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IAEA-TECDOC-719

Defining initiating eventsfor purposes of

probabilistic safety assessment

JAIASeptember 1993

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DEFINING INITIATING EVENTS FOR PURPOSES OFPROBABILISTIC SAFETY ASSESSMENT

IAEA, VIENNA, 1993IAEA-TECDOC-719

ISSN 1011-4289

Printed by the IAEA in AustriaSeptember 1993

FOREWORD

Safe operation of nuclear power plants is achieved by good design and prudentoperational practice. Both these aspects should be considered in the safety assessment of anuclear power plant. A plant specific probabilistic safety assessment (PSA) models a plantby addressing both design and operation and, in addition, takes account of operatingexperience.

PSA can be used to quantify the safety of a nuclear power plant and it has become awidely used tool for the safety assessment of nuclear plants. Although its methodology isfairly standard, the complexity of the full scope plant specific PSA is such that guidance isstill needed to ensure its completeness. Within the framework of its programme on theinternational promotion of PSA, the IAEA has issued numerous documents on topics rangingfrom guidelines on how to perform PSAs to specific case studies and generic reliabilitydatabases.

The PSA model is structured so as to permit consideration of a spectrum of possibledisturbances to a plant's normal operation (in PSA terminology, initiating events), theavailability of systems designed to cope with a particular disturbance and operator actions inthe course of the event. One of the areas where the level of completeness and the accuracyof the analysis could greatly influence a PSA model is the selection of the initiating events(IBs). The IAEA has developed this document in order to summarize and explain thedifferent approaches to the selection of IBs. It is based on examples taken from several PSAstudies.

This document is primarily directed towards technical staff involved in the performanceor review of plant specific PSAs. It highlights different approaches and provides typicalexamples useful for defining the IBs. The document also includes the generic initiating eventdatabase, containing about 300 records taken from about 30 plant specific PSAs. In additionto its usefulness during the actual performance of a PSA, the generic IE database is of theutmost importance for peer reviews of PSAs, such as the IAEA's International Peer ReviewService (IPERS) where reference to studies on similar NPPs is needed.

In the preparation of this document, the IAEA received support from several MemberStates in the form of material to be included. The main author of the document wasD. Ilberg from Israel, assisted by B. Linquist from Sweden and J. Pereq from Israel.A. Bareith from Hungary wrote the section on WWER reactors. The document wasreviewed by specialists from France, Germany, Hungary and the United Kingdom and byIAEA staff. The IAEA project officer responsible for this report was B. Tomic from theSafety Assessment Section of the Division of Nuclear Safety.

EDITORIAL NOTE

In preparing this document for press, staff of the IAEA have made up the pages from theoriginal manuscript (s). The views expressed do not necessarily reflect those of the governments of thenominating Member States or of the nominating organizations.

The use of particular designations of countries or territories does not imply any judgement bythe publisher, the IAEA, as to the legal status of such countries or territories, of their authorities andinstitutions or of the delimitation of their boundaries.

The mention of names of specific companies or products (whether or not indicated as registered)does not imply any intention to infringe proprietary rights, nor should it be construed as anendorsement or recommendation on the part of the IAEA.

CONTENTS

1. INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 71.1. Purpose of the report . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7

1.1.1. Purpose of defining the initiating events . . . . . . . . . . . . . . . . . . 71.1.2. Using this document for the evaluation of operational occurrences . . 8

1.2. Scope of the report . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 81.3. Structure of the report . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8

2. HISTORICAL BACKGROUND . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 102.1. Development of data on transient initiating events . . . . . . . . . . . . . . . . . 102.2. Development of data on LOG A initiating events . . . . . . . . . . . . . . . . . . 132.3. Development of data on common cause initiators . . . . . . . . . . . . . . . . . 14

3. SELECTION OF INITIATORS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18

3.1. Definition of initiating events . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 183.2. Main categories of initiating events . . . . . . . . . . . . . . . . . . . . . . . . . . 183.3. Methodology for the identification of initiating events . . . . . . . . . . . . . . 19

3.3.1. Identification of LOCAs . . . . . . . . . . . . . . . . . . . . . . . . . . . . 193.3.2. Identification of transients and special initiating events . . . . . . . . . 20

3.4. Examples of initiating event lists . . . . . . . . . . . . . . . . . . . . . . . . . . . 263.4.1. Initiating event lists for PWRs . . . . . . . . . . . . . . . . . . . . . . . . 273.4.2. Initiating event lists for BWRs . . . . . . . . . . . . . . . . . . . . . . . . 40

3.5. Completeness of IE lists . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40

4. GROUPING OF INITIATING EVENTS . . . . . . . . . . . . . . . . . . . . . . . . . 504.1. General principles for grouping . . . . . . . . . . . . . . . . . . . . . . . . . . . . 504.2. Grouping of initiating events for PWRs . . . . . . . . . . . . . . . . . . . . . . . 50

4.2.1. LOCAs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 504.2.2. Transients . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 514.2.3. Special common cause initiators . . . . . . . . . . . . . . . . . . . . . . . 52

4.3. Grouping of initiating events for BWRs . . . . . . . . . . . . . . . . . . . . . . . 594.3.1. LOCAs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 594.3.2. Transients . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 594.3.3. Special common cause initiators . . . . . . . . . . . . . . . . . . . . . . . 64

4.4. Impact of various transients on core damage frequency . . . . . . . . . . . . . . 64

5. DETERMINATION OF THE FREQUENCY OF INITIATORS . . . . . . . . . . . 665.1. Approaches to quantification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 66

5.1.1. One and two stage Bayesian updating analyses . . . . . . . . . . . . . . 685.1.2. Mean frequencies of frequent operational occurrences . . . . . . . . . . 685.1.3. Expert opinion on rare events . . . . . . . . . . . . . . . . . . . . . . . . 685.1.4. Frequency estimation by evaluation of failure rates and

mission times . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 68

5.1.5. Fault tree analysis for special rare events . . . . . . . . . . . . . . . . . 695.1.6. Use of experience from other plants . . . . . . . . . . . . . . . . . . . . . 695.1.7. Special attributes of plant and location . . . . . . . . . . . . . . . . . . . 69

5.2. Some examples of IE frequencies in PSA . . . . . . . . . . . . . . . . . . . . . . 71

6. EXAMPLES OF FREQUENCY DETERMINATION . . . . . . . . . . . . . . . . . . 74

6.1. Example of the use of the mean frequencies approach . . . . . . . . . . . . . . 746.2. Examples of determination of the frequency of common cause initiators ... 756.3. Examples of determination of the frequency of LOG As . . . . . . . . . . . . . 77

6.3.1. Maine Yankee PSA . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 776.3.2. Interfacing LOG A frequency based on pipe and valve rupture/

leakage failure rates . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 776.4. Examples of determination of the frequency of transients . . . . . . . . . . . . 78

6.4.1. Maine Yankee PSA . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 786.4.2. German Risk Study (GRS) PSA phase B . . . . . . . . . . . . . . . . . . 80

6.5. Events in shutdown . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 80

7. INITIATING EVENTS FOR WWER PLANTS . . . . . . . . . . . . . . . . . . . . . 83

7.1. Selection of WWER initiating events . . . . . . . . . . . . . . . . . . . . . . . . . 847.1.1. Selection of LOCAs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 917.1.2. Selection of transients and common cause initiators . . . . . . . . . . . 92

7.2. Grouping of WWER initiating events . . . . . . . . . . . . . . . . . . . . . . . . 967.2.1. Grouping of LOCAs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 967.2.2. Grouping of transients and common cause initiators . . . . . . . . . . . 99

7.3. Determination of the frequency of initiators for WWER plants . . . . . . . . . 99

8. HAZARDS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1058.1. Main categories of hazards . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1058.2. Evaluation of hazard frequencies . . . . . . . . . . . . . . . . . . . . . . . . . . . 105

8.2.1. Seismic analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1058.2.2. Fire analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1078.2.3. Flood analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 107

9. DESCRIPTION OF THE DATABASE OF INITIATING EVENTS . . . . . . . . . 1129.1. General definitions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1129.2. Summary of database status . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 113

APPENDIX: PRINTOUT OF THE DATABASE ON GENERICINITIATING EVENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 117

REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 146

LIST OF ABBREVIATIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 149

1. INTRODUCTION

Several reports have been prepared [1] within the framework of the IAEA's programmeto provide Comprehensive Guidelines for Conducting Probabilistic Safety Assessment (PSA).The selection, grouping and frequency evaluation of initiating events (IBs) is one of the mostimportant tasks to be accomplished during a PS A Level 1 study, and this document isintended to provide additional guidance on this issue.

The importance of the determination of IBs has been shown by performing uncertaintyand sensitivity analyses of past PSAs and by some peer reviews of PS A studies. Theinclusion of additional relevant IBs for completeness or revision of the estimates of theirfrequencies may change the results of a PS A study.

This document is intended to aid in the conduct and review of PSAs in Member Statesby providing reference information that can help the specialist in defining IBs. It providesguidance by describing available methods and providing tables that compare approaches usedin various past PSAs and a database of IE lists and frequencies taken from many PSAs whichare available in the open literature.

1.1. PURPOSE OF THE REPORT

The determination of initiating events is an important part of a PS A. IBs directly affectthe core damage frequency in PSAs. They are also a class of operating occurrences reportedat nuclear power plants worldwide. Processing of these occurrence reports on a plantspecific basis provides information on the actual operating experience of that plant. Thus,the information provided in this document allows comparisons of plant specific operatingexperience with ranges and frequencies of IBs considered in PSA studies of similar types ofplant.

This report is intended to provide the reader with a comprehensive overview ofdifferent approaches to defining IBs which have been taken in a number of PSA studies. Itdoes not recommend the use of any specific approach. It is left to those performing the PSAto select the approach which is the most appropriate for their particular application, and then,if necessary, to acquire more information from relevant literature listed in the references.

1.1.1. Purpose of defining the initiating events

An IE is a postulated event that could occur in a nuclear power plant. It is anoccurrence that creates a disturbance in a plant and has the potential to lead to core damage,depending on the successful operation or failure of the various mitigating systems in theplant.

The performance of PSAs to develop a comprehensive plant model requires ascomplete a list of IBs as possible. This list determines the points of departure of the accidentsequences that would be studied in the search for the dominant sequences that may lead tocore damage. Thus the frequency of IBs has a direct impact on the results for core damagefrequency, as well as on the spectrum of importance of individual components or actions.

The consequences of ill defined IBs are various. A missing IE in a PSA means thatthe core damage frequency would be underestimated by the value of the IE frequencymultiplied by the conditional probability of safety system failure given the occurrence of the

IE. A larger list of IBs than necessary (for example, due to inappropriate grouping) wouldresult in waste of resources because of the analyses of additional unnecessary accidentsequences. An IE list that is incomplete or is insufficiently precise in its frequencydetermination would generally result in an incorrect estimation of the core damagefrequencies.

1.1.2. Using this document for the evaluation of operational occurrences

This document can assist in evaluating operational occurrences in three ways:

(a) It describes approaches used for identifying occurrences and assigning them to knownor new basic initiating events, or broader categories of initiating events.

(b) It describes methods for evaluating IE frequencies from experiences of operatingoccurrences in a specific plant over a defined period of tune and provides examples ofthe application of these methods. It can be used for selecting generic prior distributionfor Bayesian updating of plant specific experience.

(c) It provides an extensive database of data from other power plants to help put the plantspecific experience in perspective.

1.2. SCOPE OF THE REPORT

This report covers the topics of defining IBs for nuclear power plants with PWR andBWR type reactors (including WWERs). It does not cover CANDU type reactors or oldertypes or types of relatively limited distribution.

The "Procedures for Conducting PSAs of Nuclear Power Plants" [1] classify the IBsinto internal IBs and hazards1 (internal and external hazards). Internal IBs are hardwarefailures in the plant or maloperation of plant hardware through human errors or due toman-machine interface problems.

External hazards (often called external events) are events originating outside the NPPthat create extreme environments common to several plant systems. Internal hazards, whichare originated within the station boundaries, create similar extreme environments, and includeinternal flooding, fire and missiles. Loss of connection to the grid (complete or partial) isconsidered here as an internal IE. The scope of the document is confined to internal events.However, an introductory section on hazards is given in Section 8.

1.3. STRUCTURE OF THE REPORT

The report is organized as follows:

Section 2 gives the historical background for defining IBs by considering variousapproaches that have been used in previous PSAs. It covers the methods that form the basisfor approaches described in Sections 3-5.

1 'Hazards' is the term used in this report for all types of 'external events' such as fires, floods (inside andoutside the plant), earthquakes, etc.

Section 3 provides a review of selected approaches in some recent PSAs and describesthe main methods used to ensure completeness in IE selection. A discussion of completenessof the IE list is given and tables of several lists that have been used in the past are provided.

Section 4 similarly deals with the approaches to IE grouping. A review of severalgrouping structures is provided by comparing several recent PSA approaches for BWRs andPWRs. Lists of IE groupings are combined from some of these PSAs to provide the PSAspecialist with 'maximal' and 'minimal' lists to compare. The relative importance of someof the IE groups used in various PSAs is shown for BWRs and PWRs respectively.

Section 5 covers the determination of IE frequencies. The various methods aredescribed and some examples are given. Several sources of data for the evaluation of IEfrequencies are outlined.

Section 6 provides examples of derivations of the frequency of initiators taken fromseveral PSAs. It includes examples of estimations of the frequency of:

- LOCA initiators;- transients;- special common cause initiators;

ATWS initiators.

An example of the treatment of initiators for non-power operating modes is alsoprovided in Section 6.

Section 7 is entirely devoted to WWER reactors. These reactors are treated separatelybecause the level of development of definition, grouping and frequency determination is, forthe time being, somewhat lower than for PWR reactors operated in other countries mentionedin this report.

A brief introduction to the treatment of internal and external hazards is provided inSection 8.

Section 9 covers the initiating event database. It describes the structure, the recordformat and the data sources used for the database.

The Appendix provides the printout of the database of initiating events compiled by theIAEA.

2. HISTORICAL BACKGROUND

The concept of initiating events (IBs) was introduced in the United States NuclearRegulatory Commission's Reactor Safety Study (RSS) in 1975 [2] together with the event treemethodology. Twenty transient IE categories and four LOCA related categories were selectedfor BWRs. Twenty-three transient and six LOCA related categories were selected for PWRs.The basis for the selection of the LOCA IBs was the plant response (based on systems/trainswhich are required to work and/or estimated timing of the accident). Consistent with plantresponse three sizes of LOCAs were selected:

- Large LOCA: 6 inches (15.2 cm) up to double ended largest pipe diameter — denoted"A";

- Small LOCA: 2-6 inches (5.1-15.2 cm) equivalent diameter — denoted SI;- Very small LOCA: 1/2-2 inches (1.2-5.1 cm) equivalent diameter — denoted S2.

In addition, reactor pressure vessel (RPV) rupture was considered for both PWRs andBWRs.

For PWRs two additional LOCA related IBs were selected:

Interfacing LOCA (known as the RSS 'event V);- Steam generator (SG) tube rupture.

The definition for the various LOCA categories was used in the event tree analysiswithout further grouping. Not so for the transients. A general division into anticipated andunanticipated transients was made. Almost all anticipated transients were modelled as onegroup in both PWR and BWR event trees. Similarly, unanticipated transients were modelledas a second group. An additional special initiating event was considered, namely the loss ofoff-site power for more than 30 minutes.

2.1. DEVELOPMENT OF DATA ON TRANSIENT INITIATING EVENTS

Following the use of operating experience in the RSS, the Electric Power ResearchInstitute (EPRI) published in 1978 its first study of anticipated transients [3] referred to asNP-801. NP-801 includes a compilation of operational occurrences from 12 BWRs and 30PWRs. For BWRs it reported 459 events in 37 selected categories of different transient IBs(compared to 20 in the RSS) and for PWRs it reported 1000 events categorized in 41 differentIBs (an expansion of the 23 found in the RSS). The data in the NP-801 covered NPPexperience up to 1978, which was the equivalent of 41 and 131 plant-years for BWRs andPWRs respectively. The NP-801 data was used extensively in probabilistic risk assessment(PRA) performed in the period 1978-1983 when the update of the NP-801 report, theNP-2230 [4], became available.

The NP-801 has many drawbacks that the NP-2230 has treated partially and the laterupdate of transients event data by EG&G [5] further improved:

(a) NP-801 used an 'effective in-service date' as supplied by the utilities. NP-2230uniformly used the first day of commercial operation as the starting point for reportingplant anticipated transients. Because of this change, 137 events of NP-801 wereexcluded from the NP-2230 update.

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(b) NP-801 reports 191 events within 37 plant-years that occurred in the years subsequentto the first year of plant operation (less than half the total number of events). NP-2230reports 647 events in 85.5 plant-years (70% of the total).

It is clear that the NP-801 included very early periods of plant operation, i.e. fromcriticality to commercial operation, whereas the NP-2230 included events that occurred onlyafter commercial operation was initiated. In general this is about half a year later. Table 2.1gives a comparison of the evaluation of selected initiator frequencies based on these two datasources [6].

The NP-2230 reported 903 events for BWRs and 2093 events for PWRs under the sameIE categories as in NP-801. The number of plants covered increased to 16 BWRs and 36PWRs with 101.5 plant-years and 213 plant-years for BWRs and PWRs respectively. Thisdata was used in PRAs after 1983.

The EG&G study [5] updated further the transient IE database for BWR and PWRplants. It included for the same IE categories 251 BWR plant-years with 1832 events and 423PWR plant-years with 3574 events. There most of the events come from the years followingthe first two years of plant operation. This database was selectively used in the NUREG-1150[7] type PRAs performed in the period 1986-1989.

The first four columns of Table 2.1 show original BWR-PRA estimates, based on NP-801 [3]. The next four columns represent results obtained from applying the samemethodology to the more recent data source (NP-2230). The two last columns present resultsusing the updated source and the two stage Bayesian methodology [8]. It can be seen thatmost of the increase in the BNL initiator frequencies derives from the updated experience ofBWR related events, rather than from the use of the Bayesian methodology.

The BWR-PRA [12] differentiated between the impact of failures during the first yearof plant operation and failures occurring in later years. However, in the review [6] it wasargued that the database used in NP-801 was not sufficiently refined for this purpose. Thelater update, given in NP-2230, showed that the impact of ignoring the first year of plantoperating experience causes a reduction of about 20% in initiator frequencies (see last twocolumns of Table 2.1). The 'weighted average' approach utilized in the BWR-PRA weightedthe data from the first year as (1/35) and the data from subsequent years as (34/35) (thisapproach does not consider ageing).

Table 2.1 shows that the number of shutdowns due to anticipated transients is higherthan that experienced in recent years. This is apparently because the NP-2230 databaseextends to 1981 only. The updated review of the experimental data [6] published in 1985may show a reduction in the frequencies of IBs. The trend of reduction in transient IEfrequency continues today in many power plants, and therefore the latest data sources shouldbe utilized whenever possible.

The EPRI reports were made available as part of the EPRI programme on anticipatedtransients without scram (ATWS). For this purpose they include a section for lEs that occurat a power level above 25% of full power. This is done because transients that occur at alower power level do not challenge the reactor shutdown system. This is further discussed inan example in Section 5.1 which treats frequency determination for ATWS.

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TABLE 2.1. COMPARISON OF SELECTED INITIATOR FREQUENCIES AND SOURCES OF DIFFERENCES

SNPS-PRA [12]EPRI-NP-801 Data

Transient

Loss ofCondenserVacuum (2,4,8)

Turbine Trip

MSIV Closure (5)

Loss of FW (22)

LOOP (31)

IORV (11)

CRW (27,28)

TOTAL

1stYear

1.6

16.9

2.2

0.6

0.4

0.7

0.1

22.5

Subseq.Years

0.38

4.14

0.19

0.16

0.11

0.08

0.03

5.09

AllYearsAverage

0.67

7.3

0.67

0.27

0.16

0.20

0.04

9.3

SNPS-PRAWeightedAverage*

0.41

4.46

0.24

01.8

0.08+

0.09

0.03

5.49

1stYear

1.0

13.4

1.67

0.27

0.13

0.53

0.13

17.1

Subseq.Years

0.38

6.39

0.27

0.11

0.12

0.15

0.10

7.52

BNL Review: [6]EPRI-NP-2230 Data

AilYearsAverage

0.47

7.39

0.47

0.13

0.12

0.21

0.11

8.9

WeightedAverage

0.40

6.59

0.31

0.12

0.08+

0.16

0.10

7.76

BNL Review: [6]Two-StageBayesian

Subseq.Years

0.40

6.85

0.29

0.11

0.12

0.19

0.11

8.07

AllYears**

0.50

7.89

0.57

0.13

0.15+ +

0.25

0.12

9.65

Numbers in parentheses correspond to EPRI NP-801 [3] categories+ Based on SNPS grid data * Used in the PRA+ + Based on NSAC-80 report [10] ** Used in the BNL review

The validity of the NP-2230 [4] data was reviewed in two cases:

(a) In the Oconee PRA [9] the data of NP-2230 was checked against the licensee eventreport (LER) data of the Duke Power Co., the owner of the plant. It was found thatin most of the cases a good agreement exists. Only in the division between 'partial lossof MFW and 'turbine trips', significant differences were found. Many more eventswere categorized as turbine trips in the Oconee PRA than as the partial loss of MFW.This was based on the in-plant data records, which was considered more accurate thanthe EPRI data.

(b) A thorough comparison was performed in the EG&G study [5] of the NP-2230 databaseand the 'NRC Gray Book' [11] database for 11 plants which were selected for thecomparison. For each plant selected, the events that occurred during the third andeighth year of operation were carefully compared. It was found that 66 (27%) eventswere categorized differently based on the event description in each of the two sourcescompared. However, about one third of the discrepancies were because the Gray Bookevent description contained less information than the NP-2230 description. The finalconclusion was that on the whole the NP-2230 data was found to be valid and isindicative of US commercial NPP experience. This is because the deviations weresmall and, in general, did not cross 'borders' of the broad groups of transients used inthe PRA studies. Another important conclusion was that a sufficient amount of detailsin the event descriptions, provided by the plants, is crucial for a correct categorization.

2.2. DEVELOPMENT OF DATA ON LOCA INITIATING EVENTS

Unlike transients, the categorization of LOCA IE categories has not much changed sincethe original RSS definitions. The main changes in definitions were the inclusion, in most ofthe newer PSAs, of the SG tube rupture (rather than the SG rupture in the RSS which wasnot further analysed there) and a group of very small LOCAs at various locations (rather thancontrol rod drive (CRD) pump leakage in the RSS). Table 2.2 compares several frequenciesused in PSA for the same LOCA IE category. It should be noted that break size definitionsof various size LOCAs are not uniform in PSAs and different for PWRs and BWRs inparticular.

The Reactor Safety Study (RSS) estimated LOCA frequencies by inference from genericdata from pipe breaks in the non-nuclear industries. This is the basis for the mean valuesshown in Table 2.2 for the RSS. The reactor pressure vessel (RPV) rupture probability wasalso based on non-nuclear vessel experience. While the frequency of the latter did not changemuch in newer PRAs (most of them still use the RSS value), the LOCA frequencies havebeen reevaluated in the newer PRAs (e.g. the Midland PRA [13]).

Oconee [9] and Seabrook [14] PRAs used experiential data for the evaluation of partof the LOCA frequencies rather than the pipe break data used in the RSS. The Oconee PRAconsidered the following events in a population of 35 plants:

- Large LOCA (A): No event occurred;Small LOCA (S): One event that occurred at Zion Unit 1 in 1975;

- SG rupture (R): Three events of SG tube ruptures with leakage ratesgreater than 100 gpm occurred: Surry Unit 2 (Nov.1972), Point Beach Unit 1 (Feb. 1975), and PrairieIsland Unit 1 (Oct. 1979).

13

A two-stage Bayesian analysis was applied to the above generic data andto the Oconee plant specific experience which reflects none of the aboveevents in any of the three units on-site. A review of the Oconee PRA [15]added another relevant event:

Very small LOCA (VS): One event that occurred at H.R. Robinson Unit 2(May 1975).

This has added a frequency of 3 x 10~3 (see Table 2.2). The B/W owner group[17] based their estimate of 'VS' on the precursor study [55] which introduced the 'Robinsonevent' mentioned above.

To summarize, the development of frequency evaluation for RPV and large LOCA hasnot changed significantly since the RSS was made. On the other hand, very small LOCAsand interfacing LOCAs received additional attention and some new studies have been madewhich are discussed further in Sections 5.1.4 and 6.3 which provide examples of the treatmentof the frequency of LOCA type IBs.

2.3. DEVELOPMENT OF DATA ON COMMON CAUSE INITIATORS

The RSS has already treated some IBs as special common cause initiators (CCIs). Twoexamples are the loss of off-site power for more than 30 minutes and the 'V event. ThePSAs that followed the RSS have added more CCIs, and in general the CCIs are of a plantspecific nature. Some of the CCIs treated in PSAs are:

- Loss of instrument air;- Loss of DC power bus;- Loss of service water or component cooling water system;- Loss of AC power bus(es);- Steam line break;- Reactor water level instrument line failure.

This section describes the treatment of one specific CCI common to all PSAs, namelythe loss of off-site power initiator. Other approaches to CCI evaluation are covered inSections 5.1.5 and 6.2.

Loss of off-site power (LOOP) experiential data have been reviewed in four studiessince 1980:

(1) Scholl [24] reviewed the data received from licensees following a June 1980 NRCrequest to submit licensee experience with LOOP events. This review includes a listof 109 occurrences of LOOP events.

(2) The results of a LOOP study were summarized in EPRI-NP-2301 [25] which uses datacollected from 47 nuclear power plant sites. The report presents frequency and durationof LOOPs based on 45 occurrences through April 1981, representing 375 plant-yearsof experience.

(3) A NSAC/ORNL study was reported in NSAC-80 [10] which covered 52 nuclear powerplant sites, for the period prior to December 1983. It summarizes 47 LOOP events in530 plant-years.

14

TABLE 2.2. A COMPARISON OF LOCA FREQUENCIES IN VARIOUS PRAs [16]

LOCA TypeInitiator (*)

PRA

ARKANSASIREP f!8lMIDLANDPRA f!31B/W OwnerGroup T171OGONEEPRA f9lNUREG-4550PWR [401NUREG-4550BWR 1361LIMERICKPRA 1201SHOREHAMPRA F121BWR-6

1211SEABROOKPRA f!4lRSS-PWR

F21RSS-BWR

f2lPALUEL 1300PSA T231German RiskStudv F221

Very Small SmallLOCA LOCA (**)(<0.5 or 1") (0.5/1" to 2/3")

VS

0.0205.0E-3

8.3E-3—

0.0200.030—

————

CRD Pump—2.8E-3

* VS: Very Small - less than 1Greater than 3 inch diameter ifnoted that the values of breakgeneral indication only.** Note: 2/3 should read 2 or

S

6.9E-43.3E-34.0E-4

3.0E-3l.OE-3

3.0E-30.0108.0E-31 . 2E-3

0.0172.7E-32.7E-32.0E-31.4E-4

MediumLOCA(2/3" to 6

M

1.6E-44 . 7E-4

—

—l.OE-33.0E-42.0E-33.0E-36 . 7E-44.7-48.1E-48.1E-43.0E-47.5E-5

LargeLOCA

") (>6")

A

8.7E-52.0E-4

—

9.3E-45.0E-4l.OE-44.0E-47.0E-42.1E-42.0E-42.7E-42.7E-4IE-4

<E-7

RPV Interfacing SG TubeRupture LOCA Rupture

RPV ISLOCA R— — —

7.7E-7 0.014-- — 0.0171.1E-6 1.4E-7 8.6E-3l.OE-8 l.OE-6 —l.OE-8 <3.0E-7 —— — —

3.0E-7 1.8E-7 —1.7E-7 —

2.7E-7 1.8E-6 0.014l.OE-6 1.1E-5 —l.OE-6 — —— 8.6E-3— <E-7 6.5E-3/1E-5

.5 inch diameter break; S: Small LOCA - less than 3 inch diameter break; L .-Large LOCA -"Medium" is not considered, otherwise greater than 6 inch diameter size. It should be

sizes can vary from PWRs to BWRs and among different PSAs, and are given here for3" diameter break.

(4) An USNRC study [26] for the resolution of the ' Station Blackout' issue was reportedin NUREG-1032. The study covered 52 NPPs (all the US NPP sites of December 1983excluding three with one off-site power connection). It summarizes 55 events in 533plant-years.

The review of the data sources has resulted in several findings:

(a) The Scholl database is rather conservative and needs additional evaluations prior to itsutilization in PRAs.

(b) The NP-2301 database is more realistic. A few events are apparently missing from thissource. Its recovery probability information is relatively conservative for use in PRAs.

(c) The N SAC-80 database appears to be suitable for realistic PRA analyses. Itrecommends exclusion of several total LOOP occurrences during shutdown which itjudges to be 'impossible' during operation. It can however be assumed that these areinadvertent human errors that should be included in LOOP frequency evaluation forcompleteness. The later NUREG-1032 considered them in its statistics.

The above three studies reported the LOOP events by plant and per geographicalregions having similar weather conditions and an interconnection agreement with respect tokeeping a reliable electric supply in that region. Another approach was proposed by a laterstudy:

(d) The NUREG-1032 data [26] are based on almost the same database as the NSAC-80.It includes the 'shutdown' events as well. The main improvement of this study is thatit provides a breakdown of all the LOOP events into well defined causes which allowstailoring of the LOOP frequency of a new plant according to its design and also allowsfor evaluating the improvement that may be expected by a design change in an olderplant (see Section 5.1.7 for more details on this approach).

All these data sources and approaches have been used by PSAs in the USA. The LOOPis considered a CCI when the conditions of no electric power are continuing for a long timeperiod (no recovery of off-site power). When off-site power is recovered within a short time,it is considered a transient similar to loss of condenser, because condenser cooling would belost in this case. Therefore, also data on recovery times of LOOP are part of the informationneeded for quantification of its CCI frequency. The data for recovery times in the USA havealso been developed in parallel to the LOOP event data discussed above. NUREG/CR-5032[27] is the more recent one for the USA.

One of the first works providing LOOP recovery data was the EPRI study [25]. TheNSAC/ORNL study [10] provided more precise recovery information on events in the USAuntil December 1983. The NUREG-1032 [26] study divided the NSAC/ORNL [10] data(withsmall modifications) into three subgroups.

- Severe weather type LOOP;Grid related LOOP;

- Plant centred (hardware failure related) LOOP.

16

For each of these subgroups the study provides a frequency of occurrence versusduration plot. The plant centred LOOP occurrences are the most frequent cases but most ofthem are recovered within half an hour [27]. Therefore, they may be treated like internalevents. The grid and weather related LOOP have lower frequency and larger duration andare therefore treated more like the special CCI group of IBs.

17

3. SELECTION OF INITIATORS

3.1. DEFINITION OF INITIATING EVENTS

An initiating event (IE) is one that creates a disturbance in the plant and has thepotential to lead to core damage, considering successful operation or not of the variousmitigating systems in the plant. The following definition of IE has been taken from Refs [1]and [28].

"An initiating event is an incident that requires an automatic or operator initiatedaction to bring the plant into a safe and steady-state condition, where in theabsence of such action the core damage states of concern can result in severe coredamage. Initiating events are usually categorized in divisions of internal andexternal initiators reflecting the origin of the events".

This section deals only with internal events. Internal hazards (such as internal floods)and external hazards (such as seismic events) are covered in Section 7. The initiating eventsconsidered here are only those normally resulting in an automatic or manual scram andoccurring above a certain power level (usually 5 to 25%).

3.2. MAIN CATEGORIES OF INITIATING EVENTS

The internal initiating events may be looked upon as consisting of three main categories:

(A) LOCAs;(B) Transients;(C) Special common cause initiating events (common cause initiators (CCIs)).

Some of the recent PSAs (for example the Paluel PSA [23]) devoted substantial effortto the events during plant shutdown. A brief introduction to these specific events is providedin Section 6.5.

(A) LOCAs

The loss of coolant accident (LOCA) initiators include primary system breaks resultingin loss of primary coolant. Pipe breaks and ruptures of different sizes, inadvertent openingand failures to re-close (stuck open) of valves are being considered in this category.

(B) Transients

The transient initiating events are those which introduce the disturbance in normal plantoperation, without loss of primary coolant and which require an automatic or manualshutdown of the reactor. Typical examples of transient initiators include disturbance infeedwater flow, turbine/condenser, reactivity control, reactor recirculation, etc. Certaindisturbances in some of the support systems will also fall into this category.

(C) Special common cause initiating events

Special initiating events are events which, in addition to requiring reactor shutdown,simultaneously disable one or more of the mitigating systems required to control the plantstatus following the initiator. Typically, they are unique to the plant being analysed.

18

Initiating events which are common to most plants and have a typical CCI character areevents such as:

- Loss of off-site power;- Loss of DC power;- Spurious containment isolation;- Loss of instrument air;- Loss of component cooling.

Even though some of the LOCAs may have CCI character (damage to equipment dueto pipe whip, environmental influence (e.g. high temperature or humidity) on safety systems),LOCAs are not normally looked upon as CCIs.

3.3 METHODOLOGY FOR THE IDENTIFICATION OF INITIATING EVENTS

3.3.1. Identification of LOCAs

Loss of coolant accidents (LOCAs) are usually identified by studying the primary andsecondary system equipment, piping and valve arrangement. Due to the different sizes ofpiping and valves as well as different mitigating systems available, a number of LOCAs breaksizes are normally assumed. The specific boundaries of the LOCA event categories for eachplant are established on the bases of overall mitigating system performance requirements (orsafety system success criteria) for various cases of LOCAs break sizes.

The mitigating systems usually provide protection for different sizes of steam versusliquid breaks. Therefore, different boundaries may be used for small and medium LOCAs ofliquid and steam line breaks (for BWRs).

In general the categories used for LOCA initiators in PSA reflect small variation amongpast PSAs. There are in use between three to five basic size categories taken from thefollowing list:

- Very very small LOCA (e.g. less than 1 inch equivalent inside diameter);- Very small LOCA (e.g. 1-2 inches equivalent inside diameter);- Small LOCA (e.g. 2-3 inches equivalent inside diameter);- Medium LOCA (e.g. 3-6 inches equivalent inside diameter);- Large LOCA (e.g. greater than 6 inches equivalent inside diameter).

Differentiation between hot leg and cold leg break locations are generally ignored inPSA studies.

In addition, the following LOCA initiators are also being considered in most PSAs:

- Reactor pressure vessel rupture;- LOCA outside containment (loss of coolant accident via the interface of low pressure

system with RCS. In this case, in addition to unrecoverable loss of coolant, thecontainment is bypassed);

- Steam generator tube rupture (PWRs only).

In several cases of steam generator tube rupture only one tube rupture is considered.Even though this is a very small LOCA the plant response is in general different from the

19

very small LOCA case (due to filling of affected SG and eventually overpressurizing it) and,in addition, a path to bypass containment is created in this case, which makes this initiatorunique.

Some PSAs consider specifically the following initiators:

- Reactor coolant pump seal leakage or failure (in PWRs);- Control rod drive system leakage or failure (in BWRs).

In other PSAs they are a part of the 'very small LOCA' initiator presented above.

3.3.2. Identification of transients and special initiating events

In order to obtain a comprehensive list of transients and special IBs, a number ofmethods and approaches have been used in PSAs. An overview of the methods utilized isprovided in Tables 3.1 (BWR) and 3.2 (PWR). The tables include the following mainmethods:

(A) Engineering evaluation or technical study of plant;(B) Reference to previous PSAs;(C) EPRI list of IBs (such as EPRI-NP-2230 or NUREG/CR 3862);(D) Logical classification: MLD, energy balance, barrier analysis;(E) Plant energy balance fault tree;(F) Analysis of operating experience for actual plant;(G) Failure mode and effect analysis;(H) Other methods.

These methods are described below.

(A) Engineering evaluation or technical study of plant

In this approach, the plant systems (operational as well as safety) and majorcomponents are systematically reviewed to see whether any of the failure modes (e.g. failureto operate, spurious operation, disruption, collapse) could lead directly, or in combination withother failures, to significant disturbances of plant operation, requiring operation of mitigatingsystems. Partial failures of systems should also be considered since, although they aregenerally less severe than complete failure, they are of higher frequency and are often lessreadily detected. This is, in principle, similar to a classical engineering approach taken duringthe plant's design aimed towards providing safety systems to cope with the spectrum of designbasis accidents. For PSA purposes, the initiators which are beyond design basis (for example,due to their low frequency of occurrence) are also taken into account.

In order to aid the engineering evaluation it may be helpful to structurize the initiatingevents by means of categorization trees, as has been done in the Swedish projectSUPER-ASAR [29]. The basis for this structurization is the availability of main functionssuch as the primary system integrity, off-site power, feedwater and condensation in the maincondenser.

This approach was at least partially utilized in many PSAs, and it is an indispensablemethod in identifying special, common cause IBs which are unique to a particular plant.

20

TABLE 3.1 APPROACHES USED FOR THE SELECTION OF lEs IN VARIOUS PS As (BWRs)

Method for Engineering Reference EPRI ListNPP Identifying Evaluation to previous of IBsPSA lEs: or Technical PSAs (such as

Study of Plant EPRI-NP-2230)

LogicalClassification:MLD, Energy

Analysis ofOperatingExperience

Balance, Barrier for ActualAnalysis Plant

OtherIdentifi-cationMethods

Peach Bottom, [19]Unit 2(-89)

Grand Gulf, [36]Unit K-89)Ringhals 1 [29]

Barsebaeck 1 [29]Oskarshamn 1 [29]Oskarshamn 2 [29]Forsmark 3 [29]Forsmark 1/2 [29]Shoreham [12]

Browns Ferry [30]K-82)Limerick 1 [20]

Big Rock Point[46]* For CCI only

GE-NEDO 2A708AASEP ListGE-NEDO 24708AASEP List

The Swedish Energy Conn.F3 - Study

FSAR

toto TABLE 3.2. APPROACHES USED FOR THE SELECTION OF IBs IN VARIOUS PSAs (PWRs)

Method for Engineering ReferenceNPP Identifying Evaluation to previousPSA IBs: or Technical PSAs

Study of Plant

EPRI Listof IBs(such asEPRI-NP-2230)

LogicalClassification:MLD, EnergyBalance, BarrierAnalysis

Analysis ofOperatingExperiencefor ActualPlant

OtherIdentificationMethods

Ringhals 2 [29] +Oconee NSAC [9] +Seabrook (-83)[14]Midland [13]Surry, Unit 1 (-86) [40]

Galvert Cliffs (-84) [41]Arkansas 1(-82)[18]Sizewell B(-82) +*[50]

FMEAFSAR

List of SubtleInteractions (SNL)ASEP ListFMEAFMEAFSARRESAR

German Risk [35]Study (Phase A)German Risk [22]Study (Phase B)Zion [44]

Maine Yankee [45]Yankee Rowe [31]Connecticut Yankee(56)

FSAR & IndustryExperienceFMEA

FMEA

For CCI

(B) Reference to previous PSAs

A large number of PSAs as well as PSA reviews are available today. It is very usefulto refer to the lists of IBs presented in these PSAs, especially those for similar reactors.Reference to available IE lists may serve as the starting point in compiling a list of plantspecific IBs. This approach was utilized in many PSAs.

(C) EPRI list of IBs

The lists of IE categories for BWRs and PWRs provided in EPRI-NP-2230 [4] andNUREG/CR-3862 [5] was used as the starting point for IE selection in a large number ofPSAs. The lists are derived by analyzing operating experience of a few hundred reactor-yearsin the USA. Therefore, the lists can be considered as one of the best sources for providinga generic IE list for PSA of a new plant of similar design and reasonably similar operatingpractice.

(D) Master logic diagram

The so-called master logic diagram (MLD) is similar to a fault tree. It presents a modelof a plant in terms of individual events and their combinations. A typical top event may bethe 'significant release of radioactive material'. It develops into a plant level logic structurewhose basic input events are the initiating events. One example of an MLD is shown inFig. 3.1 taken from Ref [7]. The initiating events are the lowest level of the tree.

The particular advantage of the MLD method is that the issue of completeness is putinto a more tangible perspective compared to other methods. However, in the cumbersometask of listing the specific causes of initiating events, the possibility of incompleteness stillexists. Examples of the use of MLD for selection of IBs can be found in Seabrook andYankee Rowe PSAs [14, 31].

(E) Plant energy balance fault tree (EBFT)

A transient condition in a nuclear reactor implies an imbalance in the state ofequilibrium in the transfer of thermal energy from the reactor core to the environment.Therefore, a fault tree analysis of the plant energy balance (sometimes referred to as heatbalance) is a valuable tool for deriving initiating events. The energy imbalance fault treeanalysis may be carried out for any degradation in thermal equilibrium including those fromfull power, partial power, hot standby and cold shutdown. The typical top event is 'imbalancein energy transfer causing a plant's IE to occur'. One example of an energy balance fault treeis shown in Fig. 3.2. As for the MLD, the basic input events derived from the analysis arethe initiating events at the equipment level.

Relative to the MLD it is unnecessary to include additional system failures in this typeof fault tree. There is no need to introduce associated simplifications in the logic and, hence,the deductive power of fault tree analysis need not to be diminished.

