NUREG-1407

Procedural and Submittal

Guidance for the Individual

Plant Examination of External

Events (IPEEE) for Severe

Accident Vulnerabilities

Final Report

U.S. Nuclear Regulatory Commission

Office of Nuclear Regulatory Research

J. T. Chen, N. C. Chokshi, R. M. Kenneally, G. B. Kelly,

W. D. Beckner, C. McCracken, A. J. Murphy, L. Reiter, D. Jeng
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NUREG-1407

Procedural and Submittal

Guidance for the Individual

Plant Examination of External

Events (IPEEE) for Severe

Accident Vulnerabilities

Final Report

Manuscript Completed: April 1991

Date Published: June 1991

J. T. Chen, N. C. Chokshi, R. M. Kenneally, G. B. Kelly,

W. D. Beckner, C. McCracken, A. J. Murphy, L. Reiter, D. Jeng

Division of Safety Issue Resolution

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

ABSTRACT

Based on a Policy Statement on Severe Accidents, the

licensee of each nuclear power plant is requested to perform an individual plant examination. The plant examination systematically looks for vulnerabilities to severe accidents and cost-effective safety improvements that reduce

or eliminate the important vulnerabilities. This document

presents guidance for performing and reporting the results of the individual plant examination of external

events (IPEEE). The guidance for reporting the results of

the individual plant examination of internal events (IPE)

is presented in NUREG-1335.

i°. NUREG-1407

CONTENTS

Page

Abst ract............................................ iii

Executive Sumay..................................... y.... ix

Acknowledgements ................................ .... .... xiii

1 Introduction ........................................................................ 1I

1.1 Background . . . . . . . . . . . . . . . . .. . .. . . . . . . . . . . . . . . . . .. 1

1.2 IPEEE Objectives ................................................................ 1

1.3 Purpose of Document ............................................................. 2

2 Events Evaluated for Inclusion in the IPEEE................................................ 3

2.1 Seismic Events .................................................................. 3

2.2 Internal Fires *....................................................... 3

2.3 High Winds and Tornadoes ......................................................... 4

2.4 External Floods................................................................... 4

2.5 Transportation and Nearby Facility Accidents............................................ 4

2.6 Lightning....................................................................... 4

2.7 Severe Temperature Transients (Extreme Heat, Extreme Cold)............................... 5

2.8 Severe Weather Storms ............................................................ 5

2.9 External Fires (Forest Fires, Grass Fires) ............................................... 5

2.10 Extraterrestrial Activity (Meteorite Strikes, Satellite Falls) .................................. 5

2.11 Volcanic Activity ................................................................. 5

2.12 Summary....................................................................... 5

3 Acceptable Methodologies for Performing the Seismic TPEEE.................................... 6

3.1 Seismic PRA .................................................................... 6

3.1.1 New Seismic PRA Analysis .................................................... 6

3.1.1.1 General Considerations................................................ 6

3.1.1.2 Hazard Selection .................................................... 7

3.1.1.3 Fragility Estimation................................................... 7

3.1.1.4 Seismic PRA Methodology Enhancements.................................. 8

3.1.1.5 Containment Performance ............................................. 8

3.1.2 Use ofan Existing PRA ...................................................... 8

3.2 Seismic Margin Methodologies....................................................... 10

3.2.1 General Considerations........................... .................. 10

3.2.2 Review Level Earthquake and Associated Spectral Shapes............................. 11

3.2.3 Reduced-Scope Margins Method ................................................. 12

3.2.4 NRC Seismic Margins Methodology ............................................. 12

3.2.4.1 Walkdown ......................................................... 12

3.2.4.2 Relay Evaluation .................................................... 12

3.2.4.3 Soil Failures........................................................ 13

3.2.4.4 Screening Criteria (Use of Screening Tables)................................ 13

3.2.4.5 Seismic Input....................................................... 13

3.2.4.6 Evaluation of Outliers-HCLPF Calculations ............................... 13

3.2.4.7 Nonseisniic Failures and Human Actions................................... 14

3.2.5 EPRI Seismic Margins Methodology ............................................. 14

3.2.5.1 Selection of Alternative Success Paths............................... &..... 14

3.2.5.2 Walkdown ......................................................... 14

3.2.5.3 Relay Evaluation .................................................... 14

3.2.5.4 Soil Failures........................................................ 14

3.2.5.5 Screening Criteria (Use of Screening Tables)................................ 14

v NUREG-1407
3.2.5.6 Seismic Input ............................... 14

3.2.5.7 Evaluation of Outliers-HCLPF Calculations ................................ 14

3.2.5.8 Nonseismic Failures and Human Actions .................................... 14

3.2.6 Containment Performance ...................................................... 14

3.3 Optional Methodologies.............................................................. 15

4 Acceptable Methodology for Performing the Internal Fires IPEEE ................................. 16

4.1 New Fire PRA Analysis .............................................................. 16

4.1.1 Identify Critical Areas of Vulnerability ............................................. 16

4.1.2 Calculate the Frequency of Fire Initiation in Each Area................................ 16

4.1.3 Analyze for the Disabling of Critical Safety Functions ................................. 16

4.1.4 Identify Fire-Induced Initiating Events/Systems Analysis................................ 16

4.1.5. Perform Containment Analysis................................................... 16

4.2 Use of anExisting Fire PRA........................................................... 16

4.3 Optional Methodologies.............................................................. 16

5 Acceptable Methodology for Performing the High Winds, Floods, and Transportation and Nearby

Facility Accident IPEEE...................................................... 17

5.1 Introduction ....................................................................... 17

5.2 Analytical Procedure ................................................................ 17

5.2.1 Review Plant-Specific Hazard Data and licensing Bases................................ 17

5.2.2 Identify Significant Changes Since OL Issuance ...................................... 17

5.2.3 Determine if the Plant/Facilities Design Meets 1975 SRP Criteria ........................ 17

5.2.4 Determine if the Hazard Frequency is Acceptably Low (Optional Step).................... 17

5.2.5 Perform a Bounding Analysis (Optional Step)....................................... 18

5.2.6 Perform a Probabilistic Risk Assessment (Optional Step)............................... 18

5.3 Optional Methodologies.............................................................. 18

6 Coordination with Ongoing Programs........................................................ 19

6.1 Introduction ....................................................................... 19

6.2 Description of Ongoing Programs....................................................... 19

6.2.1 IPE Program Related to Internal Events ........................................... 19

6.2.2 Programs Related to External Events .............................................. 19

6.2.2.1 Seismic Programs ...................................................... 19

6.2.2.2 Internal Fires Programs ................................................. 20

6.2.2.3 External Flooding Program............................................... 20

6.3 Approach on Coordination with Ongoing Programs ......................................... 20

6.3.1 Coordination With Internal Events Program (IPE).................................... 20

6.3.1.1 Preanalyses Planning ................................................... 20

6.3.1.2 Plant Modifications..................................................... 20

6.3.1.3 Accident Management .................................................. 21

6.3.2 Coordination Among External Events Programs...................................... 21

6.3.3 Coordination With Seismic Programs .............................................. 21

6.3.3.1 USI A-45 and GI-131................................................... 21

6.3.3.2 The Eastern U.S. Seismicity Issue (The Charleston Earthquake Issue) .............. 21

6.3.3.3 USI A-46............................................................ 22

6.3.4 Coordination With Other Issues.................................................. 22

7 Peer Review........................................................................... 24

7.1 Seismic Related Insights.............................................................. 24

NUREG-1407 vi
8 Documentation and Reporting ................................................................. 25

8.1 Information Submitted to the NRC ......................................................... 25

8.2 Information Retained for Audit ............................................................ 25

9 References .................................................................................. 26

Appendices

A Review Level Earthquake ..................................................................... A-i

B Comparison Between a Reduced-Scope and Full-Scope Seismic Margins Evaluation ................... B-1

C Detailed Documentation and Reporting Guidelines ............................................... C-1

D NRC Response to Comments and Questions ..................................................... D-1

Figure

5.1 Recommended IPEEE Approach for Winds, Floods, and Others ................................... 18

Tables

3.1 Review Level Earthquake-Plant Sites East of the Rocky Mountains ............................... 9

3.2 Review Level Earthquake-Western United States Plant Sites .................................... 9

vii NUREG-1407

EXECUTIVE SUMMARY

Background

In the Commission policy statement on severe accidents

in nuclear power plants issued on August 8, 1985, the

Commission concluded, based on available information,

that existing plants pose no undue risk to the public

health and safety and that there is no present basis for

immediate action on any regulatory requirements for

these plants. However, the Commission recognized,

based on NRC and industry experience with plantspecific probabilistic risk assessments (PRAs), that systematic examinations are beneficial in identifying plantspecific vulnerabilities to severe accidents that could be

fixed with low-cost improvements. As part of the implementation of the Severe Accident Policy, the Commission

issued Generic Letter 88-20 on November 23, 1988, requesting that each licensee conduct an individual plant

examination (IPE) for internally initiated events including internal flooding.

Many PRAs indicate that, in some instances, the risk from

external events could contribute significantly to core

damage. However, the examination for externally initiated events is proceeding on a later schedule to allow the

staff to carry out additional work to (1) identify which

external hazards need a systematic examination, (2) identify acceptable examination methods and develop procedural and submittal guidance, and (3) coordinate the individual plant examination of external events (IPEEE) with

other ongoing external event programs. In December

1987, an External Events Steering Group (EESG) was

established to make recommendations regarding the

scope, methods, and coordination of the IPEEE. The

EESG completed its task in May 1990. Based on the

EESG recommendations, the staff prepared this report

to provide detailed guidance to the licensees on the conduct of the IPEEE and on the structure and content of the

IPEEE submittal. The staff issued a draft of this report

for public comment in July 1990. It held a workshop in

September 1990 to explain the IPEEE process and to

obtain specific comments and questions on draft Supplement 4 to Generic Letter 88-20 and the draft of this

document). In addition to numerous comments and questions raised during the workshop, the staff received written comments from 16 organizations.

This revised report reflects the staff's thorough consideration of the public comments received. It provides,

specifically, the guidelines defining the IPEEE objectives; identifies external events that should be included in

the IPEEE; identifies acceptable methodologies; identifies coordination between the IPEEE and the ongoing

NRC programs; and provides the staff's responses to public comments and questions.

Objectives of the IPEEE

The general objectives of the IPEEE are similar to that of

the IPE-that is, for each licensee (1) to develop an

appreciation of severe accident behavior, (2) to understand the most likely severe accident sequences that could

occur at its plant under full-power operating conditions,

(3) to gain a qualitative understanding of the overall likelihood of core damage and fission product releases, and

(4) if necessary, to reduce the overall likelihood of core

damage and fission product releases by modifying, where

appropriate, hardware and procedures that would help

prevent or mitigate severe accidents. The key outcome of

an IPEEE is the knowledge and appropriate improvements resulting from such an examination process. The

examination can be conducted using any of the acceptable

approaches.

Identification of External Events

Included in the IPEEE

In supporting the implementation of the Severe Accident

Policy, a study was performed to determine which external initiators could be a potentially important accident

initiator that may pose a threat of severe core damage or

of a large radioactive release to the environment. The

external events considered, consistent with past probabilistic risk assessments (PRAs), are those events whose

cause is external to all systems used during normal and

emergency operations. The external events evaluated include seismic events, internal fires, high winds and tornadoes, external floods, transportation and nearby facility

accidents, lightning, severe temperature transients (extreme heat, extreme cold), severe winter storms, external

fires (forest fires, grass fires), extraterrestrial activity (meteorite strikes, satellite falls), and volcanic activity.

Based on the results of that study, the staff has concluded

that five external events need to be included specifically

in the IPEEE: seismic events, internal fires, high winds,

floods, and transportation and nearby facility accidents.

However, licensees should confirm that no other plantunique external events with potential severe accident

vulnerability are being excluded from the IPEEE.

Examination Methods

Seismic Events

A seismic PRA (Level 1 plus containment performance

analysis) or a seismic margins methodology (SMM) is

considered a viable approach to identify potential vulnerabilities. Guidance is provided for licensees performing a

new seismic PRA or updating an existing seismic PRA;

ix NUREG-1407
Executive Summary

emphasis is placed on the identification and ranking of

dominant plant sequences that could lead to seismically

induced core damage rather than on the numerical estimate of absolute frequency of occurrence. Methodology

upgrades include plant walkdowns, evaluation of relay

chatter, and evaluation of the effects of soil liquefaction.

Guidance is also provided for licensees using either the

NRC- or EPRI-sponsored seismic margins methodology.

The margins methodology screens components according

to their importance to safety and seismic capacity. By

design, the methodology utilizes two review or screening

levels geared to peak ground accelerations of 0.3g and

0.5g. Review level earthquakes (RLEs) were assigned

based on the Lawrence Livermore National Laboratory

(LLNL) and Electric Power Research Institute (EPRI)

hazard estimates, sensitivity tests, seismological and engineering judgment, and plant design considerations. The

use of the 0.3g full-scope and focused-scope RLE for

most plants in the Central and Eastern United States

would meet IPEEE objectives. The level of effort in the

analysis of relay chatter is the major difference between

these two categories. For some sites where the seismic

hazard is low, a reduced-scope margins methodology emphasizing plant walkdowns is considered adequate. For

sites in the Western United States, except California

coastal sites, the 0.5g RLE should be used. Methodology

upgrades include relay chatter, liquefaction, and plant

walkdown enhancements for the NRC method; guidance

on alternative success paths for the EPRI method; and

evaluation of nonseismic failure and human actions for

both methods.

Internal Fires

The internal fires IPEEE can be accomplished by performing a Level 1 fire PRA. Those issues identified in the

Fire Risk Scoping Study (NUREG/CR-5088) should be

addressed using plant-specific data and a specially tailored walkdown procedure.

The guidance does not address the fire vulnerability

evaluation (FIVE) methodology currently being developed by the Nuclear Management and Resources Council

(NUMARC) and EPRI. This methodology is being reviewed by the staff; when the review is completed, the

staff will issue an evaluation report on its acceptability for

use in the IPEEE.

High Winds, Floods, and Transportation and

Nearby Facility Accidents

The recommended overall approach consists of a progressive screening. The screening criterion for reporting

potential severe accident sequences is consistent with

that used for internal event IPEs. The steps in the progressive screening approach represent a series of analyses

in increasing levels of detail, effort, and resolution. However, the licensee may choose to bypass one or more steps

so long as it identifies the vulnerabilities or demonstrates

that they are insignificant. The screening approach consists of the following steps:

All plants:

1. Review plant-specific hazard data and licensing

bases.

2. Identify significant changes since the operating license was issued.

3. Determine if the plant and facilities design meets

the 1975 Standard Review Plan (SRP) criteria.

If the 1975 SRP criteria are not satisfied, or if it is known a

priori that they will not be satisfied, one or more of the

following approaches should be taken to further evaluate

the situation.

Optional:

4. Determine if the hazard frequency is acceptably low.

5. Perform a bounding analysis.

6. Perform a probabilistic risk assessment (PRA).

Alternative Methods

The staff recognizes that other methods capable of identifying plant-specific vulnerabilities to severe accidents

may be acceptable. A licensee may request that the staff

review any other systematic examination method to determine if it is acceptable for IPEEE purposes.

Coordination with Ongoing Programs

Guidance is provided on coordinating the IPEEE process

with ongoing programs. The first level of coordination is

among the major elements related to the implementation

of the Severe Accident Policy, that is, coordination among

the IPEEE, the internal events IPE, containment performance improvements, and accident management. The

second level of coordination is among the major elements

of the IPEEE, that is, seismic events, fires, and high

winds, floods, and others. The third level of coordination

is within each major element of the IPEEE.

Programs subsumed into the IPEEE include the external

event aspect of Unresolved Safety Issue (USI) A-45

(decay heat removal), Generic Issue (GI)-131 (in-core

flux mapping system) and the Eastern U.S. -Seismicity

(Charleston earthquake) Issue. Programs that need to be

coordinated with the IPEEE include USI A-46 (Seismic

Equipment Qualification, which also covers the seismic

spatial systems interaction of USI A-17 and the concern

NUREG-1407 X
Executive Summary

of USI A-40 for the seismic capability of large safetyrelated above-ground tanks), and GI-57 (Effects of Fire

Protection System Actuation on Safety Related Equipment).

Peer Review

The licensee should conduct a peer review by individuals

who are not associated with the initial evaluation to ensure the accuracy of the documentation package and to

validate both the IPEEE process and its results.

xi NUREG-1407

ACKNOWLEDGEMENTS

This document represents the staff position on the Individual Plant Examination for severe accident vulnerabilities due to external events (IPEEE). Representatives

of both the Office of Nuclear Regulatory Research and

the Office of Nuclear Reactor Regulation were active

contributors to the development of this guidance document; they are named below. In addition, significant input

was received from consultants and contractors to the

NRC, who are also named below. Edward Hill, of NRC,

provided technical editing.

NRC

Charles E. Ader

Goutam Bagchi

T. Y. Chang

Nilesh C. Chokshi

R. Wayne Houston

Roger M. Kenneally

Conrad McCracken

Thomas E. Murley

Thomas M. Novak

Robert Rothman

Jack Strosnider

Consultants

R. J. Budnitz

G. E. Cummings

R. P. Kennedy

M. K. Ravindra

Guy Aý Arlotto

Demetrios L Basdekas

John T. Chen

Adel A. El-Bassioni

David C. Jeng

Thomas L King

Warren Minners

Andrew J. Murphy

Leon Reiter

Lawrence C. Shao

Rex G. Wescott

William D. Beckner

Kazimieras M. Campe

Thomas M. Cheng

John H. Flack

Glenn B. Kelly

P. T. Kuo

Jocelyn A. Mitchell

David P. Notley

James E. Richardson

Brian Sheron

Future Resources Associates, Inc.

Lawrence Livermore National Laboratory

Structural Mechanics Consulting, Inc.

EQE Engineering, Inc.

Contractors and Subcontractors

P. Amico

D. L Bernreuter

M. P. Bohn

R. J. Budnitz

D. H. Chung

B. C. Davis

G. S. Hardy

J. R. McDonald

R. C. Murray

A. M. Nafday

P. G. Prassinos

M. K. Ravindra

J. Savy

Science Application International

Lawrence Livermore National Laboratory

Sandia National Laboratories

Future Resources Associates, Inc.

Lawrence Livermore National Laboratory

Lawrence Livermore National Laboratory

EQE Engineering, Inc.

Texas Tech University

Lawrence Livermore National Laboratory

EQE Engineering, Inc.

Lawrence Livermore National Laboratory

EQE Engineering, Inc.

Lawrence Livermore National Laboratory

xiii NUREG-1407

1 INTRODUCTION

1.1 Background

On August 8, 1985, the Nuclear Regulatory Commission

issued a policy statement on severe accidents (NRC,

1985). The Commission concluded, based on available

information, that existing plants pose no undue risk to the

public health and safety and that there is no present basis

for immediate action on any regulatory requirements for

these plants. However, the Commission recognizes,

based on NRC and industry experience with plantspecific probabilistic risk assessments (PRAs), that systematic examinations are beneficial in identifying plantspecific vulnerabilities to severe accidents that could be

fixed with low-cost improvements. As part of the implementation of the policy statement, the Commission issued Generic Letter 88-20 (NRC, 1988 and 1989), requesting that each licensee conduct an individual plant

examination (IPE) for internally initiated events.

Risk assessments indicate that the risk from external

events could be a significant contributor to the core damage in some instances. However, licensees were requested to proceed with the examinations only for internally initiated events (including internal flooding) in

Generic Letter 88-20. Examination of severe accident

vulnerabilities due to externally initiated events (IPEEE)

is proceeding separately and on a later schedule to allow

the staff to carry out additional work (SECY-88-147) to

(1) identify which external hazards need a systematic

examination, (2) identify examination methods and develop procedural guidance, and (3) coordinate the IPEEE

with other ongoing NRC programs that deal with various

aspects of external event evaluations to ensure that there

is no duplication of industry efforts.

To accomplish these objectives, the staff established the

External Events Steering Group (EESG) in December

1987 to make recommendations regarding the scope,

methods, and coordination of the IPEEE (Beckjord,

1987, 1988). Specifically, the EESG is responsible for

developing broad guidance for dealing with (1) external

events on a generic basis both organizationally and technologically and (2) the implementation of the severe accident policy with respect to external events. The EESG

established three technical subcommittees dealing with

earthquakes (seismic events), internal fires, and high

winds, floods, and "other" external events. The subcommittees were chartered to define the scope of the external

events examination, identify acceptable examination

methodologies, and coordinate ongoing issues and activities (for example, Unresolved Safety Issues and Generic

Issues).

The EESG completed its task in May 1990. Based on the

EESG recommendations, the staff prepared this report

to provide detailed guidance on the conduct of the IPEEE

and on the structure and content of the IPEEE submittal.

It issued a draft of this report for public comment in July

1990. It conducted a workshop in September 1990 to

explain the IPEEE process and to obtain specific comments and questions on draft Supplement 4 to Generic

Letter 88-20 and the draft of this guidance document. In

addition to the comments raised during the workshop, the

staff received written comments from 16 organizations.

This final report includes changes resulting from the resolution of these comments.

1.2 IPEEE Objectives

The objectives of the IPEEE, which are similar to the

objectives of the internal event IPE, are for each licensee:

1. to develop an appreciation of severe accident

behavior,

2. to understand the most likely severe accident sequences that could occur at the licensee's plant under full power operating conditions,

3. to gain a qualitative understanding of the overall

likelihood of core damage and fission product releases, and

4. if necessary, to reduce the overall likelihood of core

damage and radioactive material releases by modifying, where appropriate, hardware and procedures

that would help prevent or mitigate severe accidents.

However, the staff recognized at the outset that the external initiators could not necessarily be treated in exactly

the same way as internal initiators in the implementation

of the Severe Accident Policy because the sources and

treatment of uncertainties in estimates of core damage

frequencies for external and internal events can be quite

different. In addition, some methods endorsed by the staff

for evaluating external hazards and identifying vulnerabilities do not produce estimates of core damage frequency. For example, seismic margins methods produce

estimates of seismic hazard levels of high confidence-low

probability of failure (HCLPF) for a plant rather than

estimates of core damage probability.

Therefore, the staff determined that an explicit estimate

of core damage frequency was not needed to meet the

intent of the Severe Accident Policy and would not be a

condition of the IPEEE. Thus, Objective 3 above would

be addressed only indirectly by some methods acceptable

1 NUREG-1407
1. Introduction

for use in the IPEEE. Nevertheless, the key objective of

gaining an understanding of plant behavior through the

examination process could be met.

1.3 Purpose of Document

The purpose of this document is to provide guidelines for

conducting the IPEEE and on the structure and content

of the IPEEE submittal. It is not the intent of

NUREG-1407 to go beyond the information request contained in Supplement 4 to Generic Letter 88-20. The

external events recommended for inclusion in the IPEEE

are identified in Section 2. Acceptable methodologies for

performing an IPEEE along with upgrades to reflect

state-of-the-art improvements are identified in Sections 3

through 5. Section 3 addresses the seismic portion; Section 4 the internal fires portion; and Section 5 the high

winds, floods, and other portion of the IPEEE. Coordination between the IPEEE and the internal events IPE,

other external events, and ongoing programs within each

external event are provided in Section 6. A discussion of

the peer review is provided in Section 7. A summary of

documentation and reporting guidelines is provided in

Section 8. The staff's responses to public comments and

questions raised during the IPEEE workshop held in September 1990 and the written comments received soon

afterward are given in Appendix D.

NUREG-1407 2
2 EVENTS EVALUATED FOR INCLUSION IN THE IPEEE

The external events considered, consistent with past

probabilistic risk assessments (PRAs), are those events

whose cause is external to all systems used in normal

operation and emergency operation situations. Internal

fire and internal flood are external to the "system" and

therefore have been considered as external events in past

PRAs. However, internal floods are being considered in

the internal events IPE process (NRC, 1988).

In supporting the implementation of the Severe Accident

Policy, a study of the risk of core damage to nuclear power

plants in the United States due to externally initiated

events was performed. The objective was to determine

which external initiators have the potential of initiating an

accident that may lead to severe reactor-core damage or

large radioactive release to the environment. Seismically

initiated events are investigated in NUREG/CR-5042,

Suppl. 1; internal fires, high winds/tornadoes, external

floods, and transportation accidents are investigated in

NUREG/CR-5042; "other external events" are investigated in NUREG/CR-5042, Suppl. 2. The "other external events" covered are nearby industrial/military facility

accidents, on-site hazardous material storage accidents,

severe temperature transients, severe weather storms,

lightning strikes, external fires, extraterrestrial activity,

volcanic activity, earth movement, and. abrasive windstorms.

