutron scattering by nuclei of the mate-
rial, the cross sections of the nuclear reactions (n, 2n) and (n, a), etc., and also the form
of the cascade function.
According to recent recommendations [1], calculations of ?the damaging dose in terms of
the displacements per atom may be based on the US data system of ASTM [2] or the DAMSIG-81
data library [3]. In both libraries, the displacement cross section is given in 640-group
representation: 620 groups in the energy scale of the SAND II program [4] and 20 additional
groups with uniform energy divisions in the range 18-20 MeV. The ASTM data system is based on
the ENDF/B IV library of neutron cross sections and the DAMSIG-81 library mainly on the ENDF/
B III library. The aim of the present work is to develop a system of group neutron displace-
ment cross sections that is as close as possible to the international standard, for the basic
energy scales used in reactor calculations in the USSR.
Tables 1 and 2 show the system of constants developed. The mean (over the neutron group)
displacement cross sections have been determined for Al, Si, Cr, Fe, Ni, Cu, Zr, Mo, W, C, V,
Nb, Ta, Pb, and also Kh18N8 stainless steel. For the first nine materials, the displacement
cross sections obtained in [5] provided the starting point; for graphite, the data of [6],
based on the cascade function of [7]. For the first ten chemical elements, the results of
[5, 6] were also used in the comparison with the DAMSIG-81 library. For V, Nb, Ta, and Pb,
multigroup constants were developed on the basis of the data of [8], recommended by [9]. The
group cross sections for steel were obtained by a method described earlier on the basis of
the corresponding cross sections for Fe, Cr, and Ni. The effective threshold energy of dis-
placement in the cascade function in the TRN model, as in the DAMSIG-81 library, was taken to
be 40 eV for all the materials considered, except graphite, for which it was 60 eV. In aver-
aging the cross sections within the energy groups, the recommended procedure of averaging by
means of a standard neutron cross section is of the form
cp(E)= 11E E.2.5,
where E is expressed in MeV. Variation in form of the fission spectrum and the binding energy
within finite limits (2.0-2.5 MeV) has no considerable influence on the results of the calcu-
lations.
The number of displacements per atom at a specified point of the active zone of the re-
actor is defined as
(2)
D = criFi)11),
(3)
Translated from Atomnaya inergiya, Vol. 56, No. 3, pp. 139-141, March, 1984. Original
article submitted June 29, 1983.
144 0038-531X/84/5603-0144$08.50 ?1984 Plenum Publishing Corporation
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TABLE 1. Group Displacement Cross Section for 26-Group Division, 10-28 m2
F67;71737]
Neutron
energy
C
.
Al
Si
V
Cr
Fe
Ni
Cu
Zr
Nb
Mo
Ta
W
Pb
con

?=71 v
-? i
Utt
?
?
--1 14,5-14,01vielf649,4
1444
1683
2592
2417
1913
2055
2511
3088
2590
2938
2227
2387
2028
2015
0 14,07-10,5 619,4
1316
1353
2329
2145
1913
1761
2208
2438
2246
2461
1892
2024
1847
1937
1 10,5-76,5 620,4
1298
1049
1966
1846
1767
1487'
1899
1661
1863
1798
1350
1337
1343
1759
2 6,57-4,0 702,0
1340
1132
1750
1577
1535
1297
1557
1276
1426
1453
997,0
973,3
1149
1524
3 4,07-2,5 810,3
1349
998,0
1515
1344
1323
1148
1145
976,5
1090
1143
761,0
832,0
1034
1313
4 2,5-71,4 688,5
1104
llov
1054
1010
911,3
924;6
812,7
811,8
817,5
836,3
570,8
615,7
596,2
930,2
5 1,4-70,8 768,3
893,5
849,2
714,1
640,8
474,9
665,2
835,7
709,9
614,7
651,9
377,3
388,1
300,8
519,9
? 6 0,87-0,4 772,1
728,9
578;4
421,8
385,4
343;9
409,3
455;3
619;9
513,3
433,1
234,2
262,3
185,2
356,6
7 0,4-0;2 650,4
451,7
453,1
335,8
191,1
207,3
323,6
292,4
376,6
352,6
353,0
137,5
168,2
158,6
213,7
8 0,2-0,1 453,3
400,5
214,4
249,0
226,1
143,4
210,8
168,1
205,3
200,8
212,3
84,00
104,1
104,9
163,7
9 100-46,5KeW258,4.
180,1
40,53
156,3
98,71
103,0
151,0
130,9
110,1
73,30.
117,5
50,13
62,68
62,62
106,1
-10 46,5--21,5- ? 139,0
116,4
17,84
92,11
38,58
119,4
91;71
66,78
54,13
34,72
53,67
24,30
33,46
28,92
102,6
11 21,5-10,0 70,16
7,774
10,05
234,2
19,77
10,95
170,5
43,77
25,77
16,70
23,86
9,596
17,58
13.53
25,30
12 10,0-74;65 32,97-
5;782
6,232
164,7
35,76
21,38
29,21
30,80
13,36
8,920
10,94
2,743
10,78
6,356
24,59
?13 4,657-2,15 15,93,
-.14
2,728
3,223
76,45
13,88
7,089
20,79
17,00
9,914
8,321
6,336
0,126
10,56
2,171
9,407
2,15-1,0 . 5,704
1,439
1,634
2,534
2,892
5,977
8,643
8,296
2,818
0,332
4,939
0
10,93
0,356
5,635
45 1000465 eV 3,475
0,747
0,811
0,398
0,954
1,818
2,276
7,969
0,240
0
1,937
-
11,93
0,003
1,699
16 ,465-215 0 -
0,334
.0,128
0,006
0,101
0,075
0,495
11;281
1;066-
',-
5,724
-
38,82
0
0,090
.17 215-100 ? --
0,144
.0,008
0 ?
0,135
0,109
0,288
0,101
0,196
-
8,519
-
164,9
-
0,128
,18 100-746,5 -
-19
0,085
0,012
--
0,193
0,460
0,420
0,129
0,007
-
2.959
-
26,96
-
0,187
46,57-21,5 -
.0,085
0?018
--
0,288
0,237
0,619
0,245
0,010
-
38,21
-
530,3
-
0,277
20 21,5-10,0 -
0.086
0,026
-
0,430
0,348
0,913
0,435
0,014
-
1,009
-
1835
-
0,408
21 10,0--4,65 7-
0,131
0,030
-7.
0;616
0,509
1,330
-0,678
0,020
.L-.
2,231
-
44,85
-
0,594
22 4,65-2,15 --
0,188
0,057
0,908
0,749
1,958
1,052
0,030
-
0,360
-
63,88
-
0,874
23 2,15-1,0 . . 7-
0,217
0,083
-7
1;361
1,101
2,885
1,613
0,044
-7.
0,505
--
3,472
--
1,291
24 1,07-0,465 -7
0,382
0,120
-
1,942
1,604
4,209
.2,375
0,064
-
0,637
-
3,709
-
-
250,465-0,215' -
?---
7--
-
-
- --
--
- -7
---
-
?--
-
-
-
.
--
T ? - 0,020
1,470
1,170
-49,20
18,30
10,70
23,50
14,40
0,260
1,126
2,650
0,940
2,410
0,240
13,09
TABLE 2. Group Displacement Cross Sections for the WIMS and DLC-23/CASK Energy Di-
visions
Group
VilIMS
DLC-23/CASK
,cross section, 10-28 m
cross sec.
1028m2
Fe
Zrc
K8N8
staihlnless
Fe
Khl8N8
stainless
No.
neutron energy
neutron energy
steel
steel
..... cqcsi
10,0-76,066 Nfiev
1755
1588
615,2
1742
15,0-12,2 WV
1913
1980
6,066-3,679
1519 ?
1230
741,3
1502
12,2-10,0
1913
1923
3,769-2,231
1256
928,6
770,7
1256
10,0-8,18
1877
1863
2,231-71,353
843,1
795,7
694,8
864,0
8,18-76,36 ,
1732
1729
1,353-0,821
452,4
707,3
771,1
502,3
6,36-4,96
1633
1612
0,821-0,500
357,0
648,6
786,8
374,6
4,96-4,06
1459
1458
0,500-0,3025
296,5
496,1
719,9
297,9
4,06-3,01
1442
1413
0,3025-0,183
190,9
321,9
607,7
194,4
3,01-2,46
1180
1198
0,1837-0,111-
127,5
206,5
455,1
154,3
2,46-72,35
1147
1154
111-67,34 keV
131,6
133,9
313,0
135,4
2,35-1,83
1024
1055
67,34-740,85
65,57
85,42
208,2
72,13
1,83-1,11
663,4
691,4
40,85-24,78
165,9
53,45
135,7
136,5
1,11-0,5
384,4
414,0
24,78-15,03
7,194
32,99
87,86
23,96
0,5-0,111
209,2
220,4
15,03-9,118
13,71
20,26
54,59
22,79
111-3,35 key
59,97
61,94
9,118-5,53
27,12
12,98
33,48
28,88
3,35-0,583
5,235
5,132
5,53-73,519
8,349
12,01.
19,77
13,83
583-101 eV ?
0,091
0,111
3,5197-2,239
6,449
9,445
13,87
7,505
101-29,0
0,182
0,212
2,239-1,425
5,522
3,720
7,262
5,506
29,0-10,7
0,318
0,373
1,4257-0,907
6,178
1,545
4,586
5,505
10,7-73,06
0,560
0,654
907,5-367,3 eV
1,034
0,168
2,100
0,994
367,37-148,7
0,087
1,000
0
0,102
where cr. am.d. F. are the neutron cross section of displacement and the normalized energy spec-
]. ?
trum of the neutrons for the i-th group, respectively.
