SOVIET ATOMIC ENERGY VOL. 57, NO. 6

• Document Type: CREST
• Collection: General CIA Records
• Document Number (FOIA) /ESDN (CREST): CIA-RDP10-02196R000300050006-1
• Release Decision: RIFPUB
• Original Classification: K
• Document Page Count: 81
• Document Creation Date: December 27, 2016
• Document Release Date: February 22, 2013
• Sequence Number: 6
• Publication Date: December 1, 1984
• Content Type: REPORT

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~, ~ ISSN 0038-531 Y.
Russian Original Vol. 57, No. 6, December, 1984
June, 1985
5ATEAZ 56(6) 803-880 (1984)
SOVIET
ATOMIC
ENERGY
ATOMHAI:I 3HEP~VIF1
(ATOMNAYA ENERGIYA)
TRANSLATED FROM RUSSIAN
CONSULTANTS BUREAU,NEW YORK
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 Declassified and Approved For Release 2013/02/22 :CIA-RDP10-021968000300050006-1
SOVIET
ATOMIC
ENERGY
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SOVIET ATOMIC ENERGY
A translation of Atomnaya Energiya
Volume 57, Number 6 December, 1984
CONTENT8
Engl./Rues.
ARTICLES '
Major Safety Provisions in Nuclear-Powered Ships - N. S. Khlopkin,
O. B. Samoilov, V. M. Belyaev, A. M. Dubrovin, ~. M. Mel'nikov,
and B. G. Pologikh ... . . .. .... .... . .... .... .. .. .. .... .... .. 803 379
Tests on Improved Steam Separators in the Third Unit at the Chernobyl Nuclear
Power Station-O. Yu. Novosel'skii, V. B. Karasev, E. V..Sakovich,
M. A.Lyutov,andV.I.An'kov ?. ~ 807 382
Peculiarities of the Distribution of Phases in the Updraft Section of a Housed
Boiling Reactor - V . N. Fedulin, G. G. Bartolomei, V. A. Solodkii,
and V. E. Shmelev ... .. .. .. .. .. .. .. .. .. .. .. .. .. ......... ....... 811 385
Effects of Steam Generator Sectioning on the Reliability of a Nuclear Power Station
Containing a Fast Reactor - A. I. Klemin, O. B. Samoilov,
and E. V . Frolov ..... .... .. ...... ...... ...... .. .... .......... .. 815 388
Statistical Analysis of Reactor Thermal Power by the Use of Thermal and Radiation
Methods in the First Unit at the Armenian Nuclear Power Station - F. D. Barzali,
L. N. Bogachek, V. V. Lysenko, A. M. Muradyan, A. I. Musorin, A.I. Rymarenko,
I. V. Sokolova, and S. G. Tsypin .. .. .. .. . . .............. 821 393
Test Stand for Research on the Physics of High-Temperature Gas-Cooled Reactors
- A. M. Bogomolov, V. A. Zavorokhin, A. S: Kaminskii, S.V. Loboda,
A. D. Molodtsov, V. V. Paramonov, V. M. Talyzin, and A. V. Cherepanov . ? . ? ..... 825 397
Study of Model Coils Made of Suprerconductor Intended for the Winding of the T-15 Tokamak
- I. O. Anashkin, E. Yu. Klimenko, S. A. Lelekhov, N. N. Martovetskii,
S. I. Novikov, A. A. Pekhterev, and I. A. Posadskii ................ ....... 830 401
Oscillations in the Concentration of Artificial Radionuclides in the Waters of the Baltic
and North Seas. in 197?-1982 - D. B. Styro, G.I. Kadzhene, I. V. Kleiza,
and M. V . Lukinskene ? ............................... ....... 835 405
REVIEWS
Enhancement of Heat Transfer in the RBMK and RBMKP - A. I. Emel'yanov,
F. T. Kaman'shchikov, Yu. M. Nikitin, V. P. Smirnov, and V. N. Smolin ? ~ 839 408
LETTERS TO THE EDITOR
Stresses in Spherical Fuel Elements of aHigh-Temperature Gas-Cooled Reactor (VTGR)
as a Result of the Heat Load and Radiation Shrinkage of Graphite - V. S. Egorov,
V. S. Eremeev, and E. A. Ivanova ...................... ... .... ?'? 846 415
Development of an Oil-Free Forevacuum Unit for an Operating Pressure of 150-300 Pa
for the Tokamak-15 - I. A. Raizman, V. A. Pirogav, L. G. Reitsman,
E. A.Maslennikov,andV.V.Martynenko 850 417
Surface Tension of Molten Mixtures of Fluorides of Lithium, Beryllium, and Thorium
- A. A. Klimenkov, M. N. Kurbatov, S. P: Raspopin, and Yu. F. Chervinskii ? 853 419
Experimental Study of the Interaction of Pulsations of the Neutron Flux and the Coolant
Flow in a Boiling-Water Reactor - P. A. Leppik ....... ? . ? . ? ? 855 420
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CONTENTS
(continued)
Engl./Buss.
Control Experiment on Critical Heat Transfer during Water Flow in Pipes
-P.L.Kirillov,O.L.Peskov,andN.P.Serdun'?? 858 422
INDEX
Author Index, Volumes 56-57, 1984 ....................................... 863
Tables of Contents, Volumes 56-57, 1984 ................................... 869
The Russian press date (podpisano k pechati) of this issue was 11/28/1984.
Publication therefore did sot occur prior to this date, but must be assumed
to have taken place reasonably soon thereafter.
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MAJOR SAFETY PROVISIONS IN NUCLEAR-
POW.ERED SHIPS
N. S. Khlopkin, O.B. Samoilov, V. M. Belyaev,
A. M Dubrovin, ~. M Mel'nikov,
and B. G. Pologikh
Considerable experience has been accumulated in this country on the design, construction, and operation
of nuclear-powered civilian ships: the icebreakers Lenin, Leonid Brezhnev, and Sibir'. The nuclear steam
plants (NSP) used on these as the main energy source have been found to be highly reliable and safe, and it is
desirable to use them in the future not only in icebreakers but also in transport ships for use in ice fields.
The Soviet program for building and developing nuclear-powered ships has involved careful attention to
safety in ships containing NSP. The experience with the design and operation of nuclear icebreakers in recent
years has led to the revision of safety standards for the nuclear ships and correspondingly ship NSP [1, 2] and
international guidelines have been developed. If one meets the requirements of these documents, one has a
safe basis for future Soviet nuclear-powered ships.
Basic Safety Principles. Basic safety problems implied by the features of NSP are related to screening
the radiation sources and preventing the escape of radioactive substances, in addition to reliable core cooling
preventing hazardous situations by transferring the reactor to a subcritical state, and maintaining it in that
state for an appropriate time. The execution of these functions is delegated to independent safety systems. The
composition, structure, and mode of operation in these systems are chosen in accordance with the following
basic principles:
1) high fault-free operation parameters for all units;
2) the safety system should perform its functions when there is a single failure in any component;
3) a reliable and fast emergency reactor protection system;
4) the conditions should rule out damage to the fuel pins in all modes of operation;
5) prevention of coolant loss from the core in emergencies related to loss of coolant from the first loop;
6) fission products should not be allowed to pass beyond set barriers; and
7) preventing the pressure in the first loop from rising above a permissible level set by the strength
in any design emergency.
To provide reliable NSP safety, one combines the deterministic method of meeting the standard safety re-
quirements with a probability method, in which one considers not only the consequences of emergencies but also
the possibilities of their occurrence; quantitative criteria from this method are currently being formulated.
Two methods are to be considered as complementary in researching and providing safety. We note that a prob-
ability analysis is obligatory in designing nuclear, power stations [3] .
Basic Technical Specifications in Safety Provision in Future Nuclear-Powered Ships. -The NSP in nu-
clear ships under design or construction usually employ well-tested designs and units, which have been thorough-
ly checked out on nuclear icebreakers and which have shown excellent safety and working life: During the manu-
facture, installation, and testing of NSP equipment, one uses well- evaluated technology and quality- control
methods. Changes and improvements are made in the design of the equipment, particularly to increase the re-
liability and the monitoring performance and completeness. Provision is made for ongoing automatic monitor-
ing of all technological parameters during operation, including the radiation environment in spaces within the
ship, as well as periodic examination and checking of equipment, pipelines, and systems as a whole. These
measures together ensure reliable and safe operation.
Translated from Atomnaya Energiya, Vol. 57, No. 6, pp. 379-382, December, 1984. Original article sub-
mitted August 27, 1984.
0038-531X/84/5706- 0803$08.50 ?1985 Plenum Publishing Corporation 803
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One of the major means of attaining high reliability parameters in NSP is the provision of equipment
backup.
The reactor core has a negative temperature coefficient of reactivity in the working temperature range,
which provides self-regulation and dynamic stability. This self-regulation is a factor that restricts the in-
crease in reactor power and thus tends to alleviate the consequences of emergencies associated with the power
production rate exceeding the rate of core cooling.
The control and protection system provides for multichannel parameter monitoring under working condi-
tions. The emergency-protection circuits are galvanically decoupled from the measurement channels. The
discrete signals that operate the protective equipment are produced independently in each channel and pass to
the effectors along duplicated circuits . The links between systems viathe protection signals are also duplicated in the
control system. Built-in monitoring devices at various levels enable one to detect and localize faults in individual units,
which can be promptly replaced. The response rates of the electrical and mechanical emergency-protection
components provide for shutting down the reactor in any emergency while maintaining permissible temperatures
in the fuel pins .
Under normal working conditions, with the circulation pumps running, the heat from the core is carried
off by the coolant in the first loop passing through the steam generators and onward to contact with the coolant
in the second loop. When the reactor is shut down, the residual heat is removed from the core by two indepen-
dent cooling channels, which are activated automatically or remotely from the emergency protection signal.
The first is the cooling channel for the water in the second loop passing through the steam generators when the
circulation pumps in the first loop are working together with the pumps in the second loop. The second is the
cooling channel with water in the third loop, which operates when it is impossible to supply feedwater to the
steam generator.
The emergency feed pumps in the second loop with their pipes and equipment form an independent cooling
system, which is not dependent on the main condensate-feed system, and this provides ways of cooling the re-
actor in any design emergency situation.
The users of the NSP have reliable uninterrupted power supplies under all normal and emergency condi-
tions, in accordance with the requirements of [1]. Experience with the power plants in nuclear icebreakers in-
dicates that complete current failure in the NSP needs to be envisaged only for the time required for the emer-
gency power supplies to be activated. On the other hand, calculations show that the systems remain viable with
much longer current failures, in spite of the severe working conditions assumed for emergencies. It is to be
considered only as a purely hypothetical situation that the NSP fails to provide power for a prolonged period
(i.e., loss of the main power supply and failure to obtain a supply from the backup and emergency sources), as
there is multiple backup in the electrical power system. Nevertheless, this is considered in designing NSP.
The main task in that case is to provide ongoing reactor cooling by means of facilities independent of the state
of the ship's electrical power supply.
Foreign standardization documents dealing with safety in nuclear ships envisage installing safety valves
in the first loop as a passive means of preventing overpressure. However, this is not an obvious solution, be-
cause of features of the first loop: It is a powerful source of radioactivity, and there is the potential hazard
that the core will be deprived of coolant and thereby damaged, with its serious radiation consequences, while. in
addition there is the low reliability of such safety devices. The available experience with them confirms this
viewpoint. One possible reason for pressure rise in the first loop is that the coolant acquires excess heat, and
this makes it the most effective approach to provide high reliability in the facilities for shutting down the reac-
tor and cooling the NSP, which should radically prevent impermissible pressure increase and eliminate the
need to fit safety valves [4] . This is shown to be correct by prolonged experience in the accident-free operation
of NSP in nuclear icebreakers.
New designs envisage automatic connection of an independent system to the steam generator, which con-
sists of tanks containing water, a gas cylinder, pipelines, and other equipment designed to supply feedwater to
the steam generator and to discharge the steam to the atmosphere without resort to external power supplies, in
order to provide emergency cooling in the first loop when there is complete power failure (or failure of all the
cooling plant). The heat is removed from the core by the natural circulation of the coolant in the first loop.
These cooling provisions restrict the pressure increase in the first loop under all normal and emergency
conditions and ensure that it retains its integrity.
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Fig. 1. Scheme for supplying the first loop and
for providing emergency flushing to the core in a
container transport ship: 1) reactor; 2) flushing
pumps; 3) reserve tank;. 4, 5) feedwater tank
and pumps, correspondingly; 6) emergency flush-
ing valves.
u
e~
F
5
200 - 4
n,
10
I'Is
.\
~ - Core .~ P
i i ~
S00 1000 ~ sec
Fig. 2. Change in coolant parameters when the
first loop fails; solid line and dot-dash line) pipe-
line failure in the feed and compensation systems,
correspondingly; P) pressure in reactor; Vr)
coolant volume in reactor; Tsh) maximum tem-
perature of fuel-.pin sheaths.
One of the most serious emergencies is a leak in the first loop. To reduce the probability of this occur-
ringandtorestrict the consequences, there are special design measures such as minimal length in the pipeline
in the first loop and the pipelines connected to the reactor above the core. In new reactor systems such as for
transport ships, there is provision for constriction nozzles in the reactor pipes. The systems adjoining the
first loop are also fitted with a double pressurizing system. All the equipment and pipes in the first loop are
located in a hermetic container. In accordance with the standardization documentation, it is conservatively as-
sumed in considering emergencies involving leakage in the first loop that there is instantaneous failure in some
pipe connecting the body of the reactor to the first-loop equipment. It should be noted that the estimate of the
probability is less than 10-s event/reactor ? yr for an emergency involving failure in the structures within the
containment .(reactor, steam generator, and power pipes in the reactor), which represents a practical impossi-
bility, and the consequences are not examined.
The following are the major operational measures to be taken when the coolant escapes. from the first
loop: Shutting down the reactor, identifying and localizing the leak, feeding the loop and flushing the core, and`
emergency cooling.
The signal indicating pressure drop in the first loop below the permissible value leads to the reactorbe-
ing shut down automatically, and at the same time the purification system is shut off along with the working
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groups of cylinders in the pressure-compensating system, and the backup pumps are started up along with the
backup and emergency feed pumps, and the equipment in the flushing system is activated. When the pressure
in the first loop falls to the pressure provided by the flushing pumps, controlled nonreturn valves on the flush-
ing pipes open, and water begins to enter the reactor through two channels (Fig. 1). All these automatically
executed operations can be duplicated remotely from the control panel. The flushing water is supplied via two
independent branches, while the backup in the pumps and other equipment enables one to meet the above failure
principles in the safety system. If the pipe fails in a part that cannot be shut off, one cannot localize the leak,
and the main task in that case is to ensure that the core is supplied with coolant in order to provide acceptable
temperature conditions in the fuel pins. The output of the flushing pumps is chosen accordingly.
As one provides temperatures such that the fuel pins remain viable, this prevents failure in the sheaths,
which represent the first barrier to the escape of fission products, while the cooled core geometry is main-
tained, and thus any change in the state of the fuel-pin cores is ruled out such that the radionuclides contained
in them are rapidly released.
If there is a leak in the first loop, which is the second barrier to the fission products, the protective con-
tainment comes into action (third barrier). To prevent the pressure in the containment rising to the permissi-
ble limit, provision is made for passing the steam-air mixture into a special ballast volume through a bubble
tank or into a tank within the shielding.
A study of the consequences of leaks shows that the core remains flushed with water if there'is a failure
in any pipe in the first loop if the above protection facilities are fitted, and the fuel pins do not attain a critical
state, while the pressure in the containment remains below the design value (Fig. 2).
This conception of nuclear ship safety is based on experience accumulated in the construction of power
plants for nuclear icebreakers and the operating experience with these of total extent about 80 reactor ears.
This experience has been utilized in drawing up standardization documents and in designing new nuclear ships.
1. Rules for Classifying and Constructing Nuclear Ships [in Russian], Leningrad (1982).
2. Nuclear Safety Rules for Ship-Borne Nuclear Power Systems (PBYa-08-81) [in Russian], Atomizdat,
Moscow (]981).
3. General Safety Provisions for Nuclear Power Stations daring Design, Constr_ uction, and Operation (OPB-
82) [in Russian], Moscow (1982).
4. F. M. Mitenkov, L. A. Zvereva, B. I. Motorov, ~. M. Mel'nikov, and O. B. Samoilov, "The use of asafety
valve in the first loop of a nuclear power system containing a VVER reactor," At. Energ., 50, No. 5, 308-
310 (1981).
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TESTS ON IMPROVED STEAM SEPARATORS IN THE
THIRD UNIT AT THE CHERNOBYL NUCLEAR
POWER STATION
O. Yu. Novosel'skii, V . B. Karasev,
E. V. Sakovich, M. A. Lyutov,
and V . I. An'kov
Improved separator drums (Fig. 1) have been tested at the Chernobyl nuclear power station with its
RBMK-1000 reactors, these differing from those used at Leningrad power station and the first units at Kursk
and Chernobyl power stations [1] in having a larger internal diameter (2600 instead of 2300 mm) and a different
design for the internal devices (IIID). Features of the new ID design include the drainage diffusors 1 and the
base of-the compartments 2, the inclined perforated sheets 3, and the breather tubes 4 for taking off the steam
from the space 5 between the compartments into the steam volume 6. Also, the separator characteristics have
been improved by increasing the distance between the immersed perforated sheet 7 and the upper perforated
shield 8. The increased diameter has reduced the mean speed at the evaporation surface by about 7.5%? The
drainage diffusors enable one to use the water in the segments in emergency and transitional states. The in-
clined perforated sheet of section 12.5% equalizes the velocity pattern at the tapoff from the segments into the
spaces between them, which tends to increase the water retained in the separator.
The design of the drainage channel in the immersed perforated sheet (unit A in Fig. 1) incorporates the
advantage from modifyingthe ID in the 2300-mm-diameter separator [2]: The side piece 9 has a height of 170
mm, while the height of the drainage slot is not more than 55 mm. At the surface of the sheet there are stif-
fening ribs 10 with holes of height 100 mm and working section 18%.
The. immersed .perforated sheet in the improved separator is fitted with the end pieces 1 ] of height 485
mm. The brackets 12 in the segments have windows allowing the steam-water mixture to flow along the drum
and which.thus~ equalize the load along the length. These modifications were incorporated into the design when
the separator drum of diameter 2300 mm constituted the bottleneck in the forced circulation loop in the RBMK-
1.000. The improved drum has four rows of steam~vater tubes 13, whose diameter has been increased to 100
mm .
Fig. 1. Cross section of the separator drum in
the third unit at Chernobyl nuclear power station.
Translated from Atomnaya Energiya, .Vol. 57, No. 6, pp: 382- 385, December, 1984. Original article sub-
mitted March 26, 1984.
0038-531X/84/5706- 0807$08.50 ?1985 Plenum Publishing Corporation
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Fig. 2. Location of separator drums, steam pipes,
balancing vessles, and steam sampling points: 1)
pipes supplying steam to the machine hall; 2) steam
collectors; 3) steam tapoff pipes in separator drum;
4) level gauges with balancing-vessel baseline of
630 mm to measure water level above immersed
perforated sheet; 5) end level gauges with 1600-mm
baseline of balancing vessels; 6) steam sampling
lines in steam-collecting tubes; 7)' steam sampling
from steam pipe of diameter 600 mm; and a, c,and
b) end and central steam-collecting tubes, corre-
spondingly.
This third unit also employs a new assembly for the main equipment. In particular, the .separator drums
are turned through 90? and are set parallel to the machine bay (Fig. 2). The steam-pipe -layout has also been
altered. In the new layout, two collectors of diameter 400 mm from adjacent separator drums are combined in
one steam pipe of diameter 600 mm, which goes to the turbine. The steam from the collectors is taken from the
middle of the separator drum, which produces a more uniform distribution of the steam flow over the steam
tubes. Anew system is also used to measure the water level. All the balancing vessels are set up in the
staffed (cold) location, and use is made of vessels with partially heated central chambers as devised by the
Dzerzhinskii All-Union Thermal Engineering Institute. In each separator drum, one measures the water level
above the immersed perforated sheet in three sections along the length in vessels with a baseline of 630 mm
and the general level at the end in vessels with a baseline of 1600 mm.
During the installation and the runup to the nominal power, we commissioned and checked the level-mea-
suring system, which involved checking that the balancing vessels had been correctly manufactured and in-
stalled, that the connecting and pulse lines had been correctly installed, checking the blowers, and checking the
readings of the level gauges with a pressure difference OP = 0. The readings of the gauges were checked be-
fore commissioning the unitwith a water temperature of 20?C with the level in the separator drum varying in
the range from -350 to +400 mm. In the period of runup to the nominal power, the scales of the secondary in-
struments ranging from -100 to +315 mm inthe630-mm-baseline gauges were replaced by scales of from
-200 to +315 mm. The 0 level on a secondary instrument in that case corresponds to a mass level on the im-
mersed perforated sheet of 150 mm.
To test the separators, steam samples were taken from the steam pipes and tapoffs in accordance with
the scheme shown in Fig. 2. The water samples were taken from the body of the drum, while the steam. was
taken from the central tubes in each separator drum, and also from the twotubes in the BS 11 separator drum.
There were also steam sampling points in the machine hall before the turbines, from which one could deter-
mine the average water content. However, the cooling system envisaged in the design did not provide iso-
kinetic sampling ahead of the turbine with a thermal power of more than 80% of nominal. To determine the
average water content of the steam at the exit from all the separator drums at power levels of over 80%, the
samples were taken after the turbine condenser. The water content was determined from the ratio of the 24Na
concentrations in the condensate and in the circuit water by the method given in [3] .
The tests were performed at a pressure of 7 MPa and thermal power levels of 65, 83, 93, and 100% 'of
nominal. The purpose of these was to determine the dependence of the water content at the exit from the sepa-
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Fig. 3 Fig. 4
Fig. 3. Comparison of readings for gauges measuring the mass water level above
the immersed perforated sheet at various reactor thermal powers in MW: O) 2100;
~) 2660; D) 2960; Q) 3200; ~) 3350; ?) 3200.
Fig. 4. Dependence of the steam wetness 1 - x in the central steam-tapoff tubes in
the separator drums on the mass water level above the immersed :perforated sheet
(from the middle gauge) at the nominal thermal power: 1) O BS 11; 2) ~ BS 21;
3) O BS 22; 4) Q BS 12; -- central level gauge off scale; ~ water content less than
U .019'0
U1
O,OB
.o, o~
0, 04
-zoo 0 200
HgSn, mm
Fig. 5. Dependence of the average steam
wetness after the separator drums on the
mass water level above the immersed
perforated sheet (from readings of middle
gauge in BS 11) as indicated by taking
samples in the turbine condensers at the
nominal reactor thermal power; ? and O)
condensers K 5 andK6, correspondingly.
rator on the mass level at a given thermal power (steam production). The tests began at the nominal level, and
then the level was raised in steps of 50 mm until the water content of the steam at the exit exceeded 0.19'0 or
the upper limit was reached for the level gauges with scales of -200 to +315 mm. The limiting height of the
mass level of water above the zero mark at which the upper sampling point was reached for the gauges with
scales of -200 to +315 mm was determined by calculation from the formulas of [3], and it was 285 mm for the
nominal unit power. We also made tests with a level of -100 mm indicated by the instruments with scales of
-200 to +315 mm, i.e., +50 mm above the immersed perforated sheet.
When the steady state had been reached (constant power, level in the separator drum, pressure, and flow
rate), we established isokinetic flow rates for the steam samples, which were calculated from the measured
steam flow rates in: the pipes. These conditions were maintained for 30 min, and during this time we recorded
the basic working parameters. The levels in both pairs of separator drums were raised simultaneously in
order to measure the average water content after all four. In the pair of drums 11 and 12, the level was main-
tained from the gauge installed at the middle of drum 11, while in the other pair 21 and 22 it was maintained
from the gauge installed at the end of drum 21 remote from the steam pipes. During the commissioning period,
the readings of these gauges were found to be the most reliable. Comparison of the readings (Fig. 3) showed
that they agreed well (remained within the accuracy class) before and after commissioning the sets of level
gauges in August 1982. Straight lines 1 and 2 in Fig. 3 define the boundaries within which the set of gauges with
baseline 630 mm remained within the class. The deviations in the readings of the gauges with 630- and- 1600-
mm balancing-vessel baselines did not exceed the absolute error indicated by the accuracy class of various
reactor power levels.
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Fig. 6. Steam wetness distribution in steam tapoff
tubes along the length of separator 11 for various
values of the thermal power Nt in MW and various
mass water levels Hm in mm above the immersed
perforated sheet (as indicated by the middle gauge);
OO) HM = 0, Nt = 2960; O) 10, 3350; O) 140, 3200;
~) 200, 3200; O) 250, 3200; O) 260, 3200; i wet-
ness less than 0.01?Jo; 1-3) sections of separator
11 at steam tapoff tubes a, b, and c, correspondingly.
Figure 4 shows that the wetness of the steam in the central tubes in all the separator drums. did not exceed
0.02% up to levels of 200 mm above the nominal, but there was a sharp increase at 250-270 mm. An exception
was represented by the separator 12 (curve 4 in Fig. 4), which was evidently due to overestimated readings from
the middle gauge in this separator. Similar results were obtained at power levels of 83 and 939'0 of nominal.
However, as the power level decreased, the sharp increase in the water content began at higher levels in the
separators. With a thermal power of about 65% of nominal, we measured the steam wetness in front of the tur-
bines. The measurements agreed with those on the central tubes in the separators.
The wetness data obtained at the nominal power by measuring the 24Na activity in samples from the tur-
bine condensers (K 5 and K 6) are given in Fig. 5 by reference to the readings of the middle gauge in separator
11. With mass levels above the immersed perforated sheet of from -100 to +200 mm, the wetness was less
than 0.02%. When the mass level rose above +200 mm, the average wetness began to increase, and it attained
0.1% at +255 mm.
Comparison of Figs. 4 and 5 indicates that the limiting level corresponding to a wetness of 0.1% on mea-
surements in the central tubes is higher than that from the determination of the average wetness, which is ex-
plained by the measurements on the three tubes in separator 11, which illustrates the wetness distribution
over the length (Fig. 6). For a level less than 200 mm, the wetness in the central or end tubes does not exceed
0.02- 0.030J0a although there is a tendency for the wetness to be higher in tube a. With levels over +200 mm, the
wetness in the end tubes is higher than that in the central one, and the difference is by two orders of magnitude
with levels between +250 and +260 mm. The reason for the elevated wetness at the ends of the separator when
the water level exceeds +200 mm is not due to the nonuniformity in the steam and water loads along the length,
since the maximum occurs in the central part of the drum. With the working mass level in the separator, the
wetness near the ends does not exceed 0.02%, and consequently this does not make an appreciable contribution
to the average wetness ahead of the turbine. An average exit wetness of 0.1% is attained with the level lower
than the wetness in. the central tube on account of the elevated wetness at the ends of the separator
drum when the level is more than +200 mm as indicated by the instruments with scales from -200 to +315
The test results led us to increase the nominal mass level in the separator by 100 mm (to the +100 mm
mark as indicated by the gauges from -200 to +315 mm), which increased the water stock by 7 m3 while keep-
ing the steam wetness below 0.19'0.