Some external events may not pose a significant threat of

a severe accident to all plants, some events may have been

considered in the plant's design to a sufficient degree, and

some events may have been or will be reviewed under

ongoing programs at some plants. The staff's evaluation

and recommendations are contained in the following

sections.

2.1 Seismic Events

The following are based upon an examination of current

seismic design criteria, previous and ongoing seismic issues and programs, and'seismic PRAs:

1. Mean seismic core damage frequencies calculated

from past PRAs (NUREG-I150, NUREG/CR5042, Suppl. 1) have been found to be in the range of

10-4 to 10-e per year. Identified vulnerabilities are

plant specific and include yard tanks, electrical

equipment, diesel peripherals, structural failures,

and equipment anchorages.

2. New data such as the occurrences of larger than

anticipated earthquakes and the development of

new hypotheses indicate that the plant-specific seismic hazard may be quite different from that envisioned at the time of licensing and make it difficult to

rule out seismic events on the basis of initiating

event frequencies.

3. Based primarily on their vintage, the current population of plants exhibit various levels of seismic design requirements and margin. Some of the very

early plants have been backfitted under the Systematic Evaluation Program to ensure certain margins

for safe shutdown using criteria different from current licensing criteria.

4. There have been modifications to plants since their

original designs; for instance, the reduction of snubbers at some plants. These changes, in part, have

relied on existing conservatism or risk-based arguments (e.g., LOCA + SSE combinations). The systematic-examination of plants by the IPE and IPEEE

will give an integrated picture of plants as they exist.

It will also allow an integrated evaluation of the

effects of individual changes made to plants over

time.

5. There are unresolved safety issues and generic issues (e.g., USI A-45, USI A-46) that are in various

stages of implementation. The IPE/IPEEE provides

a convenient as well as meaningful framework for

addressing many of these issues.

6. PRAs and seismic margins evaluations have resulted

in cost-effective plant-specific improvements.

Therefore, the seismic external hazard should be included in the IPEEE.

2.2 Internal Fires

Based upon the examination of past fire PRAs, the contribution of internal fires to the probability of core damage

may be significant and is very plant specific (NUREG/

CR-5042). However, the numerical results always contain large uncertainties. The fire risk scoping study

(NUREG/CR-5088) further confirms the following:

1. The overall fire-induced core damage frequency for

the four plants studied (Seabrook, Oconee, Limerick, and Indian Point) increased from the original

PRA studies even though, for certain fire scenarios,

there was a net decrease. For all plants reviewed,

fire continues to represent a dominant risk contributor.

2. Most initiating event frequencies were increased

based on a much more complete data base available

on fire occurrences in nuclear power plants. Under

currently applied risk assessment methodologies,

this increase in initiating event frequency alone

3 NUREG-1407
2. Events Evaluated for IPEEE

results in a direct increase in overall fire-induced

core damage frequency with all other factors remaining constant.

3. Use of an expanded data base on historical fire suppression times for nuclear power plants resulted in a

suppression probability distribution with a lower

probability of suppression within a given time than

that assumed in the original risk assessments. Under

current methodologies, this again results in an increase in fire-initiated core damage frequency with

all other factors remaining constant.

4. Updated information on the ignition and damage

thresholds of cable insulation in some cases resulted

in lower thermal damage limits. In some cases, no

change in damage limits was required. A decrease in

the assumed thermal damage limits would, in general, be expected to lead to increased estimates of

fire-initiated core damage frequency.

5. Plant modifications made as a result of Appendix R

requirements reduced the core damage frequency at

Indian Point and Limerick for the requantified areas

by factors of ten and three, respectively. For

Seabrook, the identified Appendix R plant modifications did not affect the requantified core damage

scenarios for internal fires. The Oconee PRA had

already incorporated Appendix R modifications and

no modifications subsequent to its performance

were identified. Hence no Appendix R impact could

be identified for either Seabrook or Oconee.

6. A number of issues that were not addressed in the

past fire PRAs (effectiveness of fire brigade, effectiveness of fire barrier, seismic/fire interactions,

control system interactions, and effects of fire suppressants on safety equipment) could increase the

vulnerability to fire.

Therefore, based on the above studies, the internal fire

hazard should be included in the IPEEE.

2.3 High Winds and Tornadoes

For plants designed against NRC's current criteria, these

events pose no significant threat of a severe accident

because the current design criteria for wind are dominated by tornadoes having an annual frequency of

exceedance of about 10-7. However, older plants and

some modern plants having facilities not designed against

these criteria need a systematic examination to identify

plant-specific vulnerabilities (NUREG/CR-5042).

2.4 External Floods

For plants designed against current criteria as described

in Regulatory Guide 1.59 and applicable Standard Review Plan sections, particularly Section 2.4, floods pose no

significant threat of a sievere accident because the

exceedance frequency of the design basis flood, excluding

floods due to failure of upstream dams, isjudged to be less

than 10-5 per year (Chery, 1985), and the conditional core

damage frequency for a design basis flood is judged to be

less than 10-1. Thus core damage frequencies are estimated to be less than 10-8 per year for a plant designed

against NRC's current criteria. However, the latest probable maximum precipitation (PMP) criteria published by

the National Weather Service (NWS) call for higher rainfall intensities over shorter time intervals and smaller

areas than have previously been considered; this could

result in higher site flooding levels and greater roof ponding loads than have been used in previous design bases

(GI 103). Licensees are requested to assess the effects of

applying these new criteria to their plants in terms of

onsite flooding and roof ponding. Also, some older plants

may have higher potential risk and need systematic examinations for plant-specific vulnerabilities.

2.5 Transportation and Nearby

Facility Accidents

These events consist of accidents related to transportation and accidents at industrial and military facilities.

Plants designed against NRC's current criteria (NUREG/

CR-5042) should have no significant vulnerability to severe accidents from these events because the initiators

considered in the design should have a recurrence frequency less than 10-6 or have been shown through a

bounding analysis not to affect the plant. However,

changes may have occurred since the original design and

there may be exceptions that need some systematic

examination. Also, some older plants may not meet the

NRC's current criteria and need systematic examinations

for plant-specific vulnerabilities.

2.6 Lightning

Lightning has been experienced at many nuclear power

plants in the United States (NUREG/CR-5042, Suppl. 2;

AEOD, 1986; ACRS,. 1989). The impact of lightning on

plant operation and the vulnerability of plants to a severe

accident due to lightning has been examined. The major

conclusion is that the primary impact of lightning on nuclear power plants is the loss of offsite power. The loss of

offsite power is included as part of the internal events

IPE, and examination for vulnerabilities due to this aspect

of lightning is therefore already included in the IPE process. The staff has concluded that, in general, other effects

of lightning on nuclear power plants are insignificant.

However, further examination of lightning effects may be

warranted for certain sites where, based on past operating

experience, lightning strikes are likely to cause more than

just loss of off site power; for example, they may also affect

safety-related instrumentation and control systems.

NUREG-1407 4
2. Events Evaluated for IPEEE

Based on an examination of historical data on lightning, as

well as knowledge of plant systems, the staff concludes

the following:

1. lightning has typically caused partial or complete

loss of offsite power, which is the main impact of

lightning and which is already being examined as

part of the internal events IPE.

2. lightning is much less likely to affect the onsite

power system.

3. Lightning has affected safety-related equipment and

has caused reactor trips, but these events have not

been significant in terms of impact on the plant.

4. Safety systems (e.g., diesel generators, electrically

powered pumps) are not normally in operation.

Thus, while control systems may be damaged, the

safety systems are less susceptible to damage and

may be manually activated.

5. Redundancy of safety systems and the capability for

recovery of systems (replacing fuses or resetting

breakers) further reduce the likelihood that the low

frequency of lightning damage events will result in a

severe accident.

The staff has judged that the probability of a severe accident caused by lightning (other than one due to loss of

offsite power) is relatively low and further consideration

of lightning effects should be performed only for plant

sites where lightning strikes are likely to cause more than

just loss of offsite power or a scram.

2.7 Severe Temperature Transients

(Extreme Heat, Extreme Cold)

Severe temperature transients may affect nuclear power

plants in the United States (NUREG-1032). However,

the effects are usually limited to reducing the capacity of

the ultimate heat sink and loss of offsite power (NUREG/

CR-5042, Suppl. 2). The capacity reductioni of the ultimate heat sink would be a slow process that allows plant

operators sufficient time to take proper actions such as

reducing power output level or achieving safe shutdown,

if necessary, and maintaining the plant in a safe shutdown

condition. The other potential impact on the plant, loss of

offsite power, will be considered within the realm of the

station blackout rule (NRC, 1988b) and the internal event

IPE. Therefore, the temperature transients need not be

addressed in the IPEEE.

2.8 Severe Weather Storms

Severe weather storms (ice storm, hailstorm, snowstorm,

dust storm, sandstorm) accompanied by strong winds have

caused several complete and partial losses of offsite

power (NUREG/CR-5042, Suppl. 2). The potential effect of loss of offsite power and station blackout will be

addressed in the internal event IPE; thus severe weather

storm evaluations need not be repeated in the IPEEE.

2.9 External Fires (Forest Fires, Grass

Fires)

These are fires occurring outside the plant site boundary.

Potential effects on the plant could be loss of offsite

power and forced isolation of the plant ventilation and

possi'ble control room evacuation. Usually, external fires

are unable to spread onsite because of site clearing during

the construction stage (NUREG/CR-5042, Suppl. 2).

However, there has been one instance during which a

nearby forest fire caused a partial loss of offsite power.

The effect of loss of offsite power will be addressed in the

internal events IPE and need not be repeated in the

IPEEE. The other effects have been evaluated during

operating license (OL) review against sufficiently conservative criteria; thus they do not need to be reassessed

in the IPEEE.

2.10 Extraterrestrial Activity

(Meteorite Strikes, Satellite Falls)

Extraterrestrial activity is considered to be natural satellites such as meteors or artificial satellites that enter the

earth's atmosphere from space. Because the probability

of a meteorite strike is very small (less than 10-9)

(NUREG/CR-5042, Suppl. 2), it can be dismissed on the

basis of its low initiating event frequency.

2.11 Volcanic Activity

Most nuclear power plant sites are too far away from

active volcanoes to expect any effect at the plant, so most

licensees need not consider the volcanic effects. However, those sites in the vicinity of active volcanoes should

assess volcanic activities (NUREG/CR-5042, Suppl. 2) as

part of the IPEEE process.

2.12 Summary

In summary, based on the above evaluation, five events

need to be included by all licensees in the IPEEE: seismic

events, internal fires, high winds, floods, and transportation and nearby facility accidents. All licensees should

confirm, however, that no plant-unique external events

known to the licensee today with potential severe accident vulnerability are being excluded from the IPEEE.

5 NUREG-1407
3 ACCEPTABLE METHODOLOGIES FOR PERFORMING

THE SEISMIC IPEEE

For the purpose of performing an IPEEE, two methodologies are considered acceptable to identify potential

seismic vulnet-abilities at nuclear power plants. The first is

a seismic probabilistic risk assessment (PRA) (NUREGI

CR-2300; NUREG/CR-2815, Vol. 2; NUREG/CR4840). In addition, an evaluation of the reliability and

usefulness of results and insights obtained from external

event PRA methodologies is contained in NUREG/

CR-5477. The second method is one of the seismic

margins methodologies (SMM) described in NUREG/

CR4334, NUREG/CR-4482, and NUREG/CR-5076,

EPRI NP-6041 or the reduced SMM described later in

this section.

In meeting the objectives of the IPEEE, the examination

should focus on qualitative insights from the systematic

plant examination rather than only on absolute core damage frequency estimates. Guidance for performing the

seismic IPEEE using a PRA or margins methodology is

provided below.

3.1 Seismic PRA

This discussion deals with the use of PRA techniques in

the seismic IPEEE. The PRA should be at least a Level 1

plus containment performance analysis. The basic elements are (1) hazard analysis, (2) plant system and structure response analysis, (3) evaluation of component fragilities and failure modes, (4) plant system and sequence

analysis, and (5) containment and containment system

analysis including source terms, to identify unique seismic

sequences or vulnerabilities different from the internal

event analysis. Specific guidance and enhancements are

provided for licensees performing a new PRA or updating an existing seismic PRA.

3.1.1 New Seismic PRA Analysis

3.1.1.1 General Considerations

Licensees choosing to do a seismic PRA built on an internal events PRA should be aware of important considerations that, if incorporated in the planning of the internal

events PRA, will minimize their resource expenditure

and speed the staff reviews. For example, (1) a wellorganized walkdown team and a properly planned walkdown will enable many issues to be addressed at the same

time; (2) the peer review team should consider the need

to review both internal and external event analyses;

(3) fault tree analysts for internal events should be aware

of spatial interactions (including internal flooding effects), failure of passive components such as structures

and supports, and common-cause effects (the culling or

pruning of trees should be done with these considerations

in mind); and (4) internal event models should be developed knowing that, in the seismic analysis, the fragilities

of a component are sensitive to elevation. Also, a component and its peripheral equipment may have different

fragilities. Additional discussion of this subject can be

found in NUREG/CR-4840.

PRA calculations that account for all uncertainties are

clearly acceptable. However, the staff believes that, for

the seismic IPEEE, it is not necessary to carry out complete uncertainty quantifications defining a distribution

of core-damage frequencies in order to identify vulnerabilities. Mean point estimation using a single hazard

curve (rather than a family of hazard curves) and a single

fragility curve (rather than a family of fragility curves) for

each component is sufficient to get insights into potential

seismic vulnerabilities. To highlight the most pertinent

results/insights from the seismic portion of the IPEEE,

mean point estimates using hazard curves described in

NUREG/CR-5250 and EPRI NP-6395D should be obtained. Further discussion on the use of hazard curves is

contained in Section 3.1.1.2.

The above point estimation approach is valid only because of the IPEEE objective: to identify dominant sequences and components and where possible rank them.

(This point estimate should not be confused with a "Phase

I" type PRA analysis where point estimate calculations

are used only to define scopes for more detailed Phase II

and Phase III studies). Fragilities used in this point estimate, where possible, should be plant specific and rigorous to be able to identify dominant components and rank

them. Correlations and other aspects of analysis should

be treated so that, when a mean seismic hazard curve is

used with the mean plant fragility curve, the resulting

core damage estimate approximately represents the

mean estimate that would be derived from the full uncertainty analysis.

The recommendation of performing point estimation

type calculations is made primarily to highlight insights

needed for the severe accident behavior perspective. This

should not be construed as de-emphasizing or ignoring

uncertainties. Analysts are encouraged to make careful

study of the origins of the possible uncertainties, including those that are hardest to quantify. Many of. the insights

obtained from a PRA analysis are obtained by trying to

gain a better understanding of the uncertainties. Consideration of uncertainties may affect how the results of a

PRA are implemented in plant changes.

NUREG-1407 6
3. The Seismic IPEEE

3.1.1.2 Hazard Selection

For the United States east of the Rocky Mountains, two

highly sophisticated seismic hazard studies were conducted by Lawrence Livermore National Laboratory

(LLNL) (NUREG/CR-5250) and the Electric Power Research Institute (EPRI) (EPRI NP-6395-D). For many

sites, these studies yield significant differences at the lowprobability and high-level ground motions. The initial

PRAs carried out using these estimates (Surry and Peach

Bottom in NUREG-1150) indicate that, despite large

differences in absolute numerical estimates, the identification, ranking, and relative contributions of the dominant seismic sequences are virtually the same for both

LLNL and EPRI hazard estimates. This equivalence is

apparently due to the fact that the slopes of the seismic

hazard curves are not significantly different over those

ground motion levels that, in conjunction with the fragilities, control the relative distribution of seismically induced core damage frequencies. Although these results

are very encouraging, there is no guarantee that this will

be true for all sites in the Central and Eastern United

States.

While a full seismic hazard uncertainty analysis is not

necessary in performing a seismic PRA for the IPEEE,

the staff prefers that mean (arithmetic) hazard estimates

from both the LLNL and the EPRI studies should be used

to obtain two different point (mean) estimates. If a licensee elects to perform only one analysis, it should use the

higher of the two mean (arithmetic) hazard estimates.

The use of both the LLNL and EPRI mean hazard curves

has another advantage in that the extent of uncertainty

will become obvious and the emphasis on the bottom line

numbers is reduced. The use of both of these estimates

(LLNL and EPRI) will serve to identify differences, if any,

in the delineation of dominant seismic sequences (minor

variations in contributions and rankings are anticipated).

Such differences would have to be addressed by the licensee in its identification and listing of vulnerabilities.

For plants in the Western United States, for which there

are no counterparts to the LLNTL and EPRI studies, a

licensee should conduct its own study to define the mean

hazard estimate for use in the IPEEE. The licensee

should also provide reasonable assurance that any significant uncertainty in those elements of hazard (for example, slope) that control the identification, ranking, and

relative contribution of seismic contributors to core damage is addressed in sensitivity studies. As in the Central

and Eastern United States, the identification and listing

of vulnerabilities should take this uncertainty into account.

Most seismic PRAs use peak ground acceleration as the

hazard parameter. If this is done, spectral shapes that are

consistent with current estimates of ground motion

should be used. In the Central and Eastern United States,

current spectral estimates can be found in the LLNL and

EPRI studies. Since similar spectral shapes are obtained

from LLNL and EPRI hazard studies, separate analyses

using both spectral shapes are not needed. Median spectral shapes of 10,000 year return period provided in

NUREG/CR-5250 along with variability estimates are

recommended for use in the analyses. Other site-specific

spectral shape estimates may be proposed (that is, derived

from a suite of appropriate recorded earthquakes). For

the Western United States, site-specific spectral shape

should be established and used. Since only one spectral

shape is used for both hazard analyses, two separate plant

response and fragility analyses are not needed.

If an upper bound cutoff to ground motion at less than

1.5 g peak ground acceleration is assumed, sensitivity

studies should be conducted to determine whether the

use of this cutoff affects the delineation and ranking of

seismic sequences.

3.1.1.3 Fragility Estimation

The following guidance on fragility estimation is included

to clarify the use of fragility in the context of the "point

estimation" approach discussed above. Details and

methods for fragility and high confidence of low probability of failure (HCLPF) calculations are discussed in a

number of references, for example, NUREG/CR-2300,

NUREG/CR-4334, NUREG/CR-4482, NUREG/CR5076, NUREG/CR-4659, Vols. 1-3, EPRI-NP 6041, and

NUREG/CR-5270. It is recognized that large uncertainties exist in fragilities estimation (NUREG/CR-5270). A

perspective on how this uncertainty affects the results of

analysis (numerical and other insights, for example, dominant sequences and components) should be maintained.

Consistent with the point estimation approach, one can

use a single mean component fragility curve for each

component and hence for sequence-level and plant-level

assessments. This mean curve is defined by the median

capacity, F, and composite uncertainty, pc, where pc

2 =

Pr2 + pu

2 , when Pr and Pu are estimated separately. pr

and Pu represent random uncertainty and modeling uncertainty, respectively. It is also acceptable to use a family

of fragility curves instead of a single curve.

When a single mean fragility curve is available, HCLPF

capacity for a component (sequence, or plant) can be

approximated by -2.3 pc below the median (ie., 1% composite probability of failure is essentially equivalent to

95% confidence of less than 5% probability of failure).

While developing sequence-level and plant-level fragilities, the licensee should retain the ability to report

HCLPFs with and without nonseismic failures and human

actions.

7 NUREG-1407
3. The Seismic IPEEE

3.1.1.4 Seismic PRA Methodology Enhancements

Review of past seismic PRAs indicates that certain areas

have been treated either inconsistently or not at all.

Therefore, the following areas should be included:

1. Plant Walkdowns. Walkdowns are performed to find

as-designed, as-built, and as-operated seismic weaknesses in plants. Each licensee should perform a

walkdown consistent with the intent of the guidelines described in Sections 5 and 8 and Appendices

D and I of the EPRI Seismic Margins Methodology

(EPRI NP-6041).

2. Relay Chatter. Relays, in this context, include components such as electric relays, contactors, and

switches that are prone to chatter. Additional guidance is given in NUREG/CR-5499. The scope of the

relay chatter evaluation should be consistent with

the site's seismic margin review level earthquake

classification (full scope or focused scope) as identified in Tables 3.1 and 3.2 and discussed in Sections 3.2.4.2 and 3.2.5.3. That is, a more complete

evaluation is to be performed for sites in the fullscope category than for sites in the focused-scope

category. It is anticipated that chatter and recovery

actions will be modeled as necessary. The focusedscope evaluation can be limited to a review of low

seismic ruggedness relays for plants that are not

included in the USI A-46 program.

The examination of the relay chatter effects (for

example, the Hatch margin evaluation) has resulted

in large resource expenditures in terms of staffhours. Therefore, careful planning and use of generic insights, if they are applicable to the plant, are

necessary. Additional guidance on this topic is also

included in Section 3.2.4.2.

3. Liquefaction. The potential for soil liquefaction and

associated effects on the plant need to be examined

for some sites because of specific site conditions.

The impact on plant operation should be assessed

from the point of view of both potential for and

consequences of liquefaction. Procedures for assessing soil liquefaction are described in EPRI NP-6041.

4. HCLPF Calculations (Optional). Licensees can report plant-level, sequence-level, and componentlevel HCLPF values and use this information to

support decisions related to the identification and

listing of vulnerabilities. In several PRAs (for example, Millstone 3 and Diablo Canyon), HCLPF estimates are reported along with other PRA results.

These PRAs can be used as guidance for deriving

HCLPF values from fragilities. HCLPF values are to

be reported both with and without the effects of

nonseismic failures and human actions. If a licensee

does not supply HCLPF calculations, the staff will

calculate the HCLPF values based on information

provided in the IPEEE submittal and will use them

in the evaluation of the submittal. Note that plantlevel, sequence-level, and component fragilities are

to be documented.

3.1.1.5 Containment Performance

The primary purpose of the containment performance

evaluation is to identify sequences and vulnerabilities that

involve containment, containment functions, and containment systems (e.g., igniters and suppression pools)

seismic failure modes or timing that are significantly different from those found in the IPE internal events evaluation. Additional guidance is presented in Section 3.2.6.

3.1.2 Use of an Existing PRA

The use of an existing seismic PRA to address the seismic

IPEEE is acceptable provided the PRA reflects the current as-built and as-operated condition of the plant and

some of the deficiencies of past PRAs discussed below are

adequately addressed.

1. Hazard Selection. For PRAs at sites east of the Rocky

Mountains that did not use the LLNL and EPRI

mean hazard estimates, sensitivity studies should be

conducted to determine if the use of these results

would affect the delineation or ranking of seismic

sequences. For PRAs in the Western United States,

the sensitivity studies should be carried out to determine the effect of uncertainty in hazard on the delineation and ranking of seismic sequences.

2. Walkdowns. Since a walkdown is considered to be

one of the most important ingredients of the seismic

IPEEE, a supplementary walkdown in conformance

with the intent of the procedures described in Sections 5 and 8 and Appendices D and I of the EPRI

margin methodology (EPRI NP-6041) should be

performed. It may be necessary to amplify the earlier analysis based on the walkdown outcome. These

results should be reported.

3. Relay Chatter. Relay chatter effects either have not

been considered or were assumed fully recoverable

in past PRAs. Relays, in this context, include components such as electric relays, contactors, and

switches that are prone to chatter. Licensees should

analyze the effect of relay chatter or determine that

the type of relays used in the safety systems are not

subject to relay chatter. The scope of the review

should be consistent with the site's seismic margins

review level earthquake classification (full scope or

focused scope) as identified in Tables 3.1 and 3.2 and

discussed in Sections 3.2.4.2 and 3.2.5.3. Additional

guidance is provided in NUREG/CR-5499. Results

of this effort that lead to a PRA revision or plant

fixes should be reported.