For alloys, the group displacement cross section may be found from the expression
(4)
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TABLE 3. Number of Displacements per Atom
(DPA) Produced by a Flux of 1022 neutron/cm2
in Kh18N8 Stainless Steel for Reactors of
Various Types
Reactor
DPA
Source of
data on neU -
tron spec-
trwn
.
E>.-
E >
>o,t
MeV
E>
> 0,4
MeV
BOR-60 (CAZ?)
4,84
5,43
8,14
1111
BN-600 (ZHEt)
3,03
4,50
8,96
1111
BN-1600 (ZHE)
2,45
4,30
9,90
1111
EBR II (CSZ)
4,81
5,45
8,19
1111
ETR
?
7,01
9,,36
1121
GTE
?
7,61
9,51
1121
VVER-440
?
6,50
9,66 Z '
1121
MR
?
5,65-7,1
8,45-9,0
Present wor14
RBMK
?
5,0-7,1
8,3-8,5
Presentwork
Inner wall of
?
6,63
9,45
1131
VVER body
*CAZ, center of active zone.
IZHE, zone of high enrichment.
+E > 0.5 MeV.
(j)
where ai and p(j) are the group displacement cross section and atomic fraction of the j-th
chemical element in the alloy. It is supposed that, at the current level of knowledge of the
cascade function, it is sufficient to take account solely of the basic components of the al-
loy. Thus, for stainless steel and other alloys of the system Fe-Cr-Ni, only the three prin-
cipal components need be taken into account. The neutron cross sections shown in Tables 1
and 2 for Khl8N8 steel are obtained in this approximation. It is assumed here that the atom- ,
ic fractions of Fe, Cr, and Ni do not differ from their mass content (for steel Khl8N8, val-
ues of 0.74, 0.18, and 0.08, respectively, are taken). For 15Kh2MFA steel, from which the
body of water-water reactors is made, it is sufficient to use the displacement cross sections
for Fe, since the content of alloying components is small (note that this is as in the US
standard [2]); for zirconium alloys, the constructional material of the active zones of therm-
al-neutron reactors, it is sufficient to use the displacement cross section of pure Zr.
Tables 1 and 2 give the multigroup constants of the displacement cross sections of mate-
rials for the 26-group representation, for the energy scale of the WIMS program, and for the
energy divisions of the DLC-23/CASK program used in the calculation programs (ANISN, DOT) for
neutron fluxes in the body and indicator samples of water-water reactors. Table 1 also gives
the neutron displacement cross sections for thermal neutrons [10]. Note that, in reactor
spectra, as a rule, -98% of the total number of displacements per atom are the result of neu-
trons of energy >0.01 MeV. The contribution of thermal neutrons to the formation of displace-
ments need only be taken into account in those cases where the flux intensity of thermal neu-
trons exceeds the intensity of the fast-neutron flux by a factor of ten or more. Comparison
of the number of displacements per atom obtained from Eq. (3) with the result given by Eq.
(1), on the basis of initial data on the displacement cross sections, shows that, for the
neutron spectra in the active zone of fast, water-water, graphite, heavy-water, and certain
other reactors, the discrepancy in the results is no more than 1%.
Table 3 compares the damaging dose produced by a flux of 1022 neutron/cm2 in various
reactors. It follows from these data that the damaging power of neutrons in various reactors
(and at various points within the active zones of several reactors) is different. In connec-
tion with this, the use of the neutron cross sections in Tables 1 and 2 is recommended. The
damaging doses, expressed in numbers of displacements per atom, for Kh18N8 stainless steel
according to the data of Tables 1 and 2 are approximately 10% less than those calculated on
the basis of the damage cross sections recommended in [14].
LITERATURE CITED
1. Proceedings of Advisory Group Meeting on Nuclear Data for Radiation Damage Assessment and
Related Safety Aspects, IAEA-TECDOC-263, Vienna (1982), p. 329.
2. Standard Practice for Characterizing Neutron Exposures in Ferritic Steels in Terms of
Displacements per Atom, ASTM E693-79, USA (1980).
146
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3. w. Lijp et ai., Damage Cross Sections Library DAMSIG-81, Netherlands, ECN-104 (1980).
4. E. McElroy et al., A Computer Automated Iterative Method for Neutron Flux Spectra Deter-
mination by Foil Activation AFWL-TR-67-41, USA (1967).
5. M. Lott et al., in: Proceedings on Applications of Nuclear Data in Science and Technology,
Vol. 1, IAEA, Vienna (1973), p. 89.
6. D. Reed, The Comparison of Carbon Atom Displacement Rate in Graphite in the Dragon Reac-
tor, the Petten HFR and Low Enrichment HTR. Dragon Project Report, DP-559, Great Britain
(1967).
7. M. Thompson and S. Wright, J. Nucl. Mater., 16, 146 (1965).
8. D. Doran and N. Graves, in: Irradiation Effects on the Microstructure and Properties of
Metals, ASTM STP 611, American Society for Testing and Materials (1976), P. 463.
9. "Recommendations for the calculation of materials irradiation exposures," Nucl. Technol.,
37, 358 (1978).
10. L. Greenwood and R. Smither, Displacement Damage Calculations with ENDF/B V, IAEA-TECDOC-
263, IAEA, Vienna (1982), p. 185.
11. V. V. Orlov, A. A. Proshkin, and A. N. Tuzov, in: Collected Papers of COMECON Conference
on the Exchange of Experience on the Development and Introduction of Apparatus with Fast
Reactors Based on the BOR-60 Reactor [in Russian], Dimitrovgrad (1972), p. 30.
12. R. D. Vasil'ev et al., in: Metrology of Neutron Measurements at Nuclear-Physics Instal-
lations. Proceedings of the First All-Union School [in Russian], Vol. 1, Izd. TsNIIatom-
inform, Moscow (1976), p. 226.
13. E. B. Brodkin et al., in: Collected Abstracts of the Proceedings of the Third All-Union
Scientific Conference on the Protection of Nuclear-Engineering Installations from Ioniz-
ing Radiation [in Russian], Tbilisi (1981), p. 67.
14. V. N. Bykov and Yu. V. Konobeev, At. Energ., 43, No. 1, 20 (1977).
IN SITU EXAMINATION OF RADIATION-INDUCED INTERNAL STRESS RELAXATION
IN THE COLUMN OF A HIGH-VOLTAGE ELECTRON MICROSCOPE
I. I. Novikov, V. A. Ermishkin,
V. G. Zharkov, E. N. Samoilov,
I. S. Lupakov, and B. S. Rodchenkov
UDC 621.039.531
Internal stresses produced mechanically or thermally in metals have substantial effects
on physical processes, in particular on recovery during annealing [1], fast-electron scatter-
ing [2], radiation growth [3], and so on.
Size change occurs in a deformed body on annealing on account of thermally activated
motion of lattice defects produced by internal stresses. The deformation rate for a crystal
is determined by the activation energy (if a single mechanism applies) and is exponentially
dependent on temperature. It is also proportional to the bulk density of the defects whose
motion produces the deformation. Therefore, by varying the defect density (for example, by
irradiation) one can alter the components of the deformation rate due to the motion of these
defects, in particular point ones.
This made it of interest to examine the internal-stress relaxation under irradiation in
a material exposed to high-energy electrons. We used specimens of Zr + 1% Nb, Zr + 2.5% Nb
alloys and aluminum. The zirconium-based alloys were deformed by rolling followed by anneal-
ing at 550?C for 5 h. The aluminum specimens were prepared from rolled foil of thickness
10 pm.
The specimens were irradiated with electrons in the column of a JEM-1000 high-voltage
electron microscope HVEM at an electron flux density of 1?1019 an-2?sec-1 and an accelerating
vbltage of 1 MV. We measured the distances between a series of reference points rigidly
linked to the matrix to determine the tensor for the deformation arising during irradiation
by the use of the negative images. These reference points for the zirconium alloys were
small spherical deposits of a second phase that did not alter during irradiation. The refer-
Translated from Atomnaya Energiya, Vol. 56, No. 3, pp. 142-144, March, 1984. Original
article submitted December 14, 1982.
0038-531X/84/5603-0147$08.50? 1984 Plenum Publishing Corporation 147
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Fig. 1. Irradiated parts of specimens of
Zr + 2.5% Nb (a) and aluminum (b). The ar-
rows indicate the reference markers.
ence points for the aluminum specimens were small copper particles previously deposited under
vacuum (Fig. la and b), which were firmly attached to the surfaces. The deformation in a giv-
en part of the structure was determined from the change in distance between the reference
markers.
To correct for possible change in microscope magnification during the experiment, an un-
irradiated part of the specimen was photographed almost at the same time as the irradiated
part. Any change in distance between the reference markers in this part was due to change in
magnification and was correspondingly incorporated in deriving the deformation tensor. This
method of determining the local deformation provided an overall accuracy of about 0.05% when
we used an 1818 stereometer made by Carl Zeiss (Jena) with standard negatives of size 6 x 9
CM.
No substantial changes in the phase distributions in the zirconium alloys under irradia-
tion were observed. After irradiation for 1.5h, there were also no resolved dislocation
loops. The defect structure was not altered in the zirconium alloys or in the aluminum.
