JPRS ID: 10068 USSR REPORT ENGINEERING AND EQUIPMENT

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• Document Page Count: 55
• Document Creation Date: November 1, 2016
• Sequence Number: 45
• Content Type: REPORTS

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JPRS L/ 10068
23 October 1981
USSR Re ort
_ p
ENGINEERING AND EQUIPMENT
tFOUO 6/81)
FBIS FOREIGN BROADCAST INFORMATION SERVICE
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JPRS L/10068
_ 23 October 1981
USSR REPORT
ENGINEERING AND EQUIPMENT
(FOUO 6/81)
CONTENTS
NUCLEAR ENERGY
Modeling ~ansient and FSnergency Modes on Simulation Tx�ainer 1
Performance of Leningrad Nucleax Electric Power Station Revi.e wed 6
Analysis of Equipment Failure at Active Soviet Nuclear Power
Stations Wit~ WER-L~10 Reactors 12
Design Measures To Ensure Operability of Nuclear Electric Plants
With RBMK Reactors Under ESnergency Conditions 1?
Investigation of Efficiency of Decontaminating RHMK-1000 Coolant
of Transuranium Elements 24
Thermophysical Properties of Working Fluids of Gas-P'hase
Nuclear Reactor ...e 27
INDUSTRIAL TECHNOLOGY
Modeling on Computer System for Automated Design of Industrial
Robuts .......o 29
Classif ication and Technological Criteria for Controlling
Industrial Robots ..........e 38
~ Technological MHD lnstallations and Processes !~6
TESTING AND MATERIALS
Antifriction Coatings on Titanium Alloys Products 48
C~rying Capacity of Working Blades of Gas-Turbine ~gines
Under Vibration Loads 51
- a- LIII - USSR - 21.F S&T FOUO]
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NUCI,iaR. ENERGY
UDC [621.311.25:621.039].007:658.386
MODELING TRANSIENT AND EMERGENCY MODES ON SIMUZATION TRAINER
Moscow ELEKTRICHESKIYE STANTSII in Russian No 8, Aug 81 pp 9-11
[Article by candidate of the engineering scieri~es S.G. Muradyan and engineers
A.A. Ayrapetyan and O.S. Babadzhanyan, Yerevan Affiliate of the All-Union Scientific
Research Institute for Nuclear Power Stations]
[Text] Unforeseen deviations from the specified operating mc+de are possible at the
Ar~S's in service because of disruptions in the operation of the production process
systems, equipment or power unit monitoring and control systems [1].
In such situations, which are called nonstandard in the f.ollowing, the operational
personnel of the power unit should qui~kly determine the source of the deviation
and take the requisite steps to restora normal operating conditions for the power
unit. IncorEect actions by the operating personnel i~t nonstandard situations fre-
quently lead to serious emergencies, and for this reason, reaction speed and error
free actio~i on the part of the operators when controlling the operational modes of ~
AES power units take on special significance, and special teaching and training
centers (uTTs) are being ~reated for them, where these centers are equipped with
trainers and other technical instruction tools. ~
Paults occur randomly in the systems of AES powe*- units and constant readiness.on
the part of the operators during their shift work is required for the timely deteG-
tion and elimination of the faults, something which also explains the stressful ~
nature of their activity.
Operators go through training in the simulators in both teaching and testing modes.
In the teaching mode, operator training in the simulator is carried out in accord-
ance with previously composed scenarios, in which specific nonstandard situations
are studied. The entire instructional course program is brokQn down into topics,
upon the completion of which there is a test of the mastery of the material which
has been covered. Upon completing the training course, operators in the teaching
and training centers go through final certification and receive appropriate skill
certifications [2].
In order to reproduce as completely as possible the work conditions of the opera-
tors on sliift duty in the training simulaLor (the stressful psychological and
physiological state and the constant readiness for intervention in the power unit
- control system), it is expedient during testing to simulate the random nature of
the occurrence of defects in the power unit systems. -
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A method of simulating nonstandard situatiions in trainers for situations which occur
in a power unit, taking into account the random nature of their occurrence, is
treated in the following.
Nonstandard situations in a power unit can be caused both by defects within a power
unit (defects in equipment, in control and monitor systems) and by external effects
(for example, a sudden drop in the electrical load, operator errors in controlling
the power unit, etc.) [3]. As a rule, operators using the power unit monitoring
system do not detect the basic cause of the occurrence of nonstandard situations,
but rather a certain condition of the production process system which is a conse-
quence of the disruptions or deviations which have occurred.
Logic and dynamic ~athematical models of a power unit are reproduced in the simula-
tion trainer, where these models are developed on the b3sis of the calculated de~-
sign configur~tions of the production process systems, including a specified com-
plement of modeled equipment. The locations and nature of the defects being simu-
lared in the trainer are ascertained by means of preliminary analysis in the calcu-
lated design configurations. As a result of doing this work, the set of those para-
meters is determined (the coefficients and variables) for the mathematical models,
_ a change in which within a specified range (either continuously or in discrete
_ steps) leads to the reproduction of the nonstandard situations in the trainer. It
is not p~recluded that it will be necessary to incornorate supplemental logic or
differential equations to introduce a certatn class of defects into the mathematical
models of the power unit.
Defects in equipment, operational failures and deviations from specified operating
conditions of the production process systesns of a power unit, which are termed
perturbations in the following, can be broken down qualitatively into three classes.
The first class includes perturbations which as a rule do not lead to emergency
situations. For example, the fai?.ure of individual controls, transducers, etc.
Such perturbations are most frequently encountered during the process of operating
AES power units.
The second class includes perturbations which, if there is no operator intervention,
can lead to emergency situations. Examples of such perturbations are the loss of
reactor controls, failures in opening or closing gate valves, in turning pumps on
or off, etc.).
The third class includes perturbations which necessarily produce emergency situa-
tions. For example, a break in the pipes of the primary loop of power units with
WI?R's [water-moderated, water-cooled power reactors], the sudden disconnection of
the main circulation pumps, the loss of power by the station, etc. Perturbations
of this kind are encountered comparatively rarely. Since the simulation trainers
are designed for the purpose of instruction, it is apparent that the probability
of occurrence of perturbations must be changed in them as compared to actual power
units.
The occurrence probability for the first class of perturbations can be reduced in
the simulation trainer, since operators can acquire the requisite operational ex-
perience for such situations on an actual power unit. And vice-versa, the proba-
bility of the occurrence of the third class of perturbations must be increased
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in the trai.ners, since their occurrence entails seriuus consequencer., and in this
case, the probab ility of teaching operators the actions to take in such situations
is small with an actual unit.
The random nature of the occurrence of nonstandard situations can be simulated by
means of a random number generator [4].
A random number generator periodically (for example, with a period equal to the
integration step) generate~ numbers in an interval (a, b), which is broken down
into two parts (a, c) and (c, b), in which case, the size of the interval (c, b)
is much greater than the size of the interval (a, c). If the random number falls
in the interval (c, b), it is assumed that there is no perturbation. In turn, the
i.nterval (a, c) is broken down into small intervals, the number of iahich corresponds
~ to the niunber of possible perturbations, while the width of the intervals corres-
' ponds to the probability of their accurrence. If the randam number generator has
produced a number in the interval (ai, ci) E(a, c), this means that the perturba-
tion with the corresponding number i has been selected.
~s was noted earlier, both discrete and continuous input parameters are included
in tlie set of perturbations. If a discrete input parameter corresponds to the
interval (ai, ci), then when the random number x fa11s at any point in the range
(ai, ci), the given discrete perturbation is selected. However, if
a continuous
var.iable corresponds to the interval (ai, ci), the probability of the appearance
of values of which likewise changes continuously over this range, then the choice
of the values of the variable can be made in two ways. The input of the values of
the input parameter can be digitized, breaking the interval (ai, ci) down into sub-
intervals, the width of which is proportional to the probability of the appearance
of di.screte values of the input parameter. The choi~e of the values of the input
parameter can likewise be made by means of a certain function of the random,,number
x, similar to the distribution function for the probabilities of the appearance of
the values of th is parameter.
As an example of the second method, we sha11 cor~sider the case where with an in-
crease in the value T[T E(Tp, T1)], the probability of its appearance falls off
continuously:
7'~7'oex~, ~u(.r-n,)'),
where u-1n(Ti~To)~(c~--ar)2.
The probability of the occurrence of an arbitrary perturbation is:
n-c
9- n-h'
while the probability of the occurrence of the perturbation corresponding to the
interval (ai, c~) is: ~z;-c;
9r==, a--b-.
t
It is appar~nt that q-= U q;, where Z is the number of perturbations.
r i
The monitoring process in the simulation trainer using the method considered here
for perturbation input is accomplished as follows.
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. The specified input mode is set in the trainer and the random nwnber generator is
- triggered. The input of an arbitrary perturbation (from the selected set) into the
production process being simulated should be accomplished during a time to, deter-
mined from proced~iral and psychol~gical considerations. If we assume that there are
n drawings over the time t, then the probabilit,y that one arbitrary perturbation
will be selected during this time is
Q = 1 - (1-Q) n
It is apparent that with a sufficiently large n, the probability of selecting any
perturb ation tends to unity. �
The coordinate of the point c on the segment (a, b) for specified va~ues of to
and the drawing rate k is determined from the following expression:
c = a - (a - b)[1 - (1-Q)l~n�],
where no is the integer part of the expression kto.
The production process parameters begin to change the moment the p~rturbations are
fed in; the operator should be capable of eliminating the defects and restoring
the normal operating conditions of the power unit through his own actions.
The entire course of the check exercise is recorded and the corresponding printout
is fed out at the end of the exercise.
In A~S operational practice, it is possible for not just one emergency (nonstandard)
situation to arise, but a chain of emergency situations, one after the other. Two
methods ean be used to simulate such superimposed complex nonstandard situations.