Therefore, the separators of diameter 2600 mm have a working margin as regards limiting permissible
wetness. By maintaining the level over the immersed perforated sheet 100 mm higher than the design value,
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one can increase the water stock in the forced circulation loop by 28 m3 without interfering with increasing the
steam output.
LITERATURE CITED
1. V .. B. Karasev, Yu. M. Nikitin, O. Yu. Novosel'skii, and E . V .Sakovich, "The performance of steam sepa-
rators at power units containing RBMK reactors," At. Energ., 53, No. 2, 70-74 (1982).
2. O. Yu. Novosel'skii, V. B. Karasev, E. V. Sakovich, et al., "E~tperience with operating and modifying
separator drums in the first unit at Kursk nuclear power station," in: Nuclear Power Stations [in Rus-
sian], Issue 3, ~nergiya, Moscow (1980), pp. 101 105.
3. A. G. Ageev, V. B. Karasev, I. T. Serov, and V. F. Titov, Separation Devices at Nuclear Power Stations
[in Russian], Energoizdat, Moscow (1982).
PECULIARITIES OF THE DISTRIBUTION OF PHASES
IN THE UPDRAFT SECTION OF A HOUSED
BOILING REACTOR
V . N. Fedulin, G.. G. Bartolomei, UDC 621.039.536.2
V. A. Solodkii, and V. E. Shmelev
In order to increase the driving pressure head of the coolant natural circulation loop, one can mount a
large-diameter updraft section above the active zone of a boiling reactor. The motion of a steam~.vater mixture
in such a section is characterized by a complicated distribution of phases, and consequently the steam content. .
Nonuniformity of the energy liberation in the active zone and the velocity field of the phases at the entrance to
the updraft section exerts a decisive influence on the structure of the two-phase flow.
At present there are no reliable recommendations on the calculation of the true volume steam content in
alarge=diameter channel, which is what the updraft section of a boiling reactor is. The value of the average
steam content in the updraft section of a VK-50 reactor calculated by the VTI procedure [1] is 4090 higher than
the experimental values obtained on the basis of the measurement of the hydrostatic pressure differential. The
main cause for this discrepancy consists, in our opinion, of the fact that the existing procedures for calculation
of the steam content are based on the data of experimental investigations obtained in channels of small cross
section and. at a rather large distance from the entrance to the channel (l/d > 20), i.e., upon the motion of a hy-
drodynamically stabilized two-phase flow in which no significant transformation of the phase velocities and the
steam content occurs from cross section to cross section in the channel.
The updraft section of boiling reactors is characterized by commensurable height and diameter dimen-
sions (H/D ~ 1). -This specifies incompleteness of hydrodynamic stabilization of the flow and. makes inadmissi-
ble the use for calculation of the true volume steam content in them of the same relationships as for hydrody-
namically stabilized flows. It is necessary for the improvement of the methods for calculation of the volume
steam content in large-diameter channels to use data on the structure of hydrodynamically unstabilized two-
phase .flows. This is explained by the fact that one-dimensional computational models cannot take account of all
the peculiarities of the spatial distribution of the phases.
An investigation of the structure of a two-phase flow in alarge-diameter updraft section (D = 2 m, H = 3
m) using the electroprobing method was carried out on a VK-50 boiling reactor. The structure of the reactor,
its operating regimes, and the layout of the placement of local steam content sensors have been described in
[2, 3], and the procedure for analysis of the experimental results i.s discussed in [4].
The following peculiarities in the structure of a two-phase flow and in the hydrodynamics of the updraft
section have been revealed in the analysis of the results:
1. The radial distribution of the steam content cp at the entrance to the updraft section is determined by
the complex influence of the energy liberation q and the coolant flow rate W across the .radius of the active zone
Translated from Atomnaya Energiya, Vol. 57, No. 6, pp. 385-388, December, 1984. Original article sub-
mitted March 11, 1984.
0038-531X/84/5706-0811$08.50 ?1985 Plenum Publishing Corporation
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zo t~
~ 0,2 0,6
4~
Mw
2s
~-o.zs~
Fig. 1. Distribution of the energy liberation (O), steam
content (O), and coolant velocity (^) across the radius of
the active zone for a pressure of 2.0 MPa in the reactor
and a thermal power of 80 MW.
i ~ i
7,0 74 7B
Fig. 2. Peculiarities of the s-team content distribution with height of the updraft sec-
tion for (a) p = 6.0 MPa and N = 170 MW and (b) p = 3.2 MPa and N = 150 MW: O,
x, ^, and O) R =0.92, 0.74, 0.56, and 0.37 m, respectively.
R and, in connection withthis,is of a sharply expressed nonuniform nature (Fig. 1). In particular, the steam
content at the exit of the heat-generating assemblies (HGA) depends significantly on the position of the com-
pensating rods (CR) in the active zone. Upon a rise in the reactor power, when the peripheral CR are essential-
ly extracted, an increased steam content is: noted above the HGA of the peripheral part of the active zone. The
maximum of the steam content is shifted towards the center upon the subsequent extraction of the central CR.
2. The structure of the two-phase flow in the lower part of the updraft section, which is adjacent to the
active zone, is determined to an appreciable extent by the jet nature of the coolant outflow from the HGA heads.
To a large extent the jet nature of the outflow appears above the central HGA, and to a lesser extent above the
peripheral HGA. The height of the jets depends on the velocity of the exiting two-phase flow, the working pres-
sure p, and thehydrodyna.micsof the updraft section. For the VK-50 reactor the height of the jets does not ex-
ceed 0.4 m.
The jet nature of the outflow of the two-phase flow is exhibited in: a corresponding variation of the steam
content above the HGA heads. A maximum steam content is noted on axial lines of the jets. In power regimes
(p = 6.0 MPa) with a supply of feed water to the "cold"-drop section (i.e., under the central part of the active
zone) the following effect of jet outflow of coolant from the central HGA has been noted: The steam content in
the jet is higher at a distance of 280 mm from the HGA heads than at adistance of 80 mm (Fig. 2a). This ef-
fect is evidently explained by a "pressing down" of the steam-water flow, an increase in the rate of its move-
ment, and a migration.of the steam of its periphery to the jet axis. The jet effect disappears at a large height,
which indicates a local equalization of the steam content.
At a low reactor power (< 100 MW), a lowered working pressure, and a change in the scheme of the feed-
water supply the steam content in the jet at a distance of 280 mm from the HGA does not increase-(Fig. 2b).
This can be explained by a decrease in the velocity of the steam-water flow at the exit from the HGA and a
breakdown of the jets at a lower level. This effect has not been detected in all regimes in the peripheral part
of the updraft section, which may be caused by a deflection of the steam-water jets toward the center due to
recirculation flows.
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~i
~o
~s n, ~s g ~s > ~s z, 2s y,
Fig. 3. Distribution of the steam content with
height in the updraft section: O, O, ^) p = 6.5
MPaandN=180 MW; ~)p=6.4MPaandN=
181 MW; ~) p = 6.4 MPa and N = 180 MW; ~, ?,
^)p= 6.0 MPa and N= 170 MW; O,~)R= 0.74
~m; ^,~)R= 0.37 m; and O, ?,~,~)R=0.92 m.
m.
3. On the whole, the steam content distribution over the volume of the updraft section appears as follows.
Near the axial line of the updraft section a .coolant flow saturated with steam is formed with the maximum steam
content at a distance of about 1-1.5 m from the active zone. The steam content decreases towards the output
cross section of the updraft section. The steam content on the periphery of the updraft section first decreases
sharply with height, reaching a minimum value at a height of 0.8 m, and then it gradually increases (Fig. 3).
Complete equalization of the steam content in the output cross section of the updraft section is not attained [4] .
4. It has been established on the basis of hydrodynamic measurements that the pressure differential on
the periphery of the updraft section is higher than in its central region. In particular, the following values of
the pressure differentials are obtained at a power of 168 MW with a pressure of 6.0 MPa: ' on the periphery,
.12.8 kPa, and in `the center; 11.22 kPa. Consequently, a definite radial pressure gradient exists in the updraft
section:~which acts on the corresponding radial flows of the coolant.-
Taking into account that in the output cross section of the updraft section equalization of the steam con- ,
tent (although incomplete) has been noted, one can assume in the first. approximation that the static pressure on
this cross section is constant. Then one can conclude that the maximum radial pressure gradient is observed
in the lower part of,the updraft section. An intensive migration of steam from the- periphery of the updraft
section to its central part has been noted precisely in this zone.
5. The 'values of the steam content averaged over the height of the updraft section which are calculated
on the basis of the measured pressure differentials are related to each other and the calculated value as fol-
lows:
where ~Pc and Wp are the experimental values of the steam content in the central and peripheral regions of the
updraft section and ~calc is the average value of the steam content calculated by the procedure of [1]. This re-
sult shows that the volume steam content in the updraft section above the central HGA is close to the calculated
value and decreases significantly towards the periphery.
6. The nonuniformity of the steam content distribution in the updraft section determines the effectiveness
of the individual updraft tubes mounted on the HGA heads. The isolated individual updraft tubes on the VK-50
reactor (height, 1.5 m, diameter, 150 mm) are mounted above different rows of HGA of the active zone. The
investigations have shown that such mounting increases the coolant circulation velocity in the fifth (the peri-
phery) row of HGA by 0.2 m/sec and in the fourth row by 0.1 m/sec. No increase has been noted in the coolant
circulation velocity in the third row of HGA, which is determined by the rather high steam content above the
HGA in question without installation of an individual updraft tube.
The experimental data obtained permit suggesting the following qualitative physical model of two-phase
flow in the updraft section (Fig. 4). A jet outflow of a steam-water mixture exists at the exit from the HGA of
the active zone. Due to the enhanced output .of steam from the central HGA and the existence of a radial pres-
sure gradient the steam-water mixture from the periphery of the updraft section is directed into its central
region. In this connection an accelerated steam-water flow with an enhanced steam content is formed in this
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Fig. 4. Scheme of the flow of the steam- water mix-
ture in the updraft section of the VK-50 reactor.
region. Upon the attainment of a specified height a gradual widening of the steam-water flow begins, which
leads to equalization of the steam content over the cross section of the updraft section. A recirculation flow of
coolant which deflects the steam-water jets from the peripheral HGA to the center is formed on the periphery
of the updraft section in the zone of low steam contents.
The proposed physical model of the flow of a steam-water mixture and the experimental data obtained on
the distribution of the steam content permit calculating more accuratelythe hydrodynamics of the updraft section
of boiling reactors.
LITERATURE CITED
1. A.Ya.KramerovandYa.V.Shevelev,EngineeringCalculationsofNuclear Reactors [in Russian], Atomizdat,
Moscow (1964), pp. 320-322.
2. G. G. Bartolomei, V. A. Solodkii, V. N. Fedulin, et al., "Determination of the steam content in boiling
reactors by the electroprobe method," Teploenergetika, No. 10, 15 (1980).
3. V. A. Solodkii, G. G. Bartolomei, V. N. Fedulin, et al., "An investigation of the hydrodynamics of an adia-
batic two-phase flow in the updraft section of a VK-50 reactor," Teploenergetika, No. 6, 73 (1981).
4. V. N. Fedulin, V. E. Shmelev, V. A. Solodkii, and G. G. Bartolomei, "The steam content of a two-phase
flow in the updraft section of a VK-50 boiling reactor," Preprint NIIAR-33 (598), Dmitrovgrad (1983).
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EFFECTS OF STEAM GENERATOR SECTIONING ON
THE RELIABILITY OF A NUCLEAR POWER
STATION CONTAINING A FAST REACTOR
A. I. Klemin, O. B. Samoilov,
and E . V . Frolov
One must provide adequate reliability in the individual units and in the station as a whole in order to meet
current specifications for the economic performance and safety in nuclear power stations. In the case of sodi-
um-cooled fast reactors, a current problem is to improve the reliability in the steam generators (SG) of sodi-
um-watertype. Various Soviet and foreign designs for such reactors use sectional SG, which raise the reli-
ability and safety of the station because they localize possible faults caused by leaks in the SG heat-exchanger
tubes to a single steam-generating section [1] .The SG sectioning makes it necessary to incorporate factors in-
to the reliability model such as the SG repair strategy, the number of steam- generating sections in one cooling
loop, and the working constraints on the available* unit power. We give below a model that has been used to ex-
amine the effects of these factors on the reliability of afast-reactor system.
Model for the Structural Reliability of a Nuclear Power Station Unit Containing a Sectional SG. We con-
sider asimplified structural scheme for the unit with one turbine and a fast reactor having m cooling loops;
each of which includes pumps in the first and second loops, an intermediate heat exchanger, and a sectional SG
(Fig: 1). The latter-consists of a large number of components, including tube bundles, module bodies, steam-
water chambers, sodium, steam, and water pipes, and cutoff equipment for the sodium and water (steam). The
reliabilitycalculationsarebased on two types of enlarged element in the sectional SG (Fig. 1): the element 5,
which includes components whose failure makes it necessary to disconnect the entire SG and run down the cor-
responding cooling loop (for example, the sodium-handling gear), and element 6 (steam-generating section),
which combines components whose failure leads to the disconnection of a separate section and corresponding
reduction in the power of the given SG (tube bundle, module body, and seals on the covers on the steam-water
chambers). Clearly, each of the elements 5 and 6 consists of corresponding SG components in series connection
in the sense of reliability.
The reliability of the unit is most fully characterized by the power availability factor gip, which is calcu-
lated from the following formula for the basic mode of operation:
~W
~= 1 ~ Na (i) dT, (1)
t ~?
0
where Na(T) is the mean available power from the unit (as a fraction of the nominal power) at times TEty? while
tv, is the relevant working period.
A basic stage in constructing a model for power station reliability is to determine the dependence of Na .
on the state y, which in turn is dependent on the states of various items. The basic principles have been con-
sidered in [2]; according to which Na(y) is defined by means of special characteristics that are functions of the
states of the individual units or sets of equipment .(for example, cooling loop), which are subsequently called
component parts. The state function for a component part characterizes the contribution to the station power
(as a fraction of the nominal power) in a given state of the component part when the rest of the power station
unit is completely free from faults. The state of element l (equipment unit) is defined by means of the binary
variable x~, which takes the value 0 in fault free operation and 1 when the element fails.
The state functions gl(xl), l = 1, 7 for the equipment in this station are defined as follows:
*The maximum power that the unit can provide with all items fully viable.
Translated from Atomnaya Energiya, Vol. 57, No. 6, pp? 388-393, December, 1984. Original article sub-
mitted August 9, 1982; revision submitted November 15, ] 983.
0038-531X/84/5706- 0815$08.50 ?1985 Plenum Publishing Corporation 815
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Fig. 1. Simplified structural scheme for a unit in
a nuclear power station containing a fast reactor;
1) core; 2) first-circuit pump; 3) intermediate
heat exchanger; 4) second- circuit pump; 5) en-
larged element including the SG sodium equip-
ment; 6) SG section; 7) turbine generator.
1-xl
~It(xl)= m (1-xi) l
m \ 1 nl /
R
where ns is the number of steam-generator sections in a cooling loop. The values of the subscript l of 1 and 7
correspond to the core and the turbine generator, while the values of l of Z-6 correspond to the pump in the
first circuit, the intermediate heat exchanger, the pump in the second circuit, and the enlarged elements of the
SG, correspondingly.
As the sections are indistinguishable from the viewpoint of reliability (the elements 6), the state function
qs(j) for a sectional SG and a separate cooling loop are defined by
plc (1) _ ~ m l ns
l O+ if 1=7cr+ ne+
where j is the number of failed sections in the SG for a cooling loop and jcr is the critical number of SG sec-
tions whose failure causes the operation of the corresponding loop to be halted.
The state function for cooling loop j (i = 1, m) takes the following form because of the serial structure:
II1t = Illlil lI]9 ('1'^)+ ~13 (~'a)+ Q4 (xa)+ Q5 (xr,)+ qR ()i)}.
We use the state functions for the equipment in the cooling loop to represent the state vector for a unit in
enlarged form:
If the reactor is operated with SG sections switched out (for overhaul), there may be differences in sodi-
um temperature in the first circuit at the entry to the core resulting from the loops with different numbers of
fault-free (or failed) SG sections. When certain limiting values of the temperature differences are reached,
one usually has to impose constraints on the available power. To characterize such working constraints, we
introduce the parameter o(y) for the state vector y, which is equal to the maximum difference between the
power values (in relative units) for any two cooling loops:
~(!1)=nl{ nias ~V',l- nli^ [il"z]}, ~;t>U. (6)
We note that in (6) we envisage loops for which the values of ~i are different from zero, since otherwise
the corresponding loop is taken out of service.
The constraints on the available power are determined by the limiting permissible value [~ of o(y), which
canbe estimated as follows:
aR
OR= n ~
where do is the maximum permissible difference in the numbers of fault-free (failed) SG sections in any two
operating loops. The value of do is defined from heat-engineering and strength calculations.
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In accordance with (5), the desired Na(y) for the unit is
where Q is the state function for the system of m cooling loops:
a~ =min {~~, ~b-I- e? 1 ~;
~,~
b=min [~t1; i=1, na, 11>z>0.
Expression (9) -has been derived on the assumption that if there is a difference in sodium temperatures in
the first circuit between any two cooling loops that exceeds the limiting permissible value, that difference is
reduced to the permissible value by switching out the necessary number of sections in the loop with the higher
power in order to provide the maximum unit power.
We now calculate the probabilities of realizing various states in the unit. We assume that planned pre-
ventative measures applied to the equipment are combined with reloading the reactor and form a periodic
sequence, where the instants at which they are performed are points at which the equipment viability is com-
pletely restored. Also, we restrict consideration to the integral (tank) style for the equipment in the first cir-
cuit, which is very characteristic of fast reactors [1], where unplanned (emergency) repairs to the first-cir-
cuit pump and the intermediate heat exchanger make it essential to shut down the reactor. We subsequently as-
sume that the equipment in the first circuit (pumps and intermediate heat exchanger) is not repairable in the in-
terval [0, T] between planned preventative. repairs, while unplanned repairs are combined with the next planned
ones: Parts of 'the equipment that are. repairable in the interval between.planned" preventative repairs include
the ?core, the pumps in the second circuit, the SG elements, and the turbine generator. The probability that ele-
ment l? (Fig: 1) is fault=free at time T is given by the standard formula in the case of exponential distributions
for the time to failure and the recovery time:
Pr (ti) - Ni~-~i {?~ -~ die-~~+?~~s}, ti E I0, T ], (l = 1, 7) (10)
where ~l and ?l are the rates of failure and recovery for element 1, correspondingly. We note that (10) applies
also when the equipment is not repairable, when ?l = 0.
On the basis of the above assumptions about the repairs, (1) becomes"
T
~ ' L T , Na(ti)dti,(1-b), (11)
n
where T is the interval between planned preventative repairs and 8 is the mean proportion of station shutdowns
related to planned preventative repairs and' unplanned repairs combined with them for equipment in the first
circuit not repairable in the interval [0, T] (pumps and intermediate heat exchanger).
We consider two types of repair strategy for a sectional SG in an individual loop:
1) strategy Cl, in-which an SG section that' fails is disconnected and repaired while the other sections
operate; and
2) strategy C2, where when a certain (critical) number of sections fail one switches out the correspond-
ing cooling loop in order to repair the SG.
.Within the framework of strategies C1 and Cz there is also the particular case where the repair of failed
SG sections is combined with. the next planned preventative repair to the unit.
Let gj(T) be the probability that at time TE(0, T] the sectional SG in a loop has j failed sections. For re=
pair strategy C1, the probability gj(r) is
~; (i) = C;,l.~s.4_; (i)11 - Pa (i)1', TE [0, Tl, (12)
where C~ is a binomial coefficient and Ps(T) is-the probability of a viable state in the steam-generator section
ns
.at time ,z, which is calculated from (10) . .
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For strategy C2, the operation of the sectional SG is described by a Markov random
determined as the solution of a system of differential equations:
dQ~ (T)
dei (i)_
dT -ai-i6'i-i(i)-a;g;("r), 1=i+1cr-i;
icr,- t
~ r=n
where ? is the rate of SG recovery as a whole (with allowance for the time consumed in preparatory operations
in disconnecting and connecting the loop), ?j = (ns - j)7~, and 7~ is the failure rate in SG sections.
According to (2)-(4), the state function ~i for loop i takes jcr+1 discrete values 0,...,gs(j), ...,1/m. The
probability that ~i is equal to one of these discrete values at an instant TE [0, T] is defined as follows:
Pt{Y~t=40(1)}=~ icr-t
l 1 -z. ~--o
5
gi (i) l~ Pt (i), if 'fi't = 9s (1) > 0;
i=^_
where gj(T) is calculated from (12) and (13) correspondingly for strategies C1 and C2.
One way of raising the reliability of a unit having sectional SG is to employ backup steam-generating .sec-
tions. Let the SG in each loop have r backup sections, i.e., ns= ns+ r, where ns is the minimum necessary
number of working sections in the SG for an individual loop. We assume that the backup sections work in hot
backup, and the loop is shut down for repairing the SG when jo = r + 1 sections fail (we envisage only strategy
,CZ). Clearly, the SG state function qs(j) and that for the cooling loop ry take two discrete values: 1/m and 0.
The probability of realizing these states for loop i is
6 r
1 =2 7==0
5 r
1=2 7=0
where gj (T) is calculated as the solution to (13) for
1cr=10=r~-1,
Formulas (10) and (12)-(15) enable one to calculate the probability of realizing state
process, and gj(T) is
P: (y) _ ~'i (ti) 1'~ (T) i11i P,, {,pi _ 9s (li)}+
where ji is the number of failed SG sections in loop i, and the probability P~{~i = ?0 (Ii)}is defined by (14).
The mean available unit power is
!SEE
where E is the set of unit states for which Na(y) > 0.
Dependence of cp on Various Factors. The .effect of using a sectional SG to improve unit reliability is
dependent on various factors: the number of the sections and the reliability characteristics in the SG elements,
the type of repair strategy, and the mode constraints on the available unit power. In what follows we consider
the effects of these factors on the power use factor rP for a unit in a nuclear power station containing a fast re-
actor, having four cooling loops with sectional SG. The- calculations are based on the above model, which has
been implemented as a computer program.. Table 1 gives the equipment reliability parameters used in the cal-
culations. The values have been taken on the basis of operating- experience with the corresponding equipment
or data on the reliability of similar equipment (these may be considered as the authors' approximate estimates).
The failure rates are taken as constant, which is usually correct for the period of normal operation, when the
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 Declassified and Approved For Release 2013/02/22 :CIA-RDP10-021968000300050006-1
TABLE 1. Equipment Fault-Free Parameters
and Repairability
Fault
rate, h-t
Recoveryt
rate, h'
First-circuit pump
3,7.1(:-
Second-circuit pump
2
2.10-5
Intermediate heat ex-
,
1
4.10-~
changer .
Steam generator(r~=16):
element 5
,
1,8.10-'~
fi
7.10-g
element 6 (SG section) 4,5.10-~
,
running-in period has ended but the period of rapid aging has not begun. To simplify the calculations, the
reliability in the core and turbine generator has been taken as substantially higher than that in the other equip-
ment (Table 1) .
We estimated the failure rate for element 5 (Fig. 1), which combines elements in the sectional SG whose
failure leads to the entire SG and the corresponding loop being shut down, on the assumption that the main con-
tribution comes from the sodium-handling equipment (failures of leak type). The failure rate for element 5 was
estimated from
where y is the number of sodium-handling items per SG section and ~a is the rate of failure of leak type for the
sodium-handling- equipment (~a~ 10-7 1/h).
We also considered types of SG with different degrees of sectioning (ns = 16 and ns = 4), where we incor-
porated the dimensional effect, which is that the fault-free level of a section decreases as the power increases,
which isdueto'variations in various factors (heat-transfer surface, number of tubes, and number of welded
joints), all of which influence the fault experience in an SG section [3]. We assumed that this scale effect may
be approximated by a power function of the form
~r
?~a=fir,(','-.< 1 ~ (19)
ns 1
where 7~ and ~ are the- failure rates in an individual section for an SG with the numbers of sections (per loop)
ns and ns, correspondingly, while R = const is the parameter of the scale effect. The calculations were per-
formed, for R= 1 and R = 0.5.
The recovery rate ? for the SG with repair strategy CZ was determined from data on the repairability of
SG sections (Tabled) and the mean times involved in the preparatory operations, which was taken as about 70 h.
Table 2 gives calculations on cp for a system containing a fast reactor in the interval between reloading
operations (planned preventative repair) t~ 3000 h with the proportion of planned shutdown time b ~ 0.127. The
values of ~P in Table 2 have been derived for do = ns and ~ = 1, i.e., on the assumption that there are no mode
constraints on the available power due to possible temperature differences. It is evident that strategy C1 pro-
vides higher values of rp than does C2, which occurs because it most fully uses the advantages of the sectional
SG structure. The advantage thereby gained is largely determined by the degree of SG sectioning, as well as
the scale effect, and the maximum values occur for an SG with a small number of sections (ns = 4) when
reliability is low (R = 1). The calculations also showed that when the section reliability is high (1/7~ > 7?
104 h), the type of SG repair strategy has virtually no effect on cp.
It should be noted that implementing strategy C1 involves technical difficulties of the shutoff equipment,
which has to enable one to repair a failed section while the other sections of the SG in the loop are operating.
Therefore, one usually employs strategy CZ in practice. We considered two particular cases: jcr = 1 and jcr =
ns. In the first case, failure in any section leads to the corresponding loop being shut down for SG repair,
while in the second case the loop is not shut down when sections fail (one merely shuts down the failed sections)
and the repair is combined with the next reactor reloading. Table 2 shows that the latter strategy (jcr = ns) is
preferable for an SG with a large number of sections (ns = 16) since it provides a substantial increase (about
109'x) in the equipment reliability parameter. We note that this repair strategy, with failed sections shut down,
is possible if a certain specification for the level of repairability in the SG sections is met: .The mean.. .
section repair (replacement) time should not exceed the reactor reloading time.
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 Declassified and Approved For Release 2013/02/22 :CIA-RDP10-021968000300050006-1
TABLE 2. Values of cp in Relation to Repair
Strategy and Degree of SG Sectioning
Steam gen-
erator form
R=0,5
tta=4 R=1
0,710 I n,71`6
11,644 _ U,62Ei
TABLE 3. Values of Win Relation to Mode
Constraints
Steam ggen-
~
I do-,~F
erator {orm
I `I0 -
I
1 `'u
= ~
I
~
~o = s
cno - 1)
ns = 1('i
I (1,73
4 ( 11,
74fi
I
1),7511
I (1,751
=4 N=11
n
'
5
11,(iA
1 I
11,
7114
I
(1,7116
I 11,7116
y
R=1
U,:~f3
5
0,
620
1,,626
11,626
Two factors determine the effects of SG sectioning on cp: the scale effect for the reliability parameters
of a section, which is due to the change in unit power, and the scheme effect, which is related to the change in
the number of SG sections in the loop. Also, according to (18) any change in the number of SG sections leads
to a change in the amount of shutdown equipment in the loop and to a corresponding change in the reliability
characteristics of the enlarged element 5 in the SG (Fig. 1). The calculations showed (Table 2) that enlarging
the SG sections reduces cp for both repair. strategies. As would be expected, the reduction in cp is more sub-
stantial for strategy CZ when there is a substantial scale effect (R = 1). Also, when the number of SG sections
is small (ns = 4) and the section reliability is low (R = 1), there isapronounced optimum for gyp, which is at-
tained if one switches out a loop for SG repair when two sections fail (jcr = 2), Clearly, this optimum is de-
pendent on the degree of SG sectioning, the fault experience level,. and the repairability, as well as on the value
of the interval [0, t] between reactor reloads (planned preventative repair), and the value in each particular
case can be determined from a more detailed analysis.
