JPRS ID: 10143 USSR REPORT ENGINEERING AND EQUIPMENT

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• Sequence Number: 58
• Content Type: REPORTS

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- JPRS L/ 10143
30 No~~ember 1981
- USSR Re ort
p
~ ENGINEERING AND EQUIPMENT
(FOUO 7/81)
Fg~$ FOREIGN BROADCAST INFORMATION SERVICE
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- JPRS LJ10143
30 November 1981
USSR REP~ORT ~
ENGINEERING AND E~UIPMENT
(FOUO 7/sl)
_ CONTENTS
AERONAUTICAL AND SPACE ~
= Thermal Design c,f Spacecraft Units.....~ 1
Power Supply and Control Systems for Electrical Rocket
Engines 9
NUCLEAR E[dTERGY
Experience Wi~i~ Bringing on Line and Operating Third Power
ilnit With Fast Neutron Re~ctor at Beloyarskaya AES............ 12
Selection of Safety Devices To Counter Excessive Pressure in
Primary and Secondary Loops of Nuclear Electric Ppwer
Stations With Water--Cooled Reactors......~ .................e.. 17
Seismic Stability Evaluation of Nuclear Power Station
Electrical Equipment 24
Brushless Exciters of 500-MW Turbogenerators in Atomic P~wer
Stations 32
Optimization of WER-440 Reactor Energy Distribution............ 39
Thermohydraulic and Physicochemical Processes in Nuclear
Power Plants 42
- NON-NUCLEAR ENERGY
General and Theoretical Magnetohydrodynamics 50
Magnetohydrodynamic Machines 55
- Reliability Prediction and Diagnosis of Power Equipment......... 59
- a- [III - USSR - 21.F S&T FOUO]
,
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I~iDUSTRIAL TECHNOLOGY
Adaptive Contral of Dynamic Systems 67
TURBINE AND ENGIAIE DESIGN
Automated Marine Power Installations........�� 76
NAVIGATION AND GUIDANCE SYSTEMS
Gyrostabilizers for Inertial Guidance Systems .............o..... 100
= Mot~on Stability, Analytical Mechanics~ Motion Control.......... 103
FLUID MECHANICS
Elastic Shell Penetration Into Compressible Liquid 109
MECHANICS OF SOLIDS
Generalized Functions in Thei~atoelasticity (Collection of
Scientific Papers) 111
Dynamics and Stability of Mechanical Systems 120
- b -
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AERONAUTICAL AND SPACE
UDC 629.78
THEft1~',AL DFSIGN OF SPACECRAFT UNITS
Moscow TEPLOVO~IE PROY~;[CTIROVANIYE AGRHGATOV L~'TATII,'NYKH APPARATOY in Russian
- (signed to press 5 Dec 80) pp 2-8, 17~-175
[Annotation, introduction a.nd tab~e of contents from the book "Thermal Des~gn of
Spacecraft Units'; by Boris Mikhaylovich Pankratov~ Izdatel'stvo "Mashin~stroyeniye",
1100 copies, 176 pages]
[Text] This book presents modern concepts on the thermal design of.spacecraft
units (systems for therma,l prntection of apparatus for reentry, thermostatic
control of fuel tanks, power plants etc.)~. Papers on heat exchange~ thermody-
namics, aerodynamics~ space physics, ballistica, dynamics, nuclear phyaics etc.
were widely usea.
The book is intended for engineers wha design he~t-intensive units for spa.cecraft.
Introc:uction
The investigation of thermal modes of uni:ts and aystems and the selection of
their characteristics, takin~ into account the effect of the environment and the
limitations imposed by the parametera of the spacecraft (LA), are a part of the
~r~eral problem in designing heat-intensive designs and the apparatus itself~ i.e.,
thermal design.
_ When the LA operates for a long t~me in apace, ita design and several unita and
systems are subjected to the effect of the environment and, as a rule, to rela-
tively high thermal loada from external and ir.ternal hea.t sources. The problem
of optimal thermal design may be represented conditionally in several stages.
- The basic design parameters of the system are determined to the first approxima-
tion by taking into account the special features of the tra,3ectory and the dura-
tion of the flight. Then~ for the selected LA parameters, the motion tra~ectory
of the spacecra~'t and the optimal chax$cteristics of the units under the effect
of the external and internal thermal loads and of the euvironment are defined
more accurately.
Here the problem of the s3mthesis of the ~ptimal control of the spacecraft may be
posed.~ using the criteria of the minimum xeight of one unit or another, the mini-
mum of integral heat floxa durin~ the flight, not exceeding the permitted
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temperature of the unit structures and providing for the optimal thermal mode of
their operation. Sometimes a comprehenaive funct~anal is uaed that takea into
account, for example~ the surfaae temperature as xell as the thermal flo~+s that
act on the unit etc.
The presence of radiation zones of solar and cosmic radiations surrounding the
earth poses an ent;re series of problems for the dynamic flight of spacecraft
(KLA). The consideration of such proble~s is relatad to the necessity of taking
into account the radiation dangers ta xhich the crex~ the units and the devices
aboard the spa,ce craft axe subjected.
The choice of optimal poxer engineering in these problems muat be made by taking
� into aceount the effect of intensiva radiation zones. Of interr~st is the fact
_ that~ from the mathematical viexpoint~ the nature of these limitations is such
that they either eliminate entirely the possibility of using the classical methods
of the calculus of variations and the maximum principle, or require their essen-
tial modification. Of interest in this case is~ the consideration of several
methods of calculation~ for example~ dynamic programing etc.
At ~he first stage of thermal design, it is neceasary to a first approximation~
- but to a sufficient degree of accuracy, to determine the basic poxer characteris-
tics of the spacecraft and the engine that Nill be the initial ones for the folloW-
ing analysis and will be made more precise at the folloxing stagea.
Essentially the problem is divided into dynamic and xeight. ~ome thermal KLA mode
is established under the effect of external(sun radiation, direct or re~lected
from the planets~ its oxn radiations of lanets and the earih~ the effect of
streams of micrometeorites, protons etc.~ and internal (engines~ power systema~
appaxatus etc.) heat sources, by which~ for example, may be unders~ood the thermal
fiald~ variable in time~ in individual design components of the units~ systems
and the spacecraft as a Whole.
The ~nergy from the external heat sources and the energy emitted in the spacecraft
from internal sources are converted. into thermal radiation. The latter is prac-
tically the only method for heat removal in spa.ce from heat-loaded KLA units, if
the methods based on ejecting matter are not considered.
Due to the sma,ll density of the micrometeoritea in space~ they have little effect
on the thPrmal mode of the spacecraft and should be tak~en into account only xhen
the KLA passes through a region xith a very high meteorite content. Hoxever,
micrometeorites and other cosmic particles, xhen striking the skin of the apace-
craft, may change the optical properties of the surfac~ of the KLA (due to their
hlgh energy) which~ in its turn~ may have conaiderable effect on the thermal mode
- of the spacecraft. In a number of casea, even xhen flying at great heights, it 18
necessary to take into account the heating of the sheath of the spacecraft by the
impacts of atoms and molecules in the residual at*osphere of the planet. Thus,
the temperature of the spacecraft depends to a certain extent on ite position
in space and its orientation xith respect to thermal energy sources~ and may vary
within fairly Wide 1lmits.
