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SAFETY SERiES No. 50-SG-D15

Seismic Designand Quaiificationfor Nuciear Power PiantsA Safety Guide

A PUBLiCATiONW!TH!N THE NUSS PROGRAMME

) N T E R N A T ! O N A L A T O M t C E N E R G Y A G E N C Y , V t E N N A , 1 9 9 2

This publication is no longer valid Please see http://www-ns.iaea.org/standards/

CATEGORIES IN THE IAEA SAFETY SERIES

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Safety Fundam entals (silver cover)

Basic objectives, concepts and principles to ensure safety.

Safety Standards (red cover)

Basic requirements which must be satisfied to ensure adequate safety for particular activities or application areas.

Safety Guides (green cover)

Recommendations, on the basis of international experience, relating to the fulfilment of basic requirements.

Safety Practices (blue cover)

Practical examples and detailed methods which can be used for the application of Safety Standards or Safety Guides.

Safety Fundamentals and Safety Standards are issued with the approval of the IAEA Board of Governors; Safety Guides and Safety Practices are issued under the authority of the Director General of the IAEA.

An additional category, Safety Reports (purple cover), comprises independent reports of expert groups on safety matters, including the development of new principles, advanced concepts and major issues and events. These reports are issued under the authority of the Director General of the IAEA.

There are other publications of the IAEA which also contain information important to safety, in particular in the Proceedings Series (papers presented at symposia and conferences), the Technical Reports Series (emphasis on technological aspects) and the IAEA-TECDOC Series (information usually in a preliminary form).


SEISMIC DESIGN AND QUALIFICATION FOR NUCLEAR POW ER PLANTS

A Safety Guide


The foHowing States are Members of the International Atomic Energy Agency:

AFGHANISTAN HAITI PANAM A

ALBANIA HOLY SEE PARAGUAY

ALGERIA HUNGARY PERU

ARGENTINA ICELAND PHILIPPINES

AUSTRALIA INDIA POLAN D

AUSTRIA INDONESIA PORTUGAL

BANGLADESH IRA N, ISLAM IC REPUBLIC OF QATAR

BELARUS IRAQ ROM ANIA

BELGIUM IRELAND RUSSIAN FED ERA TIO N

BOLIVIA ISRAEL SAUDI ARABIA

BRAZIL ITALY SENEGAL

BULGARIA IA M A ICA SIERRA LEONE

CAMBODIA JAPAN SINGAPORE

CAM EROON JORDAN SOUTH AFRICA

CANADA KENYA SPAIN

CH ILE KOREA, REPUBLIC OF SRI LANKA

CHINA K UW AIT SUDAN

COLOM BIA LEBANON SW EDEN

COSTA RICA LIBERIA SW ITZERLAND

CO TE D 'IV O IRE LIBYAN ARAB JAM A HIRIY A SYRIAN ARAB REPUBLIC

CUBA LIECHTEN STEIN TH A ILA ND

CYPRUS LUXEM BOURG TUNISIA

CZECHOSLOVAKIA M ADAGASCAR TURKEY

DEM OCRA TIC PEO PL E'S M ALAYSIA UGANDA

REPU BLIC OF KOREA M A H UKRAINE

D ENM ARK M AURITIUS UNITED ARAB EM IRATES

D OM IN ICAN REPUBLIC M EXICO UNITED KINGDOM OF GREAT

ECU A DO R M ONACO BRITAIN AND NORTHERN

EGYPT M ONGOLIA IRELAND

EL SALVADOR M OROCCO UNITED REPUBLIC OF

ESTONIA M YANM AR TANZANIA

ETHIOPIA NAM IBIA UNITED STATES OF AM ERICA

FINLAN D N ETH ERLAN D S URUGUAY

FRAN CE N EW ZEALAN D V ENEZUELA

GABON NICARAGUA VIET NAM

G ERM ANY N IGER YUGOSLAVIA

GHANA NIGERIA ZAIRE

G REECE NORW AY ZAMBIA

G UATEM ALA PAKISTAN ZIM BABW E

The Agency's Statute was approved on 23 October 1956 by the Conference on the Statute of the IAEA held at United Nations Headquarters, New York; it entered into force on 29 July <957. The Headquarters of the Agency are situated in Vienna. Its principal objective is "to accelerate and enlarge the contribution of atomic energy to peace, health and prosperity throughout the wortd

O IAEA, 1992

Permission to reproduce or translate the information contained in this publication may be obtained by writing to the International Atomic Energy Agency, Wagramerstrasse 5, P.O. Box 100, A-1400 Vienna, Austria.

Printed by the IAEA in Austria October 1992


SAFETY SERIES No. 50-SG-D15

SEISMIC DESIGN AND QUALIFICATION

FOR NUCLEAR POWER PLANTS

A Safety Guide

INTERNATIONAL ATOMIC ENERGY AGENCY VIENNA, 1992


THIS SAFETY SERIES PUBLICATION IS ALSO ISSUED IN FRENCH, RUSSIAN AND SPANISH

VIC Library Catatoguing in Publication Data

Seismic design and qualification for nuclear power ptants : a safety guide. — Vienna : International Atomic Energy Agency, 1992.

p. ; 24 cm. — (Safety series, ISSN 0074-1892 ; 50-SG-DI5) STI/PUB/917 ISBN 92-0-103592-6 Includes bibliographical references.

1. Nuclear power plants—Earthquake effects. 2. Nuclear engineering— Safety measures. 3. Earthquake resistant design. I. International Atomic Energy Agency. II. Series.

VICL 92-00036


FO R E W O R D

by the D irecto r G eneral

Nuclear power is well established and can be expected to become an even more significant part of the energy programmes of many countries, provided that its safe use can be ensured and be perceived to be so ensured. Although accidents have occurred, the nuclear power industry has generally maintained a good safety record. However, improvements are always possible and necessary. Safety is not a static concept.

The International Atomic Energy Agency, recognizing the importance of the safety of the industry and desiring to promote an improving safety record, set up a programme in 1974 to give guidance to its Member States on the many aspects of the safety of nuclear power reactors. Under this Nuclear Safety Standards (NUSS) programme, some 60 Codes and Safety Guides dealing with radiological safety were published in the IAEA Safety Series between 1978 and 1986. The NUSS programme was developed for land based stationary plants with thermal neutron reactors designed for the production of power but the provisions may be appropriate to a wider range of nuclear applications.

In order to take account of lessons learned since the first publication of the NUSS programme was issued, it was decided in 1986 to revise and reissue the Codes and Safety Guides. During the original development of these publications, as well as during the revision process, care was taken to ensure that all Member States, in particular those with active nuclear power programmes, could provide their input. Several independent reviews took place including a final one by the Nuclear Safety Standards Advisory Group (NUSSAG). The revised Codes were approved by the Board of Governors in June 1988. In the revision process new developments in technology and methods of analysis have been incorporated on the basis of international consensus. It is hoped that the revised Codes will be used, and that they will be accepted and respected by Member States as a basis for regulation of the safety o f power reactors within the national legal and regulatory framework.

Any Member State wishing to enter into an agreement with the IAEA for its assistance in connection with the siting, design, construction, commissioning, operation or decommissioning o f a nuclear power plant will be required to follow those parts o f the Codes and Safety Guides that pertain to the activities to be covered by the agreement. However, it is recognized that the final decisions and legal responsibilities in any licensing procedures rest with the Member States.

The Codes and Safety Guides are presented in such a form as to enable a Member State, should it so desire, to make their contents directly applicable to activities under its jurisdiction. Therefore, consistent with the accepted practice for codes and guides, and in accordance with a proposal o f the Senior Advisory Group,


'shall' and 'should' are used to distinguish for the user between strict requirements and desirable options respectively.

The five Codes deal with the following topics:

— Governmental organization— Siting— Design— Operation— Quality assurance.

These five Codes establish the objectives and basic requirements that must be met to ensure adequate safety in the operation of nuclear power plants.

The Safety Guides are issued to describe to Member States acceptable methods o f implementing particular parts o f the relevant Codes. Methods and solutions other than those set out in these Guides may be acceptable, provided that they give at least equivalent assurance that nuclear power plants can be operated without undue risk to the health and safety of the general public and site personnel. Although these Codes and Safety Guides establish an essential basis for safety, they may require the incorporation of more detailed requirements in accordance with national practice. Moreover, there will be special aspects that need to be assessed by experts on a case by case basis.

These publications are intended for use, as appropriate, by regulatory bodies and others concerned in Member States. In order to comprehend the contents of any o f them fully, it is essential that the other relevant Codes and Safety Guides be taken into account. Other safety publications of the IAEA should be consulted as necessary.

The physical security of fissile and radioactive materials and of nuclear power plants as a whole is mentioned where appropriate but is not treated in detail Non-radiological aspects o f industrial safety and environmental protection are also not explicitly considered.

The requirements and recommendations set forth in the NUSS publications may not be fully satisfied by older plants. The decision of whether to apply them to such plants must be made on a case by case basis according to national circumstances.


CONTENTS

1. INTRODUCTION .............................................................................................. 7

Background (101-102) ............................ ............................. ........................ 7Objective (103) .................................................................................................... 7Scope (104) .......................................................................................................... 7Structure (105-106) ........................................................................................... 8

2. GENERAL CONCEPT ...................................... .............................................. 8

Earthquake levels (202-203) ............................................................................. 8Seismic categorization for structures, systems and

components (204-211) ................................................................................. 9Combination of earthquake loads with other plant process loading

conditions (212-215) ........................................................................... . 10Allowable limits for stress, strain and deformation (216-220) ............... 11

3. SEISMIC DESIGN ............................................................................................. 12

General approach to seismic design (301-305) ...................................... . 12Civil engineering structures (306) .................................................................. 13Earth structures (307) ........................................................................................ 15Piping and equipment (308) ............................................................................. 15Improvement of the vibration resistance of components (309) ............ 16Possible modes of failure of new types of equipment (310) ...... ............ 17Functional aspects of seismic design (311) .................................................. 17Effects of vertical ground motion (312) ........................................................ 17

4. SEISMIC QUALIFICATION: ANALYSIS, TESTING,EARTHQUAKE EXPERIENCE AND INDIRECT METHODS ............ 18

Seismic input (402) ............................................................................................. 18Direct qualification (403) .................................................................................. 18Qualification by analysis (404) ........................................................................ 19General and analytical methods (405-408) .................................................. 19General modelling techniques (409-411) ...................................................... 20Civil engineering structures (412-423) ......................................................... 20Mechanical and electrical components (424-439) ...................................... 22Distribution systems (440-446) ....................................................................... 24Seismic qualification testing (447-468) ......................................................... 25Seismic qualification by earthquake experience (469) ............................... 29Indirect method (470-473) ................................................................................ 30

DEFINITIONS ............................................................................................................... 1


5. SEISMIC INSTRUMENTATION .................................................................. 31

Type, location and number of instruments (502-504) ............................... 31Maintenance of instrumentation (505) ........................................................... 31Data evaluation (506) ........................................................................................ 32

APPENDIX I: METHODS OF SEISMIC ANALYSIS ..................................... 33

APPENDIX II: MODELLING TECHNIQUES ..................................................... 34

ANNEX I: EXAMPLES OF LISTS OF SEISMIC CATEGORY 1AND OTHER CATEGORY ITEMS ................................... .......... 41

ANNEX II: SLOSHING AND IMPULSE EFFECTS INLIQUID CONTAINERS ....................................................................... 53

REFERENCES ............................................................................................................... 55

BIBLIOGRAPHY .......................................................................................................... 59

CONTRIBUTORS TO DRAFTING AND REVIEW ........................................... 61

LIST OF NUSS PROGRAMME TITLES .............................................................. 67

SELECTION OF IAEA PUBLICATIONS RELATING TOTHE SAFETY OF NUCLEAR POWER PLANTS .............................................. 71


DEFINITIONS

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Operational States

States defined under Norma! Operation or Anticipated Operational Occurrences.

Normal Operation

Operation of a nuclear power plant within specified operational limits and conditions including shutdown, power operation, shutting down, starting, maintenance, testing and refuelling.

Anticipated Operationa! Occurrences*

All operational processes deviating from Normal Operation which are expected to occur once or several times during the operating life of the plant and which, in view of appropriate design provisions, do not cause any significant damage to items important to Safety nor lead to Accident Conditions.

Accident (or Accident State)

A state defined under Accident Conditions or Severe Accidents.

Accident Conditions

Deviations^ from Operational States in which the releases of radioactive materials are kept to acceptable limits by appropriate design features. These deviations do not include severe accidents.

' Examples of Anticipated Operational Occurrences are toss o f normal electric power and faults such as a turbine trip, malfunction of individual items of a normally running plant, failure to function of individual items of control equipment, loss of power to main coolant pump.

A deviation may be a major fuel failure, a loss of coolant accident (LOCA), etc.

1


Design Basis Accidents

Accident Conditions against which the nuclear power plant is designed according to established design criteria.

Severe Accidents

Nuclear power plant states beyond Accident Conditions including those causing significant core degradation.

Accident Management

Accident management is the taking of a set of actions

— during the evolution of an event sequence, before the design basis of the plant is exceeded, or

— during Severe Accidents without core degradation, or— after core degradation has occurred

to return the plant to a controlled safe state and to mitigate any consequences of the accident.

Plant states

Operational states Accidents

Anticipated Normal operational

operation ) occurrencesAccident

conditionsSevere

accidents

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Safety — See Nuclear Safety

Nuclear Safety (or simply Safety)

The achievement o f proper operating conditions, prevention o f accidents or mitigation of accident consequences, resulting in protection of site personnel, the public and the environment from undue radiation hazards.

Safety Systems

Systems important to Safety, provided to assure the safe shutdown of the reactor or the residual heat removal from the core, or to limit the consequences of Anticipated Operational Occurrences or Accident Conditions.

Protection System

A system which encompasses all electrical and mechanical devices and circuitry, from sensors to actuation device input terminals, involved in generating those signals associated with the protective function,

Safety Actuation System

The collection of equipment required to accomplish the necessary safety actions when initiated by the Protection System.

Safety System Support Features

The collection of equipment that provides services such as cooling, lubrication, and energy supply required by the Protection System and the Safety Actuation Systems.

Safety Systems consist of the Protection System, the Safety Actuation systems, and the Safety System Support Features. Components of Safety Systems may be provided solely to perform Safety Functions or may perform safety functions in some plant Operational States and non-safety functions in other plant Operational States (see diagram at the end of the Definitions).

3


Plan t equipm ent

Items important to safety Items no? important to safety

system actuation systemsystems support

features

Construction*

The process of manufacturing and assembling the components of a nuclear power plant, the erection o f civil works and structures, the installation o f components and equipment, and the performance of associated tests.

Design Floor Response Spectrum

Response spectrum defined at a particular building elevation and obtained by modifying one or more floor response spectra in order to consider the variability and uncertainty of input ground motion and of the characteristics of both building and foundation.

Design Floor Time Histories

Time histories of floor motion of structure under consideration derived from the design basis ground motion, including the variability and uncertainty in input ground motion and in building and foundation characteristics.

** The terms Siting, Design, Construction, Commissioning, Operation and Decommissioning are used to delineate the six major stages of the licensing process. Several of the stages may coexist; for example, Construction and Commissioning, or Commissioning and Operation.

