SOVIET ATOMIC ENERGY VOL. 54, NO. 4

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• Collection: General CIA Records
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• Document Page Count: 84
• Document Creation Date: December 27, 2016
• Document Release Date: February 6, 2013
• Sequence Number: 4
• Publication Date: April 1, 1983
• Content Type: REPORT

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October, 1983
SATEAZ,54(4) 243-318 (1983)
Russian Original Vol. 54, No. 4, April, 1983
SOVIET
ATOMIC
ENERGY
ATOMHAA 3HEPIVIH
(ATOMNAYA ENERGIYA)
TRANSLATED FROM RUSSIAN
CONSULTANTS UUREAU, NEW YUKK
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SOVIET
ATOMIC
ENERGY-
Soviet Atomic Energy is abstracted or in-
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Editorial Board of Atomnaya Energiya:
Editor: 0. D. KazaChkovskii
Associate Editors: N. A. Vlasov and N. N. Ponomarev-Stepnoi
Secretary: A. I. Artemov
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M. F. Troyanov
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SOVIET ATOMIC ENERGY
A translation of Atomnaya Energiya
October, 1983
Volume 54, Number 4 April, 1983
CONTENTS
EXPERIENCE ACCUMULATED BY SOVIET NUCLEAR POWER ENGINEERING
Nuclear Power in the USSR - A. P. Aleksandrov, A. S. Kochenov,
E. V. Kulov, A. G. Meshkov, V. P. Ryazantsev, and V. A. Sidorenko.
Experience with the Creation, Operation, and Means of Improvement
of Nuclear Power Plants with Water-Cooled-Water-Moderated Reactors
VVER) - F. Ya. Ovchinnikov, Yu. V. Markov, V. A. Sidorenko,
V. A. Voznesenskii, G. L. Lunin, V. V. Stekol'nikov,
G. G. Bessalov, and Yu. V. Vikhorev.
Some Characteristics of and Experience with the Operation of Nuclear
Power Plants with RBMK-1000 High-Powered Water-Cooled Channel
Reactors (RBMK) - N. A. Dollezhal', I. Ya. Emel'yanov,
Yu. M. Cherkashov, V. P. Vasilevskii, L. N. Podlazov, V. V. Postnikov,
A. P. Sirotkin, V. P. Kevrolev, and A. Ya. Kramerov. . . . . . . . .
Development and Experience of Operating Fast Reactors in the Soviet Union
- 0. D. Kazachkovskii, A. G. Meshkov, F. M. Milenkov, V. P. Nevskii,
L.
A.
Kochetov,
V.
I.
Kupnyi, B. I. Lukasevich, V. M. Malyshev,
V.
V.
Pakhomov,
F.
G.
Reshetnikov, A. A. Samrkin, M. F. Troyanov,
V.
I.
Sh.iryaev,
V.
A.
Tsykanov, and D. S. Yurchenko. . . . . . . . .
Paths for the Development of Fast Power Reactors with a High Breeding
Factor - S. B. Bobrov, A. V. Danilychev, V. A. Eliseev,
0. A. Zhukova, Yu. A. Zverkov, V. G. Ilyunin, V. P. Matveev,
A. G. Morozov, V. M. Myrogov, A. I. Novozhilov, V. V. Orlov,
I. S. Slesarev, S. A. Subbotin, M. F. Troyanov, and B. F. Shafrygin.
Standards for Safety of Atomic Power Plants in the USSR
- V. A. Sidorenko, 0. M. Kovalevich, and A. N. Isaev . . . . .
Radiation Safety of Atomic Power Plants in the USSR - E. I. Vorob'ev,
L. A. Il'in, V. D. Turovskii, L. A. Buldakov, N. G. Gusev,
0. A. Pavlovskii, and G. M. Parkhomenko . . . . . . . . . . . . . . .
Extraction and Processing of Uranium Ore in the USSR - B. N. Laskorin,
V. A. Mamilov, Yu. A. Koreisho, D. I. Skorovarov, L. I. Vodolazov,
I. P. Smirnov, 0. L. Kedrovskii, V. P. Shulika, B. V. Nevskii,
and V. N. Mosinets . . . . . . . . . . . . . . . . . . . . . . .
Experience in Handling Spent Fuel from Nuclear Power Stations
in the Soviet Union, Including Storage and Transportation
- V. M. Dubrovskii, V. I. Zemlyanyukhin, A. N. Kondrat'ev,
Yu. A. Kosarev, L. N. Lazarev, R. I. Lyubtsev, E. I. Mikerin,
B. V. Nikipelov, A. S. Nikiforov, V. M. Sedov, B. I. Snaginskii,
and V. S. Shmidt . . . . . . . . . . . . . . . . . . . . . . . . . .
Engl./Russ.
243
243
251
249
263
257
270
262
279
269
285
273
290
277
301
286
309
293
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Engl./Russ.
Problems of Radiation Safety of Atomic Power Plant Personnel and the Public
- E. I. Vorob'ev and 0. A. Pavlovskii. . . . . . . . . . . . . . . 314 303
The Russian press date (podpisano k pechati) of this issue was 3/ 24/ 1983.
Publication therefore did not occur prior to this date, but must be assumed
to have taken place reasonably soon thereafter.
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EXPERIENCE ACCUMULATED BY SOVIET NUCLEAR
POWER ENGINEERING
A. P. Aleksandrov, A. S. Kochenov, E. V. Kulov,
A. G. Meshkov, V. P. Ryazantsev,
and V. A. Sidorenko
The Soviet Union has no equal as regards reserves of organic fuel. The total reserves
of coal in the USSR are 5.5-6 x 1012 tons and constitute about half of world reserves. In
1981, the volumes of organic fuel production were as follows: oil (including gas condensate)
609 million tons (first place in the world), gas 465 billion m3 (second place in the world),
and coal 704 million tons (second place in the world).
The development of power engineering in the USSR is supported by our own resources.. Also,
oil and gas are exported to countries in Eastern and Western Europe. On the other hand, most
of the organic fuel reserves are located in the Asiatic part of the country, while four-
fifths of the fuel demand lies in the European part. For this reason, the shipping of fuel
from eastern regions into western ones constitutes about 40% of the total goods carried by
the country's railroads.
In recent years, the Asiatic part has provided almost all the increase in the extraction
of organic fuel. Therefore, if power engineering developments were to be based only on the
use of organic fuel, there would be an increasing disproportion in the location of the extrac-
tion and use of fuel. There would be also an appreciable increase in the costs for trans-
porting fuel resources to the European part of the country. This disproportion can be sub-
stantially relieved by developments in nuclear power.. Also, nuclear power enables a reduc-
tion in the cost of producing electricity in the European part, while relieving railroads
from transport load and improving labor productivity mainly by reducing the number of workers
required in the extraction industry and in transportation, while also modifying the fuel and
power balance by reducing the proportion accounted for by oil and gas.*
Oil and gas cannot remain in the basis of the world's fuel and power for long, since re-
serves are limited, and therefore oil and gas should be considered primarily as valuable
chemical raw materials and eliminated as far as possible from the fuel balance [1].
There are also difficulties with other traditional forms of fuel in many parts of the
world. On the other hand, the increasing nuclear power during the 1970s was much less than
was forecast at the end of the 1960s. Major factors that hold back development at the pres-
ent time are the following: inadequate development in specialized engineering, the. lack
of a practical solution to fuel breeding, delay in certain aspects of fuel reprocessing and
in storage of high-activity wastes, and finally lack of preparation in public opinion. On
the other hand, the efforts being made in certain countries lead one to believe that all
these problems will be resolved in the next 10-15 years
In the Soviet Union, the first designs for nuclear power stations began to be devised
at the end of the 1940s. In 1950 it was decided to construct the first nuclear power station
at Obninsk on the basis of a channel uranium-graphite reactor. This was commissioned on
June 27, 1954. Operating experience showed it was reliable and safe for the staff and for
the surrounding population. This gave a clear demonstration that nuclear power can be used-
to produce electricity.
The State program for the development of nuclear power in the USSR could not be based
on nuclear power stations of a single type, since this would not provide the necessary re-
liability and stability. On the other hand, the exploitation of each type of power reactor
*
In this section the editors publish in journal form some of the papers presented by Soviet
researchers at the IAEA International Conference on Experience Accumulated in Nuclear Power
(Vienna, September 13-17, 1982). In addition to the papers, the editors print two surveys
of the current state of foreign nuclear power engineering.
Translated from Atomnaya Energiya, Vol. 54, No. 4, pp. 243-249, April, 1983.
0038-531X/83/5404-0243$07.50 ? 1983 Plenum Publishing Corporation 243
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on a commercial scale requires a considerable time interval and considerable financial and
material resources.
A scientific program was drawn up by the State Committee on the Use of Atomic Energy in
the USSR to define the most suitable and economical effective power reactors for this country.
The program envisaged research on various power reactors: pressurized-water and boiling-
water ones, channel uranium-graphite ones, reactors with organic moderators and coolants,
etc. Much attention was given to fast reactors, since from the start it was clear that large-
scale nuclear power is inconceivable without them.
During the research, some forms of power reactor were not brought to the stage'of pro-
totype building for various reasons. It was found, e.g., that reactors with organic modera-
tors and coolants are suitable only for low-power stations. Research in this area led to
the creation in 1963 of a block-transportable nuclear power station of electrical output 750
kW. On the other hand, research on certain types of thermal power reactor have led to the
building of prototype commercial units. A water-cooled and water-moderated reactor (the
VVER-210) was built at Novyi Voronezh nuclear power station with an electrical output of 210
MW, while a pressurized-water reactor (the VK-50) was built at Dmitrovgrad with electrical out-
put 50 MW, and at Beloyarsk nuclear power station a channel uranium-graphite reactor was
built (with nuclear steam superheating and an efficiency of 36%).
In the course of the fast-reactor program, several experimental reactors and test facil-
ities were constructed for simulating and researching the physical and engineering charac-
teristics. The first experimental reactor with plutonium fuel was built in 1955. In the
later reactors, the fuel has been metallic plutonium, plutonium dioxide, uranium monocarbide,
and uranium dioxide. The main attention has been given to designs with sodium cooling. The
power outputs of the experimental reactors have gradually increased. In 1969, an experimental
reactor was commissioned at Dmitrovgrad that employed fast neutrons with sodium cooling - the
BOR-60 (electrical power'12 MW).
The types of power reactor optimum for this country were defined during the development,
construction, and operation of these prototype units. During the second half of the 1960s,
the accumulated experience led to the decision to develop nuclear power engineering on the
basis of two types of thermal reactor: pressurized-water ones and channel uranium-graphite
ones cooled by boiling water, since at that time several VVER units had been built together
with the second unit for the uranium-graphite channel reactor at Beloyarsk power station.
The decision provided, firstly, the fullest use of the country's engineering facilities and,
secondly, a more flexible design for the fuel cycle, since in particular the RBMK reactors
can utilize fuel from the VVER. The VVER-440 and RBMK-1000 units were adopted as the standard
ones.
The first two units at Novyi Voronezh power station were prototypes for nuclear power
station units containing VVER-440 [2]. The units are highly reliable and the load factors
TABLE 1. Working Parameters of Nuclear
Power Stations with VVER Units in Individ-
ual Year
Power station
Year
Installed
power,
MW
Energy
produc-
tion,
billion
kW?h
Load
coeffi-
cient,
Novyi Voronezh
1979
1409
9,92
78,9
1980
1409
11,35
82,7
1981
2409
15,56
73,7
Kola
1979
880
5,90
76,5
1980
880
7,22
93,5
1981
880/1320 *
7,39
69/83 *
Armenian
1979
407,5
2,39
66,8
1980
815
4,74
66,3
1981
815
5,50
77,0
*The denominator gives the value without al-
lowance for the introduction of the new unit.
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are high (about 80%, and appreciably higher in some years), which provided a basis for the
design of routine units. In the development of the VVER-440, the equipment was substantially
upgraded, and major changes were made in the reactor design. The first two units containing
the VVER-440 were also installed at Novyi Voronezh power station. Minor changes were made
to the designs of the subsequent standard units. Subsequently, similar units were introduced
at other nuclear power stations in the country (Kola, Armenian, and Rovensk), as well as in
the German Democratic Republic, Bulgaria, Czechoslovakia, and Finland. At the Armenian power
station, the equipment in the first loop was designed to withstand seismic shocks.
Table 1 gives some parameters of nuclear power stations containing VVER-440. The high
load coefficients are due to the successful design of the main unit.
The length of a fuel cycle in the VVER-440 is 3 yrs, with intervals of 1 yr between par-
tial reloads. The annual partial reload along with planned prophylactic maintenance occupies
about 30 days. Periodic checks are made on the state of the main equipment in the first loop,
and, in particular, the pressure-vessel metal is checked once every 4 yrs with the core and
devices within the pressure vessel unloaded. The length of the shutdown necessary for checking
the pressure-vessel metal is about 60 days. Some VVER vessels have been made of heat-re-
sisting steel without anticorrosion stainless-steel coating. The first such vessel was
installed in the second unit at Novyi Voronezh power station. The ammonia-potassium water
treatment provides a satisfactory corrosion state in the pressure-vessel metal.
The work on upgrading the VVER-440 led to the creation of the VVER-1000 reactor system,
of which the first unit was commissioned at Novyi Voronezh power station. The following
problems were overcome in the design of this: The equipment must be transportable by rail-
road, the economic parameters should be improved, and the units should meet the latest safety
requirements.
Transportability for the pressure vessel restricted the diameter to 450-460 cm, with
the diameter of the core zone, correspondingly, 310-320 cm. The energy production density
in the core is 110 kW/liter or 30% higher than that in the VVER-440. This requires special
measures to equalize the distribution. The main circulation loop was enclosed in a protec-
tive shield of prestressed reinforced concrete designed for a pressure of 0.55 MPa. Popula-
tion safety was also provided for instantaneous failure in DU-850 pipeline coinciding in time
with the completely idle state of the power station.
Commissioning operations and operating experience with a fifth unit at Novyi Voronezh
power station confirmed that the basic designs were correct. On the other hand, certain
changes were made in the design of the VVER-1000 to be located at the South Ukraine, Kalinin,
and Rovensk power stations. Projects for power stations containing VVER-1000 were devised
for regions with seismicity 5-6 points on the MSK 1964 scale. No substantial changes are
proposed in the VVER-1000 reactor system for nuclear combined heat and power stations.
Table 2 [3] gives some parameters of nuclear power stations containing RBMK-1000 units.
At such power stations, the fuel is changed on load by a charging and discharging machine.
Experience has shown that about 1000-1500 such operations can be performed at such a power
TABLE 2. Working Parameters of Nuclear
Power Stations with RBMK-1000 Units in In-
dividual Years
Energy Load
Installed produc- coeffi-
Power station
Year
power, lion, cient,
MW billion a/b
kW ?h
Leningrad
1979
2000
13.1
711
1980
3000
18.82
71,4
1981
4000
24,1
73,8
Kursk
1979
2000
10,35
64.1
1980
2000
13,89
79,1
1981
2000
13,54
77,3
Chernoby
1979
2000
12,23
69,8
1980
2000
14,21
80,9
1981
2000
13,44
75,2
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station during a calendar year. The condition of continuous reloading enables one to approxi-
mately double the burnup by comparison with the state of simultaneous reloading for the entire
core.
The experience with the RBMK-1000 confirmed preliminary theoretical conclusions that the
void and temperature coefficients of reactivity would increase as the fuel is burned and ab-
sorbers are extracted from the core, which reduces the stability of the energy production
distributions. Efforts to stabilize these distributions were concentrated on improving the
level of automation by means of a branched controlled system for the reactor and changing
the fuel isotope composition. A new local automatic-control system was devised and imple-
mented, which showed high reliability and performance. This provided characteristic deforma-
tion times for the energy-production patterns of not less than 6-8 h, which do not cause any
difficulty in managing the reactor.
In accordance with these studies, the initial fuel enrichment was raised to 2%. This
not only improved the dynamic performance of the reactor, but also raised the economic pa-
rameters by increasing the extent of burnup and reducing the specific fuel consumption.
On account of elevated safety specifications, special systems were designed to provide
acceptable temperature conditions in the fuel pins and to localize the escape of coolant on
failure of any pipeline (including a maximum diameter one of 900 mm). Such safety systems
have already been implemented at the Leningrad power station and are envisaged for all power sta-
tions that are being constructed with RBMK-1000 units.
The successful operation of the RBMK-1000 at its nominal power revealed considerable
reserve margins in the design, so it was possible to incorporate heat-transfer intensifiers
in the fuel-pin assemblies to increase the power in each channel by a factor 1.5 without
changing the dimensions or numbers of the fuel channels. The design of the new fuel-pin
assembly for the reactor, namely RBMK-1500, enables one to increase the thermal loads while
maintaining a high level of unification with the RBMK-1000 fuel-pin assemblies.
The first section of the Ignala power station with two RBMK-1500 units is now being
built. The commissioning of the head unit will lay the basis for the new generation of channel
reactors, which are more economical and should replace the well-recommended RBMK-1000. The
construction of power stations containing the RBMK-1500 will reduce the specific capital cost
by comparison with the RBMK-1000 and will reduce the costs assigned to the electricity.
In the next stage of development for channel power reactors, one may go to the design
of sectional block reactors with nuclear steam superheating (RBMKP) with unit powers of 1200
and 2400 MW. The gross efficiency of a nuclear power station containing RBMKP-1200 and-2400
units is expected to be about 37%.
The building of nuclear power stations containing thermal reactors has been accompanied
by the construction of two commercial units containing fast reactors: the BN-350 and the
BN-600 [4]. Over nine years have elapsed since the power commissioning of the BN-350. The only
major defect in the equipment has been repeated failure of the sealing between loops in the
steam generators. After the completion of repair in the damaged steam generators, the reac-
tor power was raised to 520 MW (thermal) in 1975, 650 in March, 1976, and 700 in September,
1980, which provides an electrical power of 125 MW and a daily production of 85 x 103 tons of
distillate. The design burnup of 5% of heavy atoms was attained in 1976. At present, the
burnup is being maintained at the level of 5.8% of the heavy atoms, which is related to the
acceptable shape changes in the six-faced jackets of the fuel-pin assemblies. During the
operation, the fuel pins were unified on outside diameter and sheath thickness with the pins
in the BN-600. The resulting increase in the gas compensating volume reduced the pressure
in the sheath, which reduced the number of cases of sheath failure by an order of magnitude.
The BN-600 power unit differs from the BN-350 in having and integral (tank) style for
the equipment in the first cooling loop. Steam generator of modular type also improve the
reliability. The run-up to power began in April, 1980, and by December 18, 1981, the reactor
was brought up to the nominal power of 1470 MW (thermal). On January 1, 1982, the unit con-
taining the BN-600 had produced 3.7 billion kW-h, and it had operated at power for over 104 h.
The reactor is readily controlled. Four fuel changes have been made during the operation.
The maximum fuel burnup attained 7% of the heavy atoms. On the whole, the operating experi-
ence has confirmed that the actual parameters correspond to the design ones.
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TABLE 3. Major Operating Nuclear Power
Stations in the USSR
Dates of major
stages
Power
Electrical
power of units,
Reac-
tor
physi
ener-
gy
re-
ache
station
MW
type
cal
com-
nom-
coxn-
mis-
inal
mis-
Sion-
power
pion-
ing
ing
Novyi
1st unit - 210
VVER
12.63
09.64
12.64
Voronezh
2nd ? - 365
?
12.69
12.69
04.70
3rd ? - 417
?
12.71
12.71
06.72
4th ? - 417
?
12.72
12.72
03.73
5th -1000
?
04.80
05.80
02.81
Beloyarsk
1st ? - 100
UGR
09.63
04.64
09.67
2nd - 200
10.67
12.67
12.69
3rd ? - 600
13N
02.80
04.80
12.81
Kola
1st ? -- 440
VVER
06.73
06.73
12.73
2nd ? - 440
?
11.74
12.74
02.75
3rd ? - 440
02.81
03.81
Leningrad
1st 1000
RBMK
09.73
12.73
11.74
2nd ? - 1.000
?
05.75
07.75
01.76
3rd ? --1000
09.79
12.79
06.80
4th ? -- 1000
12.80
02.81
08.81
Armenian
1st ,> - 407,5
VVER
12.76
12.76
10.79
2nd ? - 407,5
>
01.80
01.80
05.80
Kursk
1st ? -.1000
RBMK
09.76
12.76
10.77
2nd ,> - 1000
?
12.78
01.79
08.79
Chernobyl'
1st ? -1000
RBMK
08.77
09.77
05.78
2nd 1000
09.78
92.78
05.79
3rd 1000
06.81
12.81
06.82
Rovensk
1st ? -- 440
VVER
12.80
12.80
2nd ? - 440
12.81
12.81
The building and commissioning of nuclear power station units, containing BN-350 and
BN-600 is an important stage in solving the problem of fuel breeding, whose final purpose
will include the design of a standard fast-reactor unit to be built on a large scale.
As of December 31, 1981, the installed power of nuclear power stations in the USSR was
about 16 GW (Table 3). The production of electrical energy at nuclear power stations in 1980
was 73 billion kW-h, as against 86 in 1981.*
The safety of nuclear power stations in use and under construction in the USSR [5] is
supported by a wide range of measures, of which the following are the main ones:
1) high equality in equipment manufacture and installation;
2) state monitoring for the equipment at all stages in use;
3) the definition and implementation of efficient protective measures to prevent emer-
gencies, compensate for any faults arising, and reduce the consequences of possible emer-
gency situations;
4) the definition and implementation of facilities for localizing radioactive materials
in the case of an emergency;
5) logical execution of all engineering and organizational measures to provide for safe-
ty at all stages in the building and operation of nuclear power stations;
6) standardization of engineering and organizational aspects in the provision of safety;
7) the Government supervision system.
Throughout the period of nuclear power station operation in this country, there have
been no instances where an emergency has led to the need to take measures to protect the pop-
ulation, although. much attention has always been given to preparing such measures. For ex-
*At the start of 1983, the installed nuclear power station capacity in the USSR was 18 GW
(electrical), and the electrical energy production was about 100 billion kW?h - Editor.
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 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196ROO0300020004-6
TABLE 4. Releases of Gases and Aerosols
Containing 131, from Two Units of Nuclear
Power Stations Containing VVER-440 and
RBMK-1000 Units
Release
Year
Kola
Novyi I
Voronezh
I
Kursk
Chernob
C
,
V
Gas, Ci/yr
1979
2.103
2.4.103
67.9.103
133.103
1980
2-10.3
2,13103
88,9.103
280.103
,3il mCi/yr
1979
1
4
66
290
1980 .
1
11
458
5000
ample, for each nuclear power station there is a plan of measures to protect the staff and
population in the case of hypothetical emergencies going outside the framework of the design
ones.
A comparison of the health risks for the staff and population in the production of elec-
trical power by nuclear power stations and thermal ones indicates that the former are pre-
ferable. When organic fuel is used, there is considerable environmental pollution from
release of ash and gases. Current coal-fired stations consume over 2x 106 tons of coal per
GW (elec.). This results in about 4 x 10' tons of ash, of which about 8 x 103 tons is released
into the atmosphere. The sulfurous gases are particularly harmful, and these constitute
tens of thousands of tons per GW (elec.). A difference of a nuclear power station from a
coal one is that there is no such release. Also, a nuclear power station does not consume
oxygen from the atmosphere.
Observations have been performed for many years on nuclear power stations containing
VVER-440 units, not only in this country, but also in the German Democratic Republic, Bulgaria,
Czechoslovakia, and Finland, as well as on stations with RBMK-1000 units, and these indicate
that radiation safety for the staff and population is reliably provided. The levels of pene-
trating radiation at permanently staffed and largely unstaffed locations do not exceed 1.4
and 2.8 mrem/h, which have been set as the permissible levels. The average individual dose
of external radiation to the staff does not exceed 15-18% of the maximum permissible for a
year, which is 5 rem/yr. The release listed in Table 4 are substantially below the level
permitted by the health rules of 1979 and are 183 Ci/MW (elec.). yr on gases and 3.65 mCi/
MW (elec.). yr on 131I. With such small releases to the environment, the radiation back-
ground in the locality is determined by natural sources of ionizing radiation as well as by
the artificial radionuclides formed by nuclear weapons tests. It is impracticable to distin-
guish the radionuclides derived from the station against the background of the global radio-
nuclides. The y-ray dose outside the nuclear power station does not increase with operating
time and does not vary with distance from the station.
. In the Soviet Union, nuclear power is considered as a very important means of solving
major problems in the fuel and energy balance over a long period. The safe and reliable
operation of existing nuclear power stations goes with their minimal effects on the environ-
ment and their high economic performance, and at the 26th Congress of the CPSU a decision was
therefore taken to provide for the increase in production of electricity in the European
part of the country mainly by the construction of nuclear power stations and hydroelectric
ones. Correspondingly, nuclear power stations are being constructed at over 20 sites and
will gradually displace base-load stations employing organic fuel in the northwest, west,
center, and south of the European part of the country. Nuclear power stations are being
constructed along the Volga and in the Ural. The installed power of the individual stations
is 4-6 GW. The rise in installed nuclear power station output in 1981-1985 is planned to be
provided mainly by the introduction of RBMK-1000 and VVER-1000 units.
This decision substantially eases the problems in the fuel and energy balance. On the
other hand, less than 25% of the energy resources go to the production of electricity, and
during the next five-year period nuclear power stations can provide electrical energy only
to base-load users in the European part of the country, so the contribution of nuclear power
to the fuel and energy balance can hardly exceed 10-15%. This means that nuclear power sta-
tions, while substantially alleviating the problem of the power and fuel balance, cannot
provide a radial solution. Solutions to the problems can come only from substantial extension
of nuclear-power applications.
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 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6
In the USSR, about 20% of the organic fuel used is employed in centralized heat supply,
and this is mainly the scarcest form, namely gas and oil. The main users of centralized heat
supply are located in the European part of the country, i.e., in regions furthest from the
sources providing for an increment in organic fuel production. Therefore, the extension of
nuclear power to centralized heat supply is a major task in solving the fuel and energy prob-
lems. The first steps have already been taken. The Bilibinsk combined heat and power station
has been operated since 1973 and provides heat to the population, while the Shevchenko nuclear
power station provides fresh water, and the heat from the Beloyarsk, Leningrad, Kursk, and
Chernobyl' nuclear power stations is also utilized.
To reduce the consumption of organic fuel in centralized heat supply, two major stations
have been built for domestic heat supply close to the cities of Gor'kii and Voronezh, which
will supply users with hot water. These heat stations are reliably protected from accidents
such as explosions, aircraft crashes, etc. It is impossible for radioactivity to reach the
users, because there is an intermediate circuit in which the pressure in the coolant is less
than that in the heat-user circuit. These features of the heat-supply stations make them a
reasonably powerful (300-500 MW) and safe source of heat supply, which can be located in
major inhabited areas. Under these conditions there is no need to lay long and expensive
heat-carrying pipes.
The first major combined heat and power station is being constructed near Odessa, in
which the production of heat will be accompanied by the production of electrical power. The
energy source is provided by a VVER-1000 reactor. Studies are being made on the scope for
building nuclear power stations to supply steam for industrial purposes.
The rates of introduction of nuclear heat sources are planned to increase substantially
in subsequent five-year periods.
Over 15% of the organic fuel is used directly in industry, including chemistry, metal-
lurgy, etc. The introduction of high-temperature reactors will further extend the use of
nuclear power, including the production of synthetic fuel.
Of course, it is essential that the general use of nuclear power in branches of the econ-
omy using substantial amounts of energy must be reliably supported with nuclear fuel. With
the existing thermal reactors, the energy yield from a ton of natural uranium is not more than
7.5 X 103 MW-day/ton, i.e., the extent of use of the natural uranium is not more than 1%. A
solution to the fuel problem requires a substantial increase, namely by about an order of
magnitude by comparison with the existing level. This is possible, in particular, by the
introduction of fast reactors, in which the use of natural uranium can be increased by almost
two orders of magnitude. Therefore, considerable attention is being given to fast-reactor
development in the USSR. As developments proceed, the proportion of these in the structure
of nuclear power will increase. One of the future problems is to prepare various branches
of the industry for the routine introduction of fast reactors. Unfortunately, such reactors
cannot take on the role of basic energy sources in many areas. It is undesirable to operate
them with a variable load graph. Also, they can hardly provide the basis for centralized
heat supply, since, firstly, the unit power is too large (in order to increase the economic
performance and improve the neutron balance, fast reactors of unit power 800-1600 MW (elec.)
are devised, while to provide heat one requires sources mainly of 300-500 MW (thermal) and
less), and, secondly, for reasons of safety they have to be located at considerable distance
from heat users. They also cannot produce high-potential heat (about 1000?C) efficiently.
At present, it is difficult to establish with certainty the optimum proportion of fast
reactors in the structure of future nuclear power. This will be dependent on the structure
of energy use, and, in particular, on the level of electrification. However, one assumes
that the reasonable proportion of such reactors can hardly exceed about 0.5. The problem
is that when the proportion is less than 0.5, it is still necessary to provide developing
nuclear power with artificial fuel in the necessary amount.
Table 5 gives theoretical values for the rate of accumulation of excesss plutonium in
fast reactors with sodium cooling for external fuel-cycle durations of 1 and 3 yrs. The less
the duration of the external cycle, the higher the rate of plutonium accumulation, but also
the more complicated the shipping and reprocessing of the spent fuel. Table 5 shows that
with an external fuel-cycle duration of 1 yr, the BN-800 and -1600 (fuel reproduction coeffi-
cient RC = 1.3) can provide a rate of accumulation of plutonium of about 0.05 yr-3, while
the improved BU units (RC = 1.55) can provide about 0.08 y-' [6].
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 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6
TABLE 5. Rates of Accumulation of Excess
Plutonium
External fuel cycl
Reactor duration, yrs
Rate of accumula
tion of excess
plutonium, yr
BN
3
0,028
BN
1
0,051
BU
1
0,077
TABLE 6. Equilibrium Proportion of Fast
Reactors
Nuclear power
Devel-
opment
Length
of ex-
ternal
Input of natural
uranium, ton/
GW (elec.) ? yr
structure
rate,
yr-1
fuel cy-
cle, yrs
0 I
10
20
BN + LVR
0,01
3
0,83
0,71
0,60
BN + LVRU
0,01
3
0,73
0,54
11,38
BN + LVR
0,01
1
0,79
0,68
0.57
BN + LVRU
0,01
1
0,67
0,48
0,30
BU + LVR
0,01
1
0,64
0,55
0,46
BU + LVRU
0,01
1
0,49
0,35
0,22
BU + LVR
0,03
3
*
0,90
0,77
BN + LVRU
0,03
3
*
0,84
0,63
BU + LVR
0,03
1
0,90
0,78
0,66
BN + LVRU
0,03
1
0,83
0,64
0,45
BU + LVR
0,03
1
0,75
0,65
0,55
BU + LVRU
0,03
1
0,63
0,49
0,34
*Fast reactors are not capable of provid-
ing the required rate of accumulation in
nuclear fuel.
Table 6 gives the equilibrium proportion of fast reactors in relation to the rate of
development of nuclear power, the duration of the external fuel cycle, and supplies of natural
uranium. By LVR is meant a water-cooled reactor with the characteristics of the VVER-1000
(the RBMK has similar characteristics), while by the improved LVRU is meant a reactor with
characteristics analogous to those of the LVR but with the oxide fuel replaced by a uranium
of elevated density, while the improved fast BU reactors are ones with heterogenous struc-
tures for the core and sodium cooling [6].