In the first method, the complex nonstandard situation is included as one pertur-
bation in the interval (a, c). In this case, the choice of the complex nonstandard
situation has a probability
m
Eqi
i=1
of being made, where m is the number of complex perturbations. With the choice
of the i.-th camplex perturbation, the program is started which specifies the seque-
nce for the input of the simple perturbations.
In the second case, a complex nonstandard situation can be created by repeatedly
feeding in perturbations, turning on the random mnnber generator at a different
frequency for each subsequent step of the drawing (the time segment from one selec-
ted nonstandard situation to th e next). In this case, it becomes possible to
- create complex perturbations as the need arises. It is apparent that the choice
uf the same perturbation in different steps of the drawing cannot be tolerated,
- and for this reason, following the choice of a perturbation, the interval (ai, ci)
corresponding to it is added to (c, b).
An advantage of this method is the fact that one can create a rather large ntunber
oF camplex nonstandard situations with the superimposition of two, three or more
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perturbations from the specified set of perturbations. For example, with the
superimpositi.on of three perturbations, the overall nimmber of complex nonstandard
situations will be equal to Z(Z - 1)(Z - 2).
The time allocated in the trainer for each student in the teaching mode is limited,
and for this reason, the trainees during the training process can in practice study
only a porti~n of the ncnstandard and emergency situations modeled in the simula-
~ tion trainer. All of the situations provided for simulation can "e reproduced in
the testing modes using the praposed method, i.e., the traineee will sometimes be
confronted by situations with which they are familiar only from theoretical exer-
cises. In conclusion, it must be noted that the method considered here provides for
simulation of operator's work conditions in simulator for on-line power units,
samething which allows us to hope for 3 more objective estimate of their level of
training. Moreover, during the process of training in the simulator, the operators
will also acquire definite psychological and physiological skills w~en working in
nonstandard situations.
BIBLIOGRAPHY
"Ob avarii na AES 'Tri-Mayl-Aylend-2"' ["On the Accident at 'Three Mile Isian~'
Nuclear Power Station"], ATOMNAYA ENERGIYA, 1979, Vol 47, No 1.
2. Muradyan S.G., Kashirin V.M., "Sistemy podgotovki operatorov AES" ["A Training
System for Nuclear Power Station Operators"], ELEKTRICHESKIYE STANTSII
[ELECTRICAL POWER STATIONS], 1981, No 1.
3. "Analiz avariy" ["Accident Analysis"], in the book, " Bezopasnost' yadernoy
_ energetiki" ["Nuclear Power Engineering Safety"], edit~d by J. Rast and L.
- Weaver, rioscow, Atomizdat, 1980.
"SUornik nauchnykh programm na FORTRANE" ["A Collection o� Scientific Programs
in FORTRAN"], Moscow, Statistika Publis~iers, 1974, No 1.
COPYRIGHT: Energoizdat, "Elektricheskiye stantsii", 198i~
8225
CSO: 8144/1824-B
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UDC 621.311.25:621.039.004
, t~ERFORMANCE OF LENINGRAD NUCLEAR ELECTRIC POWER STATION REVIEWED
Moscow ELEKTRICHESKIYE STANTSII in Russian No 8, Aug 81 pp 6-9
[Article by engineer A.P. Yeperin, Leningradskaya AES]
[Text] The 1970's were marked by the rapid exnansion of a new area
in the field of power. engineering. This new area wa~ born in the Soviet Union in
1954 caith the start of the first nuclear electric power station in the world in
Obninsk.
A widescale program of nuclear power station construction in the European region
of the USSR was planned through the directives of the 24th and 25th CPSU Congresses.
An even more ambitious program of AES construction was approv~ed at the 26th CPSU
Congress.
The progress in the development of our nation's nuclear power engineering is linked
to no small extent with the creation of the Leningradskaya Nuclear Electric Power
_ Station imeni V.I. Lenin.
a The construction of the station was started in 1967, while the main unit with an
- electrical capacity of 1,000 MW having an RBMK-1000 reactor and two 500 MW turbine
generators was brought on line in December of 1973. In July of 1975, the second
unit of the station with the sa~re reactor went on line at the industrial load.
In 1975, the construction of the second stage of the station was started: the third
and fourth power sets with capacities of 1,000 Mf~l each. The third power unit was
placed in service in December of 1979, while on February 9th, 1981,' the creators
of the Leningradskaya AES (LAES) marked a new victory: the fourth power unit was
brought on line at the industrial load. The socialist obligations assumed by
the collectives of builders, installers, aligr~ment workers as we11 as operational
worlcers in honor of the 26th Congress of the Communist Party of the S~viet Union
_ were successfully completed with this event. The Leningradskaya AES became one of
the largest nuclear power stations in the world when the fourth unit was brought
on line.
The operation of the first and subsequent units of the station has made it possible
to acciunulate invaluable experience which was used in the design of other units
and nuclear electric pawer stations with RBMK [channel type, graphite moderated '
high power] reactors. In this case, deficiencies were found in the project plan
and design solutions, and ways of eliminating them were determined, Custom-made
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designs were mastered in a short period of time, which wer~ related to the repair
of the reactor and turbine equipment; the operational documentation was worked up.
The first loading and unloading machine (RZM) underwent its testing successfully.
~ An analysis of the o~eration of the first unit of the Lenin.Qradskaya AES has made
i.t possible to begin the design of the next series of AES's with RBMK reactors,
but now with a capacity 50 percent higher (the R$MK-1500). In this case, the core
dimensions of the reactor will remain practically the same as for the 1,000 MW
reactor (the RBMK-1000). This has become possible because of. th e development of
a new type of fuel cassette with heat exchange intensifiers, which have been
tested successfully at the Leningradskaya AES.
Special attention is being devoted to questions of safety in servicing the reactor
installation and the production pracess systems related to it, as well as to issues
of envirornnental protection in the design of nuckear power stations.
- Al1 of these questions have been successfully resolved in the design of the power
generation units of. the Leningradskaya AES imeni V.I. Lenin.
Safety in reactor control is assured by a highly reliable control and protection
system (SUZ), which incorporates three systems of automatic controllers for the
overall reactor power and a system of local automated controllers and protection
to maintain the'power within the specified range in individual regions of the reac-
tor core. A SFKRE system (physical monitoring of the liberated energy distribution)
provides for monitoring the energy liberated in each production process channel.
'This system, as a part of the complex with the "Skala" centralized monitoring sys-
tem, which is designed around specialized computers, continuously performs calcula-
tions to determine the thermal engineering reliability of each production process
channel of the reactor. The operator who controls and monitors reactor operation
receives continuous data from this system in the form of light signaling and print-
out charts. This makes it possible to perform a timely analysis of changes in the
operational mode of the reactor installation and take t11e appropriate steps.
A separate system provides for continuous monitoring of the hermetic seal integrity
of the jackets (KGO) of the fuel rods in which there is nuclear fuel in the form
of uranium dioxide and highly radioactive solid and gaseous uranium fission prod-
ucts. This system makes it possible to determine the number of a production pro-
cess channel in which there is a cassette with a loss of seal of the fuel rod.
When such a cassette is detected, the loading and unloading machine comes to as-
- sist, which installs a new cassette i:~ place of the one with the seal failure,
while the spent cassette in unloaded in 3 holding tank in a separate sealed cylin-
drical case. Thus, the cladding seal stat�:_is monitoring system, as part of a com-
plex with the loading and unloading machine, makes it possible to maintain the
purity in the circulation loop through the reac*or and provide for a normal radia-
tion status for the servicing personnel.
Because of the fact that the reliability and safety of reactor operation depend on
the cor?dition (integrity) of the production process channel piping, to ~nonitor their
condition, the appropriate continuous monitoring KTsTK system has been developed
(monitoring the integrity of the production process channels), which reacts to the
slightest leaks of coolant fram the channel into the graphite brickwork of the
_ r~actor, and makes it possible to determine the number of this channel. The �
existing equipment permits the replacement of a channel where necessary in a
shutdown reactor.
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In case of an emergency shutdown of a unit, the designers and project plann~ers have
provided systems for scrat~aning the reactor and subsequently cooling t~e co~~e. In
- this case, special atteneion has been devoted to the reliability of tfie ele~trical
power supply for the station shutdown mechanisms which participate in L~~?ese opera-
tions. Besides the traditional ways of boosting the electrical power s~sppl~ relia-
_ bility for the station, a diesel electric power generator is additional~y installed
which is started automatical.ly by the actuation signals for the emerg~ncy protection
and provides electrical power for the reliable power supply section.
Despite the fact that the w.:ter flowing through the reactor core has a relatively
high radioactivity, this nonetheless is no threat to the safety of 'the servicing
personnel. Such a situation is assured by the fdct that all of t'he production pro-
cess systems with the radioactive coolant are made as tight systems amd housed in
special protective boxes. Possible leakage of the radioactive w~ater is localized
by means of a system for trap water collection and processing, as well as intent-
ional leaks. The water which is processed and purified by this system is returned
to the production process loop.
The liquid radioactive wastes which appear during the ~rocess of c.leaning and re-
processing the production process loop water are directed via a special route into
special tanks for further reprocessing and subsequent burial of the highly radio-
active residue, The atmospheric purity of the amb ient air basin and the air in
the rooms is maintained by means of production process and o~verall ventilation ex-
change systems, the operational mode of which provides for clean air delivery to
attended rooms and exhaust with subsequent purification in special filters from
unattended rooms. Following the f ilters, the air is ejected through a ventilation
pipe 150 meters high.
Despite the reliability of the steps taken to protect the servicing personnel and
the environment against radiation, there is an automated dosimetric monitoring sys-
tem at the station fur the gamma radiation level, as well as the concentration of
radioactive gases and aerosols in the most important production rooms and in the
air ejected through the ventilation stack. The cleanline~s of the wall surfaces
of the rooms and the production process equipment is checked periodically as well
as the presence of radioactivity in the atmospheric air., the water basin, soil,
, vegetation and foodstuffs. The monitoring system confirms the reliability of the
adopted protective measures.