In the case of Zr + 1% Nb, measurement of the distances between the reference points in
the irradiated areas showed that there was no deformation. In the aluminum, the irradiated
region expanded. On the other hand, the sizes of the parts of the Zr + 2.5% Nb alloy on ir-
radiation either decreased or increased along both axes in theplane of the foil (i.e., perpen-
dicular to the beam, while the dependence of the absolute deformation on dose or time was
the same (Fig. 2). This planar expansion or contraction of the irradiated areas is difficult
to explain in terms of point-defect condensation in particular crystallographic planes [4].
Also, the deformation did not alter on subsequent annealing, which distinguishes this effect
from radiation growth [5].
148
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42
zei
IC 15 20
t .10 electrons/cm2
Fig. 2. The observed c(t) dependence: ex) one of the princi-
pal values in the deformation tensor; ) theoretical rela-
tion for k 10-3 sec-1 in (7); 0 ) experiment.
One naturally explains the difference in behavior between the irradiated parts in terms
of differences in physical state before irradiation. The state of a part of a crystal at a
given instant is completely described by the distributions of mobile defects and potential
barriers, together with the stresses in the bulk, which together determine the subsequent be-
havior. As no qualitative differences were found in the microstructures of the irradiated
parts, the differences in behavior under irradiation were ascribed to differences in internal
stress.
A simple quantitative description in good agreement with the experiments can be obtained
by assuming that the deformation under irradiation is viscous. There are internal tensile or
compressive average stresses co in an irradiated disk region, while the planar surfaces of the
disk are free from stresses, and in the relation between the deformation rate c in viscous
motion and the stress a is [6]:
; (t) =Ba (t), (1)
where B = B(t) is a coefficient defining the capacity of the material for plastic flow and
which is proportional to the point-defect concentration n(t) at a given instant. Then
13(t)..knn(0, (2)
where kB is a quantity dependent on the diffusion characteristics. We assume that the point-
defect concentration was linearly dependent on the irradiation time (we neglected point-defect
annihilation and escape from the boundary of the region):
n (t) -=kt,
(3)
where kn is dependent on the irradiation intensity, and the inelastic scattering cross section
related to the formation of Frenkel pairs.
The stresses acting in a given region were taken also as linearly dependent on the de-
formation in the radial direction:
where ka is the elastic characteristic of the material.
From (1)-(4) we get an integral equation for a:
a (t) ao?k ta (t) dt ,
where k = kBknka; the solution to (5) is
(t) = a, exp [ ?
and then from (4) we have
21
exp(-40)].
Formula (7) describes the experimental curves. for Zr + 2.5% Nb with k 11X-3 sec-1.
We note that the character of these curves and the dose- corresponding to the inflection
'region are the same as .those for the dependence of radiation growth of this alloy on dose
(4)
(5)
(6)
(7)
149
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(expressed in displacements/atom) on neutron irradiation. However, in the saturation region,
the effect is larger in magnitude by a factor 2-4 than the radiation growth [5].
The following conclusions are therefore drawn. A deformed material has internal stresses
whose absolute value and sign may vary from one part to another. These stresses may locally
exceed the macroscopic yield point. The principal values ex and e in the deformation tensor
are equal in each particular case (within the limits of the errors). The shear components of
the stress tensor, which provide dislocation movement, are practically zero, which indicates
that the internal stresses are of hydrostatic type, and that they relax only by point-defect
motion, i.e., by a diffusion mechanism. The relaxation rate is proportional to the point-de-
fect concentration or the radiation intensity at a given internal stress. The ratio of the
number of stretched regions to the number of compressed ones is evidently determined by the
character of the thermomechanical treatment, and this ratio is less than one for treatments
that on average produce compressive stresses (it is about 0.2 as indicated by the experiments
for Zr + 2.5% Nb). The initial internal stresses determine the deformation at saturation.
LITERATURE CITED
1. T. Hasegawa et al., Acta Metall., 30, 235 (1982).
2. S. Amelinckx, The Direct Observation of Dislocations, Academic Press (1964).
3. P. Kelly, "Irradiation growth in Zr," in: Proc. Int. Conf. on Physical Metallurgy of
Reactor Fuel Elements, Berkeley (1973).
4. J. Nucl. Mater., 90, 1 (1980).
5. R. Adamson, Am. Soc. Test. Mater., 326 (1977).
6. J. G. Loitsyanskii, Mechanics of Liquids and Gases [in Russian], Nauka, Moscow (1970),
p. 450.
OCCURRENCE OF GAS POROSITY ON ANNEALING NICKEL. CONTAINING HELIUM
E. Ya. Mikhlin, V.F.Chkuaseli,
Yu. N. Sokurskii, and G. A. Arutyunova
UDC 621.039.531
Introduction. There are many theoretical and experimental studies on gas porosity, which
has a considerable effect on the behavior of materials under irradiation. However, at present
there is no agreed view on which mechanisms play the decisive part in the development of gas
porosity and gas release. The observed mobility of gas bubbles [1] is the starting point in
concepts in which the porosity growth is determined by the collision and fusion of such-bub-
bles (pores) [2, 3]. These concepts have been used in kinetic models for gas swelling [4-7]
and gas release [4, 5], where it is assumed that the pore motion .is produced mainly by surface
diffusion, with the diffusion coefficient for a pore related to the radius r by an expression
of the type
3 04/3
Dis) (r) ._D, 14.
(1)
There are,however, numerous papers in which the pore evolution is related to the diffusion of
single atoms [8-10].
One can eliminate the uncertainty on the roles of these mechanisms only on the basis of
new and revised concepts on the pore mobility as affectedby radius [11, 12]. On these con-
cepts, the diffusion coefficient for equilibrium gas pores, which is related to surface dif-
fusion, is defined by
Q4/3 W(r),
2n 6 r4
(2)
where W(r) is a factor describing the suppression of the mobility because the gas density in-
creases as r decreases. It should be noted that W(r) decreases from 1 at large r to 0 at r
1 nm [11, 12]. Then D1(r) at first increases as r decreases, but after the maximum falls
Translated from Atomnaya Energiya, Vol. 56, No. 3, pp. 144-148, March, 1984. Original
article submitted February 17, 1983.
150 0038-531X/84/5603-0150$08.50? 1984 Plenum Publishing Corporation
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Declassified and Approved For Release 2013/09/14: CIA-RDP10-02196R000300040003-5
rapidly to O. For r s 1 nm, Db(r) Dr(r), since the diffusion ccefficient for not very
large pores is made up of contributions from surface diffusion D(s) and bulk diffusion D07).
b '
D (r) =DW(r)+D(P(r). (3)
Th ()
e contribution from Dv
b decreases rapidly (as r-3) as r increases. The detailed values of
Dv and Ds (volume and surface diffusion) determine whether the mobility is a nonmonotone func-
tion of r and has a minimum (sometimes deeper) at r 1 nm. These concepts agree with the
experimental data available for some materials, according to which the mobilities of small
pores are 'quite small [13-15]. Equations (2) and (3) describe these data satisfactorily.
Also, (2) and (3) imply that a marked reduction in mobility occurs only in a comparatively
narrow size range, outside of which it remains fairly high, which disposes of the assumption
that the porosity evolution is due to collision and fusion .of moving pores. The contribution
to the porosity from the diffusion of single gas atoms can be incorporated within the frame-
work of this approach as a particular case of the diffusion of pores containing single gas
atoms.
This, however, isnot the only significance of (2). One usually employs (1) in models,
which greatly overestimates the mobilities of small pores, and therefore such models are
hardly reliable in predicting fuel behavior. Therefore, a quantitative model for gas porosity
and gas release has been devised [16] in which the pore mobility is defined by (2) and (3).
Calculations from this agree well with the available data on swelling and gas release for
oxide fuel, and also with the pore-size distribution within grains. This model [16] also
provides quantitative description of details of the gas porosity such as the size distribu-
tion for pores in defect-free parts of the grains and at dislocations and grain boundaries
and edges. We examine below how well the model describes these details by comparing the
distributions calculated separately for pores in defect-free parts of grains and pores at
dislocations with the corresponding distributions obtained by electron microscopy of helium-
irradiated nickel specimens after annealing.
Experiment. Nickel foils of thickness about 10 pm were bombarded at 100?C with helium
ions to a concentration of 0.15 at.%. Then these foils were annealed at 700?C for 1, 3, 6,
12, 100, and 1000 h. At the end of annealing, the foils were thinned down and an electron
microscope was used to examine the gas porosity. We processed the electron micrographs for
various parts to derive 33 histograms for the size distributions at various annealing times.
Some of the photographs provided separate histograms for pores at dislocations and away from
them. These 17 paired histograms are particularly informative in interpreting porosity de-
velopment. There were variations within fairly wide limits in the pore density and size dis-
tribution even for pictures relating to the same annealing time, which may be due to physi-
cal inhomogeneity (for example, in the diffusion parameters or thickness), but which also
may be of purely statistical character. Therefore, the comparison with the calculations was
based on histograms for 3 and 100 h of annealing, since these were statistically the most
representative (6 and 5 correspondingly). We then added the numbers of pores in a given size
range from all the primary histograms with a given annealing time to derive the resultant
histogram, which was normalized to the total number of pores (Figs. 1 and 2).
Basic Model Concepts. These have been presented previously [16] and amount to the fol-
lowing:
1. Porosity develops because of the collision and fusion of moving gas pores [2, 3].
The motion of the pores may be due, for example, to a temperature gradient. Also, there is
random wandering.
2. All the pores are spherical and equilibrium ones, while the gas they contain is de-
scribedby the equation of state for a nonideal gas.