(F) Analysis of operating experience for actual plant

In this approach the operational history of the plant in question and of similar plantsis reviewed to search for any events which should be added to the list of IBs. This approach

23

tLAAGCLOO

3MfOiuM100

3SHALLLOCA

4LfAKAGCtoSCCOftOAAVCOOLAKT

fLOB OFPHIMA8YCOO LAN!FLOW

•LOS5OFFtfOFLOW

7lOSt OFSTEAMFLOW

mtifiNM.INITIATOAS

IHtTl TORS

lHti*NALtINITIATORS

ihlllA OKS

tMKKNAVINITIATORS

*

IKIT1ATOH&•«

IhTtRhAl.IHlTlATOftS

*•

IKtTtATORS

mTEAHALINITIATORS

•

INtTlATOHJ

•

INYCnNALINITIATORS

•

INITIATORS

1 NT E PIN XVINITI TORS

iNtTI TORS

INTERNALINITIATORS

•••

INITIATORS

INTERNALINITIATORS

*

INITIATOR4

*

INTERNALINITIATORS

»•

*

INITIATORS

•

INTERNALINITIATORS

INITI TORS

INITIATORS

INITI TORS

INITIATORS

•

INITIATORS

•

INITIATORS•*

INITIATORS

INITIATORS

•••

INITIATORS••

FIG. 3.1. Example of a master logic diagram (MLD).

LEVEL 1

LEVEL2

1ENERGYIMBALANCEBETWEENCORE & RCS

*

ENERGYIMBALANCEBETWEENRCS & SCS

ENERGYIMBALANCEBETWEEN

SCS &PLANT OUTPUT

LEVELS CHANGE IN COREHEAT GENERATION

CHANGE IN RCSHEAT REMOVAL

FROM RCS

LEVEL4

1INCREASE INCORE HEAT

GENERATION

1DECREASE INCORE HEAT

GENERATION

LEVEL5

Reactor TripLEVEL 6 Manual Shutdown

Control Rod Drop

IINCREASE INRCS HEATREMOVALFROM CORE

1DECREASEDRCS HEATREMOVALFROM CORE

INWARDMOVEMENT OFCONTROL RODS

DECREASE INNEUTRON

MODERATION

INCREASE INCOOLANT BORONCONCENTRATION

RCS - Reactor Coolant SystemSCS - Secondary Coolant System

None CVCS Malfunction

toL/i FIG. 3.2. Energy balance fault tree.

is to be looked upon as supplementary and would not, of course, be expected to reveal lowfrequency events but could show common cause initiating events of higher frequency. Suchan approach is used also in precursor studies [32] to identify the IBs and to estimate thecorresponding frequency.

(G) Failure mode and effect analysis

Failure mode and effect analysis (FMEA) is a powerful technique used extensively bythe aerospace industry to reveal component failures that can impose critical effects on systemperformance. Dealing at the component level it becomes very detailed and laborious.Therefore, it is not commonly used in PSA.

Some PSAs have used this methodology to identify IBs within the plant control systemsthat have the potential to fail a mitigating system due to dependencies [31].

(H) Other methods

The most commonly used methods or approaches to selecting IBs were covered in theprevious subsections. In special cases some other plant specific methods were used. One ofthese is the NUREG-1150 approach that has created a generic list of IBs in a special studyand this IE list (ASEP list) was used as a generic list of IBs in all "1150" PSAs [33].

Another special approach is the Swedish SUPER-ASAR [29] project approach whichis presented in Section 3.4.2.

3.4. EXAMPLES OF INITIATING EVENT LISTS

Section 3.3.2 presented several methods for identifying transient and special IBs. Twoof the methods referred to give lists of IBs from other studies. This Section providesexamples of the EPRI lists (see Section 3.3.2.C) and three specific lists that were used inprevious PSAs (two for PWRs and one for BWRs):

For PWRs:

(A) The EPRI list (considers transients only);(B) The Oconee list;(C) The German Risk Study (Phase B) list.

For BWRs:

(A) The EPRI list (considers transients only);(B) The Grand Gulf list.

As stated before, the EPRI lists for PWRs and BWRs are well known and are referredto by a number of PSAs, especially US studies.

The PSAs for Oconee and Grand Gulf provide IE lists that are partly based on the EPRIlist, but in addition considerable work has been done using a few of the other methodsdiscussed in Section 3.3.2 in order to find any additional IE which should be considered.

26

3.4.1. Initiating event lists for PWRs

(A) The EPRI list

The EPRI list of transient initiating events [4] is an expansion of the list presented inthe RSS. The expansion is based on US reactor operating experience data. The data forPWRs includes 2023 events occurring over 213 plant-years at 36 plants. The EPRI PWR listis given in Table 3.3. The number associated with each initiator is referred to in Section 4.

(B) The Oconee list

The selection of the lEs used in the Oconee PSA [9] was performed in several steps.Initially, candidate IBs were proposed, and engineering evaluation performed for each IE.The IE list generated from those two activities was compared to IE lists available from EPRI,RSS, and the RSSMAP study for the Oconee NPP [34]. The LOCA events and theirrespective sizes were selected on the basis of mitigating system requirements. Finally, plantsystems were reviewed to determine relevant special common cause initiators.

The resulting list of initiating events for Oconee is given in Tables 3.4-3.6. Table 3.6shows the categorizations of the Oconee initiators and includes comments from the initiatorselection process, which should clarify how a particular initiator has been treated. Table 3.6indicates the final category (based on mitigation requirements) under which a particularinitiator has been considered in an initiating event broad group that was further treated in anevent tree analysis.

(C) The German Risk Study list

The German Risk Study (GRS) transient IBs [22, 35] are treated in two frequencyclasses: anticipated transients that are derived from operating experience and unlikelytransients. The unlikely transients mainly include cases of steam line related events whichhave large effects on the plant systems. In comparison to these effects, the reactivity typetransients inflict very small impact on the plant systems, and, therefore, they are notconsidered in phase B of the GRS. The list of transient initiators in the GRS are shown inTable 3.7. It should be noted that in many cases the anticipated transients considered in theGRS lead to the opening of the pressurizer relief valve and, in the case of its failure to close,a LOCA event occurs. These cases of LOCA are treated within the LOCA initiating events.

LOCA type IBs are covered in the GRS in great detail as shown in Table 3.8. Itincludes LOCAs of various sizes in the primary cooling system, in the pressurizer, LOCAsin the annulus between the circular containment liner and containment shielding (because theycan impact on important additional systems by flooding) and steam generator tube ruptures.

For cases of failure of the reactor scram, ATWS events are also treated in the GRS.

Text cont. on p. 40.

27

TABLE 3.3. LIST OF PWR TRANSIENT INITIATING EVENTS

1. Loss of RCS flow (one loop)2. Uncontrolled rod withdrawal3. Problems with control-rod drive mechanism and/or rod drop4. Leakage from control rods5. Leakage in primary system6. Low pressurizer pressure7. Pressurizer leakage8. High pressurizer pressure9. Inadvertent safety injection signal10. Containment pressure problems11. CVCS malfunction-boron dilution12. Pressure, temperature, power imbalance-rod-position error13. Startup of inactive coolant pump14. Total loss of RCS flow15. Loss or reduction in feedwater flow (one loop)16. Total loss of feedwater flow (all loops)17. Full or partial closure of MSIV (one loop)18. Closure of all MSIVs19. Increase in feedwater flow (one loop)20. Increase in feedwater flow (all loops)21. Feedwater flow instability-operator error22. Feedwater flow instability-miscellaneous mechanical causes23. Loss of condensate pumps (one loop)24. Loss of condensate pumps (all loops)25. Loss of condenser vacuum26. Steam-generator leakage27. Condenser leakage28. Miscellaneous leakage in secondary system29. Sudden opening of steam relief valves30. Loss of circulating water31. Loss of component cooling32. Loss of service-water system33. Turbine trip, throttle valve closure, EHC problems34. Generator trip or generator-caused faults35. Loss of all off-site power36. Pressurizer spray failure37. Loss of power to necessary plant systems38. Spurious trips-cause unknown39. Automatic trip-no transient condition40. Manual trip-no transient condition41. Fire within plant

28

TABLE 3.4. LOCA INITIATORS INCLUDED IN THE OCONEE PRA

Event Description

1. Small-break LOCA

2. Large LOCA

3. Interfacing-system LOCA

4. RPV rupture

5. Steam generator tuberupture

A break or leak 1/2 to 4 inches in effective diameter.These are spontaneous events: induced LOCAs weretreated directly.

A break or rupture greater than 4 inches in effectivediameter except those noted below.

A large loss of coolant through the valves acting as aboundary between high and low RCS pressure.

A loss of reactor-vessel integrity precluding the ability tomaintain coolant inventory.

A rupture of a steam generator tuberesulting in an RCS leak greater than 100 gpm.

TABLE 3.5. SPECIAL INITIATORS INCLUDED IN THE OCONEE PRA

Event Description

1. Loss of instrument-airsystem

2. Loss of service-watersystem

3. Loss of integrated andauxiliary control systems

4. Loss of DC power system

The system may cause a reactor trip and afailure of instrumentation and equipment that may beneeded for a successful response to the trip.

The system may fail by a pipe break or pumpfailure and in addition prevent other safety systemoperation that depend on service water cooling supply.

The Integrated Central System (ICS) iscontrolling feedwater, pressurizer heaters etc., and maycause a transient with loss of a protecting system.

Bus 3TC failure during power operation may result infailure to supply power to a number of pumps in train Asupplying several mitigating systems.

29

TABLE 3.6. CATEGORIZATION OF THE OCONEE INITIATING EVENTS INTOTRANSIENTS AND LOCAs

Potentialinitiating event

Treatment in refinementprocess

Final category

1. Rod drop

2. Inadvertentrod withdrawal

3. Rod ejection

4. Inadvertentboration ordilution

If shutdown is required, the con-ditions after the trip are thesame as those for the reactor tripcategory.

It was determined that the reasonfor rod withdrawal would notaffect the other reactor systems.ICS failures are treated elsewhere.Conditions after the trip are thesame as for the reactor-trip cate-gory since the additional reactivitybefore the trip would not signifi-cantly change boundary conditions.

Detailed analysis not performed.Probabilistic consideration isincluded in the frequency of non-mitigatable vessel ruptures.

Boration would force the reactortoward shutdown. Credibledilutions would result in reacti-vity effects judged to be insigni-ficant with respect to mitigation.Dilutions resulting in substantialreactivity were judged to be ofsmall probability due to the timerequired and the amount of nonbor-ated water required. Core melt dueto a dilution accident was judged tobe dominated by more frequent transients.

Reactor/turbine trip


Event frequencyjudged to be verylow. Consequenceslimited to brief RCSoverpressure.

Vessel rupture


5. Reactor trip Reactor/turbine trip

30

TABLE 3.6. (cont.)



Final category

6. Cold-wateraddition

7. RCPtrip

8. RCP seizure

9. Flow-channelblockage

10. Loss of mainfeedwater

11. Excessfeedwater

The conditions at shutdown arenot significantly differentfrom those of a reactor trip.

Four-pump trip assumed to bedominated by the loss of off-site power or the loss of ser-vice water.

Only four-pump seizure was judgedto be significant with respect toreactor conditions. Simultaneousfour-pump seizure was judged to beof low probability and thereforewould not result in dominantaccident sequences.

The potential for blockage was exa-mined and found to be small due tothe presence of internals at the bottomof the vessel which wouldprevent large objects from block-ing flow channels. Smaller block-ages were judged to not affectdominant core-melt risks.

Due to its gross effects on heatremoval through the secondaryside, this was treated as aninitiator.

Excessive feedwater that resultsin overcooling transients weretreated as a transient category.Other excessive feedwater eventsthat result in a loss of mainfeedwater as the cause of tripwere included in the loss ofmain feedwater category.


Loss of off-sitepower, loss ofservice water

Loss of mainfeedwater

Excessivefeedwater

31

TABLE 3.6. (cont.)



Final category

12. Loss ofcondenservacuum

13. Inadequatemainfeedwater

14. Feedwater orcondensateline breaks

15. Steam linebreaks

Treated separately due to itslong-term affects on main-feedwater availability andbecause it changes the avail-ability of the condenser as aheat sink whereas the loss ofmain feedwater does not.

Some trips result from feedwaterinstabilities that do not de-grade the reliability of feed-water after the trip. These areincluded in the reactor-tripcategory. Another category wasdefined, partial loss of mainfeedwater, which describes thesituation when the operabilityof the main feedwater system isdegraded but not lost (e.g.,one-pump trip or condensate-pump trip).

Treated as an initiating eventdue to the effects on theavailability of condensate tothe main and emergency feed-water systems.

Kept as a separate category dueto its effects on both the pri-mary and secondary systems in-cluding overcooling and lossof steam-dump capability. Sub-categorieswere used to differentiate the responsesto breaks inside or outside thereactor building.

Loss of condenservacuum

Reactor/turbinetrip and partialloss of mainfeedwater

Feedwater linebreak

Steam line break

32

TABLE 3.6. (cont.)



Final category

16. Turbine andcontrol valvemalfunctions

17. Turbine-bypassvalve inad-vertentopening

18. Turbinemalfunction

19. Loss of con-denser cir-culatingwater

20. Small RCSpipe breaks

21. Large RCSpipe breaks

No significant effects on re-sponding plant systems.

The inadvertent opening was in-cluded in the steamline breakcategory.

The only turbine malfunctionsingled out for special studywas turbine-missile generation

Effects of the transient werefound to be included in theloss of condenser vacuum.

Grouped in one category based onan analysis of the responserequired. Breaks less than4 inches in diameter were in-cluded in the small-LOCA-category

All breaks greater than 4-inchdiameter were grouped into onecategory based on required res-ponse and success criteria.Special locations for these LOCAswere reviewed, but the effectswere judged not to be importantwhen compared to the importantmitigation system failure modes.


Steam line break


Loss of condenservacuum

Small LOCA

Large LOCA

33

TABLE 3.6. (cont.)



Final category

22. InadvertentPORV orsafety-valveopening

23. Reactor cool-ant pump sealfailure

24. Control roddrive sealleakage

25. Interfacingsystem lossof coolant

26. Reactor vesselrupture

27. Steam genera-tor tubeleak/rupture

Included in the small LOCA cate-gory. It should be noted thatPORV or safety-valve LOCAs resultingfrom other transients are modelledexplicitly in the event tree.

Included in the small LOCA cate-gory if the seal failure is aspontaneous initiating event.Seal failures resulting frominadequate protection after atrip are modelled explicitly.

This event would have no signifi-cant differences from thesmall LOCA initiating event.

Treated in a special study due tothe specificity of the eventand its effects

Treated as a separate initiatingevent. The separate category isdefined to include ruptures thatcannot be mitigated

The event is defined as its owncategory due to unique operatoractions and differentradiological considerationsfrom small LOCAs

Small LOCA

Small LOCA

Small LOCA

Interfacing systemsLOCA

Reactor vesselrupture

Steam generatortube rupture

28. Chargingexceedsletdown

Inadvertent HPI operation wasjudged to be the most restric-tive case and is treated as aninitiating-event category.Other possibilities aretreated as subsets.

Spurious engineeredsafeguards signal

34

TABLE 3.6. (cont.)



Final category

29. Letdownexceedscharging

30. Inadvertenthigh pressureinjection

31. Failure on oroff ofpressurizerheaters

32. Failure on oroff ofpressurizerspray

33. Loss of off-site power

This transient was judged to bea subset of the small LOCAcategory.

Treated as a separate initiat-ing events category due to dif- engineeredferent boundary conditionscreated by the event.

A review of the ICSidentified the potential for fail-ure causing heaters to fail onand the PORV to fail closed.This event was treated as aseparate initiating event.

Sprays failing off would result ina high pressure trip, but condi-tions after trip should not begreatly different from those ofa reactor trip. Sprays failingon would have effects similarto those of event 26.

Treated as a category due to itseffects on nearly all event-treetop events. The event was sub-divided into switchyard faultsand grid loss due to a differ-ence in the availability of theKeowee overhead for the twocases and because of the differ-ent recovery potentials.

Small LOCA

Spurious

safeguards signal

Spurious lowpressurizerpressure signal


Loss of substationswitchyard andloss of grid orfeeders

35

TABLE 3.6. (cont.)



Final category

34. Loss of powerto necessarysystems

35. Loss of powerto controlsystems

36. Loss ofservice water

Because of the large number of po-tential initiating events in thiscategory, a special study wasconducted to determine the mostsignificant events with respectto their effects on plant sys-tems (see Table 3.5).

These events were examined assubsets of event 29 and event 1of Table 3.5.

This category was analysed forspecific effects.

Loss of servicewater

37. Loss ofcomponentcooling

38. Loss of instru-ment air

The service-water system atOconee provides the most impor-tant needs. Component coolingrequires service water for theheat exchangers. The loss ofcomponent cooling results fromthe loss of service water andwas treated in that category.

Loss of the instrument air systemaffects a number of top events airsignificantly and also changesthe availability of control-room indications to the operator.The air system was reviewed andthe loss of system pressure wasjudged to be bounding case overless severe malfunctions.

Loss of servicewater

Loss of instrument

36

TABLE 3.6. (cont.)



Final category

39. Integratedcontrolsystem power

40. Fires affect-ing necessarysystems

41. Internal flood-ing affectingnecessarysystems

42.

43.

Generatorfaults

Grid distur-bances

44. Administrativeshutdowns

45. MSIV closure(1 or all)

46. ATWS

The ICS analysed in great detailboth as a support system and anevent initiator.

See Section 6 on internal/externalhazards.

The potential for internal plantflooding was recognized and re-viewed. See Section 6 on internal/external hazards.

Does not affect systems requiredfor adequate response.

Grid disturbances that result ingenerator trip do not changeplant response. Other gridfaults were assigned to thegrid-failure initiating event.

These shutdowns were included inthe reactor-trip frequency ifduring the shutdown a trip wasrequired. Shutdowns that pro-ceed orderly and do not requirea trip are considered successfulresponses.

Not relevant to the Oconee plants.

Treated separately (see Section 6.1).

ICS initiators




37

TABLE 3.7. TRANSIENT INITIATING EVENTS AND THEIR ESTIMATEDFREQUENCY OF OCCURRENCE IN THE GERMAN RISK STUDY (PWR)

Transient initiating events GRS phase B GRS phase A[22] [35]

(D) Operational transients (anticipated transients):

14. Loss of emergency power 0.13 0.1

15. Loss of main feedwater withoutloss of heat sink 0.15 0.8

16. Loss of main feedwater andloss of heat sink 0.29 0.3

17. Loss of heat sink withmain feedwater on-line 0.36 -

(E) Steam line break transients (unlikely transients):

18. Large leakage in steam lineinside containment 1.6 x 10~4 -

19. Large leakage in steam lineoutside containment 4.8 x 10~4 -

20. Intermediate leakage in steam lineinside containment 2.7 x 10~5 -

21. Intermediate leakage in steam lineoutside containment 1.1 x 10"4

(F) ATWS

22. ATWS by loss of main feedwater 4.7 x 10"6 4 x 10"5

23. ATWS by loss of emergency power 3.4 x 10"6

24. ATWS by loss of circulating water 7.5 x 10~6 2.5 x 10"5

25. ATWS by other transients 2.3 x 10~5

38

TABLE 3.8. LOCA INITIATING EVENTS AND THEIR ESTIMATED FREQUENCY OFOCCURRENCE IN THE GERMAN RISK STUDY (PWR)

LOCA initiating events GRS phase B GRS phase A[22] [35]

(A) Leakage in the primary cooling system piping:

la. Large LOCA (>10") <10'7/year 2.7 x l(T4/yearIb. Intermediate LOCA (4-10") <1(T7

2. Small LOCA 1 (1.6-4") 9 x 1(T5 8 x 1(T4

3. Small LOCA 2 (1-1.6") 7.5 x 1(T5

4. Small LOCA 3 (0.5-1") 7.5 x 10~5

5. Small LOCA 4 (0.25-0.5") 1.4 x 10"4 2.7 x 10'36. Small LOCA 5 (0.04-0.25") 2.8 x 10-3

(B) Pressurizer leakages following transients:

7. Small leakage from pressurizer(0.4") following loss of mainfeedwater 3.2 x 10~5

8. Small leakage from pressurizer(0.4") following loss of heat sink 3.3 x 10~5

9. Small leakage from pressurizer(0.4") following other transients 1.2 x 10"4 1.3 x 10"3

10. Small leakage from pressurizer(0.8") by failure-open of safety valve 8.5 x 10~4

11. Leakage in the primary system in theround-space ('ring-space') <10"7 3 x 10~8

(C) Steam generator tube leakage:

12. Steam generator tube leakage(0.12-0.25") 1 x 10

13. Steam generator tube leakage(0.02-0.12") 6.5 x 10'3

-5

39

3.4.2. Initiating event lists for BWRs

(A) The EPRI list

Since there is a considerable difference between IBs in PWR and BWR plants, EPRIdeveloped a separate IE list based on operating experience of plants with BWRs. The datafor BWRs include 903 events occurring over 101 reactor years. The EPRI BWR list is givenin Table 3.9. As for the PWR list, the number associated with each initiator is given inSection 4.

(B) The Grand Gulf list [36]

This list was obtained by making use of the IE lists in about 10 available PS As for USreactors. Generally, the initiators identified in these studies incorporate all events that hadoccurred in US nuclear power plants by 1980. A summary of these actual transient eventsis reported in EPRI-NP-801 [3] and its update EPRI-NP-2230 [4].

The above information was supplemented with actual plant trip data for Grand Gulf fora period of about 2 years of operation.

A review of the Grand Gulf design for special initiators was also undertaken.

The loss of coolant events were categorized into break size ranges, based on successcriteria for the safety injection systems. The resulting list of initiating events for Grand Gulfis given in Tables 3.10 and 3.11.

3.5. COMPLETENESS OF IE LISTS

The first step in order to get a list of initiating events as complete as possible is to useseveral of the methods presented in Section 3.3. As an obvious minimum, one should includethe following methods:

- Engineering evaluation;- Reference to previous PSAs and other available IE lists (e.g. EPRI list);- Reference to operating experience (if any relevant is available);- Logical evaluation (MLD or EBFT).

Each approach examines potential IBs from a different perspective yielding a highdegree of confidence that all risk significant IBs have been identified.

The more frequent IBs should, in general, be based on collected operating experience.This group of IBs is therefore considered to be relatively well covered.

The problem of completeness arises mainly with respect to the low frequency IBs. Lowfrequency IBs may be considered in three areas: pipe breaks or ruptures, other componentfailures and rare human errors. Pipe breaks and ruptures are considered to be relatively wellcovered if a proper engineering evaluation of the plant is performed. An important point isto examine piping outside the containment which has connections to the primary system, inorder to reveal possible interfacing LOCA initiators. In that respect, the reliability of isolationvalves has to be adequately considered.

Text cont. on p. 49.

40

TABLE 3.9. LIST OF BWR TRANSIENT INITIATING EVENTS

1. Electric load rejection2. Electric load rejection with turbine bypass valve failure3. Turbine trip4. Turbine trip with turbine bypass valve failure5. Main steam isolation valve (MSIV) closure6. Inadvertent closure of one MSIV7. Partial MSIV closure8. Loss of normal condenser vacuum9. Pressure regulator fails open10. Pressure regulator fails closed11. Inadvertent opening of a safety/relief valve (stuck)12. Turbine bypass fails open13. Turbine bypass or control valves cause increase in pressure (closed)14. Recirculation control failure—increasing flow15. Recirculation control failure—decreasing flow16. Trip of one recirculation pump17. Trip of all recirculation pumps18. Abnormal startup of idle recirculation pump19. Recirculation pump seizure20. Feedwater—increasing flow at power21. Loss of feedwater heater22. Loss of all feedwater flow23. Trip of one feedwater pump (or condensate pump)24. Feedwater low flow25. Low feedwater flow during startup or shutdown26. High feedwater flow during startup or shutdown27. Rod withdrawal at power28. High flux due to rod withdrawal at startup29. Inadvertent insertion of control rod or rods30. Detected fault in reactor protection system31. Loss of off-site power32. Loss of auxiliary power (loss of auxiliary transformer)33. Inadvertent startup of HPCI/HPCS34. Scram due to plant occurrences35. Spurious trip via instrumentation, RPS fault36. Manual scram; no out-of-tolerance condition37. Cause unknown

41

TABLE 3.10. LOCAs AND SPECIAL INITIATORS INCLUDED IN THE GRAND GULFPRA [36]

LOCA SIZE

A (large LOCA)

51 (intermediate LOCA)

52 (small LOCA)

53

R

V

Steam >0.4 sq.ft. (370 cm2)Liquid >0.4 sq.ft.

Steam 0.13-0.4 sq.ft.Liquid 0.007-0.4 sq.ft.

Steam <0.13 sq.ft. (100 cm2)Liquid O.007 sq.ft.(6.5cm2)

An S3 LOCA is a recirculation pump seal break and isisolable.If the operator does not recognize the break and fails toisolate, it is categorized as an S2 LOCA.

Vessel rupture.

Interfacing system LOCA.

SPECIAL INITIATOR

Loss of instrument-air

Loss of emergency ACor DC vital bus

DESCRIPTION

The IE is included because a loss of instrument-airresults in a plant trip and degrades one or more of thesystems required to respond to the initiator.

The IE is included because a loss of any AC or DC busresults in a plant trip and degrades one or more safetysystems. In most BWRs, the drywell coolers are anon-safety system, but are powered by the safety buses.Loss of one safety bus may eventually lead to highdrywell pressure trip and loss of a safety systempowered by that bus.

42

TABLE 3.11. TRANSIENT INITIATORS IN THE GRAND GULF PRA

INITIATINGEVENT GROUP

EPRI TRANSIENTCATEGORY RATIONALE FOR INCLUSION

Tl(Transients that

cause LOSP)

31 A loss of the offsite grid w i l l result In a reactor scram and lossof normal AC power The on-site emergency diesel generators arerequired to start to supply AC loads.

T2(Transients with

loss of PCS*)

Loss of the normal and preferred station transformers result in areactor scram and loss of normal AC power The on-site emergencyciiesel generators are required to start and supply AC loads.

A generator load rejection results In fast closure of thp turbinecontrol valves which in turn scrams the reactor A subsequentfailure of the turbine bypass valves to open results in completeIsolation from the condenser.

A turbine trip results in closure of the main turbine stop valveswhich in turn scrams the reactor. A subsequent fa i l u r e of theturbine bypass valves to open results in complete i s o l a t i o n fromthe condenser.

MSIV closure results in a reactor scram and complete Isolationfrom the condenser.

Closure of one MSIV may result In a high steam line flow signalthat Isolates the main steam lines from the condenser by dosingthe remaining MSIVs. MSIV closure scrams the reactor

Loss of PCS, that is, reactor Isolation from main condenser

TABLE 3.11. (cont.)



T2(Transients withloss of PCS*)(Concluded)

Partial closure of one MSIV may result in a high steam line flows-gnal that isolates the main steam lines from the condenser byclosing the remaining MSIVs. MSIV closure scrams the reactor.

A loss of condenser vacuum causes a closure of the main turbinestop valves. MSIVs. and turbine bypass valves. The turbine tripinitiates reactor scram.

A pressure regulator failure in the open position w i l l cause themain turbine control valves and bypass valves to completely openresulting in a low turbine inlet pressure isolation of the MSIVs.MSIV closure initiates reactor scram.

10 A pressure regulator failure in the closed direction will resultin closure of the main turbine control valves and inhibit openingof the turbine bypass valves. A high neutron flux signal w i l lscram the reactor.

12 An inadvertent or excessive opening of a turbine bypass valve willdecrease the main steam line pressure resulting in a low turbineinlet pressure closure of the MSIVs. MSIV closure initiates reactorscram.

13 Failure of the turbine control valves and bypass valves in theclosed position will isolate the reactor from the main condenser.Closure of the turbine stop valves initiate reactor scram.

37 All transients with unknown causes are assumed to result inisolation of the reactor vessel from the main condenser.

"Loss of PCS. that is, reactor isolation from main condenser.

TABLE 3.11. (cont.)



T3A(Transient withPCS available)

A generator load rejection causes a fast closure of the turbinecontrol valves that in turn causes a reactor scram. The MSIVswill remain open and the turbine bypass valves w i l l allow steamflow to the condenser.

A turbine trip results in closure of the main turbine stop valveswhich in turn causes a reactor scram. The HSIVs w i l l remain openand the turbine bypass valves wi l l allow steam flow to thecondenser.

A failure of the recirculation flow control which increasesreclrculation flow results in a high neutron flux scram of thereactor. The turbine control valves wi l l close upon decreasingturbine pressure. The MSIVs remain open and the turbine bypassvalves w i l l allow steam flow to the condenser

15

16

A failure of the reclrculation flow control which decreasesreclrculation flow will result in a vessel level swell to theLevel 8 set point for scram and turbine trip. The MSIVs remainopen and the turbine trip Initiates bypass valve operation.

The trip of on reefrculation pump will not result in reactorscram. However if an additional failure occurs, 1t Is assumedthat the other recirculatlon pump trips. A level swell willoccur causing a Level 8 reactor scram and turbine trip. TheMSIVs remain open and the turbine trip Initiate bypass valveoperation.

TABLE 3.11. (cont.)




(Continued)

17 A trip of all the reelrculation pumps w i l l result in a vessellevel swell to the Le/el 8 setpoint for scram and turbine tripThe MSIVs remain open and the turbine trip i n i t i a t e s bypass valveoperatton

18 Attempts to Startup a recirculat ion pump Bt hiqh power levelsw i l l result in a reactor scram on high flux Ipvel The MSIVsremain open and the turbine bypass valves open following turbinetrip

19 A recirculation pump seizure w i l l result in a level swellcausing a Level 8 reactor scram and turbine trip The MSIVsremain open and the turbine trip i n i t i a t e s bypass valve operation

20 A feedwater controller failure resulting in maximum flow to thevessel w i l l result in a Level 8 scram and turbine trip The HSiVsremain open and the turbine trip w i l l initiate bypass valve opening

21 A loss of feedwater heating can result in a high neutron flux scram,of the reactor. The MSIVs remafn open and the turbine bypass valveswill allow steam flow to the condenser

27 A rod withdrawal at power was shown in the FSAR not to result in areactor scram since the rod withdrawal l i m i t e r mode of the RodControl and Information System (RCIS) l i m i t s the withdrawal Eventhough the RCIS is single failure proof it is assumed that » controlrod withdrawal w i l l cause a high neutron flux scram of the reactorThe MSIVs remain open following turbine t r i p

TABLE 3.11. (cont.)




(Concluded)

29 Insertion of more than one control rod with failure of the RCISassumed results in a turbine trip on low r.toam flow. The MSIVsremain open and the turbine bypass valves open.

30 A fault in the reactor protection system can result in a reactorscram. The MSIVs remain open and the turbine bypass valves openfollowing the turbine trip.

33 An inadvertent startup of the HPCS can result in a reactor scramif the feedwater controller fails to respond correctly. A Level 3or 8 signal is assumed to result in reactor scram. The MSIVswill remain open and the turbine trip initiates bypass valveoperation.

34 Scrams from plant occurrences are asr.umrtl not tn cause MSIVclosures or turbine bypass f a i l u r e .

35 A spurious t r i p via reactor protection system instrumentationdoes not cause MSIV closure or failure of the turbine bypassvalves.

T3B(Loss of FW transient)

36 A manual scram w i l l not cause MSIV closure or failure of theturbine bypass valves.

22 A loss of all feedwater flow wi l l result m a Level 3 trip ofthe reactor. The level w i l l decrease to Level 2 actuating HPCSand RC1C. Operation of HPCS and RCIC w i l l depressurize the plantto below the low pressure isolation of the MSIVs hut closure of theMSIVs w i l l not occur since the operator i r, expected to turn thereactor mode switch to SHUTDOWN thus i n h i b i t i n g the low pressureMSIV closure signal .

TABLE 3.11. (cont.)


EPRI TRANSIENTCATEGORY RATIONALE TOR INCLUSION

T3B(Loss of FW transient)

(Concluded)

T3C(IORV transient)

23 A trip of one feedwater or condensate pump may not result in areactor scram if the feedwater control system compensates for thereduced flow. However an additional failure is assumed to occurresulting in a Level 3 reactor scram The MSIVs w i l l remain openand the turbine bypass valves operate following turbine trip Acomplete loss of feedwater flow is consequently assumed.

24 A transient resulting in low feedwater flow below the capabilitiesof the feedwater control system wil l result in a level 3 reactorscram. The MSIVs will remain open and the turbine bypass valvesoperate following turbine trip. A complete loss of feedwater flowfs conservatively assumed

11 An inadvertent open Safety Relief Valve (SRV) w i l l cause the suppres-sion pool temperature to increase above 105 F if the SRV can not beclosed (assumed in this category) A reactor scram is required byprocedures.

Note that suppression pool cooling will not prevent the suppressionpool temperature from increasing during this type initiator.

Among the other two groups of initiators, it is possible that those IBs were notspecifically considered in previous PSAs. Since those IBs, besides causing a disturbance inthe reactor operation, also affect the function of some mitigating systems (low frequencyCCIs), they should carefully be searched for. Naturally, such events that cause multipleeffects are more dangerous than single failures events. In a number of cases such initiatingevents have been identified during the performance of other PSA tasks (e.g. fault tree analysis(FTA)). These initiators are most often plant specific and the engineering evaluation is themost suitable approach for their identification. Another support for the completeness of IElists is obtained from the large size of the IE lists. Most full scope PSAs use a list of over40 IBs that cover LOCAs of all sizes, transients and CCIs of all relevant support systems.These IBs are then grouped into broad categories which are therefore less sensitive to thefailure to include a particular IE.

49

4. GROUPING OF INITIATING EVENTS

For each of the initiating events defined, an event tree, depicting the spectrum of plantresponses (accidents and successes) should be made. Some of the initiating events wouldinduce the same or a reasonably similar plant response. In that respect, the different IBs aregrouped in order to decrease the amount of analyses required for PSA.

4.1. GENERAL PRINCIPLES FOR GROUPING

The grouping of initiating events shall be made in such a way that all events in thegroup impose essentially the same success criteria on the frontline systems as well as the samespecial conditions (challenges to plant operators, or to automatic plant responses) and shallresult in the same core damage state. Thus the grouped events can be modelled using thesame event/fault tree analysis.

Sometimes the amount of subsequent analysis needed may be further reduced bygrouping together IBs that evoke the same type of plant response but for which the frontlinesystem success criteria are not identical. The success criteria applied to that group of eventswould then be the most demanding (in terms of required systems) for any member of thegroup. The savings in effort must be weighed against the conservatism which this approachintroduces.

Since there are some distinctive features for grouping each LOCA, transients or CCIs,each set is convened separately, for BWR and PWR type NPPs.

4.2. GROUPING OF INITIATING EVENTS FOR PWRs

4.2.1. LOCAs

As the plant response and success criteria is substantially different for LOCAs insideand outside containment, these two categories are separated. The inside containment divisionincludes different break size LOCAs where the grouping is made according to a particular setof success criteria, i.e. different sets of mitigating systems are required for various pipe breaksizes. The most common division of LOCAs caused by piping break, leakage or rupture is:

- Large LOCA (e.g. break size > 6 inch diameter, equivalent to > 300 cm2 leak area);- Medium LOCA (e.g. break size > 3 inch diameter, equivalent to 150-300 cm2 leak

area);- Small LOCA (e.g. break size < 3, equivalent to < 150 cm2 leak area).

These are very general divisions and they vary from one plant to the other. They aretypical to liquid type LOCAs and may differ for steam side breaks.

In some cases, the small LOCA category is further subdivided. Sometimes plantresponse requires also a different set of equipment to mitigate the very small size LOCA, suchas encountered when a RCP seal fails. Another reason to include a very small size LOCAis that it may affect the containment system response in another way than a small LOCA, e.g.a larger time period is needed for detection and other sensors are required to detect this kindof event.

50

In addition, the rupture or large leak of a reactor pressure vessel (RPV) is consideredin most PSAs as a separate group. It is a case in which all available plant mitigating systemsare considered ineffective.

The loss of coolant events 'outside containment' are interfacing LOG A and the steamgenerator tube rupture initiators. Although the coolant leaks for those events are comparableto some of the LOG As inside containment, there are two principal differences between the twogroups. A LOCA outside containment would, in addition to bypassing the containment andconsequently creating the path for possible release of radioactivity to the environment, involvean unrecoverable loss of cooling water which becomes critical for the long term cooling.

Two examples of LOCA IE groups are given in Tables 3.4 and 3.8.

4.2.2. Transients

Grouping of transient initiators is not only plant specific, it also depends heavily on thepurpose and scope of a PSA study. To illustrate possible approaches, Tables 4.1 and 4.2compare the groupings chosen in several plant specific PSA studies with the EPRI list oftransients for PWRs [4] grouped in 'Functions' as in NUREG/CR-3862 [5]. It can be seen thatthere are some substantial differences in groupings.

There are several reasons for these differences. The most important ones are:

- Different design operation features that affect plant response and success criteria;- Different level of detail in PSA;- Progress made in the PSA modelling in order to obtain more realistic results.

As a general rule, the amount of analysis (and cost) need to be weighted against theincrease in modelling detail and in more realistic results. The approaches taken would varyfrom rather coarse grouping, where due to scoping analysis a lot of conservatism is involved,to precise specific grouping, in order to achieve best estimate results.

The minimal grouping would include:

- Transients with main feedwater (MFW) initially available (turbine/reactor trips);- Transients with loss of MFW;- Loss off-site power.

When more precise grouping is preferred, the following subgrouping would fit in theabove indicated broad groups:

- Reactor trip ) Transients with MFW- Turbine trip ) available

Partial loss of MFW )- Excessive FW flow )

51

Total loss of MFW )- Loss of condenser vacuum ) Transients with

Closure of one MSIV ) loss of MFWClosure of all MSIVs )

- Core power excursion )

- Loss of primary flow ) Loss of off-site power- Loss of off-site power )

The increased details allow to take into account, in addition to the hardwareconfiguration changes with respect to the mitigation of the sequence, the recovery times whichmay differ for each of the transient initiators (e.g. condenser recovery is longer than feedwaterpumps recovery). Therefore, final sequences and associated probabilities are much morerealistic.

4.2.3. Special common cause initiators

This broad group of initiators includes relatively low frequency IBs that in addition toinitiating fault cause failures of mitigating systems. They are very much plant specific. Themost common initiators of this group are:

- Loss of vital AC power bus;- Loss of service water system;- Loss of component cooling;- Loss of a DC bus;- Loss of instrument air;- Loss of core level measuring instrument;- Loss of ventilation system;- Loss of room coolers;- Steam line break in locations where it causes additional effects or containment isolation.

The loss of off-site power initiator is somewhat specific. Normally it is included in thetransient group which implies that the diesel generators are available or the off-site power isbeing restored within 30 minutes. Otherwise it is called 'plant blackout' and becomes a moreserious transient which has similar characteristics to a common cause initiator. Therefore itis sometimes treated under the CCI category. An example is the RSS [2] in which LOOPlonger than 30 minutes were treated as a CCI. Several examples also exist of plant specificinitiators such as the loss of a particular AC bus in a plant that causes a transient with lossof a part or severe degradation of the mitigating systems.

Since common cause initiators are rather specific as far as exact plant response isconcerned, they are, in principle, treated individually, e.g. there is no grouping and an eventtree is created for each initiator separately. In some cases, several such initiators, havingsimilar plant response, would be grouped together (a typical example is loss of servicewater/loss of component cooling at some plants). An event tree will be constructed for theinitiators with the most difficult plant response. The treatment of CCIs is in principle similarfor BWR and PWR type plants.

52

TABLE 4.1. A COMPARISON OF EPRI PWR TRANSIENT CATEGORIES WITH SEVERAL PSA GROUPING SCHEMES (EXAMPLE 1)

Function(From [5])Main turbine/generator

Condenser vacuum

33

34

252730

IE Category(From [4])Turbine trip, throttle valveclosure, EHC problems

Generator trip or generatorcaused faultsLoss of condenser vacuumCondenser leakageLoss of circulating water

OCONEE[9]

1..10143339

2,3,6,,11,12,27,28,34,36

25*30

SEABROOK GERMAN RISK[14] STUDY [35]

,13,29,38

+

+ +

+ +18

SURRY[40]

1,2,3,8,9,1013.U,23,33,38,39,

17,18

4,5,1115,34,40

INDIANPOINT 3F421

,6,7 +,12,19,2036,

+

ZION[44]+

Condensate 23 Loss of condensate pump(One loop)

24 Loss of condensate pumps(all loops)

Feedwater 16 Total loss of feedwater flow(all loops)

19 Increase in feedwater flow(1 loop)

20 Increase in feedwater flow(all loops)

15 Loss or reduction infeedwater flow (1 loop)

21 Feedwater flow instability-operator error

22 Feedwater flow instability-miscellaneous mechanical

16,24 + + + + +19 + *

20 + *+ * + ++ +

15,21, + +22,23

causes28 Miscellaneous leakage in

secondary system

TABLE 4.1. (cont.)

Function

Reactor system flow

Reactor systempressure control

Reactivity control

Steam

Safety injectipn

Electrical power

IE Category OCONEE SEABROOK GERMAN RISK SURRYSTUDY

1 Loss of RCS flow (1 loop) * *13 Startup of inactive coolant * *

pumpU Total loss of RCS flow * + *6 Low pressurizer pressure * *8 High pressurizer pressure *36 Pressurizer spray failure * *

2 Uncontrolled rod with- * core power *drawal excursion *

3 CRDM problems and/or rod *drop

11 CVCS malfunction-boron * *dilutions

12 Pressure/temperature/power * *imbalance-rod position error

17 Full or partial closure of + *HSIV (1 loop)

18 Closure of all MSIV + * *29 Sudden opening of steam * +

relief valves

9 Inadvertent safety injection + -f *signal

10 Containment pressure problems * *

35 Loss of all offsite power +. + + •+•37 Loss of power to necessary + Loss ofplant systems AC Bus

INDIAN ZIONPOINT 3

+ ++ ++ +

corepower

increase

++ +

+

•*• +

TABLE 4.1. (cont.)