Table 6 shows that the minimum proportion of fast reactors is 0.55 (if the BU is used)
if these work together with the LVR, or 0.34 with the LVRU even for the comparatively low
rate of development in nuclear power of 0.03 yr-1, and with a consumption of natural uranium
of 20 tons/GW (elec.). yr, which corresponds to an average consumption of the uranium of
about 5%. The equilibrium proportion is appreciably higher if there is a lower RC and the
length of the external fuel cycle is 3 yrs.
Therefore, these preliminary calculations show that a solution to the problem of reli-
able nuclear fuel supplies requires fast reactors with RC > 1.5, and external fuel cycles of
1 yr, and also improved thermal reactors with RC = 0.7-0.75.
In principle, all three conditions can be met. Research show that fast reactors with
RC > 1.5 are possible, e.q., upon using carbide fuel and a heterogenous core construction with
sodium cooling. A possible competitive form may be fast reactor with helium cooling. There
are no theoretical obstacles to the creation of thermal reactors with RC = 0.7-0.75. If for
some reasons it does not prove possible to attain such RC in LVR with uranium fuel, then
upon using thorium one can attain RC= 1 even in reactors with light-water coolant. For ex-
ample, the VVER-1000 can provide RC= 0.7 while maintaining the diameter of the fuel pins,
the lattice pitch, and the reactor power if uranium dioxide is replaced by a mixture of
thorium dioxide with enriched uranium.
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 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6
Therefore, there is every reason to assume that the fuel problems of nuclear power will
be solved, and therefore solutions will become available to the basic problems of the fuel
and power balance over a long time-scale.
1.
A. P. Aleksandrov, in: Nuclear Power in the 20th Century [in Russian], Atomizdat, Mos-
(1974)
cow
.
2.
F. Ya. Ovchinnikov et al., (this issue).
3.
N. A. Dollezhal' et al., (this issue).
4.
0. D. Kazachkovskii et al., (this issue).
5.
_V. A. Sidorenko and 0. M. Kovalevich, Papers at the IAEA International
Nuclear Power and Its Fuel Cycle, Salzburg, May 2-13, 1977.
Conference
on
6.
S. B. Bobrov et al., (this issue).
EXPERIENCE WITH THE CREATION, OPERATION, AND MEANS
OF IMPROVEMENT OF NUCLEAR POWER PLANTS WITH WATER-
COOLED-WATER-MODERATED REACTORS (VVER)
F.
Ya.
Ovchinnikov, Yu. V. Markov,
V.
A.
Sidorenko, V. A. Voznesenskii,
G.
L.
Lunin, V. V. Stekol'nikov,
G.
G.
Bessalov, and Yu. V. Vikhorev
After the start-up in 1964 of the first unit of the Novovoronezh Nuclear Power Plant
(NNPP) with a gross electric capacity of 210 MW, 10 VVER reactors intended to produce 365-
440 MW of electrical power were subsequently placed into operation at the Novovoronezh,
Kol'skaya, Armyansk, and Rovensk Nuclear Power Plants. During this same period the design
power has been utilized in 12 units with VVER in the German Democratic Republic, Bulgaria,
Czechoslovakia, and Finland (Table 1). Further construction of nuclear power plants in the
USSR has continued mainly with the use of a new series of water-water reactors,(VVER-1000)
with a thermal capacity of 3000 MW and a nominal value of 1000 MW for the electric capacity
of the unit. The first unit with a VVER-1000 at NNPP was connected to the supply system on
May 30, 1980 and reached its nominal power on February 20, 1981.
The range of their application in power engineering has expanded simultaneously with
the increase in the individual capacity of the reactors: Designs have been created and
construction has proceeded of nuclear power plants with VVER in seismic regions, and the
operation of VVER has been proposed under conditions of regulation of the frequency and
capacity in power systems as well as for the combined generation of electrical power and heat.
IMPROVEMENT OF THE BASIC ENGINEERING SOLUTIONS FOR
NUCLEAR POWER PLANTS WITH VVER
The basic engineering characteristics of nuclear power plants with VVER are given in
[1-4] and in Table 2.
Nineteen years after the start-up.of the first unit of the Novovoronezh Nuclear Power
Plant, the individual electric capacity of the units has increased from 210 to 1000 MW, the
specific intensity of the active zone from 47 to 111 kW/liter, the pressure in the reactor
(absolute) from 100 to 160 kgf/cm2 (1 kgf/cm2 = 9.8 x 10? Pa), and the steam pressure in the
steam generators from 32 to 64 kgf/cm2. The following solutions have remained unchanged:
Six-sided heat-generating assemblies (HGA) with cylindrical fuel elements containing U02 and
covered by an alloy of Zr +1% Nb were used in the active zone; high-strength chrome-molyb-
denum steel was used for the reactor housing; steam generators of the horizontal type [5]
Translated from Atomnaya Energiya, Vol. 54, No. 4, pp. 249-262, April, 1983.
0038-531X/83/5404-0251$07.50 ? 1983 Plenum Publishing Corporation 251
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 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6
TABLE 1. Sequence of Entry into Service
of the Power Units of Nuclear Power Plants
with VVER
Power plant, Attain- Connec- utiliza-
power unit Reactor m~qt of tion to tion of
crrucalr- supply 100% of
y system capacity
Novovoronezh, VVER-210 17.12.63 30.09.64 .31.12.64
I
Rheinsberg
VVIr R-70
11.03.66
06.05.66
10.10.66
Novovoronezh,
VVE R-365
23.12.69
27.12.69
14.04.70
Novovoronezh,
VVE R-440
22.12.71
27.12.71
29.06.72
III
ovoronezh,
N
25.12.72
28.12.72
24.03.73
ov
Kol'skaya, I
?
26.06.73
29.06.73
28.12.73
Nord, I
?
02.12.73
13.12.73
11.07.74
Kozlodui, I
30.06.74
17.06.74
28.10.74
Kol'skaya, II
30.11.74
09.12.74
21.02.75
Nord, II
? ?
02.12.74
23.12.74
16.04.75
Kozlodui, II
?
28.08.75
26.08.75
05.11.75
Armyansk, I
>
22.12.76
28.12.76
06.10.79
Lovisa, 1
20.01.77
08.02.77
09.05.77
Nord, m
06.10.77
03.11.77
03.05.78
Y'aslovske-Bo-
?
27.11.78
17.12.78
30.03.79
gunitse, I
Nord, IV
22.07.79
02.08.79
31.10.79
Armyansk, II
>
04.01.80
06.01.80
31.05.80
Yaslovske-Bo-
15.03.80
20.03.80
25.05.80
gunitse, II
Novovoronezh,
VE R-1000
30.04.80
30.05.80
20.02.81.
V
Lovisa, II
MR-440
17.10.80
04.11.80
22.12.80
Kozlodui, 111
?
04.12.80
16.12.80
27.01-81
Rovensk, I
?
17.12.80
22.12.80
Kol'skaya LII
07.02.81
24.03.81
Rovensk, II
?
19.12.81
27.12.81
*The nominal electric capacity of the unit
is utilized at a thermal capacity of 92%.
were used for production of saturated steam; and transportability of the reactor housing over
the railroads of the USSR was ensured.
The layouts of nuclear power plants with VVER and the designs of the absorbers and the
actuators of the control elements, the intrahousing mechanisms and the main circulating
pumps, steam generators, and turbines have differed in their appreciable variety. For ex-
ample, along with the single-unit (one reactor + one turbine) Rheinsberg nuclear power plant
in the German Democratic Republic, the second unit of the Novovoronezh Nuclear Power Plant,
in which there are eight circulation loops and five turbogenerators, is successfully oper-
ating with a high usage coefficient of the installed capacity (Tables 3, 4).
The main changes in the equipment and systems of nuclear power plants with VVER are
discussed below.
Reactor and Intrahousing Mechanisms. Questions of vibration stability under the dynamic
action of the coolant flow have exerted a dominant effect on selection of the design of the
intrahousing mechanisms. A shift of the thermal shield in the first unit of the Novovoronezh
Nuclear Power Plant in 1969 led to a reconsideration of the streamline flow conditions and the
securing of all elements of the intrahousing mechanisms. The thermal shield for VVER-365 and
the first WER-440 units was mounted on the hollow shaft of the reactor with welding in the
upper part around the entire perimeter (Fig. 1). For VVER-1000 and later modifications of
the VVER-440, the thermal shield was eliminated as a structural element, due to thickening
of the walls of the other intrahousing mechanisms.
It was repeatedly necessary in 1974-1975, when a defect in the zirconium covers of the
fuel parts of individual control elements was discovered on some VVER-440 at the places at
which they contact the steel spacer networks, to return to questions of the interaction of
the coolant flow with the structural elements in the housing of VVER. The cause of the de-
fects is fretting corrosion, which had arisen due to increased vibrations of the control HGA
(Fig. 2) and thickening the wall of the zirconium cover from 1.5 to 2.1 mm.
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 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6
TABLE 2. Basic Engineering Characteris-
tics of Reactor Installations with VVfR
Parameters
r
a
w
?
c~7
w
co
_
?
w
ow
00
a..
w[w
> ,z
Thermal capacity 265
760
1320
'1375
3000
of reactor, MW
Number of circu- 3
6
8
6
4
latibn loops
Pressure,
2
kgf/ cm
:
in reactor 100
100
105
125
160
in steam gener 32
32
33
47
li4
ator
Temperature, C
at reactor en- 250
245
248
268
288
trance
at reactor exit 266
266
274
296
317
Flow rate of coal- 16 000
33 000
50 000 `
45 000
88 000
ant through re-
3-
/ h
actor, m
Inner diam, of re- 2640
3560
3560
3560
4139
actor housing,
mm
Equivalent diam. 190
288
288
288
311
of active zone,
cm
Height of active 250
250
246
246
356
zone in opera-
tional state, cm
Power intensity of 38
47
83
86
111
active zone,
kW/liter
Number of HGA 148
343
349
349
151
in active zone
Number of fuel 90
90
126
126
317
elements in a HGA
Outer diam. of 1(1,2
10,2
9,1
9,1
9,1
fuel element, m
Thickness of fuel- 0,6
0,6
0,65
0,65
0,67
dement c4ver-
rng made from
loy of Zr+1o
mm
Av. 1'n ar capaci 80
ty oaf fuel ele-
99
122
127
176
ments, W / cm
Fuel charge into 17,0
40,0
41,5
41,5
66,0
reactor, tons of
metal
Specific capacity 15,5
of fuel, kW./ kg
19
32
33
45,5
of U
Enrichment of 2,0
2,0
3,0
3,5
4,4
makeup fuel
upon relace-
lI 0At of 1/3
Fuel depletion 13
14
28
30
40
depth, MW/ day
Number of SCR 19
37
73
73 or 3
1()9
assemblies
*With the operation of seven loops (one
is a reserve loop).
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 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196ROO0300020004-6
TABLE 3. Expenditures for Construction
and the Average Technico-economic In-
dices of the Operation of Nuclear Power
Plants with VVER in 1977-1981
Indices
Novovoronezh, units
Kol'
skaya,
units
Arm-
yansk,
units
I
I II
IIII11V
1 V
I, 1,
I, II
Installed capaci-
210
365
880 *
1000
880
815
ty, MW (elec.)
gross
Specific capital
b
326
256
200
308
263
327
costs, ru
res/
kW
Electrical power
7,4
13,9
30,1
6,0 1
31 ,0
15,4
output, billions
k
Average usage
0,80
0,87
0,81
-
0,80
0,62
coefficient of
installed capac-
ity
*Since 1979 the total installed capacity
of units III and IV has decreased to 834
MW in connection with the high average
temperature of the water which cools the
turbine condensers.
tData for 1980-1981.
TABLE 4. Technicoeconomic Characteristics of Nuclear Power Plants with VVER during
the 1977-1981 Period
Novovoronezh, units
Kol'skaya, units
r ,
yansk
units
ts
Characteristics
Year
I
II I
III
IV
V
I
II I
III
I
II
Usage coefficient of in-
1977
0,76
0,82
0,78
0,79
-
0,59
0,73
-
0,208
-
stalled capacity
1978
0,86
0,90
0,82
0,75
-
0,83
0,83
-
0,535
-
1979
0,818
0,909
0,758
0,707
-
0168
0,85
-
0,668
-
1980
0,714
0,856
0,843
0,843
0,22
0,962
0,907
-
0,774
0,551
1981
0,847
0,859
0,838
0,898
0,562
0,802
0,855
0,338
0,778
0,762
Electrical power output,
1977
1397
2614
3013
3057
-
2288
2823
-
834
-
millions of kWh
1978
1584
2892
3150
2891
-
3205
3199
-
1909
-
1979
1505
2905
2850
2656
-
2616
3285
2386
-
1980
1317
2745
3079
3088
1112
3717
3507
-
2772
1974
1981
1558
2745
3061
3280
4918
3091
3297
1001
2779
2720
For further improvement of the hydrodynamic conditions of operation of the active zone
on all VVER, a special perforated bottom which equalizes the flow distribution through the
HGA has been installed in the lower space of the reactor, starting from the first unit of
the Armyansk Nuclear Power Plant. When the new units are started up, expanded operating
programs are performed for measurement of vibration and stresses in the structural elements
of the reactor and the intrahousing mechanisms.
At present, a great deal of experience has been accumulated on the operation of 10 VVER-
365 and VVER- 440 housings made out of steel 15Kh2MFA without noncorroding planting. A sat-
isfactory corrosion state of the inner surface of the housings is provided by the obser-
vance of an ammonia-potassium aqueous chemistry regime in the course of operation and by
the creation of an increased ammonia concentration and the execution of measures for reduc-
tion of the nitrate concentration during recharging periods. Moreover, noncorring plating
has again been introduced in accordance with universal practice on the VVER-1000 and
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 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6
Fig. 1. Construction of the thermal shield in: a) VVER-210; b) MR-365; S)
MR-440 of the third and fourth units of the NNPP; d) in the commercial VVER-
440; 1) thermal shield; 2) shaft.
and VVER-440, starting from the first unit of the Lovisa Nuclear Power Plant. Maintenance
of the required aqueous chemistry regime has been simplified somewhat.
It has become necessary for recent modifications of VVER in connection with the increase
in capacity of the systems for emergency cooling of the active zone, which supply a relatively
cold solution of boric acid directly to the reactor, to consider, in addition, the question
of the calculated reserve of the reactor housings from the standpoint of resistance to
brittle fractures. A detailed analysis of the dependence of the radiation resistance of the
housing materials on the flux of fast neutrons has been performed for all VVER housings on
the basis of an investigation of the properties of irradiated test samples. Additional mea-
sures, which provide a guaranteed calculated reserve of housing operation, were applied for
several VVER-440 housings in connection with an increase in the amount of phosphorus and
copper impurities in a welded seam in the region of the active zone. Model HGA which permit
reducing the maximum irradiation of housing sections by a factor of three were mounted on the
periphery of the active zones of these reactors in place of 36 HGA with fuel.
The Active Zone and the Control and Protection System. Fuel HGA with a covering of
Zr + 2.5% Nb alloy were used for assembly of the active zone in all VVER operating on July 1,
1982. Spacing of the fuel elements in the beam was accomplished by 12-16 steel grids. The
size of the "turned on" HGA for all VVER-1000 is 144 mm. In order to increase the specific
capacity of the active zones, the outer diameter of a fuel element was reduced from 10.2
(MR-70 and MR-210) to 9.1 mm with a simultaneous increase in the number of fuel elements
in a HGA.
The ratio of the number of hydrogen atoms to the number of uranium atoms in the opera-
tional state is 4.2-4.7, which provides a negative reactivity coefficient with respect to
coolant temperature during operation of the reactors. The temperature coefficient of reac-
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 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6
Fig. 2. Alteration of the design of the intrahousing mecha-
nisms of the reactor to decrease the dynamic action of the
coolant on the control HGA of the active zone in: a) VVER-
365; b) VVER-440 of the third and fourth units of the NNPP;
c) the commercial VVER-440; 1) shaft; 2) bottom of the shaft;
3) openings for passage of coolant into the control assemblies;
4) perforated bottom.
tivity is slightly positive only at the start of operation of the first loadings of VVER-440
and VVER-1000, which operate with boron regulation. The depletion depth of the fuel in a
VV9R-440 is ti30 MW?days/kg of U for an average enrichment of the makeup of 3.5%. One-third
of the HGA which have reached maximum depletion are unloaded each year. Fresh fuel is in-
stalled on the periphery of the active zone.
Regulating safety-and-control-rod (SCR) assemblies equipped with actuators are provided
on VVER-70 and VVER-210 to act on the reactivity. Part of these assemblies, which are in-
tended mainly for compensation of the total reactivity reserve, contain absorbers made out
of boron steel in the upper part and an assembly with nuclear fuel which is similar in con-
struction to a fuel HGA in the lower part. In addition, there are assemblies for emergency
protection which are intended for rapid shutdown of the reactor. They do not contain fuel
and have their own actuator construction.
All the SCR assemblies and their actuators have become identical in the subsequent de-
velopment of VVER. The problem of increasing the efficiency of the SCR system was initially
solved by increasing the number of assemblies with absorbers (the second unit of the NNPP).
Starting from the third unit of the NNPP, the reactivity reserve against fuel depletion and
slow changes in the reactivity have been compensated by the introduction of a boric acid so-
lution into the coolant of the first loop.
The application of boron regulation has permitted reducing the number of regulating SCR
assemblies on VVER-440 from 73 to 37. Control elements in the form of bunches of 12-18 fuel
elements containing europium dioxide or boron carbide are used in the reactor of the fifth
NNPP unit and other VVER-1000. The transition to a new construction of the absorbers has per-
mitted reducing the height of the reactor housing by virtue of a decrease in the volume under
the active zone in which the fuel portion of the SCR assemblies is positioned in VVER of
smaller capacity. The larger number of control elements, including elements with absorbers
one-half the length of those in the first MR-1000 reactor in the USSR, permits effectively
influencing the energy distribution throughout the active zone if necessary.
lifliniff
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 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6
TABLE 5. Design Engineering Characteris-
tics of the Main Circulation Pumps (MCP).
Ty
pes of
reactors and pu
mps
VVER
VVER
VVER-440
VVER-
Characteristic
210
365
100.0
MCP-
MCP-
MCP
MCP-
MCP
309
310
317
195
195
Witho
ut a st
uffing box
with
.pressur
ized
with pac
shaft an
king of
d in-
elec
tric m
otor
creased
moment
of inerti
a
Flow rate, m3/ h
5250
5600
6500
7100
20 000
Coolant tempera-
250
250
270
270
300
ture, ?C
Intake pressure,
100
'105
925
125
156
kgf / om2
4
5,5
5.3
43
6,75
Number of rpm
nchrono s1)
(s
1500
1500
1500
1500
1000
y
Consumablecapac-
ity (no more
than), kW
in cold water
--
-
--
1600
7000
in hot water
11150
1500
2000
1400
5300
Mass of MCP with
42
36
51
55
150
auxiliary equip-
ment, tons
including elec
-
-
-
15
48
tric motors
`Replacement of MCP-138 by MCP-309A was
performed in 1972-1975.
TABLE 6. Design Engineering Characteris-
tics of SG
Characteristics
VVER-
210
VVER-
365
VVER-
440
VVER-
1000
Steam capacity,
230
325
452
1468
tons/h.'
Steam pressure,
32
33
47
64
kgf / cm2
Temperature, ?C;
of coolant of
first loop
at entrance
273
280
301
320
at exit of sup-
252
252
268
290
ply water
189
195
. 225
220
Humidity of steam
0,2
.0,2 .
0,2
. 0,2
at exit no more
than, 1
H atin surface
m
n outer di
(
1300
1810
2510
6115
.
e
o
of pipes), m
outer diam, and
21x1,5
16X1,4
16X1,4
161,5
wall thickness of
pipes, mm
Mass of dry steam
104
112
155
292
generator, tons
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 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6
TABLE 7. Design Characteristics of Volume
Compensators
Characteristic
VVER-
210
VVER-
365
VVER-
440
VV@R-
1000
Type
Gas
Steam
Total internal vol.
4x17,5
4x10,7
38 or 44
79
VOL, m
Inner diam. of cy-
1800
1500
2400
3000
lindrical part,
mm
Volume of nitro
g-
t
52
18
16 or 18
24
no
or steam a
inal capacity,
m3
O perating pressure
100
105
125
160
kgf/cm
Operating tem-
260
313
325
346
perature, ?C
Total capacity of
-
1680
1620
2520
heaters, kW
Actuators of the "screw-nut" (VVER-70, VVER-210, and VVER-365) and the "rack-and-pinion,"
(VVER-440) types and several kinds of electromagnetic actuators (VVER-1000) were developed
for movement of the SCR absorbers.
Intrareactor Control. In order to control the energy distribution in the active zone
on VVER, systems of intrareactor control are provided. The temperature is measured on VVER-
210 and VVER-365 at the exit of about two-thirds of all the HGA with fuel, and, in addition,
12 or 36 measuring channels containing 5-7 neutron-flux detectors each .are mounted in the
HGA on VVER-440. The temperature at the exit of all the HGA is controlled on the VVER-1000
of the fifth NNPP unit, and neutron-flux detectors are mounted in 31 HGA. Analysis of the
data of the intrareactor measurements and presentation of the results to the operator is
presently accomplished by a special system linked to the information-computation complex com-
mon to the unit as a whole. In the event of a breakdown of this complex, the system switches
to an automatic operating mode in which information is processed by simplified algorithms.
Main Circulation Pumps (Table 5). Low-inertia pressurized pumps with a synchronous ro-
tation frequency of the rotor of 1500 rpm are used on 17 of the 24 operating VVER units. In
order to cool the active zone in the case of disconnection from the, power system, the opera-
tion of the MCP for 100 sec after shutdown of the reactor is provided on these units by means
of the energy of electromechanical coasting of the main generators or special internal-dis-
charge generators located on the same shaft with the turbines. Pumps with the rotor of the
electric motor extended beyond the confines of the first loop are used for the VVER-1000 and
also in the new designs of nuclear power plants with VVER-440. A special flywheel provides
for a slow decline of the flow rate when the MCP are disconnected
A test of the operation of the MCP at nuclear power plants in the USSR has shown that
they are one of the most reliable pieces of equipment of a reactor facility.
Steam Generators (Table 6). Steam generators (SG) with a horizontally positioned housing
and a pipe bunch are used at nuclear power plants with VVER, which provides for moderate
loads on the surface of the evaporation mirror. Cylindrical collectors of the primary coolant
are located in the surroundings of the second loop. The pipe bunch is fabricated out of
OKhl8NlOT steel, and the housing material is carbon steel.
The principal structural change in the course of the improvement of SG is the realiza-
tion for VVER-440 and VVER-1000 of access from above to the collectors of the first loop. In
this case it proved to be possible to reject special subshaft access spots for disposition
of equipment servicing the SG. Maintainance and a surveying of the sites where pipes are
sealed into the collectors are accomplished from above directly from the central room of the
nuclear power plant. Since a flanged joint is situated in the surroundings of the second
loop, special attention to the behavior of the metal of the collectors of the first loop is
necessary in the phase partition zone.
Pressure Compensators. A nitrogen pressure compensator is used in the first unit of
the NNPP and the Rheinsberg Nuclear Power Plant. Steam pressure compensators having better
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 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6
weight-size characteristics (Table 7) are used in the equipment complement of the remaining
VVER reactor facilities. The housings of the pressure compensators are fabricated out of
carbon steel. On units in which the reactor housing has no noncorroding plating, it is ab-
sent on the pressure compensator.
Main Circulation Pipelines and Slide Valves. For all the reactors except VVER-1000, the
main circulation pipelines (MCP) with an inner diameter of 500 mm (Du 500) are fabricated out
of lKhl8N9T and 1Kh18N12T steels. For VVER-100 the MCP are made two-layered (on the outside -
carbon steel; inside - noncorroding steel), and their straight-through cross section is Du
850. The possibility of cutting off loops with the help of slide valves having an electric
actuator which closes in a time no greater than 90 sec is provided for. Rapid-acting slide
valves with a hydroactuator, which have been used on the MCP of the first unit of the NNPP
and the Rheinsberg Nuclear Power Plant, were excluded in the subsequent designs.
Safety Systems. As VVER have developed, safety mechanisms intended to limit the con-
sequences of accidents and to localize radioactivity which has leaked from the main circula-
tion loop (MCL) have steadily been improved [6]. The increased reliability of the emergency
protection systems of the reactor, emergency cooling of the active zone, diversion of heat
from the steam generators, and localization of fission products has been attained both by
providing the necessary emergency arrangement of these systems and by applying better en-
gineering solutions.
If an instantaneous break in a pipeline with a diameter of about 100 mm with one-way
outflow were the maximum design emergency for the first VVER, then the protective and local-
izing mechanisms for contemporary VVER-440 (Rovensk Nuclear Power Plant) and VVER-1000 prov-
ide for safety in the event of accidents right up to instantaneous fracture of the MCP coin-
ciding in time with conditions of complete deactivation of the nuclear power plant. An in-
crease in the temperature of the fuel element casings above 1200?C is prevented with the help
of the provided systems of emeregency cooling of the active zone (water tanks connected in
pairs to the entrance and exit space of the reactor; high- and low-pressure pumps). Localiza-
tion of fission products which escape from the MCL is accomplished for new nuclear power
plants with VVER-440 by the traditional method for VVER - with the help of a system of pres-
surized rooms; the reactor room remains accessible for servicing. A condenser-bubbler pro-
vides for steam condensation during the first period of a maximum design emergency. The
maximum pressure in the chambers in the course of eliminating the emergency does not exceed
2.5 kgf/cm2.
For nuclear power plants with VVER-1000, construction of a shell is provided which en-
closes all the rooms of the MCL and the reactor room and is calculated for the total pressure
arising upon outflow of all the coolant (5 kgf/cm2) with subsequent reduction in the pressure
of the sprinkler system. In order to prevent the escape of activity during an accident, the
installation in sequence of three rapid-acting pneumatic valves is provided on the pipelines
which connect the shell to the external systems; each valve closes from its own high-pres-
sure air system. A high degree of independence of the redundant protective and localizing
systems is provided in the designs by means of placing them in different rooms, a separate
electrical supply, etc.
POWER PLANTS WITH VVER
The cumulative duration of the operation of nuclear power plants with VVER from the time
they were included in the grid to July 1, 1982 amounts to 150 reactor-yrs. Analysis of the
operation of VVER during this period permits asserting that nuclear power plants with such
reactors are capable of providing a reliable supply of electrical power to consumers, with
high technicoeconomic indicators . As follows from Table 3, the specific capital expenditures
for the construction of nuclear power plants decreases at first and reached a minimum sum
for the third and fourth units - 200 rubles per 1 kW of installed capacity. The increase
in the cost of construction for the subsequent units is explained both by factors of local
significance (construction of nuclear power plants at Zapolyar'e - the Kol'skaya Nuclear
Power Plant; provision of seismic-resistant buildings and equipment - the Armyansk Nuclear
Power Plant) and by general tendencies: the increase in costs to provide for the safety of
nuclear power plants and the rise in prices of energy resources.
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 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6
In the USSR 106.6 billion kWh of electrical power was produced at nuclear power plants
with VVER during 1977-1981, which amounts to about 2% of the total amount of electrical power
produced during this period in the entire country. The average cost of the production of
electrical power in 1981 is significantly lower than at thermal power plants. For units
which have operated longer than a year at the design capacity, the usage coefficient of
installed capacity exceeds, as a rule, the design value of 0.8 (see Table 3, 4).
A continuous effort is being made at nuclear power plants to increase the reliability
and safety of the equipment. Within the framework of the system for collection of informa-
tion on equipment failures which has been operating since January 1, 1977, the causes for the
failures are investigated and classified, and equipment with reduced reliability is revealed.
The information is directed to the factories-manufacturers and to the design organizations
for adoption of the necessary measures.
Quantitative reliability indices of the first and second units of the Kol'skaya Nuclear
Power Plant during the period from the start of its operation to June 30, 1980 without taking
account of the failures during the shakedown period are given below in hours:
Unit as a whole
Operating time to failure
1980
1902
Average recovery time
24
8.5
Shakedown period
3406
1490
Reactor with control and protection system
Operating time to failure
10,000
4692
Average recovery time
2.6
4.8
Shakedown period
2923
2000
Main circulation pump
Operating time to failure
57,230
20,020
i
388
250
me
Average recovery t
Shakedown period
0
0
Steam generator
Operating time to failure
6804
11,190
Average recovery time
149
294
Shakedown period
3000
5000
These data indicate the stable operation both of nuclear plants as a whole and of the
basic equipment. One should note that the failure of a unit as a whole is an event leading
to complete degradation of the charge, but failure of the MCP or an SG is an event which
results in the disruption of their work capacity, which leads to disconnection of the loop.
Among the problems which must be solved for operating units, one should note the devel-
opment of measures for constant upgrading of the safety of the units in connection with the
change in the operating norms and rules which regulate safety questions. In connection with
the expiration in 1984 of the design term of service of the reactor housing of the first
unit of the NNPP, the question of an operation extension by means of a possible annealing
of the housing, with simultaneous replacement of the reactor cover, the actuators of the con-
trol elements, and part of the intrahousing mechanisms, is being considered.
METHODS OF FURTHER IMPROVEMENT OF VVER
The subsequent development of nuclear power plants with VVER will be accomplished by the
application of more refined equipment, simplification of the layout of the MCL and nuclear
power plants as a whole, optimization of the thermal engineering parameters and the fuel cy-
cle, and improvement of the reliability of the systems for provision of safety. The range
of possible usage of VVER in power engineering will be simultaneously expanded.
A modified VVER-1000 reactor assembly which is distinguished by an improved layout of
the MCL and the absence of the main shutoff slide valves is being used in the majority of the
units which will be placed into operation up to 1990. The extent of the MCL and the protec-
tive shell is reduced by more than 20%. The rejection of the use of covers for the HGA per-
mits placing 163 HGA instead of 151 (fifth unit of the NNPP) in the active zone.
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 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6
The number of control elements is decreased from 109 to 61 with a simultaneous increase
in the number of fuel elements in a bunch from 12 to 18.
