SOVIET ATOMIC ENERGY VOL. 46, NO. 6

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• Document Creation Date: December 23, 2016
• Document Release Date: February 12, 2013
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• Publication Date: June 1, 1979
• Content Type: REPORT

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SATEAZ 46(6) 427500 (1979)
SOVIET
ATOMIC
ENERGY
ATOMHAH 3HEPflI}1
(ATOMNAYA ENERGIYA)
i~rv UU. O-Da I A
Russian Original Vol. 46; No. 6, June,, 1979
December, 1979
TRANSLATED FROM RUSSIAN
CONSULTANTS BUREAU, NEW YORK
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SOVIET
ATOMIC
ENERGY
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SOVIET. ATOMIC ENERGY
A translation of Atomnaya Energiya
December, 1979
Volume 46, Number 6 June, 1979
CONTENTS
Engl./Russ.
ARTICLES
Twenty-Five Years of Nuclear Power - A. G. Meshkov ........................ 427 .371
Status of the First Nuclear Power Station - V. S. Sever'yanov' and V. B. Tregubov ..... 430 375
Radiation Safety of Fast-Reactor Fuel Cycles - O. D. Bakumenko, E. M. Ikhlov,
M..Ya. Kulakovskii, M. F. Troyanov, and A. G. Tsikunov .................. . 433 378
Neutron-Physical Characteristics of BATE Ts Reactors.(Based on the Results
of the Manual Start-Ups of Four Reactors) A. A. Vaimugin, P. G. Dushin,
0. V. Komissarov, A. G. Kostromin, N. I. Logosha, V. F. Lyubchenko,
M. E. Minashin, G. E. Soldatov, and V. N. Sharapov ............. .... 437 382
Ways of Altering the Coefficients of Reactivity in RBMK Reactors - V. I. Pushkarev,
A. D. Zhirnov,. and A. P. Sirotkin .................................. 441 386
Calculation of Coolant Flow Rate by Radiation Methods and Power in First Unit
of Armenian Atomic Power Plant - L. N. Bogachek, A. L. Egorov, V. V. Lysenko,
A. I. Musorin, M. M. Parsadanyan, O. P. Prudnikova, A. I. Rymarenko,
A. G. Tevanyan, V. L. Timchenko, S. G. Tsypin, and V. A. Shmondin ......... 445 390
Physical Start-Up of IBR-2 Pulsed Research Reactor - V. D. Anan'ev, V. A. Arkhipov,
A. I. Babaev, D. I. Blokhintsev, Yu. M. Bulkin, B. N. Bunin, E. D. Vorob'ev,
N. A. Dollezhal', L. V. Edunov, V. S. Lavrukhin, V. L. Lomidze, V. V. Melikhov,
Yu. I. Mityaev, Yu. N. Pepelyshev, V. P. Plastinin, A. D. Rogov, V. S. Smirnov,
I. M. Frank, N. A. Khryastov, E. P. Shabalin, and Yu. S. Yazvitskii ........... 449 393
LETTERS
Application of Computer Tomography. for Fuel-Element Inspection
- E. Yu. Vasil'eva and A. N. Maiorov ............................... 458 403
Prospects for the Use of Carbon-Carbon Type of Materials in Nuclear Power
Engineering - K. A. Andrianov, K. P. Vlasov, L. L. Razumov, S. A. Kolesnikov,
V. I. Kostikov, I. I. Fedik, and L. M. Khananashvili ...................... 461 406
Electromagnetic Converter for Flow Rate of Liquid Metal in Fuel Assemblies
- V. P. Kornilov and N. I. Loginov ................................. 464 408
New Method for Detecting Boiling of Water in a Reactor - I. I. Zakharkin ....... ... 467 410
Energy Distribution of 235U Fission-Product Radiation for a Short Irradiation Time
- M. A. Harkina, E. S. Stariznyi, and A. Kh. Breger ..................... 469 411
Interaction of the Coolant for Prolonged Flow around Bunches of Rods
- V. K. Rukhadze ............ ............................ 471 413
Resonance Integral for Neutron Capture by 244PU - A. A. Druzhinin, N. G. Krylov,
A. A. Lvov, Yu. M. Odintsov, and V. L. Sumatokhin ... ....:......... . 473 414
Absolute Measurements of the Cross Section for the Fission of 241Am by 2.5-MeV
Neutrons - B. M. Aleksandrov, Yu. A. Nemilov, Yu. A. Selitskii,
S. M. Solov'ev, V. B. Funshtein, S. V. Khlebnikov, and B. M. Shiryaev.......... 475 416
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CONTENTS
(continued)
Engl./Russ.
SCIENCE ARCHIVES
History of the Startup of the First Nuclear Power Station (Documents and
Information) .................................................. 477 419
INTERNATIONAL COLLABORATION
Conference of the International Working Group on INTOR - V. I. Pistunovich ........... 483 423
Conference of the Working Group on Power Generation - M. V. Agranovich ............ 484 423
CONFERENCES, SEMINARS, AND SYMPOSIA
Conference on Heat Exchange and Hydrostatic Resistance during the Motion of a
Two-Phase Flow - P. A. Ushakov and A. A. Ivashkevich ..................... 485 424
Seminar on the Reliability of Nuclear Power Generating Facilities - A. I. Klemin........ 486 425
Seminar on Steam-Generators for Fast Reactors - B. I. Lukasevich ................. 487 425
French-Soviet Seminar on Reactors for Heat Supply - S. A. Skvortsov ............... 489 427
Seminar on the Water Treatment, Water Cycle, and Corrosion Protection in
Thermal and Nuclear Power Stations- Yu. V. Balaban-Irmenin ...... . ......... 490 .428
Soviet -American Symposium on Hybrid Thermonuclear Reactors - G. E. Shatalov ..... . . 491 428
Twelfth European Conference on the Interaction of Laser Radiation with a Substance
V. Yu. Baranov and A. Yu. Sebrant ..... 493 429
NEW BOOKS
A. M. Petros'yants. Problems of Nuclear Science and Technology
- Reviewed by Yu. I. Koryakin .......... . . .......................... 495 430
A. A. Vorob'yen, N. S. Rudenko, and V. I. Smetanin. Spark Chamber Techniques
- Reviewed by B. P. Maksimenko ............. ...................... 496 431
E. E. Kovalova. Atlas of the Dose Characteristics of External Ionizing Radiation
(Handbook) - Reviewed by Yu. V. Sivintsev .................. ......... 497 432
I. Ya. Emel'yanov, V. V. Voskoboinikov, and B. A. Maslenok. Design Principles
of Nuclear Reactor Control Mechanisms - Reviewed by I. T. Gusev ............... 497 432
ANNOUNCEMENTS
Fifth Scientific Conference on Power Generation in the High School of Engineering
at Zittau (German Democratic Republic) ............... ................ 499 433
The Russian press date (podpisano k pechati) of this issue was 5/29/1979.
Publication therefore did not occur prior to this date, but must be assumed
to have taken place reasonably soon thereafter.
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ARTICLES
TWENTY-FIVE YEARS OF NUCLEAR POWER
Twenty-five years ago, on June 27, 1954, the first nuclear power station was started up. Nuclear energy
was being used for the first time for a peaceful purpose, the generation of electricity. That this occurred here
in the Soviet Union was vivid evidence of the peaceful aspirations of our country and of the desire to make
nuclear energy serve the welfare of mankind as a new, practically inexhaustible energy source. This note-
worthy event was the beginning of the extensive development of nuclear power in the whole world. In the second
half of the 1950s the first nuclear power stations in Great Britain, the U.S.A., and France were opened.
. The power of the first nuclear power station was 5000 kW (electrical). The present installed power of
nuclear power stations over the world has already exceeded 100 million kW and continues to grow at a rate
considerably greater than the rate at which electrical power generation is growing as a whole. Despite several
objective circumstances which have caused a reduction in the absolute value of nuclear generated power com-
pared to the values projected several years ago, the share of nuclear power stations in the'electrical generat-
ing power of the world should be about 45% by the end of the century [1]. The extensive development of nuclear
power is due primarily to the limited supplies of organic fuel and their nonuniform distribution over the earth.
This is especially true of liquid fuel, whose share in the energy balance of most countries has grown rapidly.
According to estimates [21, a growth in the demand for oil at the rate established over the last 20 years will
lead to the exhaustion of extractable reserves by the end of the century. The technically extractable reserves
of coal are not much greater than those of oil [2, 31. Despite the increased cost of equipment and prolonging
of construction periods due to stricter safety requirements, nuclear power stations are successfully competing
in most countries with organically fueled power stations, a competition which is, of course, aided by the rising
cost of the organic fuels. In addition, the use of organic fuels may be limited in the future because of pollution
of the atmosphere by sulfur dioxide and solid particles and the expected long term ecological effects of the ac-
cumulation of carbon dioxide in the atmosphere. Nuclear power stations substantially reduce the harmful effect
on the environment. The vast potential reserves of energy included in nuclear fuel, the ecological aspects, and
the possibility of locating nuclear power stations at the most convenient points without territorially linking them
to the fuel supply bases make nuclear power the most realistic energy source in the future.
The Soviet Union is one of the countries that has assured reserves of organic fuels. However, the energy
resources are not uniformly distributed over the country's territory: about 90% of the fuel and 80% of the hy-
droelectric resources are in the Asian part [41. At the same time the bulk of the demand for electrical energy
is in the European part of the USSR and this results in a need to transport large volumes of fuel there and to
construct expensive, very long electrical transmission lines to carry electricity from the large hydroelectric
and thermal power stations in Siberia. Technicoeconomic studies of the competing ways to cover the energy
deficit in the European part of the country have demonstrated the economic practicality of building nuclear
power stations there. It is in that part of the country that they are now being constructed. The directives of
the Twenty-Fifth Congress of the Communist Party of the Soviet Union [5) provide for the "rapid development
of nuclear power in the European part of the USSR" in the Tenth Five-Year Plan. These plans are successfully
being brought to life. At the end of 1978 nuclear power plants with a total installed power of about 10,000 MW
(electrical) were in operation. The advance of nuclear power in this country since the start up of the first
nuclear power station has proceeded with the' development of two types of thermal power reactors: channel,
with graphite moderator (along the lines of the first nuclear power station), and vessel, with pressurized water
(the VVER) [61. The use of the two types of reactor provides a variety of experience and, because of the sub-
stantial design differences, enhances the prospects for industrial production.
Based on the experience gained in designing and using the first nuclear power station, as well as on the
scientific research carried out on the cooling of the reactor channels by boiling water and steam, a channel re-
actor with nuclear superheating of steam has been built that has a power of 100 MW (electrical). This reactor
was the first unit of the Beloyarsk nuclear power station. It was put into operation in .1964 and followed (in 1967)
by the second unit with a power of 200 MW (electrical). Later on, a design for a powerful commercial power
Deputy Chairman of the State Committee for the Use of Atomic Energy of the USSR. Translated from
Atomnaya I`nergiya, Vol. 46, No. 6, pp. 371-374, June, 1979.
0038-531X/79/4606-0427$07.50 ?1979 Plenum Publishing Corporation 427
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reactor was worked out, that for the RBMK-1000 channel reactor with a power of 1000 MW (electrical) [71
which uses new design features that permit better realization of the advantages of channel reactors. As op-
posed to the Beloyarsk nuclear power station reactors, a zirconium alloy (instead of stainless steel) is now
used as the material for the channels and the fuel element cladding. In the RBMK heat transfer is by boiling
water and separated dry steam is fed to two turbines with a power of 500 MW each. Reactors of this type are
operating successfully at the Leningrad, Kursk, and Chernobylsk nuclear power stations and will be installed at
other nuclear power stations. Operating experience at the Leningrad nuclear power station has demonstrated
the feasibility of increasing the power of the RBMK-1000 reactor and constructing a higher power reactor based
on it. The RBMK-1500 was created in this way. Further improvement of this type of reactor in raising the
unit power, increasing the thermodynamic parameters, and standardizing the units has led to the design of the
RBMKP-2400 with nuclear superheating of steam [8]. One feature of this reactor is unit construction which
allows the power of a single reactor to be raised by adding standard structural components and equipment that
are used in smaller reactors. From this standpoint channel reactors have significant advantages over the other
widely used type of power reactor, the pressurized water vessel reactor. The choice of refuelling regimes is
also simplified in the RBMK since it does not involve stopping the reactor. In addition, channel reactors have
more material volume than the VVER and have a strongly branched piping network in the first loop.
Pressurized water vessel reactors are the most widespread type of power reactor in the world. They
are very compact, have a fairly simple engineering arrangement which makes it possible to isolate the radio-
active loop from the steam power portion, and have a high specific heat release rate. The first power reactor
of this type (with a power of 210 MW (electrical)) was started up in the USSR in 1964 at the Novovoronezh nu-
clear power plant. This was followed by a second unit of power 365 MW (electrical) (in 1969) and a third and
fourth with powers of 440 MW (electrical) each (in 1971 and 1972). A fifth unit is being installed which will
employ a prototype VVER-1000. Standard units with VVER-400 reactors are used in the Kolskaya and Armyan-
skaya nuclear power plants which have been constructed in the past few years and in nuclear power plants con-
structed with Soviet aid overseas. After 1980 a transition to a new size and type VVER-1000 units will be
made [91. Later on units will be installed with the VVER-1000 at the Yuzhno-Ukrainskaya, Kalininskaya,
Rovenskaya, and other nuclear power stations.
Further increases in the unit power of the VVER is made difficult by rail transportation of reactor ves-
sels. This problem can be solved either by seeking new methods of transportation or by delivering the vessels
in parts and assembling and welding them directly at the construction site. The use of fundamentally new tech-
nologies for fabrication of vessels at the assembly site is not excluded in the future.
At the present time and until the end of the century the basis of nuclear power are and will be thermal
reactors of these two types. Use has demonstrated their radiation safety for personnel and the environment as
well as their economic competitiveness relative to organically fuelled thermal power plants. However, thermal
reactors have a significant drawback: a large consumption of natural uranium, which impedes the development
of large scale nuclear power using these reactors alone that could significantly expand the fuel base. Thus, in
the program for expanding the country's power production, an important. place is assigned to the development
and construction of fast reactors. Their principal advantage consists of a fundamental improvement in the effi-
ciency with which natural resources of uranium are used and thereby ensuring broad possibilities for the de-
velopment of nuclear power. In addition, combined systems of fast and thermal reactors offer more manifold
possibilities for the satisfaction of different sorts of energy demands. The development of fast breeder reac-
tors and their use for power generation began in this country about 30 years ago under the direction of
Academician A. I. Leipunskii of the Ukrainian Academy of Sciences. A large amount of scientific and engineer-
ing research was done which confirmed the validity of the basic idea of the fast breeder power reactor: ex-
panded production of nuclear fuel and the practical feasibility of such a device.
In the past the scientific-technical foundations for the industrial use of fast power reactors have been
laid. The efforts of Soviet scientists and engineers led to the power start up in June 1973 of the first large fast
power reactor with sodium cooling, the BN-350 with a thermal power of 1000 MW. This reactor has a loop de-
sign and is intended for production of electricity and fresh water. Heat removal is by liquid sodium through a
three loop scheme. Enriched uranium dioxide is used as a fuel. At present the reactor is in a continuous re-
loading regime and the maximum burnup of the fuel has exceeded the design value by roughly 20%. Although its
arrival at full design power was delayed due to defects in the steam generators, it has been successfully used
now for almost six years and has demonstrated the reliable operation of the other equipment. Its thermal
power is now 650 MW, its electrical power is 120-125 MW, and its distillate production is as much as 80,000
tons/day. At the site of the Beloyarsk nuclear power station the installation of a second BN-600 fast power re-
actor with a thermal power of 1470 MW and an electrical power of 600 MW is now being completed. Compared
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with the BN-350, this reactor has a higher energy release rate, higher steam parameters, and integral assem-
bly, that is, the first loop is in the same tank as the reactor. An improved (faster and more sensitive) system
for monitoring the formation of leaks in the steam generators is being used.
The experience of building and using the BN-350 and BN-600 make it possible to evaluate the advantages
and disadvantages of their design and to gain the industrial and technical experience needed for building com-
mercial high power fast reactors. At present work is under way on such a reactor with a power of about 1600
MW (electrical), the BN-1600, which is intended for nuclear power stations and which should later become the
commercial prototype for fast power reactors. The introduction of the first standard assemblies with the
BN-1600 will be an important milestone on the way to the large scale construction of fast sodium reactors.. If
needed, the unit power of fast reactors may be increased to 2000-2400 MW (electrical), and the problem of on-
site assembly of the reactor vessel can be solved much more simply than in the case of the VVER since the
sodium pressure in the vessel is less than 0.1-0.15 MPa as opposed to 12.5-16 MPa in the VVER.
The program for work on fast reactors also includes research on possible applications of helium or the
dissociating gas N2O4 as a coolant. Gas-cooled fast reactors could have several advantages over sodium-cooled
reactors including a simplified design and a somewhat higher breeding coefficient (for helium). However, there
are serious problems of reliability and safety in such systems which require further study and engineering de-
velopment and testing before a well-based decision can be made about building them. One important problem
is to speed up the mastery of the industrial technology for reprocessing used fuel which is necessary for reuse
of the original material, for isolating secondary fuel, and for realizing the plans for extensive production of
fuel. At present various methods for reprocessing spent fuel are being developed which enable this operation
to be carried out after the fuel is held.for 6-8 months in order to reduce its radioactivity. This is especially
important for breeder reactors. Although this problem is difficult, a certain amount of success has already
been attained [101.
The operation of the external fuel cycle is one of the basic conditions for the development of large-scale
nuclear power. In the future the relative contribution of nuclear power to the overall energy balance of this
country will rise due both to the expanded sphere of applications for electrical energy in ordinary life and in
industry and to the possibility of using nuclear power plants for producing industrial and space heating. The
BN-350 reactor at Shevchenko and the reactors at the Bilibinsk nuclear thermal and electrical station on
Chukotka (with a total power of 48 MW (electrical)) are typical examples of dual purpose reactors that produce
both heat and electricity. It seems reasonable in the future to use high-temperature graphite thermal reactors
as a heat source in synthetic fuel production and in the metallurgical industry for primary reduction of iron
ore [ilk.