4


Floor Response Spectrum

The floor response spectrum at a particular level o f a structure for a given ground motion.

Item

A general term covering materials, parts, components, systems or structures and including computer software.

O peration (see footnote 4)

All activities performed to achieve the purpose for which the nuclear power plant was constructed^ including maintenance, refuelling, in-service inspection and other associated activities.

Regulatory Body

A national authority or a system o f authorities designated by a Member State, assisted by technical and other advisory bodies, and having the legal authority for conducting the licensing process, for issuing licences and thereby for regulating nuclear power plant Siting, Design, Construction, Commissioning, Operation and Decommissioning or specified aspects thereof^.

Residual Heat

The sum of the heat originating from radioactive decay and shutdown fission and the heat stored in reactor related structures and in heat transport media.

Site

The area containing the plant, defined by a boundary and under effective control of the plant management.

Siting (see footnote 4)

The process of selecting a suitable Site for a nuclear power plant, including appropriate assessment and definition of the related design bases.

This national authority could be either the government itself, or one or more departments of the government, or a body or bodies specially vested with appropriate legal authority.

5


A time record of earthquake motion or the earthquake response motion on a time base of a particular floor or at some level on a structure resting on the ground.

Time History

NOTE ON THE INTERPRETATION OF THE TEXT

When an appendix is included it is considered to be an integral part of the document and to have the same status as the main text of the document.

However, annexes, footnotes and bibliographies are only included to provide additional information or practical examples that might be helpful to the user.

In several cases phrases may use the wording 'shall consider...' or 'shall... as far as practicable'. In these cases it is essential to give the matter in question careful attention, and the decision must be made in consideration of the circumstances of each case. However, the final decision must be rational and justifiable and its technical grounds must be documented.

Another special use of language is to be noted: " 'a ' or 'b ' " is used to indicate that either 'a ' or 'b ', but also the combination of both 'a ' and 'b ', would fulfil the requirements. If alternatives are intended to be mutually exclusive, "either... o r... " is used.

6


1. INTRODUCTION

BACKGROUND

101. This Safety Guide, which supplements the IAEA Code on the Safety of Nuclear Power Plants: Design (IAEA Safety Series No. 50-C-D (Rev. 1)), forms part of the Agency's programme, referred to as the NUSS programme, for establishing Codes and Guides relating to land based stationary thermal neutron power plants. A list of NUSS publication titles is given at the end of this book.

102. The present Guide was originally issued in 1979 as Safety Guide 50-SG-S2 within the series of NUSS guides for the siting o f nuclear power plants, extending seismic considerations from Safety Guide 50-SG-S 1 into the design and verification Held. During the revision phase in 1988-1990, this emphasis on design aspects was confirmed and consequently the Guide has been reclassified as a design Guide with the corresponding identification number 50-SG-D 15. The general character of the Guide has not been changed and it still relates strongly to 50-SG-S 1, which gives guidance on how to determine design basis ground motions for a nuclear power plant at a given site.

OBJECTIVE

103. The purpose of this Safety Guide is to provide details of a generally accepted way to design a nuclear power plant such that earthquakes at the site determined according to IAEA Safety Guide 50-SG-S1 will not jeopardize the safety of the plant. It is also intended to give guidance on methods and procedures for analysing, testing and qualifying structures and equipment such that they fulfil the purpose foreseen in the design. The Guide addresses designers of nuclear power plants and also safety assessors and regulators concerned with the licensing of plants.

SCOPE

104. The guide is applicable to the design of nuclear power plants to withstand earthquakes irrespective of the severity of the earthquake ground motion or the risk to individual plant items. It is recognized that there may be simplified procedures for some of the recommended design and verification methods. The adequacy of such procedures for the safety objective would have to be determined for the individual circumstances. It is also recognized that there is more than one possible engineering solution to problems and the approach adopted for one nuclear power plant may result in significant differences in the design of that plant compared to that of another for which a different approach has been adopted.

7


STRUCTURE

105. The general concept in Section 2 defines two or more categories of items in the plant for which, depending on their importance to safety, different guidance and recommendations are set out. This section also discusses combinations of loads and allowable stresses or deformations. In Section 3 the primary design principles for the achievement of the protection objective are presented. Guidance on analytical methods for structures and mechanical or electrical equipment is given in Section 4. This section also contains advice on qualification and the conduct of tests. Section 5 presents information on recommended seismic instrumentation, its location and use.

106. A number o f appendices and annexes expand on specific topics such as seismic categorization of items, analysis and modelling techniques.

2. GENERAL CONCEPT

201. The specific objective of this section is to assist in categorizing structures, systems and components of a nuclear power plant in terms of their importance to safety in the event of a design basis earthquake, to set out guidance and to provide recommendations on seismic design, analysis and testing to ensure safety.

EARTHQUAKE LEVELS

202. For safety design purposes the SL-2 earthquake level' shall be used, where SL-2 is a level o f extreme ground motion that shall have a very low probability of being exceeded during the lifetime of the plant^. The recommended minimum level is a peak ground acceleration of O .lg (see Safety Guide 50-SG-S1 (Rev. 1), para. 504).

203. For certain conditions, including some event combinations, post-accident inspection, national licensing requirements and economic considerations, a second earthquake level^' known as SL-1 may be considered in the design. This level corresponds to a less severe, more likely earthquake load condition with different safety implications from SL-2.

' This level corresponds to an earthquake level often denoted as SSE (safe shutdown earthquake).

See Safety Guide 50-SG-S1 (Rev. 1), Section 5 for methodological background for the determination of SL-2 and SL-1.

This level corresponds to an earthquake level often denoted as OBE (operational basis earthquake).


SEISMIC CATEGORIZATION FOR STRUCTURES, SYSTEMS AND COMPONENTS

204 Structures, systems and components (called hereafter 'item s') may be divided into two or more categories in terms of their importance to safety in the event of an earthquake. The purpose of this categorization is to facilitate the protection of the public and the environment against radioactive releases.

205. A jeMwz'c 7 shall be established for the plant. Items within thiscategory shall be designed or demonstrated to withstand the consequences of ground motions associated with earthquakes of level SL-2. Category 1 shall include'*:

(1) Items whose failure could directly or indirectly cause accident conditions (see Definitions).

(2) Items required for shutting down the reactor, monitoring critical parameters, maintaining the reactor in a shutdown condition and removing residual heat over a long period.

(3) Items that are required to prevent radioactive releases or to maintain releases below limits established by the regulatory body for accident conditions (e.g. the containment system).

206. As a conservatiye measure, it is recommended that category 1 include those items which are designed to mitigate the consequences of design basis accidents which may be postulated to occur in the primary pressure boundary, despite the fact that the primary pressure boundary is designed to withstand earthquake loads.

207. .s /.svTMc 2 may be established for the plant. Items so categorizedshall be designed to withstand the consequences of ground motions associated with earthquakes of a severity as defined below (para. 208). Category 2 should include:

(1) Items, not in category 1, which are required to prevent the escape of radioactivity beyond normal operational limits.

(2) Items, not in category 1, required to mitigate those accident conditions which last for such long periods that there is a reasonable likelihood that an earthquake of the defined severity will occur during this period.

208. The design earthquake level for category 2 shall be defined on the following basis. The additional effort to protect the items against this level of earthquake shall be commensurate with the potential reduction in risk to plant personnel or the public due to the earthquake. Nationally established acceptable limits for the release of radioactivity shall be complied with. In many cases it may therefore be reasonable to choose SL-1 as the design basis for this category. Such an approach would

" Typical list of components or systems are given in Annex I, which describes categorizations used in some Member States.

9


minimize the need for plant shutdown, inspection and reiicensing, thus allowing the plant to continue to operate.

209. Nuclear power plant items not included in categories 1 or 2 should be designed for earthquake resistance in accordance with national practice for non-nuclear applications.

210. When, as the result of an earthquake, the collapse, falling, dislodgement or any other spatial response of an item is expected on the basis of analysis, test or experience to occur and could jeopardize the functioning of items in a higher category:

(1) Such items shall be classified in the same category as the endangered items;(2) Under the reference ground motion, the absence of collapse or loss of function

of the lower category items shall be demonstrated; or(3) The endangered items shall be suitably protected so they are not jeopardized.

Since only the structural integrity of items reclassified because of their potential to jeopardize higher category items needs to be assured, less rigorous seismic evaluation criteria may be used.

211. The inclusion of an item in seismic categories 1 or 2 shall be based on a clear understanding of the functional requirements which shall be assured for safety during or after an earthquake or after an accident not caused by an earthquake (see para. 207, item 2). According to their different functions, parts o f the same system may belong to different categories. Tightness, degree of damage (fatigue, wear and tear, etc.), mechanical or electrical function capability, maximum displacement, degree of permanent distortion and preservation of geometrical dimensions are examples of aspects which shall be considered (see paras 216-220).

COMBINATION OF EARTHQUAKE LOADS WITH OTHER PLANT PROCESS LOADING CONDITIONS

212. Plant process loads are grouped as follows:

L I: loads due to normal operationL2: additional loads due to anticipated operational occurrences L3: additional loads due to accident conditions.

213. Seismic loads should be calculated for the specific location of the item under consideration, taking into account the characteristics of the soil and plant structures, including the masses and stiffnesses and the distribution of equipment within the plant. Care should be taken to ensure that the controlling loading combinations are considered in order that maximum stresses in structural elements are identified and included.

10


214. For the seismic design, it is recommended that loads from earthquakes be combined with plant process loads as follows:

(1) For category 1 items, LI be combined with SL-2 loads;(2) For category 2 and uncategorized items, LI be combined with the loads from

the respective earthquakes according to paras 208 and 209;(3) For all items, L2 or L3 be combined with the respective earthquake loads if

the L2 or L3 loads are caused by the earthquake and have a high probabilityof coinciding with the earthquake loads or if the L2 loads occur sufficientlyfrequently, independently of the earthquake^.

(4) For category 1 items, relevant L3 loads should be combined with SL-2 loads, unless L3 is not correlated with SL-2 loads.

It should be noted that an L3 load for one group of items (e.g. the reactor coolant system) can be an LI load for another group of items (e.g. containment system or safety injection system).

215. For the seismic design o f equipment and civil structures, ambient environmental conditions and other natural phenomena such as floods or fires assumed as a consequence of the earthquake shall be taken into account. They could be identified on the basis of probabilistic considerations [1].

ALLOWABLE LIMITS FOR STRESS, STRAIN AND DEFORMATION

216. Behaviour limits for load combinations of sets of events, including the effects of SL-2 with LI or L2, and SL-1 or SL-2 with L3, should be the same as those adopted in related practices for L3 acting without an earthquake. In load factor design, behaviour limits are defined by variable load factors with set limits on stress, strain or deformation. This is in contrast to working stress design where variable limits are applied to stress, strain or deformation for a fixed set o f loads. Increasing allowable stress, strain or deformation limits has the same effect as reducing load factors.

217. Items whose active safety functions require stricter behaviour limits (e.g. control rod drive mechanisms) shall be designed within appropriate limits for load combinations, which include the SL-2 earthquake, to ensure their operability during or subsequent to the earthquake. .

218. The design margin for seismic classified items for LI and L2 combined with earthquakes up to the SL-1 level, when SL-1 is considered as a design basis, should

Typica) L2 loads induced by the seismic event could be toads created by tripping of the reactor or a pressure peak in the primary system due to a tripped turbine in a BWR with a smatt steam bypass capacity to the condenser.

11


be greater than the design margin according to para. 216. For both passive components (e.g. piping, support members) and active components (e.g. valves, pumps and control rod drives whose operation is important to safety), all behaviour due to applicable load combinations should stay within the appropriate limits.

219. With respect to behaviour limits for seismically classified items, earthquakes up to SL-1 shall be assumed to occur more than once in the plant lifetime. When cyclic analyses which include the SL-1 as design basis are performed, approximately five SL-1 earthquakes and one SL-2 earthquake should be used for design purposes, assuming ten peak amplitudes corresponding to each earthquake [2, 3].

220. The behaviour limitation principle for uncategorized items should follow applicable national codes.

3. SEISMIC DESIGN

GENERAL APPROACH TO SEISMIC DESIGN

301. Seismic input motion is normally defined in the free Held at: (a) the surface of the ground; (b) the level of the foundation; or (c) on bedrock (see Safety Guide 50-SG-S1 (Rev. 1), para. 508). A more specific identification of seismic input motion depending on its use in design is defined as a 'control motion' and 'control point' (see Safety Guide 50-SG-S8, Section 3.5).

302. In the early stages of the design of the plant, a preliminary arrangement of the main facilities should be prepared; this should subsequently be periodically reviewed to achieve the most suitable solution for the seismic design. All procedures for seismic design must be firmly based on a clear appreciation of the results of past destructive earthquakes and must adopt and realistically apply this knowledge. In this preliminary work, the following principles (paras. 303-305) should be taken into account for reducing earthquake effects on structures and components.

303. In general terms, seismic effects^ shall be minimized by:

(1) Locating the centre of gravity of all structures as low as practicable;(2) Selecting a plan and elevation that is as simple and regular as practicable;(3) Avoiding protruding sections (lack of symmetry) as far as practicable;

Items (t)-(4 ) reduce seismic forces and unwanted torsional or rocking effects.

12


(4) Making the centre of rigidity at the various eievations as close to the centre of gravity as practicable;

(5) Using antiseismic systems and devices (e.g. base motion isolators ).

304. To reduce undesirable differential movements between structures, consideration shall be given to building these structures, to the extent practicable, on a common foundation structure. In locating the plant, significant differences in soil properties below the foundation structure should be avoided. If spread footings are used, the need to connect these to adjacent foundation elements should be investigated. Design loads for such connecting elements are defined according to established practice.

305. The adoption of very simple layouts and connections to structures will facilitate the seismic analysis and improve the seismic behaviour of piping and equipment appended to buildings. Care shall be taken in crossing structural boundaries, (e.g. expansion or construction joints), in making connections between buildings or in bringing services to and from a building through underground conduits to avoid damage or failure due to differential movements. AH equipment shall be positively anchored unless a different approach can be justified.

CIVIL ENGINEERING STRUCTURES

306. Particular attention shall be paid to the following points in the design and design review of structures:

(1) The adequacy of the supporting soil (see Safety Guide 50-SG-S8);(2) The suitability of types of foundation supports or of different types of

foundations under interconnected structures (e.g. part of the foundation of one building being supported on piles or rock and part being set directly on soil should be avoided);

(3) A balanced and symmetrical arrangement of structural frames and shear walls to achieve optimum stiffness, load and weight distribution, with minimum torsional effects;

(4) The need to prevent collision between adjacent buildings as a consequence of their dynamic deformations (this phenomenon may also occur in weakly coupled structures);

(5) The adequacy of the connections of annexes and appendages to the main structure (see also item (4));

' Base motion isolators shoutd be used with caution to avoid increasing the retative movements or disptacements described in paras 304 and 305.