NUREG-1407 8
3. The Seismic IPEEE

Table 3.1 Review Level Earthquake--Plant Sites East of the Rocky Mountains

Reduced Scope

Big Rock Point

Comanche Peak

Crystal River

0.3g Focused Scope

Arkansas #2

Beaver Valley

Bellefonte

Braidwood

Browns Ferry

Brunswick

Byron

Callaway

Calvert Cliffs

Catawba*

Clinton

Cook

Cooper

0.3g Full Scope

Arkansas #1

Indian Point

Duane Arnold*

Grand Gulf

River Bend

Davis-Besse

Dresden

Farley

Fermi

Fitzpatrick

Fort Calhoun

Ginna

Haddam Neck

Harris

Hatch

Hope Creek

Kewaunee

LaSalle

Maine Yankee

Oconee*

South Texas

St. Lucie

Limerick

McGuire

Millstone

Monticello

Nine Mile Pt.

North Anna*

Oyster Creek

Palisades

Peach Bottom

Perry

Point Beach

Prairie Island

Quad Cities

Turkey Pt.

Waterford

Salem

Shoreham

Summer*

Surry

Susquehanna

Three Mile Island

Vermont Yankee

Vogtle

Watts Bar

Wolf Creek

Zion

Robinson

Sequoyah

Yankee Rowe

Committed to Perform a Seismic PRA

Pilgrim** Seabrook*

Notes:

*Special attention to shallow soil conditions is appropriate for these locations (see Section 3.2.2).

"Relay chatter evaluation should be similar to a full-scope review.

Table 3.2 Review Level Earthquake--Western United States Plant Sites

0.5g*

Trojan

Washington Nuclear

Rancho Seco

Palo Verde

Seismic Margin Methods Do Not Apply to the Following Sites:

Diablo Canyon San Onofre

Notes:

*Indicates a site in the Western United Stateswhose default bin is 0.5g, unless the licensee can demonstrate that the site hazard is similar to

that at sites east of the Rocky Mountains that are found in the 0.3g bin.

Changes in the review level earthquake from 0.5g to 0.3g should be approved before doing significant analysis.

9 NUREG-1407
3. The Seismic IPEEE

4. Nonseismic Failures and Human Actions. In several

seismic PRAs, nonseismic failures (e.g., failures of

the auxiliary feedwater system and failure of feed

and bleed mode of core cooling, battery depletion,

power-operated relief valve failures) and human

actions (e.g., delays or failures in performing spedfled actions or operator misdiagnoses a situation and

takes ar improper action that is not related to the

actual, current plant situation) have been important

contributors to seismically induced core damage

frequencies or risk indices. Unless nonseismic failures are considered, improper decisions may be

made regarding plant modifications or procedural

changes.

The licensee has the option to expand its PRA or

demonstrate that the exclusion of nonseismic failures will not significantly alter the PRA results or

insights. The scope of nonseismic failures and human interactions that might affect seismic sequences

should be defined by the licensee based on the internal events analyses.

5. Liquefaction. The potential for soil liquefaction and

associated effects on the plant need to be examined

for some sites because of specific site conditions.

The impact on plant operation should be assessed

from the point of view of both potential for and

consequences of liquefaction. Procedures for assessing soil liquefaction are described in EPRI NP-6041.

6. Containment Performance. Licensees should ensure

that the performance of containment and containment systems are addressed. Section 3.2.6 contains

guidance.

7. HCLPF Calculations (Optional). Licensees can extract and report plant-level, dominant-sequencelevel, and dominant-component-level HCLPFs both

with and without the effects of nonseismic failure

and human actions and use this information to support decisions related to the identification and listing

of vulnerabilities. If the licensee does not supply

HCLPF calculations, the staff will calculate HCLPF

values based on information provided in the IPEEE

submittal and will use them in the evaluation of the

submittal.

3.2 Seismic Margin Methodologies

This discussion deals with the use of the seismic margin

methodology in the seismic IPEEE. Specifically, guidance

and enhancements are provided for a licensee using

either the NRC or EPRI margins methodology.

3.2.1 General Considerations

The seismic margin methodology is considered acceptable for addressing seismic concerns in the severe accident policy implementation. Two methodologies are currently available: one developed under NRC sponsorship

and the other developed under EPRI sponsorship. The

staff has determined that both methods (with the noted

enhancements) will adequately address IPEEE objectives.

The two methods use different system analysis philosophies. The NRC method is based on an event/fault tree

approach to delineate accident sequences. For example,

for PWRs, two safety functions are considered to be most

important to plant seismic safety. reactor subcriticality

and early emergency core cooling. If these functions are

ensured for a given earthquake, there is high confidence

that core damage would not occur at that level. The EPRI

methodology is based on a systems "success path" approach. This approach defines and evaluates the capacity

of those components required to bring the plant to a

stable condition (either hot or cold shutdown) and maintain that condition for at least 72 hours. Several possible

success paths may exist. Both the NRC and the EPRI

methods were used in the trial application at the Hatch

plant. Application of the NRC method was limited to a

systems review. Insights gained from the use of these two

methods and the differences between them are discussed

in reports by Davis (1990), Orvis et al. (1990), and Shao et

al. (1990).

Each licensee should examine its plant critically to ensure

that the generic insights used in margin methodology

development to identify critical functions, systems, and

success path logic are applicable to its plant. This is particularly vital for older plants where systems and functions may differ greatly from the plants considered in the

development of the margins methodologies (NUREG/

CR-4334, NUREG/CR-4482, NUREG/CR-5076, and

EPRI NP-6041).

Based on written comments on draft of NUREG-1407

and public meetings, the staff has defined three categories of margin studies requiring varying levels of effort.

The three categories are full-scope, focused-scope, and

reduced-scope. The focused-scope category is new, while

the full-scope (although not identified as such in draft

NUREG-1407) and reduced-scope categories are retained. The primary purpose of this further subdivision is

to reduce the level of effort for some plants. Licensees of

plants with relatively higher hazard and a lower seismic

design basis will conduct more detailed studies than will

licensees of other plants (grouping of plants into various

categories is discussed in Section 3.2.2 and Appendix A).

The examination scope in each category for NRC and

EPRI methods is discussed in detail in Sections 3.2.4 and

NUREG-1407 10
3. The Seismic IPEEE

3.2.5. An introductory discussion of the rationale and the

general characterization of full-scope and focused-scope

follows.

The major difference between the full scope and the

focused scope is the scope of relay chatter evaluation.

Based on detailed studies carried out at the Hatch and

Diablo Canyon plants at considerable resource expenditure, it appears that the problems that could be caused by

relay chatter at these plants were recoverable with existing procedures. However, there is a concern among the

staff and industry consultants that such conclusions can

not be considered generic without some additional plant

reviews. Additionally, both the NRC-sponsored and the

EPRI-sponsored relay tests indicate that relay performance is very sensitive to variables such as spring tension,

orientation, and mounting. Tests further indicate that a

significant number of relays may have capacities between

SSE and RLE levels. USI A--46 evaluations are to be

performed only at the SSE levels. Therefore, the staff is

recommending that the full-scope plants evaluate relay

chatter in a manner consistent with the approach suggested in EPRI NP-6041 or equivalent. Note that considerable efficiency can be achieved using lessons learned

from the Hatch experience (Moore et al., 1990). For

plants in the focused-scope category, a low level of effort

is recommended. The lessons learned from the full-scope

reviews may necessitate reexamining the relay chatter

issues through the generic issues process (i.e.,

NUREG-0933).

Other differences between the full- and focused-scope

relate to the level of effort for evaluating soil failure

modes and the number of margin calculations (HCLPFs)

to be reported (Reed et al. 1990; Rasin, 1990). The difference in the level of effort in these areas stems primarily

from a perceived need for more accurately characterizing

plant behavior and numerical results for plants in the

full-scope category. It is also perceived that licensees of

plants in the focused-scope category will be able to identify important vulnerabilities with more liberal use of

fewer, approximate, and bounding-type analyses without

adverse impact. However, the actual level of effort in

these areas is very much site and plant dependent and

should be determined on the basis of plant-specific considerations. For example, a plant in the full- scope category that is located on a rock site will not perform any soil

failure evaluation, while a plant in the focused-scope

category that is located at a coastal plain site may require

more sophisticated investigations. In any case, discussions

here do not preclude the use of properly substantiated

judgments to define the scope and depth of an examination.

3.2.2 Review Level Earthquake and

Associated Spectral Shapes

The seismic margins methodology was designed to demonstrate sufficient margin over SSE to ensure plant safety

and to find any "weak links" that might limit the plant

shutdown capability to safely withstand a seismic event

bigger than SSE. The seismic margin method utilizes two

review or screening levels geared to peak ground accelerations of 0.3g and 0.5g. It is the staff's judgement that

the use of a 0.3g review level earthquake (RLE) for most

of the nuclear power plant sites in the Central and Eastern United States (east of the Rocky Mountains) would

serve to meet the objectives of the IPEEE. However, all

sites east of the Rocky Mountains are not subject to the

same level of earthquake hazard. For some sites where

the seismic hazard is low, a reduced-scope margin approach centered on walkdowns is acceptable. For western

sites other than California coastal sites, a 0.5g RLE

should be used for the margin approach. The RLEs

defined for U.S. sites, as well as sites that can perform a

reduced-scope SMM, are presented in Tables 3.1 and 3.2.

The seismic margin evaluations should utilize the

NUREG/CR-0098 median rock or soil spectrum

anchored at 0.3g or 0.5g depending on the g level and

primary condition at the site. Further discussion on the

review level earthquake is presented in Appendix A.

Plants in the 0.3g bin are further subdivided into full- and

focused-scope categories, as discussed earlier. This categorization is based on consideration of hazard as well as

the seismic design basis. Additional consideration is also

given to the outlier plants resulting from the Eastern U.S.

Seismicity (the Charleston Earthquake) Issue resolution.

Additional details are given in Appendix A.

The ground motion should be considered at the surface in

the free field. If secondary conditions such as shallow soil

conditions are being considered, appropriate procedures

should be used to determine the free-field motion in the

vicinity of those affected structures and components, and

the capacity evaluation of structures and components

should take into account the effects of soil-structure interaction.

Because recent ground motion estimates, such as those

included in the LLNL and EPRI hazard studies, indicate

relatively higher ground motion at frequencies greater

than 10 Hz than that shown in the NUREG/CR-0098

spectrum, the margin evaluation of only nonductile components (if appropriate)-for instance, relays-that are

sensitive to high frequencies should be performed as discussed in Section 3.2.4.2. No plant-specific response

analysis is anticipated to address concerns related to highfrequency ground motion. However, if a licensee decides

to evaluate plant response for high-frequency ground

motion, the response spectral shapes derived from the

11 NUREG-1407
3. The Seismic IPEEE

appropriate site-specific median uniform hazard response spectra (10,000-year return period) shown in

NUREG/CR-5250 anchored at 0.3g or 0.5g should be

used.

3.2.3 Reduced-Scope Margins Method

For sites where the seismic hazard is low, a reduced-scope

seismic margins method emphasizing the walkdown is

adequate. Well-conducted, detailed walkdowns have

been demonstrated to be the most important tool for

identifying seismic weak links whose correction is highly

cost effective. Applicable sites are identified in Table 3.1.

The initial steps of the full-scope margin methodology up

to and including the initial plant walkdown are performed

regardless of method selected (NRC or EPRI). Basically,

pertinent activities up to and including the initial plant

walkdown need to be performed. These activities include

gathering system information, classifying front-line systems and identifying front-line components, classifying

support systems and identifying support system components, and identifying plant-unique features.

Further guidance on the differences between the

reduced-scope and full-scope seismic margins methods,

that is, elements preserved and elements eliminated are

provided in Appendix B.

The evaluation should be documented in a walkdown

team report and subjected to a peer review (see Section 7).

3.2.4 NRC Seismic Margins Methodology

The guidance in NUREG/CR-4334, NUREG/CR-4482,

and NUREG/CR-5076 is supplemented by that in the

following sections to (1) reflect the partitioning of the

0.3g screening criteria into the reduced-scope, 0.3g

focused-scope, and 0.3g full-scope bins identified in Table 3.1 and (2) identify enhancements so that the method

can be used for IPEEE implementation.

3.2.4.1 Walkdown

Reduced Scope

See Appendix B, Sections B.1 (1) through (4) for guidance.

Emphasis on walkdowns also applies to containment and

containment systems (that is, containment functions required to prevent early failure, containment integrity,

isolation, and prevention of bypass), USI A-45, and

GI-131.

Focused Scope, Full Scope, and O.5g

The walkdown should be performed and documented in

accordance with the recommendations contained in EPRI

NP-6041.

3.2.4.2 Relay Evaluation

Relays in this context include such components as electric

relays, contactors, and switches that are prone to chatter.

The following paragraphs define the scope of the relay

chatter evaluation for each of the three bins:

Reduced Scope

USI A-46 Plants-Completion of the USI A-46 review

will satisfy the IPEEE intent.

Non A-46 Plants-No action is needed.

Focused Scope

USI A-46 Plants-Follow USI A-46 procedures. If lowseismic-ruggedness relays are discovered during the USI

A-46 review, the relay review should be expanded to

include relays outside the scope of USI A-46 but within

the scope of the IPEEE.

Non A-46 Plants-Locate and evaluate low-seismicruggedness relays (bad actor list).

Full Scope and O.5g (Including Western U.S. Plants)

USI A-46 Plants-Follow USI A-46 procedures for the

review; review systems within the scope of the IPEEE,

including those that are also within the scope of USI

A-46, using appropriate margin (EPRI NP-6041) or USI

A-46 procedures at the RLE.

Non A-46 Plants-Review the relays in all systems within

the scope of the IPEEE, using appropriate margin (EPRI

NP-6041) or USI A-46 procedures at the RLE.

The NRC method as originally developed did not address

the relay chatter issue because information on the subject

was lacking. Hardy et al. (1989), summarized the results of

several efforts in this area and provided guidance to address this issue in an IPEEE context. Relay chatter analysis could be resource intensive, and careful planning and

use of generic insights, if they are applicable to the plant,

are desirable. Insights and recommendations based on

the Hatch experience are available in Moore et al. (1990).

Attempts to address the concerns related to highfrequency ground motion by analysis is very likely to entail

extensive efforts, including the development of new and

much more complex building models that transmit and

amplify high-frequency input and generate accurate and

meaningful floor spectra at high frequencies. Estimates

NUREG-1407 12
3. The Seismic IPEEE

of high-frequency amplification in cabinets containing

relays will also have to be developed. Rather than using

analysis, the following approach is more suitable:

1. Prepare a list of relays that are known to have highfrequency sensitivity.

2. Screen relays that are known to have very high

HCLPFs (that is, eliminate them from further consideration without performing specific response calculations).

3. Assume that the remaining relays will chatter at the

review level earthquake.

4. Screen the remaining relays by showing either that

the electrical circuity is insensitive to high-frequency

chatter or that they can be recovered from changes

of state and associated false alarms.

5. Finally, replace the remaining relays with relays that

are not sensitive to high frequency (an alternative

approach is to show that the remaining relays are

rugged by conducting tests).

Although stated in the context of high-frequency ground

motion, this approach can be used to address the relay

chatter issue.

3.2.4.3 Soil Failures

Soil failure analyses include an evaluation for instability,

settlement, and liquefaction.

Reduced Scope

No evaluation is necessary.

Focused Scope

EPRI NP-6041 contains guidance; a review based on appropriate design and construction records is considered

adequate. A detailed analysis, as necessary, will be performed if soil failure is deemed significant.

Full Scope and 0.5g

EPRI NP-6041 contains guidance; it is anticipated that

existing soil test data will be adequate. An evaluation of

plant site conditions using state-of-the-art approaches

should be performed if soil failure is deemed significant.

3.2.4.4 Screening Criteria (Use of Screening Tables)

The screening guidance given in the Generic Implementation Procedure for Seismic Verification of Nuclear

Power Plant Equipment (GIP) may be used, provided a

review is conducted at the appropriate RLE, caveats included in margins reports are observed, and limitations on

the use of the generic equipment ruggedness spectrum

(GERS) are observed. Also, spatial interaction evaluation, such as assessing the effects of flooding, as noted in

EPRI NP-6041, is retained.

Reduced Scope

Appendix B contains guidance.

0.3g Bin--Focused and Full Scope

The criteria in the < 0.3g column of NUREG/CR-4334,

Table 5.1, or EPRI NP-6041, Tables 2-3 and 2-4, should

be used.

o.5g Bin

The criteria in the 0.3-0.5g column of NUREG/

CR-4334, Table 5.1, or EPRI NP-6041, Tables 2-3 and

2-4, should be used.

3.2.4.5 Seismic Input

Reduced Scope

For the evaluation, the SSE ground response spectra and

in-structure spectra should be used. New in-structure

response spectra, if developed, should be mean plus one

standard deviation to be consistent with the conservatism

in the design input. Any differences between the Final

Safety Analysis Report (FSAR) and new response spectra

should be highlighted and discussed.

Focused-Scope, Full-Scope, and 0.5g

For the evaluation, the NUREG/CR-0098 median rock

or soil spectrum (depending on primary condition at the

site) anchored at the assigned review level earthquake

should be used.

3.2.4.6 Evaluation of Outliers-HCLPF Calculations

Two approaches, fragility analysis (FA) and conservative

deterministic failure margin (CDFM), for computing

component and plant HCLPFs are acceptable. For the

NRC margins method, if the licensee initially chose the

CDFM method to calculate component and plant

HCLPF values, it is possible to make plant HCLPF statements with and without the inclusion of nonseismic failures and human actions by developing complete fragilities for the few components that remain in the plant-level

Boolean equations (optional).

As noted in EPRI NP-6041, use of the Generic Equipment Ruggedness Spectrum (GERS) to estimate

HCLPFs should take into account the latest results from

ongoing work on the reconciliation of GERS and HCLPF.

Reduced Scope

Outliers should be evaluated for the provisions in the GIP

if the plant is also in the USI A-46 Program. For elements

outside the USI A-46 scope (structures and piping) the

13 NUREG-1407
3. The Seismic IPEEE

requirements of the plant FSAR should be used in the

evaluation. Since the evaluation is done at the design

level, the outliers should be addressed in accordance with

10 CFR 50.72(b).

Focused Scope

The seismic capability evaluation engineers/seismic review team may use judgment to rank the unscreened

structures and equipment from the lowest to the highest.

The licensee should determine the number, scope, and

type of HCLPF analyses with the aim of identifying vulnerabilities and ranking them. Reed et al. (1990) and

Rasin (1990) suggest that HCLPF capacities should be

calculated for the lowest one-third of the ranked components; the remaining components should be assigned a

conservative HCLPF based on the highest calculated

HCLPFs. The assumed and calculated HCLPFs should

be reported.

Full Scope and O.5g

HCLPFs for unscreened structures and components

should be calculated as needed to accurately characterize

plant HCLPFs and vulnerabilities and rank them.

3.2.4.7 Nonseismic Failures and Human Actions

These activities should be included; guidance on including nonseismic failures and human actions is provided in

NUREG/CR-4826 (Maine Yankee evaluation) and in

two draft reports by Budnitz (1987 and 1990).

3.2.5 EPRI Seismic Margins Methodology

The guidance provided in EPRI NP-6041 is supplemented by that in the following sections to (1) reflect the

partitioning of the 0.3g screening criteria into the

reduced-scope, 0.3g focused-scope and 0.3g full-scope

bins identified in Table 3.1, and (2) identify enhancements so that the method can be used for the implementation of the IPEEE.

3.2.5.1 Selection of Alternative Success Paths

The EPRI SMM as currently constituted calls for evaluation of a preferred path and an alternative path. The

NRC panel that reviewed the EPRI methodology recommended:

"... that a reasonably complete set of potential success

paths be set down initially, rather than a very small number, since limiting the number of success paths too quickly

can prevent the identification of some plant-level

HCLPF insights, and can mask plant differences regarding defense-in-depth. The Panel believes that preliminary

analysis to narrow the number of paths to the required

two or three should begin with the fuller set, and it recommends that this narrowing be documented in detail."

For IPEEE purposes, it is desirable that, to the maximum

extent possible, the alternative path involve operational

sequences, systems, piping runs, and components different from those used in the preferred path. The procedure

used in the trial application of the EPRI methodology

(EPRI NP-6359) provides an acceptable approach for use

in selecting success paths (preferred and alternative).

3.2.5.2 Walkdown

Same as Section 3.2.4.1.

32.5.3 Relay Evaluation

Same as Section 3.2.4.2.

3.2.5A Soil Failures

Same as Section 3.2.4.3.

3.2.5.5 Screening Criteria (Use of Screening Tables)

Same as Section 3.2.4.4.

3.2.5.6 Seismic Input

Same as Section 3.2.4.5.

32.5.7 Evaluation of Outliers-HCLPF Calculations

Same as Section 3.2.4.6.

3.2.5.8 Nonseismic Failures and Human Actions

Success paths are chosen based on a screening criterion

applied to nonseismic failures and needed human actions.

It is important that the failure modes and human actions

are clearly identified and have low enough probabilities to

not affect the seismic margins evaluation. The screening

criteria used in the Maine Yankee margin evaluation

(NUREG/CR-4826) addressing both single-train and

multi-train systems is an acceptable approach. The redundancies along a given success path should be specifically

analyzed and documented when they exist. (In a complementary sense, where a single component is truly "alone"

in performing a vital function along a success path, this

should be highlighted too). This information will serve to

indicate the extent to which a single failure would or

would not invalidate the plant's ability to respond safely

to a given earthquake level.

3.2.6 Containment Performance

The primary purpose of the evaluation for a seismic event

is to identify vulnerabilities that involve early failure of

containment functions. These include containment

integrity, containment isolation, prevention of bypass

functions, and some specific systems depending on a containment design (e.g., igniters, suppression pools, ice baskets). The analyses performed for internal events IPE

NUREG-1407 14
3. The Seismic IPEEE

should be used to determine the scope of systems for the

examination.

Each licensee should develop a plan to address containment performance during a seismic event consistent with

the above-defined purpose. Additional guidance (no requirements implied) on extending margin-type approaches to obtain containment insights is contained in

Budnitz 1991a and 1991b, and Reed, et al., 1990. Some

general guidance is provided here based on past PRA

experience and some generic capacity estimates of typical

components involved in containment systems. From a

survey of past PRAs (Amico, 1989), it appears that highconsequence sequences involve gross structural failure of

the containment itself or failure of major equipment or

structures within the containment at very high accelerations (HCLPF values greater than 0.5g) and isolation

failure due to seismically induced relay chatter.

Generally, containment penetrations are seismically rugged; a rigorous fragility analysis is needed only at review

levels greater than 0.3g, but a walkdown to evaluate for

unusual conditions (e.g., spatial interactions, unique

penetration configurations) is recommended. An evaluation of the backup air system of the equipment hatch and

personnel lock that employ inflatable seals should be

performed at all review levels. Also, some penetrations

need cooling, and the possibility and consequences of a

cooling loss caused by an earthquake should be considered.

Valves involved in the containment isolation system are

expected to be seismically rugged (NUREG/CR-4734). A

walkdown to ensure that they are similar to test data and

have known high capacities and that there are no spatial

interaction issues will sufrice. Seismic failures of actuation and control systems are more likely to cause isolation

system failures and should be included in the examination. For valves relying on a backup air system, the air

system should also be included in the seismic examination.

Components of the containment heat removal/pressure

suppression functional system that are not included elsewhere and are not known to have high capacities should

be examined. An example of such a component might be a

fan cooler unit supported on isolator shims. The walkdown should include examination of such components

and their anchorages. Similarly, support systems and

other system interaction effects (e.g., relay chatter)

should be examined as applicable.

For Mark I and ice condenser containments, it is not

feasible to screen out components (e.g., torus, ice basket

support) on a generic capacity basis. The potential for

accident sequences initiated by a containment functional

failure should be examined.

3.3 Optional Methodologies

A licensee may request the staff to review any other

systematic examination method to determine its acceptability for IPEEE purposes.