We now consider the dependence of cp on the available-power constraints due to the possibility of the above
temperature differences. The calculations were performed for wide ranges in do and ~, which determine
these constraints. The limiting value do = ns (moo = 1) corresponds to the case where there are no constraints
of this type, while do = 1 corresponds to the case where the latter are considerable. The effects of the con-
straints on cp are slight for strategy C1. Table 3 gives calculations for these forms of SG with strategy C2 (jcr =
ns) for t = 3 ?103 hand a = 0.127.
Table 3 shows that the constraints may 'reduce W if the SG has a small number of sections because these
constraints begin to make themselves felt as shutdown states with small numbers of failed SG sections as the
unit power of an SG section increases. Because of the scale effect, the probability of such states is fairly high,
and therefore the effects of the constraints (or of do and ~) on cp are most pronounced.
This model was also used in analyzing the scope for increasing the reliability of a system containing a
fast reactor by using r backup sections in each cooling loop. It was assumed that the backup sections operate in
hot backup, while a loop is shut down for repair of the corresponding SG when r + 1 sections fail. For the SG
with ns = 16 working sections, we obtained the following values of cp (t = 3.103 h, S = 0.127):
0,659, if r = 0;
W = 0.742, if r = 1;
0,773, if r=L,
where the case r = 0 corresponds to strategy CZ (jcr = 1)?
-These results showed that backup sections in the SG provide an effective means of raising the system
reliability parameters. The largest cp is attained when the number of backup sections is small (r = 1 or 2).. A
real SG design always has some spare power margin, which canbe considered as hot backup in relation to the
Declassified and Approved For Release 2013/02/22 :CIA-RDP10-021968000300050006-1
c1
trategy
0,789
0,774
n,72s
f'I,fi60
0,724
Q 675
 Declassified and Approved For Release 2013/02/22 :CIA-RDP10-021968000300050006-1
~ ;~-__
nominal SG power. Therefore, the above advantages due to section backup can be realized in practice for ex-
ample as follows: When one SG section fails, the loop continues to operate at its nominal power because the
power levels in the fault-free sections are raised. ` ~~
LITERATURE CITED
1. O. L. Kazachkovskii et al., At. Energ., 43, No. 5, 343 (1977).
2. R. A. Peskov and E. V. Frolov, Aspects of Nuclear Science and Engineering: Series Nuclear Reactor
Physics and Engineering [in Russian], Issue 1 (10) (1980), p. 29.
3. H. Procaccia, Reliability of Nuclear Power Plants, lAEA, Vienna (1975), p. 351.
STATISTICAL ANALYSIS 'OF REACTOR THERMAL POWER
BY THE USE OF THERMAL AND RADIATION METHODS
IN THE FIRST UNIT AT THE ARMENIAN NUCLEAR
POWER STATION
F. D. Barzali, L. N. Bogachek, V. V. Lysenko,
A. M. Muradyan, A. I. Musorin, A. I. Rymarenko,
I.V. Sokolova, and S. G. Tsypin
Avery important factor in providing safe operation in a nuclear power station unit is to determine the
thermal power of the reactor W with an error of 1 2q [1-4]. Operational methods have been proposed [1, 2, 4]
for;.measuring the thermal power by statistical computer calculation methods; in [3, 5], operational correlation
dispersion methods were considered for determining the thermal power by means of radiation meters (with an.
error of 1-2qo at the nominal power), these meters operating with the neutrons leaking from the reactor con-
tainment [power meters (PM)] or on the y radiation from 16N in the coolant [loop detector (LD)] . The data en-
tering these meters are used in determining discrete quantities of statistical nature, and this enables one to ob-
tain given errors.in the readings of the PM (NpM) and LD (NLD). Also., a major feature of a radiation meter
is that NpM and N)~ are strictly proportional to W, as has been confirmed by experiment [3].
In view of these advantages, it is of interest to perform a combined statistical analysis of the data ob-
tained by these methods and those provided by the standard thermal detectors (TD) in order to define the reac-
tor thermal power more accurately and to determine the errors.
Table 1 gives data for one-factor variance analysis, which have been obtained with the first unit at the
Armenian nuclear power station (the ratio of the thermal reactor power measuredby the'TD and expressed as a
percentage of the nominal power to the readings of the PM in count/sec). This approach supplements traditional
analysis of measurement methods in enabling one to analyze results corresponding to several levels of thermal
power (which may differ substantially). The dots in ~V.j and -Wi, indicate averaging correspondingly with re-
spect to the subscripts i and j .
The thermal power was determined by the following thermal methods:
1) i = 1, from the sum of the products of the differences of the enthalpies for the hot and cold .parts of the
loop by the coolant. flow -rates in the corresponding main circulation loops;
2) i = 2, from the product of -the mean difference in enthalpies. of the hot and cold parts of the loop by the
total coolant flow through the reactor;
3) i = 3, from the thermophysical_ coolant parameters in the second circuit beyond the feed pumps;
4) i = 4, from the steam in the steam generators;
5) i = 5, from the feedwater in the steam generators;
Translated from Atomnaya Energiya, Vol. 57, No. 6, pp. 393-39~, December, 1984. Original article sub-
mitted September 5, 1983.
0038-531.X/84/5706-0821$08.50 ?1985 Plenum Publishing Corporation '821
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 Declassified and Approved For Release 2013/02/22 :CIA-RDP10-021968000300050006-1
TABLE 1. Initial Data for One-Factor Variance Analysis of the Ratio Wij/NpMj of the Ther-
mal Detector Readings on the Reactor Power to the PM Readings
Number of measure-
ment method and
statistical-character-
istic estimator for ther-
mal power level j
W.i= 1 ~ Wii
INpMj i_1
Esflmate of the vari-
ance $~ ? 103
Weight of mean
i
8i~ S~
Nph1 ?, count/sec; W ?, %
7 ~
Mean thermal
power for method
Varianceesttmator
ofineasurementi
f
:1.,93
70,911
ri9,55
93,13
T w
~~
[ili
W~,=
or measurement
method
$i'10S
4s.1;7
cs,7s
sc.7z
92,72
N
J
i=t PMj
0,823f_~
0,9317
0,9485
0,9787
0,9205
4,6'111
0,8639
(1.9647
0,9749
1,0037
0,9518
3,7075
0,8951
1,0333
1,0172
1,0336
0,9948
4,4796
0,819(1
0,9234
0,9174
0,9685
0,9071
3,970(1
0,7849
0,9306
0,9358
0,9417
0,8982
5,8304
0,9561
0,9942
0,9903
1,0254
0,9915
8,0496.10-1
1,0137
1,0127
0,9949
1,0171
1,0096
9,8724.10-9
0,8794
0,9701
0,9685
0,9955
-
-
6,6924
1,9381
1,3776
1,1306
-
-
149,4232
515,9692
782,7176
884,4861
-
-
*Here i is the number of the method of measuring the thermal power, 1 ~i ~ I-7 while j is the
thermal power level 1 < j ~ J = 4.
6) i = 6, from the coolant parameters in the second circuit in the turbine control-stage chambers; and
7) i = 7, from the coolant parameters in the second circuit after the high-pressure heaters.
The variance analysis is performed with the method of measuriig the thermal power as the factor. The
Bartlett test [6] and the Cochran test [Z] enable one to determine whether there is equality in the variances?in
the data groups .
The Bartlett test is based on comparing the calculated value of XZ with the Xtab distribution having. I - 1
degrees of freedom. In that case, with probability 1 -c? = 0.95, Xtab (I - 1 = 6) = 12.592 > ~: _ 9.338, so the
variance series is homogeneous.
The Cochran test is based bn considering the quantity
u si
t=t
The value found G~ =max{Gi} =.0.245 is less than the tabulated one Gtab (I = 7, J - 1= 3) = 0.48 for 1 -
~ = 0.95, i.e., the series of variances Si is homogeneous.
We test the hypothesis Ho : wi ? = W2. = Ws ? _ ... = W7?? Adopting the hypothesis corresponds to the method
of measuring the thermal power not being significant as a factor. The quantities needed to .test the hypothesis
are given in Table 2. -
We also used the Fisher test [8] ,whose statistic is the ratio F = SB/S$. If F < Fib, the hypothesis is
accepted with probability 1 - a, where Ftab is the value of the F distribution with vB and "R degrees of freedom.
`Table 2 shows that F < Fts,b, i.e., hypothesis Ho is true with probability 1 - a = 0.95.
As the method of measuring the thermal power is not significant as a factor, the overall mean, which is
an estimator for the true value of a ratio, is [6] given by
W4.
i=1
I
The estimator for the variance of the overall mean is given by [6]
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 Declassified and Approved For Release 2013/02/22 :CIA-RDP10-021968000300050006-1
TABLE 2 . Quantities Needed to Test the Hypothesis Ho: Wl? = Wz ? _ ? ? ? = W7
Variance
Sum of the squares of
Degree of
3
F ratio (Fisher
I
source
the difference
I
freedom
Me
I
an square, x 10
test)
t _ _
J
z
I
~, (Wi.-iV)z
to
Sz
I;
Between
J~ (Wi.-W)z=0,0503
vB=I-1=6
B
S
=
=5,3855
F=
=2,507
ga
levels
i=t
r
yp
J
R
,
Within
-W
0702
)z=0
~, ~~ (W
v
=I7- I=21
\\
,,
i=t
,
``
i=t =3
46
(6
2!)=2
57
F
levels
ii
i.
,
i=t i=1
R
Sit=
,
vR
,
,
tab
I J
(W;;=w)z
'kFor simplicity, the symbol Wij/NpMj is replaced by Wij.
1 )
sw=lJ(iJ-l)L(.1-1) ~ Sep-T~' (W~-iF)2~~
=t i=t
J
where Si = J 1 1 ~ (W,i - W,.)~; Sw = 7.38.10-5.
i=1
Further, we use the same data in aone-factor variance analysis now assuming that the PM reading is the
factor, or, which is the same, the reactor thermal power level.
We checked 'the variances for equality on the Bartlett and Cochran tests, which showed that with 1 -a =
0..95,_ Xz = 6.3 Gtab (J = 4, I - 1 = 6) = 0.6.
In the case of the more searching Cochran test, we assume that the variance series is inhomogeneous.
On the Fisher test, the hypothesis Ho: W.1= W+z = W.s = Wa is rejected, i.e., the thermal power level factor
is ?ignificants
SN = O.U1811, JH . U.U027(i,
F=6.SG; Ftab (ti'i~=~3> ve?~~!')=3.01 andF>Ftab?
One.-also has to consider whether there is a monotone trend (bias). The check was made via the Abby
test [6], which showed that there was no systematic bias.
Therefore, as the variance series is inhomogeneous here, the overall mean is calculated from the follow-
ing formula [6]:
st_. = i I
J
~, s~
~t
./ J
-1 v~ nisi -~ LI k'i (W.i - Wa)3J = $.`~(i~)li ? 11)-''.
I
i =t i=1
Table 1 shows that the data group corresponding to the 45.790 thermal power level has a variance devia-
ting from homogeneity. In fact, if we exclude this level, we find that the Bartlett and Cochran tests. are obeyed:
Xz = 0.462 < Xt b (J -1 = 2) = 5-.991; Gm~ = 0.45 < Gtab (J = 3; I -1 = 6) = 0.68, and the hypothesis Ho ~ W.t =
W?z = W?s is a~opted, since F= 0.12 < Ftab ("B ? 2; vR = 18) = 3.55.
As the overall mean we take W = 0.97802 as calculated from (1) with the variance W = -6.9787 ?10-5 de-
rived from (2).
There is agreement between Wv (for four levels) and W (for three levels) within limits of 0.59'0, which
means that any of the methods can be used to determine the overall mean to represent the data.
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 Declassified and Approved For Release 2013/02/22 :CIA-RDP10-021968000300050006-1
TABLE 3. Initial Data for One-Factor Variance
Analysis of the Ratio (NLD j/NPMj) of the Readings
of the LD and PM*
LD number and
statisticalcharac-
teristics at ther-
malpower level j
5
(i
7
8
Arithmetic mean
r
N.; _~ ~ nLDij
PM i'
i=- t ~
Variancezestima-
tor On Sj
,s,c;
I cs,;a
,r,~s I
a~,~s
98,78513
'51,11217!)
21,21145(1
21,i352U4
16,9I12u6
15,415n:S
15,76677
1!),3974
99,67.5165
'L1 ,51Lt79
21,911823
21,7:3/i:iL
18,1!18(17
19,3118!14
1!),32148
19,711fi1i!I
18,4738
1!),11873
18, 79!)37
1!1,4758
16,86981
97,13:117
17,6121114
17,1','5577
21,(144:58
3L,4:;;r7:{
_"2,7!1485
2L,~~7595
98,965112
19,8117(16
3(1,111x92
211,73:3.97
18,71121153
19, 8!d189fi
19,9571126
i
211,3'1548
*Here 1 < i s 8 is the LD number and 1 s j ~4 is the
reactor thermal power level.
The thermal power level factor is significant in analyzing four levels but is not significant for three,
which indicates considerable error in measuring low thermal power either with TD or by radiation methods
since the data set is represented by the ratio Wij/NPMj?
To elucidate what influences the error in measuring the relative thermal power, we checked the readings
of the PM and LD against one another when they were working simultaneously in the first unit at the Armenian
nuclear power station, i.e., we performed aone-factor variance analysis of the data given in Table 3.
On testing the hypothesis Ho: N,1 = N,2 = N,3 = N.4, we found that the thermal .power level factor is not
significant. In fact SB = 3.9696, SR = 2.8841, while F = 1.38 < Ftab (vB = 3; vR = 28) = 2.95. Hypothesis Ho
is adopted with probability 1 - a = 0.95, so the significance of this factor for the previous. set of data with four
power levels is due to the larger error in measurement by means of the TD at power levels below the nominal
value.
This was confirmed by regression analysis in deriving the dependence of Wi on the readings NPM i [3].
According to this evaluation, the confidence limit with probability 1 -0!/2 = 0.975 is 5`~o for a thermal power of
about 50oI~ nominal, while the limit is 2.59x', for the thermal power determined from the dependence of NLD i on
NPM i
Therefore, combined statistical analysis of the thermal power determined by the standard thermaldetec-
tors and by radiation methods firstly enables one to calculate the means and the errors of these independently
of the thermal power level and secondly enables one to check radiation meters against one another during
operation.
I
1. V. A. Voznesenskii, "Some results from operating nuclear power stations containing WER-440 reactors,"
At. Energ., 44, No. 4, 294 (1976).
2. F. Ya. Ovchinnikov, L. I. Golubev, V. D. Dobrynin, et al., Working Conditions in Pressurized-Water Power
Reactors (in Russian], Atomizdat, Moscow (1979). !
3. L. N. Bogachek, K. A. Gazaryan, and V. V. Lysenko, "A correlation-variance method of measuring VVER
power," At. Energ., 50, No. 6, 420-422 (1981).
4. E. I. Igna.tenko, V. V: Zverkov, and V. N: Dement'ev, "Computer use in calculating reactor thermal- power,"
Elektricheskie Stantsii, No. 2, 9-12 (1982).
5. L. N. Bogachek, A . L. Egorov, V . V . Lysenko, et al., "Measuring coolant flow rate and power by radiation
methods in the .first unit at the Armenian nuclear power station," At. Energ., 46, No. 6, 390-393 (1979)..
6. K. P. Shirokov (ed.), Methods of Processing Observational Results from Measurments, Trudy Metrologi-
cheskikh Institutakh SSSR, Issue 134(194), Standartov, Moscow-Leningrad (1.972).
Declassified and Approved For Release 2013/02/22 :CIA-RDP10-021968000300050006-1
 Declassified and Approved For Release 2013/02/22 :CIA-RDP10-021968000300050006-1
7. A. S. Zazhigaev, A. A. Kish'yan, and Yu. I. Romanikov, Methods of Planning Physics Experiments and of
Processing the Results [in Russian], Atomizdat, Moscow (1978).
8. A. Afifi and S. Eisen, Statistical Analysis: a Computer Approach [Russian translation], Mir, Moscow
(1982).
TEST STAND FOR RESEARCH ON THE PHYSICS OF
HIGH-TEMPERATURE GAS-COOLED REACTORS
A.
M.
Bogomolov,
V. A. Zavorokhin,
A.
S.
Kaminskii,
S.
V.
Loboda,
A.
D.
Molodtsov,
V.
V.
Paramonov,
V.
M.
Talyzin, and
A.
V. Cherepanov
During the development of new reactors their physical characteristics are investigated on critical assem-
blies. High~emperature gas-cooled reactors (HTGR), under development in our country and abroad, have many
new aspects. The I.V. Kurchatov Institute of Atomic Energy has built a critical stand GROG which is used to
study the physics of HTGR, i.e., the physics that is common to all modifications and the specific physics tha
takes into account the features of different versions of HTGR; primarily the VG-400 [1].
The breadth of the studies of the HTGR physics on the GROG stand as well as economic and time con-
siderations have made it necessary to employ modeling principles, making it possible to create a simple and
flexible design that provides a capability for forming critical assemblies with different parameters quickly and
at low cost and performing a large number of experiments on them. Experiments on one-zone and two-zone as-
semblies can be carried out on the stand. In the two-zone assemblies the zone under study is formed of full-
scale reactor elements and the seed zone is formed of model elements and as a result the criticality of the as-
sembly of the whole is produced. This ensures wide variation of the neutron-physical characteristics of the
seed zone and their similarity to the system of full-scale elements as far as .properties are concerned. This.
similarity eliminates boundary effects and this makes the test zone more representative. The capability for
wide variation of the neutron-physical parameters in the critical assemblies using model elements and the
similarity of such assemblies to the reactors under study make it possible to solve many problems on model
systems. Therefore, in the initial stage, when the full-scale elements are absent, the physics of HTGR are
studied on critical assemblies formed of model elements. Since products of the. reaction of the fuel with neu-
trons are.difficult to use in assemblies, problems pertaining to the burn-up of fuel are also considered on sys-
tems containing model elements.
The necessary zones can be built from a combination of elements that are simple and convenient to use:
fuel elements (FE) of two types -with enrichment to 109'0 2ssU (at a uranium dioxide density of 0.5 g/cm3 or with
the dioxide of natural density at a density of 0.5 and 1 g/cm3), dummy moderating elements (DE) (graphite blocks
and inserts into them), and absorbing elements (AE).
The full-scale HTGR fuel elements use microfuel dispersed in a graphite matrix. The model fuel elements
of the stand GROG, universal physical imitators of HTGR fuel elements, are made in the form of a homogeneous
mixture of uranium dioxide and Teflon [2]. This composition of the fuel-element imitators allows them to be
prepared by means of a substantially simplified technology and at comparatively low cost.
The fuel-element imitators and graphite inserts are cylinders of diameter 50 mm and height 10, 25, or 50
mm. Graphite sleeves are also used to produce the necessary porosity. Continuous variation of the amount of
neutron-absorbing material in the assemblies is effected with the aid of a set of absorbers of diameter 50 mm,
made of boron-containing water or aluminum-boron alloy, as well as boron carbide rods of diameter 12 mm.
The necessary combination of cylindrical elements is placed in the channels of graphite blocks with a 250 x 250
mm cross section. The graphite blocks have nine 55-mm holes, located at the lattice points of a square lattice
with a pitch of 83.5 mm.
Translated from Atomnaya Energiya, Vol. 57, No. 6, pp. 397-400, December, 1984. Original article sub-
mitted February 21, 1984.
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~~-
Fig. 1. Critical stand GROG for research on the physics of HTGR:
1) spherical fuel elements of the zone under study; 2) fuel-assem-
blystorage; 3, 4) graphite blocks of the seed zone and reflector;
respectively; 5) experimental.channel; 6) rod of control and safety
' system; 7) oscillator.
The neutron-physical characteristics of fresh fuel elements are modeled on the critical assemblies of the
stand under the following conditions:
the model system has the same principal isotopes and the same ratios of their nuclear concentration as
does the full-scale system;
the model and full-scale systems approximate a homogeneous system for nonresonance neutrons (the
blocking in the fuel is less than 100);
the resonance effects in the model and full-scale systems are the same. The necessary ratio of the nu-
clear concentrations of the isotopes and the required resonance effects are attained by means of an appropriate
set and combination of fuel elements and graphite inserts.
The considered set of elements makes it possible also to simulate burn-up on the assemblies of the stand.
The products of the reaction of the fuel with neutrons are simulated by using zasU~ zsaU, and boron. The com-
bination of fuel elements and absorbing elements (fuel=assembly module) required for this purpose is selected
from the condition that the fission and absorption integrals be preserved for the.full-scale and model systems
. and that the macroscopic cross sections coincide in the energy range of the neutrons.
Computational studies showed that the necessary neutron-physical properties of an HTGR over the entire
range of parameters considered are ensured in the critical assemblies of the stand. In particular, the stand
provides a good simulation of the field of energy release and spectral characteristics, even in a highly inho-
mogeneous system with the principle of OPAZ in the equilibrium state.
Structurally; the critical assembly of the stand GROG has been built as follows: A set of graphite blocks
forms a cubic stack with a 450-cm face. Different geometrical and physical parameters of the core and reflec-
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Parameter
2
02
-
2
03
~
t-O1 1
2-01
1
I A
B I
A
1
B I
3-OS I
~-O1
'
Size of core, mm
z
75
75
75
25
75
25
75
75
y
92
75
75
25
75
25
75
75
Z
SO
110
110
1!0
!!0
!10
170
120
Ratio of carbon and uranium nu-
250
500
500
500
51H)
500
250
250
clear concentrations
Structure of fuel-assembly module
FE-50
FE -50
FE-50
FE 30
FE -50
F)r-10
FE -25
FE-50
DE-50
DE-50
FE-30
DE-50
DE-10
AE-0.5/5
FE-50
FE -50
FE -50'
FE-30
FE -50
FE;-10
FE -25
Composition of the elements of the module: FE-50 (30, 10, 25) - a universal physical imita-
tor of height 50 (30, 10, 25) mm with logo uranium enrichment; DE-50 (30, 10) - a graphite
insert of height 50 (30, 10) mm; AE-0.5/5 (0.5/10) -a disk of boron-containing paper of
thickness 0.5 mm and boron content 5 and 10 mg.
for are provided by placing different combinations of cylindrical elements into the channels of the graphite
stack. Control rods can be placed in the central channels of the columns of the graphite stack. Cylindrical
channels of diameter 15 mm in the corners of the graphite columns can hold boron carbide absorbing elements,
of diameter 12 mm for simulating perturbations, as well as sensors of the measuring system. When some of
the graphite blocks are removed their place can be taken by a fragment, or a mockup of a fragment, of the reac-
tor under study. In particular, mockups of spherical fuel elements, absorbing elements, and dummy elements
of the VG-400 reactor are used in the stand.
Figure 1 shows the makeup of the critical assembly of the stand GROG, including a central zone of spheri-
cal elements that is under study, a seed zone, and a graphite moderator surrounding them. As in high-tem- .
perature gas-cooled reactors, there is a space between the active zone (core) and the upper end reflector.
The design of the stand allows sensors of the measuring system to be arranged in any way desired. Be-
sides in-line acquisition and processing of experimental results, the multichannel automatic system of data
acquisition, storage, and processing, including a computer and terminals, makes it possible substantially to ex-
tend the application of techniques directed at fuller and higher-quality research.
No. of graphite columns
324
Maximum linear size of core, mm
4000
Maximum number of channels
2304
Uranium enrichment, 96
0.7-10
Maximum zssU charge, kg
250
Volume porosity of core, qo
5-40
Ratio of carbon and uranium nu-
clear concentrations
200-2000
Simulated regimes of HTGR operation
initial, transient,
Number of randomly arranged rods of
control and safety system
equilibrium
up to 24
The experiments on GROG were carried out in a certain sequence determined by the aims and logic of the
development of the investigations as well as by the methodological and technical preparedness for each stage.
The experiments .started with a study of small, methodological critical assemblies that were simple in compo-
sition and configuration; in the first place this decreases the number of indeterminacies upon comparison with
the results of calculations and, second, makes it possible for experiments with a large number of versions to be
carried out rapidly and simply. Subsequent experiments were aimed at verifying the simulation of the required
neutron-physical properties and studying the distinctive features of the HTGR studied, primarily the VG~00.
The scope of the investigations on the. methodological assemblies encompassed the development of experi-
mental, methods and the technique of work on the stand, verification of the initial (specification) data on the ma-
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TABLE 2. Composition of Measurements on
Methodological Assemblies
Parameter studied
1-OII
P-OII2-U2~
2-011
3-O1I
.-OS
Keff
-f-
-E-
-F-
-I-
-t-
~-
Reactivityeffect
-}-
-~
-I-
-{"
-{-
-~-
Eff. of control and safety
-}-
-{-
+-
-{-
-{-
-}-
rod system
R
eaction rate
Thermal-neutron activation:
tuaDy
...~
Abs, densityof neutron flux
-}-
Flux distribution:
resonance neutrons
-{-
-~-
fast neutrons
-{-
Spectral index
-{-
-{-
-}-
-{-
-~-
S (Lu, 1)y)
Rid Dy
-{-
1s~Au
Rate of 238U(n, y) reaction:
along radiusofphysical
~-
-{-
-{-
-{-
-{-
-}-
imitator
alon height of physical
~
-~-
-~-
.-f-
tor
imit
Plutonium coefficient f
+
.+
+
.+.
.~
.~-.
Pen
~'
-)-
~-
~'
~
~-
Reactton rate measured with
-~-
-{-
-}-
"}-
-~-
--~
fission and boron chambers
~
i
TABLE 3. Effective Neutron~Vlultiplication Fac-
tor in Assemblies
1-01 I
2-01 (
9-01 1
4-01
1,00661
1,00461
1,0551
1,01051
0,0002
0,0002
0,003
0,0005
0,997
1,006
1,046
1,015
terial composition of the assemblies, and approval of some methods and programs for the physical design of
HTGR and methodological assemblies (Tables 1 and 2). Most of the measuring methods are used in experiments
on different critical assemblies. Two original features of GROG .are scanning of the neutron flux over the
volume of the core with the aid of small fission and boron chambers [3] and measurement of appreciable varia-
tions of reactivity in physically large breeding systems.
The neutron-physical design of the assemblies was carried out in two stages. First we calculated unit cells
to obtain homogenized macroscopic constants, which were then used to determine the neutron-physical param-
eters of the assemblies. With allowance for the distinctive features of the assemblies of the stand GROG the
group constants were calculated with the aid of the programs WIMS [4, 51, MONR1 [6], and PTT [7]. The neutron-
multiplication factor Keff and the space-energy distribution of neutrons in the assemblies were determined
from the three-dimensional program QUM-3-HER [8]. Some distributions of the reaction rate and estimates of
the effects were found from the two~iimensional program PENAP [9] .