The successful development of nuclear power in this country will promote the creation of the necessary
machine construction base founded on the "Atommash" ! factory, whose first line has already gone into operation,
and an expansion of the capacity of other machine construction firms. The creation and development of a new
branch of the energy industry based on advanced technology such as nuclear power is one of the important con-
ditions for realizing the plans for progress in the country's economy adopted at the Twenty-Fifth Congress of
the Communist Party of the Soviet Union.
The Soviet Union shares its experience in the field of nuclear power with other countries. The A-1 and
B-1 nuclear power stations in Czechoslovakia, the Reinsberg and Bruno Leuschner nuclear power stations in
the GDR, the Kozlodui nuclear power station in Bulgaria, and the Lovisa nuclear power station in Finland have
been built with technical help from the USSR. In the future the Soviet Union will assist in the development of
nuclear power in other countries.
The principles of socialist integration among the member-countries of the Council for Mutual Economic
Aid have made it possible to organize multilateral cooperation in the design and construction of equipment for
nuclear power stations and in the creation of large machine construction enterprises. The international eco-
nomic unions "Interatomenergo" and "Interatominstrument" have been created for cooperation among member-
countries of the Council for Mutual Economic Aid in the areas of nuclear power machine construction as well
as instrument manufacture and the fabrication of special equipment for nuclear power stations [121.
The Soviet Union, which now has rich experience in the design, construction, and use of power reactors
of different types, fosters the development of nuclear power on a world scale. As a member of the IAEA, our
country actively participates in the development and realization of an extensive program of research to promote
the development of nuclear power and in offering technical assistance to developing countries for the peaceful
utilization of nuclear energy.
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In a welcoming message to the Twenty-First Jubilee Session of the General Conference of the IAEA, the
General Secretary of the Central Committee of the Communist Party of the Soviet Union and Chairman of the
.Supreme Soviet of the USSR, L. I. Brezhnev, wrote that "the International Atomic Energy Agency has been called
upon to play an important role in solving this vital problem and we hope that the agency will strive as hard as
possible to make the atom serve only the interests of peace."
LITERATURE CITED
1. Yu. I. Koryakin, "Nuclear energy at the MIREX X," At. Tekh. Rubezhom, No. 2, 3 (1978).
2. A. Giraud, Trans. Am. Nucl. Soc., 25, 17 (1977).
3. A. A. Beschinskii and K. D. Labrinenko, Teploenergetika, No. 3, 2 (1978).
4. L. A. Melent'ev and E. O. Shteingauz, The Economics of Energy in the USSR [in Russian, Gosenergoiz-
dat, Moscow (1963).
5. Principal directions for development of the national, economy of the USSR over 1976-1980 [in Russian,
Politizdat, Moscow (1977).
6. A. M. Petros'yants, in: Proc. Int. Conf. on Nuclear Power and its Fuel Cycle, IAEA, Vienna (1977), p.
130.
7. A. M. Petros'yants et al., At. Energ., 31, 333 (1971).
8. A. P. Aleksandrov and N. A. Dollezhal', ibid., 43, 337 (1977).
9. A. P. Grigor'yants et al., in: Proc. int. Conf. on Nuclear Power and its Fuel Cycle, IAEA, Vienna (1977),
Vol. 1, p. 328.
10. I. K. Kikoin et al., Experimental reprocessing of irradiated uranium fuel from a BOR-60 reactor by
means of the fluoride method [in Russian, Talk at a meeting of IAEA experts, Leningrad, May 17-21,
1976.
11. N. A. Dollezhal' et al., At. Energ., 43, No. 6, 432 (1977).
12. I. Barbur et al., ibid., No. 5, p. 402.
V. S. Sever'yanov and V. B. Tregubov UDC 621.039.5.56
The failure-free operation of the first nuclear power station for 25 years was an indisputable proof that
the scientific-technical problems asked as it was being built have been successfully solved, that the installed
equipment is reliable, and that the operating regimes have been correctly chosen and has led to a number of
important practical generalizations about the future development and use of power reactors. Besides producing
power, the reactor has been the subject of numerous scientific and applied studies which have now become the
principal purpose for which it is used.
Over these years the steam generator has worked in the specified regimes and the status of the graphite-
water power reactor and the systems and equipment for servicing it have therefore become, after long use, of
great interest for experts working on the manufacture of nuclear equipment. Only once, in 1971, was the plant
shut down for major repairs of the upper communications of the reactor. At that time a detailed inspection of
the metal construction of the reactor and the graphite pile was made and the state of the equipment and of the
plumbing in the main and auxiliary systems was studied. In 1976 it was decided to use the steam from the
steam generator for heating rather than electricity generation, and since the end of that year all the heat pro-
duced by the reactor is used in the heating system of the city.
The accident-free operation of the reactor [1[, the experimental devices mounted on it [21, the improved
quality and reduced duration of preventive repairs, and the precise organization of refuelling operations have
made it possible in recent years to substantially increase the operating time of the reactor at power. Thus,
since 1974 the first nuclear power station in the world, operating as an experimental reactor, has been at full
power 80-90% of the calendar time (see Fig. 1).
The rather high operational, characteristics of the first nuclear power station are primarily due to the
reliable operation of the fuel elements and of the entire fuel channel. During its years of operation, several
Translated from Atomnaya Energiya, Vol. 46, No. 6, pp. 375-377, June, 1979. Original article submitted
March 15, 1979.
430 0038-531X/79/4606- 0430$07.50 ?1979 Plenum Ppblishing Corporation
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moiB
1964 1955 1966 19671968 1969 1970 19711972 1973 1974 1975 1975 1977 1978
Fig. 1. Operation of the reactor at power (as a percent of cal-
endar time).
thousand fuel elements have been used up in the reactor. When they are properly cooled there has not been a
single case of destruction of the inner tube of a fuel element bearing the water pressure in the primary loop.
The average burnup of the fuel composition has been 2-2.5 times above the design value. Damage to the seal
of the outer shell found by the system for monitoring the intactness of fuel elements represents hundredths of a
percent and, since it appears while there is still little burnup, cannot be attributed to radiation damage. Thus,
it can again be said that the design and development. of the technology for making cylindrical fuel elements of
the dispersed type with a stainless steel shell are one of the greatest achievements in building the reactor for
the first nuclear power station.
In 1971 when major repairs were being done, the reactor was unloaded for the first time and it was pos-
sible to inspect the state of the entire graphite pile at once. The reactor elements were examined with a peri-
scope, the diameters of the elements were measured and their change along the height of the reactor pile de-
termined. Samples of graphite were removed from different parts of the pile to study their physical and chem-
ical properties under laboratory conditions. The maximum fluence of thermal neutrons in the core was
5.52.1021 neutrons/sect. The flux of intermediate and fast neutrons in the center of the reactor is roughly equal
to the flux of thermal neutrons and is about 30% of it in the outer elements.
During the examination of the graphite pile It was found that some blocks of the pile, both in the central
and peripheral elements, have a crack or even several branched cracks. They are mostly through cracks,
passing through the entire block, and are longitudinal. The damage was greatest in the center of the core. An
examination of the inner surface of the blocks showed that the graphite in the core has an oxidized surface.
The locking joints of the blocks were in good shape. The outer surface of the graphite blocks was less oxidized
although its temperature was higher by about 30?C. This is explained by the fact that oxidation is mainly due to
oxygen impurities in fresh nitrogen which is sent into the reactor along the central drop tube of the fuel channel
and comes into contact with the inner surface of the block.
Shrinkage of the graphite, which was greatest during the first 10 years of operation, caused a reduction
in the diameters of the elements along the center of the core and the structural gap between the assembly.and
the fuel channel. This required a mechanical reworking of the elements to their nominal dimensions, which,
of course, caused additional damage to the blocks.
An analysis of the results of the inspection and selective examination of elements in later years have
shown that the elements which were used throughout the years of operation to assemble the fuel channels and
in which the temperature did not exceed 700?C are in satisfactory shape. These data make it possible tocon-
clude that operation of a graphite pile for many years at temperatures of up to 700?C is feasible.
The case of the graphite pile is made of sheet carbon steel of thickness 7.5 mm with a zinc coating. The
maximum temperature of the case is 360?C. It was possible to examine and evaluate the state of the outer sur-
face of the case over a limited area through a ventilation hole in the water protection tanks. It was shown dur-
ing this inspection that the zinc coating was preserved, over the bulk of the visible surface, is in a good state,
and has not peeled off due to mechanical action. Where the coating has been destroyed the structure is covered
with a thin even layer of rust but no traces of local corrosion pitting were observed. During an inspection and
testing of the sealing of the compensator gasket of the upper plate with the reactor case no cracks in either the
metal or the seams were observed.
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The water protection tanks for the reactor were made of 15-mm sheet stainless steel. During all the
years of operation no leaks occurred in the tanks either along the welding seams or in the metal. On the outer
walls of the tanks, on the side of the concrete shaft, there are insignificant traces of corrosion. The outer wall
of the tanks which comes up under the protective plates of the reactor have kept their zinc coating in places
and are coated with a thin layer of corrosion products in other places. Pitting was not observed under the layer
of corrosion products. The temperature of the water in the tanks, measured in the upper parts, varied from
40 to 900C depending on the reactor power and temperature of the cooling water.
A survey showed that the inner surface of the tanks as well as the reinforcing ribs were covered with a
layer of iron oxides consisting mainly of magnetite. The most corrosion products are observed at the boundary
where the water level oscillates. In this zone on the hotter wall nearer to the reactor traces of local corrosion
in the form of pits were observed under the oxide layer. The rate of corrosion of the protection water tanks is
0.04 mm/yr, or 0.13 mm/yr including penetration by pits.
The pipes, heat-exchange equipment, pumps, and fittings of the primary loop are made of stainless steel.
An examination of the main pipes of the first loop and the entire operating experience with them confirm their
high reliability. Despite the extensive branching due to the large number of steam generators and circulation
pumps, no defects associated with damage to the water-tightness of the metal itself were observed in any piping.
Individual leaks in welding seams which appeared early in the operating period can be fully explained by the
imperfect welding techniques and methods of checking the welds that were used when the station was being
built.
The "weakest" point in the first nuclear power station is the small diameter pipes in the primary loop,
i.e., the branching system of separate loops leading to the fuel channels and the pulsed lines to the thermal
control equipment. Many of them are so located that defects cannot be repaired without major disassembly and
it is this which led to the need for major repairs in 1971 due to the complete (except for the output collector)
reassembly of the upper portion of the reactor.
As studies showed, the main reason for the failure of the loops was corrosion cracking under stress due
to increased concentration of chlorides on the outer surface. The chlorides precipitated when water which fell
on the leads during leaks evaporated; The first leaks appeared on the demountable joints (ten joints to each
channel under the reactor plate) and also along the seal of the cutoff valve located on the outlet lead. During
the major repairs the number of demountable joints was greatly reduced: they were replaced by welds, the
outlet valve was eliminated, and at present there are no leaks in the leads to the fuel channels.
Operating experience with the heat-exchange apparatus in the first loop (steam generators, refrigerators,
etc.) indicates that it is highly reliable. The four steam generators in the first nuclear power station have
operated 85,000, 86,000, 100,000, and 15,000 h, respectively, in all regimes since the start up day. During this
time 1100 cycles of rapid change of the thermal load on the steam generators have been completed. The cus-
tomary water regime for nuclear power stations has been maintained on the first and second loops and this has
prevented incrustation and deterioration of heat exchange at the transfer surfaces. Cases of loss of sealing in
the steam generators have mainly occurred in the steam superheating sections. During the operation period
the. piping in the steam superheaters of.the three steam generators has been completely replaced twice. There
have been no losses of sealing in the economizer piping (similar in design to the steam superheaters). During
the first years of operation there were two losses of sealing - in the vaporizers (most likely due to technical
defects) along welding seams of single pipes and in the place where the bottoms are attached to the collectors.
Metallographic examinations of the defective pipes of the steam superheaters showed that the cracks have a
transcrystalline character and are formed due to corrosion cracking under stress. One proposal about fatigue
destruction of bracket clamped pipes due to vibration was not verified.
One of the most complicated parts built for the first nuclear power station was the packing gland-type
circulation pumps for the first loop with hydraulic seals to prevent the escape of radioactive water from the
loop. The system which maintained the pressure drops on the packing gland seals required qualified super-
vision. Constant addition of water to the loop due to leaks through the packing glands required constant release
of water from the loop which led to additional expense for cleanup of the excess water. In 1963 a circulation
pump without a packing gland of the vertical partition type with a floating rotor was installed. This pump is
reliable and convenient to use and requires minimum supervision during operation.
The equipment in the secondary loop has also shown high efficiency. To monitor the state of the carbon
steel steam lines of the secondary loop, a section of the main pipe in the region of the temperature expansion
compensator located in the lower point was cut out. Then it was found that the inner surface of the pipe is
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covered with a layer of corrosion products. Its lower part suffered considerable pitting. In several cases the
largest pits have joined and can be characterized as cavities. The state of the metal was also tested by ultra-
sonic defectoscopy. No cracks were found and the depth of pitting (up to 2.6 mm) agreed with the mechanical
measurements. In the upper part of the tube numerous pits of diameter up to 4 mm were observed but their
depth does not exceed 0.05 mm which indicates that corrosion is uniform. This sharp difference in the state of
the upper and lower parts of the steam pipe indicate that corrosion occurs mainly during stops when condensate
accumulates in the lower part of the steam pipe. Conservation of the steam pipe by means of short duration
stops was not tried. Numerous defects in the form of honeycombs appeared on the cooling water lines and in
the fire and drinking water pipes in places with high humidity and temperature. These pipes were completely
replaced in 1971.
In summing up the 25 years of operation of the first nuclear power station, it should be noted that all the
main components of the station have maintained their operating ability and continue to be used successfully.
1. G. N. Ushakov, The First Nuclear Power Station in Russian], Gosenergoizdat, Moscow (1959).
2. The Tenth Anniversary of the World's First Nuclear Power Station in the USSR, A Collection of Articles
[in Russians, Atomizdat, Moscow (1964).
RADIATION SAFETY OF FAST-REACTOR FUEL CYCLES
O. D. Bakumenko, E. M. Ikhlov, M. Ya. Kulakovskii, UDC 621.039.58
M. F. Troyanov, and A. G. Ts.ikunov
The use of plutonium fuel in fast power reactors requires a special approach to the problems of radia-
tion safety in the manufacture of fuel elements and the handling of fuel assemblies (FA). The level of neutron
and y radiation resulting from the natural activity of plutonium fuel limits access to equipment in performing
technological operations with fuel. The high specific energy release rate (250 kW per kg of fuel) and the high
burnup (10% of the heavy atoms) result in an appreciable afterheat and a high fission-product activity. Irradia-
tion of the higher isotopes of plutonium in a reactor leads to further sources of neutron and -y radiation which
must be taken into account in the development of systems for transporting and reprocessing spent fuel.
The purpose of the present paper is to examine the main physical characteristics of fast power reactor
fuel and to determine the radiation environment in handling fuel at various stages of the external fuel cycle, in.
particular in making fuel from reprocessed material and in transporting spent fuel from a nuclear power plant
to a chemical reprocessing plant.
Radiation Environment in the Manufacture of Fuel Assemblies from Reprocessed Material. After an ap-
propriate cooling period the spent fuel is subjected to chemical reprocessing to remove the fission products
and to recover the unburned reusable fissionable materials. Depending on the fission-product decontamination
factor, the fuel will have a certain residual fission-product activity. If decontamination is complete, the activ-
ity of the reprocessed fuel is determined solely by its natural activity. The sources of this activity are the
following:
1. a activity which results from the decay of plutonium nuclides. The overall a activity of uranium-
plutonium fuel used in plutonium thermal reactors with the isotopic composition 236pu, 238pu, 239pu, 240pu, 241pu,
and 242Pu equal to 10-', 1, 58, 23, 14, and 4% [1[ is 230 Ci/kg of PuO2 and is determined by 238Pu; the a activity
of plutonium of equilibrium composition (according to estimates of the authors 236pu, 238pu, 239pu, 240pu, 241pu,
and 242Pu comprise 16-1, 0.2, 62, 27, 6, and 5%) from a fast reactor is half as large. The high a activity of re-
processed material requires leak-tight technological equipment.
2. Neutron and y activities of the fuel, which determine the external irradiation of personnel in the fab-
rication of fuel elements and FA of reprocessed material. The neutron activity of the fuel results from the
spontaneous fission of 238Pu, 240Pu, and 242Pu, and from the.(a, n) reaction in the oxygen of uranium and pluto
Translated from Atomnaya Energiya, Vol. 46, No. 6, pp. 378-381, June, 1979. Original article submitted
May 24, 1978.
0038-531X/79/4606-0433$07.50 ?1979 Plenum Publishing Corporation
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TABLE 1. Neutron Yield from Plutonium Dioxide (neutrons/sec?g of isotope)
Source of neutrons
236PU02
268PU02
239Pu02 240PU02
241PU02
242Pu02
241AmO2
(a, n) reaction in
oxygen
5,80.105
1,38.104
3,30.101
1,26.102
1,00
1,60
2,70. 103
Spontaneous fission
3,60.104
2,50.103
2,20.10-2
9,00.102
2,30.10-2
1,57.103
1,24
Total yield-
6,16.105
1,63.104
3,30.101
1,03
1,02
1,57.103
2,70.103
TABLE 2. Intensity of Characteristic Radiation from Internal Conversion of y Photons
in the K and L Shells, (y photons/sec?g of isotope)
Energy, keV
236PU
238pU
239pu
240PU
241PU * I
242PU 241Am
12-17 2,15.1012
7,03
9,90.107 7,84.108 1,08.108
1,16.107
4,53.1010
20 2,80.1011
9.100
1,18.107 9,54.107 1,40.107
1,42.106
6,52.109
98 2,26.107
1,20.106
2,50.108 6,95.103 7,20.107
-
3,92.106
114 6,80.106
3,60.105
9.104 2,10.103 2,28.107
-
1,20.106
*Including characteristic radiation of 231U which is the product of the a decay of 241Pu.
nium oxides. The specific neutron yields for various isotopes of plutonium dioxide are listed in Table 1. The
241Am formed in the reprocessed fuel by the P decay of 241Pu has little effect on the neutron activity, although it
has a high specific neutron yield. The neutron activity of plutonium from a thermal reactor is 4.2.105 neutrons/
sec-kg of Pu02 [1], with the contribution from spontaneous fission being approximately equal to that from the
(a, n) reaction in oxygen. The neutron activity may be appreciably increased by the (a, n) reaction in admix-
tures of light elements such as beryllium, boron, fluorine, etc.