Operational experience with the first two units of the station, and the increase
in the stringency of the requirements placed on assuring the safety of AES's and
protecting the environment, as well as the achievements of science and engineering,
- have necessitated new design solutions for individual equipment and production
_ process questions in the design of the third and fourth units of the station.
Al]. of this was responsible for a number of special features of the third and fourth
units as compared to the RBMK--1000 reactors which had gone on line previously:
--TY:e modular layout of the main and auxiliary installations and production process
systems, which makes it possible to resolve questions of repairing the equipment
and fittings in a nonoperating unit relatively simply and preclude errors by
personnel during changeovers;
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--The improvement of the system for containing the products from a loss of seal
accident by containers and tanks in emergencies, related to pipe breakages in the
multiple forced circulation (MPTs) loop;
--The introduction of an improved system for emergency cooling of the reactor;
--The use of new types of main circulation pumps with a new end seal for the shaft;
--The cooling loop of the protection and control system is made in a gravitational"ly
driven configuration using an independent operational loop.
The structural design and production process features o~ the second stage units have
- brought about a marked change in the layout of the main building, the systems housed
in it and have predetermined specific features of the start-up and aligrtment work
as well as the physical and power starts of a unit.
A major features of starting a reactor of the third and fourth units is the fact
that two percent enriched fuel is used in the initial charge of all fue~ rods in
them for the first time in reactors of this type.
In the process of running the physical starts of the RBMK reactors of the first
stage of the Leningradskaya AES, extensive experimental data was obtained on the
physics of the reactor core. The analysis of the work done in this case and the
restilts obtained showed that differences are possible between identical reactors
which are due to the scatter in some of the technological parametArs such as the
graphite density, the fuel content in a fuel assembly (TVS), etc. Moreover, there
- is necessity of an experimental check of new equipment or operational solutions
with each start. Taking this into account, as well as the fact that the third
unit reactor was the first of th e RBMK series reactors in which on~y two-percent
enriched fuel was used for the initial charge, the physical start of the reactor
was accomplished in accordance with a program which provides f or an experimental
checlc of the design data and an actual estimate of the main neutron physics
characteristics, the compensating capability of the protection and control system
rods and the supplemental moderators (DP's), as well as setting up a full scale
load while meeting the requirements of nuclear safety.
- 'The excess reactivity which is due to the increase in enrichment is compensated
- by incr~~asing the number of control and protection system rods from 179 to 211
and the length of the mudera`ing portion of these rods from 5 to 6 meters, using
"heavy" su~plemental modera~ors, containing inserts made only of boron steeJ.. A
provision was made for tiie use of moderator rods which can be installed in the
central cavity of a f.uel cassette as an additional means of com;~ensating for
excess reactivity i.n the physical start program. In contrast to all of the
previous RBMK-1000 reactors, the charging of the third unit reactor took place
with water present in the production process channel (TK's) of the multiple
forced circulation loop.
Critical experiments were performed during the charging which made it ~ossible
to correct the computational procedures for predictin~ the initial charge of the
reactor; measurements were made of the bulk neutron fields for the full scale
charge for the purpose of determining the nonuniformity coefficients of the cold
unpoisoned reactors and determine the precision of the neutron physics
9
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calculations based on two-dimensional programs; experiments were run to determine
reactivity effects in order to obtian data for running three-dimensional computa-
tional programs.
A set of start-up and aligr~ment operations and experiments was performed during
the power start period, which made it possible to check and more precisely specify
the characteristics and operational modes of the equipment, the production process
systems at various power levels, including the neutron physics characteristics of
the core, the temperature conditions of the metal structures and the grphite
brickwork, as well as effects caused by activation and reactor radiation, i.e,,
check the quality of the biological shieldings; the radio-chemical composition of
the coolant for all loops; the radiation status in the rooms and on site, as well
- as the dynamic characteristics of the unit. The control point settings f or the
protection, interlocking and signaling were also made more precise during this
- same period; the safety valves and the automatic control systems were adjusted;
the equipment and instruments of the physical monitoiing system for the energy
distribution, monitoring the hermetic seal of the jackets of the fuel rods as
well as monitoring the integrity of the production process channel were calibrated
and aligned~.
The programs which were carried out in executing the physical and power start of
the third unit reactor makes it possible to accomplish the start-up and alighment
operations as well as charge tlie core and bring the reactor up to power in the
L-ourth unit a shorter period of time.
The experience of builders, installation and set-up workers as well as operational
worl:ers, supported by the requisite organizational and technical measures (the
implementation of joint construction and installation work, a significant consoli-
dation of the main metal structures anc~ assemblies outside the installation area,
_ as well as the pre-assembly examination of the pumping equipment, fittings, elec-
trical equipment, etc.) and their self-sacrificing labor made it possible to
bring the fourth unit of the station on line with in a curtailed timeframe. The
connection of the first turbine generator of the third unit took place three
months after the physical start and only 1.5 months after the physical start of
the fourth unit.
a The mastery of the design capacity of the f.ourth unit of the Leningradskaya AES
is a complex and important task, on which the station staff is now working. The
experience acciunulated and the successes achieved in the start and mastery of the
- first three units inspire confidence in the successful resolution of this problem.
Th e station staff finished 1980, the concluding year of the lOth Five-Year Plan,
_ with good results:
--The five-year plan for electrical power output was fulfilled on November 4,
1980, and the plan for power generation was fulfilled on November 17, 1980;
--The annual plan for electrical power output was fulfilled on December 18 and
that for power generation on December 20;
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--The installed capacity utilization factor was 73 percent, despite the fact that
the installed capacity of th e third power unit was mastered in the first half-year
fol.lowing start-up;
--The production cost per KWH of electrical power was reduced by 4.1 percent and
amounted to 0.797 kopeck/KWH.
Based on the results of the All-Union Socialists Competition of 1980, the station
- staff was awarded the challenge Red Banner of the CPSU Central Committee, the USSR
Council of Minis ters, the All-Union Central Trade Union Council and the Komsomol
Central Committee, as well as the memorial medal "For High Work Quality and Effic-
iency in the lOth Five-Year Plan", which was entered on the All-Union Board of
Honor of the USSR Exhibition of National Economic Achievements.
The Teningradskaya AES imeni V.I. Lenin has become the major supplier of electric
po~~er to the Leningrad Power Administration system. The plan for electric power
gene ration in 1981 is 22.6 b illion KWH. By the end of the year, the total amount
_ of el.ectrical power generated b~~ the station from the time it was started will
reach 100 b illion KWH.
The successes of the staff of the Leningradskaya AES imeni V.I. Lenin once again
underscore the correctness of the party and goverrnnent policy related to the further
development of power engineering in our nation and inspire confidence that the taslcs
ahead in this field will be carried out successfully.
_ COPYRIGHT: Energoizdat, "Elektricheskiye stantsii", 1981.
8225
CSO: 81.44/1824-B
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UDC 621.039.58'68
ANALYSIS OF EQUIPMENT FAILURE AT ACTIVE SOVIET NUCLEAR POWER STATIONS WITH VVER-440
_ REACTORS
~ Moscow ATOMNAYA ENERGIYA in Russian Vol 50, No 4, Apr 81 (manuscript received
14 Nov 80) pp 248-250
[Article by F. Ya. Ovchinnikov, L. M. Voronin, B. E. Baturov, A. A. Abagyan and
S. A. Lesnoy]
[Text] One of the major trends in development of nuclear power in the Soviet Union
is construction of nuclear e"lectric facilities with water-cooled water-moderated
power reactors [WER]. The first power facility with a WER-210 was put into in-
dustrial operation in 1964 at the Novocherkassk Nuclear Electric Plant. A large
series of power facilities with VVER-440 reactors has been put into operation since
1971 in the USSR and several other nations (East Germany, Bulgaria, Finland, Czecho-
slovakia) with technical assistance from the Soviet Union.
Operation has confirmed the correctness of engineering decisions in desigr. develop-
ments, as well as conformance of the actual working characteristics of power facili-
ties to the projected levels. Operational experience has also enabled determina-
tion of ways to further improve equipment and technological sy~tems in accordance
with increasing requirements for safety and reliability of nuclear electric plant
- operation.
Safety problems for normal working conditions can be considered completely solved.
However, for emergency conditions these problems need further research. Interna-
tional experience in the development of nuclear power shows that the very concept
of "safety" and the methods of achieving it are undergoing continuous changes in
connection with massive construction of nuclear power facilities and the search
for rational, technically feasible mea~s of ensuring safety. WER reactors of
the first generation had shielding and localizing systems corresponding to the
limited scale of a maximum credible accident that was accepted at the time. Con-
siderable emphasis was placed on the factor of keeping the nuclear electric plant
far from populated areas. The safety systems in power facilities with unified
WER-440 reactors that are now being introduced are designed for counteracting
more extensive damage up to and including a break in the pipelines of the main
circulation loop with maximum diameter for which the consequences of an accident
are potentially more serious. Obviously it is very difficult to solve this problem
- by technical means alone, i. e. by using hardware to meet all safety requirements
in building a nuclear electric plant that does not subject the surrounding environ-
_ ment to at least a slight risk of contamination. All efforts in the area of safety
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should be directed toward reducing the degree of risk. Such efforts will be fruit-
ful only upon condition of inseparable combination of techni~al safety facilities
with a high level of organization in utilizing these facilities.
