SOVIET ATOMIC ENERGY VOL. 56, NO. 5

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• Document Page Count: 79
• Document Creation Date: January 4, 2017
• Document Release Date: September 14, 2013
• Sequence Number: 5
• Publication Date: May 1, 1984
• Content Type: REPORT

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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
November, 1984
Volume 56, Number 5
May, 1984
ARTICLES
CONTENTS
Engl./Russ.
Quantitative Estimation of"NUclear Safety ? B.P. ShiShin. .... ? ?
Emissions of " ilom .
Co, Ag and 541ih at the Armenian'NuClear'Power
287
275
Plant, and Their Content in the Surrounding Atmosphere
? E. S. Saakov, A. A. Avetisyan, and K. T. Pyuskyulyaw. . . ...
?
291
278
Unloading Additional Absorbers from the RBMK1000 Core
? N. V. Isaev, V. E. Druzhinin, and Yu. V. Shmonin
294
280
Dissolution of Oxide Films on Constructional Steels
? Yu. G. Bobrov, G. M. Gueyanov, A. P. Kovarskii, Yu. P. Kostikov
and A. V. Mbtornyi
298
282
Influence of Cold Deformation on the Behavior of Helium in Steel
OKh16N15M3B ? A. G. Zaluzhnyi, M. V. Cherednichenko,
O. M. Storozhuk, V. F. Reutov, and G. T. Zhdan
304
286
Isotopic Composition of Fuel in the Blanket of a Hybrid
Thermonuclear Reactor with a Thorium Cycle S.-V. Marin
and G. E. Shatalov
307
289
Recuperator with Inhomogeneous Electric and Magnetic Fields
? S. K. Dimitrov and Ya. A. MWnik
311
291
Calculation of Model High-Level Wastes in a Horizontal Apparatus
? V. V. Kulichenko, V. F. Savel'ev, V. A. Prokhodtsev,
and A. A. Ryabova
314
293
Concentration Ratios for Radiogenic Lead and Uranium in Aureoles
around Hydrothermal UraniumMineralization ? XT M. Ershov
319
298
Effect of Gamma-Neutron Radiation from a Nuclear Reactor on the
Electrical Stability of Microlite 7 N. S. Kostyukov,
M. I. Muminov, and V. M. Lanskov... . ... . . .........
322
300
REVIEWS
The New Generation of Highly Charged ion Sources
? K. S. Golovanivskii . . . . . .. ? ? ? ?, ? ? . .......
326
303
LETTERS
Small-Scale System for the Formation of a Field of Irradiation
with Accelerated Electrons ? O. A. Gusev, S. P. Dmitriev,
A. S. Ivanov, V. P. Ovchinnikov, M. P. Svin'n, and M. T. Fedotov.
336
311
Density and Surface Tension of Molten Mixtures of Uranium
Tetrafluoride with Lithium and Sodium Fluorides
? A. A. Klimenkov, N. N. Kurbatov, S. P. Raspopin,
and Yu. F. Chervinskii
339
312
Influence of Grain Size and Doping with Boron on the Behavior of
Helium in Stainless Steel 16-15 ? A. G. Zaluzhnyi,
M. V. Cherednichenko-Alchevskii, O. M. Storozhukm
and A. G. Zholnin
341
314
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CONTENTS
Consideration Of the Decay of 23312u When Determining the Isotopic
(continued)
Engl./Russ.
Composition of Uranium Fuel of a Hybrid Thermonuclear Reactor
? S. V. Marin and G. E, Shatalov,??????? OOOOOO "??? ? 6
343
315
Static Instability of Once-Through Steam Generators with Convective
Heating ? I. I. Belyakov, M. A. Kvetnyi, D. A. Loginov,
and S. I. Mochan
347
317?
Cryogenic Loop for y-Ray Sources I. I Buzukashvili,
and, G. S. Katumidze O ... ..? . .....
351
319
Influence of Neutron Spectrum on Formation of .283U from 232Th
Gerasimov, G. V. Kiselev, and A. P. Rudik
353
320
Spectral-Angular Distribution of yRadiation behind an. Instrumentation
? Unit? P. A. Barsov, V. M. SakharoV,? and V. G. SezenOv
357
322
The Russian press date (podpistuto k pechati) of this issue was 4/20/1984.
Publication therefore did not occur prior to this date, but must be assumed
to have taken place reasonably soon thereafter.