We list below the specific neutron yields from the (a, n) reaction in various light elements contained in
the amount of 10 4 Wt.%p in plutonium with the.isotopic composition 236Pu, 238pU, 239Pu, 240Pu, 241Pu, and 242PU
equal to 10-5, 1, 58,
23, 14, and 4%, neutrons/see-kg of Pu:
Beryllium ........... 2.76- 103
Silicon .......... 5.30
Boron .............. 6.45.102
Carbon .......
4
Fluorine ............ 3.03.102
Sodium..........
42
Magnesium .......... 42
Lithium .........
84
Aluminum . .......... 22
In calculating dose characteristics it is necessary to take account of the fact that the average energy of
neutrons from the (a, n) reaction is - 2.5 MeV higher than the average energy of neutrons from spontaneous
fission.
The y activity of uranium-plutonium fuel is determined by the y radiation of plutonium and its daughter
elements accumulated after reprocessing of the fuel. In addition to y rays from plutonium, the characteristic
x rays from the internal conversion of y photons in the K and L shells play a significant role. For most plu-
tonium isotopes the intensity of this radiation is two orders of magnitude higher than that of the y radiation.
In 241Pu (in equilibrium with 2"!U) and 241Am the intensity of the characteristic radiation is approximately equal
to that of the y radiation (Table 2).
The radiation dose rate from unshielded plutonium fuel is determined by the low-energy characteristic
radiation (13-20 keV). The dose rate is 3000 ?R/sec on the surface of plutonium dioxide powder from thermal
reactors, and 10001AR/sec for plutonium dioxide powder of equilibrium composition from a fast reactor. This
results from the difference in 238Pu content. After storage for 1 year the buildup of 241Am in reprocessed
plutonium increases the -y dose rate by a factor of 1.5.
The dose rate at the unshielded surface of fast-reactor fuel pellets enriched 18% in plutonium is
400 ?R/sec. In working with a small amount of plutonium, the dose rate from neutron radiation is low. It ap-
proaches the dose rate from y radiation when working with 30-100 kg of plutonium, depending on the isotopic
composition of the plutonium.
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TABLE 3. Yields of Principal Plutonium isotopes and 241Am (photons/sec?g of isotope)
Energy range, ke
236Pu
238PU
239Pu
240Pu
241Pu
242PU
241Am
26-52
47 $eV ;
44 keV ;
39keV ;
45 keV;
26,3keV; 2,26.106
45 keV; 4,7.104
26 keV; 3,2.109
6, 1.109
2,44.108
5,2.105
3.106
52 keV;
8,1.105
44,7 keV; 1,6?105
33,2 keV; 1,4.108
51 keV; 2.105
43,5 keV; 9.107
56,6 keV; 1.105
52-80
69 keV;
1,8.104
60 keV; 3,4.10'
-
4,57.1010
65 keV- 1,2.106
/
5
76,9ke
; 6,9.10
80-130
2,37.109
5,35.107
5,4.105
7,5.105
9,9.107
1.104
6,64107
(including charac-
teristic X radia-
tion) u
130-200
1,3.108
312.107
2,5 104
4,2.104
1.107
6.102
4.105
200-300
-
2,56.104
2,2.104
6.103
2,27.101
-
1.106
300-400
-
-
7,45.104
3.103
1,48.106
-
2.106
400-600
5,3.107
-
3,2.104
-
-
-
2,7.105
600-800
4,7.107
3.2.105
2.103
1,5.103
-
-
1,6.106
*Taking account of y radiation from 237U which is a product of the a decay of 241Pu.
The large y dose rates cited above necessitate special shielding when handling plutonium material. The
use of a remote-control device can decrease the dose at the hands appreciably. Thus, moving away from the
surface of a pellet by 10 cm decreases the dose rate by a factor of 1000. In working in glove boxes with G-13.
gloves 2 mm thick the y dose rate to the hands is halved.
Only y and characteristic radiation from internal conversion in the K shell need to be considered in cal-
culating the dose rate from a fuel element, since the characteristic radiation from conversion in the L shell is
of such low energy that it is almost completely absorbed by the fuel element cladding. Table 3 lists the y yields
of the principal y radiators in reprocessed plutonium fuel from which the fission products have been completely
removed.
Neutron radiation does not contribute more than 7-15% to the dose rate. The principal contribution to the
dose rate from a fuel element comes from the 241Pu decay products 2rU and 241Am. The dose rate at the sur-
face of a plutonium fuel element from a thermal reactor is 50 ?R/sec, and at a distance of 5 cm it is decreased
to the maximum permissible value for the hands and forearms. The storage of a reprocessed plutonium fuel
element for a year doubles the dose rate as a result of the buildup of 24tAm which has a high specific yield of
60-keVy rays. .
The dose rate from a plutonium FA used in a thermal reactor is also determined by the y and neutron
radiations. For plutonium of equilibrium composition used in a fast reactor the dose rate from neutron radia-
tion is twice as large as that from y radiation, and approximately equal to the neutron dose rate from plutonium
of a thermal reactor.
They dose rate is determined by 241Pu; the buildup of 241Am during the storage of reprocessed material
for 1 year increases the total dose rate by only 5%, since the low-energy radiation from 241Am is absorbed by
the FA wall. The dose rate at the surface of a plutonium FA from a thermal reactor is 80 ?rem/sec, at 0.3 m
from the surface it is 8 ?rem/sec, and at 2 m from the surface of a FA it does not exceed the maximum per-'
missible value.
Radiation safety of personnel during. operations with fuel assemblies requires local hydrogenous neutron
shielding material up to 15 cm thick along the active part of the FA. Then the head and tail of the FA, where
the principal operations are performed, become freely accessible.
Radiation Environment in the Transportation of Spent Fuel. For fast reactors the length of the external
fuel cycle is of fundamental importance; the shorter it is the more efficiently such reactors generate nuclear
power. One of the stages of the fuel cycle is the cooling time of the spent fuel before chemical reprocessing.
The length of the cooling period determines the conditions for transporting the fuel and performing subsequent
operations. A decrease in the cooling time leads to an increase in the afterheat and the fission-product activ-
ity, which complicates transport operations. Using the BN-1600 reactor as an example, Table 4 lists the phys-
ical characteristics of spent uranium-plutonium fuel made of plutonium from thermal reactors.
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TABLE 4. Characteristics of Spent Fuel
Parameter
Cooling time, months
1
3
6
12
Fission product activity;
I
Ci/ton (U, Pu) 02
5,00.107
2,5&107
1,42.107
7,70.106
Ci/FA*
3,40.106
1,76.106
9,66.105
5,25.108
After-shutdown heat-release rate of
fission products
kW /ton (U, PU) 02
235
126
75
38
kW/FA
16,00
8,65
5,10
2,46
Neutron activity
neutrons /[see ton (U. Pu) 021
4,0.106
3,1.108
2,5.106
1,2.106
neutrons/sec-FA
2,7.108
2,1.108
1,7.108
8,1.107
*68 kg of fuel in FA.
Table 4 shows that spent fuel from fast reactors has a high activity and a high heat-release rate. The
neutron activity of spent fuel is determined by the 242Cm and 244Cm contents, which build up to 0.1 kg/ton of fuel,
and have high specific neutron yields of 2.5.107 and 1.2.107 neutrons/sec?g of isotope, respectively. An increase
in the storage time of reprocessed fuel before loading into the reactor leads to an appreciable increase in the
neutron activity of spent fuel as a result of the increase in the 242Cm content. Thus, 1-yr storage is approxi-
mately equivalent to doubling the neutron activity. In addition, an increase in the 242Cm content makes an ap-
preciable change in the heat release in spent fuel. If the reprocessed fuel is stored for 1 yr, the 242Cm con-
tributes 25% of the afterheat for 6 months cooling of the fuel. To eliminate the contribution of transplutonium
elements to the radiation characteristics of plutonium fuel, these elements must be removed in the chemical
reprocessing with a decontamination factor no smaller than 100.
The high fission-product and neutron activities require the construction of special transport facilities,
and necessitate taking measures to limit the radiation to personnel. According to the safety rules, the maxi-
mum permissible equivalent radiation dose rate is 200 mrem/h at any point on the surface of a container, and
10 mrem/h at a distance of 2 m from transport facilities [21. In addition, the rules for transporting radioac-
tive materials [31 impose certain restrictions on the number of FA which can be transported. These restric-
tions amount to limiting the temperature of the surface of the container and the FA. Theoretical and experi-
mental studies show that the optimum container is one designed for the transport of FA with a total heat-release
rate of 20-40 kW [41. For the BN-1600 reactor this amounts to 10-15 FA of spent fuel which have cooled 12
months. Such an arrangement requires a y shield of 20-25 cm of lead, and a neutron shield of 15-20 cm of
hydrogeneous material. It should be noted that the thickness of the y shield is only slightly dependent on the
number of FA being transported and on the cooling time of the spent fuel, because of the large self-absorption
of y rays in a FA.
LITERATURE CITED
1. 0. D. Bakumenko et at., At. Energ., 44, No. 2, 140 (1978).
2. Safety Rules for the Transportation of Radioactive Materials [in Russian, Atomizdat, Moscow (1974).
3. Safety Rules for the Transportation of Radioactive Materials. Series of publications on' safety in Russians,
No. 6, IAEA (1973).
4. Decision of a Conference of IAEA Experts on Problems of Reprocessing Fast-Reactor Fuel in Russian,
Leningrad, May 17-21,.1976.
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 Declassified and Approved For Release 2013/02/12 : CIA-RDP10-02196R000800010006-0
NEUTRON-PHYSICAL CHARACTERISTICS OF BATETs
REACTORS (BASED ON THE RESULTS OF THE
MANUAL START-UPS OF .FOUR REACTORS)
A. A. Vaimugin, P. G. Dushin, O. V. Komissarov, UDC 621.039.524.2:621.039.519
A. G. Kostromin, N. I. Logosha, V. F. Lyubchenko,
M. E.. Minashin, G. E. Soldatov, and V. N. Sharapov
The neutron-physical characteristics of Bilibinskaya ATETs (BATETs) reactors during the period of its
design have been investigated experimentally on a uranium-graphite assembly with subcritical insertion which
simulated .the active zone [1, 21. The subcritical insertion was small, and its volume was less by a factor of
-20 than the volume of the active zone of the reactors. The assembly was used to check the accuracy of the
methods which were utilized in connection with the calculation of some neutron-physical characteristics of the
lattice of the reactors, in particular, the neutron multiplication coefficient K... For this purpose the spatial
distribution of the fluxes of thermal and resonant neutrons and the cadmium ratios with respect to 238'235U,
197Au, 55Mn, and In were measured for all types of layouts of the active zone.
The principal neutron-physical characteristics which must be known for the operation of the reactors
were determined during manual start-ups, when the effects, reactivity margin, and effectiveness of the control
and safeguard system (CSS) units were measured. A large number of measurements were made to check and
refine the methods of controlling the energy distribution in an operating reactor. The investigation of the phys-
ical characteristics during the start-up of the first reactor was communicated in [31. The remaining three
power units were introduced into the system in 1974-1976 (Table 1). After the introduction of the last reactor
into the system it became possible to compare the neutron-physical characteristics of all four reactors.
The BATETs reactors have the identical electrical capacity (12 MW each), thermal power takeoff
(25 Gcal/h), and essentially identical design. They belong to the channel-type of reactors with graphite moder-
ator. Heat removal from the tubular fuel elements is accomplished with boiling water by means of its natural
circulation. The layout of the reactor installation is single-loop with separation of the steam in a drum and
supply of it to the central-heating turbine. Some difference of the reactors of the third and fourth power units
from the first two is associated with a change in the arrangement of the four emergency protection (EP) rods
in the active zone, due to which the lattice of absorbing rods has a regular appearance (Fig. 1).*
Probing of the Pile. Probing of the graphite pile (within the boundaries of the active zone) was performed
in the reactors prior to charging the engineering channels (EC) with the use of a neutron source and boron
counters. The aim of the probing is to check the uniformity of the pile and to estimate its diffusion properties.
No appreciable nonuniformities were detected in the reactors of the first three units, and the properties of the
pile corresponded to the design parameters. Defective graphite elements were detected in the pile of the fourth
unit, and they were removed and replaced by new ones prior to the start of charging the EC of the reactor.
Charging of the Reactor. A complete charge of each reactor consisted of 217 EC with uranium of 3%
enrichment (EC-3) and 56 EC with uranium of 3.3% enrichment (EC-3.3). Since the EC-3 differed somewhat in
uranium content, they were placed in the reactor in a specified order in order to produce a certain shaping of
the charge, which resulted in an increase of the uranium charge with reactor radius. The EC-3.3 were loaded
into 56 peripheral cells arranged on the boundary with the reflector (Table 2).
Reactivity Margin and Compensation for It. With a complete charge the reactivity margin Ok/k was
measured for two states of the reactors: with water and without water in the EC. As a result it has been es-
tablished that zk/k differs inappreciably for these two states (Table 3). When water is present in the EC, the
average value is Ak/k = 0.112, and without water in the EC-0.107, which agrees with the design value of 0.11
[4, 51. In the case of a uniform distribution of absorbing rods in the active zone the reactivity margin of the
*A map of the reactor of the first two power units is given in [ 31 .
Translated from Atomnaya Energiya, Vol. 46, No. 6, pp. 382-386, June, 1979. Original article submitted
August 7, 1978.
0038-531X/79/4606-0437$07.50 01979 Plenum Publishing. Corporation 437
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 Declassified and Approved For Release 2013/02/12 : CIA-RDP10-02196R000800010006-0
Fig. 1. Map of the reactor of the third and fourth BATETs
units: 0. ?. and a) rods of the manual and 'automatic control
and the emergency protection; ---) boundary of the active
zone.
TABLE 1. Start-up Date of the BATETs TABLE 2. Charging of 235U in BATETs
First at-
Duration
First
Duration
tainment
of ma-
switching
of main-
Unit
of criti-
nual
on of the
tenance
cality
start-up,
turbogen-
of
deesign
days
erator un-
t
da
s
der load
y
First
11.12.73
21
12.01.74
98
Second
7.12.74
16
30.12.74
47
Third
6.12.75
10
22.12.75
55
Fourth
17.12.76
7
27.12.76
14
Amount of 235[7, kg
Max. scat-
235
Unit
I
U
ter of
217
56 1
'
total
content in
EC-3
EC-3.3
EC-3, olo
First
165,74
48,06
213,80
7,5
Second
164,52
46,76
211,28
5,5
Third
165,48
47,60
213,08
5,0
Fourth
165,54
47,03
212,57
4,6
reactor charged by EC with water is compensated by the complete insertion on the average of -42 CSS rods.
When water is absent in the EC, -31 CSS rods are necessary to compensate for the reactivity margin. The
appreciable difference in the number of rods necessary for the compensation of reactors with and without
water in the EC is explained by the fact that due to the difference in the diffusion properties of the zones the
average effectiveness of a single CSS rod in a zone loaded with EC with water is less by - 30% than in a zone
with EC without water.
Since provision is made in each BATETs reactor for 52 absorbing rods in addition to the EP rods for
compensation of the reactivity margin (the rods of manual (MC) and automatic (AC) control), there is a suffi-
cient margin (- 10 rods) even for the state of a reactor with the maximum multiplication coefficient (fresh EC
with water, a cold nonfunctioning state). Subcriticality of the.reactor with all the MC and AC rods (the EP rods
are inserted) rods completely inserted into the zone amounts to-1.3.10 for the first two reactors and
-2.0.10-2 for the other two. The increase in the subcriticality is caused by a difference in the position of the
rods in the active zone of the reactors.
Emergency Protection. Emergency stopping of the reactors is accomplished by the rapid insertion of
eight EP rods into the zone in -1.6 sec, of which the four central ones are arranged at a distance of 0.15-0.4
Ra.z. from the center of the active zone, and the outer four - at a distance of 0.7Ra.z. (Ra.z. is the radius of
the active zone, which is equal to 206 cm). With such'an arrangement the efficiency of the EP rods depends
significantly on the configuration of the radial neutron field, which is determined primarily by the charging of
the EC and the arrangement of the MC rods compensating for the reactivity margin. Since excess MC rods
remain when the reactivity is compensated, it is possible to change the'shape of the radial neutron field by ar-
ranging the compensating rods in different ways. The effectiveness of the EP rods is - 1.8 times higher in the
case of a field which has a strong rise at the center of the active zone than in the case of a decrease in the field
at the center of the zone (Table 4).
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TABLE 3. Experimental Data on the Reac- TABLE 4. Effectiveness of the Eight EP
tivity Margin Rods, Ok/k?102
Presence
Ak/n? J02
No. of ab-
Unit
of water
the
i
for each
av
sorbing
n
unit
,
rods
~C
First
Yes
10,5+1,5
11,2?1,0
43,0?0,5
Second
>
11,5?1,5
41,5?0,5
Third
?
11,7?1,5
41,5?0,5
Fourth
11,0?1,5
41,0?0,5
First
No
10,4?1,5
10,7+1,0
32,0?0,5
Second
10,6?1,5
31,0?0,5
Third
11,1?1,5
31,0?0,5
Arrangement
EC with water
EC without water
of the MC
expt. I
calc.
expt. I
calc.
rods
Uniformly
1,38?0,07
1,48
2,03?0,10
1,75
throughout
the active
zone
All extracted
rods are:
in center
1,79?0,09
1,86
2,70?0,14
2,53
on periphery
1,08?0,05
1,05
1,57?0,08
1,43
It should be noted that positive interference is observed among the EP rods of the peripheral group: the
overall effectiveness of this group of rods is higher by 20% than the sum of the effectivenesses of the remain-
ing rods. The reverse pattern is observed for the EP rods of the central group: in this case mutual interfer-
ence among the rods results in a decrease of the overall effectiveness of this group by 20%. When the MC rods
are arranged uniformly, the central and peripheral groups of EP rods are identical with respect to their own
effectiveness. The total effectiveness of both groups agrees with the overall effectiveness of all eight EP rods,
i.e., the interference between these groups of rods is not observed. The overall effectiveness of these rods is
reduced by 15% in the case in which one of the rods of the central or peripheral groups is absent ("rejected"),
and by 25% when two rods (one from the central and one from the peripheral group) are absent ("rejected").
