JPRS ID: 10461 WORLDWIDE REPORT NUCLEAR DEVELOPMENT AND PROLIFERATION

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• Collection: JPRS
• Document Number (FOIA) /ESDN (CREST): CIA-RDP82-00850R000500050037-1
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• Document Page Count: 25
• Document Creation Date: November 1, 2016
• Sequence Number: 37
• Content Type: REPORTS

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JPRS L/ 10461
15 April 1982
Worldwide Re ort
p
NI.~CLEAR DEVELOPMENT AND PROLIFERATION
CFOUO 4/82)
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JPRS L/10461
15 April 1982
WORLDWIDE REPORT
NUCLEAR DEVELOPMENT AND PROLIFE RATION
(~'OUO 4/82~
CONTENTS
' EAST ECJROPE
�
YNTF.RNATIONAL AFFAIRS
Development of CEMA Nuclear Power System Viewed
(Yevgeniy P. Vlasov, et al.; JADERNA ENERGIE, No 12,
1981) 1
CZECHOSLOV.AKIA
Reactor Simulation Program Results Swmnarized 11
(Ivan Kinka, et al.; JADERNA ENERGIE, No 12, 1981)..
- a - [III - WW - 141 FOUO]
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INTERNATIONAL AFFAIR5
UDC 621.039.001
DEVELOPMENT OF CEMA NUCLEAR POWER SYSTEM VIEWFD
Prague JADERNA ENERGIE in Russian No 12, 1981 pp 439-443
[Article by Yevgeniy P. Vlasov, International Scientif ic Research Institute of Con-
trol Problems, Moscow; Milos Dragny, Czechoslovak Atomic Energy Commission, Prague;
and Yuriy A. Tyurin, Department of Scientific and T.echnical Cooperation of the CEMA
Secretaxiat, Moscow: "A Study of Methodological Problems in the Forecasting and
Optimal Development of a Nuclear Power System for CEMA MPmber-Nations"]
f,TQxt] One of the important directions in solving the fuel-
_ and-power problem of CEMA member nations is the accelerated devel-
opment of th~ nuclear power industry. The present stage of the
cooperation among these countries in this area is characterized

w by the large-scale and comprehensive ~:nplementation of mQasures,
th;: considerable influence of scientific and technical progress
on production efficiency and the great contribution of these
countries to the overal? issue of realizing the nuclear power-
production portion of the long-term specif ic program of coopera-
- tion (DTsPS) to provide for the economically founded requirements
of the CEMA member-nations f or the basic types of power, f uel
and raw materials to the year 2000. The indicated conditions ~
' for t',he development of the nuclear power industry are dictated ~
by th~~ necessity of improving methods of contro].lin~ and planning
the development of the entire nuclear power complex (YaEK) of
~ the socialist countries (1,2].
Figure 1 presents the interindustry structure of the nuclear power-production com-
plex. It includes the fuel-cycle industries (extraction., processing and enrichment
of fuels, the manufacture of fuel elements, chemical processing of spent nuclear
fuel and processing and storage of ra3ioactive wastes), nuclear stations and in-
stallations for the generati~~i of electric and thermal power (nuclear electric-
power stations, nuclear heat-and-power stations and nuclear power-production instal-
lations) and all capital-generating sectors of the nuclear power complex (metallurgy,
heavy, chemical and power-engineering induetries, the electrical equipment induatry,
instrument engineering, power industry consttuction and Che construction i*_:~lustries) .
Of course, t~e nuclear power complex of a given country can and must be considered
- a subsystem of the integrated nuclear power complex of the CEMA member-nations.
zt is possib~e to rationally solve uroblems which wiChin the framework of a single
country cannot be solved at all or, at ieast, cannot be solved eff iciently. This
1
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N~UK U!'MII:IAL U,r. VI~1LY
can be accomplished througi? international cooperation, di~isio~ ot la,bor, special-
ization of production and construcCion through combined effort or cooperation with-
in the scope of capital investment in construction.
The indus*ries in the nuclear power complex are embrac ed by complex interrelation-
ships, inc].uding reciprocal relationships (for example, the chemical processing of
spent fuel) . The formation of this aggregate of industries into a unified nuclear
power complex is di~tated by the degree of specialization, the influence they exert
on one another as well as by the averall production cycle.
The characteristic integrity of the nuclear power complex which we have pointed out
determines the nec~ssity for applying a systematic approach to examining the pros-
- pects f or its dev~lopment and t!~e introduction of inethods for multisector plann in g
and con trol.
The overall orQ~nizational structure for planning the long-term development of the
nuclear power complex can be presented in the followin g manner (fig. 2). This dia-
gram displays the operational makeup and the interrelationships of these operations
in the forecasting and formation of nuclear power programs and a comprehensive long-
range p lan of. development for those industries included in the nuclear power c omplex.
The close interrelationship between the planning of sc ientif ic research and industry
is shown by a block for the determination of the needs of industry in the develop-
ment of equipment and technology ar~d effective directions f or the utilization of
the new results from scientific research and experimen tal design work. A positive
feature of this long-term planning program is the comb ination of the processes f or
forecasting and forming programs and plans for the development of the nuclear power
complex into a unified cycle of interrelated operations. The results from forecast-
ing programs and the coordinating of these programs with specialized plans should
insurP the best continuity. ~
At the present time, we have good reason to believe that the CEMA member-nations
- will in sure the realization of a comprehensive and specific approach to managing
the development of the nuclear power industry up to the year 1990. Within the scope
of the previously mentioned long-term specific program of cooperation, we have con-
cluded the Agreement on Multilateral International Specialization and Cooperation
in the Production and Mutual Supply of AES EquipmPn~ f or the Period 1981-1990. We
are creating the necessary organizational prerequisites and are solving three of
- the most important problems in scientific and technical cooperation within the nu-
clear power industry: the mastery of water-cooled water-modulated reac~ors with
outputs on the order of 1,000 MW; the development of h igh-output fast reactors;
and the development of nuclear heat and power stations and nuclear heat plants.
' However, features of the process for the integrated deve"lopment of a nur:lear power
complex--a considerable rate of growth, a broad spectrum of scier.tif ic research
and f easible structural transformations, the time lag and the capital-intensive
nature of the problem--dictate that technical and economi~ studies of th~ prospects
- for a CI:MA member-nation nuclear power complex be r.!snducte~ for the years beyond
1990 and that corresponding long-term cycles for managing its development and pro-
gress be organized.