The design characteristics of the MR-1000 (the fifth unit of the NNPP and the modern-
ized unit, respectively) are given below:
Thermal capacity, MW
Coolant pressure, kgf/cm2
Average coolant temperature, ?C
Coolant flow rate, m3/h
Outer diameter of the reactor
housing, mm
Height of assembled reactor, mm
Equivalent diameter of the active
zone, cm
Height of the active zone in the
operating state, cm
Power intensity of the active
zone, kW/liter
Number of HGA
Shape and type of HGA
Size of an HGA with cover when
"turned on", mm
Fuel (U02) charge in the active
zone, tons
Outer diameter and spacing of the
fuel elements, mm
Average thermal flux, W/cm2
Operating period of the fuel, yrs
Number of rechargings per operating
period
Enrichment of fresh fuel in the
steady-state recharging regime,
%
Average depletion depth of the
fuel, MW-days/kg of U
Number of control elements
Number of fuel elements in a con-
trol element
Number of MCP
Number of revolutions of the MCP,
rpm
Presence of shutoff slide valves
in the MCP loops
Number of SG
Type of SG
Steam capacity of a single SG,
tons/h
Steam pressure at the exit from
the SG, kgf/cm2
Steam temperature, ?C
3000
160
306
80,000
4535
22,592
311
356
111
151
Six-sided
3000-3200
160
306-307
80,000
4535
19,137
316
356
107-115
163
Six-sided
with cover
without cover
238
75.5
9.1/12.75
176
2 or 3
2 or 3
234
80
9.1/12.75
166-177
3
3
3.3 or 4.4
4.4
27 or 40
40
109
61
12
18
4
4
1000
1000
Yes
No
4
Horizontal
4'
1469
1469-1575
64
64
278.5
278.5
As the development is accomplished, new more refined equipment will be used in the VVER-
1000. More compact SG constructions are being designed: vertical with natural circulation
of the evaporator, and direct-flow. Steam generators with higher steam parameters (305-310?C,
70-74 kgf/cm2) are being considered in a number of possible alternatives. The development
of a pump assembly with a shaft
rotation of
3000
rpm and with mass characteristics approximate-
ly twice as good as those of
individual subassemblies.
the MCP-195
is in
the stage
of experimental checking of the
The use of a nuclear fuel other than uranium dioxide is not anticipated for VVER reac-
tors up to 1990. In order to improve the characteristics of the fuel cycle, it has been
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 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196ROO0300020004-6
proposed to increase the average fuel depletion depth to 40 MW-days/kg of U by increasing the
enrichment of the makeup fuel with simultaneous optimization of the recharging regime and a
restriction in the structure of the active zone on the number of construction materials with
large neutron absorption cross section (steel). The efficiency of the fuel elements of the
VVER-1000 at such a depletion has been confirmed in an experimental microroentgenometer loop
There has also been a successful test of the attainment of a depletion of ti 50 MW?days/kg of
U in the regular HGA of the VVER-440 with them located in the active zone for 5 yrs. The
gradual conversion of VVER to fuel of increased density or the incorporation of thorium into
the fuel cycle is advisable in the future in the interests of the economy of inexpensive
natural uranium. The scientific-engineering bases for such a change in the fuel cycle should
be worked out in the forthcoming decade.
An increase in the safety level of VVER reactor installations is mainly guaranteed by
means of the further development of the system of intrareactor measurements and the incorpora-
tion of operational systems for control of the state of the equipment and the metal of the
MCP. At present, the USSR COMECON member-nations are working on a cooperative program which
includes tests on operating reactors for working out methods of recording variations in the
noise spectrum during disruptions of the normal operation of a reactor facility, including
the onset of boiling in the active zone. Scientific principles and instrumentation for the
detection and classification of defects in the materials of MCL.equipment using the acoustic
emission method are being developed.
As ordinary fuel becomes more expensive, the number of regions in the USSR in which it
is economically advisable to construct nuclear power plants will increase during the suc-
ceeding decades. In this connection, as well as with possible deliveries for export taken
into account, VVER-1000 reactor assemblies will be calculated to withstand an earthquake
with a force up to a reading of 9 (maximum acceleration at ground level of 0.4 g). The pos-
sibility of using seawater for cooling of the auxiliary equipment and the placement of nu-
clear power plants at sites with a humid tropical climate is foreseen.
It has been proposed that nuclear power plants with VVER can participate in providing
for the variable loading diagram of power systems. The requirements on such nuclear power
plants at present provide for the possibility of daily disconnections from the grid for 5-8
h and weekly ones for 24-55 h, enhanced rates of change of the loading from 1-4% of the
nominal capacity per min, and keeping the units in operation in the event of short-term de-
creases in the frequency right down to 46 Hz. It has become necessary to solve a number of
problems and first of all to produce a design for the fuel element which is efficient under
long-term cyclic loads and to verify it experimental'. It is possible that VVER intended
for control of the power and frequency in a system will operate at a lower power intensity
than baseline reactors but with higher coolant parameters.
An important aspect of the use of VVER is their application for a centralized heat sup-
ply. The technicoeconomic discussion carried out up to the present of the alternatives for
the heat supply of a number of large cities in the European part of the USSR from sources
based on nuclear and organic fuel indicates the advisability of the application of nuclear
heat supply plants (NHSP) for the generation of heat and of nuclear heat-power plants (NHPP)
for the combined generation of heat and electrical power in comparison with boilers and heat
and electric power plants operating on organic fuel. The choice of NHPP or NHSP depends on
the conditions of a given city. No significant variations of any kind are assumed in the
VVER-1000 reactor assembly for NHPP. The heat supply for the consumers is accomplished from
diversions of the steam of the TK-450/500-60 central-heating turbines, which provide a maxi-
mum heat production of 450 Gcal/h at an electrical load of about 450 MW each.
1. A. Ya. Kramerov et al., Third Geneva Conference, USSR Lecture R/304 [in Russian] (1964).
2. V. P. Denisov et al., At. Energ., 31, No. 4, 323 (1971).
3. V. A. Voznesenskii, At. Energ., 44, No. 4, 299 (1978).
4. Yu. V. Vikhorev et al., At. Energ., 50, No. 2, 87 (1981).
5. V. F. Titov et al., The main trends in the development of steam generators for nuclear
power plants with VVER in the Soviet Union," Lecture at the Soviet-Italian Seminar "Con-
temporary Problems of Power Engineering," November 21-24, 1977, Moscow.
6. V. A. Sidorenko et al., At. Energ., 43, No. 6, 449 (1977).
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 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196ROO0300020004-6
SOME CHARACTERISTICS OF AND EXPERIENCE WITH THE OPERATION
OF NUCLEAR POWER PLANTS WITH RBMK-1000 HIGH-POWERED
WATER-COOLED CHANNEL REACTORS (RBMK)
N.
A.
Dollezhal',
I.
Ya. Emel'yanov,
UDC 621.039.56:621.039.524.2.
Yu.
M.
Cherkashov,
V.
P. Vasilevskii,
034.44
L.
A.
N.
P.
Podlazov, V.
Sirotkin, V.
V.
P.
Postnikov,
Kevrolev,
and A. Ya. Kramerov
During the period from the end of 1973, when the first power unit with an RBMK-1000 en-
tered into service at the Leningrad Nuclear Power Plant, to January 1, 1981, the installed
capacity of nuclear power of the Soviet Union grew from 3.2 to 14.6 GW. During this period
8 GW of the 11.4-GW growth in nuclear capacities, i.e., about 70%, was due to the fraction
of nuclear power plants with RBMK-1000. Now nine power units with this reactor are being
operated: four at the Leningrad Nuclear Power Plant (LNPP), three at the Chernobyl Nuclear
Power Plant (ChNPP), and two at the Kursk Nuclear Power Plant (KNPP). In all, nuclear power
plants with RBMK-1000 have generated about 200 billion kWh of electrical power. The signi-
ficant increase in the capacities of the nuclear power plants of the country based on the
RBMK-1000 which has been achieved in a comparatively short period of time, the successful
utilization of their nominal capacity, and the reliability and safety of their operation
testify to the promising outlook for channel uranium-graphite reactors, on which the develop-
ment of nuclear power in the USSR will be based in the succeeding decades. One can add the
following to the list of factors in favor of channel boiling reactors of the RBMK type, which
have been taken into account in the process of the design and construction development and
which completely support the practice of their construction and operation:
The RBMK-1000 is manufactured at operating factories and does not require the construc-
tion of new industrial enterprises with unique equipment;
there do not exist limiting values of the individual capacity associated with the manu-
facture, transporting, and maintenance of the equipment used;
shattering of the main loops increases the overall safety of the reactor, since there is
not complete dehydration of the active zone;
due to the good physical characteristics of the reactors and the continuous fuel re-
charging, the prerequisities are created for highly efficient utilization of weakly enriched
fuel, the attainment of a small content of fissionable uranium isotopes in the exhausted fuel,
and the production of a sufficiently large increase in the depletion due to consumption of
the plutonium made in passing; and
a high thermal engineering reliability of the power units is provided by a broad range
of regulation of parameters by in-channel control.
Approximately 1660 fuel channels and more than 200 special channels of the monitoring,
control, and protection system are positioned in vertical openings of the graphite reactor
stack in a square array with a 250-mm spacing. Two heat-generating assemblies (HGA) with 18
fuel elements in each are mounted inside the zirconium tube of the fuel channel. Pellets
made out of uranium dioxide with an enrichment of 2% in 235U are used as the fuel. The
casings of the fuel elements with an outer diameter of 13.5 mm and a thickness of 0.9 mm are
fabricated out of a zirconium-niobium alloy. Water from the distributing collectors under-.
heated to boiling is individually supplied to each channel. The necessary flow rate is
established with the help of a channel flowmeter and a regulating valve. In the active zone
%15% of the water is converted into steam. The steam-water mixture from each channel is
also diverted into separators through individual pipelines. Saturated steam at a pressure
Translated from Atomnaya Energiya, Vol. 54, No. 4, pp. 257-262, April, 1983.
0038-531X/83/5404-0263$07.50 ? 1983 Plenum Publishing Corporation 263
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 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196ROO0300020004-6
//4\ \Y
H S t //~\\\~,/ s~~ 90 v~
o
a b c d a b c d
Fig. 1 Fig. 2
Fig. 1. The duration of construction of a nuclear power plant
until power-up of the second units: a, b) first and second
turns of the LNPP, respectively; c, d) first turn of the KNPP
and ChNPP, respectively.
Fig. 2. Duration of reactor assembly until the start of the
flushings of the reactor systems (the notation is the same as in
Fig. 1).
7.7
9
0
~~ss~, s9 6
T ;r n r '~\r n H 2
a b c d a b c
Fig. 3 Fig. 4
Fig. 3. Duration of the start-up adjustment operations from the start of
flushings of the systems until power-up of the unit (the notation is the
same as in Fig. 1).
Fig. 4. Duration of utilization of the capacity from power-up of the unit
attainment of nominal capacity (the notation is the same as in Fig. 1).
of 7 MPa is directed into two turbines with a capacity of 500 MW each. The separated water,
mixing with the supply water, is again fed by the main circulation pumps to the entrance of
the channels. The reactor has two circulation loops whose equipment is mounted symmetrically
with respect to the vertical plane passing through the reactor axis in the direction of the
machine room [1].
Paired layout is used in the design of nuclear power plants with RBMK-1000. i.e., two
power units are located in the main building of the nuclear power plant. Each reactor,
mounted in its own shaft, has an independent circulation loop. The four turbines of the two
units are arranged in series on the same axis in the common machine room which is adjacent
to the main building. The units do not depend on each other, but they have a series of aux-
iliary interchangeable systems, which creates definite advantages in the course of operation
of the nuclear power plants and especially in the maintenance of the equipment. The adopted
layout of a nuclear power plant provides for start of the construction of the main buildings
of the first (I) and second (II) units and maintenance of the equipment practically simulta-
neously (Figs. 1-4). As follows from the diagrams given in Figs. 1-4, the even units are
activated appreciably more rapidly. If assembly of the odd units is accomplished in 1.5-2
yrs, it has proven possible to shorten this time to 8-10 months for the even units. On the
average, the time to carry out the start-up adjustments on the even units is shortened by a
factor of two. The nominal capacity of the leading units was utilized in 8-10 months, and
in 5-6 months on the succeeding even units.
This is explained to a significant extent by the fact that readiness of a large number
of auxiliary systems which are common to both units is required for start-up of the leading
units. In addition, the experience of performing the start-up adjustment on the first power
units has shown that their extent can be reduced on the subsequent nuclear power plants.
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TABLE 1. Some Operational Indices of Nu-
clear Power Plants with RBMK-1000 in
1980
Installed
Production
of electrical
Nuclear power
capacity,
power, kWh
CUC,
plant
MW
Leningrad
3000
18,82
73,0
Kursk
2000
13,89
79,2
Chernobyl
2000
14,2t
80,9
Total
7000
46,92
-
The start-up adjustment operations should be determined by tests of the installed equipment
and complex tests of the systems, but it is advisable to perform only those of the investi-
gative efforts whose results obtained on the preceding units turn out for this or the other
reason to be unsuitable for the subsequent ones. This has served as the basis for the crea-
tion of optimal standard procedures and programs of performing the adjustment of the systems.
In the standard plot which has been developed for utilization of the capacity of a commercial
unit with an RBMK-l000, six months is provided. Further reduction of the periods for utiliza-
tion of the capacity of commercial units has been acknowledged as being inadvisable for the
following reasons: In the first place, a specified time is required for checking the fitness
of the equipment in intermediate stages which permits predicting a safe elevation of the ca-
pacity at the next stage; secondly, experience in control of the unit is acquired by the
operating personnel.
In 1980, 73 billion kWh of electrical power was produced at the nuclear power plants of
the country, and of this amount 47 billion kWh or 64.5% was produced at nuclear power plants
with RBMK-1000. In comparison with the previous year, the production of electrical power at
nuclear power plants increased by more than 35%. This has been achieved not only due to the
introduction of new capacities, but also owing to the high value of the capacity utilization
coefficient (CUC). The average value of the CUC for nuclear power plants with RBMK-1000 ex-
ceeds 76% (Table 1).
The data on the operational readiness of the reactor equipment of the power units is of
interest. In 1979 the operational readiness coefficient (ORC) on the first turn of the Lenin-
grad Nuclear Power Plant reached 85%, at the KNPP 84.5%, and at the ChNPP 88.7%, with CUC
values of 74.4, 73.1, and 74.6%, respectively. These same high values of the ORC also charac-
terize the operation of RBMK-1000 in 1980. For example, at the ChNPP the number of hours of
operation of the reactors of the first and second units was 7522 and 7622 h, and out of these
the reactors operated for 6899 and 7313 h at nominal capacity. At the KNPP the number of
hours of operation of the reactors of the first and second units was 7677 and 7642 h, and the
reactors operated for 6999 and 6890 h at nominal capacity. The cited operational indices of
nuclear power plants with channel reactors are not inferior, judging from the published data,
to the best operational indices of foreign nuclear power plants with reactor housings of
equal capacity, both boiling and with water under pressure.
Deep depletion of the nuclear fuel with a low initial enrichment is characteristic of
RBMK-type reactors, which is provided for by continuous fuel recharging at the operating facil-
ity. Fuel recharging at capacity is constantly accomplished at all nuclear power plants
with RBMK-1000 with the help of an unloading-loadirig machine. A regime of continuous re-
chargings permits increasing by approximately a factor of two the fuel depletion depth in
comparison with the regime of one-time complete recharging of the active zone. The 235U con-
centration decreases from 18-20 to ,, 3.7 kg/ton of U, and the amount of fissionable plutonium
reaches ti 2.8 kg/ton of U. With such a change in the isotopic composition of the fuel, the
neutron-physical characteristics of the cell are significantly altered. If in the steady-
state regime of fuel recharging only the local characteristics (e.g., the power) of the chan-
nels are altered but the characteristics of the reactor as a whole remain practically con-
stant, then the most important changes in its physical characteristics, in particular, the
reactivity coefficients (steam, thermal of the graphite, thermal from heating up) occur
during the initial period of operation of a reactor loaded with fresh fuel and additional
absorbers. The values of these coefficients depend not only on the isotopic composition of
the fuel, but also on the presence of absorbers in the active zone.
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Experience with the operation of the RBMK-1000 has confirmed the theoretical conclusions
that as the fuel is depleted and the absorbers are withdrawn, the reactivity coefficients in-
crease and the stability of the energy distribution decreases. A radial-azimuthal energy
distribution, for which the form of the nonsteady deformations is determined by several of
the lowest harmonics, turned out to be the least stable. Measures related to stabilization
of the energy distribution have been carried out in two directions:
an increase in the automation by virtue of the creation of a branched system for regula-
tion of the reactor; and
a purposeful change in the composition of the nuclear fuel.
As a result of the development of measures of the first direction, a qualitatively new
system of local automatic regulation of the energy distribution (LAR) and local emergency
protection (LEP) which operates from intrazonal detectors [2] has been created and introduced
into operational practice. The LAR system fulfills the function of automatic stabilization
of the lowest harmonics of the radial-azimuthal energy distribution. Maintaining a specified
capacity of the reactor, this system can, by virtue of auxiliary elements operating in the
individual mode, automatically regulate the capacity in individual regions of the active zone.
The LEP system accomplishes emergency power reduction in the case of local bursts of it, which
can arise due to the failure of LAR elements or for other reasons. A structural peculiarity
of the LAR and LEP consists of the use, for regulation of the capacity and protection of the
reactor, of groups of (from 7 to 12) slave mechanisms with a regulating rod uniformly posi-
tioned in the active zone and surrounded by two LEP detectors and four LAR detectors. The
average correction signal of the LAR detectors is used to control the rods. Triaxial chambers
located in the central hermetic sleeves of the HGA serve as the detectors of the LAR-LEP sys-
tem. As follows from operating experience the LAR-LEP system has exhibited high reliability
and effectiveness.
Computational investigations of the effectiveness of the measures of the second direc-
tion have shown that when the initial enrichment of the fuel in 235U is increased, not only
do the dynamic properties of the reactor improve, but its technicoeconomic indices also in-
crease due to an increase in the depletion depth and a decrease in the specific consumption
of nuclear fuel. An important dependence of the variation of the time constant of the first
azimuthal harmonic of the deformation of the energy distribution (To,) on the steam reactivi-
ty coefficient has been established. The smaller the value of the positive steam reactivity
coefficient, the higher the stability of the energy distribution and the simpler the moni-
toring of the reactor. The most rational method for decreasing the steam coefficient is an
increase of the ratio of the concentration of 235U nuclei and the moderator nuclei in the
active zone. A decrease in the steam coefficient due to a change to a fuel of 2% enrichement
is estimated to be approximately 1.3 8, where 8 is the effective fraction of delayed neu-
trons. These conclusions have served as the basis for the adoption of the solution of in-
creasing the enrichment of the RBMK-1000 fuel to 2% (Table 2).
The 8-yr operation of systems which provide for the control and regulation of the energy
distribution in RBMK-1000 has confirmed the correctness of the engineering solutions which
have been taken as the basis for their development. The combined and consistent functioning
of the three systems the monitoring and protection system (MPS), which operates off lateral
ionization chambers; the system for physical control of the energy distribution (SPCED) with
respect to radius and height of the active zone, which uses 8-emission neutron detectors of
the cable type; and the SKALA system for centralized control (SCC) - has facilitated the
reliable control and regulation of the energy distribution in all operating modes of the re-
actor. The accumulated experience of the assimilation and subsequent operation of the moni-
toring and control systems has permitted developing and incorporating measures directed at
a further increase in the reliability of their operation. Among these measures one can count
the conversion of the logic portion of the MPS to more reliable integrated circuits, which
have permitted appreciably developing its functional possibilities with a reduction by se-
veral times in the dimensions of the electronic equipment, the replacement of the cable link
in the slave mechanisms of the MPS by a belt link to increase their operational reserve, and
the introduction of noncontact thyristor circuits for strong control of the MPS servomecha-
nisms. The service term of the detectors for control of the energy distribution with respect
to the radius of the active zone exceeds the operating time of the HGA in which they are
mounted. In order to increase the reliability of operation of the detectors, soldered connec-
tions have been replaced by welded ones. The detector assemblies for control of the energy
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 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6
TABLE 2. Basic Characteristics of the RBMK-1000 Fuel Cycle
Characteristic
Initial 235U enrichment, %
1,8
2.0
Uranium depletion, MW-days/kg
18.5
22.3
Final 235U content in unloaded fuel, kg/ton
3.9
3.5
Decrease in steam reactivity coefficient, $
-
-1.3
Ann. cnsmptn. of enriched uranium per
50.5
41.5
priming, tons/GW?yr
Ann. cnsmptn, of fuel elements per priming
16.3
13.3
of reactor, thousands/GW?yr
Ann. cnsmptn. of natural uranium per
169
159
priming, tons/GW?yr
Oper. period of fuel, eff. days
1000
1350
distribution with respect to the height of the active zone preserve their effectiveness for
4 yrs.
A great deal of attention has been devoted to the perfection of thermal automation and
emergency protection systems in the interests of increasing the reliability and safety of
the operation of nuclear power plants with RBMK-1000. The equation of kinetics, hydrodynam-
ics, and heat transfer and algorithms of the operation of the equipment and systems for
automatic regulation of the parameters of a nuclear power plant are used in a mathematical
model which has been developed for the investigation of transition and emergency conditions
[3]. Upon comparison of the results of calculations with the data of the dynamic processes
on operating units with RBMK-1000, it has been established that the model satisfactorily de-
scribes the dynamics of the power unit. Some emergency conditions associated mainly with the
transition to natural circulation of the coolant have been studied on special test stands.
In order to justify the reliability of the cooling of the active zone under conditions of
natural circulation, three series of experiments have been performed under natural conditions
on the first and third units of the LNPP and the second unit of the KNPP in steady-state and
transitional regimes. As a result of the computational-experimental investigations, a set
of measures have been developed which raise the reliability and safety of operation of the
unit. One should point out the main ones:
Automatic reduction of the reactor power right down to its complete shutdown is intro-
duced with emergency reduction of the flow rate of the supply water;
the modes of operation of the automatic steam-discharge devices and the number of main
steam safety valves are optimized;
supplementary emergency protection of the reactor for a number of engineering parameters
(reduction of the flow rate in the circulation loop, an increase in the pressure in the re-
actor space, and dehydration of the MPS channels) is introduced; and
a system of automatic regulation of the level and pressure in the separators is con-
verted into a new element base of the KASKAD type which possesses the best characteristics,
and the structure of the regulation system has been improved.
Based on the results of the start-up adjustment operations, experimental investigations,
and operating experience, some changes in the construction of the individual reactor subas-
semblies and the equipment of the circulation loop have been introduced. A large part of
further structural improvement has been accomplished not only on the nuclear power plants
being designed with RBMK-type reactors, but also on the nuclear power plants which are
operating and being constructed. For example, a redesign of the pipelines of the steam water
communications is being performed, a rearrangement of the steam pipes in the space of the
separator rooms is being carried out, and optimal shimming of the steam-discharge fittings
of the separators has been introduced for equalization of the steam loads and elimination of
misalignments of the levels lengthwise and between adjacent separators.
Experience with systematic preventive and capital maintenance of the equipment of oper-
ating power units with RBMK-1000 has shown that in order to shorten the periods for carrying
out the maintenance operations and to increase their quality, it is necessary to improve the
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 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6
1000
N P 0 PD ffi Ig
1980
Fig. 5. Plot of the operation of the second
unit of the ChNPP in 1980: 1) disconnection
of the turbogenerator (TG) for the elimina-
tion of flaws on the T-junction of the se-
parator-steam generator (SSG); 2) disconnec-
tion of the TG for replacement of a section
of pipeline of warming steam condensate of
the SSG; 3) disconnection of the TG due to
failures of the automatic equipment of the
pumps of the machine room; 4) shutdown of
the unit for the elimination of leaks in the
condensers; 5) planned shutdown of the unit
for average maintenance; 6) shutdown of the
unit for elimination of a leak in the MPS
cooling loop; 7) planned shutdown of the
unit for current maintenance.
maintenance technology and the methods of cooling a shutdown reactor. A method of prelimina-
ry formation of icy stoppers in the underwater communications has been developed for masssive
replacement of the pressure-regulating valves and the detectors of flowmeters of the fuel
channels after they have exhausted their reserve capacity. With this method of refriger-
ating, the necessary operations for maintenance of the indicated subassemblies is accomplished
4-5 times faster than when the design technology is used. A special system for diverting
the residual heat generation with forced circulation of the coolant has been developed for
the maintenance of the pipelines of the circulation loop without unloading the fuel from the
active zone.
The intensive development of nuclear power has not only raised a series of immediate
problems concerning providing for the safety of nuclear power plants, but has also required
a tightening up of the requirements which are presented to the technical means for safety
provision. First of all, this refers to emergencies associated with depressurization of the
pipelines of the circulation loop. An instantaneous complete break in the pipeline with a
maximum diameter of " 900 mm is adopted as the maximum design emergency. The technical means
of safety provision, the principal ones of which are the emergency reactor cooling system and
the accident localization system [4], are calculated for this emergency. In order to deter-
mine the parameters and characteristics of these systems, a set of scientific research and
experimental-structural operations is performed. As a result, systems have been designed
which provide for an acceptable thermal regime of the fuel elements upon a fracture of any
pipeline of the circulation loop and for localization of coolant ejections.
It is well known that the fuel temperature, the temperature of the graphite stack and
the metal structures, and the margin of heat exchange until a crisis are the determining
parameters which limit the power of channel uranium-graphite reactors with a boiling coolant.
These parameters in an operating RBMK-1000 do not reach the limiting permissible values.
Thus, the maximum power of a fuel channel at the nominal reactor power is about 2600 kW with
a permissible value of 3000 kW, the maximum temperature of the graphite stack is 550?C with
a permissible value of 750?C, the maximum temperature of the metal structures is 300?C with
a permissible value of 350?C, and the margin of heat exchange until a crisis is no lower
than 1.05-1.06. A plot of the load of the second unit of the ChNPP by months is presented
in Fig. 5.
Experience with the successful operation of power units with RBMK-1000 at nominal capa-
city and the presence of reserves in the operation of the reactor equipment indicate that one
can increase the reactor power without changing the dimensions and number of fuel channels
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 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6
by increasing the critical power of the channels at which a heat-exchange crisis arises [5].
This problem has been solved by means of applying heat-exchange intensifiers in the HGA.
Test-stand experiments have shown that the power of an RBMK channel with intensifiers in-
creases by approximately a factor of 1.5. The construction of a new HGA, with special me-
chanisms permitting an increase in the thermal loads, which has been developed for the RBMK-
1500, has a high level of unification of the individual subassemblies with the HGA of the
RBMK-1000.
At present, the first turn of the Ignalina Nuclear Power Plant with two RBMK-1500 units
having an electric capacity of 1500 MW each is being constructed.. Power-up of the leading
unit will mark the start of the creation of a new generation of channel reactors, which,
being more economical, should be a replacement for the well-recommended RBMK-1000. The con-
struction of nuclear power plants with RBMK-1500 will permit reducing by 20-30% the specific
capital expenditures in comparison with nuclear power plants with the RBMK-1000 and reducing
the cited costs for electrical power.
Experience with the operation of channel uranium-graphite RBMK-1000 with boiling coolant
confirms the validity of the adopted solution of creating in the USSR a large series of nu-
clear power plants with reactors of a given type. The accumulated experience in creating
powerful channel power reactors is a good basis for their further refinement and development.
1.
N. A. Dollezhal' and I. Ya. Emel'yanov, A
Atomizdat, Moscow (1980).
Channel Nuclear Power Reactor [in Russian],
2.
I. Ya. Emel'yanov et al., At. Energ., 49, No. 6, 357 (1980).
3.
I. Ya. Emel'yanov, S. P. Kuznetsov, and Yu. M. Cherkashov, At. Energ., 50, No. 4, 251
(1981).
4.
I. Ya. Emel'yanov et al., At. Energ., 43, No. 6, 458 (1977).
5.
N. A. Dollezhal' and I. Ya. Emel'yanov, At. Energ., 40, No. 2,
117 (1976).
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 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6
0.
D.
Kazachkovskii, A. G. Meshkov,
F.
M.
Milenkov, V.
P.
Nevskii,
L.
A.
Kochetov, V.
I.
Kupnyi,
B.
I.
Lukasevich,
V.
M. Malyshev,
V.
V.
Pakhomov, F.
G.
Reshetnikov,
A.
A.
Samarkin, M.
F.
Troyanov,
V.
I.
Shiryaev, V.
A.
Tsykanov,
and D. S. Yurchenko
The concept of the construction of fast reactors in the Soviet Union was promoted at the
end of the 1940s by A. I. Leipunskii. It was based on the supposition about the more advan-
tageous neutron balance, from the physics point of view, in a reactor with a "hard" spectrum.
More than 20 yrs of intensive scientific-research and experimental-constructional work have
been required in order to traverse the path from physical guesses to the first BN-350 reac-
tor prototype. In order to consolidate the new trend in nuclear power generation, weighty
arguments concerning the advantages of fast reactors by comparison with the simpler reactors
were necessary, their success already having been demonstrated at this time, but such success
was also due to the large efforts ground of nuclear power generation based on thermal reactors.
A number of fundamental problems had to be solved in the fields of physics, heat-mass ex-
change, material behavior, chemistry, economics, and also time, in order to realize more
clearly the acute necessity for the development of fast reactors.
Among the most important questions requiring resolution at that time, the following may
be mentioned:
What realistic value of the breeding factor can be obtained in future large-sized power
reactors?
It is possible to ensure the necessary safety of fast reactors and is control (automatic
or manual) at all possible with such a reactor?
What specific power intensity must be ensured and what coolant most completely meets the
demands of a fast reactor?
What structural and fuel materials can satisfy the requirements of fast reactors?
How practicable is the industrial chemical reprocessing of spent fuel and the commercial
manufacture of fuel elements based on plutonium, and what are the fuel losses in this manu-
facture?
What are the reserves of nuclear natural fuel and how much time will be available to the
community before the mass construction of fast reactors?
In order to answer these questions, the painstaking efforts of theoreticians and experi-
menters were necessary - it was necessary to construct a powerful experimental base. Thermo-
hydraulic and material testing rigs were constructed, upon which research on thermophysics,
hydraulics, and material behavior of liquid metal coolants (sodium, sodium-potassium, and
mercury) were conducted; chemicotechnological rigs were constructed for the development of
monitoring and purification of coolants; physics rigs (BR-1, BFS-1, BFS-2, and "Kobra") were
constructed for the study of physics problems; and experiments on the fuel and the whole fuel
cycle were developed. The joint work of the theoreticians and experimenters allowed the ini-
What grouping (design) of the reactor will ensure the optimum value of the breeding fac-
Translated from Atomnaya Energiya, Vol. 54, No. 4, pp. 262-273, April, 1983.
0038-531X/83/5404-0270$07.50
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 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196ROO0300020004-6
tial concepts to be refined significantly and allowed convincing proofs of the long-term
outlook of the trend to be obtained. It was essential to proceed to the next stage of the
work -the construction of experimental facilities and design work on prototype plants.
The construction of the research reactors BR-2 in 1956 (capacity 100 kW, coolant mercu-
ry), BR-5 in 1968 (5 MW, sodium), and the experimental power reactor BOR-60 in 1969 (60 MW,
sodium) was an extremely important stage in the history of fast reactors. The development
and operation of these reactors allowed the choice of the principal decisions to be con-
firmed finally for the BN-350 and BN-600 demonstration reactors: oxide core, sodium coolant,
dense packing of the fuel elements in the fuel element assembly. Very valuable experience
was obtained in the operation of sodium radioactive, circuits. The feasibility emerged of
proceeding to bulk radiation tests and investigations of different structural materials and
types of fuel (mixed oxide, carbide, and carbide-nitride), and to tests of steam generators.
The feasibility was shown of extended nuclear fuel breeding, using already tested decisions
of the core with a breeding factor of 1.30-1.45. A great deal of experience was built up on
transition regimes and safety; the high degree of safety and the excellent controllability
of fast reactors with sodium coolant were confirmed (BN reactors). To summarize, all this
allowed, with the necessary degree of justification, complete conversion to design work on
demonstration facilities, and the gradual solution of problems of reliability and efficiency
and of those scientific-technical questions which most of all are associated with large-scale
facilities (fuel cycle, verification of different concepts of plant grouping and different
plant designs, especially of the steam generators).
The current stage of development of fast reactors in the Soviet Union if characterized
by the buildup of experience in the operation of three successfully active power reactors:
the experimental BOR-60 and two large commercial reactors, the BN-350 and the BN-600. Opera-
tion of the BR-10 research reactor is continuing. In the present paper it is proposed to
discuss only the results of recent years, as earlier information has been given repeatedly
[1, 2].