Gxperience in the operation of nuclear electric plants throughout the world shows
that while we are developing and constantly improving hardware for preventing and
localizing large ~ccidents (up to and including instantaneous transverse rupture
of maximum-diameter pipelines), we need to be just as serious in working out and
perfecting methods and means of preventing and clearing up so-called "minor acci-
dents." In large part, these methods and means are the same as those aimed at
ensuring the reliability of nuclear power plants as sources of energy, since any
disruptions in the operation of ma;~r equipment that are due to failures and de-
fects lead to power limitations foi ~urposes of preventing deviations of the param-
eters of the facility beyond safe limits. In other words, to ensure safety, a
reduction in reliability predetermines the necessity of placing constraints on
the working conditions of a facility.
Comprehensive in-depth analysis of equipment operation to find the weakest links
in technological systems of nuclear electric plants and improve their reliability
tias been a matter of course since the startup of the first WER power facilities.
Since 1977, a unified system has been in operation in the USSR for collecting data
on failures and defects of nuclear electric plant equipment. The acquisition of
reliable information enables isolation of the most typical failures that lead to
emergency outages, unplanned down time and reduced economic efficiency of nuclear
~ electric plants. Timely determination of the causes of equipment failures and
defects (especially for the equipment of systems having to do with the safety of
- a nuclear electric plant) means that effective work can be done on improving this
equipment from the design stage to final operational use. For the sake of'conveni-
- ence of such analysis and evaluation of the influence of failures and dPfects on
the operational reliability and safety of nuclear electric plants, equipment has
been divided into groups in accordance with functional designation. The table
shows the spectrum of distribution of failures of equipment by percentages as typi-
cal of power facilities with VVER-440 reactors.
In analyzing the information, consideration was taken of all kinds of failures,
both complete and partial, that lead or may lead to limitations in the operation
of. major equipment, as well as those failures that do not affect normal operation
oE power facilities due to the secondary nature of equipment, or built-in redundancy.
Ttie tablc shows that 11.3% af the failures pertain to equipment of the primary
circuit having the greatest significance from the standpoint of ensuring reliability
and saEety of the nuclear electric plant. Failures of reactor equipment, including
the control system, amount to ~4%,.failures of steam generators 3.5%, and of pipe-
lines--less than 1%. This shows the fairly high level of reliability of the main
circulation loop. The remaining 88.7% of failures and defects pertains mainly
to equipment that is not specific to nuclear electric plants and is typical of
convenLional power facilir..ies.
- A large number of failures (-38%) pertain to monito.ring and control instrumenta-
tion. However, this has almost no effect on the reliability of the power facility
since it most frequently involves instruments and communication lines with adequate
backup. Analysis of the statistics of equipment failures for 1977-1979 showed that
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Distribution of Failures and Defects of Equipment in Nuclear Electric Plants
With VVER Power Facilities According to Functional Groups
and Individual Kinds of Equipment
- Equipment Group Number of Failures percent
per Unit per Year
Primary Circuit Equipment:
reactor equipment 5 4.3
steam generator equipment 4 3.4
main circulation pumps 2 1�~
main shutoff gates 1 0.9
pipelines 1 0.9
- For the Group 13 11.3
Turbine-Unit Group:
turbines 1 0.9
condensers 3 2.6
steam-superheater separators 3 2�6
regenerative heaters 6 5.2
For the Group 13 11.3
Pump Equipment of All Kinds 9 7.8
Fittings (Except for Main
Shutoff Gates 10 8.7
- Blower Equipment 7 6.1
Compressor Equipment 4 3.5
Electrical Equipment:
turbogenerators 1 0.9
electric pump drives 3 2.6
electric drives of control
assemblies 2 1�8
breakers, disconnects 9
For the Group 15 13.1
Monitoring and Control
Instrumentation:
primary instruments 9
secondary instruments 23 20.0
communication lines 12 10.4
For the Group 44 38.2
T~TAL 115 100
the most characteristic defects are welding flaws (up to 32%) and hidden flaws
- in materials (up to 28%). Failures and damage through fault of servicing personnel
amount to less than 7%, which is evidence of a rather high skill level.
To work out requirements for equipment reliability (which are especially necessary
on the stages of design and manufacture), reliability indices are calculated on
the basis of statistical data on equipment failure. One such very important index
for restorable items is the parameter of failure rate w(t). Because of the com-
paratively small volume of the statistical sample, only point values of w(t) have
been obtained, without estimation of the confidence leve].. Nevertheless, these
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values of c~~(t) establish a lower limit of reliability and can be used as primary
~ioi-malized data in the design, manufacture and utilization of equipment. For
. example ~~~(t) in hr-1 for the reactor is (1.6-2.1)�10-5, for the steam generators
--8.2�10-5, for the reactor control system--6.2�IO-4, for the main circulation
pump--2.6�10-5, for the turbine--8�10-5. Cumparison of these data with those avai].-
able for equipment of non-Soviet nuclear electric plants shows that they are com-
pletely comparable iF consideration is taken of the fact that ordinarily only total
failures are considered in calculations of w(t) in non-Soviet practice.
The given information shows that failures and flaws apply mainly to subsidiary
- equ~pment, or to auxiliary systems of major equipment. Therefore these is nQ re-
duction in the reliability and safety of the nuclear electric plant as a whole.
This is evidenced by the stable and high level of the coefficient of utilization
oI installed power of VVER-440 power facilities: in 1970 this index was 72.6%,
in 1978--80.7%, in 1979--73.8%.
We point out that the optimum coefficient of utilization of installed power for
power Faci].ities with WER-440 reactorG in the USSR is 80~, which corresponds to
7000 hours of operation of the equipment at rated power per year. This is deter-
mined by estahlished periodicity and standards of duration of repairs of major
equipment (reactor equipment, turbines and so on).
Analysis of the struct�re of the coefficient of utilization of installed power
shows that underuse of installed capacities associated with unplanned repairs and
equipment deFects, i. e. due to down time having a direct relation to reliability
' and safety of the nuclear electric plant, amounts to no more than 3e7%, whereas
rhis index was 8% in the early years of operation of WER-440 reactors. This is
evidence of an appreciable improvement in the reliabi]_ity of major equipment over
the el.apsed period.
Coi~tir~ited work in the following major areas will ensure retention of the attained
reliability level and Curther improvement:
per.fection of equipment design; .
improving the quality of the equipment during manufacture and the quality of in-
stallation as a basis for operational safety and reduction of the probability of
failures and damage. Programs of quality control have been developed and imple-
mented at the manufacturing plants for major equipment. All equipment arriving
at ttie nuclear electric plant goes through a pre-instllation entry inspection.
Improvements are being made in the technology of installation and welding processes,
and in methods and equipment for quality control on welding jobs;
checking the condition of equipment during utilization with application of up-
to-date methods for early detection of defects;
improving and increasing the technical level of utilization;
improving the eff.icacy of supervision zor observance of directive and normative-
technical documents in the process of manufacture, installation and utilization
of nucLear etectric plant equipment;
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improving the skill of personnel working in the nuclear electric plant, and syste-
matic personnel safety training in accordance with specially developed compre-
hensive programs.
On the whole, experience in utilization of WER-440 power facilities brings us
to the conclusion that they are sufficiently highly reliable and safe.
COPYRIGHT: Energoizdat, "Atomnaya energiya", 1981
6610
CSO: 8144/1354
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UDC 621.039.58
DESIGN MEASURES TO ENSURE OPERABILITY OF NUCLEAR ELECTRIC PLANTS WITH RBMK REACTORS
UNDER F1~IERGENCY CONDITIONS
- Moscow ATOMNAYA ENERGIYA in Russian Vol 50, No 4, Apr 81 (manuscript received
14 Nov 80) pp 251-254 .
[Article by I. Ya. Yemel'yanov, S. P. Kuznetsov and Yti. M. CY~erkashov] `
[TextJ A number of advantages of RBMK uranium-graphite boiling-water channel reac-
tors have brought about their extensive use in nuclear power in the U.SSR [Ref. 1].
RBMK reactors are characterized in particular by high reliability thanks to moni-
toring and control of parameters of individual process channels, and also replace-
' ment of the fuel assembly without shutting down the reactor.
In successful operation at the present time are seven power facilities with RBMK
reactors having 1000 MW of electric power; a program has been worked out and is
being implemented on constructing and utilizing several more reactors. Design,
construction and utilization experience has enabled us to go on to construction
of the RBMK-1500--a channel reactor of the same design and dimensions with power
of 1500 MWe.
S
_ _ 2 4
- ~ 6
~ Y~ + ~d
oo--
~ i - - 7
i 14 ~ - -
11
~Z 9
~p B �
t
9~ �
Fig. 1. Schematic diagram of the circulation loop of the RBMK~1000: 1--reactor;
2--separator; 3--main circulating pump; 4--level regulator; 5--pressure regulator;
_ 6--turbogenerator; 7--water tank; 8--pumps of self-contained emergency reactor
cooling system (ERCS); 9--pumped storage unit of ERCS; 10--fast-action valve of
ERCS; 11--pressurized collector; 12--distributing group collector; 13--ERCS col-
lector; 14--flow limiters
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The fundamental heating arrangement of the RBMK-100Q (Fig. 1) is typical of one-
loop boiling-water reactors. The coolant is circulated through the reactor by
eight main circulating pumps, two of which are backup units. Steam separation
is done in four separators of gravity type, and the saturated ateam then goes to
the turbines and to the intermediate steam superheaters. Two turbogenerators are
installed on each power unit with RBMK-1000 reactor.
The RBMK parameters are monitored by a centralized control system based on computer
equipment. The system provides operating personnel with visual and recorded in-
- f ormation on the operation and condition of components of the structure.
Energy distribution through the volume of the core is monitored by a physical con-
trol system that includes radial and heightwise intrareactor sensors. An algorithm
, for calculating the power of all process channels with respect to discrete reference
points, i. e. b~� channels with sensors, is realized in the PRIZMA program for the
plant computer that is part of the centralized control system. The program uti-
lizes characteristics of radial energy distribution ~btained by a program of physi-
cal calculation on an external computer, and also the position of the control rods
at the same instant of time when the physical calculation is done. These data
are fed to the plant computer via punched tape.