2
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Control over the development and progress of the nuclear power complex and the real-
ization of the corre,aponding industrial-economic and scientific-technical activities
are contained in the iield of activity of a number of agencies and orgenizations,
primary among which are:
a) at the national level:
- the highast organs (government and ~oint agencies for planning, f inance, technical
development, foreign trade, etc.)
- special ministries (ministries of power, metallurgy, machine construction, instru-
ment engineering, construction and the construction industry, etc.)
- the organization of a scientific-research and experimental-design base;
b) at the CEMA level:
- supreme organs (Supreme Soviet, Executive Committee),
-co~ittees for cooperation in the area of p.'~anned activity and for scientific and
technical cooperation, as well as their working agencies (namely, the Working Group
- for the Fuel and Power Balance and the Council for Scientif ic and Techn:Ical Coopera-
tion in the Area of Fuel and Power Problems--the TEP Council),~
- special Permanent Committees (namely, those for the utilization of nuclear power
for peaceful purposes, for electric power, for machine building and for construction)
and their working agencies,
- international economic organizations (namely, international economic organiza-
tions of Interatomenergo and Interatominstrument),
- international organizations of the scientific research base (namely, MNIIPU--
the International Scientific Research Institute for Control Problems).
Within the oLgan izational plan, integrated programmatic-specific planning and man-
agement of the CEMA member-nation nuclear power complex according to diagram 2 should
be constructed on one interrelated plan of harmonious work coordinated by the above-
mentioned agencies and organizations aC the national and international levels.
Let us now dwell upon the status of the work being done in the area of forecasting
the long-range development of the nuclear power industry in socialiat countries.
We refer here to~the work being carried out within the s~ope of activity of the
Co:nmittee of the Scientif ic and Technical . Council which, on behalf of the 24th meet ing
of the CEMA Session, organizes the development of a scientific and technical forecast
for the soluti~n to fuel and power problems in Che period to the year 2000 and for
the long t.erm. This work, being carried out from 1980 to 1984, is based upon data
from inter::~ted CEM'1 member-nations regarding the planned development of natiional
fuel and power c mnplexes.
The first meeting of the Provisional Collective of Scientists and Specialists (VKUS),
_ created by the TEP Co�lncil to develop a f orecast foY the development of the nuclear
power industry, took place from 27 to 30 January 1981 in the ci~y of Ostrava in
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tl~e Czechoslovak Socialist Republic. The meeting examined materials prepared
jointly by the Czechoslovak and Soviet sides with the part~cipation of a department
of the Secretariat's Scientific and Technical Council and M~IIIPU. T~ao ma~or docu-
ments were submitted for approval:
1. ~ program of development and
2. organizational and methodological statutes regarding the development of a fore-
cast for the growth of the nuclear power industry in ~EMA member-nations in the
- period to the year 2000 and for the long term.
The sixth session of the TEP Council which t~ok place in Berlt.n, East Germany f�tom
16 through 19 June 1981 approved both documents and noted that the Soviet Union had
assumed the duties of the coordinating nat:ion.
This program of development and the organizatio~nal-methodological statutes determine
the makeup and the inter3ependence of work being done at the national and CEMA 1ev-
els. Work at the CEMA level is oriented toward correlating and systematically com-
bining national forecasts and forming the best stra.te~y for the intensive develop-
ment of the nuclear power compl~x af CEMA member-nations together with the drawing-
up of recommendations for the individual countries. Forecast studies at the national
level are oriented toward the formation of an optimum :long-term strategy for the
� country which is based on its specific situation and which insures the participati.on
of national organizations in i~ternational cooperation for the realization of inte-
grating measures in the area of nuclear power generation and related areas of the
~ nuclear power complex. Studies carried out at the national level are the chief
and initial basis for forecasting. Two ma~or stages have been proposed: the develop-
ment of a forecast plan and, naturally, the f orecast itself. During the f irst stage,
alternative strategies for the development of the power industry itself are examined
and evaluated. During the second stage, these strategies are refined and ad~usted.
In addition, suggestions are developed regarding integrating measures and the con-
tribution of each country toward their realization.
At first, the cauntries develop a structure f or the development of the national
nuclear power industry and carry out an initial evaluation of the long-term require-
ments for external deliveries of the basic types of equipment and scarce resources
and the feasibility of making their o~m deliveries to the other countries. On the
basis of these data presented in CEMA, the Provisional Collective of Scientists and
Specialists work out alternatives for an interindustry structure and scenarios for
the development of the nuclear power complex so as to satisfy ~he countries' re-
quirements for the deveZopment of the national nuclear power industries to the max-
imum degree possible. As a result, the basic directions��or the development of
the nuclear power complex are determined. They expand and deepen the sphere of
_ scientific-technical and industrial~economic cooperation among the countries and
contribute to the achievement of their national goals.
At the next stage, the structure for the devel~pment of the country's nuclear power
industry is refined, with consideration being given to the basic directions for
the development of the CEMA member-nations' nuclear power complex on the whole.
Specif ic alternatives are being deveXoped for the realizaCion of long-term goals
- for the growth of the national nuclear power base.
4
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At the same time, on the interr.ational level, alternatives are worked out for the
_ realization of integrating measures f or CEMA member-natione as w~ll as mutually ad-
vantageous alternatives for scien~ific-technical and industrial-economic cooperation.
- At the concluding stage of forecast development, the specif ic structure is submitted
for approval and a plan of the decisions is prepared for examination by the corres-
- ponding CEMA agencies with regard to integrating measures which insure the long-term
growth of the CEMA member-nation nuclear power complex.
Let us dwell in more detail upon the development of a forecast plan for the growth
of the national nuclear power industry. The development of a forecast begins with
the determination of the scope of the national nuclear power industry fur the years
1990-2000-2010 and the selection of types, unit outputs and the stages of utiliza-
tion of nuclear reactors. Then studies are developed for determining the basic
tasks of scientific research, alternatives for the development of an industrial
and constxuction base and ttie demand for scarce materials and the basic types of
J equipment. Simultaneously, as a result of the determination of tt!e structure of
AES's, ATETs's and AST's planned for construction in the period to 2010 and the
evaluation af the prospects for the utilization of nuc:lear power installations in
metallurgy, chemistry, agriculture, transportation and in other sectors of the
economy, we are conducting:
- an analysis of ecological factors and the identification of t~e corresponding
limitations on and requirements for the construction of nuclear power installations;
- the development af alternative meth~ds for maintaining, storing and transporting
- spent fuel and for processing and burying radioactive wastes;
- the development of alternative methods for decommissioning nuclear power installa-
- tions.
