Analysis of failures in concrete containments

Directorate-General For Translation European Commission

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Nuclear energy and safety

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• Publié par: EUROPEAN-COMMISSION
• Nombre de lectures: 1
• Langue: English
• Poids de l'ouvrage: 3 Mo

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Commission of the European Communities
Nuclear Science and Technology
Shared Cost Action
Reactor Safety Programme 1985-1987
Analysis of Failures in Concrete Containments
Final Report
Directorate-General for Science Research and Development
Joint Research Centre - Ispra Site
September 1989 EUR 12390 EN Commission of the European Communities
Nuclear Science and Technology
Shared Cost Action
Reactor Safety Programme 1985-1987
Analysis of Failures in Concrete Containments
Final Report
Work performed in the frame of the
Shared Cost Action (SCA) programme 1985-87
Research Contract No. 3300-87-12 EUSPE
Authorfs): A. Moreno-Gonzalez
Empresarios Agrupados S.A.
Glorieta de Quevedo 9
E-28015 Madrid
Directorate-General for Science Research and Development
Joint Research Centre - Ispra Site
September 1989 Published by the
COMMISSION OF THE EUROPEAN COMMUNITIES
Directorate-General
Telecommunications, Information Industries and Innovation
Bâtiment Jean Monnet
LUXEMBOURG
LEGAL NOTICE
Neither the Commission of the European Communities nor any person
acting on behalf of the Commission 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 92-826-0756-9 Catalogue number: CD-NA-12390-EN-C
© ECSC - EEC - EAEC, Brussels-Luxembourg, 1989
Printed in Italy CONTENTS
1. INTRODUCTION 1
2. OBJECTIVES OF CONTAINMENT OPERATION AND DEFINITION
OF FAILURE 2
3. PRESENT STATUS OF CONTAINMENT FAILURE ANALYSES 3
3.1 FAILURE ANALYSIS PARAMETERS
3.2 EXPERIENCE
4
3.3 LINES OF ACTION
5
FACTORS THAT DETERMINE CONTAINMENT BEHAVIOUR 5
4.1 CONTAINMENT LOAD DEFINITION
4.1.1 Steam Spike or Steam Explosion Loads 6
4.1.2 Concrete Attack Loads S
4.1.3 Hydrogen Burn Loads 7
4.1.4e Liner Attack
4.1.5 Diffusion Flame Loads
4.1.6 Direct Heatings 8
4.1.7 Seismic-Induced Severe Accident
4.1.8 Impact Loads
4.1.9 Table of Load Conditions
4.2 EVALUATION OF LEAKS ASSOCIATED WITH CONTAINMENT OPERATION 11
4.2.1 Leak Rate through Penetrations, Valves and
Hatches H
4.2.2 Effluent Rate through the Cracked Concrete 13
4.2.3 Containment Venting ¡5
4.3 AGING OF CONTAINMENT MATERIALS 17
4.3.1 Experience
4.3.2 Degradation Mechanisms in Concrete Containments -,-j
4.3.3 Summary of Conclusions and Recommendations 3¡
ANALYSIS OF FAILURE IN CONCRETE CONTAINMENTS 3
5.1 CONTAINMENT STRUCTURAL FAILURE MODES
5.2 HANDLING OF THERMAL LOADS 40
5.2.1 Effect of Thermal Loads on Concrete
4
5.2.2t ofls on the Liner 4
5.3 METHODOLOGY OF STRUCTURAL FAILURE ANALYSIS
5.4 FAILURE CRITERIA AND PARAMETERS6 5.5 UNCERTAINTIES 48
5.6 COMMENTS AND CONCLUSIONS9
6. STRUCTURAL BREAKAGE AND LEAKAGE CRITERIA 50
6.1 CRITERIA FOR BREAKAGE DUE TO GLOBAL FAILURE
6.1.1 Failures due to Membrane Stress
6.1.2s due to Radial Shear4
6.1.3s due to Tangential Shear5
6.1.4 Failures due to Crushing 57
6.1.5 Soil Failure9
6.1.6 Basemat Failure
6.2 CRITERIA FOR BREAKAGE DUE TO LOCAL FAILURE 60
6.2.1 Failure due to Radial and Tangential Shear
6.2.2e due to Crushing 61
6.2.3 Failure due to Punching or Peripherical Shear 6
6.2.4e due to Torsion2
6.2.5e of Containment Equipment3
6.2.6 Failure of Gaskets and Seals5
6.2.7 Failures including the Liner as a Structural
Element 66
6.2.8 Failure of Liner Anchorage7
6.2.9e ofr due to Buckling 6
6.3 INTERPRETATION OF GLOBAL AND LOCAL FAILURE CRITERIA 68
6.3.1 Parameters and Mechanisms that Characterize
Structural Failure8
6.3.2 Relation between Structural Failure Criteria
and Leak Criteria
6.4 CRITERIA FOR FAILURE DUE TO LINER BREAK (LEAKAGE) 70
6.4.1 Leakage in Concrete Containments without Liner
6.4.2e ines with Liner1
6.4.3 Determination of Liner Deformation Process 7
6.4.4 Criteria for Leakage and Crack Growth6
6.5 INTERPRETATION OF LEAK FAILURE CRITERIA 82
6.5.1 Leak Parameters and Leak Rates in Unlined Containments 8
6.5.2 Leaks and Leak Rates in Lineds 8
6.6 GENERAL CRITERIA FOR CONCRETE STRUCTURE FAILURE 8
6.6.1 Failure Criteria according to Codes4 6.6.2 Failure Criteria in Reinforced Concrete Elements 86
6.6.3ea in Prestressedes 9
6.6.4 Special Issues 9
6.6.5 Failure Criteria in Plain Concrete 105
6.6.6ea Application Methodology for Reinforced
and Prestressed Concrete Structures 12