Function

Reactor integrity

Miscellaneous

IE Category OCONEE SEABROOK GERMAN RISKSTUDY

k Leakage from control rods not Incl. in Incl. in5 Leakage in primary system included the LOCA the LOCA7 Pressurizer leakage in PSA categories categories26 Steam generator leakage * *- Rod ejection +31 Loss of component cooling + +32 Loss of service water + +

systems41 Fire within plant * *

SURRY INDIAN ZIONPOINT 3

***

+ + ++ + +

+

Spurious trips 38 Spurious trips -cause unknown

39 Auto trips - notransient condition

UO Manual trip - notransient condition(planned shutdown)

** *

*

Non EPRIList Categories

Non EPRIList Categories

Integrated Control System +Failures

- Loss of Instrument Air +Spurious low PressurizerPressure signal +

- Feedwater condensate line +break

- Steam line break - large + + + + +- medium +

- Loss of DC Bus + +

Notes: "+" taken into account in the same group"*" taken into account but in a different group"Blanks" indicate that the category was not considered In that PSA (in several cases the basis for the omissionis provided in the PSAs)

TABLE 4.2. A COMPARISON OF EPRI PWR TRANSIENT CATEGORIES WITH SEVERAL PSA GROUPING SCHEMES (EXAMPLE 2)

Function(From [5])Main turbine/generator

Condenser vacuum

Condensate

Feedwater

IE Category(From [ 4 ] )

33 Turbine trip, throttle valveclosure, EHC problems

34 Generator trip or generatorcaused faults

25 Loss of condenser vacuum27 Condenser leakage30 Loss of circulating water23 Loss of condensate pump

(one loop)24 Loss of condensate pumps

(all loops)16 Total loss of feedwater flow

(all loops)19 Increase in feedwater flow

(1 loop)20 Increase in feedwater flow

(all loops)15 Loss or reduction in

feedwater flow (1 loop)21 Feedwater flow instability-

operator error22 Feedwater flow instability-

miscellaneous mechanicalcauses

28 Miscellaneous leakage insecondary system

SIZEWELLBF501

FC4FC4

FC4*FC4FC4FC4

FC4FC3FC3FC4FC3FC3

*

RING- ANO-1HALS-2f29l F181

CAT. 2 +CAT.2 1,2,3,6,8-10,

13,14,15,17,23,34

CAT.2 *CAT.2CAT.2 *

CAT.2 *CAT.3A *

CAT.3A +.24,25,29,30,17,18

CAT.3BCAT.3B +CAT.2CAT.3B +CAT.3B +

CAT.2,9,18,24,32

CALVERTCLIFFS 1

F411+, 37

***

+, 18, 24,2530

**

*

TABLE 4.2. (cont.)

Function

Reactor system flow

Reactor systempressure control

Reactivity control

Steam

Safety injection

Electrical power

Reactor integrity

IE Category

1 Loss of RCS flow (1 loop)13 Startup of inactive coolant

pump14 Total loss of RCS flow

6 Low pressurizer pressure8 High pressurizer pressure36 Pressurizer spray failure2 Uncontrolled rod with-drawal

3 CRDM problems and/or roddrop

11 CVCS malfunction-borondilutions

12 Pressure/teraperature/powerimbalance-rod position error

17 Full or partial closure ofMSIV (1 loop)

18 Closure of all MSIV29 Sudden opening of steam

relief valves9 Inadvertent safety injectionsignal

10 Containment pressure problems35 Loss of all offsite power37 Loss of power to necessary

plant systems4 Leakage from control rods5 Leakage in primary system7 Pressurizer leakage

26 Steam generator leakage- Rod ejection

SIZEWELLB

FC6

FC6

FC-11

FC-7FC-7FC-7*

FC-4FC-4FC-3

FC-8

FC-5FC-5

FC-9FC-9FC-9FC-9

RINGHALS-2 ANO-1 CalvertCliffs-1

CAT. 2 * +.3,9,11,12CAT. 2 15,17,19,20,23,27CAT.l * +28,29,38,

39,40CAT. 2 * +CAT.l * +CAT.l +

CAT.l *

CAT. 2 * *

CAT.l *CAT. 2 *

CAT. 2 * *CAT.3B * *CAT. 2 * *

CAT.3b * *

CAT. 2 *

CAT. 4 + +CAT/2 loss of AC

Bus *CAT. 3 incl.CAT. 2 in theCAT. 2 LOCACAT. 2 categories*

TABLE 4.2 (cont.)

Function

Miscellaneous

Spurious trips

IE Category

31 Loss of component cooling32 Loss of service water

systems41 Fire within plant38 Spurious trips -

cause unknown39 Auto trips - no

transient condition40 Manual trip - no

transient condition("planned shutdown)

SIZEWELLB

FC-11FC-11

FC-13

FC-2

RING- ANO-1HALS-2CAT. 2 +CAT.3C +

CAT. 2 **CAT. 2CAT. 2

CalvertCliffs-1

++

***

Non-EPRIList Categories

Integrated control systemfailureSpurious low pressurizerpressure signalLoss of Instrument AirLoss of core coolingduring steam-gen, inspec.Feed water condensateline break

Steam Line Break -Large-Medium

Loss of DC BusCold overpressurization

FC-4FC-3

FC-8

CAT.O

CAT.3ACAT.3A

Notes: "•*-" taken into account in the same group"*" taken into account but in a different PSA IE group"Blanks" indicate that the category was not considered in that PSA (in several cases the basis for the omissionis provided in the PSAs)"FC"- Fault CategoryCAT.O: Not regular (faulty) reactor shutdown conditionsCAT.l: Transients challenging RCS integrityCAT.2: Transients that do not challenge -RCS integrity or auxiliary core cooling systemsCAT.3: Transients challenging auxiliary core cooling systems with offsite power availableCAT.4: LOSP

4.3. GROUPING OF INITIATING EVENTS FOR BWRs

4.3.1. LOCAs

As mentioned in Section 4.2.1 LOCAs are grouped in different break size categoriesaccording to the different success criteria for the safety injection system. A typical divisionwould correspond to the following [36]:

A Large LOCA Automatic depressurization of primary system is not required.

S, Medium LOCA Automatic depressurization necessary in order to make itpossible for low pressure ECCS injection

S2 Small LOCA Low pressure ECCS injection not required.

In addition to the above, some of the PSAs also define a small-small LOCA (S3). Theinitiator behind that definition is a recirculation pump seal leak or pipe leak of similar size.Such leaks have occurred in power plants, primarily from the wearing out of these pump sealsduring normal operation. Sometimes the control rod drive (CRD) system leakage isconsidered in the small-scale LOCA category. As in the case of PWRs, interfacing LOCA isconsidered as a separate category. The same is true for reactor vessel rupture. In BWRs steamis generated directly in the reactor vessel, and LOCA events may be on steam and waterpiping. Since the loss of coolant rate will depend on rupture being either on the steam orwater side, actual break sizes (or equivalent flow areas) for the steam and water side will bedifferent for each of the LOCA groups. Again the need for depressurization and successcriteria for low pressure safety injection will determine actual grouping. Table 3.9 providesan example of LOCA grouping used for the Grand Gulf PSA.

4.3.2. Transients

As for PWRs, grouping of initiating events depends on the plant response and scope ofthe PSA study. Table 4.3 compares grouping in PSA studies for Shoreham, Grand Gulf,Forsmark and Browns Ferry NPPs with EPRI Generic IE categories [4] and functions fromNUREG/CR-3862 [5].

Table 4.3 shows that there are some major differences in grouping done in plant specificPSA study. One obvious reason is differences in design and logics of these NPPs. Anotherreason is that considerations regarding the severity of the individual initiators did influencethe chosen grouping scheme. Decisions to utilize many IE groups mean that increased effortto obtain more realistic modelling was judged to be cost effective.

In the Reactor Safety Study (WASH-1400) [2] the BWR transients were originallygrouped into three major categories:

Tl: Transients involving loss of off-site power;T2: Transients involving loss of the power conversion system (PCS) (MSIV closure, loss

of condenser vacuum, etc.);T3: Transients with the PCS initially available (turbine trip, etc.).

59

TABLE 4.3. A COMPARISON OF EPRI BWR TRANSIENT CATEGORIES WITH SEVERAL PSA GROUPING SCHEMES

FunctionMain turbine/generator

Condenser vacuum

IE Category1 Electric load rejection2 Electric load rejection with

turbine bypass valve failure3 Turbine trip4 Turbine trip with turbinebypass valve failure

8 Loss of normal vacuum

SHOREHAMrial1 35,36,37(Tt)* 32,33,34

27,28,29,303 24,25,26(Tt)

* 19,20,21,2314,15,16,1718,6,7,9,1012,13 (Tt)

8 (Tc)

GRAND GULFT361

1 (T3A), 14-212 (T2), 4-10,12, 13, 373 (T3A),27,29,3033-364 <T2)

*

FORSMARK 3 BROWNS FERRYF291 T301

^s 22 (Tt) T23 (T8),6, 1312,57,8,9,1039,40

*(Tt) TICondensate

Feedwater

Reactor system flow

20 Feedwater-increasing flowat power

21 Loss of feedwater heater22 Loss of all feedwater flow23 Trip of one feedwater pump

(or condensate pump)24 Feedwater-low flow25 Low feedwater flow during

startup or shutdown26 High feedwater flow during

startup or shutdown14 Recirculation control

failure-increasing flow15 Recirculation control

failure-decreasing flow16 Trip of one recirculation

pump17 Trip of all recirculation

pumps18 Abnormal startup of idle

recirculation pump19 Recirculation pump seizure

*

22 (TF)****

*

*

*

*

*

*

22 (T3B)23 (T3B)24 (T3B)

*

*

*

*

*

*

* TI* -22 (Tf) T!23 (Tf)* _

it —

* _

* _

* _

* _

TABLE 4.3. (cont.)

FunctionReactor systempressure control

Reactivity control

Steam

Safety injection

Electrical power

Miscellaneous

IE Category9 Pressure regulator failsopen

10 Pressure regulator failsclosed

12 Turbine bypass fails open13 Turbine bypass or control

valves cause increasedpressure (closed)

27 Rod withdrawal at power28 High £lux due to rod with-

drawal at startup29 Inadvertent insertion of

rod or rods5 Main steam isolation valve6 Inadvertent closure of oneMSIV (rest open)

7 Partial MSIV closure11 Inadvertent opening of a

safety/relief valve (stuck)33 Inadvertent startup of

HPCI/HPCS31 Loss of offsite power32 Loss of auxiliary power34 Scram due to plant

occurrences30 Detected fault in reactor

Protection system

SHOREHAM GRAND GULF* ** ** ** *

* ** _* *

5 /m \ i^\Tjj) *5* *

* *11 (TI) n (T3c)* *

31 <TE) 31 (T!)* 32 (TX)* *

* *

FORSMARK 3 BROWNS FERRY* (Tt) T2* (Tt>* (Tt)* (Ts) T2

* __

* _

* dt) T!* (T8) T2* _Small LOCA

_

31 <Te) T332 (Tt) T!

* <TS>30 16,17,18 -19,27,29,20,21,22,23,24,14,15 (Ta)

o\to TABLE 4.3. (cont.)

FunctionSpurious trips

IE Category35 Spurious trip by way of

instrumentation, RPS fault36 Manual scram-no out-of-

tolerance condition37 Cause unknown

SHOREHAM*

*

GRAND GULF*

*

FORSMARK 335 (Ts)36 (T8>37,34 (T8)

BROWNS FERRY

-

Not included inEPRI List

- Planned shutdown- Loss of DC bus- Loss of instrument air

Notes: "+" taken into account in the same group"*" taken into account but in a different category"Blanks" indicate that the category was not considered in that PSA (in several cases the basis for the omissionis provided in the PSAs)TT - Turbine trip; TQ - Loss of condenser; Tg - LOSPjTM - MSIV closure; Tj - Inadvertent open relief valve;TJ- - Loss of feedwater flow,(T!, T2, 13 - same as in WASH-1400),

In some later PS As, the number of categories was expanded. However, in the morerecent PSAs for Peach Bottom and Grand Gulf, a review was conducted on the interim ASEP[33] IE list, in order to determine whether expansion of these categories was necessary. Inaddition, the actual operating history for Grand Gulf and Peach Bottom was reviewed (asgiven in post-trip analysis reports which summarize, among other things, the causes for plantshutdowns). This information was combined into the list of transient initiators. In general,it was found that transient events listed in EPRI NP-2230 [4] could remain grouped into thethree main WASH- 1400 [2] transient categories, but with a small modification: in the PeachBottom and Grand Gulf PSAs, the T3 events category (transients with PCS available) isfurther grouped into three subcategories:

T3B: Loss of feedwater;T3C: Inadvertent open relief valve;T3A: All other T3 events.

The categorization used for Peach Bottom and Grand Gulf is also in general agreementwith the categorization principles developed especially for the Swedish SUPER- AS AR project[29], where the categorization of transients is primarily based on the availability of the mainfunctions: (1) off-site power, (2) feedwater, (3) the main condenser. The intention is that theIBs within each group should be similar with respect to affected process parameters andcountermeasures. The following main groups are considered:

(1) Planned shutdown;(2) Scram due to small disturbances;(3) Loss of main condenser;(4) Loss of feedwater;(5) Loss of feedwater and main condenser;(6) Loss of 400 KV outer grid;(7) Pipe break of different sizes and locations.

The logic of this categorization is shown in Fig. 4.1.

The main groups are then further subdivided with respect to availability of (battery)supplied grid and the status of all systems covered by technical specifications (main group 2),availability of primary system decay heat removal (groups 3 and 5), partial or total loss offeedwater (group 4) and the availability of off-site grid and generator house load operation(group 6).

The Shoreham PS A [12] considers the following groups:

(1) Transients that result in turbine trip (TT);(2) Transients caused by main steam isolation valve (MSIV) closure which lead to isolation

of the reactor vessel from the main condenser (TM);(3) Transients following loss of feed water flow (Tf);(4) Transients resulting from loss of condenser (Tc);(5) Transient resulting from loss of off-site power (TE);(6) Transients resulting from inadvertent open relief valve (IORV);(7) Orderly and controlled manual shutdown

63

INTACTPIPE SYSTEM

OUTER GRID400 KV

FEEDWATER

MAINCONDENSER

PLANNEDSHUT DOWN

YES

NO

PQ PLANNED*P SHUT DOWN-0 SCZIAM DUB TOLS SMALL DISTURBANCE

,(3 LOSS OFAT MAIN CONDENSERO LOSS OF'•P FEED WATERQ LOSS OF FEED WATER*ATF +MAIN CONDENSER0 LOSS OF OUTER

AE GRID 400 KV

S,A,R, PIPE BREAKSV,AV,SY,AY

FIG. 4.1. Initiating event classification tree.

Compared to earlier PS As, the Shoreham list has been expanded. The explanation isthat the Shoreham PS A considers the severity of the different initiating events in more detailin order to achieve more realistic risk spectra. For example, event Tc (loss of condenser) andevent TM (closure of MSIV) are identical in the early stage of the event, and were treated asone group in some PSAs. Since the recovery time for loss of condenser is considered to besubstantially larger than the necessary recovery time after MSIV closure, separating theseinitiators in two groups means getting more realistic results.

These examples are intended to give some guidance for grouping. But it has to beremembered that plant specific design or logics may necessitate different approaches. And,as mentioned earlier, the effects of the increased amount of work and the possible realism orconservatism in the result, have to be weighted against each other.

4.3.3. Special common cause initiators

Special common cause initiators such as loss of DC power, loss of instrument air or lossof component cooling influence the plant in different ways. For that reason they are normallynot grouped. The discussion in Section 4.2.3 applies here too.

4.4. IMPACT OF VARIOUS TRANSIENTS ON CORE DAMAGE FREQUENCY

The impact of initiating events on core damage frequencies of the power plant can bemeasured by recording the conditional probability of core damage given a certain IE.Table 4.4 provides one example for a BWR plant taken from the Shoreham PS A [12] and itsreview [6]. Table 4.5 provides another example for a PWR plant taken from the German RiskStudy, phase B [22].

64

TABLE 4.4. IMPACT OF TRANSIENT IBs ON THE SHOREHAM CDFs [12]

Transientinitiator

Turbine tripLoss of condenser

and MSIVLoss of FWLoss of off-site

powerInadvertent open

of relief valve

Conditionalprobability ofcore damage

5.5 E-69.7 E-6

4.4 E-69.6 E-6

4.8 E-6

Initiatingeventfrequency(per year)

6.850.7

0.10.15

0.2

Core damagefrequencycontribution(per year)

3.8 E-56.8 E-6

4.4 E-71.4 E-6

9.6 E-7

TOTAL 8.0a 4.8 E-5a

a Does not include all contributions.

TABLE 4.5. IMPACT OF TRANSIENT IBs ON THE GERMAN RISK STUDY CDFs [22]

Transientinitiator

Loss of emergencypower

Loss of FW withoutloss of heat sink

Loss of FW withloss of heat sink

Loss of heat sinkwithout loss of FW

Large steam linebreak outsidecontainment

SG tube leakage(1-6 cm2 area)

TOTAL

Conditionalprobability ofcore damage

1.7 E-5

2.1 E-5

2.3 E-5

8.0 E-6

2.1 E-3

1.5 E-4

-

Initiatingeventfrequency(per year)

0.13

0.15

0.29

0.36

5 x E-4

6.5 E-3

0.95a

Core damagefrequencycontribution(per year)

2.2 E-6

3.2 E-6

6.7 E-6

2.9 E-6

1.0 E-6

1.0 E-6

1.7 E-5a

Does not include all contributions.

65

5. DETERMINATION OF THE FREQUENCY OF INITIATORS

Several methods for the quantification of IE frequencies were employed in past PS As.Each of the methods has its own advantages and disadvantages and is to be used in specificcases.

5.1. APPROACHES TO QUANTIFICATION

It is possible to distinguish between two extreme cases of PS A application:

(1) New plants for which no plant specific operating experience exists.(2) Old plants for which a number of reactor-years of operating experience exist and

occurrences have been experienced in a few IE categories. If the time period of plantoperation is long enough, cases in which no occurrence has been recorded in a certaincategory may also be of significance for the quantification of the IE.

Different quantification approaches may be used for different groups of initiators. Theexample of division into IE groups for which different approaches may be used is as follows:

(1) General transients initiators;(2) Large LOG As, large steam line breaks (in containment), RPV failures;(3) Small break LOG As, small steam line breaks (in containment), interfacing LOG As;(4) SG tube rupture;(5) Common cause initiators;(6) ATWS initiators;(7) Loss of off-site power.

The quantification approaches covered in this document are based on various PS As andtheir reviews [9, 12, 14, 20, 29, 36]. These approaches are delineated in the followingsections. For convenience they are listed here:

(1) One stage Bayesian methodology using generic experience data;(2) Two stage Bayesian methodology using both generic and plant specific data;(3) Mean frequencies from large operating experience data gathered over a long period;(4) Expert opinion on rare events;(5) Failure rates and mission time (valves, pumps, expansion joints);(6) Failure rate per pipe length and for different pipe categories of quality (standard),

diameter and environmental conditions;(7) Fault tree analysis for special rare events such as common cause initiators. The top

event being the occurrence frequency of this initiator;(8) Use of similar plants' experience, other PS As or generic studies such as the EPRI [4]

and EG&G [5] studies;(9) Special plant attributes and characteristics of the geographic location of the power

plant.

Table 5.1 provides information on the initiator types for which the various approacheshave been applied in the past.

66

TABLE 5.1. APPROACHES TO IE FREQUENCY DETERMINATION USED FOR VARIOUS TYPES OF IE CATEGORIES

Quantification Approach

IE One StageBayesian

Two StageBayesian

MeanFrequ-encies

Failurerates &missiontimes

Failurerates &pipinglength

Fault Expert Similar SpecialTree Opinion experience plant and

from other locationplants attributes

Frequent Transient IE:new plants +old plants

Non Frequent Transients:old plants

Large LOCAsLarge SLBRPV failure

Small LOCAsSmall SLBInterfacing LOCAs

SG Tube Rupture

Common CauseInitiators . +

ATWS Initiators:new plants +old plants

Loss of Offsite Powers:new plants +old plants

+

+

OS

5.1.1. One and two stage Bayesian updating analyses

The Bayesian method has been used in several PSAs for evaluating the frequency ofIBs. Two of the available references on the methodology are Appendix B of the Oconee PRA[9] and Chapter 5 of the PRA Procedures Guide [28]. When only generic data is availableand plant specific information is not (e.g. a new plant) a one stage Bayesian analysis isperformed. In this analysis the number of failures and the duration of time in which thefailures occurred are the input evidence used to update a log-uniform or other type of priordistribution. The resulting posterior distribution is a generic distribution of the number offailures that can be expected for a population of plants of which the 'new plant' is a member.

When plant specific data on a number of failures and duration of time in which theywere recorded is available ('old plant') a two stage Bayesian analysis can be made. Thedistribution of the population of plants obtained before is now taken to be the prior genericdistributing and the plant specific data is the new evidence used to update the genericdistribution in order to obtain a plant specific distribution of the failure frequencies. TheOconee PRA [9], the Indian Point-3 PRA [42] and other PRAs used this methodology toobtain plant specific IE frequencies of all events for which substantial plant specificoccurrences were available.

5.1.2. Mean frequencies of frequent operational occurrences

This is a common approach used for quantification of anticipated transients and manualshutdowns. Limerick and Shoreham PSAs are examples were this approach was used ratherthan the previously described Bayesian approach. For the cases of large amount of data andwhen uncertainty bounds are not required, this approach may be effective.

In this approach the event records for a particular plant are searched for all relevantevents that occurred during a given time period.

5.1.3. Expert opinion on rare events

Expert opinion may be used to obtain an estimate on the frequency of a rare event thathas not been observed in nuclear power plant experience. A case in which expert opinion wasused in the past is in the evaluation of seismic hazard to nuclear power plants in easternUSA. This was done because some of the data needed could not be obtained otherwise.

The frequency of RPV failure is also based in part on expert opinion. In this case theexperts had to evaluate to what extent nuclear power plants are more reliable than the pipingused by the non-nuclear industry, in which failure statistics could be obtained and evaluated.Similarly, the large LOG A frequency was obtained in the RSS.

The PRA Procedures Guide [28] provides some additional information on the treatmentof expert opinion. A word of caution is always in place, that large uncertainties can beassociated with expert opinion, so that the use of this approach requires careful planning anddocumentations to allow peer review.

5.1.4. Frequency estimation by evaluation of failure rates and mission times

This is an approach frequently used in PSA to evaluate various plant specific initiatingevents. The cases of valve and pipe breaks or leakages such as interfacing LOG A, a steam

68

line break in a certain plant location and internal flooding [9, 15] are examples. In all thesecases sources of valves and pipe failure rates are searched for the applicable values for theparticular plant, based on piping/valve quality, type and size. Table 5.2 provides acomparison of the motor operated valve (MOV) failure rates taken from various sources,including PS A applications and other special studies.

For an interfacing LOCA frequency determination, additional information is neededon valve failure by rupture or leakage. Table 5.3 provides a comparison of several sourceson valve failure, rupture and leakage rates as used in several PRAs and other specialprobabilistic studies. While the RSS used data from non-nuclear sources, NUREG/CR-1363[37] uses LERs from many plants, most of them including small as well as very smallleakages (technical specifications exceedance). The last two sources in Table 5.3 [38, 39]use more recent LERs. The data provided in the tables is used to estimate interfacing LOCAfrequencies according to the number of valves available in every leak path identified in theparticular plant under review, and on the basis of the testing intervals (e.g. relevant missiontime to be applied to the failure rates) used on each leak path's through valves.

5.1.5. Fault tree analysis for special rare events

If a plant never experienced an initiating event, its frequency can be estimated usinga fault tree approach. A fault tree is then constructed to include all equipment (and possiblyhuman errors) contributing to an initiating event. Using plant design information and plantspecific failure data, the frequency of IBs will be generated. A few examples taken fromPS As are given in Section 6.2.

5.1.6. Use of experience from other plants

This approach is used for plants having a similar design or vendor. In the case of newplants from the same vendor, the studies that are available can be applied to a group of plantsof the same design. This group can be used as a source of data for the new plant. Theapproach is used in PSA performed for new plants of a design similar to those plants thathave been in operation for some time.

Another utilization of this approach is for comparison or for obtaining a range ofvalues to estimate the frequency of an initiating event. This is especially important when anIE is considered for exclusion from a specific PSA (for example, because the particular plantunder consideration has several mitigating systems to render any accident sequence of thisIE).

5.1.7. Special attributes of plant and location

For obtaining loss of off-site power (LOOP) frequencies a specific approach has beendeveloped. An extensive study of factors contributing to the frequency of LOOP in nuclearpower plants was performed in the USA with respect to the power plant blackout study. Inthe study the relation between LOOP frequency and plant specific station switchyard designfeatures were investigated. Similarly, the correlation between LOOP frequency and numberof off-site lines connected to the station switchyard were established. The impact of weatherconditions in various locations on the LOOP frequency, as well as some additional factorswere studied.

69

The NUREG-1032 report [26] is a manual to determine a frequency of LOOP for acertain plant out of plant specific information and its general attributes.

This approach is particularly suitable for new plants. For the older plants that haveaccumulated specific evidence of LOOP in their area, the two stage Bayesian approach usingthe NSAC-80 [10] could also be a suitable approach.

TABLE 5.2. MOTOR OPERATED VALVES FAILURE RATES

Source Failure mode Assessed range Mean value

RSS [2] Failure to operate(include command)

3E-4 to 3E-3/d 1.3E-3/d

NUREG/CR-1363(for BWRs) [37]

Failure to operate(include command)

8E-3/d

NUREG/CR-1363(for BWRs) [37]

Failure to operate(w/o command)

6E-3/d

Command failure Failure of inboardof both MOVs and outboard MOVs(inboard andoutboard)

2E-3/d

NUREG/CR-4050Generic data [6]

SNPS-PRA[12] App. A.2a

CHU-1987 [39]

CHU-1987 [39]

CHU-1987 [39]

Failure to operate IE-3 to 9E-3/d(include command)

MOV spurious —opening

MOV Spurious —opening

Failure to —close

Inadvertent —opening

3E-3/d

1.6E-/h

9.2E-8/h

3.9E-6/d

1.2E-3/d

Based on GE evaluation.

70

TABLE 5.3. CHECK AND MOTOR OPER VTED VALVE RUPTURE OR EXCESSIVELEAKAGE RATES

Source

RSS [2]

SEABROOKPRA [38]

NUREG/CR/1363 [37]

CHU-1987 [39]

RSS [2]

NUREG/CR-1363 [37]

SEABROOKPRA [38]

CHU-1987 [39]

Failure mode

Internal leakage(severe)

Internal leakage(severe)(all sizes)

Internal leakage(all sizes)

Internal leakage(all sizes)

Rupture

Externalleakage/rupture

Rupture

Disk separation

Assessed range Mean value[IT1] [IT1]

1E-6-1E-7 2.8 E-7

1E.95E-8-1E-105E-6-5E-8 5E-7

IE-6

3.4E-7

1E-7-1E-9 2.7E-8

7E-8

<5E-9a

1.4E-7

Never occurred in more than 10 000 valve-years.

5.2. SOME EXAMPLES OF IE FREQUENCIES IN PSA

The Appendix provides a database print )ut which includes IE lists and their frequenciesfrom a large number of published PS As. Ir this section several examples of such IE listswith frequencies of occurrence are given and ] irovide cross-reference to additional lists of thistype given in previous sections.

Table 2.1 provides the Shoreham (BWR) tra isient IE frequencies.Table 2.2 provides several PSA LOG A Freqaencies for BWR and PWR plants.Table 3.7 provides the German Risk Study (5WR) IE frequencies.Table 5.4 provides the Grand Gulf (BWR) 1)1 frequencies.Table 5.5 provides the Big Rock Point (BW1L) IE frequencies.Table 6.1 provides average B/W (PWR) frecuencies for ATWS IBs.

71

TABLE 5.4. GRAND GULF [36] INITIATING EVENT GROUPS1 AND FREQUENCIES

Initiatornomenclature

Description MeanFrequency(per year)

TlT2

T3AT3B

T3C

TIAS

A515253

Loss of off-site power (LOOP) transientTransients with loss of powerConversion system (PCS)Transients with PCS initially availableTransients involving loss of feedwater(LOFW) but with the steam side of thePCS initially availableTransient caused by an inadvertent openrelief valve (IORV) on the reactorvesselTransient caused by loss of instrumentairLarge loss of coolant accident (LOCA)Intermediate LOCASmall LOCASmall-small LOCA (recirculation pumpseal LOCA)

0.111.62

4.510.76

0.14

8.1E-4

l.OE-43.0E-43.0E-33.0E-2

a For the distribution of the individual transient initiating events over the groups of transients given above, seeSection 3.4.2 (Table 3.11).

72

TABLE 5.5. INITIATING EVENTS FOR B]IP PRA [46] FOR WHICH EVENT TREESWERE DEVELOPED

Initiating event Frequency(per year)

Turbine trip 1.4Loss of main condenser 6.0E-2Spurious closure of MSIV 6.0E-2Loss of feed water 1.6E-1Loss of off-site power 1.3E-1Loss of instrument air 6.0E-2Spurious opening of turbineBypass valve l.OE-1Spurious opening of RDSIsolation valve 1.2E-3Spurious closure of both

recirculation line valves 1.7E-2Stuck-open safety valve 2.6E-4Interfacing LOCA 2.0E-3High energy line break in

recirculation pump room 3.9E-7High energy line break in pipe tunnel 3.8E-6Small LOCA l.OE-3Medium LOCA l.OE-4Large LOCA l.OE-5Small steam line break

inside containment l.OE-3Medium steam line break

inside containment l.OE-4

73

6. EXAMPLES OF FREQUENCY DETERMINATION

This Section gives several selected examples of how some of the approaches discussedin Section 5 have been used in past PS As and in the evaluation of operating experience.

6.1. EXAMPLE OF THE USE OF THE MEAN FREQUENCIES APPROACHAnticipated transient without scram (ATWS) events are sometimes modelled as a

separate class of initiating events [9, 12, 20, 22]. One of the reasons for their treatment asa separate IE (rather than branching out from transient IBs) is the fact that not all transientscan lead to an ATWS. Transients that start at a low power level in the core would in mostcases be controlled (because of negative reactivity coefficient feedback) at a power levelwhich is within the cooling capabilities of the core cooling systems. Only transients startingat a high power level have a high chance of threatening the core integrity.

The EPRI and EG&G reports [4, 5] include separate tables on transients experiencedat power levels of 25 to 110% of full power. Estimation of ATWS IE frequency can proceedusing those tables as the source for operating experience. For Oconee (a PWR plant) 41different IBs are reported in the EPRI tables. They should be reduced into a smaller groupof transient initiators.

The Oconee PRA [9] starts with the data for B/W plants rather than the total PWRdatabase in Ref [4]. The B/W plants have a relatively higher frequency of shutdowns andmany more cases of loss of main feedwater than other types of PWR. The EPRI reportincludes 500 events in 48 plant-years for eight B/W plants. This data for 41 IBs wascondensed into 12 transient categories. Out of these 12 categories, eight have been discardedfor the following reasons:

MSIV closure and RCS loop startup:

Load increase and excessive cooldown:

Control rod withdrawal and boron dilution:

Loss of RCS flow and RCS depressurization:

Not applicable to Oconee-3.

Insignificant contributorsbecause of their low frequencyand comparable consequences to theother.

Bounded in their consequences byother more frequent transients.

Bounded by LOOP discussed laterand small LOG A without scram.The latter is a less severe ATWScase.

The four ATWS categories retained were:

- Loss of condenser vacuum- Turbine trip- Loss of main feedwater- Loss of off-site Power

Total

0-110% 25-110% power

0.2/year5.7/year0.7/year0.2/vear

0.1 /year3.9/year0.2/year0.1/vear

6.8/year 4.3/year

74

In deriving the above frequencies the EPl'I data was used and average frequencies wereobtained from using the number of events that occurred in each category and the timeduration in which these occurrences happened.

Table 6.1 is a summary of the collap ;e scheme of the 41 transients into the 12transients considered first (four of them retained for the ATWS analysis). The table displaysthe frequencies evaluated in the Oconee PRA ft >r the ATWS initiators based on data for B/Wplants taken from the EPRI data [4].

6.2. EXAMPLES OF DETERMINATION OF THE FREQUENCY OF COMMON CAUSEINITIATORS

Common cause initiators (CCIs) are events which, in addition to plant trip, result inthe degradation of one or more mitigation systems. A list of CCIs evaluated in various PS Aswas given in Section 4.2.3. Here several examples of CCI frequency determination aresummarized.

(a) Loss of 480 V bus (Surry-NPP [40])

Here the failure rate of the transformer is used (10~6/h) with a mission time of 8760hours to calculate a mean frequency of k ss of 480 V bus which is al to 9 x 10"3/year.The transformer failure rate was evaluate 1 by reviewing several sources [9, 14, 41, 42,44] to obtain a range of data. The median of this range was chosen. Since thetransformer failure rate was significantly larger than the bus failure rate, the latter wasneglected. Similar analysis was performed in the Oconee PRA where a value of5.4 X lO'Vyear was found for IE 'loss rf power to 4-KV switchgear 3TC'.

(b) Loss of charging pump cooling (Surry-NPP [40])

The charging pump cooling depends (in component cooling/service water systemoperation. The frequency of the event ^ /as derived by comparing generic data fromRef. [5] with Surry plant specific data f om an LER search.

(c) Loss of instrument air system (Oconee 1 'RA [9])

In this case a fault tree for the top event 'instrument air system failure' wasconstructed. Failure rates for all basic events were generated and the top eventfrequency determined. Loss of instrumer t air event is a combination of several causes;the frequency is given below:

Contamination:Pipe rupture:Loss of station air and oneinstrument-air compressor:

Other combination of stationand instrument-air failures:

1.7 x 10-5/h8.3 x lO'Vh

1.1 X 10-6/h

9.1 X 10-?/h

Total: 2.7 X 10"5/h or about 0.2/year.

75

-o TABLE 6.1. COLLAPSE OF 41 PWR IE CATEGORIES INTO 12 ATWS GROUPS USED IN OPRA [9, 15]

Transient Category Group

Loss of Condenser Vacuum

Turbine Trip

Loss of main feedwater

Loss of offsite power

Load increase

Loss of RCS Flow

Control rod withdrawal

RCS depressurization

Boron dilution

Excessive cooldown

MSIV closure

Inactive-RCS-loopstartup

TOTAL

EPRI Initiating EventCategories Type(numbers correspondto events in table 4.1)

25, 27, 30

3, 8, 12, 15, 19, 23, 28,33, 34, 36, 38, 39, 40

16, 22, 24

35

21, 26, 29

1,14

2

4 ,5 ,7

11

6, 9, 20

17, 18

13

Number andB/W Plants

0.110%power*

8

314

30

4

5

13

3

2

0

0

0

0

379

Frequencyfor Second

25-110%power**

3

104

7

1

0

5

2

2

0

0

0

0

124

of OccurrencesYear and Later

Frequency0-110%

0.20

7.90

0.75

0.10

0.13

0.33

0.08

0.05

<0.01

<0.01

N.A.

<0.01

9.54

in

Frequency25-110%

0.10

3.61

0.24

0.03

<0.01

0.17

0.07

0.07

<0.01

<0.01

N.A.

<0.01

4.29

* Number of events (0 to 110% power) for the second year of operation and later on (until 1982 - 39.8 year of operation)** Number of events (25 - 110% power) for the second year of operation and later on (until 1982 - 28.8 years of operation)

(d) Steam line break (Oconee PRA [9])

This event was evaluated on the basis o f operating experience and two stage Bayesianupdates: one event in 205 reactor-yea *s was assumed as prior and no event in 19reactor years as updating evidence Phis resulted in an estimated frequency of3 x 10~3/years for this IE in Oconee. ' Tie same approach was used to determine thefrequency of the 'steam generator tube rupture initiator'.

6.3. EXAMPLES OF DETERMINATION OF THE FREQUENCY OF LOG As

Table 5.1 lists several approaches to the determination of LOG A frequencies. Section5.2 discusses these approaches further. In the following we present several examples takenfrom actual studies:

6.3.1. Maine Yankee PSA [45]

The approach applied for the Maine Y* nkee PSA was to use similar experience fromother plants. It used the WASH-1400 generic frequency for medium and large LOCAs(8.0E-4, 3.0E-4 respectively). For small LO< 'As (three different subgroups) and for ruptureof single SG tube, it utilized a combination (f generic and plant specific data as follows:

(a) Small LOCAs can be the results of sm ill pipe breaks from failure of reactor coolantpump seals [52] or can be induced by i transient event that challenges the PORV orother safety valves. Based on previous PSAs a generic value of 7 x 10~3 per yeartotal was used. Based on plant spec!ft; information, a further breakdown was madeas follows:

- Very, very small LOCA 5.10~3/year;- Very small LOCA 1 X 10-3/year;- Small LOCA 1 X 10-3/year.

all of them have a range factor (uncert dnty factor) of 5.

(b) Steam generator tube rupture is a verr, very small LOCA that has the potential tobypass containment and is therefore treated separately. The frequency of this IE wascalculated as follows:

Maine Yankee has 17 100 tubes of which 70 are plugged. A generic frequency of1.4 x 10"6 per tube-gear was adopted from a previous PWR PSA which results in aSG tube rupture initiator probability oil:

1.4 x 10'6 X 17030 = 2.4 x 10-Vjear.

6.3.2. Interfacing LOCA frequency based on pipe and valve rupture/leakage failurerates

The interface (or intersystem) LOCA i( defined as a leak from RCS due to rupture oflow pressure piping connected to the primaiy system as a result of overpressurization whichis due to loss of isolation between low and high pressure piping. An estimate of thefrequency of interfacing LOCA may be deten lined based on the actual length of low pressure

77

piping and the number of pipe sections between the valves. In such an evaluation the detailof the length of all piping paths from the RCS to the area outside containment is listed. Foreach path the pipe sections and associated valves are identified. Data on valverupture/leakage rates such as in Tables 5.2 and 5.3 are used. Table 6.2 summarizes variouspaths to outside containment including their estimated frequencies (from Ref. [48]).

6.4. EXAMPLES OF DETERMINATION OF THE FREQUENCY OF TRANSIENTS

6.4.1. Maine Yankee PSA [45]

Maine Yankee is a PWR type plant which has been in operation since 1972. Therefore,the approach used for determining transient frequency is relying on the plant's own operatingexperience (i.e. 'old plant approach').

Three data sources were reviewed in the process of determining the transientfrequencies:

(a) Outage and load reduction information (1/1/73 to 3/28/87);(b) Scram database (for a similar period);(c) The NUREG/CR-3862 [5] transient occurrences for Maine Yankee.

List (a) was used from the second year of operation (1/1/74). It correlates well withthe other two sources in most cases. In general the in-plant records were the main sourceused in the frequency determination of the more frequent transient IBs. The list of plant tripswas reviewed to identify events that occurred at a power level greater than 15% and wereunscheduled. Over the thirteen years considered in the PSA, a relatively constant averagefrequency of nine events per year was observed.

Based on a review of plant response to transient initiators, it was concluded that threetypes of event should be considered in the Maine Yankee PSA:

(a) Normal plant trip (do not fail any additional equipment);(b) Closure of all flow check valves or loss of condenser vacuum;(c) Loss of all condensate pumps.

The frequency of IE (a) was determined for plant records as discussed above. Thefrequency of IE (b) was determined based on NUREG/CR-3862 [5] data for Maine Yankeewhich indicated a mean value of 0.04 per year for the closure of the check valves. Thefrequency that the NUREG/CR-3862 indicated for Maine Yankee on loss of condenservacuum was too high compared to actual plant specific experience. Therefore the samefrequency was used for both contributors in that category. It resulted in the total frequencyof 0.08 per year for this class of IBs. The frequency of case (c) was determined to be 0.02per year based on NUREG/CR-3862 data but modified to a somewhat higher value accordingto engineering judgement of the uncertainty involved in this event.

In addition to the above mentioned four categories of IBs additional special plantspecific IBs considered in the Maine Yankee PSA include:

(a) Loss of service water 2.0 X 10~4/year;(b) Loss of secondary component cooling 2.3 x 10"2/year;(c) Loss of primary component cooling 2.3 x 10~2/year;(d) Loss of control air 3.2 x 10~2/year.

