Practices and rules applied for the design of large dry PWR-containments within EC countries

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Nuclear energy and safety
Energy research

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• Publié par: EUROPEAN-COMMISSION-DIRECTORATE-GENERAL-FOR-THE-INFORMATION-SOCIETY-AND-MEDIA-DI
• Nombre de lectures: 25
• Langue: English
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Commission of the European Communities
nuclear science
■
and technology
Practices and rules applied for the design of large
dry PWR-containments within EC countries Commission of the European Communities
nuclear science
and technology
Practices and rules applied for the design of large
dry PWR-containments within EC countries
H. Karwaţ
Technische Universität München
D*804§ Qarchjng
Contract No ECI­1462­B 7221­86 D
PADA. lïH ?." Moth.
Directorate-General
Science, Research and Development NX./-- *
^7^ËÛR"12251EN| 1989 Published by the
COMMISSION OF THE EUROPEAN COMMUNITIES
Directorate-General
Telecommunications, Information Industries and Innovation
L-2920 Luxembourg
LEGAL NOTICE
Neither the Commission of the European Communities nor any person acting
on behalf of then is responsible for the use which might be made of
the following information
Cataloguing data can be found at the end of this publication
Luxembourg: Office for Official Publications of the European Communities, 1989
ISBN 92-826-0605-8 Catalogue number: CD-NA-12251 -EN-C
© ECSC-EEC-EAEC, Brussels • Luxembourg, 1989
Printed in Belgium Abstract
The containment system is an essential component of nuclear power plants
and its behaviour plays an important role in the evaluation of accident
consequences of pressurized water reactors.
A variety of constructive solutions exist for PWR containments (the pre­
ferred reactor type in European countries) this being partially the result
of different national practices and rules applied in design and safety as­
sessment of containment structures. The rules and guidelines applied dif­
fer from country to country and are sometimes also linked to national or
international research experience.
The design of PWR containments is based on the prediction of thermalhy-
draulic and external initiator Loads which the containment shell and the
internal structure must withstand. The analytical methods for the simula­
tion of the relevant phenomena and the experimental background to sup­
port these methods has been illuminated.
Main purpose of this report was to compare the rules and practices ap­
plied within the member countries of the European Community to determine
the thermal-hydraulic local and global design loads. This comparison was
based on applicable national rules and guidelines as far as such were
available. The report does not discuss the methods and rules applied to
fix the dimensions of containment structures. This matter has been treat­
ed in other reports drafted earlier for the European Commission.
The comparison shows in general that the analytical methods to determine
the design loads and the specific instructions on how to apply these
methods are similar throughout the Member countries surveyed and result
to a large extent from a similar degree and approach to conservatism in
the determination of design loads. Major differences exist in the structur­
al layout of PWR containments. Spherical steel shells, cylindrical rein­
forced as well as prestressed concrete shell structures used in both
single and double structural barriers are in application.
Although the load determination methods for these three typical groups of
PWR containment structures are nearly equal the evaluation of their
structural design is quite different. However, within the frame of this
study it was not possible to compare in detail e.g. the required methods
for the stress analysis of a steel shell, a reinforced and a prestressed
concrete containment shell structure. If the practically existing margins of
conservatism for each of these typical containment categories were to be
quantified, more in-depth studies would be necessary going beyond the
scope of earlier CEC reports mentioned above.
Concerning the structural design of the internal compartments of a PWR
containment the differences in structural layout are much smaller as all
PWR containments are subdivided into the typical sections housing the re­
actor pressure vessel, the secondary steam generators, pumps, pressur-
izer etc. In general, all compartment structures are reinforced concrete
structures requiring similar stress analysis methods.
Ill As far the thermalhydraulic load prediction procedure is concerned only
small differences have been found, particularly in the required safety ad­
dition factors covering the uncertainties of the predictive analytical simu­
lation procedures.
Albeit carefully designed a small probability exists that a containment may
fail during a severe accident sequence. The containment behaviour essen­
tially determines the consequences of such accidents which are assessed
in terms of the residual risk of nuclear power plant operation. To fully
merit this risk it is very useful to have some measures on the available
margins between the level of the design load and the level of thermalhy­
draulic loads at which a major failure of the main structures of the con­
tainment shell is to be expected. This is a very important aspect for the
assessment of those accidents which albeity unlikely are anticipated
not to be mitigated by the existing safety systems (beyond DPA condi­
tions). Fig. 12 of sect. 7 of this report illustrates the situation for a
particular accident sequences. Major failures of the containment shell or
essential parts of it (e.g. penetrations, access locks etc.) are expected
to occur at a range of pressures well above the design pressure. Compo­
nent testing and several integral containment load experiments have in­
creased the confidence in predictions of failure loads of such structures.
A continuous monitoring of the associated activity is highly recommended
to broaden the confirmatory basis of this observation«
On the other hand, considerable uncertainties must be taken into account
for the predictability of thermalhydraulic loads which may accompany se-
vere accident scenarios going beyond the Design Basis Accident. This is
largely due to the variety of important boundary conditions deduced from
the probabilities of analytical component failures or malfunctions describ­
ing the risk-dominating accident sequence. The analyses of severe acci­
dent phenomena also require predictive simulation models to be applied for
many situations for which they have only partially been verified by ex­
periments. As far as the containment behaviour is concerned the conse­
quences of a hydrogen burn and the mechanical and chemical response to
a highly energetic coremelt ejection from the reactor pressure vessel are
issues which dominate the uncertainties in the prediction of the contain­
ment response and hence the associated radiological source term
discussion.
Several experimental activities which address,, some of these problems are
still on the wşy worldwide with a considerable financial Involvement on the
national and international level. The analytical assessment of many valu­
able expepimepţs however is not very well balanced, Financial and man­
power commitment devoted to analyze experiments has been noted to
scarcely amount more than 5­10% of the commitment associated to the ex­
periments. A considerable increase in analytical assessment of experimen­
tal evidence should be considered as the most effective opportunity for
improving the situation and reducing at least some of the uncertainties, A
specific task group should be formed which should generate a code vali­
dation matrix for the most important severe accident related phenomena
requiring improvements.
­ IV CONTENTS
Page
1. INTRODUCTION 1
2. SCOPE OF THIS REPORT 8
3. THE DESIGN BASIS ACCIDENT PHILOSOPHY
4. THERMOHYDRAULIC LOADS 10
4.1. Local and global pressurisation1
4.2. Jet impingement and pipewhip effects5
5. ANALYTICAL SILULATION MODELS FOR THE THERMOHYDRAULIC
LOAD PREDICTION6
5.1. Important phenomena for the analytical simulation 1
5.2. The analytical basis 2
5.3. The experimental background2
5.4. The importance of international standard problems 27
6. RULES AND GUIDELINES 31
6.1. Specific requirements for prediction of the maximum
global containment pressure4
6.2. Specifics forn of the compartment
pressurisation process7
6.3. Specific requirements for the prediction of jet
impingement loads 40
6.4. Containment heat removal systems
6.5. Assurance of containment isolation2
7. MARGINS BETWEEN DESIGN AND ULTIMATE LOAD CAPACITY 43
8. CONCLUSIONS AND RECOMMENDATIONS8
9. REFERENCES 5
ANNEXES 67
Annex A - Rules and guidelines (Excerpts) 69
Annex B - The specific phenomena of severe accident scenarios 81. Introduction
Nuclear power plants and their associated protection and safety systems
are designed to operate so that even in the event of a major (albeit very
improbable) accident, the reactor core will not melt. This emphasis on re­
actor safety is justified, since so long as the reactor core remains essen­
tially intact, eve