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The problem arises of maintaining the thcirmal modea of preserving fuel components
aboard the KLA ~hich may be reduced to optimizing the characteristics of the .
thermostatic control syst~m, the mass of the fuel compastments or the mass of the
entire spacecraft. Here it is necessary to take into account the heating of tl~e
fuel components by the exte~nal and internal heat sources from xhich there axiae
a number of effects related to the heat exchange betKeen the liquid, the wall of
the tznk and the gas, the boiling, the stratification of the component~ steam
conde:~sation, superchaxge of tl~e diffusion gasas etc. (aeo chapter 5).
To preserve the cryogenic fuel components in the liquid atate during the entire
_ flight, it is expedient to use several types of thermostatic control simultaneous-
ly ("passive" and "active") or one of the more efficient of them. Many paPers
are devoted to the problem of various thermostatic control systems. We used the
results of these for a thermal design to obtain approximate relationships. By
using these relationships to the first approximation, it is possible to determine
the axea of application of the various thermostatic control systems and the mass
relationships between them.
In some caseso thermal currenta from engine and power installations affect con-
siderably the thermal mode of the KLA. Here heat transmission is implemented
primarily from the jet of the operating engir.e and, in some cases, becomes the
_ determining factor in the structure of the ionizin~ radiation of the engine or
the poxer installation. The calculation of the heat exchange systems necessaxy
for the operation of the poxer installations that affect the thermal mode of the
spacecraft as a Whole is a complex problem.
Already, at the very early stages of thermal design of the KLA units and systems,
there arises the necessity of carrying out a fairly detailesi calculation of the
design of the thermal mode. Thus, frequently xe approach the probiem of the
theoretical numerical simulation of the thermal mode of the design of the space-
craft. Txo basic approaches are possible here. The first approa,ch is investi-
gating the thermal mode of individual design componen,ts. In this case, as a rule~
it is possible xith some difficulty to take into account the mutual thermal ef-
fect of the design components on each other. The seeond approach involves an
investigation of the thermal mode of the spacecraft by means of a thermal model
deter:nined in a certain ma.nner. The model deacribes the thermal model of the
most important :,A design components and takes into gccount their mutual effect on
each other (chapter 4).
In the procesa of mathematical simulation to describe the thermo-physical process-
es~ experimental relationshi.~s and methods obtained previously are used, as a
rule. The mathematical model prepared for investigating the LA thermal mo~e
makes it possible to investigate it for various conditiona of external loading,
to take into account the variability of the tharmo-physical characteristics of the
materials and coatings and, moreover, the changes introduced in the design solu-
tion.
Various proposals and methods are used in preparing the mathematical models. Re-
cently, methods based on using the theory of graphs components xere introduced
in mathematical modeling. These methoda may ba used successfully also in the
autoraated design of LA and other ob,jects.
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In a number of cases~ mathematical aimulation makea it possible to investiga,tp
thermo-physical pr~cesses, whose phyafcal modeling cannot be done undar terres-
trial conditions. Amon~ such processes are therao-gravitational convection of lox
bailing point liquids in LA fuel tanks under Keak field conditions and several
others.
Mathema.tical modeling plays an important role in investig~2.ting problems of inter-
dependent heat exchange in the proposed complex configuration design because it
permits the consideration of the progress of the thermo-physical processes, in-
vestigate~ their mutual effect~ solves the problem of optimizing the thermal mode
of the desi~ etc.
It should be noted that the LA thera~al mode optimization is closely related to
the solution of inverse problems of heat er.change in complex shape designs.
5uch proble~us include~ for example~ those of restoring the termperature profile
in mutually related components of the considered design in accordance Kith the
known values of temperatures in individual co~ponents. Although these problems
are complex even in the formulation plan, it is to be hoped that their solution
will be obtained in the very nez.r future.
A completely special ~ea is the part of the thermal design of KLA operating under
- the effect of intensive ionizing radiation from several poxer installation of the
spacecraft or cruisin~ engines using nuclear fission energy.
Using the solutions of the problems for optimizing the mass and shape of multi-
layer protection from the reactor radiation, taking into account the evaporated
fuel component, it is possible, in genera'l, to solve the comprehensive problem
on choosing the optimal pa.rameters of radiational protection, the geometry of the
fuel compartment~ the ~rass of the adapter between the KLA booster units and the
caxrier and evaporated component.
Of considerable interest are problema rel~ted to the selection of basic paramet~rs
for thermal protection systems for reentry, designed to de~iver the crew and sci-
- entific equLpment to the surface of the planet. Tl~e apacecraft upon reentry to
the earth's aurface is subjected to the strong effect of the environment. In
- developin~ the design of such ~pacecraft a thermal protection aystem that pro-
tects the structure, units, devices~ crerr etc. from temperature loads playa a
considerable role. The investigation of ths operation of thermal protection and
- the choice of its parameters, taking into account the special feature of the en-
vironmental effects ~,nd limitations imposed by the garameters of the reentering
spacecraft, is one of the main parts of the problem of the thermal design of the
units and the spacecraft as a Whole.
Amono the many limitations primarily imposed on the choice of spacecraft param-
eters in therma.l design are the limitations on the choice of the shape. As a
x~ule~ the shape of the spacecraft is selected to provide a given lift-drag ra-
tio~ stability, ease of control, maximum density of the axrangement, minimum mass
01' thermal protecti~n and the spacecraft as a whole~ its ,greatest reliability and
efficiency, comfort of the crax etc. Hera it should be taken into account that
_ the shape of the spacecraft directly affects the type of thermal protection coat-
in~ used. ~
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A number of complex problems arise in developing the thermal protection ayatem.
(~ne of them is the study of the interaction betw~en the heated gas and the thermal
prot.ection materials and the selection of the most efficient of them. The inter-
~ction betwe~n the heated gas and the thermal protection materials is due to many
- mutually r~lated processes. The solution of this problem in the general ca.se is
based on the solution of a system of differential equations that describe the
phenomenon of nonstationary heat-mass transfer in a gas-body system and~at present~
represents great difficulties from the mathematical viexpoint. Therefore, in
studying such complex phenomena (especially of nonstationary processea)~ the role
of experimental investigations~ xith the wide appZication of automatic complexes
for gathering and processing data, increases essentially. In many cases, the
bases of software for nonstationary thermal experiments are methods for solving
inverse problems of heat exchangei the determination of limitin~ thermal modes,
identification of the processes of heat and mass transfer, the restoration of ex-
ternal and internal thermal fi.elds etc.
The thermal de5ign of units~ systems and of spacecraft as a Hhole~ is onYy one
of many pa,r+s of the ~eneral design of the a~a.cecraft. However~ this paxt differs
essentially from all the others by the fact that besides the solution of basic
pr~blems of LA design~ related to mass and ballistic desi~, it demanda from the
investi~ator an extensive analysis of physical phenomena and the application to
t!-,is analysis of many related sect�ions of science and technol.ogy?.
~~ren a short list of basic pr~blems related to thermal design attests to the neces-
sity of applying a wide range of problema in therual physics, heat-mass exchange,
thermodynamics~ space physics, ballistics~ dynamics~ nuclear physics, mathematics
etc., needed at the present time to solve the design problems.