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ARTICLES
QUANTITATIVE ESTIMATION OF NUCLEAR SAFETY
B. P. Shishin UDC 621.039.58
Ensuringnuclear safety of a critical stand or nuclear reactor, i.e., preventing un-
controlled increase in the breeding coefficient of neutrons in a breeder system above unity,
presumes the solution of a series of technical 'problems: the development of special control
and protection systems, training of staff, and the production of documentation coordinating
and limiting the functional activity of the staff and systems of the equipment. The de-
signers, users, and inspection personnel must evaluate the nuclear-safety level of nuclear-
physics equipment in the same way. In such conditions, it is important to formulate and
employ objective methods of estimation as an instrument of, analysis, as a supplement to
the Rules of Nuclear Safety. In' the present work, the possibility of quantitative estima-
tion of the nuclear safety of apparatus, primarily critical stands, is demonstrated; it is
reduced to the problem of analyzing the reliability of functioning of the systems and ele-
ments of the equipment, using the apparatus of mathematical logic and probability theory.
The nuclear installation takes the form of an active region in combination with systems
for emergency protection, flooding the active region with water, parameter monitoring, etc.
The service personnel are regarded as one of the elements of the systems. Each system con-
sists of elements (sensors, valves, tubes, mechanisms, etc.), which may be in one of several
possible States (operative, inoperative, emergency, etc.).
The set of states of the elements determines the set of states of the system, which, in
turn, are components of the set of states of the installation. The states if the elements
may be taken in accordance with some value of the probability, i.e., the relative proportion
of time in which the element is in a particular state. Transition of the system from one
state to another in the course of the operating time is also characterized by a probability
[1]. Emergency states of the installation correspond to some probability value, which is
a measure of the "nuclear hazard," while its inverse is a measure of the "nuclear safety."
Estimation of these quantities is an important problem of nuclear safety. Nuclear emergency
of the installation may be regarded as the intersection of two events: the nuclear installa-
tion is in a supercritical state, and increase in the nuclear flux above a specified limiting
value in the active zone does.not trigger the emergency-control (EC) devices.
The probability of nuclear emergency 35 (NE) is written in the form
35(NE.)=34(p>0)q(EC), (1)
where g(p:>0) is the probability of a supercritical state of the reactor; o is the reac-
tivity of the reactor; q(EC) is the probability of EC failure. The probability g4(1))..o)
consists basically of two components. One -0,1(p>0) is determined by the operating program.
In the course of this program, the reactor is sometimes in subcritical, critical, and super-
critical states. The probability that the reactor will be in a supercritical state may be
estimated if the operating program and the methods of its implementation are known. The
second component 052(p>0) is determined by random unplanned processes in reactor control,
when it becomes supercritical as a result of operator error or by breakdown or loss of func-
tioning in the systems of the installation (systems of moderator and fuel filling, control
and safety system, etc.).
The probability of failure of the emergency-control system q(EC) is determined by the
specific logic circuit and elements of the installation. If a limiting value of the nuclear-
emergency probability is specified, then variation of the values of gi(fl>0) and q(EC) is
permitted. The nuclear-safety level may be increased by decreasing the probability that the
reactor will reach a supercritical state and increasing the reliability of operation of the
EC system.
Translated from Atomnaya Energiya, Vol. 56, No. 5, pp. 275-277, May, 1984. Original
article submitted July 6, 1982.
0038-531X/84/5605-0287$08.50 01984 Plenum Publishing Corporation 287
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Pip,>nd
0,3
I ? 1 1 1 1 1 1
-5fi
-10 47A rioo
Fig. 1. Dependence of the
probability of the reactor
state on the reactivity.
Fig. 2. System for flooding CP with
water.
In the present work, an example of estimating the nuclear safety of a critical stand
(critical pile with a water moderator, a system for flooding with water, and a control and
safety system) is considered, when experimental programs assocated with the attainment of a
supercritical state are employed, with possible disruption of the functioning of the water-
flooding system and EC failure.
Estimating .951 (ID> 0).
It is assumed that the critical pile (CP) in the flaw of working mixture first moves
from a deep subcritical state (pp> ? (3 , it is approximately 0.1. In such conditions,
the probability of random emergency flooding of the CP with water is estimated as
(E) 0) --- pa pb+ pc? 840-5+5-10-7
+8-10-5=1,6.10-4.
Estimating q(EC)
In the given example, it is assumed that the EC system (Fig. 3) has two power-measure-
ment channels (1-2 and 3-4), a reactivity-measurement channel (5-6), two groups of EC work-
ing components, (8 and 9), elements of manual (10) and automatic (11) regulation, and a
compensating unit 12. Sensors 1, 3, 5 send signals to the comparative instruments 2, 4, 6.
When the signal exceeds a specified level, it is fed to commutating device 7, resulting in
the triggering of all the working elements of the safety and control system (8-12).