15 NUREG-1407
4 ACCEPTABLE METHODOLOGY FOR PERFORMING

THE INTERNAL FIRES IPEEE

For purposes of an IPEEE, a Level 1 probabilistic risk

assessment (PRA) is considered acceptable to identify

potential internal fire vulnerabilities at nuclear power

plants. Some fire issues identified in the Fire Risk Scoping Study, (1) seismic/fire interactions, (2) effects of fire

suppressants on safety equipment, and (3) control system

interactions, should be addressed in the IPEEE. The

walkdown procedures of the IPEEE should address the

above issues and should be specifically tailored to assess

the potential vulnerabilities related to these issues. The

licensee should use a plant-specific data base on fire brigade training in the IPEEE to assess the effectiveness of

manual fire fighting to determine the response time for

the manual fire fighters. The licensee should also show

the effectiveness of fire barriers in the IPEEE. The current fire PRA method has its limitations (NUREG/CR5088, 1989) and significant "engineering judgment" must

be brought to bear once the PRA has been accomplished

to allow for sensible application of the results. The staff

believes that the type of "engineering judgment" needed

to interpret the results of a PRA is fully within the competence of most fire-safety experts, including experts within

the regulatory staff. Further, despite current limitations

in the methodology, a fire "vulnerability search" in the

spirit of the Severe Accident Policy Statement and the

IPE exercise is feasible, and such a vulnerability search

need not wait for the completion of further methodology

development. Finally, in meeting the objectives of the

IPEEE, it is desirable to focus on relative insights rather

than on absolute core damage frequency.

4.1 New Fire PRA Analysis

There are several different approaches for the analysis

of fires (NUREG/CR-2300, 1983, NUREG/CR-2815,

1985, NUREG/CR-4840, 1990, and NUREG/CR-5259,

1990). Although not all fire PRAs delineate their analysis

steps in exactly the same way, the following steps, in one

form or another, should be part of any analysis.

4.1.1 Identify Critical Areas of Vulnerability

The criterion is whether a fire could compromise important safety equipment. Emphasis should be placed on

areas where multiple equipment could be compromised,

in particular, several trains of redundant equipment to

perform the same safety function. Attention should be

given to the potential for cross-zone spread of fire and the

likelihood that transient fuels might supplement fuels

already present in a zone.

4.1.2 Calculate the Frequency of Fire

Initiation in Each Area

This calculation is sensitive to location within a larger

area, particularly if fuel loading conditions, cross-zone

spreading potential, or other idiosyncrasies are considered. Also, the data base on fires in various areas should

be coupled with location-specific information obtained

from the plant walkdown and other experience to account

for uncertainties.

4.1.3 Analyze for the Disabling of Critical

Safety Functions

Determine the likelihood of equipment being disabled by

a fire. The areas to be addressed include:

1. Fire growth and spread, including the treatment of

hot gases and smoke.

2. Detection/suppression effectiveness and reliability.

3. Component fragility to fire and combustion products.

4. Probability estimates (distributions) for fault tree

quantification.

4.1.4 Identify Fire-Induced Initiating Events/

Systems Analysis

Perform the analysis to determine the frequency of fire

initiated accident sequences leading to core damage.

4.1.5. Perform Containment Analysis

Perform containment analysis if containment failure

modes differ significantly from those found in the IPE

internal events evaluation.

Perform in a fashion similar to an internal-initiator PRA.

4.2 Use of an Existing Fire PRA

The use of an existing fire PRA for the internal fires

IPEEE is acceptable provided the PRA reflects the current as-built and as-operated status of the plant and the

licensee addresses the deficiencies of past PRAs that are

identified in the Fire Risk Scoping Study (NUREG/

CR-5088). Deficiencies may include the use of low conditional failure probabilities for dampers and penetrations,

no consideration of damage from the use of fire suppressants, inappropriate estimates of the effectiveness of

manual fire fighting, and no consideration of seismic/fire

interactions.

4.3 Optional Methodologies

A licensee may request the staff to review any other

systematic examination method to determine its acceptability for IPEEE purposes.

NUREG-1407 16
5 ACCEPTABLE METHODOLOGY FOR PERFORMING THE HIGH WINDS,

FLOODS, AND TRANSPORTATION AND NEARBY FACILITY ACCIDENT

IPEEE

For the purposes of an IPEEE, the staff recommends a

progressive screening approach to identify potential vulnerabilities at nuclear power plants due to high winds,

floods, and transportation and nearby facility accident.

The owners of Trojan and Washington Nuclear Plant 2,

who are requested to evaluate the effects of volcanic

activities in assessing severe accident vulnerabilities,

should determine if the recommended screening approach is applicable to their unique situation.

5.1 Introduction

It is assumed that the IPE for internal events will be in

progress or completed when the portion of the IPEEE

pertaining to high winds, floods and transportation and

nearby facility accident is being performed. Some external events will be addressed in the internal events IPE

analyses (e.g., the primary effect of lightning is loss of

offsite power, which is included in the internal events

analyses); other external events will have been screened

from further consideration by the staff. For those external

events not in either of these categories, further consideration using the progressive screening approach shown in

Figure 5.1 is recommended.

5.2 Analytical Procedure

The steps shown in Figure 5.1 represent a series of analyses in increasing level of detail, effort, and resolution.

However, the licensee may choose to bypass one or more

of the optional steps as long as the 1975 Standard Review

Plan (SRP) (NUREG-75/087) criteria are met or the

potential vulnerabilities are either identified or demonstrated to be insignificant.

In general, the containment structure, equipment hatch,

personnel air lock, and other penetrations are designed

and constructed to have high capacities in resisting the

effects of high winds, floods, and overpressure induced by

transportation or nearby facility accidents. Therefore, no

additional containment performance assessment (beyond

that discussed for the seismic portion of the IPEEE in

Sections 3.1.1.5, 3.1.2, and 3.2.6) is needed unless a licensee predicts or identifies plant-unique accident sequences

different from those determined by the internal events

IPE.

5.2.1 Review Plant-Specific Hazard Data and

Licensing Bases

All licensees should review the information on plant design hazard and the licensing bases, including the resolution of each event.

5.2.2 Identify Significant Changes Since OL

Issuance

All licensees should review the site for any significant

changes since the operating license was issued with respect to (1) military and industrial facilities within 5 miles

of the site, (2) onsite storage or other activities involving

hazardous materials, (3) transportation, or (4) developments that could affect the original design conditions.

5.2.3 Determine if the Plant/Facilities Design

Meets 1975 SRP Criteria

All licensees should compare the information obtained

from the review discussed in Sections 5.2.1 and 5.2.2 for

conformance to 1975 SRP criteria and perform a confirmatory walkdown of the plant. The walkdown would concentrate on outdoor facilities that could be affected by

high winds, onsite storage of hazardous materials, and

offsite developments. If the comparison indicates that the

plant conforms to the 1975 SRP criteria and the walkdown reveals no potential vulnerabilities not included in

the original design basis analysis, it is judged that the

contribution from that hazard to core damage frequency

is less than 10-e per year and the IPEEE screening criterion is met.

Otherwise or if a licensee knows that the 1975 SRP criteria will not be met, it should take one or more of the

optional steps given in Sections 5.2.4, 5.2.5, and 5.2.6 to

further evaluate the situation.

5.2.4 Determine if the Hazard Frequency is

Acceptably Low (Optional Step)

If the original design basis does not meet current regulatory requirements, the licensee may choose to demonstrate that the original design basis is sufficiently lowthat is, less than 10-5 per year, and the conditional core

damage frequency is judged to be less than 10-1.

If the original design basis hazard combined with the

conditional core damage frequency is not sufficiently low

(i.e., less than the screening criterion of 10-8 per year),

additional analysis may be needed.

17 NUREG-1407
5. The Accident 1PEEEs

1) Review Plant Specific Hazard Data and Ucensing Bases

( R(FSAR) I

1 (2) Identify Significant Changes, if any, since OL Issuance I

SNO J (3) Does Plant/Facilities Design Meet 1975 SRP Criteria?

"I (Quick Screening & Walkdown)

YES

OR .] (4) Is the Hazard Frequency Acceptably Low? I

NOj

OR >J (5) Bounding Analysis P I (Response/Consequence)

NO~

OR 30-, 1 (6) P R A I

(7) Documentation

(Incl. Identified Reportable Items and Proposed Improvements

Figure 5.1 Recommended IPEEE Approach for Winds, Floods, and Others

5.2.5 Perform a Bounding Analysis

(Optional Step)

This analysis is intended to provide a conservative calculation showing that either the hazard would not result in

core damage or the core damage frequency is below the

reporting criterion. The level of detail is that level needed

to demonstrate the point; judgment is needed for determining the proper level of detail and needed effort.

5.2.6 Perform a Probabilistic Risk

Assessment (Optional Step)

A probabilistic risk assessment (PRA) consists of the following key elements: hazard analysis, fragility evaluation,

plant systems and accident analysis (event/fault trees),

and radioactive material release analysis. The detailed

procedure is described in NUREG/CR-2300, NUREGI

CR-2815, and NUREG/CR-5259. If the core damage

frequency is less than 10-3 per year, the event need not be

considered further. The level of detail is that level needed

to conclude that the core damage frequency is low or to

find vulnerabilities.

5.3 Optional Methodologies

A licensee may request the staff to review any other

systematic examination method to determine its acceptability for IPEEE purposes.

NUREG-1407 18
6 COORDINATION WITH ONGOING PROGRAMS

6.1 Introduction

If unnecessary duplication of effort is to be avoided, coordination with ongoing programs is necessary. The first

level of coordination consists of the three major elements

related to the implementation of the Severe Accident

Policy, that is, coordination of the IPEEE with the internal events IPE and accident management. The second

level of coordination consists of the three major elements

of the IPEEE, that is, seismic events, internal fires, and

high winds, floods, and others. The third level of coordination consists of each major element of the IPEEE, for

example, seismic events, and the ongoing programs related to that element.

6.2 Description of Ongoing Programs

6.2.1 IPE Program Related to Internal

Events

In Generic Letter 88-20, the NRC requested that the

licensee of each plant to perform a systematic examination to identify any plant-specific vulnerabilities to

severe accidents and to report the results to the staff. The

process was defined as an individual plant examination

(IPE). Licensees were requested to proceed with the examinations for internally initiated events only (including

internal flooding). Examination of externally initiated

events would proceed separately and on a later schedule.

However, while performing the IPE for internally initiated events, licensees were advised to document and retain plant-specific data relevant to external events so that

they can be readily retrieveu in a convenient form when

needed for later external event analyses.

6.2.2 Programs Related to External Events

6.2.2.1 Seismic Programs

The following is a brief description of the programs related to seismic events:

1. USI A-17, "System Interactions in Nuclear Power

Plants," addresses NRC's concerns regarding the

interaction of various systems with regard to

whether actions or consequences could adversely

affect the redundancy and independence of safety

systems. The evaluation of system interactions related to internal events and internal floods is included in the IPE (GL 88-20). The evaluation of

spatial system interaction under seismic conditions

(the SSE) is included in USI A-46.

2. USI A-40, "Seismic Design Criteria," investigates

selected areas of the seismic design process. The

staff identified alternative approaches to certain design procedures and modifications to the NRC criteria in the Standard Review Plan to reflect the current state of the art and industry practice. The

concern for the seismic capacity of safety-related

above-ground tanks (at the SSE) is included in USI

A-46.

3. USI A-45, "Shutdown Decay Heat Removal

Requirements," has the objective of determining

whether t6e decay heat removal function at operating plants is adequate and if cost-beneficial improvement could be identified. USI A-45 was subsumed in

the IPE (GL 88-20); therefore, the external event

aspects including the seismic adequacy of the decay

heat removal system should be included in the

IPEEE.

4. USI A-46, "Verification of Seismic Adequacy of

Equipment in Operating Plants" has developed an

alternative method and acceptance criteria (to current licensing requirements) to verify the seismic

adequacy of equipment in some operating plants

with construction permit applications docketed before about 1972. All these plants will be reviewed to

the existing safe shutdown earthquake (SSE). The

scope of USI A-46 has been expanded to cover the

seismic spatial system interaction of USI A-17 and

the concern of USI A-40 for the seismic capability of

large safety-related above-ground tanks.

5. GI-131, "Potential Seismic Interaction Involving

the Movable In-Core Flux Mapping System Used in

Westinghouse Plants," was identified because portions of the in-core flux mapping system that have

not been seismically analyzed are located directly

above the seal table. Failure of this equipment during a seismic event could cause multiple failures at

the seal table and could produce an equivalent

small-break LOCA.

6. The "Eastern U.S. Seismicity Issue" (formerly the

Charleston Earthquake Issue) came about as a result of a U.S. Geological Survey letter in 1982 that

pointed out the possibility that large, damaging

earthquakes have some likelihood of occurring at

locations that had not been considered in licensing

decisions. The staff initiated the Seismic Hazard

Characterization Project at JLLNL, which provided

probabilistic seismic hazard estimates for all nuclear

power plant sites east of the Rocky Mountains. A

similar project was carried out by EPRI for the electric utility industry. The staff's purpose in evaluating

the probabilistic studies has been to identify plants

in the Central and Eastern United States where past

19 NUREG-1407
6. Coordination with Ongoing Programs

licensing decisions may have resulted in their being

outliers with respect to seismic hazard, that is, the

likelihood of exceeding their design bases. As a result of the probabilistic analyses performed, eight

plants at five Eastern U.S. sites were identified as

being outliers. The IPEEE will provide a resolution

for the outlier plants with no need for additional

analyses or documentation from the licensees.

6.22.2 Internal Fires Programs

The following is a brief description of programs related to

internal fires:

1. NUREGICR-5088, "Fire Risk Scoping Study,"

identifies some fire issues that had not previously

been addressed in the fire PRAs: fire growth code,

seismic/fire interaction, fire barrier effectiveness,

manual fire fighting effectiveness, effects of fire suppressants on safety equipment, and control system

interactions. A plant-specific analysis (including a

specifically tailored walkdown) should be performed

to assess the actual risk impact of these issues at a

plant.

2. GI-57, "Effects of Fire Protection System Actuation

on Safety-Related Equipment," assesses the impact

of inadvertent actuation of fire protection systems

on safety systems. This is one of the issues identified

in the Fire Risk Scoping Study. The industry,

through EPRI, has a program collecting data on the

effects of suppressants on the safety equipment.

3. USI A-45, "Shutdown Decay Heat Removal Requirements," was initiated to determine if the decay

heat removal function at operating plants is adequate and if cost-beneficial improvement could be

identified. The USI A-45 was subsumed in the IPE

(GL 88-20); therefore, the adequacy of the decay

heat removal system under internal fires should be

addressed in the fire IPEEE.

6.2.2.3 External Flooding Program

GI 103, "Design for Probable Maximum Precipitation

(PMP)," is a related external flooding issue. The staff

provided the resolution of this issue to all licensees in

Generic Letter 89-22, dated Oct. 19, 1989. Specifically,

the NRC requested that future plants be designed against

a new PMP criterion. For existing plants, the NRC recommended that licensees review the information contained

in Generic Letter 89-22 and determine if they need to

take additional action. For the IPEEE, the severe accident risk from PMP should be assessed. The licensees

should assess the effects of applying this new PMP criterion to their plants in terms of onsite flooding and roof

ponding to determine whether that would lead to severe

accidents.

6.3 Approach on Coordination with

Ongoing Programs

If duplication of effort by the staff and licensees is to be

avoided, it is important that the above ongoing programs

be coordinated.

6.3.1 Coordination With Internal Events

Program (IPE)

The coordination between the internal events IPE and

the IPEEE can be categorized into three phases:

preanalyses planning, plant modifications, and accident

management.

63.1.1 Preanalyses Planning

Considerations include (1) definition of elements and

their boundaries, (2) walkdown procedures and spatial

interactions, and (3) composition of the peer review

group. It is likely that the IPE will precede the IPEEE.

Careful planning, taking into account the above considerations, will avoid a duplication of effort by the licensee.

63.1.2 Plant Modifications

Since the IPE and the IPEEE are likely to be performed

separately, it is imperative to examine the impact of modifications identified during the IPE on external events and

vice versa. The staff examined several PRAs that included

both internal events and external events (Bohn, 1989) to

identify possible interactions. Highlights of the major

findings in the seismic area (which to some extent are

applicable to fire), are the following:

1. In general, the modifications proposed as a result of

the internal events analysis would not adversely

affect the seismic or fire risk, provided the modifications do not become weak links.

2. In general, the modifications made could potentially

contribute to an increase in risk at the plant in the

following ways:

a. Many of the modifications proposed may involve adding valves or suction lines to existing

systems. Thus, the possibilityi of violating the

pressure boundary and creating a potential diversion exists if the modification were to fail

during an earthquake. Also, modifications may

involve routing different trains of electrical

power or power from adjacent units. The possibility exists that the circuitry could be designed

in such a way that failure of non-safety-related

electrical components could actually defeat the

circuitry that was desired to provide redundancy, and

NUREG-1407 20
6. Coordination with Ongoing Programs

b. The possibility exists that inadequate anchorage could defeat the planned redundancy during a seismic event.

3. The potential adverse effects of the modifications

include:

a. Poor accessibility for maintenance,

b. Stiffening of systems leading to higher stress

due to thermal cycles during normal plant operation.

The cited study (Bohn, 1989) provides specific examples

of modifications and their effects on other initiating

events.

6.3.1.3 Accident Management

Guidance on the integration of findings from the IPEEE

and accident management is being developed (SECY89-308, Oct. 1989).

6.3.2 Coordination Among External Events

Programs

The issue of integration between external events primarily involves interactions between seismic events and fires

and seismic events and floods. Seismically induced fires

and floods are to be addressed as part of the IPEEE. The

effects of seismically induced fires and the impact of inadvertent actuation of fire protection systems on safety systems should be addressed. The effects of seismically induced external flooding and internal flooding on plant

safety should be included. The scope of the evaluation of

seismically induced floods, in addition to that of the external sources of water (e.g., tanks, upstream dams), should

include the evaluation of some internal flooding consistent with the discussion in Appendix I of EPRI NP-6041.

The coordination between the seismic and the fire or

flood analysts should be based on the following:

1. The seismic analysts should generally search for and

identify the initiating events (certain specific seismically initiated failures of equipment or structures)

that can cause fires or floods, and

2. The seismic and fire or flood analysts should also

discuss other concurrent seismically induced failures or possible effects on human actions and then,

proceed to complete the rest of the IPEEE analysis.

The coordination should include a meeting, prior to seismic walkdown, in which the fire and flood analysts discuss

the key issues, how the analysis will be done, and what to

look for. The fire or flood analyst may need to participate

in parts of the seismic walkdown or revisit the areas identified during the seismic walkdown to grasp the issues

from the seismic-capacity point of view.

6.33 Coordination With Seismic Programs

A number of programs related to seismic events requiring

licensee action have been identified. Many of these programs have arisen as a result of the changing perception

of hazards and revisions in the design and qualification

criteria. There are two categories of seismic programs as

they relate to the seismic IPEEE. The first category involves programs, e.g., USI A-45, "Shutdown Decay Heat

Removal Requirements", GI-131, "Potential Seismic Interaction Involving the Movable In-Core Flux Mapping

System Used in Westinghouse Plants," and "the Eastern

U.S. Seismicity Issue," that have been subsumed into the

IPE/IPEEE. USI A-45 and GI-131 should be specifically

addressed as part of the seismic IPEEE. The Eastern U.S.

Seismicity Issue requires no additional licensee actions or

reporting. The second category involves programs (e.g.,

USI A-46, "Verification of Seismic Adequacy of Equipment in Operating Plants,") that can be coordinated with

the seismic IPEEE. The coordination of these programs

with the seismic IPEEE is most beneficial in reducing the

resources expended by the licensee and the staff.

6.33.1 USI A-45 and GI-131

The methodology used in the seismic IPEEE can also be

used to address USI.A-45 and GI-131. The systems and

components for addressing USI A-45 will have been determined by the internal events IPE, and the purpose of

the seismic IPEEE is to identify any significant and

unique seismic vulnerabilities in the decay heat removal

function. In addition, the seismic IPEEE will evaluate the

potential seismic interaction of the movable in-core flux

mapping system used in Westinghouse plants.

Capacities of decay heat removal components can be

established using either the fragility analysis (FA) or

conservative deterministic failure margin (CDFM)

approaches depending upon the methodology chosen to

implement the seismic IPEEE. Thus resolution of these

issues can be easily accomplished during the seismic

IPEEE evaluation.

The potential interaction between the seal table and nonCategory I seismic systems associated with the movable

in-core flux mapping system can be identified during the

seismic walkdown of the IPEEE. If needed, the component capacities or consequences of component failure can

be evaluated using the same procedures that are used in

the seismic IPEEE.

6.3.3.2 The Eastern U.S. Seismicity Issue (The

Charleston Earthquake Issue)

As a result of work carried out to resolve the Eastern U.S.

Seismicity Issue (Charleston Earthquake Issue), probabilistic seismic hazard estimates exist for all nuclear power

21 NUREG-1407
6. Coordination with Ongoing Programs

plants east of the Rocky Mountains. These should be used

directly by any licensee in that region opting to satisfy the

seismic IPEEE with a seismic PRA. The hazard estimates

also played a key. role in determining the review level

earthquake used in the seismic margins methodology option. The IPEEE will provide a resolution for this issue

without requiring additional analyses or documentation

from licenseesL

6.3.3.3 USI A-46

Implementation of the USI A-46 program involves plants

with construction permit applications docketed before

about 1972. The USI A-46 plants thus form a subset of all

the nuclear power plants in the U.S. that are requested to

perform the seismic IPEEE.

The most efficient way to address the ongoing seismic

programs for USI A-46 plants is to conduct the A-46

review and walkdown to gather relevant information for

the seismic IPEEE. In order to facilitate this approach,

the activities of USI A-46 and the seismic IPEEE need to

be coordinated, and the plant walkdown needs to be well

planned. Several inherent differences between the A-46

program and the seismic IPEEE should be noted at the

outset before attempting to coordinate the two programs.

First, the objectives are quite different. The USI A-46

program has licensing implications on plant operation;

this program will assess and ensure the seismic ruggedness of safety-related equipment in a plant to withstand

the SSE. The seismic IPEEE, on the other hand, generally tries to identify plant vulnerabilities when subjected

to earthquake levels higher than the SSE design basis.

Second, the scope of the reviews are different. USI A-46

is concerned with only one success path (with some requirement on equipment redundancy) of equipment

needed to bring the plant to safe shutdown in the event of

an earthquake and maintain it there for at least 72 hours.

The scenario considers an earthquake of the SSE level

with a possible loss of offsite power because of this earthquake. The probabilities of a seismically induced LOCA

(small or large) and a high-energy line break (HELB)

occurring are judged to be low enough that their consideration at this earthquake level is not warranted. Piping,

tubing, and structures will be examined during a walkdown only if they have the potential to cause seismic

interaction with the equipment reviewed and cause damage to this equipment. The review of above-ground tanks

(as part of USI A-40) is an exception. The seismic IPEEE

is concerned with the vulnerabilities of the whole plant,

not just the equipment. Also, evaluations are generally

made at levels above the design basis. At this level of

earthquake, seismically induced LOCAs are considered,

and mitigating systems and equipment to address this

initiator are reviewed. Therefore, even if the EPRI seismic margins methodology is utilized to implement the

seismic IPEEE (since it is quite similar to the USI A-46

evaluation), it would need additional equipment to be

reviewed over that required for implementing USI A-46.

Third, the levels of review and walkdown are different.

The Seismic Qualificatiori Utility Group (SQUG) and

EPRI have developed a detailed Generic Implementation Procedure (GIP) for the USI A-46 review and walkdown that was reviewed by the NRC staff, and a Safety

Evaluation Report (SER) was issued. The GIP should be

followed in performing the USI A-46 review and walkdown. The guidelines associated with the seismic PRA or

seismic margins methodology are not as specific as those

in the GIP. To illustrate this point, in the walkdown review of expansion anchor bolts, GIP calls for the use of a

wrench test for the bolt tightness check, whereas the

margins walkdown ensures only that the anchor bolts are

adequate to hold down the equipment as designed with no

specific testing requirements to confirm anchor capacity.

The completion of the seismic IPEEE does not automatically mean that the USI A-46 review is satisfactorily completed.

There may be overlaps or differences in the equipment

scope for USI A-46 and the seismic IPEEE. For equipment that is within the scope of USI A-46 or the seismic

IPEEE only, it is clear that either GIP or IPEEE guidelines, respectively, should apply. For the overlapping

equipment, the efficient approach is to use the GIP for

both walkdowns; however, the IPEEE should use the

review level earthquake. Caveats and interaction (such as

flooding) provisions of EPRI NP-6041 should be observed.