Naturally, these problems in the development of the nuclear power industry give
rise to their own scientif ic-technical and industriai-economic questions. A study
" of the alternative methods for developing production and scientific research in
individual sectors of the nuclear power complex concludes~with a system f or evaluat-
ing alternative methods for the development of the power industry. If it becomes
necessary, the previously selected structure and alternative methods for the utili-
zation of nuclear reactors can be adjusted according to the results of this evalua-
tion in order to improve the effectiveness of the solutions obtained. Af ter satis-
- factory results are obtained, a selection is made of the ma~or directions for the
scientific and industrial work being done to insure the development of the national
nuclear power industry. The country then draws up proposals for international coop-
eration in the given area. ~
It must be noted that the initial selection of the structure and ~lternative methods
of utilization of nuclear reactors is a most critical stage in the development of
a national forecast. This selection essentially determines future scientific and
industrial activity both within the scope of the national nuclear power program
as well as international programs for the integrated development of the nuclear
power complex. Thus, we msst develop this stage with particular care, enlisting
- the aid of leading scientists from various sectors of the nuclear power industry.
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In order to carry out systematic research and optimize strategies for the long-term
development of the CEMA member-nations' nuclear power complex, MNIIPU together with
the interested national organizations is formulating a system of mathematical eco-
nomic models. This system includes models f or the development of f uel-cycle sectors,
a model for the development of the power-production base, a model f or the develop-
fient of the machine-construction base and the related nuclear-power production in-
dustries as well as a model of the interrelations among sectors of the nuclear power
industry. This sy stem of models reflects the multistage process of production (for
example, the extraction, enrichment, processing, etc. of ores) and considers the
- basic production parameters which determine its tQChnological feasibility and its
technical and economic efficiency. Moreover, the models directly reflect the re-
sults of the development of new equipment and techniques and their introduction into
industry.
The realization of tasks regard~ng research and the optimization of long-term strat-
egies for the development of a nuclear power complex utilizing the indicated models
= is accomplished on the basis of the SOPOT [expansion not provided] man-machine sys-
tem developed at the International Scientif ic Research Institute f or Control Prob-
lems. This system is problem-oriented toward the solution af tasks involving simu-
lation and the optimization of intersector industrial complexes [3]. The SOPOT
system consists of: a data bank; a generstor of model programs; a numerical-method
library; a model-research and problem-solving block; and an interactive-procedure
block.
The features of the SOPOT system provide:
- the capac ity f or system correlation and problem decomposition in multilevel
organizational systems;
- facility in adjusting marhematical economic models and problem conditions;
- the solution to complex data problems, including the answers to nonoperational
queries.
At the present time, MNIIPU is organizing the exchange of SOPOT system software
with interested national organizations. The planned exchange wi11 enable the coun-
tries to reduce considerably the time and resources spent on creating a system for
' simulating the growth of national fuel-and-power complexes, as well as to prepare
a basis for combin ing national systems of power-industry models in order to conduct
~oint investigations of alternative strategies for integrated development of fuel--
and-pr~wer complexes.
r
In conclusion, we must emphasize: ,
1. the importance of data formulated at the national level and, thus~ the complex-
, ity of the makeup of national working groups. It is desirable that all interested
_ agencies and organizations in the scientif ic-technical and economic spheres of each
country participate within the scope of these working groups. The participation of
administrative and economic agencies (the State Planning Committee, the ministries,
etc.) is necessa?}? in view of the fact that these agencies are the ones which will
subsequently utilize the results of the f orecast and can from the very outset intro-
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Figure 1. Diagram of Interindustry Structure
� of the Nuclear Power Complex '
Key:
B. Radioactive discharges and waste
1. Revenue-generating sectors of the nuclear power complex: metallurgy, heavy,
power, chemical and electrical machine construction, instrument engineering,
the construction industry, etc.
2. Equipment for extracting and preparing ore
3. Reactor units
4. Thermonuclear equipment
S. Chemical equipment
6. Electrical equipment
7. Instrumentation and computers
8. Hoisting and transporr.ing equipment e
9. Construction and prefabricated materials
10. Natural reserves of uranium
11. Natural reserves of thorium
12. Extraction and preparation of uranium ore
13. Extraction and preparation of thorium ore
14 . Uran ium enr ichmen t
15. Enriched uranium
16. Depleted uranium
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Figure 1. Key (con'd.)
17. Natural uranium
18. Production of fuel elements
- 19. Burial of radioactive wastes
20. Processing of spent f uel elements
21. Fresh fuel elements
22. Spent fuel elements
23. Radioact~ive wastes
24. AES's, ATETs's and AST's with thE:rmal reactors
25. AES's, ATETs's and nuclear power installations with fast rea~tors
26. Nuclear power installations wiCh ltigh-temperature reactors
27. Electric power
28. High-potential heat
29. Low-potential heat
30. Thermal wastes
31. Non-power materials
duce the necessary corrections in the studies being conducted and orient these stu-
dies in accordance with their own requirements. The immediate users will be the
organizations in the scientif ic-technical sphere which will introduce the desired
scientif ic approach and methodological apparatus into forecast studies;
2. the important role of ineetings of the Provisional Collective�of Scientists and
Specialists which will observe and evaluate the progress of the work, insure t~+e
exchange of information and experience as well as consolidate individual points
of view and discuss all data p~esented by the member-nations and the coordinator;
3. the extraordinarily vital and complex role of the coordinator. In contrast to
the Provisional Collective of Scientists and Specialists, which can operate only
periodically during tiie course of its Meetings and will be limited to the discussion
of the materials presented, the coordinator will continuously carry out the follow-
ing activities on an international scale: ~
- prepare proposals for refining methodological instructions and working programs,
- develop a conceptual plan f or an integrated nuclear power compl~x (at the f irst
stage of development),
- evaluate materials presented by the countries regarding national nuclear power
complexes and develop plans of the corresponding recommendations,
- evaluate the countries' considerations regarding an integrated nuclear power com-
plex, summarize them and, finally, ~
- complete editing of the forecast plan and the final forecast for the int~grated
nuclear power complex. ~
Naturally, in order to successf u11y implement such broad and multilateral activity,
the coordinator must make extenstve use of its own scientific-technical potential
and accumulated expertise in the area of organization of forecasC work.