The power of the process channels is periodically calculated upon order by the
operator. The results are used to correct the energy distrlbution. On-the-spot
control of energy distribution is done by operational personnel directly in accor-
dance with the readings of the intrareactor sensors.
The distribution of coolant flowrate through the process channels is monitored
by flowmeters installed at the inlet to each channel. The reactor design provides
for on-the-spot control of the distribution of water flowrate through the process
channels by changing the position of the channel regulating and cutoff valves.
The capability of ineasuring and controlling the water flowrate in each process
channel is a distinguishing feature of the RBMK; this capability ensures the re-
quired maneuverability of the distribution of reserves before a heat exchange crisis
upon a change in the power of the reactor or radial energy distribution.
Leaky fuel elements and loss of integrity of channel pipes are detected in ample
time by systems for monitoring gas tightness.of fuel element cladding and the integ-
rity of each process channel pipe.
The reactor control system is designed with consideration of requirements of nuclear
- and radiation safety. For example, local automatic controller and local scram ~
systems are provided to prevent unsteady deformations of energy distribution. For
= purposes of maintaining power conditions of operation of the facility when complete
shutdown of the reactor is not required, provisions are made in the reactor control
system for rapid controllable power reduction in some cases, in addition to the
conventional facilities for complete damping of the chain reaction. These provi-
- sions include disconnection of some main circulating pumps, reduction of the water
level in the separators, reduction of feed water flowrate, partly disconnecting
or dumping the load by turbogenerators in the unit, increasing the energy intensity
of the fuel elements and the like. The various types of emer~ency protection are
actuated by signals of equipment malfunction.
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To study emergency conditions, a mathematical model was developed including equa-
tions of kinetics, hydrodynamics and heat exchange, and algorithms for operation
of aquipment and systems for automatic control of nuclear electric plant parameters.
Comparison of the results of calculations done with the aid of the model, with
data on individual dynamic processes that have occurred in operating nuclear elec-
tric plants showed that the model satisfactorily describes the dynamics of a power
facility. Some emergency conditions, mainly involving a transition to natural
circulation, were studied on special model stands.
- This paper discusses the results of investigation of emergency conditions caused
only by total de-energizing of the power facility, disruptions ia the feed water
- sugply system, and breaks in large pipelines of the circuit.
j~lten the facility is de-energized, emergency equipment operates that scrams the
reactor; early in the accident, the core is cooled by main circulating pumps that
do not immediately stop, and then by natural ciLCUlation of the coolant. For re-
liable cooling of the reactor during this period, the main circulating pumps must
have fairly large flywheel masses, and for this purpose special flywheels are in-
corporated into their design. A diesel generator facility is automatically switched
in to supply emergency feed pumps, the pumps of the emergency reactor cooling system
and the like, that are necessary for carrying off the residual energy release in
the reactor.
The major parameters of the power fa-
N~~~aINGQ~ cility in the state of de-energizing
i _ _ are sho~an in Fig. 2. At the beginning
o,z ~?4 ~~?5 I of the transient process, Q is somewhat
higher than G, i. e. in this period
a ~o o - - - the average subcritical power reserve
I through the reactor is less than the
-az s,s o~s - - - - . s rated value. However, calculations
( show that the reserve to the heat ex-
-a,y 62 ps - change crisis in the most stressed
I 4 channels does not fall below unity.
'J - Thus it can be stated that in the early
-O,B S,B 0,25 - - - -
1 ~ Z period of accidental de-energizing
~ the reactor parameters do not go beyond
p >0 20 ,to 4o so Fa ~o g~ safe limits.
time, s
Fig. 2. Emergency state of de-energizing Within 30-35 s after de-energizing
of the facility: 1--relative neutron of the main circulating pumps of~ the
power N; 2--relative thermal Qower Q; power facility, the core is being cooled
3--relative coolant flowrate G; 4--level by natural circulation of the coolant,
H in the separators; 5--pressure P in and the stability and intensity of
- the separators this circulation depend in large mea-
sure on several factors su~h as the
design of the loop, the pressure therein, change in temperature and flowrate of
the feed water and so on. To determine the reliability of core cooling in tl~e
mode of natural circulation, a set of experiments was done both on the techno-
logical model stand, and on the reactors of the first and third power units at
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the Leningrad Nuclear Electric Plant. Steady-state and transient conditions were
studied with various parameters influencing the development and intensity of natural
circulation. As a result, the p3rameters were determined that ensure reliable
cool.ing of the reactor with natural circulation.
It should be noted that reliability and safety of reactor cooling in such a mode
has been confirmed by work experience with operating nuclear electric plants with
RBI~IIt reactors.
Dynamic studies of the characteristics of a power unit during emergencies in the
system of feed water supply have enablad determination of the conditions of safe
operation of the facility, and elaboration of requirements for the equipment and
for action by operational personnel in such emergency situations.
Deactivation of one of f our operating feed pumps has demonstrated that in this
state there are slight and even changes in technological parameters, and therefore
it is not required to introduce any auxiliary protection of the equipment. To
bring the reactor power into line with the altered feed water flowrate it is suf-
ficient to reduce the power manually by adjusting the controller setting. When
two or more feed pumps are disconnected at the same time, the emergency protection
ensures safety of the facility by automatically reducing the reactor power to a
safe level.
M~~ , ~J The transient rocess with disconnec-
`c ~ G' ~ I ,s P
~~o~ ~ tion of two feed pumps and operation
~
~ 70 S of emergency protection in accordance
_o~- ~ b y with a signal of reduction in the flow-
s,6 o,~s - - rate of feed water by 25% of the current
2
-o~'- value (Fig. 3) is characterized by
s,2 c,s ' ~ the following: an automatic regulator
-U6 - 0 10 20 30 40 SO 60 70 BO
reduces reactor power to 60% of the
time, s rated level with slight overcontrol;
Fig. 3. Emergency state of disconnection maximum deviation of the level in the
- of two feed pumps (1-5, see Fig._2; 6-- separators is observed within 50 s,
relative flowrate of feed water Gnoe~ and amounts to 150 mm, after which
the level is recovered; the steam pres-
sure in the separators is maintained at a level close to the nominal value by means
of a pressure regulator that unloads both turbogenerators in the facility.
SaFety of the tacility in the case of total cessation of feed wa~er supply in the
power units is ensured by emergency scramming of the reactor upon a signal of flow-
rate reduction below 50% of the current value. In this state, the water supply
to the circulation loop by emergency feed pumps is ~10% of the nominal level; these
pumps are energized within 10-20 s after cessation of the supply of feed water.
Studies have shown that cessation of the feed water flow leads to a reduction of
the level in the separators. This can cause undesirable trapping of steam on the
downcomers of the loop, cavitation cutoff of main circulating pumps, and impediments
- to the development of natural circulation. To prevent the level from falling in
the separators, the main circulating pumps are disconnected, which slows down the
- rate of reduction of steam content in the core and outlet lines of the reactor.
As a result, less water is required from the separators to replace the steam in
the circulation loop.
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'I'he state of cessation of the supply of
f~M.p, ~ G-- - feed water with disconnection of the
5 main circulating pumps is analogous
o ~o ~,o to the state of de-energizing of the
i facility, which has been found to be
- "02 safe. The main parameters of the power
6,6 0,75 -
-oy I unit in this state are shown, in Fig. 4.
4- Analysis shows the following:
-qs - -o,s - ~ - I--
~ ~he rate of reduction in thermal power
-0,8 S,B 0,75 -a- - - --1-- -
-2~ I~ is greater than the rate of falloff
; i ~ ~
..~~G ~ in water flowrate throughout the entire
o ~0 2~ .~0 4o s~ so ~o so transient process, which is evidence
of reliable cooling of the core;
time, s
Fig. 4. Emergency state of disconnection maximum drop in the level in the sepa-
oE all four feed pumps (1-5, see Fig. 2) rators is observed within 75 s, and
then the level begins to rise; the
steam pressure in the separators decreases at the beginning of the process, then
becomes somewhat higher than nominal, and after 72 s the pressure stabilizes at
the nominal level.
Thus these results show reliability of cooling of the core in a state of complete
instantaneous cessation of the supply of feed water accompanied by disconnection
of the main circulating pumps. Therefore the main circulating pumps are disconnected
with a delay of ~9 s after emergency protection operates in power facilities with
RBMK reactors upon a signal of reduction in the feed water flowrate below 50% of
the current value. The core is cooled down by natural circulation of the coolant.
It is assumed that the most serious emergency situations can arise when large pipe-
lines of the power facility are ruptured. The design provides for technical facili-
ties that prevent discharge of the steam-gas mixture ix~to the service areas, and
especially beyond the limits of the nuclear electric plant. Most typical damage
to the circulating loop is breaks in small tubing (drains, impulse Iines and the
like). Rupture of a large pipeline is extremely improbable. Experiments on full-
scale specimens have shown that a leak is possible in pipelines with diameter
of ~800 mm at a pressure of 8.5-9.5 MPa if fatigue cracks are ~75~ of the wall
thickness in depth, and ~470 mm long [Rf~f. 2]. Inspection of the metal guarantees
- that there will be no sudden rupture of the pipeline, since the critical dimensions
of defects a:e large, and they should be detected in planned shutdowns of the fa-
cilities. During inspect~.on, the ~netal is examined and checked by special tech-
niques (ultrasonic f.law detection, acoustic amission). In spite of this, the de-
sign of the nuclear electric plant provides for measures to ensure safety in case
of instantaneous transverse rupture of the largest pipeline.