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(6) The need to ensure sufficient resistance of essentia! structural elements, especially to lateral shear forces;

(7) The need to ensure sufficient ductility and avoid brittle failure by shear or compression, for example by ensuring that there is an adequate amount of reinforcement steel, in particular enough hoop ties for columns (adequate confinement);

(8) The arrangement and distribution of reinforcing bars: too high a concentration o f bars may cause cracking of concrete along the lines of the bars;

(9) The need for the design of joints between structural elements and the anchorage o f items cast into concrete to ensure ductile failure modes (for example, anchor lengths must be sufficiently long to avoid pull-out and adequate transverse tie reinforcement must be provided) and, whenever practical, for connections between members to be made as strong and as ductile as the members they connect;

(10) An evaluation of the bending moments arising from the vertical forces and the lateral deformation due to the P-A effect of the earthquake on the structures when large deformations are permitted;

(11) The additional effect of groundwater buoyancy on the foundation;(12) The possibility of lateral sliding, during an earthquake, o f structures on

waterproofing (especially if wet);(13) The effect of 'non-structural' elements, such as partition walls, on structural

elements: cracks have sometimes occurred at column-beam connections as a result of in-plane forces in a partition wall and this may particularly occur at the highest floor where the beneficial effect of such 'non-structural' elements on the vertical loads is reduced;

(14) The detailed design of construction joints and thermally induced stresses in large, integrated, monolithic structures designed to resist differential earthquake motions;

(15) The effects of the transfer of forces in cases when the stiffness of a containment vessel is greater than that of the surrounding concrete structures, and they are interconnected or may interact so that the earthquake loads of the concrete structures may be transferred to the containment vessel; where, owing to the complexity of the interacting structures, it is too difficult to evaluate such forces, it is recommended as far as possible to decouple such structures above the foundation level.

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EARTH STRUCTURES [4]

307. The following safety related earth structures may be encountered at nuclearpower plant sites:

— Ultimate heat sinks — dams, dikes and embankments— Site protection — dams, dikes, breakwaters, seawalls, revetments— Site contour — retaining walls, natural slopes, cuts and fills.

These earth structures should be designed for the following seismic related effectswith adequate strength against failure:

(1) Slope failure induced by SL-2 vibratory ground motions, including liquefaction;

(2) Sliding of structures on weak foundation materials or materials whose strength may be reduced by liquefaction;

(3) Piping failure or seepage through cracks induced by ground motions;(4) Overtopping o f the structure due to tsunamis on coastal sites or seiches in

reservoirs, slides or rock falls into reservoirs or failure of spillway or outlet works.

Design methods and allowable values are subject to the criteria of each MemberState.

PIPING AND EQUIPMENT

308. With regard to equipment and piping supports:

(1) Care should be taken in the design of these supports to ensure that all jointsare designed to behave as assumed in the analysis of the support and transmitthe full range of loads determined in the connected members. In particular, if integral supports are used, they should be designed, manufactured and installed so as to minimize the danger that any unexpected failure or crack initiated in the supporting element propagates to the functional parts, such as the pressurized shell and the primary piping.

(2) Care should be taken in the design of devices for anchoring equipment, forexample the possible use of hook-shaped or end-plate anchor bolts, to ensurethat all potential forces and moments are fully evaluated and that anchoring materials are suitable for their purpose. It is o f particular importance to ensure that base plates are sufficiently stiff to avoid prying effects and that anchor bolts are adequately tightened to avoid rocking effects, lowered frequencies, increased response levels, higher-than-design loads and increased risk of loosening, pull-out or fatigue. Overdesigned or redundant anchors, preloaded to close to their yield point on installation, are therefore recommended.

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IMPROVEMENT OP THE VIBRATION RESISTANCE OF COMPONENTS

309. The following points should be taken into account to improve the resistance ofcomponents to vibration:

(1) Equipment support legs should be cross-braced unless their dimensions warrant departure from this recommended practice. In most cases, stiffness can be increased to avoid resonance; however, in some cases (e.g. for core internals for which it is difficult to avoid resonance by design modification), the vibration characteristics of the reactor building internal structure itself might be modified to prevent resonance effects. If systems are made stiffer for this purpose, particular attention should be paid to thermal stresses, other dynamic loads and differential motions of the supporting points, because of the increase in stress caused by high rigidity.

(2) It is important to avoid, as far as is reasonably practicable, resonance of equipment such as piping, instrumentation and core internals at the frequency of the dominant modes of supporting structures. In some cases, where the response of equipment, although significant, cannot in practice be reduced by other means, the damping of the system may be increased by suitable design modifications.

(3) To provide seismic restraints for piping and components and at the same time allow freedom for thermal deformations, dampers or motion limiting stops should be used. Excessive use of snubbers should be avoided. Realistic damping values to define seismic design input should be used, since overdesign for seismic loads can reduce design margins for thermal (restraint of free displacement) loads [5, 6].

(4) Particular attention should be paid to the possibility of collision between adjacent components, or between components and adjacent parts of a building, as a consequence of their dynamic displacement. It is also important to allow for flexibility of connections between such components, between components and building penetrations and underground connections to buildings, as well as between buildings.

The advantage of motion limiting devices or stops for piping systems is that normal thermal expansion is permitted, without the potential o f overrestraint, as with locked snubbers. In addition, motion limiting stops are less likely to fail when called upon to operate than inoperative snubbers (free to move but offering no resistance to dynamic or shock forces) as a result, for example, o f corrosion or of loss o f oil or broken shear pins.

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POSSIBLE MODES OF FAILURE OF NEW TYPES OF EQUIPMENT

310. It is important to recognize which part (or parts) o f a piece of equipment are essential for structural integrity and functional capability. This means that a systematic evaluation of the possible modes of failure related to earthquakes is important and has to be performed on hew types of equipment for which there is no relevant experience. Seismic qualification tests can identify the modes of failure and, for this reason, such tests are recommended. With more sophisticated computer simulation analysis techniques available, even the performance of 'active' equipment under earthquake conditions can be predicted with some confidence (e.g. pumps, valves and diesel generator sets). Use of such analytical techniques is recommended when considered applicable. However, it must be understood that even a high level of analytical sophistication still requires a number of assumptions and produces at best only an indicator; of seismic behaviour. Real seismic behaviour of a component, particularly with regard to functionality, can only be determined by test or experience of the actual or similar components in a strong motion earthquake.

FUNCTIONAL ASPECTS OF SEISMIC DESIGN

311. The designer should have a thorough and practical understanding of the functional requirements and potential modes of failure of a mechanism when considering its seismic design. It is also recommended that there be co-operation between the designer and the analyst in developing analytical models which adequately represent the behaviour of the structure and in interpreting the results.

EFFECTS OF VERTICAL GROUND MOTION

312. Vertical motions produced by an earthquake have the primary effect of cyclically increasing and decreasing the vertical loads on structures, as well as potentially amplifying the motion of floors on which equipment may be supported. Critical conditions can be reached during the phase in which load is decreased for those structures or pieces o f equipment whose stability depends on frictional or confinement effects (e.g. non-anchored pieces of equipment and vertical reinforced concrete structural elements — see item (7) of para. 306). The design shall take such loads into account in protecting plant items against failure.

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4. SEISMIC QUALIFICATION:ANALYSIS, TESTING, EARTHQUAKE EXPERIENCE AND

INDIRECT METHODS

401. Seismic qualification of items important to safety^ can be performed by the use of one or more of the following methods: analysis; testing; earthquake experience; or comparison. It is also possible to use combinations o f these methods or indirect means as shown in Pig. 1. Seismic qualification also generally includes structural integrity qualification as well as operability or functional qualification. Seismic qualifications are made directly on actual or prototype items, or indirectly on a reduced scale model, a reduced scale prototype or a simplified item, or by means of similarity where this can be established between a candidate and a reference item and direct qualification has been performed on the latter.

SEISMIC INPUT

402. The horizontal and vertical components of the earthquake motion should be considered individually or in combination as inputs for the analysis. When the components are used individually, the corresponding structural responses should be suitably combined. In some countries, horizontal seismic input is defined as a resultant, rather than as a component (see Appendix I). Dynamic input motions used to qualify items are conservatively yet realistically defined by either time histories or response spectra. In the case of response spectra, it is necessary to define not only the spectrum shape and the zero period peak ground acceleration, but also the duration of the motion as discussed in Safety Guide 50-SG-S1. In some countries, a constant acceleration input is used for a preliminary design or a simplified design, especially for items of lower importance, and is also used in the vertical direction [7],

DIRECT QUALIFICATION

403. Direct seismic qualification impiies that the actual or a prototype item is being qualified by either analysis, test or experience.

Testing or qualification of soils, with structures and foundation media, is outside the scope of this Guide. This subject is treated in Safety Guide 50-SG-S8.

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F/G. 7. &wwMfy !?f s'e;',s'w;'t' or ver!/:c<3fi'oM

QUALIFICATION BY ANALYSIS

404. Seismic qualification by analysis is generally applied to items which are unique and are of such a size or scale as to preclude qualification by testing. Civil engineering structures, tanks, distribution systems and large items of equipment are usually qualified by analytical methods.

GENERAL AND ANALYTICAL METHODS

405. A dynamic modal analysis, as well as a dynamic direct time integration analysis o f a lumped mass model or other models with many degrees of freedom, may be used. The modal analysis can be applied using either time histories or response spectra as input, as discussed in Appendix I.

406. Linear dynamic analyses are generally adequate for most items. Alternatively, non-linear dynamic analysis may be used where appropriate or necessary (e.g. structural lift-off or non-linear load dependent support or foundation properties).

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407. An equivalent static analysis of items may be performed using the peak of the applicable acceleration response spectrum as input for cantilever and other items supported at no more than two points along their long axis. For items supported at more than two points along their long axis, a factor of 1.5 times the peak of the applicable acceleration response spectrum is recommended, unless a lower value can be justified by comparison with dynamic analysis results. The resultant accelerations shall be applied to the mass distribution of the structure along the orthogonal axes of the item. Resultant seismic inertia forces, moments or stresses thus determined may be combined by calculating the square root of the sum of the squares (SRSS) or other appropriate means [8, 9].

408. A dynamic analysis is recommended for seismic category 1 structures to determine seismic response, displacements or loads. Both the time history method and the response spectra method are acceptable. An equivalent static load method, without a parallel dynamic analysis, may be used only if its conservatism is satisfactorily demonstrated.

GENERAL MODELLING TECHNIQUES

409. Typically, nuclear power plant items can be modelled in one of the following ways, according to their structural characteristics: lumped mass models, one dimensional, axisymmetric, two dimensional or three dimensional finite element models in two or three dimensional space. Rigid mass models with spring support may also be used to represent foundation-structure interaction, as discussed in Appendix II.

410. Consideration should be given to translation, rocking and torsional modes, individually or coupled, as appropriate.

411. In the lumped mass or finite element formulation, consideration should be given to the adequacy of the number and distribution of masses chosen to represent the item (see Appendix II).

CIVIL ENGINEERING STRUCTURES

412. In the modelling of the building or large ground mounted tank structures, the soil-structure interaction should be taken into account (see Appendix II).

413. Subsystems, which may include the inertial effect o f the liquid contained in tanks and pools, should be taken into account in the model of the structures (see Annex II).

414. Adjacent buildings or components on the same foundation structure should be included in the same model when relative motion of the foundation is included in the analysis.

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415. Lateral earth pressures induced oh underground portions o f structures or foundations by ground motion should be evaluated. These procedures usually provide values of the dynamic pressure increment induced by the earthquake; this increment is then added to the applicable static lateral earth pressure. Simplified procedures may also be used to calculate this dynamic pressure increment [9], These simplified procedures, however, do not take into account the effect of adjacent structures on this pressure increment; the effect may, in some cases, be significant, depending on the plant layout.

416. The liquefaction potential of saturated granular soil layers during the design basis vibratory ground motion should be evaluated for the limiting earthquake, SL-2 or SL-1 ground motion, and the existence of appropriate safety margins should be demonstrated [10, 11]. Appropriate measures to prevent liquefaction in these layers should be undertaken if the results of the evaluation indicate a lack of the required safety margins [4]. These measures may include in-place treatment (e.g. compaction or grouting), dewatering, excavation and replacement with properly compacted fill or permanent surcharging, or a combination of these measures. Field verification to ensure attainment of the desired improvements in soil characteristics should be made.

417. The potential for a loss of the bearing capacity of the soil during or after an earthquake should be investigated [12, 13]. Normally, the bearing capacity factor of safety under static loading conditions is reasonably high. Therefore, unless there is a significant reduction in the strength of the soil as a result o f seismic loading (e.g. liquefaction in granular soils, or reduction in strength due to development of high shear strains in softer clays), the bearing capacity factor of safety under the combined static and seismic loads is expected to remain adequately high. If significant reduction in the bearing capacity is identified, such reduction should be accounted for in design.

418. Potential settlements (especially differential settlements) caused by shaking may be of significance, particularly in non-uniform soil conditions where they should be evaluated [14, 15]. Judgement should be exercised in estimating settlements in the vicinity of and beneath the structures, on the basis of the results o f the free field calculations.

419. Foundation analysis for structures is also discussed in Safety Guide 50-SG-S8.

420. The following earthquake effects on buried long structures (e.g. buried pipes,ducts and well casings) should be taken into account:

(1) Deformations imposed by the surrounding soil during the earthquake;(2) Differential displacements or loads at end connections to buildings or other

structures;(3) Effects of contained fluids (impulse loads, hydrostatic pressure and sloshing

effects).

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421. AH effects can be evaiuated by appropriate modelling o f the soil-structure complex [15], However, simplified conservative methods can also be applied [9].

422. Modelling of concrete structures is usually undertaken assuming uncracked section properties. However, when the dominant frequency of the input motion falls within the range of fundamental frequencies of the structure, the cracked section properties should also be evaluated. Assumptions as to the stiffness of shear wall type structures can be found in Ref. [9].

423. The adequacy of shake or rattle space in structural joints between adjacent structural parts or between adjacent buildings should be checked, taking into account the need for an adequate safety margin.

MECHANICAL AND ELECTRICAL COMPONENTS

424. Mechanical and electrical components are usually represented in the analysis by a single mass or a multimass system attached to the supporting building. As discussed in Appendix II, their dynamic coupling to the main building can usually be neglected.

425. Either dynamic methods of analysis, as discussed in para. 405, or an equivalent static analysis method, as discussed in para. 407, may be used.

426. A variation of the static method discussed in para. 407 has been used in someMember States. Here, the maximum acceleration of the design response spectrum in the frequency range between 0.5ff and 2 .Off (when f? is the fundamental frequency of the equipment) is taken as the design acceleration. Again, a suitable factor, typically 1.0 to 1.5 times such a maximum, depending on the number of supports, has been applied. A further variation of this method is to determine, statistically, the percentage of total stress attributable to seismic loading and analyse the horizontal type vessel and piping for normal dead weight loads, allowing a conservative margin for the dynamic earthquake components [16].

427. The calculated stresses and reaction loads in the equipment and equipmentsupports can be a direct output of either dynamic or static analysis. It should be noted that electrical equipment, exclusive of anchorage or support, is generally evaluated only for operability by testing or using experience data. For this reason, analysis of these components is limited to assuring elastic response of the electrical cabinet, panel or rack structure and evaluation of support or anchorage loads.