78

TABLE 6,2, ESTIMATED FREQUENCIES OF BREAKS OUTSIDE CONTAINMENT

BREAK LOCATION

Reactor CoreIsolation Cooling(RCIC) Steam Line

RCIC/HPCI SteamDrain Line

Reactor WaterCleanup System(RWCU) SupplyLine

CASE

I

II

III

I

I

II

HI

BREAKSIZE

4"

3"

1"

1-1/2

6"

6"

3"

NUMBER OF:LINES

1

1

1

1

1

1

2

SECTI0NS

1

6

14

1

1

1

3

VALVES(*)

1

6

14

0

1

0

3

ISOLATION VALVESVALVESDESIGNATORS

IE51-MOV041

IE510MOV041and MOV042

IE51-MOV048and

it Ji-nu»w**

IE51/IE41AOV-081 orAOV-082HOV033 (F001)and MOVF102and MOVF100

F106The above andIG33-MOV034or(F004)same as above

ASSUMEDFAILUREPROBABILITY

1.0

1.0

1.0

1.0

1.0

1.0

1.0

INITIALBREAKFLOW:STEAMOR LIQUID

steam

steam

steam

steam

Liquid

Liquid

Liquid

ESTIMATEDFREQUENCYOF BREAKOCCURRENCE

2.1E-6

5.8E-6

1.2E-3

5.E-5

2.1E-6

7.5E-6

5.4E-4

DESCRIPTION OF THE CASE ANALYZED

Break exclusion section and valve between ReactorBuilding penetration and the outboard MOV 042.(Elevation 87).Non break exclusion sections and valves from out-board isolation valve up to RCIC turbine. Fouropenings per year of valve -042 are assumed (Allelevation below elevation 87 down to elevation 8).Non break exclusion section and valves from Reac-tor Building Penetration up to the i-l/£ HPCi/IAI« tv u i a i i i iiuc uu vuimcii3ci» iw» ma i i j vpcn(Elevation 87 down to Elevation 8).Section between HPCI and RCIC drain lines connec-tion and the penetration to the main steam tunnel.(Between elevation 11 and 70).First section 1n Reactor Building. It is breakexclusion and normal operating. (Elevation 112)

Section from outboard isolation valve to the 6x3"reducer. Non break exclusion (Elevation 112).Section and valves from reducer up to RWCU pumps.(Elevation 112)

The frequency of these events was estimated using special plant specific models suchas fault trees.

6.4.2. German Risk Study (GRS) PSA phase B [22]

The German PSA (phase B) included two main types of transient initiators:

(a) Anticipated transient

- Loss of AC power 0.18/year- Loss of main feedwater w/o loss of heat sink 0.1/year- Loss of main feedwater with loss of heat sink 0.29/year- Loss of main heat sink w/o loss of FW 0.36/year

Total: 0.93/year

(b) Unlikely transients

- Large leakage in steam line within containment -1.6 x 10~4/year- Large leakage hi steam line outside containment -4.8 x 10~4/year- Medium leakage in steam line within containment -2.1 x 10~5/year- Medium leakage in steam line outside containment -1.1 x IQ'Vyear

Total: 7.8 x 10'Vyear

(c) ATWS (various transient sources)

The frequency of case (a) were calculated using the two stage Bayesian updating withplant specific data. The plant data was based on the Biblis B experience since it wasconnected to the grid in 1977. Some modifications hi the system design that weremade in Biblis were taken into account. The prior distribution was assumed to be non-informative (uniform). The update was made using Gama distribution. The frequencyof loss of AC power was calculated based on a database larger than the Biblis Bdatabase.

6.5. EVENTS IN SHUTDOWN

Traditionally, PS As have been modelling plants assuming full power operation. Severalincidents involving loss of decay heat removal capability during shutdown pointed out thevulnerability of the plant in the shutdown mode, and eventually prompted the performanceof PSA studies for other operating modes than full power.

Two of the PSA studies for French 900 and 1300 MW(e) reactors performedrespectively by the Commissariat a 1'energie atomique (CEA) and Electricite de France(EDF) indicated that a high percentage of the total risk is actually coming from the shutdownstate. Similar events are currently being studied elsewhere.

Although loosing decay heat removal is the major event to be analysed hi shutdown,if the analyses covers other non power states, additional events should be considered. In theCEA study of French 900 MW(e) reactors [49] the following non-power states wereconsidered:

period while the reactor is shut down, with temperature below 280°C and pressurebelow 133 bars, but above residual heat removal (RHR) system parameter;

80

primary circuit full of water, RHR conrected and in operation;primary circuit partially drained;refuelling, with reactor vessel head open and pond above reactor vessel full.

The list of initiators considered for sorn; non-power states in this study includes:

loss of primary coolant;SG tube rupture;LOCA outside containment;loss of electric power supply;dilution accident;loss of RHR.

In a recent study [47] performed at the Sandia National Laboratories, USA, initiatingevents for the operating modes other than full power operation for a BWR plant have beendetermined. The operating modes considered 'vere:

low power operation (less than 15% power);startup;hot shutdown;cold shutdown;refuelling.

The general criteria for determining the i litiators were: (1) any disruption from normaloperation requiring rapid shutdown and challer ging safety systems to remove decay heat (formodes 1 and 2); and (2) any event which requires an automatic or manual action to preventcore damage (modes 3, 4 and 5).

The initiating events identified were div ded into five major groups:

(1) Transients;(2) LOCAs;(3) Decay heat removal (DHR) challenge events;(4) Special events;(5) Hazards.

The first step in identification of initiati tig events was the evaluation of applicabilityof initiators considered in full power mode foi each of the non-power modes. Additionally,initiators unique to modes other than full powei were postulated. New types of initiators havebeen identified in shutdown modes (modes 3, 4 and 5) and were basically due to changingplant states (equipment out of service, etc.) aid increased human activities in those modes.Table 6.3 summarizes potential initiating events identified for each of the non-poweroperating modes.

81

TABLE 6.3. POTENTIAL INITIATING EVENTS AND OPERATING MODESTO WHICH THEY APPLY

INITIATING EVENTS

TRANSIENTSTl: Loss of Off -Site PowerT2: Loss of Power Conversion SysT3A: Power Conversion Sys. Avail,T3B: Loss of FeedwaterT3C: Inadvertent Open Relief Vlv.

LOSS OF COOLANT ACCIDENTS (LOCAs)A: Large LOCASI: Intermediate LOCAS2: Small LOCAS3: Small-Small LOCAHI: Diversion to Suppression PoolH2 : Diversion to CondenserJ: LOCA in Connected SystemK: Test/Maintenance -Induced LOCA

DHR CHALLENGE INITIATORSE1A: Loss of Condensate (DHR)E1B: Loss of RHR- Shutdown CoolingE1C: Loss of RWCU (DHR)E1D: Loss of Alternate DHR SystemE1E: Loss of RCICE2A: Isolation of CondensateE2B: Iso. of RHR- Shutdown CoolingE2C: Isolation of RWCU (DHR)E2D: Iso. of Alternate DHR SystemE2E: Isolation of RCICSPECIAL EVENTSCriticality Events:T4A: Rod Withdrawal ErrorT4B: Refueling AccidentLoss of Support Systems:T5A: Loss of Standby Serv. WaterT5B: Loss of Turb Bldg Cool WaterT5C: Loss of Plant Service WaterTIAS : Loss of Instrument Air Sys .TEP: Loss of electric powerOther Events :TORV: Inadvertent Open Relief

Valve - ShutdownTIOP: Inadvertent

OverpressurizationHAZARDS EVENTSInternal FireInternal Flooding

MODE 1L

XXXXX

XXXX

X

T2

XX

MODE 2

XXXXX

XXXXXXXX

X

X

X

X

X

XX

X

X

XX

MODE 3

X

XXXXXXXX

XX

XXX

X

X

XX

XX

X

X

XX

MODE 4

X

XXXXXXXX

XX

X

XX

X

X

XXXXX

X

XX

MODE 5

X

XXXXXXXX

XXX

XXX

XX

XXXXX

XX

T2 - Included Under T2 Transient

82

7. INITIATING EVENTS FOR WWER PLANTS

Most of the eastern European countries operating WWER units have now their ownPS A programme. Two Level 1 PS A studies lave been completed to date: the GreifswaldNPP PSA, Germany, and the Loviisa NPP PS A, Finland. In addition to these analyses, a so-called 'Top Level Risk Study' has been performed recently in Bulgaria for the KozloduyNPP, Units 1 to 4 [60],which involves limitec scope PSA as well. The other ongoing PSAanalyses for WWER plants show variety with "espect to the status of completion: some are'semi-finished' and some are in a preliminary state. While the other examples of PWR IBsappearing in this document are taken from completed studies, the information given in thisSection represents the current status of selec ion, grouping and quantification of IBs forWWERs.

The collective term 'WWER' covers three types of PWR with different designcharacteristics:

(1) WWER-440/230;(2) WWER-440/213;(3) WWER-1000/320.

Due to the fact that at present PSA of' ' VWER-440 type reactors is better developedthan the WWER-1000 analyses, qualitative information and quantitative data provided in thisSection primarily relate to WWER-440 reactor types, except for some references to WWER-1000 data, such as the accumulated operating experience for WWER-1000 plants inSection 7.3.

The 230 and 213 models represent two generations of the WWER-440 type reactorshaving distinctive safety features. The main differences in safety characteristics of the basic230 and 213 models are briefly described bek w.

The maximum design basis accident is ; LOG A with break size 100 mm for the 230reactors and large LOG A with break size 50C mm (equivalent to the diameter of the mainprimary coolant pipeline) for the 213 reactors. The primary cooling circuit equipment andcomponents of the 213 model are housed in a leaktight containment to cope with 0.15 MPaoverpressure. No real containment but a press ire resistant confinement exists for the W230type. The 213 type plants are equipped with (ore flooding hydroaccumulators as well as abubbling condenser-vacuum system, while tl e 230 plants are not. The emergency corecooling system of the basic 230 model have a redundancy of 2 x 100%, designed to copewith the 100 mm LOCA. The emergency core cooling system of the 213 type consists of thehigh and low pressure systems, both with redundancy of 3 X 100% designed to cope withthe 500 mm LOCA.

In the following sections the 213 and 230 types of WWER-440 reactors are treatedtogether in order to cover the issue of IE defii ition for both models. Therefore the WWERinitiating list (definition and grouping) and the associated frequencies should be handled withcare to avoid misinterpretation of the inform; tion, e.g. use of an IE irrelevant to a givenWWER type plant.

83

Within the framework of an IAEA project set up to help PSA activities of WWERowners and to promote the exchange of information among them, a specific workshop washeld in Rez, Czechoslovakia, in February 1992. At that workshop a section was devoted tothe definition, grouping and frequency determination of IBs for WWER type reactors.Sections 7.1 to 7.3 summarize the achievements of that common effort, reflecting also thecontributions from the participating countries (Bulgaria, Czechoslovakia, Germany, Hungary,Poland and Russia). Regarding the determination of IE frequencies, some additional data isprovided in Section 7.3 taken from the Kola-1 PSA study and from the 'Top Level RiskStudy for Kozloduy' [60] besides the IE frequencies made available during the Rezworkshop.

7.1. SELECTION OF WWER INITIATING EVENTS

The WWER operators had developed their own initiating event lists. At Rez theselection methodologies used by the participating teams were discussed in detail. The selectiontechniques followed by them are given below.

(A) Approach used for Kozloduy NPP, Bulgaria

In the identification of IBs for the Kozloduy NPP, the actual plant operating experiencewas used as much as possible. However, the major procedural steps were as follows:

(1) Generic IE lists and the experience gained throughout the world helped to forman initial cut of IBs and their sub-events.

(2) Plant specific information from the operating history was then used to modify thislist.

(3) A list of IBs was derived as a result of steps (1) and (2) above.

(B) Approach used for Dukovany PSA, Czechoslovakia

(1) Selection of IBs (sources):

(a) Accident analysis in CSFR safety analysis reports.

(2) Selection criteria:

(a) IE must cause reactor scram; and/or(b) A selected IE must cover several individual similar IBs.

(3) Based on the above methodology a list of 14 IBs was derived.

(C) Approach used in Greifswald PSA, Germany

(1) Reference to previous PSAs (especially to the German Risk Study (GRS) [22]).

(2) Use of the EPRI list of IBs [4].

(3) Use of the draft of German PSA guidelines.

84

(4) Engineering evaluation; technical study of the plant.

(5) Analysis of the plant specific ope 'ating experience.

(6) Performing a limited scope PSA, which excludes many of the initiating events.

(D) Approach used for Paks PSA, Hungary

(1) As a first step a generic list was gathered from publicly available PWR PS As.This step considered the following sources:

(a) EPRI-NP-2230, Interim Re] >ort (1982) [4];(b) Results of the German Risk Study, Phase B [22];(c) Studvik SUPER-ASAR Rej ort, 1987 [29];(d) Oconee PRA-NSAC-60, 19 54 [9];(e) Zion PRA, 1981 [44];(f) Seabrook station PRA, PLC -0300, 1983 [14];(g) IAEA Initiating Events Datibase, 1988;(h) A list of IBs especially mide for WWER type reactors by the Soviet

specialists for the former C VlEA programmes.

(2) As a second step, the above sourc 3s were used to construct a comprehensive IElist of 96 IBs.

(3) The list was then discussed in an sxpert group consisting of specialists in PSA,thermohydraulic analysts, system analysts and experts from the operatingpersonnel. As a result of this carei al selection process a list of IBs was generatedconsisting of 66 different initiatoi s.

(4) A further step was taken, of engir eering evaluation, using master logic diagram(MLD) methodology, operating < xperience and comparison with results fromother PSA studies.

(5) Further reduction of the IE ist will be carried out on the basis ofthermohydraulic analyses as they become available.

(E) Zarnowiec approach

(1) Several generic PSA studies, demoted to IBs, and several plant specific PSAreports were reviewed in order to identify candidates for WWER plant specificIBs. These sources included:

(a) EPRI categorization schen* as described in NUREG/CR-3862 [5];(b) IAEA IE database (Status I )ec. 1988);(c) PRAs for NPPs with PWR type reactors such as Oconee PRA [9], Surry

[40], Borselle;(d) Available information on P SA results for WWER-440 power plants such

as Loviisa, Kola, Dukovan?.

(2) The information from all the abovi: sources was carefully evaluated, and a generictype list of IBs was generated.

85

(F) Approach used for WWER IE selection, Russia

(1) In general, the approach is based on IAEA and NRC PSA procedure guides.

(2) As a first step a comprehensive list of possible IBs is composed by using:

(a) A detailed analysis of the plant;(b) Operating experience from similar PSAs or NPPs;(c) Environmental studies;(d) Results from probabilistic assessment of equipment rupture possibilities;(e) Generic databases (IAEA and others);(f) The list of design basis accidents;(g) Supervision organization (licensing authorities) requirement.

(3) All the above information is evaluated taking into account:

(a) Is the IE suggested a simple or a complex IE including several dependentor independent events?

(b) Where does the IE occur (reactor, spent fuel storage pool, etc.)?(c) Duration of the IE. This applies to loss of off-site power and spurious

opening of valves.(d) The size of the failed device.(e) The state of the NPP when the IE took place (power operation, refuelling,

startup after refuelling, shutdown).(f) Is the IE within the design basis? If it is, the plant response is known from

safety analysis. If not a design basis IE then the plant response needs firstof all to be understood, in order to facilitate evaluation of the suggested IE.

(g) Only IBs requiring a safety function performance are included in theRussian list of IBs.

As a result of the individual activities in the field of IE selection, the following sourceswere used to develop a generic IE list for Level 1 PSA of WWER-440 type reactors:

- List of IBs for Kozloduy PSA activities, Bulgaria;- List of IBs selected for the preliminary NPP Dukovany PSA Study, Czechoslovakia;- List of IBs used for Greifswald PSA, Germany;

List of IBs for Level 1 PSA of Paks NPP, Unit 3, Hungary;- Generic list of IBs, Poland;

List of IBs for WWER-440 reactors, Russia.

The IE lists mentioned above were compared and evaluated in order to create ascomplete as possible a WWER-440 initiating event list. The IE list derived from this processis provided in Table 7.1 for LOCAs and in Table 7.2 for transients and CCIs, both taken fromRef. [59]. The list covers 80 internal initiators during full power operation. External andinternal hazards as well as events during shutdown or low power operation are not included.

86

TABLE 7.1. CLASSIFICATION OF LOCA;, FOR PSA OF WWER-440 PLANTS

Initiating event

1. GROSS REACTOR VESSELRUPTURE

2. LARGE LOSS OF COOLANTACCIDENTS (OVER DN 200 mm)

2.1. Large LOCA: loops 2, 3 and 5cold leg

Rationale for inclusion

It causes an immediate loss of allcooling water in the reactor vessel. Themitigating systems are considered to beineffective. Note: in case of smallerreactor vessel damage the low pressureinjection can be effective.

The break location affects theeffectiveness of the high pressureemergency core cooling system due tothe fact that the high pressure ECCSlines are connected to the cold leg ofloops 2, 3 and 5 respectively. Bothhydroaccumulators feeding to thedowncomer are thought to beineffective.

2.2. Large LOCA: loops 2, 3 and 5hot leg

In this case the water fed by thehydroaccumulators connected to theupper plenum is lost. With respect to thecore flooding system only the other twohydroaccumulators can mitigate theconsequences of the initiating event. Theeffectiveness of high pressure injectionis not affected as much as in the case ofbreaks on the cold leg of loops 2, 3, 5.

2.3. Large LOCA: loop 4 cold leg One train of the low pressure ECCS isconnected to loop 4, therefore this lowpressure line is considered to beineffective in addition to the hydro-accumulators feeding to the downcomer.

87

TABLE 7.1. (cont.)

Initiating event Rationale for inclusion

2.4. Large LOCA: loop 4 hot leg Similarly to the break on the cold leg ofloop 4, one low pressure pump is almostlost in the case of this initiating event.O n t h e o t h e r h a n d , t h ehydroaccumulators feeding to the upperplenum cannot be effective.

2.5. Large LOCA: loops 1 and 6cold leg

No mitigating system line is connectedto loops 1 and 6. Thus, this breaklocation reduces the effectiveness ofonly the hydroaccumulators connected tothe downcomer.

2.6. Large LOCA: loops 1 and 6 hotleg

The mitigation provided by the hydro-accumulators that are connected to theupper plenum is degraded by thisinitiating event. The effectiveness of theother mitigating systems are not affectedby the location of the break itself.

2.7. Rupture of connection line ofhydroaccumulator feeding down-comer

One low pressure ECCS line isconnected to this hydroaccumulator line.Thus, one low pressure pump is lost inaddition to the two 'lower' hydro-accumulators if this event occurs.

2.8. Rupture of connection line ofhydroaccumulator feeding upperplenum

Concerning the effectiveness of themitigating systems the initiating eventimplies the same consequences as theother hydroaccumulator line break (seeIE 2.8). The only difference is that theother two hydroaccumulators are lost inthis case.

88

TABLE 7.1. (cont.)

Initiating event

3. MEDIUM LOSS OF COOLANTACCIDENTS (DN 32-200 mm)

3.1. Medium LOCA not affectingemergency core coding systemoperation


No degradation is considered in themitigating systems owing to the locationof the break.

3.2. Medium LOCA affecting highpressure ECCS operation

3.3. Medium LOCA affecting lowpressure ECCS operation

Some portion of the cooling water fedby one of the high pressure pumpsimmediately leaks through the break, sothat pump cannot perform its function.

This initiating event causes degradationin the redundancy of the low pressureECCS.

3.4. Medium LOCA affecting bothhigh and low pressure ECCSoperation

4. SMALL LOSS OF COOLANTACCIDENTS

There are certain pipelines in theprimary system the rupture of whichreduces the effectiveness of both highand low pressure injection.

4.1. Small LOCA initiating ECCSoperation

Compared to IE No. 4.3 the break sizein this case is such that it requiresECCS operation regardless of the powerrate of the make-up water pumps

4.2. Inadvertent opening ofpressurizer relief valve(DN<25mm)

The stuck open relief valve on thehighest point of the primary circuitresults in a specific non-isolable LOCAsituation that should be handled as aseparate IE.

89

TABLE 7.1. (cont.)

Initiating event Rationale for inclusion

4.3. Small LOCA, that can becompensated by make-up water(DN<13mm)

The make-up water system is sufficientto keep the balance of coolant loss andcoolant addition giving a wide timewindow to the operator to takecorrective actions.

LOSS OF COOLANTACCIDENTS OUTSIDECONTAINMENT5.1 .a. Steam generator tube

rupture (ca. 5 cm2)The main difference why the singlesteam generator tube rupture is handledseparately from other more extensiveleakages between the primary and thesecondary circuit is that in this case theoperation of the ECCS is not required orcan be prevented by correct operatoractions.

S.l.b. Several steam generatortubes leakages

In addition to the ECCS operation thisinitiating event results in greaterradioactivity release to the secondarycircuit/environment than IE No. 5.1.a.

5.2. Steam generator collectorrupture

The loss of coolant caused by this IE isconsidered to be equivalent to the lossin case of a medium LOCA in spite ofthe fact that the break size correspondsto a large LOCA (DN 400 mm),because the primary coolant goesdirectly to the secondary side and not toa low pressure space (i.e. thecontainment) in this case.

90

TABLE 7.1. (cont.)

Initiating event

5.3. Interfacing-systemLOCA towards highpressure ECCS


The basis for the distinction betweeninitiating events Nos 5.3 and 5.4 are thedifferent rates of flow bypassing thecontainment (DN 111 mm).

5.4. Interfacing-systemLOCA towards lowpressure ECCS

The loss of primary system flow ismuch higher than in the case of IE 5.3.There is a factor of about 2 between therelevant nominal pipeline diameters forthe two cases.

7.1.1. Selection of LOCAs

The LOCA type IBs on the WWER lisi are broken down into three main categoriesbased on the break size. The categories are as follows:

Large LOCA;Medium LOCA;Small LOCA.

It should be noted that the break sizes; appearing in Table 7.1 reflect the state ofclassification as of February 1992. Verification and finalization of the various break sizeboundaries are still going on as discussed in S action 7.2.1.

The large LOCA events are further subc!Hot leg and cold leg locations are treated silocations. The reason for this detailed classificalof the primary circuit can impact on the effeclThe medium size LOCAs are categorized like1;4.3 within the small LOCA category does notbut implies different break sizes. The appro*Table 7.1.

ivided according to the location of the break,parately in addition to other specific breakion is the fact that breaks at different locationsiveness of the mitigating systems differently,vise. The differentiation between IBs 4.1 andnecessarily indicate different break locationsimate small LOCA break sizes are given in

In addition, three main LOCA categories outside the containment, including ruptureswithin the steam generator and other inter! icing system LOCAs, are also covered inTable 7.1. Reactor vessel rupture is treated as a separate IE.

91

7.1.2. Selection of transients and common cause initiators

The WWER initiating event list of transients and CCIs (given in Table 7.2) issubdivided into nine major groups as follows:

1. Reduction in primary system coolant flow;2. Loss or reduction in feedwater flow;3. Reduction in steam flow;4. Loss of steam;5. Transients causing turbine(s) trip;6. Failure in electric power supply and I&C systems;7. Failure of auxiliary systems;8. Unplanned reactor trip;9. Reactivity transients.

The above classification follows the technological plant features. Thus, these categoriesdo not agree totally with PSA needs and do not imply final grouping of transients and CCIsfor WWER plants.

Group 1

Initiating events 1.1-1.4 do not require reactor shutdown if additional malfunctions donot occur, because the primary circuit is equipped with six loops. Even the simultaneous lossof three reactor coolant pumps — which is the strongest transient of those mentioned above— requires only the reduction of reactor power down to 50%. Due to the structure of theprimary circuit these transients have moderate effects and are not likely to endanger coreintegrity. The simultaneous loss of all reactor coolant pumps is the only event in group 1 bywhich reactor shutdown is needed.

Group 2

The transients listed in group 2 include events causing partial or total loss of feedwaterflow (see IBs 2.1 and 2.2), feedwater line ruptures (see IBs 2.2-2.10) and other eventsintroducing disturbances in feedwater flow (see IBs 2.10-2.15). Loss of one feedwater pumpinitiating event does not require reactor shutdown. This initiating event usually creates littledisturbance in the normal operation of the plant because the feedwater system has fivefeedwater pumps (with four of them normally operating).

The feedwater line ruptures are categorized according to break size and location. Thisgrouping has been done by detailed investigation of the feedwater system piping arrangement.The primary basis for the classification is the degree of degradation in the secondary sidecooling caused by the various feedwater line ruptures. This degradation strongly affects theavailability of the main and/or emergency feedwater systems.

The initiating events 2.12 to 2.15 affect the feedwater availability only in the long term.However, these events to some extent affect the availability of the condenser(s) as a heat sinkas well.

92

TABLE 7.2. LIST OF TRANSIENT AND COMMON CAUSE INITIATING EVENTSFOR LEVEL 1 PROBABILISTIC SAFETY ASSESSMENT OF WWER-440 PLANTS

REDUCTION IN PRIMARY SYSTEM COOLANT FLOW1.1. Trip of one or two reactor coolant pumps1 .2. Simultaneous trip of three reactoi coolant pumps (230)1.3. Inadvertent closure of main gate /alve (primary circuit)1 .4. Reactor coolant pump seizure1.5. Total loss of reactor coolant

2. LOSS OR REDUCTION IN FEEDWATER FLOW2.1. Loss of one feedwater pump2.2. Loss of all feedwater pumps (mo -e than 50 %)2.3. Feedwater collector rupture2.4. Feedwater line rupture that can b J isolated by separation of one steam

generator and compensated by st; rtup of reserve feedwater pump2.5. Feedwater line rupture that can b ; isolated by separation of one steam

generator and cannot be compens ated by startup of reserve feedwater pump2.6. Feedwater line rupture that can b j isolated by separation of one half of

feedwater collector and compens; ted by startup of reserve feedwater pump2.7. Feedwater line rupture that can b J isolated by separation of one half of

feedwater collector and cannot b( compensated by startup of reserve feedwaterpump

2.8. Rupture of feedwater pump discharge line (before check valve)2.9. Rupture of feedwater pump suctii >n line2.10. Feedwater line rupture inside containment2.1 1. Feedwater flow instability2.12. Loss of condenser pump (one)2.13. Loss of condenser pumps (all)2.14. Loss of condenser vacuum2.15. Loss of circulating water

3. REDUCTION IN STEAM FLOW3.1. Inadvertent closure of main stean i isolation valve3.2. Miscellaneous leakages in secondary system

4. LOSS OF STEAM4.1. Inadvertent opening of steam generator relief valve4.2a. Inadvertent opening of steam dut ip valve (bypass)4.2b. Inadvertent opening of steam dm ip valve (to the atmosphere)4.3. Steam line rupture inside contain nent4.4. Steam line rupture outside contai iment4.5. Rupture of main steam collector4.6. Turbine control valve malfunctio i

93

TABLE 7.2. (cont.)

5. TRANSIENTS CAUSING TURBINE IS TRIP5.1. Turbine trip (one turbine)5.2. Turbine tripe (both turbines)5.3. Total loss of electric load5.4. Generator faults/trip

6. FAILURES IN ELECTRIC POWER SUPPLY AND I&C SYSTEMS6.1. Loss of off-site power6.2. Loss of all 6 kV bus-bars6.3. Loss of one 6 kV bus-bar6.4. Loss of bus-bar EV (non-interruptible AC power)6.5. Spurious 'Pbox > 1.1 bar' signal6.6. Spurious 'large LOCA' signal (213) or spurious 'low pressure in the

primary circuit' signal (230)6.7. Spurious 'main steam collector rupture' signal (213) or spurious 'low pressure

on the steam header' signal (230)6.8. Loss of normal grid connection

7. FAILURES OF AUXILIARY SYSTEMS7.1. Loss of service water system7.2. Loss of intermediate cooling to reactor coolant pumps7.3. Loss of intermediate cooling to control rods7.4. Loss of ventilation systems7.5. Loss of high pressure air7.6. Loss of room cooling in a vital instrumentation compartment7.7. Primary system purification system fault

8. UNPLANNED REACTOR TRIP8.1. Spurious reactor trip8.2. Reactor trip due to administrative procedure such as simultaneous unavailability

of two safety systems

9. REACTIVITY TRANSIENTS9.1. Uncontrolled single control rod withdrawal9.2. Uncontrolled control rod group withdrawal9.3. Inadvertent boron dilution9.4. Control rod ejection with reactor vessel damage9.5. Anticipated transient without scram (ATWS)9.6. Control rod drop (single rod)9.7. Control rod ejection without reactor vessel damage9.8. Restart of isolated primary system loop (opening of main gate valve) with low

boron concentration

94

Groups 3 and 4

The initiating events grouped under ca egories 'reduction in steam flow' and 'loss ofsteam' are in accordance with IE groups v s :d in most PS As. These events include valvemalfunctions and steam line breaks. The pifing and valve arrangement on the steam side ofthe secondary circuit is necessarily reflected in groups 3 and 4. Steam line breaks inside andoutside the containment are treated separa e y. In addition, the main steam line break isdistinguished from the main steam collector rupture due to different sizes. Various valveopenings leading to inadvertent steam dump/i elease are identified as separate IBs.

Group 5

Transients listed in group 5 are events r ;sulting in turbine trip. A unique feature of thiscategory is that it includes trip of one turbine and trip of two turbines. Trip of one turbinedoes not require reactor shutdown but reduce i power (50%) operation.

Groups 6 and 7

The initiating events of groups 6 and 7 , ire mostly common cause initiating events and,apart from a few exceptions such as 'loss of o Jf-site power', 'loss of normal grid connection'and 'loss of service water system' which are generic IBs, they are specific to WWERs.

As an example of how these events affe ct several systems/components simultaneously,the consequences of two CCIs resulting from bus-bar failures are described below.

(a) Loss of one 6 kV bus-bar

There are four vital 6 kV AC buses in the power supply system. Depending on which busis lost, the initiating event can cause the Li >ss of electric power supply to a make-up waterpump, reactor coolant pump(s), feedwalei pump(s) and condense pumps at the same time.In addition, the three safety bus-bars are also connected to the 6 kV buses.

(b) Loss of bus-bar EV (non-interruptible A 3 power)

The initiating event causes substantial oss of information in the control room. If itoccurs, the state indicator lamps of the i sactor coolant and feedwater pumps disappeartogether with the indicators of the 6 kV < ircuit breakers. Most of the controls cannot beoperated from the control room. Some of the control circuits and interlocks on thesecondary side are disabled too.

Group 6 also includes three initiators leimed 'spurious signal'. These three events arealso well supported by WWER operating ex pi Tience. Loss of the component cooling systemis broken down to two events, namely the 1 DSS of intermediate cooling to reactor coolantpumps and the loss of intermediate cooling I o control rods. The failures of other auxiliarysystems such as the ventilation systems are n 3t further subdivided in the list.

Group 8

Two reactor trip categories are defined, The first category, the spurious reactor trip,includes both manual and automatic trips. The other category covers reactor trips required bythe technical specifications.

95

Group 9

The last group of the WWER initiating event list covers the reactivity transients. Mostof the initiating events in this category result in core power distribution asymmetry throughreactor power increase (IBs 9.1, 9.2, 9.3, 9.4, 9.7, 9.8). The control rod drop IE has justopposite effects on the core. The control rod ejection with reactor vessel damage initiatingevent (9.4) causes substantial loss of primary coolant, i.e., a LOG A situation in addition tothe reactivity transient. Anticipated transient without scram (ATWS) events are also includedin group 9. Those transients by which the ATWS events should be taken into account are notidentified in the list.

7.2. GROUPING OF WWER INITIATING EVENTS

7.2.1. Grouping of LOCAs

In the WWER initiating event list the division of LOCAs is based on the followingbreak size boundaries:

Large LOCA: break size > 200 mm DN;Medium LOCA: 32 mm DN < break size < 200 mm DN;Small LOCA: break size < 32 mm DN.

As regards large LOCAs, some additional thermalhydraulic analyses are still under wayto investigate the break size limits with respect to plant response and mitigating systemrequirements. Grouping of large LOCAs based on break location, e.g. breaks on hot and coldlegs, etc. are not fully justified by thermal hydraulic calculations either.

The medium LOCAs may have to be further subdivided. The latest calculations suggestthat the broad spectrum of 32-200 mm should be split into 2-3 subcategories. However, theseresults are not reflected in the current WWER list because some more plant response analysesare needed to finalize the break size boundaries. The location of the medium LOCAs affectsthe availability of the emergency core cooling systems; they are therefore differentiatedaccording to break location as well.

Within the small LOCA category a very small LOCA with break size < 13 mm DN isidentified besides the pressurizer valve malfunction. The latter is a specific LOCA situation(see Table 7.1).

The break size boundaries related to the subcategories in the small LOCA group shouldalso be verified by plant response analysis.

The loss of coolant accidents outside containment includes three steam generator failures(tube(s) or collector rupture) and two additional interfacing system LOCAs. These events arehandled as separate groups from other LOCAs because they require specific operator actions.Moreover, the unrecoverable primary coolant released to the environment can have substantialradiological effects in the case of steam generator failures.

Rupture of a reactor vessel is treated as a separate group on the WWER IE list asdescribed in Table 7.1.

96

To demonstrate the efforts being made ii this area, Table 7.3 shows the results of someplant response analyses for WWER-440/2131 /] >e reactors. The categorization of LOCAs givenin Table 7.3 is based on thermalhydraulic ami rses performed in Czechoslovakia, the RussianFederation, Poland and on engineering judge r snt. The rationale of this classification and themitigating system requirements for the various LOCA categories are briefly described below(based on Ref. [63]).

TABLE 7.3. SYSTEM SUCCESS GRIT E RIA FOR LOCAs CLASSIFIEDACCORDING TO BREAK SIZE RANGE (BASED ON REF. [63])

LOCAcategory

LI

L2

L3

L4

L5

L6

Break size range(mm)

10 < D < 20

20 < D < (50-70)

(50-70) < D <(120-150)

( 120-1 50) < D < 2 0 0

200 < D < 300

300 < D < 500

HP

1

1

I

1

1

1

1

System Success Criteria

HPR

1

1

1

1

LPI

1

1

1

1

1

LPR

1

1

1

1

1

11

11

1

CFS

0^a

0-4a

1+1"

1+1"

Hl>

EPS

1

1

1

scsPM

PM

HR

HR

HPIHPRLPILPRCFSEPSSCSPMHR

- High pressure injection system.- High pressure recirculation system.- Low pressure injection system.- Low pressure recirculation system.- Core flooding system (hydroaccumulator ).- Emergency/auxiliary feedwater system- Secondary pressure control system.- Pressure maintenance mode.- Heat removal mode (30 K/h).

a Availability of CFS affects timing requiremenls for secondary side cooling initiation and relatedprobability of operator error.b 1 out of 2 hydroaccumulators connected to upi >e r plenum and 1 out of 2 hydroaccumulators connected todowncomer are required to operate.

97

LOCA group LI

Breaks of size 10-20 mm DN are considered as the smallest LOCA category. The lowerboundary is estimated as the highest break size that can be compensated by normalmakeup water system. For this category energy removed by coolant flow through thebreak is insufficient to ensure depressurization if secondary side cooling is not provided.

Regarding the mitigating system requirements the high pressure injection/recirculationsystem (HPI/HPR) is required to operate for the initial phase of the accident and thelow pressure recirculation system (LPR) can provide long term cooling. If LPR failsHPR can be used for long term cooling too. In both cases secondary side cooling isnecessary besides ECCS.

If HPS is operable but secondary side cooling is not provided, primary feed-and-bleedis the only way of accident mitigation.

If HPS fails in the early phase of the accident, secondary feed-and-bleed and LPSshould provide RCS depressurization and long term cooling respectively. In this casethe availability of core flooding hydroaccumulators (CFS) has substantial effect on thetime window for the operator to initiate depressurization.

LOCA group L2

The lower break size boundary (~20 mm DN) is based on requirements againstsecondary side cooling. To mitigate LOCAs of size larger than this lower limit onlyHPS is required even if secondary side cooling is not provided.

Upper boundary (~50-70 mm DN) of this category is considered to be the break sizeby which sufficient time margin is available for correct plant state diagnosis and formanual alignment of secondary side cooling. Accident mitigation related to this categoryis similar to that of group LI except that secondary side cooling is not needed if HPSdoes not fail.

LOCA group L3

The estimation of upper boundary for this category (-120-150 mm DN) is based onengineering judgement.

As regards the mitigation system requirements HPS is the only system that can establishstable core cooling in the initial phase and RCS depressurization down to LPSoperational pressure limit. Secondary feed-and-bleed is considered to be unachievablebecause of timing requirements. Accident mitigation related to this category is similarto that of group L2 except that in the case of HPS failure secondary feed-and-bleedwith subsequent use of LPS is not taken into account.

LOCA group L4

The upper break size boundary (200 mm DN) is estimated on the basis of existingthermal hydraulic evidence.

98

For LOCAs of this category the operatic i of CFS and LPS provides an alternative wayof successful accident mitigation besids1 the mitigation measures described for groupL3.

LOCA group L5

The upper break size boundary (300 mm DN) is estimated on the basis of existingthermal hydraulic evidence.

For LOCAs of this category the operatio i of CFS and LPS is needed in the early phaseof the accident while LPR is required f 01 long term cooling. In contrast to group L4 inthis case the HPS alone is insufficient for successful accident mitigation.

- LOCA group L6

This category covers all break sizes larj er than 300 mm DN.

For LOCAs of this category the operatio i of CFS and LPS is the only measure that canprovide successful accident mitigation

7.2.2. Grouping of transients and commcm cause initiators

The breakdown of 69 WWER transien ts and CCIs to nine main categories representsa classification primarily based on engineering judgement. A master logic diagram typescheme was the basis for the categorization.

All the disturbances that could potentia 11 f lead to loss of core cooling without primarycoolant boundary failure or to an increase in core power were taken into account byidentifying their main causes. These causes ani the investigation of plant specific potentialsfor CCIs then formed the basis for the nine groups listed in Table 7.2. Within each categorythose events requiring different plant response i.e. different event tree models according tothe plant response analyses and engineering knowledge, were treated separately.

This approach appeared to be an effect] v ; means for both establishing a comprehensivelist of WWER IBs and performing an initial g 'ouping of transients and CCIs. Nevertheless,since the transient list is rather complex, comprising a large number of initiators, the currentgrouping on a judgemental basis can only be regarded as a preliminary categorization thatneeds improvement/finalization to be supported by thermal hydraulic analysis and eventtree/fault tree modelling as well. The refineme it of the initial grouping is in a rather differentstate in the individual WWER PS As, therel o -e no further categorization of transients andCCIs is given here in addition to the one sus'f.ested by Table 7.2.

7.3. DETERMINATION OF THE FREQUENCY OF INITIATORS FOR WWER PLANTS

At present the IE frequencies used in the individual PS As in countries operatingWWER type units are mainly based on their own operating experience and, for rare events,on generic data not specifically developed fcr WWER plants. However, for the majority of

99

IBs listed in Table 7.2 and for some events of Table 7.1 a substantial amount of operatingexperience has been accumulated in the last twenty years, i.e.:

(A) 190 reactor-years of type WWER-440 (both 230 and 213) and of type WWER-1000in the former USSR and in its successor countries;

(B) 50 reactor-years of type WWER-440 (230) in Kozloduy, Bulgaria;

(C) 50 reactor-years of type WWER-440 (230) in Greifswald, Germany;

(D) About 30 reactor-years of type WWER-440 (213) in Paks, Hungary;

(E) Similar experience of type WWER-440 (both 230 and 213) and of type WWER-1000,in Czechoslovakia.

The ongoing elaboration of this experience and the exchange of information amongWWER operators aim to facilitate the development of a generic WWER IE frequency list.Most of the transients ('likely' transients) and some LOG As, such as very small LOG As orPORVs stuck open, can also be quantified on the basis of the existing operating experience.

A list of IE frequencies prepared in the manner described above massively supports theuse of a variety of approaches, mainly the Bayesian updating techniques and the meanfrequency concept to determination of IE intensities for the individual plants.

Concerning rare events, especially LOG As, the frequencies that have been used forPSAs of WWERs are mostly based on generic data. However, some fracture mechanicstudies have recently been carried out in Russia to determine LOG A frequencies. Accordingto these calculations, the LOG A frequencies are as follows:

Large LOCA (break size > 150 mm DN) 4.0 X 10"6/year;Medium LOCA (50mm DN < break size < 150 mm DN) 4.2 x 10"4/year;Small LOCA (break size < 50 mm DN) 3.0 x 10"3/year.

In addition to fracture mechanics, engineering judgement is needed to assess to whatextent generic nuclear IE data or even non-nuclear data can be used for WWER plants. Theresults of these analyses provide frequency of rare events figuring in the WWER IE list.

The fault tree analysis is undertaken in parallel with the event tree/fault treedevelopments in the individual PSAs for those plant specific CCIs that occur due tosystem/subsystem failures (e.g. loss of intermediate cooling to reactor coolant pumps) andfor which the observed failure data is scarce.

Table 7.4 summarizes the IE event data made available at the Rez workshop (datacolumns 1 to 5), supplemented by the IE frequencies used for the Level 1 PS A of Kola-1NPP and also by the IE frequencies used in the top level risk study for Kozloduy Units 1 to4 [60] (last two data columns of Table 7.4). The first two data columns of the table containthe IE that was used in the two completed WWER PSAs, in the Greifswald and Loviisa PSAsrespectively.

100

TABLE 7.4. FREQUENCY OF INITIATING EVENTS FOR PSA OF WWER REACTORS

liL0CAeo

Initiating Event

Reactor pressure vessel rupture

Large LOCA

Medium LOCA

Small LOCA

Very small LOCA

Hydroaccumulator LOCA

Single SG tube rupture

Several SG tubes leakages

SG collector break

SG header cover rupture

LOCA outside containment

[nadvertent opening of?ORV/pressurizer valve leakage

Steam line rupture insideconfinement

GreifswaldPSA

(I/year)I.OxlO'*"

i.oxio-*21

1.2X10"1

S.OxlO"1

Z.OxlO'*31

l.OxlO-*"

1.0x10-'

3.0X10"1

l.OxlO"3

Loviisa PSA

(I/year)5.5xlO-KW

5.3X10"1

5.3X10"4

4.0xlO'3

3.1xlO'5

1.8x10''

53X10-"

2.3xlO'(

3.4x10"'

BulgarianOper. exp.

and Gen. Data(I/year)

l.OxlO'^

1.0X10-*51

l.OxlO-*51

3-SxlO-"*

1.0xlO'!