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~ Itponto~ aaxcr_a I ITOts.aoKo-cYp~- ,zv~r-rA F- ~~70R7101~ RO tC~tO.T:~ ~~Uf~ /
,eloO duY. I R.K ~ czs i~t~- I~e~a~om Mo-, cocrue (
~ I e~piewA eaa~ ! ~ ~a~~MOm wu~- ~ ~~nQ ( eoC rewc- : fi3K
l I ~ ~ ~ROC'p:!!01lI I I I ~ ~ 'CSJORS ~ :R~ M ~
~ ~ ~ ~ z tp.; ' iaaato:~ ;,..xno:or I I ~i8~
, I~
r,por,~o~ aa~,ar~u ~ ~ 13 ) L `T--~ ~ ' ' I
,~~~o-~~,~.~a�.~~ i~ i.~,..~~a , ~ i I I(16 ) ~ ~(17) I 1
9 c co.. oan.ete ` -
CO]~i1Ci1 ~tOfUC RS"'~YC" 7'~~''.,'.~:Jt /IS~JM~' ~ ~ � f
~ 7N!'il:if~l:lt7! ~ ~ � ~ ~
i 7:I._...~5~
I sgarcce a--r.�~:.:: �.eoz. ~r.~- , i ~ r
y;::;z:;,..,,:, i ?a~pao~-va.:LlaN:r'''9~K~
a z ca-:acoar.F oi:aaa-:n ^ ! :ep.oipoW~r.a ~
I L�]t~�"C: i].u 1 t~:i:M! fI1pV7t0-::CY.n!:RG.. J ~ 0.t':LletfLSl ~
~ ~ � 1:11~^..4NOC~N C~Y6't:f 7 ~ 1
i eac-us A~K I ~ or~ai!1 i
, ~ ,
rtporNO~ :s~:z:u - ;sr ~ ~19~ ~ ~ ~2~~ ,
~ 10~ we~extit .~oN~aotei~suxa:.~ j ~
ea~~ A3K , ~ 1 ~ , ~
i i
I rio7Dexcw;oa~a ~:~~Fy~ ~
Axa.m rei~e~eaEi pa~- ' "
~ 11 ~ ~Rrx~ A3'F. ~ a;; ::::,s ~ J ~ 21 ~ Orpuane~ xa D'~fi~+
~ ~ (22)
~t7D~00~R~ ~MT'~~9~~ ft :~:MfIpOl811M! IIOIff~C OWIYX Pa~naOO?lt~ RQpCRtNt111NYx :UI~Nq~ N~tll?fUt
~12~ L~-n~;e^eK'~ue.^ ;av�r.:z- ---.~rt~c'-}uw yut~or~ ~:I~ ~ crn~~c~eN r eoer.ys fiJn
(23) (24) .
Figure 2. Overall Structure of a System for the Long-Range
~ Developmental Planning of a Nuclear Power Complex
Key: ~
1. Joint program of scientific and technical progress
2. Goals o~ the country's social and economic .development
3. Basic directions for the country's economic and social development
4. Limitations on resources
5. Program adjustment
6. Formulation, balancing and optimization of alternative long-range plans
for the development of industrial production within the nuclear power
complex
7. Forecast of economic requirements for the outputsof the nuclear power
complex
8. Forecast for the development of the raw materials base ~
9. Forecast for the development of acientific and technical progress in the
creation of new nuclear power equipment and technology
10. Forecast for the development and situation of the nuclear power complex's
production base
11. Analysis of foreign trends in the development of the nuclear power complex
12. Formulation of forecasts and goals in the long-range developmenC of the
nuclear power complex
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Figure 2. Key (con' d. )
13. Integrated forecast for the development of the nuclear power industry
and its raw-material, scientific-technical and industrial base
14. Determination of goals according to the basic directions for the develop-
ment of the nuclear power complex (the creation of AES'.s, development of
the f uel and raw-material base, nuclear equipment construction, etc.)
15. Formation of long-term scientific-technical and industrial programs for
the development of the nuclear power complex
16. Industry requirements for the development of equipmerrt~and technology
17. Proposals f or the utilization of new equipment and technology
18. Formulation of an integrated long-range plan for development of industries
within the nuclear power complex
19. Formulation, balancing and optimization of alternative plans for scien-
_ tific-technical activity in industries within the nuclear power complex
20. Formulation of five-year plans for industry
~ 21. Program adjustment
22. Limitations on resources
23. Formulation of long-term programs for the development of the nuclear power
complex
24. Formulation of long-range plans for the development of industries within
the nuclear power complex
BIBLIOGRAPHY ~
1. Dragny, M., and Svetlik, Ya., "The ~Juclear Power Complex and the Main Directions
' of Scientific and Technical Progress," ATOMNAYA ENERGIYA, Vol 49, 1980.
2. Dragny, M., "The Integrated Nuclear Power Complex of the CEMA Member-Nations,"
JADERNA ENERGIE, No 26, 1980. �
- 3. Vlasov, Ye. P., ~urin, Yu. A., "The Study of Methodological Questions and the
SOPOT Man-Machine System for the Forecastin g and Optimal Planning of the Devel-
opment of the CEMA Member-Nation Fuel and Power Complex (Based on the Example
of the Nuclear Power Complex)," International Scientific Coordinating Conference
on the Problem of Improving the Control System for the Development of the CEMA
Member-Nation Fuel and Power Complex, Moscow, 1-3 December 1980.