3. Even the smallest gas-vacancy clusters down through complexes containing only one
atom of gas each may be considered as pores, which enables one to incorporate the diffusion
of single gas atoms into the general scheme. Any pore can then be considered as due to fu-
sion of n one-atom pores (pores of the initial size).
Structural defects such as dislocations and grain boundaries or edges retard the dis-
placement of the pores they trap and thus act as sites for more rapidpore enlargement. There-
fore, there are distinctive regions associated with each type of defect, and there are also
defect-free regions, for each of which the porosity development is described separately.
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6.0
90
0
8 40
0.
g?.
20
0
2,0 0 ;run
Fif. 1. Size histograms after 3 h of annealing at Dv = 10-16
cm ibsec-2 in both regions: a) for pores in the defect-free
part of thegrains (FW = 3.001016 cm-3; FIC= 5.5.10" cm-3);
b) for pores at dislocations (F(2) =1.1.1016 cm-3; F(21 =
exp ca c
4.0.1016 cm-3); here and in the subsequent figures, each his-
togram has been normalized to the corresponding total number of
pores; ------) calculation; ) experiment.
J
^
--n
--
- -t...
'
0
0
20
42 ? 10
0
4) 0
to 0
E. 20
o ? 10
r-
r
" .r.
?LI% ?0 ?0-
2
S rinm
Fig. 2. Size histograms after 100 h of annealing with Dv =
10-16 cm2esec-1 in both regions: a) for pores in tie defect-
free
part of the grains (Fkl) = 3.5e1023 cm-3, Flc = 2.4.
1016 cm-3); b) for pores at dislocations (F(2) = 2.9?1012
exp
cm-', F(2) = 1.5?1016 cm72).
calc
In Our experiments, the annealing was performed in the absence of temperatuTe\gradients,
so the pores moved only on account of their Brownian motion. The concentration F111 of pores
of size i in the defect-free region (first region) is described [16] by
4 N
(1) F(I) NI 2 v(t, 0),F(0) (1,13)
i)F0>F0). Fm 2 v(il, t)F ?
dt =-- {-3 i 2 i ; ? i ?
1=1 11=2 1=1 13=2
(4)
Here the first sum characterizes the increase in Fil) due to collisions between p9rcs in re-
gion 1 of size less than i. The second and third sums describe the decrease in F"/ due to
collisions of pores of size i from the first region with pores of other sizes from the first
region and from all other regions correspondingly (0 = 2 for dislocations, 0 = 3 for boundar-
ies, and 8 = 4 for grain edges). The last sum describes the flow of pores of size i to free
parts of dislocations, boundaries, and edges. In the case of isothermal annealing,
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4a [Db 1) Db(rt)] (r1 ? r (5)
where D (0 is given by (3), in which DP)(r) is defined by (2) and
The change in the concentration Fi2) of pores at dislocations (region 2) is defined by
dF(..12)
dt
1-1
0 vu
2)740 v(2) v(2) vi _,2)L4it) :s q!.12wwj_F12):s
3-I 3-1 3=1
where the first two sums characterize the changes in F12) due co collisions of pores from .hq.
first region with pores associated with dislocations and (Dc1
,2 is the rate of change in Fe)
due to diffusion flow of pores of size i to free parts of Aislocations. The third and fourth
sums describe the changes in Ff2) due to collision of pores involved in one-dimensional ran-
dom wandering along dislocations. Consideration of the contributions of these processes to
the porosity kinetics [7] gives
(6) ?
(2, 2) / (I)
q [1 (t) dt' .
(7)
Results and Discussion. Figures 1 and 2 show the size histograms for pores from the
first and second regions obtained from calculations for annealing for 3 and 100 h together
with the observed histograms. The calculations were performed for 700?C with the following
values for the parameters: atomic volume Q = 114,10-24 cm3, dislocation density d = 3.1010 cm_2,
grain size Z = 20 pm, surface energy y = 2400 erg/cm2 (1 erg = 1.10-7 J), van der Waals con-
stant b = 40.10-24 cm3 (for helium), suppression factor W determined for q = 400.10-24 cm3
[11, 12], surface self-diffusion coefficient Ds = 5.8.10-3 cm21sec-1,* and volume self-dif-
fusion coefficient Dv = 10-14 cm2?sec-1.t
The histograms show that the model calculations for both regions give size distributions
close to the observed ones. There is good agreement, particularly for the first region, for
specimens annealed for 100 h. However, on annealing for three hours the calculated distribu-
tion is somewhat to the left of the experimental one. Also, the calculated distribution for
pores at dislocations is somewhat displaced to smaller sizes by comparison with that for the
defect-free parts of the grains, whereas the observed one is slightly displaced towards larg-
er dimensions.
The latter feature indicates that the mobility of the pores along the dislocations (at
least for small ones) is somewhat higher than in the defect-free part of the grain. This is
fairly reasonable, since point defects diffuse near the core of a dislocation much more rap-
idly than far from it [22]. . Therefore, the pore displacement along dislocations may be some-
what more rapid, at least-by the self-diffusion mechanism. Of course, this applies to small
pores, whose displacement is associated with volume self-diffusion and whose dimensions do
not exceed the cross section of the area of elevated mobility near the core.
On this basis we performed the calculation in which it was assumed that in the second
region (dislocations) Dv = 3.10-13 cm2,0sec-1; Figs. 3 and 4 show the results, which indicate
that the size distribution for pores at dislocations is displaced to the right, while the
distribution for the defect-free part is virtually unaltered. This confirms the above as-
sumption about increased mobility of small pores along dislocations.
Conclusions. When helium-saturated nickel is annealed, the gas porosity is closely de-
scribedby the model, as in various cases considered previously [16], if the mobility is de-
scribedby expressions (2) and (3). There is fairly good agreement between the calculated and
2
*The value of Ds for 700?C according to [17] lies in the range 1.6.10-13-1.2?10-3 cm2esec-
,
while according to [18] it is in the range 4'10-3-9.3.10-8 cm2esec-1. The value used is
approximately in the middle of the range in Ds common to these two papers.
tDv at 700?C may have a value from 10-15 to 2.10-14 cm20SeC-1 [19-21].
153
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60
to
20
? n
? u
'5380
Foi^ 3O
20
a
? ? _
^
1 I - I
2,0 4c, 0 inm
Fig. 3. Size histograms after 3 h of annealing when D is in-
creased for pores at dislocations (Dv = 3.10-13 cm2eseC7-1): a) for
the defect-free parts of grains (F) = 3.0*1016 cm-3, Fc = 5.5*
1017 cm-3); b) for pores at dislocations (F(2) = 1.1,1016 cm-3,
F(2l ) = 1.8.1016 cm exp-3).
cac
?
o
oJO
t
0
co
...
0
20
10
0
0
13
0
JO
20
10
0
a
I
2, 0 +,0 40 nm
Fig. 4. Size histograms after 100 h of annealing when Dv is in-
creased for pores at dislocations (D, = 3*10-13 cessec-1): a)
for pores in the defect-free parts of grains (FR I, = 3.5.101; ,
cm-3 = 2.210" cm-3; b) for pores at disi8cations (V2/ =
ca c 3 (2
Fk = 1.34,1016 cm-3). exp
calc
observed size distributions for the defect-free region and for pores at dislocations. The cal-
culations from the model also confirm the assumption that there is elevated mobility for small
pores along dislocations.
LITERATURE CITED
1. R. Barnes and D. Mazey, Proc. R. Soc. A, 275, 47 (1963).
2. R. Barnes, J. Nucl. Mater., 11, 35 (1964).
3. V. M. Agranovich, E. Ya. Mikhlin, and L. P. Semenov, Third Geneva Conference 1964, USSR
Paper No. 338a.
4. H. Warner and F. Nichols, Nucl. Appl. Technol., 9, 148 (1970).
5. C. Dollins and F. Nichols, J. Nucl. Mater.., 66, 143 (1977).
6. E. Mikhlin and V. Chkuaseli, Phys. Status Solidi, A, 29, 331 (1975).
7. E. Ya. Mikhlin and V. F. Chkuaseli, in: Reactor Material Sciences (Proceedings of the
Conference on Reactor Material Science, Alushta, 29 May-1 June 1978) [in Russian], Vol.
2, Izd. TsNIIatominform, Moscow (1978), p. 124.
8. P. Prajoto, A. Wazzan,and D, Okrent, Nucl. Eng. Des., 48; 461 (1978).
9. J. Turnbull, J. Nucl. Mater., 62, 325 (1976).
10. 11. Wood, ibid., 82, 257 (1979).
11. E. Mikhlin, Phys. StatusSolidi, A, 56, 763 (1979).
154
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Declassified and Approved For Release 2013/09/14: CIA-RDP10-02196R000300040003-5
12.
E. Mikhlin, J. Nucl. Mater., 87, 405 (1979).
13.
R.
Cornell and G. Bannister, Proc. Brit. Ceram. Soc., 7, 855 (1967).
14.
M.
Gulden, J. Nucl. Mater., 23, 30 (1967).
15.
L.
Willertz and P. Schewmon, Metall. Trans., 1, 2217 (1970).
16.
E.
Mikhlin and V. Chkuaseli, J. Nucl. Mater., 105, 223 (1982).
17.
Yu.
S. Kaganovskii and Do Van Hai, Fiz. Met. Metalloved., 39, 828 (1975).
18.
P.
Maiya and J. Blakely, J. Appl. Phys., 38, 6981 (1967).
19.
M.
Feller-Kniepmeier, M. Grundler, and H. Helfmeier, Z. Metallk, 67, 533
?
(1976).