(7)

3.0xlO'!

Data Source

Polish GenericData

(I/year)

2.0x1 0'*"

7.5xio-<<5)

3.6x\V™

2.4x1 0'J

Russian Data

(I/year)< l.OxlO'5

4.0X10-*3

4.2xlO-4<a

3.0xlO-XJ)

Kola-1 PSA(8)

(I/year)

3.7xlO'3 «32mm)

2.8x1 0"1 (32 10 «0 mm)

UxlO-'cKHolOOmm)

l.OxlO'2

7.1x10-'

1.6xlO'J

4.8X10'4

Top LevelRisk Study forKozloduy'00'

(I/year)

l.OxlO'4

1.0x10''

l.OxlO-'/l.OxlO-*9

3.4x10-'

5.0x10'' (leakage)

1. Oxl O'2 (rupture)

Comments

(1) Based on

(2) Based on

mechanics

(3) Based_oi\operatingexperience

(4) 17 events

r. yrs.... _ . ii

information (notfromWWERS)

(6) Includes PTS

(7) Experienced but

yet

(8) Based onoperating

Bayesianupdating

(9) Units 1,2/3,4

(10) Based onoperatingexperience andengineeringjudgement

s TABLE 7.4. (cont.)

Initiating Event

TRANSIENTS

&

>••("iIS

Tola] loss of RCS flow

Partial loss of RCS flow

RCP failure (incl. RCP shaftsreak)

Core flow blockage

Total loss of main feedwater

Partial loss of main feedwater

7eedwater line rupture

"eedwater header rupture

Main steam line break

tfain steam header break

nadvertent opening of SO relic:valve

nadvertent opening of steamdump valve to atmosphere

[nadvertent opening of steamdump valve to condenser

"urbine trip (one turbine)

'urbine trip (both turbines)

Generator trip

feactor trip

Data Source

GreifswaldPSA

(I/year)

1-5x10''

l.OxIQ-'

Loviisa PSA

(I/year)

2.8x10''

2.3X10'2

.7x10-'

BulgarianOper. exp.


1.0xlOJ-1.0xlO<3)

1.7x10-'

i.Oxio-M.Oxio-<!)

8-5x10-'

3.0x10''

8.5x10-'

Polish GenericData

(I/year)

2.5xlO-*!>

Russian Data

(I/year)

Kola-1 PSA(8)

(I/year)

8.8X10'3

7.1X10'3

7.1x10''

4.4x10-"

1.6X10'2

7.5x10-'

2.1

Top LevelRisk Study foKozloduy^10'

(I/year)1.0x10-'

5.0x10''

I.OxlO'Yl.OxlO-'^

1.0x10-'

l.OxlO"'

5.0x10''

5.0x10''

1.0x10''

i.oxio-'1.0X10'2

5.0X10'2

5.0xlO'J

.0

.0x10''

.0 (spurious trip)

Comments

(1) Based onembrittlement

(2) Based onfracturemechanics

(3) Based onoperatingexperience

(4) 17 eventsoccurred in 50r. yrs.

(5) Genericinformation (no!fromWWERS)

(6) Includes PTSconsiderations

(7) Experienced butnot quantifiedyet

(8) Based onoperatingexperience andBayesianupdating

(9) Units 1,2/3, 4


TABLE 7.4. (cont.)

T

ANS

II T

ENTS

&

CcIS

Initiating Event

Control rod withdrawal (bank)

Control rod withdrawal(ejection)

Inadvertent boron dilution

Single control rod drop orinsertion

ATWS

Reconnection of idle loop

Inadvertent high pressure safetyinjection

Loss of off-site power for morethan 2 Hrs

Loss of normal grid connection

Total loss of DC

Loss of AC/DC bus

Loss of component/service wateicooling

Loss of instrumentation roomcooling

*CS purification fault

GreifswaldPSA

(I/year)

l.OxlO'5

1.0x10-*

1.2X10"1

Loviisa PSA

(I/year)

S.OxlO'1

2.5X10'2

7.0x10''

3.2x10-'

4.0x10'*

-

BulgarianOper. exp.


6.0x10 2

3.0X10'2

9.0xlO':

9.0x10 2

Data Source

Polish GenericData

(I/year)

Russian Data

(I/year)

Kola-1 PSA(8)

(I/year)

8.0x10 2

Top LevelRisk Study forKozloduy^10'

(I/year)l.OxlO'2

l.OxlO'/l.OxlO'5"'

l.OxlO-1

l.OxlO'1

1.0x10-'

1.0X10'2

1.0X10'2

1.0x10-'

l.OxlO'Vl.OxlO^"

5.0xlOJ

Comments

(1) Based onembrittJement

(2) Based onfracture

(3) Based on

experience

(4) 17 eventsoccurred in 50r. yrs.

information (noifrom WWERS)

(6) Includes PTSconsiderations

(7) Experienced butnot quantifiedyet

(8) Based onoperatingBayesianupdating

(9) Units 1, 2/3, 4


Data provided by the Bulgarian team for Kozloduy is a mixture of generic IEfrequencies and data derived from the quantification of the operating experience. The latteris mainly related to 'likely transients' while LOG A frequencies are based on generic dataexcept very small LOG A.

The Polish data originates from a collection of generic IE frequencies for LOG As.

Frequencies given in the Russian data represent the results of the fracture mechanicstudies for rare LOG As. The vast operating experience is not represented in that datacolumn. However, the Kola PSA data is mainly based on operating experience from theWWER type of NPPs.

In the Kola PSA initiating event frequencies were estimated by the use of Bayesianupdating. For the calculations both WWER and non-WWER sources (primarily the GermanRisk Study, Phase B [22]) were taken as prior distributions. The updating was done on thebasis of in-country operating experience mainly for WWER-440 and in some cases forWWER-1000 type plants. Feedwater line ruptures (two of them are analysed in the study),however, were estimated by using Loviisa NPP data.

The top level risk study data for Kozloduy was determined by considering genericwestern PWR frequencies and by modification of those using applicable site specificoperating experience provided by the plant and engineering judgement as well. Engineeringjudgement was also considered essential to determine the applicability of western PWRfrequencies.

104

8. HAZARDS

When estimating the risks of nuclear power plants, contributions from both internalinitiating events and external initiators (or hazards) need to be considered. While early PS Ashad mainly been addressing internal initiator!!, the current practice worldwide is to includehazards into full scope PS As. Some of the P'JA studies which include hazards have shownthat these have a larger potential risk for the snvironment than internal initiators [53, 54].

8.1. MAIN CATEGORIES OF HAZARDS

Hazards are usually classified under two main categories: internal and external hazards.

The list of hazards is quite established i: id most PSAs include a sublist of the followinghazards:

Internal hazards:

Fires;Internal floods;Turbine missiles.

External hazards:

Forest fires;External floods, high waters;Airplane crash;Seismic events;Extreme winds and/or tornados;

Release of chemical or toxic gas.

Some of the hazards have subcategori;:• or an entire magnitude frequency distributionof initiating events.

Internal and external hazards may als :> be divided according to their origin:

Natural (floods, seismic, tornados, et:);Human (turbine missiles, airplane eras lies, chemical release, etc.).

8.2. EVALUATION OF HAZARD FREQ TENCHES

8.2.1. Seismic analysis

Unlike internal IBs, the occurrence of the seismic initiator is completely independentfrom the plant. The seismic hazard at a »iven power plant is characterized by a hazardfunction which gives the probability of except lance (per year) of a ground motion parameter,such as peak ground acceleration. Figure 8.1 shows a representative hazard curve for a plantsite (nominal with upper and lower uncerta nty bounds). The curves are derived from a

105

1.0

CUMULATIVE NUMBER OF EARTHQUAKES PER YEAR

(NORMALIZED TO 8,800 SQ. MU

PIEDMONT AND UPPER COASTAL PLAINDEFORMED APPALACHIAN HIGHLANDS

0.1

0.01

0.001

\ \\\\ \v\&£

n nr vr

MODIFIED MERCALLI INTENSITY

FIG. 8 1 Example of seismic hazard curve

combination of recorded seismic data, estimates of the magnitudes of the known events,geological information and use of expert opinions.

In addition, the frequency characteristics of the earthquakes is required in order togenerate artificial acceleration time histories for two horizontal and one vertical component.

106

8.2.2. Fire analysis

Fire analysis consists of the following main steps:

Identification of critical locations susc< ptible to fire;Estimations of the potential frequency of fire in these locations;Analysis of fire propagation, taking ir to account mitigating systems like fire alarmsand fire suppression;Identification of the potential for causing LOG As or transients and the continuationfrom that point to core damage using tlie common PS A methodologies.

When fires are treated as initiating evt r ts, then the first two steps are performed. Fireanalysis is different from seismic hazard an li ysis but similar to internal flood analysis (seeSection 8.2.3). The main fire locations that v. ere considered in past PS As are given in Table8.1. For these locations it is also possible tj use historical data from generic nuclear powerplant fire experience (LERs).

The determination of the frequency (f fires in a chosen location can proceed in twoways:

(a) Using the generic operating experience in nuclear power plants. This is a relativerough estimation of the frequency of f r :s, that does not include plant specific features.

(b) Considering the plant specific amoui t of combustible material in each location andusing probability of fire ignition based on the amount of combustible material and thedesign of the area with respect to existing fire ignition sources.

Table 8.1 provides examples of fire lr< quencies used in several PS A studies.

8.2.3. Flood analysis

Flood analysis consists of the followii,; main steps (for internal floods):

Identification of critical locations susce ptible to floods;

Estimation of the potential frequency c f floods in the critical location, from internalfloods;

Identification of floods flow rates ir .hese locations based on pipe and value sizecharacteristics;

Determination of the critical water \e\ e heights at which plant safety systems will failto perform their functions;

Computation of time required to reaci the critical levels;

Determination of corrective action an i time available to their successful performance(and success probability based on this Time periods);

107

oooTABLE 8.1. HAZARD FREQUENCIES IN SEVERAL PS As

Hazard

Earthquake:- Intensity 1- Intensity 2- Intensity 3- Intensity 4- Intensity 5- Intensity 6- Intensity 7- Intensity 8

High Wind/Tornado

Airplane Crash:- On containment- On Control room- On Aux. Building

Frequency of Occurrence (per year):

Oconee Zion German MidlandRisk Study

[9] [43] [22] [13]

7.0E-4 1.5E-49.5E-5 2.0E-55.0E-6 6.6E-7

3.9E-8

3.5E-3 l.OE-3

6.3E-7

Seabrook Big Rock IndianPoint Point

[14] [46] [42]

l.OE-6 l.OE-38.2E-7 l.OE-43.6E-4 l.OE-51.1E-44.3E-52.0E-52.0E-52.5E-6

7.8E-5 — l.OE-4

1.2E-81.4E-72.0E.7

Limerick

[20]

2.3E-4

External Flood:- In Building- Loss of Heat Sink

Internal Flood:(Containment)

(Turbine Building)

2.3E-5 3.0E-31.6E-6

9.0E-74.0E-6

3.2E-4

TABLE 8.1. (cont.)

Hazard Frequency of Occurrence (per year):

Oconee

[9]

Zion

[43]

GermanRisk Study

[22]

Midland

[13]

Seabrook

[14]

Big RockPoint

[46]

IndianPoint[42]

Limerick

[20]

Fires:- Vital Bus or Panel- Cable Spreading Room 5.0E-4 4.2E-3- Turbine Building 6.4E-4- Control Room 1.3E-5 l.OE-4- Containment Penetration 1.8E-3- Aux. Building 4.0E-3

Turbine Missiles — 7.4E-8

Hazardous Chemicals — 6.0E-3

Performing flood event tree analyses of the various flood sequences that may lead tocore damage;

Identification of the critical sequences, their consequence in term of core damagecategory and their probability. Summarizing impact on core damage frequency.

The frequency of the flood IBs may be determined by calculating the frequency of pipebreak and maintenance errors probabilities (e.g. improper isolation following maintenance).This is done for any building (turbine, auxiliary or containment building). Table 8.2provides an example of flood frequencies reported in the Oconee PS A review [15, 51].

no

TABLE 8.2. FLOOD INITIATOR FREQUENCIES AND THEIR CONTRIBUTION TO CORE DAMADE FREQUENCY [15, 51]

FloodInitiators

FVL1NFVL1IIFVL1IO

FVL2NFVL2IIFVT 9TO

FL1NFL1IIFL1IO

FL2NFL2IIFL2IO

FMNFMIIFMIO

TOTAL

FloodFlow Rate(1000 gpm)

350350350

170-349170-349T7O_^4Q

120-169120-169120-169

60-11960-11960-119

12-5912-5912-59

Type ofIsolationPossible

non isolableisolable inletisolable outlet

non isolableisolable inlet-!cr\1oKlo rvsiflof




TotalConditionalProbability ofCore Damage

0.240.0140.012

0.190.010

0.0808.3E-41.6E-3

0.0161.2E-33.3E-4

4.4E-36.6E-4l.OE-4

2.4E-3

Contributionof Flood Initiatorto Core DamageFrequency (Yy-1)

2.4E-62.5E-62.3E-6

3.4E-51.8E-7

3.0E-64.3E-78.8E-7

6.6E-66.5E-61.7E-6

1.1E-55.3E-6LIE-6

7.8E-5

Frequency ofFlood Initiator(Yy-1)

l.OE-51.8E-41.9E-4

1.8E-41.8E-5

3.8E-55.3E-45.5E-4

4.1E-45.4E-35.1E-3

2.7E-38.0E-31.1E-2

3.4E-2

9. DESCRIPTION OF THE DATABASE OF INITIATING EVENTS

9.1. GENERAL DEFINITIONS

For the purpose of providing examples of IE selection, grouping and frequencydetermination, the Appendix summarizes IE data from a number of PS As. This IE dataallows the user of the report to implement the methodology of the use of past experience indetermining the IBs for his study. To ease the use of the database, two printouts have beenprovided: (1) database sorted in accordance to data source, and (2) database sorted inaccordance to initiating event category.

The format of the data and the contents of each of the database field in the Appendixis as follows:

No:

PLANT TYPE:

IE SET:

The sequential number of a record. More than 270 entries are given for25 different studies.

Reactor type (BWR or PWR only).

Basic general category to which the IE belongs (LOCA, transient orcommon cause initiator).

IE CATEGORY: Specific category where an initiator belongs. It provides the grouping ofthe IE which was used in the PS A study (the individual event tree is, inprinciple, developed for specific IE category listed here).

INITIATORS:

FREQ DESC:

FREQ:

RANGE VAR:

UPPER END:

LOWER END:

SOURCE:

Basic initiating event considered before the grouping (for transients) andthe defined break size range (for LOG As).

Frequency descriptor or units used (mean event per reactor-year, or pointestimate events per reactor-year, or median events per reactor-year.

Frequency of the initiating event or of a group of initiators.

Range of variation in IE frequency. An error factor (EF 10 means errorfactor of 10) or a range of variation (5% to 95% confidence level) isprovided.

Upper bound of the frequency distribution. In most cases this is the 95 %upper bound. In case EF is used, this field remains blank.

Lower bound of the frequency distribution. In most cases this is the 5 %lower bound. In case EF is used, this field remains blank.

Provides the name of the data source. It also provides in some cases theexact table in which the data appears in the source.

112

ULT SOURCE: Ultimate source that was us sd to generate the specific data for a particularPS A study. The different ;ases include:

Generic data: data or ginating elsewhere (calculation of operatingexperience, expert opir ion) and generally representative of a spectrumof different plants;Generic data updated i;/ith plant operating experience: generic dataused as a prior, and pi, .nt specific data used in Bayesian updating, togenerate IE frequency;Plant operating experk ace: plant data gathered from events reportedduring plant life;Engineering evaluation: expert opinion of individuals familiar withplant design and opera lion;System analysis using; fault tree model: one of the methodologiesdescribed in this manu il for determination of CCI frequency.

The references of the various sources ;u 5 as follows (in the order of appearance in theAppendix):

Reference

German Risk Study [35]IREP ANO-1 [18]Limeric PRA [20]NUREG/CR 4550 [7, 19]Oconee PRA [9]Sequoyah NPP RSSMAP [34]Shoreham PRA [12]WASH-1400 [2]IREP Calvert Cliffs 1 [41]IREP Browns Ferry 1 [30]Barseback 1 & 2 [29]Forsmark 3 [29]Oskarshamn 1, 2 & 3 [29]Ringhals 1 & 2 [29]Zion NPP PSS [44]Angra NPP PSA [58]Caorso NPP PSS [57]

9.2. SUMMARY OF DATABASE STATUS

INTERFACING LOCA

Number of records: 11Highest value: 4.6E-6/yearLowest value: l.OE-8/year

113

LARGE LOCA


MEDIUM LOCA


REACTOR VESSEL RUPTURE

Number of records: 7Highest value: l.lE-6/yearLowest value: l.OE-7/year

SMALL LOCA

Number of records: 26Highest value: 2.3E-2/yearLowest value: 1.8E-3/year

SMALL SMALL LOCA


STEAM GENERATOR TUBE RUPTURE


TRANSIENT CAUSED BY INADVERTENT OPENING OF RELIEF VALVE

Number of records: 5Highest value: 1.4E-l/yearLowest value: 1.3E-2/year

TRANSIENT CAUSED BY INADVERTENT SAFETY INJECTION


114

TRANSIENT CAUSED BY ISOLATION EVENTS

Number of records: 13Highest value: 2.3E-0/yearLowest value: 1.5E-l/year

TRANSIENT CAUSED BY LOSS OF AC ElUS


TRANSIENT CAUSED BY LOSS OF DC EiUS


TRANSIENT CAUSED BY LOSS OF FBI I WATER


TRANSIENT CAUSED BY LOSS OF OFF SITE POWER


TRANSIENT CAUSED BY LOSS OF POWER CONVERSION SYSTEM


TRANSIENT CAUSED BY LOSS OF SEE VICE WATER


TRANSIENT CAUSED BY STEAM LINE JIREAK


115

TRANSIENT CAUSED BY TURBINE OR REACTOR TRIP OR WITH FRONT LINESYSTEMS AVAILABLE


116

Apaundix

PRINTOUT OF 11 CE DATABASE ONGENERIC IMITATING EVENTS

B?:?H;i1-.

3_;

567g9

10

11121314151617IB19202122232425262728293O313233343536373839

40414243444546474S49SO5152S35*55»57585960

616263646566676369707171737475767778

Pii&NT-'TK P££PWRPWRPWRPWRPWRPWRPWRPWRPWR

PWR

BWRBWRBWRPWRBWRPWRBWRPWRBWRPWRBWRPWRPWRPWRPWRPWRBWRBWRBWRBWRBWRBWRBWRBWRBWRBWRPWRPWRPWR

PWRPWRPWRPWRPWRPWRBWRBWRBWRBWRBWR

PWRABWRPWRABWR

BWRPWRBWRPWRPWRPWRPWRPWR

BWRBWRBWRBWRBWRBWRPWRPWRPWRPWRPWRBWRBWRBWRBWRBWRBWRBWR

i&SEf?LOCALOCALOCALOCALOCALOCALOCALOCALOCA

LOCA

LOCALOCALOCALOCALOCALOCAUOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCA

LOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCA

LOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCA

•:Si¥i;:;:tBS5§A ebR**:S:S:S;s"wS:s:'Urea LOCAmedium LOCAsmall LOCAsmall Icajc on nressina*largeLOCAUrge LOCAmedium LOCAsmall LOCAsmall LOCA

small small LOCA

lam LOCAmedium LOCAsmall LOCAlargeLOCAlargeLOCAmedium LOCAmedium LOCAsmall LOCAsnail LOCAsmall small LOCAsmaU small LOCAhuge LOCAmedium LOCAsmall LOCAsmalt small LOCAmtenystem LOCAlargeLOCAmedium LOCAsmall LOCAsmall small LOCAmtersynem LOCAtune LOCAmedknaLOCAsmalt LOCAsmall small LOCAmteravnem LOCAlargeLOCArejectee vessel rupturesmall LOCA

steam generator tube rupmroUrge LOCAmedium LOCAsmall LOCAmtersystem LOCAreactor vessel ruptureUrge LOCAmedium LOCAsmall LOCAreactor vessel rapturemcerevnetn LOCAreactor venal ruppvetarg»LOCAmedium LOCAmedium LOCAsmaaLOCAsuit LOCAintersvnem LOCAlargtLOCAmedtaiLOCAemail LOCA

kncLCCAO)IsrgaLOCAtt}large LOCA O)medium LOCA (1)medium LOCA (2.)small LOCAUrn LOCAmedium LOCAsmall LOCA (noniwlable)Small LOCA OsolabJe)steam generator tube rupturelargeLOCAmedium LOCAsmall LOCAreactor vessel rupcurelarge LOCAmedium LOCAsrosllLOCA

: ;S:jffi;;:::S:;:;SS:; ^

break with cocuvea greater than 4OO sq.cmbreak with equ .area 80 lo4OD sq.cmbreak wito eou.area2-ta 8OsQ.cmsuet men relief or saietv valvebreak with caujtze treatcr nan 13.5 inchesbreak with cau-sizc 10 to 1 3.5 mchesbreak with am -tar 4 to 10 inchesbreak with tqu-sae 1.66 » 4 inchesbreak witi) equoe 1.2 to 1.66 mchesstuck open prsn safety valv«break witti equivakacwe 38 to L2 fflttesstock opea relief valveRCP seal failurebrat with equ -rae urour than 4 inchesbreak with eqiLsiz* 1 to 4 incheibreak with eqiLsia up to 1 inchbreak with equivalent srze > 6 inchesbreak with equivalent area > .1-.3 aq ftbreak with equivitent size 2 U>6 inchesbreak with equivalent area -BO4-.3 sqfUliql .l-.Jsqftfseam)break with equivalent rize -5-2 inchesbreak wim owiv»lenisrei<X)03sftOiq) or <-! sqft/sani)break with «TU.SJM >S inch or flow 50-100 gpmbreak with flow 50 to 100 <pmbreak with equ intent me 6 to 29 inchesbreak with equivalent size 2 to 6 inchesbreak with equivalent size 1/2 to 2 inchesbreak with equivalent itze <J inches

break with equivalent area >. 1 softbreak with equivalent acea £10* -.1 ssftbreak with equivalent area < .004 sqft(liq) or 01 sqftfiteazn}Iraki with 5O-100 gpm flow (recircolKion pump seal)

break wid] equivalent size >3 sq ftbreak with equivalent area JOOS-.3 aqftfliq) and .U3sqft(steiim)break with equivalent area -c.005 sqftfKq) jod -c.1 mfi(rteam)leaks whh 50 to 100 cm {lowjrecrmilgtoo pump seal)

break with effective diameter > 4- inches

break with equivalent size 1/2 u> 4 inchesinadvenxnt PORV or safety valve openingRCP seal failurecontrol rod drive seal Icalugetube rupture widi teak greater than 100 Rpmbreak with equivalent size >6 incheabreak with equivalent sue 2-6 mcoexbreak with equivalent size .5-2 inches

break with equivalent size >4 inchesbreak with equivalent stu 1-4 inchesbeak. with eotiivakM sac <l iacb

break with eqiuvakot in* > 6 inchesbreak with equivalent size 2SX.5 at (lip.) and 4.7-6 b (swam)break with equivalent size 2-6 mchesbreak with equivalent sit* 0.6-2.6 in (lu?) and 1 0-t.TiA (stem)break with equivalent siu 1/2-2 iachei

breofc with equivalent ate *43 incfae*break with wurviboi stze 1.9 WJ 4.3 ijichtsbreak with equivalent size J to 1.9 mchesreactor coolact pump seal ntptorebreak with eocivilau «m J u> 4 J sq ft liquid sucrico sidebreak with equivalent area J to 4.3 sq ft liquid discharge sidebreak with couivaleni area 1.4 to 4. 1 sq ft sttuabreak wid) equivalent KLC .11 to 3 SQ ft liquidbreak wids equivaket area .12 lo 1.4 sq ft steambreak with equivalent area <, 12 sq ft

Break with equivalent Cow >2000 kt>Break with equivalent flow 30 lo 2000 kg/ibreak with eoijiva leu flow 10 to 3O Ict/i

break with ecruivalcnl Me* :>45O «) cmbreak with ecrutvalcnt area 80 u> 450 sq cmbreak with eouivalenl are« < SO so cm

VKKQfOtSEfmm^-fmean evemn-.yr.mean eveniA-.vrmem eventfr.vr.mean evepi^.vr.mean e*em/r.vr.mean evewA'.vr.mean evemAp.yr.meaa evenln-.yr.mean event^r.yr.

mem even/r .yr.

mean evem/r.vr.mean eveiaA.VT.mean eveotrV.vr.mean evewA.vr.meaa event/r.vr.mean evonMvr.mean evenMr.yr.roKio ntajt.fr.men ev*nWr yr .meaa eveetfr.yr.meaa eveu .yr.mna eventfr.yr.meaa cvcntA'.yr.mean evem/r.yr.raetn eveiaVt.yr.mean tveot/t.yr.mean eveni .yr.otan eweaMr.yr.mean eveal/r.yr-meao cveflt^.yr.mean cventA.yr.mean eventA*.Tr.mcsii eveot/r.yr.BKffi cvootA.rr.mean eventfr.yr.mem eventfc.vr.mBita evcxtt^r.vr.mean eventA.vr.meta tvadjt.yr.

mean event*.yr.mean event/r.yr.mean eveatA-.yr.meaa evem/r.yr.mean evertUr.yr.mesa cvtBi/r.yr.point est event/ir.yr.poise est. tffosft.yT,paint est. enaji.yi.poiac «SL eveBt/r.yr.point tsL event/r.yr.medtu, evemrrcacuryr.median, ev«u/re*cioryr.median, eventsAcacwr YImedian, eveatsAvactor nmedian, eventt/reactor yimedi»fl, ovce4»A»actpc VImedian, evcmfteactoryr.mean evsitfr.yr.meaa evcoUr.yr.mean evenUr.yr.

mean eveoA.yr.mean event>r.yr.mean eventAr.yr.meaa eveu/r.yr.meta evem^r.yr.meaa evenUr.yr.mean evenu/ryr.mean events/r.yr.mean eveois/r.n.mean eventstr.yr.mean events/r.yr.men evtnt/r.yr.mean event/r.vr..•nean eVMBA-.yr.racafi cvtffcrc.yr..-n««n evatfi.YT.nean «iventfr.yr.mean eveni/rrv.

^REQj:Z7E-4/vr8.0EJM2,7E- Tt.3&3^^7JrV5NrIZE-Stvr1.6E4rvr3.8&4\rllE-4/yr

ZDtyyr

4.0E-4>yrZOE-3fVrl-OE-2/yr5.0E-4fvr1.0&4rvr

l.OE-3/vr3.0r>4/yrJ£&3iT,lOBJ/yr2JDE-2/yr3J3E.2/yr5.064/yflJDE-3/yrlJDB-3fVrXOE-2/yrl.OE-6/rr2.7&4Wyr8.06-tryr2.7E-3rrrZ7E-2ft»1J3&«/YI3J)&4/*rg-OE-ifirr3.0B-3rtlIDGrVyrl^&l/rr9.3B-»/vrl.lE-6/yrlAE-Xtf

8.6E-3AT4.7E-5/yr9J&^/yrl.SB-3/yr4.6E-6/yrlXH>7flir7J»4/yr3X)B-3/yr&0&3/TT3.0E-7/nliB-7/yrl.OE-7/yrl.OEU/yr3.06J/yr3.06-WyrlJDS-3/yrUOB-3Arr4JJ&«fyr23&4ftiZ4&4ftr2.1bV2^yr

9.9&SryT3.9B-5/yrS_2E-S/»r9.0&Vyr2.1E-*/yrI.OE-3/yr10&4/»r4.6lU/yr4X)E-3ftr2.3E-2AyrME-3/yr3.0&4/yr9.0E-4ryr3XiE-3/yr2.7E-7Nrrl.O&Wvr3.0E4/Wl.OE-3/vr

...AANCEVAR iFJfJGFlOEF10FJ=6nAn/aoAnAD/l

BA

nAnAn/aEF3EF3EF3EF3EF3EF3EF3EF3nAnAn/inAnAnAnAnAnAn/anAnAnAnAnA

95*. i*9S«.5%-a,

954,5%nAnA

nAnA

EF10EF3EF3

EF10nA

90* lOKDOnatl di«r90% fasDorraal distr90%lojBiomialdat.90%to*nonnald«r90% Jognormal dim-90% lognormal disir

EF10nAnAnA

nAnAnAnAnAnA

95%: 5% of distribution9Sft,J«°fiii*noBt«fl95%.5WdtHnbtitwo95%, 5% of distribution95%. 5* of duttibuuca

nAnAnA.iAr.ii•tfl

-'.

jffEBBrtt): Low«tBN»iotiHCKn/an/afjtn/aa/inAan/tn/an/a

a/I

a/aiAa/aiAiAiAn/an/anAiAaAiAn/an/an/an/aa/anA-nAnAnAnAnAa/anAnA

lSE-3/yr4.1&6Ar

"*17E-I/yr

nAnAnAn/anAa/»a/ia/anAnA

1-OE-cVyrUOE-y^rlOE-3/rr30E-3ftrIOE-2/yrl.OE-1/yr

nAnAnA

nAn/anAnAa/aaA

5iE-d/yr1.2E-3AT14E-2/WJOE-2/yrZlE-2/vr

aAnAn/an/in/anA1/1

aAnAnAnAaAnAnAnAnA

DA

iAnAnAnAnAnAaAaAaAnAnAaAnAaAnAiAnAaAnAnAaAnAnAnAnAnA

9.4E-7AT6.0E-X/yr

"*ioBJfrr

nAnAnAnAnAnAnAnAnAnA

tO&S/yrIJDK-Styr3.0E-S/yrSuOE-5/yr1B&4/VT

ljO&4/yr

nAnAnA

nAnAnAnAnAnA

76E-6/yr2JE-S/yri.2E4/yrJJE-3/yr3.1&4/yr

nAnAnAnAnAnAnA

German Risk Study, table Fl . 4-5German Risk Study table Fl . 4-5German Risk Study, table Fl , 4-5German Risk Study, table Fl, 4-5IREP-ANOt able 4-7IREP-ANO1 table 4-7IREP.ANOIttcJe4-7IREP-ANO1 table 4-7IREP-ANOIUbte4-7

lREP-ANOIuW«4-7

Lua me PRA. uble A.I 6Umcnc PRA, table A.1 6UaencPRA.tab)eA.16NUREG/CR 4550, VoLl,Men>odotogy_ a le Vffl 1-1NUREG/CR 4550, VoU,Methodoto«v_ ta le Vffl 1-1romEGYCR 4550, VoLl, Methodology., la leVmi-lNUREG/CR 4550, VoU, Methodology- a le vm 1-1NUREG/CR 4550, VcU,Meth*ioIotv_ »•!» VEI.1-1ramEG/CR 4550, Veil. Methodology- ta leVQI.l-lNUREG/CR 4550. VoLl.Methodotoo- a leVdLl-lNIFREOCK4550, VoLl.MethodoIofy- ta'toVEU-1NUREG/CR 4550, VoU,S«ry Una 1_ akUIV.3-1NUREG/CR 4550, VoU, Sutry Unit 1 _ at 1 5 IV.3-1NUREG/CR 4550, VoU.Smy Unit !„, it 1 j rV.3-1NUREG/CR. 4550, VoU, Suny Unit I... aklsIVJ-INUREG/CR 4550. VoU. Sony Unit 1_ atliIVJ-1NURECrCR^SSO^VoU, Peach Bottom.., i Me IV.3-1NUREG/CR4559, VoU, Peach Botton. , 1 1 W« tV^lNUREC/CR-4550, VoU, Peach Bottxn. ., i We 1 V.3-1NUREG/CR45S), VoU, Peach Bottom. ^11 W«IV3-1NUREG/CR.-4S5), VoU, Peach Bottom. , 1 1 bte IV3-1NUREC/CR-4550, VoUS, Gotnd Gutf !_ a te FV.MNURECrCR-4JS>,VoUS,GtaadGutf i_ a 4. IV.3-1NURECVCR-4550, VoLS, Grand Golf !_ u teIV3-lNUREGrCR-4559, VoLS, Grand Golf 1_ ti to IV3-1NURECrOUSSD, VoLS, Grtod Golf 1_ tj le IV3-1OCOOM PRA. table 59OcooeePRA.tabie5.9Oconee PRA. table S.9

OGonecPRA,ubleS9Seqowab NPP RSSMAP. taWe 7 JSeoQoyah NPP RSSMAP. tabte 7-4Sequovah NPP RSSMAP, tacU 7-4Seceovah NPP RSSMAP, table 7-4Seqnovah NPP RSSMAP, tabU 7-4Shorehaa PRA, Appendix A.1ShcreiBn PRA, ApoendtxA.1Shcreh«a PRA, Appendix A.1Sborehm PRA, Appendix A.1ShcreJiao PRA, Appendix A.IWASH 1400L Reactor Siter Scary , App id ; i V, cbapur 4 5WASH- 1400. Reactor 5«fetr Sudr. Appi alit 01 Uote m 6-9WASH-1400, Reactor Safety Study, App. at it IE, ubl« m 6-9WASH-140a Reactor Safety Stttdy, App ttdn HI table m 6-9WASH 1400. ReactorSafetyStody.Appi ad t ffl. table ffl 6-9WASH- 1400. Reactor Safety Study. App at i « m, table m 6-9WASH- 1400. Rercsor Safety Stndy, Appi nd i » V, chapter 4 4CarratCuffii Ua»t 1 DtEP,tabU42Calven Caffs Unit 1 IREP, table 42Calvm Cttffii Ua» 1 KEP, table 42

Brown* Ferry Unit 1 IREP, tabta 5Brown* Ferry Umt 1 KEP, tabk 5Brown* Ferry Umt 1 IREP. table 5Brown* Ferry Umt 1 IREP. table 5Brown* Ferry Umt 1 IRBP. able 5Browns Ferry Unit 1 IREP. table 5OldPWROldPWROldPWROldPWROldPWRBanebadcl&lNPPBanebactl 42NPPButebact 1 A 1 NPPForsma*3NrTFbrsmaric 3 NPPForsmar* 3 NPPFonmaric 3 NPP

KtTSOllRCE... ......... . ,, "generic dau source*-enenc dila source*feune dau sourceswenanff exDeneaee and enemeenn^ aaalvsuRSS d*» eombned with ANO-I break range*RSS dau combated with AN'O-1 break raniiesRSS dau combtaed wjft ANO-! break raaw*RSS dau combtaed with ANO-1 break rangesRSS data cotabwd with AND- 1 break range*

RSS dau coolnocd with AiNO- 1 break ranges

enniaeennt; evaluauon and different daa sourcesenmneennit evaluation and different data souredengaieennjt evaluation and different data souredaverage of oast PRA*RSS valueaverate o past PRA*assoed from unclear, industrial and other dan sourcesaverage of pwt PRA*

NPP operatmK expeneocerenew of past PRA*Based on NURBG/CR -4550 Vol. 1Based on NURBGCR - 4SSO Vol 1Bastd on NUREG/CR- 4550 Vol 1Based on NUREG/CR-4SSO Vol 1RSS dauput PRA* and other source*past PRAa and other sourceipast PRA* and other source*pan PRA* and other source*analysis of the system interface* tamj generic Mure dau

Analyst* of the system interfaces utmg generic failure dataacaenc poor updated with plant specific operating experiencevarutu source*update cf fenenc prior

jcenenc dau updated wuh plain soeoftc ooenung cxpcnenceengiwermt evataatioa niamenenc dataengeeenni evaktatioauskn i-enenc daaemtneermt evaktatm vtmt generic daucntnwXTmg evaauoonosmg generic daugenetic dauceaenc dau b«*ed OB coDecuon of operating expeneocetgeoenc data bawd on actual reactor operating npenencegeneric dau bajed on reactor operating experiencegeaenc dau source*system evahiatwnbaMd on noo-*uc!e«r cqjerteccebased on number of Mek«rja<iujtnal «ctd «4ber dau minx*uacneat based en atdear, tadmtnal and other data Morceaasteanem based on nuclear industrial and other dau KoreaaneseBett based oa nuclear, industrial and other data mace*asseaed based oa tmdear industrial and other data sources•uryu of th« aynean interface o*tnj getienc faUnre me*geaencdaugenencdaugenencdau

geaencdatageaenedaugeneric daugeaenedaugteenedaugenencdauGeaene dau updated w*h plant speafie operating experienceCeaenc dau updauil with plaai cxcific ooenutf excenewtGenenc dau updated with plant sxcfic coeratmg excentnccGenenc dau updated with plant specific operating exoeneoceGeneric dau updated with plant sceafic coenung exoenenceS tame source*literature source*literature diuImmure sonrcciIneranire source*loemure sources'ilerature sources 119

&g;gs]fri&r **^*;sRSSi3femf7930SI528384858687888990919293949S96979899100

101

102

10310*105106107ice109110111112113114.US116117US119120

121

12Z

123

124125126

120

BWRBWRBWRBWRBWRBWRBWRBWRB\rnBWTtBWRBWRBWRBWRBWRPWRPWRPWRPWRPWRPWRPWR

PWR

PWR

PWRPWRPWRPWRPWRPWRBWRBWRBWRBWRBWRBWRBWRBWRPWRPWRPWR

PWR

PWR

BWR

BWR

BWRBWRRWR

LOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCA

LGCA

LOCA

LOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCATi in ill attnnsiealtnnskaE

tnesicm

tnniinl

trmsjeattnasiesttrmsiear.

large LOCAmedium LOCAsmall LOCAlatersvstem LOCAlarreLOCAmedium LOCAscull LOCAlane* LOCAmedium LOCAsmall LOCAlares LOCAmedium LOCAsmall LOCAIntersystem LOCApressure vessel rupturelarge LOCAmedium LOCAsmall LOCAreactor vessel ruptureuuersystera LOCAHf ani generator tube rupturetote LOCA

median LOCA

small LOCA

steam generator tube rupturesmall LOCAsteam generator tube rupturemedium LOCAtug* LOCAmtenystcin LOCAttncc LOCA inside dryweUbrae LOCA outside drywetlmedium LOCAsnail LOCAmtersystem LOCA-LPCI bretkintenysuxn LOCA-CS breakECCS breaksreactor vessel ruptureloss of fcedwater(FW)loss of off rite powerf.LOOP)loss of off site power

odMr (tun toss of PCS (tucbuM trip)

aohtioB ifreacux-iron ooodcnstr

turbine trip

madveront open relief valveloss of feedwiierloss of of frit* power

break with eauivalent flow 600 to 2000 ks/sbretk with equivalent flow 35 to <00k?/sbretk with eouivalent flow <3J kc/s

break with etraivalenl now > 2000 kg/sbretk with equivalent flow 30 to 2000 kg/sbreak with eouivalent flow 10 to 30 kt/jbretk with eqaivshou flow >30QO kjt/t

break with eomrnkm flow >]200 kjr/tbretk with eouhmleni flow 35 to 1200 kg/sbretk with equivalent flow 5 to 35 kgh

break with eauivalent diameter zreiier thin 15 onbreak with equivalent diameter between 5 nad 15 cm

Break wait equivalent diameter pester man 6 inchesRPV failureBreak win eqirnlant SIZM teawea 2 tod 6 iacaecMuWrte pressomer safety and icncf valve failureBreak win equiviltnt diiaeier «***it»«- than 2 'm^^fPresstkizer safety and tel ief valve failureCRDMftUuroRCP seal failure

break siu greater thai J MI ftbreak size greater than J tq ftbreak sixes between .1 and 3 so. ftfacanOxnd JX>4 and ^Oio.uid)break size op to .Isq ft (steam) or .00) «q ft (liquid)

component ftihires in FW systemkm of vottare on more titan 1 power bus

Teal lots «TFW flowFull cr psrtisl dosure of MSIV<1 loop)QcnraoftllMCV-tIncrease b FW flowFW flow iastibi% (operator error)

Loss of cogdensate pomps (aS)Loss of condenser vacuum (total)Opeaitt( of steam rdisf valvesLots of drculMinj water (CW)La« of RCP flow <1 loop)Uscootrolled rod witboVawiiCRDM problaBS and/or rod dnoHifs cr tow £x jv prcuun

Full crpanal dosure of MSIVTurbina trip, throtto valve closure, EHC problemsGenerator trip and (eae/aior caused problemsSpurious auto tripManual trip due to filse signalsQosureofallMSIVTurbine trip without bypassLot* of condenserPartial closure of MSIVTurbine trip with bypassSunup of idle recirc.toopPttsmre regulation failureuudvertant opening of bypassRod withdrawalDismrbtnce of FWElectric load rejection

LoaJ loss of reedwater

mean evenur.vrmean cvenUr.vrmean cventA-.vrmean «\«nt/r.vr.mean event/r.yr.mean eventA-.yr.mean evenOr.vr.mean event/r.yr.mean event/r.yr.mean eventAr.yr.mean eventA-.yr.mean eventjSr.yr.mean event/r.yr.mean evem^r.yr.mean event/r.yr.mean eventA.vr.mean event/r.yr.mean evenW.yr.mean eventA.yr,mean event/r.yr.mean event/r.yr.mean events/r.yr.

me«n event/r.yr.

mean eventrr.yr.

mean «vettc/r.yr.mean evcnt/r.yr.mean event/r.yr.mean cventAr.yr.mean tvtot/r.yr.mean tvent/r.yr.mean, event/r.yr.mean eventA'.yr.meaa eventA'.yr.mean event/r.yr.mean evmt/r.yr.mean eveat/r.yr.mean eventrr.yr.mean eventA'.yr.mean event/r.yr.reeaft event/r.yr.mean eventA'.yr.mean evcotft'.yr.

meaa cvoa/r.yi.

meaa evcd/r.yr.

mean evemyr.yr.

mesa evem/r.yr..•nean evest/r.yr.niens evart/r.yr.