BR-10 Reactor. From 1973 to 1979, the reactor has been operated at a power of up to
7.5 MW; plutonium dioxide was used as the fuel. On October 1, 1979 the reactor was shut down
for overhaul. At this time, the maximum fuel burnup attained 14.2%.* All standard and ex-
perimental fuel element assemblies were removed from the reactor and the hermeticity of the
fuel element cans was tested. According to preliminary data, the hermeticity of the cans was
destroyed for "1% of the fuel elements. Nevertheless, this did not prevent the completion
of the planned program and subsequent conducting of the repair work. After draining the so-
dium from the primary circuit, it was washed out three times by the steam-gas method, with a
final washing with distillate. Samples were cut out at various points of the primary circuit
for investigations, which confirmed the satisfactory state of the material of the circuit
(steel OKh18N9T). At the instant of shutdown, the main reactor vessel was irradiated with
a fluence of 8.1022 neutrons/cm2, which corresponds to "-40 displacements/atom. Measurements
showed that at the site of maximum fluence, the diameter of the central duct was increased by
3.10? 0.27 mm (swelling of the steel AV/V = 2.8%). The high fluence received by the material
of the reactor vessel also was one of the principal reasons for shutting down the reactor for
overhaul with a replacement vessel. By means of a specially developed tool and protective
facilities, the vessel was cut off from the primary circuit and withdrawn from the reactor
shaft. The overhaul also provided for the replacement of part of the main plant (cold trap,
pump), reactor monitoring and control systems, and electrical heating and emergency cooling
systems. At the end of 1981, the new vessel was installed in the shaft and completely joined
to the primary circuit. The overhaul work is continuing. The next fuel charge is being pre-
pared, based on uranium nitride.
BOR-60 Reactor. Major work has been carried out recently on the BOR-60 reactor, namely:
An extensive material testing program has been carried out, intensive investigations
have been carried on astudy of the radiation effect on the behavior of austenitic and fer-
rite-inartensitic steels and fuel and moderating materials;
*Here and below, we have in mind the fraction of heavy atoms.
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a micromodulular steam generator of Czechoslovakian construction has been tested (1973-
1981), and investigations of a large-scale model of the BN-600 steam generator have been con-
ducted since 1978; in 1981, tests of the so-called reverse steam generator, also developed
by specialists of Czechoslovakia, were started; and a study of the monitoring and safety sys-
tems of the steam generators is continuing;
a program and experimental equipment for carrying out work on a study of sodium boiling
in the core* have been developed, a study of the behavior of radioactive corrosion products
and fission products in the circuit has been continued, and methods of purifying the sodium
of the primary circuit from the most dangerous radionuclides have been investigated.
Recent work has been completed on the creation of a design for a trap based on graphite,
which has undergone tests, in the BOR-60 and then in the BN-350. By pumping the sodium through
the trap in the BOR-60, 1,20 TBq (540 Ci) of cesium were removed from the circuit, and the
y background in the primary circuit compartments was reduced by a factor of ti2.6 [3].
Among the overall achievements in the field of material behavior investigations, the as-
similation of vibration technology in the manufacture of the fuel elements should be mentioned,
which is interesting from the point of view of setting up an automated process for the produc-
tion of fuel elements of mixed uranium-plutonium oxide fuel. This required the carrying out
of an extensive complex of technological investigations, involving the following:
determination of the conditions for achieving the required density values of the mixed
fuel and uniformity of distribution of plutonium oxide in the oxide mixture;
a study of the dynamics of structural changes in the initially molded filling and vibro-
compacted fuel column during raising of the reactor power;
a study of the redistribution of the fuel components over the height and radius of the
fuel elements during its lifetime tests;
a study of the temperature conditions and swelling of the fuel elements;
a study of gas release [4].
As a result of the investigations carried out, confirmation emerged that fuel elements
prepared from a mixture of oxides by vibration technology are able to provide the same power
intensity of the core and the same burnup as pelleted fuel elements. For the final verifica-
tion of these preliminary conclusions an additional program was planned, according to which
in 1981 a set of fuel element assemblies was prepared, based on mixed fuel, by vibration
technology; these fuel element assemblies were loaded into the BOR-60 reactor.
For the purpose of determining the prospects for increasing the breeding in fast reac-
tors, the investigations of a metallic fuel in the BOR-60 are of important value. The in-
vestigations of a metallic uranium and uranium-plutonium fuel were directed at the prevention
of large swelling of the metallic fuel and its significant interaction with the cladding.
As a result of many years of functioning in the BOR-60 reactor, a large number of tests of
experimental fuel elements with metallic fuel have been conducted. A burnup of '6% was
achieved in the experimental fuel elements with.uranium-plutonium fuel, in conditions similar
to those characteristic for the present-day fast reactors with oxide fuel.
BN-350. Since the time of the power generation startup of the BN-350, the first in the
Soviet Union and the most powerful commercial fast reactor at that time, 9 yrs have elapsed.
The only major plant defect which appeared during the whole process of assimilation of the
power of the station was a defect of the steam generators: repeated breakdown of the inter-
circuit sealing [5]. The principal cause of this was the poor quality of manufacture and
welding of the lower end components of the heat transfer tubes. Because of the special fea-
tures of the circulation from the direction of the tertiary circuit (natural circulation in
the Field tubes), concern was caused by the primary pores and by the quality of the feed
water, particularly the iron content in it (15-20 pg/kg). The overhaul of all the damaged
(five of the six) steam generators was completed in 1975, and the power of the facility was
raised to 520 MW (thermal); in March, 1976 it was raised to 650 MW (thermal), and in Septem-
ber, 1980, to 700 MW (thermal). At 700 MW the reactor provides an electrical capacity of
125 MW (elec.) and additionally generates 85,000 tons per day of distillate. In May, 1980,
`Specialists of the German Democratic Republic participated in these investigations.
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assembly was completed of the first steam generator of Czechoslovakian design and it was
brought to the condition of normal operation.
The time-utilization factor over the period from the instant of startup to 1977 amounted
to 86%, and when operating at a power of 650 MW (thermal) and above, it was 88%, which cor-
responds to "7700 h of operation of the facility on power annually.
Other important economic indexes are the attained fuel burnup and the operating lifetime
of the main plant. For the first time, the planned fuel burnup of 5% in the central section
of the fuel element assemblies was attained in 1976. At the present time, this index is
equal to 5.8% and is dependent on the permissible dimensions of the change of shape of the
hexagonal sheaths of the fuel element assemblies.
The initial operating lifetime of the major part of the newly developed nonstandard plant
was exceeded (with the exception of the steam-generator evaporators.). The operating life-
time of the plant as of January 1, 1982 is shown below:
Steam-generator evaporators (after
overhaul)
Steam-generator steam superheaters
Control and safety equipment
Sodium-sodium intermediate heat
exchange
Primary circuit slide valves with
diameter 500 and 600 mm
Cold traps of primary and second-
ary circuits
Lifetime increased from 20,000 to
50,000 h
Lifetime increased from 20,000 to
50,000 h. Maximum operating period
amounts to 57,000 h
45,000-55,000 h
51,000-57,000 h
Operating up to now without replace-
ment
Ensure normal recharging cycle; ob-
served jamming of plugs eliminated
by increasing the sodium tempera-
ture from 200 to 250?C.
Operating without breakdown of seal-
ing for more than 60,000 h
Operating without faults; provide
absolute sealing when closed
Operating up to the present time
without replacement
It can be seen that all the main plant has operated for more than 9 yrs without replace-
ment, including all the steam-generator steam superheaters. The steam-generator evaporators
after overhaul also demonstrated the considerable lifetime of accident-free operation up to
55,000 h. The evaporators of one regular steam generator operated accident-free for 56,000
h.. At the present time the generator is dismantled and has been transferred-to research; a
second generator of Czechoslovakian design has been installed in its place.
Almost 10 years of operating experience also confirms the high degree of safety of the
facility. Thus, during the whole of this time, there was not one case of sodium leakage from
the primary circuit; in the secondary circuit during the same period, two leakages were re-
corded (in the sampling and oxide indication systems). In each case the leakage did not ex-
ceed 10 liters.
The radioactivity of the discharges into the ventilation duct is determined by 41Ar and
amounted-to not more than 7.4.1011 Bq/day (20 Ci/day), and the radioactivity of discharged
aerosols was a factor of 106 less than the argon activity. One shutdown of the facility oc-
curred during operation, in which all safety devices functioned normally.
Recently, the following systems and plants have been modernized:
The geometrical dimensions of the fuel elements have been unified with the fuel elements
of the BN-600 reactor (diameter 6.9x 0.4 mm); at the same time, the gas compensation space of
the fuel elements has been increased, which has led to a reduction of the pressure. under the
cladding, and to a reduction by a factor of 10 of cases of depressurization of the fuel ele-
ment cans;
the control and reactivity compensation rods have been modernized, the efficiency of the
rods has been increased, and the operating time of the reactor on power between two shutdowns
for recharging has been increased from 55 to 73.5 days;
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the steam generator feed has been converted to water with total and extreme purifica-
tion (desalination), and a cycle of complex feed water processing has been introduced;
the return valves at the head of the primary circuit pumps have been redesigned.
At the present time the reactor is also being used for experimental work in physics,
material behavior and sodium technology. Among the most important tasks and achievements
here the following should be mentioned:
a cycle of experimental work on refining the breeding parameters (conversion). Measure-
ments carried out have allowed the experimental value of the conversion factor 1.05? 0.05 to
be established, which agrees quite well with that predicted by a computational method (1.03);
a cycle of work to study the changes of shape and mechanical properties of materials
in conditions of irradiation with high fast neutron fluences; in individual fuel element as-
semblies, a burnup of 6.6% has been achieved and the maximum burnup in an experimental fuel
element assembly attained 7.7%. Investigations of the change of shape of spent fuel element
assemblies have shown that as a consequence of radiation swelling and radiation creep, the
diameter of the hexagonal can is increased on the average from 96 to 97.2 mm, with a sag of
15-17 mm.
These and other experimental studies have been conducted in the BN-350 and not to the
detriment of the planned tasks on the generation of electric power and distillate, the ful-
fillment of which is the main criterion for assessing the activity of the staff and the ef-
ficiency of the facility.
NB-600. In contrast to the BN-350, the grouping of the plant of this reactor is inte-
gral; the diameter of the vessel is 12.8 m and the height 13.0 m.
Moreover, in contrast to the facility with the BN-350, in the unit with the BN-600
straight-through modular steam generators are used, which significantly increase the thermo-
dynamic parameters of the steam and the temperature of the sodium circuits.
The design characteristics of a unit with the BN-600 have been described in quite some
detail in [6]. Therefore, we shall confine ourselves only to the most important general con-
clusions which can be made on the basis of the experience gained with the construction and
startup of the third unit of the Beloyarsk nuclear power station.
The vessel and the intravessel structure, including the whole of the primary circuit,
were assembled in an assembly area. Experience has shown that the preparation and assembly
of such a unique plant could be carried out successfully, and any difficulties encountered
as a result of this could be overcome.
The high saturation of the reactor with metal structures and plant caused concern be-
cause of the possible vibration of the intravessel equipment, access to which after filling
the reactor with sodium, and especially after bringing the reactor to power, is extremely
difficult. Therefore, design measures were taken for the installation of a large number of
vibro- and strain-sensors, and special programs have been compiled for the stage-by-stage
verification of the vibrocharacteristics of the reactor. Not one sensor has recorded vibra-
tion in excess of the limit of sensitivity (0.6 mm).
Among the problems associated with the sodium coolant, two have required close attention.
The first was the assurance of the necessary quality (impurities, suspended matter) of the
coolant; concern was caused by the narrowness of the reactor space, especially in the last
stage of assembly work, and by the large volume of welding operations inside the vessel. The
measures taken were successful in providing a blocking temperature of 150-155?C. The second
problem was the assurance of safety during acceptance, cleansing, and pressure transfer from
railroad tank cars to the primary and secondary circuit of 1800 tons of sodium. It should
be mentioned with satisfaction that during these operations with sodium, and also during the
startup-adjustment operations, there was only one case of sodium leakage through the sealing
of the detachable section of the tank to the intake pipeline, and a few leakage droplets were
noted through the sealing of the'sodium slide valves of the steam generators. None of the
personnel came to harm.
In order to refine the physical characteristics of the reactor, a program was prepared
and executed for the physics startup and physics measurements. As a result, the critical
loading was refined, the efficiency of the control rods and the fuel element assemblies and
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 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6
n
F1 I
J,
Fill
i I II
I I I I
i II I II
0.4 0.6 0..8 10 12 0.1 0.3 0.5, 0.7 0.9 1112
Months
Fig. 1. Enlarged graph of the loading of
the third unit of the Beloyarsk nuclear
power station [- -- -) according to plan;
)): a) 1980, power utilization factor=
26.8%, breeding factor = 68%; b) 1981, PUF=
50.1%, BF = 71.1%.
the power distribution over the core volume was measured, and the reactivity coefficients,
essential for operation, were obtained. The measurements showed the excellent agreement of
these parameters with the calculated values. The efficiency of the scram rods was found to
be somewhat less than the design figure; however, it ensured completely safe operation of the
reactor. A slight nonagreement of the temperature and power effects of reactivity also ap-
peared. The series of physics measurements carried out allowed the power startup of the re-
actor to proceed and allowed the subsequent assimilation of the operating and power regimes
of the facility.
The conclusions, made on the basis of experience of more than two decades of operation
of the facility on power, reduce to the following:
The plant of the facility operates stably and reliably. The operating characteristics
of the plant and circuits correspond mainly to the design values. This has allowed the power
to be raised systematically up to 90% at the end of 1981 (see Fig. 1). The power restrictions
of the facility were due to the permissible operating parameters of the fuel elements with
fresh fuel during transition of the core to the steady state, which was achieved in the first
half of 1982. On December 18, 1981, the reactor was brought to nominal power and it was
operated at this level during three days for the series testing of all systems of the facili-
Investigations of the hydraulics of the primary circuit indicated excellent agreement
between the design and actual characteristics in all operating regimes of the facility. As
a result of measurements by means of a flow meter device, installed in the rotatable plugs,
it was found that the sodium flow rate through the fuel element assemblies was close to the
design figure.
During operation of the BN-600, the inadequacy of mixing of the relatively cold coolant
leaving the fuel element assemblies of the radial breeding zone, with the hot coolant leaving
the fuel element assemblies of the core, was verified. The observed laminar flow of coolant
at a different temperature leads to temperature instability in the region of the thermocuples
installed in the reactor mixing chamber. Therefore, in order to regulate the reactor, thermo-
couples installed above the caps of the fuel element assemblies are used.
The hydraulic resistance of the loop of the secondary circuit is somewhat lower than the
design value. Nominal flow rates in the loops are ensured with a speed of rotation of the
pumps of ti720 rpm. Because of this, the operating speed range was limited to 250-720 rpm,
which is achieved without difficulty with systems for regulating the number of revolutions
and with a control system.
Fuel rechargings showed that the combination of mechanisms of the fuel element assembly
recharging system allows the necessary number of fuel element assemblies to be replaced in
the remote and automatic control regimes with small losses of time (not more than 1 h per re-
placement of one fuel element assembly).
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The startup-adjustment operations revealed the advisability of certain changes in design
of the scram rod slave mechanisms in order to increase their reliability in the case of erro-
neous withdrawals of the mechanism pickup into the rigid support. The design studies of the
BN-800 and BN-1600, taking account of the experience in operating the BN-600, confirmed the
feasibility and advantage of using, for all three of these facilities, a unified design of
the scram rod pickup mechanisms.
In the initial period of operation, several cases of leakage in the steam generator
tubes were noted. The leakage monitoring system detected a looseness in the early stages of
their occurrence.
The separation of the steam generator into sections and modules was justified, since
when individual modules were out of action, operation can be continued after cutting off the
defective modules by means of the armature.
The most probable cause of leakiness in the steam generator tubes is the development of
microdefects in the welded joints and in the material of the tubes in operating conditions.
These defects are below the threshold of sensitivity of the instruments for monitoring the
quality of the materials during manufacture, because of which the necessity arises for in-
creasing the sensitivity of the monitoring devices and for constructional measures to reduce
thermocyclic stresses in operating conditions.
For the initial period of operation, taking account of the design of the fuel element
assemblies and the materials used, the maximum values of the fuel burnup were established:
5% for zones of low enrichment and 7% for zones of high enrichment. When achieving this fuel
burnup, no difficulties arose due to radiation phenomena of swelling and creep of the struc-
tural materials. For the purpose of increasing the fuel burnup, jacketed tubes for the fuel
element assemblies and the fuel element cans were provided, manufactured in the cold-deformed
state, and also some variations of their dimensions, in order to increase the compensation
power of the reactor core by comparison with the radiation change of dimensions of the
jacketed tubes. The use of new, more radiation-resistant structural materials is also being
considered, which will allow the fuel burnup to be additionally increased and thereby improve
the economic indices of the facility.
In conclusion, we note that the new design solutions on the BN-600 reactor (integral
grouping, design of the governing plant, and decisions on the main systems) were justified.
The facility can operate reliably at a power of up to 100% nominal, ensuring the buildup of
much practical experience for the improvement of both the BN-600 reactor itself, and the
newly developed facilities BN-800 and BN-1600.
The successful operation of the BN-350 and BN-600 has shown the practicability, reliabil-
ity, and safety of the new prospective trend in nuclear power generation. The BN-800 and
BN-1600 designs being developed are intended for commercial introduction. The basis of these
reactors is the experience and achievements obtained with their forerunners. The main prob-
lems of the new developments, in addition to large-scale changes, reduce to the following:
a further increase of plant reliability and safety of the station as a whole, improvement of
the economic indices, and increase of secondary fuel breeding.
At the present time, the technical designs of the BN-800 and BN-1600 have been developed.
Their principal design parameters, with the exception of power, are similar.
The major part of the structural-grouping decisions on the BN-800 is similar to the de-
cisions on the BN-600. The principal differences between the BN-800 and the BN-1600 reduce
to the following:
The volume of the core is increased because of the increase of its height from 750 to
950 mm and the increase in the number of fuel element assemblies; the use of a three-zone
scheme of power smoothing in the core (owing to fuel of different enrichment);
the gaps between the fuel element assembly jackets are increased in order to increase
the fuel burnup (up to 'l0%);
the number of recharging mechanisms is reduced (one instead of two) due to the increase
of the number of rotatable plugs (three instead of two);
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the design of the bearing subassembly of the vessel has been changed, which will allow
the strength and vibroresistance of the bearing collar to be increased, the stresses in the
bottom of the vessel to be reduced, the dimensions of the forging to be increased, and the
number of welded seams to be reduced;
an additional system of reactor cooling is provided, by means of air coolers, connected
in parallel with the main coolant course of the secondary circuit;
the scheme for connecting the steam generators to the sodium circuit has been changed
(the sodium intermediate steam heater has been excluded, and in place of it live steam heating
in the superheaters, installed in the machine hall of the station, has been introduced; the
surface area of the heat exchanger has been reduced because of the reduction of the number
of modules in the steam generator section - up to 20 per steam generator instead of 24.
The increase of electric power of the BN-800 by comparison with the BN-600 is provided
with approximately the same capital expenditure on the plant, which is one of the main fac-
tors for the improvement of the economic indices. Moreover, a considerable economy is
achieved because of the use in the BN-800 of the major part of the plant developed for the
BN-600.
The BN-1600 reactor is also being constructed for series commercial nuclear power sta-
tions with a large capacity. The thermal layout and the installation decisions are mainly
similar to the decisions of the BN-600 and BN-800. The grouping of the reactor is integral.
The plant of the primary circuit is disposed in the reactor vessel, with a diamter of til9 m.
Preference is given to mounting of the reactor vessel in the upper part to a shielded strong
structure, absorbing the mass of the entire plant and coolant of the primary circuit. This
has necessitated the construction of screens, providing shielding of the reactor vessel roof
and the upper strong ceiling from the action of the high temperature and thermal cycling.
The structural changes in the BN-1600 are related mainly with the increase of power and
the corresponding increase of dimensions of the fuel element assemblies.
For the purpose of increasing the operating efficiency of the BN-800 and BN-1600, the
experience of the BN-600 is being analyzed industriously and, taking it into consideration,
the necessary improvements are being introduced. Work must be continued on raising the quali-
ty of manufacture of the steam generators over the whole cycle - from the choice of materials
to pilot tests of the steam generators ready for delivery. Special attention should be paid
to the welding to tubes with the pipe panel (improvement of both the welding technique and
the procedure for subsequent inspection). Although a modular design of a straight-tube steam
generator has been chosen for the BN-800 and BN-1600, other designs are nevertheless being
developed. Specialists of Czechoslovakia are involved in this. Monitoring systems for
leakage in the steam generators are being modernized, and the quality of the accessories is
being increased. Since, in the BN-1600, it is proposed to use higher efficiency equipment
(e.g., pumps), it is planned to construct the necessary experimental base of its development.
The safety of the BN-800 and BN-1600 facilities is being increased by means of the fol-
lowing measures:
Additional improvements in the electromechanical and electronic parts of the reactor saf e-
ty system are being introduced.
Monitoring of the state of the core and the reactor protection units is being modernized.
In particular, it is acknowledged as advantageous to develop a visual observation system
below the sodium layer, similar to the system developed by the French and American specialists.
The fire safety system is being modernized and new methods of monitoring in the case of
the occurrence and extinction of a sodium fire are being developed. In this respect there
is special interest in the latest achievements in the field of passive means of fire extinc-
tion (unified means of sealing off potentially dangerous compartments are being developed;
light substances which quench the sodium which has overflowed over the floor surface have
been well recommended - the main advantage of which is the possibility of their disposal be-
forehand over the surface of the compartment floor).
The system for the removal of residual heat release, due to the introduction of auton-
omous contours of the natural circulation, including a third, air, circuit, is being modern-
ized.
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It is proposed once again to return to the discussion of the advisability of an addi-
tional external housing, calculated for an aircraft crash and the internal pressure of a maxi-
mum sodium fire.
In the present-day stage, until relatively inexpensive uranium is available for the fuel
cycle, better economic indices than in thermal reactors can hardly be achieved successfully
on BN pilot plants. This is due in the first place to the relatively high capital costs on
the BN facility. Thus, the specific capital costs on power plants with BN-600 is higher than
the similar costs on the fifth unit of the Novovoronezh nuclear power station with VVER-1000
by a factor of 1.6. Although here the difference in capacity and climatic conditions is not
taken into account, nevertheless these costs are quite high. Therefore, one can understand
the tendency to reduce capital costs on future facilities with BN reactors. However, this
natural desire contrasts with the tendency to examine excessive means of monitoring, protec-
tion, inspection, and safety on pilot models. Therefore, the planners at the present time
use only those methods of reducing costs which do not enter into contradiction with the as-
surance of safety. Among the decisions to reduce costs on the BN-800 and BN-1600 pilot
plants may be mentioned the use of cheaper materials, where this is possible, to replace
austenitic steels by low-alloy steels (shielded casing, internal neutron and thermal shields);
increase of the capacity of individual plants and stations; and increase of the fuel burnup.
The reduction of the construction times of the stations and transition to series construction
may make a marked contribution to the solution of the economy problem.
Conjunctural considerations at present allow a low breeding coefficient to be accepted
and, moreover, for the present this is even advantageous from the point of view of economy.
However, in proportion to the utilization of the fuel cycle factories and the reduction of
the lifetime of the nuclear fuel, the problem of breeding develops into the category of im-
mediate and most important problems. The dynamics of the development of nuclear power genera-
tion in the Soviet Union indicates the necessity for the development of breeder reactors with
a breeding factor of 1.6. It seems that this breeding factor can be obtained only with the
use of a fuel which is more dense than oxide. For this, a program is being implemented to
study the feasibilities of using metallic uranium in the reactivity control agents, in the
end and lateral breeding zones, and in the core of a heterogeneous reactor. The possibili-
ties of using a uranium-plutonium metallic fuel in the core are also being studied, and a
study is continuing of carbide and carbide-nitride fuel. However, the primary problem is
the use of a mixed oxide fuel (and its reprocessing), which will allow the economic indices
of facilities with BN reactors to be improved markedly. After tests of the experimental fuel
elements with mixed oxide fuel in BR-10 and BR-60, the first 10 full-scale fuel element as-
semblies with this fuel will be manufactured and tests will be run in the BR-350 reactor, in
order to refine the breeding factor and to obtain additional data about the effect of dif-
ferent technological factors on the efficiency of the fuel elements. In the future, a
thorough economic comparison is proposed of the different technological alternatives for the
manufacture of fuel elements from mixed oxide fuel for their mass remote-controlled produc-
tion.
Thus, considerable experience has been built up in the Soviet Union on the development
of sodium-cooled fast reactors. This experience confirms the principal theoretical prere-
quisites for the new trend and indicates the feasibility of constructing reliable commercial
facilities, and also shows the paths for ensuring the required nuclear fuel breeding charac-
teristics. The purpose of this is the achievement of the necessary tempos of the economy of
natural uranium and the gradual transition of nuclear power generation to self-provision with
fuel.
1. 0. D. Kazachkovskii et al., At. Energ., 43, No. 5, 343 (1977).
2. L. A. Kochetkov and Yu. E. Bagdasarov, Report at the International Symposium on Design,
Construction and Experience in the Operation of Demonstration Sodium-Cooled Fast Reac-
tors [in Russian], IAEA-SM-225/78, Bologna (1978).
3. N. V. Krasnoyarov and V. M. Polyakov, Preprint of the Scientific-Research Institute of
Nuclear Reactors [in Russian], Dimitrovgrad, 6(459) (1981).
4. V. A. Tsykanov, Preprint of the Scientific-Research Institute of Nuclear Reactors [in
Russian], Dimitrovgrad, 1(454) (1981).
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 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196ROO0300020004-6
5. D. S. Yurchenko et al., Report at the International Symposium on the Design, Construc-
tion and Experience in the Operation of Demonstration Sodium-Cooled Fast Reactors [in
Russian], IAEA-SM-225/60, Bologna (1978).
6. V. M. Budov et al., The BN-600 Fast Neutron Nuclear Reactor - Facility for Power Genera-
tion of the Near Future [in Russian], Nuclex-75, Basel (1975).
PATHS FOR THE DEVELOPMENT OF FAST POWER REACTORS
WITH A HIGH BREEDING FACTOR
S.
B.
Bobrov, A. V. Danilychev, V. A. Eliseev,
UDC 621.038.526
0.
A.
Zhukova, Yu. A. Zverkov, V. G. Ilyunin,
V.
A.
S.
P.
I.
A.
Matveev, A. G. Morozov, V. M. Myrogov,
Novozhilov, V. V. Orlov, I. S. Slesarev,
Subbotin, M. F. Troyanov, and B. F. Shafrygin.
The assurance of nuclear power generation being developed with fuel resources over a
long time is possible with the presence of fast breeder reactors in the system. Estimates
show that for the necessary rate of growth of this multicomponent nuclear power generation,
fast reactors are required with high breeding factors (BF --1.6; T2 %---7-9 yrs). The construc-
tion of these reactors usually is associated with the use of prospective types of fuel - carb-
ide, nitride, and metallic. At the same time, it will be very tempting to consider the pos-
sibility of the partial use of uranium metal fuel in reactors with a well-studied and sound
oxide fuel, within the scope of the heterogeneous core concept. Such a core does not require
high thermal loadings on the metallic fuel elements and, above all, large burnups in them.
In the majority of cases, the burnup for a metallic fuel can be limited to the value pmax
1-2%.* An important question, determining the operating efficiency of metal fuel elements
in these conditions, is the interaction of the metallic core with the cladding at the working
temperature values of the fuel elements. However, there is a path for solving this problem,
consisting in the construction of barrier blankets.
The investigations carried out in recent years on the improvement of the breeding charac-
teristics of fast reactors have stimulated practical interest in the use of these hetero-
geneous cores [1-4]. Heterogeneous cores, in which fuel element assemblies are used with
fuel elements containing oxide-enriched and dense-metallic fuel, operating in these conditions,
may have marked advantages by comparison with the conventional homogeneous cores with oxide
fuel with respect to the principal breeding characteristics [5, 6].
In the results of investigations of heterogeneous oxide-metal compositions, presented
earlier, the dependence of the principal breeding parameters of the fuel on the fraction of
metallic fuel em and the configuration of the inner breeding zones has been analyzed. These
parameters are the breeding gain (BG), the excess plutonium made r, the excess plutonium pro-
duction R in the breeder reactor system, developing at a specific rate w (%/yr), the doubling
time T2 (for a duration of the external fuel cycle Tex = 1 yr), and also certain parameters
determining the safety of the breeder reactor (change of reactivity due to the Doppler DER
effect in the working range of variation of temperature, and the total sodium void reactivity
effect MKNa).
The investigations were conducted on different models of heterogeneous cores. In all
cases the BNAB-78 library of group constants was used in the calculations [7].
The core was represented by a multiplicity of two-dimensional (r, z) cylindrical cells,
containing oxide fuel in the form of "islets" surrounded by a layer of metallic uranium. The
dimensions of the islets were chosen equal to the dimensions of the fuel element (quasihomo-
Here and in the future, burnup refers to heavy atoms.
Translated from Atomnaya Energiya, Vol. 54, No. 4, pp. 269-273, April, 1983.
0038-531X/83/5404-027W7.50 ? 1983 Plenum Publishing Corporation
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AKeff, x 10?
BG BG, x 102
--
1 'I
~
T, Y BG* x9'"; % (DER)
I ~~
Fig. 1 Fig. 2
Fig. 1. Dependence of T2, BG, BG* = BG (1 - In 22 ), and BG-1 on the frac-
tion of metallic fuel cm for reactors with a fine heterogeneous structure
(No. 0).
Fig. 2. Dependence of BG, DER, and X9+41 on the fraction of metallic
fuel cm and the heterogeneity dimensions +41
m (values of X9 of composition
No. 0 are shown for the homogenized model).
geneous composition) No. 0, one assembly (with a diameter "below the key" of 150 mm) No. 1,
a symmetrical module of 7 assemblies (No. 7), 19 assemblies (No. 19), and 61 assemblies (No.
61). The latter composition simulated a small homogeneous reactor with oxide fuel and a me-
tallic shield. The cells comprised the end shields with a thickness of up to 50 cm. The
volume fractions of the core components and the end shield were: fuel 0.40, sodium 0.22.
The density of the oxide fuel in the core was 8.6 g/cm3, and of the metallic fuel 15.5 g/cm3.
The isotopic composition of the plutonium corresponds to that made in thermal reactors.
By varying the fraction of metallic fuel, it is possible to convert from a conventional re-
actor with oxide fuel to a high power homogeneous reactor with metallic fuel, operating in
the sparing regime. For this transition (in the calculations T2) with composition No. 0, the
enrichment of the oxide fuels in Pu did not exceed 30% (all the additional plutonium necessary
or criticality in this computational model was disposed uniformly in the metallic fuel).
Figures 1 and 2 show the dependences of the fuel enriched in the isotopes 239Pu and 241PU
(X9+41), the breeding characteristics, and the safety on the fraction of metallic fuel in the
composition of the reactor core. The following limitations are assumed for the oxide fuel:
maximum burnup pmax = 10%, maximum. linear power of the fuel element g1ax = W/cm.
Analysis showed the following:
With the introduction, into the reactor with oxide fuel, of a small fraction of metallic
fuel, a rapid increase of BG and BF is observed, but with further increase of Em the indica-
tions of saturation of these parameters is appreciable. This particularly concerns the BG,
determining the change of r and R. The retardation of the increase of BG with a large frac-
tion of metal cm in relation to the increase of BF - 1 begins because of the increased con-
tribution of fissions in the raw isotopes (S) and the decrease of a = \61, . As a result, the
specific (per unit of power) production of plutonium increases in the region of large values
of Em more slowly than BF - 1.