In summary, it is recommended that licensees coordinate

the information collection for the USI A-46 and seismic

IPEEE review and walkdown in order to minimize or

avoid duplication of effort by the licensees and staff. Care

should be exercised in the coordination to ensure that the

objectives of both programs are fulfilled. Coordination of

the two programs has been shown to be feasible in the

trial evaluation of the Hatch plant using the EPRI seismic

margins methodology.

6.3.4 Coordination With Other Issues

In addition to the specific USIs and GIs discussed above,

if, during its IPEEE, a licensee (1) discovers a notable

vulnerability that is topically associated with any other

USI or GI and proposes measures to dispose of the specific safety issue or (2) concludes that no vulnerability

exists at its plant that is topically associated with any USI

or GI, the staff will consider the US! or GI resolved for a

plant upon review and acceptance of the results of the

IPEEE. The following should be discussed:

NUREG-1407 22
6. Coordination with Ongoing Programs

a. The ability of the methodology to identify vulnerabilities associated with the USI or GI being addressed.

b. The contribution of each USI or GI to core damage

frequency or unusually poor containment performance, including sources of uncertainty when PRA is

used.

c. The technical basis for resolving the issue.

23 NUREG-1407
7 PEER REVIEW

In Supplement 4 to Generic Letter 88-20, the staff requests that each licensee conduct a peer review by individuals who are not associated with the initial evaluation

to ensure the accuracy of the documentation and to validate both the IPEEE process and its results. The submittal should contain, as a minimum, a description of the

review performed, the results of the review team's evaluation, and a list of the review team members.

The purpose of the peer review is twofold. The first purpose is to provide quality control and quality assurance to

the IPEEE process. Independence of the review team is

desirable because it reflects a quality control and quality

assurance attitude. This does not imply that the staff is

seeking a review or document control similar to that

specified in Appendix B to 10 CFR.Part 50. The staff does

seek to ensure that the IPEEE process produces reliable,

factual information. If it is necessary to use a reviewer

who is not totally removed from the plant-specific IPEEE

process, the licensee should be confident that the reviewer can be objective and capable of providing a critical

review. The "in-house" team can be supplemented by outside consultants as determined appropriate by the licensee.

The second purpose of the review relates to the importance of having utility personnel cognizant of the IPEEE.

The maximum benefit to the utility would occur if the

combination of persons involved in the original analysis

and peer review, taken as a group, provide both a cadre of

utility personnel to facilitate the continued use of the

results and the expertise in the methods to ensure that the

techniques have been correctly applied. The staff expects

all utilities to have in-house personnel who have the most

expert knowledge of their plants, system configurations,

and operating practices and procedures.

The staff recommends that the peer review team have

combined experience in the areas of systems engineering

and specific external events. For example, the seismic

peer review team should have combined experience in the

areas of sytems engineering, seismic capacity engineering,

and seismic PRAs or seismic margins methodologies.

7.1 Seismic Related Insights

Three trial plant applications of the seismic margins

methodologies have demonstrated that considerable

judgment is involved in applying these methods, and that

peer review groups have considerably aided this judgment

and have also enhanced the overall credibility of the studies (Davis, 1990).

A meeting of the peer review team with the seismic review team prior to the plant walkdown will provide insights into the appropriateness of the proposed plant

walkdowns. Also, the peer review team's endorsement of

the final results will add credibility to the HCLPF

estimates.

NUREG-1407 24
8 DOCUMENTATION AND REPORTING

The IPEEE should be documented in a traceable manner

to provide the basis for the findings. This can be dealt with

most efficiently by a two-tier approach. The first tier

consists of the results of the examination, which will be

reported to the NRC for review. The second tier is the

documentation of the examination itself, which should be

retained by the licensee for the duration of the license

unless superseded.

The information submitted to the NRC should be organized and presented in accordance with Appendix C. The

submittal may enable many issues to be dealt with in the

IPEEE review. Pertinent issues are discussed in Section 6.

For some issues, for example, USI A-46, a detailed documentation requirement exists, and it should be followed

in the broad framework of IPEEE submittals. Specific

information relevant to particular issues, e.g., USIs and

GIs, should be identified.

8.1 Information Submitted to the NRC

A detailed list of information to be submitted to the NRC

is provided in Appendix C.

8.2 Information Retained for Audit

Retained documentation should include applicable event

trees and fault trees, current versions of the system notebooks (if applicable), walkdown reports, and the results of

the examination. In general, all documents essential for a

practitioner in the field to understand what was done in

the IPEEE should be retained. In addition, the manner in

which the validity of these documents has been ensured

should be documented. If credit is allowed in the IPEEE

for any actions taken by the operators, the licensee should

have established plant procedures to be used by the plant

staff responsible for managing a severe accident should

one occur. Procedures should provide assurance that the

operators can and will take the proper action.

25 NUREG-1407
9 REFERENCES

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1987.

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-- , EPRI NP-6395-D, "Probabilistic Seismic Hazard

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NUREG-1407 26
9. References

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-- , NUREG-1032, "Evaluation of Station Blackout

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1978.

--- , NUREG/CR-2300, "PRA Procedures Guide,"

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-- , NUREG/CR-2815, "Probabilistic Safety Analysis

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--- , NUREG/CR-4334, "An Approach to the Quantification of Seismic Margins in Nuclear Power Plants,"

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March 1986.

-- , NUREG/CR-4659, "Seismic Fragility of Nuclear

Power Plant Components, Phase I," Vol. 1, June 1986.

--- , NUREG/CR-4659, "Seismic Fragility of Nuclear

Power Plant Components, Phase II, Motor Control Center, Switchboard, Panelboard and Power Supply," Vol. 2,

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--- , NUREG1CR-4659, "Seismic Fragility of Nuclear

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-- , NUREG/CR-4734, "Seismic Testing of Typical

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1986.

- -, NUREG/CR-4826, "Seismic Margin Review of the

Maine Yankee Atomic Power Station," Vols. 1-3, March

1987.

-- , NUREG/CR-4840, "Recommended Procedures

for the Simplified External Event Risk Analysis for

NUREG-1150," Sandia National Laboratory, September 1989.

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-- , NUREG/CR-5042, Supplement 1, "Evaluation of

External Hazards to Nuclear Power Plants in the United

States-Seismic Hazard," April 1988.

-- , NUREG/CR-5042, Supplement 2, "Evaluation of

External Hazards to Nuclear Power Plants in the United

States-Other External Events," February 1989.

NUREG/CR-5076, "An Approach to the Quantification of Seismic Margins in Nuclear Power Plants: The

Importance of BWR Plant Systems and Functions to Seismic Margins," May 1988.

--- , NUREG/CR-5088, "Fire Risk Scoping Study,"

January 1989.

--- , NUREG/CR-5250, "Seismic Hazard Characterization of 69 Nuclear Power Plant Sites East of the Rocky

Mountains," Vols. 1-8, January 1989.

-- , NUREG/CR-5259, "Individual Plant Examination

for External Events: Guidance and Procedures," draft,

March 1989.

-- , NUREG/CR-5270, "Assessment of Seismic Margin Calculation Methods," March 1989.

--- , NUREG/CR-5477, "An Evaluation of the Reliability and Usefulness of External Initiator PRA Methodologies," January 1990.

--, NUREG/CR-5499, "Guidance on Relay Chatter

Effects," 1991.

--- , NUREG/CR-5501, "Selection of Review Level

Earthquake for Seismic Margin Studies Using Seismic

PRA Results," June 1991.

--- , "Policy Statement on Severe Reactor Accidents,"

Federal Register, Vol. 50, p. 32138, August 8, 1985.

--- , SECY 88-147, "Integration Plans for Closure of

Severe Accident Issues," May 25, 1988.

27 NUREG-1407

APPENDIX A

REVIEW LEVEL EARTHQUAKE

APPENDIX A

REVIEW LEVEL EARTHQUAKE

The seismic margins methodology was designed to demonstrate sufficient margin over the Safe Shutdown Earthquake (SSE) to ensure plant safety and to find any "weak

links" that might limit the plant shutdown capability to

safely withstand a seismic event larger than the SSE or

lead to seismically induced core damage. The methodology involves the screening of components based on their

importance to safety and seismic capacity. The seismic

margins method utilizes two review or screening levels

geared to peak ground accelerations of 0.3g and 0.5g. In

areas of low to moderate seismic hazard, most plants that

have been evaluated using PRAs or margins studies have

been shown to have HCLPFs at or below 0.3g. Past experience indicates that, at the 0.3g screening level, a small

number of "weak links" are likely to be identified, efficiently defining the dominant contributors to seismically

induced core damage. It is the staff's judgment that the

use of a 0.3g review level earthquake for most of the

nuclear power plant sites in the Central and Eastern

United States (east of the Rocky Mountains) would serve

to meet the objectives of the IPEEE.

All sites east of the Rocky Mountains, however, are not

subject to the same level of earthquake hazard. The recent studies by LLNL (NUREG/CR-5250) and EPRI

(EPRI NP-6395-D) show significant differences depending on location and specific site conditions. Because the

two studies do not necessarily agree with each other, it

was deemed necessary to use both studies in determining

which review level earthquake should be assigned to each

site. Hazard comparisons were made using the median,

85th percentile, and mean from the site-specific results

provided by the LLNL and EPRI studies. Based on the

sensitivity tests and engineering and seismological judgment, the staff has defined the review level earthquake

for each site (0.3g, 0.5g, or reduced scope) in Table 3.1. A

second criterion, plant design basis, was used to subdivide

the 0.3g bin. The subdivision, based on a composite conditional probability of exceeding the SSE for each nuclear

power plant, resulted in plants within the 0.3g bin being

assigned a full-scope or focused-scope review.

The sites in the Western United States (west of the Rocky

Mountains) are treated differently. Those sites in coastal

California where the seismic hazard is much higher and

the resulting design bases are greater than 0.5g cannot

make use of the margins methodology. The other plant

sites in the West should use a 0.5g review level earthquake unless it can be demonstrated that the seismic

hazard level at a particular plant site is consistent with the

seismic hazard at the 0.3g bin plant sites east of the Rocky

Mountains. Western sites that show such a consistency in

seismic hazard will conduct the full-scope 0.3g margins

review. The results of the binning for the plants in the

Western United States are presented in Table 3.2.

The rationale for the selection of the review level earthquakes (RLEs) and the grouping of the plants east of the

Rocky Mountains is discussed below.

A.1 Introduction

The specification of a review level earthquake (RLE) for

use in carrying out an individual plant examination for

external events (IPEEE) has been a complex problem

involving the search for consistency. It would be preferable if the selection of the RLEs were completely consistent with the individual plant examination (IPE) for internal events and the inherent strengths of the seismic margins methodologies, but it is very difficult to satisfy all of

these elements in any rigorous quantitative sense. Thus,

for example, attempting to equate the review level earthquake to the reporting criteria in the IPE (mean sequence

frequency leading to core damage of 10-e per year) is

fraught with difficulties because of the large uncertainties

in numerical estimates of seismically induced core damage, the inappropriateness of a comparison between numerical estimates of seismically and internally induced

core damage (the source and treatment of uncertainty can

be quite different), and the inherent difficulties in relating the output of a seismic margins study (HCLPF) to

estimates of core damage frequency. For some of the

same reasons, it was recognized that external initiators,

including earthquakes, need not necessarily be treated in

the same manner as internal initiators in implementing

the Severe Accident Policy. It should be noted that the

RLE defines a reporting level. A HCLPF value lower

than RLE does not necessarily represent a plant vulnerability. However, the licensee should assess the significance of HCLPF values lower than RLE and take any

necessary actions and make other improvements that are

deemed appropriate by the licensee.

A.2 General Evaluation Procedure

A.2.1 Data Evaluated

The staff has recommended three review level earthquakes to be used when applying the seismic margins

methodology to nuclear power plants east of the Rocky

Mountains for the IPEEE. The review levels or "bins" are

0.5g, 0.3g, and a reduced-scope level. The basic information used was the Lawrence Livermore National Laboratory (LLNL) hazard study (NUREG/CR-5250) and the

Electric Power Research Institute (EPRI) hazard study

(EPRI NP-6395-D). These studies represent state-ofthe-art estimates of seismic hazard. Because the two

A-1 NUREG-1407
Appendix A

studies do not necessarily agree with each other, it was

deemed necessary to use them both in determining which

bin a particular site belonged in.

In the LLNL study (NUREG/CR-5250), it was noted

that, for some sites, the mean estimates of seismic hazard

were dominated by the input of one ground motion expert

(No. 5). This, dominance was caused by the low attenuation, high uncertainty, and relatively high motion on rock

found in this expert's input. This input has received a

great deal of attention, and some have argued that it is

inconsistent with the data. The staff requested LLNL (as

a sensitivity study) to calculate the hazard at nuclear

power plant sites east of the Rocky Mountains leaving out

the input of this expert.

Data from the Saguenay Event in Quebec, Canada (November 1988), the largest earthquake in eastern North

America in 50 years, appears to be quite different from

previous data sets and has not helped to resolve the controversy. At this time, in order to avoid relying exclusively

on the LLNL results that include the input of expert No.

5, the staff is treating the LLNL hazard estimates based

on the other four ground motion experts as a separate

study when binning nuclear power plant sites for IPEEE.

A.2.2 Comparison Procedure

Hazard comparisons were made using the mean, median,

and 85th percentile from the site-specific results provided

by the LLNL and EPRI studies. Each of these pieces of

information represents a different way of characterizing

the distribution of seismic hazard estimates at each site as

determined by a particular study.

Mean: The mean is a commonly used statistic that can be

assigned actuarial significance. However, because of the

skewed nature of the distribution, it is also a highly unstable (with respect to methodology and input assumptions)

view of hazard. The mean is highly sensitive to the characterization of the extremes of the distribution.

Mediaw The median is more stable than the mean and

shows the greatest agreement between the LLNL and

EPRI studies. However, it is only the 50th percentile of

the hazard and is insensitive to the extent of uncertainty.

85th Percentile: An alternative candidate to the mean is

the 85th percentile. It reflects uncertainty in that it indicates the breadth of the distribution, but it is less sensitive

to extreme outliers.

A.2.3 Weighting Criteria

In the past, great emphasis has been placed on the likelihood of exceeding peak ground acceleration (PGA). In

this evaluation, site hazard comparisons were made using

response spectra and PGA. The likelihoods of exceeding

spectral response accelerations in the 2.5 to 10 Hz range

were examined because these frequencies are more

closely related to the types of motion that could cause

damage at nuclear power plants. Unit weights (2/7th

each) were assigned to the likelihoods of exceeding spectral response ordinates at 2.5, 5, and 10 Hz. One-half unit

weight (1/7th) was assigned to the likelihood of exceeding

the PGA.

A.2.4 Ranking Criteria

Emphasis was placed on the relative ranking of sites with

respect to other sites using the same seismic hazard study,

statistic, and ground motion measures. Extensive use was

made of a clustering methodology developed by LLNL

for the NRC (Bernreuter et al., 1989a, 1989b). For a given

hazard study, statistic, ground motion measure and reference level, this methodology divides the ensemble of sites

into groups so that the sites in any one group are "close"

to each other with respect to seismic hazard. For example,

the sites may be divided into groups based on mean estimates of exceeding 0.5g PGA from the EPRI study or

median estimates of exceeding the 2.5 Hz spectral response (associated with the NUREG/CR-0098 response

spectrum anchored at 0.3g) from the LLNL five-expert

study. Although there were a fixed number of groups, no

minimum number of sites were in a group, and indeed

some groups contained only one site.

A.2.5 Spectral Shape

The spectral shape associated with the 03g screening

level was assumed to be the median NUREG/CR-0098

spectrum anchored at 03g. There has been some discussion that the screening level should actually be associated

with a somewhat higher ground motion (the Seismic

Qualification Utility Group (SQUG) bounding spectrum)

but in this relative comparison, the use of this alternative

spectrum would make little or no difference.

A.3 Specific Binning Procedure

A.3.1 Initial Binning Evaluation

As the first step, sites that consistently fell into the group

that had the highest likelihood of exceeding the 0.3g

NUREG/CR-0098 5% damped median spectrum were

conditionally assigned to the 0.5g bin. Sites that fell into

the group that had the lowest likelihood of exceeding the

0.3g NUREG/CR-0098 5% damped spectrum were assigned to the reduced-scope bin.

The ground motion measure compared was the weighted

combination of 2.5 Hz, 5 Hz, 10 Hz and PGA. The individual consistency criteria used were:

1. Agreement among the LLNL five-expert, LLNL

four-expert, and EPRI studies, and

NUREG-1407 A-2
Appendix A

2. Agreement between the median and either mean or

85th percentile statistics.

This resulted in a comparison of nine separate hazard

groupings (three pieces of information for each of the

three studies).

For example, if a particular site fell in the top group (0.5g

bin) for all of the criteria except the EPRI median, it

remained in the 0.3g bin. The conclusions must be supported by all the hazard studies. On the other hand, if a

particular site fell in the bottom group for all of the

criteria except for the LLNL four- and LLNL five-expert

mean estimates, it was included in the reduced-scope bin.

Only one measure of uncertainty, mean or 85th percentile, needs to be satisfied.

A.3.2 0.3g Bin Subdivision

The staff investigated the potential for using the seismic

design basis as a parameter for making the initial binning

assignments. There was insufficient technical basis for its

use; thus it was not used for the initial binning. However,

when combined with hazard and engineering judgment,

the use of the seismic design provided a basis for an

overall cost-effective reduction in the scope of the 0.3g

margins review. The staff repeated the process that was

used to obtain the initial binning with the sole change that

instead of factoring in only the seismic hazard, the seismic

hazard and the seismic design basis were used.

Composite conditional probabilities were obtained for

the three seismic hazard curves (EPRI, LLNL with four

experts, and LLNL with five experts) and the three statistical measures of the hazard curves (mean, median, and

85%)-nine separate probabilities for each site. A composite conditional probability was formed by adding the

weighted conditional probabilities of exceeding the uniform hazard spectra at a particular ground motion frequency;, Le., the intersection of the plant-specific seismic

design spectrum for the particular frequencies with the

uniform hazard spectra yields the conditional probability.

The frequencies were those used for the initial binning

(2.5Hz, 5H1z, 10Hz, and PGA). The were also weighted

the same (2/7, 2/7, 2/7, and 1/7).

Using the same agreement criteria as in the initial binning, six sites were identified, i.e., consistently fell into

the top group. These are listed in Table 3.1 to do the

full-scope 0.3g seismic margins review.

As a "sanity check" of this approach, the list derived from

this approach was compared to the list derived from seismic hazard alone. The six full-scope plants were among

the top ten seismic hazard sites.

The staff's resolution of the Eastern U.S. Seismicity Issue

(Charleston Earthquake Issue) has identified eight plants

at five sites as outliers. The staff determined that these

plants should be assigned to the full-scope category. This

action added a single additional plant, Arkansas Nuclear

One, Unit 1, to the list derived on the basis of seismic

hazard and seismic design.

The candidates for the 0.5g and reduced-scope bins were

then subjected to additional evaluation by the staff.

A.3.3 Subsequent Binning Evaluations

The candidates for the 0.5g bin were first examined to

provide some assurance that, although the hazard was

relatively high, it was high enough to warrant inclusion in

this bin.

As a test, it was considered appropriate that a site belonged in the 0.5g bin if a hypothetical nuclear power

plant at that site was assumed to have a HCLPF of 03g

and the mean annual core damage frequency associated

with that hypothetical plant was 1O-6 or higher. The work

cited in NUREG/CR-5501 showed that the mean annual

core damage frequency was roughly an order of magnitude lower than the mean annual likelihood of exceeding

the plant HCLPF and very roughly equal to the median

annual likelihood of exceeding the plant HCLPF.

Based on these estimates, the staff assumed that inclusion

in the 0.5g bin would be supported if:

1. The mean or 85th percentile annual likelihood of

exceeding the 0.3g spectrum from all three studies

was 10-4 or greater, and

2. The median annual likelihood of exceeding the 0.3g

spectrum from all three studies was 10-5 or greater.

This evaluation should be viewed as a "sanity check"; it

should not be viewed as a plant-specific statement on core

damage frequencies. The reasons are:

1. The uncertainty and generic nature associated with

the correlation in NUREG/CR-5501,

2. The use of spectral estimates rather than peak

ground acceleration,

3. The inclusion of the 85th percentile estimates, and

4. All the previously mentioned problems associated

with bottom line numbers.

Finally, the staff examined the candidates for the 0.5g and

reduced-scope bins to assure itself that the classification

made good seismological sense and there was no need to

include additional sites in these bins. In conjunction with

A-3 NUREG-1407
Appendix A

this examination, limited sensitivity tests were als. carried out to determine the impact of slight relaxations in

the consistency criteria.

A.4 Binning of Sites -Results

A.4.l Reduced-Scope Margins Methodology

Bin

The consistency criteria outlined in Section A.3.1 were

slightly modified to identify sites for the reduced-scope

bin. The two bottom median groups were included rather

than the bottom group alone. When this was done, five

sites (South Texas, Comanche Peak, Waterford, River

Bend, and Crystal River) were identified as belonging to

the reduced-scope bin.

Also added to this bin were several sites for which no

EPRI calculations were available but were in the bottom

groups in both the LLNL four- and five-expert studies.

They are Duane Arnold, Big Rock Point, Grand Gulf, St.

Lucie, and Turkey Point. The ten candidate sites in the

reduced-scope bin lie in areas of low seismic hazard along

or near the Gulf and Florida coasts and in the upper

Midwest.

A.4.2 0.5g Bin

As a result of the evaluations cited above, two sites (Pilgrim and Seabrook) were identified as belonging in the

0.5g bin.

A.43 0.3g Bin

All sites not identified as belonging in the 0.5g or

reduced-scope bins were assigned to the 0.3g bin.

A.4A Other Considerations

The grouping was made assuming that each location was

associated with one site condition (rock or varying depths

of soil). Some twelve plant sites east of the Rocky Mountains whose main Category I structures are located on

rock also have some Category I structures or components

located on shallow or intermediate depths of soil. Since

shallow soil, less than about 80 feet thick, can significantly

amplify ground motion, these sites should perform soil

amplification studies to determine the effect.

In particular, for four of the sites included in the 0.3g bin

(on the basis of their primary site conditions), the hazard

for structures or components on the secondary site conditions is equal to or higher than the hazard associated with

those plants in the 0.5g bin. Licensees should, if the

site-specific analysis indicates, use the 0.5g screening tables for elements affected by soil amplification. Similarly,

for one site in the reduced-scope bin, site-specific analysis

should be carried out to determine the effects on those

elements affected by soil amplification.

NUREG-1407 A-4
APPENDIX B

COMPARISON BETWEEN A REDUCED-SCOPE

AND FULI-SCOPE SEISMIC MARGINS EVALUATION

APPENDIX B

COMPARISON BETWEEN A REDUCED-SCOPE

AND FULL-SCOPE SEISMIC MARGINS EVALUATION

There are differences between the reduced-scope and

full-scope margins evaluation both in the extent of the

systems analysis and in the amount of quantification of

HCLPF values for equipment identified in the walkdown.

The comparison is presented in Table B.1. The emphasis

on walkdown and not on quantification also applies to the

performance of containment and containment systems

(that is containment performance analysis should concentrate on identifying seismically induced vulnerabilities

and sequences different from those obtained from the

HE), USI A-45 (Decay Heat Removal Requirements),

and GI-131 (In-Core Flux Mapping System).

B.1 Elements Preserved

The following elements of the seismic margins methodology must be preserved; that is, they must be identical in

the reduced-scope and full-scope evaluation:

1. For either the NRC or EPRI methodology, the systems engineers must perform significant prewalkdown work that should be preserved in a

reduced-scope evaluation. In the NRC methodology, this involves defining initiating events, defining

event trees and the safety functions involved, and

identifying systems and components necessary to

carry out these functions. In the EPRI methodology,

this involves defining success paths (primary and

alternative) and the systems and components involved in these paths. For both methodologies, the

thrust of this work is to narrow the scope and focus

the effort of the key element of the review, the

walkdown.

2. For either the NRC or the EPRI methodology, the

seismic capability evaluation engineers must perform significant pre-walkdown work that should be

preserved in the reduced-scope evaluation. In each

methodology, this involves developing an understanding of the seismic input to the plant and the

seismic design basis and realistic ground and floor

response spectra. It also involves pre-walkdown

screening of the key systems and components identified by the systems engineers so as to make the

walkdown itself most efficient. The thrust of the

screening is to identify items thought to have very

high HCLPF values, items suspected of having low

HCLPF values, and therefore lists of items to be

examined at various levels of detail during the walkdown.