This book is Qne of the first that generalizes the numerous individual papers in
the area of thermal desi~n of spacecraft units snd~ to a certain extent, has the
nature of a compendium.
'I'he main purpose of ~he book is nnt onJy to recite the bases of the thermal de~ign
of some components of spaGecraft and thermal-intensive units, but also to provide
the methodology for solving similar problems and provide relatively simple and
convenient relationships~ useful in determining a number of basic eharacteristics
of sgacecraft systems and the calculation of power and thermal parameters at +,he
initial stage of design.
Whe~ever necessary~ concrete numerical values are given in the book. This makes
it possible to use the book as a useful reference mar?ual also.
Along with the bases and problems of thermal design of spacecraft, the book pre-
sents the concept ~f the modern level of investigations on the considered ques-
tLons. The bLbliography given in the book is intended basically for further,
m~re extensive study of the individual chapters.
At the re:~uest of the author, subchaptera 4.3~ 4.4 arid 4.5 have been xritten by
the senior staff xorker~ candidate of technical sciencas, Y. S. K:hokhulin~to Nhom
he expresses his sincere gratitude.
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~ vl\ VI'cal.~t1L VJC. VI\LL
- The author is especially grateful to academician V, P. Mishin, xho helped greatly
in the preparation and writing of this book, as well as to reviewer professor
Yu. V. Polezhayev~ doctor of technical sclences~ for valusble comments and sug-
gestions he made When examining the manuscript.
We are grateful beforehand to all readers xho xill make critical remarks and sug-
gest improvements to the book. ~
Table of Contents page
Introduction 3
Chapter 1. Basic power characteristics of the engine
ir,stallations of the ascent units of spacecraft 9
1.1. General statement of the problem 9
1.2. D~uations of motions 9
l.j. Basic mass relationships of spacecraft 11
1.4. ~elative masses of engine installations~ fuel compartments
and other units 13
1.5. Power characteristics ~f engine installations and their
effect on spacecraft 17
I.6. 5election of optimal control program for fixed parameters
of the engine installation 19
Chapte.r 2. Effect of external thermal conditio:zs and cosmic
radiati~n on the system characteristics for a long
~ flighi 23
2.1. Thermal currents acting on tt~e spacecraft 23
2.2. Thermal mode of units z7
2.3. Primary cosmic radiation 30
2.4. Radiation zones 32
2.5, Solax cosmic raafation 35
2.E. Determination of characteristics of the syatem for
protection against cosraic radiation 3?
2.7. :alculation of poWer characteristics of the system for
minimal radiation danger 51
Chapter 3. Basic thermal loads acting on reentering apacecraft 54
3.1. Motion equations of reentering spacecraft 5~
3.2. Ballistic reentry ~
3.~. Planning reentry
3.4. Aerc~dynamic heating of the reentering spacecraft 61
3�5� Some problems in the thermal protection theory ~
3.6. Approximate evaluation of thermal currents acting on a
reentering spa::ecraft 67
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Chapter 4. Effect of currents from internal hmat sources an
the efficiency of unit design 72
4.1. General statement of the groblem 72
4.2. ~hgine installations sources of ra:diation 72
4.3. Design of a thermal model anci calculation of the
temperature distribution in the design from thermal floxs
of engine and power systems 75
4.4. Algorithm for the solution of single-dimeneional heat
conduction equations given in graph 79
4.5. Method for calculating the temper~.ture distribution in the 83
spacecraft design at an arbitraxy thermal loa.d.
Chapter 5. Thermal design of units ~8
5.1. General statement of thermal de~ign problem of a complex 88
system
5,2. Special features of tl~erma3 deeign of fuel compartment 93
5�3� Passive systems of thermmstat control
5.4. Certain problems of heat and mass exchange in tanks for
fuel components being h~ated by external and intern~l
heat sources 97
5.5� Certain parameters of the fuel systems (active ayatems 1~5
of thermostat control) 108
5.6. ~'ficiency of thermostat control system
- Chapter 6. System analysis in thermal design of booster units lil
for spacecraft and methods for automa.tion
6.1. N~a.ss relationships of the sgacecraft taking into account ili
certain limitations
6.2. Protection of spacecraft compartments from the radiation
zones of the earth~ solar bursts~ primary coamic 112
radiation and other types of ra.diation
6,3. Optimization of basic characteriatica for thermal design 117
of a flat shield
6.k. The determination of the optimal shape of the ahield 123
6.5. Selectinn of efficient parameters and ahapes of certain 128
units of the booster stages of the apacecraft
6.6. Automation of the thermal design procesa of spacecraft 133
units
Chapter F'rinciples of thermal design of reent~tiring apacecraft 138
7.1. Basic assumptions i38
7.2. (;eneral atatement of the problem 1~
7.3. Certain systems for cooling units of reentering spacecraft 148
- 7.4. Selection of the reentry tra~ectory and control of the 1~
spacecraft
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7.5. Mass analysis of parameters of the spacecraft and its units 156
7.6. Possible areas of application of inveree problems o~ heat
conductivity in the investigation of thermal modes of '
reentering spacecraft 158
7.7. General statement of the problem of e.~cperimental
investigatfons on nonstationary deatruction and heating of
thermal protection coatings (TZP). Use of computers in
carrying out the thermal experiment 161
Bibliography i69
COYYRIGHT: Izdatel'stvo "I~Iashinostr~yeniye", 1981
2291
CSO: 1861/21
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*JDC 6zi.455.32
~'OW~t 3UPPLY AND CCNTRCL 5YSTr:M.S FOR II,ECTRIC,AL ROCKGT ENGINES
- Mc,scaw SIST "~MY PITAN IYA I UYRAVLII~IIYA ~,IICTRICH~iKIMI RAK~TNYMI DVIGAT~,YAMI in
Russian 1981 (signed to press 30 Jan 81~ pp 2-4. i36
[Annotation~ foreword and table of conten+.s from the book "Power 5upply and Con-
trol ~ystems for r;lectrical Rocket ~gines'; by Marks Mikhaylovich Glibitskiy,
Izdatel'stva "Mashinostroyeniye"~ 930 copies, 1~6 pages~
~Text 1 Annotation
This book presents theoretical and engineering problems involved in producing
- optimized systems for the power supply and control (SPU) of electrical rocket
engines (T~' D) and spacecraft (KLA). The most developed types of ERD and functional
circuits for their power supply and control xere considered. Questions of protec-
tion for various types of SP'U and ERD were elucidated under anomalous and emer-
gency situations.
The book is intended for specialists in the investigation and development of elec-
trical rocket engines.
Foreword
The assimilation and peaceful utilization of space for practical purposes was be-
~ by Soviet cosmonauts on 4 October 1957 and are b~ing developed and improved
at a rate unseen in cther sectors of ecience and technology. I,ox motive power
engines electric rocket enginea (ERD) occupy a apecial and very important place
in the means and devices for developing,jet enginea for epacecraft (KLA). The ERD
have a number of essential differences that place them outside the competition with
other types of jet engines. The advantages of the FRD Were confirmed by long and
successful experiments in space on axtificial earth satellites launched in the
USSR and the United States.