428. The modelling of equipment is typically divided into several categories as shown in Fig. 11.2 of Appendix II.

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429. For components not modelled together with-the supporting structure, the input for analysis is the floor response, expressed in terms of either design floor time histories or design response spectra. In some Member States, an equivalent static input is used in the vertical direction.

430. The design floor response spectra, typically used as seismic input for equipment, can be obtained on the basis o f structure response to design basis ground motion. Time history motions which can be shown to generate response spectra at least as conservative as the design ground motion response spectra can be used for this purpose [17].

431. Alternatively, direct methods may be used to calculate design floor response spectra [8, 9], It is recommended that before the direct method is used, it should be ascertained for representative cases that the conservatism of the results is comparable to that provided by the time history solution.

432. Once the floor response curve has been obtained for a particular floor of a building, a critical review of the calculated floor response spectra should be made to assess their reasonableness (on the basis of sound engineering judgement), the relationship between the vibration characteristics of the building and the supporting foundation materials, and the shape of the floor response curve.

433. The calculated floor response spectrum should be broadened to account for possible uncertainties in the evaluation of the vibration characteristics of the building components. Alternatively, segments of a broadened spectrum can be used sequentially to analyse the components. Examples of current procedures are given in Ref. [2j and Section 3.7.2 of Ref. [8], For systems having closely spaced frequencies, the use of such segmented response spectra will avoid overconservatism.

434. Consideration should be given to the modification of floor response spectrum input for equipment attached on very flexible structural members (vertical amplification due to floor flexibility) and, when significant torsional motion of the building is present, for those items located away from the centre of shear of the building, when the centre of shear and centre of mass are significantly different and such a difference has not already been considered in the modelling of the building structure.

435. The design flodr time histories can be obtained from the structural response of the building to design basis ground motion. Uncertainty effects, similar to those associated with response spectra, should be considered in analyses in which the floor motion time histories rather than the response spectra are used by altering the time scale to define the time history motion (see Section 3.7.1 of Ref. [8] and Refs [9, 17]).

436. The damping factors used in the analysis of equipment should be based on field testing and experience, as discussed in Appendix II. The quantity of insulation, the

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size, location and the number o f support gaps, the frequency of response, and the use of elastopiastic or energy absorbing support devices may all have an effect on the damping to be considered in the design of components.

437. Equivalent linear methods have usually been used to represent the response of elastopiastic support devices.

438. For vessels and tanks which contain liquids, sloshing and impulse load effects, including frequency effects, should be considered. The effects o f liquid motion or pressure changes on submerged structures should also be considered, as discussed in Annex I. These effects may involve hydrodynamic loads from the fluid, the reduction of functional capability (e.g. loss of shielding efficiency of fuel pools or disturbance of instrument signals).

439. The operability of active components may be analysed also when their potential failure modes can be identified and described in terms of stress, deformation (including clearances) or loads. Otherwise, the use of testing or earthquake experience is required for the qualification of active components.

DISTRIBUTION SYSTEMS

440. This section covers the seismic analysis o f piping, cable trays, conduits, tubing and ducts and their supports. Modal response spectrum analysis is often used for the seismic design of large bore (Do > 6.0 cm) category 1 piping, but the static method discussed in paras 407 and 426 is often used for the analysis of small bore and category 2 or other category piping. Spacing tables and charts based on generic analysis or testing are also used in the evaluation of small bore and category 2 or other category piping and are typically employed to evaluate cable trays, conduits, tubing and ducts. Simplified analytical or design procedures based on earthquake experience data may also be used (see para. 472 for the application to cable trays). In some Member States, modelling and analysis of cable trays and ducts, similar to that performed on large bore piping systems, has been conducted.

441. When distribution systems are connected to two or more points having different movements and applicable response spectra, the use of a single response spectrum to define input at a particular support point requires some modification. The usually accepted method is to apply at all the support points an envelope spectrum of the inputs at each support. The results of this method are usually rather conservative. Other methods which consider independent support motion in conjunction with modal analysis may be used where design conservatism can be demonstrated [18].

442. In addition to inertia effects, careful consideration shall be given to the effects o f differential motions between supports since earthquake experience has demon

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strated that this phenomenon can be a major contributor to seismic induced failure of piping systems.

443. A static analytical method frequently used with particular application to small bore piping of 6 cm or less in diameter is to apply the peak of the applicable design floor response spectrum acceieration, multiplied by a suitable factor, to the mass distribution of the piping system in order to determine an inertial force to be applied to the system. A factor o f 1.5 has typically been used in the past (see Section 3.7.2 of Ref. [8]) for systems supported at more than two points. However, lower values may be used provided that such lower values can be shown to give conservative answers when compared to dynamic (time history) analysis results.

444. Methods commonly used for modelling piping systems include:

— One dimensional finite element in three dimensional space— Transfer matrix method, with the matrix coefficients calculated according to

beam theory [16],

445. The flexibility or stiffness of elements of piping systems such as elbows, tees and spring hangers shall be considered in the model. Constant spring hangers usually have no stiffness effect but they behave as a constant external load on the system throughout their travel range. Variable spring hanger effects are usually ignored in the seismic analysis of piping. If there is a pump or a valve in the piping system, its contribution to the response should be evaluated. All additional masses, including eccentricities such as valve operators, pumps, the liquid inside the pipe and thermal insulation, should be taken into account.

446. It should be understood that the response of distribution systems to earthquake excitations tends to be quite non-linear. Therefore computations of stresses and support reactions by means of linear elastic analyses are at best an indication of stresses and support loads suitable for comparison with allowable behaviour criteria to determine design adequacy but are not accurate measures of actual stresses and support reactions. For this reason, nominally fixed supports for distribution systems with some limit on deflections may be considered rigid for modelling purposes.

SEISMIC QUALIFICATION TESTING

447. Another method of direct seismic qualification of items is the testing of the actual item or prototype [19, 20]. If the integrity or functional capability of an item cannot be demonstrated with a reasonable degree of confidence by analysis, a test may be needed to prove or to assist directly or indirectly in qualifying the item.

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448. The types of testing can be iisted as foHows:

— type approval test (fragility test)— acceptance test (proof test)— low impedence test (dynamic characteristic test)— code verification test.

449. Test qualification of seismic category 1 items is required when failure modes cannot be identified or defined by analysis or earthquake experience. Direct qualification by testing employs type approval and acceptance tests. Low impedence (dynamic characteristic) tests are normally used to identify similarity or verify or help develop analytical models. Code verification tests are used to develop generic verification of analytical procedures which typically use computer codes. The methods of testing depend on the required input, weight, size, configuration and operating characteristics of the item, plus the characteristics of the available test facility. Type approval and qualification acceptance tests include:

— shake table, either one, two or three dimensional— hydraulic actuator— electric actuator— mechanical actuator (unbalanced mass type)— impact hammer— blast.

Low impedence and code verification tests typically use the last three of these methods. In addition, they may also include self-excitation or ambient vibration excitations.

450. A meaningful test, performed with the purpose of assessing the integrity or functionalcapability of an item, requires that the conditions existing for this item in the plant during an earthquake are correctly or conservatively simulated or that any departure from them will not significantly influence the result. Among these conditions, the most important are:

— input motions— boundary (support) conditions— environmental conditions (e.g. pressure and temperature)— operational conditions (if functional capability has to be assessed).

451. The concept of the testing procedure is based on subjecting the item to conservatively derived test conditions in order to produce effects at least as severe as those o f the design basis seismic event concurrent with other applicable operating or design conditions.

452. The functional and integrity testing of complex items, such as control panels containing many different devices, may be performed on the prototype of the item

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itself or on individual devices with seismic test input scaled to allow for the location and attachment of the device within the item or on the item.

453. Account should be taken of such effects as radiation and ageing, or other conditions which may cause deterioration or otherwise alter the characteristics of the item during its in-service life.

454. Low impedence (dynamic characteristic) tests are usually made on items in situ. Items are typically tested by:

— mechanical actuation— impact— blast

and other low energy exciters as well as ambient excitation. These tests cannot be used for direct seismic qualification of the item but can be used to define dynamic, including support, characteristics which can then be used in analysis or other tests to qualify the item.

455. The type approval (fragility) test is typically used for standard electrical components and some mechanical components. In this case, design margins to failure, damage or non-linear response and identification of the lower bound failure mode can be obtained. Such testing is typically carried out by means o f a shaking table. The fragility test is useful for finding unexpected failure modes or potential malfunctions, because the test conditions can cover a wider spectrum of loading than those required as a design basis. .

456. The acceptance (proof) test is also used for electrical and mechanical components to demonstrate seismic adequacy. It is typically performed by manufacturers to demonstrate compliance with procurement specifications and does not give data relative to seismic design margins or failure modes. Such testing is typically carried out by means of a shaking table.

457. The code verification test is important for reliable analytical work. Computer codes should be verified before their application by using an adequate number of test results or results obtained from other appropriate computer codes or analytical procedures. The use of a number of test results which cover the range of interest and have been correlated with analytical results is strongly recommended.

458. Seismic tests may be performed on the item itself or on a full scale model or, where appropriate, oh reduced scale models. However, for qualification purposes, it is strongly recommended that the component itself or a full scale model without any simplification should be tested; if there is no other practical alternative, the careful use of a reduced scale model may be permitted for qualification purposes. Such tests include:

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— functional tests intended to ensure the required safety function of the component or absence of transient malfunctioning during and after an earthquake; and

— integrity tests aimed at proving the mechanical strength of the component.

When reduced scale testing is performed, similarity requirements associated with indirect methods of seismic qualification, as discussed in para. 471, must be considered.

459. The number of repetitions of testing or cycles of loading per test is dependent on the application, but accumulation of various types of damage associated with fatigue or ratchetting phenomena should be taken into account for the evaluation of the result and to permit qualification for the operating life of the item.

460. For components whose functional capability has to be demonstrated by testing under earthquake conditions, excitation in one direction at a time can be considered adequate if one of the following conditions applies:

(1) The component design review and visual inspection or exploratory tests clearly demonstrate that the effects o f excitation in the three directions on the component are sufficiently independent of each other.

(2) The severity of shaking table tests can be increased in such a way as to take into account the interaction effects from simultaneous excitation in the three directions. (For example, the amplitude of excitation can be increased in one direction to envelope the response in another direction due to coupling effects.)

Otherwise, simultaneous multidirectional input should be applied [20],

461. If random vibration or multifrequency input motion are used, detailed recommendations such as those listed in Section 3.10 o f Ref. [8] should be followed. The duration of the input motion should be decided on the basis of anticipated earthquake duration (see Safety Guide 50-SG-S1).

462. A sinusoidal or sinusoidal beat motion can be used for qualification testing of stiff systems at a frequency significantly lower than the first mode eigen frequency o f the system. This results in a test response spectrum which envelopes the reference response spectrum required to qualify the item. If no adequate shaking device is available, a sinusoidal motion can be used at resonance to obtain the necessary qualifying level of response of the item.

463. When the system has one or more vibrational resonances in the frequency range of interest, the test input motion should have a response spectrum not less than the required design response spectrum. This can be achieved by using a time history input whose test response spectrum envelopes the reference response spectrum required to qualify the item.

464. When the eigen frequencies of the item are well separated, independent tests can be made, for example with suitably scaled sinusoidal input at the given frequency

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with a half-sine or other time envelope of interest. 'Suitably scaled' means that the amplitude of the test spectrum at the eigen frequency is higher than the amplitude

of the required spectrum. However, it is recommended that tests be made with two or more time histories or natural time histories whose response spectra are not less than the required design spectrum. The use of several different time histories helps to overcome the deficiencies which could arise from the peculiarities of a single time history.

465. Natural frequencies and other vibration characteristics of the components may generally be assessed by a low impedence vibration characteristic test (for which a low level — 0.01 to 0.05g — input can be used).

466. It should be noted that the results o f the low impedence or excitation level test may be different from those of the test carried out under higher seismic levels for non-linear systems. To be of use in seismically qualifying equipment, low impedence tests require the equipment response to be essentially linear up to potential failure mode levels of excitation so as to be able to identify design margins.

467. In general it is necessary to establish functional requirements for active items (i.e. those which move or otherwise change state) in advance, as part of the test procedure. Most active items are required to perform their active function after the earthquake excitation has passed. However, if they have to perform such active functions during the earthquake excitation or during potential aftershocks, this must be considered in establishing functional test requirements. Care must also be taken that functionality tests are consistent with the required safety functions in service. For example, the lighting of an indicator lamp as a result of relay chatter during an earthquake would be inappropriate for cases where a 20 ms opening of the relay would be required for the circuit to change state.

468. Functional requirements for a digital computer used for control or data evaluation are of particular concern. The seismic resistance of such equipment is very complicated and the detection of malfunction or failure may be difficult.

SEISMIC QUALIFICATION BY EARTHQUAKE EXPERIENCE

469. The direct seismic qualification of items by the use of experience from strong motion seismic events has seen limited but growing application. It has only been within the past ten years that data from strong motion earthquakes have generally been collected in the detail and quality necessary to provide the information required for direct application to individual items. Such direct qualification requires that seismic excitation o f the item at its point o f installation in the building structure effectively envelopes the reference or required seismic design input motion. It also requires that the item being qualified and the one which underwent the strong motion

29


earthquake be the same model and type or have the same physical characteristics and have similar support or anchorage characteristics. In the case of active items it is also necessary in general to show that the item in the earthquake performed the same functions during or following the earthquake, including the potential aftershock effects as would be required of the category 1 item. In general the quality and detail o f the information,used to directly qualify individual items on the basis of experience data should not be less than those required for direct qualification by analysis or testing. As is the case for direct qualification by analysis or testing, earthquake experience may also be used as the basis for qualification by the indirect method.

INDIRECT METHOD

470. The indirect method of seismic qualification employs the analysis, test or earthquake experience methods applied to the direct qualification of the reference item. The indirect method involves establishing the similarity of a candidate item to a reference item previously qualified and thereby seismically qualifying the candidate item.

471. Similarity requires that the seismic input to the candidate item essentially envelopes the reference or design requirement for that item and that the seismic input to the reference item also equals or exceeds that required for the candidate item. Input to scale models of items shall consider proper similitude relationships. Similarity also requires that the physical and support conditions, the functional characteristics for active items and the requirements of the candidate item closely resemble those of the reference item. Guidance on the limits of similarity between reference and candidate items is given in Ref. [21].

472. To some degree, large quantities of earthquake experience data, in particular those applicable to the seismic qualification of distribution systems, have been used to justify simplification of the analytical evaluation and seismic qualification of such systems. The use of a factor of three times the dead weight capacity, with respect to normal behaviour criteria limits, is applicable to cable trays with ductile supports (which permit large lateral movement without failure). Seismic qualification of cable trays is an example of a simplified analytical evaluation based on earthquake experience data.

473. Because of the large numbers of potential seismic induced interactions between structures, equipment and distribution systems and the importance of adequate anchorage and support of structures, equipment and distribution systems, it is recommended that all seismically qualified items in the nuclear power plant be walked down by qualified seismic design and earthquake experienced structural engineers prior to operation to ensure that the 'as installed' items are capable of withstanding the design basis seismic effects without loss of structural integrity, with anchorage and seismic interaction effects taken into account.