In the region of small values of cm, the parameters BF and BG and almost independent of
the geometrical heterogeneity dimensions. With relatively large fractions Em of metallic
fuel, the physical effect of the additional increase of BG, noted earlier by S. M. Feinberg
and due to "rugedization" of the neutron spectrum is small oxide modules with increased en-
richement of X9 41 in them, is important. As a result, an additional increase of BG occurs;
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however, for the stated reasons, it is small. Moreover, in these high-enrichment "modular"
reactors, the problem arises of the rapid increase of the fuel charge, which leads to a signi-
ficant worsening of R and T2.
The doubling time T2 of compositions with a fine heterogeneity structure (No. 0) has a
significant minimum for EoptT2 = 0.3 to 0.4, owing to the rapid increase of BF for small
values of cm and the insignificant increase of the specific fuel charge. A homogeneous reac-
tor with metallic fuel in the sparing regime of operation is significantly inferior with
respect to T2 than reactors with a heterogeneous core and is approximately equivalent to a
homogeneous reactor with oxide fuel.
The optimum compositions with respect to R of heterogeneous reactors are slightly biased
relative to coPt T2 to the side of large values (E?Pt R = 0.5-0.6). Thus, according to the
breeding indices, determined by the combination oT the parameters R, r, and T2, heterogeneous
compositions with metallic fuel fractions in the range Eopt R,r,T2 = 0.3-0.7 should be as-
sumed to be optimal. In this case, by comparison with reactors with oxide fuel, an almost
twofold increase of the excess fuel production and a marked (up to 30-40%) reduction of T2
can be ensured.
The strong dependence of the Doppler reactivity effect on Em associated mainly with
change of "hardness" of the neutron spectrum, will allow preference to be given to hetero-
geneous cores with a small fraction of metallic fuel.
Compositions with a fine heterogeneity structure are most desirable in power-generating
breeder reactors, because of the better power uniformity and the possibility of stabilization
of the power of the fuel element assemblies.
When analyzing the structure of developing nuclear power generation, the following must
be taken into account: If it appears that the increased fuel charge of heterogeneous reactor
breeders will limit the increase of the power of breeder reactors in the initial stage of
their development, because of a shortage of plutonium, then it is economically advantageous -
as shown in [3] - to introduce, into the U-Pu fuel charge of the breeder reactor, slightly
enriched 235U.
OXIDE--METAL CORES OF RADIAL-ANNULAR GROUPING
The groupings of heterogeneous cores are very varied: islet, symmetrical rings, modules,
etc., types. However, from the point of view of the BG their differences are immaterial, as
the main contribution to an increase of the BG is made by an increase of the average density
of the breeder material in the core. When considering a heterogeneous core, the simplest
radial-annular grouping was chosen. It supposes the presence of a central insert consisting
of seven fuel element assemblies and three concentric rings, each formed with one array of
breeder fuel element assemblies. The heterogeneous core of the type considered, with oxide-
metallic fuel, allows the BG to be increased up to 0.46 and T2 to be reduced to 9 yrs. It
may be noted that the fuel charge of the heterogeneous core is somewhat larger by comparison
with the conventionel core (by approximately 12%).
The parametric investigations of the heterogeneous core were carried out for different
cases. When determining the optimum with respect to T2, the number of breeder fuel element
assemblies in the core was varied. Two volume fractions of fuel in the core were considered
(Ef = 0.45 and Ef = 0.35) with two plutonium buildups in the breeder fuel element assemblies
(2 and 4%). For comparison, versions with UO2 in the breeder fuel element assemblies were
also analyzed. The volume fraction of the breeder material in these fuel element assemblies
was assumed to be equal to 0.55 for Ef = 0.45 and 0.35, and the volume fraction of metallic
uranium in them was the same as the volume fraction of fuel in the fuel element assemblies
of the core. The results are shown in Fig. 3. In order to determine the feasibility of re-
ducing the specific fuel charge in the heterogeneous core, the diameters of the fuel elements
and the density of the fuel were varied. When varying the diameter of the fuel elements,
their pitch and linear power were preserved, and when varying the density of the fuel (e.g.,
because of a change of the central opening in the fuel), the pitch, diameter of the fuel ele-
ment, and the linear power were preserved. When determining T2 in these investigations, the
possibility was taken into account of the high fuel burnup in the heterogeneous versions,
because of the reduction of the neutron fluence by comparison with the fluence in a conven-
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V. . , --T--
0 5 10 15 20 25 eibz' %
Fig. 3. Dependence of T2 on the
fraction of the inner breeding zones
Cibz in a heterogeneous radial-an-
nular core: 1) U02; 2) metallic
uranium; -) 2% Pu; - - - -) 4% Pu.
tional core. The safety parameters were also investigated: the sodium void reactivity ef-
fect (SVRE) and the Doppler effect (DER). A model of the BN-1600 reactor was considered,
similar to the foregoing, but with two rings of breeder fuel element assemblies, and also a
model with only one central breeding zone with a radius of 46 cm (34 fuel element assemblies).
In the calculations the volume fractions of material were varied for both the core and the
breeding zones. Versions were analyzed with uranium dioxide and metallic uranium in the
breeding zones. The results of the calculations are shown in Fig. 4.
The following can be seen from the data obtained:
The improvement of the breeding index in a heterogeneous core depends on the volume frac-
tion of the fuel. The effects of heterogeneity are expressed more strongly in less dense
zones, which can be explained by the large neutron leakage from the fuel zones into the
breeding zones. In the "pure" form, i.e., with identical volume fractions and identical ma-
terial in the fuel and breeder fuel element assemblies, the effect of heterogeneity on the
breeding parameters is revealed very weakly: the BG is increased by 0.01-0.02, but T2 is
almost unchanged. Only an increase of the volume fraction (or density) of the breeder mate-
rial affects T2. In dense cores (ef ~ 0.45) a heterogeneous grouping with oxide breeder mate-
rial reduces T2 only insignificantly, so that an appreciable effect is obtained only when
metallic uranium is used. In all cases, the buildup of plutonium in the breeder fuel ele-
ment assemblies has a marked effect on the reduction of T2. The minimum of T2 is shifted to
the side of a large value of the breeder fuel element assemblies in both the case of a denser
core and in the case of using a denser (metallic) breeder material. The minimum of T2 for
oxide-metal cores of fast reactors is obtained with a 30-40% fraction of breeder fuel element
assemblies in the core.
Iti heterogeneous cores, the dependence of the breeding index on the volume fraction of
the fuel is more weakly expressed by comparison with the similar dependence for a conventional
core. This gives the possibility for optimizing the heterogeneous core from the point of view
of the specific fuel charge. For example, in the reactor considered with an oxide-metal
heterogeneous core, a reduction of the fuel density in the fuel elements from 8.6 to 7.1 g/
cm3 reduces the specific fuel charge to the level of the charge in a conventional core, when
the gain in T2 is maintained at the previous level. The gain in T2 is also maintained with
a marked reduction of the diameter of the fuel elements. In this case, together with a re-
duction of the specific charge, an additional effect appears due to the improvement of the
hydraulic characteristics of the fuel element assemblies.
The sodium void reactivity effect depends significantly on the volume fraction of the
fuel in the core. In heterogeneous cores of the annular type, the oxide breeder material
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0,B
03 04 0,5 0,6 0,3 0,4 0,5 Ef
Fig. 4. Sodium void reactivity effect as a
result of the removal of sodium from the
whole of the reactor, as a function of the
volume fraction of the fuel ef: a) U02; b)
metallic fuel; ) conventional core;
-?-.-.) heterogeneous core with central
breeding zone; -- -) heterogeneous core with
annular breeding zone.
reduces the sodium void reactivity effect (SVRE), and a metallic material increases it. Group-
ings of a heterogeneous core are also possible with metal breeding zones, for which the' SVRE
will be lower than in the conventional core. One version of this grouping corresponds to
the arrangement of the breeder fuel element assemblies with uranium metal in the form of a
symmetrical central zone.
The Doppler effect in heterogeneous cores is reduced because of the increased fuel en-
richment. In oxide-metal heterogenous cores, it is reduced additionally because of the
ruggedness of the neutron spectrum, related with the introduction of metallic uranium into
the core. On the whole, the safety parameters in these cores are somewhat worse than in the
conventional core. For the radial-annular type of groupings considered, the SVRE is increased
by til5%, and the Doppler effect over the whole reactor is reduced by 25%. However, this cir-
cumstance can hardly be a serious limitation on the path of development and the introduction
of such cores.
The concept was considered of a core in which the temperature of the fuel element cans
with metallic fuel can be reduced appreciably, without deterioration of the thermodynamic
parameters [6, 8].
A reactor is considered where, from the direction of entry of the "cold" coolant, fuel
elements with metallic fuel are distributed. The relatively low coolant temperature in this
part of the reactor (400-480?C) determines favorable conditions for the operating efficiency
of these fuel elements. In the region of higher coolant temperatures (500-560?C), fuel ele-
ments with oxide fuel are distributed. With the metallic fuel arranged in the low-tempera-
ture region of the core and the oxide fuel in the high-temperature region, the temperature
parameters of a purely oxide reactor are maintained and, at the same time, the BG is in-
creased by a factor of approximately 0.15. The increase of the BG is due mainly to an in-
crease of the internal breeding coefficient of the core, which is important not only from the
point of view of the rate of nuclear fuel breeding, but also optimization of the operating
regime of a high-capacity fast reactor, taking account of the change of reactivity during its
continuous operation.
The application of this design may be different. It can represent a single fuel element
with different fuel along the height in a single cladding. Another version is the use of
fuel element assemblies with two lattices of different fuel elements: with metal and oxide
fuel. In this case it is advantageous to separate the different fuel elements by a filler.
In the reactor being considered with axial arrangement of the oxide and metal fuel of
high enrichment, the subzone with oxide fuel, to a considerable degree, plays the role of a
"seeding" subzone. As a result, despite the relatively low fuel burnup of the metal subzone
(4-5%), the fraction of the depleted initial charge in it is quite high. The corresponding
burnup of the oxide fuel amounts to 8-10%, with an irradiation time of the composite fuel
element assemblies equal to 1.5 yrs. We note that an increase of the average burnup of the
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metallic fuel of more than 5% does not lead to a significant increase of the rate of nuclear
fuel breeding.
A higher rate of breeding in the metal subzone, with a power fraction in it
of
40-50%,
and with a sufficient fraction of burnup of the initial charge for approximately
the
same
conditions, provides a total breeding rate in the composite core that is a factor
1.35 higher than in the reactor with oxide fuel.
of
1.25-
It follows from the analysis presented that the use of heterogeneous oxide-metal cores
is promising for the improvement of fast reactor fuel breeding.
Finally, this solution will require additional investigations on the justification of
the working efficiency of the fuel elements in a heterogeneous core, and it will possibly
lead to some complexity of design and technology for the manufacture of the fuel elements by
comparison with the conventional oxide fuel. However, as a rule, any path for increasing
the breeding rate will give rise to similar problems.
1. D. Thornton et al., Proceedings of an International Conference on "Optimization of So-
dium Cooled Reactors," London, Vol. 11 (1977).
2. A. N. Shmelev and D. C. Yurchenko, Physics of Nuclear Reactors [in Russian], Atomizdat
Moscow (1978), p. 42.
3. V. V. Orlov et al., Report IAEA-SM-244/78 at the Symposium on the Physics of Fast Reac-
tor [in Russian] Aix-en-Provence, France, Vienna (1979).
4. V. V. Orlov et al., Report IAEA-SM-244/76 at the Symposium on the Physics of Fast Reac-
tors [in Russian], Aix-en-Provence, France, Vienna (1979).
5. M. F. Troyanov et.al., Report IAEA-SM-244/81 at the Symposium on the Physics of Fast
Reactors [in Russian], Aix-en-Provence, France, Vienna (1979).
6. V. V. Orlov et al., Report IAEA-SM-244/77 at the Symposium on the Physics of Fast Reac-
tors [in Russian], Aix-en-Provence, France, Vienna (1979).
7. L. P. Abagyan et al., Group Constants for the Calculation of Reactors and Shielding [in
Russian], Energoizdat, Moscow (1981).
8. M. F. Troyanov et al., At. Energ., 49, No. 5, 275 (1980).
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 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6
IN THE USSR
V. A. Sidorenko, 0. M. Kovalevich, UDC 621.039.68(47-57)
and A. N. Isaev
Setting up safety standards in official documents is one of the main courses taken in
ensuring the reliability of atomic power plants (APP) in the USSR [1, 2]. The system of
standards-technical documents (STD) on the structure and modes of introduction at various
stages in the construction and operation of atomic power plants corresponds to the national
conditions of the organization of the national economy and the established division of func-
tions among state supervisory organs regulating the development of nuclear power. Naturally,
the given STD system is not definitively formulated and frozen but is developing continually
in keeping with the growing scales on which nuclear power develops and its range of applica-
tion expands.
The safety of atomic plants is supervised by the following:
The State Committee on the Supervision of Industrial Work Safety and on Mining Supervi-
sion at the Council of Ministers of the USSR (Gosgortekhnadzor SSSR) checks that the construc-
tion of an APP and its equipment are in accordance with the technical safety standards during
the design, erection, and operation.
The State Nuclear Safety Inspectorate of the USSR (Gosatomnadzor SSSR) supervises the
observance of nuclear safety standards during the design, construction, and operation of APP.
The State Sanitary Inspectorate of the USSR at the Ministry of Public Health of the USSR
supervises observance of the sanitary regulations and the radiation safety standards during
the design, construction, and operation of APP, establishes permissible standards of irra-
diation of the plant personnel and the local population, as well as of environmental contami-
nation with radioactive products, and takes the necessary measures that must be carried out
in order to guarantee that these standards are met.
The system comprised of these three supervisory organs in great measure determined the
structure of the set of standards-technical documents on APP safety; in it, documents that
gravitate toward the problems dealt with by the aforementioned supervisory organs emerge from
Ministry of
public health
Gosgortekhnad-
zor
` I
General safety control
regulations [3]
-~
Nuclear safety Radiation safe-
regulations [ i ty standards
[6]
Regulations on
construction
and safe opera-
tion [4]
Fig. 1. Structure of standards-technical doc-
umentation of APP safety in the USSR.
Translated from Atomnaya Energiya, Vol. 54, No. 4, pp. 273-277, April, 1983.
0038-531X/83/5404-0285$07.50 ? 1983 Plenum Publishing Corporation 285
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the main document in three directions. In individual cases this does not rule out horizontal
links, i.e., the incorporation of lower stages of problems - pertaining to different super-
visory organs - into some documents (Fig. 1).
The main standards document on APP safety in the USSR are the "General Safety Regulations
for Atomic Power Plants during Design, Construction, and Operation" [3], introduced in 1973.
This document formulated the main principles concerning the construction of APP and laid down
the basis for coordinated execution of technical and organizational measures to ensure safety
at all stages of construction and operation. The scope of this document extends to all atomic
power plants with all types of reactors intended for commercial use in the nuclear power in-
dustry of the USSR in the nearest future (VVER, RBMK, BN, AST). This approach in great mea-
sure determined the character of the exposition of the requirement in general form without
sufficient specificity. In most cases the "General Regulations" only pose the problems which
must be solved in order to ensure safety (which must be done), while not dictating definite
solutions (as should be done). Other standardizing documents (regulations, standards, pro-
cedures) develop the General Regulations and make them more specific in a certain direction,
providing a basis for the work of designers and the pertinent supervisory organs.
One of the principal documents in the domain of Gosgortekhnadzor concerning technical
safety is the document "Regulations on the Construction and Operating Safety of the Equipment
of Atomic Power Plants and Experimental and Research Nuclear Reactors and Facilities" [4].
These regulations extend to reactors, steam generators, vessels, the housings of pumps, and
fitting and tubing operating under pressure in the primary and secondary loops of atomic power
plants with water-moderated-water-cooled and uranium-graphite reactors. The document presents
the main requirements on the design of the housings, tubing, and welded joints, it discusses
the requirements on the materials used to fabricate, assemble, and repair the equipment and
tubing, and it indicates the characteristics of the mechanical properties that must be deter-
mined upon introducing new materials into production. Moreover, these regulations establish
the requirements on the fabrication and assembly of the equipment and tubing, the methods and
the scope of the methods used to check welded joints, the characteristics of the fittings,
monitoring and measuring instruments, and safety devices, and determine the order of recording,
the technical inspection and operation of the equipment, as well as the authorization of per-
sonnel to operate the equipment. These regulations constitute one of the main documents on
which Gosgortekhnadzor bases its operations.
The document that forms the basis of the work of Gosatomnadzor is the document "Nuclear
Safety Regulations for Atomic Power Plants, PBYa-04-74" [5], which was introduced in 1975
and which regulates APP safety problems associated with preventing loss of control and moni-
toring of the chain fission reaction in the reactor core and eliminating the possibility of
a critical mass being formed during recharging, transportation, storage of fuel bundles, and
assembly and repair work. It contains the main engineering and organizational requirements
for ensuring nuclear safety during the design, construction, and operation of atomic power
plants, as well as the requirements for the training and qualification of personnel engaged
in the operation of a reactor installation. These regulations establish the main technical
requirements as to the design of the reactor plant and systems that ensure nuclear safety,
and in doing so they specify the minimum composition and number of channels for monitoring
the capacity of the emergency protection system of the reactor, and the minimum list of sig-
nals for tripping the emergency protection system. This document describes (from the point
of view of nuclear safety) the sequence in which the APP is to be put into operation and also
gives a list of the documentation necessary for starting up and operating the plant.
In the realm of radiation safety, the main document, on which the organs of sanitary
supervision are based, is the document "Radiation Safety Standards NRB-76" [6]. These stan-
dards have been developed on the basis of the recommendations of the International Commission
on Radiological Protection and set up a system of dose limits and the principles of their
application. This is the main document regulating the level of exposure to ionizing radiation.
On the basis of the possible effects of ionizing radiation on the organism, these standards
establish the following categories of irradiated persons: personnel, individuals from the
general population, and the population as a whole, in assessing the genetically significant
dose of irradiation. The standards define the maximum allowable dose of irradiation of per-
sonnel under normal and emergency conditions and also set the maximum irradiation doses for
individuals from the public and for the population as a whole. The "Sanitary Regulations
for the Design and Operation of Atomic Power Plants, SP-AES-79"' [7], reflecting the specific
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features of APP, expand and supplement the radiation safety standards. This document pre-
sents the requirements on the siting of atomic power plants, the layout and finish of the
production premises, the organization of the technological process, the biological shielding,
and the sanitary and dosimetric monitoring.
This system of standards-technical documents of APP safety functions along with the sys-
tem of documents of the State Committee of the USSR on Standards (Gosstandart SSSR), which
is responsible on the national scale for the creation, introduction, and implementation of
standards in various areas of science and technology. The system of standards (state and
industrial standards, technical specifications, sanitary measures) supplement the systems of
standards-technical documents in the matter of ensuring the safety of APP by guaranteeing
the quality of numerous elements, materials, processes, etc., tried and tested in various
branches of industry and used in the nuclear power industry. These documents play a signifi-
cant role in solving the problem of ensuring the quality of APP, as this is understood in
many other countries [8], and which is discussed in detail in [2].
SOME CHARACTERISTIC ASPECTS OF THE PRACTICE OF SETTING
STANDARDS FOR APP SAFETY
Let us dwell in some detail on approaches to the solution of individual present-day
problems of APP safety as handled in national standards documents.
Differentiation of Requirements on the Safety and Siting of APP for Different Purposes.
The safety of an APP along with the quality of its construction and technical equipment is
determined by the choice of its site, the main role in this choice being played by the
distance of the APP from densely populated areas. The siting of APP is regulated by the
"Sanitary Regulations for the Design and Operation of Atomic Power Plants, SP-AES-79" [7]
and "Requirements on the Siting of Atomic Central Heating Plants and Atomic Heat and Power
Plants with Regard to Radiation Safety" [9]. The documents give three types of atomic
plants depending on their purpose: atomic power plants (APP) for the generation of electri-
city, atomic heat and power plants (AHPP) for the production of thermal and electrical energy,
and atomic heating plants (AHP) for producing hot water for domestic purposes.
The Sanitary Regulations stipulate that an APP with a rating of 440 MW(elec.) or more should
be sited no closer than 25 km to'cities with a population of more than 300,000 and no closer
than 40 km to cities with a population of more than 1 million. Analysis of actual sites of
operating APP shows that the number of inhabitants within a 30-km radius of the plant, in-
cluding the rural population and small populated localities, usually does not exceed 109,000-
200,000.
The use of atomic plants for supplying heat requires that, in order to obtain acceptable
economic indicators, they be put closer to the consumers, i.e., that they be built at a sub-
stantially smaller distance from large population centers. The increase in the risk to the
public as a result of the atomic plants being sited closer to the cities is compensated by
the imposition of additional safety requirements, ensuring protection of the APP from a
broader class of internal damage and external actions. When these requirements are met, the
risk for inhabitants of the city from the AHP is at least no greater than for an APP further
removed from the city [2]. The document [9] allows AHP to be sited no closer than 2 km from
the prospective boundary of the residential area of the city. Further development of the
city should take place with allowance for the presence of the AHP.
When atomic heat and power plants (AHPP) are used as a source of heat they may be sited
closer than APP to larger cities without the imposition of requirements for additional pro-
tection from internal damage and external actions, but with an additional requirement, placed
on AHPP, as to the maximum irradiation of large populations under normal conditions and
during failures.
Any atomic source of heat supply (AHP and AHPP) must meet requirements designed to pre-
vent radioactive substances from reaching the consumer of the heat with coolant from the
plant.
Specific safety requirements for APP are envisaged in four areas (Fig. 2):
1) In this case, during the construction of APP, AHPP, and AHP one must be guided by
the requirements of the "General Regulations" [3] and SP-AES-79 [7];
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E
x
v i
o 2
? 20- //
r
4-. . d
10 3
S
2
Ca 0 500 1000
Population, x 103
Fig. 2. Areas of specific safety require-
ments imposed on APP. The boundary marked
by == = is arbitrary.
TABLE 1. Limits of Dose of Irradiation of a Limited Part of the Population
Group of critical organs`
Irradiation
Source of
1
2
1
2
radiation
during normal operation
during maximum design
failure
Individual
Total sourcet
0.5 rem/yr 1.5 rem/yr
- -
Irradiation
20 mrem/yr 60 mrem/yr
10 rem 30 rem
from APP
Collective
Irradiation
10? man?rem/yr -
105 man?rem $
from AHPP,
AHP
'`The first group comprises the whole body and the second group is for the
thyroid.
tNot including medical radiation and the natural radiation background.
The collective dose of irradiation of the population of a large city near
the APP.
2) in this case, AHPP and AHP can be built with the additional requirements concerning
the standardization of the collective dose of irradiation of the population;
3) construction of only AHP is allowed on condition that requirements of a constructional
character are satisfied ("General Regulations" [3], SP-AES-79, and the requirements of the
document [9]);
4) atomic plants cannot be sited here ("forbidden area").
Within the framework of the. requirements of the "General Regulations," the maximum de-
sign failure adopted for APP is an instantaneous transverse rupture of a tube of maximum
diameter, and the design should take account of the action of all natural phenomena inherent
to the given site. External effects due to human activity (explosions in nearby industrial
plants and in transport, possible airplane crash) are taken into account by an appropriate
choice of site for the plant, making it possible to eliminate the possibility of such action
on the plant. Within the framework of the additional requirements of the document [9], the
maximum design failure envisaged in the AHP design is damage to any vessel of the reactor
facility that leads to loss of hermeticity. Measures should be envisaged for preventing
melting of fuel elements in the reactor core. Allowance must also be made for such external
actions as the crash of an airplane and a shock wave in the event of explosions in the vicin-
ity of the plant, with calculated parameters of the action as stipulated in the regulations.
Limits of Irradiation Dose for the Population. The "Sanitary Regulations" [7] establish
a certain dose of irradiation of a limited part of the population because of gas-aerosol dis-
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 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6
charges during normal operation of the APP and also establish allowable doses of individual
irradiation during a maximum design failure. The "Requirements on the Siting of Atomic Cen-
tral Heating Plants" [9] place additional limitations on the collective dose of irradiation
of the surrounding population (see Table 1).
The Role of the Quantitative-Probabilistic Approach to Setting Safety Standards. The
question of what approach, deterministic or quantitative-probabilistic, is employed in the
country when setting safety standards cannot be answered unambiguously, since one and the
other are used at different stages. The extent to which one approach or the other is used
is determined by what problems, in ensuring the safety of atomic power plants, are being solved,
and when. In the light of the problem posed, it is desirable to divide the activity of en-
suring APP safety into two stages: development of approaches and formulation of the require-
ments as to APP safety, and elaboration of a design, constructing, and operating the APP.
Each stage is characterized by its own formulation of problems for a certain circle of per-
sons or organizations and these problems are solved independently, although, naturally, not
in isolation from each other, since there is a strong interrelation between these problems.
In the early period of the development of the atomic power industry, the first stage
was characterized by purely intuitive and engineering approaches. At the present time the
quantitative-probabilistic approach is increasingly becoming the basis. The studies being
developed and expanded in the country on quantitative-probabilistic analysis are directed
primarily toward these goals. The elaboration of additional safety requirements for atomic
heating plants has been based in great measure on the quantitative-probabilistic approach.
For reliable application of quantitative-probabilistic analysis of safety in the design
stage of APP it is necessary to have the pertinent statistical data. Such data can be ob-
tained in sufficient number for most natural phenomena. However, statistical data about the
reliability of specific equipment used in the atomic industry are limited at this time. This,
in the main, is responsible for the deterministic approach in the second stage. Certain ele-
ments of the quantitative-probabilistic approach, however, do exist here and they are laid
out in the standards-technical documents. We note the principal ones.
As a rule, the parameters of the natural phenomena taken into account in the design are
chosen on the basis of a quantitative-probabilistic analysis. For example, the design for
the construction of an APP makes provision for an earthquake with an average recurrence
period of up to 100 yrs, and the maximum design earthquake is assumed to have parameters
which, according to the calculations, have a probability of 10-4 yr-' [10]. The choice of
the design values for the wind, snow, and other loads when taking the meteorology into ac-
count is also based on statistical data.
There are direct indications for the use of the quantitative-probabilistic approach
during designing of equipment and systems for APP. Thus, the "General Regulations" envisage
a quantitative analysis of the reliability of the systems, which leads to a search for the
most reliable schemes, quantitative analysis of the probability of damage to the equipment,
and realization of various failure situations considered in the design stage. Special pro-
cedures have been developed for these purposes. In addition to the postulated failures, the
APP design may not take account of failures of systems (elements) whose reliability is fairly
high according to estimates.
As statistical data are accumulated and the pertinent methods are approved, the domain
of application of the quantitative-probabilistic approach in the process of APP designing
and monitoring on the part of the supervisory organs will grow.
1. V. A. Sidorenko et al., At. Energ., 43, No. 5, 360 (1977).
2. V. A. Sidorenko and 0. M. Kovalevich, At. Energ., 50, No. 2, 93 (1981).
3. General Safety Regulations for Atomic Power Plants During Design, Construction, and
Operations, Izd. Minenergo SSSR, Moscow (1973).
4. Regulations on the Construction and Operating Safety of the Equipment of Atomic Power
Plantas and Experimental and Research Nuclear Reactor and Facilities, Metallurgiya,
Moscow (1973).
5. Nuclear Safety Regulations for Atomic Power Plants, PBYa-04-74, Atomizdat, Moscow
(1976).
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 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6
6. Radiation Safety Standards, NTB-76, Atomizdat, Moscow (1978).
7. Sanitary Regulations for the Design and Operation of Atomic Power Plants, SP-AES-79
[in Russian], Energoizdat, Moscow (1981).
8. Summary of IAEA Regulations 50-S-OA "Safety of Atomic Power Plants - Ensuring Quality
in APP," Vienna (1979).
9. Requirements on the Siting of Atomic Central Heating Plants and Atomic Heat and Power
Plants With Regard to Radiation Safety, At. Energ., 49, No. 2, 150, (1980).
10. Provisional Standards for the Designing of Atomic Power Plants for Seismic Regions,
VSN-15-78, Izd. Mindnergo SSSR, Moscow (1979).
E. I. Vorob'ev, L. A. I1'in, V. D. Turovskii,
L. A. Buldakov, N. G. Gusev, 0. A. Pavlovskii,
and G. M. Parkhomenko
At the beginnin~ of 1982 the total power of atomic power plants (APP) in our country
was as 16 GW (elec.) and this is to be raised to 40 GW (elec.) by the end of 1985 [1]. The
nuclear power industry is expected to develop at an even greater rate in subsequent periods:
according to the estimates of specialists, the power of APP is to reach 90 GW (elec.) by
1990 [2] and 200 GW (elec.) by the end of the century [3].
The setting of sanitary standards for radiation factors has developed in our country
in advance of the construction of large APP. The first sanitary regulations and standards
were drawn up in 1953 during the period leading up to the start-up of the world's first
atomic power plant at Obninsk. Later, during the period when the Beloyarsk and Novovoronezh
APP were being designed, the "Sanitary Regulations for Work with Radioactive Substances and
Other Sources of Ionizing Radiation, SP-333-60" were issued. In 1965, the Ministry of Public
Health of the USSR set up the National Commission on Radiation Protection (NCRP), which was
given the task of generalizing data on the scientific substantiation of the principles of
radiation protection and developing radiation safety standards and regulations. The NCRP
prepared the "Radiation Safety Standards, NRB-69," the NRB-76 standards that are in force
at present, as well as the "Fundamental Sanitary Regulations with Radioactive Substances and
Other Sources of Ionizing Radiation, OSP-72/80" [4]. In the course of developing the main
propositions of these documents, the "Sanitary Regulations for the Design and Operation of
Atomic Power Plants, SP-AES-79" [5] were issued, containing requirements on ensuring the
radiation safety of APP personnel and the population living in the region of the APP, as
well as on protection of the environment from contamination with radioactive waste and dis-
charge of waste heat. It is necessary to point out that unlike similar international norms,
our documents are legislative in character and not just recommendations.
For the protection of the population and the environment, SD-AES-79 sets dose limits
(DL) for the dose caused by the total gas-aerosol emissions and liquid radioactive discharges
from APP, which constitute 5% of the dose limit for the limited part of the population (DLB
according to NRB-76). The setting of a 5% dose limit on radioactive wastes from an APP is
consistent primarily with the well known ALARA principle [6], which is particularly timely
under the conditions of a nuclear power industry developing in densely populated regions of
the country. Moreover, the actual dose in a locality from radioactive wastes of operating
APP [7, 8] is substantially lower than the dose limits envisaged in SP-AES-79 for a limited
part of the population (Table 1). And, finally, these limitations are in accord with the
thresholdless dose-effect concept. The limits of the individual equivalent doses given in
Table 1 are for inhabitants on the boundary of the sanitary-protection zone or outside it at
a distance at which the highest radiation dose is expected. This refers to the limits of
*The installed capacity of APP in the USSR was 18 GW (elec.) at the beginning of 1983.
Translated from Atomnaya Energiya, Vol. 54, No. 4, pp. 277-285, April, 1983.
290 0038-531X/83/5404-0290$07.50 @1983 Plenum Publishing Corporation
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 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6
TABLE 1. Limits of Equivalent Dose for the
Inhabitants (Category B) of the Region of
an APP Site, mrem/yr
Form of radioactive
wastes
Fraction of Group of critical
DL (NRB- organs
76), %
TABLE 2. Average Daily Permissible Gas-
Aerosol Emission from an APP
APP power, GW (elec.