«.*SFEEOJ.'I.OE-7Mt.lE-SfVrl.lE-I/vr1.0E-7 rl-OB-T./vrLOtVS/vrl.OE-2/yr1.0E-4M5.0&4rvr1.0&3ryr3.O&J/yr9.0E4ryr3.0E-3ryr!.9E-7ryr2,7B-7/yrd.OBJ/vrS.1&4/VT

!.lE-2rvr2.7E-7ryr4JlS*ryr9.7E-3/yr9"*64/yr

9^*r

IS&Vyr

2.4&7/vr3JB-2Vyr3.7B-2/yr9.4B-t/yr9.4E-4rVr<.&E-7rrr2.7E-Wyr].OE4/vr2.7E-3/yr2.7r>7Jyr7JB*ryr13E-7/yrl.JB-5/yr3J>B-7/yrg.OB-1/yr1.0B-l/yr3.2S-l/yr

l.lEtdryr

^.0&2/\T7.0& I /yr5JE-2/VT

^lUMSSfVAR ewERBNa LOWER B>r&SGBjMX' ..^...... ...nMnitnMnMnMnMBMBMBM

nMnMBMsMBMnM95*95*95*95%nM95*

95*.5*

96*. 3*

95*. 5*

95*. 5*nMaftnMaMnMnMnMnMnMaMnMnMDM

EF3EF3nMnM

n/a

DM

nM

nMsitaM

n/an/anMn/an/anMn/an/an/an/a&MnMa/an/an/a

UE-3/yr3.0E-3/5TX5E-2/JTl.OE-6/yr

nM10&-2/JT3.6E-3*

3-6E-3/ir

7.2E-2/yr

77E-2/JTnMnMnMnMnMnMnMnMnMnMn/an/an/an/an/inMnM

nM

n/a

n/a

I/a3M

-it

nMnMntanMnMnMnMnMnMnMnMnMnMnMnMnMnMnMnMnMDM

3J3E-S/yr

3JE-5/yr

t.3&2/yr

2^E-3/yrnMnMnMnMnMnMnMnMnMnMBM

nMnMnMnMnM

DM

nM

nMnMnM

Osknhactt I NPPOskarshamn 1 NPPOslorstumn 1 VPPOikantamn 1NPPQtkanhantn2NPPCMonhama 2 NPPOtbnbami)2NPPOskanhamn 3 NPPOdcarshams 3 NPPOskanfaamn 3 NPPRiatMh 1 NPPRmrfiali J NPPRinghalilNPPRin*hal* 1 NPPRinitnits 1 NPPRon (halt! NPPRiagflal«2 NPPRiB(*alj2NPPRinejiab2MPPRmajialt2NPPRinsb»li2NPP25od NPP PSS table 11.1-2

ZioB NPP PSS table 1 1.1-2

ZSooKP? PS5 able J U-2

2Jce NPP PSS tatto l.I.J-2Anura NPP PS A, Sanmtary reportAn pa NPP PS A, Summary reportAn«T» NPP PSA, Swmnary reportAagra NPP PS A. Sanaurf reportAnitra NPP PSA, Smmarr reponCacnoNPPPSSUblaA-2Caono NPP PSS table A-2Caorao NPP PSS table A-2Caono NPP PSS table A-2Cacno NPP PSS able A-2Cacno NPP PSS table A-2C»coo NPP PSS table A-2Caono NPP PSS Bbte 3JGerman Ri* Stady, table Fl, 4-5German Rbfc Study, table Fl, 4-5IREP-ANO1 table 4-71REP-ANO1 table 4-7

REP-ANO1 tabie4-7

UnWKPRA.ubtoXU

L4nencPRA. table A.U

LunencPRA Ut4eA.t3UmencPRA.tableA.t3LunencPRA UbleAU

UtT SOURCEItteanin uurctj (aophcauon of LBB cnuna)literature sourcetlitftwture wxircaIttenntre tourcettiteratmc sourcaloeranm aourcecliteraton source*liunnm aowcf*liter XtUTQ 9DQTCCS

IHeraluxc sourtesIttenTOnnourtMliterabire sourcaloenomiourcctIttenoi)* souredIiuratu»dJUalicencure svurcctliterature aoorcesliterature sourcesliterature sources

litearture source*

Uenora loorce* aad plant oporatin( experience

Ijwntum WUTCM aad ptol open 004 expeneoce

literature source* aad pUni optntian experience

literature source*

literature soorcex.,tfuf alure source*Uteratnre soorcet

Uec^ure sowrte*

btentora databteraom SDurcettiunam source*IfUntum source*gennan operating exoencacesystem aaahni*EPRI NP-801 and data from lie utilityEPRI NP-801 and ptwt ipcolic tola

EPRI NP-801 «nd ptux. ipeafic mfo

BWR operabnt expenenc*

BWR openung experience

BWR ooentmt exoencnceBWR operaao; exoeneoceregional ^nd data

121

127

128

129130

131

132

133

134

13S136

137

133

139

140141

142143144145

PtANTTYPE IB SETPWR traanenl

BWR transient

PWR 4 BWR traamotPWR Trantisnt

BWR tnamaeot

Bw*R transxd

PWR liaasnK

BWR trimm*

BWR tnama

PWH tTXaaatnfald

PWR Iniimal

PWR transarat

BWR transucBWR tnniitM

BWR tranoeMBWR iraoaea*HW.H LimalHV.R iranaamt

" TE. JCATEGOBYtoss of PCS

Iou of power conversion system

loss of a DC busother than loss of PCS

loss of FW (steam nde PCS available)

other than iocs of PCS

Ion of off site power

Iocs of off ma power

atadvertaat ooenmn of relief valvetnrknoa tnp with MFW avnUble

toss of mam feedwaaer

losa of charging pump cooling

loss of off nte power

Iocs of PCSother than to** of PCS

loss of off site powerloss of PCSloss of FW and steam side PCS availableother than tan of PCS

tNfTUTORSTotal loss of FW flow (luted at DESCRJidd)

Full or partial closure of MSIV(ooe loop)aocureofaliMSrV-*Loss of condenaate pomps (ill locos)Loss of condenser vacuumCoodmsfr Ink ageSG leakageOpening of steam relief valve

Loss of CWLoss of CCLonofSWElectric load rejection with bypaM failureTurbine tnp wtth bypass failureFull MStV closurePartial MSIV damnLost of condaucr vacuumPressure regulator fails openPressure regnlazar fails closedTurbme bypaaa &3s openTorbuw bypass faus dot*Tarbine bypass or control valve increase pressure

Loss of RCS ftow<OM loop); Total loss of RCS flowUncontroltd rod withdrawal^rVCS problems-baron dilutionCRDM mechttuoJ problenu or »d dropLalayt from: cootrot rod. pnmary system. [•••••""•'"•trf triMJZtc jjiujuia, low or mgh^ firmu iif i spray £athircInadvertant SI AffnalConUiniBcot pressnrBprcvlenSpun ous tnp; aato or manual tnpfno traosuat ccoddioni}

Startup of macnve RCS loopLost or redttcticn of FW fio»(aa* loop)Increase n FW Omrune locopul loops[ f^f nf i^trrfmntto ^f^rr^nf tooa aH loopsTizbiae cnp.Renerator mpnhroDle valve ciosureiEHC problemslots of all FW flowFW low flowTurbine tnp kctnc k>ad rejectionRecirculabon cootrot failure: dfcrming flow, rncrotunj flowRedrculatlon puotp m 111 one. allSiarojp rf idle ITI ni'TtrTTm pumpRtcireulacoa pomp senuraFW mcreasmg flow, lots of FW heaterTnp of one FW or condes*ate pompRod withdnrml^Budvertaot msemon of rodsInadverunt sortnp of HPCI/HPCSSpunous tnp na sutramentatiaaRPS&ttteSCRAM dne to otier plant occarancesMaoual SCRAM -oo transient conds&onSCRAM cao» BBkoownIocs of aD oflsue power!oss of power to neccessary plant systemloss of off stta powerlots of auxiliary pow« {traosf onner)

otrbm* tnp; reactor tnpSots of loadMSIVclonucIocs of turbine controlfailure of mam FWhigh SO water levelmxdvertant SI signalloss of charging pump CCkws of charging pomp SWfailure of oSsite power gndion of sutxm reserve powerloss of power to the switchyard

madvertant open relief vaive in pnenary systemiou of FW but steam side of PCS initially availableother initiators with same consooeoces

FMEOMESCmean evera/ryr

men eveot/ryr

mean event/ryrmeaa eveot/ryr

mean evaajryr

mean evenlA'yr

mean •vent/ryr

meaa eveot/ryr

mean event/ryrmean events yr

mean evtafryr

mean event/ryr

mean event/fyr

mean cventfr yrmean evenl/r yr

mean event/r yrmean event/r vrmean evenUr \rmean evenlA-vr

FRE<i14E-HYyr

16E+Cryr

30E-3/yr

56E.1/JT

10E-l/yr

IDE-l/yr

14E-l/yr73E+<Vyr

94E-l/yr

30E.2/yr

705-1/yr

15E+Cryr26E+Wyr

70E-2Ayr( 5E*(VvT70E-2/V54F*O/\r

122

KAKGSVABEF3

EFJ

EF3EF3

EF3

EF3

EF3

EF3

EF3n/a

n/a

n/i

n/a

n/an/a

n/an/a^«n/>

WPPEHBNQ tOWHlBS» SOfiltCK ." . . . . . . . . . . . . . . . . . . . . .~. . . . . . . . . . . . . .n/a n/a NUREG/CR 4550. VoLl. Methodology-, table VTH 1-1

a/a n/» mmEG/CR4S50.VoU.Mahodolojy_,ubl«Vnn-l

gya n/a NURHVCR 4550. VoLl. Methodology., able VTO 1-1n/a n/a NUREG/CR 4550. VoLl. Mahodoloty-, table Vffl 1-1

g/a n/a NURBG/CR 4550, VoLl, Mahodototy-., Hole Vm 1 1

n/a n/a m]RBOCR4550,VoLl,Maboaaloiy_.ubleVm.Ul

I/a n/a NUREG/CR 4550. VoLl, Methodology-, tabte Vffl 1-1

n/a n/a NURBG/Ol 4550, VoLl, Methodology table Vm.1-1

tit n/a NUREG/CR*SSO,VoLl,Me*o**)«y-,taM«Vin.l 1I/a n/a NUREG/Ol4550.VoL3,Surr7Unal_ubkIV.3-l

i/a n/a NURBG/CJl 4550. VoL3. Snnr Uo* 1 taWelV 3-1

n/« n& ^aJREG«^4S50.V»U.Syrry^tal_Ublc[V.3-l

a/a n/a NUREG/CR4550, VoU.Surr Umt l-.,ublt [V3-1

i/a 1/1 NUREG/CR-45SO. VoU, Ffeaca Booom. .able IV 3-1I/a nM NUREG/CH-t550.Vol4.PociBoBom_...ubltIV3-l

-.11 aft NUREC/CR-4550. Vol 4 Peaca Bonom—. able IV 3-txra nM NUREG/CR-«5SO. Vol 6. Grand Calf 1 ,ubb>IV31.„! n/a NUREG/CR-1550. VolA Cnad Gulf 1 .ublelVil-/a nh VUREG/CR-45SO. Vol 6. Grand Gulf 1 table IV 31

tlLT SOURCEBased on NURBG/CR 3862

Bued on NUREG/CR 3862

DC power ojpprjr anidy, NUREG 0666Baaed aa NURBG/CR 3861

Based co NURfcO/CR 3862

>*———— ———*-»»"«

PUW Bddmui inaly« NUREG 1032

Band on NUREG/CR 3S62plant ipeofic opemat experience

plant apeofic openung experience

Comparuon of NURBG/CR 3862 and openung experience

NUREG 1032

nuclear sources and plant openung exDenencenuclear sources and plant operating experience

nuclear sources and plant operating exoenencedifferent nuclear sources ud plant oocmxit. exoenenwdifferent nuclear sources and plant ooeraline; exoenencedifferent nuclear sources and plant uneraltnK exoenence

123

•B;-W-P146147148

149150

151

152153154155

156157158159160161162

163164165166167168

169170171

172

173

BWRBWRPWR

PWRPWR

PWR

PWRPWRPWRPWR

PWRPWRPWRPWRPWRPWRBWR

BWRBWRBWRBWRBWRBWR

BWRBWRBWRBWR

PWR

ES;IB!SBf£gJtransienttransiesttnmiest

trantieatransit*

iransiet.

trntnenttranxieettransie*traroksc

tnmrie*transienttraoie*transienttrantiesttransieMtrantiesA

trance*tnowettransienttramieattrajxsicattrtnsiett

trawkMtrtntteatransie*

transieafi

wms$i&e&W!®siij333;gtmSWMloos of off lice powerinadvertant ooenin^ of relief valve[caaorAurbine trip

spurious SI ritualstectn line break

lots of condenser

FW line break (large)loatofmainFWpartial lot* of main FWIMS of off site power

loo of off rite powtr {jobjnfion &»lta)excetttofeedwau*spurious tow presranzer risiallots of off site powerlots of main FWtransient with main FW availablelots of condenser

MSIVdotnrelotsoffeedwaterlots of off site powerinadvenast opening of rdief valvecofltfoi rod wtiodrawai

aam tteam toe breakfoedwater line breakHPdrROC line breaktitpid shutdown

rapid shutdown

^SiOTstitOH^s^^^^^^^^t;^^^;^^^

Rod dropInadvenant rod withdrawalInadvenant boration or dilutionReactor tripCold water additionInadequate main FWTurbine and control valve malfunctionPlWBStttr spray fainnTurbmetripGenerator faultsdid distrmtneeesAdmnistritivfi shutdownsHP! fk>wsie&m Uneittptaretmbine bypass valve inadvenant openinKlott of condenser vtcuumkm of condenser circulating water

faikreof |ridfnktre of feeders

lott of normal coodenter vacuumturbine trip with bypan valve failureelectric load rejection with bypass vatve faihire

Ekctric load rejectionTurbine tripInadvertant dotnre of one MSIV; partial MSIV clocurefteuujc regulator fiilK open; dotedTurbine bypass fails openTubine bypass of cootrol valve catisc increase uiusure

Trip of ncircalation pumpc one; allAhnonnal startup of uQa reciroalMioo pumpReciicnlation pump seimreLOB of FW beuerTrip of one FWor coodensate pump

Rod withdrawal at powerFW conlroler failure; lots of FW flowReciTctUatiofl flow cuotroi laiktrerdecreasingiAcre&singStartup of idl« rectrculation pumpLots of FW hearingInadrenant HPO startLoa of auxiliary powerTurbine tnp<turbine valve clotoreXload rtjeokx^uop valve cl)MSIV closureRecirculation pump trip (one pump)RecucuaUion pump aeizitrerone pump; two punpcT-O pressure regulator failure-rapid openingRed ejection; rod drop accidentStartup of id)« recutx.pump with simultaneous turbine tripTurbine trip; lots of main generator; LOOP

Inadvertant MSIV closureLots of main FW; lost of main CW; lots of condensate pumpsInntvertint openig of SG PORV-sIncrease in MFW flow .Open ing of all bypass valveXsteam dump)Unconiroled rod withdrawal; control asxinbiy dropBoron dilution (CVCS malfunction)Stamp of inactive RCS loopOpening of preHUriKsr SRV or R VLoss of RCS coolant How; siezure of all RCP-sRupture of FW piping or main steam lines; rupture ot SGRuclure of CRDM housing

?;;S5ER«OSii£5SfCjg:si»|;KSmean evemfr.vr.mean eventfr.vr.mean event/r.yr.

mean evenWr.yr.mean event/r.yr.

mean event/r.yr.

mean event/r.yr.mean eveni/r.yr.mean event* .yr.mean evcm/r.yr.

mean evem/r.yr.mean event/r.yr.mean evem/r.yr.mean evem/r.yr.mem event/r.yr.mean evenut.yr.mean evem/r.yr.

mean event/r.yr.mean eventA-.yr.mean event/r.yr.mean evem .yr.mean evestir.yr.

point esx. eveiu/r.yr.point etc event/r.yr.point est event/r.vr.median, event/reactor yr.

ma^t^7.0E-2Nt!.4E-Uvr

lDE-2/yrIOB.3

2.1&1/)T

93E4/yr6.4B.1AT6\9E-l/yr«JE.2Ar

UB-l/yr92&2/yr44E-2ATZOE-l/yr3De«<Vyr4-OE+tVyr41&Uyc

X4E-l/yr1.8E-lfrraOE-2/yr9.0E-2AI3.0&2/yr

3^&S/yr8.7E-»/yr1.4&S/yr

IJJ&fl/yr

median, events/reactor yi l.OE+1/yr

124

;...ftAJse»y*it.n/an/a

95%, 59,

95%, 5%95%, 5%

95%, 5% '

95%, 5%95*, 5*95%, S%95%, 5%

95%, 5%95%, 5%95%, 5%

n/an/an/a

EF3

EF3EF3EF3EF3n/a

EF3

n/an/tn/a

EF2

EF2

tJPPERBMB L&W£Jt: BMy littCRCE.'** -..-.:•;., . ,-• '••*.••. • . . . , , • • .- . ... <-.n/a n/t NUREG/CR-«SS!>. Vote. Grand Gulf 1 , t bto IV.3-1n/t n/a NUREG/CR-45SO. Vol.6, Grand Gulf 1 , ta le IV 3-1

5.7E+0/yr JlE-tOir Occoee PRA. table 5 9

4JE-2/yr 78E-6Vyr Oconee PRA, table 5 9L2E-2/yr 10&«ryr Ocunee PRA, table 5.9

3£E+0/yr SJE-2/yr Oconee PRA. table 5.938E-]/yr BNL Review, taMe4JZ8E-3/yr 6.9E-7/yr Oconee PRA, taMeJ.99iE-l/yr 36E-l/rr Oconee PRA, table 5.99.7E-1M 4.0E-1/1T Ocooee PRA. table 5.99.7E-2/yr 7.1E-3/yr OCCOM PRA, table 5.9

3-OE-l/yr Z2E-2/VT Oconee PRA, table 5.9llE-l/vr l^B-2/rr Oconee PRA, table 5 917E-l/vr 1.7E-3/yr Ocoow PRA, table 5.9

n/a n/J Seoooyth NPP RSSMAP, table 7-4n/a at* Sequoyah NPP RSSMAP, table 7-4n/a n/a Seqnorah NPP RSSMAP, table 7-4n/a n/a SoonnamPRA.AppendixA.tableA.l-'

n/a n/t Sboreham PRA, Appetdix A,taWe A.1-'n/a n/« Sborehara PRA, Appendix A, table A 1-n/t n/a Sboreham PRA, Appendix A, table A 1-n/t nA Snorenao PRA. Appendix A. table A 1-n/t n/> Sboreham PRA, Appendix A. Uble A. 1-n/t n/a Shorenaa FRA.AppendixA.tabteA.l-

n/t n/a Sboreham PRA, Appendix A.1n/t n/t Sboreham PRA, Appendix A.In/t n/t Shoreham PRA, Appendix A.1

WASH-1400, Reactor Safety Study, Apt tn « V, chapter 43A Appendot I, Table 1 - 4.12

WASH-1400. Reactor Safety Study, API er tx V, chapter 4 34 Appendix I, Table I - 4 9

ULT soun.ce. ' ,. '- ~, -- , .„ ', ;assesraent of different nuclear sourcesdifferent nuclear sources and olant operating exoenencegeneric data updated with plant specific operating experience

generic data updated with plant specific openung experiencegeneric data updated wrth plant specific operaung experience

generic data updated with plant specific openung experiencerecalculation made by the reviewers •generic data updated with plant specific openunj experiencegeneric data updated with pUia specific operatmK experiencegeneric data updated wadi plant specific operauag experiencegeneric data updated w«h plant specific operaunf experience

generic data updated w«h plant ipeafic operating experiencegeneric data updated with plant specific operauig exepeneocesystem analysis

generic data based on nuclear operating expetrencetaken from EPRI NP-W1

generic data based on nuclear operating expetrence etc.generic data based on nuclear operating expetrence etc.utility specific datageneric data and eagmeenag enbuomgeneric data based on nuclear operating exoetrence etc.generic ^*fi based on nuclear operating expetrence etc.

engmeuujj. evaluationengineering evaluationengmeennx evaluationNPP operating mpntm&t engmeeraiB ffrrmatt

NPP operating experience, enguieenng estimate

125

' f TptA«S*aT TYI*E IK Sti" *" Us- CftltEGORV "* Gflf JiAjORS ]". . ^, A . ... . f .174175

176

177

178

179

180181182183184185186187its189190191192193194195196197198199200201202203204205206207208209210211212213214215216217218219220221222223224225

PWRPWR

PWR

PWR

BWR

BWR

PWRPWRPWRPWRPWRPWRPWRPWRPWRPWRPWRPWRPWRPWRBWRBWRBWRBWRBWRBWRBWRBWRBWRBWRBWRBWRBWRBWRBWRBWRBWRBWRBWRBWRBWRBWRBWRBWRBWRBWRPWRPWRPWRPWRPWRPWR

triMmt loss of off sue oowertrsnsanc tool intcrupuon of PCS

miiimt transient requinat RCS pressure relief

T«aJ Iocs of FW flowClosure of all MSIVsFW Oow nsubuiryLoss of all coodeestie pumpsLoss of coadeaaer vacuumLots of CWBjrbatetnp or tkrottk valve doture(caemor tnp or tawnter caused EMihstat of power K> aecceoary plant systems

tnaaat transient which do not affect from line sysierns Loss of RCS flow (OB* loop); total loss of RCS flowCRDM problem*, rod drop

tmac* loss of PCS

•tnastoK otberthanloss of PCS

trmneat turbine trtpUantmt loss of main FW (total loss)uautitBt toss of main FW (partial)Uinum excessive FW flowlunnut tost of condenser vacuumtnormt closure of one MSIYcransJeM closure of all MSIVscransmt core power yycufsujptrsnwnjii tit**" line brealt (smriit mmtiwnent)CI'USMH ]Wirri une î'p**f (outsuje contaoflient/Ii'inum main stem relief valve openmicstransaott hadvertanl SIn nniiinl tool lose of reactor coolant flowUanskat tot* of off-tun powertrantaeal transient with front In* systems availabletranstttf loss w comleusertfinnm loss of FW (total loss)tranonat loss of FW and condensertaaaat loss of off-site powertfanum tniosteat with front lae systems availablecraatMal loss of coodenaerrnasm loss of FWtnnskat lossof FW and condenserimutai loss of off-site powertnaMC rraosjentwith front IBM systems availabletrmoBBt lotsofcoddenjertransiot looof FW <nd coodenaertrmtirai lossof off-site poweruaniaeal transient with front tine systems avatbbkuanmei loss of condenseriranneM loss of FWirantxa lost of off-site powertrustctt tracnert with front Ime systems availabletransiat lossof coodensertraaneK lossof FW and coodensertransK« loss of off site powertnoswB transient with front line svstems availabletranswat lossof condensertrartnaat loss of FW A condenseriransMat lossof off si te poweriraraiot transient with PCS isolationtransm transient with front line svsrasavailabktransicB loss of off site powerirxnsiot transient reouinn^ RCS pressure relieftranswal inmdvertant safety miecuonmnsteai. neam line break

eVCS malfunction- boron dihmoa

Lots or redaction ai main FW(l loop)FttH or partial dosaw of one MSJVbcreas* in FW flow: one loop; all loopsLoas of ceadtBsaM pntnp

Sudden cpeam of relief vafvt

Spanou tmr Manul tnojsnai tnp-No transient coaditioasMSIV closure

Preuare regulator fafls openLcoorieedwMeraowLoa of off sue powerLoss of auoEarv powerbcreased flow at powerEkctnc load rejeeoonElectric load rejectwa wtfh bypass failureTurbine mpTurbine tnp with bypass fiulureInidvcnau opeoica; of MStVPressure rejoaoor fails closedBrpus/conirol valve causing pressure increaseRecuruUnoo control fails aaant uxrtued flow

. . fRKQ DESCmean evenifrvrmean evtnt/ryr

mean evenlAyr

mean evejftSryr

mean event/ryr

mean evcnlAyr.

mean events/r yr.mean events/ryrmean eveats/ryrmean eventirVyrmean eventsfryr.mean cvcntsfryrmean events/ryrmean eventsfryrmean eveoWryr.mean events/ryr.mean events/r.yr.mean eventi/ryr.mean evcntxrryrmean evenufryr.mean event/r yrmean evenl/ryrmean event/r.yrmesa even(/ryr.mem eveat/ryi.mean eveotA-yr.mem eveaifr.vr.mean event/ryrmem event/ryrmean eventAyrmean event/ryrmean evcnVrvr.mean event/r yrmean eventAyrmean eventAyrmean event/ryrmean event/ryrmean eventAyr.mean evenlAvrmean eventAyrmean eventAyrmean eventAyrmean eventA.yr.mean eventAyrmean eventAyrmean eventA vrmean eveniA yrmean evefttA.vrmean eventA vr-nean evenlAvrmean cvenlA vrmean evcnlA vr

FREQ1 JE-lAr

l9E^Vrr

t7E+<Vyr

1.7E*<Vyr

10&OM13E-IAyr4.9E-l/yrilE-l/yrS.lEr2M7JE-2/yr2DB-3/VTl-SB-2/yr4.<E-4/vr4J2E-1V168-2/rrS3B-2/vr6i9&2/yr49B-2/yr2JD6+Wyr3JE-l/yr1.6K-l/yr3-3E-l/vtISE-l/vrITE-l/yr2JB«<VyrITE-l/yrITE-l/wISE-l/vr1.9G»Wvr12E-)/vr12E-l/yr3J2E-!/yr16E«<VvT7Ê-1^T7Ê-l/vr2.0E-2/yr9S6-I/VTUB-I/yr9JKB-l/yr1.6E-1/VT9iE-l/vr2-OE-»«yr3.7E.l/yrl8E-lryrUE-l/yr60€+OSr70E-1M4 OE-lArt 2E*(V»T44E-4/vr

126

sijJSiiJi GE;: ^n/an/a

n/a

n/i

aft

of*

95*: 5* of dbtribwion95%; 5% of dbtribntioo9S«;S* of distribution95*; 5% of d bmbutioo9S*;S*ofdisoib«i«i95%; 5% of distribution.95%: 5* of distribution95%; 5% of distribution95%;S*ofdinribauoo95%; 5% of (ttaribuioa95%; 5% at distribution.95%;S*trfdi«riboooo95%; 5% of distribution95%; 5% of dasributiaa

aftn/l

n/iaimn/in/aR/10/1

n/in/aa/an/in/aeftn/in/an/ia/an/in/an/aa/ioftn/aa/ao/»

95%95*95*95*95%95*

n/an/a

n/a

0/1

n/a

n/a

1.4EXVyr4.9E-l/yr6iSE-Wyr4.9E-t/yrl.SE-1/yr

1-SB-l/yr5.7E-3/JT3.6E-2/yr1.2E-3/yr13E-2ftr4»E-2/yrL2B-l/yr1.5E-l/yrl-lE-I/yr

n/an/an/an/an/in/i0/1n/in/in/an/an/an/>n/in/in/i0/1n/ln/in/in/i0/1n/in/tn/in/l

4.0E-l/yr9.6E*fl/vr15EtO/vrS.6E-l/vr2.0E+(tfvr1.2B.3/vr

nAn/i

0/1

n/i

n/l

n/l

7.1E-l^r1.7E-t/yriSE-l/^t1.4E-I/yrlÊ-2/yrlÊ-2/rr3.9E-S/yr6.7&4/1T13E-5/yrI.1E4/J-T5.7EJ/yrlAE-2/yrlÊ-2/yr7.1E-3/1T

n/inAmiD/l

n/in/in/in/in/an/ln/an/a0/1

n/in/in/in/in/i0*

n/ln/in/i0/1

0/1n/in/ln/in/in/in/in/an/i

CalvenCIiffe Unit ! IREP. table 43divert Clifb Unit 1 IREP, table 4J8

Catvcrt CliHi Unit 1 IREP, table 4 a

divert Clifls Unit 1 IREP, table43

Brown* Ferry Uoit 1 IREP, table S

Brawns Ferry Unit 1 IREP, able 6

OMPWROtdPWROldPWROWPWROMPWROldPWROldPWROWPWROldPWROUPWROldPWROldPWROldPWROldPWRBuneoKklNPPBuwbuk: 1 NPPBvnebKfclNPPBanrfwdclNPPBanebKfclNPPOtkanhinm I NPPOttanhimn 1 NPPOdanfaanin 1 NPPOdanhamn 1 NPPOdonhaout 1NPPOstanJiirmn 2 NPPOdc«nbamn2NPP0»lcartB*nul2NPPOckmfaamn 2 NPPOtkxnhamn 3 NPPOjfctrshunn 3 NPPOtkanhanm 3 NPPOikanhamn 3 NPPB*nebKk2NPPBanebwk2NPPBanebackZNPPB««b»cfc24NPPRinghals 1 NPPRinghali 1 NPPRlnidtali 1 NPPRinjnaU 1 NPPKingBiti 2 NPPRtofShalj 2 NPPRinghiOs 2 NPPRunhali 2 NPPRin?hals2NPPRinRhils2NPP

EPRINP-2230EPRINP-223O

EPRINP-2230

EPRINP-2230

plant specific daU

piutt specific data

Geoeric data updated wih plant nxcific opcnlint experienceGeneric dtu updated w»h pUu setafic ooentinic tjtperieoctGeneric data updated wSa plaat specific opening experienceGen eric data updated with plant specific operatinit experienceGeoeric data updated wit p|a« tpectfic opentmg experienuGeneric data updated with platt specific operatiaR experienceGeneric data updated with plant specific opentiaE experienceGeneric d«u updated with plant ipecific operacioit experienceGeneric data updated with plant specific operatiDR experienceGeneric data updated witn plant tpecific opoatimt axperienc*Gen eric data updated with plaat specific opemint experienceGeneric data updated with plant fpeeifie operating experienceGeneric data updated with plant specific operatinc experienceGeneric data updated with plant specific operating experienceplant oc«nitiDt experience (&. 11 rean)potal specific operating experience {8.1 1 yean)plant specific operatinit experience (8.1 1 yean)ptuc specific eventing experience (8.1 1 rem)plant specific operatnt experiencepkBt opentmc cxperxw* O.7 yean)plant operating experienc* {5.7 yean)plant specific operation experience (5.7 yean)phat specific operatinit experienced? yean)plant specific operatinit experience (5 .7 yean)plant opentiniE experience (&2 yean)plant specific operating experience (63. ytar»Jpitat apecifie opemtinn experience (62 yean)plaat specific experienceplant specific operation experience <6.4 yean)plant specific opentmK experetnce (6.4 tern)plant specific operating experience (6.4 years)literature sourcesplant specific opentmg experienceplant specific operating experienceplant specific operating experienceplant specific operating experienceplant, specific operating exoerience (5.44 vein)plant specific operating experience (5.44 years)plant specific operaliaK experienceplant specific operating experience (5.44 van)plant specific operating experienceplant specific operating exoerienceplant soeafic operating exoerienceolant specirlc operating experienceDtant specific ODeratin^ experienceLiterature sources

127

1MLANT TtrK IK SOT- tKCXtEGOKY . ,r . . Csm*TOR!f ... . . . F3U5Q BESC FSEO226227228

229

230

231232233

234

235

236

237238239240241242

243

244

24524o

247

PWRPWRPWR

PWR

PWR

PWRPWRPWR

PWR

PWR

PWR

PWRPWRPWRPWRPWR

BWR

BWR

BWR

BWRBWR

BWR

trtnsjeat transient oocurnnt; after shutdowntraonot loaofRCSflowInnsieA loss of feedwater Dow

tnotxat partial loo of item flow

transKK torbmetnp

transient too of off rile powertransKat spunoos safety injectiontraocMBi reactor trrp

ii imifpi km ftf iTmin iwtmlT nmtaiRrn mt

™-

TjMincar Mum fT«*m lm% bf*JV

Mar** IMS of cfl-aj* powertEBflUm. JOM OC QUUS XCCQWMCT

tmMM TCMCforinptransMat mrtmemp

transient MSIVdacure

tranueat madvertantopemnftof relief valveuanswt Iocs of feedwaier

inraiea lota of of f sue power

meanmean

FW pipe rapture ouuide coatammeai meanLoss / reduction of FW flow m one SOLots of FW flow » all SGFW fk)w instability - operator error

Lora of on coodeotatt pumpLoss of all candeuate pomps

Other secondary leakageMSTV Closure meanFull closure of MSIVPartial closure of MSIVOther losm of steam flowaosureofallMSIV-s meaaIncrease of FW flow m one SGL«Q6X o/ coodcoscr vxcutmLoaofCWThrottle valve dostire/Elecirohydranlic control problemGenerator np / tenarator caused fmkiTurbw onp du« to over speedOther turbine trnM

meanmean

CRDM problem /rod drops meanHigh and low pretiunzer pretsureHifh prrnumrr levelSpunooc «iif»n»n« tnp - DO transient coodmooAutomaacAnaniat tro> - operator errorManna! tnp dnv to false ym*lfSpunous np - cause unknownPrimary jyttoa pretsure temperattire or power inbalaoceSteam pip* rupture innda containment meanFW pipe rupture toside r*TTTt*innmitStem relief or safety vihre open tnadvercwtty (leak upnreua of MSIV)Other Siena losses made contxmnent

Throttle valve/Etoorohydnalic control problemlSteara and dump valve fail openOther steam losse* outside containmentUncontrolled rod withdrawal meanBoron dihmorj - CVCS DulbncoonCore inlet umnerature dropOdier posibve reactivtty additions

meanmeanmeaameanmean

Spunoua tnp via nstruoentauon meaaScram due to plant occurencesDetected fault m RPSInadvertent insertion of rod or rodsManual shuulownTurbinetnp meanGenerator tripHigh FW durmi startup or shutdownuadvcrtant startup of HPQReeoxsUaaco eomrol failure mcreara j flowTurbuw bypass or control vatve failuresTnp of one or all radrculanoa pumpsLossofFWaeucrRearcunuion pump aenturePressure refuhuor fails openTurbine bypass failc openFW flow aKreatnf ot powerFW controller mumuin demandMSrVcJewe meanPartial MSIV danreInadveiunt dosure of one MSTV

Turbine tnp wtth bypaaa failureGenerator tnp with bypass failurePressure regulator full closed

meanLots of all FW flow meanTnp of one of FW pump*FW tow Row1 an FW Tow during startuo or shutdown

mean

eventA-yrevenWvrcventjryr

eventfr yr

events yr

eventfryrevent/ryrevent/ryr

event^ yr

evcnt^r yr

event* yr

event* yreventAr.yr.

eventn yrevent>rvrevent^ yreventAyr

event/ryr

event/f yr

event/rvrcveni/ryr

eveniyr vr

106-fOrtT3 6E-1/YT

2JE-lftT

37E+<Vyr

S8E.2/yreUE-l/yr

94E4/yr

94E4/yr

2J&2/yr

3.9E-2/yrIJE-l/yr3.2Et<Yyr31E*<VvT30E*<Vyr

IJErtfyr

4.1E-1/VT

13E-2/yrZlE-l/yr

fi.7D2/vr

128

,,iu3<c$,y.*tt,,,.,.. SOURCE BET SOtlttCg,

95*4*95%. 5%

ITE-tO/yr nja Ringtail 2 NPP phot specific operaunii expenc6-OE-l/yr 1 9E-l/yr Zion NPP PSS table t 1.1-2 literature saarea and pUal operating experience64E+0/yr 4lE+0/yr Zioo NPP PSS table I 1 1 2 Idcraturc source! (pnor) and plant operating experience

9S*,S* SiE-l/yr 9 6E-2/yr Z3oo NPP PSS table 1 11-2 Locntun jourca (pnor) and plant openung expeneoce

2^E40/yr Zjoo NPP PSS table 111-2 Uurattue wureet<pncr) *nd plane operating cxpenence

95%, S% 17E-l/yr Son NPP PSS labte 1.11-2 ItUfature JOUTCCT md plant operating e%penenoe95%, 5% Son NPP PSS table 1.1 1-2 Itttraoire Kiurca and plant operating openepce95%, 5* 2.9E*0|yr 23on NPP PSS table 1 11 2 literature source! and plant operating experience

95*. 5* 3.6E-3/yr 3.3E-5/JT Zion NPP PSS table 111-2 Uuntnre aoorces and plant operating experience

95*, 5% 3.6E-3/JT 3JE-5/yr Zioa NPP PSS table 111-2 loerature sources and plant operating expeneooe

95%,5* 6.1E-2/yr 4 66-3/yr Hon NPP PSS table ! 1 1-2 bteramre aowtc* aod plant opera bug expcneDce

tt/a n/a n/a____AnpaNPPPSA,Scmauryiepon Ueracore snrceaa/a a/a n/i ApciiNPPPSA.S<iBn«aryieport ftegoency b«ied on ipeafic studyn/a n/a n/a____Ang» NPP P3A. Suinany itport litcrac&rB sovrcesa/a n/a n/i____Ang»NPPPSA,Samiuryreeort Ifterabm soaredn/a n/a nfi____Anpa NPP PS A. Somncny report Utcracare wurcefn/a ort n/a CaonoNPPPSStabkA-Z pfant spectfic operating expeneace

a/a n/a n/a CaonoNFPPSSlaMe.Vl pjjut yr^f-'fif cpcratmc ctpcrciflc&

n/a n/a Caono NPP PSS table A-2 plant specific openung expeneoce

n/i Camo NPP PSS table A-2n/a n/a ofe Caorao NPP PSS table A-2

study for the-pfaart with sunjUf valves (Atto Latio)pbust specific operating experience

n/a n/J CaoreoNPPPSSlableA-2 Htstoncat cata for nonh lutv

129

7487497.S)711252

2512542tf25675725«259260261262263264265266267268269270

271

272

mwfroEPWRPWRPWRPWRPWR

BWRBWRPWRPWRPWRPWRPWRBWRBWRBWR

PWR 4 BWRPWRPWRBWRPWRPWRPWRBWR

BWR

BWR

1&SSK ,CCIaiMJorCCIaituttrCCItutuk*CCInitnttrCC Motor

CCInitMorCCIrutsBorCCI&dsMrCCmaiJBrCCbita**CChfcatwCCInitiaorCCbutottrCdoiMnrCCIniteorOCImtutorCCInaiaorCCIuMorCC imdMorCCTtutuiDrCCImtMorCCIu&MorCC Initiator

CCIutMor

CCImtuaor

'lEeATEGOBY "tow of MAC boslotfofiDCbuslots of stcvua wneriOttoflDCbuilotfofsnACbui

ioosofiDCbutoit of to AC battotstternet-mutoo of power bos (OaCSwppiy)low of » AC bustotf of unutinmt »rloHof4kViw«di|!enreactor wrter ierd mwrujneotuioo foknlots of MAC busloo of • DC bosLoM of offsite power >30 tnimitelots of service wuerfiflunofDCbusIon of debated DC powerIon of Krvice weer systemIMS of service wBerJotfof conipoocot coolinglc**ofTBCCW«yjton

^ofinarume*,.

tots of 4cV tmtrnency bus

urtmroR*'short to pxwnd on AC bus

tuaire of ncnnaJty open discbirge MOV

thort on 4160V butshort oa 480V butfuture of 41 60M80 V trenSonner

Lon of coolmg wsttr to TBCCW tynemLea of coadeosxe pumpiLow of service waterLon of sstnancQl urLOB of RBCCW jyjton

TREQDESCman evdt^.yrman cvouA-yrman ev«m/ryrman eveni^ryrman *v«oWryr

man evenUryrman enaUr.fraam evenOr.yrman event/r.yrman cvtot/r.yrmcQ enajt.yr.man even^vrpoint eitevtnt/r.yrpoint est. event/r yrpoint estevta/r ye.

mem eventfryrman evosAyrmaa evtntfryr.maa ewotAyrman evratAyr.man eveat/ryrman event/r yr.

man evcn&yr.

man eventfr.yr.