Broad programs were developed and published recently for further investigations
- and assimilation of space, including the creation of orbital technological shops
laboratories, global communications systems satellites, navigation and meteorolog-
ical satellites etc., in xhich it is proposed to uae ERD. It is planned to use ERD
for cruising propulsion enginea for interplanetary flights, and ~o investigate
distant space and comets. It is planned to use groups of ERD xith a total poxer
counted in hundxeds of kiloxatts as motive power engines of booster units
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for the delivery of sections of solar spa.ce electric poxer plants into geosta-
tionary orbits, called upon to supply electric power to industry on the earth by
the end of the txentieth century. It is planned to launch scale models of such
150 kw power plants in 1983 L40].
Broad periodical literature and a number of monographs of Soviet and foreign scien-
tists axe devoted to the physical bases and principles of ERD operation~ These
cover theoretical and engineerin~ problems of developin~ tt~e F~D proper and the
primary power sources for thE KLA, While the problems of the inveatigations and
developments of ERD power supply and control syatems Were cor?sidered extremely
_ brlefly only in a small number of journ.al articlea. This book gives a systematic
presentation of the theoretical and practical problems that axise in the develop-
ment and design of ~D power supply and control systems.
The author is sincerely grateful to professor A. I. Morozov, doctor of technical
sciences,whose initiative and constant attention facilitated the appearance of the
book~ to reviewer professor D. D. Sevruk. doctor of technical sciences, as well as
to R. K. Chuyan~ doctor of technical sciences, and assistant professor V. Kim for
their helpful advice and fruitful discussions of the manuscript that facilitated
its improvement.
_ Table of Contents
Page
Foreword 3
Chapter 1. Power supply and control systems for electric rocket
engines 5
- 1.1. r~ ectrical engines in space 5
1.2. ~chematic diagrams of r'~D and their classification 6
_ 1.3. r~ ectric rocket engines $
_ 1.4. System approach and problems of optimizing control of
- ~D installations (':~?DU) 10
Chapter 2. ~tD types and requirements of power supply and control
_ syatems 14
2.1. Thermoelectric plasma enginea (P'D) 14
2.2. ~,'lectric heatin~ rocket engines 14
2.3. Arc electric rocket engines 15
?_.4. Pulse plasma en~ines 17
2.5. Ionic engines (ID) 19
= 2.6. ID power supply and control s stema z3
2.7. 5tationary plasma en3~ine (SPD~ 27
2.8. Front Ha,ll eng.ine (TKhD~ z9
2�9� Grouped electric r~~cket enginea 31
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Page
Chapter 3. Primary power supplies of spacecraft 32
- 3.1. Chemical prima,ry poxer supplies ~ 3~+
3.2. Atomic primaxy power supplies 35
3�3� Solax batteries 39
Chapter ~econdary EftD power supplies 42
4.1. Switching installations~ devices and apparatus 43
4.2. Thyristor switches and inverters 48
4.3. Transistor switches 52
4.4. Trar;sistor inverters and converters 55
4.5. Transformers and magnetic switches for F~D poxer supply
and control systems (SPU) 63
4.6. ,~pecific chaxacteristics of electronic systems for
energy conversion ?6
Chapter 5. L~RDU control and protection systems 78
5.1. Optimal control of KLA witn ERD 7a
- 5.2. ERDU optima.l control 80
5�3� Anomalous and emergEncy ,a~~ies in ERDU poxer supply systems 84
5.4. Methods and means for protecting inverters in II3DU 88
5�5� ~'rotection of power rectifiers in nucleax turbogenerator
'r:RD power supply installations 93
Chapter 6. C?ptima.l design of SPU for 1~D 96
6.1. Problems of optimizing KLA with II3D 96
6.2. Criteria of quality in aptima.l design of ERDU and
question of reliability 97
- 6.3. Problems and methods of seaxchin~ for optimal design
solutions for SPU r~iD 108
Chapter 7. apecial features of F~RD power supply and control systems 122
7.1. ~'fect of the space environment on materials and c~mponents
of SPU ERD 122
7.2. Heat exchange under space conditions and thermal modea
of SP'U ~D 124
7.3. r`~ectric rocket engines in the very near future 130
Bibliography 133
COPYRIGHT~ Izdatel'stvo "Mashinostroyeniye"~ 1981
2291
Cso, 1861/20
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NUCLEAR ENERGY
UDC 621.311.25:621.039.004.17
EXPERIENCE WITH BRINGING ON LINE AND OPERATING THIRD POWER U?~TIT WITH FAST
NEUTRON REACTOR AT BELOYARSKAYA AES
Moscow ELEKTRICHESKI.YE STANTSII in Russian No 9, Sep 81 pp 12-17
[Article by V.I. Kupnyy and V.V. Budziyevskiy, engineers, Beloyarskaya AES]
[Text] Power unit No. 3 of the Beloyarskaya AES with electrical capacity of
600 MW has three loops and is designed with a triple loop conf iguration for heat
removal from the reactor; in this case, metallic sodium serves as the coolant for
loops I and II, while water (steam) is used as the working materiaZ of loop III.
The ty~e BN-600 fast neutron reactor with a thermal capacity of 1,470 MW has an
- integral layout (tank configuration) for the loop l equipment. The reactor takes
the form of a cylindrical tank 12.8 m in diameter and 13.0 m high , which is mounted
on roller supports in the reactor shaft. The core is positioned in the~reactor
tank as well as the three main circulation pumps of loop I, the six process heat
exchangers of loops 1- 2(two each for ~ach loop), th~ internal tank protection
with the thermal shields and the neutron channels. The control drives for the
- contro] and protection system mechanisms and reloading are positzoned on rotating
plugs installed in the upper portion of the reactor.
Each of the three loops includes the following: a once-through modular eight
section type PGN-200M steam generator with a steam output of 660 t/hr having steam
parameters of 505 �C and 140 kgf/cm2, a K-200-130 turbine installation with a
capacity of 200 MW, a TGV-200M generator, a DSP-800 water feed system with a
deaerator and three PE-380-200/185 electrical feed pumps.
The sodium is circulated in loops I and II by single stage centrifugal pumps. The
rotor of the pump is mounted in a cantilever bracket fashion on the lower end of
the shaft and has a hydrostatic bearing. The utilization of a main circulating pump
with an asynchronous rectifier stage makes it possible to continuous~y vary *_he
r.p~m~'s of the pumps in a range of 250 ta 970 r.p~m.
The high parameters of the live steam and the superheated pracess steam made it
possible in the design of the unit to employ seandard turbine room equipment with
the 200 MW units of thermal electric power stations.
- Power unit No. 3 is outfitted with a"Kompleks-Uran" information computer system,
which consists of the M-60 data subsystem, a computer subsystem designed around
the series produced M-7000 computer and "Orion" data displays.
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r Nn~ ,
- ao ,
60 �
yo ~
~0 ' months
a
~ 2 3 4 S 6 7 9~~eca~ta
- Figure 1. Consolidated graph showing the bringing of the
- capacity of power unit No. 3 on line at the
Beloyarskaya AES.
The "Kompleks-Uran" system calculates and analyzes the technical and economic indi-
cators (TEP) of the operation of the mai.n and awciliary equipment of the power
unit, records the preliminary and postaccident values of the digital and analog
parameters and the production process protection which has actuated. Special
programs provide for printout of the results.