< 6, Ci/GW
(elec.) ? day
Radioactive noble
gases
1311 (all forms)
Mixture of long-
lived nuclides
Mixture of short-
lived nuclides
1Ci=3.7.1010Bq.
0,01
0,015
0,06
0,09
the annual dose reached under the conditions of radioactive equilibrium in the environment.
For liquid discharges the values of DLB are given with allowance for different ways in which
water is used: fishery, fish breeding, irrigation, and drinking water supply.
Although all possible safety measures are taken when designing an APP, a failure is not
ruled out theoretically. The regulations SP-AES-79 provide for the creation of a technical
safety system that would protect the population in the event of a maximum design failure
(MDF). In this case the following is the expected dose received by the population from a
failure: external whole-body irradiation 10 rem and internal irradiation of the thyroid of
children and of any other organs, 30 and 10 rem, respectively.
The choice of the dose of failure irradiation was based on the assumption of: first,
an extremely small probability of a failure that would lead to such a dose; second, compar-
ability with the dose limits for individual inhabitants as set in NRB-76 (0.5 rem/yr, or 30
rem in the case of chronic lifelong irradiation), which, in the opinion of the ICRP, is com-
parable, in regard to biological effect, to 10 rem in the case of irradiation of short dura-
tion; and, third, comparability with the maximum permissible dose (MPD) for professional
workers during normal operation of the APP (5, 15, and 30 rem/yr for critical organs in
groups I, II, and III, respectively).
In addition to limitations put on the main characteristics, i.e., the dose limits, deri-
vative quantitites have also been introduced: permissible emissions (PE) and permissible
discharges (PD), as well as control emissions and discharges. Strictly speaking, the DE
should correspond to discharges which, under conditions where in radiological equilibrium is
reached in the environment, should not exceed the DL given in Table 1. In SP-AES-79, however,
the permissible emissions were calculated with allowance for an additional requirement: They
should not significantly exceed the actual emissions of operating APP, the latter usually
being lower than the calculated values of MPE (Table 2). The average monthly permissible
gas-aerosol emission of 9OSr from an APP with a power < 6 GW (elec.) is 1.5 mCi/'GW (elec.).
6, Ci/
APP ? day
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 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6
TABLE 3. Average Annual Personal Radia- TABLE 4. Radiation Dose APP Personnel
Period of Average
Standardized collec-
APP
1977
1978
1979
Novovoronezh
0,78
0,61
0,6
Kol'skaya
0,84
0,59
0,68
Armyansk
0,33
0,14
0,62
Chernobyl
-
1,2
1,0
Kursk
0,23
0,33
0,38
Reactor
observation
reactor-yrs
I
close,
rem
tive dose, man -rem/
MW (elec.)?yr
VVER-440
16
0,56
1,1
RBMK-1000
11
0, 68
1,3
BN-350
6
0,01
7.10-3 *
month, and from an APP with a power 6 GW (elec.) it is 9 mCi/APP?month, and the figures
for 137Cs, 6OCo, 34Mn, and 31Cr are 15 rCi/GW (elec.)-month and 90 mCi/APP?month, respective-
ly.
An inseparable part of the system of radiation protection of atomic power plant person-
nel and the population is that of radiation monitoring. In an APP this monitoring includes
measuring the personal doses of external radiation, the equivalent dose rate of y rays and neu-
trons, as well as the neutron and a-particle flux density, the concentration and composition
of radioactive gases and aerosols in the air in the production premises, the radioactive
contamination of working surfaces, structures, and equipment, leather covers, and working
and personal clothing of the personnel, the activity and composition of gas-aerosol emissions
and liquid discharges into the environment, and the content of radioactive substances in
various objects in the environment.
Special service teams, provided with dosimetric and spectrometric equipment, monitor
the environment for the dose rate and the annual radiation dose in a locality, and determine
the concentration of radionuclides in the atmospheric air, soil, vegetation, the water of
open bodies of water, foodstuffs, and animal feed produced locally.
The requirements of radiation protection of the personnel of an atomic power plant are
met by organization of biological shielding of the equipment, zone layout of the premises,
ventilation, organizational and technical measures to reduce the y radiation due to the equip-
ment of the reactor core during planned preventative maintenance, and constant radiation
monitoring. Operating experience shows that y rays are the main harmful factor for atomic
power plant employees. The total annual dose of y rays, however, does not exceed 5 rem/yr,
and in the overwhelming majority of cases it is below this value (Table 3).
The average annual radiation dose of the personnel of atomic power plants equipped with
reactors of different types and powers ranges from 0.14 to 1.2 rem. The data of Table 4 in-
dicate that the conditions created in the APP in our country are such that the latest recom-
mendations of the ICRP are fulfilled: The average radiation dose of the personnel is one-
tenth the value adopted for the MPD. Comparison of the irradiation received by personnel
with the materials published by the United Nations Scientific Committee on the Effects of
Atomic Radiation shows that with respect to both the personal and collective doses, the irra-
diation in domestic atomic power plants is close to those in other countries.
The irradiation is due mainly to maintenance work. In an atomic power plant with an
RBMK-1000 reactor, such work consists of inspections and maintenance of drum separators, re-
placement of ball-type flow regulators for water and shut-off valves, and maintenance of the
main circulating pumps (all told, 32% of the collective dose; in an APP with a VV R-400 reac-
tor, this work encompasses fuel recharging with the attendant maintenance work on the reac-
tor vessel, steam generators, and main circulating pumps (62% of the collective dose)).During
maintenance work, in all the APP the number of persons under radiation-hazardous conditions
is increased 2 to 2.5 times. The entry of radionuclides into the human body, which is pos-
sible during fuel reloading and maintenance and repair work, is at a much lower rate than
the permissible value, and the content of radioactive substances in the body of professional
workers (in 90% of them) is less than 0.02 of the limiting permissible content indicated in
NRB-76, and only in individual cases are 137Cs and 6OCo found to be present in quantities of
60-180 nCi.
Analysis of the data of large-scale polyclinical examinations of the personnel (periodic
medical check-ups), as well as the rate of sickness with temporary incapacitation, shows that
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* Normalized to 1 MW (thermal).
 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6
TABLE 5. Emissions Normalized per Unit
Electricity Generated
RBG, Ci W 131 I, pCi/MW
APP, reactor (elec.) - yr (elec.) . yr
type 1977 1978 11979 1977 1978 1979
VVE R
Novovoronezh
Kol'skaya
Armyansk RBMK-
1000
Leningrad
Chernobyl
Kursk BN-350
Shevchenkov?
skaya (4-yr
average)
5,6
2,8
288
5,8
2,8
4,9
191
60
35
4,3
3,0
6,0
145
82
80
8,4
36
4,6
1,4
281
1234
178
21
5,5
1.2
441
2340
180
78
* Only the aerosol phase of 1311 was measured in this
period.
the health of APP employees does not exhibit any deviations attributable to the effect of
ionizing radiation. No increase in the rate of sickness is observed in personnel who have
been working in atomic power plants for 10-15 yrs.
One of the achievements and advantages of nuclear power from the point of view of the
radiation safety of the population is that under normal operation APP do not endanger the
environment with radioactive substances. Gas-aerosol emissions from domestic APP are pre-
sented in [8]. Accordingly, we give only the data concerning the main component (Table 5).
From a comparison of the actual emissions with the permissible levels established in
1979 .(up until 1979 the permissible emission had been 1.3.10 Ci/yr in the case of radio-
biological gases (RBG) and 36 Ci/yr in the case of 1311) it is seen that the actual emissions
are considerably below the limits set. This is also true of other radionuclides. Upon com-
paring the normalized emissions of domestic and foreign APP, we can notice that, first, the
emissions of RBG and 1311 from VVER and PWR reactors as a rule are lower than those from
RBMK and BWR and, second, the emissions from domestic reactors are comparable to and in some
cases lower than the average from foreign reactors [7].
In our country one of the main principles of ensuring the radiation safety of the water
environment in the region of an APP is that of a closed cycle of water utilization in the
technological loops which may contain radionuclides. Experience gained from the operation of
APP, especially those with the first, prototype units, shows, however, that during mainten-
ance and repair work, as well as during maladjustments in the reactor operating regime,
certain quantities of surplus (unbalanced) water may be produced and discharged into cooling
ponds after appropriate radiation monitoring [9]. Trap (deactivation) and special washing
water, subjected to special purification, is the source of this water. As shown by experi-
ence with domestic APP, in the first 2 yrs of operation of a plant, during the period of ad-
justment of the technological systems, including the apparatus for special purification of
salt-containing water, the volume of unbalanced water is a maximum (20,000-70,000 m3/yr per
unit). At the present time a closed cycle of water utilization is ensured over a prolonged
period (ti 2 yrs) only at the Leningrad APP, but other APP are close to switching to such an
organization. of operation.
Normal operation of power plants ensures a low concentration of radionuclides in the
unbalanced water. Constant monitoring of this water for its content of radionuclides that
are most important from the health physics point of view and that make the greatest contribu-
tion to the total activity of the discharged water makes it possible to estimate the gross
activity of the nuclides carried into the cooling pond by the water. As shown by the data
of Table 6, the main contribution to the activity of the unbalanced water comes from tritium
(tens.of Ci per year for plants with VVER and RBMK reactors). The activity of corrosion ele-
ments, mainly 60Co, is 10-6 to 10-3 Ci/yr for one unit. The content of fission fragments
is also low and is determined mainly by 137Cs and 1311. It should be pointed out that these
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 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6
TABLE 6. Gross and Normalized Activity of Liquid Discharges from APP into the Water
Environment
Nuclide
Kurskaya APP
Clernobyl APP
Armyansk APP
3H
50-60 *
(2,5=3)-10-2
20-60 *
(1,0-3).10-2 fi
40-60 *
0,1-0,15 t
54Mn
-
-
(1-6)?10-7
(0,5-3).10-10
-
-
58Co
-
-
(1-6).10-7
(0,5-3).10-10
-
-
80Co
(0,3-5).10_4
(0,1-2,5)?10-7
(1-3).10-6
(0,5-1,5).10-9
-
-
88Sr
(1,0-4).10-5
(0,5-2).10-8
-
-
-
-
'?Sr
(1,0-4).10-6
(0,5-2)?10-9
(1,2-3)?10-7
(0,6-1,5)?10-10
5.10-5
1,1.10-7
1311
(0,1-1).10-3
(0,5-5)?10-7
(0,8-6).10'
(0,4-2,8)?10-7
-
134Cs
(0,3-5).10-4
(0,1-2,5).10-7
(1,0-6).10-6
(0,5-3).10-9
4.10-6
9.10-9
' 7Cs
(2,0-4).10-3
(1,0-2)?10-6
(0,5-1)?10-4
(2,5-5).10-8
(0,5-1,5).10-3
(1,1-3,3).10-6
Discharge rate, Ci/yr.
tNormalized discharge rate, Ci/MW (elec.)-yr.
TABLE 7. Personal y-Ray Dose, rem/yr
Distance from APP, km
Nuclide
1
2 I
5 I
10
20
50
100
4'A r
6,1.10-7
3,6.10-7
1
2.10-7
4
3.10-8
1
5.10-8
2
6.10-9
2,9.10-10
85mK r
4,2.10-7
2,6.10-7
,
9,0.10-8
,
3,4.10-8
,
14.10-8
,
3,7.10-0
8,5.10-10
85K r
6,5.10-10
4,0.10-10
1,3.10-10
5,8.10-11
2,5.10-11
9,6.10-12
3,8.10-12
87Kr
6,5.10-7
3,4.10-7
1,0.10-7
3,6.10-8
91.10-8
1,4.10-9
8,4.10-11
88Kr
2,1.10-5
1,2.10-5
4,3.10-6
1,6.10-6
5,9.10-7
1,3.10-7
2,3.10-8
133Xe
135Xe
7,5.10-6
1,2.10-5
4,5.10-6
6,8.10-6
1,8.10-6
2,5.10-6
7,5.10-7
1,0.10-6
3,3.10-7
4,2.10-7
1,2.10-7
1,4.10-7
4,8.10-8
4,2.10-8
I
Total 1
4,2.10-5
2,4.10-5
8,8.10-0
3,5.10-0
9,4.10-6
4,0.10-7
9,1.10-7
data pertain to the steady-state operating conditions of the APP. In the first 1 or 2 yrs
after start-up the discharges of fission and induced radionuclides are roughly an order of
magnitude larger. For APP with an RBMK reactor, some users of the cooling water, e.g., equip-
ment in the machine shop, etc., can also be sources of radionuclides discharged into the
cooling ponds. The activity of water emerging from the cooling system is, however, difficult
to measure directly. By indirect estimates the discharges of radionuclides with the cooling
water can be compared with the discharge of radionuclides contained in the unbalanced water.
Comparison of the data shows that the activity of liquid discharges of domestic APP is
almost 1-2 orders of magnitude lower than that of similar foreign installations. With respect
to tritium this difference is smaller, but is still a factor of 2-10.
For all APP in the country, the radiation components are monitored by dosimetric, radio-
metric, and spectrometric methods within a 30-50-km radius of the plant. The results of
measurements of the radionuclide concentration in the atmosphere, water, soil, vegetation,
and food products (milk, butter), as well as the annual dose of external y radiation, are
given in detail in [8]. All of these characteristics are fully determined by the natural
background and the total radioactive fallout from nuclear explosions carried out in the 1950s
and 1960s. A slight radioactive contamination of objects in the environment with fission and
induced radionuclides formed during the operation of an APP is observed only on the industrial
site and in individual cases in adjoining territory within the limits of the sanitary-protec-
tion zone. Thus, the inhabitants of the region in which the APP is sited are not exposed to
additional radiation. This radiation cannot be measured by present-day means and can only
be estimated by calculation.
As an example, we consider the radiation conditions in the region of the Novovoronezh
APP, which has been operating for a long time in a steady-state regime and which is located
in the densely populated European part of our country with intensive agriculture. The dose
loads on the population were calculated from the formulas given in [10, 11]. In our calcula-
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 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6
10 6
10-'
0 10
C
R y
N q
9 O
a 9011 o m 90
Distance rfrrlom APP, Distance from APP,
km
Fig. 1 Fig. 2
Fig. 1. The y-ray dose at a locality as the result of
a cloud of emissions and radioactive fallout: 1) aero-
sol; 2) fallout; 3) RBG; 4) total (1 sievert = 100 rem).
Fig. 2. Dose of irradiation of the human body from
gas-aerosol emissions: 1) internal (inhalation); 2)
external irradiation; 3) internal (oral) ingestion;
4) total.
tions we took account of the actual recurrence of categories of weather (in the Pasquille
classification), the elongation of the wind rose for various sectors of the site, as well as
the data about the distribution of-the rural and urban population within a 100-km radius
around the power plant. The calculations were based on gas-aerosol emission in 1979, Ci/yr:
RBG
89Sr
0,0008
41Ar
11
90Sr
0,0039
85mKr
62
95Zr
0,0046
85Kr
fi
95Nb
0,0079
N7K r
17
103Ru
0,0033
88K r
242
106RI1
0,0)6
113X0
4230
110m
0,018
135X0
1072
1311
erosol)
0,0073
Total
5640
134Cs
0,025
Long-lived aerosols
137Cs
0,047
51Cr
0,013
141Ce
0,0005
54Mn
0,040
144Ce
0,0015
58Co
0,016
Total
0,33
60Cu
0,087
The calculated values of the y-ray dose in an open locality from the annual emissions
of both each radionuclide separately and a mixture of RBG in 1979 are presented in Table 7.
As follows from Table 7 the main contribution to the dose is made by e8Kr, 193Xe, and
135Xe, but even if we take account of the elongation of the wind rose for the site of the
Novovoronezh APP (1:15), the maximum value of the y-ray dose from a cloud of RBG in an open
locality at a distance 1 km from the APP does not exceed 63 urem/yr, i.e., is lower than the
permissible limit (20 mrem/yr) set in SP-AES-79. Besides RBG, a certain contribution to the
1-ray dose in the open locality can be made by the aerosol component of the emissions into
the air. As follows from Fig. 1, at a short distance from the APP the largest contribution
comes from the RBG cloud (up to 90% of the total dose), but with distance from the APP the
role of the y radiation from fallout radionuclides grows and at a distance of 100 km their
contribution to the total dose is 60%. In our calculations we took account of the fact that
for the Novovoronezh APP the fraction of 1311 in aerosol form comes to only about 2% of the
total activity, while the remaining 98% is accounted for by its organic compounds.
The entry of radioactive products into the human body with air breathed in or with food
products eaten causes internal irradiation of different organs and tissue in the body. Com-
parison of the dose of the external irradiation and the effective equivalent dose of internal
irradiation of the human body, calculated in accordance with the ICRP recommendations, shows
that (Fig. 2) even at a distance of 2-5 km from the Novovoronezh APP, the total annual dose
of external and internal irradiation of the rural population is only 14-40 urem and is due
mainly to the external irradiation and oral ingestion of radionuclides into the body. The
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 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196ROO0300020004-6
of the Population in 1979
Normalized
value
Absolute Normalized value
value, man
man? rem/Ci man-rem/MW
rem/yr (elec.) ? yr
External irra
0,gi
1,9 .10-1
4,(i?90-1
diation
Internal irra-
diation:
inhalation
2,4.10-2
3,7.10-6
1,6.10-6
oral in-
1,2
2,1.10-4
9,1.10-4
gestion
Total
1,8
3,2.10-1
1,4.10-3
TABLE 9. Normalized Values and Composition of Gas-Aerosol Emissions [12, 13]
Nuclide
VVER
RBMK
Nuclide
VVER
RBMK
Nuclide
VVER
RBMK
'1:1r
2,0.10-:1 *
2,9.10
""Co
-
0.10-8
4
131Cs
-
5,6.10-10
6,0.10-_
7.10-3
4
loco
5.10-6
1
,
4
I.10-9
137 CS
5,2.10-0
1,3?'10-0
B:,mKr
5,4.1(-_
,
4,(i?10--
19Sr
.
-
,
'I,(;- 0-0
1ao13a-f_1aoLa
1,7.10-0
I` Kr
88
1,0.10-2
-2
9,6.10-2
1
911s],
?
"
10.10-'
3,9.10-
li;C1,
0.10-6
-
KI.
la:l .
2,2.10
-1
I.3.10-
Zr
1,3.:10-0
-
Activity of emission,
u 5
70
,
\~
1:1.5X
7,2.10
-1
2.5-10-1
-1
1loni
A g
1
,
4,6.10
Ci/ MW (elec.)?yr
1,3,10
1,8.10
311
9,0?'10-6
5,0 10 5
51Cr
4,0.10-0
5,4.10-8
1:1:11
1.3.10-0
9,1.'10-5
11MIl
5,0.'10-6
-
1:151
4.2.10-6
6,9.10-5
*For all radionuclides we give the relative contribution in fractions of 1.
quantity of radionuclides inhaled is smaller by a factor of roughly 40, which means that this
factor of the radiation effect of air emissions of an APP need not be taken into account when
making estimates.
The total effective equivalent dose of irradiation of the population from liquid dis-
charges from an APP with VVER-440 and RBMK-1000 reactors, with the,most conservative estimates,
does not exceed 0.05 prem/MW (elec.)?yr, more than 99% of this being due to tritium. Such
irradiation can occur only if the water from the APP cooling pond is used for drinking-water
supplies. If radionuclides are not ingested with drinking water and there is only the "fish
chain," this dose is almost two orders of magnitude lower. Estimates of the dose for oper-
ating APP in the country show that the contribution from liquid discharges to the irradiation
dose of the population does not exceed 10-20%.
On the basis of data, similar to those given in Fig. 2, for various sectors of the site
of the Novovoronezh APP with allowance for the actual wind rose, as well as data about the
distribution of the rural and urban population in a 100-km zone around the plant, we calcu-
lated the collective irradiation dose of the population (Table 8). The first normalized
value is the annual collective dose, normalized to a unit of activity of the emission. It
mainly characterizes the site of the APP, since it depends primarily on the actual meteoro-
logical conditions and the distribution of the population around the APP. The second nor-
malized value is associated with the technical indicators of the APP itself, i.e., activity
of the gas-aerosol emissions per unit of electricity generated. Comparing these indicators
with the data given in [7], we can see that with respect to both the first and second indi-
cators the.Novovoronezh APP is among the best in the world. Thus, the values presented above
for the personal and collective doses of irradiation of the population show the absolute safe-
ty of the APP operation as far as public health is concerned.
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 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6
y 10-1
Q~~10?I o
1 10 100 10?0 Q
Distance from APP,
km
8 10"
a1.D
10 1
ccc 10-70
ao 10n
1012
6) 0 >.
0NH1111J
Q N 10 y
1 10 100 1000
Distance from APP, km
Fig. 3 Fig. 4
Fig. 3. Dose of y radiation of the human body from
a cloud of emissions and radioactive fallout for APP
with VVER (1) and RBMK (2) reactors.
Fig. 4. Dose of irradiation due to y rays from an
emission cloud, radioactive fallout, and inhalation
of radionuclides into the human body: 1) external
irradiation; 2) lungs; 3) effective dose; 4) thyroid;
5) red bone marrow; 6) gastrointestinal tract; 7)
gonads.
The dose loads on the population of our country have been estimated on the basis of the
same methodological principles as employed in calculating the dose around the Novovoronezh
APP. It was further assumed that the average wind speed in the region of the APP site is
2.8 m/sec and the recurrence period of weather categories after Pasquille in fractions of
a year is: 0.026 for category A, 0.114 for B, 0.230 for C, 0.322 for. D, 0.178 for E, and
0.130 for F. The effective height of the emission from the VVER reactor was 140 m (Htr =
100 m), while that from the RBMK reactor was 190 m (Htr = 150 m). The other assumptions are
in accord with the data of [11].
From the averaged characteristics of the gas-aerosol emissions of the APP, which are
given in Table 9, it follows that the main contribution to the activity of the VVER emissions
comes from 133Xe, while for the RBMK reactor the role of this radionuclide in the formation
of the activity of the emissions is comparable to that of 41Ar that is formed in the gas
loop of the reactor. Moreover, the emission of radioactive aerosol products is 4-6 orders
of magnitude smaller than that of RBG and the normalized emission from channel-type reactors
is an order of magnitude higher than the VVER emission.
The results of the calculation of the dose of external y radiation of the human body
from a cloud of emissions and radioactive fallout with allowance for the protection offered
by buildings, equipment, and the life style of the population are given in Fig. 3. The
values given in Fig. 3 for the dose correspond to the annual emissions of a typical 1000-MW
(elec.) unit with a 75% power utilization factor (PUF). The gas-aerosol emissions of the
RBMK have a substantially higher radiation significance than that of the VVER. This is at-
tributed not only to the differences in the normalized emissions, but also to their isotopic
composition, since the largest contribution to the activity of air emissions of the RBMK is
made by the "hard" y ray emitters 41Ar, 87Kr, and 88Kr. It should also be pointed out that
the main role in the formation of the dose of external radiation in the locality is played
by RBG (the contribution of radionuclides that fall on the locality does not exceed 0.2%,
while the aerosols contained in the ground layer of. air contribute ti 10-3 %). On the basis
of this, however, we cannot assert that with respect to radiation effect on the environment
and man the RBMK is worse than the VVER. The use of boron to control the VVER leads to the
formation of a certain quantity of tritium which enters the environment with gas-aerosol
emissions and liquid discharges from the APP. Estimates show that according to normalized
discharges and emissions into the environment, the RBMK is almost an order of magnitude
better than the VVER.
In Figs. 4 and 5 the values of the dose of external irradiation of the human body are com-
pared with the dose of internal irradiation of the organs and tissue because of inhalation
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 Declassified and Approved For Release 2013/02/06
9 10 100 9000
Distance from APP, km
Fig. 5
9a-si
1 90 100 9000
Distancekfmom APP,
Fig. 6
Fig. 5. Dose of irradiation due to y rays from
an emission cloud, radioactive fallout, and oral
ingestion of radionuclides into the human body:
1) thyroid, 2) external irradiation, 3) gastro-
intestinal tract, 4) red bone marrow, 5) effec-
tive dose, 6) lungs, 7) gonads.
Fig. 6. Comparative estimate of the dose loads
on the population because of gas-aerosol emissions
from different reactors: 1, 2, 5) external and
internal oral and inhalation irradiation, respec-
tively, for the RBMK; 3, 4, 6) external and in-
ternal oral and inhalation irradiation, respec-
tively, for the VVER.
-S
Fig. 7. Isopleths of collective dose (man-
sievert/yr) of irradiation within the limits of
each of 16 points of the compass for the spread-
ing of gas-aerosol emissions from the Beloyarsk
APP: 0, ?)towns with populations of 30,000-
100,000 and 10,000-30,000, respectively.
and oral ingestion of radionuclides. The dose loads on the lungs, gonads, and other organs
and tissues of the human body as the result of inhalation of radionuclides are shown as a
function of the distance from an APP with a VVER reactor. The.dose of internal irradiation
of the organs indicated and the effective dose HE of internal irradiation of the body are
two or more orders of magnitude lower than the y ray dose of the emission cloud. Thus, the
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 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6
TABLE 10. Collective Dose of Irradiation
of Population from Gas-Aerosol Emissions,
man-rem/MW (elec.)
Type
Circular
External
irradia-
Effective dose of
internal irradiatio
Total
of rear
zone, km
tion
inhala- j
oral in-
dose
tor
tion
gestion
VV R
1-10
7,2.10-5
8,9.10-7
3,7.10-5
1,1.10-4
10-50
1,6.10-4
3,0.10-6
1,4.10-1
3,0.10-4
50-100
1,1.10-4
2,9.10-6
1,6.10-4
2,8.10-4
100-1000
5,4.10-1
1,4.10-5
1,3.1()-3
1,9.10-3
1-1000
8,9.10-6
2,0.1.0-5
1,7.10-3
2,6.10-3
RBMK
1-10
3,3.10-3
1,2.10-6
4,2-10-1
3,8.10-3
10-50
5,8.10-3
4,1.10-6
1,5.10-3
7,3.10-3
50-100
1,5.10-3
3,4.10-6
1,5.10-3
3,0.10-3
100-1000
2,9.10-3
1,0.10-5
6,2.10-3
9,2.10-3
1-1000
1,4.10-2
1,9.10-5
9,7.1.0-3
2,3.10-2
TABLE 11. Collective Dose of Irradiation
of Population in 1980-2000' and the Col-
lective Risk from Irradation
Parameter
1080
1985
2000
Collective dose, man rem/yr
Gas-aerosol emission
external irradiation
96
250
1300
internal irradiation
lungs
7
18
91
gastrointestinal tract
14
36
180
skeleton
17
44
210
red bone marrow
17
44
210
thyroid
2300
6000
29000
liver
5,5
14
72
gonads
5,8
15
76
effective dose
75
200
970
total dose
170
450
2200
Liquid discharges (total dose)t
11
25
56
Total dose
180
470
2300
Collective risk, man/yr
0,026*
0,070
0,33
* It is assumed that in the period under
consideration 40% of the power will come
from VVER and 60% from RBMK.
'Calculated from the data of [9] with the
assumption that 50% of the APP will have
dry or wet cooling towers and 50% will
discharge heat and a small quantity of
radionuclides into cooling ponds.
The estimate was made on the basis of a
risk of 1.5.10-4 rem 1 [6].
role of inhalation of radioactive products of the gas-aerosol discharges from an APP during
normal operation of the reactor is altogether negligible.
A slightly greater role is played by oral ingestion of radionuclides into the body. The
dose of internal irradiation of the thyroid, especially in the case of children up to 1 yr
of age, may exceed the dose of external irradiation because of an RBG cloud, but in this
case as well HE is substantially lower than the dose of external irradiation of the body.
Similar results have also been obtained for RBMK emissions, as confirmed by Fig. 6. As in
the construction of Fig. 3, the values of the doses here correspond to emissions from reac-
tors with a power of 1000 MW (elec.) with a PUF of 75%. It must be pointed out that the dose
values given in Fig. 6 are two or more orders of magnitude lower than the permissible levels
set in SP-APS-79 and are equal to 20 mrem/yr for critical organs in group I.
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 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6
Naturally, the actual distribution of the population around the APP with allowance for
the meteorological characteristics of the locality results in an inhomogeneous collective
dose of irradiation of persons in each sector to which the emissions spread. An example of
such estimates is Fig. 7, which shows the position of the isopleths of the collective dose
of external y ray irradiation for a coordinate grid with 16 points of the compass relative
to the Beloyarsk APP. In the calculations we assumed that the isotopic composition of the
RBMK emissions corresponds to the data of Table 9, that the height of the reactor stack is
100 m, and that the emission rate is 1 Ci/yr. As is seen from Fig. 7, populated points are
peculiar centers that "attract" collective-dose isopleths to themselves. In this respect,
both small populated points which are close to the APP (the settlement Zarechnyi) and large
towns at a considerable distance from the APP (Sverdlovsk, Pervoural'sk) are of major impor-
tance. It should also be pointed out that the collective dose of irradiation of the popula-
tion for the entire 100-km zone around the Beloyarsk APP, calculated with allowance for the
actual data on the meteorology and the population distribution, is 2.7.10-4 man?rem/Ci, and
when we use the average values of the meteorological data and the average population density
over the provinces of Sverdlovsk and Chelyabinsk (23 inhabitants/km2), the value obtained for
the dose.is substantially lower, 6.10-5/man?rem/Ci.
The results presented here once again confirm that for a correct estimate of the collec-
tive radiation dose of the population it is necessary to use actual data about the site of
the APP.
The value of the collective radiation dose of the population, normalized to 1 MW (elec.)-
yr, for four characteristic zones around APP are given in Table 10. Using the data of this
table we can calculate the collective dose of the population in our country from air emis-
sions of APP in 1980, 1985, and 2000 and we can also estimate the risk. The results of the
calculations, which are presented in Table 11, once again confirm the safety of nuclear power
as far as the public health is concerned. Even with a total nuclear power capacity of ti 200
GW (elec.), the expected number of cases of random consequences of irradiation from emissions
of radionuclides into the environment during operation of atomic power plants is only 0.33
man/yr, i.e., a level that is absolutely undetectable against the background of the natural
level of incidence of malignant tumors. One can also point out that a collective dose of
2.3- 103 man?rem/yr is equal to the dose of radiation that the population of our country re-
ceives from the natural background in just 40 min.
We have not considered the radiation consequence of accidents in atomic power plants.
A large number of special investigations have been devoted to this important subject. How-
ever, if we consider that the probability of major accidents in an atomic power plant is low
(ti l0-6 to 10-7 yr-I[14]), then despite the considerable dose loads in the region of the plant
during such an accident, the collective risk of irradiation of the population of our country
per year of reactor operation will not change and will be close to the values given in Table
10.
Thus, the experience from the operation of APP in our country, regulated by appropriate
norms and regulations, permits us to assess the radiation conditions as being good. One can
say that the working conditions, the health of the personnel, as well as the state of the
environment around actual APP are more favorable than in other branches of the power industry.
The present authors felt it was desirable to show this positive experience in the domain of
the radiation safety of nuclear power. This is all the more important in that the view is
often expressed that nuclear power is a dangerous branch of industry and a source of harmful
effects on the personnel, the population, and the environment. Such unqualified statements
cannot bring anything else but actual harm. One of the tasks facing the authors, therefore,
is to present objective information about the true state of radiation conditions of the nu-
clear power industry of our country..
1. Proceedings of XXVI Congress of the Communist Party of the Soviet Union [in Russian],
Politizdat, Moscow (1981).