20.
A.
Wazzan, J. Appl. Phys.., 36, 3596 (1965).
21.
A.
Messner, R. Benson, and J. Dorn, Trans Am. Soc. Met., 53, 227 (1961).
22.
R.
Balluffi, Phys. Status Solidi, 42, 11 (1970).
SMALL INDUCTION MOTORS FOR NUCLEAR POWER STATIONS
K. A. Alikhanyan UDC 621.313.13+621.311.2:621.039
Nuclear power engineering in the Soviet Union is a field of industry requiring induction
motors for various mechanisms, in company with a wide range of other components. Improvements
in the structural schemes of nuclear power stations and experience in their operation are de-
pendent upon the creation of new motors for use under nuclear-power conditions. Small (frac-
tional horsepower) induction motors are used to drive various mechanisms involved in the
functioning of a number of systems under both normal and fault conditions. These motors are
principally used for driving slide valves and stop valves on water lines in the reactor and
for various actuator mechanisms. It is not possible to use ordinary general-purpose motors in
these applications.
The first series-A4 motors were developed in three structural designs and installed on
the Lovisa nuclear power station in Finland. They were installed in clean reactor buildings
and in contamination boxes (0.12-0.55 kW) in the upper (0.18 kW) and lower (0.12, 0.18 kW)
levels of the reactor hall. The motors are continually subjected to high ambient temperatures
and humidities at high pressures and at significant levels of radiation. The highest doses
of radiation(100 rad/h) are to be found at the lower level of the reactor hall. These con-
ditions, plus the demand for high reliability under all conditions of operation, including
fault conditions, dictate the choice of structural and insulating materials.
Studies have shown that despite the high temperature (120?C) and humidity (saturated
steam), they can have the same power ratings as their parent machines.
Motors with ratings of 0.18 and 0.25 kW, developed and brought into mass production for
power stations equipped with the VVER-1000 reactor, are also intended for driving slide and
stop valves. The development of new motors is called for by the specific features of opera-
tion of the power stations furnished with WEIR-1000s: elevated temperatures (70?C), relative
humidities (up to 100%), pressures (up to 1 kPa), radiation effects (5?105 mrad/h), service-
ability under all fault conditions, characterized by sharp increases in temperature (up to
150?C), pressure (490 kPa), and radiation (50107 mrad/h). There are also severe restrictions
relating to reliability, earthquake resistance, hermitic sealing, operation without servicing
and repair for sustained periods (up to 2 years).
For use inside the motors of insulating materials of heat stability class F, conductors,
coatings, and varnishes for impregnation have to be selected bearing in mind their subsequent
use on a nuclear power station. Tests and operating trials have confirmed their quality.
These machines were developed in 1979-1981 and unified as far as possible with machines
created earlier for nuclear power stations in Cuba and Libya. These were subject to even
more rigorous demands on earthquake resistance.
The motors for experimental mechanisms comprise a special group in themselves. They have
relatively low electromechanical time constants (tm = 0.03 sec) for rapid operation of ac-
. /
Translated from4tomnaya Energiya, Vol. 56, No. 3, pp, 148:149, March, 1984. Original
article submitted August 30, 1983.
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Induction motor,
screened protected
Induction motor, totally
enclosed: t = 60?C, n =
9050, p = 0.085-
0.1008 MPa
Clean en-
closure
with con-
tamina-
tinn henepc
Upper level
of reactor
hall
Lower level
of reactor
hall
Shut-off
valve
VVg R-1000
Shut-off
valve
VVgR-440
Actuator
drive
r= 0.5 rad/
h, t = 45?C
r= 0.5, rad/
h, t= 40?C,
= 90;
r=100rad/
h, t = 50?C,
n = 99*
r= 5.105
mrad/h, c =
force 9
r=5?104
mrad/h, c =
force 9
r= 5104
mrad/h, c =
force 9
t= 60-120?C,
p = (1.56-
2.27)? 10-4
Pa G
t= 60-120?C
p= (1.56-
2.27)? 10-4
Pa G
MPa A
MPa G
t= 70-90?C,
p= 0.17
MPa G
3
-x
a:
0,015
0,05
0,11
0,18
0,25
0,37
?44411111' -
0,55
v,
r.p.m.
1000;1500
3000
1500
1500;1800
750; 900
f, Hz
50
50,50
I
Fig. 1. Classification of conditions of use of induction motors and their para-
metricseries. Induction motor; t) temperature; n) relative, humidity; p)
pressure; 0 radiation; c) earthquake resistance; P) power; v) speed; 0 frequency;
G) steam-gas mixture; A) steam-air, mixture; 1,11) rated and fault conditions of
operation; Ll, 2, 3) are operating conditions: Si) continuous, S2) intermittent;
S4) continuous-intermittent.
Fig.
2. Small induction motors for nu-
clear power stations.
tuators (particularly shut-off valves). The electromechanical time constant is found by means
of expression
tm= J (wo/Mh),
where J is the moment of inertia of the rotor, coo is the angular velocity of the rotor, Mk is
the starting torque of the motor.
A batch of motors has been manufactured and brought into operation on unit 5 at the Novo-
voronezh nuclear power station. Mass production has been organized by the Soyuz4lektromash
production association.
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Auct_Lybib ul Lne contract snows that the motors have been used on all power stations
equipped with types VVfR-440 and 1000 reactors. The demand for new types of motors is being
studied. The stabilization of types ordered for nuclear power stations has enabled them to
be classified according to their conditions of use and their basic size and parameter series
(Fig. 1). Last, but by no means least, we have been able to unify all previously developed
machines, thus reducing the number of different types in production to a minimum as far as is
practicable. -The design of the motors is illustrated in Fig. 2.
AUTOMATED SYSTEM FOR MAKING OBSERVATIONS, EVALUATIONS, AND FORECASTS:
A FOUNDATION FOR COMPREHENSIVE PROTECTION OF THE ENVIRONMENT AND OF
THE PUBLIC HEALTH
E. I. Vorob'ev, V. M. Prusakov, UDC 339
and V. A. Minchenko
At the present stage of development of environmental protection and some related areas
of this science, one of the most important problems is further development of the theoreti-
cal and scientific foundations of the system for controlling the quality of the environment
in the interests of public health and maintenance of labor resources. Here, the problem of
creating automated systems for protecting the environment in regions of large-scale indus-
trial-power complexes becomes very important [1]. An analysis of the status of this problem
shows that a great deal of attention is directed in this country and abroad [2-5] toward
automating the observations and evaluations of the state of the environment on the one hand
and of public health on the other. Most existing systems and most systems under development
reflect only some one aspect of the interaction of man (or the source of the perturbation)
with the environment. The problems involved in making comprehensive evaluations and fore-
casts of the action of industrial emissions on the environment, including the working envir-
onment and the state of public health, are solved extremely slowly. The methodological and
conceptual foundations required for the solution of this problem have not yet been formulated.
In this connection, a special-purpose program for creating scientific (medical-health,
technological, and mathematical) foundations of an automated system for protecting labor per-
sonnel, the environment, and public health in regions of large-scale industrial-power complex-
es ("Angara" program) has been developed and is being realized for one of the industrial cit-
ies of Siberia.
An analysis of domestic concepts of monitoring [6] permits representing the automated
system for making observations, evaluations, and forecasts together with the organs and the
object of control as one of the variants of the most advanced system for protecting the en-
vironment and the public health. In this respect, the system for making observations, eval-
uations, and forecasts is viewed as a means for providing the required reliable data to con-
trol organs in order to make decisions concerning regulation of the sources of environmental
disturbance, the state of the environment, and the level of public health.
The .object of the investigation and of control is a complicated system of interacting
elements: production (source of action)--environment-public health. In its general form such
a system must include the interacting elements mentioned above (A1, A2, and A3, respectively),
the subsystems of factors determining the conditions of formation of the disturbance of the
environment (B1) and of man (B2) and, finally, the factors affecting (controlling) the state
and nature of the interaction of the basic elements of the system (C).
The distinguishing features of the automated information system (AIS) must be as follows:
comprehensiveness of the. estimates of the past, present, and future state of the sources
of the disturbance Of the environment and of the public health in developing solutions to
problems;
Translated from Atomnaya gnergiya, Vol. 56, No. 3, pp. 149-153, March, 1984. Original
article submitted April 8, 1983.
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functional completeness of the AIS, including dynamic observation, evaluation, and fore-
casting of the state of the object of control;
availability of the required complex of technological and programmed means for perform-
ing the investigatory functions (data bank, software, and family of models);
orientation of the system toward revealing the possible spectrum of responses and, pri-
marily,theprimary, (initial) diversions in the population under the action of harmful factors
in the environment.
This constantly operating system must raise the effectiveness and validity of the con-
trol decisions adopted for protecting the living conditions and health of the population from
undesirable actions of chemical pollution of the environment and other factors in regions of
large-scale industrial-power complexes. To create such an AIS, a system for collecting data,
data banks, and the means for analyzing and evaluating information as well as for modeling
and forecasting is currently being actively developed (Fig. 1).
The main principle of information support in creating the data collection system is re-
ferral to stations for obtaining and concentrating information on the state of the public
health, on the environment, on the work place, on the sources of environmental disturbance,
and on social factors. This comprises a constantly operating subdivision of medical, analyti-
cal, and monitoring services of the city, factory, administrative and statistical institu-
tions, etc. Plans are being made for obtaining information through the efforts of scientific-
research subdivisions in performing experiments and for adjusting separate elements and prob-
lems of the system as well as the methodology. The required documentation is developed,
adapted, and disseminated and the points of observation, population quotas, and periodicity
of the data collection are selected.