The specification data for the elements. of the assembly were checked, analyzed, and statistically pro-
cessed beforehand. Computational and experimental studies" [10] indicated the homogeneity of the properties of
the graphite stack in different directions. The agreement, within the limits of the error of measurement (2~)
between the measured neutron diffusion length in a continuous graphite medium and the corresponding calcu-
lated value, obtained by using the specification data for: graphite elements, gives confidence as to the correct-
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 Declassified and Approved For Release 2013/02/22 :CIA-RDP10-021968000300050006-1
ness of the latter. The calculations showed that the technological deviations of the parameters of the assembly
elements have a weak influence on the principal characteristics of the assembly: no more than 0.1`~ on Keff
and no more than 19'0 on the relative deviation of the neutron fluxes [ll].
As follows from Table 3, the calculated values of the effective neutron-multiplication factor in the assem-
blies are in satisfactory agreement with the experimental data (the difference does not exceed 19'0).
The experimental and calculated values of the spectral parameters agree to within 79o while the difference
between the experimental and calculated values in the distribution of the reaction rate may reach 109'0.
A transition to experiments on more complex assemblies, close in geometry and composition to the cores
of HTGR, will help improve the methods and programs for the design of such reactors.
LITERATURE CITED
1. E . V . Komarov, F . V . Laptev, V . G . Lyubivyi, F . M . Mitenkov, et al., "The VG-400 atomic-power en-
gineering facility. Possible reactor core designs," At. Energ., 47, No. 2, 79 (1972).
2. A. M. Bogomolov, A. S. Kaminskii, A. D. Molodtsov, et al., "Fuel element of the critical assembly of a
reactor," Inventor's Certificate No. 915628, Byull. Izobret., No. 30, 300 (1982).
3. A. B. Dmitriev and E. K. Malyshev, Neutron Ionization Chambers for Reactor Engineering [in Russian],
Atomizdat, Moscow (1975).
4. J. Ackew et al., "A general description of the lattice code WIMS," J. BNES, 5, No. 4, 564 (1966).
5. , N. I. Laletin and V. A. Lyul'ka, "On resonance absorption of neutrons in 238U," in: Neutron Physics [in
Russian], Part 4, Atomizdat, Moscow (1980), pp. 35-37.
6. A. G: Sboev, "Program for the calculation of resonance absorption of neutrons in reactor cells by the
Monte Carlo method MONR1," Vopr. At. Nauki Tekh., Ser. Fiz. Tekh. Yad. Reakt., No. 5(18), 91 (1981).
7. A. S. Kaminskii and L. V. Maiorov, "The PIT program for the calculation of fluxes of slow neutrons by
the Monte Carlo method with allowance for thermalization," Vopr. At. Nauki Tekh., Ser. Fiz. Tekh: Yad.
Reakt., No. 8(21), 76 (1981).
8. S. S. Gorodkov, M. I. Gurevich, and N. L. Pozdnyakov, "Instruction for the use of theQUM-3-HER program
for the calculation of athree-dimensional or two-dimensional heterogeneous reactor," Preprint 1A~~794,
Kurchatov Institute of Atomic Energy, Moscow (1977).
9. P. N. Alekseev, S. M. Zaritskii, L. N. LTsachev, and L. K. Shishkov, "The TVK-2D complex of programs,"
Vopr. At. Nauki Tekh, Ser. Fiz. Tekh. Yad. Reakt., No. 4(33), 32 (1983).
10. L. A. Anikina, A. S. Kaminskii, and E. S. Subbotin, "Measurement of the diffusion length of a graphite
.stack by an,improved prism method," Vopr. At. Nauki Tekh, Ser._Yad. Konstanty, No. 4(53); 44 (1983).
11. L. A. Anikina, A. M. Bogomolov, and A. S. Kaminskii, "Influence of spreads of parameters of elements
of critical assemblies of the stand GROG on their neutron~hysical characteristics," in: Experimentation
in Reactor Physics [in Russian], Moscow (1983), pp. 237-241.
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 Declassified and Approved For Release 2013/02/22 :CIA-RDP10-021968000300050006-1
STUDY OF MODEL COILS MADE OF SUPERCONDUCTOR
INTENDED FOR THE WINDING OF THE T-15
TOKAMAK
I.
O.
Anashkin, E. Yu. Klimenko,
S.
A.
Lelekhov, N. N. Martovetskii,
S.
I.
Novikov, A. A. Pekhterev,
The preparation for outfitting the Tokamak-l5 has included several stages of tests on a superconducting
current-carrying element (SCCE): study of the current-carrying capability of short specimens, and
tests of model coils, full-scale test coils, and regular windings of the facility. Much attention to the
behavior o~ SCCE in a winding has been paid since the decision to use niobium- tin multifilamen-
taryconductoras the basis of the current~arrying element of the superconducting toroidal-field winding
(STFW) for the T 15. Since 1978 tests on windings made of niobium-tin conductors similar in construction to
the current-carrying element of the STFW have been under way at the I. V. Kurchatov Institute of Atomic Ener-
gy. Some of the results have been published [1-3]. The testing of the model coils was preceded by the construc-
tion of special stand equipment and was carried out on a stand designed for tests of niobium-tin windings with a
diameter of ~ 1 m [4] . This, on the one hand, gave reason to hope that the tests could be performed fairly
rapidly but, on the other hand, led to a risk of damage to the niobium-tin filaments of the SCCE, designed for
use in a construction with a bending radius three times that of the model coil. This program came into being
in connection with the almost formal necessity of making certain that the behavior of the SCCE, mockup speci-
mens of which were successfully tested in 1977, would be predictable in the model winding. The first experi-
ments, however, gave some unsettling results: The model windings went into the normal state at a current that
was only half that used for short specimens of the current-carrying element (Fig. 1). As is seen, the results
did not confirm the hopes for ensuring the operating characteristics of the T-15. The main objective of the
second series of tests of model coils, which are described in this paper, was to ascertain the causes of the cur-
rent degradation, to demonstrate the possibility of obtaining in the windings currents close to the current in a
short specimen, and provisionally to verify that the design of the winding adopted for the T-15 facility is ap-
propriate for its purpose.
The following assumptions about the causes of current degradation in model coils could be made a priori:
1. The existence of macroinhomogeneities of the conductor (weak spots). It could be assumed that in
1979 since the tests were conducted on a limited number of specimens such weak spots were not detected and in
tests of the model coil it was those weak spots that determined the critical, current. (Now that 100 short speci-
mens have been tested and their total length is several times the length of the model winding and not a single
test in a strong magnetic field gave a result below the theoretical requirements, this cause need not be con-
sidered.)
2. Damage to the current-carrying element during transportation or winding.
3. Deformation of the winding under the action of ponderomotive forces, leading to heat release exceeding
the stability threshold of the current-carrying element.
In order to ascertain which of the last two causes was responsible for the observed degradation, a care-
ful study was made of the last model of the first series of tests [3], henceforth called model A; its parameters
are given in Table 1. The arrangement for the tests did not differ from that described earlier [3] (Fig. 2). A
segment of the model winding was placed in the gap of a superconducting niobium-tin solenoid of diameter 240
mm. This solenoid produced a magnetic field of up to 8 T in part of the winding and its interaction with the
model winding resulted in a force of up to 1.2 ?105 N that bent the model winding. The solenoid was cooled by
immersion in liquid helium and the model winding was cooled by a flow of helium along the hollow current-carry-
Translated from Atomnaya Energiya, Vol. 57, No. 6, pp. 401-404, December, 1984. Original article sub-
mitted March 6, 1984.
830 0038-531X/84/5706-0830$08.50 ?1985 Plenum Publishing Corporation
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 Declassified and Approved For Release 2013/02/22 :CIA-RDP10-021968000300050006-1
~~
~,
a x
?,~
Fig. 1. Results of first series of tests of
model windings: O) double pancake coils
of diameter 560 mm; O) glued pancake
coil of diameter 560 mm; CI) imitation of
STFW; model A; x) operating point of
glued pancake coil; 1, 2) characteristics
of SCCE of glued pancake coil and model
A, respectively.
TABLE 1. Parameters of Models Tested in the Second
Series
STENO-2K-9-
STENO-2K-11-
STENO-2K-11
Material of winding
9225'
9225
14641
Type of winding
Double pancake
Single pancake
Double pancake
No. of turns
11
4
8
Winding dimensions
,
mm:
inner diameter
780
761)
770
outer diameter
860
835
850
height
38
78
38
PresenceofpuLsed
Yes .
No
Yes
toroidal wtnding
Coupling factor, T/
kA:
between field and cur
0,15
0,04
0,073
rentof modelwinding
between field and cur-
rent of xternal
solenot~
between field and
current of toroidal
winding
*Superconducting current-carrying element with two cooling
channels and nine niobium- tin cores, each containing 7225
filaments .
ing element [1]. The model winding was insulated from the liquid helium in the cryostat by fiberglass textolite
tape, reducing-the heat exchange with the helium in comparison with internal cooling enough so that it could be
neglected. Wound around the model winding was an auxiliary toroidal winding, imitating the field of the plasma
column, directed parallel to the axis of the current-carrying element. The model and solenoid windings re-
ceived their power supply in series from a lO~cA generator and an auxiliary 3-kA generator made it possible
to add current to or subtract current from the model winding. This arrangement made it possible to study the
dependence of the critical current of the model on the magnetic field.
? In the second series of tests we used an additional compensator winding that increased. the sensitivity of
the bridge circuit for the detection of the normal phase so that the reverse part of the current-voltage (I V)
characteristics of the model would be observed. Tensometric displacement gauges were-.also set up to record
changes in the shape. of the winding and any possible displacements of the winding relative to the solenoid.
In tests of model A no abrupt deformations of the model were recorded up to the level of ~currerit"attained.
Recording of the I=V characteristics allowed an unambiguous conclusion to be made about the cause of the de-
gradation. The smearing of the I-V characteristic of the transition of the model was several times the smear-
ing of the I V characteristic of the SCCE specimen (the parameter of the growth of the electric field with the
current was 1050 A for the model and 360 A for the short specimen, Fig. 3). This allows the lowering of the
current-carrying capability of the winding to be attributed to local damage to it during winding, "transportation,
or previous tests.
Identification of the causes of the failure of model A led to the preparation of a new model in which the
probability of damage to the SCCE would be reduced to a minimum. The second problem in the preparation of
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 Declassified and Approved For Release 2013/02/22 :CIA-RDP10-021968000300050006-1
IT-3
IT-1 IT-2
~-r
Fig. 2. Arrangement for tests of model windings:
1) control valves; 2) cryostat; 3) power supplies;
4) solenoid setting up field in segment of winding;
5) pulsed toroidal winding; 6) model winding; 7)
bearing flange; 8) heater.
Fig. 3
4 6 B f0
Magnetic field, T
Fig. 4
3
ao
..
3 7S
f0
~ 5
m
x
1 P 6 B f0
Carrent, kA
Fig. 3. Comparison of the parameters of the growth of the electric field with
the current for the model (a) and the short specimen of SCCE from which the
model was made (b) .
Fig. 4. Results of second series of tests of model windings_O) model B; O)
model C; x) operating paint of T-15 STFW; 1, 2) characteristics of SCCE of
models B and C .
Fig. 5. Results of tests of models by the method of "extrapolation to zero
heater" (the squares correspond to the case when the critical current was
limited by mechanical perturbations in the winding):. O, O, O) models B, C,
and C';- = ?- ?-) expected behavior of model C' in the absence of mechanical
perturbations.
model B was, that :of demonstrating the possibility of producing a winding in which not only degradation but also
conditioning of ;the SCCE would be eliminated. The latter is essential since at the design current of 5.2 kA the
T 15 wind,ing_ (STFW) is not stabilized in a stationary state and the beginning of conditioning of such a large
magnet. system can be tantamount to a limitation of its working capacity. Model B (see Table 1) was aone-layer
pancake coil, glued on either side to 0.8 mm stainless steel sheet. To diminish the risk of breakage of the
niobium-tin filaments, immediately after the application of a stabilizing layer of electrolytic copper [5] the cur-
rent~arrying element was wound on a receiving cassette with an inner diameter of 0.8 m. It was not subse-
quently rewound. Before the gluing the tension in the SCCE was reduced and this permitted fiberglass textolite
insulating insert to be placed between the windings. Because of the bandaging scheme adopted model B has an
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 Declassified and Approved For Release 2013/02/22 :CIA-RDP10-021968000300050006-1
d=B,ST
~- %
~,6
U4
0,2
0
-Q2
-0,4
-J,6
-UB
-1 0 1 -1,0
Tensile_deformation, %
Fig. 7.
Fig. 7. Dependence of critical current of the SCCE on
the tensile-compressive deformation: O) extrapolated
critical current of model B.
Fig. 8. Dependence of the stability of the SCCE on the
amplitude of the pulsed magnetic field, the relative ori-
entation of I and Bo during cooling with two-phase heli-
um at a pressure of 0.13-0.15 MPa (~) and supercriti-
cal helium at a pressure of 0.4-0.5 MPa (O).
extremely great rigidity with respect to bending in the plane of the winding. The tests were carried out accord-
ing to the scheme indicated above. Without any conditioning a reversible transition to the normal state was at-
tained at a current substantially above the critical current of model A (Fig. 4). Further on we discuss why the
current of the short specimen is not attained in small~liameter models. In any case the slope of the I V curves,
which coincides with the slope of the I V curve of the short specimen, indicates that the SCCE in model B was
not damaged. This is also indicated by the results of tests using the so-called method of "extrapolation to zero
heater" (Fig. 5): the current at which a resistive zone appears in the winding is studied as a function of the
power of a local heater arranged on it or as a function of the readings of a thermometer placed together with the
heater. In the absence of damage in the SCCE and perturbations in the winding, this dependence is plotted as a
smooth, almost straight line and the point corresponding to the transition of the winding when the heater is
switched on lies on this line (models B and C).
If the SCCE has been damaged and the heater turned out not to be at the site of the damage, it is unlikely
that characteristic kinks would appear on this line. If the transition is caused by perturbations, e.g., of a
mechanical origin, then starting from a certain value of current at which the acting forces reach an appreciable
value the critical current ceases to depend on the power of the heater and the transition takes place abruptly,
without a recorded resistive segment of the I-V curve (model C').
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The fact that a current satisfying the requirements of the STFW for the T-15 tokamak was attained in
model B with rigidly fastened winding without any conditioning stimulated the construction of model C, whose
design imitated the design of the T-15 STFW. The current-carrying element was wound on a rigid stainless
steel ring with an L-shaped cross section (Fig. 6). In this winding, too, we managed to approach the current of
a short specimen, but only after a single conditioning. This is a cause of some concern since the forces acting
in the STFW of the T-15 will be several orders of magnitude greater than in the model and the number of con-
ditionings may increase sharply.
The fact that in models B and C the critical current is somewhat below the level of current in the short
specimen can be explained if the dependence of the current-carrying capacity of the SCCE on the deformation
is taken into account (Fig. 7). During winding of aheat-treated SCCE of radius 0.4 m the inner layer of the
transported cores turns out to be compressed.0.5go and their current-carrying capacity should diminish. This
decrease cannot be compensated by an increase in the current-carrying capacity of the stretched cores of the
outer layer since the cross resistance of the conductor is too high for the current to be redistributed between
the stretched and compressed cores over the transposition length (80 mm) on which the cores change places.
The critical current of the bent SCCE practically coincides with the critical current of the compressed cur-
rent carrying element.
The earlier hypothesis that the sensitivity of the SCCE to perturbations by a pulsed magnetic field de-
pends on the relative direction of transposition, the current in the winding, and the pulsed field has been checked
on models A and B. Since the transposed cores of the SCCE form helices about the axis of the conductor, a
change in the longitudinal magnetic field can induce in them a current in the same direction as the transport
current or in the opposite direction. At the same time eddy currents in the stabilizing copper warm up the
superconductor slightly. Depending on the resultant current that flows in the superconductor, this warming up
either causes the transition of the superconductor to the normal state or is insufficient to do so.
Figure 8 gives the results of a study of this effect on model A. A current slightly below the critical
value (6 kA) was introduced into the model winding. Then a current of 0 to 1200 A was introduced into the toroi-
dal winding wound around the model winding, up to a certain value, starting from 50 A. After this the current
was removed from the winding approximately exponentially with a time constant of 7 10 msec; this imitated
the nature, amplitude, and rate of change of the magnetic field of the plasma during tearing of the plasma in the
T-15 facility. If with such removal of current the model winding remained in the superconducting state, the cur-
rent in the toroidal winding was increased by 50 A and if it went into the normal state the corresponding value
of the longitudinal field was recorded as the maximum attainable for the given value of current in the model
winding. The experiments were conducted in three SCCE cooling regimes, corresponding to two~hase and one-
phase supercritical helium in the channels. It was shown that cooling with two-phase helium slightly increases
the stability of the winding..
Thus, after a second series of tests of model windings of the current-carrying element intended for the
STFW of the T-15 tokamak it has been established that the material of the winding is suitable for use and has a
considerable reserve of stability in stationary regimes and in pulsed perturbations by a longitudinal magnetic
field. The current-carrying element can be used in the STFW of the T-15 tokamak, but the requirements as to
ensuring the rigidity of the winding are much more stringent.
1. N. Chernoplekov, "Status and trends of superconducting magnets for thermonuclear research in the
USSR," IEEE Trans., MAG-17, No. 5, 2158 (1981).
2. D. P. Ivanov et al., "Study of the properties and stability of a superconducting current~arrying element
for the Tokamak-15 facility with respect to pulsed poloidal fields," in: Proceedings of the Al1~Jnion Con-
ference rnthe Engineering Problems of Fusion Reactors [in Russian], Vol. 8, A. D. Efremov Scientific-
Research Institute of Electrophysical Apparatus, Leningrad (1982), pp. 94 103.
3. D. P. Ivanov et al., "Study of model coils of the superconducting current-carrying element of the
Tokamak-15 facility," Preprint IAE-3715/7, Institute of Atomic Energy, Moscow (1983).
4. I. L. Zotov et al., "Test stand for large superconducting magnet systems," Prib. Tekh. Eksp., No. 6,
176 (1974) .
5. V. Agureev et al., "Electroplated stabilized multifilament superconductor," IEEE Trans., MAG=11, No. 2,
303 (1975).
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 Declassified and Approved For Release 2013/02/22 :CIA-RDP10-021968000300050006-1
OSCILLATIONS IN THE CONCENTRATION OF ARTIFICIAL
RADIONUCLIDES IN THE WATERS OF THE BALTIC
AND NORTH SEAS IN 1977-1982
D. B. Styro, G. I. Kadzhene, UDC 551.464.679
I. V . Kleiza, and M . V . Lukinskene
As a result of extended atmospheric tests of nuclear and thermonuclear weaponry a certain "background"
of artificial radionuclides has been formed in the environment [1] . Up to the present time this background has
decreased significantly in the waters of the Pacific Ocean. Oscillations of it have turned out to be primarily
associated with discharges of radioactive wastes by undertakings of the nuclear industry [2, 3] or with the ef-
fects of individual atmospheric nuclear explosions carried out up to the present by the People's Republic of
China [4, 5] .
The Baltic and North Seas are contaminated with various radionuclides, of which the most radioactively
dangerous are 137Cs, 90Sr, and to a lesser degree, 144Ce. If contamination of the waters of the Baltic Sea is
caused mainly by global fallout, the waters of the North Sea are contaminated predominantly by the wastes of
undertakings of nuclear industry [2, 3, 6] .
Regular observations of the radioactive contamination of the water areas of these seas have been made by
us since 1973 [7 10]. The natural data obtained have permitted estimating the possible values of the concentra-
tion of various radionuclides, investigating with the help of a mathematical method (the so-called objective anal-
ysis) the structure of the radionuclide concentration fields [10-12], and revealing the possible source of the
contamination [13] .
The isolation of radionuclides from seawater was performed by the radiochemical methods described in
[14, 15] and with the help of a very rapid sorption procedure for is7Cs [16]. We have determined the maximum
possible error of a test from the results of numerous measurements of the concentration of the radionuclides
of one and the same sample of water. In the determination of the concentration of is7Cs and 144Ce it amounted
to 2590 and for 90Sr - 159'0? The results of measurements obtained on the basis of the sorption procedure under
natural conditions were selectively monitored by radiochemical analyses during each series of observations.
The average values. of the concentration of the radionuclides investigated are not constant in the waters of the
Baltic Sea, and especially large variations in the concentration of 90Sr are characteristic (Fig. 1). The concen-
trations of the other radionuclides are subject to oscillations to a lesser extent.
It has been .established in the analysis of the published data [4, 5, 17-20] that the variations in the concen-
tration of radionuclides in the surface waters of the Baltic Sea are caused by changes in the global fallout due
to atmospheric nuclear tests conducted by the People's Republic of China. The time of the nuclear tests corre-
lates well with the time of increase in the average concentration of radionuclides in the waters of the Baltic
Sea (mainly 90Sr). Amore noticeable increase in the average concentration of 137Cs and 144Ce has been noted in
the fall of 1980 after thermonuclear tests in the People's Republic of China. [5] (see Fig. 1). One should empha-
size that the increase in the intensity of global fallout in recent years has been accompanied mainly by an in-
crease in the concentration of 90Sr in the Baltic Sea; it has reached its own "na.tural" background values in a
more extended period of time than has 137Cs.
The contamination of the North Sea waters is caused mainly by discharges of wastes by undertakings of
the nuclear industry [2; 3] . Therefore the variations in the global fallout do not cause noticeable variations in
the is7Cs concentration, as has been observed in the Baltic Sea. The average concentration of is7Cs in the sur-
face waters of the North Sea has increased especially strongly since 1976: thus from 1975 through 1976 it in-
creased from 36 to 60 Bq/m3, in 1977 it increased to 108 Bq/m3, an~1 it maintained approximately the same
value up to 1981 (106 Bq/m3). A decrease to 84 Bq/m3 was detected only in the fall of 1982.
Translated from Atomnaya Energiya, Vol. 57, No. 6, pp. 405-407, December, 1984. Original article
submitted April 4, 1984.
0038-531X/84/5706- 0835$08.50 ?1985 Plenum Publishing Corporation 835
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 Declassified and Approved For Release 2013/02/22 :CIA-RDP10-021968000300050006-1
.ra
i 200
1978 9979
Fig. 2
Fig. 1. Average values of the is7Cs (O) , 90Sr (~) , and i44Ce (x) concentrations in the
surface waters of the Baltic Sea.
Fig. 2. Variation of the is7Cs (O), and 90Sr (~) concentrations in the surface waters
of the North Sea near the bay of the Firth of Forth.
TABLE 1. Concentration of Radionuclides in
September-October 1982 in the Surface Waters
of Seas, Bq/m3*
Baltic
25/7
34/8
16/5
North
233/32
37/9
,48/6
Norwegian (south-
ern part)
8/4
9/3
33/5
*The numerator denotes the maximum concen-
tration, and the denominator denotes the mini-
mum concentration.
The most contaminated waters evidently enter the North Sea from the bay of the Firth of Forth, which is
connected by systems of canals with the Irish Sea. Many years of observations have permitted detecting a con-
stantly increased 137Cs concentration in this region of the sea [2, 10, 13]. One should note that here the in-
creases or decreases in the is7Cs and 90Sr concentrations correlates well, although the absolute values of the
soSr content are lower by an order of magnitude (Fig. 2). In this region the concentration of both radionuclides
have their largest values as a rule. Consequently, the waters of the North Sea are contaminated by both radio-
nuclides due to discharges of wastes by undertakings of the nuclear industry.
We shall dwell in more detail on the results of measurements of the concentration of radionuclides in'the
surface waters of the Baltic and North seas in the fall of 1982, since they have been published first. The total
content of the radionuclides under discussion in the surface waters of the Baltic Sea decreased significantly in
comparison with 1981 (see Fig. 1) except for 137Cs, whose average concentration remained the same (15 Bq/m3)
[10] . However, the absolute values of the concentration of these radionuclides in the surface waters measured
at different points of the Baltic Sea vary within rather wide limits (see Table 1).
Figure 3 illustrates the nonuniformity of the distribution of radionuclide concentrations in the surface
waters of the Baltic Sea. The greatest nonuniformity in concentration is characteristic of 90Sr (Fig.. 3b). Here
an interesting regularity is revealed: the is7Cs concentration decreases mainly at the points where the 90Sr and
i44Ce concentrations increase. Natural oscillations of the radionuclide concentrations in the waters of the
Baltic Sea significantly exceed the errors of the measurement results. The cited method -objective analysis
using the experimental information obtained -was applied to the investigation of the structure of the concentra-
tion fields of radionuclides in the reservoirs under investigation.-
The structure of the is7Cs concentration field in the surface waters of the Baltic Sea in the fall of 1982
is given in Fig. 4a. This structure has changed noticeably in comparison with the spring of 1981 [10], although
836
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 Declassified and Approved For Release 2013/02/22 :CIA-RDP10-021968000300050006-1
7, d a _
?3 ,~ ,'
2~
x ,,191
>0 -S 0 S 70
a
70 '.S 0
b
~_
5 70
iii i
0 dG, Bq/m3
c
Variations of the (a) 137Cs, (b) 90Sr, and
(c) 144Ce concentrations iri the surface waters of
the Baltic Sea in the fall of 1982 (vertical solid
.lines denote the maximum experimental errors).
Fig. 4. Concentration field of (a) 137Cs, (b) 90Sr, and (c) iaaCe in the surface waters of the
Baltic, North, and Norwegian (southern part) Seas in the fall of 1982 in Bq/m3.
the average value of the concentration remained the same: the absolute values of the 137Cs concentration de-
creased in the northern part and then increased in the central part.: The absolute values of the 90Sr concentra-
tion decreased significantly, and the structure of its field changed completely (Fig. 4b). In this case the greatest
values of the concentration were exhibited in the central part of the sea. The 144Ce concentration field in the
Baltic Sea appears comparatively uniform (Fig. 4c), and the average concentration of this radionuclide in the
surface waters decreased from 16 to 10 Bq/m3 fromthe spring of 1981 [10] to the fall of 1982.
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 Declassified and Approved For Release 2013/02/22 :CIA-RDP10-021968000300050006-1
The structure of the 137Cs and 90Sr concentration fields was noticeably altered in the waters of the North
Sea (Figs. 4a and b) in comparison with the spring situation of 1981. A gradual decrease inthe concentration
of these radionuclides has been noted as one moves away in all directions from the bay of the Firth of Forth.
The average value of the 90Sr concentration in the surface waters decreased somewhat in comparison with the
spring of 1981 [10] and reached 20 Bq/m3, but 144Ce increased by approximately a factor of two (21 Bq/m3).
A gradual decrease in the 137Cs and 90Sr concentrations in the northern direction is observed in the
southern part of the Norwegian Sea (see Figs. 4a and b). However, th.e structure of the 144Ce concentration
field turned out to be completely different -two regions of enhanced values in the southern parts of the North
and Norwegian Seas (see Fig. 4c).