2. I. S. Zheludev and L. V. Konstantinov, IAEA Bull.[in Russian], 22, No. 2, 34 (1980).
3. L. A. Il'in and 0. A. Pavlovskii, in: Abstracts, Second Conference of COMECON Member-
Countries on Radiation Safety of Atomic Power Plants [in Russian], Vilnius, (1982).
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 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196ROO0300020004-6
4. Radiation Safety Standards NRB-76 and Basic Sanitary Regulations of Work with Radio-
active Substances and Other Sources of Ionizing Radiation OSP-72/80 [in Russian], Ener-
goizdat, Moscow (1981).
5. Sanitary Regulations for the Design and Operation of Atomic Power Plants Sp-AES-79
[in Russian], Energoizdat, Moscow (1981).
6. Radiation Protection, Recommendations of the ICRP. Publication No. 26 [Russian transla-
tion], Atomizdat, Moscow (1978).
7. Sources and Effect of Ionizing Radiation. Report of the UN Scientific Committee on the
Effect of Atomic Radiation [in Russian], New York (1978).
8. L. A. Buldakov, D. I. Gusev, N. G. Gusev, et al., in: Radiation Safety in the Atomic
Power Industry [in Russian], A. I. Burnazyan (ed.)., Atomizdat, Moscow (1981).
9. 0. A. Pavlovskii et al., in: Abstracts, Second Conference of COMECON Member-Countries
on Radiation Safety of Atomic Power Plants [in Russian], Vilnius (1982).
10. D. I. Gusev and 0. A. Pavlovskii, in: Atomic Power Plants [in Russian], No. 2, Energiya,
Moscow (1979), p. 191.
11. 0. A. Pavlovskii, in: Radiation Safety and Protection [in Russian], No. 7, Energiya,
Moscow (1982).
12. A. I. Burnazyan et al., Proceedings of Int. Conf. on Nuclear Power and Its Fuel Cycle,
Salsburg, Austria, May 2-13, 1977, IAEA-CN-36/351 (1977).
13. N. S. Babaev,V. F. Demin, L. A. I1'in, et al., in: Nuclear Power, Man, and the Environ-
ment [in Russian], A. P. Aleksandrov (ed.), Energoizdat, Moscow (1981).
14. Reactor Safety Study. Report WASH-1400, NUREC-75/014, October (1975).
EXTRACTION AND PROCESSING OF URANIUM ORE IN THE USSR
B.
N.
Laskorin, V.
A.
Mamilov, Yu. A..Koreisho,
UDC 622.775.791:
D.
I.
Skorovarov,
L.
I. Vodolazov, I. P. Smirnov,
541.183.12
0.
and
L.
V.
Kedrovskii,
N. Mosinets
V.
P. Shulika, B. V. Nevskii,.
DEVELOPMENT OF URANIUM DEPOSITS
Uranium deposits in the USSR are situated in different locations, but genetically they
belong to two categories - hydrothermal and hydrogenic. Many consist of ores which, in addi-
tion to uranium, include other elements, but extracting all of these elements is too expen-
sive to make the enterprise worthwhile.
Development of uranium ores is a relatively recent branch of mining but it has resolved
effectively a number of difficult problems in the operation of mines stemming from extraction
of radioactive ores and the related peculiar features such as limited concentrations of re-
serves in the interior of the earth, complex morphology of most ore deposites being developed,
inadmissability of large exposures of ore blocks and accumulation of considerable amounts of
broken-off ore in the stoping space because of the radon release. Increasingly poorer ore
becomes involved in development, and efficiency can only be achieved with the use of modern
mining and processing techniques, high selectivity of extraction, field preparation of deposits,
and'high speed of mining, with safe and comfortable labor conditions for miners operating in
high temperature stopes at deep levels (1200-1700 m) in rock beds with a high burst hazard.
Open Mining. In developing hydrogenic deposits with the open method, an effective flow
process of stripping operation is practiced, using rotor equipment featuring 1000-5000 m3/h
when operating for internal and external dumps; this allows efficient uranium ore mining with
stripping ratios of 50-60 m3/ton and more. When developing strong enclosing rocks, prelimina-
ry explosive rock loosening is practiced. This increases the productivity, compared with
excavation of unloosened rocks, by 1.4-1.9 times and reduces costs by 20-60%. An effective
way of cutting transportation costs in driving openings and developing deep horizons in such
mines was the method of transportation-dump dams built perpendicular to the stope advance
front and connecting the stripping and mining benches with the inner clump stories. This re-
duced transportation costs by 25-40%.
Translated from Atomnaya Energiya, Vol. 54, No. 4, pp. 286-292, April, 1983.
0038-531X/83/5404-0301$07.50
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 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6
With opencast mining of hydrothermal deposits, high efficiency was achieved by using
combined car-conveyor or car-track transport, where the transportation cost, compared to the
cost with cars alone, was reduced 20-25%. Raising the height of drilled-off benches m and
sorting the ore in excavations after blasting on subbenches and installing equipment on the
blasted rock mass proved an efficient solution.
Subdivision into subbenches is important for stoping in deep mines, where uranium deposits
of a complex structure are worked. Developing the blasting technologies that retain the geo-
logic structure of ore bodies (placers) with the loosening coefficients of 1.05-1.12 has re-
duced the ore loss 1.5-2.0 times and curtailed ore dilution by 40-50%. This reduced the
stripping coefficient, transportation amount, benefication tails yield, and largely expanded
the scope of opencast mining. A transitional area in the mining of uranium deposits is com-
bined opencast and underground workings conducted in a number of mines with an extremely high
operation density.
Underground Mining. In underground mining of hydrogenic deposits, the common practice
is the shield driving of the shafts of capital and development openings during preliminary
mine drainage from special drainage horizons. In stoping, powered and water supported instal-
lations have been introduced, including those with active maintenance of a "false" roof auger
drilling and plow rigs with radiometric devices leading the measurement and control instru-
ments along the bed boundary. The working of hard inclusions - carbonate and sand rocks - is
done by powered roll cutters.
The high intensity of mining in the opening phase has been attained by simultaneous mul-
tistage opening of deposits and assignment of separate concentration horizons as justified
economically, with the operation conducted independently in each stage; enlarging the mine
fields, group opening and preparation of the stories, increasing the story height from 45-60
m to 90-120 m, and other technological solutions. The working of each stage with burial, in
the worked-out space of the lower horizons of the gangue, rejected ore, and benefication
tails produced when developing the upper horizons created conditions for waste-free opera-
tion. Intensification of mining at the stage of stoping excavation was attained by enlarging
the parameters of the breakage blocks, and the simultaneous panel preparation of blocks and
the use of a continuous excavation procedure, multiface operation of miners' teams and the
use of modern self-propelled drilling and loader-transportation units, vibration containers,
breakage, and advance machines.
Because of the complex structure of hydrothermal uranium deposits, they are mostly
worked with the use of a hardening filling of substory headings or crosscuts, substory ware-
housing, reducing the losses and ore dilution, containing the strain of enclosing rocks and
ore masses, and thus reducing radon release. These systems, while ensuring a complete extrac-
tion of uranium, create conditions for extensive mechanization of the basic and auxiliary pro-
duction processes, and warrant safety for the miners and high speeds of excavation advance.
Intense mechanization is conducted in workings on the surface of the deposits where automated
and mechanized sets of rail-car exchange are used and layouts of the mines have been improved.
Radiometric sensors on hoppers of self-propelled loading and delivery machines used for
sorting the flows of pay ores, substandard ore, and dead rock even in the stope at an early
stage of mixing, reduce the ore loss and dilution.
Underground and Concentrated Leaching of Uranium. Underground leaching of uranium in
hydrogenic deposits is the most effective method for mining of poor uranium ore found in dif-
ficult mining and geologic conditions where neither underground nor open cast methods would
be efficient. Developing hydrogenic deposits with easily lixiviated poor uranium ore has
changed the general estimate of uranium reserves and the very notion of raw material supplies
of nuclear power engineering in the long range. With underground leaching, no large infra-
structure, transport facilities, processing equipment and tails storage are required; condi-
tions for highly efficient development of poor but large deposits are thus provided, reducing
investment 2-3 times and tripling the labor productivity as assessed in terms of the end-pro-
duct.
Development of the underground leaching of hydrogenic deposits in the past few years
characteristically involved deep, relatively narrow and long placers concentrated in low per-
meability rocks. The use of powerful immersion electric pumps to extract solutions, im-
proved hydrodynamic regimes of pumping the solution in and out, optimized network of recon-
naissance and development holes, fluctuating operation schedules, creation of vertical and
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 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6
horizontal antipercolation shields, development of artificial cracking by using various tech-
nical facilities in low permeability rocks, increased diameters of near-filter zones, crea-
tion of new technical facilities (polyethylene and glass-reinforced plastic pipes, disk fil-
ters, and processes of pipe-free solution rise) - all have further enhanced the efficiency
of underground uranium ore leaching in the USSR.
Underground leaching in hydrothermal deposits is normally practiced in combination with
standard mining operations during the stage of deposit development, working of flanks and
substandard ore areas simultaneous to the principal operation as well as in post-operation
working of deposits. Independent underground leaching as the main method of deposit develop-
ment is rare. With the standard system of opening and preparation of ore deposits, efficien-
cy can be ensured only at the stage of cutting and preparation of blocks, when most ore is
left in place instead of being raised to the surface. With improved processes of preparation
of ore bodies with block-free excavation, optimization of the parameters and technology of
drilling and blasting, and high degrees of ore fracturing and optimum filtrational uniformity
of broken ore mass it was possible to ensure 70% uranium extraction; secondary blasting en-
hanced extraction to 90%, which, in current conditions, places underground leaching on a par
with standard mining operation and sometimes makes it even superior.
Uranium Leaching. For silicate and alumosilicate ores, leaching with a solution of sul-
furic acid with oxidizers (pyrolusite, sodium or potassium chlorates, and ferric ion) is the
principal method. Ferric iron ions are important for oxidizing the minerals of tetravalent
uranium in sulfuric acid leaching. A systematic study of the behavior of uranium minerals and
accessory minerals of the gangue in sulfuric acid leaching with the use of different oxidizers
has made it possible to determine the action of oxidizers, the ways of their efficient appli-
cation, and potential reduction of agent consumption.
Progress in the acid leaching under atmospheric pressure followed the trend of enlarging
the planned productivity to reduce the number of processing chains, spending less electricity
on pulp mixing, and less steam on maintaining a rated temperature. Equipment for leaching that
uses minimum amounts of liquid for pulp mixing has been developed.
Leaching of ores with a high content of acid-intensive components (carbonates) is done
with carbonate and sodium bicarbonate solutions. Minerals of tetravalent uranium are oxidized
in carbonate solution with the cheapest known oxidizer - the atmospheric oxygen. This process
was improved by means of better air spraying, application of industrial oxygen, selection of
oxidation catalysts, and efficient heat utilization and recuperation.
A promising method for extracting uranium from poorly resistant and complex ores is auto-
clave leaching at 100?C and higher temperatures, which raises the rate of useful reactions
that evolve too slowly at lower temperatures. The use of industrial oxygen or air as the
oxidizer in the autoclave made it possible to combine uranium leaching with sulfuric acid
production from the pyrite present in the ore being oxidized, as well as to oxidize sulfides
of other valuable metals brought into the solution. Positive features of industrial oxygen
and air used as oxidizers is their availability, and absence of noxious, gaseous, or dis-
solved environmental pollutants in reaction products. Autoclave leaching saves chemicals
and energy.
Molybdenum is a valuable component of uranium ores. Bringing molybdenum from molybdenite
into sulfuric acid solution involves certain difficulties, as oxidizers commonly used for
uranium ore leaching are inefficient with molybdenite. Leaching in autoclave under pressure
with industrial oxygen or air used with a small admixture of nitric acid or nitrates effi-
ciently dissolves molybdenite.
With acid autoclave leaching it is easier to create a closed process system for ore
treatment, preventing formation of liquid and gaseous waste which would pollute the environ-
ment. An advantage of autoclave carbonate leaching with air is the improved extraction of
valuable components and the reduced energy costs (steam and compressed air) due to process
intensification and heat recuperation.
Figure 1 shows energy costs as a function of leaching duration in a standard container
and an autoclave with mechanical mixing of the pulp. The cost of steam in autoclave leaching
is estimated taking into account heat recuperation. If the length of leaching in a regular
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 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6
0
9,5
A
0,5
0 17 29 J6 48 Z
Fig.
1.
Energy cost (C, ruble/
ton)
as
a function of leaching
time
(r,
h): 1) steam and air
in the container; 2) steam,
air, and electricity in the
autoclave; 3) electricity in
the autoclave.
container exceeds 5 h (normally 10-12 h or longer), the process is more efficient with auto-
clave leaching. Carbonate leaching in an autoclave under air pressure oxidizes completely
thionates and polythionates that can poison ionic exchange resins. The economic effect of
the use of the autoclave ensures recovering the plant investment in 1.5-2 years. Heightened'
temperatures and pressures and the use of most inexpensive and pure oxidizer - air oxygen -
allowed effective processing of various types of poorly resistant and complex uranium ores
and the extraction of valuable components while reducing the expenditure of agents and energy.
Various types of autoclaves are under serial production and in industrial use in the USSR.
Two types of autoclave are used in Soviet practice for uranium ore leaching - horizontal
four-chamber units with mechanical mixers and vertical units with pneumatic pulp agitation.
The agitators are two-stage turbine cantilevered mixers. Each agitator sucks in and disperses
some 300 m3 of gas per 1 m3 of pulp. The power of the individual electric drive is 75 kW. The
experience with the horizontal autoclave shows that the turbine aerator-agitator is an ef-
fective mixer, providing for a high oxygen utilization ratio.. The vertical autoclave with
pneumatic pulp mixing is a 100-200-m3 container rated for operation at heightened tempera-
tures up to 180?C and pressures up to 16 atm (1 atm = 101.325 kPa). The useful volume of the
autoclave is 100 m3; the diameter and the height being 3 and 18 m, respectively. An air lift
pipe for pulp circulation is located along the central axis of the plant. The air is fed
from the bottom through the distributing device (a punched plate with caps) installed beneath
the air lift tube. The outer surface is insulated to reduce heat loss.
Sorptive-Extractive Removal of Uranium. In the initial stage of the development of the
uranium industry, ore was processed by. conventional, classic techniques, using such opera-
tions as filtration, repulping, decanting, multiple precipitation, and dissolution of uranium.
For the hard-to-filtrate uranium ores of the clay type, energy- and labor-intensive opera-
tions were involved in the separation of the uranium-containing solution from the bulk of the
leached ore. Multiple filtration or countercurrent decanting resulted in substantial losses
of leached-out uranium in the moisture of filtered-off ore or condensed ore residue, with ad-
ditional expenses on decontamination of radioactive effluents.
In 1953, a nonfiltrational sorptive-extractive method for extraction and processing of
dense and viscous pulp was developed, which currently is the basic industrial process for
extraction of uranium and valuable accessories from ores and concentrates. Sorption from the
pulp allowed to combine separation of ionite from the ore mass with uranium concentration
and purification. The introduction of the sorption of uranium from pulp reduced energy costs,
enhanced productivity, and saved millions of square meters of filtering fabrics and hundreds
of tons of acids, alkali, and other chemicals, while the production capacity increased sever-
alfold. This process resulted in a review of the requirements to extracted ores and, at some
mines the reduction of the rated uranium contents resulted in an increase of the effectively,
utilized reserves.
The process of uranium sorption from pulp is more intensive than conventional filtration.,
and decanting methods, exceeding them by hundreds and sometimes thousands of times in effica-
cy. The process effectiveness increases further when sorption is combined with leaching.
If ionite is introduced at the stage of leaching, uranium extraction is enhanced, the over-
all ore processing time is reduced, and the high sorptive capacities of such ore minerals as
schist, coal, bentonite, montmorillonite, and zeolites are suppressed, so that a high con-
centration of leaching agent is maintained to keep the leached-out uranium in solution.
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TABLE 1. Efficacy of Sorptive Absorption
of Uranium from Natural Water with Salt
Content of u6 g/liter and Uranium Concen-
tration 60 mg/m3 at pH = 8.4-8.7 and Sorp-
tion and Desorption Durations of 48 and
6 h
Sorbent
Granulation,
mm
Uranium
capacity,
mg /liter
of sorbent
Alumogel
-0,10
0,21)
co
Titanogel
-0,10
0,22-0,30
?
Calciumphosphategel
-0,10
11,20
Zerolite FF
?0,20-0,40
0,21
Amberlite IRA-410
-1-0,20-0,63
0,54
Anionite AM-lOXp
-17-0,056-0,16
5,3
Z,
Anionite AM-lOXp
?0,16-0,25
5,0
w 2
Anionite AM-lOXp
?0,25-0,40
3,0-478
Anionite AMP
- 0,63-1,6
0,45-0,8
N a
Varion AP
1-0,63--1 ,6
0,5-0,7
Anionite AM1Opp
;-0,63-1 0
2,0-2,5
Anionite AM-1DXp
-i-1,0--1,6
1,5-2,0
U W
For uranium sorption, at the first stage, carboxyl cationite SG-l was commonly used,
which sorbed uranium through complex formation. Since 1948, studies were conducted in the
USSR into the specifics of uranium sorption from sulfuric acid and carbonate solutions, as
well as from nitric acid, hydrochloric acid, phosphoric acid, and fluoride media. As a re-
sult, at the second stage of development of ionic exchange technology, a wider use of anion-
ites, ampholytes, and complexites was possible. In generalizing the results, one notes the
clear connection between ionite basicity and its affinity to various uranium complexes. Nor-
mally, high-basicity, macroporous anionites, compared with gel-like anionites, have advan-
tages in uranium sorption and desorption kinetics, which determine the optimal areas of ap-
plication of the commonly known ionic exchange resins:
High-basicity anionites AN, AMP, VP-lAp - for extraction of uranium from low-salt sul-
furic acid media at pH = 1-3 and also for uranium sorption from carbonate media at pH < 10.5
(for sorption of tricarbonate complexes of uranium phosphonic and arsonic bases are promising);
low- and medium-basicity anionites AN-2F, EDE-10p, AM-3, and BP-1p, as well as nitrogen-
phosphorus containing ionites (API) of the complex type - for processing solutions with a
high sulfate content and high residual acidity (30-150 g/liter of sulfuric acid). Similar
data have been obtained for phosphoric-sulfuric acid solutions with phosphoric acid content
of 150-350 g/liter;
specific anionite AM-lOXp - for extraction of uranium from natural lake and river waters
with a content of 20-60 mg/m3 (Table 1).
In the industry, various equipment is used for sorption (at sorption and desorption
stages):
standard reactors with mechanical mixing mounted as cascades of 3-6 plants;
columns with suspended ionite layer of up to 150 m3 for processing of diluted pulp of
a density of 1.05-1.10 g/cm3;
units with pneumatic mixing of a volume from 0.3 to 500 m3 for processing of dense pulp
up to solid-to-liquid ratio of 1:1;
columns with free moving ionite layer (KDS) for washing ionite from ooze and regenera-
tion of uranium-containing ionite; and
columns of continuous sorption-desorption with pneumatic discharge (KNSPR), pressure
sorption columns (SNK), and countercurrent ionic exchange columns (PIK-1), with the uranium
desorption effectiveness ranked as follows: PIK-1, SNK> KNSPR> KDS. -
Extragents such as trialkylamines (TAA), tributylphosphate (TBP), di-2-ethylhexylphos-
phoric acid (D2EHPA) are widely used for purifying uranium from impurities in processing of
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industrial eluates and in extracting uranium from ore solutions. Synergetic mixes of acid
extragents with neutral organic compounds are efficiently used for these purposes.
The extractional method for purifying and producing uranium has been in use for 25 years.
Extractors in most enterprises are horizontal mixer-settling tank boxes with a capacity of
1000-2000 m3/day, with turbine agitators that have replaced column plants. By modernizing
the mixer-settler, the volume of extragent in the fire-hazardous diluent (kerosene) can be
reduced. A greater effect, however, can be obtained in using centrifugal extractors and
clarifiers of initial solutions; industrial development of their design will allow in the
nearest future to switch to processes with intensified phase separation.
Even more effective superextragent substances are under development, which include phos-
phinoxides, phosphoric trisamide, cyclotriphosphazotrienes, and crown ethers. Experimental
production is under way in the USSR of phosphinoxides (up to several dozens of tons annual-
ly). Satisfactory results have been obtained with selective nitrogen-phosphorus containing
extragents for uranium extraction from nitrate and nitric acid media. Very strong complex-
formants of actinoids such as phosphoric trisamide and cyclotriphosphazotriene have been
found among the substances, with effective constants of uranium extraction attaining the
values as high as 106-107. In recent years, there has been lively interest in a new type of
absorbent-impregnated sorbents (IS), often referred to as solid extragents (SE). They are
produced by injecting porous copolymers of the extragents into the granules. The resulting
absorbent can be recommended for extraction from solutions of a complex salt composition.
Extragent emulsions in mineral acid solutions, such as tributylphosphate or trialkyl-
amine in kerosene, emulsified in 3% sulfuric acid, are recommended for desorption of uranium
from high-basicity anionites. The process is conducted in mixer-settler plants with immersed
or outside chambers, with the volume ratios of anionite solution and extragent being 1:1:5.
The mixing is conducted for 1 h at each stage; ionite is then separated from the emulsion on
a sieve for 2-3 min, and a layering of the emulsion is conducted in the settling tank for
5-10 min. This completes uranium desorption from AMP anionite to residual capacity of 0.25%
of initial content in 7-8 processing stages for uranium saturation of the extragent of up to
6-7 g/liter. With extractional desorption, effective mass transfer is achieved, and the de-
sorption is completed faster than in ordinary conditions. Similarly, in four desorption
stages ferric iron can be eluted from cationite SG-l by emulsion of kerosene solution of mono-
laurylphosphoric acid in 30% sulfuric acid, and molybdenum can be desorbed from anionite
AMP by kerosene solution of the mixture of extragents D2EHPA with TBP or TAA.
With widespread introduction of nonfiltrational sorption-extraction processes, the rela-
tive cost of raw material processing is reduced continuously, and uranium extraction from
the ores is raised, while its content in raw materials is dwindling (Table 2).
The accessory valuable components are usually extracted together with uranium during
leaching and then separated during sorption, desorption, or extraction, or, less commonly,
by chemical precipitation and crystallization.
Molybdenum. Ample industrial experience has been accumulated in the USSR in combined
and separate sorption of uranium and molybdenum from acid and carbonate pulps by high-basici-
ty anionites AM, AMP, and VP-lAp. In acid solutions, molybdenum is readily hydrolized and
polymerized, so that its sorption is sensitive to structural and stoichiometric factors.
TABLE 2. Efficiency of Hydrometallurgic
Processing of Monometallic Uranium Ores
Rel. ura- I
Rel. processing
Uranium ex-
Operation
nium con
cost
10
traction, 1of
year
tent, %
,
1952-1957
figure
1952-1957
100
100
Inn
1961-1965
101)
80
106
1966-1970
90
70
109
1971-1972
75
70
111
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Thus, the sorptive capacity of highly porous anionite VP-lAp is 5.5 times that of gel anionite
AMP. From acid pulps, the maximum separation of components is attained with two-stage sorp-
tive processes where molybdenum is extracted at a relatively high acidity, followed by sorp-
tion of uranium from pulps neutralized to pH = 1.5-3.5.
Vanadium. Vanadium behavior during sorption greatly resembles that of molybdenum. How-
ever, the depressing action of sulfates is more pronounced. The oxidative potential is im-
portant in order to prevent formation of tetravalent vanadium ions existing in solutions with
emf below 500 mV that are poorly sorbable by anionites. At emf above 700 mV, vanadium is in
a heptavalent state. In real solutions, after leaching of uranium ores, vanadium concentra-
tion is 1.5-6.3 g/liter. Useful capacities in terms of vanadium for such solutions may be
as high as 420, 330, and 110 mg/g for ionites VP-lAp, VP-1p, and BPK, respectively.
Copper. In processing uranium-copper ores by acid-uranium leaching, copper passes into
solution relatively easily. It can be sorbed together with uranium or separately after re-
moval of uranium by selective ionites. The highest capacities of sorbents in terms of copper
at pH =4 are: 60-80 mg/g for ionites SG-l, AN-2F, EDE-10p, AMK, ANKB-l0, and 100-140 mg/g for
ampholytes VPK, ANKB-1, and ANKB-2. There is a class of specific sorbents which can extract
copper from more acid solutions with the capacity of 70-100 mg/g at pH = 2. These are the
ionites EDE-10p, VPK, ANKB-1, and ANKB-2.
Rhenium. For sorption of perrhenate ions, activated carbons can be used, which are
ranked by their rhenium-absorption capacity as follows: AG-N, AG-3, AG-5> KAD-iodic > AR-3,
BAU, SKT. Carbon is saturated by rhenium up to a content of 1-2%. Better results can be
achieved by using high polymer ionites at pH = 5-6, since raising acidity to 5-15 g/liter in
the area of low- equilibrium rhenium concentration lowers the ionite capacity by 8-10 times.
The efficacy of ionites in rhenium absorption decreases in the following order: VP-lp > AM-3,
AN-21> AN, amberlyte IRA-400> carbons.
Zirconium. After sulfitization of silicates for an autoclave processing of carbonate
uranium-zirconium ores, up to 3-15 g/liter zirconium passes into solution. In an aqueous
solution, zirconium and hafnium ions are liable to hydrolysis and polymerization, which com-
pounds the ionic state and diversifies the possible uses of ionic exchange materials. For
instance, for solutions with zirconium contents of 0.5 g/liter and pH = 1.8-2.0, practically
acceptable ionite saturation can be obtained, %:
Porous sulfocationite KU-23 8-11
Porous carboxycationite KM-2p 6.0
Porous anionite AM-2B, VP-lp 2.3-2.6
Porous ionite AE-14, VPK 3.3-4.0
By adding fluorides into sulfuric acid solutions, one can intensify zirconium absorption
by anionites to the following saturation, %:
AM-p, APF-2, VP-lAp, SF-5 10-12
AMP, VP-1p, VP-15p, VPK 4-8
Am-2B, BPG, KU-2, KU-23, EDE-lop 2-4
Tantalum and Niobium. In leaching uranium tantaloniobate by sulfitization, up to 1 and
10 g/liter of tantulum and niobium, respectively, in terms of their oxide, pass into solution.
They can be extracted by porous anionites AM-p and AMP-p, with total capacity of up to 50 mg/
g of sorbent. As for zirconium, tantalum and niobium absorption can be intensified three-
to fourfold by introducing fluoride anions (2-8 moles per the total of tantalum and niobium
moles). The following features of tantalum and niobium sorption are notable:
With identical matrices, the high-basicity anionites have the highest tantalum capacity
(AMP > AM > AM-2B > AM-3 > AM-4);
for identical basicity of the functional group, the highest tantalum capacity is dis-
played by anionites of the styrene divinylbenzene type (AMP > BP-lAp);
porous anionites have a higher tantalum capacity than gel anionites, while porosity does
not affect niobium sorption; and
anionite AMP-p has the maximum ratio of tantalum and niobium separation.
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 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6
TABLE 3. Efficacy of Ionite Saturation
with Gold from Cyanide Pulps after Leach-
ing Ore with Gold Content of 2 to 5 g/ton
AM-2B
AM-p
Sorbent capacity, mg/liter
of sorbent
15.3
3.1
3.1
4,3
3.5
5,5
4,7
25.6
Share of
gold in total
absorbed
components
0.55
0,17
Gold. From gold-containing uranium ores, gold is dissolved by the method of sorptive
leaching with cyanides of alkali metals before and after uranium extraction. To separate
gold from the pulp, the bifunctional anionite AM-2B is used which features high capacity and
enhanced selectivity (Table 3). Saturated anionite is regenerated by urea solutions. From
the eluates obtained after extractive purification, crude gold ingots are obtained, which
are then refined electrolytically to obtain ingots with 99.99% gold content.
Cesium. During the sulfuric acid leaching of uranium ore, up to 5 mg/liter cesium and
12 mg/liter rubidium passes into the solution from destroyed hydromica, chloride, and some
other minerals. Concomitant dissolution of cesium and rubidium is enhanced when autoclave
leaching of uranium ores is practiced. The dissolved rare alkali metals can be extracted
from the pulp with simultaneous concentration of more than 2000 times in terms of cesium by
using the mineral-organic sorbent cesiite-7. For this, countercurrent operation mode is used,
and one container with sorbent load in volume of 10% and pulp stay of 1 h is sufficient. The
capacity of saturated sorbent is 12.7-14.0 mg/g.
Thorium, Scandium, Rare-Earth Elements, and Phosphate Fertilizers. Clay uranium-phos-
phorus-rare-earth-scandium ores with considerable amounts of microscopic pyrite (melnikovite)
and uranium-phosphorus-thorium ores containing some carbonates and clay are processed in the
USSR. The ores are subjected to preliminary mechanical concentration because of a low con-
tent of such valuable components as phosphorus pentoxide (4-12%), thorium (0.03%), and rare-
earth elements (0.2%). Two-stage disintegration and multistage classification in hydrocy-
clones of the class ? 0.02 mm are practiced: The underflow of hydrocyclone is the francolitic
concentrate with a high content of all valuable components and high extraction (90%), while
the overflow is dispersed clay and microscopic melnikovite grains. A process has been devel-
oped to separate from the tails in battery hydrocyclones, pyrite concentrate suitable for
production of sulfuric acid, required for subsequent leaching of uranium-phosphorus-rare-
earth concentrate. The latter is leached by a mix of sulfuric and nitric acids, bringing
virtually all valuable components into solution and fixing almost half of the calcium as
gypsum with which the bulk of radioactive elements (mainly 226Ra and 230Th) coprecipitate.
The fertilizer obtained as a result contains up to 50% of water-soluble phosphorus and 50%
of citrate-soluble phosphorus.
From the phosphoric acid solution, after iron reduction, collective chemical concentrates
of uranium and rare earth elements and the remainder of actinium and 230Th are precipitated;
the concentrate is then dissolved in nitric acid and uranium is obtained by extraction and
subsequently separated during reextraction stage in the form of uranyl tricarbonate, leaving
rare-earth elements in the mother liquor of the extraction. Depending on the consumption of
rare-earth elements, mother liquor is purified by sorption or extraction and carbonates are
settled which are then annealed to obtain pure oxides of the total of rare-earth elements
and yttrium. The latter are processed with separation of pure yttrium and europium oxides
for luminophors, ligatures, and catalysts.
From nitric-phosphoric solutions, after separating the collective concentrate by ammon-
ization, evaporation, and granulation, one obtains nitric-phosphoric fertilizer such as nitro-
phos, which contains up to 40% nutrients in the clay form. The benefication concentrate can
be processed with a simplified technology without removing the rare-earth elements, by leach-
ing in sulfuric acid, filtering on rotary filters, extracting uranium and scandium from the
solution, and subsequently obtaining compounds of uranium and scandium (oxides). After am-
monization, evaporation, and granulation, nitrogen-phosphate fertilizer of the ammophos type
is obtained.
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Uranium-phosphorus-thorium ores are processed similarly by decomposing the concentrate
with sulfuric acid, uranium and thorium extraction from the solution, and separate reextrac-
tion of uranium and thorium. The uranium reextract is subjected to additional extractional
purification to obtain pure crystals of ammonium uranyl-tricarbonate, as well as chemical
concentrate from the thorium reextract.
From poor industrial ores of a complex composition, for the first time, uranium and
many valuable accessory components thus have been extracted with positive economic character-
istics and simultaneous utilization of spent chemicals and production of nitrogen-phosphate
ammonium fertilizers. The sale of the by-products cuts the uranium extraction costs approxi-
mately by one half.