30


5. SEISMIC INSTRUMENTATION

INTRODUCTION

501. Seismic instrumentation is installed at a nuclear power plant for the followingreasons:

(1) To check whether or not the design response spectra (i.e. free Held response spectra, floor response spectra, and response spectra associated with a number of selected items) corresponding to SL-1 and SL-2 levels have been exceeded, so that the need for post-earthquake inspection and plant shutdown can be evaluated.

(2) To collect data on the dynamic behaviour of structures, systems and components of the nuclear power plant and to assess the degree o f applicability o f the analytical methodology used in the seismic design and qualification of the buildings and equipment.

(3) To provide triggering mechanisms for alarms or for shutting down the plant (in some Member States).

TYPE, LOCATION AND NUMBER OF INSTRUMENTS

502. A minimum amount of seismic instrumentation should be installed at each nuclear power plant site as follows [22-24]:

(1) One triaxial strong motion recorder installed to register the free field motion time history;

(2) One triaxial strong motion recorder installed to register the motion of the base mat of the reactor building;

(3) One triaxial strong motion recorder installed on the most representative floor of the reactor building.

503. Additional seismic instrumentation should be considered for installation at sites having an SL-2 free field acceleration equal to or greater than 0.25g.

504. The instruments should be set to be triggered at levels of motion consistent with the seismicity of the site area.

MAINTENANCE OF INSTRUMENTATION

505. The seismic instruments installed at the nuclear power plant should be calibrated and maintained in accordance with written maintenance procedures.

31


DATA EVALUATION

506. Evaluation of data collected from the seismic instruments should be made in accordance with prescribed written procedures. The data should be considered as a part o f the records of the nuclear power plant and the period of retention should be determined on this basis.

32


Appendix I

1.101. The followingmethods are currently used to calculate the maximum relative displacements, relative velocities, absolute accelerations and maximum stresses during an earthquake; time history; response spectrum; and equivalent static method.

1.102. The standard dynamic methods for major structures, systems and components are the response spectrum and time history methods.

1.103. The equivalent static method can be used to obtain conservative results for items which have been correlated with dynamic analysis results [25].

1.104. In the response spectrum method, the maximum response of each mode is calculated by direct use of the design response spectrum. The maximum response in each direction is usually taken as the square root of the sum of the squares of each modal response. For closely spaced modal frequencies, a conservative procedure is to take the sum of the absolute values of each closely spaced modal and rigid response [26]. Missing mass as a function of the modelling detail, and cut-off of frequencies and modal participation used in the analysis must also be taken into consideration [27],

1.105. Responses due to input acceleration in the three different component directions can be combined by taking the square root of the sum of the squares of individual responses. In some countries, horizontal input motion is defined as a resultant in one of two orthogonal directions and is combined with vertical motion to determine the worst case response.

1.106. In the time history method, the response of the system is calculated as a function of time directly or after a transformation to modal co-ordinates. The input motion may be natural or artificial time histories of acceleration at ground level or a specific floor level, suitably chosen to represent the design response spectrum.

METHODS OF SEISMIC ANALYSIS

33


Appendix H

11.101. Nuclear power plant structures may be very complex and a single complete model of the entire structure would be too cumbersome or ill conditioned. Thus, the first step in the analysis o f a plant is to identify the substructures by defining main systems and subsystems.

11.102. Major structures that are considered in conjunction with foundation media to form a soil-structure interaction model constitute the main systems. Other structures, systems and components attached to the main systems constitute the subsystems.

DECOUPLING CRITERIA

II. 103. Certain criteria have to be used to decide if a particular subsystem has to be taken into account in the analysis of the main system. Such decoupling criteria are obtained by putting some limits on the relative mass ratio between the subsystem and the supporting main system, limits which have to be more severe when there is a possibility of resonance between the subsystem and the main system (see Section 3.7.2 of Ref. [8] and Ref. [28]).

II. 104. If the decoupling criteria are not satisfied, a suitable model of the subsystem should be included in the main system model. For a subsystem having all its resonant frequencies (the flexibility of the support being taken into account) higher than the amplified frequencies (15 to 30 Hz for usual design earthquakes), only the mass needs to be included in the main system model.

11.105. For detailed analysis of subsystems, if necessary, the seismic input, including differential support or attachment motion, can be obtained by the analysis of the main system. When coupling is considered, it is essential that the model of the subsystem be included in the analysis of the main system and have at least the same natural frequencies and modal masses as the detailed model of the subsystem in the frequency range of interest.

MODELLING OF SYSTEMS OR SUBSYSTEMS

11.106. The stiffness and mass characteristics of the structural systems should be adequately incorporated in the analytical models.

MODELLING TECHNIQUES

INTRODUCTION

34


II. 107. More than one model might be needed when there is some doubt about the behaviour of some part o f the structure. A sensitivity analysis which had convergence towards the same results and which varies the number and size o f the elements or masses used can provide a basis for this decision; it can also assist in the choice of the size and number of finite elements if this modelling technique is used. Models might also be validated by testing in order to resolve possible uncertainties.

Lumped mass method

11.108. The lumped mass method consists of lumping the masses (and inertial moments, if necessary) at some adequately chosen points (e.g. floor levels in a building) and determining the stiffness (or flexibility) coefficients by a static study of all the single movements corresponding to the selected degrees of freedom. This method is described in Refs [29-31].

11.109. Selection of an adequate number of degrees of freedom is often straightforward, as for example in the calculation of a conventional building with floors. In other cases, for shell or beam type structures, this selection is not obvious and depends in fact on the number of modes needed for the seismic analysis. The requirement that the number fof degrees of freedom be at least twice the number of modes is sometimes given; such a rule has to be used with care (see item II. la (4) of Section 3.7.2 of Ref. [8]). A practical way to ensure that sufficient modes (missing mass) are included in the analysis is to add a rigid body or zero period acceleration mode which corrects for the higher modes that may otherwise not be included in the evaluation. It also ensures, within the limits of the finite element model, that full reactions at supports are achieved [27, 32].

Finite element method

11.110. The finite element method is described in detail in many publications (e.g. Refs [33, 34]). Various types o f elements can be used: beam elements (for frames, pipes or long slender cylindrical structures); axisymmetric shell elements, which can account for non-axisymmetric loadings or displacements by a Fourier representation on the azimuthal variable (for containment buildings, pressure vessels, or other axisymmetric structures); shell and plate elements (for complex shell structures or box-type buildings); and three dimensional elements (for thicker wall items).

II. 111. The mesh size to obtain accurate results depends on the type of finite element used. A general requirement is that the shape of the deformed structure be well represented by the set o f elements chosen to model it.

35


Transfer matrix method

11.112. The basis of the transfer matrix method is that for some structures (pipes, axisymmetric shells and, more generally, monodimensionally modelled simple structures), it is possible to calculate analytically for each frequency the transfer matrix between the vector state (forces, moments, displacements and rotations) at a point and the vector state at the next point. Then, by successive matrix multiplication, it is possible to obtain the frequency dependent matrix of the whole system, including boundary conditions. This allows the determination of the natural frequencies, mode shapes or response to sinusoidal inputs, or, by a Fourier transformation, the response to more general inputs [16, 35], While there is no mesh size problem, care must be

M

Lumped mass or rigid body model

Ms<<< Ri

M;< <K R;

M, <R-

/ / / / / / / / / / / / / /

(b)

Lumped mass mode)

(c)

Beam or one dimensions finite element mode)

Ms< )Rs

M;<R,

<'Ri

Ke/77T 777

R,tRz

<R-

(d)Kv,

Beam or one dimensional finite element on soil springs model

7 * 7 7 * 7 7*7* 7 * 7

(e)

Two dimensional finite element mode) representing building and soil structure interaction

Two dimensional finite element or axis symmetrical building model

(g).

Three dimensional finite element model of foundation and building structure

K : 6 x 6 stiffness matrix : Horizontal spring stiffness

Ke : Rotational spring stiffness : Vertical spring stiffness

F/G. /A 7. Examples o / varioMS &MiMirtg modeis associated with dynamic or static analysis.

36


77777777777

(a)Lumped mass model

(b)

Lumped mass model with flexible foundation

(c)Beam or one dimensional finite element model

(d)

Two dimensional finite element or axis symetric equipment mode)

(e)

Combined two and one dimensional finite etement mode)

(f)Three dimensional thick walled vessel mode)

F/G. //. 2. Ecawp/M o / van'oM-s mode/^ aMocfafed dynamic or ,sfa;;c a^a/y.sM.

M

(a) Lumped m ass mode!

S: Support of static nodeM: Mass nodeK: 6 x 6 stiffness etement

(b) Beam or one dimensions) finite etement mode!

* Structure nodeo Mass node

F/G. //. J. o van'om d['jfr;' M;;o/! .system aMociafed w M dynaw;c or iMU'c

37


exercised in the use of this method because of the possible accumulation of errors in the multiplication process. A sensitivity analysis which varies the number and size of the elements or masses used can provide the basis for this decision; it can also assist in the choice of the size and number of finite elements if this modelling technique is used. Typical models for structures, equipment and distribution systems are shown in Figs 11.1-11.3.

DAMPING

11.113. Damping is applied to the analysis o f nuclear power plant items to obtain general or more exact agreement between the observed response of items to cyclic loads arid the response predicted by linear analysis. The damping used in the analysis includes hysteretic energy loss in materials in load deformation cycling plus the effect of small non-linearities due to changes in structural stiffness (concrete cracking) and boundary conditions (joint slippage and small support gaps) and may include the effects o f energy loss due to impacts as well. Damping values are normally determined experimentally [36, 37],

II. 114. The damping used in the analysis to qualify design should be conservatively yet realistically defined. (For example, experimentally determined damping for a particular type of material and structural system might range between 2 and 10%. In such a case, a conservative yet realistic value of damping to be used in the analysis to qualify a new design might be 4% .) Typical design basis damping values are given in Refs [38-40],

11.115. Damping values used in seismic analysis to verify existing designs rather than qualify a new design tend to be somewhat less conservative or more mean or median centred. In the example noted in para. 11.114, a damping value of 6% might be selected when design verification rather than qualification is being performed [41],

11.116. For each of the modelling methods described in this Appendix, knowledge o f the damping characteristics o f the main systems and subsystems is required. The way in which damping needs to be expressed depends on the method of solution used to evaluate the response of the model (see Appendix I). In the response spectrum, or time history modal analysis methods, damping is most conveniently expressed in terms of modal damping coefficients as a function of the percentage critical damping. In the time history direct integration method, it is necessary to construct a damping matrix. The commonly used procedure is to express the damping matrix as a combination of the mass and stiffness matrices of the model as follows [8, 9]:

[C] = a[M ] + <8[K]

where [C], [M] and [K] are the damping, mass and stiffness matrices.

38


11.117. Modal damping factors for composite structures can be obtained by the energy weighting technique [9].

11.118. The most important contribution to foundation-structure damping is energy dissipation within the ground. The actual radiation damping for nuclear power plants associated with lumped mass or one dimensional finite elements and spring soil- structure interaction models may be as great as 20-40% of critical damping, but lower values in the range of 10-25% are used in design practice. However, it is important to recognize that the damping of supported structures is lower than those values (2-7% ). If the structure is rigid, the radiation damping value depends on the mode of vibration of the structure and the coupling of the significant modes and the severity of the ground motion. If the structure is not rigid, it may be necessary to adopt an interaction model which couples the basement with its flexibility (deformation) taken into account and the soil represented by an elastic half-space. It is recognized that the value of damping that is used in the seismic analysis o f structures will require conservatively applied engineering judgement. Variation of damping factors with the frequency and amplitude of motion may be taken into account if warranted by experimental data [42].

39


Annex I

EXAMPLES OF U ST S OF MAIN SEISMIC CATEGORY 1 AND OTHER CATEGORY ITEMS

The foHowing are typical Member State positions on seismic design categorization or classification.

1-1. CANADIAN POSITION — CANDU REACTOR

Table 1-1 gives a list of CANDU systems and their earthquake design classifications.

TABLE 1-1. LIST OF SYSTEMS TO BE QUALIFIED

SystemEarthquake

level'

and sVrMCfMre.s

Reactor building DBE

Airlocks DBE

Turbine buitding DBE

Service buitding DBE

Secondary contro] area DBE

Spent fuel bay DBE

Emergency water and power DBE

Supply structures

Rgacfor

Fuel channel assemblies DBE

Calandria and end shields DBE

Reactivity control units DBE

41


TABLE 1-1. (cont.)

SystemEarthquake

level"

Pnm ary fra^.spor: syi!em

Main circuit and all subsystems DBE

Steam generators DBE

Shutdown cooling system DBE

/4MLXi7i'ary sysfewM

Dousing system DBE

Emergency core cooling system DBE/SDE

Emergency water supply system DBE/SDE

Liquid injection shutdown

system (SDS2)

DBE

Fuel changing DBE

Spent fuel transfer and storage DBE

JfeafM afM? wafer .sy^/e^!

Main steam supply DBE

Steam generator pressure DBErelief system (main steam safety valves)

Fuel DBE

Maw yee^w a^r c;rcM;'f DBE

R ecfnca /

Emergency power supply system DBE

Lighting DBE

Cables, conduits and cable trays DBE

42


TABLE 1-1. (cont.)

SystemEarthquake

level"

/fMfrHfHenfaMOM and con?ro/

Failed fuel monitoring DBE

Reactor regulatory system DBE

Computer system (watchdog) DBE

Secondary control area equipment DBE

Shutdown system No. 1 DBE

Shutdown system No. 2 . DBE

Process instrumentation systems DBE/SDE

Common procgMM and ggrwca?

Fire protection system DBE

Containment isolation DBE

Instrument air (inside reactor building) DBE

Solid radioactive waste DBE

" DBE is similar to SL-2 level earthquake; SDE is similar to SL-1 level earthquake.

1-2. FRENCH POSITION — PRESSURIZED WATER REALTOR

The French regulations provide that the safe shutdown of a nuclear plant reactor, the fuel cooling and the radioactive product confinement be achieved in the event of probable earthquakes on the plant site considered, and after such earthquakes.

They also define acceptable methods to determine all the motions to which the structures and civil works classified as 'seismic' will be subjected, on the basis of the motions considered as well as their corresponding load levels, in order to allow the design and checking of:

— the resistance Of the civil works subjected to loads resulting from earthquakes and other actions combined with earthquakes;

— the appropriate behaviour of the equipment connected to these civil works under combined action conditions involving earthquakes.

43


The evaluation of the motions and loads included in the structure by earthquake actions takes account of:

— the design seismic motions corresponding to the design spectra mentioned in basic safety rule 1.2.c;

— the foundation soil mechanical characteristics determined according to basic safety rule 1.3.c;

— the characteristics of the civil works, including those of the foundation system and the possible interactions between structures;

— the characteristics of the equipment connected to these structures, in so far as they influence motions.

As a result o f the standardization policy followed in France, which implies the use of standardized spectra, the computed SMS (seisme majore de securite or overvalued safety earthquake) spectra are used for each site to verify that the standardized spectrum actually provides an envelope of the SMS spectra.

Some examples of elementary systems classified as seismic are given in Table 1-2.