But in both cases the effectiveness of the remaining EP units exceeds the effective fraction of delayed neutrons.
In the case of an emergency shutdown of the reactors along with the rapid insertion of the EP rods four
AC and six MC rods are also automatically inserted into the zone. The insertion time of the first rods is
- 20 sec, and that of the second group is 110 sec. Measurements made during manual start-ups have shown
that the additional insertion of these rods increases the effectiveness of the emergency protection from 1.4.10 2
to 2.9.10 2 (for a cold zone charged by EC with water). One should note that the latter value appreciably exceeds
the sum of the power and temperature reactivity effects, which amounts to 1.2- 10 2 for BATETs reactors at the
start of a run. The emergency protection (EP) provides for a sufficiently rapid decrease in the neutron power:
it is reduced by a factor of four in 1.6 sec (Fig. 2). The effectiveness of the EP rods is approximately the
same in all four reactors.
Temperature Coefficient of the Reactivity. In order to determine the temperature coefficient of the re-
activity during. manual startups of the reactors of the second and third units, they were heated up by means of
pumping hot water (tmax = 104?C) through all the EC and the CSS channels. During the heating up the output of
the reactors was held at a level _10-5 Nwom, and variation of the reactivity was compensated for by the AC
rods. The temperature was controlled by thermocouples mounted in four EC and at two points of the graphite
pile (at the center of the active zone and on the periphery, on the boundary with the lateral reflector). Varia-
tion of the reactivity was determined from the depletion of AC calibration rods and with the use of an analog
reactivity meter (61. The initial temperature of the fuel elements and the graphite pile was24?C. Toward the
end of the measurements it increased to 104?C. The overall variation of the reactivity upon heating up was
-4.5.10 3. This corresponds to a temperature coefficient of the reactivity of -5.10 5 ?C-1 for the temperature
range of 24-104?C and the nonboiling operational mode of the reactor.
Neutron Field along the Reactor Radius. The shape of the radial neutron field in BATETs reactors is
very sensitive to the arrangement of the MC and_ AC rods in the active zone. Since the CSS channels form the
correct lattice in the reactor with a 40 x 40 cm grid, the best field is obtained with a symmetric and uniform
arrangement of absorbers in the active zone. With such an arrangement of the absorbers it is possible without
special difficulty to provide a radial neutron field with a nonuniformity coefficient Kr 1.4. As experiments
conducted in the reactor of the first unit have shown, a disruption of the symmetry in the absorber arrangement
caused by the removal of a single peripheral MC rod results in the neutron flux in the'30-40 EC situated near
the removed rod becoming higher by 20-60% than in the symmetrical EC in the other half of the reactor. An
especially great difference is observed in the neutron fluxes for the peripheral EC. Therefore, symmetry is
satisfied in BATETs reactors operating at capacity in the arrangement of the absorbers, for which the varia-
tions in reactivity (in proportion to the fuel depletion) are compensated by the shifting of four or six symmetri-
cal MC rods.
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1,2 r
48
E
v0,6
e 0,4
3
0 02 04 06 08 10 1.2 1.4 16
Channel No.
Fig. 2 Fig. 3
0 60 120 160 240 300 360 420
Distance from pile bottom, cm
Fig. 4
Fig. 2. Variation of the coolant flow rate G (1 - experiment) the thermal power N (2 - cal-
culation), and neutron power (3 - experiment) as'a function of time upon a cut-off of the EP
rods.
Fig. 3. Distribution of thermal neutrons along the diameter of a cold reactor having EC with-
out water: ?) experiment; 0) calculation.
Fig. 4. Thermal neutron distribution with respect to height in a reactor having EC without
water near the reactor center: ?) zone without CSS rods (critical assembly); A) in the zone
of the 32 completely inserted CSS rods.
The radial neutron field during manual start-up was measured by fission chambers, which were lowered
into the central tube of an EC to the level of the active zone. The calculated values of the neutron flux obtained
by the double-group method [71 agree satisfactorily with the experimental values. The neutron distribution
along one of the diametral series of EC is shown in Fig. 3. For this case the mean square deviation of the cal-
culated and experimental values of the neutron flux calculated with all EC taken into account is 4.7%.
Neutron Field with Respect to Reactor Height. Since no special measures have been adopted in BATETs
reactors with respect to equalization of the neutron flux with height, the distribution of the neutron flux with
height is close to cosinusoidal only when the state of the active zone includes completely charged compensation
units (AC, MC). The maximum of the distribution is shifted somewhat (by 10 cm) below the center of the active
zone. This is caused by the fact that upon complete charging the compensation units fall short of the lower
boundary of the active zone by 15 cm. The experimental thermal neutron distribution given in Fig. 4 is obtained
by means of moving the fission chamber in the central tube of the EC. The absence of a burst of neutrons at
the lower reflector is explained by the increased steel content in the lower part of the EC.
In order to reduce the nonuniformity in the energy yield with respect to height upon operation at capacity,
a scheme of reactivity margin compensation has been adopted at BATETs according to which there are no more
than 10 rods (4 AC and 6 MC) in the intermediate position. In this case the increase in the nonuniformity of
the neutron flux with respect to height even in the EC located near.the rods which are in the intermediate posi-
tion does not exceed 16% (one should note that the capacity of these EC is less than the average).
BATETs reactors are identical in neutron-physical characteristics and correspond to the planned imple
mentations. The increase in the number of compensating rods in the critical state for the first reactor in com-
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parison with the rest is evidently associated with the increased 235U charge in it. Therefore, one should expect
that its operating period is also somewhat greater than that of the other reactors.
The experimental data obtained in connection with the start-up of the BATETs reactors can be used to
check other methods of calculating uranium-graphite systems.
The authors thanktheirco-workers of FEI and BATETs who took part in the manual start-ups of the reac-
tors of. this plant.
LITERATURE CITED
1. I. S. Akimov et al., in: Problems of the Physics of Nuclear Reactors in Russian), Tr. FEI, No. 1,
Obninsk (1968), p. 385.
2. V. V. Orlov et al., At. Energ., 36, No. 6, 491 (1974).
3. A. A. Vaimugin et al., ibid., 39, No. 1, 3 (1975).
4. V. M. Abramov et al., ibid., 35, No. 5, 299 (1973).
5. A. A. Vaimugin et al., ibid., 39, No. 2, 110 (1975).
6. B. G. Dubovskii et al., ibid., 36, No. 2, 104 (1974).
7. I. S. Akimov et al., op. cit., No. 6, 427.
V. I. Pushkarev, A. D. Zhirnov, and A. P. Sirotkin UDC 621.039.524.2.034.3
While the RBMK reactor is being brought up to steady-state operating conditions with continuous fuel
recharging, its core undergoes significant changes. Practically from the very first days of reactor operation,
withdrawal of the auxiliary absorbers (AA), which compensate the initial reactivity, begins and the freed chan-
nels are charged with fuel assemblies. In the first period of operation the fresh fuel assemblies practically do
not differ from those of the initial charge in respect of burn-up but gradually the difference in the burn-up of
the various fuel assemblies has an increasingly pronounced effect on their individual physical characteristics,
including their contribution to the integrated coefficient of reactivity for the reactor core.- The combined
change in the structure of the reactor core, the auxiliary absorbers, and the average burn-up of the fuel pro-
duces a quite complex picture of variation of the coefficients of reactivity with the operation of the reactor.
The variation of the various coefficients of reactivity with the burn-up is shown in Fig. 1. Let us note that the
real picture of the variation of the coefficients of reactivity is considerably more complicated since during
operation, especially in the initial period of building up to the rated capacity of the power plant, the thermo-
hydraulic and other parameters of the reactor may differ substantially from the nominal values and this affects
the actual coefficients of reactivity.
Determination of the coefficients of reactivity by calculations or experiment is not the end in itself. The
reliability of the coefficients in many ways determines the accuracy of the prediction of the dynamic behavior
of the reactor and the stability of the energy distribution under steady-state and transient operating conditions.
In great measure it is the job of the control system of the reactor to ensure the stability of the energy distri-
bution. The characteristics of control systems are being steadily improved and it is proposed in future to bring
in completely automated reactor control with the aid of a computer. However, attainment of the best possible
dynamic parameters in the controlled system will greatly simplify the solution of this problem in the future,
and will'at the present time facilitate the work of the operator in controlling the reactor. Hence the need to
modify the coefficients of reactivity of reactors in operation.
As noted in [1), "... the uranium-graphite ratio adopted in the design (RBMK) is not entirely optimal
from the point of view of reactor control under transient conditions. In subsequent reactors the uranium-
graphite ratio will be close to the optimal and the control of the power distribution will be automatic." This
conclusion, arrived at on the basis of the operation of the main block of the Leningrad Atomic Power Plant, can
Translated from Atomnaya Energiya, Vol. 46, No. 6, pp. 386-389, June, 1979. Original article submitted
June 17, 1978.
0038-531X/79/4606- 0441 $07.50 ?1979 Plenum Publishing Corporation
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ac
Fig. 1. Variation of coefficients of reactivity as
the fuel burns up: aC and af, temperature co-
efficients of the moderator and the fuel, respec-
tively; ag,, steam-void coefficient;' P, average
burn-
up of the fuel
.
5 10 e, MW ? day/kg
of 10511?C
-1
also be drawn from computational investigations of the dynamic characteristics of the reactor. Indeed, as the
coefficients of reactivity change the time constant Tot of the development of the deformation of the energy dis-
tribution along the radius (its first azimuthal, the least stable harmonic) decreases, with operation, from sev-
eral hours in the initial period to several minutes under steady-state conditions of fuel recharging (Fig. 1).
The allowable decrease in rol, ensuring reliable and safe control of the reactor by an operator using means
available to him for monitoring and controlling the energy distribution, can be estimated only experimentally.
In particular, it has been shown that when a reactor is operated without local automatic controls the minimum
allowable time Toi is 15-20 min. Accordingly, an optimal set of coefficients of reactivity can be determined for
the operation of the reactor: the steam-void coefficient* aqp should lie within the interval f 1%, the temperature
coefficient of the moderator ac should not be greater than 5.10 5 (?C)-1, and the temperature coefficient of the
fuel af should lie in the interval from -1-10-' to-2.10-5 (?C)-1. As is seen, there is some divergence between
the desired and the existing coefficients of reactivity and appropriate measures should be taken to eliminate
this divergence.
Below we give the results of computational investigations to find the dependence of the coefficients of re-
activity on the principal thermohydraulic and operational parameters of the reactor core as well as to choose
the most effective ways of enhancing the stability of energy distribution in both functioning reactors and planned
reactors. Our attention was focused on the steam-void (a(P ) and temperature (aC) coefficients of reactivity
since it is precisely these coefficients which, firstly, have the greatest effect on the stability of the energy dis-
tribution and, secondly, depend strongly on the core parameters.
In relation to functioning reactors, all the measures to vary the coefficients of reactivity can be arbitrar-
ily divided into two groups:
employment of operating modes which ensure a set of reactor thermohydraulic characteristics for
which the coefficients of reactivity are maintained within the given range;
realization of such technological measures which reduce the coefficients of reactivity and improve
the stability of energy distribution.
This encompasses increasing the fuel enrichment and density, including going over to metallic fuel, in-
creasing the auxiliary absorbers in the core, raising the operational reactivity margin, etc. In the case of re-
actors in the design stage, in addition to the measures mentioned above it is possible to employ measures
whose effect on the coefficients of reactivity are irreversible and unchanging in the course of operation. Among
these are changes in the lattice pitch of fuel assemblies and/or in the effective density of the moderator.
Let us consider the effect of these factors on the coefficients of reactivity. The calculated relations be-
tween the coefficients a and etc and the moderator temperature, mean density of the water, and the number of
rods compensating the operational reactivity are quite strong dependences and must be taken into account in any
comparison of calculations and experiment. This is particularly true of the steam-void coefficient of reactiv-
ity. With small deviations from the nominal parameters these relations are practically linear and can be
represented as follows: increasing the graphite temperature by 100?C increases a (p by 0.2% and decreases aC
by 0.45-10-1 (?C)-1; raising the mean density of the water by 0.1 g/cm3 increases a by 0.37% and reduces aC
by 0.05.10 5 (?C)-1; an increase in the operational reactivity margin by 10 rods reduces a(p by 0.3%. The com-
bined deviation of the heat-engineering characteristics of the reactor core during variation of the power (with
a constant coolant flow rate) introduces a correction to the coefficient of reactivity, as shown in Fig. 2, the
main factor determining the dependence of a(p on the power being the variation in the mean density of the cool-
ant.
*The steam-void coefficient a(P of reactivity is defined as app = Op/o~o, where L.p is the change in the reactiv-
ity when the mean volume content of steam changes by Ocp.
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1 1 - ,(,,%
I ] 9 c(
~i Ei av 2 o 6
4
4 400 450 500 tz;C 44 45- 0,6 1120, g/cm3
Fig. .2 Fig. 3
Fig. 2. Dependence of coefficients of reactivity, on graphite temperature.
Fig. 3. Dependence of coefficients of reactivity on mean density of.water in
reactor core.
Increasing the mean density of the fuel pellets proves to be important not only from the point of view of
increasing the efficiency of the fuel elements but also in respect of having a favorable effect on the steam-void
coefficient of reactivity: an increase of 0.1 g/cm 3 in the mean density of the pellets. reduces a p by 0.27%. The
most operative way of preventing a rise in the steam-void coefficient of reactivity above the allowable level is
that of preserving auxiliary absorption in the reactor core. In this case, clearly, this is to. the considerable
detriment of the neutron balance as manifested by a reduction. of the burn-up fraction of the fuel discharged:
if the number of auxiliary absorbers in the reactor core is increased. by 10, a(P drops by 0.12% but the burn-up
fraction of the discharged fuel falls by 0.54 MW-day/kg.
A more effective measure improving the dynamic characteristics of the reactor is that of increasing the
enrichment of the fresh fuel. In particular, increasing the initial enrichment from 1.8 to 2.0% reduces a. by
0.6%. The attendant increase in the burn-up fraction of the discharged fuel by 3.8 MW'day/kg even somewhat
improves the technicoeconomic indicators of the reactor.
For planned reactors whose design can be changed in accordance with optimization studies the most ef-
fective measure for reducing the coefficients of reactivity is that of reducing the ratio of graphite nuclei to
235U nuclei in the reactor core. In this case the burn-up fraction changes insignificantly and because of the
more efficient use of 2'
'38U and the large build-up of plutonium the power of the freshly charged fuel assemblies
is decreased. The most obvious way of reducing the graphite-uranium ratio is that of diminishing the pitch of
the lattice of channels in the reactor. However, since a reduction of the pitch of the channel lattice is usually
accompanied by difficulties with the separation of the lines bringing in and carrying away the coolant, proposals
for decreasing the effective density of the graphite are also of interest. As shown by calculations, both of these
factors affect the coefficients of reactivity in the same way and can be represented quantitatively by the ratio
of graphite nuclei to 235U nuclei, NC/N5 (Fig. 3).
Let us note that the discussion presented here applies to a 1.8-2.0% enrichment of the fuel charged. As
the burn-up fraction increases and, consequently, the enrichment of the fuel charged is increased, say, above
3%, it may be that the lattice pitch or the effective density of the graphite must be increased and not decreased.
The steam-void coefficient of reactivity is decreased substantially in the transition from dioxide fuel to
;
fuel with a higher density, e.g.,.uranium metal. This transition also possesses many other advantages [ 21
therefore, the expected major difficulties of a technological and operational nature may be justified.
Of interest are the results which were obtained upon considering the influence of the shape of the distri-
bution of the neutron flux over the height on the steam-void coefficient of reactivity of the reactor. When the
dependence of a 9 on the density of the water is taken into account, it may be expected that the average steam-
void coefficient of reactivity for the entire reactor core should be different for different heights of the energy
distribution. According to calculations it turned out that the increase in a(P during the transition from an en-
ergy distribution with a maximum in the upper part of the core to one with a maximum in the lower part of the
core is -1.0%, which undoubtedly affects the stability of the radial energy distribution. This type of effect
must necessarily be taken into account when choosing the distribution of the absorbing properties of the auxil-
iary absorbers over the length of the channel in the case when, to reduce the coefficients of reactivity in the
reactor core, some of the auxiliary absorbers initially charged are kept for a sufficiently long time. It is a
quite complicated procedure to optimize their properties by calculation since in the process indeterminacies
of a physical as well as a thermohydraulic character are superimposed upon each other. When the channel
power is changed, the density of the water is redistributed over the entire length of the steam-producing seg-
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wm
0
OM
Gra Black ra '
0 1 2 3 4 5 6 m
a
Gmy
Mack
Gray
n
el
ac
Gra
Black
Fig. 4. Effect of properties of auxiliary
absorbers over height on parameters of
reactor core:
a)Kz=1.50; Aa(P =0;Ap=0;
b) Kz = 2.0; Aa(P = +0.35%; Ap = +0.014;
c) Kz = 1.80; Aa (P = +0.25t; AP = +0.011.
ment but the largest change in the steam content and water density occurs in a limited region, roughly in the
middle of the channel. This makes it possible to choose the distribution of the absorbing properties of the
auxiliary absorbers over the height with a view to providing the most effective action on the steam-void coeffi-
cient. It must be borne in mind in this case, however, that even a slight variation in the properties of the aux-
iliary absorbers over the height affects the energy distribution over the height and this in turn affects the
steam distribution over the length of the fuel assembly. The effect of the auxiliary absorbers with various
properties on the steam-void coefficient of reactivity and the shape of the energy distribution over the height
is illustrated in Fig. 4, which shows changes in the shape of the energy distribution over the height, its coeffi-
cient of nonuniformity Kz, the steam-void coefficient Aar, and the efficiency Ap of the auxiliary absorbers.
The last two parameters, are given in relation to the composition of the auxiliary absorbers, are shown in
.Fig. 4a.