The greatest difficulties in developing the system arose in formulating the indicators
for observing the state of the public health. To overcome these difficulties we proposed a
comprehensive approach, characterized by two basic features . First, our approach includes
a wide spectrum of indicators, which presumes that judgements concerning the dependence of
the state of the public health on environmental factors will be based on objective informa-
tion on such characteristics as reproductive function, physical development, and functional
activity of the organism, morbidity and mortality, and longevity in representative groups.
In determining the indicators we used the recommendations and documentation from the N. A.
Semashko All-Union Scientific-Research Institute of Public Health, the Institute of Medical
Genetics of the USSR Academy of Medical Science, the Central Institute for Improving Doctors,
and other organizations. Second, we combine the selection of already available indicators
with the search for and development of new indicators, including fine indicators of changes
in the health of the population, at both the individual and collective levels.
Thus a procedure for genetic monitoring is being developed following a special program
together with the Institute of Medical Genetics of the USSR Academy of Medical Science.
Based on a retrospective analysis of the dynamics of the frequencies of inherited pathology,
a new method has been proposed for monitoring the process of mutation in man: observation
of the dynamics of "units " of the inherited pathology, which reflect the contribution of
new mutations and are easily incorporated into practical work [7].
Special attention is being devoted, in the system, to the development of a method for
evaluating the state of individual health. For this purpose, a variant of the automated sys-
tem of mass predoctor examinations using a special automated complex is being introduced.
A new scheme for medical examination, based on data obtained from questionnaires (anamnesis),
inclusion of the frequency of consultation of doctors, and chronic pathology as well as in-
strumental examinations has been proposed. The system will permit estimating in a differen-
tiated manner the state of health not only for the individual but also for separate groups of
workers and will give a considerable economic savings due to a reduction of the time spent
Son medical examinations and the numbers of workers visiting- a clinic. This will raise the
functional level of the AIS under development up to the level of discovery of prepathological
states as a foundation for requirements on the need for improvement of sanitary conditions.
Other investigations, in particular immunological, are performed in order to search for in-
formative tests of the state of adaptive-compensatory mechanisms.
An important element of the system under development is the data bank. The data bank
of the Angara program, the composition of the integrated bases of which is shown in Fig. 2,
is being designed to accept 60-270 million characters of diverse information per year. The
158
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Automated system for protecting
the environment and the public
health
System for
collecting data
Data bank
Printed doc-
uments, input
and control
programs
DBCS, data
base
Analysis of
information
Analytical
programs
Work flow
Modeling and
forecasting
Models
Technical means and system
mathematical support
IComponents of
AIS
Fig. 1. Special?purpose structure of the mathematical and technical
aspects of the "Angara" program: DBCS is the Data Base Control System.
bank presently includes the data base for sources of disturbance (sanitary-technical, chem-
ical, physical, biological, nervous-emotional, and psychological); for the state of the work-
ing environment and of the atmosphere, water reservoirs, the quality of drinking water, and
weather conditions; for social-economic (and food) factors, state of health of individuals
and groups of individuals from data obtained by massive medical examinations and special
physiological clinical examinations, stationary and dispensary observations, turnover, and
demographic statistics as well as from data characterizing the reproductive function of the
population. Information on the technological equipment and process, degree of heaviness,
work stress and work regime, content of chemical and biological impurities in the air, in
the working area, and on the skin of workers, physical and microclimatic factors, content of
chemical and biological pollutants in the atmosphere and water, meteorological factors, liv-
ing conditions, education, level of medical service, wages, professional course, duration
and location of residents in the city, functional state of separate systems and of the organ-
ism as a whole, morbidity (including oncological), duration and termination of pregnancy,
birth defects, physical and functional state of newborns and children, mortality, birth rate,
migration, and other factors are incorporated as primary information into the data bank.
Based on the primary information, a large number of the so-called secondary indicators,
which increase the working efficiency of data bank users, is formed. For example, based on
printed documentation for taking into account the frequency with which workers seek medical
help, more than 100 secondary indicators, which will become the primary hierarchy of the data
bases required for solving practical and scientific problems, will be obtained. At the first
stage, the health of twenty thousand workers in the leading industrial complex and forty
thousand residents of all ages in two regions of the city (twenty-two thousand individuals
in each region) with different content of pollutants in the air is monitored.
By virtue of the particularity of the problem being solved, related to the accumulation
of a large amount of information, which is constantly being added to and changed, as well as
to the structure of this information, a hierarchical model of data bases was chosen, and on
this basis the hierarchical type system of control of data bases (DBCS) INES was chosen. In
the system under development, the means of analysis and evaluation include software, matched
with the technical, information, and program means, and a collection of solved problems and
criteria for evaluating the information obtained and the results of the analysis. It is not
useful to create a special program-algorithmic support for this type of system. It can be
equated to the development of a special computer for such purposes. For this reason, the main
path is to adapt existing software to the requirements of the specific system.
The process of developing the means for analyzing the program data is characterized by
the need to perform the work with an incomplete determination of the list of problems to be
solved. In other words, means must be provided for investigating any fragment of the data
bases and any set of the data bases using all of the modern tools of mathematical statistics.
For this reason, the determination and systematization of the basic types of problems is of
fundamental importance in selecting the mathematical support based on classification of the
starting data.
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PDB
ES
CE
WEB
CF
PF
BF
PEF
PSF
AB
MD
OL
BSEF
MS
LC
BDI
BR
Mi
SP
BPH
PS
PD
FSO
RF
T
DD
Fig. 2. System of data bases of the data bank for the Angara
program: program data bank (PDB); data base for emission sourc-
es (ES); location (L); characteristics of emissions (CE); data
base for working environment (WEB); chemical, physical, bio-
logical, psychoemotional, and psychosocial factors (CF, PF, BF,
PEF, and PSF, respectively); atmospheric data base (AB); meteor-
ological data (MD); content of impurities (I); data base for wa-
ter (WB); organoleptic properties (OL); epidemiological danger
(E); chemical impurities (C); base of social-economic factors
(BSEF); level of medical service (MS); living conditions (LC);
education (E); food (F); base of demographic indicators (BDI);
birth rate (BR); still births (S); mortality (M); longevity (L);
migration (Mi); structure of population (SP); base for data on
the public health (BPH); passport-social information (PS); phys-
ical development (PD); functional state of the organism (FSO);
reproductive function (RF); turnover (TO); data on deaths (DD).
According to their significance, all data can be referred to quantity, frequency, and
grouping factors. Thus the indicators of the atmosphere, water, and working environment bas-
es are the concentration levels, noise level, etc. The morbidity of population .groups during
a particular period of time is the frequency indicator; sex, place of work, and age Are the
grouping indicators. But this is not unique. In determining the chemical load on a person,
the concentration of impurities is viewed as a quantity, but in characterizing the quality
of work at the factory, from the point of view of protecting the environment, the same data
over a definite period of time can be represented as the distribution of the frequency with
which the measured quantities fall in the range up to 1 MPC (maximum permissible concentra-
tion), 1-5 MPC, 5-10 MPC, etc. Analogously, the grouping indicator can be used in some prob-
lems as a parameter. According to established ideas, the dynamic and statistical data are
distinguished according to the forms of observation and the characteristics of the short-time
or chronic phenomena, referred to different time intervals, are distinguished according to
duration, etc.
In order to make possible a combined analysis.of diverse data, the data are tied to the
location and time of sampling with unified coding of the data for the Angara bank. The main
means for performing the analysis is a packet of applied programs for processing biomedical'
information. Blocks are partially formed with the help of DBCS INES. The unification of
blocks for combined analysis is provided for by the original program, which permits combining
into a single realization the data from Several blocks according to a set of coinciding in-
dicators or indicators shifted by a constant number of digits. An example of such a problem
is the search for a relation between morbidity and the state of the atmosphere during a par-
ticular period of time.
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Based on the means indicated above, the set of problems that can be solved in the auto-
matic regime on a computer is being investigated. Criteria are being formulated for evaluat-
ing the results. Since one of the main purposes of the system is to obtain differentiated
estimates of indicators of the health and of the state of the environment, a complex of cri-
teria for the health estimates is being created. It will include the following indicators
(in order of decreasing significance): the actual danger, the potential danger, the special
or objective indicators. The criteria of the actual danger are indicators of the changes in
the state of health. The criteria of the potential danger are the experimentally established
MPC and the indicators based on them. The special, or objective, criteria are the usual in-
dicators.
The theoretical foundation of this proposal consists of presently established character-
istics of low intensity: nonspecific nature of the actions due to a decrease in the resistance
of the organism to different pathogenic factors; manifestation of such action on the popula-
tion in a wide range of responses (from primary shifts to distinct harm) depending on the
strength of the harmful effect.
The model which permits studying and evaluating the processes and consequences of the
solutions adopted to various problems concerning the protection of the environment and of
the public health in the production-environment-public health system must be the main element
of the system. It is evidently now possible to develop a simulation model comprising a col-
lection of different types of models, each of which realizes a specific goal: It demonstrates
the behavior of separate aspects of the real system under the action of changes introduced
into it. On the whole, they will reflect to some degree of completeness our knowledge of
the entire system under study. This collection of models is being developed using a different
language for expressing the models (block diagrams, matrices, mathematical and algorithmic
description in the form of equations, mathematical operations, program description). The
modeling is based on the conceptual model of the production (source of action)-environment-
public-health system. The general structure of the model for computer realization, the gen-
eral algorithm for forecasting public health, and the working model for forecasting in the
pollution-source-air-basin-public-health (children) system has been developed.