The operation of power unit No. 3 started April 8, 1980 following the co:.nection
_ of the turbogenerator to the Sve~rdlovsk power administrat3on power system. Fol-
lowing comprehensive testing, the bringing on line of the unit capacity was started,
which was accomplished in the following stages: operation at a reactor power of
30 percent Nnom ro check the operation of the turbogenerators under load, the
adjustment ofi th~~ water supply mode, operational adjustment of the equipment and
power unit system as well as to perform investigative work; operation at a reactor
capacity of 40 to 70 percent of Nn~ to more precisely specify the operational
modes of the core, the sodium loops and the major equipment as well as to perform
investigative work; operation at a reactor capacity of 80 percent of Nnom to more
precisely specify equipment operational modes at naminal steam parameters and to
- perform investigative work.
- A consolidated grapll of the bringing on line of the capacity of power unit No. 3
~ is shown in Figure 1. Increasing the reactor power above 80 percent of Nnom is
_ possible only after the core becomes steady-state following the execution of three
fuel reloadi:ngs because of the core physics.
Following comprehensive tests of the production process protection and interlocking
(TZB) of the unit and tests of the disconnection of the internal loads of the unit,
the reactor was brought up to 0.1 percent of the nominal load. On April 2nd, 1980
following the filling of the steam generators (PG) with feed water and the estab-
lishin~ of circulation through the third loop, the reactor power was increased up
to five percent and the steam generators were brought up to steam operating condi-
tions. At this power, the setting of a large number of steam generator safety
valves was accomplished. On April 8th, the capacity of the reactor was brought
up to 30 percent of Nnom~ the sodium temperature of loop I at the outlet fram the
reactor reached 400 �C and the turbogenerators were sequentially connected to the
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power system. To improve safety, the settings of tne rate of change in the
coolant temperature during start was reduced by a factor of seven below the
design v�lue whiZe the coolant parameters and steam were lower than the nominal
values biit provided safe operating conditions for the steam generators. During
the first, three days of steam generator opera~-ion, the water chemistry conditions
in loop ~III were established in accordance with the norms for once-through steam
generator. In the fi~st stage of bringing the unit capacity on line, the major
starting conditions were tested as well as those for shutdown and operation at the
steady-state po~aer level. Tests were made of accident modes, including the actu-
ation of the fast reactor accident protection (BAZ), slow reactor accident protec-
tion (P~,Z) and the disconnection of one loop. In a11 operating modes, special
equipmerit was used to carefully monitor vib~ation and tensile and thermal stresses
~n the ~:omponents of the major equipment and piping.
The operational setting of the automatic controllers for the production process
was accomplished during this same period. The radiation status in the prodiiction
rooms and the environment was studied. As a result of the first stage of bringing
the uni.t capacity on line, a considerable number of design errors were ascertair.ed
_ i.n the setting of the transdiicers for the rate of flow of the feed water and the
steam ~;enerator steam, something ~ohich created considerably difficulties in determ-
ining the thermal balance of. the unit. The thermal balance was determined fram
the parameters of loop II, where the electromagnetic flow rate meters were graduated
, by a correlation technique prior to the power start of the unit. Errors were also
discovered in the plan for the setting of the condensate level t:ansducers in the
- condensors of. the turbines, which caused two shutdowns of the turbogenerators
(TG) Y,ecause of overflow of the condensate collectors. The unsuccessful design
solut-Cons for the steam supply for internal loads and the steam ejector installation
made the operation of the unit considerably more difficult. The large amount of
water passed through the fast acting emergency valves for water discharge from
- the steam generators led to the undetected drying of the steam generators at low
_ power� levels. Because of this, a decision was made to dispense with ~he titiliza-
- tion of thes~ valves and to dry the steam generator throubh the dump valves at the
outlat from the evaporation modules of the steam generators. Considerable rework-
ing was done on the turbogenerator lubrication systems.
Frorirtay 14th to June 15th 1980, the unit was shut down to make changes and repairs
in ~he major auxiliary thermomechanical equipment. During this time, the feed
water and steam flow rate transducers for the steam generators were rebuilt. At
steam generator No. 6, the covers for the water and steam cavities of the evapor-
at~_on and steam superheating modules were removed to inspec~ them internally; in
t~l"LS case, no deformations were discovered or corrosion damage and deposits.
Measurements were made of the energy liberated in the fuel packets and the sodium
flow rates through the packets of the main core were measured. The energy liber-
at.ed was in line with the design values, and sodium flow rates matched those
me:asured prior to the power start. Tests were performed to sp~ci.fy the hydro-
d~~namic ctiaracter~stics of loops I and II more precisely and to study the natural
c~~rculation mode (YeTs) through loops I and II. Additional safety valves were
installed on the ~ream generators at the live steam headers because of tlle inade-
qtiate carrying capacity of the safety valves called for i.n the design.
14
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On June 15th 1980, the reactor was brought up to a power level o� 40 percent of
Nn~ and the second stage of bringing the capacity on line was begun. The gener-
ator capacity was specified from the thermal balance and was gradually increased up
to 50.6 percent of Nnom on June 21. The adjustment of the automatic controllers,
the automatic water chemistry monitoring system as well as the implementation of
programs for production process recording and calculations continued during the
operation of the unit. During the process of bringing the capacity on line, with
th2 suspicion of a break of the seal between the loops, four steam superheater
modules in stean? generators Nos. 4 and S were cutoff by means of stop valve fit-
tings and brought out for repair. The work to remove the modules for repair
revealed the insufficiEnt capacity of the system f~r feeding nitrogen irito the
steam water cavities of. the steam generators, scm?ething which following the dis-
connection of a defective module can cause soditun to get into the pipes of loop III.
This is promoted by the long duration of the discharge of sodium f.rom loop II, the
unreliable operation of the sodium fittings and thP routing of the water and steam
headers below the modules. In the process of bringing the power up, a degradati~n
was observed in the quality of loop II sodiim? because of the inadequate capacity
of the sodium cleaning system, which did not provide for good retention of the
diffusion hydrogen liberated in the sodium, because of which, the cleaning system
operates at maximum capacity with the reserve trap filters inserted,
On August 22nd, 1980, the power of the reactor was increased up to 70 percent of
Nnom~ and the temperature of the sodium in loop I at the outlet fram the reactor
*hen reached 500 �C, while the electrical capacity reached 380 MW.
The third stage of bringing the power unit on line began on September 16, 1980:
the reactor power was increased up to 80 percent of Nnom ~1,176 MW), and in this
case, the temperature of the loop I sodium at the outlet from the reactor reached
510 �C, while the electrical capacity 440 MW.
The unit was shut down on October 3rd, 1980 for scheduled preventive maintenance.
During the scheduled preventive maintenance, physical experiments were performed
on the reactor to precisely specify the effects of radioactivity, and tests were
made to determine the level of residual heat emission and heat losses in loop I.
The pi_ping bundles for t}ie high pressure~heaters (PVD's) were opened up and
flushed with feedwater, after which the high pressure heaters were connected to
th` steam generator water feed channeY. The excitation and the run-out units for
the block of turbogenerators were set up and adjusted and the pennissible dispari-
- r.ies in the rotational speeds of the main circulation pumps of loops I were
determined.
After finishi.ng the scheduled preventive maintenance, the block was sequentially
run up to the 60%, 70% and 80% of Nnom power levels.