7. METHODOLOGY FOR APPLICATION OF LEAK AND BREAK CRITERIA6
8. SUMMARY AND CONCLUSIONS 128
9. REFERENCES 135
APPENDIX A - REVIEW OF CONTAINMENT SCALE TESTS 143 1. INTRODUCTION
The first overall analysis of the extent of risks associated with
the operation of light water reactors was the Reactor Safety Study (1).
This study showed that the probability of a severe accident, an accident
involving core degradation, was higher than previously estimated.
The accident at Three Mile Island, Unit 2, and subsequently the
accident at Chernobyl, awakened the feelings of regulatory bodies,
industries and public opinion on the consequences of accidents involving
core degradation.
In the past few years, a great effort has been made to assess risks
with more precision and implement means to lessen them. The result of
this effort is the Risk Reference Document (2), which describes in detail
the sequences of possible accidents in five reference plants.
The containment is of particular significance in this study, since
it is the ultimate barrier against the release of radioactive products
and its function is exclusively that of safety. For this reason it has
been given special attention and a detailed description of the fault tree
for different containments is included (2). The results show that the
containment plays a determining role in the assessment of risks
associated with an accident involving core degradation.
Traditional design employing a philosophy of "in-depth defense"
implies designing the containment as a pressure vessel in accordance with
the ASME Code to withstand the loads associated with a Design Basis
Accident (DBA).
However, an accident involving core degradation could result in
loads on the containment greater than the design loads. Such loads could
cause damage, in turn giving rise to leak paths for radioactive products.
Analyses of accidents with core degradation do not form part of the
licensability procedures; they are used to assess the risks associated
with such accidents and are essential in the preparation of Probabilistic
Risk Analyses. However, although the probability of such accidents is
very low, public interest is focussed more on the possible consequences
than on the risks, and thus the interest in knowing how to limit such
consequences.
The problem is knowing the behaviour of the containment in the light
of possible accident sequences. Such knowledge will facilitate a more
reliable assessment of the consequences associated with failure.
As a basic criterion, it is considered that containment failure can
be defined as the incapability, in the event of a severe accident, to
fulfil its operating objectives. This does not include operational
errors. With a definition of this kind, structural failure is only a
part of the general concept of containment failure.
From this point of view,t failure does not only mean
taking into account the actual structural failure, but also carrying out
a study and integrated assessment of the behaviour of penetrations,
equipment hatches, vents forming leak paths to the atmosphere, possible
load, pressure and temperature histories, impacts, etc.
It is not intended in this study to give an exhaustive description
of the problem concerning containment behaviour under severe accident
loads involving core degradation. The study is, therefore, focussed upon
three aspects (see Fig. 1).
The first describes certain factors that determine the behaviour of
the containment and the establishment of definition criteria for possible
failures thereof. Particular attention will be paid to two
characteristic aspects of concrete containments, the problem of concrete
aging and that of leaks through the cracked concrete.
1 Two other sections deal from a structural and mechanical point of
view with possible failure mechanisms, both global and local, the
analysis models, and the parameters that control them.
The study is completed with a detailed description of possible
failure criteria.
The general work scheme corresponds to the typical scheme of
containment integrity analysis detailed in the following sketch.
CONTAINMENT INTEGRITY RESEARCH
I I I
FAILURE
| SEVERE | |
I I MODES
| ACCIDENTI |
J L 1
I I
CORE MELTING MELTHROUGH
I
1 1
I I I
PRESSURE
| | GLOBAL + LOCAL |
+ I
TEMPERATURE FAILURE MODE ITHRESHOLD
J 1
MODE
I I - I
SEISMIC EVENT | | EXISTING LEAKS |— LEAKAGE
L J L
2. OBJECTIVES OF CONTAINMENT OPERATION AND DEFINITION OF FATLURE
The basic idea of safety in the operation of nuclear power plants