Failure of the EC system may occur in the case of failure of the following system ele-
ments: 1 or 2, 3 or 4, 5 or 6, or even element 7, or element 8-12. The expression for the
289
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probability of such an event is written in the form
( EC) - II (qi- qi+1) q7 -I- [I qi, (5)
3, 5
where qi is the probability of failure of the i-th element in the system; r is the product
sign.
The failure rate of the radioelectronic equipment, its components, the relays, drives,
electric motors, is 10-4-10-6 h-1 [5-7]. The failure probability of such equipment in the
course of a workingshift is 10-2-10-2. According to Eq. (5), the probability of EC-system
failure in this case is approximately 10-2-10-3. For nuclear power plants, the approximate
failure rate of the control and safety system is a few times 10-4 h-1 [5, 7].
Probability of Nuclear Emergency
The probability of nuclear emergency in experiments associates with attainment of a
supercritical state by the reactor, in the case of possible disruption of the functioning of
the flooding system and RC system is calculated from Eq. (1) with
- (p > 0) (p >0) + 1,2(p>0). ?
With the values of fP (f) >0), g's (1) > 0) , and q(EC) determined above, the probability of
such an emergency is p (NE) = 10-4-10-6.
The maximum contribution comes from AC failure in planned movement of the reactor to
a supercritical state in the ?course of ?the experiments. The next most probable emergency is
associated with erroneous release of water from volume 5 and emergency discharge through
channel 1-7 (Fig. 1). If V' (NE) is the probability that there will be an emergency in the
course of a single operating shift, the probability of such an emergency in the course of a
year (100 operating shifts) is 10-2-10-4. A similar analysis may be conducted with respect
to other nuclear-installation systems.
The development of quantitative methods of estimating nuclear safety will facilitate
the adoption of more accurate solutions and the reduction in material costs in ensuring the
safety of nuclear-physics installations.
LITERATURE CITED
1. V. V. Frolov and V. I. Bulanenko, At. Tekh. Rub., No. 1, 3 (1981).
2. Handbook on Engineering Phychology [in Russian], Mashinostroenie, Moscow (1982).
3. R. Lloyd et al., Nucl. Technol., 42, 13 (1979).
4. S. M. Trunin et al., Reliability of Marine Machines and Mechanisms [in Russian],
Sudostroenie, Leningrad (1980).
5. F. Ya. Ovchinnikov et al., At. &erg., 50, No. 4, 248 (1981).
6. Proceedings of Seventh, Eighth, and Ninth Symposium on Natural Reliability and Quality
Control, Washington (1961-1963).
7. Proceedings of a Symposium an the Reliability of Nuclear Power Plants, IAEA, Vienna
(1975).
290
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.EMISSIONS OF "Co, 11?ThAg, AND 541111 AT THE ARMENIAN
NUCLEAR POWER PLANT, AND THEIR CONTENT IN THE
SURROUNDING ATMOSPHERE
E. S. SaSkov, A. A. Avetisyan, UDC 551.510.721.614.73
and K. I. Pyuskyulyan
Six years of operation of the Armenian nuclear power plant (NPP) with a VVgR-440 led to
an increase in the contribution of radioactive corrosion products, in particular "Co, 1"1'Ag,
and 54Mn, to the total activity of the gas?air mixture discharged into the stack (Table 1).
According to the data for January-September, 1982, these radionuclides contributed 57% of the
31
total activity of the long-lived aerosols discharged into the atmosphere, neglecting 11.
Consequently, the yearly emission of long-lived radioactive aerosols is due largely to ra-
dioactive corrosion products.
The "Co enters the coolant from abrasion resistant alloys containing this element
(pump bushings, control and safety rod actuators, etc.), and from austenitic stainless
4
steels. The S Mn comes from the same structural materials, and the 110Ag from the silver
in alloys used in the thermoelectric heater of the volume compensator and control devices.
Most of the radioactive corrosion products are deposited on the inner walls of the
primary loop system. As a result of changes in the coolant velocity and temperature and the
water-chemical condition during transient operation there is a change in the relative amounts
of corrosion products deposited on the walls of the equipment and in the coolant; during
shutdown the amount of ammonia and boric acid in the primary loop is increased. Thus, it
was determined that-with a drop in power the specific activities of "Co, 110mA8, and 54Mn
at the Kola NPP increased by factors of 50, 80, and 110 respectively. According to data in
[2], about 25% of the corrosion products in the coolant are in the form of large particles,
about 75% are colloid, and the ion fraction is less than 1%. The concentrations of the
nuclides mentioned reach a maximum in 't,46-48 h after a drop in power. Then the specific
activity decreases, and the coefficient of precipitation of radioactive corrosion products
becomes an order of magnitude, smaller ?than that obtained under steady reactor operation El].