One should note that the structure of the is7Cs and 90Sr concentration fields in the waters of the North Sea
and its oscillations are explained on the one hand by a variation in the action of contamination sources and on
the other hand by the specific properties of hydrometeorological factors. The structure of the 144Ce concentra-
tion field in this case is evidently associated with the peculiarity of global fallout. The structure of the con-
centrationfields of the radionuclides under investigation and its variation in the waters of the Baltic Sea are
related mainly to oscillations of the atmospheric fallout and hydrometeorological conditions.
1. B. A. Nelepo, Nuclear Hydrophysics [in Russian], Atomizdat, Moscow (1970).
2. H. Kautsky, Dtsch. Hydrogr., Z., No. 6, 241 (1973).
3. H. Kautsky, Dtsch. Hydrogr., Z., No. 7, 217 (1977).
4. E. Reiter, Atmospheric Transport Processes, Part 4: Radioactive Tracers, Technical Information Cen-
ter, U.S. Department of Energy, Washington (1978).
5. S. Sadasivan and U. Mishra, J. Geophys. Res., C87, No. 9, 7343 (1982).
6. G. I. Kadzhene et al., in: Radioactivity of the Atmosphere and Hydrosphere. Radioactive Tracers [in
Russian], Mokslas, Vilnius (1977), p. 207.
7. D. B. Styro et al., At. Energ., 45, No. 3, 201 (1978).
8. D. B. Styro et al., At. Energ., 49, No. 1, 43 (1980).
9. D. B. Styro et al., At. Energ., 51, No. 2, 116 (1981).
10. D. B. Styro et al., At. Energ., 55, No. 4, 238 (1983).
11. D. B. Styro and I. V. Kleiza, in: Propagation of Impurities in the Environment [in Russian], Mokslas,
Vilnius (1980), p. 148.
12. D. B. Styro and I. V. Kleiza, Okeanologiya, 21, No. 3, 464 (1981).
T3. D. B. Styro and I. V. Kleiza, in: Problems of Investigations of Atmospheric Contamination [in Russian],
Mokslas, Vilnius (1981),. p. 119.
14. L. M. Ivanova, Radiokhimiya, 9, No. 5, 622 (1967).
15. Yu. Ya. Mikhailov, "A procedure for determination of strontium-90, cesium-137, cerium-144, and lead-
210 in samples of plant origin," Inst. Exptl. Vetern., Moscow (1980).
16. D. B. Styro, in: The Propagation of Impurities in the Environment [in Russian], Mokslas, Vilnius (1980),
p. 121.
17. A. Yamato, Radioisotopes, 30, No. 2, 104 (1981).
18. J. Roy et al., Health Phys., 41, No. 3, 449 (1981).
19. R. Reiter and K. Munzert, Arch. Meteorol. Geophys. and Bioclimatol., B31, No. 3, 191 (1982).
20. L. Burchfield et al., J. Geophys. Res., C87, No. 9, 7273 (1982).
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 Declassified and Approved For Release 2013/02/22 :CIA-RDP10-021968000300050006-1
ENHANCEMENT OF HEAT TRANSFER IN THE
RBMK AND RBMKP
A. I. Emel~yanov, F. T. KamenT shchikov, UDC 621.039.553,34
Yu. M. Nikitin, V. P. Smirnov, and V. N. Smolin
The increase of unit power of nuclear reactors and the related increase in heat removal is one of the
pressing problems of the development of nuclear power [1]. An effective way of solving this problem is the
enhancement of heat transfer infuel-element assemblies (FEA).
Heat-removal enhancement is a complicated problem, related not only to an increase in heat removal
while maintaining the accepted criteria of the working capacity of the fuel elements, but also to the solution
of such concomitant problems as the decrease of the energy expended to pump through the coolant, the guar-
. antee of the necessary technological effectiveness of the proposed enhancement devices, the account of the
possible effect of these devices on the physical parameters of the core, etc.
The approaches to the increase of heat removal in FEA with two-phase and single-phase coolants are
quite different. This is a result of the different nature of the heat and mass transfer processes which deter-
mine the maximum thermal stress of thefuel-element cladding, which in turn is the main barrier to the re-
lease of radioactive fission products. For a FEA with a boiling coolant we take as an implicit parameter of
the maximum thermal stress of thefuel-element cladding the critical heat flux density characterizing the over-
all growth or local spikes of thefuel-element temperature.
The onset of aheat-transfer crisis in the disperse annular flow of the coolant, realized 'in channels with
a boiling coolant, is related to the overall or local (stable or periodic) disappearance of the liquid film from
the heated surface as a result of the imbalance of the removal processes, evaporation and deposition in the
presence of an appreciable amount of moisture in the flow core [2-4]. The cooling conditions of fuel elements
are generally :not identical over a cross section of a FEA because of possible thermohydraulic inequalities of
individual elementary hydraulic cells (both as a whole and over the azimuth of each fuel element), Such a
heat-transfer mechanism in disperse annular flow also predetermines possible methods of increasing; heat
removal in a FEA: the deposition of moisture from the flow core on heated surfaces and the equalization of the
thermohydraulic parameters of the coolant over the cross section of a FEA. In doing this it must be kept in
mind that~tliis separation of enhancement methods is somewhat arbitrary, since they may accompany and be
produced by common enhancement devices.
An effective heat-transfer enhancement method is the rotation of the coolant stream. This produces a
centrifugal acceleration field which separates the moisture from the flow core on the heated surface. In most
of the articles on the study of this method of heat-transfer enhancement, as noted earlier [5, 6], a favorable
effect is observed which increases with increasing stream content [7]. As applied to a FEA with fuel rods,
heat-transfer enhancement is accomplished by using vortex generators which produce a rotation of the flow in
each elementary hydraulic cell of astraight-through cross section of the assembly.
Research performed at the All-Union Institute of Heat Engineering (V TI) [8] on the heat-transfer crisis
in assemblies with twisted tapes over the whole length showed the possibility of achieving crisis-free operat-
ing conditions up to steam contents of 80, 70, and 60%, specific heat loads of 0.58, 0.93, and 1.16 MW/mz, mass
velocities of 800-1000 kg/mZ ? sec, and a pressure of 6.9 MPa. It was shown in [9, 10] that the steam content
at a channel exit when twisted tapes extended along the whole length could be increased from 10.5 to 25-35%.
When the tapes extended along the whole length of the assembly, technological difficulties arose, particularly
in long channels, and also the amount of absorbing material in the reactor core is appreciably increased. To
decrease the Dore absorption and the hydraulic resistance, it is more expedient to use short twisted tapes, but
this decreases the effectiveness of the heat-transfer enhancement [9].
Local heat-transfer enhancers in the form of short twisted tapes were tested experimentally at NIKI~Te
in an RBMK FEA consisting of a 19-rod bundle of fuel elements 7 m long [l 1, 12]. The structural parameters
Translated from Atomnaya Energiya, Vol. 57, No. 6, pp. 408-413, December, 1984. Original article
submitted March 2, 1984.
0038-531X/84/5706- 0839$08.50?1985 Plenum Publishing Corporation
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 Declassified and Approved For Release 2013/02/22 :CIA-RDP10-021968000300050006-1
Fig. 1. Schematic diagram of circulation of
single-phase flow in a transverse straight-
through section of a 19-rod FEA following
an axial twist enhancer.
and the positions of the short twisted tape enhancers along the height of the assembly were optimized. Tests
to estimate the effectiveness of such enhancers were performed on electrically heated assemblies with seven
rods 2.2 m long and three rods 7 m long.
The experiments were performed for the following range of parameters: pressure 7.4 and 9.8 MPa,
mass velocities from 600 to 2000 kg/m2 =sec, and heat fluxes from 0.3 to 1.6 MW/m2 with underheated water
entering the assembly at the saturation temperature.
The results showed that for the same heat flux density the absolute growth of the critical mass steam
content was 20-30%, and this growth increases with an increase in the mass velocity and pressure. Calcula-
tion of the maximum power of a FEA from these data showed the possibility of increasing the RBMK power by
50%. Further development of the heat-transfer enhancers was directed toward producing more technological
types of heat-transfer enhancers based on the RBMK-1000 FEA spacer lattices.
It should be noted-that because of the mixing of the flow and the redfstribution of phases, practically all
types of spacer lattices permit anincrease in the critical heat flux density. The heat flux density increases
appreciably immediately after the spacer elements, and gradually decreases with increasing distance from
them. Therefore, experiments were first performed on assemblies with the regular RBMK spacer lattices
for a decrease in their spacing along the height of the assembly. Shortening the distance between them from
350 to 175 mm increased the critical heat flux density by 10-20%, and the hydraulic resistance by 50-60%.
Various kinds of enhancement elements were placed on the rims and mesh of the spacer lattices to increase
their effectiveness. A similar approach toheat-transfer enhancement was discussed in [13, 14].
In placing the enhancement elements, consideration was given to their effect on the heat-transfer crisis
and its place of origin in the assembly. Tests on the electrically heated model of a RBMK FEA showed that the
heat-transfer crisis arises in the hottest triangular hydraulic cell between the inner and outer sets of rods,
and also on those parts of the perimeter of the fuel element tubes of the outer row of the assembly, which are
separated from an unheated channel by a narrow gap. This can be accounted for qualitatively by the fact that
the elementary cells are not thermally or hydraulically equivalent, and by the effect of an unheated (cold) wall
along which an appreciable part of the moisture flows, which lowers its flow rate on the surface of the fuel
elements in this region.
The variants of lattice-enhancers developed are mainly of two types. One type is based on the idea of
rotating the flow in each elementary hydraulic cell of a FEA by using straight vane-deflectors to produce
centrifugal acceleration gelds. These are substantially more effective than vortex generators in the form of
twisted tapes [15]. The other type uses structural elements between the outer set of rods and the channel,.
and also in the least-filled parts of the straight-through cross section between the inner and outer sets of rods.
These greatly assist the mixing of the coolant in a FEA cross section.
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 Declassified and Approved For Release 2013/02/22 :CIA-RDP10-021968000300050006-1
3 H so
E ~o
o `'v
c
?..
i ~~
I '? ?
.?
? ...
?s.. .
Fig, 2, Dependence of heat-
transfer coefficients on mass
steam content for an assembly
with axial twist enhancers in
an experiment with p = 7,5
MP a, pw=1000-2000kg/m2. "
sec, and q = 0.9 MW/mz,
The experiments were performed in several stages. First, the study of the spacer lattices of the heat-
transfer enhancers was completed ona stand at the EMS branch of the VNIIAES [16], The experiments on
crisis heat transfer and hydraulic resistance were performed on the RBMK FEA thermohydraulic model, con-
sisting of a 19-rod bundle with afull-scale straight-through cross section and a heated length of 1,1 m, This
model permitted the study of full-scale enhancers, and since it was more mobile than the full-scale model, it
speeded up the completion of the experimental study of the enhancers, and broadened the possibilities of
diagnosing theheat-transfer crisis. The use of thin-walled electrically heated tubes on a shortened assembly
(b = 0.2 mm) made it possible to place thermocouples in the center of a tube to measure the average tempera-
ture, and other thermocouples on the inner surface of a tube (up to five in one of its cross sections) to fix the
location of the onset and development of aheat-transfer crisis from the local wall temperature, Based on the
possibilities of the stand, tests were performed for p = 7,5 MPa, pw = 1000-2000 kg/m2: sec, and q up to 1
MW/mz for asteam-water mixture supplied at the entrance.
This approach to the study of cheat-transfer crisis in rod assemblies with heat-transfer enhancers is
justified experimentally by the agreement of results obtained on shortened and full-scale FEA models [17],
A comparison of the results of experiments performed on shortened FEA models with developed types
of heat-transfer lattice enhancers [18] permitted the determination of the most effective type of axial twist
enhancers [19], The subsequent stages of the experimental research on these heat-transfer enhancers, per-
formed on full-scale electrically heated models on a stand at the Institute of Atomic Energy (20], and also
in in-pile tests [21], showed the possibility of using such enhancers in ftBMK-1500 FEA.
For a broader understanding of the kinematics of flow beyond axial twist enhancers, the velocity field
was measured on the aerodynamic model of an RBMK-1500 assembly on a scale of 2 : 1. The results gave a
qualitative picture of the circulation of flow in a section of the assembly (Fig, 1), explaining the probable mech-
anism of the effect of enhancers on vapor-liquid flow: the equalization of the parameters as a result. of axial
rotation between assembly and channel, the.equalization of the liquid film around the periphery of a fuel ele- ,
ment as a result of the circulation around the rods, and the separation of moisture from the flow core on the
fuel elements as a result of the formation of secondary vortex flows.
In considering the effect of axial twist enhancers on the heat-transfer crisis, it should also be noted that
in addition to the increase in critical parameters, the character of supercritical heat transfer is also changed,
and consequently also the supercritical thermal conditions of thefuel-element cladding. The graph of the
supercritical heat-transfer coefficient a = f(x) in Fig. 2 shows a smooth decrease of the heat-transfer coeffi-
cients from values characteristic for the development of boiling, to values characteristic of heat transfer after
the accumulation of a sufficient number of experimental data clearly permits raising the question of the pos-
sibility of operating FEA (if only temporarily) in the supercritical region up to the maximum admissible tem-
perature for the fuel-element cladding, and the incorrect use of such a criterion as the critical heat flux den-
sity as controlling.
The production of more effective heat-transfer enhancers will, in our opinion, be determined largely
by the development of research on mass transfer processes in disperse-annular flows and supercritical
heat-transfer processes which takes account of the possible effect of the structural elements of the lattices
[22], This requires improving the diagnostics and methodology of local crisis phenomena. Such work should
culminate in programs for calculating the thermohydraulic parameters in a FEA. The development of these
programs is still in the initial stage,
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 Declassified and Approved For Release 2013/02/22 :CIA-RDP10-021968000300050006-1
Fig. 3 Dependence of heat-transfer coef$cient
and coefficient of friction on Re and height h of
roughness protrusions or s/h = 10; 4) h = 0.1;
x) 0.075; ~) 0.05; O) 0.03 mm.
Questions of heat-transfer enhancement for gas (vapor)-cooled nuclear reactor cores such as the BGR-
300 [23] and the RBMKP [1] are very important. Heat transfer in the FEA of such reactors is characterized
by greater heating of the coolant in the core and larger thermal heads between thefuel-element surface and
the coolant. The magnitude of the thermal stresses in the structural materials of the fuel-element cladding
is -the basic criterion of the effectiveness of heat-transfer enhancement. It is known that the thermal heads
for heat transfer with a gaseous coolant are determined by the thermal resistance of a thin boundary layer.
Practically the only effective way to improve heat transfer from the heat-transfer surface to the flow core
is to create detachment zones and other vortex structures in the boundary layer. Such vortex zones as a rule
are produced by regular protrusions (turbulators, arti$ciai roughness) of various heights and shapes.
Many papers have been devoted to the study ofheat-transfer enhancement by regular roughness. The
most complete summary of these papers is given in [24, 25]. Most of the experiments were performed with
turbulators which practically could not be mass produced because of the great difficulty and complexity of
ensuring stable geometry of protrusions with a height of ~ 0.1 mm. Therefore, the first step in solving the
problem of heat-transfer enhancement in FEA of gas-cooled reactors was the selection of a method of forming
regular turbulators on the surface .of the fuel-element cladding. Research abroad [26] showed that the most
promising methods of forming artificial roughness of fuel-element cladding are mechanical and electrochemi-
cal, Removal of metal in electrochemical roughening occurs mainly as a result of electrochemical processes.
The shape of the $ns is determined by the shape of the abrasive disk.
Anexperimental-industrial unit, the TE V-1, was built at the Tula Polytechnic Institute (TPI) under the
direction of E. I. Pupkov to produce artificial roughness of the outer surface of fuel-element cladding by an
electrochemical treatment (ECT) with arotating cathode. The working surface of the cathode was shaped in
the form of the hollow between the protrusions of the artificial roughness for a spiral relative motion of the
cathode and the billet [27]. The ECT method ensures that the height of the roughness profile is accurate
to within 0.005 mm in the absence of cold working and corrosion along the grain boundaries of the metal, and
can easily be realized on a large scale.
The thermohydraulic characteristicsof the artificial roughness formed by the ECT were obtained by
making a batch of samples of roughened tubes ~ 10 mm in diameter with a wall thickness of 1 mm. Most of the
.studies were on symmetrical trapezoidal shaped roughnesswith a relative pitch s/h in the range 7-35 and a
height of 0.03-0.1 mm. The experiments were performed at the VTI on an electrically heated annular channel
with a flow of air [27]. Roughened tubes were inserted into a tube with an inside diameter of 19.46 mm. The
Reynolds number was varied from 25 x 103 to 650 x 103, air was heated 30-40?K from entrance to exit, the ther-
mal head varied from 60 to 100?K, the temperature factor was 1.2-1.3, and the Mach number did not exceed
0.5.
Figure 3 shows how the ratios of the coef$cients of hydraulic resistance ~1/~sm and the Stanton num-
bers St /Stsm depend on the relative pitch s/h for the region of the annulus adjacent to the inner roughened
tube. The experimental data for this region were scaled by the Wilki method [28]. The relations shown in
Fig. 3 con$rm the earlier conclusions [24] for roughness of the .nontechnological type that the maximum
increase in the Stanton number Stl over Stsm for a smooth tube at the same values of the Reynolds number
Rel occurs for s/h = 10-12. The dependences of ~ 1 /~ sm and' St1/Stem on Rel (Fig. 3) show that artificial, rough-
ening of the fuel-element cladding by ECT can increase the heat-transfer coefficient by a factor of 2-2.5 while
increasing the hydraulic resistance by a factor of 2.7-4.
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While ECT produces highly accurate roughness, it is a slow process (5 mm/min), but it can be speeded
up by using a multifilament cathode to produce roughness in the form of multiturn helical grooves chile main-
taining between protrusions a distance equal to the pitch of single-turn roughness. Experiments on the VTI
stand using samples of tubes with four-turn artificial roughness made at the TPI showed that such tubes have
approximately the same thermohydraulic characteristics as tubes with single-turn roughness [27]. However,
the ECT speeded up the process by a factor of.3.5. A further increase in the number of cathode filaments
should show how much the ECT can speed up the process while maintaining the thermohydraulic character-
istics of the roughness.
The problem of heat-transfer enhancement in an overheated RBMKP FEA is related to the problem of
the buildup of artificial roughness by the precipitation of salts. According to recommendations in [24J, for
Reynolds numbers of (5-8) ? 105, which are characteristic for an overheated assembly, the optimum height of
roughness protrusions is ~ 0.03 mm, which can be commensurate with the thickness of a layer of precipitated
salts. In addition, the flow around protrusion type roughness is accompanied by the .formation of stagnation
zones near the protrusions. More intense precipitation of salts.may occur, in these zones, which leads to a
leveiing off of the protrusions and a decrease of the heat-transfer enhancement effect, In our opinion these
unfavorable effects can be avoided by using wave-type roughness:
Experiments [29J on the VTI stand with tubes having artificial wave-type roughness which were made at
the TPI by ECT showed that for Re= 7.105 the height of this type of roughness could be increased by a factor
of five to six, while preserving the values of Stl and ~1, as also for roughness of the rectangular protrusions
type. By increasing the height of the roughness to 0.15-0.2 mm and removing the stagnation zones, wave-type
roughness offered better resistance to the precipitation of salts on thefuel-element cladding in the steam
flow.
The high sensitivity of the thermohydraulic characteristics of gas (vapor)-cooled FEA with rod fuel ele-
ments, depending on the characteristics of their construction and on the distribution of fuel elements over the
cross section, introduces increased requirements in the calculation of the thermal state of the fuel elements.
As also for evaporative assemblies, further experimental research with careful diagnostics and the introduc-
tion of the results into programs for calculating the thermohydraulic parameters of the assemblies is pro-
posed, In addition it should be noted that for single-phase coolants great success has been achieved in pro-
ducing such computational programs [30], although there are some complications. For example, the use of
the ordinary cellular methods [31] for FEA with few rods, and also for the peripheral cells of multirod FEA
may lead to inaccuracies in the calculation of the temperature of thefuel-element cladding of up to 20-30?C.
In our opinion cellular methods can be refined by solving the thermohydraulic problem for a number of
FEA cross sections while maintaining the average values of the thermohydraulic parameters in the cells found
by the cellular method of calculation. In such methods of analysis of multirod FEA, the cross section is as-
sumed uniform in each elementary cell (although mixing factors are also taken into account). In addition, for
cells of complex shape, which are typical for peripheral cells, and very full bundles, the azimuthal nonuni-
formities of the tangential stresses and the velocity fields, which in turn affect the coefficient of"friction and
the heat-transfer coefficient in a given cross section, must be very important.
Therefore, after calculations of the average parameters in the cells, it is expedient to perform additional
thermohydraulic calculations in the chosen cross sections of the FEA under the assumption of stabilization of
the flow, as for a channel of complex shape, using, for example, methods of the type proposed in [32]. In doing
this, iterative return to cellular methods is possible to correct the coefficient of friction and heat-transfer. co-
efficient, if this is necessary. The realization of this method is logical for FEA of evaporative channels also.
But once again it should be emphasized that the development of any cellular methods, including refined meth-
ods, requires careful experimental research with accurate diagnostics, which also may serve at the present
stage of development as the only means of producing high-stress FEA.
1, N. A. Dollezhal~, ~~Nuclear power and the scientific-technical problems in its development,n At. Energ.
44, No..3, 203 (1978).
2, V. E. Doroshchuk, Heat Transfer Crises during Boiling in Tubes [in Russian], Energiya, Moscow (1970).
3. V . N. Smolin, ^ Model of the heat-transfer crisis mechanism during the motion of a steam-water mixture
and methods of calculating crisis conditions in tubular fuel elements,n in: Seminar TF-74 [in Russian],
IAE, Moscow (1974), pp. 209-224.
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 Declassified and Approved For Release 2013/02/22 :CIA-RDP10-021968000300050006-1
4. P . L. Kirillov, "Contemporary ways of developing theories of heat-transfer crisis during boiling in
channels,^ in: Proc. Physics and Power Institute [in Russian], Atomizdat, Moscow (1974), p. 242,
5. A. V. Zhukovskii, V. B. Khabenskii, and G. S. Shvertsman, "Enhancement of heat transfer in two-phase
flows," Sudostr. Rubezhom, No, 2, 68 (1969).
6. N. I. Perepilitsa and A. P. Sapankevich, "Methods of increasing parameters during aheat-transfer
crisis," (Analytic review). Preprint FBI [in Russian], Obninsk (1976).
7. L. S. Tong, Boiling Heat Transfer and Two-Phase Flow, Krieger (1975),
8? Yu. D. Barulin, A. S. Kon'kov, and A. I. Leont~ev, "Study of heat-transfer enhancement and hydraulic
resistance in a model of a fuel element assembly of a boiling reactor," [3], pp. 323-334.
9. I, Vollradt, "Boiling water reactor with twisted tapes between the fuel rods," Atomwirtschaft, 12, 595
(1967).
10. A. Rosuel and X. Rouvillois, "Use of twisted tapes for increasing the power density of boiling-water
reactors," Energ, Nucl., 9, 496 (1967).
11, V. N. Smolin and V. K. Polyakov, "Heat-transfer enhancement in a rod bundle using local vortex gen-
erators," [3], pp. 307-312.
12. V. N. Sm~lin, V. K. Polyakov, R. E. Polyakova, and L. I. Ezhova, "Study of heat-transfer
enhancement in rod bundles with local vortex generators," Vopr. At, Nauki Tekh,, Ser. Reaktorostr,,
No, 1 (12), 15 (1976).
13. J. O. Cermak, R. F. Farman, L. S. Tong, et al., "The departure from nucleate boiling in rod bundles
during pressure blowdown," Trans. ASME, J. Heat Transfer, 92, 621 (1970).
14, E. Rosal et al,, "High-pressure rod bundle DNB data with axially nonuniform heat flux," Nucl. Eng.
Design, 31, 1 (1974),
15. F. T, Kamen shchikov, V. A. Reshetov, A. I. Emel~yanov, and T, A. Pikuleva, "Optimization of vortex
generators in centrifugal heat-transfer enhancers and centrifugal steam separators," Vopr. At. Nauki
Tekh,, Ser. Fiz. Tekh, Yad. Reaktorov, 3 (16), 15 (1981),
16. A. N. Ryabov, F. T. Kamen shchikov, V. N. Filippov, et al,, "Experimental study of heat-transfer crisis,
temperature behavior, and hydraulic resistance in rod bundles with heat-transfer enhancers," Vopr.
At, Nauki Tekh., -Ser. Reaktorstr,, 1 (12), 36 (1976).
17. A, I. Emel~yanov, F. T. Kamen shchikov, Yu. M. Nikitin, et al., "Experimental evidence of the possibility
of investigating heat-transfer crisis on shortened models of fuel element assemblies," Vopr, At, Nauki
Tekh,; Ser. Fiz. Tekh. Yad. Reaktorov, 4 (26), 14 (1982).
18. A . N. Ryabov, F. T. Kamen shchikov, V. N. Filippov, et al., "Investigation of heat-transfer crisis and
hydraulic resistance in rod bundles with heat-transfer enhancers," in: Fifth Minsk Conference on Heat
and Mass Transfer, Vol. 3, Part 2 [in Russian] (1976), pp. 28-39.
19. V. G. Aden and V. N. Filippov, "Fuel element assemblies of RBMK type reactors,^ Vopr, At. Nauki Tekh,,
Ser, Fiz. Tekh. Yad. Reaktorov. 1 (21), Part 1, 28 (1978).
20, V. Aden, V..Asmolov, T. Blagovestova, et al., "The study of heat exchange enhancement in models of fuel
element bundles with boiling coolant," in: Proc. Sixth Int, Heat Trans. Conf,, Vol, 5, Hemisphere Pub.
Co,, Toronto (1978), pp. 41-46,
21. V. G. Aden, I. A. Varovin, B. A. Vorontsov, et al., "Testing of RBMK-1500 cassettes in the reactor at the
Leningrad power station," At. Energ., 51, No, 3, 150 (1981),
22, A. I. Emel+yanov, F. T. Kamen shchikov, E. V. Stolyarov, et al., "Effect of heat-transfer spacer lattice
enhancers on heat-transfer crisis in steam generator channels," Vopr. At. Nauki Tekh.r Ser. Fiz. Tekh.
Yad. Reaktorov, 2 (31), 49 (1983).
23. V. V. Gur'ev, "High-temperature gas-cooled nuclear reactors," Vopr. At. Nauki Tekh., Ser. Fiz. Tekh.
Yad. Reaktorov, 1 (21), 117 (1978).
24. ~ . K. Kalinin, G. A. Dreitser, and S. A. Yarkho, Heat Transfer Enhancement in Channels [in Russian],
Mashinostroenie, Moscow (1981).
25. A. A. Zhukauskas, Convective Transport in Heat Exchangers [in Russian], Nauka, Moscow (1982).
26. M. Peehs, J. Lindgren, and P. Moser, "Gas-cooled fast breeder core element fabrication technology,"
Nucl, Eng. Design, 40, 101 (1977).
27. V. L. Lel' chuk, Yu. M. Nikitin, E. I. Pupkov, et al., ^Enhancement of convective heat transfer," Teplo-
energetika, 2, 57 (1980).