Conclusions. In RBMK reactors there are many possibilities of acting on the coefficients of reactivity,
primarily on the steam-void coefficient a(P. Some of these methods can be implemented only by building new
reactors and are irreversible (e.g., changing the lattice pitch of the fuel assemblies). Other ways of great in-
terest make it possible to operatively act on the coefficients of reactivity even in existing reactors. These in-
clude such strong, but economically acceptable, measures as keeping some auxiliary absorbers in the reactor
core or increasing the operational reactivity margin as well as economically effective measures involving an
increase in the density of the fuel and in the initial enrichment. An extremely great effect on the steam-void
coefficient of reactivity is displayed by such operating modes as maintaining the mean water density in the re-
actor and the energy distribution over the height at a required level.
LITERATURE CITED
1. A. P. Aleksandrov and N. A. Dollezhal', At. Energ., 43, No. 5, 337 (1977).
2. A. D. Zhirnov et al., At. Energ., 34, No. 6, 479 (1973).
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 Declassified and Approved For Release 2013/02/12 : CIA-RDP10-02196R000800010006-0
CALCULATION OF COOLANT FLOW RATE BY RADIATION
METHODS AND POWER IN FIRST UNIT OF ARMENIAN
ATOMIC POWER PLANT
L. N. Bogachek, A. L. Egorov, V. V. Lysenko,
A. I. Musorin, M. M. Parsadanyan, O. P. Prudnikova,
A. I. Rymarenko, A. G. Tevanyan, V. L. Timchenko,
S. G. Tsypin, and V. A. Shmondin
The coolant flow rate of the primary circuit is one of the parameters which are necessary for determin-
ing the thermal power of a unit of an atomic power plant [1]. Exact and operational measurements of the flow
rate make it possible to establish-constant monitoring of the thermal operating conditions of a reactor and to
ensure optimal operating conditions for the atomic. power plant. Conversely, the lack of applicable instruments
for measuring the flow rate of the coolant, e.g., in the San Onorf Atomic Power Plant [2], was one cause of the
long delay in bringing the plant up to nominal operating conditions.
In operating the units. of an atomic power plant the thermal power must be measured with sufficient ac-
curacy [31. Unfortunately, atomic power plants hitherto did'not have a simple device for measuring this power
operatively and accurately (with an error of 1-2%). One of the most promising methods for use in atomic power
plants is the radiation method of measuring the coolant flow rate in the primary circuit [4] and the thermal
power [5]. The possibility of using these methods was first shown in [6] on the basis of studies carried out
on the first and third units of the Novovoronezh Atomic Power Plant.
The present paper discusses the results of measurements of the coolant flow rate in all six of the main
circulating loops (ML) and the thermal power of the first unit of the Armenian Power Plant. The coolant flow
rate was measured by a method based on the registration of the decay of 16N activity with coolant circulating in
the ML; the method used to measure the thermal power was based on registration of the neutron flux density
under the reactor vessel. For practical implementation of these methods we developed and set up in the first
unit of the power plant a flowmeter in, the form of a system of 12 experimental channels above the pipes of the
primary circuit with sensors installed in them and a power meter was set up under the reactor.(Fig. 1).
Measuring the Flow Rate of Coolant in the Primary Circuit. High-stability SI-3BG Geiger-Muller coun-
ters were used as sensors. The sensors were installed above the primary-circuit pipes in special collimators
to isolate a definite, strictly specified volume of coolant.
The first sensor records the intensity of the 'IN y rays of the primary-circuit coolant in the segment of
the ML at the outlet from the reactor (first detection point DP-1); the second sensor, after the coolant has
passed through the steam generator at the inlet of the. ML to the reactor (second detection point DP-2).
The coolant flow rate G in each ML was found from the ratio of the counting rates Ni and N2 of the sensors
installed on it according to the refined formula l 02GO (p2)
G=2XVI[ln `v -,Z --ln + N21 2in PjGo(PI) J
where A is the 16N decay constant (0.0971 see-1); V, effective volume of coolant in the ML between the pipe cross
sections passing through the detection points (13.6 m); N121 counting rate of the second sensor at the first point
DP-1; N21, counting rate of the first sensor at the second point DP-2;* pi, coolant density at the i-th detection
*Introducing the term In N12/N21 in the denominator of the function given above takes account of the differences
in the efficiency of -y-ray detection by the sensors.
Translated .from Atomnaya Energiya, Vol. 46, No. 6, pp. 390-393, June, 1979. Original article submitted
September 11, 1978.
0038-531X/79/4606-0445$07.50 ?1979 Plenum Publishing Corporation 445
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 Declassified and Approved For Release 2013/02/12 : CIA-RDP10-02196R000800010006-0
PGO(A
^
260 280
Fig. 2
Fig. 1. Positions of flowmeter sensors and power meter: 1) flowmeter sen-
sors; 2) to steam generator; 3). main circulating loop; 4) power meter.
Fig. 2. Attenuation factor pG0(p) vs coolant temperature.
point; and Ga(p), attenuation factor of the y-ray source as a function of the density of the material of the
source* (coolant).
The factor G0(p) was calculated from the NTs program (71 making it possible to calculate the y-ray
field from a set of thick-walled cylindrical layers in an arbitrary geometry. In this case the relative values
of G0(p) are found with a mathematical error of no worse than 0.1%. The temperature dependence of G0(p) is
shown in Fig. 2 for the range from 230 to 300?C; it follows from Fig. 2 that for the nominal power of a unit in
an atomic power plant with a VV-9R-440 reactor (t1 = 297"C for DP-1 and t2 = 268'C for DP-2) the ratio
piG(p2)/p1G9(p1) is 1.04. Failure to take 21n (p2G0p2/p1Gop1) into accountin the refined formula results in the
flow rate being overestimated by roughly 7%.
Let us note that in deriving the formula given above we assumed that both sensors were set up in an iden-
tical geometry with respect to the ML pipe and that they record y rays from equal "visible" volumes of coolant.
-Measures were undertaken to ensure high accuracy in setting up the sensors. Nevertheless, because of some .
differences in the geometry of the positioning of the sensors and collimators relative to the Du-500 pipe, a cer-
tain error was observed in the measurement of the coolant flow rate in ML. Analysis of the factors which are
the sources of this error showed that they include differences in the positioning of the two sensors relative to
the pipe- and the sensitive volumes of the sensors relative to the collimators, different background contributions
by y rays from the ends of the pipes passing through the shields of the collimators, etc. Corrections were
calculated to take account of the y ray background. Finally, the total error in the measured values of the cool-
ant flow rate in the ML of the first unit of the Armenian Atomic Power Plant and was found to be 6-7%.
A series of experiments were carried out using the flow meter to measure the steady-state flow rates in
the ML of the first unit of the Armenian Power Plant at reactor power levels at 20-50% of the nominal value.
At the same time we determined the coolant flow rate G(Ap)1 from the head characteristics of the main circu-
lating-pumps (see Table 1). The results of the experiments are compared with data obtained from the head
characteristics of the main circulating pumps (MCP).
From the data in Table 1 it follows that the maximum deviation of the flow rate Gi from the mean value
G = 1 E for each mode is t7%. This does not exceed the calculated error in the flow rate. The coolant flow
n 1
rate through the reactor core for various modes remains constant to within 2% and differs from the total flow
rate obtained from the head characteristics by no more than 2%.
To obtain additional information about the characteristics of the flow meter we performed a special ex-
=periment. In one of the ML (ML No. 2) we gradually reduced the coolant'flow rate by partially closing the main
c.*Source is taken to mean the volume of coolant in the primary-circuit pipe from which y rays can enter the
tsensor without interacting with the material of the collimator.
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TABLE 1. Coolant Flow Rates in Primary
Circuit in Five Modes, m3/h
Mode
L?o?P
G (OP)i
Gi
c
cx6
6
G (Ap)i
1
1
7600
8280
7860
47140
46750
2
7850
-
3
7590
7930
4
7860
7360
5
7870
6
7980
2
1
7570
8330
7700
46200
46720
2
7900
-
3
7600
7860.
4
7870
7200
5
7870
7410
6
7910
-
3
7660
8370
7670.
46030
46750
2
7940
-
7580
7890
4
7850
7100
5
7930
7330
6
7790
4
7590
8440
7860
47140
46530
2
7970
7900
3
7620
-
4
7630
7310
5
7900
7800
6
7820
-
5
7620
8470
7700
46230
46480
2
7960
7660
3
7580
-
4
7630
.7270
5
7870
7410
6
7820
-
shut-off valve. The flow meter was used to measure the coolant flow rate Gi in all the loops of the five posi-
tions of the shut-off valve. Each position was set for 10 min, the time necessary for establishing steady-state
operating conditions and carrying out measurements (Fig. 3). It follows from Fig. 3 that at low flow rates the
characteristics from the specifications, which were determined for the main circulating pumps at a pressure
drop Ap on a test bed, differ from the real characteristics determined with the VVER-440 in the range under
7000 m3/h.
It was possible with the flow meter to determine the flow rate (reverse flow rate) of the coolant in the ML
arising when the MCP were cut off. For this purpose one of the MCP was disconnected, upon which the valve
was opened temporarily and.the time dependence of the indications N(t) of both sensors positioned on the given
ML. The interval of time for which the values of the function N(t) for both sensors are constant is character-
ized by the constancy of the flow rate in the ML. The reverse flow rate, found by this method is 3600 m3/h,
which is 44% of the forward flow rate through this loop. There are other possible ways of using the flowmeter,
especially in studying unsteady flow rates.
Measuring the Thermal Power of a Nuclear Reactor. The sensor of the power meter was placed beneath
the reactor at some distance from sources of perturbation to the neutron flux emerging from the reactor. This
sensor detects neutrons which emerge from the reactor vessel in the region of the .core and, reflecting from
the. concrete. shield, travel through an annular gap down through a labyrinth to the sensor of the power meter.
The neutrons are recorded by the power meter .by. conversion to y rays in a cylindrical polyethylene tar-
get with a diameter of 200 mm in which Geiger-Muller y ray counters are arranged. External y rays are cut
off from the target and the counters by a lead screen 200 mm thick. If necessary, the y ray counters are re-
placed by means of a special channel which is led out into a semiservice room. The counter indications are
recorded with analog and digital indicators located on the control board of the particular unit of the power plant.
For absolute calibration of the relative indications of the power meter in units of thermal power we car-
ried out joint measurements of the thermal power WT of the reactor (in the range from 40 to 85% nominal
power) by the method of making a thermal balance, and of the power-meter indications Npm. A linear depend-
ence of WT on Npm was obtained from these data by the least-squares method. In this. case the correlation
coefficient is rN W = 0.9997. This value makes it possible to obtain (with calibration for n > 4 joint meas-
urements at varPous levels of thermal power) a confidence interval of ?0.02WT for the values of the thermal
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Ss
x
4000 5000 6000 7000 0, m?Jh
Fig. 3. Head characteristics of MCP-2 of first unit of
Armenian Atomic Power Plant: 0) flow meter data;
specifications.
power of the reactor in the range of thermal power variations from 20 to 100% (with a significance level of 95%)
with the power meter. Thus, the above method of correlation measurements of the thermal power of a reactor
of the water-moderated-water-cooled (VVER) type makes it possible to significantly refine the thermal power
of the reactor of the VVER-440 unit.
The authors wish to thank V. A. Sidorenko for his interest in the work and for his valuable comments.
LITERATURE CITED
1. V. A. Sidorenko, Problems. of Operating Safety of VVER Reactors in Russian), Atomizdat, Moscow
(1977).
2. J. Ortega, C.. Johnson, and K. Baskin, Nucl. Safety, 11, No. 2, 142 (1970).
3. F. Ya. Ovchinnikov et al., Operating Modes of Water-Moderated Water-Cooled Power Reactors [in Rus-
sian), Atomizdat, Moscow (1977).
4. D. Howard, U.S. Patent No. 2. 841.73. (1958).
5. S. G. Tsygin et al., in: Digest of Papers of All-Union. Scientific Conference on Protection from Ionizing
Radiation of Nuclear Engineering Facilities fin Russian, Izd. Moskovsk. Inzh.-Fiz. Inst. (MIFI),,Moscow
(1974), p. 103.
6. A. I. Rymarenko, A. A. Bolberov, and V. V. Lysenko, in: Proceedings of the All-Union Institute of Heat
Engineering, No. 2, Energiya, Moscow (1974), p. 31.
7. Ya. A. Bychkov et al. in: Radiation Safety and Protection of Atomic Power Plants [in Russian], No. 3,
Atomizdat, Moscow (1977), p. 141.
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 Declassified and Approved For Release 2013/02/12 : CIA-RDP10-02196R000800010006-0
PHYSICAL START-UP OF IBR-2 PULSED RESEARCH
REACTOR
V. D. Anan'ev, V. A. Arkhipov, A. I. Babaev, UDC 621.039.516.621.039.55-
D. I. Blokhintsev,* Yu. M. Bulkin, B. N. Bunin,
E. D. Vorob'ev, N. -A. Dollezhal', L. V. Edunov,
V. S. Lavrukhin, V. L. Lomidze, V. V.. Melikhov,
Yu. I. Mityaev, Yu. N. Pepelyshev, V. P. Plastinin,
A. D. Rogov, V. S. Smirnov, I. M-. Frank,
N. A. Khryastov, E. P. Shabalin, and Yu. S. Yazvitskii.
The physical start-up of the IBR-2 periodically functioning pulsed reactor took place in Dubna At the end
of 1977 and the beginning of 1978; the reactor was designed for research on nuclear physics and the physics of
condensed media in extracted beams of slow neutrons at a mean power of .4 MW and a maximum thermal-neu-
tronflux density of - 1016 neutrons /cm2-sec [1, 21. The investigations were carried out without coolant at an
average reactor power of up to 500 W in both steady-state and pulsed modes of operation. The IBR-2 was
brought up to the critical state for the first time on Nov. 30, 1977, and pulse criticality was achieved on Jan.
13, 1978. The present paper gives a review of the principal experiments performed during the physical start-
up of the IBR-2.
Critical Assembly. Before the fuel assemblies were loaded all the cells of the reactor core were filled
with fuel-assembly imitators which differed from the fuel assemblies only in that the fuel elements in them
contained copper instead of plutonium dioxide. A Po-Be source with an intensity of 107 neutrons/sec was set
up in the center of the reactor core. The neutron flux was monitored with the regular start-up instrumentation
with three 235U fission chambers set up outside the reactor at a distance of 1150 mm from its center (Fig. 1, 1)
as well as two auxiliary experimental channels with replaceable detectors set up both in the Plexiglas block of
the cold-moderator imitator (see Fig. 1, 11) and in the reactor core 8. The counting rate of the regular and
experimental detectors with the reactor core loaded with imitators was 5-50 counts/sec. Moreover, for linear
monitoring of the power we used an experimental current channel with a boron chamber in the CMI, analog
reactimeter, and recording potentiometer.
Fuel was charged into the reactor core by successively replacing the imitators with fuel assemblies
which were loaded in order of increasing calculated efficiency. The extrapolation of the curve of inverse
counting therefore always showed a critical-mass value that was lower than the actual value; this provided
additional safety in the start-up operations. The critical state was attained for two variants of reactor charge
(Fig. 2): "central" (70 fuel assemblies) and "peripheral" (74 fuel assemblies). The calculated critical charge
of these variants was 71 ? 2 and 75 f 2 fuel assemblies, respectively.
Effect of Reactivity of Control and Safety Elements and Other Reactor Elements. The reactivity effects
were measured by both the method of reverse multiplication of the counting rate in the start-up and experi-
mental channels in the subcritical state (with a multiplication factor of 50-2000) and dynamic methods while
varying the reactor power in the range up to 100 W. In the measurements of the reactivity according to the
multiplication we took account of the different sensitivity of the detector to the neutrons from the source and
the fission neutrons, the different neutron importance of the neutrons from the source and from the fission, the
decay of the Po-Be source, and the contribution from spontaneous fission of 240Pu. The last factor was found
experimentally.
The reactivity was measured by dynamic methods with the analog reactimeter as well as by recording
the signals from the chamber with a recording potentiometer or a loop oscillograph and subsequent processing
of the data by computer. In some cases the reactivity was found from the established reactor riding-up time.
The differential efficiency of the control and safety elements was measured by the overcompensation method.
Translated from Atomnaya Energiya, Vol. 46, No. 6, pp. 393-400, June, 1979. Original article submitted
December 4, 1978.
0038-531X/79/4606-0449$07.50 ?1979 Plenum Publishing Corporation 449
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 Declassified and Approved For Release 2013/02/12 : CIA-RDP10-02196R000800010006-0
0'1
Fig. 1. Cross section of IBR-2: 1) regular fission chambers; 2) water moderator behind moving
reflector (MR); 3) beryllium insert of auxiliary moving reflector (AMR); 4) blade of main moving
reflector (MMR); 5) rod of automatic controller (AC); 6) scram protection (SP-2); 7) water mod-
erator behind reactivity compensator (RC-2); 8) reactor core; 9) reactivity compensator (RC-2);
10) slow emergency protection (SEP-2); 11) cold-moderator imitator (CMI); 12) experimental
chambers; 13) moderator of regular chambers; 14) SEP-1; 15) RC-1; 16) experimental cham-
ber in target channel; 17) SP-1; 18) intermediate controller (IC); 19) water moderator behind
RC-1.
Fig. 2. Recorder chart of critical charge of reactor with cells containing fuel assemblies for:
1) "central" and 2) "peripheral" charges.
Analysis of the data from measurement of the efficiency of one and the same segments of the reactivity
compensators (RC) by different methods reveals that the relative deviation of the results does not exceed 5%,
i.e., is smaller than the error of the dynamic methods (?5%) and the error of.the reverse multiplication method
(f15%). It must be noted that the result of measurement of the detector position. Thus, the total efficiency of
the SEP (see Fig. 1, 14) measured with a chamber in the CMI proved to be 1.5 times the efficiency measured
by a chamber in the. center of the reactor core. The efficiency of the RC and the moving reflector (MR) (see
Fig. 1, 9, 15, 4) measured with the regular chambers situated behind these elements was below the efficiency
measured by other chambers. This was caused by local perturbations in the neutron flux and spectrum.
The most complete measurements were made with a peripheral charge of fuel assemblies (Tables 1 and
2). With a central charge of fuel assemblies the efficiency of the RC and SEP units is almost 10% higher and
the effect of the MR reactivity is also about as much higher than with a peripheral charge (Table 3). This. is
explained by the fact that with a central charge the reactor core is somewhat displaced towards the moving re-
flector and is further separated from the compensators and the SEP.