In view of this, one should expect that during reactor startup and shutdown the emission
of radioactive corrosion products would be appreciably increased. Using the daily measure-
ments of the total activity of the discharge of long-lived aerosols during the period of
planned preventive maintenance on the first block of the Armenian NPP, we plotted a graph
(Fig. 1) of the time variation of the activity of the discharge. The figure shows that
during cooling, hydraulic testing, and reactor startup, the total activity of the discharge
increase sharply. The same conclusion can be drawn from spectrometric analysis (Table 2).
TABLE 1. Contribution of Radionuclides to
Total Activity of Discharge of Long-Lived
Aerosols, %
Period
"Co
I"mAg
'Mn
1918
2,4
1,9
1,7
January-September 1982
12,0
14,0
5,0
Translated from Atomnaya gnergiya, Vol. 56, pp. 278-280, May, 1984. Original article
submitted September 20, 1983.
0038-531X/84/5605-0291$08.50 01984 Plenum Publishing Corporation
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36
40 -
III i I i
27 1 S 9 13 17 21 2329 2 5 10
July August
Fig. 1. Total activity of discharge of
long-lived aerosols into the stack during:
I) cooling; II) unpacking main joint,
opening Volume compensator; III) refueling;
IV) hydraulic testing, output at power.
TABLE 2. Rate of Discharge of Radionuclides
in 1982, 107 Bq/month
Period
61410
".?mAg
541Wn
January
1,65
1,30
'0,69
February
3,43
.
? 3,20
. 1,30
March
2,60
1;42
0,84
April ?
20,80
6,75
7,22
May
7,45
7,22
? 2,37
June
5,21
7;58
? 1;78
J
4,00
1,85
13,02
August
4.02
2,96
? 0,91
September
. 1.66 .
1,54
0,71
TABLE 3. Concentration of Radionuclides
in Air At 1 km from NPP in 1982, 10-4
Bq/liter
Period, . ?
6IiCo
110MAg
-54Mn
1st quarter
0,62
-0,64
Traces
2nd quarter
1,40
3,00
1,10
July
.1,70
4,80
1,00
August
0,80
.0,80
Traces
:September
Traces
.
Traces
Traces
The increase in the discharge of 60co, ilomAg, and "Mn coincides with the periods of planned
preventive maintenance at the NPP. The time variations of the discharge of "Co and 5411a are
identical, since they have a common source. The behavior of 11?mAg is somewhat different;
In particular there was a sharp increase in the discharge of silver in July due to opening
the manhole of the volume compensator and repairing it during planned preventive maintenance
on the first block.
292
Air samples were taken at seven check points in various directions from the Armenian NPP
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TABLE 4. Comparison of Calculated and Mea-
sured Maximum Concentration of Radionuclides
During July, 1982, 10" Bq/liter
Radionuclide
6410
110mAg
542.1n'
Cale. values
1,10
3,60
0,51
Meas. values
1,70
4,80
1,00
and at distances of 1, 5, 6, 11, 14, 15, and 50 km from it. The spectrometric analysis was
performed by using a DGDK-32A-1 semiconductor detector of AI-4096 and UNO-4096-90 analyzers.
The minimum detectable concentration of aerosols estimated by the formula in [3] is 0.4x 10-5
Bq/liter with an error of 30% (Table 3). Only traces of radionuclides were found in samples
taken at 5 and 6 km, and none were found in the samples taken at the remaining points. The
results are confirmed by analysis of sedimentation samples. Samples of top soil showed only
insignificant amounts of 157Cs and 50Sr due to global fallout.
Evidently the recording of these radionuclides in the air near the Armenian NPP is due
to specific climatic conditions in the region. The climate of the Ararat valley is clearly
continental, with an arid summer and a short rarely cold winter. The atmospheric precipita-
tion is 250-300 mm. During the long calm summer with an abundance of sunny days, the air
is heated up to 40?C. The wind velocity in the industrial region of the Armenian NPP does
not exceed 3 m/sec. Under the strong solar heating of the earth's surface, the air temperature
decreases with height, so that the vertical temperature gradient is larger than adiabatic.
Such conditions lead to intense vertical displacement of the air. The maximum concentration
near the ground for the present stack height is relatively large, but it decreases rapidly
with distance from the source, depending on the wind direction.
The minimum value of Kd, the average monthly dilution ratio of the contaminant, was
calculated from metereological data obtained at stations in the NPP region, and the formula
recommended in [4]. The calculated value of Kd was 1.4 x 106 m5/sec. For comparison we
point out that the minimum yearly average dilution ratio for the center of our European
territory is 3.4 x 106 m5/sec [4].