COPYRIGHT: SNTL, n.p. 1981
9512
cso: 51oo/3oi2
. ~
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CZECHOSLOVAKIA
REACTOR SIMULATION PROGRAM RESULTS SUNINIARIZED
Prague JADERNA ENERGIE in Slovak No 12, 1981 pp 428-433
~ [Article by Ivan Kinka, Energopro~ekt, Prague; Rostislav Pernica, InStitute
~ of Nuclear Research, Rez; Jozef Misak, Research Institute of Nuclear Power
Stations, Jaslovske Bohunice; and Bedrich Hermansky, Faculty of Nuclear
Science and Physical Engineering, Prague: "Evaluation of Results of Test
Runs of the REPAID, WER-D, SICHTA and DYN79 Programs: Part II. Summary of
Results"]
[Text] The first part of this article gave a concise
description of the programs tested (REPAID, WER-D,
SICHTA and DYN79) and discussed the formulation of the
simulation problems and the input data. This second
part is devoted to the computation results and evalua-
tion of the individual programs. The most important
parameters that were monitored are presented in figures
and tables.
5. Results of Individual Simulation Runs
In addition to the results presented in this article, a considerab],y greater
amount of information can be obtained from references 5-11, which report the
capabilities of the individual programs on the basis. of practical results.
The relative deviations given in the tables are always calculated reZative
to the 'lowest value obtained.
5.1. Uncontrolled Removal of Control Rod Groups at Zero Reactor Power
The course of the process as determined by the individual programs is sum-
marized by the selected parameters given in Figs. la-lc. Because the results
were obtained taking the programs in pairs, i.e., REPAID with WER-D and
SICHTA with DYN79, it was quite apparent that the reason for the considerable
differences in the results for these two pairs was the fuel-reactivity
coefficients, which were temperature-dependent in the first pair and constant
in the second. In spite of this evident fact, we conducted an extra control
computation with the REPAID program using a constant fuel-reactivity coeffi-
cient of -2.17 � 10-5 deg'1, which in absolute terms i3 somewhat lower than
that used in the SICHTA and DYN79 programs. The curve of power as a function
11
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~
of time for this case, shown in Fig. 1, also indicates the effect of reac-
tivity coefficients on the process.
at -
I ' ;
_ OCO~ ~D
Q6 t~[Nii ~ I
t CYN 7D
~arrc+~r Wao~+rr ara ro
a 4~ ,r~d~+~~.~~~rvo+~~ a Ke � a. Control co utation with
~ r,~. ~o K ~ ~ Y� ~
47~ I REPAID with temperature
o - coefficient of reactivity
equal to -2.17 � 10-5 deg'1
�i i
~ ~
.
I~? ~ ~ _
. ~ . .
~a .
.
qe ~ ,
aq6~ ,l~ l ~ i
/
- ,
4~
9?
� o x~ ~o ~o a ~an. 7n
~
Fig. la. Uncontrolled removal of control rod group, zero reactor power:
reactivity and relative power as functions of time (PNp~-nominal
power).
central channel
~tau ~+^~^r,~^o~
~ ~ i
1000 - q!a !0
~e- vK4�D
$~A
81J0
~ ~ .
y, I
fQ0
70d
?IqD ~
hot
'~'0 channel
~
~ ~�a' `
~ao ~
~
o ~o m a ron nn
Fig. lb. Uncontrolled removal of control rod group, zero reactor power:
maximum fuel temperature in central and hot channels as funcCion
of time.
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~ao
central
~ ie'''''`"�~ channel
~
-M- -o �
~ ?l0 + ,
~ '
.1'1~
~q0
~0 1'10 t
~ "a^~�~"a'~ channel
~
~
w
mo
xo
r.n
aav
- o ~n ~o ao o m ~av .
~
Fig. lc. Uncontrolled removal of control rod group, zero reactor power:
maximum temperature of inneL surface of cladding for hot and
central channels as a function of time~.
It is worth noting ~gain that the interpretation of "power" is slightly .
different in the SICHTA program than in the others: in the SICHTA program
the "thermal power" figure takes account of the quantity of heat liberated
by fission products, while the neutron-flux density obtained from this pro-
gram is approximately in agreement with the interpretation of "power" in the
other programs .
Table 1 surveys the extreme parameters obtained with all four programs. For~
completeness we also give the largest absolute differences obtained in our
results, including the case in which the reactivity coefficient was altered
[REPAID program]:
--relative power, 0.185;
--maximum total reactivity, 0.004%;
--maximum fuel temperature, 396�C;
--maximum temperature of inner surface of cladding, 14.7�C;
--outlet temperature of coolant, 12�C.
13 .
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Table 1. Comparison of extreme valuea ~btained for +incontrolled removal of
control rod group at zero initial reactur power. ST--central
channel; HOR--hot channel. Meaning of numerical values: parameter
value, time at which attained (sec), relative deviation (X).
a b a a I.;
x ~ z
Parametcr ~ ~ D ~
~i a D ~ m A
vykon reaktora, ~
PlpvoM 1,06 1,06 1,238 1,20
(vztiahnuty - 133 140 160 120
k nomin~lnemu) 0 0 17,8 14,3
msx. hodnota d 0,817 0,813 0,818 0,815
celkovej reekti� - 46,8 48,7 48,8 46,7
vity Oq 0,8b 0 0,33 0,33
msximRlns tep- e 1020 1041 1281 127b
lotA palivs [�CJ ST 133 140 160 130
q+~ 0 2,08 26,9 2~,0
1809 1636 200b 1980
HOR 133 140 150 130
0 1,7 24,8 23,0
. maximgins tep- f 280,7 284,2 290 282
lota vnutorn~ho ST 133 140 150 130
povrchu polcry� Q 1,2 3,3 0,5
tia T~ [�C] 331,3 339 348 a48
HOR 133 140 180 130 �
- 0 2,3 4,4 S,4
msxim~tm vy� g 239 239,1 248 ~38
etupnejteploty 8T 133 140 1S0 13U
ohladiva [�C] 1,3 1,3 4,2 0
283,b 283,8 273 �l81
IiOR 133 140 1b0 130
0,98 1,07 4,8 0
xey :
a. Parameter e. Maximum fuel temperature Tf (�C)
b. Channel f. Maximum temperature T~ of inner
c. Reactor power P/PNOM ~relative surface of cladding (�C)
to nominal power) g. Maximum exit temperature of
d. Maximum total reactivity Op coolant (�C)
5.2. Ejection of Control Rod at Nominal Reactor Power
The graph of the parameters in question over time is presented in Figs. 2a-2d,
and the extreme values obtained are given in Table 2. At the power peak,
which in this case is produced almost exclusively by the kinetic equations,
there are only extremely small discrepancies: the .lower value from the
SICHTA program results from the interpretation of "thermal power" already
described. If we take the neutron-flux denaity in the SICHTA program as the
power, we obtain a peak value of 2.07. The greatest discrepancies are in
temperature, particularly on the inner surface of the cladding, where the
heat-transfer equations and the thermophyeical parameters of the cladding
which were used may have been responsible.