3. The reduced-scope evaluation should be identical in

quality and effort to that as for the full-scope margins methodology. One crucial feature is that it

should involve interactions among seismic capability

evaluation engineers, systems engineers, and the

licensee's plant operations personnel. The walkdown team should visually inspect pertinent structures, equipment, and anchorages consistent with

the full-scope NRC or EPRI methodology. If potentially vulnerable components are found during the

walkdown, a capacity check may be necessary using

the applicable SSE ground response spectra. These

results should be documented. Data sheets similar

to those found in Appendix I of EPRI NP-6041

should be used to document the walkdown. A review

of construction drawings for structural details that

can not be seen in the field should be performed.

4. While the post-walkdown assessment effort for a

reduced-scope evaluation should be identical in

quality to that in the full-scope margins methodology its thrust and level of effort are different because sequence-level (NRC) or success path-level

(EPRI) HCLPFs will not be computed. Instead, its

emphasis should be on identifying possible weaklink items that may need strengthening.

B.2 Reductions

The following, although needed in the full-scope margins

methodology, are not needed in a reduced-scope margins

evaluation

B.2.1 NRC Seismic Margins Methodology

1. The systems engineers need not prepare or quantify

fault trees and Boolean expressions representing

accident sequences. Also, since fault trees will not

be developed, these engineers need not combine

nonseismic failure basic events with seismically initiated failures in any rigorous fashion, although the

existence of those non-seismic failures, if identified,

should be noted and their importance assessed in

the course of the margin evaluation.

2. The seismic capability evaluation engineers need

not develop HCLPF capacity values for all of the key

equipment items that would be represented on the

sequence level Booleans (which will not be developed). It follows that they can not develop a plantlevel HCLPF capacity value.

B-1 NUREG-1407
Appendix B

B.2.2 EPRI Seismic Margins Methodology

The seismic capability evaluation engineers need not develop HCLPF capacity values for all of the key equipment

items found on the success paths (primary and alternative) being studied. It follows that they can not develop

any success-path-level HCLPF capacity values that would

be taken as representing the plant-level HCLPF capacity.

Table BA Reduced-Scope Margins Method

Based on NRC Seismic Margins Methodology (NUREG/CR-4482, Chapter 4)

Step No.

1

2

3

4

5

6

7

8

Description

Selection of Earthquake Review Level

Initial Systems Review

Initial Component HCLPF Categorization

First Plant Walkdown

Systems Modeling Finalize Event Trees:

Fault Tree Development:

Second Plant Walkdown

Systems Model Development

Margin Evaluation of Components, Plant

In Reduced Program.?

Not applicable, NRC designates sites that qualify

Yes, in entirety

Yes, in entirety

Yes, in entirety

Yes:

No

Only as needed

No

No

Based on EPRI Seismic Margins Methodology (EPRI NP-6041, Chapter 2)

Step No. Description

1

2

3

4

5

6

7

Selection of Seismic Margins Earthquake

Selection of Assessment Team

Pre-Walkdown Preparation Work

Systems and Element Selection Walkdown

Seismic Capacity Walkdown

Subsequent Walkdowns

Seismic Margin Assessment Work

In Reduced Program?

Not applicable, NRC designates sites that qualify

Yes, in entirety

Yes, in entirety

Yes, in entirety

Yes, in entirety

Only as needed

No

NUREG-1407 B-2
APPENDIX C

DETAILED DOCUMENTATION AND REPORTING GUIDELINES

APPENDIX C

DETAILED DOCUMENTATION AND REPORTING GUIDELINES

This appendix provides the guidelines for detailed documentation and reporting format and content for the

IPEEE submittals. The major parts of this appendix are

the guidelines for seismic analysis (Section C.2), internal

fire analysis (Section C3), other analyses (Section C.4),

specific safety features and plant improvements (Section

C.1.4), and the licensee review team (Section C.1.5). The

licensees are requested to submit their IPEEE reports

using the standard table of contents given in Table C.1 or

provide a cross reference. This will facilitate review by the

NRC and promote consistency among various submittals.

The contents of the elements of this table are discussed in

sections below.

The level of detail needed in the documentation should

be sufficient to enable NRC to understand and determine

the validity of all input data and calculation models used,

to assess the sensitivity of the results to all key aspects of

the analysis, and to audit any calculation. It is not necessary to submit all the documentation needed for such an

NRC review. Relevant documentation should be cited in

the [PEEE submittal, and be available in easily retrievable form. The guideline for judging the adequacy of

retained documentation is that independent expert analysts should be able to reproduce any portion of the

results of the calculations in a straight forward, unambiguous manner. To the extent possible, the retained documentation should be organized along the lines identified

in the areas of review. Any information that is comparable

to that provided under the IPE for internal events can be

incorporated by reference.

C.1 General

C.1.1 Conformance with Generic Letter and

Supporting Material

Certification that an IPEEE has been completed and

documented as requested by Generic Letter 88-20, Supplement 4. The certification should also identify the

measures taken to ensure" the technical adequacy of the

IPEEE and the validation of results, including any uncertainty, sensitivity, and importance analyses.

C.1.2 General Methodology

Provide an overview description of the methodology employed in the IPEEE for each external event.

C.13 Information Assembly

Reporting guidelines include:

1. Plant layout and containment building information

not contained in the Final Safety Analysis Report

(FSAR).

2. A concise description of plant documentation used

in the IPEEE, (e.g., the FSAR; system descriptions,

procedures, and licensee event reports); and a concise discussion of the process used to confirm that

the IPEEE represents the as-built, as-operated

plant. The intent of such a confirmation is not to

propose new design reverification efforts on the part

of the licensees but to account for the impact of

previous plant modifications or modifications conducted within the IPEEE framework.

3. A description of the coordination activities of the

IPEEE teams among the external events (e.g., for

seismically induced fires).

C.1.4 Submittal of Specific Safety Features

and Potential Plant Improvements

The licensee should provide a discussion of the criteria

used to define vulnerabilities for each external event

evaluated. The licensee should list any potential improvements (including equipment changes as well as changes in

maintenance, operating and emergency procedures, surveillance, staffing, and training programs) that have been

selected for implementation based on the IPEEE (a

schedule for implementation should be provided) or that

have already been implemented. A discussion of anticipated benefits in terms of averted potential risk or increased plant seismic capacity as well as drawbacks to any

improvements should be provided. Those improvements

that have been taken credit for in the analysis and have

not yet been implemented at the plant, should be specifically highlighted in the submittal.

C.1.5 IPEEE Team and Peer Review

The basis for requiring the involvement of the licensee's

staff in the IPEEE review is the belief that the maximum

benefit from the performance of an IPEEE would be

realized if the licensee's staff were involved in all aspects

of the examination and that involvement would facilitate

integration of the knowledge gained from the examination into operating procedures and training programs.

Thus the submittal should describe licensee staff participation and the extent to which the licensee was involved

in all aspects of the program.

The submittal should also contain a description of the

peer review performed, the same type of review as requested for the internal event IPE, the results of the

C-1 NUREG-1407
Appendix C

review team's evaluation, and a list of the review team

members.

C.2 Seismic Events

Section C.2.1 describes submittal information guidelines

for licensees who choose the seismic PRA for seismic

IPEEE, whilq section C.2.2 describes information guidelines for licensees who choose the seismic margin method

for the seismic IPEEE. The submittal should be presented in conformance with Table C.1.

C.2.1 Seismic PRA Methodology

The following information on the seismic IPEEE should

be documented and submitted to the NRC:

1. A description of the methodology and key assumptions used in performing the seismic IPEEE.

2. The hazard curve(s) (or table of hazard values) used

and the associated spectral shape used in the analysis. Also, if an upper bound cutoff to ground motion

of less than 1.5g peak ground acceleration is assumed, the results of sensitivity studies to determine

whether the cutoff affected the overall results and

the delineation and ranking of seismic sequences.

3. A summary of the walkdown findings and a concise

description of the walkdown team and the procedures used.

4. All functional/systemic seismic event trees as well as

data (including origin and method of analysis). Address to what extent the recommended enhancements have been incorporated in the IPEEE. A description of how nonseismic failures, human actions,

dependencies, relay chatter, soil liquefaction, and

seismically induced floods/fires are accounted for.

Also, a list of important nonseismic failures with a

rationale for the assumed failure rate given a seismic

event.

5. A description of dominant functional/systemic sequences leading to core damage along with their

frequencies and percentage contribution to overall

seismic core damage frequencies (for both LLNL

and EPRI hazard curves if used). Sequence selection

criteria are provided in GL 88-20 and NUREG1335. If either hazard curve causes a sequence to

meet these criteria, that sequence should be included. The description of the sequences should

include a discussion of specific assumptions and human recovery actions.

6. The estimated core damage frequency (for both the

LLNL and EPRI hazard curves, if used) and plant

damage state frequencies, the timing of the core

damage, including a qualitative discussion of uncertainties and how they might affect the final results,

and contributions of different ground motions to

core damage frequencies.

7. Any seismically induced containment failures and

other containment performance insights. Particularly, vulnerabilities found in the systems/functions

which will lead to early containment failure that

might result in high consequences. This includes:

isolation, bypass, integrity, and systems (e.g., igniters) required to prevent early failure. The computed fragilities of containment components,

systems, and functions as applicable should be provided. The licensee may submit computed HCLPFs

associated with containment performance (Optional).

8. Atable of fragilities, both generic and plant-specific,

used for screening as well as in the quantification.

The estimated fragilities for the plant, dominant

sequences, and dominant components should be reported. (OptionaL: The estimated HCLPF for the

plant, dominant sequences, and components with

and without nonseismic failures and human actions

may be submitted by the licensee.)

9. Documentation with regard to other seismic issues

(Section 6) addressed by the submittal, the basis and

assumptions used to address these issues, and a discussion of the findings and conclusions. Evaluation

results and potential improvements associated with

the decay heat removal function and movable incore flux mapping system (for Westinghouse plants)

should be specifically highlighted.

10. A Discussion of nonseismic failures and human actions that are significant contributors, or have impacts on results.

11. When an existing PRA is used to address the seismic

IPEEE, the licensee should describe sensitivity studies related to the use of the initial hazard curves,

supplemental plant walkdown results and subsequent evaluations, and relay-chatter evaluations.

The licensee should examine the above list to fill in

those items missed in the existing seismic PRA (See

Section 3.1.2).

C.2.2 Seismic Margins Methodology

The following information on the seismic IPEEE should

be documented and submitted to the NRC for a fullscope or a focused-scope SMM review:.

1. A description of the methodology and a list of important assumptions, including their basis, used in

performing the seismic IPEEE. Address the extent

to which the following were taken into account:

NUREG-1407 C-2
Appendix C

nonseismic failures, human actions, dependencies,

relay chatter, soil liquefaction, and seismically induced floods/fires. Also, a list of important nonseismic failures with a rationale for the assumed failure

rate given a seismic event.

2. A summary of the walkdown results and a concise

description of the walkdown team and procedures

used.

3. All functional/systemic seismic event trees data (including origin and method of analysis) when NRC

SMM is used.

4. A description of the most important sequences and

important minimal cutsets (for both seismic and

nonseismic failures) leading to core damage (NRC

method) or a description of the success paths and

procedures used for their selection and of each component in the controlling success path (EPRI

method).

5. Any seismically induced containment failures and

other containment performance insights. Particularly, vulnerabilities found in the systems/functions

which will lead to early containment failure and high

consequences. This includes: isolation, bypass, containment integrity and systems (e.g., igniters) required to prevent early failure. Also, computed fragilities (if used) and HCLPFs of containment

components, systems, and functions as applicable.

6. A table of fragilities (if used) and HCLPFs, both

generic and plant-specific, used for screening as well

as in the quantification. The estimated fragilities (if

used) and HCLPFs for the plant, dominant sequences, and dominant components should be reported.

7. Documentation with regard to other seismic issues

(Section 6) addressed by the submittal, the basis and

assumptions used to address these issues, and a discussion of the findings and conclusions. Evaluation

results and potential improvements associated with

the decay heat removal function and movable incore flux mapping system (for Westinghouse plants)

should be specifically highlighted.

8. For the NRC method, provide a discussion of nonseismic failures and human actions that are significant contributors to or have an impact on the results.

The following is the information that should be documented and submitted to the NRC for a reduced-scope

SMM review:

1. A description of the procedures used to identify

systems and components for the walkdown in performing the seismic IPEEE.

2. A summary of the walkdown findings and a concise

description of the walkdown team and procedures

used.

3. A discussion and the results of any specific component capacity evaluations performed, the methods

used, and assumptions.

4. Documentation with regard to other seismic issues

(Section 6) addressed by the submittal, the basis and

assumptions used to address these issues, and a discussion of the findings and conclusions. Evaluation

results and potential improvements associated with

the decay heat removal function and movable incore flux mapping system (for Westinghouse plants)

should be specifically highlighted.

C.3 Internal Fires

The information on the internal fires IPEEE identified

below should be documented and submitted to the NRC.

1. A description of the methodology and key assumptions used in performing the fire IPEEE and a

discussion of the status of Appendix R modifications.

2. A summary of the walkdown findings and a concise

description of the walkdown team and the procedures used. This should include a description of the

efforts to ensure that cable routing used in the analysis represents as-built information and the treatment of any existing dependence between remote

shutdown and control room circuitry.

3. A discussion of the criteria used to identify critical

fire areas and a list of critical areas, including (a) single area in which equipment failures represent a

serious erosion of safety margin, and (b) same as (a),

but for double or multiple areas sharing common

barriers, penetration seals, HVAC ducting, etc.

4. A discussion of the criteria used for fire size and

duration and the treatment of cross-zone fire spread

and associated major assumptions.

5. A discussion of the fire initiation data base, including the plant-specific data base used. Provide documentation in each case where the plant-specific data

used is less conservative than the data base used in

the approved fire vulnerability methodologies. Describe data handling method, including major assumptions, the role of expert judgment, and the

identification and evaluation of sources of data uncertainties.

C-3, NUREG-1407
Appendix C

6. A discussion of the treatment of fire growth and

spread, the spread of hot gases and smoke, and the

analysis of detection and suppression and their associated assumptions, including the treatment of

suppression-induced damage to equipment.

7. A discussion of fire damage modeling, including the

definition of fire-induced failures related to fire barriers and control systems and fire induced damage to

cabinets. A discussion of how human intervention is

treated and how fire-induced and non-fire-induced

failures are combined. Identify recovery actions and

types of fire mitigating actions taken credit for in

these sequences.

8. Discuss the treatment of detection and suppression,

including fire fighting procedures, fire brigade training and adequacy of existing fire brigade equipment,

and treatment of access routes versus existing

barriers.

9. All functional/systemic event trees associated with

fire-initiated sequences.

10. A description of dominant functional/systemic sequences leading to core damage along with their

frequencies and percentage contribution to overall

core damage frequencies due to fire. Sequence selection criteria are provided in GL 88-20 and

NUREG-1335. The description of the sequences

should include a discussion of specific assumptions

and human recovery action.

11. The estimated core damage frequency, the timing of

the associated core damage, a list of analytical assumptions including their bases, and the sources of

uncertainties.

12. Any fire induced containment failures identified as

being different from those identified in the internal

events analysis.

13. Documentation with regard to the decay heat removal

function and Fire Risk Scoping Study issues addressed by the submittal, the basis and assumptions

used to address these issues, and a discussion of the

findings and conclusions. Evaluation results and potential improvements should be specifically highlighted.

14. When an existing PRA is used to address the fire

IPEEE, the licensee should describe sensitivity studies related to the use of the initial hazard, supplemental plant walkdown results and subsequent

evaluations. The licensee should examine the above

list to fill in those items missed in the existing fire

PRA.

C.4 High Winds, Floods, and Others

The following information on the high winds, floods, and

others portion of the IPEEE should be documented and

submitted to the NRC:

1. A description of the methodologies used in the

examination.

2. Information on plant-specific hazard data and licensing bases.

3. Identified significant changes not reported per

10 CFR 50.71(e) (See Section 5.2.2), if any, since OL

issuance with respect to high winds, floods, and

other external events.

4. Results of plant/facility design review to determine

their robustness in relation to NRC's 1975 SRP

criteria.

5. Results of the assessment of the hazard frequency

and the associated conditional core damage frequency if step 4 of Figure 5.1 is used.

6. Results of the bounding analysis if step 5 of Figure

5.1 is used.

7. All functional event trees, including origin and

method of analysis (PRA only).

8. A description of each functional sequence selected,

including discussion of specific assumptions and human recovery action (PRA only).

9. The estimated core damage frequency, the timing of

the associated core damage, a list of analytical assumptions including their bases, and the sources of

uncertainties, if applicable (PRA only).

10. A certification that no other plant-unique external

event is known that poses any significant threat of

severe accident within the context of the screening

approach for "High Winds, Floods, and Others."

NUREG-1407 C--4
Appendix C

Table C.1 Standard Table of Contents for IPEEE Submittal

1. Executive Summary

1.1 Background and Objectives

1.2 Plant Familiarization

1.3 Overall Methodology

1A Summary of Major Findings

2. Examination Description

2.1 Introduction

2.2 Conformance with Generic Letter and Supporting Material

2.3 General Methodology

2.4 Information Assembly

3. Seismic Analysis

3.0 Methodology Selection

3.1a Seismic PRA

3.1.1 Hazard Analysis

3.1.2 Review of Plant Information and Walkdown

3.1.3 Analysis of Plant System and Structure Response

3.1.4 Evaluation of Component Fragilities and Failure Modes

3.1.5 Analysis of Plant Systems and Sequences

3.1.6 Analysis of Containment Performance

3.1b Seismic Margins Method (SMM) (NRC, EPRI, or Reduced SMM)

3.1.1 Review of Plant Information, Screening, and Walkdown

3.1.2 System Analysis

3.1.3 Analysis of Structure Response

3.1.4 Evaluation, of Seismic Capacities of Components and Plant

3.1.5 Analysis of Containment Performance

3.2 USI A-45, GI-131, and Other Seismic Safety Issues

4. Internal Fires Analysis

4.0 Methodology Selection

4.1 Fire Hazard Analysis

4.2 Review of Plant Information and Walkdown

4.3 Fire Growth and Propagation

4.4 Evaluation of Component Fragilities and Failure modes

4.5 Fire Detection and Suppression

4.6 Analysis of Plant Systems, Sequences, and Plant Response

4.7 Analysis of Containment Performance (If Applicable)

4.8 Treatment of Fire Risk Scoping Study Issues

4.9 USI A-45 and other Safety Issues

5. High Winds, Floods, and Others

5.1 High Winds

5.2 Floods

5.3 Transportation and Nearby Facility Accidents

5.4 Others

6. Licensee Participation and Internal Review Team

6.1 IPEEE Program Organization

6.2 Composition of Independent Review Team

6.3 Areas of Review and Major Comments

6.4 Resolution of Comments

7. Plant Improvements and Unique Safety Features

8. Summary and Conclusions (including proposed resolution of USIs and GIs)

C-5 NUREG-1407

APPENDIX D

NRC RESPONSE TO COMMENTS AND QUESTIONS

APPENDIX D

NRC RESPONSE TO COMMENTS AND QUESTIONS

D.1 Introduction and Summary

The NRC staff conducted an IPEEE Workshop on September 11-13, 1990 at the Pittsburgh Hilton in Pittsburgh, Pennsylvania. The objectives of the workshop

were to discuss the IPEEE process and to solicit questions

and comments on the guidance for performing the

IPEEE and for reporting the results of the review. The

schedule of the IPEEE Workshop was announced in the

Federal Register (55 FR 30332) July 25, 1990, and a

preliminary agenda of the workshop was published on

August 10, 1990, in the Federal Register (55 FR 32712).

All the questions and comments raised at the workshop or

submitted after the workshop were categorized into several major subject areas. This appendix summarizes these

questions and comments and the NRC staff responses

(SR) to them. The most significant comments, concerns,

and questions, together with staff response (SR), are summarized below. This summary also serves to highlight the

major changes made in going from the proposed to the

final documents.

1) Backfit Analysis: Should a regulatory backfit analysis of

the proposed 1PEEE effort be performed prior to issuance of the IPEEE generic letter?

SR: The staff does not believe, as a legal matter, that a 10

CFR 50.109 type backfit analysis is needed for the

IPEEE generic letter (memorandum from W. Parler

to Commissioners, dated July 27, 1990). Because the

request to perform the IPEEE is considered to be a

request under 10 CFR 50.54(f), the staff has performed a 50.54(f) analysis, which is included as Appendix 5 to Supplement 4 of Generic Letter 88-20.

However, in view of the significant licensee resource

commitment required to respond to this information

request and in the interest of a prudent policy, the

staff has completed a value-impact analysis and has

included it as an attachment to this appendix. This

analysis shows that, based on previous experience

with the evaluation of severe accidents initiated by

external events, the IPEEE has the potential to

identify items that, if corrected, would result in

substantial increases in safety, and that the cost of

corrections, including the cost of the IPEEE, would

be commensurate with averted potential risk to severe accidents.

2) Cost Estimates and Resource Requirements: The overall cost of the IPEEE was underestimated.

SR: The staff derived its estimates of cost and resource

requirements to perform an IPEEE from the actual

costs spent on two NUREG-1150 plants and the

cost spent on the Hatch seismic review extrapolated

to the IPEEE scope. At the IPEEE workshop, certain industry estimates were presented that were

either less than or comparable to the staffs estimates. The staff recognizes that there are uncertainties in the costs because there are uncertainties associated with tfie analysis of external events. However,

the staff believes that there are ways to keep the cost

under control. If additional questions arise, regarding the IPEEE process and the associated

quidelines, the staff will meet with licensees and

respond to those questions.

3) Schedule and Resource Availability Schedule and resource availability for performing the JPEEE are of

concern.

SR: The methods identified in this report are not new

and have been used and discussed extensively in the

past. Probabilistic risk assessment (PRA) procedures for assessing the risk associated with external

events have been used since the late 1970's. The

NRC seismic margins method was published in

1985; the Electric Power Research Institute (EPRI)

seismic margins method was published in 1988.

These methods were derived from the insights

gained from available seismic PRAs. They were

widely discussed at many conferences and workshops and were used at three plant sites. Procedures

for the seismic walkdown, one of the most important

ingredients in the seismic IPEEE, are similar to

those that will be used in the implementation of

Unresolved Safety Issue (USI) A-46, "Verification

of Seismic Adequacy of Equipment in Operating

Plants." A-number of trial walkdown training workshops with a number of participants from the utilities were conducted in the past by the Seismic Qualification Utility Group (SQUG), which developed

the walkdown procedures for USI A-46. Finally, the

event trees and fault trees developed for the internal event IPE, which was initiated about 18 months

before the IPEEE, will be available for use in the

IPEEE. Therefore, the staff believes that a large

pool of talent is available (as evidenced by the number [approximately 25] of consultants and consulting

firms represented at the IPEEE workshop), and that

within the 3-year period to perform the IPEEE,

licensees can develop or obtain the necessary expertise to conduct the IPEEE. However, as with the

D-1 NUREG-1407
Appendix D

internal events IPE, the staff will consider extending

this date on a case-by-case basis.

4) Licensee Response Time: The initial response time of6O

days to identify the methodologies for completing

LPEEE is too short.

SR: The staff believes that it is appropriate to extend the

licensee's initial response time from 60 days to 180

days to allow for some essential preparatory work

(i.e., the processing of bids, completion of the development of the alternative fire evaluation methodology by the Nuclear Management and Resource

Council (NUMARC) and EPRI, and the staff's review and assessment of this methodology). One hundred eighty days was selected in consideration of the

current schedule for NUMARC/EPRI to complete

the development and verification of their alternate

fire evaluation methodology, and for the subsequent

staff review. These activities are expected to be completed in July 1991 (see Response 6.1).

5) Inclusion of Issues: The inclusion of (a) lightning, (b)

volcanic activities, and (c) Generic Issue (GI) 103,

'Design for Probable Maximum Precipitation (PMP)"

in the IPEEE was questioned.