1 em*3^E-2yyUt2yvr16E-3/W90&4yyrUEJ4r

90E-4frr90E-4/yr40&3/yrlOB-2/yr5.48-3/ir1.7B-l/yr5.4E-3/VT36&2M3JB-2/yr3.0E-3/VT4J3E-2/yr1.8&3M3.6E.2/JT3-2E-l^14E-5/yr94E4/yr94&4/yr3-2&2/yr

""*HE4/jr

130

KASCfcVAft-nAn/tnAn/an/t

nAnAnA

95%, 5%95%, 5%

n/a93%, 5%

EF3EF10EF3

tiPPERBNBnAnAnAnAn/a

nAnAnA

7JE-2ftrUE-2/yr

nAUE-%r

nAnAnA

LOWER. BM>nADA

nAnADA

nAnAnA

7.5E4/yr2.7E-5/yr

nA17&3/yr

nAnAnA

soratcB-. £"* -> - - - - - vIREP-ANOl table 4-7EREP-ANO1 table 4-7IREP-ANOl table 4-7NUREG/CR 4550, VoL3,Smr)aJnu !.„,!* leIV.3-1NtflU^^VoUSanyUnuUu ^IVJ-1

NUREGCR-45S), VoM( Peach Bottom. ., ablcIV 3-1NUREC/CR-4550, VoM, Peach Bottom,,., !ib[elV.3-lOccDMPRA. table 5.9Dome PRA, table 19OcooMPRA, taWe J.9OcMMPRAj ttbto 5.9OccwePRA, ttbte 5.9Shorehtra PRA, Appendix A.1Shcntuffl PRA, Appendix A.ISboreham PRA, Appendix A.I

.'...BtTSOBfceEft'^'''^''" " ' ' 'OCONEE PRA RSSMAP dau batedata taken from other PSAmgtneenn* evaluation DMH Kenenc rebabditv diuDerived from jenenc component failure dauDenved from (eatnc conptuat Uihati dala

auclcar »URt* »d plaat opetxuoit expetreacenuclear xweei aod plant operauog experiencejyjtem analyjit as* f*ab tree modeltnitvnt of plaatt fj^tcouplant operating experienc*denved from aynem aaalytii otmg fault bte nodellystem aoalvnt4»c «BJpment faultntmitxl frera noctear opemrag expenettcesnclear opentmt expeneoce and otgmeenag evtluabaanucletr operating aperancc

WASH-MOO, Reactor Safety Study , App.tx xl. Rn.I-4 11 44.12nAnAnAnA

95*. 5%95%,5%

nA

nAnAnAnA

3.6E.3/yr3.6E-3/g^

nA

nAnAnAnA

13E-S/vr3JE-3/F

nA

CaJratCUffi Una 1 1REP, table 42Catmt Cbfi Un* 1 IREP, table 4 aOrijrah giro 1 NPPRinit)»l»2NPPZkn NPP PSS table 1.1. 1-2Zke NPP PSS table 1.1. 1-2Caooo NPP PSS able A-2

ivstem evakiatwo us«g jcnenc fajjure dataeasmeermg evalotuoa and nocletr expeneace datapbua specific opentatexpea-ence (6.2 yon)Itfenmre aoarcealaeratara aourca (pnor) aad plant operabng expeneocelitentun gouccM (pricr) and plant opera&ng expeneacemalvsu aanj ftott tree method

n/a CM«O NPP PSS table A-2 gcitvju of monrnwot air tyatm

nA Caooo NPP PSA table A- analym of emergency power synem by fault tree

131

PJLANT TYPE- aS SETS -In

34567391011

12

1314

15

16171319

20

2122232425262728293031•a.3334

35

PWRPWRPWRPWRPWRPWRPWRPWRPWRPWRPWR

PWR

PWRPWR

PWR

PWRPWRPWRBWR

BWR

BWRBWRBWRBWRBWRBWRPWRBWRPWRBWRPWRBWRPWRBWRPWR

LOCALOCALOCALOCAtnooeattraaSKatLOCALOCALOCALOCALOCA

LOCA

transienttransient

tnaneBl

CCIutiatorCC InitiatorCCainatermanege

tmneat

triaaiiia-riiMniiTtnacMBCLOCALOCALOCALOCALOCALOCALOCALOCALOCALOCALOCAtraancat

- IE e*TKSORV 'large LOCAmedium LOCAsmall LOCAsmall leak on prexaunzerkm of feedwaterCFW)kxs of off ntt power^LOO P^lance LOCAlance LOCAmedium LOCAsmall LOCAsmall LOCA

small small LOCA.

loss of off site powerloss of power conversion system (PCS)

otter man Iocs of PCS (turbine tnp)

loss of an AC busloss of a DC busIocs of service waternotation of reactor from condenser

turbmemp

andvertant open rauef valveIocs of feedwaterIocs of off sito powerlarg*LCCAmedium LOCAsmall LOCAlarge LOCAJargeLCCAmedium LOCAm*dwmLOCAsmall LOCAsmall LOCAsmall small LOCAsmall small LOCA

loss of PCS

UflFn*TORSbreak wilh ecu.area Rreater than 400 sq cmbreak wimequ are* SOto 400 tq cmbreak with equ.arai 2 to 80 sq onstuck open relief or safety valvecomponent failures m FW systemloss of voltage on more thaa 1 power busbreak with equ.ioe Kreaur than 13 j inchesbreak with eqo.xt» 10 to 13-5 inchesbreak with equ.size4 to 10 inchesbreak with eqg.sae 1 66 to 4 inches

break with equine 1.2 to 1 6* aiehesstuck open prsrz safety valvebreak with equivalent sue .38 to 1.2 inchesnude open relief valveRCP seal failure

Total loss of FW flowFall or psaal closure o/MSIV(t loop)CloturaofaUMISV-*Increase m FW flowFW flow nuttbOity (operator error)

Loss of conde&satft pumps (all}Loss of condenser vacuum (total)Opening of steam relief valvesLoss of cuculaimx water (CW)Locs of RCP flow (1 loop)Uncontrolled rod withdrawalCRDM problems and/or rod dropHigh or low prxzr pressureComamment pressure problemsFuU or partial closure of MS IVTurbine tnp, mro&k valve dome. EHC problemsGenerator trm and (enerator cauKd problemsSpunoos cuto tupManual tnp due to false signalsshort to ground on AC bus

failure of normally open discharge MOVClosure of all MSIVTurbine mp without bypassLocs of condenserPartial ckmu* of MSIVTorbme tnp woh bypassSunup of idle recce loopPressure regulation niiniTBInadveruni openmc of bypassRod withdrawalDuuitrbsnct) of FWElectric load rejection

total lots of feedwater

break with equ-soe greater than 4mcaesbreak with equAia 1 to 4 inchesbreak with equAza up to 1 archbreak with equivalent sac > 6 inchesbreak widi equivalent are* > l-.3sqft.break with equivalent tax 2 to 6 cchesbreak with eoAiivalenl area XO4-J sqft(hq) l-.33qft(!ttam)break with eqnivilmi sue -5-2 inchesbreak with equivalent area <jOOJ sqftfbq) or <-l sqtKsteam)break with equ joe >-5 inch or now 50-100 gpmbreak with flow 50 to 100 tpmTotal loan of FW now (Usted m DESCRJield)FW flow mstatnliiyopeniar errorpntchanifal causesFuU or partial closure of MSIV(ooe loop)Closure of all MSIV-sLoss of condensate pumps (all loops)Locs of condenser vacuumCondenser leakageSO leakageOpening of steam relief valveVuubmeous leakages on secondary sideLocs of CWLoss of CCLoss of SW

- FREOJBESCmean events vrmean evemA'vrmean evenUrvrmean evem/rytmean evenfAyrmean tvenl/rvrmean eveatjlryrmean eveotfryrmean eventAryrmean evem/ryrmean event* yr

mean crvcat^ryr

meaa eveat/rvrmean eveat^yr

mean eventA-yr

mean evem/ryrmean eveat/ryrmean event/r yrmean eveoAyr

mean eveot/ryr

meaa aveotA^yrmean evem^ryrmean ovtojryrmeaa eveaUryrntttfi cvcnuTi.)Tmean eventA-yrmean eveniA'yrmean ewnt/ryrmean events yrmean eveat^-yrmean evem/ryrmean eventA* vrmean evem/ryrmean cvcm/ryr

- FREQZ7E-4/VTg.0&4/vr2.7E-3/VTUE-3/yrS_OE-t/yrIXIE-l/yr7JB-5/yrUE-5/yr16&4/yr33&4/yr3.1&4/VT

ME,2^r

3^E-l/yrIKE+O/yr

7 lE*Oryr

35E-2/yrUE-2/yri6E-3/yr1 lE*0/yr

6\06-2/yr7jOE-lryr5J&2/yr4jOE-4/yr2X)E-3/yr10t>2/yrSXiB-VyrIJ)&4/VflOB-3/vr10E-4/yrl^E-3/yr3.08-3/yr2J)E.2/yrJOB-TOT14E+Oryr

132

•.BJ&GIKVXREFX)EF10EF10EF6EF3EF3aftn/an/an/an/a

n/i

n/n

85PKERBWBn/in/an/ao/an/an/an/aB/Jn/an/an/a

a/*

n/a

LOWERBNR SWHtCE- ~ ~ " ' '" "..'•'•<n/an/tn/an/anAnAnAnAn/anAnA

n/i

nA

German Ride Studr. tabfe H . 4-5German Risk Study, table Fl . 4-5GensaoRukStady.taMoFI.4-S.Cennu Ruk Study, table Fl . 4-5.German Ride Stndy. tatria Fl . 4-S.Gcman Ruk Study, tabtoFl. 4-5IREP.ANOltabl.4-7DtEP-ANOl taUe4-7REF-ANOl table 4-7DCEP-ANO1 table 4-7KEP-ANOlt»bk4-7

REP-ANOlIabte4-7

REP- ANO1 table 4-7

a'fcT SfftSlCE ^ ^ •>'" V ^j^;v— •jmenc data sourcesgenenc d«U souretsgeneric data sourcesopenoaK exDerwocc and emuneerax aoatynigetman opennnc experiencesystem anatyssRSS data coobned wt* ANO-l break raaxaRSS data conbned wuh ANO-l break ran netRSS data combined wiita ANO-l break nage*RSS data combined wirh ANO-l break rangeiRSS dau combiaed WUB. ANO-l break raaiea

RSS data combined witfa ANO-l break noee»

EPRI NP-801 and data from the utilityn/i

a/a

n/a

n/a

o/a

n/a

D/t [REP-AN01 tail* 4-7 EPRI NP-801 and plant specific info

IREP-AKOl table 4-7 EPRI KP-SOI and plant specific info

n/i nAn/a nAn/i nAB/a n/a

nAnAnAnA

IREP-ANO1 table 4-7IREP-ANO1 table 4-7mEP-ANOt table 4-7Lnnanc PRA, table A. 1 J

OCONEE PRA RSSMAP data bandata taken from other PSAeBgaeenng evahiatioei us«t; genenc reiiabiJuy dataBWR operating expenenoe

n/a Lmxnc PRA, table A.U BWR operating experience

n/inAnAnAnAnA

EF3EF3EF3EF3EF3EF3EF3EF3EF3

nAnAnAa/anAa/anAnAnAnAnA9A

n/anAn/a

nAnAnABAnAnAnAnAnAnAnAnAnAnAnA

Lfawric PRA, table A.IJUn>enc PRA, note A.l.3Lbaeric PRA, ublc A 6J.Lanenc PRA.taWe A.1.6UmericPRA.tableA.1 6Lmenc PRA. table A. 1.6NUREWCR 4550, VoLl,NUREWCR4550,VoLt,NUREQCR 4550, VoLl.NUREG)CR4SSO.VoLl.NURBQCR 4550. VoLl.NUREGrCK 4550. VoLl.NUREOrtS 4550, VoLl.NUREGCX4S50, VoLt.NURECKR 4550, VoLl,

Metbodolo».,t ^vm.i.MetSodotow. , t »le Vm.1-MBfaodotoQ.. . tMftjMdoiogy. ,tMemodolOEy. ,tMeaodeloty. .tMctfMdoloity^ , tMetBodoJow- ,tMethodoki«y- . I

>tevnia-jiaVm.1-Asvm.i->4«vra.i-*vra.i-»evmi-i>leVmi 1

BWR operating expeneoceBWR operating experienceracioaal end dataeoK^eensg evatuucfi and dunreBt uala aocifcecengneeruia; evaluation and dittereat data sourcesengtaeenng evakutioB and different data tauiaaaverage of pest PRAiRSS valueaveraRB of pstsc PRAsannand front nodcar, ndnatnal and other data soaredaverage of past PRAimeted form nadear. BdtatrfaJ and otter data sourcesNPP oe««in(( expeneaccreview of pen PKAtBawl OB NUREG/CR 3862

133

36

3738

39

40

41

42

•43444546474849

30

SI

Si

5354

55565758596061

f>2nfA

P5LANT TYPEME-SEF '

PWR A BWR li i i . i i inrPWR traisaml

BWR Ti-aarifnt

PWR trawct

BWR tnasient

BWR traaaicntPWR LOCAPWR LOCAPWR LOCAPWR LOCAPWR LOCAPWR naaninl

PWR H ••am

PWR t«te

PWR trmim

PWR CCfcMatorPWR CClsttator

BWR LOCABWR LOCABWR LOCABWR LOCABWR LOCABWR trassautBWR u aw in

HlrVR tnBQmt

BWR OCblUalorBWR CCfctruator

H& CATEGORY'-'- ' ... -loss of power conversion system

loss of » DC bosother rhan Iocs of PCS

Ion of FW (steam ode PCS available)

Who- dim lots of PCS

Iocs of offline power

lo*> of off site power

madveront ODenmK of rebef valvehue LOCAmedium LOCAsmall LOCAmail small LOCAmtenystem LOCAturbine tnp wok MFW available-

loss of main feedwater

loss of chargmf. pump cooling

Iocs of off tile power

loss ofaDC busloss of aa AC bus

large LOCAmedium LOCAmall LOCAsmall small LOCAintenvstem LOCAkm of PCSother dun Ion of PCS

IIKS of off site Dower\va of a DC bus'iyis of an AC bus

•JiNmATQSS '•• • • • . . . .Electric Load rejection wilh bypass failureTurbtoe trip with bypass failureFull MSIV clocuraPartial MSIV donre1 *«* rtf frtfft^rt iff *r»m*tvr*

Prtsore regulator fnli openPrenure regulator fiub dotedTurbi&b bypaa fafls openTurbine bypat* btteckweTorfaine bypca or oontro) vatve mcreasB pressure

Low of RCS Oow<one loop); Total loss of RCS flowUncaurokd rod wufadnwiLCVCS problean-baroa dilutionCRDM meduaucal probtons or rod dropLeakage froni: ooctrol rod. primary system, pressunzerI^ressurizer pieaoug low or high; Preuunzer spray failureInadvertant SI si(9alrfiM.iimmn prueule pmMjm

Spuriow crip; auto or manual tnp(no mosicat conditions)Presjure-umptrinn^pciwrr im balance rod poanoc errorStamp of inactiw RCS loopLoss or reduction of FW flow(one loop)Increase in FW ftownoe loop;all loopsLota of condentate pumpcona loop; all kxipsTurbine tnT>;«eneratonnp:ihroale valve dosure;EKC problemloo of all FW DowFW low flowTurbtoe trip ;ekctnc load rejectioaRcareaiatKe control ttUore: deaeastng flow, increasing flowRccirculitioD pusp tripe ooe; allStartup of idle recarcaiatHa pumpRaarcuUban pnap KimreFW iacreasmt ftow; Ion of FW beaterThp of one FW or condcnsue pumpRod wtihdrawaUiaadverDmt insemon of rodsInadveruat stamp of HPCI/HPCSSpurious trip via iostrumcatauonRFSIaviltSCRAM due to other plant ocetinacesManual SCKAM-ao Dat»unc ooodmonSCRAM cause unknownloo of all ofiiue powerloss of power ID Mccessw ptanl systemloss of off me powerVMS of aaxihary powtr (transformer)

break with eoomleBt size 6 to 29 inchesbreak witti eqojvmlera sire 2 to 6 inchesbreak wt&eonmiemsixe l/2to2mchesbreak with eoamlott sue <_5 inches

uiiUue Uy^ reactor tnptoss of loadMSIVclOBtrelow of torbiw coacralfaUtre of ntm FWiijhSG water levelioadvotamSlHninlloss of caartLDC pmp CCloss of chanOBH pompSWfiuaira of offule powe jndloss •( atauon resem powerloss of power to rite swiicayard

sBotoa41«OVbttsthononXfAVbxsfailure of 41«W*> V traatformerbreak with equivalent area > 1 sqftbreak with eoaivtlau area XXM - 1 sqftbreak with equivalent area < OO4 jqft(liq) or.Ol suftfsteam)leaks wtth SO- 100 fpm flow frearculatMX) pump seal)

madvertaot open rebef valve tn primary systemloss of FW but neaos side of PCS initially availableother mitiators with same ctHiseqcnces

RKEODE5Cmean ev<nt/r.yr.

mean evenvfrvr.meafl eveat/ryr.

mean eveut .yr.

mean emtk.yr.

mean event .yr.

mead evemA'.yr.

toeaii eveniVryr.mean eve£ft>ryr.mean evem .yr.mean evcniA'yr.mean eveoUcyt.mean eveajrvr.mesa evtntA.yr.

UKJUI eveotfr-VT.

meao evenUr.yr.

mean eveobryr.

mean event^yrmean evottA'yr.

mesn evenUr.vrmean eventA.yr,mean eventA-.yr.mean eventA-yrmean event* yr.mean evem/r.vr.mean cvemyr.yr.

mean evetu/rvrmean cveniyr yrmean evetiUrvr

FREOto&HWyr

SX)B-3^r&8EXyyr

5.6E-l/yr

4^E*4Vyr

l.OB-1/yr

l.OE-1/yr

l4B-l/yrSX>E-Wyrl.OE-JMt.OE-3/yrXOE-2ryrl^)B-6/yrTJB+flryr

9.4B-l/yr

lOB-Z-yr

7.0E-2/yr

9.0&4/yr90E-3/VT

i7rvVyr8,OE4/yr17E-3ftn-iTE-lfyrtjOE*yrl^E«<Vyr2.6E*<yyr

7.0E-2/vr9OK-4^r>OB4^vr

134

TUNOR VAK UPPERBNITEF3 a/a

EF3 n/aEF3 a/a

EF3 n/a

EF3 n/a

EF3 n/a

EF3 n/«

EF3 n/an/a n/aa/a n/an/t n/an/n n/an/a n/an/a a/a

n/a n/a

n/a n/a

n/t n/a

n/a n/an/a n/a

0/1 n/an/a n/an/a n/an/a n/«n/a. n/an/a n/an/a n/a

n/a n/an/a n/aa/i i/a

LOWER.B'Vft SOURCE •• •> -^n/a .NUREG/CR 4550. Volt, Methodology. , 1 Me Vffl.1-1

n/i NUREG/CR 4530. VoLl,Methodolo»_.tiSleVmi-tn/a NUREG/CR4S50.VoU,M«hodoloty..tiM*vm,l-L

aA NUREG/CR 4550, VoU. Methodology ., btoVnLl-l

n/a NUREC/CR4SSO,VoU,Meinodology , bUVHLl-l

n/a NUREG/CR 4550, Voi.1. Methodology , W.VHI1-1

n/a NUREG/CR 4SJO.VoLl,Mediaeoioty „ *teVnLl-l

n/a NUREG/CR 4550. VoLl.MethodolorV-, *teVmi-ln/a NUREG/CR45»,Vol3,SurrylFniil-,tib!eIV.3-ta/a NUREO/CR4S»,Vol3.SunyUmll.,tiiiJerV3-ln/a NUREG/CR4SSO.Vol3.SurryUniU_,tiliJelV3.lnA NUREG/CR4S».VoU.SnrryUnnl..ti!ileIV3-In/a NUREG/CR 4550, VoU.Snrry Una 1_ . i ibteIV-3-!n/« >OJREG/CR4iSO,VoL3>SurryUmtl...tral*rV3-I

n* NUREWat4i50,VoL3.SurryUintK.t!bl«IV3-l

n/a NUREG/CR 4550, VoL3. Suny Umt L . I ibteIV3-l

n/a NUREG/CR4SSO,VoL3,SurryUnnl.,!!t.erV3.1

n/i NUREG/CR 4550, VoJ3, Surty Unit 1 ,nb)eIV3-ln/i >flJREG/CR 4550. VoL3.Surry Unit 1 .nbtarVJ-l

art NUREG/CR-4S50, Volxt. Peach Booou _ ubk IV 3- 1a/a NUREG/CR-4550.VoU. Peach Botwci^ BbleIV3-ta/a NUREG/CR-4S50, Vol.4, Peach Bonoci- table IV 3- 1nj* NUREG/CR-4J5D. VoM, Peaca Booor^ UbkIVJ-1n/a NUREG/CR-4S50. Vol 4. Peach Bonag>_ abtelVJ-tn/a NUREG/CR-4550. VolA, Peach Bon«i_ oMeIV3-lnra NUREC/CR-4S50. VoM, Peach Boaoni^ tabkIV3-I

n/a SUREG/CR-4JSO Vol 4, Peach Bonocs_ able IV 3-1n/i SUREG/CR-4550 Vol 4 Pe»ch Honor . abteIV3-l1/1 NUREG/CR-4550 Vol4 Pesch Boaora. jblelV3-l

UtT SOURCE.Bated oa NUREC/CR 3862

DC power suoplv study, NUREG O666B*aed on NUREG/CR 3862

Based co NUREG/CR 3S62

plant blackout analyst* NUREC 1032

Plant btoetoutmaryu* NUREG 1032

Band on NUREG/CR 3862Based oa NUREGACR 45» VoL 1Bwd en NUREG«CR 4JSO VoL 1Based on NUREG/CR 4550 VoL 1Based on NUREG/CR 45» VoJ. 1RSSdnasfaral cpealic operaung expeneoce

Companson of NUREG/CR 3862 and operaunj experience

NUREG 1032

Derived from eeoene component failure dataDerived from generic component failure data

pan PRAl and other sourcespan PRA J and other sourcespan PRAi and other sourcespot PRA* aad otber sourcesaaalvsis of the svstem interfaces usnE generic failure datanuclear sources and plant openung experiencenuclear source) aad plant operating experience

nuclear sources and plant operating experiencenuclear sources and Dlant operating cxneirencenuclear sources and plant opcraunc exocncnce

135

PI65666768697071727374757677

7879

8081

82

S3848586

87888990919293949596979899100101102103lot10510610710S

109HO111112113114

US116

BWRBWRBWRBWRBWRBWRBWRBWRBWRBWRPWRPWRPWR

PWRPWR

PWRPWR

PWR

PWRPWRPWRPWR

PWRPWRPWRPWRPWRPWRPWRPWRPWRPWRPWRPWRPWHPWRPWRPWRBWRBWRBWRBWRBWRBWR

BWRBWRBWRBWRBWRBWR

HW-RBWR

i iBSEnr-LOCALOCALOCALOCALOCAl i ami intransienttr_c«tt

OW$«MtnaneatLOCALOCALOCA

LOCA

transienttransient

tfetaHlatflt

traneattrassaenttraacxnt

tn"™t

BaBSKntQaosieattrams*CC InitiatorCC InitiatorCC InitiatorOC InitiatorCC InitiatorLOCALOCALOCALOCALOCAtnssrio*tnsakntrrnaiiniLOCALOCALOCALOCALOCAnrntua.

mnm.uitnwtenttnasxxttransienttranaentDaoskoi

uannenJtransient

IE CATEGORYlarge LOCAmedium LOCAsmall LOCAsmall small LOCAtntersystem LOCAloss of PCSton of FW and steam side PCS availableother than Iocs of PCSIon of off site powertnadvertaat open ing of relief valvelance LOCAreactor vessel rapturesmall LOCA

steam generator tube rupturereactorAnrbme tnp

ipunmi 51 signalsteam line break

Jots of condenser

FW line break (large)loss of main FWpartial tot* of man FWkm of off site power

Inn of off site power (tttbstauon faute)excessive feedwatersponous tow prenaroer siimalIon of service waterloss of power bus KICKS supply)loss of as AC butIon of murttniMU airIon of 4 kV iwnchgeartame LOCAmedium LOCAsmaDLOCAuita&ysteiu LOCAreactor vessel ropeinkm of off sile power(on of main FWtransient with main FW availabletarge LOCAmedium LOCAsmall LOCAreactor vessel rupturemtenjntem LOCAIon of condenser

MSIV closure!on of feedwaterloss of aft site powermadvenant openmK of relief vmivecontrol rod withdrawalturbine tnp

mam steam line breaKfeedwiier line bresx

INITIATORS-break with equivalent size > 3 sq ftbreak with eouivalent area MS- 3 sqftfliq) and 1 3soltf steam 1break with eouivalenl area <-OOS sqlUhq) and <.! sqftf steam )leaks wni 50 lo 100 axn flow (reorculalioB pomp seal)

break with effective diameter > 4 inches

break with equivalent sna 1/2 to 4 nchesaudverlant FORV or safety valve openingRCP seal failurecontrol rod drive seal leakagetube nnxnre with leak greater than 100 gpmRod dropLttdveflani rod withdrawalInadvertanl bonbon or dilutionReactor tripCold water additionInadequate mam FWTurbine and control valve malfunctionPressunzer spray failureTurbine tnpGenerator faunsGnd disourbancesAdmnustnnve shutdownsHPIftowsteam line repttiretnrbme bypass valve utadvenant openinglost of condenser vacuumloo of condenser orcutatmn water

failOFBof gndfailure of feeders

break with equivalent sue >6 mcfaesbreak with equivalent SIM 2-6 acbesbreak with equivalent sue 3-2 inches

break with oquivaknt sue >4 inchesbreak with eonivaknt nu 1-4 Bdwsbreak wah equivalent sac <1 inch

loss of normal coodeater vacuumtuibne tnp watt bypass valve failureelectric load rejection with bypass va r»e failure

EJecsnc load rejectionTurbine tnpInadvertant closure of one MSTV, partial MSWdosurePressore regulater failc open; ckxedTurbine bypass fa ds-optaTurixne bypass or control vatve cause increase pressureRecoculauon flow future. ilecreagnc'ocreaimgTnp of recBcuUCMnpuBp: toe; allAbnonnal startup of ulk recirculatian pumpRecuculaiion pump seirureLeo of FW heaterTno of one FW or condensate cump

FRKODESCmean event* vrmean cvctitfr vrmean everu* vrmean event* yrmean event*vrmean event* yrmean event* yrmean event*vrmean event*yrmean eveu*yrmean event* yrmean event* yrmean event* yr

mean event* vrmean event* yr

mean event* yrmean event* yr

mean event* yr

mean ev«a*yrmean event* yrmean event* yrmean event* yr

mean event* yrmean event* yrmean event* yrmean event* yrmean event* yrmean event* vrmean event* yrmean event* yrmean event* vrmean event* yrmean event* yrmean event* yrmean eveat*.yr.mean ev«m*yrmcsB eveot*.yr.mean event* yrpoint esL event* yrpoint est. event* yrpoint est. event/ryrpant esL event* yrpoint est event/ryrmean eveot*.yr

mean event* yrmean event* vrmean event*yrmean ewot*yrmean event* yrmean event* yr

ooim est. event* vrpoint CSL evem/r vr

FRW}30E-Wyr>iOE-»fyrJj01v3/yr30E-2/yrtX>&SfcrlJE+<yyr70B-2/W5XE+<Vyr7XK-2/yrL4E-I/YT93B4/yr1 lE^/yr

8.6E-3/yr49E-»0/yr

lX»E-2/yrlOE-Vyr

2-lE-l/yr

93&4fycojJE-1/yr&9E-l/yr40>&2/rr

IJB-lAyr9iE-2/yr4.46.2/vr4.06.3/yr2J5&2/yr54E-3/yr175-l/yrS-4E-3/VT4.7B.5/yr93&4/yrl^E-3/yr46&6/yrl.OE-7/yr10B-l/yr3XC«<VyrIJOEtOtp7jOB4lrr3XIE-3ATS.OE-3/yr3.0&7/yrl^E-7/yr4.1&1/T,

Z4E-l/^r13E-1/JT

E.CE-2/rr9fl6-2/yrWB-i/yr

3gfiJ!/vr8.7&9/W

136

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95* 5*95*, 5*93*. 3*

93% 5%95*.S*

95* 5%93*. 5*

95*. 5*

95*. 5*95*. 5*95%. 5%95*. 5*

95*. 5%95*. 5%95*, 5*

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2-8E-3/yr43E-6/vr1-2E-2/JT

2.7E-2/yr5.7E-tO/yr

4J£-2#rl-2E-2/yr

3BE-l/vrZ8E-3/yr9.2E-l/yr97E-l/yr97Mftr

30E-l/yrXlE-l/yr17E-l/yr

n/a.7-SE-2/yr13E-2ftr

n/alJJE-2/yr

n/an/an/an/an/an/an/an/an/an/an/an/an/aate

n/an/an/an/an/an/a

"/an/a

ôwER-Rjn:n/an/»n/an/an/an/an/an/an/an/a

94E-7/JT6.0&S/yr10B-6/yr

£6E-5/yr

7.8E-6/yrU>E*yr

"&2/yr

6.9E-7/yr36E-l/yr4-OE-l/yr71&3/yr

riE-2/yrl.BE-2/yr17E-3/vr

n/a7.5&4/yr2.7E-S/yr

n/t2.7E-5/yr

n/an/an/tn&n/aHAn/»n/an/an/an/an/an/an/a

n/an/an/an/an/an/a

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k SOURCE ^ >» ,NUREG/CR.4S50.VoIo, Grand Gulf 1_ J>fc>tV3-lNUREG/CR-4550 Vol 6. Grand Gulf 1 . iMelVJ-1VUREG/CR^t550, Vol 6. Grand Gulfl . iblelV3-l\UREG/CR-<JSSO, Vol 6. Grand Gulf !_., ibie IV 3 1NUREG/CR-4SSO,Vol«,GBu>iJGam_, ibfeIV3-lâ^ÊC/CR-4SSO, Vol.6. Grand Cain--, iblelV.3-1NUREG/CR-WSO, Vd 6. Grtad Calf L_, iWeIVJ-1NUREG/CR-4S50, Vc4^. Grand Gulf !„, ibleIV3-lNUREGA31-4S50, VoL6. Grand Gatt 1^ ibb IV 3-1MTJREG/OW550. VoL6. Grand OUfL.. iWelVJ-1OcooeePRA.table59OconeaPRA,table59Ocooee PRA, table 5 9

Oconee PRA table 5 9Oconee PRA. table S.9

Oconee PRA. table 5 9GxxneePRA.uble59

Ocooee PRA. table 5.9B ML Review, t*bl* <3OcooeePRA,t»ble59Ocooee PRA, ttb Is 5 9Ocoo«PRA.table59Occ«.PRA,t^S9

Ocooee PRA, table 5 9Oconee PRA, table 5 9Oconee PRA, table 5 9Oconee PRA, table 5 9Oconee PRA, table 5 9Oconee PRA, table 5 9Ocooee PRA, table 5 9Ocooee PRA, table 5 9Seqnoyaa NPP RSSMAP, able -^.Sequoyifa NPP RSSMAP, table 7-4S«njoy*h NPP RSSMAP, table 1-4.Swuoysh NPP RSSMAP, table 7-4Sequoyah NPP RSSMAP, table 7-4Sequoymh NPP RSSMAP, table 7-4Sequoyafa NPP RSSMAP, oUe7-«Sequoyak NPP RSSMAP. table 7-4Sfioreham PRA, Appendix A.ISaorehan PRA, Apejeadtx A,lSacnhaa PRA.Aposnd:x A.ISnorehaa PRA, Appendix A 1Sborehaa PRA, Appendix A.1Shoreham PRA. Appendix A.txMe A,!- 1

Saorebaa PRA, Appendix A, ubleA.1 1Shorehara PRA. Appendix A, table A 1- 1Sboreham PRA. Appendix A.labie A.!- 1Sfcoreham PRA, Appendix A,taMeA.I 1Shoreham PRA. Appendix A. table A.1- 1Sboreham PRA, Appendix A. table A.1 )

Shorehara PRA. Appendix A 1Shoreham PRA. ^ppendlx A 1

VLT SOURCE

Analysa of the system interfaces nsto^ generic 'aiture diudifferent nuclear lources and pfant operating exoenencediffereta nuclear sources sad plant operating experiencedifferent nuclear won** tad pteat operaim* exoenenceassenn ent al different nuclear source*different nuclear sources and pbat oceratmf exceneocegeneric pnor updated with pbnt specific openmm experiencevanous sourcesupdate of generic pnor

generic data uodated with oliul sooafic ooenlne exoenencegeccnc data updated will] plint specific cpcraimj experience

generic data updated with plant specific operatm* experiencegeneric dau updated wtfh plant specific operauag expcncoce

generic dau updated wah plant specific operating experiencerecalculation made by the revtewengeneric dau updated wah ptasi specific opersung exDenencegeneric data updated \«1 plan specific operatmn expenencegeneric dtu updated witti plaal speaiicoperttnn excenencegenenc cjitt updated with ptaal specific operating expenence

generic data updated w«h plant specific operating exoen encegenenc dui updated wxll pUnt spectfK: operatmt exepeneocesystem onatvsissystem xnatvsu usntg fault tree modelanarvsu of plants systemsplant operating experiencederived Iran system analysis using fault tree mooelsystem analysis-basic equtoment fcub.eagmeermg evaluation using geaeric dataeagaieeniig evaluation using genenc dataengmeeniiK evahiauon using genenc daueagtneenns evahutm usag geoertc datageneric dau

generic data based on collection of operating experiencesgeneric d&Ut bated on ******* reactor operating expenenceseoenc data based on reactor ooenuaR experiencegeneric data source!system evaluation£C&erhC data *^ff*<i on nuclear ocentu&A jyf^>r*i*f*taken fiotn EPRI NP-801

tfotftc dj^ bdsed on nuclear operating expemoce etc.Xenenc dad bawd oa rmclear opentmR expetrtnca etc.ublity specdic d»uReacnc data aad eopneenofi evahuttcaKenenc data based oo nuclear operatun opetrcnce etc.geaenc data b«ued oo nuclear operating expetreoce etc.

«»îneenn^ evalumon>3*înccrin^ cviluiiion

137

117118119l»

171177.m124IM116127123

129

130131132133

134135

136

137

138139140141142143144145

PLANT TJfPRBWRHWRBWRBWR

PWR* BWRPWR* BWR

BWRPWRBWRPWRPWRBWR

PWR

PWR4BWR

PWRPWRPWR

PWR

PWR

PWR

PWR

PWRPWR

BWRBWRBWR

BWRRWR

RWR

IE SETtransientCC InitiatorCC InitiatorCC InitiatorLOCALOCALOCALOCALOCALOCALOCA

transteot

CC Initiator

LOCALOCALOCA

transienttraoncat

tfaimeia.

transient

CC InitiatorCCtnanaorLOCALOCALOCALOCALOCALOCA

IE CATEGORYHPO/RCIC line breaksreactor water level instrumentation failureloaofanACbusloBOfaDCbusreactor vessel rupturehow LOCAmedian LOCAm^«jiMM LOCAnull LOCAsun LOCAiDlfl system LOCArapid shutdown

rapid shutdown

Loss of offnte power >30 minutelarge LOCAmedium LOCAsmall LOCA

loss of off site powertotal imerupuon of PCS

uauaunl requinng RCS pressure relief

transient which do not affect front line systei

loss of service waterfailure of DC bustag. LOCA (1)larp LOCA (2)larreLOCAO)medium LOCA ( 1 ~imedium LOCA (11small I OCA

1MT1ATORS-

break with equivalent sue > o mcnesbreak wife equrvakotsrw 2J-8J n (liqj and 47-o in (steam)break wuh equivalent sue 1-6 mesabreak wim equivalent sue OLo-2.6 ra <hq) and ljO-4 7m rsteam)break WMh eoamleat tat 1/2-2 inches

Rod withdrawal at powerFW ccsltroJtr failure; loss of FW flowReorculat» flow control &ibredeere*»euicreasingStartup of idle recaculatMn pumpLots of FW bettingInadvtrtant HFCt stanLots of attdfitry powerTurbine traXmrbma valve deeurefcload rejeeooeCstop valve cl)MSIV dosare

Rearaatoan pomp awmreooe pomp; two pumpcT-O preaaore regulatcr bitare-rapid openingRod cjecoox. rod drop accidentStamp of idl« rearapump won smultaneous urbine tnpTurbine tnp; lost of man ffotmar, LOOPLots of cooienjer vacuumInadvcitanJ MSIV dosareLocsof Bttm FW( lose of man CWt loss of condensate pumpsInadvenaat openig of SO PORV-«Increase m MFW flowOpemng ei alt bypass valvc*(steam dump)Uncontroltd rod withdrawal: control assembly dropBoron dttadea (CVCS nulhnokn)Startup of naeove RCS loopOpening d pressonzer SRV or RVLoas of RCS coobni flow, sieznre of all RCP-sRupture of FW pa^ng or man steam ISMS: rupture of SORupture of CRDM housing

break wift equivalent sot >43 ncbeabreak writ «pivalenl son 1 9 to A3 mchesbreak wim coumltnt stu 3 to 15 mchexreactor coolaal pump seat ruDrure

Total loaaofFW flowClosure of all MSIVsFv/ flow instabilityLc0 of all condensale pumccLoca of coBdflisef vacuumLost of CWmrbne trp or tbroafc vahrn closuregenerator Dtp or geeterator caused £au Itskm of power to neccesary plant systems

us Loss of RCS Dow (one toopX total loss of RCS flowCRDM probtens, rod dropPlea ilium pressure: low^ughInadvtrtaBt SI stgoalCVCS mil fund up- boron dilution

Lots or reduction m main FW(1 loop}Full or partial closure of one MSIVIncrease e FW flow one loop; all loopsLoss of ooBoensate pump

Sudden epenng of relief valvePressun2cr spray failureSpurious trnr, Manual Inp-.iulo tnp-No transient conditions

break with equivalent area j to 4 3 sq ft liquid suction sidebreak with equivalent area jto4J sq ft liquid duchirse sidebreak with equivalentarea 1 4 u > 4 1 sq ft steambreak with equivalent sffc ' 2 lo J M It Imuidbreak with equivalent area 12 o 1 4 <q ft sieam

break with equivalent area < '.2 •*> ft

FREO DESC FREQ•XMM est event/r vr 1 4E-SMKnnt est. event/r vr 3 6E-2/vroomt esL event/r vr 3 5E-2/yroomt est. event/r ve. 3 OE-3/yrmedian, event/reactor yr 1 OE-7/yrmedian, event/reactor yr 1 OE-4/yrmedian, evenaAeacVX V» 30E-4/vrmedian, events/reactor yi 3 OE-4/yrmedian. events*eactor yi 1 OE-3/yrmedian. evrauAeaooryi 1 OE-3/vrmedian, event/reactor yr 40&6/yrmedian, event/reactor yr 1 OG+l/yr

median, evecls/reactoc yi 1 OB+l^r

4,OE-7Vyrmean eveotAyr 2.3E-4Mmean eventA yr 2.4E-4/yrmean event/r yr 2.1E-2 r

mean event/r vr 1 4E-l/yrmean evou/ryr !X)E-l/yr

mean event/r yr 1 9E«<Vyr

mean eventA yr I 8E-3/yrmean eventA yr 36E-7/yrmean event/r yr 99E-6/yrmean event/r vr 3 9E-5/yrmean evtnt/r vr 52E-S/vrmean event/r vr 90E-S/vrmean event/r yr ZlCJ'vrmean eveniA vr OF>3/vr

138

RjttWIE V*R BPKERBlSttn/a n/aEF3 n/s

EF10 n/aEF3 n/a

90* lognonnal djstr 1 OB-6/yr90* lofcoonnal distr 1 OE-3/jr90% lognormal dis_ 3J)E-3/yr90* logaocmal distr 3J5fi-3/yr90* fefttiormal dotr. l.OE-2/yr90* lognormal distr 1 OE-2/vr

EF10EF2

EF2

iOTORBJ*n/an/an/an/a

lOfcVS/yrlOE-S/yr3-OE-S/yr3£B-S/yr10&4/Win&4/yr

t sdeiicz' '"'''" ' 'Shorebatn PRA. Appendix A 1SborehamPRA. Appendix A 1Shorebam PRA. Appendix A 1Sborehani PRA. Appendix A 1WASH 1400 Reactor Safety Study, Append « V. chapter 4.5WASH-1400. Reactor Safety Study. Appenc x m. table m 6-9WASH-1400. Reactor Safety Study, Append xm. table m 6-9WASH-1400. Reactor Safely Study, Append « HI. table m 6-9WASH 1400, Reactor Safety Study, Appe id i m, table m 6-9WASH-1400, Reactor Safety Study, Appe nd ( HI, Ubie ffl 6-9WASH-1400. Reactor Safety Study, Appe ml < V, chapter 4 4WASH-1400. Reaoor Safety Study, Appe ml tV, chapter 43& Appendix 1, Table I - 4 12

WASH-1400, Reactor Safety Study, Appuxl i x V, chapter 43

OLT SOURCEengineering evaluationestimated from nuclear ooeraunig experiencenuclear operaunEexoenence and engineering eviiujEjonnuclear ooeraunr exoereincebased oo non-nuclear expemocBbased on number of nuclear .industrial and other data sourcesassexmem based on unclear, flkduslrtat and other dau sourcesxtscameM \_*"td 00 Im^lear mdu^nal xiul ot CT d*l* tf^giî

Pî*»ifCTip^T h^wrf «•> mvif^f uutusinal and other data sourcesataoi based on nuclear tadttsm*] and «lber dau sourcesaoahiii of the ivstem laterf ace usut Eeaeric failure ratesMPP operatmf expeneoce, enjioeenng *""""»

^PP opexatinz expcneooe, fflpnf^Tffiff *^fn|«<*

WAS H-1400, Reactor Safety Study, AppcaciitI.RgJ-4 11 *412

n/a n/an/a n/aa/a DA

n/a n/an/a a/a

a/a a/a

B/a a/a

n/a n/an/a n/an/a n/an/a n/an>a n/an/a a/tn/a n/an/a n/a

n/an/an/a

n/aa/a

a/a

a/a

n/an/an/tn/an/an/an/an/a

Calvm Cliffi Una 1 IREP, table 42CitrertCliffil Urut I IREP, ubl«4iCalvm Cliffi Umt 1 KEP, table 4:2