14
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- 4<
- I � _,v[M !0
~a'
' ~DYN7D l~ _
4
� 4~ ~ ~ a1i�~'~'�' central
~ ~ ~ ~zio
, channel
QI I
c ttoo }a
"`~m
z?
ran
zo "
~ A""'~ hot
?,e 2aao
� channel
~ ~ ~ ~aav
a u
~
o ~~floD
/,7 11oD
i o ~ ~ ~ ~f7 s 6. ~ ' ~ao�. : ~ ~!J � r ~
Fig. 2a. E~ection of control rod at Fig. 2b. E~ection of control rod
nominal reactor power: reactivity at nominal reactor power: maximum
and relative power as functions of fuel temperature in central and hot
time. channels as function of time.
~ro
~o central
~ _~~~o ~ channel z6 hot
tso
~ .a~,~ ~r ~,~;;o channel
~o -+~ar~rm . naridhnil
z~ 8
^a orti'70 i
yJ0 .
zo
HO Aonia Aorol .t2 .
l00 tlOt ~d '
~ channel
~ f,6
~ {i
~~0 f 1 J 6 5 6 7 ~10 I 1 ,J I 5 6 7
~ 1=1 ~ I~I
Fig. 2c. E~ection of control rod at Fig. 2d. E~ection of control rod
nominal reactor power: maximum at nominal reactor power: critical
temperature on inner surface of temperature ratio in hot channel
cladding of central and hot channels as function of time.
as a function of time.
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Table 2. Comparison of extreme values achieved in e~ection of control rod at
rated initial reactor power. ST--central channel; HOR--hot chan-
nel. Meaning of numerical values: parameter value, t3me at which
attained (sec)y relative deviation (Y).
b~ Q ~ � F' ti
. .
1 ~~ruu?ctcr c I ' J ^
,
j I w i ~
~-~�kon re~il:tura, ~I ~
1~ij'~u~1 �.IIW :.~,OJ_' I 1.!)i t ~ ~'.11;
~�ztiahimt ` - U,'3 I i?,l:t I 1~.1:{ 1~,1:1
( J
1: numinrilncmnl li.li ~ li.l ' - --5.u
~nnximRlnn 1~~d� I �
notn cclko~�cj U,37 0,37 U,:i(iG ~1,37
rcakti~�it~- d~~ - 0,1 U,1 ll,l p,l
~oo~ 1,4 1,4 U 1,4
inaYimbl~ia 1;i20 1370 1384 1380
teplota Pali~�n S'1' 7 7 2b 7
~o~~~ ;;,g 4,8 3,4
187U 2UlU 3U78 2UdU
HUli 7 7 11,2 7
I U 2,0 L,b (f,l
mn~iiiufl~in t~~p. f ~iUl ;~HF1 ~inll.'2 ~3U2
~ntii ~�ntit~n�n~~l~~~ N'1' U.i:r 11.04 .':i i
~~~~~�ruhii ~mkt�~�. U,ii :.~.4 __.11 _ II.K
t ui ; 1 :tU+; 4uU :its~,u
-I NUIt U,75 (~,lli 7 -
:l.7 S.li 11
iiuixitnum - . i111 '311 ;il I 3llU
Fti~pnrj t~�~~I~~t~� S7' ~5,9 4,4Fi 9 7
clili~di~~u ~`C] U,:i 0,(f Il.li_ 0
_ ;3'lU,;! 32(f,:i ;i_'li.~i 320,:i
HUIi U,4:4 0,22 0,9 1
U 0 U 0
miiiimdlna hcu~�
li~~tN I:ritickc~li~~ 1.54 l,,il 1,:i78 :.48
tc~pi~lni+l~u pome� HUlt 1?,43 11,2fi 111,G u
~�t? kkr 17,(i (1 ~i.2 RU,3
Key:
a. Parameter
b. Channel
c. Reactor power P/PNOM ~relative to nominal power)
_ d. Maximum total reactivity ~p (~6)
e. Maximum fuel temperature Tf (�C)
f. Maximum temperature T~ of inner surface of cladding (�C)
g. Maximum exit temperature of coolant (�C)
h. Minimum critical temperature ratio kkr
In the case of the critical temperature ratio no bene�it is gained from com-
paring DYN79 with the other programs, because the equations used differ
considerably (Bezrukov equations va Osmachkin equations). In addition, in
this situation the WER-D program's more precise description of the hydro-
dynamic ratios in the channels has a pronounced effect on the results.
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In this case, we obtained the following largest absolute differences in the
values of the parameters monitored:
--relative power, 0.131 (owing to interpretation of power in SICHTA program);
--maximum total reactivity, 0.005y;
--maximum fuel temperature, 120�C;
--maximum temperature of inner surface of cladding, 21.5�C;
--exit temperature of coolant, 2�C (apparently as a result of boiling in the
hot channel);
--critical temperature ratio, 0.23 (Osmachkin eyuations only).
5.3. Leakage of Cold Water Into Core
The qualitative difference between this process and those described above is
that the output temperature of the ~oolant and the mass flow change over time
and the change in reactivity is produced by the reactivity coefficient of the
coolant.
The dispersion of the results obtained, particularly between SICHTA and the
REPAID and WER-D pair, is partly explained by the fuel-reactivity coeffi-
cients and p artly by the heat-transfer coefficient of the crack. The most
notable differences between the DYN79 program and the others--since they show
- up even at time T= 0--are apparently produced by the heat-transfer coeffi-
cient of the crack. The power peak in the SICHTA program, with a neutron-
flux density of 1.375, is considerably lower than that obtained from the
VVER-D and REPAID programs .
rxn ,
i central
t~Gn y,~d;yFona~
channel
ao~ :~o
i
4~~ ~ -r-YY
R 0 l~
-~SKN/A � ~p
~A? I tOrMID ~,/1~ tr~!/
0
~ ~ t3G/7I .
V IA~ ~O~A
o. 0
v
' l1[0 "
-4~
teaD
-onc "�'~�'"�r ho t
' t,w~
~6 channel
' ~rao .