TABLE 1-2. FRENCH POSITION: LIST OF ELEMENTARY SYSTEMS CLASSIFIED AS SEISMIC (PWR PLANT)

Elementary systems and symbols System limits/item

Steam generator blowdown system (APG) From steam generator to containment isolation

Feedwater flow control system (ARE) From external containment to steam generator

Auxiliary feedwater system (ASG) Entirely

Circulating water filtration system (CFI) Filters and washing

Electrical building chilled water system (DEL)

Entirely

Fuel building handling equipment (DMK) Lifting chain

Reactor building handling equipment (DMR) Handling platform

Control room air conditioning system (DVC) Entirely

Auxiliary feedwater pump room ventilation system (DVG)

Unit heaters

44


TABLE 1-2. (cont.)

Elementary systems and symbols System limits/item

Charging pump room emergency ventilation Extraction of smoke fromsystem (DVH) charge pump rooms and extraction

to chimney

Fuel building ventilation system (DVK) Part concerned with reduced flow

Nuclear auxiliary building ventilation system (DVN)

Release to stack

Computer room air conditioning system (DVR) Entirely

Safety injection and containment spray pump motor room ventilation system (DVS)

Part concerned with air extraction

Electrical room ventilation (DVZ) Entirely

Containment spray system (EAS) Entirely

Containment sweeping ventilation system (EDE)

Entirely

Containment atmosphere monitoring system (ETY)

Entirely (expected blowing)

Containment continuous ventilation system — Pool well(EVR) — Extraction: dome level of

reactor building

Turbine bypass system (GCT) Connection with main steam system

Fire fighting general water distribution system (JPD)

Nuclear island

Nuclear island fire protection system (JPI) Nuclear island

Nuclear island liquid radwaste monitoring and discharge system (KER)

Storage tanks

Emergency power supply systems Entirely and including all(LHP-LHQ) annex materials

Fuel handling and storage system (PMC) Entirely

Reactor cavity and spent fuel pit cooling Pool and tank coolantand treatment system (PTR)

45


TABLE 1-2. (cont.)

Eiementary systems and symbols System limits/item

Reactor coolant system (RCP) Entirely (except pressurizer discharge tank)

Chemical and volume control system (RCV) Entirely

Reactor boron and water make-up system (REA)

Boric acid preparation equipment

Nuclear sampling system (REN) — Connection with primary circuit— Connection with the secondary

side of steam generator— Connection with gas storage

tanks

In-core instrumentation system (RIC) Partly

Safety injection system (RIS) Entirely

Nuclear island vent and drain system (RPE) Primary and gaseous wastes

Residual heat removal system (RRA) Entirely

Component cooling system (RRI) Two redundant lines

Instrument compressed air distribution Main tankssystem (SAR)

Essential service water system (SEC) Entirely

Chemical reagent injection system (SIR) From containment isolation to steam generator

Gaseous waste treatment system (TEG) Entirely

Boron recycle system (TEP) Head tank o f the chain

Liquid waste discharge system (TER) Three tanks

Main steam system (VVP) From steam generator to steam isolation (outside o f reactor building)

1-3. JAPANESE POSITION

The Japanese position in shown in Table 1-3.

46


TABLE 1-3. SEISMIC CLASSIFICATION OF MAIN NUCLEAR POWER PLANT FACILITIES IN JAPAN

Seismic classification FacilitiesExamples o f items

PWR BWR

As (i) Equipment and piping constituting the 'reactor coolant pressure boundaries' whose definition is the same as that given in 'Regulatory Guide for Safety Design of Light Water Power Reactors'

(1) Reactor vessel (2) Reactor coolant pressure boundary

(ii) Spent fuel storage facilities (1)(2)

Spent fuel pit Spent fuel storage rack

0 )(2)

Fuel poolSpent fuel storage rack

(iii) Facilities required to introduce negative reactivity into the core

(1) Control rod cluster Control rod drive mechanism

(1)(2)

Control rodControl rod drive mechanism

to affect an emergency shutdown and to maintain the core in a shutdown state

(2) Boron injection system (transfer system)

(3) Control rod drive hydraulic system

(iv) Facilities required for the (1) Main steam feedwater system (1) Reactor core isolationremoval o f decay heat from the (2) Auxiliary feedwater system cooling systemcore after reactor shutdown (3)

(4)Condensate storage tank Residual heat removal system

(2)

(3)

High pressure core spray systemResidual heat removal system

(4) Suppression pool


TABLE 1-3. (cont.)

Seismic classification FacilitiesExamples of items

PWR BWR

(v) Facilities serving as pressure boundaries which directly prevent the spread of radioactive material in the event of a failure of the reactor coolant pressure boundaries

(1)(2)

Containment vessel Container boundary

(1)(2)

Primary containment vessel Container boundary

A (i) Facilities required for the (1.) Safety injection system (1) Emergency core cooling systemremoval o f decay heat from the (2) Refuelling water storage — High pressure core spraycore in the event o f a failure (3) Residual heat removal system systemof the reactor coolant pressure — Low pressure core sprayboundaries system

— Residual heat removal system— Automatic depressurization

system(2) Suppression pool

(ii) Facilities, other than those (1) Containment spray system (1) Residual heat removal systembelonging to As class (v), which (2) Annulus seal (2) Flammability control systemserve to prevent the discharge (3) Annulus air cleanup system (3) Reactor buildingof radioactive material into the (4) Containment purge exhaust stack (4) Filtration recirculation andatmosphere after an accident (5) Safeguard component area ventilation systeminvolving radioactive material cleanup system (5) Pressure suppression system


TABLE 1-3. (cont.)

Examples of itemsSeismic classification Facilities

P W R B W R

(diaphragm floor, vent pipe)(6) Main steam isolation valve

leakage control system(7) Suppression pool

(iii) Others (1) Spent fuel pit service water (1) Fuel pool water make-upsystem system

(2) Standby liquid control system

(i) Facilities which are directly connected to the reactor coolant pressure boundaries and which contain or may contain primary coolant

(ii) Facilities containing radioactive waste material, excluding those with relatively small quantities of radioactive material or those whose rupture would lead to a considerably lower radioactive dose to the general public than the annual dose permitted at the controlled environs

(1) Chemical and volume control system

(2) Letdown system(3) Excess letdown system

(1) Main steam system(2) Feedwater system(3) Reactor coolant purification

system

(1) Waste disposal system


TABLE 1-3. (cont.)

Examples of itemsSeismic classification Facilities ---------------------------------------------------------------

PWR BWR

Facilities containing radioactive (1) Spent fuel pit cleanup system (1) Turbine, condenser and feedmaterial other than radioactive (2) Auxiliary building crane water heaterwaste whose rupture might lead (3) Spent fuel pit crane (2) Condensate demineralizerto the excessive exposure of (4) Refuelling crane systemthe general public and/or plant (3) Fuel transfer system (3) Condensate storage tankpersonnel (4) Fuel pool purification system

(5) Shielding(6) Control rod drive hydraulic

system(7) Reactor building crane

(8) Fuel handling system(9) Control rod storage rack

Facilities required to cool spent (!) Spent fuel pit water cooling (!) Fuel pool cooling systemfuel system

(v) Facilities, other than thosebelonging to classes As, A and B, which would prevent the release of radioactive material into the atmosphere in the event of an accident accompanied by the release of radioactive material


TABLE 1-3. (cont.)

Seismic classification FacilitiesExamples o f items

PWR BWR

C (i) Facilities other than those (1) Control rod drive mechanism (1) Recirculation flow controlbelonging to classes As, A and systemB, which are required to (2) Control rod drive hydrauliccontrol radioactivity system

(") Facilities other than those (1) Sampling system (!) Sampling systembelonging to classes As, A and (2) Floor drain system (2) Floor drain systemB, which contain or relate (3) Laundry drain treatment (3) Laundry liquid waste treatmentto radioactive material package system

(4) Solid waste disposal system (4) Solid waste treatment system

(5) Boric acid evaporator (5) W aste disposal system

(6) Waste evaporator (6) Hydraulic press baling machine(7) Primary make-up water system

(iii) Facilities not related to (1) Turbine system (1) Circulating water systemradioactive safety but classified (2) Component cooling water (2) Turbine building closedas reactor facilities system cooling water system

(3) Auxiliary boiler (3) Auxiliary boiler

(4) Fire protection system (4) Fire protection system

(5) Main gnerator and main (5) Main generator and transformertransformer (6) Turbine building crane

(6) Heating and ventilation system(7) New fuel storage system (7) Station air system


TABLE 1-3. (cont.)

Examples of itemsSeismic classification Facilities ------------------------------------------------------------------------------------------

PWR BWR

(8) Steam gnerator flowdownsystem

(9) Station service air system(10) Polar crane

Notes:

As: items are usually evaluated for the SL-2 level earthquake.A: items are usually designed for the SL-t level earthquake.B: items are usually designed for an earthquake of 0.5 Sl-1 intensity.C: items are usually designed for an earthquake of 0.33 SL-2 intensity.


Annex H

SLOSHING AND IMPULSE EFFECTS IN LIQUID CONTAINERS

II-J . CONCEPT

The oscillation of a free liquid surface [43] has been called 'sloshing'. This phenomenon is similar to the seiches resulting from earthquake effects in a lake or basin. For seismic design associated with the sloshing liquid, only the fundamental frequency is important. The natural frequency is about 1 Hz for a cylindrical vessel whose diameter is approximately 1 m, while for a 100 m diameter vessel it is slightly greater than 0.1 Hz. The fundamental frequency of sloshing motion therefore lies in the range of 1 to 0.1 Hz for the cases relevant to seismic design in a nuclear power plant-

Also of concern in the seismic design of liquid containers are impulsive load effects of the liquid below the sloshing liquid [44, 45].

H -2. SIGNIFICANCE

The sloshing liquid, typically in an up and down motion, can impart significant impulsive and impact loads as well as cyclic loads on a structure or portions of the structure in its path. In particular, it can cause failure o f tank roofs and attachments to walls o f tanks and pools in its path.

Impulsive liquid levels are of paramount importance in the design of the container and container anchorage, because the impulsive liquid typically generates overall loads associated with pressure and overturning or uplift moments on the container and its supports which are much larger than the sloshing loads.

11-3. DESIGN

To evaluate the motions, the concepts of sloshing liquid and an impulsive liquid are generally used [9]. The subdivision derives from the fact that only the semispher- ical or cylindrical part of the water with diameter equal to the average dimension of the free surface is sloshing (moving) in the case of an axisymmetric or rectangular vessel. Therefore, if the depth of the liquid container is greater than one half o f the diameter, the fundamental frequency of the sloshing liquid is a function only o f the diameter of the container.

The seismic response of the impulsive fluid is also a function of the dynamic characteristics of the container and can be determined as shown in Ref. [9].

53


11-4. RESONANCE SLOSHING EFFECTS

Sometimes resonance phenomena have been observed and damage to fuel stored in pools and underground water reservoirs has been reported. The ground motion of high magnitude earthquakes sometimes reaches total displacement values of up to 1 m peak-to-peak amplitude. If the sloshing resonates with the ground motion, its wave height can easily reach several metres.

11-5. DAMPING

The damping coefficient for the sloshing mode is extremely low and is typically taken as 0.5% or less. Damping associated with the impulse mode of vibration is typically that associated with the container material, and the connections and anchorage used.

11-6. VERTICAL EXCITATION

If the vertical component of the acceleration at the free water surface is greater than l.Og, additional waves at the free surface could be generated. In such cases, non-linear damping effects should be considered in the response.

54


R E FE R E N C E S

f ] ] AMERICAN NUCLEAR SOCIETY, Guidelines for Combining Natural and Externai Man-Made Hazards at Power Reactor Sites, Rep. ANSI/ANS-2.12-1978, ANS, La Grange Park, IL (1978).

[2] AMERICAN SOCIETY OF MECHANICAL ENGINEERS, "Dynamic analysis methods", Boiler and Pressure Vessel Code, Section III, Appendix N, ASME, New York, NY (1989).

[3] DUFF, C.G:, HEIDEBRECHT, A.C., "Earthquake fatigue effects on CANDU nuclear power plants", paper presented at 3rd Canadian Conf. on Earthquake Engineering, Montreal, 4-6 June 1979.

[4] AMERICAN SOCIETY OF CIVIL ENGINEERS, Guidelines for the Design and Analysis of Nuclear Safety Related Earth Structures, Rep. ANSI/ASCE N-725, ASCE, New York, NY (1982).

[5] DUFF, C.G., "Potential overdesign for the extreme load condition — A Canadian viewpoint", paper presented at the 5th Int. Sem. on Probabilistic and Extreme Load Design of Nuclear Plant Facilities, Chicago, IL, 29-30 August 1983.

[6] LU, S.C., CHOU, C.K., Reliability Analysis of Stiff Versus Flexible Piping: Final Project Report, NUREG/CR-4263, Lawrence Livermore Natl Lab., CA (May 1985).

[7] UDOGUCHI, T., OHSAKI, Y;, SHIBATA, H., "The aseismic design of nuclear power plants in Japan", Peaceful Uses of Atomic Energy (Proc. 4th Int. Conf. Geneva, 1971), Vol. 3, United Nations, New York and IAEA, Vienna (1972) 297.

[8] NUCLEAR REGULATORY COMMISSION, Seismic Design, Standard Review Plan, Section 3.7, Rev. 2, USNRC, Washington, DC (August 1989).

[9] AMERICAN SOCIETY OF CIVIL ENGINEERS, Seismic Analysis of Safety Related Nuclear Structures and Commentary on Standards 4-86, ASCE, New York (1986).

[10] SEED, H.B., IDRISS, I.M., A simplified procedure for evaluating soil liquefaction potential, J. Soil Mech. Found. Div., Am: Soc. Civ. Eng. 97 SM9 (1971) 1249.

[11] SEED, H.B., ARANGO, I., CHAN, C.K., Evaluation of Soil Liquefaction Potential During Earthquakes, Rep. EERC 75-28, Earthquake Engineering Research Center, Univ. of California, Berkeley (1975).

[12] UNITED STATES ARMY CORPS OF ENGINEERS, Engineering and Design Stability of Earth and Rock-Fill Dams, Manual EM 1110-2-1902, Office of the Chief of Engineers, Dept : of the Army, Washington, DC (1970).

[13] REDDY, A.S., SRINIVASAN, R.J., Bearing capacity of footings on clays, Soils Found. (Japan) 11 3 (Sept. 1971).

[14] SILVER, M.L., SEED, H.B., Deformation characteristics of sands under cyclic loading, J. Soil Mech. Found. Div., Am. Soc. Civ. Eng. 97 SM8 (1971) 1081.

[15] BROWN, P.T., Influence of soil inhomogeneity on raft behavior, Soils Found. (Japan) 14 1 (March 1974).

[16] SHIBATA, H., et al., Development of aseismic design of piping, vessels, and equipment in nuclear facilities, Nucl. Eng. Des. 22 2 (1972) 247.

[17] AZIZ, T.S., BISWAS, J.K., "Spectrum compatible time histories for seismic design of nuclear power plants", paper presented at 3rd Canadian Conf. on Earthquake Engineering, Montreal, Canada, 4-6 June 1979.

55


[18] BEZLER, P., WANG, Y.K., REICH, M., Response Margins Investigation of Piping Dynamic Analysis Using the Independent Support Motion Method and PVRC Damping, Rep. NUREG/CR 5105, Brookhaven Natl Lab., Upton, NY (March 1988).