28. D. Wilki, "Calculation of heat transfer and flow resistance at rough and smooth surfaces in a single
passage,^ in: Proc. Third Int, Heat Trans, Conf., Paper No. 2, Chicago (1966), pp. 20-31,
29. Y. Barulin, A. Konjkov, A. Leontiev, et al,, "Enhancement of heat transfer in core models of water-
moderated, water-cooled, and gas-cooled reactors," XX JAHR Congress Proc, Subjects, Vol. 4, Moscow
(1983), pp. 580-589.
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 Declassified and Approved For Release 2013/02/22 :CIA-RDP10-021968000300050006-1
30. L. V. Zhukov et al,, Methods and programs of In-Channel Thermohydraulic Calculation of Fuel Element
Assemblies, Taking Account of Interchannel Interaction of Coolant (Analytic Review) [in Russian], FEI,
Obninsk (1980):
31, G. S. Mingaleeva and Yu, V. Mironov, ^ Thermohydraulic calculations of multirod heat-liberation piles
cooled by single-phase heat carrier," At. Energ., 48, No. 5, 303 (1980).
32. M, Kh, ibragimov, I. A. Isupov, L. L. Kobzar~, and V. I. Subbotin, "Calculation of tangential stresses
at the wall of a channel and the velocity distribution in a turbulent flow of. liquid," At, Energ., 21,
No.2, 101 (1966).
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 Declassified and Approved For Release 2013/02/22 :CIA-RDP10-021968000300050006-1
LETTERS TO THE EDITOR
STRESSES IN SPHERICAL FUEL ELEMENTS OF A
HIGH-TEMPERATURE GAS-COOLED REACTOR
(VTGR) AS A RESULT OF THE HEAT LOAD AND
RADIATION SHRINKAGE OF GRAPHITE
V. S. Egorov, V. S. Eremeev, UDC 621.039.546
and E . A . Ivanova
The core of high~emperature gas~ooled reactors (VTGR) developed in the USSR contains spherical dis-
persion type fuel elements with graphite cladding and a core of matrix graphite [1] . The fuel is UO2. The fuel
elements fall through special locks into the upper part of a spherical charge, and move downward. The fast
neutron fluence up to the end of the reactor operating life is ~ 1.5 x 1021 cm-2. The variations of the tempera-
ture Tc at the center of a fuel element, the temperature drop OT, and the integrated flux F of fast neutrons as
functions of the displacement of a fuel element along the axis of the reactor are described bythe data inTable 1 [1] .
Stresses arise mainly from the nonuniform variation of the volume over the cross section of a fuel ele-
ment, which is determined by the temperature distribution and the graphite shrinkage caused by neutron bom-
bardment. The computational model for the analysis of thermal stresses in an elastic isotropic body is based
on the familiar linear relation between the components vik of the stress tensor and uik of the strain tensor [2]
rtik=-tiro (T ---To)Sik-~-tiu!(~ik-L '3f~ (u;k-- ~ dikui~) , (1)
where K = E/3 (1- 2v) is the bulk modulus; E, Young's modulus; v, Poisson's ratio; ?, shear. modulus; a,
volume coefficient of thermal expansion in the absence of stresses; T, temperature; and Sik Kronecker
symbol.
It can be shown that the form of Eq. (1) is preserved in the description of stresses resulting from arbi-
trary nonuniform variations of volume [3-5]. Without considering the specific cause of the nonuniform variation
of volume w(r, t), which in general is a function of the spatial coordinate r and the time t, we rewrite Eq. (1) in .
the form
fTilt =.ti lull-w)t1ik J'?.'.IL ~ uik- 3 bik ulll (2)
Equation (2) was used in [6] to analyze the stresses in graphite-based spherical fuel elements.
We assume that a fuel element is displaced in a core characterized by the parameters listed in Table 1.
Stresses are caused by thermal expansion 'and the nonuniform shrinkage of graphite under neutron bombardment.
The dependence of the relative change of the linear dimensions of graphite on the neutron fluence and the tem-
perature is shown in Fig. 1 [61.
The problem was solved by using the viscoelastic Maxwell model for a radially symmetric temperature
field. In view of the small concentration of fuel, it was assumed that the physical and mechanical properties of
thefuel-element core and cladding are those of graphite. The tangential stress v~ is the most dangerous, act-
ing in a direction perpendicular to the radius. The general solution. of the problem for a nonuniformly heated
sphere for constant values of Young's modulus, Poisson's ratio, and the viscosity has the form [2]:
a~~(r, t)=o41 (r, t)-~~-st f sc:;t'~4l (r, r")~t~~ (3)
u
where vel(r, t) is the solution of the corresponding problem for an elastic medium, s = E/6(1 - v)r~, and ~ is the
viscosity. The stress v~l(r, t) was calculated from awell-known formula [2] in which, following recommenda-
tions in the discussion of Eq. (2), the temperature variation of volume was. replaced by w(r, t):
Translated from Atomnaya Energiya, Vol. 57, No. 6, pp. 4151117, December, 1984. Original article sub-
mitted April 15, 1983.
0038~i31X/84/5706- 0846$08.50 p 1985 Plenum Publishing Corporation
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 Declassified and Approved For Release 2013/02/22 :CIA-RDP10-021968000300050006-1
TABLE 1. Calculated Variations of Tc, OT,
and F for a Fuel Element as Functions of its
Displacement Along the Axis of the Reactor
[1]
Distance from
level of upper
z
^~;
e'r, ^c
r? t u-zi,
charge,,?m
,
c
cm-z
(1,1
79(1
L10
0,06
0
3
96n
26U
0
18
,
11,5
1f~90
290
,
0,30
0,9
1220
270
0,60
1,3
1240
2(111
11,90
1,7
1230
1311
1,10
2,1 ~
1230
80
1,30
3,7
1l2u
45
1,50
S,3
1220
l5
1,55
3,95
1'110
15
1,60
0,4 QB 71 76
F~ 90-;' cm r
.Fig. 2
Fig. 1. Relative variation of linear dimensions of Mark A3
graphite under neutron bombardment [6]
Fig. 2. Variation of the' linear coefficient of thermal ex-
pansion 1) on the surface and 2) at the center of a fuel ele-
ment [6] .
R
'~ 1
~l (r, t1= F r , ~o(r, t)r~'dr-~ a J w(r, t)r=dr-ta(r. t) ,
u
where r is the running radius, and R is the outside radius of the fuel element.
The variation of volume w(r, t) in this case is made up of the thermal expansion and the nonuniform graph-
ite shrinkage. The rate of viscous flow u11 of the material under a uniaxial load under irradiation is
~ri~-~(13T~ tir~)Rii,
where BT is the creep constant in. the absence of neutron bombardment, Kp is the radiation creep constant, and
n is the neutron flux.
The viscosity is determined [7] in terms of BT and Kp by the formula
At a temperature ? 1300 1500?C graphite shows practically no thermal creep (BT = 0) [8]. Therefore,
taking account of Eq. (6) and BT = 0, the integration with respect to time in Eq. (3) can be transformed to an in-
tegration with respect to the fluence F:
o,i (r, t)='Rel (,? H') ._~,-t~H' f /~cl'r~ael (r, F')dh",
0
where p = EKp/2(1 - v).
Strictly speaking, Eq. (7) was derived by assuming that the parameters E, v, and ~ are constant in the ab-
sence of stresses at time t = 0. Calculations similar to those in [2] show that the expression under considera-
tion retains its form also when there are stresses at t = 0, if they satisfy the equations of thermoelasticity.
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~s ~o e -
Fig. 3
Fig. 3. Dependence of relative change of the elastic
modulus of A3 graphite on the fast neutron fluence for
a constant irradiation temperature [6] . .
Fig. 4. Variation of stresses 1) on the surface and 2)
at the center for motion of a spherical fuel element
along the reactor axis.
This permits astep-by-step calculation by dividing the total fluence into several intervals . The computational
procedure was reduced to the following. Suppose the stress distribution at the k-th step is known. Then the
problem is reduced to the determination of the stresses at the end of the k + 1-th interval, with the physical and
mechanical properties changed in accordance with the data of Table 1 and Figs. 1~ [6]. All the parameters were
assumed independent of the fluence within the limits of the chosen interval, which corresponds to approximating
the time dependence of the parameters by a step function. To ensure the continuity of the stresses, at each new
step the initial conditions were taken as the stresses accumulated at the end of the preceding step, multiplying
them by the ratio of the elastic modulus in the k + 1-th interval to that in the k-th interval.
Poisson's ratio and the radiation creep constant were assumed constant: v = 0.2, Kp = 5 x 105 (neu-
trons/cmZ ?MPa]-1: the radius of the fuel element R = 3 cm, and the initial value of Young's modulus E= 5
GPa. In spite of the fact that E varies strongly with the fluence, its variation with the radius was small, -and
inthe most unfavorable case for a neutron fluence of X1021 cm Zit did not exceed 20%> which lies within the limits
of experimental errors in the measurement of the mechanical properties of graphite. Thus, the maximum error
of the calculation made with the simple formula. (7), which is valid when Young's modulus is independent of the
spatial coordinate, does not exceed 20%.
The temperature distribution over a cross section of a fuel element was described by a formula corre-
sponding to the solution of the steady-state heat-conduction equation with the thermal characteristics of the fuel
element in the interval under consideration assumed approximately constant:
z (8)
The values of Tc and DT were taken from Table 1. Certain difficulties arose in estimating Young's
modulus because of its strong dependence on irradiation. The data in Fig. 3 [6] were obtained for a constant
temperature. In order to take account of the simultaneous variation of the temperature and fluence, we con-
sider avery simple model of the effect of radiation on mechanical properties.
The time dependence of Young's modulus is determined by the rate of radiation damage and the rate of an-
nealing of defects
d8 B _
dt + r (T) A (T) ~~, (9)
where r(T) is the relaxation time of the defects, A(T) is a parameter numerically equal to the change of Young's
modulus per incident neutron, and n is the neutron flux.
The solution of Eq. (9) has the form
1 ~ I T
h.' "= L'in -Fc ? J Aeon dt".
0
where A=ay.
848
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(10)
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Fig. 5
Fig. 5. Variation of thermoelastic stresses 1) on the sur-
face and 2) at the center of a spherical fuel element.
Fig. 6. Residual stresses 1) on the surface and 2) at the
center of a spherical fuel element as a function of the fast
neutron fluence.
The integral in (10) contains twotemperature-dependent constants z and A, which were found by process-.
ing the experimental results for isothermal conditions (Fig. 3). Then, using the known values of z and A,
Young's modulus was calculated from Eq. (10) as applied to the trajectory of a fuel element as given by the data
in Table 1. The change in volume was determined as the sum of the thermal expansion (Fig. 2) and the graphite
shrinkage under neutron bombardment (Fig. 1). The calculated values of the tangential stresses acting on the
surface and at the center of a fuel element are shown in Figs. 4-6.
- The calculated variation of the stresses_under the experimental conditions is illustrated byahe data of
Fig. 4. For a neutron fluence of ~5 x 1020 cm 2 there is a tensile stress of X2.0-2.5 MPa on the surface, due
mainly to the heat load. Later, as shrinkage phenomena develop, which occur at a higher rate in a hot zone,
the outer surface becomes compressed, and the central part of the fuel element is expanded, with the maximum
value of v~ = 7.5 MPa.
Additional thermal stresses appear at reactor shutdown. As aresult the residual stresses will be
The values of Q~l calculated with Eq. (4) are shown in Fig. 5, and the residual stresses in Fig... 6:
The most dangerous situation arises after the neutron fluence reaches 9x 1020 cm-2. In this case. (Fig. 6) the
central part of the fuel element is under uniform tension with v~ = 11 MPa. If E _ .7 MPa initially, the maxi-
mum value of Q~ at the center of a fuel element reaches 16 MPs.
The main error in the calculations arises from the choice of the radiation creep constant -and the values
of the graphite shrinkage under irradiation. Shrinkage phenomena depend strongly on the technology of-the
fuel element manufacture. The data in Fig. 2 refer to foreign Mark A3 graphite. Nevertheless, analysis of the
stress-strain state by the method described is of definite engineering interest during the stage of design and
development of the technology of spherical fuel-element manufacture.
Calculations with various values of Kp gave the following results. For an increase in Kpby a factor of
2-3 above the nominal value, the maximum stresses in Fig. 6 are shifted to the left to fluences of (3-4) x 1020
cm-2. The stresses themselves are only slightly changed. If Kp is decreased by a factor of 2-3, the maximum
values in Fig. 6 are shifted to the right into the region of higher neutron fluence, while the character of the
variation of Q~ with F remains unchanged.
The maximum value of Qcc~pp at the center of a fuel element, equal to 11 MPa (Fig. 6), is smaller than the
flexural-strength, which is 30-33 MPa [9]. Consequently, one should expect that a fuel element will be suffi-
ciently stable against stresses arising from temperature distributions and shrinkage phenomena. In-pile
tests of fuel elements of the construction being considered confirmed their suitability under operating conditions
in the VG-400 reactor [9] .
Thus, we have developed an algorithm- for calculating stresses in graphite-base spherical fuel elements,
taking account of the temperature distribution, the nonuniformity of the shrinkage of the material, and radiation
creep. We have determined that the maximum tensile stresses, as applied to the VG-400 reactor, develop in .
the central part of the fuel element after core shutdown, and, according to the calculations, lie within admissible
limits.
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 Declassified and Approved For Release 2013/02/22 :CIA-RDP10-021968000300050006-1
1. A. N. Protsenko,and I. G. Belousov, "Basic requirements on nuclear power sources for technological
factories and high-temperature gas-cooled reactors," in: Atomic-Hydrogen Power and Technology [in
Russian], No. 2.,Atomizdat, Moscow (1980), pp. 5-57.
2. B. Boley and J. Weiner, Theory of Thermal Stresses, Wiley, New York (1960) .
3. B. Ya. Lyubov and N. S. Fastov, "Effect of concentration pressures on diffusion processes in solid solu-
tions," Dokl. Akad. Nauk SSSR 84, 939 (1952).
4. Ya. S. Podstrigach, "Diffusion theory of the anelasticity of metals," Zh. Prikl. Mekh. Tekh. Fiz., 62,
6? (1965).
5. V. S. Eremeev, "Phenomenological analysis of diffusion in metal alloys," Fiz. Met. Metalloved., 42, 231
(1976) .
6. Elers, "Stress in spherical fuel elements of ahigh-temperature reactor," in: Physical and Mechanical
Research and Structural Design of Atomic Industry Plants, No. 1 [in Russian], NIIINFORMTYaZhMASh
(1975).
7. V. S. Eremeev and A. I. Gudkov, "On the relation of the rheological and microscopic models of sintering,"
Poroshk. Metall., No. 3, 31 (1978).
8. Properties of Carbon-Base Structural Materials [in Russian], Metallurgiya, Moscow (1975).
9. A. S. Chernikov, V. S. Kolesov, L. N. Permyakov, et al., "Design and basic characteristics of VTGR fuel
elements," Vesti Akad. Nauk BSSR, Ser. Fiz.-Energ. Nauk,. No. 3, 77 (1982).
DEVELOPMENT OF AN OIL-FREE FOREVACUUM UNIT
FOR AN OPERATING PRESSURE OF 150-300 Pa
FOR THE TOKAMAK-15
I. A. Raizman, V. A. Pirogov, L. G. Reitsman,
E. A. Maslennikov, and V. V. Martynenko
A vacuum is one of the necessary conditions for the operation of many physics facilities. Multistage vacu-
um units, in which forevacuum pumps produce a pressure of 1-100 Pa, are necessary for attaining the required
rarefaction and Roots twin-rotor, turbomolecular, booster, sorption, and other types of pumps are used for fur-
ther rarefaction. These pumps cannot operate at a pressure appreciably above T00 Pa.
The forepumps used at present operate with vacuum oil. The entry of oil vapor into the spaces undergoing
evacuation lead to undesirable consequences. To prevent oil vapor from seeping in, traps of various types are
installed in these units ahead of the forevacuum pumps but these traps are not sufficiently reliable.. The develop-
ment of oil-free forevacuum units thus is an important task for vacuum engineering.
Multistage steam ejector pumps are oil-free, but they cannot always be used; water-ring pumps are also
oil-free. The maximum vacuum that these pumps can produce, however, is 2.5~ kPa. If they are to be used as
forevacuum pumps, this limit must be lowered by a factor of roughly 50-100. This can be done by connecting a
steam-air ejector pump to a water-ring pump (Fig. 1). The ejector is connected to the intake fitting of the
pump, in which pressure is maintained at 10 kPa. Air enters into the active nozzle AN directly from the atmo-
sphere. The passive nozzle PN is connected to the space being evacuated. The two streams enter the mixing
chamber MC, in which the velocity fields are equalized. The gas then goes to the diffuser D. The ratio of the
pressure of the working gas to the pressure of the gas at the ejector exit is Pi~Pa ~ 10, which ensures normal .
operation of the ejector.
Figure 2 presents the characteristics of steam-air ejectors of some foreign firms. Their capacity is
50-70% of that of a water-ring pump. The capacity of the second stage of a two-stage ejector can also be 50-
70% of that of the first stage. As a result, the total capacity of the system is 25-50% of the capacity of the ini-
tial pump. It is evidently because of this that multistage ejectors are not made in other countries.
Translated from Atomnaya Energiya, Vol. 57, No. 6, pp. 417-419, December, 1984. Original article sub-
mitted December 12, 1983.
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 Declassified and Approved For Release 2013/02/22 :CIA-RDP10-021968000300050006-1
RC
AN
Fig. 1. Diagram of steam-air
ejector: RC) receiving cham-
ber; AN) active (operating)
nozzle; PN) passive nozzle;
MC) mixing chamber; D) dif-
fuser; 1, 2, 3, 4) cross sec-
tion of the ejector.
VN-6
Fig. 2 Fig. 3
Fig. 2. Characteristics of ejectors: 1) WN2-12 with aone-stage ejector; .2) WN2- .
12 with atwo-stage ejector; 3) LHG14 WEDAG; 4) 2BA 322 Siemens; 5) L6624 SIHI';
6) CL703 NASH; Q is.the ratio of the ejector capacity to the capacity of the water- .
ring pump at 10 kPa; here and in Fig. 4 p2 is the suction pressure.
It is our hypothesis that the low capacityof these pumps is due to the fact that their theoryf has not been
,developed sufficiently. The complexity and inadequate accuracy of the calculations for the design of ejectors
with conical mixing chambers have been pointed out in [1]. Under the. conditions under consideration the ratio
of the total pressure of the working gas and the-gas being pumped off reaches a value of 1000 or more. Without
ejector design calculations of an acceptable degree of accuracy and also without optimization of the ejector pa=
rameters, therefore, a high ejector capacity cannot be attained. We give a system of equations that describes
the operation of an ejector with a cynical mixing chamber;
~~(,'+~-1)(?U~ 1) Lz (R;~)-~ %i.a,=z(1.,) f-nOz(i.x)+ ~ * ,
2 2 pi F'~ n (~a)
1 -1-}- z(7,~)-1 1 ~ 1
(f (~~.~) :. (~t)-` z (~1) )Ill ~~ (~1)
1 4 (Rz)
I
ry(al) .I '
]7~a ~) 9(R,)
where p is the pressure, T is the temperature, F is the cross-sectional area, Q is the pump capacity, k is the
isoentropic exponent, n is the ejection coefficient, B = TZ/Ti, a = Fi/F2, Ilo = pY/pz, a is the reduced velocity,
zGl) and' q(l) are gasdynamic functions, X is the reduced length of the mixing chamber, and k is a correction
factor which, according to experimental results, is equal to 0.9-1.0 (the subscripts 1, 2, 3, 4 correspond to the
cross sections of the ejector; see Fig.l); the prime denotes the choking cross section and the asterisk de-
notes drag parameters.
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 Declassified and Approved For Release 2013/02/22 :CIA-RDP10-021968000300050006-1
~0
E .Y
~ b'
d 4
0,2 Q4 Qb ~0 9,51'.,0 .1,0 S,0
pZ, kPa
Fig. 4. Design ( )and experimental (O) char-
acteristics of the two-stage steam-air ejector
of the Tokamak-15 forevacuum unit: a, b) first
and second stages of the ejector (QZ is the flow
rate of the pumped-off gas) .
For the operating conditions under consideration we have formulated a. problem of an optimum ejector
that is different from the traditional problem. An optimum ejector is one which, for the specific pump charac-
teristic Q4 = f(pl) at a specified suction pressure, ensures the maximum volumetric capacity for the gas being
pumped off. In carrying out the mathematical optimization of the ejector, we took into account the fact that its
operation is characterized by 12 independent parameters: five parameters (pressure and temperature of the
working gas and the gas being pumped off as well as the gas flow rate through the pump) are prescribed and we
have four independent equations. Thus, three parameters are free and determine the optimality of the ejector..
These could be the ratio of the areas of the characteristic cross sections of the ejector and the reduced veloc-
ities cx, air and 1~3. The optimization determines the geometrical dimensions of the nozzles for the working gas
and the gas to be pumped off as well as the size of the mixing chamber.
A procedure has also been developed for the calculation of the characteristics of the stages of an ejector.
Besides the parameters indicated above,, alsoknownarethe cross-sectional areas F1, FZ, F3, and F4 which makes
the problem uniquely definable.
Both calculations and experiments have shown that the characteristic consists of two branches. If the
suction pressure is lower than the calculated value,-the velocity of the pumped-off gas in the choking cross sec-
tion is critical and if this pressure is higher than the calculated value, the gas velocity is subcritical. There-
fore, the characteristic has a peak at the calculated point. The second (from the pump) stage of the ejector is
calculated with the aid of the same procedure, taking the characteristic of the first stage is assumed to be the
initial characteristic.
A two~tage ejector makes it possible, using awater-ring pump, to obtain a pressure of ~ 20000 Pa in
the operating regime and 70 Pa in the limiting regime. Instead of awater-ring pump, any other forevacuum
pump that ensures a pressure of ~ 10 kPa can be used.
A forevacuum unit consisting of two RVN-6 vane-type rotary pumps and a two stage steam-air ejector
(Fig. 3) has been developed for the Tokamak-15. The casings of the ejectors were made of D 16 Dural and the
ejector nozzles were made of Kh18N10T steel. The characteristic of the unit is given in Fig. 4. Its ultimate
vacuum is ?r 100 Pa, and the pumping rate at X300 Pa is above 100 liters/sec, which is commensurate with the
pumping rate of the NVZ 150 pump at this pressure. The unit can pump off any gas, including corrosive gases.
The ejector does not require cooling and the electrical consumption is determined in practice by the RVN-8
pumps .
It is proposed that the forevacuum unit with atwo-stage steam-air ejector be used for pumping down the
T=15 cryostat with an effective volume of ~ 350 m3, from atmospheric pressure to about 100 Pa. The unit has
been pretested on a cryostat designed for testing superconducting magnetic coils of the longitudinal field of the
T=15, which has an effective volume of ~ 90 m3. A pressure of ~ 100 Pa was obtained in the test cryostat in
5 h.
The experience gained from prolonged operation of water-ring pumps with steam-air ejectors has been
positive [2]. The water-ring pumps manufactured at this time enable units with a capacity of 5000 liters/
sec to be built.
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 Declassified and Approved For Release 2013/02/22 :CIA-RDP10-021968000300050006-1
LITERATURE CITED
1. G. N. Abramovich, Applied Gas Dynamics [in Russian], Nauka, Moscow (1969), p. 457.
2. Yu. M.Prilutskikh, I.A. Raizman, A. D. Chuchuryukin, and M. V. Demchenko, "Introduction of a vacuum
system based on water-ring pumps into the production of titanium ingots," Tekhnoi. Legkikh Splavov, No.
11-12, 37-40 (1982).
SURFACE TENSION OF MOLTEN MIXTURES OF
FLUORIDES OF LITHIUM, BERYLLIUM,
AND THORIUM
A. A. Klimenkov, N. N. Kurbatov, UDC 532.61:546.161-143
S. P. Raspopin, and Yu. F. Chervinskii
Melts of mixtures of fluorides of lithium, beryllium, thorium, and uranium satisfy most completely the
many requirements imposed on fuel composites and breeder-zone materials in a liquid-salt nuclear reactor [1],
and therefore there is a need for information on the physicochemical properties of such melts.
The surface tension of three-component melts containing fluorides of thorium, beryllium, and lithium were
measured by the maximum-pressure method in a gas bubble. The salts investigated were placed in giass~ar-
bon crucibles. The material used for the capillaries was nickel, which is resistant to the action of fluoride
melts. As the working gas, we used argon from which the traces of moisture and oxygen had first been re-
moved .
In determining the surface tension, we used data on the density of melts taken from [2] . The surface ten-
sion was calculated according to the Cantor-SchrSdinger formula [3] . In our investigations, we paid close at-
tention to the preparation of the appropriate anhydrous salts. The anhydrous lithium fluoride was obtained by
air-drying the initial chemically pure salt by successive gradual heating in a vacuum to the melting point and
slow crystallization in an argon medium. The thorium tetrafluoride was obtained by the method described
earlier in [4] . The technically pure beryllium flupride was subjected tovacuum distillation at 1400?x. The
melting points of the resulting salts, determined by the method of differential-thermal analysis, are in good
agreement (f 2 ?K) with the most. reliable published data I5] .
The surface tension of the melted mixtures of the three-component system LiF-BeF2-ThF4 was studied
by the method of polythermal sections made frpm the beryllium-fluoride vertex of the concentration triangle to
the opposite side. In the system we investigated seven such sections in each of which the ratio of the lithium-
fluoride concentration to the thorium-fluoride concentration remained constant. The results of the measure-
ments were processed by the method of least squares. The temperature dependence of the surface tension of
each melt in the temperature intervals investigated is described fairly satisfactorily.by linear equations whose
coefficients are shown in Table 1. The table also gives the mean-square deviations S of the experimental values
from those calculated by using the equations, and the temperature intervals of the measurements are indicated.
The surface tension of the molten mixtures within the limits of each polythermal section decreases sub-
stantially as the beryllium~luoride concentration increases, and this suggests that it is surface-active with re-
spect to the other components of the mixture. After the minimum values have been reached (the mole fraction
of the BeF2 is 60%), the surface tension begins to increase, which is apparently due to the process of polymeri-
zation and the formation of vitreous structures in the melts with a high percentage of beryllium fluoride, char-
acterized by its polymer nature [6]. Figure 1 shows the curves of equal surface tension of the melts of the in-
vestigated system in the plane of the concentration triangle. As the concentration of thorium tetrafluoride de-
creases, the surface tension in the melts of the system decreases as well, and the slowest rate of decrease is
observed in the regions with a high concentration of lithium fluoride. The nature of the variation of surface
Translated from Atomnaya Energiya, Vol. 57, No. 6, pp. 419-420, December, 1984. Original article sub-
mitted January 1, 1984.
003831X/84/5706-085308.50 ?1985 Plenum Publishing Corporation
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 Declassified and Approved For Release 2013/02/22 :CIA-RDP10-021968000300050006-1
TABLE 1. Surface Tension of Molten Mixtures of the LiF-BeF2-ThF4 System
Mole fraction, o
?~
? - A - B ? T,
z
m1/m
Temp. in-
Mole fraction, ~o
?