At tY~e initial instant the leakags is about 6 metric tons per second in the case
_ of complete instantaneous rupture of a pipe 300 aan in diameter, and ~40 metric
tons per second for such rupture of a pipe 900 mm in diameter. As a result of
analysis of emergency situations, two independent signals have been selected for
operation of emergency reactor protection: a rise in pressure in the rooms where
- the pipelines of the loop are accommodated, and a reduction of the level in any
21
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separator to a value that exceeds
P' ~a ~ its deviation from the nominal
B,0 - - under transient conditions.
6,a - The most dangerous pipeline rup-
ture is one on the pressure side
~ of a main circulating pump, since
this instantaneously cuts off
~ the supply of coolant to the
, channels of the emergency half
2'~ i of the reactor. It is just such
ZI ~ a hypothetical accident that
has dictated the fast-action
0 20 40 6o do ~oo ~20 >90 96o emergency reactor cooling system
time, s
(ERCS), its maximum capacity
Fig. S. Change of pressure in RBIrIIC-1000 circu- ~about 1.1 metric tons per second)
lation loop upon rupture of a pressurized col- and minimum time of discharge
- lector 900 mm in diameter: 1--pressure in of all the coolant from the emer-
separators; 2--pressure in the pressurized Sency loop (10-12 s). Fig. 5
collector shows the pressure change in
the circuit with instantaneous
rupture of 900 mm pressurized collector of a main circulating pump. Coolant from
the ERCS is fed to the channels of this half to prevent damage to the fuel elements.
Water from the ERCS is sent to each distributing group collector, and to avoid
nonproductive discharge of the water through the cross section of the break in
the pressurized collector, check valves are provided at the inlet to the distrib-
uting group collector. The ERCS consists of two subsystems (see Fig. 1): the
main subsystem with pumped-storage unit, and a prolonged cooling subsystem with
special pumps and tanked water reserves. The cooling water is supplied from bot-
tles, and after they are emptied the water is supplied by pumps to the ERCS of
~ each half of the reactor, and thence through pipelines to each distributing group
collector. Installed in the water feed lines to the collectors are fast-action
v,..~lves that open when pressure rises in the rooms. T~Ihen this happens, the water
goes to the reactor loop in which the level has dropped in the separators, or the
pressure differential has decreased between the pressurized collectors and the
separators. Such an algorithm of engagement of the main subsystem of the ERCS
ensures cooling of the core with complete or partial rupture of a large-diameter
- pipeline, and precludes a false alarm in case of accidents that do not involve
loss of integrity of the circuit.
Studies have shown that acceptable temperature conditions of the fuel elements
are ensured by the speed and productivity of the ERCS in case of any pipeline rup-
ture up to and including a maximum break.
Al1 equipment and pipelines of the circulation loop of the reactor are accommodated
in securely tight enclosures that prevent emissions of the steam-gas mixture from
- the rooms of the nuclear electric plant ~nto the atmosphere in case of breaks.
The steam-gas mixture goes through special tunnels to a localizing unit where the
steam is condensed. The enclosures are designed for an excess pressure of ~0.4 MPa,
. 22
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which is not surpassed even in the case of complete instantaneous rupture of the
- largest pipeline, thanks to special condensation devices of the bubbler type, a
system of bypass valves, sprinklers and heat exchangers.
REFERENCE S
1. Dollezhal', N. A., Yemel'yanov, I. Ya., "Kanal'nyy yadernyy energeticheskiy
- reaktor" [Channel-Type Nuclear Power ReactorJ, Moscow, Atomizdat, 1980.
2. Rivkin, Ye. Yu. et al., TEPLOENERGETIKA, No 11, 1978, p 71.
COPYRIGHT: ~nergoi.zdat, "Atomnaya energiya", 1981
6610
CSO: 8144/1354
23
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- UDC 621.029,534.44
INVESTIGATION OF EFFICIEN~Y OF DECONTAMINATING RBMK-1000 COOLANT OF TRANSURANIUM
ELEMENTS
Moscow ATOMNAYA ENERGIYA in Russian Vol 50, No 4, Apr 81 (manuscript received
11 Feb 80) pp 274-275
_ [Article by A. M. Vorob'yev and N. P. Star~donovaJ
[Text] ror purposes of radiation safety at nuclear electric plants of all types,
" both Soviet and non-Soviet, provisions are made for purifying the water coolant
of radionuclides. The bypass purif ication facilities aC nuclear electric plants
are intended for extracting products of corrosion, sol~ible sa?ts and radionuclides
from the n;irging water. In nuclear electric plants with RBMK reactors, 200 cu. m
of water per hour is nurif ied with total volume of the multipleforced circulation
loop of ~1200 m3. The water is taken off f rom the pressure side of the main circu-
~ lating pumps, and after appropriate treatment and cooling to 40-50�C it is sent
- to a purification facility, where it goes in sequence through mechanical (hydraulic
perlite), mixed cationic-ani~nic exchange (KU-2, AV-17) and trapping filters. The
process includes monitoring tor pH, hardness, C1- content, products of corrosion
(iron, copper), specific electrical conductivity and content of some radionuclides
before and after purification. The purified coolant is heated to 270�C and returned
to the circuit.
_ In addition to radionuclides of corrosion or igin and fission products, the water
of the multiple forced circulation loop may be contaminated with a-emitters that
are formed us a consequence of slight surface cont~mination of the fuel elements
with urani.um o~t migration of radionuclides f rom defective fuel elements. Accumu-
l.ation of transuranium elements in nuclear fuel was studied in Ref. 1, 2. It was
_ sfiown that sequential capture of neutrons in the fuel forms a large quantity of
isotopes of plutonium, americium and curium that are quite dangerous from the radia-
tion standpoint. For example, 19 kg of 237Np, ~700 kg of isotopes of plutonium,
>100 icg oE isotopes of americium and ~2 kg of curium are formed after a three-
year run in 80 metric tons of fuel with 3.3% 235U enrichment at a reactor power
of 1000 MW [Ref. 2].
Unfortunately, we have been unable to locate any published data on the coefficients
oE diffusion of transuranium elements from intact and leaky fuel elements. However,
_ this qiiantity can be roughly estimated by considering the similarity of properties
of. lanthanides and actinides. For example, if the relative diffusion of iodine and
24
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cesium from leaky fuel elements is taken as 1, then diffusion is equal to 10-4 far
cerium [Ref. 3]. In case of severe loss of integrity of fuel elements and contact
between fuel. and coolant, there may be a considerable increase in this parameter.
In emergency situations with fuel meltdown, a-emitters may play a decisive part
iri raising the radioactivity of the coolant and formation of the radiation environ-
ment at the nuclear electric plant and its surroundings. In this connection, coolant
- purification takes on special significance.
To study the effectiveness of bypass decontamination with respect to transuranium
elements, the content of 239Np, plutonium isotopes, 2`'lAm and 2`t2Cm was determined
in the coolant before and after purification. The studies were done in different
periods of operation of the nuclear electric plant. The 239Np content was deter-
mined by a Y-spectrometric method using a semiconductor Ge(Li) detector. Identifi-
cation was done simultaneously with respect to five peaks (210, 228, 278, 316 and
334 keV). Americium and curium were determined by radiochemical methods based
on precipitation on b ismuth phosphate after the plutonium had been oxidized to
the hexavalent state. Reduced plutonium (III) or (IV) was then also coprecipitated
with bism�th phosphate. The a-activity count (100% efficiency) was then done in
a layer of solid scintillator (Zn5) by the technique of Ref. 1. With background
- ot 0.2 pulse/minute, the sensitivity of the method was 10-13 Ci/Z (1 Ci = 3.700 lOlo
disintegrations per second). The error of determinatian was �20% [Ref. 2].
Content of a-emitters in the water of the
- multiple forced circulation loop, Ci/Z Table 1
Util.ization period `_'Pu, .Vn, ~:m I =k'u ~~~Am f=~=~:,n I =~l'
~
~ { _ . _
t i r 1 f ) 11- l a ~ 11 i~
Initial (up to one year) ~ ~1 -i~)~t~~-_~~ ~~_~-,~,~.~~r?~~ (1=`,j�t~-~~ ~:t;~tn~~~
Vor.mal (1:1'?)�1n ( . 1 ' ~
- ~,uss oE integrity of fuel ~,~~rU 4,6. f(r1� 5,4�lU�1o 1~'-'''
elements ~
Table 1 summarizes averaged data of investigation of the content of a-emitters
in water for different periods of. operation of a nuclear electric plant. Calcu-
lations show that the content of transuranium elements in the initial period of
utilizati.on [(1-5)�10 -13 Ci/ZJ is due mainly to surface contamination of fuel
elements with uranium during fabrication. As the length of the run increases,
microcracks appear in the cladding, and the content of transuranium elements in
the coolant increases by two orders of magnitude. A further increase in the content
of these elements (to 10-9 Ci/Z) is observed with occurrence of defective fuel
= elements subject ~o replacement. Some time after their replacement, the concen-
- tration oE a-emitters in the coolant again stabilizes at a level of (1-10)�10-11
- Ci./Z .
Table 2 gives some results of evaluation of the eff iciency of bypass purification
of. transuranium elements on nuclear electric plants with RBMK-1000 reactors.