The maximum ground level concentrations of "Co, 116mAg, and 541111 were found form the
measured values of their emission and the calculated value of Kd. Within the limits of
error, the calculated values agree with the spectrometric analysis of air samples taken at 1
km from the NPP (Table 4).
The data obtained justify the following conclusions and recommendations:
the possibility of recording 60co, 110mAg, and 561n in the air at 1 km from the Armenian
NPP for emission which does not differ from that of other NPP with VVgR, is due to the specific
climatic conditions at its location;
the emission of radioactive corrosion products during reactor shutdown and startup can
be reduced by using filters to retain the deposits, after loosening them and scraping them
off;
in determining the isotopic composition of the radionuclides in the layer of the atmos-
phere near the ground, and the radiation doses received by people living near NPP, it is
necessary to take account of the contribution of radioactive corrosion products.
LITERATURE CITED
1. V. B. Gall' et al., in: Radiation Safety and Shielding of NPP [in Russian], No. 5,
Atomizdat, Moscow (1981), p. 5.
2. I. K. Morozova et al., Removal and Deposition of Corrosion Products of Reactor Materials
[in Russian], Atomizdat, Moscow (1975).
3. S. M. Vakulovskii et al., in: Methodological Recommendations for Monitoring the Radio-
active Content of Objects in the Environment [in Russian], Moscow (1980), p. 234.
4. N. E. Artemova et al., Admissible Emissions of Radioactive and Harmful Chemicals into
the Ground Layer of the Atmosphere [in Russian], Atomizdat, Moscow (1980).
293
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UNLOADING ADDITIONAL ABSORBERS FROM THE RBMK -1000. CORE
N. V. Isaev, V. E. Druzhinin, UDC 621.039.539.1
and Yu. V. Shmonin
In the initial loading of the RBMK-1000 reactor, the considerable reactivity margin is
reduced by loading the channels only incompletely with fuel assemblies (FA), absorbers being
inserted in the unfilled channels. The core then contains 234 additional absorbers (AA) and
44 unloaded channels containing water columns (WC).
The RBMK-1000 design envisages a strategy for reloading the channels by dividing them
up into periodicity cells (polycells) of size 4x 4 cells each [1]. One channel in each
polycell is reloaded, and to maintain the loading symmetry and periodicity, one unloads
channels identically placed in the polycell. This method has been used in reloading the
reactors at Kursk nuclear power station.
Here we examine the effects of AA unloading order on the economic parameters of the
second-generation RBMK-1000 reactors in the steady state. As the safety state is attained
at 7-8 years after commissioning, the optimum AA unloading strategy in the transition period
can be examined only on theoretical models. The unloading of the. AA involves about 670-700
effective working days.
Model Description. The REF program [2] is used in calculating the transient state in
the RBMK and envisages a specially homogenized core. This program is widely used in de-
termining RBMK characteristics, so the algorithm is not described here.
The OPERA program for optimizing apparatus reloading [3] enables one to perform a
detailed full-scale two-group diffusion calculation on the RBMK parameters under conditions
of partial and ongoing reloading, including the adjustment of the power production field by
varying the insertion lengths of the control and protection rods (CPR). The OPERA program
is based on the OPTIMA one [4]. A difference from the OPTIMA algorithm, which is described
in [4], is that the OPERA program uses an improved algorithm that instead of using the CPR
to compensate the reactivity and equalize the power production (while retaining kef) enables
one to bring the reactor to a given kef by automatic selection of the necessary rods in the
CPR set. In simulating RBMK reloading, the power production field is profiled to a certain
set field WTeg. The following constraints must be met in using the OPERA program:
1. The insertion depth hi of the CPR is in the range
0 Uo.
The mechanism of the rise in efficiency with increasing Um has not been ascertained as yet.
For comparison Fig. 3 gives the values found for the efficiency from the formulas presented
by Timofeev [1].
Thus, the following conclusions can be made. For the conversion of the energy of plasma
fluxes containing particles of both signs with a small longitudinal velocity component it is
expeditious to use the recuperator described by Timofeev. But for nonmonoenergetic beams of
Charged particles of the same sign with an arbitrary ratio cn /6,I the recuperator presented in
this paper is more efficient (11:-.30.7 for e4 /el.:0.9 and RA 0.7). Inorder to increase its
efficiency, it is necessary to increase the size of the apparatus, i.e., decrease the ratio Ro/R.
The recuperator discussed here can be used in the neutral-particle injectors of thermo-
nuclear machines, microwave devices (traveling-wave tubes, klystrons), electron-beam valves,
etc.