[19] NUCLEAR REGULATORY COMMISSION, Seismic Qualification of Mechanical Equipment for Nuclear Power Plants, Regulatory Guide 1-100, Rev. 2, USNRC, Washington, DC (June 1988).

[20] INSTITUTE OF ELECTRICAL AND ELECTRONICS ENGINEERS, IEEE Recommended Practice for Seismic Qualification of Class IE Equipment for Nuclear Power Generation Stations, Standard 344-1987, IEEE, New York (August 1987).

[21] KANA, D.J., POMERANANG, D.J., Similarity Principles for Equipment Qualified by Experience, Rep. NUREG/CR-5012, USNRC, Washington, DC (July 1988).

[22] NUCLEAR REGULATORY COMMISSION, Instrumentation for Earthquakes, Regulatory Guide 1.12, Rev. 1, USNRC, Washington, DC (April 1974).

[23] AMERICAN NUCLEAR SOCIETY, Earthquake Instrumentation Criteria for Nuclear Power Plants, Rep. ANSI/ANS-2.2-1988, ANS, La Grange Park, IL (1988).

[24] MINISTRY OF RESEARCH AND INDUSTRY, Basic Safety Rules, Rule No. 1.3b— Seismic Instrumentation, Central Service for the Safety of Nuclear Installations,

. Paris (June 1984).[25] AMERICAN SOCIETY OF MECHANICAL ENGINEERS, Alternate Method of

Earthquake Description for Class 2 and 3 Piping at Low Seismicity Sites, Code Case N-468, Section III, Division 1, ASME, New York (1989).

[26] NUCLEAR REGULATORY COMMISSION, Combining Modal Responses and Spatial Components in Seismic Response Analysis, Regulatory Guide 1.92, Rev. 1, USNRC, Washington, DC (Feb. 1976).

[27] COATS, D.W., Recommended Revisions to Nuclear Regulatory Commission Seismic Design Criteria, Rep. NUREG/CR-1161, USNRC, Washington, DC (May 1980).

[28] AZIZ, T.S., DUFF, C.G., "Decoupling criteria for seismic analysis of nuclear power plant systems", paper presented at ASME/CSME Pressure Vessels and Piping Conf., Montreal, Canada, 25-30 June 1978.

[29] NEWMARK, N.M., ROSENBLUETH, E., Fundamentals of Earthquake Engineering, Prentice-Hall, Englewood Cliffs, NJ (1971).

[30] WIEGEL, R.L., Earthquake Engineering, Prentice-Hall, Englewood Cliffs, NJ (1970).[31] HISADA, T., et al., Philosophy and practice of the aseismic design of nuclear power

plants — Summary of the guidelines in Japan, Nuc. Eng. Des. 20 2 (1972) 339.[32] DUFF, C.G., "Simplified seismic analysis methods used by AECL for the seismic

qualification of CANDU nuclear power plants", paper presented at 2nd Symp. on Current Issues Related to Nuclear Power Plant Structures, EPRI and North Carolina State University, Orlando, FL, 7-9 December 1988.

[33] ZIENKIEWICZ, O.E., CHEUNG, Y.K., Finite Element Method in Structural and Continuum Mechanics, McGraw-Hill, New York (1967).

[34] CLOUGH, R.W., BATHE, K.J., Finite Element Analysis of Dynamic Response (ODEN, J.T., Ed.), University of Alabama, Huntsville (1970).

[35] PESTEL, E.C., LECKIE, F.A., Matrix Methods in Elastomechanics, McGraw-Hill, New. York (1963).

[36] BLEVINS, R.D., Flow Induced Vibration, 2nd edn, Van Nostrand Reinhold, Princeton, NJ, and New York (1990).

56


[37] KOT, C.A., Dynamic Testing of As-Built Civil Engineering Structures: A Review and Evaluation, Rep. NUREG/CR-3649, Argonne National Laboratory, IL (1984).

[38] ATOMIC ENERGY COMMISSION, Damping Values for Seismic Design of Nuclear Power Plants, Regulatory Guide 1.61, USNRC, Washington, DC (1973).

[39] USMANI, S.A., SALEEM, M.A., SO, G., Damping Considerations in CANDU Feeder Pipe Design and Analysis, ASME PVP, Vol. 198, Nashville, TN (June 1990).

[40] JAPAN ELECTRIC ASSOCIATION, Technical Guidelines for Aseismic Design of Nuclear Power Plants, Tokyo (1987) (in Japanese).

[41] NEWMARK, N.M., HALL, W.J., Development of Criteria for Seismic Review of Selected Nuclear Power Plants, Rep. NUREG/CR-0098, USNRC, Washington, DC (May 1978).

[42] SALEEM, M.A., CHANDHOKE, P., AGGARWAL, M.L., "Assessment of seismic code case N-411 for application to CANDU piping", paper presented at ASME PVP Conf. Pittsburgh, PA (June 1988).

[43] SOGABE, K., On sloshing effects of liquid in cylindrical and spherical vessels during a strong earthquake, Bull. Earthquake Resistant Structure Research Center, University of Tokyo, No. 8 (1974) 18,

[44] DAVIDOVKI, V , HADDADI, A., "Calcul pratique de reservoirs en zone sismique'', n° 409, Serie: Theories et mdthodes de calcul 256, Institut technique du batiment et des travaux publics (1982).

[45] HAROUN, M.A., Vibration studies and testings of liquid storage tanks, Earthquake Engineering and Structural Dynamics 2 (1983) 179-206.

57


BIBLIOGRAPHY

CODE OF FEDERAL REGULATION, 10 CFR 100 Appendix, Seismic Requirements for Nuctear Power Ptants, Washington, DC (1976).

FUKUOKA, M., Damage to civit engineering structures, Soils Found. (Japan) 6 2 (1966).

HARDIN, B.C., DRNEVICH, V.P., Shear modulus and damping in soils, J. Soil Mech. Found. Div., Am. Soc. Civ. Eng. 98 SM6 (1972) 667.

JAPAN ATOMIC ENERGY COMMISSION, Regulatory Guide for Aseismic Design of Nuclear Power Reactor Facilities, Tokyo (July 1981).

KERNTECHNISCHER AUSSCHUSS, Design of Nuclear Power Plants Against Seismic Events, Part 1: Basic principles, Rep. KTA 2201, KTA-Geschaeftsstelle beim Institut fur Reaktorsicherheit des technischen Uberwachungs-Verein eV, Cologne (1975).

M.W. KELLOG COMPANY, Design of Piping Systems, Wiley, London (1956).

NUCLEAR REGULATORY COMMISSION, REPORT ON THE US Nuclear Regulatory Commission Piping Review Committee, Rep. NUREG-1061, Vols 1-5, USNRC, Washington, DC (April 1985).

RICHARD, F.E., HALL, J R., WOODS, R.D., Vibration of Soils and Foundations, Prentice-Hall, Englewood Cliffs, NJ (1970).

SEED, H.B., Landslides during earthquakes due to liquefaction, J. Soil Mech. Found. Div., Am. Soc. Civ. Eng. 94 SM5 (1968) 1053.

SEED, H., LEE, K.L., Liquefaction of saturated sands during cyclic loading, J. Soil Mech. Found. Div., Am. Soc. Civ. Eng. 92 SM6 (1966) 105.

SHIBATA, H., AKINO, K., KATO, H., On estimated modes of failure of NPP by potential earthquakes, Nuct. Eng. Des. 28 2 (1974) 257.

SISCHER, E.G., Sine beat vibration testing related to earthquake response spectra, Shock Vib. Bull. 42 Part 2 (1972).

TAJIMI, H., "A statistical method of determining the maximum response of building structure during an earthquake", paper presented at 2nd World Conf. on Earthquake Engineering, Tokyo and Kyoto, 1960.

TAJIMI, H., "Dynamic analysis of a structure embedded in an elastic stratum", paper presented at 4th World Conf. on Earthquake Engineering, Santiago, Chile, 1969.

WATABE, M., KATO, M., KUROPA, T., "Procedures, analysis and research on earthquake resistant design of nuclear power plants", paper presented at Todatsu Seminar, October 1982.

59


CONTRIBUTORS TO DRAFTING AND REVIEW

During the development phase the following experts participated in one or more of the meetings (1975-1978):

Akino, K. Japan

Beare, J.W. Canada

Becker, K. International Organization for Standardization

Burkhardt, W. Council for Mutual Economic Assistance

Candes, P. (Chairman) France

Carmona, J. Argentina

Ciement, B. France

Dtouhy, Z. Czechostovakia

Fischer, J. Germany

Franzen, L.F. Germany

Gammill, W.P. United States of America

Ganguly, A.K. India

Gausden, R.A. United Kingdom

Giuliani, P. Italy

Gronow, W.S. United Kingdom

Hedgran, A. Sweden

Heimgartner, E. Switzerland

Hendrie, J. United States of America

Hurst, D. Canada

Iansiti, E. Internationa] Atomic Energy Agency

61


Itari, O.

Jeschki, W.

Karbassioun, A.

Ktik, F.

Konstantinov, L.

Kovalevich, O.M.

Kriz, Z.

Livotant, M.

Mattson, R.J.

Messiah, A.

Minogue, B.

Mochizuki, K.

Nitson, R.

Omote, S.

Pete, J.

Petrangeli, G.

Roberts, I. Craig

Sanchez Gutierrez, J

Schwarz, G.

Sevcik, A.

Shibata, H.

Soman, S.D.

Idriss, I.M .

62

Nuctear Energy Agency of the OECD

Switzerland

Internationat Atomic Energy Agency

Czechostovakia

International Atomic Energy Agency

Union of Soviet Sociatist Repubtics

Czechostovakia

France

United States of America

France


Japan

Internationat Organization for Standardization

Japan

Commission of the European Communities

Itaty


Mexico

Canada

Czechostovakia

Japan

India

U nited States o f Am erica


Stadie, K. Nuclear Energy Agency of the OECD

Stevenson, J.D . United States o f America

Tiidsiey, F.C .J. United Kingdom

Uchida, H. Japan

Van Reijen, G. Commission of the European Communities

Velez, C. Mexico

Witulski, H. Germany

Zuber, J.F . Switzerland

D uring the rev is ion phase th e fo llow ing experts p a rtic ipa ted it

(1988 -1990 ):

Aggarwal, M .L. Canada

Conte, M. France

Culambourg, J. France

Danisch, R. Germany

Duff, G. Canada

Fischer, J. International Atomic Energy Agency

Ford, P.J. United Kingdom

Giuliani, P. International Atomic Energy Agency

Godoy, A. International Atomic Energy Agency

Giirpinar, A. International Atomic Energy Agency

Hintergraber, M. Germany

Inkester, J. United Kingdom

Jeng, P.J. United States o f America

63


Maiboroda, V.V.

Paganini, C.

Shibata, H.

Stevenson, J.

Takashima, K.

Argentina

Japan


Japan

U nion o f Soviet Socialist R epublics

NUCLEAR SAFETY STANDARDS ADVISORY GROUP (NUSSAG)

Barber, P.

Brooks, G.

Bystedt, P.

Gast, K.

Herttrich, M.

Heltemes, C.J., Jr.

Isaev, A.

Ishikawa, M.

Kovalevich, O.M.

Lee, S.H.

Lei, Y.

Lopez-Menchero, M.E.

Medina, M.

Mucskai, G.

Pele, J.

Reed, J.

France

Canada

Sweden

Germany

Germany


Russian Federation

Japan

Russian Federation

Korea, Republic of

China


Mexico

Hungary


United Kingdom

64


Ryder, E.

Sarma, M.S.R

Scherrer, J.

Versteeg, J.

Yamamoto, K.

India

France

Netherlands

Japan

U nited K ingdom

65


LIST OF NUSS PROGRAMME TITLES

A sAoMM wofeJ sowe ;nser;'cs way rewscd w fAe wear fMrg.

77iosc fAar Aave a/rcady ^ecw rcwscJ arc :W:'ca?ed &y fAe ajdi'doH o / '(Kcv. 7)' ;o yAe HMw gr.

1. GOVERNMENTAL ORGANIZATION

50-C-G (Rev. 1) Code on the safety of nuclear power plants: Governmentalorganization

&3/efy GMK?M

50-SG-G1 [Qualifications and training of staff of the regulatory bodyfor nuclear power plants

50-SG-G2 Information to be submitted in support of licensingapplications for nuctear power plants

50-SG-G3 Conduct of regutatory review and assessment during thelicensing process for nuclear power plants

50-SG-G4 Inspection and enforcement by the regulatory body fornuclear power plants

50-SG-G6 Preparedness of public authorities for emergencies atnuctear power plants

50-SG-G8 Licences for nuclear power plants: Content, format andi legal .considerations :

50-SG-G9 Regulations and guides for nuclear power plants

2. SITING

50-C-S (Rev. 1) Code on the safety of nuctear power plants: Siting

3a/efy GtH'des

50-SG-S1 (Rev. 1) . Earthquakes and associated topics in relation to nuclear power plant siting

50-SG-S3 Atmospheric dispersion in nuclear power plant siting

50-SG-S4 Site selection and evaluation for nuclear power plantswith respect to population distribution

1988

1979

1979

1980

1980

1982

1982

1984

198,8

1991

1980

1980

67


50-SG-S6

50-SG-S7

50-SG-S8

50-SG-S9

50-SG-S 10A

50-SG-S 10B

50-SG-S H A

50-SG-SHB

3. DESIGN

50-C-D (Rev. 1)

50-SG-D 1

50-SG-D2 (Rev. 1)

50-SG-D3

50-SG-D4

50-SG-D5

50-SG-D6

50-SG-D7 (Rev. 1)

50-SG-D8

50-SG-S5

50-SG-D9

External man-induced events in relation to nuclear power plant siting

Hydrological dispersion of radioactive material in relation to nuclear power plant siting

Nuclear power plant siting: Hydrogeological aspects

Safety aspects o f the foundations o f nuclear power plants

Site survey for nuclear power plants

Design basis flood for nuclear power plants on river sites

Design basis flood for nuclear power plants on coastal sites

Extreme meteorological events in nuclear power plant siting, excluding tropical cyclones

Design basis tropical cyclone for nuclear power plants

Code on the safety o f nuclear power plants: Design

Safety functions and component classification for BWR, PWR and PTR

Fire protection in nuclear power plants

Protection system and related features in nuclear power plants

Protection against internally generated missiles and their secondary effects in nuclear power plants

External man-induced events in relation to nuclear power plant design

Ultimate heat sink and directly associated heat transport systems for nuclear power plants

Emergency power systems at nuclear power plants

Safety-related instrumentation and control systems for nuclear power plants

Design aspects of radiation protection for nuclearpower plants

1985

1984

1986

1984

1983

1983

1981

1984

1988

1979

1992

1980

1980

1982

1981

1991

1984

1985

1981

68


50-SG-D11 General design safety principles for nuclear power plants

50-SG-D12 Design of the reactor containment systems in nuclearpower plants

50-SG-D13 Reactor coolant and associated systems in nuclear powerplants

50-SG-D14 Design for reactor core safety in nuclear power plants

50-SG-D15 Seismic design and qualification for nuclear power plants

4. OPERATION

50-C -0 (Rev. 1) Code on the safety o f nuclear power plants: Operation

So/i?fy

50-SG-01 (Rev. 1) Staffing o f nuclear power plants and the recruitment,training and authorization of operating personnel