Cirwrt Cliffit Unit 1 IREP. tabte4.ltCalvenCaCfi Unit 1 IREP, table 4.8

Cilvtrt Cli/S Unit 1 IREP, UW*4^

Calvert CMS Una 1 IREP, Utte 4.8

Calvat Cliffi Una 1 IREP, table 43Cat»mCli£Ei Una 1 IREP, table. 4 £Browns Ferry Unit 1 IREP. tabto SBrowns Ferry Unit 1 FREP,ubki5Brownx Ferry Unit 1 IREP. txbie SBrowns Fenr Unit 1 IREP. table 5Brown* Ferry Unit 1 IREP. table 5Browni Ferry Unit 1 IREP, able 5

generic daugeneric dau{eoencdala

EPRINP-2230EPRINP-2230

EPRINP-2230

EPRINP-2230

svstera eraktauoa Him? generic failure daueagmeenn*. evahiauoo aad nuclear exoerienee daugeneric datageneric dauzeaencdatazenencdiuênencdua;enencdata

139

PLANT TYPE IE SET146

147

148149ISO1511521531S4155156157158159160161162163164165166167168169170171172173174175176177178179180181IS21S31841851S6lg?18818919019119219319419519619719819920020!2022032CK205206207208209210211

DWR

BWR

PWRPWRPWRPWRPWRPWRPWRFWRPWRPWRPWRPWRPWRPWRPWRPWRPWRPWRPWRBWRBWRBWRBWRBWRBWRBWRBWR

BWRBWRBWRBWRBWRBWRBWRBWRBWRBWRBWRBWRBWRBWRBWRBWRBWRBWRBWRBWRBWRBWRBWRBWRBWRBWRBWRBWRBWRBWRBWRBWR

BWRBWRBW=«HWRBV.R

tratimjH

tinman

LOCALOCALOCA.LOCALOCAtraBBcatIII .MJIH

tnaxacatn,a»iiltrmaeD!

rramrr*imnniitmnenttamtaf.traaBJeottttaaatn a«aiii intsmmtattraooeatLOCALOCALOCAtranaeottnBBxmtimnxnttnavaeoiMMttslLOCALOCALOCALOCALOCALOCALOCALOCAtramsunseattnawaVMamttnflnentLOCALOCALOCAtnBMBtHUM n«i,,,mmmtram*OC UtuuorLOCALOCALOCA11 ,1,bamont

1l,.,l,llll11 ••» <

uaoacntI I IMKtnonesjtUHBEOt

LOCALOCALOCALOCALOCA

IE CATEGORYloss of PCS

other than Ion of PCS

ta*eLOCAmod mm LOCAsnail LOCA (noouolabte)Small LOCA (bobble)steam Renerator cube ruptureturbine tripIon of main FW (tout loss)Ion of muni FW (partial}excessive FWflowloss of condenser vacuumclosure of one MSIVclosure of all MSIVtcore power excursionste*m line break (inside comimmem)steam line break (outside containment)mim steam relief vthw opemnnjnud vernal SItotal lo*> of reactor coolant flowIocs of off-site powerlarge LOCAmedium LOCAsmall LOCAtransient wiih front Ime svstems availableloss of condenserIce* of FW (total lost)loss of FW and condenserIon of off-str. Dowerreactor vessel rupturelargo LOCAmedaHoLOCAsmaULOCAluge LOCAmedium LOCAsmall LOCAInteriTStem LOCAtraoswni with front line systems availableIon of condenserIocs of FWtots of FW and condenserIocs of off-silt powerhneeLOCAmedium LOCA•nail LOCAnonea with front Ime systems availablelow of condenserIon of FW and condenserloss of off 'Site powerIon of dedicated DC oowerlarge LOCAmedium LOCAsaaJJLOCAtransient with front line svstems availableIon of coodensortonofFWIMS of off-site powermount with front line systems availableloss of cood en serIon of FW and condenserloss of of f site powerlaripLOCAmedmaa LOCAsnuflLOCAIntersvstem LOCApressure vessel ruoturt

fMTIATORSMSJV closureLoss of normal condenser vacuumPressure regulator fills openLots of feedwater flowLoss of off ale powerLoss of auxiliary powerIncreased flow at powerElectric load rejectionElectric load rejection with bypass failureTurbine tnpTurbine tnp with bypass failureIntdverunt opening of MSIVPressure reguatlor fails closedBvpassAxHitrol valve causing pressure increaseRedrculauon control fails causing increased flow

Brealc with equivalent flow >2GOO k^/sBreak with eouivalent flow 3" ra 20(XI k?/sbreax wiih eauivalemfiow 10to301c^/s

brezilc with equivalent area >450 sq cmbreak with equivalent area 80 to 450 sq cmbreak with eouivikot are* < 80 » onbreak with ecuiva lent flow 600 to 2000 kg/sbreak with eouivaleat How 35 to 600 ki^sbreax with eauivalent flow <35 kg/s

break with equivalent now > 2000 Tffjtbreak with equivalent flow 30 to 2000 kg/sbreak with equivalent flow lOto 30 kg/s

break with equivalent flow >3000 k^s

break wi th eouivakrjtflow>1200 kfjibreak with equivalent How 35 to 1200 kg/sbreak with equivalcnrflowSto 35 kR/s

KREQOESCmean evenl/r \r

mean evenyr.yr

mean events/r.vrmean events/r vrmean events/r.vrmean events/r yrmean events/r vrmean eveots/r.vr.mean events/r yrmean eveots/r.yr.mean events/r.vrmean tvents/r.vr.mean events/r vrmean events/r.vr.mean events/r.vrmean eventa/r.vr.mean events/r.vrmean evencs/r.yrmean eventt/ryrmean evenu/r.vrmean events/r vrmean event^r vrmean evenyryrmean event/r vrmean evenl/r.yr.mean evemA'yrmean eveatA-.vr.mean event/r.vrmean tvaxfr.vt.mean event/r yrmean event/r.vrmean event^r vrmean events-TV.mean evened vrmean event^.vr.mean event/r vrmean evencA'.vr.mean event/r.yrmean eventA'.vr.mean event/r.vrmean event/r.vrmean eveatA'.vr.mean event/r.vr.mean eventA*.yr.mem eventA.yr.man event* yr.meai event/r.yr.mean event/r yrmean event/r.vrmean event/r.yrmean eveot/r.yrmean event/r vrmean evenl/r.yrmean evenl/r.yrmeao evem/r.yrmean eve&t/r vr.mean event/r.yr.mean evem/ryrmean event/r.yrmean event/r.yr.mean event/r.yrmean event/r.vr.Tiean evenl/r >T-lean evenl/r vr-lean cveffl/rvr— ean cvenl/r \ r

VREQ1 7E«<Vyr

UEtO/yr

2.0E-d/vr46&4AT40E.3/yrZ3E-2/yr&2E-3/yr10E-0/yr3JE-l/yr4.9E-1/JT31E-l/yr8.1E-2/yr7J2E-2/yrZOE-3/yrlJE-2/yr4.6&4/JT45E-3/JT1.6E-2/yr6.3E-2ryr6.9E-2/JT49E-2/yrJO&4/VT

90E-»/yr30E-3/yr

ZOE+flryr33E-l/yr16E-lryr33&l/yrl.SE-l/vr2.7E-7/yr10E4/yr30E-4ATlOE-3/vrljOE-7/yrl.lE-5/yr1 lE-2/vrlOE-7/vr1.7E-l/yr23E4«/yr17E.l/yr17E-l/yr1SE-1/JTl.OE-7/yr10E-6/yrlX)E-2/yr1.9E+Oryr33E-l/yr35E-l/yr3.2E-l/yr3Ê-l/yrlOE-4/vr50&4/yrlOE-3/yr

1.6E+Ofyr78E-l/vr78E-l/yrrOE-2/vr98E-l/yr16E-l/vr9Ê-l/yr16E-l/yr30&4/yr^ÔE-4/vr30R-3/VTI 9E-7/VT17F,7A-

140

RAJHSE YAK JJPPERBN&n/a

n/a

95% 5% of dismbuuon95% 5% of distnbuuon95% 5% of distribution95%. 5% of distribution95%, 5% of distnbuuon95%, 5% of d«nbuuon95%, 5% of donbuuon95%, 5 % of dtstnbulioc95%, 3% of distnbuuon95% 5%ofdistnbut]on95%, 5% of distnbuuon95%, 5% of distnbuuon95%, 3% of distribution95%, 5% of distribution95%, 5* of distribution95% 5% of distribution95%, 5* of dBtnbuuoo95%,5%ofdatnbuucn95%, J% of dtstnbuuon

n/aa/an/an/an/an/an/an/an/an/an/an/an/an/an/an/an/an/»n/an/tn/an/an/an/an/an/in/an/an/aO/I

n/aaft.altaftn/an/aalt.n/an/an/an/an/an/irt/«n/a

n/a

n/a

52E^/vrl.2E-3/yr14E,0/vr50E-2/yrilE-2/yr14E4ryr4.9E-l/yr6-SE-l/yr49E-l/yr!_5E-l/yrl-5E-l/yr57E-3/yr3.6E-2/yrliE-3/yrlJE-2/yr44E-2/yr1.2E.1/1TIJE-lVyrHE-l/yr

n/an/an/an/an/an/an/an/an/an/an/an/an/an/an/an/an/an/an/an/an/an/an/an/*n/aeftn/an/an/an/an/an/an/an/an/an/an/an/an/an/an/an/tn/an/an/a

LOWER DH& SOURCEn/a

n/a

76E-6/yrUE-5/yr1.2E-4/yr336-Vyr3,lE-4/yr/ Icrlfyr1 7&-l/yr2.9E-lfrr146-1/yr1.8&2/VT18E-2/VT39E-S/yr6.7&4/yr23£-5/yrHE4/yr5,7&4/yrljtE-2/yrl.SE-2/yr71E-3/yr

n/an/an/an/an/in/an/an/an/an/in/an/an/an/an/an/an/sn/an/attMn/aHfea/an/an/aa/an/an/an/anAln/an/an/anAn/an/an/anAn/an/an/anAn/an/anA

Browns Ferrv b'ml 1 IRCP table 6

Browns Ferry Unit 1 [REP, table 6

OldPWROldPWROldPWROldPWROldPWROtdPWROSdPWROldPWROldPWROtdPWROtdPWROldPWROMPWROldPWROidPWROldPWROWPWROWPWROldPWRBaraebtckl A2NPPBarseback 1 & 2 NPPBarsebadc 1 & 2 NPPBarsebadc 1 NPPBaneback:' NPPBarsebadc 1 NPPBaraebadc 1 NPPBarsebadc 1 NPPForsmaA 3 NPPFonm«rk3NPPForsm*rfc3NPP

OsJarahamtt 1 NPPOskarsbamn 1 NPPOjkarsnanui 1 NPPOjtamhamn 1 NPPCukarshamn 1NPPOsxanhaxna 1 NPPOsxanhamn 1 NPPOfkanhamtt 1 NPPOskanhamn 1 NPPO(kmtianii2NPPOflianhanul2NPPO*kanhasnn2NPPOfkarthamn2NPPOfkxniuBmi2NPPOsksnfaamn2NPPOskarshamn2NPPOskanbamtt 2 NPPOfkanhaaa 3 NPPOjVarshamn 3 NPPOfkanhamn 3 NPPOilcarshamn 3 NPPOstarahamn 3 NPPOsksnhanm 3 NPPOskannamn 3 NPPaanebnclc 2 VPPBane*** 2 VPP&ir*o»ck2.VPPB»»*Kk2NTPRinidials 1 VPPRmshaU 1 VPPRinghals 1 VPPR instills i vppRinihali i -??

OE/rSOtRCBplant specific oau

plant specific oau

Generic dau updated wnh plant specific operating experienceGeneric cau updated with plant specific operating expenenceGenenc dau undated with plant specific operatmt experienceGenenc data updated w*a place specdicoperainiiexpeneBceGenenc dau updated w«h plant specific operates expenenceGenenc dau updated with plant specific operauit expenenceGenenc dau updated with plant specific operataic experienceGenenc dau updated w«h otant specific operaun expenenceGeneric dau updated with plant specific operating expenenceGenenc dau updated with plant specific opentsit expenenceGenenc dau updated wmh plant specific optratxtg expenenceGenenc dau urxutted w«h plant specific opeiiuu,. r^enaxxGeneric data updated wah plant specific cperatxtg expenenceGenenc dau updated with pUnt specific oocratfix expeneoceGeneric dau updaled wmi plant specific operatmt expenenceGenenc data updated wtth plant specific opencsot expenenceGeneric data updated wtth plant specific opcntat expenenceGenenc data updated wxh piaot specific ooeraunt; expenenceGenenc dau updated with plant awafic aperxnt expenencegeneric sources1 iterarure sourcesliterature dataplant operating expenence (8 1 1 years)pbuu speci^c ooeratm; expenence (8 1 1 yean)plant specific operating expenence (8 1 1 yean)point specific openuiK expeneace (811 yean)piant specific operating expenenceliterature sourcesliterature sourcesliterature sourcesliterature source*Ittearure sources (application of LBB cntena)literature sourceslitcarture sources.literature sourcespfeni operainR experience (5 7 yean)plant operatmx cxpeneoce (5.7 yean)pta specific operatmg expenence (5 7 yean)plant specific operatnj expenence(5 7 yean)plant speafc opentstg expenence (5 7 yean)Ijterarorft source*literatim fourceaItteratnrc aourccaphmt operatatiE npcneoce (*,2 van)plant specific openuu expcneact (62 yean)plant specific opcratxtt: expenence (62 years)plant specific experienceplant specific OMratnt, expetreoce (612 years)literature sourcesliterature sourcesliterature so Jrcesplant specific operauif. expenence (6 4 yean)plant specific operatu! expereince (6.4 man)plant specific oeenuig exDeneace (6 4 yean)literature sourcesplan specific oocntzix apeneDceplant specific opentnc expert eacapbua specific oocraun expeneaceplant soeftf-e ooeratag expenencelueaiture so ircesliterature so jccslitenrure so -ccsIiLeruurc <i,_rccstucrtturc cu

141

2122!32U21521621721821922022!222223224225226227228229230

231

232

233234235

236

237

238239240

241

242

243

244245246247248249250251

JPtANTTYPE.1WRBWRnwRBWRPWRPWRPWRPWRPWRPWRPWRPWRPWRPWRPWRPWRPWRPWRPWR

PWR

PWR

PWRPWRPWR

PWR

PWR

PWRPWRPWR

PWR

PWR

PWR

PWRPWRPWRPWRPWRPWRPWRPWR

IESEEV 'traat»«•jmneottramaeatransits*LOCALOCALOCALOCALOCALOCAtnodoKnii i i iBtraakattraacnttnoBctttraniottranneMrfMnitnrLOCA

LOCA

LOCA

LOCAtnncattlXBBMSA

tnamat

trancMKtrancie*

transie*

tranmt

tram

— Taan iCCbUMrLOCALOCA[.OCALOCALOCAuanae*

• IE CATEGORYswsient with front line svsteras availableIocs of condenser'ass 01 FW 4 condenserloss ot otf site powerlam LOCAmedium LOCAsmall LOCAreactor vessel ruptureIntercvstem LOCAsteam generator tube rupturetransient wnh PCS isolationtransient with front line systems availablelots of off site powertransient requjnng RCS pressure reliefmadveram safety iniectionsteam line breaktransient occurring alter shutdownloss of service water systemlute LOCA

medium LOCA

small LOCA

Kara generator tube ruptureloss of RCS flowiocs of feed water flow

partial loss of steam flow

ttlrbffle tnp

Iocs of off site Dowerspurious safety tniecuoareactor tnp

toss of steam inside containment

Iocs of steam outside containment

core power increase

Iocs of component coolingloccof service watersmall LOCAstem generator tube rupturemedium LOCAUnteLOCAmtersvstem LOCAmam neam line nreat

IMTIATORS

break with eoutvaient diameter greater than 15cmbreak with ecuivalenc diameter between 5 nad 15 on

Break with equivalent diameter greater than 6 incheiRPV failureBreak with equivalent sizes between 2 and 6 inchesMultiple pressurizer safety and relief valve failureBreak with equivalent diameter smaller than 2 incheiPressunzer safety and relief valve failureCRDM failuresRCP seal failure

FW ptpe rupture outside containmentLoss / reduction of FW Dow in one SOLoss of FW flow to all SOFW flow instability - operator error

Loss of one condessate pumpLoss of all coodensale pumpsCondenser leakageOther secondary leakageMSIV ClosureFuU ctoswe of MSIVPartial closure of MSIVOther losses of sieam flowClosure of all MSIV-sIncrease of FW flow m one SC

Loss of CWThrottle valve dosure/Electrohydniilic control problemGenerator trip / generator caused faubsTurbine tnp due to over speedOther turbine trios

CRDM problem / rod drop*High and ]ow pressunzer pressureHigh pressunzer levelSpunous automatic mp - no transient condincoAutomatic/manual nip - operator errorManual tnp due to false ngoalsSptnouc tnp - cause unknownPrimary system premre temperature or power unbalance

FREQDESG.mean event/r.vrmean event* vrmean event/r vrmean event/r.vrmean eventA.yrmean ervent/r.vrmean event/r.yrmean event/r vr.mean eveaUr.yr.mean evenWr.yr.mean event/r.vr.mean event/r.yrmean event/r.yr.mean event/r.yrmean event/r.yr.mean eventA.yr,mean event/r.yrmean event/r.yr.mean eventi/r.yr

mean event/ryr.

mean event/r.yr

iz-m event^.yrmean eventA'.yr.mean event/r.yr.

mean evefll^.yr.

mean evealfr yr.

mean eventA'vrmean evedt/ryrmean evcct/r.yr

FREO9.2E.1/W20E+«yvr3.7E-; >Tl^E-l/vr40E4/yr8.t&4/yr1 lE-2/yr2-7E-7/yr42&«/yr97B-3/VTUE-l/yr6.0E+Cryr70&l/yr40E-l/yrliE+<Vyr44&4/yrlOE*<Vvr2.4&5/yr9.4&4/yr

°4&4ryr

3JE-2^r

2.4E-2/yr3.6E-lfrr

2JE-lryr

3.7Et<Vyr

58E-2/yr6.4E-l/yr

Steam Pipe rupture inside containment mean event^yr. 9 ^ft-4r¥rFW pipe rupture inside containmentSieam relkcf or safety valve open inadvenanlly (leak, upstream of MSIV)Other steam losses inside containmentSteam pipe rupture outside coalauuneniThrottle valve/Electrohydraultc control problemsSteam and dump valve fail openOther steam losses outside containmentUncontrolled rod withdrawalBoron dilution - CVCS malfunctionCore vita, temperature dropOther oocitive reactivity additions

mean eveot/ryr

mean rventA yr.

mean cvenLt.vr.m em even Ur.yrmean eventAyrmean event/ryr.mean evenlfr.vr.mean eventAyrnean evenl^yrinean eventA ^r

--»

2JE-2/yr

94&4/yr94M/WJJE-2/yr37E-2NT94E4/vr94E-4*vriOE-Dvr3 9F,2AT

142

n/an/arjia/a95*95*95*95*n/a95*95%95%95%95%95%95% •95%n/a

95%. 5%

95%, 5%

95%, 5%

95%, 5%95% .5%95%.5%

95%, 5%

95%, 5%

95%. 5%95%. 5%95*. 5%

95%, 5%

95%, 5%

95%, 5%

95%. 5%95%. 5%

n/an/.a/a

n/an/an/a

n/an/an/an/a

lJZE-3/yr3.0B-3/yr2_5E-2/yr1.0E-6ryr

n/a10E-2/yr4.0E-l/yr9.66-tO rliE+O/yr8.6E-1/JTlOE+Ofrr1.2E-3/yr].7E+0/yr

n/a3.<E-3/yr

3.6E-UVJT

7.2E-2/JT

7.7E-2/yr6.0E-l/yr

5.2E-l/yr

4.7E+0^yr

1.7E-l/yrUE+Wyr4.7£+Oftr

3.6E-3/yr

3.6E-3/JT

6.1E-2/yr

3.6E-3/yr3.oE-3/yr

n/an/an/an/an/an/a

n/an/an/an/an/an/an/an/an/aa/an/an/an/an/an/an/anAa/a

3JE-5/yr

^E-5/yr

UE-2/yr

2.8E-3/yr1.9E-l/yr

4.1E-40frr

9.6E-2Vyr

23E-tO/yr

8.3E.3/yr3.3E-l/yrX9E+0/yr

3JE-5/yr

3.3E-5/yr

4.6E-3/yr

3.3E-5/yr3.3E-5/yr

n/an/an/an/aa/an/i

RingDilI 1 NPPRintfiaii 1 NPPRingiali 1 NPPRan (Hah 1 NPPRinnWi2NPPRinRhali2NPPRinghali2NPPHianiaU2NPPRiaRtal»2NPPRimiaJs2NPPRjnuhalsINPPRinshals2NPPRingluls2NPPRiraghalslNPPRJUg)l»lj2NPPRini*al»2NPPRiiJl*alj2NPPRicshilt2NPPZioa NPP PSS table 1.1. 1-2

Zico NPP PSS able 1.1.1-2

Zion NPP PSS able 1.1.1-2

Zion NPP PSS able 1.1.1-2Zion NPP PSS able 1.1.1-2Zion NPP PSS able 1.1.1-2

Eon NPP PSS table 1.1.1-2

Zion NPP PSS table 1.1.1-2

Son NPP PSS table 1.1.1-2Zion NPP PSS table 1.1. 1-2Zion NPP PSS table 1.1.1-2



Zion NPP PSS uHe 1.1. 1-2

Zion NPP PSS able 1.1. 1-2Zion NPP PSS able 1.1.1-2Align NPP PSA, Summary reportAngra NPP PSA. Summary reportAngra NPP PSA, Summary reportAlign NPP PSA, Summary reportAnan SPP PSA. Summary reportAnisra NPP PSA. Summary report

olant specific operating experience (5.44 vears)3lamsoecif!cooemhigexi)erience(5.d4vears! Iplant soecillc operating experienceatant specific ooenting experience (5.4d years)literature souredliterature sourcesliterature sourcesliterature source*

lilearture sourcesplant specific operating experienceplant specific operating experienceplant specific operating experienceplant specific operating experienceplant specific operating experienceLiterature sourcesplant specific operatin!; experience 'literature sources

literature sources aod plant operating experience

literature sources and plant operating experience


literature sources and plant opentinR experienceliterature sources and plant operating experienceliterature sources (prior) and plant operating experience

literature sources (prior) and plant operating experience

literature sources (pnor) and plant operating experience

literature sources and plant operating experienceliterature sources and plant operating experienceliterature sources and plant operating experience



literature sources tad plant operating experience

literature sources (prior) tnd plant operating experienceliterature sourcesjjjrior) and riant operating experienceliterature source!literature sourcesliterature sources |literature sources 'literature sources •literature sources '

143

252:S32542552567^7232597,60261?fi?.263264

265

266

267268

269270

271

272

PLANT TYPEFWR"WRPWRPWRBWRBWRBWRBWRBWRBWRBWRBWRBWR

BWR

BWR

BWRBWR

BWRBWR

BWR

BWR

IB SETtransienttransienttransienttransientLOCALOCALOCALOCALOCALOCALOCALOCAtransient

transient

traCSttnl

transientCC Initiator

CCInuuior

CC Initiator

rer CATEGORY<oif of off-site oowerM'.s at main fccuwatcrreactor tnoturbine trmlarge LOCA made drvwellUrge LOC\ outside drvwellmedium LOCAsmsllLOCAintersvstem LOCA-LPd breiklatersvstem. LQCA-CS breakECCS breaksreactor vessel runturemanual shutdown or spurious \nf

mrbine tnp

MSIVdonure

madvenant ooenmg of relief valveloss of fccdwtuf

losiof off site txjwtrlouofTBCCWsyatm

ton of muniment iff

lost of 6k V emerqencv bus

INITIATORS

brealcsiz* groner than 3sqf tbreat sue ^TMter thin 3sqf tbrralcsizej between. 1 and 3 sq ft<ae»n)uid OCWand 2riiouid)brtik sue uo to Isq ft fsteun)«- (XM sg f t f l iouid)

Spunna tnp vu instrumenuuicoScran due to oiim occurcncesDdcaed faatl in RPSInjdvertint insertion of rod or rodsMsnual shutdownTubmetnpGeoemof tripHigh FWdunngsnrliiporihutdownfrudverumt tortup of HPdRtorculiUDO ccftuoi fmkire imeuint flowTurbme bypast or cxvtrot vitve futuresTnp of we at ill recirculauon pump!Lora of FW haterRtcuuibnmo pump seizurePressure regulator full openTurbine bypca fills openFW flow atcrasmg « powerFW coctrollc muctmum doatndMSIYdoaoeFarmaMSIVdoAireludverum doture of ooe MSIVLoss of condenser vacuumTurbine tnp with bypus fadureGeoerautr trip with bypass failurePressure regulator fills closed

Loss of all FW nowTnp of ooe of FW pumpsFW low FlowLow FW flow during surtup or shutdown

Loa of cool mg wtter to TBCCW sysamLoss of coodeosate pumpsLws of service tnaaLess of mstrumnit anLoss of RBCCW system

FREQ DESCTiein t;vent/r vr•nesn evenur\rmean cveni/r vrTiean evenlA-vr

mean event/r vrmean eveiUr vtraenn event/rvrmean evenUr vrmean eventyr vrmean eventAvrmean event/rvrmean event/ryrmem eventAyr

mem eveaiVr yr.

mean evml/ryr

mean tvem/r yrmean eveat/r yr

mean event/ryrmean event/r yr.

mean event/ryr

mean events yr

FREQ'51>l/vr>^C-nWr> lE+IVvr30E+IVVT2.7MVvrlOM/yr17E-3/yr17E-2/W73E*vr13&7Nr13B-5/VT30&7/yr4JE+Oryr

ZTE^flryr

41E-l^yr

13E-7JyrllE-l/yi

6.7E.7Jyr3OE-2/VT

l^&3/yr

1 1&4M

144

- RANGE VARnit•uaaltalt

altBit

altn/«n/in/an/in/in/«

nit

aft

aftti*

altalt

alt

alt

UPPER BNB'-yj•lititalt!>/*

n/iaftaltaltnita/tnitn/»

alt

alt

afta/t

altalt

alt

sit

tOWERBM*adini/an/aD/>

nAn/a

rut1/1n/ac/»n/ia/a

0/1

nit

n/>Hh

nAnA

n/a

n/t

SOtJRCBAniiraNPP PS A. Sumnurv reoort\n?n N PP PS A. Sumnurv retwrtAngn NPP PS A. Summuv rrooriAngri NPP PS A. Summirv reoonC2or»oNPPPSSubl*A-2CaorsoNPPPSSwbleA-2CaorsoNPPPSSuWeA-2Ciono NPP PSS able A-2OK»» NPP PSS able A-2CaonoNPPPSSubleA-2C»e»»)NPPPSSub]eA-lCi«rsoNPPPSSt»We3SOKXSO NPP PSS able A-2

Oorao NPP PSS able A-2

C*or» NPP PSS able A-2

CioreoNPPPSSuWeA-2aarsoNPPPSSiibleA2

Ciono NPP PSS able A-2C»orao NPP PSS ttble A-2

OOTJO NPP PSS able A-2

C»onoNPPPSAul)teA-2

-. UtT SOURtttfreQucncv bxscd on soecilic ^udviuemurt SOUTHSlacniure source*iilerature source!luenture sourcesliterature sourcalaenioretowcetIwntunnurceililenQiiedtUlueruurt sourcesliterature source*Ijtermire jourcaplant jpeaftt openuag experience

pluu specific; openimj opereince

ptua specific cperumj npencacc

smdvfor the clMrtwuhsmUr valves (AttoUziolplam speo/ic openung rapmcBce

Historical data for rtonti tlalvanaly$u uang fault tree method

aulystf of ouuumeQi air system

inalysii of enier^encv power system bv fau It tree

145

REFERENCES

[I] INTERNATIONAL ATOMIC ENERGY AGENCY, Procedures for ConductingProbabilistic Safety Assessments of Nuclear Power Plants (Level 1), Safety Series No.50-P-4, IAEA, Vienna (1992).

[2] NUCLEAR REGULATORY COMMISSION, Reactor Safety Study: An Assessmentof Accident Risks in US Commercial Nuclear Power Plants, Rep. WASH-1400-MR(NUREG/75-014), Washington, DC (1975).

[3] LEVERENZ, F.L., et al., ATWS: A Reappraisal, Part 3: Frequency of AnticipatedTransients, Rep. EPRI-NP-801, Palo Alto, CA (1978).

[4] McCLYMONT, A.S., POEHLMAN, B.W., ATWS: A Reappraisal, Part 3:Frequency of Anticipated Transients, Rep. EPRI-NP-2230, Palo Alto, CA (1982).

[5] MACKOWIAK, D.P., et al., Development of Treatment of Initiating EventFrequencies for Use in Probabilistic Risk Assessments, Rep. NUREG/CR-3862,EG&G 2323, Idaho National Engineering Laboratories, Idaho Falls, ID (1985).

[6] ILBERG, D., et al., A Review of the Shoreham Nuclear Power Station ProbabilisticRisk Assessment, Rep. NUREG/CR-4050, USNRC, Washington, DC (1985).

[7] NUCLEAR REGULATORY COMMISSION, Severe Accident Risks: An Assessmentfor Five US Nuclear Power Plants, Final Summary Report, Rep. NUREG-1150, 3volumes, Washington, DC (1990).

[8] KAPLAN, S., On a Two Stage Bayesian Procedure for Determining Failure Ratesfrom the Experimental Data, Rep. PLG-0191, Pickard, Lowe and Garrick, Irvine, CA(1981).

[9] ELECTRIC POWER RESEARCH INSTITUTE, A Probabilistic Risk Assessment ofOconee Unit 3, Rep. NSAC-60, EPRI, Palo Alto, CA (1984).

[10] WYCKOFF, H., Losses of Off-site Power at US Nuclear Power Plants through 1983,Rep. NSAC-80, EPRI, Palo Alto, CA (1984).

[II] NUCLEAR REGULATORY COMMISSION, Operating Units Status Reports (referredto as 'Gray Books'), published quarterly as USNRC NUREG-0200.

[12] SCIENCE APPLICATIONS, INC., Probabilistic Risk Assessment: Shoreham NuclearPower Station, Rep. SAI372-83-PA01, SAIC, NY (1983).

[13] CONSUMERS POWER COMPANY AND PICKARD, LOWE AND GARRICK,Midland Nuclear Plant - Probabilistic Risk Assessment, Irvine, CA (1984).

[14] PICKARD, LOWE AND GARRICK, Seabrook Station Probabilistic SafetyAssessment, Rep. PLG-0300, Irvine, CA (1983).

[15] HANAN, N., et al., A Review of the Oconee-3 PRA, Rep. NUREG/CR-4374,USNRC, Washington, DC (1986).

[16] ILBERG, D., A Review of initiating events selection and frequencies for PSA Studies,Evaluation of Reliability Data Sources, IAEA-TECDOC-504, IAEA, Vienna (1989).

[17] BABCOK AND WILCOX, Owners Group Probabilistic Evaluation of PressurizedThermal Shock Phase 1, Internal Report, Rep. BAW-1791 (1983).

[18] KOLB, G. J., et al., Interim Reliability Evaluation Program: Analysis of the ANO-Unit1 Nuclear Power Plant, Rep. NUREG/CR-2787, Sandia National Laboratories,Albuquerque, NM (1982).

[19] KOLACZKOWSKI, M., et al., Analysis of Core Damage Frequency From InternalEvents: Peach Bottom, Unit 2, Rep. NUREG/CR-4550, Vol.4, SAND86-2084, SandiaNational Laboratories, Albuquerque, NM, and Livermore, CA (1986).

[20] PHILADELPHIA ELECTRIC COMPANY, Probabilistic Risk Assessment: LimerickGenerating Station, Revision 4, Philadelphia, PA (1982).

146

[21] GENERAL ELECTRIC COMPANY, 3WR-6 GESSAR PRA (1982).[22] GESELLSCHAFT FUR ANLAGEN- JND-REAKTORSICHERHEIT (GRS) mbH,

German Risk Study Phase B, GRS-72, Verlag TUV, Cologne (1989).GESELLSCHAFT FUR ANLAGEN- JND-REAKTORSICHERHEIT (GRS) mbH,German Risk Study Phase B, Vol. 1 -- GRS-A-1600, Verlag, TUV, Cologne (1989).

[23] ELECTRICITE DE FRANCE, Paluel ^PP PSA, EDF, Paris (1990).[24] SCHOLL, R.F., Loss of Off-site Powf r Survey Status Report, Revision 3, Report of

the Systematic Evaluation Program Branch, Division of Licensing, USNRC,Washington, DC (1980).

[25] McCLYMOND, A.S., POEHLMAN, ] t.W., Loss of Off-site Power at Nuclear PowerPlants: Data and Analysis, Rep. EPRt-NP-2301, Palo Alto, CA (1982).

[26] BARANOWSKY, P.W., Evaluation oi Station Blackout Accidents at Nuclear PowerPlants, Rep. NUREG-1032, US Nucl< ar Regulatory Commission, Washington, DC(1988).

[27] IMAN, R.L., HORA, S.C., Modelling Time to Recovery and Initiating EventFrequency for Loss of Off-Site Powar Incidents at Nuclear Power Plants, Rep.NUREG/CR-5032, USNRC, Washing >n, DC (1988).

[28] NUCLEAR REGULATORY COMM1S SIGN, Probabilistic Risk Analysis: ProceduresGuide, Rep. NUREG-2300, Washing* n, DC (1983).

[29] BLOMQUIST, R., PORN, K., SUPEE -ASAR: Initiating Events in Swedish NuclearPower Plants, Classification and Qualit itive Comparisons, Rep. Studsvik/NP-87/103,Nykoping (1987).

[30] MAYS, S.E., et al., Interim Reliabilit; Evaluation Program: Analysis of the BrownsFerry, Unit 1, Nuclear Plant, Rep.' NUREG/CR-2802, EG&G, Idaho NationalEngineering Laboratory (EG&G, Idaho, Inc.), Idaho Falls, ID, and EnergyIncorporated, Seattle, WA (1982).

[31] YANKEE NUCLEAR SERVICES DIVISION, Yankee Nuclear Power StationProbabilistic Safety Study, Framinghwi, MA (1980).

[32] HOERTNER, H., KAFKA, P., IEICHART, G., The German PrecursorStudy — Methodology and Insights, Roliab. Eng. Syst. Saf. 27 (1990) 53.

[33] DROUIN, M.T., HARPER, F.T., CA1 IP, A.L., Analysis of Core Damage Frequencyfrom Internal Events: Methodology Guidelines, Rep. NUREG/CR-4550, Vol. 1,SAND86-2084, Sandia National Laboratories, Albuquerque, NM (1987).

[34] HATCH, S., et al., Reactor Safety Sti c y Methodology Applications Program: OconeeUnit 1, PWR Power Plant, Rep. NUR iG/CR-1659, SAND80-1897, Sandia NationalLaboratories, Albuquerque, NM (1981).

[35] DEUTSCHE RISIKOSTUDIE KERNF RAFTWERKE GRS, Verlag TUV Rheinland,Cologne (1980).

[36] DROUIN, M.T., et al., Analysis of C ore Damage Frequency From Internal Events:Grand Gulf, Unit 1, Rep. NUREG/CR- 4550, Vol. 6, SAND86-2084, Sandia NationalLaboratories, Albuquerque, NM, and '. jvermore, CA (1987).

[37] HUBBLE, W.H., MILLER, C.F., Data Summaries of LERs on Valves at USCommercial Nuclear Power Plains, Reps NUREG/CR-1363 (1980) andNUREG/CR-1963 (1985), USNRC, Washington, DC.

[38] FLEMING, K.N., Realistic Assessment of Interfacing Systems LOG A Scenarios inPWR Plants, Proc. Int. Topical Meeting on Probabilistic Safety Assessment and RiskManagement, Zurich (1987) and Verla i TUV, Rheinland, Cologne (1987).

[39] CHU, T.L., et al., Interfacing System:, LOCAs at Boiling Water Reactors, Proc. Int.Topical Meeting on Probabilistic Sale y Assessment and Risk Management, Zurich(1987) and Verlag TUV, Rheinland, C ologne (1987).

147

[40] BERTUCIO, C., et al., Analysis of Core Damage Frequency From Internal Events:Surry, Unit 1, Rep. NUREG/CR-4550, Vol.3, SAND 86-2084, Sandia NationalLaboratories, Albuquerque, NM, and Livermore, CA (1986).

[41] PAYNE, A.C., et al., Interim Reliability Evaluation Program: Analysis of the CalvertCliffs Unit 1 Nuclear Power Plant, Rep. NUREG/CR-3511, Vols. 1 and 2, SandiaNational Laboratories, Albuquerque, NM and Livermore, CA (1984).

[42] CONSOLIDATED EDISON, Indian Point Probabilistic Safety Study, Power Authorityof the State of New York and Consolidated Edison, NY (1982).

[43] WHEELER, T.A., Analysis of Core Damage Frequency From Internal Events: ZionUnit 1, Rep. NUREG/CR-4550, Vol. 7, Sandia National Laboratories, Albuquerque,NM (1986).

[44] COMMONWEALTH EDISON COMPANY, Zion Probabilistic Safety Study, Chicago(1981).

[45] NORTHEAST UTILITIES SERVICE, Maine Yankee Probabilistic Safety Assessment,Hartford (1989).

[46] CONSUMERS POWER COMPANY, Big Rock Point Nuclear Power PlantProbabilistic Risk Assessment (1981).

[47] HAKE, T.M., WHITEHEAD, D.W., Initiating Event Analysis for a BWR Low Powerand Shutdown Accident Frequency Assessment presented at the PS AM '91, LosAngeles (1991).

[48] ILBERG, D., HANAN, N., An Evaluation of Non-isolated LOCA Outside theDry well in a BWR, Brookhaven National Laboratory, Technical Report A-3740,Upton, NY (1985).

[49] LANORE, J.M., CHAMP, M., Probabilistic Safety Assessment of French PWRs:Importance of Non-Power States, Use of Probabilistic Safety Assessment forOperational Safety (Proc. Symp. Vienna, 1991), IAEA, Vienna (1992).

[50] SHEPHERD, C.H., Sizewell-B PWR Preconstruction Safety Report, Reference Report— Initiating Fault Frequencies, National Nuclear Corporation, Whetston (1982).

[51] ILBERG, D., HANAN, N., A Review of the Turbine-Building Flooding Study of theOconee PRA, Proc. Int. ANS/ENS Topical Meeting on Thermal Reactor Safety, SanDiego (1986).

[52] BOARDMAN, T., Leak Rate Analysis of the Westinghouse Reactor Coolant Pump,Rep. NUREG/CR-4294, Washington, DC (1985).

[53] KIMURA, G.Y., et al., Evaluation of External Hazards to Nuclear Power Plants inthe United States, Rep. NUREG/CR-5042, Lawrence Livermore National Laboratory,CA (1987).

[54] Treatment of External Hazards in PS A, International Atomic Energy Agency, Vienna,Austria (in preparation).

[55] MINARICK, J.N., KUKIELKA, C.A., Precursors to Potential Severe Core DamageAccidents: 1969-1979 — A Status Report, Rep. NUREG/CR-2497, Washington, DC(1982).

[56] Connecticut Yankee (Haddam Neck Plant) Probabilistic Safety Study, NortheastUtilities Service Company Report, NUSCO-149, Hartford, CA (1986).

[57] ENTE NAZIONALE PER L'ENERGIA ELETTRICA, Probabilistic Safety StudyCaorso NPP, Rep. ENEL DCO 401.V040,VR001, ENEL-DCO, Rome (1988).

[58] ARRIETA, L.A.I., ANGRA-1 Probabilistic Safety Study, Final Report, CNEN, Riode Janeiro (1988).

[59] IAEA Regional Programme on PS A for WWER Reactors, Report of a Workshop inRez, (1992).

[60] BEQE, Top Level Risk Study for Kozloduy Units 1 to 4, Report No. 58-01-R-001(1992).

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LIST OF AB BREVIATIONS

AC Alternating currentAFW Auxiliary feedwaterATWS Anticipated transient without sci amBWR Boiling water reactorB/W Babcock and Wilcox Co.CC Component coolingCCI Common cause initiatorCCW Component cooling waterCDF Core damage frequencyCFS Containment fan systemCW Circulating waterCRD Control rod driveCRDM Control rod drive mechanismCRW Control rod withdrawalDC Direct currentDHR Decay heat removalEBFT Energy balance fault treeEHC Electro-hydraulic controlFMEA Failure mode and effect analysisFSAR Failure safety analysis reportFTA Fault tree analysisFW FeedwaterGPM Gallon per minuteGRS German Risk StudyHPCI High pressure core injectionHPCS High pressure core sprayHPI High pressure injectionICS Integrated control circuitIE Initiating eventIORV Inadvertent open relief valveISLOCA Interfacing system LOCALER Licensee event reportLOCA Loss of coolant accidentLOFW Loss of feedwaterLOOP Loss of off-site power/loss of sU tion powerMFW Main feedwaterMSIV Main steam isolation valveMLD Master logic diagramMOV Motor operated valveNPP Nuclear power plantPCS Power conversion systemPORV Power operated relief valvePWR Pressurized water reactorPRA Probabilistic risk assessmentPSA Probabilistic safety assessmentRCS Reactor coolant systemRCP Reactor coolant pump

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RHR Residual heat removalRPS Reactor protection systemRPV Reactor pressure vesselRV Relief valveSG Steam generatorSIS Safety injection signalSRV Safety relief valveWWER Soviet designed PWR (water moderated, water cooled reactor)

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