During use of the AIS the models will be improved and enriched due to the internal con-
tent of separate subsystems and blocks. Analysis of the modeling has indicated the paths of
development and improvement of the models. Based on the achievements of forecasting in sci-
ence and technology, a series of assumptions for public health forecasting in the pollution-
source-air-basin-public-health system have been developed.
Experience has shown that the guiding principle for carrying out the program is the use
of modern concepts in the area of environmental protection and public health and their de-
velopment; large-scale use and testing of the latest means for recording, collecting, stor-
ing, and analyzing information; practical application of the results obtained at the develop-
ment stages, transmission, and dissemination of the elements of the system after they have
been tested and used commercially (adoption).
Thus an automated information system for protecting the environment and public health
is being developed. It is intended for studying the general and particular characteristics
of the interaction of man and environment (adaptive-compensatory processes) at the system,
individual, collective, and population levels, as well as for studying the relation between
the environment and sources of pollution.
The development of .a reasonable and efficient system requires that comprehensive medi-
cal-hygienic, social-hygienic, technical-mathematical and other forms of investigations be
performed on both methodological and theoretical levels.
LITERATURE CITED
1. E. I. Vorob'ev and V. Yu. Reznichenko, At. Energ., 50, No. 4, 234 (1981).
2. A. V. Primak and A. N. Shcherban', Methods and Means for Monitoring Atmospheric Pollu-
tion [in Russian], Naukova Dumka, Kiev (1980).
3. M. A. Murov and R. A. Uklonskaya, Experience in a Number of Foreign Countries on the Ap-
plication.of Automated Systems for Controlling Hospitals and Mass Examinations of the
Population [in Russian], VNIIMI, Moscow (1974).
4. B. Mikhov et al., in: Problems of Providing a Dispensary System for the Population
[in Russian], Meditsina i Fizkul'tura, Sofiya (1982), p. 171.
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5. T. Schneider, in: Handbook on Monitoring Air
hagen (1980), P. 241.
6. Yu. A. Izrael', Ecology and Monitoring of the
Gidrometeoizdat, Leningrad (1979), p. 375.
7. N. P. Bochkov et al., Tsitol. Genet., 16, No.
Quality in Cities [in Russian], VOZ, Copen-
State of the Environment [in Russian],
6, 33 (1982).
RADIOACTIVE CONTAMINATION OF THE SEAENVIRONMENT NEAR THE
LENINGRAD ATOMIC ENERGY PLANT IN 1982
S. M. Vakulovskii and A. I. Nikitin UDC 551.464.6.02
The Baltic countries are utilizing on a large scale the coastal waters of the Baltic Sea
for cooling the turbine condensers of atomic energy stations and for the disposal of liquid
waste with low specific activity [1, 2]. In the USSR, the coastal waters of Kopor Bay in
the Gulf of Finland are used as the cooling reservoir for the V. I. Lenin Leningrad Atomic
Energy Plant (LAEP). The low-activity disbalance water from the plant is also discharged
here [2]. The first unit of the plant incorporating an RBMK-1000 reactor was put into service
in 1973, and three additional similar units were made operational in 1981 [1]. Thus, a con-
siderable time interval has elapsed since the plant was put into operation, which permits us
to draw conclusions concerning the effect of radioactive waste from LAEP on the radioactivity
conditions in Kopor Bay.
Information on the radioactive contamination of biometers, mainly brown Fucus algae,
caused by the operation of atomic energy plants was published recently [3, 4]. According to
data from [5], the zone of contamination produced by waste from LAEP and detected by means
of modern radiometry methods is limited to a radius of a few kilometers, while the contamina-
tion zone pertaining to algae of the biometer type has a radius of up to 15-20 km.
The Institute of Experimental Meteorology completed a field investigation of Kopor Bay
in June 1982 in order to obtain data on the contamination level and the distribution of radio-
nuclides from LAEP waste in the sea environment components. During the investigation, radio-
nuclides were concentrated from samples taken from the coastal waters and from the open water
area of Kopor Bay, and samples of bottom deposits and biometers (brown and green algae) were
taken at points located at distances of up to 5.5 km from the LAEP. The position of the
sampling points is shown in Fig. 1. Water samples were taken by means of the equipment de-
scribed in [6, 7] with simultaneous separation of weighed matter and concentration of the ra-
dioactive impurity by means of a selective sorbent. A fibrous sorbent based on copper ferro-
cyanide was used for extracting the dissolved fraction [6]. The water was filtered through a
combination filter consisting of FPP-15-1.5 fabric (first layer) and a "blue-ribbon" paper
filter. After calcination at 350?C, the algae specimens, the sorbents, and the filters were
measured by means of a semiconductor gamma-spectrometer incorporating a DGDK-80 detector.
Samples of bottom deposits were measured in a similar manner after drying at 100?C.
The results obtained in determining the percentage of artificial y-rddiators in sea en-
vironment samples from Kopor Bay are given in Tables 1-3. It has been found that neutron ac-
tivation products from LAEP waste are present in virtually, all sea environment constituents
in the samples taken. The observed level of contamination indicates that the plant has been
operating in a normal manner during the period of time under consideration, which is readily
seen by comparing the level of biometer contamination near the LAEP with the data pertaining
to some of the atomic energy stations in other Baltic countries. Thus, at a distance of 1-2 km
from the Lovisa atomic energy station, the concentration of induced radionuclides "Co and
54Mn in brown Fucus algae amounted to 23-39 and 15-22 Bq/kg, respectively [4]. These values
are only slightly below the level of contamination of similar biometers near the LAEP. It
should be taken into account that the Lovisa atomic energy plant incorporates two units with
VVgRr.440,reactors, which contaminate the environment to a lesser extent than channel-type reac
tors. The contamination level of brown Fucus algae at a distance of 1-2 km from the Swedish
Translated from Atomnaya Energiya, Vol. 56, No. 3, pp. 153-155, March, 1984. Original
article submitted June 29, 1983.
162
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: CIA-RDP10-02196R000300040003-5
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Beach
3? LAEP I
? 0
2 1
?
?
Kopor Bay
/km
5
?
7
?
Fig. 1. Position of the sampling points
near the LAEP.
TABLE 1. Percentages of I37Cs and Neutron
Activation Products in Biometers from Kopor
Bay (samples taken on June 16, 1982), Bq/
kg dry weight
Point
No.
4.)
c) x
g ,
U E ti.-1
.51g n
Biometer
type
r
mi *
54mn
6000
137CS
M2
1
Greet) algae
0,12
12
54
22
2
2
Same
0,16
14
36
19
2
Brown algae
0,27
34
93
16
*M1 and M2 are the masses of dry and wet
algae, respectively.
TABLE 2. Percentages of "Co and I57Cs in
Seawater from Kopor Bay, Bq/m
oint
o.
U 5
g
4-' E PEI
2.Ensample
Sampl"m i Type of
d ate
Vol. of I
sample,
liters
owco
137C2
2
2
16.06.82
Suspension
1250
0,5
0,6
Dissolved part
750
0,7
6,0
4
5
19.06.82
Suspension
2310
0,08
0,1
Dissolved part
2170
4,0
5
20.06.82
Suspension
1720
40 eV the value of Wex is practical-
y constant, and for E > 300 eV it is somewhat larger than W. We can write
W = on (Pex/Pion)-8.";
Wex =- -B-Tx (Pton/Pex)iaon,
(3)
(4)
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?
102_
8
2-
el 7
1^ 0?
28
U5
o
w
? 2
10' 8 102 2 4 5 4
Electron energy, eV
Fig. 4. Average number of ionizations ninn- and excitations nex
and their variances ?ion' Ge, during the complete stopping of
an electron in a tissue-equivalent material (right-hand scale); F)
Fano factor; A) our calculation; ) data from [18]. The
curves in the upper part of the figure show the probability p(E)
(n = 1, ...,8) of the formation of n ion pairs when an electron
is completely stopped in matter.
10
8
6
2
10f
8
6
2
100
8
6
4
eV_
eV
90
80
70
50
SO
40
JO
20
10
0
10'
2 4 5 8 102 2 4 5 8 103 2 9 5 le
Electron energy, eV
Fig. 5. Average energy of ion formation W; ) average
energy for excitation; calculated values of W* = E/n ; A,
41) experimental values of W cited in [14, 20] for -a tissue-equiva-
lent gas; 04 ) data from [211 for a tissue-equivalent gas based
on methane and propane.
-Ion -ex
where el and el are, respectively the average energy expended per ionization and per ex-
citation (19.6 and 13.9 eV in the plateau region); _pion and pex are the fractions of the totl
number of inelasticcollisions with energy losses ern and TVc, -respectively (0.527 and O,47
in the plateau region). These relations are useful for analyzing the contributions to W and
W from various inelastic processes.
ex .
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The calculated average values of the characteristics of the slowing down of slow elec-
trons and the comparison with experimental data indicate that the modeling algorithm described
in [13] is adequate for solving a wide range of radiation physics problems related to the cor-
rect description of the transmission of slow electrons. The use of the "range" and "average
specific energy loss" concepts in such problems is evidently inadmissible.
LITERATURE CITED
1. J. Bevan, in: Proc. First Symp. on Microdosimetry, Brussels (1968), p. 161.
2. B. Grosswendt, in: Proc. Seventh Symp. on Microdosimetry, Vol. 1, Brussels-Luxemburgh
(1981), p. 319.