50-SG-02 In-service inspection for nuclear power plants

50-SG-03 Operational limits and conditions for nuclear power plants :

50-SG -04 Commissioning procedures for nuclear power plants

50-SG-05 Radiation protection during operation of nuclearpower plants

50-SG -06 Preparedness o f the operating organization (licensee)for emergencies at nuctear power plants

50-SG-07 (Rev. 1) Maintenance o f nuclear power ptants

50-SG -08 (Rev. 1) Surveillance of items important to safety in nuclearpower plants

50-SG -09 Management o f nuclear power plants for safe operation

50-SG-010 Core management and fuel handling for nuclearpower ptants

50-SG-011 Operational management of radioactive effluents andwastes arising in nuctear power plants

50-SG-D10 Fuel handling and storage systems in nuclear power plants

S. QUALITY ASSURANCE

50-C-QA (Rev. 1) Code on the safety o f nuctear power ptants: Quality assurance

1984

.1986

1985

1986

1986

1992

1988

1991

1980

1979

1980

1983

1982

1990

1990

1984

1985

1986

1988

69


50-SG-QA1 Establishing of the quality assurance programme for anuclear power plant project

50-SG-QA2 Quality assurance records system for nuclearpower plants

50 SG-QA3 Quality assurance in the procurement of items andservices for nuclear power plants

50-SG-QA4 Quality assurance during site construction of nuclearpower plants

50-SG-QA5 (Rev. 1) Quality assurance during commissioning and operation o f nuclear power plants

50-SG-QA6 Quality assurance in the design o f nuclear power plants

50-SG-QA7 Quality assurance organization for nuclear power plants

50-SG-QA8 Quality assurance in the manufacture of items fornuclear power plants ;

50-SG-QA10 Quality assurance auditing for nuclear power plants

50-SG-QA11 Quality assurance in the procurement, design andmanufacture o f nuclear fuel assemblies

SAFETY PRAC7YCES

50-P-l Application of the single failure criterion

50-P-2 In-service inspection of nuclear power plantsA manual

50-P-3 Data collection and record keeping for themanagement o f nuclear power plant ageing

1979

1979

1981

1986

1981

1983

1981

1980

1983

1990

1991

1991

1984

70


SELECTION OF IAEA PUBLICATIONS RELATING TO THE SAFETY OF

NUCLEAR POWER PLANTS

SAFETY SERIES

9 Basic safety standards for radiation protection,1982 edition

49, Radiological surveillance of airborne contaminantsin the working environment

52 Factors relevant to the decommissioning of land-basednuclear reactor plants

55 Planning for off-site response to radiationaccidents in nuclear facilities

57 Generic models and parameters for assessingthe environmental transfer o f radionuclides from routine releases: Exposures o f critical groups

67 Assigning a value to transboundary radiation exposure

69 Management of radioactive wastes from nuclearpower plants

72 Principles for establishing intervention levels for the protection of the public in the event o f a nuclear accident or radiological emergency

73 Emergency preparedness exercises for nuclear , facilities: Preparation, conduct and evaluation

75-INSAG-1 Summary report on the post-accident review meetingo n th e C h e m o b y la c c id e n t,

75-INSAG-2 Radionuclide source terms from severe accidents tonuclear power plants with light water reactors

75-INSAG-3 Basic safety principles for nuclear power plants

75-INSAG-4 Safety culture

75-INSAG-5 The safety o f nuclear power: INSAG-5

75-INSAG-6 Probabilistic safety assessment

77 Principles for limiting releases of radioactiveeffluents into the environment

1979

1980,

1981;

1982

1985

1985

1985

1985

1986 :

1987

1988-

1991

1992

1992

1986

1982,,

71


79 Design of radioactive waste management systemsat nuclear power plants

81 Derived intervention levels for application incontrolling radiation doses to the public in the event o f a nuclear accident or radiological emergency: Principles, procedures and data

84 Basic principles for occupational radiation monitoring

86 Techniques and decision making in the assessmentof off-site consequences o f an accident in a nuclear facility

93 Systems for reporting unusual events in nuclear power plants

94 Response to a radioactive materials release having a transboundary impact

97 Principles and techniques for post-accident assessment and recovery in a contaminated environment o f a nuclear facility

98 On-site habitability in the event o f an accident at a nuclear facility:Guidance for assessment and improvement

101 Operational radiation protection: A guide to optimization

103 Provision of operational radiation protection services at nuclear power plants

104 Extension of the principles o f radiation protection to sources of potential exposure

105 The regulatory process for the decommissioning of nuclear facilities

106 The role o f probabilistic safety assessmentand probabilistic safety criteria in nuclear power plant safety

TECHNICAL REPORTS SERIES

217 Guidebook on the introduction of nuclear power

224 Interaction of grid characteristics with design andperformance of nuclear power plants: A guidebook

230 Decommissioning of nuclear facilities:Decontamination, disassembly arid waste management

1986

1987

1987

1989

1989

1989

1989

1990

1990

1990

1990

1992

1982

1983

1986

1983

72


239

242

249

262

267

268

27 i

274

282

292

294

296

299

300

301

237

306

Nuctear power plant instrumentation and control:A guidebook

Qualification of nuclear power plant operations personnel: A guidebook

Decontamination of nuclear facilities to permit operation, inspection, maintenance, modification or plant decommissioning

Manual on training, qualification and certification of quality assurance personnel

Methodology and technology of decommissioning nuclear facilities

Manual on maintenance of systems and components important to safety

Introducing nuclear power plants into electrical power systems of limited capacity: Problems and remedial measures

Design of off-gas and air cleaning systems at nuclear power plants

Manual on quality assurance for computer software related to the safety o f nuclear power plants

Design and operation of off-gas cleaning and ventilation systems in facilities handling low and intermediate level radioactive material

Options for the treatment and solidification of organic radioactive wastes

Regulatory inspection of the implementation o f quality assurance programmes: A manual

Review of fuel element developments for water cooled nuclear power reactors

Cleanup o f targe areas contaminated as a result o f a nuclear accident

Matnual on quatity assurance for installation and commissioning of instrumentation, control and electrical equipment in nuclear power ptants

Manual on quality assurance programme auditing

Guidebook on the education and training of techniciansfor nuclear power

1984

1984

1985

1986

1986

1986

1987

1987'

1988

1988

1989

1989

1989

1989

1989

1984

1989

73


307 Management of abnormal radioactive wastes at nuclear power plants

327 Planning for cleanup of large areas contaminated as aresult of a nuclear accident

328 Grading of quality assurance requirements: A manual

330 Disposal o f waste from the cleanup of large areascontaminated as a result of a nuclear accident

334 Monitoring programmes for unrestricted releaserelated to decommissioning of nuclear facilities

338 Methodology for the management o f ageing ofnuclear power plant components important to safety

IAEA TECDOC SERIES

276 Management o f radioactive waste from nuclearpower plants

294 International experience in the implementation of thelessons learned from the Three Mile Island accident

303 Manua) on the selection of appropriate quality assuranceprogrammes for items and services o f a nuclear power plant

308 Survey o f probabilistic methods in safety and risk assessment for nuclear power plant licensing

332 Safety aspects o f station blackout at nuclear power plants

341 Developments in the preparation of operating proceduresfor emergency conditions o f nuclear power plants

348 Earthquake resistant design of nuclear facilities withiimited radioactive inventory

355 Comparison of high efficiency particulate filter testingmethods

377 Safety aspects o f unplanned shutdowns and trips

379 Atmospheric dispersion models for application inrelation to radionuclide releases

387 Combining risk analysis and operating experience

390 Safety assessment o f emergency electric power systemsfor nuclear power plants

416 Manual on quality assurance for the survey, evaluation andconfirmation of nuclear power plant sites

1991

1991

1992

1992

1992

1983

1983

1984

1984

1985

1985

1985

1985

1986

1986

1986

1986

1987

1989

74


424

425

443

444

450

451

458

497

498

499

508

510

522

523

525

529

540

542

543

547

Identification of failure sequences sensitive to human error

Simulation of a loss of cooiant accident

Experience with simulator training for emergency conditions

Improving nuctear power ptant safety through operatoraids

Dose assessments in nuctear power ptant siting

Some practical implications of source term reassessment

OSART results

OSART resutts II

Good practices for improved nuctear power ptant performance

Models and data requirements for human reliability analysis

Survey o f ranges o f component reliability data for use in probabilistic safety assessment

Status of advanced technology and design for water cooted reactors: Heavy water reactors

A probabitistic safety assessment peer review:Case study on the use o f probabitistic safety assessment for safety decisions

Probabilistic safety criteria at the safety function/system tevet

Guidebook on training to estabtish and maintainthe qualification and competence of nuctear power ptantoperations personnet

User requirements for decision support systems used for nuctear power plant accident prevention and mitigation

Safety aspects o f nuctear power ptant ageing

Use of expert systems in nuctear safety

Procedures for conducting independent peer reviews of probabilistic safety assessment

The use of probabitistic safety assessment in the reticensing o f nuctear power plants for extended tifetimes

1987

1987

1987

1988

1988

1988

1989

1989

1989

1989

1989

1989

1989

1989

1989

1990

1990

1990

1987

1990

75


553

561

570

581

586

590

591

592

593

599

600

605

611

618

631

632

635

640

648

550

Computer codes for Level 1 probabilistic safety assessment

Reviewing computer capabilities in nuclear power plants

OSART mission highlights 1988-1989

Safety implications of computerized process control in nuclear power plants

Simulation of a loss o f coolant accident with rupture in the steam generator hot collector

Case study on the use of PSA methods:Determining safety importance of systems and components at nuclear power plants

Case study on the use of PSA methods:Backfitting decisions

Case study on the use o f PSA methods:Human reliability analysis

Case study on the use o f PSA methods:Station blackout risk at Millstone Unit 3

Use of probabilistic safety assessment to evaluate nuclear power plant technical specifications

Numerical indicators of nuclear power plant safety performance

OSART good practices: 1986-1989

Use of plant specific PSA to evaluate incidents at nuclear power plants

Human reliability data collection and modelling

Reviewing reactor engineering and fuel handling: Supplementary guidance and reference material for IAEA OSARTs

ASSET guidelines: Revised 1991 edition

OSART guidelines: 1992 edition

Ranking o f safety issues for WWER-440 model 230 nuclear power plants

Procedures for conducting common cause failure analysis in probabilistic safety assessment

Safety of nuclear installations: Future direction

1990

1990

1990

1991

1991

1991

1991

1991

1991

1991

1991

1991

1991

1991

1992

1991

1992

1992

1990

1992

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658 Safety related maintenance in the framework of the 1992reliability centered maintenance concept

659 Reactor pressure vesset embrittlement 1992

PROCEEDINGS SERIES

STI/PUB/593 Quaiity assurance for nuclear power piants 1982

STI/PUB/628 Nuclear power plant control and instrumentation 1983

STI/PUB/645 Reliability o f reactor pressure components 1983

STI/PUB/673 IAEA safety codes and guides (NUSS) in the light o f 1985current safety issues

STI/PUB/700 Source term evaluation for accident conditions 1986

STI/PUB/701 Emergency planning and preparedness for nuclear 1986facilities

STI/PUB/716 Optimization of radiation protection 1986

STI/PUB/759 Safety aspects of the ageing and maintenance of 1988nuclear power plants

STI/PUB/761 Nuclear power performance and safety 1988

STI/PUB/782 Severe accidents in nuclear power plants 1988

STI/PUB/783 Radiation protection in nuclear energy 1988

STI/PUB/785 Feedback of operational safety experience 1989from nuctear power plants

STI/PUB/803 Regulatory practices and safety standards 1989for nuclear power plants

STI/PUB/824 Fire protection and fire fighting in nuclear installations 1989

STI/PUB/825 Environmental contamination fotlowing a 1990major nuclear accident

STI/PUB/826 Recovery operations in the event of a nuclear accident or 1990radiological emergency

STI/PUB/843 Balancing automation and human action in nuctear 1991power plants

STI/PUB/880 The safety of nuclear power: Strategy for the future 1992

77


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An exciusive sates agent for tAEA pubtications, to whom a!) orders and inquiries shoutd be addressed, has been appointed for the foHowing countries:

CANADAUNITED STATES OF AMERICA UNIPUB, 46 1 1-F A ssem bly Drive, Lanham , MD 20706-4391, USA

)n the foiiowing countries tAEA pubtications may be purchased from the sa tes agents or booksetiers tisted or through major iocai bookseiiers. Payment can be made in tocai currency or with UNESCO coupons.

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Com ision Nacional d e Energi'a Atdmica, Avenida del Llbertador 8250, RA-1429 B uenos AiresH unter Publications, 58 A G ipps S treet, Collingwood, Victoria 3066 Service Courrier UNESCO, 202, A venue du Roi, B-1060 B russels Com isi6n Chilena de Energi'a Nuclear, V enta d e Publicaciones, A m unategui 95, Casilla 188-D, Santiago IAEA Publications in C hinese:C hina N uclear Energy Industry Corporation, T ranslation Section,P.O. Box 2103, BeijingIAEA Publications o ther than in C hinese:C hina National Publications Import & Export Corporation,D eu tsche Abteilung, P.O . Box 88, BeijingS.N.T.L., M ikulandska 4, C S-116 86 P rague 1Alfa, Publishers, Hurbanovo nam estie 3, CS-815 89 BratislavaOffice International de Docum entation e t Librairie, 48, rue Gay-Lussac,F-75240 Paris C edex 05Kultura, H ungarian Foreign Trading Com pany,P.O. Box 149, H-1389 B udapest 62Oxford Book and S tationery Co., 17, Park S treet, Calcutta-700 016 Oxford Book an d S tationery Co., Scindia H ouse, New Delhi-110 001 YOZMOT (1989) Ltd, P.O. Box 56055, Tel Aviv 61560 Libreria Scientifica, Dott. Lucio de Biasio "ae io u " ,Via Meravigli 16, 1-20123 MilanM aruzen Com pany, Ltd, P.O. Box 5050, 100-31 Tokyo International Mirza Book Agency, 65, Shah rah Quaid-e-Azam, P.O. Box 729, Lahore 3 Ars Polona-R uch, C entra la Handlu Zagranicznego,Krakowskie P rzedm iescie 7, PL-00-068 W arsaw llexim, P.O. Box 136-137, B ucharestM ezhdunarodnaya Kniga, Sm olenskaya-Sennaya 32-34, M oscow G-200Van Schaik Bookstore (Pty) Ltd, P.O. Box 724, Pretoria 0001Diaz de San tos, L agasca 95, E-28006 MadridDiaz de San tos, B alm es 417, E-08022 BarcelonaAB Fritzes Kungl. Hovbokhandel, F red sg a tan 2, P.O . Box 16356,S-103 27 StockholmHMSO, Publications C entre, Agency Section,51 Nine Elms Lane, London SW8 5DRJugoslavenska Knjiga, Terazije 27, P.O. Box 36, YU-11001 B elgrade

Orders from countries where sa tes agents have not yet been appointed and requests for information shoutd be addressed directty to:

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