LITERATURE CITED
1. A. V. Timofeev, Fiz. Plazmy, 4, No. 4, 826 (1978).
313
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Declassified and Approved For Release 2013/09/14: CIA-RDP10-02196R000300040005-3
CALCULATION OF MODEL HIGH-LEVEL WASTES IN A HORIZONTAL
APPARATUS
V. V. Kulichenko, V. F. Saverev, UDC 621.039.7.14
V. A. Prokhodtsev, and A. A. Ryabova
In the processing of radioactive wastes by calcination [1], the wastes are dried and
the salts are thermally decomposed, the final product consisting mainly of oxides. As the
wastes contain many different compounds, a wide temperature range is used: first the nitrates
of the rare-earth and other trivalent elements decompose, then the nitrates of the alkaline-
earth metals, and finally at the highest temperatures and with the lowest rates the same
applies to the alkali-metal nitrates. Calcination is considered as the first stage before
vitrification, the production of ceramics, etc.
Various calcination methods have been devised based on equipment of sprayer type,
fluidized beds, and of horizontal type with a rotor or rotating body [2-4]. A horizontal
plant is much less sensitive to the waste composition than is a fluidized-bed or sprayer
one, and the volume of the gas discharged is determined only by the vapor-gas mixture formed
during the drying and calcination. However, the design is more complicated, since there are
moving mechanisms in the rotor (worm conveyer) or body. Also, the calcination is conducted
in a thin layer and heat must be supplied through the wall of the apparatus, which makes it
difficult to design a horizontal plant of large throughput.
A horizontal calcinator of throughput 40 liter/h [3-4] has been operated for several
years in France in a plant for vitrifying high-level wastes. The calcinator consists of a
heated tube of diameter 27 cm and length 3.6 in rotating at 30 rpm, which is tilted at 1047.
Within it there is a free rod to grind up the solid product. The liquid wastes contain
mainly the nitrates of aluminum (up to 81 g/liter), sodium (19-23 g/liter), and iron (15-17
g/liter) together with other salts at much lower concentrations. The product from any
apparatus must be as dry as possible and contain the minimum amount of nitrate (source of
corrosive oxides of nitrogen) and should flow freely, in order to provide for transfer to the
vitrification apparatus.
We have examined the calcination conditions in a horizontal plant operating with wastes
having elevated sodium nitrate contents.
The drying and calcination were performed in a horizontal tubular apparatus containing
a rotating worm conveyer to mix and transport the product through the working zone into the
bunker (Fig. 1). The length of the working zone was 1400 mm, internal diameter 160 mm, gap
between projections on the blades of the worm conveyer and the inner wall of the apparatus
about 2.5 mm. The apparatus was heated by six external demountable heaters of power 2.5 kW
each. The temperatures in the sections of the working zone were 250-800?C in accordance
with the required conditions.
The nitrate solutions were supplied by a dispensing pump to the input, where they were
dodatered, denitrated, and calcinatedas theymoved thorughtheworkingzone. The solid product
in the form of powder or granules was transmitted by the worm conveyer to the bunker. The
steau gas mixture containing oxides of nitrogen passed successively through a filter (to
remove aerosols), a cooling condenser (to collect condensate), and to a bubble tower con-
taining alkaline solution to neutralize the oxides of nitrogen.
The final product was analyzed for nitrogen, water, total carbon, carbon in the form
of carbonate, and iron, and some mechanical properties were determined. The specific surface
was measured by thermal nitrogen desorption; the data curves were recorded with an OD-103
instrument by the standard method. Table 1 gives data on the calcination of model solutions
of various compositions in the presence of molasses and without them at temperatures below
800?C with throughputs of 4.0-17 liter/h and time spent by the product in the apparatus of
2 and 4 min.
Translated from Atomnaya gnergiya, Vol. 56, No. 5, pp. 293-.297, May, 1984. Original
article submitted August 18, 1983.
314 0038-531X/84/5605-0314$08.50 1984 Plenum Publishing Corporation
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Declassified and Approved For Release 2013/09/14: CIA-RDP10-02196R000300040005-3
0
1==
0
0. As x2 + x3 increases the value of R grows insignificantly
and the decreases for all the regimes considered, except 0=10 neutrons/cm2..sec and y = 0.4,
its value being R< 1 at the maximum of x2 + x3 (this is attributed to the contribution of
neutrons formed in the fission of 233U). The value of R increase substantially with the
growth of 0 and y and the weakening of the blocking since in this case the 233Pa content rises
and the 233U fraction decreases correspondingly and together with it, the contribution of
233U fission neutrons. This effect manifests itself most clearly in the regime with 0=1014
neutrons/cm2osec and y = 0.4: Thenumber of 233U fission neutrons is insufficient to compensate
the external neutrons absorbed and R> 1 for all values of x2 + x3 (we note that in this case
R increases as the blocking is intensified).