S~ ,
v; ~~~oo
o f1 ~a.
t1~
4~ -
tIQD
~0 10 s1D JiD ~0 JO 00 7G
~~f, 10000 /0 7b 70 10 SO QO 70
~ [s)
Fig. 3a. Leakage of cold water into Fig. 3b. Leakage of cold water into
core: reactivity and relative power core: maximum fuel temperature in
as functions of time. central and hot channels as function
_ of time.
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~ro
_,~a p central
J6o '`"�"''`�`d t,'
M�;� channel �on;aw,no~ hot
^ '~p"'~ ~1 . channe 1
- ~~j. ~ I - NCOA 10
~ -x- w[A-0
710 I t JI[MA ~ . /
,~p 29
~ Artia:~b+o/ ~10t ~
26 ~ -
, channel ~
V jb
w ~
7~0
'~0 ro 10 aD b SO 10 ln Z~0 !0 ]iD JO 10 SO 60 70
t(a~ { ~s~
Fig. 3c. Leakage of cold water into Fig. 3d. Leakage of cold water into
core: maximum temperature on inner core: critical temperature ratio
surface of cladding of central and in hot channel as function of time.
hot channels as function of time.
Table 3. Comparisons of extreme values obtained for leakage of cold water
into core at 5/6 of nominal power at beginning of process: ST--
central channel; HOR--hot cFlannel. Meaning of numerical values:
parameter value, time at which attained (sec), relative deviation
a b A Q ~
Ke
psrameter Ve a W y~
~ w~ D ~ Q a. Parameter
b. Channel
- vy~kon resktors c. Reactor power P/PNOM ~rela-
PlPxou 1,48 1,406 1,33b 1,21
(vztiehnut~ 38,2 38,9 37,9 38 tive to nominal power )
k nomiaLtnemu) 22,3 18,1 10,3 o d. Ma.ximum total reactivity
maximLlns 0,067 0,061 0,048 0,043 ~p
hodnota celkovej - 30,7 37,4 38,4 37,6 e. M3ximum fuel temperature
resktivity dp 32,8 18,8 14,0 0 0
maxim~lna tep� . 2129 2141 2146 22b6 Tf ~
lotn palivs T~ ST 40,7 43,4 39,8 40 f. Maximum temperature T~ of
~~C~ e o o,ae o,7a 6,s inner surface of cladding
2390 2398 2612 2760 (�C)
HOR 40,2 41,9 39,7 28 Maximum exit tem erature
0 0,33 b,l 16,1 P
maxim~lna tep� 368 368 841,3 33b,b of COOlant (�C)
- lota vnutorn6ho ST 38,2 38,9 39,9 0 h. Minimum CritiCSl tempera-
povrchu o,t s,~ 1,~ o ture ratio kkr
pokrytia T, [�C] 388 388 372,8
f HOR 38,2 38,9 39,2 -
3,6 3,6 0
msximum v~� 301 301 302 299
etupnej teploty ST Q 0 0 0
chlndivo [�C] - - - -
317 318 313 311
g Hn]Z 38.~ 38.9 0 0
1,9:i 2,3 U,6 U
minim~ilnn hod� `
uut~~ ki�~tickul?u 2,Uu 2,U4 2,38 b,~5
t.cpoln~ho pn� h HUR 38,2 38,9 38,9 42,b
~i~~ri~ ,~:A., 0,49 0 1b,7 137,4
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The change of parameters over time is shown in Figs. 3a-3d, and the extreme
values obtained are presented in Table 3. On the basis ~f the results
obtained, in this type of situation we may expect the following dispersion
_ of the extreme values of the parameters monitored for the different programs
- (including the effect of the input data, namely the coefficient of reactivity
and heat-transfer coefficient of the crack):
--relative power, 0.27;
- --maximum toral reactivity, 0.014%;
- --maximum fuel temperature, 360�C;
- --maximum temperature of inner surface of cladding, 22.5�C;
--exit temperature of coolant, 7�C.
5.4. Uncontrolled Removal of Group of Control Rods at Nominal Reactor Power
The last of the test runs is the classical emergency situation, in which
limiting values may be approached (or exceeded), even in the central channel.
Accordingly, it is in this case that physical models of the individual pro-
grams have the most evident effect. This case is, moreover, the most suitable
one for comparing the REPAlD, V`IER-D and SICHTA progratas, because a
temperature-dependent fuel-reactivity coefficient was also used here for the
SICHTA program. The results are most readily apparent from the power curves
of Fig. 4, which are virtually identical for almost the entire simulation
period of the programs. The constant fuel-reactivity coefficient value again
had an evident effect in the 1~YN79 program. The abrupt increase in the
temperature of the inner surfa;.e of the cladding of the hot channel given by
the REPAID program results from the transition to film boiling, while the
VVER-D program uses equations for nucleate boiling even after the critical
neutron-flux density is exceeded. The temperature discrepancies are small on
the whole and mostly reflect differences in the thermophysical parameters and
heat transfer modes used. It is noteworthy that the SICHTA program gives a
relative power figure (represented by the neutron-flux density) of 2.38 at
time T = 60 seconds.
The variation of the values monitored over time is g~ven in Figs. 4a-4d; the
extremes at time T= 60 are given in Table 4. The time T= 60 sec is merely
an arbitrary reference point, since some of the programs followed the process
f ar ther .
On the basis of our results, in this type of situation w~ may expect about the
following dispersion in extreme values (including inconsistency of input
parameters):
--relative power, 0.49;
--maximum total reactivity, 0.034%;
--maximum fuel temperature, 184�C;
--maximum temperature of inner surface of cladding, 146�C;
--exit dryness of coolant (hot channel), 30.7%;
--minimum critical temperature ratio, 0.27 (excluding DYN79 and hot channel).
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a~l ~ ~ 1!~
- Qa ~ ~ .~,.s~r'~r ~ central
~a~o
~ ~ ~ channel
qa~ 1- :
~ ~ f100
o l _ CC0/ tp ~
� ap~ -.-vKO�D �1
t St[N~a ~ fJLG
tOnr7D j ~ t7D?I
- Q01 q00 -~-Atv1D
o _ .__~i
- ~
z~ - - rra~o
; e,,;a ,?,~cr ~lot
z� '~1D channel
; p , t~ao
4
a ~6 ...tE1D
~y 1Q10
u "~0
reoio
;�o ro lo ~o ~o so ac m o ~o to ao ao ao ~v
~ .