This table shows that the water purification factors with respect to the indicated
transuranium elements are fairly close, and are equal to 3000 in the case of fresh
25
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Table 2 charging of the filters with ion-exchange
resins, about 1000 after 5-8 months,
Efficiency of bypass removal of and 25-100 after 9 months on this charge.
transuranium elements from coolant Individual values of KPur for a filter
- service period of 5-8 months amount to
~Coo ant activity purifica- 1025 for neptunium, 985 for plutonium
Radio- Ci/Z tion and 750 for curium.
nuclide Ibefore I after I factor
- The degree of removal of transuranium
~~I'n. ,~m, t;r~~ 1,2� tn-p 3,1 � to-~~ 37iN1 __lements from the coolant on ion-exchange
1,2�1ir' 6�1u-11 ~~~xx~ filters is higher than for radionuclides
S~�~'i~-;; ~5~ of corrosion and fission origin. These
tcr a�t~r scxi
4�i~~-~~ ~.i~r?:~ results are to be expected since it is
5�lcr~t t.fu-~s 5~x1 known that the stren th of sor tion of
4,4�1(1-~~ ~1�!(r~a $ P
3� 1~r1J < 1� tcrlY cations on KU-2 resin increases with
increasing cation charge. The radionu-
Kav- i~�5 clides of interest to us are arranged
in the following sequence with respect
`'F'~~ ~-~oF~u, 3,5�tir~~ 7�i~rla 5~x~ to the strength of bonding to the cation
""p~~ etc 4,G� 1lT-1U 3. f~i-i~ tb35
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FOR OFF[CIAL USE ONLY
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FOR OFFICIAL USE ONLY
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Control systems i.ntended to solve multipurpose control problems should be called
universal systems; they axe implemented primarily by developing and atoring special
control functions~ i.e., by developing means of programing. Under concrete condi-
tions~ the development of special softwaxe may also be required~
In creating special control systems, the designer determines, based on the required
control functions, the control volume, convenience in servicing, as well as produc-
tion possibilities etc,, the necessary groups of structural components and units.
Thus, for simple cases, pneuma,tic control systems axe specified. New developments
of special ChPU systems should include minicomputers.
Universal control systems should ha,ve built-in mini- or microcomputers or be con-
nected to a central computer.
Programed control systems in principle are not meant for processing pa.ths and mo-
tions on a digital uasis as compaxed to freely programed ChPU computer systems.
Programing problems on moving pa,rts span all actions for prepaxing motion programs
and introducing them into the control system. The motion program contains all the
control functions intended to implement the conerete problem of PR control.
The control system must provide economical programing according to the requirements
for their application and readjustability when complying with special features of
motion. The basic requirements must include convenience for the consumer and, if
possible, the least laborious preparation of the program and the easiest correct-
ability and optimization. Methods ~or data introduction and programing should be
classified according to the place of programing and the type of control system.
Programing methods axe selected according to conditions for the use of PR and, at
the same time, must satisfy the requirements mentioned above. Programing at the
site of use is very convenient because the programer usually does not need special
knowledge, and special auxiliary mea,ns for programing axe not required.. However,
in a number of cases of PR use, especially for comprehensive automa.tion of assembly,
programing outside the boundaries of the work position (external programing) is re-
quired. In this case, the program is made up when prepaxing for work and is recorded
on the program carrier assigned to the control system. Such programing must be
used only in those cases where programing at the site is impossible or inexpedient
(for example, the motion task due to its complexity, or switching with the process
at the working position does not lend itself to description; implementing control
functions which may be programed only when prepaxing for work; or the PR receives
control data from central control~.
When the PR is programed at the worker position~ after the motion cycle is deter-
mined, all the necessaxy data, for designing the program is fed directly into the
control system.
Programing by motion cycles of parts is limited b;~ the program cycle, i.e., the
motion cycle is designed as a sequence of instructions for direct control of motors
of axes and grips, inputs and outputs etc., and is implemented manually. The
motion paths are adjusted by means of stops or position sensors directly on the PR;
therefore, they axe not a component part of the motion program.
L~2
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43
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The input of data into the program memory may be done manua,lly by a.djusting, con-
necting, pushing an operator's pushbutton etc. with the motion reading exhibited
on the control panel, and automa,tically by functional switches due to which data
is transferred, to the addressed places in the memory, i.e., by the self-tea,ching
method. A PR programed at the site whose program also includes path conditions
in the component parts of the program, along with the motion cycle, is called
freely programable.
Specified motions are adjusted maxiua,lly in the required sequence. In this case,
- the motion cycle and the path conditions axe stored. step-by-step. The PR position
is determined by the pa,th sensors which feed data on the path.
To produce motion along the loop, the grip(tool) is directed manually along the
desired motion loop during the adjustment mod.e and the trajectory is automatically
stored as a sequence of points. The number of points stored per unit of time must
be such that a sufficiently precise reproduction of the programed motion line is
made automatically.
As a rule, programing requires special documentation, equipment and devices for
_ prepaxing, converting, checking, correcting programs and transferring da.ta to the
memory. Al1 tasks on motion data must be compiled fully at the stage of work
prepaxation.
The improvement of external PR programing methods is achieved by a similar method
of programing ma,chine tools with ChPU.
The use of inechanical program carriers cam dises, rigidly secured by cams~ is a
special case of external programing only for simple PR. ilhen readjusting, the
program carrier is replaced or rea,d.justed anew.
An approxima,te control system using coded data may be implemented by special cod-
ing on program carriers, punched tape, diode matrices or semiconductor memories.
Specialized program languages which axe subdivided into simple languages oriented
toward a specific problem and langua,ges of a higher order are used; they differ
in complexity of control functions and in the volume of reprocessed data.
The motion program recorded in the corresponding language, up to the time that it
may be placed in the program memory, must be translated into the elementaxy control
language. There is also the possibility of converting instructions in specialized
languages directly into control instructions (by means of interpreters). Prob-
lem oriented languages axe used primar~ly in connection with universal control
systems for the solution of complex problems for multima,chine tool servicing and
_ assembly operations.
In external programing, problems on the motion of an industrial robot on the site
of its use, mixed programing is also utilized.
44
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Conclusions
An analysis of control systems and grograming methods for PR tmakes it possible to
establish a relationship between the control system type~ methods of programing~
the control application field and the content of motion grograms. Based on these
general criteria, it is possible to classify control syste~s. In practice, some
control system criteria may be found in combination when only a part of the de-
grees of motion belong to positioning. The type of control system is determined
according to the optimal vaxiation of a positioning robot with a given accuracy
and quick action which may be achieved by mathematical simulation of its motions.
COPYRIGHTt Izdatel'stvo "Mashinostroyeniye", "~~estnik mashinostroyeniya", 1981
2291
CSO: 1861/194-A
= 45
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UDC 532.538.4
TECHNOLOGI~AL MHD INSTALLATIONS AND PROCESSES
Kiev TEKHNOLOGICHESKIYE MGD USTANOVKI I PROTSESSY in Russian 1980 (signed to press
21 Oct 80) pp 22, 189-190
[Annotation and table of contents from book "Technological MHD Installations and
Processes", by Anatoliy Fedorovich Kolesnichenko, Institute of Electrodynamics,
UkSSR Academy of Sciences, Izdatel'stvo "Naukova dumka", 1000 copies, 192 pages]
[Text) This monograph examines qualitative and spatial transformations in bounded
regions filled with an electrically conductive liquid or gas and placed in physical
fields: electric, magnetic and gravitational. Included among such transformations
are changes in the form of energy ansl energy transfer--conversion of electremagnetic
energy to heat and mechanical work, transfer of heat and mass of the conductive
fluid under the influence of various combinations of electromagnetic, capillary
and thermoconductive factors. Resulta of theoretical and experinnental research
are given on a new class of MHD phenomena--capillary effects that arise during
arc and induction treatment of inetals. New techniques for processing alloys are
described that are based on the use of I~ effects.
The book is intended for scientists, engineers and technicians interested in the
development and use of MHD installations.
Figures 82, table l, references 119.
Contents page
Preface 3
Principal symbols 6
Chapter 1: Principal Equations of Magnetohydrodynamics 9
Chapter 2: MHD Phenomena in Electric-Arc Welding lg
1. Body and surface forces in electrically conductive droplets, ~ets and
plasma 18
2. Shape of the boundaries of phase transformations as a cylindrical electrode
is melted by a welding arc 30
3. Quasi-steady states of the free surface of droplets and ~ets of electrode
metal 40
4. Drop formation and transport during arc melting of a cylindrical electrode 46
5. Drop formation and transport of electrode metal in welding by direct
current with superposition of current pulses 54
46
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6. Drop formation and transport of inelt in electric arc welding with unsteady
electrode feed 5~
7. Drop formation in modulation of magnetic pressure of noncontracting ares 59
Chapter 3: MHD Granulation of Metals 61
1. MHD methods of controlling dissociation of free electrically conductive
jets 61
2. Free electrically conductive jets under the effect of sign-alternating
electromagnetic forces in the sources 69
3. Conditions of dissociation of free jets placed in a longitudinal alternating
magnetic field 81
4. Conditions of production of spherical particles 83
5. MHD dispersers of liquid metals and alloys 90
Chapter 4: M~ID Heat and Mass Transfer in Induction Melting Furnaces 102
l. Eddy pattern of electromagnetic forces and mechanisms that produce uni-
directional motion 104
2. Pressure developed by active sections of the channel 108
3. Hydraulic drag of induction channels 113
4. Simulation of heat and mass transfer processes in the channels of induction
furnaces 115
5. Modeling of industrial induction furnaces 134
Chapter 5: MHD Devices Based on Piston Gas-Liquid FlowS 139
1. Physical essence of processes in accelerated gas-liquid flows 141
2. Formation of piston flows 144
3. Stable forms of gas-liquid interfaces. Change in mass of a liquid
batch during acceleration 150
4. Dispersal of a liquid batch and mass transfer 158
5. Interphase heat exchange in the channels of acceleration devices 162
6. Model of motion of liquid batches of variable mass 167
7. Losses during acceleration of piston flow. Efficiency and other param-
eters of acceleration devices based on piston flows 171
8. Example of calculation of a.device for accelerating piston flows 174
References 182
COPYRIGHT: Izdatel'stvo "Naukova dumka", 1980
6610
CSO: 18E~1/154
47
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uDC 621.793.669.248
ANTIFRTCTI~1 CQATINGS QN TITANNM ALLOYS PRC~UGTS
Moscox V~TNIK I~USHINUSTROYFNIYA in Russian No 5, May 81 p 47
~Article by V. I. La,tatuyev~ cand. of tech, sciences anc~ G. N. Ganay, engineer]
[Text~ It is Well known [1-5] that it is extremely difficult to deposit good
bonding coatings on titanium and its alloys due to the formation of oxide film on
their surfaces.