Following the connection of th e high pressure heaters of the steam turbines,
because of the increase i.n the feed water temperature up to 220 - 230 �C, the
- iron content in the feed water ahead of the steam generator increased from 5 to 7
ug/kg up to 17 - 23 ug/kg by virtue of the naturally occurring process of the
forn~ation of the protective magnetitefilm on the internal surfaces of the pipe
bundles of the high pressure heaters. The connection of the high pressure heaters
increased the elec.trical capacity of the unit up to 460 MW.
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In the process of bringing the capacity on line, tests were made of the accident
and transient modes of the unit, including a test for the run-out of turbogenerator
Nd, 6 with the internal load applied, as well as tests of the skewing of the liber-
ated energy field, studies of the natural circulation conditions for loop 2; a
nondesigned circulation mode was tested for loop III with. the return of the conden-
sate of the RR-13 expander to the unit deaerators.
- Operational experience with power unit No. 3 with the BN-600 reacto~ confirmed
that the actual parameters matched the design values. The major power engineering
characteristics of the power unit equipment matched ~he design values, the unit is
easily controlled and tl:~ operational modes are stable.
Production process protection systems, interlocking and the automatic regulation
systems were incorporated in the operation step by step in accordance with the
rise in the power level, providing for safe unit operation in each of the steps.
On April lbth, 1981, the power unit operated in a power generation mode for 240
days, and generated 1,841,522 KWH of electrical power. The nuclear reactor oper-
ated for 147 effective days and the maximum fuel depletion reached 3.5 percent of
the heavy atoms.
The radiation status is good in the unit: no loss of seal in the fuel rods was
detected, the ionizing radiation fluxes correspond to the design values and the
emission of radioactive isotopes into the ventilation pipes when the unit is in
operation amounts to 1- 3 Curie per 24 hours.
During the operation of the unit, systems were prepared for operation which were
not included in the operation at the outset of the power start because of the
lack of the need for them: the gas purification system, the system for cooling and
cleaning the drum of spent packets, the system for cooling and cleaning the water
of the holding tank, the equipment of the holding tank and the fuel reloading
rooms.
The siiccessfulmastery of the nominal power level of the first reactor charge was
the result of the combined labor of the operating and repair personnel, as well as
the personnel of the set-up and research organizations. The success of the collec-
tive is the successes of the senior operator of reactor shop No. 2, Yu.D. Krutikov,
senior machinist of turbine shop No. 2, Yu.P. Fitsko, electrical shop electrician
L.V. Okulov, welder of the thermal automation equipment shop, V.PI. Malygin, shift
chief of the electrical shop V.I. Batanov and shift ehief of the station V.I. Anikin.
Experience with bringing on. line and operating power unit No. 3 with the BN-600
reactor made it possible to check and work out the design and production process
solutions for the design of even higher capacity power units w~th sodium cooled
fast neutron reactors.
COPYRIGHT: Energoizdat, "Elektricheskiye stantsii", 1981.
8225
CSO: 8144/062
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UDC [621.311.25:C2Z.039]-'
SELECTION OF SAFETY DEVICES TO COUNTER EXCESSIVE PRESSURE IN PRIMARY AND SECONDARY
LOOPS OF NUCLEAR ELECTRIC POWER STATI~NS WITH WATER-COOLED REACTORS
Moscow ELEKTRICHESKIYE STANTSII in Russian No 6, Jun 81 pp 11-14
[Article by A.M. Bukrinskiy, candidate of the engineering sciences and E.Ya.
Chernikova, engineer, All-Union Institute of Thermal Engineering imeni F.E.
Dzera_hinsky]
[TextJ The ~,rimary and secondary loops of a reactor installation are protected
against excessive pressure build-up by safety valves, which are installed on the
volLUne equalizer and the steam generators. The capacity o� the safety valves is
chosen in accordance with thn regulations of the State Committee of the Council
of Ministers for the Supervision of Industrial Safety and Mining Inspection [1].
This capacity should be sufficient, so that with the actuation of the valves, the
pressure does not exceed the design value by more than 10%.
. The safety function assigned to relief valves is an extremely important one. But
since it is possible for them to fail to close after actuation, the capacity of
relief valves should not be excessively high.
Primary Loop Safety Valves. We shall consider the major possible reasons for a
pressure increase in the primary loop. In general, the following three cases are
possible:
--Primary lcop overheating due to a mismatch between the heat produced in a reactor
and the heat delivery in the steam generators;
--Steam generation in the velume equalizers by the electric heaters;
--Pressurization of the primary loop by the pumps during its filling or during a
make-up feed.
The latter case is practically nonexistent at present AES's as well as those now
being planned, since the capacity of the piston make-up pumps being used, as a
rule, is low while centrifugal pumps are chosen from design calculations so that
the pressure head developed by them does not leac~~to the actuation of the safety
valves. Of the remaining two cases, the first is the most dangerous.
A mismatch between the heat delivery in a reactor and its removal in the steam
generators occurs in the case of disturbances which lead to an excess reactor
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power or to a drop in the turbine load. However, the greatest mismatch bet~een
the heat delivery in a reactor and the heat removed in the secondary loop will
appear with a complete loss of the capability of removing heat from the secondary
loop.
- The elimination of the mismatch between the heat delivery and removal in the pri-
mary loop is accomplished by means of reducing the energy produced in the reactor,
something which is realized by tl:e fast response accident protection for the reac-
tor. The fast response reactor accident protection (AZ1) performs a safety func-
tion, and for th is reason its reliab ility should be no less than the reliability
of the safety valves. Therefore, at AES's, in contrast t~ haw this is handled in
boiler facilit~es, there is no basis for failing to take into account the action
of the reactor accident protection when determining the capacity of the relief
valves.
Thus, the emergency reactor protection should be used for the reactor installations
as the first stage of protection against excessive pressure, while volume equalizer
relief valves should be used for the second stage when the heat delivery to the
primary loop is determined by the level of residual energy generation in the reac-
tor.
For emergency heat removal from the secondary loop, it is necessary to vent the
steam from the steam generators and deliver a make-up feed by means of the emer-
gency feed pumps. Since this system performs an important safety function, it
is usually designed in accordance with the requirements of the general principles
for the assurance of the safety of nuclear power stations in their design, const-
ruction and operation [2].
It is necessary to choose the capacity of relief valves by working from an analysis
of accidents with a complete loss of feed water. Such an accident can happen in
the case of a break in any feed water pipe. In the worst case, it occurs with a
- break in the feed water pipe between the check valve and the stea~n generator, as
shown in the schematic of I'i~ure 1. In this case, besides cutting off all of the
main feed pumps because of the pressure drop in the main pressure line, there will
be a leak from one steam generator. If for any reason there is a total failure of
the emergency make-up feed system for the secondary loop, as happened at the
T'hree Mile Island 2 AES [3], the accider.t will develop under conditi~ns of a total
feed water loss.
Such an emergency situation has been analyzed for a power unit with a WER-400
reactor [440 KW water-moderated, water-cooled power reactor] using the mathematical
models described in [3, 4] as coell as the "Kontur", "Kompensator" and other pr~-
grams based on them.
The results of the analysis are shown in Figure 2.