Fig. 4a. Uncontrolled removal of con- Fig. 4b. Uncontrolled removal of
trol rod group at nominal reactor control rod group at nominal reactor
power: reactivity and relative power power: maximum fuel temperature in
as functions of time. central and hot channels as func-
tion of time.
'fn0~"'`�"�~ central
;1 channel
~ -arwro
Zo wra-o �
+rxNu
~ ~aNn.
'0� �na`~"�'�~ central
~ ~ channel ~
xa,a ~
~ ~ r,tv.o ; nw~c;~;r hOt
,UCN71
prN A
channel
no~v:; rond~ S6 '
.pp _ _ - t10 C ~
i ~ channel
~ 60.0
~ I j ~ , t?
i ( t0
1 .
. 7:' .c '~x s`o eo Q~o ~o n~o a so aa n
~ ~f~ . ~Cr1
Fig. 4c. Uncontrolled removal of con- Fig. 4d. Uncontrolled removal of
trol rod group at nominal reactor control rod group at nominal reactor
power: maximum temperature on inner power: critical temperature ratio
surface of cladding for central and for central and hot channels as
hot channels as function of time. function of time.
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,
Table 4. Comparison of extreme valuee achieved for uneontrolled removal of
co*~trol rod group at nominal initial reactor power. ST--central
cnannel; HOR--hot channel. Meaning of numerical values: parameter
value, relative deviation Reference time is T= 60 sec for all
cases except for reactivity and exit temperatur~ of coolant, where
the time of attainment is shown in parentheses.
a b a ? a
: c; H ~
P;~rnmeter ~ ~ ~ V ~
,edi I D A Q
vykon reaktora
I'IPxoat (vztiah-
nut~ k nomi- - 2,$0 2,14 2,29 1,81
nglnomu) 27,1 18,2 ~8,0 0
msxim~lna hod� 0,086 0,073 0,093 0,069
nota celkovej - 44a1 23a7 ~3 4 ~60 ~
reaktivity de [ '
znaYimglna tep� S1' 1938 1890 20�l0 1838
lota paliva T~ b,0 2,9 10,0 0
~~C~ HOR 2705 2880 - 275Q
4,8 0 8,u
_ muxiiniUnu tc~~� >'C :iUU ~3RU :itf:l :!'J3
tola vnutorn8ho 1,8 l,lf U 'l,0
povrchu pokry� HOR S78 429 - -
tia T~ [�C] f 34,0 0
msximum vy- ST 320,3 328,3 329,3 32a,3
etupnej teploty (52) (49) (63) (73)
chladiva [�CJ HOR 328.3 328,3 - 326,3
S (20) (20+8)
vfretupn~ eu-
ahoati chladiva IiOR 23,2 42,0 - 11,:3
[ lOb 272 0
minimglna hod- ST 1,89 1~72 1,4b 3,714
note? kritiokbho 18,8 18,8 0 lb8
tepaln6ho po� HOR 0,86 O,b - 3,71
maru kk, i ~0~0 0 842,0.
Key:
a. Parameter
b. Channel
c. Reactor P/PNOM ~relative to nominal power)
d. Maximum total reactivity ~p
e. Maximum fuel temperature Tf (�C)
f. Maximum temperature T~ of inner surface of cladding (�C)
g. Maximum exit temperature of coolant (�C)
h. Exit dryness of coolant (X)
i. Minimum critical temperature ratio kkr
- 6. Conclusions
On the basis of the results from our test runs, we may state that all of the
programs produce equivalent results in Che solution of the basic sets of
equations, and that the discrepancies result by and large from the capabilities
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of the programa (aelectability of input parameters) and from physical charac-
teristics built into them. Other than the reactivity values used, the deci-
sive f igures affecting the process over time are the fuel-reactivity coeffi-
cients, for which the temperature dependency of the Doppler effect generally
Cannot be ignored. The maximum fuel temperatures are affected not only by
the development of power over time, but also to a substantial degree by the
heat-transfer coefficient of the crack. The equations for heat transfer to
the coolant have a quite prominent effect in the region of heaE transfer to
cooled liquid, while certain discrepancies appear in the equations after the
beginn.ing of boiling. The critical heat-flux density is highly dependent on
the equations used, with the Osm~chkin equations giving considerably lower
values. The results for crisis heat-transfer conditions are greatly affected
by the precision with which the pressure and flow ratios in the core are
described. In spite of these differences, which increase in such a way that
they could be eliminated (or reduced) without great difficulty, each of the
programs tested may be considered capable of performing computations for
safety documentation, since we are dealing with a point model of the reactor
and since the redistribution of coolant flo~v between channels under different
temperature loads can be ignored.
BIBLIOGRAPHY
5. Tinka, I. "Description Procedure and Parametric Study of the WER-440
Reactor Emergencies With Disruption of Reactivity," Energopro3ekt Techni-
cal Report 11-6-19420, Prague, 1979.
6. Pernica, R. "Methodology of WER-S and WER-D Computation Programs,"
UJV [Institute of Nuclear Research] Report 5084-T, Rez, 1979.
7. Pernica, R. "Studies of Selected Emergency States of WER-440 Type 213
Nuclear Power Stations," UJV Report 4553-T, Rez, 1978.
8. Misak, J. "The SICHTA Computer Program for Analyzing the Temnerature
History of a VVER Reactor Core," Sbornik seminare RVHP TF-78 [Documents
from CEMA Seminar TF-78], Budapest, 1978.
9. Misak, J. "Analysis of Selected Fmergency Modes of the V2 Power.Station,"
WJE [Research Institute of Nuclear Power Stations] Report No 118/79.
10. Hermansky, B. "The DYN79 Computer Program and Performance of Simulations
for the WER-440," Vyzk. sprava ve spolupraci FJFI-WJE* [FJFI-WJE
Cooperative Research Office], 1979.
11. Hermansky, B. "Th~ DYN80 Program and Simulations of Uncontrolled E~ection
of Control Rods," FJFI-WJE Cooperatl.ve Research Office, 1930.
*FJFI: ~aculty of Nuclear Science and Physical Engineering
COPYRIGHT: SNTL, n.p. 1981
8480
CSO: 5100/3011 END
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