- AZ B? T,
~fm
Tem
in-
BeFg
I ThFq
I LiF
A I
B
s
terVal, ?K
BeFg I
ThP'q I
L1F
A I
B
s
p.
[erval,'K
10
9
81
333,0
0,1084
0,6
970-1170
40
24
36
329,4
0,1323
0
9
1(170-1220
20
8
72
322,0
0,1095
0,8
870-1170
~
20
311
323,6
0,1321
,
0
8
1020-1170
30
7
63
322,5
0,1212
0,9
870-1170
GO
16
24
328,8
(1,1412
,
0
1
1020-1170
40
6
54
313,0
0,1254
0,8
870-1170
10
45
45
398,2
0,1544
,
0
6
11711-1220
50
5
45
283,4
0,11(',8
0 8
'
810-1120
20
4U
40
393,4
0,1622
,
(1,7
1170-1220
60
4
36
312,7
0,1392
1
0
870-1120
311
35
35
378,0
0,1602
0,8
1170-12211
70
3
27
340,9
0,1618
1,4
870-1120
40
30
311
356,7
0,1514
0
1
1170-1220
10
18
72
358,0
0,1284
(1,8
970-1170
5(1
25
2Ci
339,5
11,1435
,
0,9
109;1-1170
20
16
64
352,0
0,1345
0,8
970-1170
Glt
211
Zit
:;53,8
11
1593
1
2
1070-117(1
30
14
56
353,7
0,1451
1,0
970-1170
10
54
:3G
397,6
,
it,1512
,
11,8
1170-1271)
40
12
48
364,9
0,1682
0,9
970-1120
211
48
32
374,1
11,1427
1,11
117)-1270
50
10
40
336,1
0,1527
0,8
920-1120
30
42
28
361,0
0,1412
0,8
1121-1220
60
8
32
351,8
0,1693
1,0
920-1170
4(1
36
24
353,It
0,1445
0
i
1120-1220
1(t
27
63
390,7
0,1561
O,G
1021-1220
50
; 0
3
20
335,4
O,t358
,
1
1
1120--1220
20
24
56
378,9
0,1554
0,8
102n--1170
Grt
24
16
351,5
0
1526
,
0
1
1120-122~t
30
21
49
363,8
0,1538
(1,7
920-1170
10
67,5
22,~i
409,1
,
11
1564
,
0
8
1220-1321
40
50
18
15
42
35
345,9
0,1490
0,8
920-1171)
20
3
611
20
396,11
,
O,i575
,
0,9
1220-131111
326,0
0,1384
1,0
'970-1120
0
52,5
17,5
376,9
0
1524
1,0
1170-12711
6U
i2
28
338,8
0,1532
1,1
970-1120
40
45
i5
353,5
,
0,1412
0
8
1150-1270
10
36
54
383,5
0,1445
1,1
10211-1220
50
37,5
12,5
338,9
0,1324
,
0
1
1120-f221~
20
32
48
362,6
0,1378
0,8
1070-122(1
60
3a.
111
338,8
0,1358
,
1
1
1120-1221
30
28
42
345,3
0,4344
1,0
1070-1220
70
22,5
7,5
358,5
0,1513
,
1,0
1070-122'1
LiF Mole fraction, ?Jo fieFz
Fig. 1. Curves of equal sur-
face tension (v, mJ/m2)of melts
of the Lif-BeF2-ThF4 system
at 1130?x.
tension as a function of the concentration indicates that in relation to the thorium tetrafluori'de, the beryllium
fluoride has a higher surface activity than the lithium fluoride.
LITERATURE CITED
1. V. L. Blinkin and V. N. Novikov, Liquid-Salt Reactors [in Russian], Atomizdat, Moscow (1978).
2. A. A. Klimenkov, N. N. Kurbatov, S. P. Raspopin, and Yu. F. Chervinskii, Density of Molten Three-Com-
ponent Mixtures of Fluorides of Lithium, Beryllium, and Thorium [in Russian].. Submitted article No.
578KhP~2. Cherkassy: Division of the Scientific-Research Institute of Technical and Economic Inves-
tigations of the Ministry of the Chemical Industry, May 17, 1982.
3. V. K, Semenchenko, Surface Phenomena in Metals and Melts [in Russian], Gostekhteorizdat, Moscow
(1957) .
4. V. N. Desyatnik, A. A. Klimenkov, N. N. Kurbatov, et ai., "Density and kinematic viscosity of the melts
NaF-ThF4 and KF-ThF4," At. Energ., 51, No. 6, 390-392 (1981).
5. Handbook of Molten Salts [Russian translation], Vol. 1, Ktlimiya, Leningrad (1972). ,'
6. A. Rahman, R.Fower,and A. Norten, J. Chem. Phys. 52, No. 7, 3010-3012 (1972).
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 Declassified and Approved For Release 2013/02/22 :CIA-RDP10-021968000300050006-1
EXPERIMENTAL STUDY OF THE INTERACTION OF
PULSATIONS OF THE NEUTRON FLUX AND THE
COOLANT FLOW IN A BOILING WATER REACTOR
In [1], based on calculations utilizing a refined model of the dynamics of a boiling-water reactor [2, 3], it
is concluded that the interaction of the neutron flux and the coolant flow plays an important role in the mecha-
nism of high-frequency (HF) resona.nt instability of the VK-50 boiling-water reactor. In order to confirm this
conclusion and to check the working model, signals from probes measuring the flow rate of the coolant and the
neutron flux were recorded simultaneously (with the help of a magnetograph) in experiments performed in 1981
on driving the VK-50 reactor into the HF resona.nt instability regimes. Estimates were then obtained for the
statistical characteristics of the, pulsations of the flow rate and of the neutron flux, including the cross-correla-
tion functions and coherence functions. The basic results of these studies are reported in this paper.
The reactor was driven into the HF resonant instability regime by gradually increasing its head load while
maintaining the pressure in the range p = 1.0-3.0 MPa. As the so-called extrapolated linear boundary of in-
stability was approached [ 4] ,the amplitude of the pulsations of the neutron flux reached 2030%. Two-blade
flowmeters (TF), based on the work performed at the G. M. Krzhizhanovskii Power-Engineering Scientific-Re-
search Institute [5], were used as probes for measuring the flow rate of the coolant. The signal from the TF is
excited with the help of a permanent magnet, fastened onthe axis of the rotor, and an induction coil with a mag-
netic core. We positioned the flowmeters at the inlet to several fuel assemblies at different distances from the
axis of the active zone. The rate of rotation of the rotors of the TF varied from 15 to 30 Hz under the experi-
mental conditions.
To convert the primary signal from the TF into aflow-rate signal, we used a fast frequency-to-voltage
converter developed at the I. V . Kurchatov Institute of Atomic Energy, whose principle of operation is based on
the measurement of each half-period of the input signal: The converter contains an internal digital computing .;
setup and analog setups at the input and output. The neutron-flux signals were obtained simultaneously from an
ionization chamber (IC), positioned outside the reactor, and from adirect-charging probe (DCP), positioned at.
the center of the active zone.
We performed the sampling estimates of the main statistical characteristics of the pulsations of the flow
rate and of the neutron flux with the help of apparatus manufactured by the Honeywell Company (SAI-43A correla-
tion meter with aSAI-470 Fourier transformer). We estimated the coherence functions using the technique de-
scribed in [6]. The duration of the realizations processed varied from 7 to 20 min and the sampling time was
0.05 sec. At the input to the correlation meter, all signals were passed through identical low-frequency and
high-frequency filters with a transmission band of 0.1-4 Hz.
Figures 1 and 2 show typical examples of the estimates of the spectral power densities (spd) and the co-
herence functions of the flow rate and of neutron flux for a state close to the linear stability boundary.* It is
evident that the spd of the flow rate, just as for the neutron flux, has a resonant peak at-the frequency fr ~ 0.85
Hz, but does not decrease as rapidly away from the resonant peak. The flow rate and the neutron flux have a
significant coherence at a frequency close to the resonant frequency. At f = 0.75-0.95 Hz, the coherence func-
tion reaches a value of 0.8-0.95, but away from the resonant frequency the coherence of the pulsations of the
flow rate and of the neutron flux decreases markedly. This, in all probability, is associated with the presence
of components in the flow- rate signal measured which are not correlated with the neutron flux (turbulent pulsa-
tions of the flow velocity of the coolant in the fuel assembly, noise in the measuring channel, etc.)
*The form of the autocorrelation function (ACF) of the neutron flux is close to o2Ne-?~T~ coswoT (for the state un-
der studyo2N = (15%)2, ~ = 0.22 sec-1; Wo = 5.6 sec-1). The ACF of the flow rate and the cross-correlation func-
tion of the flow rate and neutron flux differ in form in that the first has a sharp .peak near the point r= 0, where-
as the peak in the second is shifted to the right by 0.25 sec.
Translated from Atomnaya Energiya, Vol. 57, No. 6, pp. 420-422, December, 1984. Original article sub-
mitted March 14, 1984.
0038-531X/84/5706-0855$08.50?1985 Plenum Publishing Corporation
Declassified and Approved For Release 2013/02/22 :CIA-RDP10-021968000300050006-1
 Declassified and Approved For Release 2013/02/22 :CIA-RDP10-021968000300050006-1
y 2 3
Frequency; Hz
Frequency, Hz
Fig. l Fig. 2
Fig. 1. Normalized estimates of the spectral power densities of the neutron flux (1)
and coolant flow rate in the fuel assembly (2) (obtained using the signals from the
TF ,in the third row of fuel assemblies and from the IC for a state with a pressure
of 2.5 MPa and a heat load of 137 MW).
An analysis of the estimation obtained for the statistical characteristics of the pulsations of the flow rate
and of the neutron flux made it possible to check the working model of the reactor dynamics [2, 3] used in [1].
For this, with 'the help of the working model, we checked the ratio of the amplitudes and phases of periodic
oscillations of the flow rate of the coolant and neutron flux in states at the boundary of linear stability (the
pressure and heat load were given in accordance with the linear boundary of stability of the reactor determined
from the experiment using the procedure described in [4]; exact coincidence of the computed stability bound-
ary with the experimental boundary was achieved by adjusting the thermal resistance of the gaseous gap in the
fuel elements) . The computed ratio was compared with the measured ratio of the amplitudes and phases of
the components of the pulsations of the flow rate and neutron flux with frequency fr for a number of states close
to the stability boundary. The latter was estimated from the equations:-
AN ? I ScN ~If[) I .~. 'CNN ~f [)_
KNC= f1 c; "' Scr.Ur) I S.vr,Ulr) I
Dy'~,r=arg ScN (II r) z -'Ln,Di /r
~SrPNU - sQ~N - S(('Qi
AN and AG are the amplitudes while ~ and cpG are the phases of the oscillations of the neutron flux and flow
rate; SGN is the cross spectral power density of the flow rate and of the neutron flux; SrJrJ and SGG are the
spd of the neutron flux and of the flow rate; Oz is the time shift, corresponding to the peak in the cross-cor-
relation function of the flow rate and the neutron flux; b ~ and bcpG are the phase shifts in the measuring chan-
nels for the neutron flux and flow rate*; and, the symbol ~ denotes estimated quantities.
*For the channel from the IC bcpN ~5?; for the channel from the TF bye ~5?; and, for the channel from the DCP
N
bcpN ~ 20-25?. The large phase shift in the signal in the channel from the DCP is a result of the uncertain selec-
tion of the parameters of the correction (phase-lead) feedback device at the output of the channel (we estimated
b~ after the experiment by a test method on a stand using a sinusoidal signal generator). The dynamic attenua-
tion of the signal amplitudes at the frequency fr for all measuring channels was insignificant ( 42 -S. Mukhammedov,
A. Vasidov, and E. Pardaev ....................................... .
LETTERS TO THE EDITOR
Results of Measurements of the Neutron Field in the Channels
of the VVi`ft-1000 - S. S. Lomakin, A, G. Morozov,
G. G. Panfilov, V.P. Kruglov, and G. M. Bakhirev ......................... .
Amplitude Characteristic of Pyroelectric Detectors - V. A. Borisyonok
and E. Z. Novitskil .............................................. .
Engl./Rues.
13
15
19
20
26
25
30
29
34
39
32
36
44
40
48
43
52
47
56
50
59
54
61
55
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 Declassified and Approved For Release 2013/02/22 :CIA-RDP10-021968000300050006-1
Engl./Ruse..
Control of the Energy Distribution and Safety of a VVER-1000
Reactor When Operating in the Variable Mode
- E. V. Filipchuk, V. A. Voznesenskii, V. G. Dunaev,
I. A. Luk'yanets, V. I. Mitin, P. T. Potapenko,
and E. S. Timokhin
65
67
Multichannel Correlation System for Measurement of the Coolant
Flow Rate for High-Powered Water-Cooled Channel Reactors
(RBMK) - V. M. Selivanov, B. V. Lysikov, N. P. Karlov,
B. A. Kuznetsov, A. D. Martynov, V. V. Prostyakov,
D. Pallagi, Sh. Khorani, T. Khargitai, and Sh. Tezher.
70
71
Stability of Natural Circulation in a Loop with Boiling of
the Coolant - V. I. Budnikov and E. F. Sabaev.
74
74
Methods of Calculation of the Stress-Strain State of Microfuel
Elements of High-Temperature Helium-Cooled Reactors and the
Choice of Their Design - N. N. Ponomarev-Stepnoi,
T. A. Sazykina, and N. I. Tikhonov
79
77
Sorption of Radioactive Iodine by Stainless Steel from a
Flow of Nitrogen Tetroxide Vapor - V. M. Dolgov
and M. A. Drugachenok.
84
81
Damage of the Surface of Structural Materials by the Action
of Plasmoids - V. I. Pol'skii, B. A. Kalin, P. I. Kartsev,
D. M. Skorov, E. P. Fomina, Yu. V. Skvortsov, N. Pi. Umrikhin,
F. R. Kh~..:idullin, S. S. Tserevitinov, 0. A. Kozhevnikov,
A. N. Lapin, and N. B. Odintsov
87
83
A ~~ectronhotometric Study of the Equilibrium in the Reaction
PuOz"~' + C1- ( + 1/2 Clz in a NaCl-2CsC1 Melt - S. K. Vavilov,
G. N. Kazantsev, and 0. V. Shishalov
.
93
88
Neutron Leakage from a Channel with a Low Density of
Fissionable
Material - B. P. :'ochurov.
98
91
New Radiation Safet;~ Standards for Tritium Compounds
- M. I. Balonov, I. A. Likhtarev, and Yu. I. Moskalev.
102
94
Radiation Capacity of a River Containing Little Drift on
Short-Tern .^.adionuclide Discharge - A. L. Kononovich
107
98
LETTERS TO THE EDITOR
Diamond Dosimeter of Pulsed Gamma Pays - Z. A. A1'bikov
and N. I. Terent'ev.
111
101
Error of the .Reich-Moore Formula for Neutron Cross Sections
- I. I. Surina
113
102
Application of Gas-Controlled Heat Pipes in Nuclear Technology
- I. G. Gverdtsiteli, A. G. Kalandarishvili,
V. A. Kuchukhidze, and P. D. Chilingarishvili.
115
103
Reactor Tests of a Neutron-Flux Regulation System Based
on_Layered Compounds of Pyrolytic Graphite with Cesium
- I. G. Gverdtsiteli, Yu. A. Bunchin, A. G. Kalandarishvili,
M. N. Korotenko, Yu. N. Krasik, S. D. Krivonosov,
B. A. Mskhalaya, and A. V. Nikonov
118
105
Hydrodynamic Characteristics of a Dispersely Annular Gas-Liquid
Flow in Annular Channels - G. V. Alekseev, N. A. Gromov,
Yu. I. Dzarasov, and S. Yu. Orlov. .'.
120
106
Radioactive Contamination Spread by Transportation Moving along
Roads - V. E. Vostrukhov and A. F. Lyzlov.
123
108
Determination of the Industrial Contamination of Soil with
137Cs on the Global Background by Paeans of the Analysis
of the Depth Distribution of Activity - A. N. Silant'ev,
I. G. Shkuratova, and G. N. Neushkina.
126
109
Behavior of Wa"ter-Cooled dater Moderated Fuel Elements in a
Y.ypothetical Accident with the Ejection of Control Rods
- A. I. Mysenkov and V. N. Proselkov
128
111
ARTICLES
Material Behavior Investigations of the VK-50 Reactor Fuel
Element Assemblies - V. A. Tsykanov, V..Y.. Shamardin,
A. B. Andreeva, G. P. Kobylyanskii, G. I. Maershina,
Yu. D. Goncharenko, R. E. Fedyakin, and V. P. Sadulin. .
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Engl./Rues.
Determining the Frictional Characteristics of Reactor Materials
- V. M. Shchavelin, A. V. Kostochka, A. M. Bolobolichev,
A. A. Kuznetsov, I. S. Golovnin, and Yu. K. Bibilashvili .
Modeling the Gaseous Swelling of Fuel Elements
- Yu. G. Degal'tsev, V. F. Kuznetsov, and N. N. Ponomarew
Stepnoi.
Neutron Cross Sections for the Calculation of the Damaging, Dose
in Reactor Materials - V. I. Avramenko, Yu. V. Konobeev,
and A. M. Strokova :
In Situ Examination of Radiation-Induced Internal Stress
Relaxation in the Column of a High-Voltage Electron
Microscope - I. I. Novikov, V. A. Ermishkin, V. G. Zharkov,
E. N. SamoiloV, I. S. Lupakov, and B. S.:Rodchenkov. .
Occurrence of Gas Porosity of Annealing Nickel Containing helium
- E. Ya. Mikhlin, V. F. Chkuaseli, Yu. N. Sokurskii,
and G. A. Arutyunova
Small Induction Motors for Nuclear Power Stations
- K. A. Alikhanyan
Automated System for Making Observations, EValuations,.and
Forecasts: a Foundation for Comprehensive Protection of the
Environ:acnt and of the Public liealth - E. I. Vorob'ev,
V. M. PrusakoV, and V. A. Minchenko. .. .
Radioactive Contamination of the Sea Environment Near the
Leningrad Atomic Energy Plant in 1982 - S. i4. Vakulovskii
and A. I. Nikitin.
Neutron Inspection of Moisture in Slightly Enriched UOz
- V. V. Frolov, V. I. Bulanenko, and V. V. Charycliansl:ii
Average Characteristics of the Slowing Down of Logo-Energy Electrons
in a Tissue-Equivalent Material - V. A. Pitkevich. .
Measurement of the Ratio of the Fission Cross Sections of aseU and
z3sU for Pleutron Energies in the Range 5.4-10.4 MeV
- A. A. Goverdovskii, B. D. Kuz'minov, V. F. Mitrofanov,
A. I.. Sergachev, S. M. Solov'ev, P. S. Soloshenkov,
and A. K. Gordyushin
Measurement of the Fission Croas?Section of 238U and?Z3bU :luclei
by 14 MeV Neutrons - A. A. ('.overdovskii, A. Y.. Gordyushin,
B. D. Kuz'minov, A. I. Sergachev, S. M. Solov'ev,
.and P. S. Soloshenkov.
"Gas Target" in the Divertor of a Tokamak - M. Z. Tokar'.. .
LETTERS TO THE EDITOR
y Radiation Field Generated by Neutrons in an Unbounded Uniform Air
24edium - A. V. Zhemerev,
Determining the Density of Pyrocarbon Coatings on Micropins by
Gasification in a Glow 13ischarge - A. A. Babad-Zakhryapin
and I. S. Alekseeva ...
Graphite Capsules for Accommodating Indicators of the Temperature
and the Neutron Flux in Irradiation Units
- T. N. Shurshakova, V: V< Gundorov, and K. V. Grigor'eva. .
Radiation-Induced Changes in the Thercaal Conductivity ,and the
Electrical Resistivity of pyrolytic Graphite
- Yu. S. Virgil'ev and I. A. Dmitriev.
Preliminary Filtering of the Results of Measurements Performed
in Exploratory Gamma-Logging - I. M..Y.haikovich and `J. N. Popov. .
Calibration of Individual Dosimeters by the Absorbed Dose of Photon
Radiation - Yu. P. Bakulin, V. P. Bashmakov, E. A. Bopdanov,
T. I. Gimadova, N. I. Muratov, and A. V. Tultaev .
Comparative Investigation of the Process of Defect Formation in SiO~
under Gamma and Gamma-tleutron Irradiation ~- I. Kh. Abdukadyrova. .
Vohime 56, Nmnbet 4 Apn'1, 1984
.
138
'134
. ~
141
137
.
144
139
147.
142
.
150.
144
155
148
157
149
162
153
165
155
16 8
158
173
162
176
164
17a
16s
1.88
17 3
.
192
175
194
177
196
17 7
198
179
201
180
203
182
ARTICLES
Physicochemical Interaction of Oxide Fuel with the Cladding
of the Fuel Elements of a Fast Reactor - V. A. Tsykanov,
E. F. Davydpv, E. P. Klochkov, V. R. Shamardin, V. N.
Golovanov, and F. N. Kryukov ............................................ 207 195
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 Declassified and Approved For Release 2013/02/22 :CIA-RDP10-021968000300050006-1
Engl./Ruse.
OSKh14MF Steel.Uaed in Nuclear Power - A. K. Mednikov, P. G.
Krutikov, V. M. Sedov, V. P. Borisov,: L. I. Loshkova,
and A. S. Zlobin ........................................................ 213 199
Analysis of the Leakage of Steam-Water Flow with Rupture
of the Heat-Carrier Circulation Loop of Nuclear
Reactors - Yu. V. Mironov and T. I. Fomicheva ........................... 217 202
Dynamic Model of a Mixture of Nonequilibrium Steam-Water
Flow, Uncondensed Gas, and Boric Acid - N. I. Kolev .................... 222 205
Passive Stabilization of Vertical instability in a Tokamak
Reactor - S. G. Bespoludennov, S. A. Galkin, L. M.
Degtyarev, and V. I. Pistunovich ........................................ 227 210'
THermal-Neutron Fission Cross Section of the Short-Lived
Isomer of 36Np - E. A. Gromova, S. S. Kovalenko,
Yu. A. Nemilov, Yu. A. Selitskii, A. V. Stepanov, A. M.
Fridkin, V. B. Funshtein, V. A. Yakovlev, G. V.
Val~akii, and G. A. Petrov .............................................. 230 212
Measurement of Burnup of .Nuclear Fuel in a Reactor
by Neutrino Emission - V. A. Korovkin, S. A. Kodanev,
A. D. Yarichin, A. A. Borovoi, V. I. Kopeikin, L. A.
Mika~lyan, and V. D. Sidorenko .......................................... 233 214
Irradiation of the USSR Population in Medical Diagnostic
Procedures - E. I. Vorob'ev, R. V. Stavitskii, V. A.
Rnizhnikov, R. M. Barkhudarov, V. N. Korsunskii, V. I. .
Popov, Yu. I. Tarasenko, V. A. Postnikov, N. V. Frolov,
and V. P. Sidorin ....................................................... 239 218
Behavior of Thorium in the Laterite Process - V. A. Kopeikin ................ 242 221
LETTERS TO THE EDITOR
Excitation of Surface Vibrations of Drops of a Radioactive
.Liquid - V. I. Ralechits, I. E. Nakhutin, and P. P.
Polu6ktov ............................................................... 246 224
Heating of the Focusing Optics of Laser Thermonuclear
Reactors by X-Ray Radiation from the Target - M. I.
Belousov, P. A. Grishunin, V. Z. Subbotin, and V. V.
,. .Kharitonov ........:..........:.......................................... 248 225
~An Evacuated Emission Detector with External Supply Source
for Recording y Radiation in Nuclear Reactors - S. V.
Chuklyaev and 0. Z. Shchetinin ????????????????????????????????????????? 250 226
Influence of y-Quantum Scattering upon the Development
of the Radiation Field of a Cobalt Isotope Unit -
V. F. Zinchenko and M, N. Lebedev ....................................... 253 228
Estimate of the Corrosion Resistance of Chrome~iicicel Steels
in Molten Fluorides - V. B. Kirillov, I. E. Lyubllnskil,
and N. M. Beskorovainyi?.?...??.?????????????????.?????????????????????? 256 230
Kinetics of the Behavior of Sodium Acetylide in a Sodium-
Mineral Oil System - Yu. P. Kovalev and N. V. Gavrilova ................ 258 231
Analysis of an Emergency Involving Rupture of the Main
Circulation Pipeline in the WER-1000 - A. I. Boyadzhiev
261 232
and S. I. Stefanova ............................ ......................
Shock Loads of Units within the Contai~ent in the VVER-1000
in the Initial Stage of an Emergency Involving Failure
in the Main Circulation Pipeline - A. I. Boyadzhiev
and S. I. Stefanova. .. .. ... .......... ........ 264 234
Neutron Energy Spectra in?BFS?Critical?Assemblies -?V.?V.
Vozyakov, E. N. Kuzin, and A. V. Shapar' ................................ 267 236
Measurement of the Thickness of the Deposits of Nuclear Fuel
- P. S. Otstavnov and V. P. Roroleva .................................... 270 238
Radiation-Stimulated Diffusion of Aerosols - I. E. Nakhutin,
P. P. Polu~ktov, and G. Yu. Kolomeitsev .......................:......... 271 239
Possibilities of Using Neurons of the Cosmic Background
for the Investigation of the Salt Content of Seawater
- E. M. Filippov .............................. ......................... 274 240
Combined Shielding for Monochromators of Crystal-Diffraction
Instruments - E. A. Garusav, I. V. Golosovskii, N. K.
Pleshanov, V. A. Trunov, A. K. Tsytaenko, and V. B..
Shchebetova .....................................:...................... 279 243
Measurement of Energy and Intensity of Principal '46Cm
a-Groups - V, M. Shatinskii ..................:.......................... 282 245
Application of Ferrosulfate Solution in Dgsimetric Research
on Reactor Beams - S. P. Kapchigashev, V. I. Potetnya,
and 0. I. Potetnya ...................................................... 283 246
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Engt./Rues.
ARTICLES
Quantitative Estimation of Nuclear Safety - B. P. Shishin.
287
275
Emissions of 6OCo, 11OmAg, and '4Mn at the Armenian Nuclear Power
Plant,. and Their Content in the Surrounding Atmosphere
- E. S. Saakov, A. A. Avetisyan, and K. I. Pyuskyulyan.
291
278
Unloading Additional Absorbers from the RBMK-1000 Core
- N. V. Isaev, V. E. Druzhinin, and Yu. V. Shmonin.
294
280
Dissolution of Oxide Films on Constructional Steels
- Yu. G. Bobrov, G. M. Gur'yanov, A. P. Kovarskii, Yu. P: Kostikov_
and A. V. Motornyi.
298
282
Influence of Cold Deformation on the Behavior of Helium in Steel
OKh16N15M3B - A. G. Zaluzhnyi, M. V. Cherednichenko,
0. M. Storozhuk, V. F. Reutov, and G. T. Zhdan.
304
286
Isotopic Composition of Fuel in the Blanket of a Hybrid
Thermonuclear Reactor with a Thorium Cycle - S. V. Marin
and G. E. Shatalov.
?
-307
289
Magnetic Fields
Recuperator with Inhomogeneous Electric and
- S. K. Dimitrov and Ya. A. Melnik :
?