Unti1 tile complete evaporation of the water from the secondary loop, no dangerous
situation arises since after the actuation of the reactor accident protection, a
sufficient removal of heat from the primary loop is assured. Howev~r, after the
evaporation of the secondary loop water, the cooling of the primary loop ceases
and it begins to heat up rapidly. The heating of the pr3.mary loop is accompanied
I8
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Figure 2. Feed pipe break between the check valve and the steam
generator.
Key: 1. Water mass rate of flow through the reactor;
2. Pressure in the operating steam generators;
3. Total steam mass rate of flow;
4. Thermal power of the reactor;
~ S. Quantity of water in the operating steam generator~;
6. Quantity of water in the emergency steam generator;
7, 8. Waten temperature at the outlet and inlet of the core
respecti~ely;
9. Pressure in the volume equalizer;
_ 10. Water level deviation in the volume equalizer from the
nominal value; ~
A. Bare exposure of the feedwater branch delivery pipes;
B. Actuation of thp fast response reactor accident
protection.
by a rise in the pr~ssure. This leads ro the op~ening of the relief valves in the
volume equalizer. Initial.ly, the ste.am relief valvea actuate and then after the
steam cushi.on is forced out through the relief valves, the hot water runs out. The
most dangerous moment begin~ when the ~rimary coolant starts to boil. At this point
in time, the greatest volumetric rate of coolant flow is dumped through the relief
valves and this is the governing factor in the selection of the capacity of the
volume equalizer relief valves.
- Because of the large water reserve in the secondary loop of AES's with WER-440
reactors, a dangeruus situation can arise no earlier than 2.5 hours after the onset
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of the feedwater loss accident. At AES's with the WER-1000 reactor, the relative
water reserve in the secondary loop per unit capacity of the reactor is somewhat
less, and for this reason, a dangerous situation with the boiiing of the primary
loop coolant can begin approximately one hour after the start of a similar accident.
The operational personr~el have this time availab le to take steps to restore the
feed for the secondary loop. We will note that at the Three Mile Island 2 the
_ emergency feed was restored eight minutes after the start of the accident.
The restoration of the feed is one of the conditions for assuring the operational
reliability of AES's, since otherwise the water will boil out of the primary loop,
something which in the f inal analysis can cause damage to the fuel assemblies in
the reactor core.
Settings were recommended for the actuation of the reactor accident protection with
respect to primary loop pressure and with respect to the water level in the steam
_ generators on the secondary loop side based on an analysis of accident situations
for AES's with WER-440 reactors.
Since relief valves perform an important protective function, the nimiber of them
and their capacity should be chosen by working from the requirements of the general
principles for assuring AES safety.
The Capacity of Secondary Loop Safety Valves and BRU-A [Quick-Acting Reducers for
Atmospheric Steam Venting] Valves. The number and capacity of secondary loop safety
valves are likewise selected in accordance ~with th e requirements of [1]. Usually,
two valves each with an overall capacity equal to the capacity of the steam genera-
tor are installed on each steam generator. If it is assumed that such a capacity
is needed to carry out the safety functions assigned to the valves, then the instal-
- lation of two valves with a capacity of 50 percent each is inadequate to satisfy the
unit failure principle. However, this is not so in this case.
Ntm?erous calculations perfornied at the All-Union Institute of Thermal Engineering
imeni F.E. Dzerzhinskiy for the engineering substantiation of the safety of AES's
in the process of designing them show that in all cases, the indicated capacity
of a single valve is sufficient to protect the secondary loop against excessive
pressure. Consequently, the second valve can be treated as a standby for the case
- of �ailure of the first valve. This is due to the fact that in secondary loop ex-
cessive pressure modes, when the necessity of s~fety valve operation arises, the
capacity of the steam generators does not remai;i constant, but is substantially re-
duced.
If the steam generators are not disengaged in the emergency mode, then considering
the fact that one backup valve is installed on each of them, we come to the conclu-
sion that protection against excessive pressure in the secondary loop in this in-
~ stance is assured with a considerable redundancy. From this standpoint, a more
dangerous case is steam generator operation with false actuation of the cutoff
valve (see the schematic of Figure 1). However, in this case too, a reduction in
the temperature head by virtue of the pressure rise in the secondary loop reduces
the requisite capacity of the safety valves to such an extent that one valve proves
to be sufficient to limit the pressure in accordance with the requirements of [1].
The transient process in this mode is shown in Figure 3.
21
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kg f /~n ~ ,
ti~, r, cn ;,i Pnat~ 1' lpd.esign
- ~
5
� _ J1aBne~~ue u ~8upuic~rpn
~ ~nrcpoze~:epamnpe (A)
45 i ~ i i ~ i~i~~ i~
. ; 4 ~ ?C Z~ 30 YO f00 400 c
' seconds
- Figure 3. False actuation of the cutoff valve o�
an AES with a WER-440 reactor.
Key: A. Pressure in the emergency steam generator.
Kli~~r; Z ~ ~ ~l
J:' ~~i~
- SJ- a~-
i.
s~-cn
~~s - ~fn - ' ~
S
47 - 20 ~
4$
~r 2 s~~ ~n rs za ,,o an ~~,u 4no ~
�c
310 - S ~
- 300 - ~ !
290 -
~ 2~i0 -
270 l~
ZE!' ~ ~ ~ ~ ~ ,
1 2 5'r 10 2U 30 40 100 4(~S C
- xrc/cn2 -
1Z,5
121 - ~
111 -
113 ~ ~ ~ ~ ~ ~
1 2 3 4 10 ~0 ?c1 ~r0 100 '~GOc
Figure 4. Scramming an AES with a WER-440 reactor.
Key: 1. Thermal power of the reactor;
2. Total steam mass rate of flow;
3, Mass rate of flow of water through the reactor;
4. Pressure in the main steam header;
5. Water temperature at the outlet from the core;
6. Water temperature at the inlet to the core;
7. Pressure in the volume equalizer.
In addition to safety valves, fast-acting reducers are installed for dinnping the
steam into the turbine header and the atmosphere, the BRU-K and BRU-A respectively,
_ to protect against excessive pressure in the secondary loop.
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As nnmerous calculations have shown, the necessity of the actuation of the relief
valves in the secondary loop can arise only when the BRU-K do not work. Such a
situation is camparatively rare. Moreover, the probability of safety valve failure
can be substantially reduced if their operation in a pulsating mode is preventedi
a condition similar to that shown in Figure 2. It follows from this that it is
- expedient to assign this f unction specifically to a BRU-A.
As can be seen from the results of calculating total AES shutdown, which are shown
in Figure 4, the safety valves close 40 seconds after their actuation following the
start of the accident. The residual energy lib eration amounts to no more than 5
percent by this time. This value can also b e recommended as the requisite minimal
capacity of the BRU-A valves. Th e setting for the opening of the BRU-A can in prac-
tice match the setting for the actuation of the relief valves, while the setting far
the closing of the BRU-A should be significantly lower than the setting for the
closing of the safety valves, something which provides the capability of removing
residual heat and cooling through BRU-A following the closure of the safety valves.