The measured total efficiency of the control and safety elements is lower than the design values by a fac-
tor of 1.5-1; this is explained by the fact that the calculations did not take account of some design details.
Notwithstanding this, the efficiency of the control elements can be considered satisfactory whereas the effi-
ciency of the SEP is inadequate since, should one element of the SEP fail, it would not ensure compensation of
the temperature effect of the reactivity. Before the power start-up of the reactor the efficiency of the control
and safety elements will be increased by changes in their design.
Figure 3 shows how the power of the reactor is affected by the tripping of one element of the SP. In 0.02
sec after a scram signal one element of the SP introduces a reactivity of 8.10 4 keff, although its total efficiency
is much higher (see Table 1). In the pulsed mode of operation the tripping of one SP element results in a de-
crease in the pulse energy by a factor of roughly 50.
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TABLE 1. Efficiency of Control and Safety
Elements
Elements
Total, 10'2 k
eff
Differential,
4
10" , keff/mm
RC-1
1,50+0,12
.0,50+0,03
RC-2
1, 55?0,12.'.
0,52?0, 03
IC
0,26T-0 ,02
0,11+0,01
AC
0,032?0,002
0,011?0,001
SP'
0,14?0,01
-
SEP'
0,60?0,06
TABLE 2. Effect of Reactivity of Main and
Auxiliary Moving Reflectors with
Peripheral Charge of Fuel Assemblies*
Position of
a)
r
Error,
-
MR
z
x
?;
23 v
x- 6
6
a
I
et
4
4 ,-i
ZS v
b
b
AMR.at phys, c.t
1,76
1,01
-
-
3
1
AMRwithdrawn
t
2,09
1,6
-
-
2
1
MMR at phys. c
-
-
0,39
0,1
2
20
MMR withdrawn
-
-
0,72
-
2
-
*See Eq.(1) and the relevant text in
.reference to the coefficient a.
tThat is, in the physical center which is
taken to mean the position of the AMR or
MMR corresponding to their maximum
efficiency.
TABLE 3. Effect of Reactivity of Main and
Auxiliary Moving Reflectors of IBR-2 with
Central Charge of Fuel Assemblies
Position of
MR
Error.
zi
?
6
a
I e
d ,-1
~ p
b
b
AMR at phys.c.
2,04
1,4
2
10
AMR withdrawn
2141
-
-
-
I
-
MR at phys.c.
-
-
0,45
-
2
-
IM MR withdrawn
-
-
0,82
0,4
2
12
Fig. 3. Oscillogram of IBR-2 power in
operation in steady-state mode and with
one SP element dropped.
Experimental assessment of the effect of external hydrogeneous moderators, which are used to form the
spectrum in neutron beams, on the efficiency of the control and safety elements showed that the greatest effect
is exerted by the CMI (see Fig. 1). When it is moved further away, the efficiency of the SEP increases by 20%
and the. efficiency of the RC decreases by the same proportion. Other moderators (7 and 9 in Fig. 1) have prac-
tically no influence on the efficiency of the control and safety elements.
The MMR and AMR (Figs. 1 and 4) are used for periodic modulation of the reactivity and production of
power pulses at a frequency. of 50 and 5 Hz. Accordingly, particular attention is paid to measuring the effects
of their reactivity. In measuring the angular dependence of the reactivity of the IBR-2 on the position of the
moving reflector we rotated the reflector at a speed of 0.03-0.4 deg/sec and monitored the position with an
error of no more than 0.07? for the MMR and 1? for the AMR.
Besides the dependence of the reactivity effects of the MR on the charge of the reactor core, the effect of
the position of one of the reflectors on the efficiency of the other (see Tables 2 and 3) proves to be very sub-
stantial. The reason is that the AMR shades the MMR. The shadow effect has a particularly strong influence
on the parabola coefficient which describes the behavior of the reactivity when the MMR is moved through
small angles (f 3-4?) relative to the physical center:
e (y) = sP - atpz, (1)
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Fig. 4. Main (a). and auxiliary (b) moving reflectors of IBR-2.
where Ep is the maximum prompt reactivity.
The value of a turned out to be smaller by a factor of 3 or 4 than the value found with allowance for the
effect of the AMR. Withdrawal of the beryllium insert (especially the AMR) from the. reactor, even through a
small angle (-5?) results in a substantial increase in a. Indeed, in special experiments with an MMR we ob-
tained a = 2.5-10-4 deg -2 and the total efficiency proved to be practically equal to the calculated value of
2..65.10-2 keff (Fig. 5).
Of the other measured effects mention should be made of the change in reactivity during axial displace-
ment of the MMR, (dk/dx)MMR; this determines the effect of the vibrations. of the MMR rotor on the power
fluctuations:
(dkldx)MMIf= (4.5:j-_ 0.3) ? 10-4keff/ mm
The measured differential efficiency when the entire MR machine is removed from the reactor core is
(dkldx)Mp = (9.0 ? 0.3). 10-4kff/mm
and the total effect is -0.06keff? The reactivity effect of the MMR counterweight proved to be .substantial, i.e.,
0.006keff (see Fig. 4).
In measurements of the reactivity effects of the fuel assemblies and their imitators (Table 4) the effi-
ciency of the imitators on the periphery of the reactor core proved to be unexpectedly high. Experiments on
measurement of the efficiency of metal specimens in the region of the MR confirmed the assumption concern-
ing the comparatively. high efficiency of copper in the spectrum of the IBR-2. The efficiency of specimens of
steel, copper, tungsten, and beryllium with the same volume are in a ratio of 1.0:1.4:1.5:1.8, respectively.
Lifetime of Neutrons in Reactor. The mean lifetime T of a generation of prompt neutrons in many ways
determines the length of the power pulse at half maximum which can be calculated for the IBR-2 from the,
formula
0 1.4 (i/aV2)ii3, (2)
where v is the speed of rotation of the MMR. The mean lifetime was measured by four independent methods.
The least error was yielded by the Rossi a method [ 31: 83 ? 2 nsec. The value of a measured from the shape
of the power pulse turned out to be 90 ? 9 nsec whereas the values found from the power fluctuations in the
pulsed [4] and steady-state [3] modes of operation were 80 ? 10 and 130 t 10 nsec, respectively. Since the
last result had a considerable systematic error due to the frequency characteristic of the measuring channel,
it was not taken into account in our calculations of the mean value T. The mean value T found from
T tiici 2/ (j72,
i i
where Ti and ci are the mean value and the variance of T in the ith method coincided with the value obtained by
the Rossi a method and amounted to 83 ? 2 nsec.
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 Declassified and Approved For Release 2013/02/12 CIA-RDP10-02196R000800010006-0
-30 -20 -10 m-101 10 20 30 W
Fig. 5. Dependence of reactivity of IBR-2
on angle of rotation of MMR: 1) with regu-
lar AMR;'.2) without AMR; and 3) in com-
parison with measurements on BF-S stand
t31~
TABLE 4. Efficiency of Fuel Assemblies
and Fuel-Assembly Imitators of IBR-2
Reactor, 10.2keff
U
Fuel As-
sembly
Fuel-as-
sembly
imitator
o
r.
..
U
Fuel As-
sembly
Fuel-as-
sembly
imitador
1
1,16+0,04
0,13?0,02
58
0,96?0,03
0,31?0,04
4
1,21?0,04
0,10?0,02
59
1,02?0,03
0,30?0,03
5.
1,20+0,04
0,10?0, 02
61
1,07+0,03
0,30?0,03
8
1,16?0,03
-
63
0,92?0,03
0,27?0,03
14
1,07?0,02
.0,09?0,02
74
0,78+0,04
0,26?0,03
20
1,28?0,05
0,10?0,0
75
0,65?0,03
0,20?0,03
31
0,99?0,02
0,28?0,0
77
0,86?0,03
0,27?0,03
35
1,18?0,05
0,15?-0,0
The great difference between the measured T and the value calculated by the .Monte Carlo method
(43 nsec) [1) is attributed primarily to the effect of the AMR disk, the CMI, and the MR vessel, which was not
taken into account in the calculations. Measurements by the Rossi a method, carried out without these ele-
ments, yielded T = 47 nsec.
Neutron Spectrum and Flux in External Beams and Power Distribution in Reactor. The spectrum of fast
and resonance neutrons was measured from, the activation of threshold detectors with subsequent reconstruc-
tion of the initial energy distribution of the neutrons by computer [5). In the intermediate portion of the spec-
trum satisfactory agreement was obtained with calculated data (Fig. 6).
The density of the thermal-neutron flux at the external surface of the moderator was found by the cad-
mium difference method from the absolute activity of copper and gold tracers [5). The absolute activity of the
tracers was measured by calibrated sensors and refined on a 0-ycoincidence arrangement. The measured
density of the thermal-neutron flux, (3.0 ? 0.2)?106 neutrons/(cm2 ? sec ? W), proved to be roughly double the calcu-
lated value.
The power distribution in the reactor core was found from the y activity of the fission products. The ac-
tivity of the fuel assemblies was measured over a period of two months after irradiation. The measured dis-
tribution in the main is in agreement with the design data; the coefficients of nonuniformity of the power dis-
tribution over the height and volume of the reactor core were 1.3 and. 1.6, respectively. The largest local
deviations of the measured and calculated distributions are in the region of the reactor core adjacent to the
MR as well as in the region in which slow neutrons from the moderator enter the core.
Pulse Criticality of the Reactor. In comparison with the'attainment of delayed criticality under steady-
state operating conditions, the attainment of criticality under pulsed operating conditions is characterized by
distinctive features. The ratio of the mean power of the subcritical pulsed reactor to the power of the external
source is given by [4)
y = W/So = (1 /n) (Kp /1- Kp),
(4)
where Kp is the so-called pulsed neutron-multiplication factor which depends in an involved manner on the
maximum prompt reactivity Ep [see Eq. (1)1; Kp = 1 when Ep = Epo (here Epo is the equilibrium pulsed critical-
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 Declassified and Approved For Release 2013/02/12 : CIA-RDP10-02196R000800010006-0
y(E;
1012
Fig. 6. Neutron spectrum (neutrons/cm2?sec?MeV?W)
in external beam of IBR-2: 1), at surface of modera-
tor; and 2) at distance of 8 in from it; A, 0. O, ? ) ex-
periment; ) calculation.
ity), and Ki av2T2) it is
nT 1
I/op 1-Kp lfpl ,. 111F I-
It follows from Eqs. (4) and (5) that WO depends more on Ep than W does {Kp(Ep) is a very weak function
of 6p for Ep < 01. Moreover, for pulsed operating conditions are characterized by the following interesting
features: The reactor state extrapolated from the region of deep subcriticality is one of prompt criticality:
Ep= 0 [see Eq. (6)].
The reactor is brought up to the critical state (Ep = Epo - 10-3keff) in the following manner. First, with
the control and safety elements withdrawn, the reactivity modulator was started up. When the MR had reached
the nominal number of revolutions, the RC were introduced gradually. After, each rise in reactivity the count-
ing rate in a pulse in 500 sec, i.e.,, the multiplication yo, was measured and the position of the RC correspond-
ing to the expected criticality was estimated. With a pulse-repetition frequency of 5 Hz, when both MR are
rotating, and with a small multiplication yo we made a rough synchronization of the rotors by finding an AMR
rotation phase such that corresponded to the maximum counting rate. The final synchronization was carried
out in the critical state.
Good agreement was obtained between the measured and calculated values of the function yl(Ep) (Fig. 7).
The calculated relations were obtained on the basis of Eq. (5) by using the measured values of a. A pulsed
mode was achieved only with peripheral charging of fuel assemblies. At a pulse-repetition frequency of 50 Hz
the equilibrium pulsed supercriticality was 0.9.10-3keff whereas at a frequency of 5 Hz it was 1.3.10-3keff
(see Fig. 7).
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 Declassified and Approved For Release 2013/02/12 CIA-RDP10-02196R000800010006-0
tP,10-3 0 200 400 600 t
Fig. 8
Fig. 7. Dependence of inverse multiplication in pulsed mode on prompt reactiv-
ity: ) calculation at frequencies of (1) 5 Hz and (2) 50 Hz; o, o) ex-.
per iment.
Fig. 8. Measured (0) and calculated ( ) shape of IBR-2 power pulse at fre-
quency of 50 Hz.
Shape of Power Pulse. The shape of the power pulse from the IBR-2 is one Of its principal characteris_
tics which determines, along with the intensity of the neutron flux, the efficiency of the reactor. as a source of
neutrons for time-of-flight neutron spectroscopy.
The shape of the spike of fast neutrons was measured in detail in different states of the reactor by the
counting method with a multichannel time analyzer and by the current method by recording the detector signal
from an individual pulse of power on the screen of a storage oscillograph. The fast-neutron detectors used
were 238U fission chambers in the reactor core and thorium fission chambers in the extracted neutron beam as
well as a plastic scintillator with a photomultiplier which were also set up in the beam.
In the process of bringing the reactor up to criticality the shape of the pulse was measured by the count-
ing method. At a subcriticality of 1.10-2 and 2.10-2keff the pulse width at half maximum 0 was equal to 720 and
824 ?sec, respectively. In the case of prompt criticality 0 = 240 sec at maximum reactivity, which is in good
agreement with the calculated value of 244 sec obtained by using the experimental values of a and T.
In the state of pulsed criticality the measurements were carried out mainly by the current method which
in the given case is more exact than the counting method (error no greater than 2%). At a pulse frequency of
50 Hz (AMR immobile) the value of 0 for fast neutrons was 220 psec and at a frequency of 5 Hz it was 198 ?sec.
The fact that the measured value of 8 considerably exceeds the design value (92 ?sec for 5 Hz) is due to the
difference between the actual values of a and T and the calculated values. Numerical solution of the one-point
kinetic equation for the reactor with experimental a and T gives a pulse shape which practically coincides with
the measured shape (Fig._ 8).
Clearly, the pulse is lengthened by those elements of the motor construction which have a considerable
effect on a and T. It was established that the greatest effect is exerted by the AMR. The measured dependence
of 0 on the position of the AMR is plotted in Fig. 9. The AMR position at the physical center, i.e., correspond-
ing to the highest efficiency, was taken to be zero. Withdrawal of the AMR from the core results in a marked
shortening of the pulse length 0, from 190 to 150 gsec. Special experiments established that this occurs mainly
because of an increase in aMMR and partially because of a shortening of T. The pulse length is also reduced
considerably when the external moderators are withdrawn. The moderator behind the AMR increases by
30 f 4 ?sec and the moderator behind the blocks of the SEP, by 8 ?.4 ?sec.
Power Fluctuations of IBR-2 in the Pulsed Mode. Because of the high sensitivity of the pulsed reactor to
reactivity variations during physical start-up a detailed study was made of the character of the power fluctua-
tions as well as their correlation with other random processes affecting the reactivity (MR vibrations, etc.).
Data acquisition took place by recording discrete signals from the sensors on magnetic tape. The recorded
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 Declassified and Approved For Release 2013/02/12 : CIA-RDP10-02196R000800010006-0
d2
qz
6
i I i I I tP I , I 7 I I I I 11
-20 -10 0 10 .20 30 W 0 41 42 43 0,4 45 0,5 47. 48. 0,9 W-'
Fig. 9 Fig. 10
Fig. 9. Dependence of length of 0 of IBR-2 power pulse on AMR position ~p at frequency of
50 Hz: o ) no moderator behind AMR; ') no moderators at all.
Fig. 10. Dependence of relative variance oqz on inverse mean reactor power W at 50 Hz:
A) presence and o) absence of water moderators; at 5 Hz in the absence of water modera-
tors ^ ); ) approximation according to Eq. (6).
information was processed by computer with. the aid of a special program for the analysis of steady-state ran-
dom processes in a periodically operating pulsed reactor. With.a low mean reactor power (under 10 W) stoch-
astic fluctuations of the .pulse energy predominate (4). The experimental dependence of the relative variance
Oqz of the pulse energy on W is, in accordance with the theory, linear in character (Fig. 10) and for W > 0.5 W
is approximated by the function
AV \ W ! = wQ (0) + 750 .50 ,10-*,
where A2 z(0) is extrapolated value of the variance, characterizing the fluctuations, which is due to the devia-
tions of the reactivity owing to vibrations of the elements of the reactor construction. At a frequency of 50 Hz
the variance A2gz(0) (3 f 2)-10-4 and at 5 Hz it is (5 f 2)-10-4. The stochastic fluctuations reach a maximum of
in the region of prompt criticality for W 0.02 W; (Ogz)m = 0.30. All the experiments on studying O2 qz(0)
were performed at a power of 300 W, which made it possible not to take account of the stochastic fluctuations
of the pulse energy. It turned out that the dominating contribution to Agz(0) is made by the transverse dis-
placements of the MMR. It is seen from Fig. 11, which gives the results of spectral-correlation analysis, that
the interrelation between the pulse energy and the MMR displacement is due to the resonance peak (in the spec-
trum of MMR vibration frequencies) at 16 Hz, which is exactly one-third the rotational frequency of the MMR.
This is -exactly the transmission ratio of the MMR reducing gear train which apparently is the source of vibra-
tions of the rotor and, consequently, the fluctuations of the reactor power. However, the fluctuations caused in
the pulse power by these vibrations are extremely small (-2%). A detailed spectral-correlation analysis was
also carried out for other processes which affect the reactor power. As a result of these investigations it was
established that:
a) the oscillations in the rotational velocity of the MMR and the corresponding power fluctuations are
very small (the relative standard deviation does not exceed 0.2%);
b) the circulation of water in the moderators results in reactivity fluctuations not exceeding 5.10-6keff,
which corresponds to power fluctuations -2.5%. A large part of the oscillation spectrum lies in the range of
low frequencies (f < 0.5 Hz);
c) with the reactor operating with a pulse frequency of 5 Hz, when the AMR also rotates, additional power
fluctuations appear because of torsional oscillations of the AMR disk which are almost periodic. The spread
they cause in the power-pulse amplitudes is no greater than 0.4%. The character of the MMR oscillations in
5-Hz modes differs substantially from that in 50-Hz modes.