SR: Licensees need to confirm that lightning or volcanic

activity is not a dominant contributor to severeaccident risk at their nuclear power plant sites. The

determination should be based on plant-specific experience. The concern related to lightning (as

pointed out by the ACRS) is that lightning strikes, in

addition to causing loss-of-offsite power, may damage instrumentation and control systems. If this had

happened before at a site, the staff would expect the

IPEEE for that specific plant to address this concern. In regard to volcanic activity, only two sites

would be affected. In either case, a simple discussion

will be sufficient for those plants not affected by

these events. For plants that may be affected, a

success screening process, such as the one described

in this report can be used.

With regard to GI-103, the NRC acknowledged the

importance of this new PMP criterion in Generic

Letter 89-22 by requiring that future plants be designed against it (i.e., design basis). For existing

plants, the NRC recommended that licensees review the material contained in GL 89-22 to determine whether they believe additional action is necessary; however, licensees need not change their

design bases. For the IPEEE, the staff believes that

assessing the potential for a PMP to cause a severe

accident is justified, since the National Weather

Service PMP data are being applied to future designs. However, the staff has clarified it's recommendation by limiting the assessment of the effects

to onsite flooding and roof ponding.

6) Fire Evaluation

6.1) Expeditious NRC review of the NUMARCIEPRI alternate fire-evaluation methodology was requested.

SR: The staff had previously committed to review

an alternate methodology being developed by

NUMARC and EPRI for evaluating fires. Currently, the staff is reviewing a NUMARC document

describing the methodology and is waiting for more

information from EPRI and NUMARC on the results of demonstration applications of this methodology at two nuclear plants (the staff anticipates this

will come in April 1991). Following receipt of this

information, the staff plans to complete its review,

including discussions with the ACRS and the

NUMARC. As stated in Response 4, these activities

are not expected to be completed until July 1991.

Therefore, the staff plans to respond separately to

the NUMARCIEPRI alternate methodology, so as

not to delay issuing the IPEEE generic letter and

guidance document. However, as discussed in Response 4, additional time was given to licensees, so

that they have the results of the staff review before

they commit to a fire evaluation methodology in

their IPEEE submittal plans.

6.2) GI-57, "Effects of Fire Protection System Actuation on

Safety-Related Equipment:" The relationship of this GI

to the IPEEE was questioned.

SR: The effect of fire suppressants on safety equipment

is one of the safety issues identified in the Fire Risk

Scoping Study, (NUREG/CR-5088) and may be a

significant contributor to risk. Accordingly, it was

raised as a generic safety issue and was also included

in the IPEEE. The staff expects that if a licensee

discovers a significant vulnerability in this area

through the IPEEE, the licensee would address the

problem and not await the GI-57 resolution. During

the walkdown, licensees can collect relevant information on whether actuated fire protection systems

would spray safety-related equipment, and can institute some protective measures to prevent the safety

equipment from being sprayed by fire suppressants.

The additional effort to collect this information during the walkdown should not be a burden. However,

the formal resolution of GI-57 does not have to be a

part of the IPEEE.

7) Seismic Events: Treatment of seismic events needs clanificatiort

71) The need to use both the LLNL and EPRI seismic

hazard curves was questioned.

NUREG-1407 D-2
Appendix D

SR: The staff considered the difference between these

two curves in specifying the enhancements for the

seismic iiiargins methods and the seismic PRA.

However, based on the available information to

date, the staff is unable to dispute the merit of either

curve and considers both of them to be valid. The

staff also believes that the added cost of using two

curves should not be burdensome, based on what

was spent on two NUREG-1150 plants. A utility

may choose to use only a single curve, provided the

higher one of the two curves is chosen.

7.2) The use of a site-specific seismic curve (in lieu of the

LLNL and EPRI curves) was suggested.

SR: The LLNL and EPRI curves are "site-specific" seismic curves. Each used its standard methodology and

uniform interpretation of data bases to calculate the

seismic hazards for power plant sites in the Eastern

United States. The use of other site-specific seismic

hazard curves is an acceptable option subject to review and acceptance by the staff. However, the staff

believes that the cost associated with the development of new site-specific seismic hazard curves

could be very high and time consuming.

7.3) The use of plant design bases in the seismic binning

process was suggested.

SR: The staff investigated the potential of using the seismic design basis as a parameter for making the initial

binning assignments. Because there was insufficient

technical basis for its use, it was not used initially.

However, when considered in conjunction with the

seismic hazard, the use of the seismic design bases

provided a means for reducing the scope of the 0.3g

margins review. Specifically, plant sites in the 0.3g

bin were assigned to a full-scope or a newly defined

focused-scope category. The full-scope category is

essentially the review specified in the draft generic

letter and guidance document that were distributed

for public comment, whereas the focused-scope review represents a reduced scope review. The primary purpose of this further subdivision is to reduce

the level of review effort, mainly in the relay chatter

area, for plants with a lower hazard or higher design

basis. Plants with a relatively higher hazard and

lower seismic design basis should perform a more

detailed study than the other plants. (Grouping of

plants into various categories is discussed in Section

3 and Appendix A of this report.) Of the 56 sites

originally assigned to the 0.3g category, 7 remain in

the full-scope category, the remainder moved to the

focused-scope category.

7.4) The scope of the relay chatter evaluation was questioned,

SR: Detailed relay chatter studies conducted at the

Hatch and Diablo Canyon plants showed that considerable resources were expended to perform the

relay chatter review;, and using existing procedures,

operators could solve the relay chatter problems

identified at these plants. However, the staff and

industry consultants are concerned that such conclusions cannot be considered generic without some

additional plant reviews.

Therefore, as discussed above, the staff is recommending that the 0.3 g bin be subdivided into two

categories, a full-scope and a focused-scope category. For plants in the full-scope category, licensees

will have to evaluate the relay chatter consistent

with the approach discussed in EPRI NP-6041 or its

equivalent. For reduced-scope review, the implementation of USI A-46 program will provide information for satisfying the IPEEE provisions. Note

that licensees can perform the IPEEE with considerable efficiency, if they take advantage of the lessons

learned from the Hatch and Diablo Canyon relay

chatter evaluations. For plants in the focused-scope

category, a lower level of effort is recommended;

this would entail looking for and addressing lowcapacity relays only.

D.2 Detailed Response to Public

Comments and Questions

This section paraphrases, summarizes, and categorizes

into subject areas, questions and comments either raised

at the workshop or received by the staff (16 parties submitted written comments, see sources of comments). The

NRC staff response is also provided. Table D.1 contains a

listing of the subject areas discussed in this section. The

workshop transcript and a copy of the comments received

are available in the NRC Public Document Room.

Table D-1 Categorization of Question and Comments

1. IPEEE, IPE, CPI, and Accident Management

2. Backfit analysis

3. Cost estimates and resource requirements

4. Schedule and response time

5. High winds, floods, and transportation and

nearby facility accidents

6. Internal fires

7. Seismic events

D--3 NUREG-1407
Appendix D

1. IPEEE, IPE, CPI, and Accident Management

1.1 How do the IPE, IPEEE, and the accident management all relate to the design bases of the plant in

terms of identified plant vulnerabilities, improvements, and potential increase in risk? How do the

plant operators make the day-to-day decisions when

PRA insights and Tech Specs are in conflict? (Ref.

D.16, p. 40)

SR: The thrust of the whole severe accident program is

to reduce the likelihood of severe accidents and

their consequences. As such, they are looking at

accident scenarios beyond the traditional design basis envelope. If a vulnerability is identified and a fix is

proposed, it is important to make sure that the proposed fix has no adverse effect on the plant. If a

licensee makes modifications to the plant, which

resulted in a change of the plant design basis, then

that must be documented, tracked, and accounted

for, in accordance with the provisions of 10 CFR

50.59. If the PRA identifies a" conflict with the Technical Specifications or operating procedures, the licensee should examine the reason for the conflict. It

is important that the licensee not make plant or

procedural modifications without understanding the

basis behind the PRA conclusions. For example, if

the underlying model in the PRA was developed

with simplified assumptions and modeling techniques, it might be prudent to perform a more realistic evaluation to assure that the modeling assumptions have not biased the results in an inappropriate

manner.

1.2 After the utilities had factored vulnerabilities identified through IPE and IPEEE in their emergency

operating procedures (EOPs) and the Tech Specs

that support the EOPs, what else would the utilities

be required to do for the accident management?

Any specific example of a guideline that the staff

would be putting forth as part of accident management? (Ref. D.16, p. 46)

SR: It is important to recognize that accident management responses are not just limited to emergency

procedure guidelines (EPGs) or EOPs. Technical

support and the kind of guidance and personnel

training that are needed should be part of accident

management. IPE and IPEEE results should be considered as an information source which provides inputs to training programs and to the development of

emergency preparedness exercises. Accident management takes that information and uses it in the

planning, training, exercises, and to establish the

communication and the feedback mechanism at the

utilities. Specific guidance and examples are still

being developed by industry (NUMARC) and NRC,

and will be addressed in a future generic letter on

accident management.

1.3 The Severe Accident Policy Statement is silent on

external events, thus, there is no need to do much

beyond what is already done for internal event IPEs.

(Ref. D.16, p. 53)

SR: The Commission Policy Statement identified the

need to seek vulnerabilities systematically at all operating plants. It didn't distinguish between internal

and external events. However, PRA studies have

shown that external events, in particular seismic and

fire, are principal contributors to overall risk. Accordingly, the staff recommended to the Commission in SECY-86-162, dated May 22, 1986, that

external events be included in implementation of

the Severe Accident Policy Statement.

1.4 IPEEE, basically, is an evaluation looking at a point

in time, a snapshot in time, is there an intent of

keeping it living? (Ref. D.16, p. 78)

SR: The staff is treating the IPEEE as a one-time evaluation. There is no requirement to keep it living.

However, based on utilities' experience, once one

has gone through the process and invested the resources and constructed the PRA or equivalent, it

would be useful to keep it up to date. However, it is

really up to the utility whether or not to keep it

living.

1.5 Is it correct to assume that there is no requirement

for the pedigree of the program, that it can be basically be a study without a QA type of pedigree? (Ref.

D.16, p. 80)

SR: There is no requirement for an Appendix B-type

QA program to check the IPE or IPEEE. The licensees should perform an ongoing internal quality assurance effort to ensure that the results of the IPE

and IPEEE are factual and represent the as-built,

as-operated plant. Typically, licensees will define "pinch points" in their performance of a risk assessment to stop and assess the progress and quality of

their effort to date. As in the internal event IPE, the

staff is asking for a peer review as part of the IPEEE.

The peer review provides a type of QA function.

1.6 In the area of other external events that are not

included in the IPEEE, does the utility need to develop a hazard curve associated with that particular

event? (Ref. D.16, p. 81)

SR: No. The staff is not asking the utilities to justify not

including those events. However, if a utility knows of

a particular hazard that is greater than what the staff

has considered in the generic study, the utility

should consider including it in the IPEEE.

NUREG-1407 D--4
Appendix D

1.7 Is sabotage included in the IPEEE? (Ref. D.16, p.

82)

SR: Sabotage is not included.

1.8 Florida Power & Light has a policy that upon approaching hurricanes in S. Florida, the unit will be

shut down and the unit will go to a Mode 3 or a Mode

4 in advance of the hurricane, does that mean the

hurricane need not be considered in the IPEEE?

(Ref. D.16, p. 82)

SR: In general, the shutdown mode is outside of the

IPEEE scope. However, in a case like this, the licensee should make sure that the plant can be shutdown

and maintained in a safe shutdown condition (USI

A-45 requirements). In other words, the combined

frequency of the hurricane and failure to shutdown

and to maintain the plant in a safe shutdown condition needs to be assessed.

1.9 Faced with large uncertainties, how are risk, human

reliability, operational reliability, maintenance reliability, etc. associated with external events to be

quantified? (Ref. D.16, p. 84)

SR: The staff recognizes that there are significant uncertainties in quantifying risks associated with both internal and external events. Thus, the staff has deemphasized the bottomline numbers in both the IPE,

and more importantly, for the IPEEE.

1.10 When should the improvements resulting from the

IPE process be carried out, right after the IPE or

wait until the IPEEE is completed? (Ref. D.16, p.

87)

SR: The staff has looked at the interaction between the

internal and external events. The staff has concluded that it is unlikely that the cost effective improvements based on internal IPEs would have a

negative impact on safety for external events (see

Section 6.3.1.2). The Generic Letter states that it

expects each licensee to move expeditiously to correct any vulnerabilities that it determines warrants

correction.

1.11 Since the purpose of the IPEEE is to gain a qualitative understanding of core damage frequency, not

quantitative, is it good enough for seismic PRAs to

just report seismic risk in terms of high, medium, or

low, instead of putting in numbers? (Ref. D.16, p.

89)

SR: No. Core damage frequencies are requested from

the seismic PRA so as to obtain insights and the

relative ranking of the accident sequences, and

dominant components as well as assess the findings

against the reporting criteria.

1.12 In the draft generic letter, how will this information

be used is rather general. It seems that it is up to the

utilities to determine what to do with the results,

how they see fit; make the change or justify not

making the change. However, in the generic letter it

also says that the NRC will assess whether the conclusions the licensee draws from the IPEEE regarding changes to the plant systems or components are

adequate. May be you should include some kind of

an example of either a positive or a negative finding

that you've made that can provide some sort of guide

as to what is adequate and what is not adequate.

(Ref. D.16, p. 92)

SR: If the staff disagrees with what a licensee did, the

options that are open to the staff are contained in

the regulations. The staff may request additional

information via questions pertaining to the submittal, or may impose plant modifications via the backfit

rule.

1.13 Vulnerabilities need to be tied to core damage risk.

If you can't tie a vulnerability to a core damage risk,

then it's not a vulnerability. The Severe Accident

Policy Statement says that if you identified a vulnerability, then what you do next is to see if by fixing that

vulnerability you can reduce the risk of core damage.

So it seems that you have to have a quantitative

number for core damage before and afteryou fix the

identified vulnerability. (Ref. D.16, p. 94)

There are several stages in the process: (1) To identify a vulnerability;, (2) To identify fixes for that vulnerability;, (3) To determine if it's substantial; and

(4) To determine if it's cost-beneficial. The costbeneficial side of it has to be determined by the

reduction in risk to the population outside the plant,

which, in effect, requires a containment failure.

So we havelto identify the vulnerability list and then

follow these steps. Because the IPEEE implements

severe accident policy, we all have to make sure that

we're implementing it in the way intended.

SR: The process described is what the staff would go

through if we chose to backfit a plant. It's not necessarily what the utilities might do in their plant. We

leave it up to the utilities to decide what process they

would use and how they would define a vulnerability.

1.14 Clarify the IPEEE Objectives. (Ref. D.1)

SR: The purpose of the IPEEE is to gain a qualitative

understanding of core damage frequency, not quantitative. Section 1.2 provides the discussion of this

D-5 NUREG-1407
Appendix D

aspect and also points out that some methods have

been developed for evaluating external event hazard

and identifying vulnerabilities that do not produce

estimates of damage frequency. Seismic margins

method is cited as a specific example. The objectives

were reworded to emphasize this point.

2. Backfit Analysis

2.1 The staff has stated that the Office of the General

Counsel is looking into whether the request for an

IPEEE should be under 50.54(f) or the backfit rule.

If the Office of the General Counsel does determine

that the backfit rule applies, would it be correct to

assume that this supplement to the generic letter

will be issued similar to Generic Letter 89-16, the

Hardened Vent of the Mark I's, where doing the

action or performing the IPEEE would be voluntary,

and for those utilities who do not volunteer to perform it, plant-specific backfit analyses would be performed using plant-specific values and criteria in

terms of our own resources required and the scope

of the analyses that an individual plant would have to

perform? (Refs. D.13 and D.16, pp. 90-91)

SR: The staff has determined that a backfit analysis is not

needed for the IPEEE.

2.2 If the General Counsel does determine that this can

go ahead under a 50.54(f) request for information,

will the utilities have an opportunity to look at that

for themselves and perhaps appeal that decision or

do their own analyses, whether we feel that backfit

rule actually applies or not? (Ref. D.16, p. 92)

SR: All utilities will be required to respond to the IPEEE

50.54(f) request.

2.3 Without some kind of numerical or specific figure of

merit, how can one really say that it's cost-effective

to implement one type of fix over another, or even to

make a fix at all? (Ref. D.16, p. 92)

SR: The staff traditionally uses a cost benefit figure of

merit of $1000/man-rem in imposing new requirements.

That does not constrain a licensee to use the same

cost benefit as their criterion for what to fix or what

not to fix. Any criterion that the licensee chooses to

use should be justified.

2.4 What are the criteria that NRC will use in determining what to fix? (Ref. D.16, p. 154)

SR: If the staff determines that a fix is required to bring

the plant into conformance with the regulations, no

cost-benefit or other analysis is required, and the

staff can seek to have the fix implemented, if necessary by Order. If the staff determines that a fix desired by the staff is beyond the design bases or Regulations of the Commission, the staff must prepare a

backfit analysis and submit it to the Commission. A

back!it analysis does not have to include a probabilistic risk evaluation, but can instead be presented

relying primarily on engineering judgment.

2.5 The Severe Accident Policy Statement doesn't give

you any probabilistic numbers that you can really

work with. However, in June 1990, the Commission

directed the staff to consider 10-4 core melt frequency as safe enough as in the Safety Goal Policy

Statement.

We propose to link "how safe is safe enough" with

the Safety Goal Policy Statement about undue risk

in severe accident with adequate protection in the

backfit rule. We could make a very good case that

plants currently are safe enough especially if we use

that number on the individual plant level. (Ref.

D.16, p. 156)

SR: The Commission Safety Goal is not just 10-4 peryear

for core melt. The Commission safety goals are the

quantitative health objectives for fatalities. The staff

had recommended that subsidiary objectives (e.g.,

10-4 CDF) be established as a way to implement the

Safety Goals in a practical manner.

In addition, the Safety Goals are not to be used to

judge individual plants. They are to be used to judge

the acceptability of the NRC regulations.

2.6 There is a concern about the closure process, particularly in relation to the fact that both the EPRI

and the Livermore hazard curves will be used in a

seismic PRA.

What's going to happen is ultimately, if a seismic

vulnerability would be expensive to repair, you are

forced into doing some probabilistic type cost benefit analysis. Even though you may have done a seismic margins assessment to identify that vulnerability, I see no way that you are able to avoid not

reverting to using the Livermore hazard curves in

your ultimate decision-making process. I think this is

the time that there ought to be some effort to resolve the difference between these two curves. (Ref.

D.16, pp. 162-164)

SR: When we do backfit analysis and regulatory analysis,

it's not uncommon to have areas of great uncertainty, even as large as the difference between these

two curves. The EPRI and LLNL hazard curves display a level of that uncertainty. Some backfit evaluations are evaluated without having core damage/

risk values, but rather rely on engineering judgment.

NUREG-1407 D-6
Appendix D

The Commission makes use of all the information

available in making its decisions. HCLPF s may also

be an important consideration in making decisions

regarding backfit.

If a licensee prefers to use a single curve in the

seismic review, that is also acceptable provided the

higher of the two curves is chosen. See response 7.8.

3. Cost Estimates and Resource Requirements

3.1 The staff's resource estimate for IPEEE is low.

(Refs. D.9, D.13, and D.16, p. 42)

SR: The staff believes that the cost estimate for IPEEEs

was developed conservatively. Obviously, there are

uncertainties in the costs because there are uncertainties in how a licensee will choose to analyze

external events. The staff has used previous utility

and NRC experience in conducting external event

analysis in arriving at the cost estimate and has attempted to clarify the scope of the IPEEE where

questions on scope were raised. However, where

questions still exist, licensees should come in and

talk to the staff to make sure that they have a clear

understanding of the IPEEE process.

3.2 Six person years was the estimate for the IPEEE cost

or manpower resource requirement, what is the basis? (Ref. D.16, p. 155)

SR: The basis is discussed in Appendix 5 to Generic

Letter 88-20, Supplement 4. Basically what the staff

did was to estimate the overall effort required to

perform the IPEEE and to use the cost spent for

NUREG-1150 plants and the Hatch seismic review

to estimate the IPEEE cost. We also received estimates from a few PRA companies. We do recognize

that there are some costs that licensees will have to

spend over and above just what it will cost the PRA

company. We do believe that the staff's estimates

are reasonable. K. Fleming of PLG Inc., whose firm

has done most industry-sponsored PRAs, and D.

Dube of NE Utilities provided estimates that are

either less than or comparable to the staff's estimates. Table D.2 provides a comparison of the

staff's and the industry resource estimates.

4. Schedule and Response Time

4.1 What is the schedule for IPEEE, staff review, and

fixes? (Ref. D.16, p. 50)

SR: The staff intends to complete closure of severe accident issues in 1995. Accordingly, a three year completion period for the IPEEE is specified so as to

give the staff time to review the submittals.

5. High Winds, Floods, and Transportation and

Nearby Facility Accidents

5.1 The flooding criterion screens at a frequency of 10-s

per year, but the wind screens at 10-8 per year, why is

there an inconsistency between flooding and wind?

(Ref. D.16, pp. 144 & 150)

SR: The screening criteria are essentially a reporting

criteria, which are consistent for all external events

and internal events.

We're not using 10-6 per year flood frequency alone

to screen out floods. Based on a number of flood

studies, the judgment was made that the probable

maximum flood has a 10-5/yr. or less frequency. But

that alone does not screen out the flood concern.

Judgment was made that the conditional failure

probability for a plant designed against the probable

maximum flood is somewhere around 10-1, so that

the combined probability is 10-e or less per year. A

similar approach was used for the wind, where the

design basis wind was usually selected to have a

probability of less than 10-e per year.

5.2 The item 2 on the flowchart for screening external

events, identifying significant changes, does that refer to the hazard at the site or does that refer to the

design of a plant? (Ref. D.16, p. 147)

SR: It refers to the hazard on the site and the land use of

the general vicinity, for example, a new airport built

nearby.

53 For item 3, review plant against the current standard

review plan criteria, why do older plants need to do

that? Obviously, some older plants were not de- signed using current methods and codes. (Ref. D.16,

p. 147)

SR: Item 3 is a screening criterion only. If you know

already that your plant does not meet the 1975 Standard Review Plan (SRP) criteria, you should move

on to the next step in the evaluation process.

5.4 The inclusion of lightning and volcanic activity is

questioned. (Refs. D.1, 8, & 11)

SR: The confirmation that lightning or volcanic activity is

not a significant contributor to severe accidents at

the nuclear power plant, should be assessed to the

satisfaction of the utilities. A relatively simple discussion by the licensee to state why a licensee believes these issues to be unimportant, will be sufficient for these issues for most plants.

D--7 NUREG-1407
Appendix D

Table D.2 IPEEE Resource Estimates (Manhours)

Hatch' Future PLG

Task NUMARC2 Actual Today* SMM+ SPRA3

1. Select SME 0 200 200 0-200

2. Select success paths 600 850 700 600-950

3. Mod. seismic bld'g model 600 430 430 0-950

4. Perform SSI & Dee. FRS 2000 1020 1020 0-2000

5. Soil liq. evaluation 1000 500 500 0-1000

6. Relay chatter evaluation 2000 2800 1700 1400-2100

7. Pre-walkdown 160 160 100-200

8. Walkdown preparation 200 200 100-300

9. Seismic walkdown 900 900 900 600-1400

10. SMM of outliers 1800 2680 2000 1200-2700

11. Report & documentation 700 1320 1000 980-1200

12. Walkdown travel expense 500

13. Containment review isolation,

bypass, etc 800

Long term mitigation 400

14. Misc. cost (startup, plant

support, training, peer review,

NRC interaction) 1200

Total Seismic IPEEE 13000 11060 8810 4980-13000 1700-2700

Surry Peach Btm PRA PLG

Seismic event 1400 1320 1500-3000 1100-1800

Internal fires 350 450 750 900-1500

High winds - - 0-200

External floods 150 250 0-300

Others - - 0-650

QA 450 450 200-500

Plant support 1200 1300 1300-2500 1000-2000

Sources:

1. + D. P. Moore of Southern Company Services, Inc.

2

NUMARC cost estimate of full-scope SMM of Nov. 13, 1990.