311
291
a Horizontal Apparatus
Calculation of Model High-Level Wastes in
- V. V. Kulichenko, V. F. Savel'ev, V. A. Prokhodtsev,
and A. A. Ryabova
314
293
Concentration Ratios for Radiogenic Lead and Uranium in Aureoles
around Hydrothermal Uranium Mineralization - V. M. Ershov
319
298
Effect of Gamma-Neutron Radiation from a Nuclear Reactor on the
Electrical Stability of Microlite - N. S. Kostyukov,
M. I. Muminov, and V. M. Lanskov.
322
300
REVIEWS
The New Generation of Highly Charged ion Sources
- K. S .. Golovanivskii
326
303
LETTERS
Small-Scale System for the Formation of a Field of Irradiation
with Accelerated Electrons - 0. A. Gusev, S. P. Dmitriev,
A. S. Ivanov, V. P. Ovchinnikov, M. P. Svin'n, and M. T. Fedotov.
336
.311
Density and Surface Tension of Molten Mixtures of Uranium
Tetrafluoride with Lithium and Sodium Fluorides
- A. A. Klimenkov, N. N. Kurbatov, S. P. Raspopin,
and Yu. F. Chervi.nskii.
339
312
Influence of Grain Size and Doping with Boron on the Behavior of
Helium in Stainless Steel 16-15 - A. G. Zaluzhnyi,
M. V. Cherednichenko-Alchevskii, 0. M. Storozhukm
and A. G. Zholnin
341
314
Consideration of the Decay of ?BPu When Determining the Isotopic
Composition of Uranium Fuel of a Hybrid Thermonuclear Reactor
- S. V. Marin and G. E. Shatalov.
343
315
Static Instability of Once-Through Steam Generators with Convective
Heating - I. I. Belyakov, M. A. Kvetnyi, D. A. Loginov, --
and S. I. Mochan.
347
317
Cryogenic Loop for y-Ray Sources - I. I. Buzukashvili,
and G. S. Karumidze
,
.
?
351
319
Influence of Neutron Spectrum on Formation of
'B9U from
"'Th
- A. S. Gerasimov, G. V. Kiselev, and A. P. Rudik
353
320
Spectral-Angular Distribution of y-Radiation behind an Instrumentation
Unit - P, A. Barsov, V. M. Sakharov, and V. G. Semenov.
357
322
State-of~the-Art and Development Prospects for Nuclear Power Stations.
Containing Pressurized-Water Reactors (W BR) - G. A. Shasharin,
~. I. Ignatenko, and V. M. Boldyrev. 361 353
State-of-the Art and Development Prospects?for?Nuclear Power?Stations
Containing RBMK Reactors - E. V. Kulikov 368 359
Staterof-the Art and Development Prospects for Nuclear?Power?Stationa
Containing Fast Reactors - 0. D. Kazachkovskii 375 365
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PAGES OF HISTORY
Critical Assembly of the World's First Nuclear Power Station
- M. E. Minashin .
ARTICLES
Heat Accumulators at Nuclear Power Stations - V. M. Chakhovskii .
A Load-Following Atomic Heat and Power Plant - V. M. Boldyrev
and V. P. Lozgachev.
Computer-Assisted Radiation Tomography of Spherical Fuel?Elements
- E. Yu. Vasil'eva, L. I. Kosarev, N. R. Kuzelev, A. N. Maiorov,
and A. S. Shtan'
Physicochemical Approach?to the Description of?the?Distribution
of Macroquantities of Pu(IV) in Extraction by Tributylphosphate
from Nitrate Solutions in the Presence of Complex Formers
Applicable to the Regeneration of Spent Nuclear Fuel from Fast
Reactors - A. S. Solovkin and V. N. Rubisov. .
LETTERS TO THE EDITOR
Contribution of Nuclear Interactions to the Distribution of Absorbed
Energy in Thin Plates Bombarded with Fast Charged Particles
- S. G. Andreev, I. M. Dmitrievskii, and I. K. Khvoatunov. .
Stability of Scintillation Detectors Vis-A--Vis y Radiation
-'V. V. Pomerantsev, I. B. Gagauz, Yu. A. Tairlin,
and 0. V.:Levchina
Radiation Stability of Scintillating?Polystyrene -?I. B. Gagauz,
A. P. Meshman, V. F. Pererva, V. V. Pomerantsev,
and V. M. Solomonov.
Yield of Electron Bremsstrahlung?from Thick Targets - V.?I. Isaev
and V. P. Kovalev
Volume 57, Number 1 Jniy,1984
Engl./Ruse.
381 382
?
,
388
389
397
396
402
400
410
406
418
413
421.
415
423
416
425
417
ARTICLES
An Estimate of the Errors of a Method for Determining the Depletion
of Spent Mixed Uranium-Plutonium Fuel of Fast Reactors
- V. Ya. Gabeskiriya, V. I. Borisenkov, V. V. Kalygin,
M. N. Maslennikova, V. S. Prokopenko, V. M. Prokop'ev,
V. L. Chernyshev, and A, P. Chetverikov. ..
429
8
Nuclear-Physical Investigation of Mass Transport of Carbon
and Nitrogen by Sodium Coolant - 0, V. Starkov, I. V, Istomin,
V, A. Karabash, M. Kh. Kononyuk, A. N. Sosnin, and V. S. Shorin.
432
10
Critical Transportation Velocity of Suspensions in Heat-Carrier Flows
- V. V, Alekseev, F. A. Kozlov, and Yu. I. Zagorul'ko
437
14
Use of Detectors External to Reactor to Determine the Reactor Power
and the Mean Energy Distribution over the Height of the Active
Zone - A, N, Kamyshan, A, M. Luzhnov, A, S. Makhon'kov,
V. V. Morozov, N. S. Orekhova, and S. G. Tsypin
443
18
Buildup of Gaseous Nuclear Reaction Products in Chromium and Nickel
Cawed by High-Energy Electron Lrradiation - A, G, Zaluzhnyi,
O, M, Storozhuk, M. V. Cherednichenko-Alchevsk,ii, N, L, Emets,
L, Z, Ozhigov, Y~i, N. Ranyuk, and V, A. Yamnitskii.
448
21
Corrosion-Test Methods for Zirr_onium Alloys in a Research Reactor
- G, V, Samsonov, S, V, Seredkin, and V, N. Shulimov.
454
25
Purif.ir_ation of. the First Circuit of Nuclear Power Systems
with a Two-Component Coolant-Gas Flow - A. V, Beznosov,
P, N, Martynov, S, Yu. Orlov, and V. E. Serov
458
29
Construction of Automatic Systems for Radiation Monitoring
of the Environment of Nuclear Power Stations - E. P, Volkov,
A, I, Glushchenko, V, N, Durnev, V. V, Zhabo, M, I. Saparov,
and L. P, Kham'yanov,
462
32
Effect of Aluminum, Beryllium, and Trivalent Chromium Nitrates
on -the Extraction of Uranyl Nitrate with a 10% Solution
of Tri-n-Butylphosphate in n-Paraffins - B. S. Zakharkin,
T. A. Rumyantseva, and D. P. Adaev.
?
465
34
Uranium Blankets
Neutron Parameters Which Can'Be Reached in Fast
of Hybrid Fusion Reactors - G, E. Shatalov.
468
36
REVIEWS
Radiation Background in Living Accommodations - ~, M, Krisyuk
and V, I. Parkhomenko
475
42
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 Declassified and Approved For Release 2013/02/22 :CIA-RDP10-021968000300050006-1
Engl./Rues.
LETTERS TO THE EDTTOR
Change of Mechanical Properties of Loaded Steel Specimens Under
Electrolytic Saturation with Hydrogen - G. Biggiero
and A. Rorruto.
483
49
Er_onomic_ Aspects of the Use of Bremsstrahlung in Radiation
Technology - V. V, Krayushkin, A, V, Larichev,
and M. E. Tulupov
486
50
Effect of Lithium on Hot-Pressed Boron-Carbide Parts
- P, D. Kervalishvili and Sh. Sh, Shavelashvili
489
52
Diffusion of 1O3Ru and 'gZr Fission Products in Monocrystalline
Tungsten - D. K. Daukeev, Zh. R. Zhotabaev, R. T. Musurmankulov,
and N. A. Reutova
491
53
Optimizing Reactor Power Distribution with a Restricted Number
of Simultaneously Displaced Control Rods - A. M. Afanas'ev.
493
54
Anomalous Effects of Small Doses of Ionizing Radiation in Metals
and Alloys - I. P. Chernov, A. P. Mamontov, A. A. Botaki,
P. A. Cherdantsev, B. V. Chakhlov, and S. R. Sharov
497
56
Changes in the Structure of VK Alloy Produced by Low y-Ray Doses
- I. P, Chernov, Yu. A. Timoshnikov, A. P, Mamontov,
V, A. Korotchenko, I, A. Lapsker, and B. S. Semukhin.
499
58
Parameters of Neutron Resonances of 1OBCd - V. A. Anufriev,
S. I. Babich, V. N. Nefedov, and N. G. Kocherygin
502
S9
Measurement of the Fission Cross Section of 2 "Cm with Fast Neutrons,
Using a Nanogram Quantity of Isotope - P. E. Vorotnikov, -
L. I). Kozlov, and Yu. D. Nlolchanov.
504
61
Volume 57, N~ber 2 August, 1984
ARTICLES
An Experimental Study of Emergency Cooling Conditions in RBMK Reactors
on Load Disconnection - V, N. Smolin, V. I. Esikov, V. P, Shishov
S, P, Kuznetsov, and V, S. Grioor'ev.
Safety under Servicing Cooling Conditions for?RBMK Reactors
- V, N, Smolin, V, I, Esikov, V, P, Shishov, V, P, Vasilevskii,
and V? S. Grigor~ev
Gas Phase in Experimental?Fuel Elements with Compact Uranium Dioxide,
Irradiated in the Sm-2 Reactor - A, P, Kirillovich,
V, Sh. Sulaberidze, Yu, I, Pimonov, V. N, Shulimov, .
Yu, G. Lavrinovi,ch, A, S, Biryukov, and V, N, Kupriyanov. .
Influence of the Texture of Prismatic Planes on the Anisotropy
of Deformation of Irradiated Zirconium Alloys - Yu. N. Knizhnikov
and V, V, Kolomytkin.
Ion- and Electron-Stimulated Low-Temperature Desorption of Gases
Dissolved in Metals - N. P, Katrich and V, N. Kanishchev. .
Absorption Parameters of Deuterium Ions in Molybdenum - A. A, Pisarev
and V. N. Tsyplakov
Dehydration of the Steam-Generating Channel?Due to Cooling Loop?
Depressurization - E. I, Liverant, A, P, Proshutinskii,
and E. M, Staviskii
Studying the Electrophysical Parameters of Piezoceramics of?Various?
Types in an IW- 2M Reactor - Yu. P, Meleshko, S. V, Babaev,
S, G, Karpechko, V, I. Nalivaev, Yu, A, Safin, and V, M, Smirnov,
Use of Monoisooctylmethylphosphonic Acid and Its Trivalent Iron Salt
in Determining Radionuclides in Effluents - N, E. Tsvetaeva,
V, M, Filin, L, A, Ivanova, V. N. Revnov, E. P, Rodionov,
L, Ya, Rudaya, I, A. Suslin, and K. Yu. Shapiro .
A Method of Calculating Membrane-Element Cascades for Separating
Multicomponent Mixtures - E. B, Gruzdev, N, I, Laguntsov,
B, I, Nikolaev, A, P, Todosiev, and G, A, Sulaberidze .
LETTERS
Density of Melts of the Ternary Mutual System K, Kr//F, C1
- S. E, Darienko, N. N. Kurbatov, S? P, Raspopin,
and Yu. F, Chervinski i.. ... ..
Heat Exchange during the Flow of a Melt of LiF-NaF-KF Fluoride Salts
in
a Circular Tube - V, V, Ignat~ev, S,
V, Keronovskii,
A,
I, Surenkov, 0. P, Shcherbanyuk, S.
P, Manchkha,
and
Yu. B. Smirnov.
507
83
512
87
517
.91
523
95
?
527
99
?
534
104
?
540
108
?
544
711
?
548.
114
?
553
117
?
558
122
.
560
123
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 Declassified and Approved For Release 2013/02/22 :CIA-RDP10-021968000300050006-1
Engl./Rues.
Effect of Reactor Irradiation on the Microstructure of Pyrocarbon
Coatings - I. S. Alekseeva, A. A, Babad-Zakhryapin,
Yu. G, Degal'tsev, L. A, Elesin, Yu. M. Utkin, and Yu. N. Yurovskikh.
562
124
High-Speed Bulk Analysis of Uranium in Geological Samples by the Delayed
Neutron Method - V. R. Burmistrov
565
126
Isotopic Composition of Cesium Built Up in VVER Nuclear Fuel
- V. Ya. Gabeskiriya, V. I, Borisenkov, V. V, Kalygin,
V. M, Prokop'ev, V, S. Prokopenko, M, N. Maslennikova,
and A, P. Chetverikov
567
127
Saturation of the Surface Layer of Carbon Denitrified Glass, Boron
Nitride, and Quartz Glass with Ion-Implanted Deuterium
- V, V. Bandurko, V, V. Kulik, A, A, Pisarev, and V. N. Tsyplakov
569
128
Yields of `ZK and 43K Upon Irradiation of Calcium by Protons and Deuterons
- P. P. Dmitriev and M. V. Panarin.
571
130
Implantation of Low-Energy Hydrogen Ions in Lithium - 0. G. Voronkov,
V. F. Zubarev, and L. M. Frantseva.
573
131
Vobnne 57, Number 3 September, 1984
ARTICLES
Cost of Information in Nuclear Power - Ya. V. Shevelev.
577
147
Using In-Reactor Measurements to Reduce the Indeterminacy of the
Physical Calculation of Fields of Energy Release - V. K. (:oryunov
and Ya. V. Shevelev. ..
587
153
Incorporating Reliability in Optimizing Units in Nuclear Power Stations
Containing Water-Cooled-Water-Moderated Reactors - N. E. Buinov,
S. M. Kaplun and L. S. Popyrin.
593
157
Effects of Irradiation on the Elastoplastic Deformation of Zr + 1ti Nb
Fuel Pin Cladding - V. A. Matushkin, A. A. Medvedev, Yu. V. Miloserdin,
B. D. Semenov, Yu. K. Bibilashvili, and I. S. Colovnin
599
]62
Damage Summation in Annealing and Repeated Irradiation of Pressure-Vessel
Steel - V. A. Nikolaev, V. I. Badanin, and A. M. Morozov.
603
]65
Effect of Chemical Composition and Annealing Conditions on the Radiation
Embrittlement of the Metal of Low-Alloy Welded Seams
- V. A. Nikolaev, A. M. Morozov, V. I. Badanin, A. S. Teshchenko,
and R. P. Vinogradov ..
606
]67
Erosion of Fe-Cr-Ni Alloys and Vanadium Alloys during Bombardment with
helium Ions - B. A. Kalin, I. I. Chernov, V. L. Yakushin, V. 1. Badanin,
I. P. Kursevich, V. A. Nikolaev, and V. N. Kulagin
61.3
173
Resonance Effects in the Interaction of 0.2-0.8-MeV Neutrons with
s6Fe Nuclei - A. A. Sarkisov, I. N. Martem'yanov, A. M. Buguslavskii,
V. N. Ivanov, -and C. N. Ivanov
620
1.79
Application of Gamma Spectrometry in Integrated Experiments on Reactor
Physics - A. V. Bushuev and V. N. Ozerkov.
625
182
Penetration of Radioactive Ind??trial Waters from the North Sea into Central.
Regions of. the Baltic - S. M. Vakulovskii and A. I. Nikitin.
631
l86
Underground Low-Background-Level Laboratory for Radiu~;eochemica.l
Investigations - Vu. A. Surkov, 0. P. Sobornuv, 0. P. Shcheglov,
G. Sh. Shengelaya, K. K. Daneliya, and 7'. (:.,Khundzhc~a
633
7.88
Measurement of the Ranges of Recoil Nuclei of Heavy Actinides Formed in
Multinucleon Transfer Reactions initiated by 22Ne Ions - A. G. Demin,
V. A. Druin, Yu. V. Lobanov, R. N. Sagaidak,. and V. K. Utenkov
637
1.91.
Energy and Angular Distributions of Electron Bremsstrahlunk from Thick
Targets - V. I. Isaev and V. P. Koval.ev. ..
643
1.95
LETTERS TO THE EDITOR
Calculation of Equivalent and Absorbed Photoneutron Doses - V. L. Isaev and
V. P. Kovalev.
649
199
Investigation of the Influence of the Surrounding Material. on the Activation
of Samples in a Neutron-Activation Apparatus with Central.l.y Located
Source - B. S. Vakhti.n, V. S. Ivanov, and C.. A. Kuznetsov.
651.
ZOO
Effect of Temperature on the indications of Magnetic Flowmeters in the Sodium
Loops of the Power Block of BN-600 - A. I. Karpenko, A. A. Lyzhin,
V. P. Minin, and A. G. $heinkman
654
202
Influence of a Resonance Buffer Gas on the Characteristics of the Process
of Infrared Multiphoton Dissociation of 2-Chloroethenyl
Dichloroborane - G. I. Abdushelishvili, T. G. Abzianidze,
A. S. Egiazarov, G. I. Tkeshelashvili, and T. B. Tsinadze.
656
203
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 Declassified and Approved For Release 2013/02/22 :CIA-RDP10-021968000300050006-1
Engl./Ruse.
Effective Fission-Product Yields in VVER-365 and VVER-440 Fuel-
- V. Ya. Gabeskiriya and V. I. Borisenkov. .
Effect of Geometrical Parameters on the Circulation Characteristics of an
Evaporative Element - V. A. Farafonov and V. I. Churyumov. .
Temperature Intervals of Si(Li)p~-n. Detector Use - R. A. Muminov,
B. N. Zaveryukhin, V. D. Krevchik, Kh. Kh. Ismailov,
and A. Sh. Shamagdiev.
Density Distribution of the Photoneutron Flux in a Thick Lead Target
- V. I. Kasilov, N. I. Lapin, and S. F. Shcherbak. .
Cross Section of the 11'Cd(n, n')133mCd Reaction in .the Internal
Extending from the Threshold Energy to 2.2 MeV - Yu. N. Trofimov .
ARTICLES
Power Startup of the IBR-2 Reactor and., the First Physics Investigations
in Its Beams - V. D. Anan'ev, V. A. Arkhipov, A. I. Babaev,
Yu. M. Bulkin, B. N. Bunin, V. S. Dmitriev, N. A. Dollezhal',
L. V. Edunov, A. D. Zhirnov, V. L. Lomidze, V. I. Dushchikov,
Yu. I. Mityaev, Yu. M. Ostanevich, Yu. N. Pepelyshev,
V. S. Smirnov, I. M. Frank, N. A. Khryastov, Yu. M. Cherkashov,
E. P. Shabalin, and Yu. S. Yazvitskii.. .
Nuclear Data Requirements for Fast Reactors - V. N. Manokhin
and L. N. Usachev.
Measurement of the Fission Cross Section of the z3sU Isomer by Thermal
Neutrons - V. I'. Mostovoi and G. I. Ustroev. .. .
Comparative Analysis of Estimates of Neutron Radiative Capture Cross
Sections for the Most Important Fission Products
- T. S. Belanova, L. V. Gorbacheva, 0. T. Grudzevich,
A. V. Ignatyuk, G. N. Manturov, and V. I. Plyaskin .
Absolute Measurements of the 239Pu Fission-Cross Section for
8.5-MeV Neutrons - R. Arlt, H. Bohn, W. Wagner, M. Josch,
G. Musiol, H.-G. Ortlepp, G. Pausch, K. Herbach (GDR),
I. D. Alkhazov, E. A. Ganza, L. V. Drapchinskii, V. N. Dushin,
S. S. Kovalenko, 0. I. Kostochkin, V. N. Kuz'min,
K. A. Petrzhak, B. V. Rumyantsev, S. M. Solov'ev,
P. S. Soloshenkov, A. V. Fomichev, and V. I. Shpakov (USSR). .
Measurement of the a Value at Z35U Resonances
- Yu. V. Adamchuk, M. A. Voskanyan, G. V. Muradyan,
P. Yu. Simonov, and Yu. G. Shchepkin .
Experimental Investigation of the Form of the Energy Distribution of
Neutrons in the Spontaneous Fission of zszCf - M. V. Blinov,
G. S. Boikov, and B. A. Vitenko.
Effects of the Fluctuation of the Resonance Parameters in the Average
Neutron Cross Sections - N. Koyumdzhieva, S. Toshkov,
and N. Yaneva.
Total Neutron Cross Sections of Radioactive 153Gd and Stable.'SZGd
- V. P. Vertebnyi, P. N. Vorona, A. I. Kal'chenko,
V. G. Krivenko, and V. Yu. Chervyakov. .
Cross Sections of the Interaction of Fast Neutrons with Chromium and
Its Isotopes - I. A. Korzh, V. A. Mishchenko, M. V. Pasechnik,
and N. M. Pravdivyi.
Spectrum of Secondary Neutrons and Cross Section of the (n, 2n)
Reaction at Niobium - A. A. Lychagin, V. A. Vinogradov,
0. T. Grudzevich, B. V. Devkin, G. V. Kotel'nikova,
V. I. Plyaskin, and 0. A. Sal'nikov. .
Neutron Generator with Yield of 10'Z sec-' - G. G. Voronin,
A. N. Dyumin, A. V. Morozov, V. A. Smolin, G. V. Tarvid,
and B. B. Tokarev.
Using a Linear Polarimeter for Investigating the y Radiation of an
(n, n'y) Reaction - L. I. Govor, A. M. Demidov,.
0. K. Zhuravlev, V. A. Kurkin, and Yu. K. Cherepantsev .
Mathematical Modeling of a Nonequilibrium Flow Consisting of Water,
Steam, and Air - N. I. Kolev
LETTERS TO THE EDITOR
Buildup of Radionuclides in Nickel as the Result of Electron and y
Irradiation - N. L. Emets, V. G. Batii, Yu. V. Vladimirov,
Yu. N. Ranyuk, E. A. Shakun, and V. A. Yamnitskii. .
,
.
.
660
662
665
668
670
204
206
207
208
210
.
673
227
.
683
234
.
692
241
.
694
243
.
702
249
705
251
714
257
716
259
718
260
721
,262
726
266
729
268
.
732
270
734
272
742
278
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 Declassified and Approved For Release 2013/02/22 :CIA-RDP10-021968000300050006-1
Engl./Rues.
Calculation of the Absorbed Dose of Electron Bremsstrahlung
- V. I. Isaev and V. P. Kovalev. 745
280
Tritium Balance in the Baltic Sea during the Years 1972-1982
- S. M. Bakulovskii and I. Yu. Katrich 747
281
Volume 57, Number 4 November, 1984
ARTICLES
Promising Role of Nuclear Power in an Organie-Fuel Economy
- S. Ya. Chernavskii. 751
328
Optimizing the Parameters of a Ship-Borne iduclear Power Plant on a
Combined Criterion - V. N. Dolgov 762
336
Continuous Markov Processes in Working-Life Estimation with Reference to
an RBMK-1000 Control Valve - A. I. Klemin, V. S. Emel'yanov;
and A. V. Rabchun 768
341
Apparatus for Autoradiographic Monitoring of the Distribution of Plutonium
in Mixed Oxide Fuel - I. A. Golenishchev, I. G. Isaev, V. id. Lyubakov,
A. N. Maiorov, and ,V. V. Yampol'skii. 774
346
Liquid-Metal Transducers for Examining the Deformation of i4aterials in
Reactor Experiments - V. A. Neverov and Yu. L. Revyakin 778
349
Radiation Changes of the Properties of Carbon Pyroceramic - Yu. S. Virgil'ev
and E. I. Kurolenkin. 783
353
Spontaneous Fission Half-Life of 24OPu - A. A. Androsenko, P. A. Androsenko,
Yu~. V. Ivanov, A. E. Konyaev, V. F. Kositsyn, E.. i4. Tsenter,
and V. T. Shchebolev. 788
357
A Relationship between the I7ormalized Equivalent and Exposure Dosages of
Photon Radiation - Yu. G. Kostyleva and I. P. 14ysev 791
359
LETTERS TO TI{E EDITOR
Kinetic Functions of the Delayed Neutrons from a iliicture of the iduclides
sszTh and zaeU - P. P. Ganich, I1. V. Goshovskii, A. I. Lendel,
V. I. Lomonosov, D. I. Sikora, and S. I. Sychev 797
363
Radioelectrochemical Conversion of the Energy of Ionizing Radiation in a
Cell with a Semiconductor Electrode - 14. D. Krotova, A. A. Revina,
Yu. V. Pleskov, A. 14. Morozov, G. E. Zakharov, T. B. Ashrapov,
and E. F. Kalinichenko. 800
364
Vohune 57, Number 6 December, 1984
ARTICLES
Major Safety Provisions in Nuclear-Powered Ships - N. S. Khlopkin.
O. B. Samoilov, V. M. Belyaev, A? M. Dubrovin, E. M. McPnikov, -----
and B. G. Pologikh . .... .. ......... . ...... ........ .. .. .. .. .. .. .. 803
---
379
Tests on Improved Steam Separators in the Third Unit at the Chernobyl Nuclear
Power Station-O. Yu. Novosel'skii, V. B. Karasev, E. V. Sakovich,
M. A. Lyutov; and V. I. An'kov ~ ~ ~ - ~ ~ - - - - - - - - - - - - - - - ~ 807
382
Peculiarities of the Distribution of Phases in the Updraft Section of a Housed
Boiling Reactor - V. N. Fedulin, G. G. Bartolomei, V. A. Solodkii,
and V. E. Shmelev ... .... ........ .. ............ .. .... ........... 8.11
385
Effects of Steam Generator Sectioning on the Reliability of a Nuclear Power Station
Containing a Fast Reactor - A. I. Klemin, O. B. Samoilov,
..................... 81~
and E. V . Frolov .......................... ''
388
Statistical Analysis of Reactor Thermal Power by the Use of Thermal and Radiation
Methods in the First Unit 'at the Armenian Nuclear Power Station - F. D: Barzali,
L. N. Bogachek, V. V. Lysenko, A. M. Muradyan, A. I. Musorin, A.I. Rymarenko,
I. V.Sokolova,andS.G.Tsypin ~?. ?. -??.... 821
393
Test Stand for Research on the Physics of High-Temperature Gas-Cooled Reactors
-A. M. Bogomolov, V. A. Zavorokhin, A. S. Kaminskii, S.V. Loboda,
A. D. Molodtsov, V. V. Paramonov, V. M. Talyzin, and A. V. Cherepanov .........
825
397
Study of Model Coils Made of Superconductor Intended for the Winding of the T-15 Tokamak
- I. O. Anashkin, E. Yu. Klimenko, S. A. Lelekhov, N. N. Martovetskii,
S. I. Novikov, A. A. Pekhterev, and I. A. Posadskii ................ .......
'830
401
Oscillations in the Concentration of Artificial Radionuclides in the Waters of the Baltic
and North Seas in 1977-1982 - D. B. Styro, G. I. Kadzhene, I. V . Kleiza,
a~1M.V.Lukinskene ????????