BIBLIOGRAPHY
1. "Pravila ustroystva i bezopasnoy ekspluatatsii oborudovaniya atomnykh elektro-
stantsiy, opytnykh i issledovatel'skikh yadernykh reaktorov i ustanovok" ["Regu-
lations for the Installation and Safe Operation of the Equ~pment of Nuclear
Electric Power Stations, Dxperimental And Research Nuclear Reactors and Instal-
lations"], Moscow, Metallurgiya Eublishers, 1973.
2, Bukrinskiy A.M., Tatarnikov V.P., "Osnovnyye printsipy obespecheniya bezopasnosti
pri proyektirovanii i ekspluatatsii atamnykh elektrostantsiy" ["Fundamental Prin-
ciples of Safety Engineering in the Design and Operation of Nu`lear Electric
Power Stations"~, ELEKTRICHESKIYE STAI3TSIT, 1978, No. 8.
3. Cas~o W.R., Cottrell Wm.B., NUCLEAR SAFETY, 1979, 20 (4).
- 4. Bukrinskiy A.M., 'Ladorozhnaya I.N., "Matematicheskaya model' dlya rascheta
perekhodnykh protsessov s glubokimi~vozmushcheniyami~dlya bloka AES s vodovodya-
nym enerteticheskim reaktoram" ["A Mathematical Model for the Calculation of the
Transient Processes with Severe Disturbances for a Nuclear Power Station with
a Water-Moderated, Water-Cooled Power Reactor TRUDY VTI [PROCEEDINGS OF THE
ALL-UNION INSTITUTE OF THERMAL ENGINEERING IMENI F.E. DZERZHINSKIY], 1977, No 11.
5. Bukrinskiy A.M., Kol'tsova N.V., "Matematiche'skoye modelirovaniye protsessov
v parovom kompensatore ob"yema" ["Mathematical Modeling of the Processes in a
Steam Volume Equalizer"], TRUDY VTI, 1977, No 11.
COPYRIGHT: Energoizdat, "Elektricheskiye stantsii", 1981.
8225
" CSO: 8144/1794-A
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UDC [621.311.25:621.039]:699.841
- SEISMIC STABILITY EVALUATION OF NUCLEAR POWER STATION ELECTRICAL EQUIPMENT
Moscow ELEKTRICHESKIYE STANTSII in Russian No 6, Jun 81 pp 8-11
~ [Article by V.V. Piskarev, candidate of the engineering sciences, and D.K.
Ponomarev~, engineer, Scientific Research Department of the All-Union Planning,
Prospecting and Scientific Research Institute imeni S.Ya. Zhuk]
[Text] One of the major requirements for the reliable operation of nuclear elec-
tric power stations being built in seismically active regions is the preservation
of the operability of the category I electrical equipment (ETO) during and after
an earthquake. The level of seismic activity can reach such values that. taking
- into account the dynamic chara~teristics of the s~pport structures for the electri-
cal equipment, there is the probability of failure of individual constituent com-
ponents and instruments. Methods exist for the evaluation of the seismic stability
of electrical equipment, which consist in vibration load testing of the equipment
using vibration test stands. The wide scale dissemination of these techniques is
difficult because of the use of special vibration test platforms, the necessity of
shipping the e~uipment to the.test site and the simulation of the equipment oper-
ating conditions.
A method is described in this paper which makes it possible to.operationally evalu-
ate the seismic: stability of electrical equipment directly in an AES. The tech-
nique permits the determination of equipment operability under seismic loads, speci-
fied in the form of accelerometer graphs of earthquakes.
Both in domestic practice and abroad [1, 2], seismic stability testing of electri-
cal equipment, as a rule, reduces to the following tests:
1. Panels, control boards and stands which are included in the electrical equip-
ment complement are mounted on a base made of steel channels which are rigidly
fastened directly to a VPZK-100 vibration test platform. The platform is ceupled
to a hydraulic drive, which generates harmonic vibrations in a frequency range of
1 to 50 Hz. The equipment being tested is connected to electrical circuits which
simulate operational conditions, and its operability is registered during this.
The dynamic (vibration) load conditions are specified by working from the nature
of the design accelerometer recordings of an earthquake. In the case of failures
in the operation of the electrical circuits, the stiffness of the support struc-
ture is increased and the tests are repeated. As experience has shown, the costs
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of imparting additional rigidity to the electrical equipment bein~ tested is signi-
ficantly less than the cost of engineering design work related to the replacement
~ of component parts or re-equipping.
2. Individual types of component parts (relays, contactors, instruments) which are
_ incorporated in a panel under test are mounted on a special vit+ration test stand,
for example, the VEDS-400 using the same method as when mounted in the panel. Then
the product being tested is connected to the electrical circuits and its operating
parameters are recorded. The vibration test stand sets accelerations in a range
of 2 to 50 Hz with a step of 1 Hz, at which the component fails. As a result, the
amplitude-frequency limit of operability of the given product is determined. The
panel in which the previously tested components are installed is rigidly secured to
a VPZK-100 test platform. The vibrational platform sets the oscillatory load con-
ditions in accordance with the character of the design earthquake acceleration
graphs. The accelerations are recorded at the points where the components are fas-
tened. The seismic stability of the panel is evaluated by comparing the operability
limit of the product and the maximum acceleration recorded in the region where it
is secured during tests of the panel itself.
Tests of electrical equipment using the procedures given here are undoubtedly an
essential part of the work which mal~es it possible to evaluate equipment seismic
stability, and will serve as the basis for making the necessary recommendations.
A drawback to the indicated techniques is the fact that at each point in time, a
single frequency sinusoid acts on the product being tested rather than a spectrum
of frequencies, as occurs under actual conditions.
Th.~ fastening ef panels in an AES is quite different from the fastenings on test
stands. This in turn leads to a variation in the dpnamic properties of the support
structures of the electrical equipment~ and is reflected in its seismic
stability.
In the f inal analysis, a comprehensive test of the electrical equipment installed
in the assigned locations in an AES building is required, and where necessary, ad-
ditional steps must be taken to assure its seismic stability.
To resolve the problem posed here, work was done to design a comprehensive test
procedure for equipment under AES conditions. This method includes both the ex-
perimental and computational portions.
The tests of electrical equipment for seismic stability under AES conditions pro-
- vide for the performance of the following work:
--The testing of relays, contactors, instruments and other products included in the
complement of electrical equipment under harmonic loads for the purpose of deter-
mining their amplitude-frequency limit of operability. Thus, for the RP-23 relay,
the curve of the operability of this product was obtained as a function of the
amplitude loading and frequency (Figure 1). The failure curve was plotted starting
at 8 Hz, since the technical capabilities of Che VDES-400 do not allow for set-
ting adequate accelerations in the range up to 8 Hz. However, under those
accelerations which were obtained using the vibration stand in this range, the
relay maintained its operability. As can be seen from the graph of Figure 1,
, exceeding a dynamic mode of 4 m/sec2 leads to failure of the RP-23 relay at a
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frequency of 39 Hz. Working from this, the permissible loads on this relay for
its reliable functioning during a seismic event should not exceed 4 m/sec2 through-
out the entire frequency range;
a, m/sec2 -
!t, rr~C Z
ru~ f
~
SO j .
- 40'
30 1 4 2
i0- _ _ , .
f0 , ~
.~r- - -r- -
U S 10 15 20 25 30 35 40 . 45 f;~~ 5- 3;
f , Hz .
,-~-,r--