Energy Release E. During the fission of the 233U formed energy is released in the 232Th
charged. In Fig. 6 we give the dependence of E on x2 +x3 per ton of the initial 232Th.
The energy released in one 232Th fission event was assumed to be 191 MeV [7]. For a fixed
x2 + x3 the energy release E decreases as and y increase and the blocking becomes weaker;
this is explained by the lowering of the fraction of 233U in the mixture of 233Pa and 2331J.
Clearly, the energy release is proportional to the number of neutrons produced during the
fission of 233U.
Conclusion. The results of the calculations are of interest from two points of view:
First, the main physical characteristics of the 232Th 233U process are presented as a func-
tion of the flux density and the energy spectrum of thethermal neutrons, making it possible
to evaluate the efficiency of one reactor design or other; second, the decrease in the
232Th consumption and the increase in the neutral consumption R as y grows give reason to
hope that the problem of economic optimization for the 232Th 233U process will have a
nontrivial solution.
LITERATURE CITED
1.
2.
V. M. Murogov,}4. F. Troyanov,
[in Russian], Energoatomizdat,
A. S. Gerasimov, A. K. Kruglov,
and A.
Moscow
and A.
N. Shmelev, Use of Thorium in Nuclear Reactors
(1983).
P. Rudik, At. nerg., 51, No. 4, 237 (1981).
356
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Declassified and Approved For Release 2013/09/14: CIA-RDP10-02196R000300040005-3
3. S. Mughabhab and D. Garber, Neutron Cross Sections, BNL-325, 3rd ed. (1973), Vol. 1.
4. G. N. Gusev and P. P. Dmitriev, Radioative Chains (Handbook) [in Russian], Atomizdat,
Moscow (1978).
5. L. V. Matveev and g. M. Tsenter, At. Tekh. Rubezhom, No. 4, 10 (1980).
6. T. S. Zaritskaya et al., At. gnerg., 48, No. 2, 67 (1980).
7. I. V. Gordeev, D. A. Kardashev, and A. V. Malyshev, Nuclear-Physical Constants
Russian], Gosatomizdat, Moscow (1963):
SPECTRAL-ANGULAR DISTRIBUTION OF y RADIATION BEHIND
AN INSTRUMENTATION UNIT
P. A. Barsov, V. M. Sakharov, UDC 621.039.538.7
and V. G. Semenov
The weight fraction of the various equipment and devices is appreciable on transporting
apparatuses with nuclear reactors; therefore taking correct account of the shielding proper-
ties of these items is rather important. The possibility of using an effective attenuation
coefficient to estimate the shielding properties of the equipment, which has been modeled
in such a way that a binomial distribution of the units of material was realized, has been
verified experimentally in [1, 2]. It has been shown that the dose is attenuated exponential-
ly along the axis of a wide radiation beam with an effective coefficient determined by the
binomial law.
The possibility of using an effective attenuation coefficient in calculations by the
Monte Carlo method of the differential characteristics of y radiation behind actual equip-
ment units is discussed in this paper. With this goal we have performed experiments and cal-
culations in the identical geometry.
In the experiments monoenergetic y radiation of a point isotropic 137Cs source with a
photon energy Ey = 662 keV was directed at a rectangular equipment unit with a geometrical
thickness at the irradiation site of Z = 53 cm and an average density of p = 0.43 g/cm3.
We used a scintillation spectrometer with a CsI(T1) crystal 16 mm in diameter and 40 mm in
height placed in a conical collimator as the le-radiation detector. The isotropy of the de-
tector is no better than ?5% within the limits of the collimation angle. The distance be-
tween the equipment unit and the detector was selected so that all the y radiation emergent
from the unit in the direction of the detector was recorded. Measurements were made for
angles of 0, 30, 60, and 90? to the inner normal relative to the irradiated surface. The
error of the instrumental spectra was q,30% in the region of Ey ' 130-250 keV and 15-20% for
higher values of the energy. It did not prove possible to make measurements for Ey < 130 keV
due to the high background.
The calculations were performed by the Monte Carlo method using the BSERAD program [3],
whose geometrical section simulated the geometry of the source, detector, and equipment unit
made out of aluminum. The effective attenuation coefficient of the unscattered y radiation
used to determine the mean free path of a photon by simulating the trajectory was found using
the formula [4]:
Translated from Atomnaya gnergiya, Vol. 56, No. 5, pp. 322-323, May, 1984. Original
article submitted October 12, 1983.
0038-531X/84/5605-0357$08.50 ?1984 Plenum Publishing Corporation 357
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Declassified and Approved ForRelease2013/09/14 :
.a
--, c...) 1
' 60 - I 1s
a c 1
ON ....-? ?.