When the oxide film is ree~oved by etching by uaing a mixture of nitric and hydro-
t'luoric acids there is observed a atrong hydrogen abaorption on the titanium surface
due to the formation of a hydride (0.069~ and more of hydrogen in the outer layers).
The presence of the hydride film reduces sharply the adhesion of the coating to
the titanium and deteriorates the strength characteristics, i.e., increases the
brittleness of the metal. Tt is proposed [2] that annealing (400-500~C) removes
the hydrogen, and the adhesive properties of the electroplating are improved.
A new and very promising method ,,ts one for depositing nickel-phosphorus coatings
by chemical nickel plating (KhN) xith acid and alkali (ammonium) solutions [6-9] ~
on the titanium and its alloys.
The phosphorus content (2-15~) in the coating imparts specific propertiea~ great
hardness and higher antifriction properties xhich makes it possible to utilize
KhN to solve the problem of acrexing parta made of titanium alloys together xhen
they are sub~ected to high loads. Tests Were a~ade on samples of the VTZ-1 and the
VT-20 alloys. Preliminary testa on the KhN in solutiona of xell-kno?rn compounds
(GOST 9.047-75) indicated that titanium alloya either cannot be nickel plated or
form a coating xith poor a~hesion.
On the basis of recommendations ~7-8], inveatigationa xere made on the preparation
of the surface of the titanium alloys before chemical nickel plating (before each
chemical operation the parts xere washed thorou~rhly in running xater for 2 to 3
minutes) by first degreasing in organic solvents (gasoline,dichloroetha,ne etc.)
for 5 to 10 minutes~ then in an alkaline aolution for 5 to 10 minutes at a tem-
perature of 60 to 70~C.
l~8
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The oxide film Was removed in solutions of sulfuric or hydrochloric acids at vari-
ous temperatures and times. The best results xere obtained by pickling in concentra-
- ted hydrochloric acid at 50 t 5~C for 8 to 10 minutes (2 to 3 nicrometers are etched
off). Too lengthy etching deatroys the surface due to autocatalysis. For example,
is to 20 micrometers are etched off.
The most critical operation is the activation of the titanium alloy surface
the deposition of a thin film (approximately 0.1 micrometer) of nickel that cata-
lyzes the KhN process.
A solution* consisting of 150 to 250 grams~liter of nickel aulfate and 100 to 120
milliliters~liter of hydxochloric acid at 20 to 25�C xas taken as a basis. The
activation time was 5 to 10 seconda.
After thorough ~+ashing in running xater~ the saaples xere immediately immersed in
the KhN solution. In this case, solution "D" xas used Whfch coneisted of 20 to 30
grams/liter of nickel sulfate, 20 to 25 grama~liter of sodium hypophosphite~ 10
to 15 grams~liter of sodium acetate, 0.003 grama~liter of thiourea and 2 ndlliliters/
liter of acetic acid~ at a temperature of 90 2~C and a apeed of 20 to 25 micro-
meters~hour.
For removing hydrogen and increasing adhesion~ the samples xere heat-treated
in air at 400�C for 2 hours. The adhesion of the plating met the C06T requirementa.
The developed technology solves the problem of depositing an aatiacuff nickel-
phosphorus plating on threaded parts made of titanium alloy8.
BIBLIOGRAPHY
i. Colner, A et al. J. Electrochem. SoC. N il~ 100, 1953�
2. Layner~ V. I. "Modern Pslectroplating." Moscox. Metallurgiya, 19~7r 3~ P~e~�
3. Ginberg~ A. M. "r3ectrolytic Deposition of Metals on Aluminum." Leningrad.
Sudpromgiz, 1957� 167 pa.ges.
4. Vernik~ 5.; Pinner, R. "Ch~mical and ~Lectrochemical Treatment of Aluminum and
its Alloys." Tran slated from the I~glish. Leningrad, Sudpromgiz, 19~. 322
pa~es .
5. Vayner, Ya. V., D~soyan, M. A. "Technology of Plectrochemical Coa.tings." ~
Leningrad. Mashinostroyeniye, 1977� 468 pages.
6. Gorbunova~ K. M.; Nikiforova, A. A. "Physio-cheonical Principles of the Chemi-
cal Nickel-Plating Process." USSR Acaden~y of 5ciences, 1970~ 207 FeBes�
~Val'syunene~ Ya. N. 3olution for activating titanium before chemical i~ickel
plating. Author's certificate 185182.
49
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7� Latatuyev~ V. I.; Ganay~ G. N.; Denisov, A. D. "Metal Coating by the Chemical
Process. Alt. kn. izd-vo~ 19~. 208 pages.
8. Nikandrova, I,. I. "Chemical Methods for Metal Coatings." 8dited by P. M.
- Vyacheslavov. Moscow. Mashinostroyeniyeo 1974. 104~ pagea.
9. Vishenkov, S. A. "Chemical and ~lectrochemical I~Iethods for Depositing Metal
Coatings." Moscox. Mashinostroyeniye~ 1975~ 312 pages.
CQPYRIGHT: Izdatel'stvo "Mashinostroyeniye","Vestnik mashinostroyeniya", 1981
2291
CSO: 1861/194-A
50
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UDC 621.438.62-226-2
CARRYING CA,PACITY OF WOR:ZING BLADES OF GAS-TURBINE ENGINES UNDER VIBRATION LOADS
Kiev NESUSHCHAYA SPOSOBNOST'RABOCHIKH LOPATOK GAZOTURBINNYKH nVIGATELEY PRI VIBRATSI-
ONNYKH NAGRUZHENIYAKH in Russian 1981 (signed to press 29 Dec 80) pp 4, 313-314
[Annotation and table of contents from book "Carrying Capacity of Working Blades
of Gas-Turbine Engines Under Vibration Loads", by Valeriy Trofimovich Troshchenko,
academician, UkSSR Academy of Sciences, Valentin Vladimirovich Matveyev, Boris
Alekseyevich Gryaznov, Sergey Semenovich Gorodetskiy and Anatoliy Beynaminovich
Roytman, Institute of Strength Problems, UkSSR Academy of Sciences, Izdatel'stvo
"Naukova dumka", 1000 copies, 316 pages]
[Text] This monograph gives the procedure and results of investigation of fatigue
of high-temperature alloys subjected [o complex thermomechanical loading, and also
- the fatigue of working blades of turbines and compress~r:, i*~ gas-turbine engines
with c~nsideration of their changes during utilization. A me~.hod is described
f or modeling experimental loads for turbine blades (exp~rimental facilities). Data
are given on the damping properties of turbine blade materials, lock joints of
compressor and turbine blades, and also on the damping capability of working blades
that is due to aerodynamic dra~ of the ambient flow and structural hysteresis with
consideration of operating conditions. An examination i~ made of ~ethods of re-
~ ducing the dynamic stress on hlades by using structural damping and dissipation
~ of the energy of vibrations in the material. Reco~endations are rnade for increas-
ing the carrying capacity of ~he turbine blades in gas�turbine engines.
~ For scientists, engineers and technicians dealing with problems of developitzg and
_ using gas-turbine engines.
Figures 245, tables 38, references 209.
Contents page
Preface 5
- Chapter 1: General Information on Operational Damage to C~npressor and Turbine
Working Blades ~
1. Major factors that influence the carrying capacity 3nd service life of
working blades ~
2. Typical types of damage to the blades of the axial compressor and turbine 22
_ 3. Indices of operational reliability of gas-turbine aircraft engines 37
51

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Cliapter. 2: Durability of Turbine Blade Materials 44
l. Methods of studying the durability of high-temperature alloys 44
2. Influence o~ high temperature on the durability of turbine blade alloys 57
3. Durabil.ity of high-temperature alloys under conditions of combined action
of inechanical and thermal cyclic stresses 64
Chapter 3: Methods of Studying the Carrying Capactiy of Working Blades 78
_ 1. inethods of testing i~lades for durability 78
2. The UL-1 facility for studying durability under room-temperature and high-
temperature conditions 90
~ 3. The VL-2 automated vibration stand for studying fatigue under conditions of
programmed load and temperature variation 95
_ 4. Statistical processing of the results of inechanical tests 103
- Chapter 4: Investigation of Fatigue of Working Blades 22g
1. Influence oi operational and technological factors on fatigue of working
blades 118
2. Damage to turbine blades after operational failure 129
3. Influence of testing conditions on the microstructure and type of fracture
of working blades 144
Chapter 5: Metl~ods of Studying the Damping Capacity of Working Blades 155
- 1, Damping characteristics and methods of determining them 156
Systems for automating the process of determining characteristics of
vibration damping 166
3. Investigation of the damping properties of turbine blade materials 175
4. Investigation of the damping capacity of lock joints 187
5. Investigation of aerodynamic damping of vibrations 204
_ 6. Investigation of damping of vibrations in a centrifugal force field 213
Chapter 6: Investigation of the Damping Capacity of Working Blades 220
1. Damping properties of turbine blade materials 22p
2. Damping capacity of lock joints 23~
3. Aerodynamic vibration damping 25g
- 4. Vibration damping in a centrifugal force field 270
Chapter 7: Improving the Vibrationa.l Reliability of Working Blades 281
1. Optimizing the damping properties of turbine blade materials 281
Improving the characteristics of aerodynamic vibrational damping 286
3. Increasing the damping capability of lock joints 291
4. Im~roving the vibrational reliability of pairwise shr~uded blades 298
Re.Eerences 3Q2
COPIRIGHT: Izdatel'stvo "Naukova dumka", 1981
6610
CSO : 1861 / 153 ErID
- 52
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