Neutron-physical investigations with a "dry" (no coolant) start-up confirmed the possibility of realizing
the principal design characteristics of the IBR-2 as a source of neutrons for physics research. The difference
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 Declassified and Approved For Release 2013/02/12 : CIA-RDP10-02196R000800010006-0
20
16,16
116,53
Fig. 11. Spectral densities Gq of pulse energy
and transverse displacements Gz of MMR and
intercorrelation function rqz of processes q
and z at frequency of: 1) 48.5 and 2) 49.6 Hz;
t is the time between pulses.
0,2 0,3 0; 45 t, sec
of some physical parameters is due in the main to the fact that the calculations did not make allowance for
auxiliary elements of the reactor construction. Measures for further improvements in the reactor character-
istics have been planned on the basis of the results of the physical start-up and are being implemented.
In conclusion, the authors express their profound gratitude to all the teams and individual workers who
participated in the preparation and execution of the physical start-up of the reactor.
1. V. D. Anan'ev et al., Prib. Tekh. Eksp., No. 5, 17 (1977).
2. V. D. Anan'ev et al., At. Energ., 31, No. 4, 352 (1971).
3. R. E. Uhrig, Statistical Methods in Nuclear Reactor Physics [Russian translation, Atomizdat, Moscow
(1974).
4. E. P. Shabalin, Fast-Neutron Pulsed Reactors in Russian), Atomizdat, Moscow (1976).
5. Metrology of Neutron Measurements in Nuclear-Physical Installations. Proceedings of the First All-
Union School in Russian], Vols. 1 and 2, Moscow (1976).
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 Declassified and Approved For Release 2013/02/12 : CIA-RDP10-02196R000800010006-0
E, Yu. Vasil'eva and A. N. Maiorov UDC 621.039.546:621.039.548
The technological process of fabricating rod-type fuel elements includes a number of operations of qual-
ity control and inspection of a number of parameters of the internal structure of the product. With traditional
industrial radiography the two-dimensional image of the macrostructure of a product makes it difficult to in-
terpret the results. Because of the complexity of the technique and the apparatus involved, stereoradiography
has not hitherto found widespread application [1]. New possibilities are opened up by tomography, which pro-
duces sectional images of the internal macrostructure of the product. The method of transverse transmission
computer tomography developed in 1971-1973 has been extensively used in medicine [2-5]. In this method the
object is irradiated by a beam of radiation at various angles and then recording the information obtained and
processing it by computer [3, 61. This method makes it possible to obtain fundamentally new information about
the internal structure of the object, both quantitative information and information shown on a display [ 3-5] .
Radiographic data obtained by repeated x-raying of a rod-type fuel element with the element rotating in
coordination with the motion of the film may be a basis for reconstructing an image of the internal structure.
The x-ray photograph made with the fuel element turned through various angles to its longitudinal axis consists
of a set of projections of the distribution function of the density of material inside the object on the plane of the
film, provided that the beam of radiation can be assumed to be. parallel. This condition will hold for an x-raying
arrangement with F ? h, where F is the focal distance and h is the height of the object in the x-raying direc-
tion. Reading the information from the film -along a chosen direction AAA, for one and the same object we get
projections of the function from various angular directions a in the plane sought (Fig. 1). In this case the num-
ber of planes is determined by the scanning pitch of the reader. The information obtained can be written as
S (a, P) ='T 1 P (x, y) dl,
L(r, p)
where S(a, p) is the projection of the function (density of the film blackening); p(x, y), sought function (distribu-
tion of density of material inside object); L(a, p), a straight line along which the projection of the function is
determined; T, a factor coupling the density of the material with the blackening of the film; a, angle of projec-
tion; and p, a parameter giving the straight line L(a, p) in the normal form.
Equation (1) is a Radon. transformation of function p(x, y) with respect to the hyperplane L(a, p). Integral
geometry considers the relation between functions in space. and the integrals of these functions over all possi-
ble hyperplanes [7]. Equation (1) can also be written as -
S(a, p)=t p(x,y)S]P-(xcosa+ysina)]dxdy, (2)
where 6(x) is the Dirac delta function and pa = x cos a + y sin a is a parameter giving the straight line L(a, p)
which passes through the point with coordinates (x, y).
The problem thus is to find the function p(x, y) from the values of the function S(a, p) taken along any
straight line L(a, p). The sought function can be calculated from the values of its integrals, taken over all pos-
sible hyperplanes, from the formula for the Radon transformation [6, 71 recast as
(
P (x, y)= J da S p, 12S (a, Pa)-S (a, Pa+P)-S (a, Pa-P)] dP?
0 0
Equation (3) formed the basis of an algorithm developed for reconstructing the images of the internal structure
of a body with transmission computer tomography [8]. The inner integral was taken by the trapezoid method
and the outer integral by the rectangle method and Eq. (3) was rewritten as
Translated from Atomnaya Energiya, Vol. 46, No. 6, pp. 403-406, June, 1979. Original article submitted
May 15, 1978.
458 0038-531X/79/4606-0458$07.50 ?1979 Plenum Publishing Corporation
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 Declassified and Approved For Release 2013/02/12 : CIA-RDP10-02196R000800010006-0
d
Fig. 1. Arrangement for obtaining radiographic information for recon-
structing tomographic image: a) geometry of experiment; b) radiograph
at various angles of rotation of fuel element [1) x-ray tube; 2) fuel; 3) fuel
element; 4) x-ray film) ; c) reconstruction along AN direction for one fuel
element ( 5) projection of p (x, y) function]; d) explanation of Eq. (1).
1 f 2S (Pa) 1 (2S (Pa)-S (Pa+h)-S (Pa-h)
ti2wn l D + 2h L 1 +
J=1
.n-1 11 _
2 c 2S(Pa)-S(Pa+ih)-S(Pa-ih) + 2S(Pa)-S(Pa+nh)-S(Pa-nh)JJ
i2 n2 ) .
where D is the diameter of the region of reconstruction; m, number of intervals with respect to a from 0 to. 7r;
h, step of the partition of the integration range; and n, number of partition points on the line of projection of
the function.
A program for the M-222 computer was devised on the basis of Eq. (4). With an input array (number of
counts over all projections of the function) of 4096 numbers and with reconstruction picture dimensions of
64 x 64 cells the reconstruction time is 7-10 min with a computer speed of 20,000-25,000. operations per sec-
ond. Increasing the capacity is related to the computer speed and at I million operations per second the proc-
essing time amounts to only a few seconds.
The function is also reconstructed from its projections by other mathematical methods of processing, in-
cluding Fourier analysis, solving systems of linear equations, direct iterative methods, etc. [9-121. They have
become an object of extensive discussion, the central point being that of comparing methods as to accuracy and
speed. The lack of standard comparison techniques has hitherto not allowed an optimal method to be chosen.
The use of the Radon transformation apparently is preferable for the following reasons:
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 Declassified and Approved For Release 2013/02/12 : CIA-RDP10-02196R000800010006-0
. ? ? ?
. ?
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??????? 0 0 0 0 0 0 0 0 0 0 0 0 0 0
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? ? ? ? ? ? ? ?? ? ? ? ?
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? ? ? ? ? ? ? ? ? ? ? ? ?
? ? ? .? ? ? ? ? ? ? ?
0
? ? 0 0
0
00
0
0
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0
? ?
? ?
? ? ? ? ? ? ? ? ? . ? ? ?
? ? ? ? ? ? ? ? ?-?
0 0
? ? ? ? ?
0
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0
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0 0
? ? ?
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? ? ? ? ?
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? ?
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? ? ?
? ?
? ?
Fig. 2. Reconstructed image of cross section of rod-type fuel element with Aa = 30?,
Ap 0.055 mm (pitch of reconstruction grid); ?) fuel-element can; 0) filling of fuel
element.
a) being in a single-valued relation with the Fourier transformation, the Radon transformation does not
require a two-fold procedure in reconstruction, i.e., a transition to the space of spectra and then back
again, and consequently this algorithm is more economical in respect of computational operations;
b) the use of the method of solving systems of linear equations puts increased requirements on the com-
puter since the computational matrix comprises ^-108 numbers, whereas the computational algorithm
based on the Radon transformation can be realized on Soviet-made computers of the M-222 type;
c) the time required for data processing with iterative methods is many times longer than the recon-
struction time necessary with the Radon transformation.
The accuracy of reconstruction for all these algorithms is the same. Analytical comparison of the re-
construction methods [13) showed that the Radon transformation is the ultimate form of the solution to this
problem.
The tomographic method of inspection was verified experimentally with several imitators of fuel elements,
resembling the fuel elements of the WER-1 water-moderated-water-cooled power reactor [14). The diameter
of the sand-filled imitators was 10.2 mm with a can thickness of 0.65 mm. Radiographic inspection was car-
ried out by x-raying the imitators. The x-ray tube of the RUP-150/300-10-1 x-ray machine was -700 mm
above the film. The imitators were placed right next to the film and turned through an angle a = 30?. A spec-
ial reader was used to process the radiographic data [151. The aperture of the reader in the scanning direction
had an opening of 50 ?m. The scanning pitch was varied from 0.2 to 0.055 m. The number of points in the
collection of data for each projection of the function from the x-ray picture varied from 30 to 180 counts.
Figure 2 shows the can and the filling of a rod-type fuel element in cross section with 0.1-mm cells in the re-
construction grid. The results confirm that the internal structure of fuel elements can be reconstructed by
transverse transmission tomography. Further studies on the resolution and the sensitivity (contrast) to various
defects are necessary, however, for practical application; moreover, it is necessary to optimize the discre-
tization of data acquisition over angular and pitch intervals as well as the conditions for obtaining radiographic
information for reconstruction of the tomographic image.
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 Declassified and Approved For Release 2013/02/12 : CIA-RDP10-02196R000800010006-0
Research on computer tomography (body-section radiography) for medical purposes [3-5, 6, 8, 12, 131
permitted the conclusion that the minimum size of a detected defect can be comparable to the size of the count-
ing window, i.e., the resolution will be 50-100 gm.
The method considered here can be used to analyze the uniformity of the distribution. of fuel inside fuel
elements in developing new designs or new technologies for fabricating fuel elements. Such information can
also be obtained by using neutron radiography which is employed in a number of research centers [16-18).
LITERATURE CITED
1. V. A. Dobromyslov and S. V. Rumyantsev, Radiation Introscopy [in Russian), Atomizdat, Moscow (1972).
2. R. V. Sinitsyn et al., Byull. Izobret., No. 41, 200 (1975).
3. G. Hounsfield, Brit. J. Radiol., 46, No. 552, 1016 (1973).
4. Z. Cho et al., Phys. Med. Biol., 19, No. 4, 511 (1974).
5. "X-ray scanning apparatus," Elektronika, 48, No. 26, 21 (1975).
6. E. Yu. Vasil'eva and A. D. Vasil'kova, Med. Radiologiya, No. 4, 52 (1978).
7. I. M. Gel'fand et al., Generalized Functions, Academic Press (1966).
8. E. Yu. Vasil'eva et al., Program for Reconstructing Body Section by the Inverse Radon Transformation
Method. File of Algorithms and Programs [in Russian), No. P002746, All-Union Research Institute of
Medical Instruments and Equipment (VNIIMI) (1977).
9. B. K. Vainshtein, Usp. Fiz. Nauk, 109, No. 3, 455 (1973).
10. R. Mercereau and N. Oppenheim, TIIER, 62, No. 10, 29 (1974).
11. Intern. Workshop on Three-Dimensional Image Reconstruction Techniques, BNL, U.S.A., July (1974).
12. Z. Cho, IEEE Trans. Nucl. Sci., NS-21, No. 1, 218 (1974).
13. H. Barret and W. Sandel, TIIER, 65, No. 1, 109 (1977).
14. V. S. Belokopytov et al., At. Energ., 30, No. 2, 211 (1971).
15. A. N. Maiorov et al., Radioisotopic Inspection in Russian, Atomizdat,.Moscow (1976).
16. C. F. Barton, Trans., 27, 212 (1977).
17, V. A. Karpeikin et al., in: Problems of-Atomic Science and Engineering. Radiation Technique [in Rus-
sian], No. 15, Atomizdat, Moscow (1977), p. 163.
18. V. A. Karpeikin et al., in: Problems of Atomic Science and Engineering. Radiation Technique [in Rus-_
sian], No. 16, Atomizdat, Moscow (1978), p. 202.
PROSPECTS FOR THE USE OF CARBON-CARBON-TYPE
OF MATERIALS IN NUCLEAR POWER ENGINEERING
K. A. Andrianov,* K. P. Vlasov,
L. L. Razumov, S. A. Kolesnikov,
V. I. Kostikov, I. I. Fedik, and L. M. Khananashvili
Carbon (in the form of artificial graphite) has found wide application in nuclear reactors. Besides its
traditional use as moderator and reflector of neutrons, it has been proposed to apply different technological
modifications of graphite also as a construction material - in the blankets of thermonuclear reactors and in
high-temperature gas-cooled reactors [1-31 for the jackets of fuel elements, matrix material, etc. Therefore,
the need arises for a broader consideration of carbon materials from the point of view of the optimal choice of
the kind of graphite for a specific technical problem. The principal types of carbon-graphite materials and
their properties are given in Table 1 [4-6). An appreciable effect of the original properties and peculiarities
of the technology on the characteristics of the irradiated material has been noted [5-71. The most significant
alteration of the properties is observed already for a small flux of fast neutrons (-1-2.1020 neutrons/cm2),
after which the properties are relatively stabilized, and new changes arise at a higher flux (1022 neutrons /CM2).
Translated from Atomnaya Energiya, Vol. 46, No. 6, pp. 406-408, June, 1979. Original article submitted
June 20, 1978.
0038-531X/79/4606-0461$07.50 ?1979 Plenum Publishing Corporation 461
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TABLE 1. Basic Properties-of Carbon-Graphite Materials
Strength limit,
M \
Thermal expansion6
tt 10'
coefficient
Thermal conducti-
k /mm2
gf
,
de-
vity, x (W /m ? deg K)
o
2
A
?
w
x
?
Material
01-4
a
7; c:
'
0
0
`
y
g
E
v
a
o
o
. d
w
a
o
0
0
4
b
0
?
a
Ll
0
U b
Q)
.
8 b
? V .~4
Q' N
.r
m
m
w
0
a?
-
N
GMZ
1
6-1
7 1
3
30
1
0
0
01
0
0
1,111
1
1
1
5
1
5
graphite
'
'
,
,1
,7
,5
11,3
4,1
5,4
6,1
6
,3
103
2
32
27
ARV
7 1
1
6-1
4,5
1,5
1,2 1
0,60 1
1,361
14,8 1
4
5
6
1
6 7
86
58
42
1
g aphite
,
,
3
4
5
,1
5
97
69
48
4
1
72-1
851
1
10,0
5,3
3,31
1,05 1
1,111
6 '
8
81
43
37
graphite
,
,
6
7
1
8:
rlphite
U
231
15-2
2
30-35
10,0 1
5,0 1
2,7 1
25
4,3
1-0
4
1
1
1
13
12
1
(
PV
,
,
60
:
210
24,0
25,0
25,0
3,2
1,7
1,4
1,45
carbon
as
11,45-1,52
26,0
8,35
1 5,2
1 2,7 1
1,0
1 .4,4
12
3
1
3,51
4,05
1 4,4
1 7,0
1 8,3
110,0
111,0
1c0)
l
2
4
Carbon-carbon-type of
1,40-1,45
42,0
50,0
29,0
13,5
10
15,8
-0,5
0,6
,0
1
1,4
21,0
14,0
10,0
9,0
material (KUP-VM)
5,6
7,0
.
-
-
1,
1,8
2,00
2,20
CSF (U.S.A.)
1 1,66
1
1
i
76
1
1 2,95
1
1
1
1
1
1
1
1
1
4,55
1
0,7
0
516
.
PGA (England)
1,65-1, i5
1
'
1
1
1 0
30
1,80
1
6
1
~
1
1
1
1
1
1
30
31
1,05
,
4'7
Notes: 1. The values of the characteristics oc, ?b, ct+ E, al/all, and p are given for
room temperature.
2. The data of the measurement in the direction parallel to the bonding (deposi-
tion) axis are given in the numerator, and that for measurement in the perpendicular
direction are given in the denominator.
As the temperature increases, the effect of irradiation decreases, although, e.g., in the case of pyrographite
[81 joining of defects occurs at an irradiation temperature >800?C, and a change in shape increases. One can
expect that graphite having the highest mechanical strength, a high modulus of elasticity, enhanced isotropicity
of properties, and a small thermal expansion coefficient, and which is subject to heat treatment at a tempera-
ture higher than 2000?C [6, 71, will have the best durability. Thus one should assume the application of car-
bon-carbon-type of materials with a high-modulus fiber, pyrolytic graphite, and glass carbon to be promising.
Only pyrographite from among the indicated materials has been investigated under conditions of irradia-
tion [5, 8-101. Therefore, it is advisable to investigate the basic possibility of using carbon-carbon-type ma-
terials (KUP-VM) under conditions of irradiation. Other materials were irradiated simultaneously, namely:
pyrolytic graphite (PGV), glass carbon (SU-850, SU-1300), and two domestic graphites, ARV and MPG-6. Tests
were conducted in two temperature ranges: 100-200?C ("cold" tests) and 850-950?C ("hot" tests). The dimen-
sions and shape of the samples, as well as the test conditions in loop channels, have been given earlier [101.
In the first stage of the investigations the values of the flux were 2.1020 neutrons/cm2, which usually corre-
sponds to the onset of stabilization of defect formation.
After the tests the samples were visually inspected, their geometrical dimensions were measured, and
they were weighed. The density of the material was determined by the method of hydrostatic weighing. In the
case of measurement of the electrical resistivity of rod samples the error was not greater than 1.5%. X-ray
structural analysis of the samples was performed on a DRON-1.5 apparatus with the application of a quartz
monochromator. The lattice parameter was calculated from the (006) and (110) reflections. The error in the
measurement of the interplanar distance was ?0.005X. The density of all the investigated materials declines
after the radiation tests. The largest density changes (-6%) are noted for pyrographite. Upon an increase in
the temperature of the radiation tests the density of the majority of the materials increases again. High-tem-
perature irradiation of PGV results in a decrease in the tetragonality of the structure: "a" decreases, and "c"
increases by ft1.1%. Metallographic investigations of PGV (Tir = 100?C, Tir = 900?C) did not allow revealing
any kind of. significant changes in the microstructure after irradiation. Thus, material based on a high-modulus
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Fig. 1. Effect of radiation dose on
changes in the linear dimensions
Da/ and 0lc/ (-), density Ad/d
(---), and the parameters of the
crystalline lattice Ac/c,and Da /a
(- ? -) of pyrographite and of a car-
bon-carbon-type of material Ad/d
in the case of an irradiation tem-
perature of 100-150?C (data of this
paper: o, o) pyrographite; 0) KUP-VM).
fiber obtained by the winding method showed good working qualities after irradiation both at 100?C and at
9-00-950?C (it preserved its integrity, stiffness, and mechanical strength with a minimal change in shape).