EXPERIENCE IN HANDLING SPENT FUEL FROM NUCLEAR POWER
STATIONS IN THE SOVIET UNION, INCLUDING STORAGE
AND TRANSPORTATION
V.
M.
Dubrovskii, V. I. Zemlyanukhin,
A.
N.
Kondrat'ev, Yu. A. Kosarev,
L.
N.
Lazarev, R. I. Lyubtsev,
E.
I.
Mikerin, B. V.'Nikipelov,
A.
S.
Nikiforov, V. M. Sedov,
B.
I.
Snaginskii, and V. S. Shmidt
The development of nuclear power generation in the Soviet Union from the very start has
been planned on a closed fuel cycle, i.e., with regeneration of the spent fuel and the use
of plutonium produced in fast reactors. The spent fuel of nuclear power stations constructed
in other countries with the cooperation of the Soviet Union will also be processed in the
Soviet Union, and therefore the Soviet Union should be considered as the-regional center for
the reprocessing of Nuclear power station fuel.
In the designs of nuclear power stations with VVER, storage vaults are provided, calcu-
lated on cooling of the fuel for up to 3 years, which are attached directly to the reactors.
However, this cooling is advantageous only when the mass construction of fast reactors starts,
in which plutonium will be used and which represents the principal value of the spent fuel.
Taking account of the delay in the construction of commercial fast reactors, the creation
of additional, individually standing storage vaults for spent fuel, calculated on an a 10-yr
operation of the nuclear power stations, is being considered.
However, the solution does not eliminate the necessity for transportation and regenera-
tion of the spent fuel, but only reduces their pace somewhat.
Storage of Spent Nuclear Fuel. The storage of spent fuel is provided for directly in
the case of the VVER-440, VVER-1000, RBMK-1000, and RBMK-1500 reactors, and in individually
standing building in the case of nuclear power stations. The solutions with respect to these
storage vaults are, in principle, approximately identical, and therefore the principal solu-
tions are being considered by the example of a standard design of an additional, separately
standing storage vault for nuclear power stations with VVER-440 (Fig. 1).
The additional storage vault for the spent fuel from nuclear power stations with VVER-
440 has been developed on the basis of experience in the designing and operation of similar
vaults in the Soviet Union and in other countries. This vault has been calculated on the
total receipt of spent fuel from four reactors over 10 years, which amounts to IU5000 fuel
element assemblies or " 600 tons of fuel.
The grouping,' technological process, and plant have been solved by taking account of the
maximum utilization of the standard type and nonstandard plant, used in similar storage vaults.
Translated from Atomnaya Energiya, Vol. 54, No. 4, pp. 293-297, April, 1983.
0038-531X/83/5404-0309$07.50 ? 1983 Plenum Publishing Corporation 309
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Fig. 1. Spent fuel storage vault (longitudinal section): 1)
bridge crane 125/20 tons; 2) bridge crane 15 tons; 3) coffin
with fuel element assemblies; 4) container wagon.
The additional storage vault consists of a section for the admission, transshipment, and
distribution of the transport containers, a section for the storage of the fuel, and a sec-
tion for technological systems and services for providing the fuel-storage conditions.
The section for the receipt and transshipment of the transport containers includes a
transport corridor and a transshipment room. The transport corridor is located below the
transshipment room and is joined to it by a shielded opening for transfer of the containers.
In the transshipment room are located: the compartment for transshipment of the containers,
a compartment for washing and discharging the fuel element assemblies, pits for washing out
and decontamination of the containers, pits for the storage of equipment and instruments,
and also a section for small repair work of technological plant. The transshipment and
washing compartments are joined by a transfer corridor with storage compartments for spent
fuel element assemblies.
The transshipment room is equippped with a crane with a load lifting capacity of 125/20
tons (with supplementary lifting values of speed) and other devices for operating with spe-
cialized equipment, benches for the technological monitoring of contaminated containers, and
also a tool for transfer operations.
The compartment for the storage of the spent fuel consists of a tank filled with water,
and a transport room. There are four sections in the tank, in which the coffins with the
spent fuel element assemblies are installed for storage and a transfer corridor joining the
tank compartment and the transshipment and fuel element assembly washing compartment.
In the tank compartments, a slit cover is provided, which ensures normal working condi-
tions for the personnel. The slits are the transport paths for the conveyance of the fuel
coffins and they provide the necesary orders of separation of the coffins in the compartments.,
Below the tank is located the transport room of the storage vault. The spent fuel is stored
without the constant presence of servicing personnel. At the panel of one of the nuclear
power station units, secondary monitoring signals are displayed, about the state of the prin-
cipal technological parameters determining the normal storage conditions of the fuel element
assemblies: temperature of the water in the tank compartment, the level of the water in the
compartments, data about the functioning of the cooling and purification systems, and the
conditions of the air medium of the tank. Storage is carried out under a protective layer
of water (ti 3 m above the active part of the fuel element assemblies).
The compartment for the technological systems and the service for ensuring the storage
conditions of the spent fuel is an annex, in which are located the cooling and water purifi-
cation systems of the tank, the technological and sanitary-technical ventilation, electrical
supplies, washing and decontamination, and also a desk, operator controlled, medical check
point, administrative and other rooms.
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ni
25440
Fig. 2. Railroad container wagon for transporting spent fuel
from nuclear power stations with VVER-440: 1) transporter; 2)
auxiliary compartment; 3) load compartment; 4) container; 5)
ventilation and heating system.
The fuel is conveyed from the nuclear power station units in a TK-6 transportation con-
tainer, in which a transport coffin with 30 spent fuel-element assemblies are placed. After
a 3-yr cooling in the reactor pond, the fuel is loaded into a container which is transferred
to a specially equipped trailer in the transportation corridor of the storage vault. The
container is lowered into the room with the bridge crane, and installed in the transshipment
compartment which is filled with water. The transportation coffin with the spent fuel ele-
ment assemblies is withdrawn from the container with the bridge crane, and transferred to the
tank compartment where it is placed in storage. The container from the transshipment compart-
ment is transferred to washing and decontamination, after which the empty coffin is placed
in it. The prepared container is taken away for a new load.
Transportation of Spent Nuclear Fuel from Nuclear Power Stations. The spent fuel from
nuclear power stations with VVER-440 is transported by train, consisting of four to eight
TK-6 container-wagons (Fig. 2) and two VS-TK-3 and VS-TK-4 escort-wagons. For the transporta-
tion of fuel from nuclear power stations with RBMK, a TK-11 container wagon is being devel-
oped, and from nuclear power stations with MR-1000 - a TK-10 container-wagon.
The TK-6 container-wagon is a 12-axle railroad transporter, equipped with a body inside
which the shielded container is installed. The container (load) compartment of the body is
provided with thermal insulation. There is a ventilation-heating facility in the body, by
means of which the load compartment can be cooled or heated up. This solution ensures the
necessary thermal conditions for the loaded spent fuel container under all possible transpor-
tation conditions.
The shielded container, with the coffin installed in it, comprises the transportable
packing assembly. The State Committee for the Utilization of Nuclear Power in the Soviet
Union issues a certificate, according to which this packing assembly loaded with fuel ele-
ment assemblies with a burnup of up to 24 GW?day/ton, with the inside cavity of the container
filled with gas, is defined in accordance with the IAEA classification as a type B(U) package.
For this, the maximum heat release of an individual fuel element assembly is limited to 340
W, and the total heat release of the container is 8 W. Calculations and tests of the pack-
ing assembly have shown that with the stated values of heat release the temperature of the
gas coolant in the container does not exceed 175?C, and the temperature of the fuel element
cans does not exceed 200?C.
When transporting the spent fuel of VVER-440 with a burnup of more than 24 GW?day/ton
in containers filled with gas, an additional biological shielding from neutron radiation is
required. It can be filled on the outside (water jacket, hydrogen-containing materials), and
also it can be accomplished by filling the container with water.
Different methods of loading the spent fuel element assemblies into the container have
been tested at different nuclear power stations: the "dry" loading of the coffin with 30
fuel element assemblies into the container, previously removed from the transport facility;
dry loading of the coffin with the fuel element assemblies into the container without re-
moving it from the wagon; "wet" piecemeal loading of the fuel element assemblies into the
container installed in the cooling pond; and wet loading of the coffin with the spent fuel
element assemblies into the container.
From the point of view of the duration of operations, the dry loading scheme must be
accepted as the most suitable. Despite the fact that additionally a transfer container must
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be used, and time is involved in the additional operations, the overall duration of loading
is found to be less than for the wet version. The surface of the container in the case of
wet loading is inevitably contaminated with radioactive substances, present in the water of
the storage vault tank, and the complex surface configuration (the presence of fins, acces-
sories, etc.) makes decontamination a time-consuming and prolonged operation, the maximum
duration of which amounted to about 24 h per container. As a result of this, the overall
duration of the presence of the railroad echelon at the nuclear power station during loading
is increased significantly.
The containers, after the fuel element assemblies are loaded, are stood until the tem-
perature inside the container has stabilized. The accompanying documents are drawn up on
finalization of the monitored measurements of the level of radiation, the temperature of the
medium inside the container, and other parameters. The conveyance of the spent fuel from
the nuclear power station is effected mainly by railroad transport.
In practice, different schemes of conveyance are used, in which in addition to railroad
transport, truck and water transport also are used. Conveyance is carried out in accordance
with "Regulations for the Safe Conveyance of Spent Nuclear Fuel from Nuclear Power Stations,"
and "Technical Conditions for the Assembly of Spent Fuel Elements."
On the whole, all types of trnasportation of spent fuel are conducted without incident.
and unforseen holdups. The parameters of the packing outfit have corresponded to the tech-
nical conditions and instruction for operation. The temperature was below the maximum per-
missible.
Reprocessing of Spent Nuclear Fuel. From the nuclear power station the spent fuel ele-
ments are transported to the reception department (storage vault) of the reprocessing facility,
similar to their solutions for storage vaults in the case of nuclear power stations. The
wagons, with the containers, are conveyed in turn into the transport corridor. The scheme
for discharging the wagons without removing them from their container is adopted.
The coffins with the fuel elements are transferred from the storage vault to a room
where the batches of fuel are grouped according to the fissile material content, using non-
destructive monitoring. The methods for recording the y-quanta of the fission products or
the neutrons of the natural radiation are used for this.
The grouped batches of spent fuel elements in the coffin are transferred to the next
room where, by means of the electric arc method and with a disk saw, the shafts of the fuel
element assemblies are cut off and directed to the burial ground; the cores of the fuel ele-
ment assemblies are directed to the chopping machine for size reduction. The fuel fragments
from the chopping machine enter the dissolver.
A continuous action vibration dissolver and a periodic action dissolver with pneumatic
charging of the fragments have been developed.
In order to catch and localize 1291 (ti 130 mCi/ton* of uranium) from the gas phase, two
technological processes have been developed: the use of an organosilicon liquid and alkali.
The latter allows the main bulk of the 14C also to be caught simultaneously with the,129I.
The tritium formed during operation of the reactor, due to the ternary fission of 235U,
239Pu, and 241Pu nuclei, the content of which in the spent fuel may attain 700 Ci/ton of U,
more than 95% remains in the graphite during extraction reprocessing and, during concentra-
tion of the latter, it is distributed between the regenerated nitric acid and the condensate.
In this technological process, one version for the concentration of the tritium is to return
the tritium-containing acid and condensate to the process with part of the condensate led
away to burial ("recycle" version).
Another method of localizing the tritium is the thermochemical processing of the dis-
sected fuel in an atmosphere of hydrogen or water, at a temperature of about 723?K. This
operation, before dissolution of the fuel, allows more than 99% of the tritium to be con-
centrated into small volumes. In order to purify the solution of nuclear fuel obtained from
solid impurities, cartridge metal-ceramic filters and centrifuges have been developed.
The extraction process lies at the basis of the nuclear fuel reprocessing technology,
in which tributylphosphate (TBP) in a diluent is used as the extraction mixture. The choice
*1 Ci = 3.700.1010 Bq.
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of the diluent is determined from the results of an investigation of its physicochemical
properties, radiation and chemical stability, the presence of impurities and their effect on
the extraction efficiency of the TBP; the flash point and melting point, viscosity, density,
and surface tension.
In the first extraction cycle, the uranium and plutonium being jointly removed separate
from the main bulk of the radionuclides - fission products and impurities remaining in the
aqueous-tail solution, which proceed to concentration and subsequent extraction of valuable
elements. Then, from the extract by means of reducing reactions, the plutonium is reextracted,
after which the uranium is reextracted with dilute nitric acid.
The aqueous flows, containing the uranium and plutonium, are guided to the corresponding
cycles for final purification and subsequent separation of solid compounds. In the final
plutonium purification cycles, a reflux-process can be used, which promotes an increase of
the purification coefficients with a significant additional concentration of valuable elements.
The extraction process is optimized by means of mathematical modeling on a computer.
The purpose of automatic control is to ensure the conditions which are maximally favorable
to the purification and concentration of the valuable components with a guaranteed retention
of other indices (for example, extraction) at the required level.
For extraction processes, mixer-settlers have been developed best of all; these are
equipments most easily modelled and are amenable to dimensional changes during construction.
They ensure stable operation, even in the case of periodic shutdowns of the process, and they
allow the flows of solutions to be withdrawn from any stage of the reflux extraction. Cen-
trifugal extractors (CE) possess these same merits, which are characterized by a short dura-
tion of contact between the extractant and the highly active aqueous solutions, and small
product and extractant volumes in the cycle, which increases the safety of the process and
reduces the volume of incomplete production. At the same time, the complexity of design of
centrifugal extractors increases the demands on the quality of their manufacture, servicing
and repair.
The purified plutonium from the solution can be converted to the form of the final solid
product by precipitation of the oxalate and then, by calcining the latter, the dioxide can
be obtained. The uranium can be precipitated in the form of ammonium diuranate and, in the
case of necessity for producing mixed uranium-plutonium fuel, in the form of ammonium di-
uranate -Pu(VI) hydroxide, with subsequent calcining and reduction to (U, Pu)02. Processes
for obtaining the oxides (individual and mixed) by direct thermal denitration of the reex-
tracts after their concentration have also been investigated.
For the future use of the final regenerated products, monitoring of the isotopic content
of the latter is required and, in particular, the content of nuclides - sources of the daugh-
ter elements of highly active s-emitters.
A three-zoned grouping of the process has been adopted, according to which all the plants
with the radioactive solutions must be allocated to the first zone behind a biological shield.
Compartments intended for maintenance work must be referred to the second zone, and operator
and shielded compartments, etc. to the third zone. Ventilation of all zones is effected by dif-
ferent systems. In order to ensure the gravity flow of movement of solid and liquid products,
a cascade arrangement of the plant is used.
All plants with a short operating lifetime must have a stand-by. Disabled monitoring-
measurement devices, valves, batchers, etc. must be replaced by remote control, by means of
special mechanisms and without shutting down the process.
The regeneration technology established for the fuel elements of thermal reactors, pro-
vies for their complex reprocessing and the production, together with uranium and plutonium,
also of concentrates of rare-earth elements (REE) and strontium. In order to extract these
elements from the aqueous tailings-solution the first cycle containing all the fission pro
ducts, evaporation (in order to remove nitric acid) and partial neutralization are necessary.
In order to separate the valuable elements from the concentrated raffinate, two tech-
nological processes have been developed, in which a solution of tributylphosphate or di-2-
ethylhexyl phosphoric acid (D2EHPA) in a diluent is used as the extractant. These schemes
ensure more than a 90% yield of rare earths in the concentrate, the mutual separation by a
factor of 20-30, and a purification coefficient from other elements (fission products) of
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more than 100. The use of D2EPHA as the extractant also allows a concentrate of strontium
to be obtained, with a yield of more than 95% and a purification coefficient from REE of
50-100.
The necessity for reducing the acid concentration in the solution before extraction re-
quires special distillation operations, or neutralization of the HNO3. Therefore, a search
is being conducted for extractants capable of extracting the trivalent platinoids, strontium
and cesium directly from solutions of 2-3 M HNO3. Although there are technological difficul-
ties still limiting the feasibility of the practical application of the new extractants, it
has been established that the REE are extracted well from acid solutions by phosphine dioxides
and bidentate extractants, containing simultaneously the groups P=0 and C=O, and cesium and
strontium from 2-3 M HNO3 by dicarbolides, products of the carboranes.
In perpective, in proportion with the emergence of sufficient demand in the national
economy, certain other valuable elements can also be recovered from the raffinates of the ex-
traction schemes for reprocessing fuel elements, for example, technetium and palladium.
The theoretical investigations in the field of the kinetics of extraction processes,
mathematical modeling of counterflow extraction, forecasting the properties of extractants
of different structure and diluents of different nature, and work on the preparative chemistry
of solid compounds of plutonium and other actinides, should serve as the basis for the further
improvement of fuel element reprocessing technology. Successes in the investigations of the
scientific bases of regeneration technology for the fuel of thermal reactors, are also the
basis for the creation of a technological scheme for reprocessing the fuel of fast reactors,
meeting present-day demands. Research work has been directed at the creation of processes
and equipments of very high speed, with a short retention time of the products and with a
high unit output. These investigations are the principal reservation for increasing the labor
productivity of personnel and plant, and the reduction of capital and operating costs, which
is an urgent problem related with the steady growth of capacity of nuclear power generation
and, correspondingly, with the increased demand for fuel regeneration.
PROBLEMS OF RADIATION SAFETY OF ATOMIC POWER PLANT
PERSONNEL AND THE PUBLIC
The great deal of attention devoted to the problem of ensuring the radiation safety of
the personnel and the public at various stages in the nuclear fuel cycle (NFC), which was
discussed at the IAEA conference on experience acquired in nuclear power generation (September,
1982), is well illustrated by the fact that these topics were dealt with at the plenary ses-
sion "Experience in the Domain of Nuclear Safety," the technical meeting "Radiation Protec-
tion in an Atomic Power Plant and Fuel Cycle Plants," as well as a special evening meeting
"Nuclear Power, Energy, and the Health-Physics Conditions in the Environment." These meetings
heard 19 papers.
The main paper at the plenary meeting was t;z ICRP Report "International Recommendations
on Radiation Protection: Five Years of Experiences after Release of ICRP Publication No. 26."
The report discussed the influence of the ICRP recommendations on the administrative aspects
of radiation protection and described the experience gained from the practical application
of this document during the 5 years. Total recognition of this ICRP Publication has been
discussed broadly in all countries, although many of its concepts and definition such as the
effective equivalent dose, collective dose, expected dose, and optimization of protection,
which had at first been received with doubt and scepticism, are gradually being accepted.
The IAEA Report "Present-Day Problems of Nuclear Safety" considered the role of interna-
tional organizations. It devoted much attention to the program of IAEA publications on nu-
clear safety such as "IAEA Safety Standards," "IAEA Safety Handbook," "IAEA Recommendations,"
Translated from Atomnaya Energiya, Vol. 54, No. 4, pp. 303-305, April, 1983.
314, 0038-531X/83/5404-0314$07.50 ? 1983 Plenum Publishing Corporation
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as well as "Informational and Procedural Documents." The program encompases the preparation
and publication of more than 50 publications and is known as the PDSS APP (program for the
Development of Safety Standards for Atomic Power Plants). Of these, 26 publications have
already been published in English and many of them in other IAEA working languages (Spanish,
Russian, and French).
The paper "Principle of Designing for Safety" was devoted to the principal concepts of
safety of atomic power plants in the Federal German Republic. These are the high quality of
the materials, the minimum number of welded seams, optimization of the structural strength
of materials, limitation of the work load on the operator, and monitoring of his working con-
ditions, as well as monitoring and periodic inspection of the sites of coolant leaks. It
is pointed out that careful design is a reliable obstacle to the emission of radioactive sub-
stances into the environment during accidents in atomic power plants.
The utmost interest was aroused by a paper by W. Marshall (Great Britain), "Major Nuclear
Accidents" which pointed out that the discussion on the safety of nuclear power generation is
due primarily to the accumulation of a vast number of unlikely events which might cause a
major accident during the operation of a reactor, with a deadly outcome among the public.
The improper use of terminology has given the public a warped picture of the safety of the
nuclear power industry. The paper demonstrated that if one assumes a major accident in an
atomic power plant, as a result of which the 10 million inhabitants of London will receive
a dose of 1 rem, then this could subsequently cause 1250 cases of cancer with a fatal out-
come. However, this same injury will be caused if every Londoner smokes only 1/20-th of a
cigarette every Sunday. In the last paper of the plenary session ("Processing of Informa-
tion Concerning Experience in Operation and Operator Training," France) it was noted that the
training of operators and their response actions under extraordinary circumstances may have
a major influence on the consequence of an accident in a reactor.
The technical meeting heard 12 papers.
The main approaches to the standardization of the dose loads on the personnel and the
public during the operation of an atomic power plant were presented in an Argentinian paper.
As it pointed out, a reduction in the irradiation is conditional upon improvement of the
monitoring in atomic power plants and the nuclear fuel cycle. In the opinion of the ICRP,
in order for operations involving the action of radiation to be legitimate it is necessary
that the radiation protection be optimal and that in all cases the limits on the individual
dose be observed. However, legitimacy is not a criterion that is used directly in designing
the shielding. The aim of the optimization of the radiation shielding is to obtain adequate
equilibrium between the cost of the shielding and the damage inflicted by the residual action
of the radiation so that the total time of harmful action be reduced to a minimum.
The use of dose limits in the optimization procedure, however, poses difficulties in the
case of irradiation of the population. In view.of this it is proposed to use upper dose
limits, established by component bodies; these upper limits would be such a small part of the
corresponding dose limits so as to avoid overlapping of doses of irradiation from different
sources and to ensure a certain margin for unforeseen circumstances. On the basis of this,
Argentina adopted a limit on the dose of irradiation to"the population from discharges and
emissions from an atomic power plant according to two criteria at once: the individual dose -
30 mrem/yr and standardized collective radiation dose - 1.5 man?rem/MW(E)?yr.
Another Argentinian paper described the experience from the operation of the Atchua
atomic power plant with a power of 367 MW(E) and estimated the irradiation of the public from the
600-MW(E) Cordova atomic power plant commissioned in 1983. It was emphasized that the radia-
tion dose received by the public upon implementation of the plans for nuclear power develop-
ment in the country by 40-50 GW(E) will be 200 man-rem and will be smaller by a factor of
almost 20 than the radiation dose of atomic power plant personnel (3500 man-rem) and by a
factor of 50 than in the production of uranium (about 10,000 man-rem).
In a paper on the technique of optimization and limitation of the dose in underground
uranium mines in France attention is drawn to the necessity of ensuring optimum radiation
safety. In this case an endeavor must be made to make sure that the striving for economy
should not lead to a worsening of the protection. The optimization took account of the fol-
lowing factors: the dose-effect relation, the dosimetry, allowance for the radiometric char-
acteristics in the uranium deposits, the mechanism of irradiation of the personnel, and pro-
tective measures.
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In producing dose loads on personnel engaged in the nuclear fuel cycle in France, the
individual radiation dose received by uranium miners will be in first place (ti 3 rem/yr) while
the collective dose is 0.57 man?rem/MW(E)?yr, i.e., will be equal to ti 40% of the collective dose
of the entire cycle.
A Soviet paper on the practice of ensuring radiation safety in atomic power plants and
predicting the dose loads on the population in connection with the development of nuclear
power generation presented the main premises of Soviet legislation ensuring protection of
the health of personnel and the public as well as reducing the environmental pollution. It
was shown that the mean annual dose received by the personnel of atomic power plants with
VVER, RBMK, and BN reactors with various powers is 0.14-1.2 rem and the standardized collec-
tive dose is 1.1-1.3 man?rem/MW(E)?yr. Thus, conditions have been created in atomic power
plants in the Soviet Union so as to carry out the recommendations of the ICRP that the mean
radiation dose of the personnel should be one-tenth that of the maximum allowable dose of 5
rem/yr. The collective radiation dose of the population upon implementation of the plans
for the development of nuclear power generation is 2300 man?rem/yrand about 98% of this will
be accounted for by gas-aerosol emissions from atomic power plants. The collective risk is
0.33 man/yr, i.e., absolutely undetectable against the background of the natural level of
malignant tumors. This collective dose is equal to the radiation dose received by the popu-
lation from the natural background radiation in only 40 min.
Specialists from Britain characterized the conditions at several nuclear fuel cycle
plants in their country. They pointed out that since 1976 cases of a dose of 5 rem/yr for
employees of the Sellafield plant have become only single cases or do not occur at all whereas
previously there had been 100-140 cases a year. The collective radiation dose has been
stabilized at a level of 5000-5500 man?rem/yr, although the number of the personnel has in-
creased steadily, reaching 10,000 in 1980. The mean individual dose is 0.53 rem/yr, which
is in accord with the ICRP recommendation. They reported that in 1971 tests revealed that
14 workers had 50% of the maximum content of plutonium in their bones.
An analysis was made of the death rate due to cancer and other diseases among profes-
sional workers of three plants in the period 1962-1978. It was shown that the 283 deaths
among workers. and 202 among pensioners correlate well with the average indicators for Great
Britain, thus indicating that work-related irradiation does not cause an increase in the
cancer incidence among employees of the nuclear fuel cycle plants in the country.
The Central Electricity Authority of Great Britain devoted its report to 20 years of
experience in the domain of radiation safety. In the fifties, certain difficulties were en-
countered in establishing the allowable dose limits. In 1960 generally accepted dose limits
were introduced: 5 rem/yr for professional workers and 0.5 rem for the population, although
a dose of 1.5 rem/yr was allowed for individual critical groups near the controlled zones.
Consideration is being given to the development of views on the ICRP allowable dose limits
and their reflection in standardizing documents. Although a dose of 5 rem/yr is the estab-
lished dose, less than 15% of the personnel receives a dose exceeding 0.5 rem/yr and the
mean dose is a mere 0.24 rem/yr. The collective dose is 0.4 man?rem/MW(E)?yr. Gas-aerosol
emissions and liquid discharges from atomic power plants are carefully recorded in annual
reports.
In Japan, eleven BWR and ten PWR with a total power of 16.5 GW(E) are in operation. The
Japanese Nuclear Safety Commission set limits of a dose of 5 mrem/yr for the whole body and
15 mrem/yr for the thyroid of children. The limits set for emissions are 5.104-3.105 Ci/yr
and (3-9).104 Ci/yr for PWR, which leads to a maximum calculated dose of external radiation
at a level of 0.3-3.5 mrem/yr for the whole body and 0.4-13 mrem/yr for the thyroid of chil-
dren with allowance for inhalation and peroral intake of radionuclides by children. The cal-
culated collective radiation dose is 0.02 man?rem/MW(E)?yr for PWR and 0.04 for BWR. The
actual emissions from atomic power plants in Japan, however, are almost two orders of magni-
tude lower than the maximum values. As a result, the collective dose of radiation received
by the population of Japan in 1980 was only 2 man?rem/~yr and with the start-up of atomic
power plants with a total capacity of 51 GW(E) in Japan this will rise to only 10 man-rem/yr.
Such low emissions (a few curies per year) are dueto the'fact that gases are held prior to
discharge into the ventilating stack of the atomic power plant as well as to the low percen-
tage of nonhermetic fuel elements.
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In its paper the German Democratic Republic presented extensive material on the collec-
tive radiation dose of personnel of the Bruno Loischner Atomic Power Plant. It was shown
that as in other atomic power plants, e.g., in the U.S.A., the dose increases with the oper-
ating time. For the second unit of the plant it grew from 0.8 to 2 man?sievertpreactor?yr
for the period 1975-1981. The paper pointed out that the contribution of repairs to the
total radiation dose of the personnel is extremely large, reaching 86-96%.
The report from the USSR generalized research done on such water systems as the Baltic
Sea, the Danube basin, and the Black Sea. In the period from 1974 to 1980 the Baltic dis-
played a tendency toward a decreasing concentration of 137 Cs and 90Sr. The tritium concen-
tration in recent years has been close to 0.2 nCi/'liter. It has also been found that on the
whole processes reducing the concentration of artificial radionuclides predominate. The ob-
served concentrations and dose loads are much lower than the allowable levels. In the near
future radioactive contamination will not restrict the development of nuclear power on the
shores of the Baltic and the Danube basin if the requirements and recommendations on limiting
discharges and emissions are met.
The paper from the Federal German Republic gave data about the radioactive emissions
from reactors and the radiation dose. A tendency has been observed towards reduction of
discharges, especially from reactors built in recent years, which can be explained by ad-
vances in technology. The results attest to the satisfactory state of radiation protection
near atomic power plants in the Federal German Republic.
Great interest was aroused by a paper from the USSR on comparative assessment of the
injurity to the health of the personnel and the population during the generation of electrici-
ty in atomic and thermal power plants. It was shown that the risk due to the deleterious
effect on the health of the personnel and the population during operation of a coal-fired
power plant is several tens of times that from an atomic power plant with the same power.
And when the chemical components of the discharges are taken into account it is less dan-
gerous to live near an atomic power plant than near a thermal plant by a factor of 36,000.
The last paper presented in the given section emphasized that the technical features of
the CANDU reactor (Canada) are characterized by a low defectiveness of the pipes in the cooling
system, which was a mere 0.02% for 12 such reactors as against 2.1% for 85 LWR. It was
pointed out that the radiation dose of the personnel in the Pickering-A and Bruce-A atomic
power plants was 0.43 and 0.2 man?rem/MW(E)?yr, respectively, in 1981, which is below the
corresponding indicators for LWR.
The evening session of the conference heard a report from the United Nations Scientific
Committee on the Effects of Atomic Radiation, prepared by the General Assembly in 1982. In
this document, as in the previous report in 1977, general information is presented about
sources of radiation and the effects produced in the human body and animals under the action
of ionizing radiation. A great deal of attention was devoted to analysis of the dose loads
on the personnel and population at various stages in the nuclear fuel cycle. Another paper
made a comparative assessment of nuclear and other sources of energy (especially coal and
wood) from the point of view of their impact on the environment and man. In the discussion
that followed much attention was paid to the contribution by a Soviet specialist concerning
the health physics and ecological aspects of obtaining electricity from atomic, thermal, and
hydroelectric plants as well as solar power plants. It was demonstrated convincingly that
as far as minimal impact on the environment and lack of influence on public health are con-
cerned, nuclear power is unmatched by any other developed energy source. At the same time
all of the participants in the discussion pointed out that in the nuclear power industry,
especially in the handling of radioactive wastes, there still are quite a few problems that
require serious attention and study.
On the whole, it can be said that the papers read at the conference contained useful in-
formation about:
standards requirements on ensuring the radiation safety of the personnel and population;
practical means of ensuring the radiation protection of the personnel and population at
individual stages of the nuclear fuel cycle;
the actual dose loads on the personnel and population as a result of discharges and
emissions of radioactive products at various stages of the nuclear fuel cycle;
Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196ROO0300020004-6
 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6
estimates of the development of nuclear power in different countries in the next few
decades and the attendant dose loads on the population of these countries;
comparison of the nuclear fuel cycle with other methods of obtaining electrical energy
from the point of view of protection of public health and protection of the environment. It
has been shown that improvements in the technology, automation, and increase in the technical
level of operation, and higher qualifications-of the personnel leads to a reduction in the
radioactive emissions and the radiation dose of both the personnel and the population.
The conference proceedings are of unquestionable interest for application in the do-
mestic practice of ensuring radiation safety in atomic power plants and other enterprises of
the nuclear power industry.
Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6
 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6
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