3. B. Grosswendt and E. Waibel, in: Proc. Sixth Symp. on Microdosimetry, Vol. 1, London
(1978), p. 311.
4. B. Grosswendt and E. Waibel, Nucl. Instrum. Methods, 155, 145 (1978).
5. M. Terrissol, J. Fourmenty, and J. Potau, in: Proc. Fifth Symp. on Microdosimetry, Vol.
1, Brussels ?Luxemburgh (1976), p. 393.
6. K. Maeda, Atmos. Terr. Phys., 27, 259 (1965).
7. M. Berger, S. Seltzer, and K. Maeda, ibid., 32, 1015 (1970).
8. M. Berger, S. Seltzer, and K. Maeda, ibid., 36, 59 (1974).
9. M. Berger, in: Proc. Second Symp. on Microdosimetry, Luxemburgh (1969), p. 541.
LO. M. J. Berger, in: Proc. Fourth Symp. on Microdosimetry, Vol. 2, Luxemburgh (1973),
695.
Ll. H. Paretzke, ibid., Vol. 1, p. 141.
1.2. M. Terrissol and J. Patau, [2], p. 411.
1.3. V. A. Pitkevich, V. G. Videnskii, and V. V. Duba, At. Energ., 52,190 (1982).
1.4. B. Smith and J. Booz, [3], Vol. 2, p. 759.
L5. H. Paretzke, G. Leuthold, and W. Wilson, Extended Abstract for the Meeting of European
Dosimetry Group, Dundee, April 10-11 (1979).
L6. A. Cole, Rad. Res., 38, 7 (1969).
17. H. Iskef, D. Twaites, and D. Watt, [2], Vol. 1, p. 201.
18. H. Paretzke and M. Berger, [3], Vol. 2, p. 749.
L9. U. Fano, Phys. Rev., 63, 222 (1946); 70, 20 (1947).
20. A. Waker and J. Booz, in: Proc. Second Symp. on Neutron Dosimetry in Biology and Medicine,
Luxemburgh, EUR-5452 (1975), p. 455.
U. D. Combecher, Rad. Res., 84, 189 (1980).
lEASUREMENT OF THE RATIO OF THE FISSION CROSS SECTIONS
)F 238U AND 235U FOR NEUTRON ENERGIES IN THE RANGE
.4-10.4 MeV
A. A. Goverdovskii, B. D. Kuz'minov,
V. F. Mitrofanov, A. I. Sergachev,
S. M. Solov'ev, P. S. Soloshenkov,
and A. K. Gordyushin
UDC 539.185
The main raw material in making nuclear fuel is 238U, so that the nuclear constants of
his isotope must be known very reliably. In accordance with the international list of re-
uired nuclear data [1], the error in the ratio of the fission cross sections of 238U and
35U nuclei (4/4) for reactor applications constitutes 2% with a neutron energy of 5-10 MeV.
he disagreements in the results of measurements obtained in different laboratories with neu-
ron energies exceeding 5 MeV reach 10-15%. This circumstance makes it necessary to obtain
dditional independent experimental data on al/ai:
In this work, we measured the fission cross sections of 238U and 23513 nuclei on the EGP?
OM electrostatic charge-exchange accelerator at the Physical Power Institute, operating in
he pulsed mode with a repetition frequency of 5 MHz. The average deuteron current on target
Translated from Atomnaya EnergiYa, Vol. 56, No. 3, pp. 162,464, March, 1984. Original
rticle submitted March 15, 1983.
0038-531X/84/5603-0173$08.50 ?1984 Plenum Publishing Corporation
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TABLE 1. Isotopic Composition of Uranium
Targets (%) and the Results of Measurements
of the Ratio of the Fission Cross Sections
of 23811 and 235U Nuclei with 7.3-MeV Neu-
trons*
Target
, number
2361J
236U
2385J
8,...5
aff"7
1
99,992?0,0010,004?0,0005
0,020?0,0005
-
2
6,864?0,0070,020?0,002
93,116+0,007
0.583?0,006
3
3,213?0,0100,007?0,002
96,78?0,01
0,584?0.006
*Content of 233U and 234U is negligibly
small.
TABLE 2. Results of Measurements of Fis-
sion Cross Section of 2 38u and 2 35U Nuclei
E, MeV
AEnAvtev
40,
Error, '
statisti-
cal
total
5,44
0,15
0,491
0,83
1,56
5,64
0,14
0,521
1,03
1,69
5,89
0,14
0,548
1,00
1,68
6,14
0,13
0,575
0,90
1,59
6,38
0,12
0,597
0,89
1,59
6,50
0,12
0,601
0,80
1,52
6,61
0,11
0,609
1,13
1,72
6,84
0,11
0,620
1,20
1,76
7,07
0,10
0,598
1,20
1,76
7,19
0,10
0,581
0,90
1,57
7,30
0,10
0,584
0,20
0,68
7,52
0,10
0,576
1,21
1,77
7,74
0,09
0,572
0,86
1,55
7,96
0,09
0,570
4,20
1,77
8,17
0,09
0,564
1,16
1,74
8,33
0,08
0,570
1:1,91
1,58
8,49
0,08
0,562
0,90
1,56
8,65
0,08
0,555
0,80
1,52
8,81
0,08
0,569
0,87
1,56
8,97
0,08
0,565
0,87
1,56
9,12
0,07
0,557
1,30
1,83
9,28
0,07
0,568
0,81
1,53
9,44
0,07
0,556
0,78
1,51
9,59
0,07
0,560
0,80
1,52
9,80
0,07
0,559
1,05
1,67
10,00
0,07
0,578
0,85
1,55
10,41
0,06
0,567
1,05
1,67
3;
was 1.2 IA with a current pulse duration of about 1 nsec. The reaction D(d, n) 3He served a
the source of neutrons. A gaseous deuterium target - a cylinder with a diameter of 10 mm an
a length of 40 mm -- was used. The output window of the target was a molybdenum foil 17 pm
thick. A platinum foil with a surface density of 200 mg?cm-2 was fastened to the bottom of
the cylinder. The deuterium pressure in the target was 1.18,103 Pa.
The fission fragments were recorded with a fast double ionization chamber. The target
foils consisting of the fissioning isotopes were placed orthogonally to the direction of thE
neutron flow. The chamber was filled with argon with a 10% addition of carbon dioxide gas'
at a pressure of 1.8,103 Pa. The distance between the electrodes in the chamber was 2 mm.31
The construction of the chamber included the necessity of decreasing the scattering materilJ
to a minimum. In the measurements, we used layers of uranium oxide whose thicknesses variec
from 200 to 500 pg?cm-2 on thin aluminum substrates. ii
We simultaneously measured two temporal and two amplitude spectra of the fission fragmt
pulses. The amplitude spectra of the pulses were used to determine the efficiency with whij
the fission fragments were recorded. The loss of counts created by the fission fragments
the chamber constituted -1%. The temporal spectra permitted separating out fission eventsl
due to the background neutrons. The ultimate temporal resolution was -3 nsec. The transitJ
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laseline of the neutrons (50 cm) was chosen so as to achieve a maximum rate of fissioning in
:he uranium targets while retaining the possibility of separating the primary and background
Leutron groups. The starting pulses were the current pulses created in the ionization cham-
)er by the fission fragments. The stop signal entered from the output of the master oscilla-
:or determining the frequency of the system forming the pulsed ion current.
In the course of the measurements, the 238U and 23513 targets were practically in an
.dentical neutron flow. To eliminate the effects due to the finite distance between the 2381J
:rid 235U layers and the blocking of one layer by the other, we performed the measurements
rith MO positions of the chamber. In one position, the 23811 layer was turned toward the
leutron source and in the other the 235U layer was turned toward the source. The results ob-
:ained were averaged. In this case, the influence of kinematic effects on the efficiency
rith which the fission fragments were recorded was likewise eliminated.
We performed the measurements in two stages. The first stage involved the determination
)f the energy dependence of the ratio of the fission cross sections of 238U and 23513. For
:his, we used layers with high isotopic purity (99.99%). At the second stage we performed
:he absolute normalization of the dependence obtained. In this case, the 23813 target con-
:ained an impurity of 235U nuclei. We determined the ratio of the numbers of 23513 and 238U
Luclei (n) in the target by a mass spectrometrical method. We determined the ratio of the
Lumber of 235U nuclei in the 238U target and the number of nuclei in the pure 295U target
r.om the count of the fission acts in the thermal neutron flow. Taking into account the dif-
:erence in the efficiency with which the fission fragments were recorded in the two chambers,
re can write:
=
there n5,8 is the number of fission acts in the 235U and 238U layers in the flow of fast and
:hermal neutrons.
We performed the normalization with a neutron energy of En = 7.3 MeV using two 238
U
:argets with different impurity concentrations of 235U nuclei (Table 1).
In analyzing the results of the measurements, we took into account a number of effects,
thich either distorted the quantity being measured or decreased the reliability of the value
)f a8/a5 obtained: incomplete recording of ?fission acts due to the finite level of the ampli-
f f
:ude discrimination (6) and the total stoppage of some of the fragments in the target itself
E62); scattering of neutrons by the substrates of the targets (63); scattering of neutrons by
,tructural parts of the chamber (64); fissioning of impurity isotopes (65); analysis of the
)eak of basic neutrons on a temporal scale (66); error in normalizing the curve of the energy
lependence of 4/4 (67).
, We present below the typical corrections (A) introduced into the final results and the
!rrors corresponding to them (B), %:
A B
61 1 0.74
82 0.5 0.05
63 0.05