0
0
cl?P+0-. 10
m
41':1.i..- r.
v o 41
?-
s..a .4
- a
-0 a?0 a. a
J `1
r? ...,
L. n
..
? I 1 b
.., 0
CIA-RDP10-02196R000300040005-3
5"E
a
1810
150
270 2/.7 bV
E keV
Fig. 1. Calculated and experimental instrumental spec-
tra of the y radiation behind an equipment unit at
angles to the beam axis of (a) 6 = 0, (b) 60, (c) 30,
and (d) 900 with a photon energy of 662 keV. ---) ex-
periment; - - - -) calculation for the quipment unit;
) calculation for a layer of aluminum 7.2 cm thick
with a density of 2.7 g/cm3.
Iii 11.i 0yAlc,, 714,
(1)
The
The distribution function of the material thickness f(x, Z) was obtained experimentally
by the method of y transillumination of the unit in different directions [5]. We used the
sensitivity matrix determined experimentally as the sensitivity function of the detector in
the calculations. The error of the spectra calculated for the indicated angles does not ex-
ceed 15% over the entire energy range.
The calculated and experimental instrumental spectra normalized per photon emitted from
the source in the direction of the equipment unit are similar in shape for all detection
angles (see Fig. 1). A certain systematic understating of the computational results in com-
parison with the experiment;11 results in the region of the single-scattering peak is evidently
associated with the use of an attenuation coefficient for aluminum, whereas the unit being
investigated contains an appreciable amount of lighter elements (hydrogen, carbon, oxygen).
The total mass attenuation coefficient for these elements is smaller than for aluminum, which
leads to the indicated difference of 20-307..
The dependence of the y-radiation intensity on the recording angle is of an exponential
nature. The value of the characteristic angle 6 = 47? turned out to be close to its value
for a homogeneous layer of aluminum and a point isotropic 137Cs source [6]. This fact served
as the reason for doing similar experiments and calculations for a homogeneous layer of alu-
minum with PM, = 2.7 g/cm3 and Z = 7.2 cm, which corresponds to an optical thickness of the
equipment unit, which is defined as the product of the effective attenuation coefficient by
the geometrical thickness of the unit in the direction of the incident beam of y-photons.
The thickness of the layer, which is calculated as pZ/PA1, is equal to 8.4 cm.
The experiments and calculations for a layer of aluminum 7.2 cm thick have shown that
the spectral-angular distributions behind the layer and behind the equipment unit are prac-
tically identical. At the same time the calculations for a layer of aluminum 8.4 cm thick
give results which are understated by a factor of 1.5 in comparison with those given for
the equipment unit over the entire energy range.
358
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Declassified and Approved For Release 2013/09/14: CIA-RDP10-02196R000300040005-3
The calculations performed for an optical thickness of the unit of px = 4 have shown
that the use of a homogeneous layer with thickness peffx leads to an understating by a factor
of 1.5 of the flux of y-photons in the region Ey > 400 keV and to some softening of the spec-
trum in contrast to the direct calculations, in which the distance between successive in-
teractions is determined with the use of veff.
The investigations carried out have led to the conclusion that the passage of y radiation
in equipment with a thickness Z < 30 g/cm2 (px,2) can be calculated as in a homogeneous mate-
rial with the use of the usual interaction constants, determining the optical thickness of
the layer with the help of the effective attenuation coefficient. For equipment of greater
thickness it is necessary to use the effective attenuation coefficient. The latter asser-
tion requires further experimental verification however.
LITERATURE CITED
1. V. V. Bolyatko et al. in: Problems of Dosimetry and Shielding from Radiation [in Rus-
sian], No. 8, Atomizdat, Moscow (1968), p. 80.
2. P. A. Barsov et al., in: Abstracts of Lectures at the 2nd All-Union Scientific Conference
on Shielding from Ionizing Radiations of Nuclear Engineering Facilities [in Russian],
Atomizdat, Moscow (1978), p. 40.
3. P. A. Barsov et al., in: Radiation Safety of Nuclear Power Plants [in Russian], No.
26, Moscow (1979), p. 75.
4. A. V. Kolomenskii et al. At. Energ., 44, No. 6, 517 (1978).
5. V. V. Bodin et al., in: Abstracts of Lectures at the 2nd All-Union Scientifiec Con-
ference on Shielding from Ionizing Radiations of Nuclear Engineering Facilities [in
Russian], Atomizdat, Moscow (1978), p. 103.
6. Yu. A. Kazanskii et al., Physical Investigations of Reactor Shielding [in Russian],
Atomizdat, Moscow (1966).
359
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Declassified and Approved For Release 2013/09/14: CIA-RDP10-02196R000300040005-3
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