Pyrolytic graphite exhibited a significant change in shape (see Fig. 1) [6, 81 as well as splitting and lamination
failure of ring samples. Glass carbon underwent definite shrinkage, although it preserved its structure.
And so one can speak of the prospect of the use of carbon-carbon-type of materials in high-temperature
and thermonuclear reactors.
LITERATURE CITED
1. N. A. Dollezhal' and Yu. I. Koryakan, At. Energ., 40, No. 2, 133 (1976).
2. Coal Age, 7-9, No. 4, 104 (1974).
3. D. Bedinig, Gas-Cooled High-Temperature Reactors [in Russian], Atomizdat, Moscow (1975).
4. V. P. Sosedova (editor), Nature of Construction Materials with a Carbon Base [Russian translation],
Metallurgiya, Moscow (1975).
5. K. P. Vlasova (editor), Graphite as a High-Temperature Material [Russian translation], Mir, Moscow
(1964).
6. V. V. Goncharov et al., Effect of Irradiation on Graphite Nuclear Reactors [in Russian], Atomizdat,
Moscow (1978).
7. I. P. Kalyagina and Yu. S. Virgil'ev, At. Energ., 43, No. 2, 106 (1977).
8. P. A. Platonov, 0. K. Chugunov, et al., Preprint IAE-2247, Moscow (1972); Preprint IAE-2266, Moscow
(1973).
9. T. N. Shurshakova, Yu. S. Virgil'ev, and I. P.. Kalyagina, At. Energ., 40, No. 5, 399 (1976).
10. N. Abdusalyamov et al., in: Proceedings of the All-Union Institute for Investigating In-Reactor Methods
[in Russian], Scientific Testing Institute of Aviation Instruments (1978), p. 388.
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 Declassified and Approved For Release 2013/02/12 : CIA-RDP10-02196R000800010006-0
OF LIQUID METAL IN FUEL ASSEMBLIES
.V. P.. Kornilov and N. I. Loginov UDC 681.128
In investigations on the hydrodynamics and heat exchange in models of reactor cores it is necessary to
measure the flow rate in the cells of the lattice of a fuel assembly. The device described in [1) can be used to
successively determine. the flow rate in cells adjoining a given fuel element. To do this the fuel element must
be turned. about its axis along with the converter. A major disadvantage of the converter is that its output sig-
nal depends on the flow rate in neighboring cells as well as the given cell. It is difficult to take account of the
contribution from neighboring cells to the output signal,, especially under nonstationary conditions. A converter
with a copper ring, shunting the induced voltage in. neighboring cells [21, allows the influence of the neighboring
cells to be eliminated, thus making it possible to measure the azimuthal distribution of the velocity in the cells
by rotating the fuel element with the converter about its axis.. This converter cannot be used to measure the
flow rate in cells under nonstationary conditions.
A converter with a multipolar magnet (Fig. 1) has been proposed for simultaneous measurement of the
flow rate in several cells [31. The magnetic system has an alternating polarity and the number of poles is
chosen to be equal to the number of cells adjoining the fuel element in which the converter has been set up.
The electrodes are welded to the wall between the poles of the magnetic system. Their ends are led out beyond
the limits of the experimental segment and are connected to a secondary instrument. As it moves through the
space between the fuel elements the liquid interacts with the magnetic field and a potential difference is set up
between each pair of neighboring electrodes, this potential difference being in a single-valued relation with the
flow rate in the respective cell. The values of the flow rate in each cell at any moment of time can be assessed
from the simultaneous recording of signals from each pair of electrodes.
By using such a converter we can take account of the effect neighboring cells have on each other and ob-
tain a true picture of the distribution of flow rates over the cells. In view of the fact that an analytic solution
of the problem for a given geometry is very difficult to get, let us consider the equivalent electrical circuit of
the converter (Fig. 2). Here E1-E6 are the emfts induced in the liquid in the respective cells and are propor-
tional to the flow rate in them, r1-r6 are the internal resistances of the sources of emf, i.e., the liquid metal,
R1-Rg are the resistances of the load, i.e., of the walls of the converter and the adjoining fuel elements, shunt-
ing the emf, and U1-U2 are the measured voltages; the emf induced in a given cell is proportional to the flow
rate and. the magnetic induction in the cell, i.e., Ei = kB1Qi. If the geometric dimensions of the cells, as well
as the magnetic induction in them, are the same, then the coefficient k is the same for all the cells. Its value
can be found by calibration by measuring the total flow rate through a bundle (e.g., in the feed pipe) and as-
sociating it with the emf of all the cells. Then, in order to get the numerical values of the flow rate in each
cell it is sufficient to know the induced emf.
Setting up and solving the system of Kirchhoff equations for the equivalent circuit, we can find the rela-
tion between the measured voltage U1-U6 and the sought emf:
R't = U1 (1 +Rt/ri)+ n~/?nj (- 1)1-1 Ui (Rilri)?
Ri i=1
i=1
This formula is valid for any odd value of n. Since no constraints were placed in Ri and rl in the derivation of
the formula, it is also valid for cells of different shape.
The converter is intended mainly for measuring the flow rate in bundles with identical cells, i.e., can be
used in all cells except for peripheral cells. If the cells in the bundle are identical, then
R1 Re=...Re; ri Rilr1=R/r;
Rel R, = 1/6;
Translated from Atomnaya Energiya, Vol. 46, No. 6, pp. 408-410, June, 1979. Original article submitted
July 17, 1978.
464 0038-531X/79/4606-0464$07.50 ?1979 Plenum Publishing Corporation
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Fig. 1 Fig. 2
Fig. 1. Schematic diagram of converter: 1) magnetic system; 2) electrode; 3) can of
fuel-element mockup; 4) neighboring fuel elements.
Fig. 2. Equivalent electrical circuit of converter.
Et=Ut{1+R/r)+ (-61ri R (-1)t-1Ut.
It is not possible with this method of problem solving to calculate the values of Ri and ri, or their ratio, on the
basis of the known geometric dimensions of the cells and the physical properties of the liquid and wall. The
converter therefore requires calibration.
. The flow rate through the cell can be expressed as
t ``6 g
t=EilkBt= kBtt (1+R/r) + tfikBtrR LI (-1)t-t Ut=t1Ui f czUt.
1 1
The constants c1 and c2 are found in.the course of the calibration. To this end, we measure the total flow rate
through the bundle, all of whose cells are the same, and is set equal to the sum of the flow rates in the indi-
vidual cells:
`~n~ n `~6~
Qt0t Q1-~1 LJ Ui+cz LJ LJ (-1)t-1 Ut
1 1 i=f
Since the converter is calibrated at several values of the total flow rate, by choosing two values we can use
Eq. (4) to get a system of two equations with two unknowns c1 and c2 and to determine those unknowns. Substi-
tution of the values of c1 and c2 in Eq. (3) permits the flow rate through any cell to be found.
This converter with multipolar magnets can be employed to measure the flow rate in large-diameter
pipes. The converters with cylindrical magnets [4) used for these purposes yield quite reliable results only
in an undistorted axisymmetric flow since the flow rate is found from the known (e.g., logarithmic) velocity
profile. The true profile of the velocity in the pipe, however, is not known as a rule. The multipolar-magnet
converter makes it possible to determine whether the flow is axisymmetric. If the flow proves to be such, then
the doubts as to the reliability of the results are dispelled. Otherwise, the converter makes it possible to as=sess the degree of asymmetry of the flow and the possible error of measurement.
When the multipolar-magnet converter is in a circular pipe the problem lends itself to analytic solution.
In the two-dimensional approximation, i.e., with the assumption that the magnetic field extends along the axis
of the pipe to a considerable distance, we can find the potential distribution in the liquid and in the wall of the
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converter as well as the output voltage of the converter. The potential difference between two neighboring
electrodes of the converter is found from
r0/2+i (1+B U) Bo n/2 R1 W (r)
AU;a = 2/3Bm
ac-bd (RIB )n r'n*1
R
(5)
In the case of the problem of a six-pole converter the radius appears in the sixth power in the denominator of
the integrand and the value of the integral is determined mainly by the region of the flow around the converter.
The velocity profile in the, central part of the pipe is quite flat and W(r) WO = const. Then, for n = 6 we have
AUa,=2/3BmWoRo aecbdRe Ri (1-Ro/Ri)?
Since R/Rj usually has a value -0.1, the value of this ratio to the sixth power can usually be neglected:
AU,a- 7BmWoRol(riR /RB)/a],
where
a (1-l-Rn/Rn)+(awlot) (1-Ro/Rn);
Bm is the maximum value. of .the, magnetic induction on the surface of the magnet pole, and vv, and of are the
electroconductivity of the wall and the liquid (fluid.). Equation (7) is the fundamental basis for measuring the,
velocity (and flow rate) in a pipe.
The converter should be calibrated before being used. Since the device is sensitive only to the velocity
in the region adjoining it, the calibration during which the relation is established between the output voltage
and the velocity of the liquid can be carried out in a small-diameter pipe. The converter can then be used in
large-diameter pipes to measure the axial velocity and, in the case of an undistorted profile, to calculate the
flow rate.
LITERATURE CITED
1. V. I. Subbotin, M. Kh. Ibragimov, and N. I. Loginov, At. Energ., 25, No. 2, 150 (1968).
2. V. 1. Subbotin, M. Kh. Ibragimov, P. A. Ushakov, V. P. Bobkov, A. V. Zhukov, and Yu. S. Yur'ev, Hydro-
dynamics and Heat Exchange in Nuclear Power Plants [in Russian), Atomizdat, Moscow (1975).
3. N. I. Loginov, Inventor's Certificate No. 444938, Byull. Otkrytiya, Izobret., Prom. Orbaztsy, Tov. Zn.,
No. 36, 91 (1974).
4. N. I. Loginov, Magn. Gidrodin., No. 2, 128 (1971).
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 Declassified and Approved For Release 2013/02/12 : CIA-RDP10-02196R000800010006-0
IN A REACTOR
It is necessary not only to detect the boiling of water in the core, but also to determine where it is oc-
curring. This is important to prevent the creation and development of a heat-removal crisis which might dam-
age the fuel elements. Another problem involves monitoring the steam content of the steam-water mixture at
the core outlet. Boiling can be detected by ultrasonic, electrical, optical, and other methods based on a statis-
tical analysis of reactor noise power. We propose an optical method which employs the effect of Cerenkov
radiation of electrons in water.*
It is known that a charged particle moving in a dielectric medium with a velocity greater than the phase
velocity of light in that medium emits electromagnetic Cerenkov radiation which is propagated at an angle 9
with.the direction of its trajectory, where
/3 = v/c is the velocity of the particle in units of the velocity of light and n(v) is the index of refraction of the
medium as a function of frequency v.
The intensity of the radiation in the visible region is given by the expression
I - 450 sine 0 photons/cm. (1)
The threshold for the Cerenkov effect is given by the condition /3t > 1/n(v). The limiting angle of radiation is
reached for (3 = 1. In this case (cos 0)Z = 1/n(v). The threshold kinetic energy of a particle Er is related to
the index of refraction by the expression
E_ = mo?Q t {n2 (v)(v)1j1/2 -1} , (2)
where m0c2 is the rest energy (0.51 MeV for an electron). Thus, the threshold energy for the production of
Cerenkov radiation by a given kind of particle is determined by the index of refraction of the medium.
In an aqueous coolant having subcritical parameters the density of the liquid phase (water) is higher than
that of the gaseous phase (steam). The index of refraction for each phase is given by the Lorenz-Lorentz
equation
n (v) = I W +2PA (v)IW - PA ,(v)11/2,
where W is the molecular mass of the medium; o, density of the medium; A(v), molar refractivity (A(v) = 3.7
for sodium D light). Because of the difference in densities, the index of refraction of water is higher than that
of steam. Consequently, the threshold energy Et for the radiation of electrons is lower in water than in steam.
The graph of Fig. 1, which was obtained from Eqs. (2) and (3), shows that for water Et is practically independ-
ent of pressure, while for steam it varies rapidly.
Most of the electrons in an aqueous coolant are produced as a result of the interaction of rays with core
materials and the coolant. For simplicity, we can assume that in an operating reactor the energy distribution
of the flux density of the electrons which give rise to Cerenkov radiation is determined by the corresponding
distribution of the flux of prompt y rays. Above ti 0.5 MeV the prompt y flux falls off with increasing energy.
For example, fewer than 5% of the y photons per fission have energies above 2.5 MeV. The energy distribution
of the flux of electrons producing Cerenkov radiation on the whole also falls off with increasing energy.
*I. I. Zakharkin, Inventor's Certificate No. 448770. Byull. Otkrytiya, Izobret., Promysh. Obraztsy, Tov. Zn.,
No. 40 (1974), p. 143.
Translated from Atomnaya Energiya, Vol. 46, No. 6, pp. 410-411, June, 1979. Original article submitted
August 7, 1978.
0038-531X/79/4606-0467$07.50 ?1979 Plenum Publishing Corporation 467
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4 6 8 102 2 4 101 2 4 6 8 102 2 4
P. kgf/cm2 p, kgf/cm2
Fig. 1 Fig. 2
Fig. 1. Threshold kinetic energy of electrons. for water and steam at the satura-
tion line.
Fig. 2. Ratio of intensities of radiation in steam and water at the saturation line
for electrons with /3 1.
The number of electrons giving rise to radiation in water will be appreciably larger than the number
producing radiation in steam. Electrons with energies below the limit for steam will radiate only in water.
Thus, as a result of the difference in indices of refraction of water and steam and the shape of the energy
spectrum of the electrons producing Cerenkov radiation, the total intensity of the radiation per unit path length
will be higher in water than in steam.
It must also be remembered that even for electrons moving with a velocity /3 1, when the limiting angle
of radiation is reached both in water, and in steam, the intensity of the radiation will be higher in water than in
steam. Actually, for /3 1 the ratio of the intensities of the radiation per unit path length in steam and water
on the basis of Eq. (1) is determined by the expression
I slI w ^-' (n2 (v) -1) nw (v)/(n,$, (v) -1) %2(v),
where ns(v) and nw(v) are, respectively, the indices of refraction of steam and water. Figure 2 for Is/Iw as a
function of the pressure of the steam-water mixture at the saturation line shows that, e.g., for a pressure of
100 kgf/cm2 Is/Iw - 0.1, i.e., the radiation intensity inappreciably lower in steam than in water. Estimates
show that for the pressure chosen, taking account of the actual electron spectrum, the radiation intensity in
steam can be neglected. In this case the radiation intensity in the steam-water mixture will be proportional to
the liquid phase fraction, i.e., to the water.
It should be noted that the present discussion does not take account of the slowing down of electrons in the
medium, which occurs mainly because of ionization losses. The radiation intensity in the medium decreases
with decreasing electron energy. The radiation vanishes for electron energies below the threshold. A more
accurate analysis would require taking account of the slowing down of electrons in the medium, but the main
conclusions about the possibility of using Cerenkov radiation to detect boiling are not changed.
Light from the region of the coolant being monitored can be led out through hollow metal light pipes. A
measurement of the.Cerenkov light emerging from a light pipe gives the necessary information about the state
of the coolant. This method can be applied to an experimental thermophysical channel by using y rays to ex-
cite the Cerenkov radiation. In addition, by choosing a y source with energies below E~ in steam, it is possible
in principle to eliminate Cerenkov radiation in the steam fraction of the coolant.
Declassified and Approved For Release 2013/02/12 : CIA-RDP10-02196R000800010006-0
 Declassified and Approved For Release 2013/02/12 : CIA-RDP10-02196R000800010006-0
ENERGY DISTRIBUTION OF 235U FISSION-PRODUCT
yRADIATION FOR A SHORT IRRADIATION TIME
M.. A. Markina, E. S. Stari,znyi,
and A. Kh. Brege.r
In connection with a detailed study of the radiation and technological characteristics of complex-power-
chemical installations with uranium radiation loops (URL) as y sources [11, and the development of the VGR-50
Fig. 1. Time variations of total (o) specific
dose rate of fission-product y radiation and
various energy groups for irradiation times of
a) 300 sec: b) 1 h; c) 10 h; 1) 0.04-0.4; 2) 0.4-
.0.8; 3) 0.8-1.2; 4) 1.2-1.6; 5) 1.6-2.0; 6) 2.0-2.4;
7) 2.4-2.8; 8) 2.8-3.2 MeV; 9) results of [8].
Translated from Atomnaya E`nergiya, Vol. 46, No. 6, pp. 411-413, June, 1979. Original article submitted
September 1, 1978.
0038-531X/79/4606-0469$07.50 ?1979 Plenum Publishing Corporation 469
Declassified and Approved For Release 2013/02/12 : CIA-RDP10-02196R000800010006-0
 Declassified and Approved For Release 2013/02/12 : CIA-RDP10-02196R000800010006-0
Fig. 2. Dependence of average energy of energy
spectrum on cooling time for tr = 300 sec ~? ) our
data; 0) [41]; for tr 1 h [-) our data; 0) 151; 0) [41;
A) [71, for tr = 10 h ---) our data; x) [51; V) [711.
10 10-1 100 101 102 10, t, h
installation in our country f2, 31, information is required on the energy distribution of y spectra for short.
(from a few minutes to several hours) fuel residence times tr in the core.
The published time variations of y spectra were obtained either by processing [4] known calculated [ 51
and experimental [61 curves for prompt fission without taking account of the burnup of fission products in the
core, or by calculations [7]. The. values of tr 1 by using the effective viscosity (6v+v) instead of 1'v in [1) and
adopting the indicated expression for Zij. For liquids with Pr