3

PLG letter to NRC, dated Oct 9, 1990

5.5 The requirement for assessing GI 103, Design for

Probable Maximum Precipitation (PMP) and lack of

specific guidance in the generic letter are questioned. (Refs. D 1, 8, 11, & 13)

SR: IPEEE imposes no requirements. With regard to

the PMP, NRC acknowledged the importance of this

new PMP criterion in generic letter 89-22, dated

Oct. 19, 1989, by requiring that future plants be

designed against this new criterion. For existing

plants, NRC recommended that the licensees review the material contained in GL 89-22 to determine whether they believe additional action on their

NUREG-1407 D-8
Appendix D

part is necessary. However, this previous review was

directed toward assessing the adequacy of the design

basis, whereas the IPEEE is directed toward severe

accidents. Therefore, the staff is requesting that

PMP be looked at to assess the effects on plants in

terms of onsite flooding and roof ponding to determine whether that would lead to a severe accident.

This is consistent with the staff's request that licensees confirm that no plant-unique external events

with the potential to initiate severe accidents have

been excluded from the IPEEE, as stated in Section

3 of the Generic Letter 88-20, Supplement 4. The

general procedure can be found in section D. 2.4 of

SRP and section 11.4 of NUREG/CR-2300. The

staff believes that this information is readily available per GL 89-22 recommendation.

6. Internal Fires

6.1 The generic letter doesn't state whether the Fire

Vulnerability Evaluation (FIVE) methodology and

the associated database, being developed by

NUMARC and EPRI, are acceptable or not, either

for meeting the IPEEE requirements or for satisfying the issues in the Sandia fire risk scoping study.

Are you planning to put that in the final generic

letter and guidance document? Since we don't know

whether the FIVE methodology is acceptable or not,

we can't really make our decision on what method to

use and respond in the 60-day timeframe as requested in the generic letter. (Ref. D.16, p. 96)

SR: The development of FIVE has been acknowledged

in the generic letter supplement. However, the staff

is unable at this time to endorse the FIVE method

for the use in IPEEE, because NUMARC and EPRI

have not completed its validation and documentation. The staff intends to review the NUMARC/

EPRI methodology and, if it is acceptable, endorse it

as an acceptable way to deal With fires under the

IPEEE. However, final review will not be completed

in time to incorporate a final position on FIVE into

the generic letter, therefore, the staff will address

the acceptability of the FIVE methodology in a separate document.

With regard to the 60-day initial response time, that

has been extended to 180 days after the issuance of

the final generic letter, primarily to allow time for

completion of the FIVE methodology.

6.2 What are the procedurally directed walk-downs in

terms of addressing seismic-fire interaction. Do

they pertain to walk-downs for the fire or walkdowns for the seismic IPEEE. (Ref. D.16, p. 130)

SR: The procedurally directed walk-downs associated

with internal fires vulnerability evaluation can be

planned as part of the seismic walk-downs that

would specifically look for the seismic-induced fire

vulnerability issues. The idea is to first identify those

areas that could be vulnerable so that they can be

brought into focus during the walkdown.

For example, if a plant didn't have its diesel fuel tank

strapped down properly one could postulate a large

fuel source for fire as a result of a seismic event.

Other similar seismic/fire interactions were summarized in Section 7 of NUREG/CR-5088.

6.3 If the utilities had already assessed the safe shutdown capability, are spot checks acceptable for the

cable routing verification? (Ref. D.16, p. 130)

SR: licensees should rely on previous assessments for

IPEEE information, provided the information is upto-date. The licensee does not have to perform any

design verification, or retrace cables if that had been

done previously.

6.4 Will the fire database be updated and when will it be

available? (Ref. D.16, p. 132)

SR: EPRI is updating the fire database through 1988. It

is expected to be available to staff sometime in the

December-January timeframe.

6.5 How is safety system separation to be assessed in the

fire analysis? (Ref. D.16, p. 132)

SR: Separation should be modeled as it exists and the

fuel sources as they exist in order to understand,

using the codes (propagation analysis), the effects of

fire on redundant trains of equipment.

6.6 In treating of transient combustibles for fire, would

it be sufficient in the IPEEE fire analysis to state

that administrative control for transient combustibles takes care of this or would additional quantitative analysis be needed? (Ref. D.16, p. 148)

SR: Transient combustibles cannot be ignored. If they're

procedurally controlled, a fire protection engineer

should be involved in doing the examination, so that

a determination can be made as to whether procedural control will really limit significant transient

combustibles in a given area.

6.7 The requirement stated in Sec. 4.3 Item 5, "provide

documentation in each case where the plant-specific

data used is less conservative than the approved data

base" is a disincentive to use plant specific fire

initiation data. The IPEEE results will capture generic vulnerabilities instead of plant specific fire

vulnerabilities by preferring generic data (Ref. D.8)

SR: In most cases, plant-specific data are rather limited.

The use of generic data, which usually have a much

D--9 NUREG-1407
Appendix D

broader base, will provide valuable insights about

what could happen at a plant. The awareness of this

kind information is very important to the plant operating staff. This is consistent with the purpose of the

Commission's Severe Accident Policy, "to understand the most likely severe accident sequences'that

could occur at its plant."

6.8 Water as a suppression agent may cause potential

damage to safe shutdown components. What about

CO2 and Halon? The Sandia study states that there

are no data available to quantify damage from these

sources. Is a simple statement or identification as

potential damage adequate? (Ref. D.16, p. 149)

SR: In the majority of instances seen to date in power

plants, water caused the damage. Therefore, the

staff believes water damage is the most probable. If

there is an area where suppression damage from

CO2 or Halon is likely, however, one should not

ignore it.

6.9 A major emphasis of the IPEAIPEEE is to have licensee staff involved to help ensure the most benefit. NUREG-1407 states that should involve engineering judgments of the fire safety experts. Does it

mean a fire protection engineer is adequate? (Ref.

D.16, p. 166)

SR: The staff would consider a fire protection engineer

of a utility to be a fire safety expert.

6.10 The inclusion of GI-57, "Effects of Fire Protection

System Actuation on Safety-Related Equipment,"

was questioned. (Ref. D.1)

SR: The effect of fire suppressants on safety equipment

is one of the safety issues identified in the Fire Risk

Scoping Study, NUREG/CR-5088. Relevant information on whether the actuated fire protection systems would spray safety-related equipment or not

and some protective measures to prevent the safety

equipment from being sprayed by fire suppressants,

can be collected during the walkdown. The additional effort to collect this information during the

IPEEE walkdown should be minimal. We also want

to point out that the resolution of this issue does not

have to be a part of the IPEEE.

7. Seismic Events

7.1 What is the justification for including Earthquakes

in IPEEE? (Ref. D.7)

SR: Based upon the examination of NRC's and industry's plant-specific probabilistic risk assessments

(PRAs), the mean core damage frequencies at some

plants could be relatively high, in the range of 1E-4

to 1E-6 per year (rable D.3). Many cost-effective

improvements that would reduce the potential risk

were identified from these PRAs; many were implemented at plants as discussed in Appendix 5 of Supplement 4 to Generic Letter 88-20. This finding is

Table D.3 Summary of PRA Results of Core Damage Frequency (ME-5)

Total Total Int'l Extn'! High LightPlant Total Int'l External Seismic Fire Flood Floods Winds ning

PWR

Pt. Beach 31.3 13.9 17.4 6.1 3.3 7.7 - 0.4 0.006

Turkey Pt. 23.6 7.1 16.51 .7 7.5 - 4.6 2.4 0.26

St. Lucie 7.44 1.4 6.04 1.3 4.4 - 0.32 - 0.02

ANO 1 17.9 8.8 9.15 7.3 0.58 - 0.72 0.53 0.02

IP2 43.5 6.0 37.5 14.0 19.2 - - 4.3 -

IP3 15.7 9.0 6.7 0.31 6.3 - - 0.13 -

Zion 34-40 34.2 0.1-6 < 0.1-6

MS3 15-23 14.7 0.8-8Est - - - -

Oconee 15-28 7.4 8-21 6.0 1.0 - 0 1.3

10.0 (NRC) - 2.5 (NRC) 2.3 (NRC) -

BWR

Quad City 19.7 9.9 9.8 8.3 1.3 - 0.01 0.01 0.2

Cooper 43.7 28.9 14.8 8.1 1.1 - 5. 0.4 0.2

Limerick 9.2 8.4 0.8 0.5 0.3 - - -

Shoreham 7.4 5.4 2.0 - - 2.0 (NRC) -

IP2/IF3--Indian Point 2/Indian Point 3

MS3-Millstone 3

NUREG-1407 D-10
Appendix D

consistent with the statement in the Commission's

Severe Accident Policy that systematic examinations

are beneficial in identifying plant-specific vulnerabilities to severe accidents that could be fixed with

low-cost improvements. References D.2, D.4, D.6,

and D.11 all acknowledged that systematic examinations are valuable tools for gaining insights into the

plant operation and identifying cost-effective plant

improvements.

Another indication that earthquakes can be important risk contributors can be seen from the earthquake experience on foreign nuclear power plants.

For example, (1) On April 22, 1987, Fukushima

units 1, 3, and 5 in Japan, tripped as a result of an

earthquake with a magnitude of 6.6; and (2) On

March 4,1977, one steam generator at the Kozloduy

nuclear power plant in Bulgaria was displaced by 5

inches. This earthquake experience caused a major

overhaul in the seismic design of later Russian

plants. It also resulted in a major backfit at the

Armenia unit 2 in Soviet Russia.

Therefore, based on risk considerations and the potential for identifying cost-effective improvements,

the NRC concludes that seismic events should be

included in the IPEEE.

7.2 The use of both Seismic Hazard Curves (LLNL and

EPRI Seismic Hazard Curves) was questioned

(Refs. D.1, 6, 7, 11, & 20)

SR: There is not enough earthquake data at this time to

determine precisely the validity of a single curve. In

other words, there is no way to conclusively validate

or dispute either curve. Therefore, both the LLNL

and EPRI seismic hazard curves are recommended

for use in the seismic PRAs. This is consistent with

the recommendation of the NUREG-1150 peer review group. An acceptable alternative for licensees

choosing to perform only one analysis is to use the

higher curve. Also see response 7.8.

7.3 Are extensive margin calculations needed for all

components? (Ref. D.7)

SR: Extensive calculations of HCLPFs are not needed in

order to achieve the NRC's goals for the seismic

IPEEE. Refer to response 7.17 for the scope of

margins evaluations. Also, see revised Sections

3.2.4.6 and 3.2.5.7 of this report.

7.4 IPEEE Objectives and Methods: ... recommend

that the objectives of the IPEEE be modified to

better delineate IPEEE objectives for each of the

accepted methodologies, or other non-probabilistic

methods that may be proposed by licensees." (Ref.

D.1)

SR: References to probabilities are now eliminated from

the objectives. The IPEEE objective is to gain a

qualitative understanding of the overall likelihood

of core damage and fission product releases. This is

different from that of the IPE, where quantitative

understanding is emphasized. In Section 1.2 of this

report, the staff has acknowledged specifically that

"some methods have been developed for evaluating

external hazards and identifying vulnerabilities that

do not produce estimates of core damage frequency.

.Thus, objective 3 above would be addressed only

indirectly for some methods to be used in the

IPEEE" It should be noted that the seismic margin

methods were derived from seismic PRAs. The objectives have been reworded to emphasize this point.

Also, see the staff response to items 1.11 and 1.14.

7.5 Scope and Depth of Assessment: "The scope of

evaluations requested for seismic IPE is more extensive than needed to satisfy the Severe Accident Policy Statement." (Ref. D.1)

SR: The staff, based on a NUMARC recommendation,

has defined three review categories with varying levels of effort. This approach leads to an overall reduction in the effort to carry out the examination. Additional details are discussed in response to comment

7.17.

7.6 Seismically induced floods are mentioned for the

first time in draft NUREG-1407, Section 6.3.2 and

not in Generic Letter 88-20 Supplement 4. We understand that the scope of review for seismically

induced external flooding is limited to a review of

external sources of water (e.g., tanks, upstream

dams, or other significant structures) and not internal water sources such as piping. This should be

clearly stated in Generic Letter 88-20, Supplement

4, in order to avoid possible confusion in future

interpretations. (Ref. D.1)

SR: The scope of the seismically induced floods, in addition to the external sources, includes the evaluation

of some internal flooding consistent with the discussion in Appendix I, Check Lists and Walkdown Data

Sheets, of EPRI NP-6041. Section 6.3.2 will be

modified to include reference to EPRI NP-6041. In

addition, the generic letter has been modified.

7.7 Draft Generic letter 88-20, Supplement 4, Section 5

lists the three related programs subsumed in the

IPEEE: (1) the external event portion of USI A-45,

"Shutdown Decay Heat Removal Requirements,"

(2) GI-131, "Potential Seismic Interaction Involving

the Movable In-Core Flux Mapping System Used in

Westinghouse Plants," and (3) the "Charleston

Earthquake Issue." When the IPEEE is utilized for

closure of a subsumed issue, we understand that no

D-11 NUREG-1407
Appendix D

special evaluation, documentation, or reporting will

be needed beyond the program defined by IPEEE

(Ref. D.1).

SR: It is generally true that evaluation and reporting

beyond that identified in this report should not be

required for subsumed issues except for any additional information that may be needed as a part of

the normal staff review of the IPEEE submittal.

Note though, that the IPEEE submittal is to address

specifically USI A-45 and GI-131. Closure of the

IPEEE also means closure for these issues. No specific reporting requirements are identified for the

Eastern U.S. Seismicity Issue, formerly identified as

the Charleston Earthquake Issue.

7.8 Seismic hazard related comments from Ref. D.1, p

8, Comment 1, are summarized as follows:

a. Use of two hazard curves is not needed. Additional expenses for two analyses not justified.

b. The EPRI methodology has been reviewed and

accepted by the USGS. The LLNL results are

not realistic.

SR: a. The staff has revised its position regarding the

use of hazard curves for a PRA analysis in response to this and other similar comments. The

staff still prefers that both LLNL and EPRI

hazard curves be used in an analysis as this will

serve to highlight uncertainties in the bottom

line numbers as well as robustness in the

identification of vulnerabilities. However, an

option of using one hazard curve is now included provided the higher hazard curve of the

two is used. The reasons for using the higher

hazard curve are twofold. One, as discussed in

more detail in b. below, is that the validity of

one curve over the other has not been determined yet. The second is that the use of the

higher hazard curve will ensure that all potential seismic sequences are identified.

Comments related to expenses appear to result, to a certain extent, from a misunderstanding of the scope of analyses required to obtain

results using two hazard curves. Similar comments were also made at the workshop specifically suggesting that two hazard curves will necessitate two separate plant response and

fragility analyses. However, the staff never intended that two separate response or fragility

analyses would be needed and therefore specified only one spectral shape. This point is further emphasized in Section 3.1.1.2 of this report by categorically stating: "Since only one

spectral shape is used for both hazard analyses,

two separate plant response and fragility analyses are not needed." The additional effort to

generate results from an additional hazard

curve is relatively trivial, and requires convolution of a hazard curve with the existing plant

level and sequence level fragility curves.

b. Licensees east of the Rocky Mountains using a

seismic PRA for the IPEEE examination are

requested to use the results of two seismic hazard studies. These studies, conducted by the

Lawrence Livermore National Laboratory

(LLNL) and the Electric Power Research Institute (EPRI), represent a state-of-the-art developmental effort. However, for reasons associated with methodology development, these two

studies can produce significantly different seismic hazard curves.

The uncertainty associated with seismic hazard

assessment is clearly demonstrated when one

compares the vast differences between the

mean, median, 15th, and 85th percentile estimates associated with one curve. The differences between the LLNL and EPRI hazard

curves further demonstrate the large uncertainty.

In an attempt to resolve the differences between the two curves the staff requested assistance from the National Academy of Sciences.

The Academy has criteria it uses to evaluate

scientific and technical approaches used in research. It concluded that both studies followed

good scientific procedures and practices, and

therefore both studies are credible. Therefore,

the staff is encouraging licensees to use both

LLNL and EPRI seismic hazard curves in the

IPEEE evaluation. However, if only one curve

is used, it should be the higher one. As stated

earlier, the use of both will serve to identify

differences, if any, in the delineation of dominant seismic sequences. Although NUREG1150 studies did not identify any significant differences in ranking and contributions, variations in contributions and rankings could occur when curves have markedly different slope

characteristics. Taken together, these pieces of

data (contributions and rankings from both

hazard estimates) give a complete representation of the seismic event. These data will be

extremely useful to the licensee in identifying

plant vulnerabilities and deciding if plant modifications are warranted.

The NUMARC claim about the review and acceptance of the EPRI methodology by USGS appears to

mischaracterize the extent and intent of the USGS

NUREG-1407 D-12
Appendix D

review. Certain observations need to be made regarding the USGS review of the EPRI methodology.

First of all, the USGS reviewed the EPRI methodology as a staff contractor and their findings were

incorporated in the staff Safety Evaluation Report

(SER) addressing this methodology. Second, the review was limited to the methodology and did not

include geology, tectonics, ground motion, or site

specific results. The staff SER (Richardson, 1988)

clearly stated that acceptance was limited to the

methodology, and any application to regulatory issues was not part of the approval. Finally, there were

a number of caveats in both the USGS and the staff

evaluations and neither indicated a blanket acceptance of the EPRI results. The staff conditioned its

approval by noting several areas in which problems

may arise if certain precautions were not observed

based on the USGS review. The staff concluded that

"... . the staff intends to use seismic hazard calculations resulting from the application of the SOG/

EPRI methodology in conjunction with similar

results obtained from LLNL Seismic Hazard Characterization Program (SHCP). If significant differences are observed that can not be resolved, the

NRC staff will use the tw6 sets of calculations to

define the range of seismic hazard to be used in the

decision making process. In any case, these uncertainties are such that the specific calculation of seismic hazard, be it that obtained by EPRI or LLNL,

should be viewed with some caution. The staff finds

that seismic hazard calculations are better used for

making relative comparisons than for placing reliance upon the specific numerical estimates."

7.9 The two Eastern U.S. plants that were placed in the

0.5g RLE bin need not be reviewed at that level.

Recommend adding a footnote that should read

"... indicates an Eastern U.S. Site whose RLE is

greater than 0.3g unless the licensee can demonstrate on site specific examination that the plant's

seismic exposure is similar to, or less than, those

plants assigned to 0.3g RLE". (Refs. D.1, 10, and 14)

SR: Since the publication of the draft generic letter and

the guidance document, both plants placed in the

0.5g RLE bin have committed to either enhance the

existing PRA (Ref. D.2) or to perform a new seismic

PRA (Ref. D.19); therefore, they have been removed from this category.

7.10 Reporting of HCLPF values for components, sequences, and the plant for both new and existing

PRAs should not be required. (Ref. D.1)

SR: The staff has accepted this comment and the reporting of HCLPF values for licensees using PRA methods is now optional. The staff has also indicated that

this information is readily available from PRAs, and

the staff intends to use HCLPF or margin related

insights in the evaluation of the IPEEE submittals. It

should be further noted that it is a current practice

to include HCLPF information in PRA submittals.

7.11 Draft Generic Letter 88-20, Supplement 4, Appendix 4, Section 4.2.2, Item 6, for SMA method, calls

for calculation of HCLPF values "with or without

nonseismic failures and human actions." This item

should be clarified to state that it does not apply to

the EPRI SMA methodology. We understand that

was the intent because, in the EPRI SMA method,

success paths are chosen avoiding unreliable equipment and unrealistic human recovery actions. (Ref.

D.1)

SR: The staff agrees and will clarify the noted item. Also,

see response to item 7.6 of Section D.1 of this Appendbx.

7.12 With regard to the containment performance evaluation (Ref. D.1):

a. It should be clarified that only systems required

to prevent early containment failure need to be

assessed.

b. For reduced-scope plants, we recommend no

containment performance evaluation.

SR: a. The staff has now included this clarification

(see the staff response to item 7.7 of Section

D.1 of this Appendix).

b. The staff is still recommending retention of the

walkdown of containment systems necessary to

prevent early failures because the walkdown

will identify anchorage and spatial interaction

problems that can occur. Such a review of containment is consistent with the defense-indepth philosophy adopted in other parts of this

program.

7.13 The draft generic letter states that the Charleston

earthquake issue is subsumed in the IPEEE and that

completion of the IPEEE will constitute a resolution

of the Charleston earthquake issue. We believe the

Charleston earthquake issue should be closed based

on the information contained in EPRI Report,

NP-6395-D. (Ref. D.1, & Ref. D.5, Attachment 1)

SR: The issue of the 1886 Charleston earthquake has

been resolved. The issue of eight outlier plants identified thru the Eastern U.S. Seismicity program has

been subsumed in the IPEEE and no specific reporting is required to close this issue. The staff will

review the IPEEE results for the affected plants.

7.14 Median estimates of seismic hazard curves should be

used rather than mean values. Mean values are

D-13 NUREG-1407
Appendix D

unduly influenced by outlier experts and thus are

unstable. Median values are less affected by the

extreme estimates and thus provide more stability.

(Ref. D.1)

SR: The staff is recommending the use of mean hazard

curves for the following reasons: (1)The use of mean

hazard curves and mean fragility curves will lead to

approximate mean level frequencies for core damage. No statistical meaning can be attached to a

point estimate obtained through use of median hazard curves. (2) The instability and uncertainties are

better presented by displaying mean results from

two hazard studies as recommended by the staff. The

use of median curves is tantamount to ignoring uncertainties and some expert opinions without an

adequate technical basis.

7.15 Scope of relay chatter evaluation. (Ref. D.1)

SR: See the staff response to item 7.17 of this section,

response to Attachment 1 of Ref. D.1.

7.16 If the intent for the NRC SMA method is to require

the development of level 1 and 2 functional sequences from event trees, the cost of SMA using the

NRC method would be substantially increased.

(Ref. D.1)

SR: The staff does not require the development

of functional event trees beyond that contained

in NUREG/CR-4334, NUREG/CR-4482, and

NUREG/CR-5076, to address containment performance issues when the margin approach is used.

(Note that for the PRA approach the tie-in between

Level I and Level 2 is quite clear, and this should not

be an issue). As stated in this report, the licensee

should develop its own containment performance

plans based on the IPE results. What is required is to

examine containment functions (regardless of the

plant damage states that may be indicated by Level 1

margin sequences or success paths) required to prevent early failures and report HCLPFs for these

functions and components if below the RLE. Obviously, a licensee has an option to develop Level 2

trees at its discretion. Discussions of various ways to

extend Level 1 margin analysis to Level 2 are contained in Budnitz 1991a and 1991b for both NRC

and EPRI methods. (This is suggested for general

guidance, no specific requirements based on these

references are implied.) A success path oriented

approach is also discussed in Ref. D.1 and Reed, et

al., 1990. In summary, the cost of the NRC SMA

method need not be greater than the EPRI SMA

method.

7.17 Summary of NUMARC recommendations for the

implementation of the seismic aspects of Generic

Letter 88-20, Supplement 4. (Ref. D.1, Attachment

1)

SR: The staff responses to these summary recommendations are provided section by section in accordance

with Attachment 1 of Ref. D.1. Many of these recommendations have been discussed earlier. Detailed discussion is included only when the staff does

not agree totally with a recommendation.

Selection of Full-, Focused- and Reduced-Scope

Plants

Reduced-scope, 0.3g RLE Review (Full-scope and

Focused-scope SMA):

The staff has accepted the NUMARC suggestion of

creating full- and focused-scope categories in the

0.3g bin. The 0.3g bin is subdivided into the full and

focused scope based on the NUMARC suggested

approach of using both hazard and seismic design

basis as parameters. Additional consideration was

also given to the identification of outlier plants resulting from resolution of the Charleston earthquake issue (see Appendix A of this report for more

discussion).

.1 0.5gRLE Review: Provide opportunity to two Eastern U.S. plants in the 0.5g bin to submit site-specific

justification for a binning change from 0.5g to 0.3g,

similar to consideration given to Western US plants.

SR: See response to item 7.9 of this section.

.2 Multiple Units at a Site: Lessons learned in evaluating

the first unit may be used in examining the other

unit(s); in particular, any areas of concern that may

be identified during the evaluation of the first unit

would be examined in the other unit(s). Otherwise,

the scope of review for the other unit(s) can be

reduced accordingly.

SR: The staff agrees'that results and findings from the

first unit should be used to help in the evaluation of

other units, provided appropriate similarities exist.

However, such judgements can be made only on a

case-by-case basis. This report has been revised to

include a statement to this effect. In any event, walkdowns of all units will have to take place to ensure

similarities. It is very likely that the greatest reduction would be achievable in analytical effort.

.3 Scope of Detenninistic Seismic Review: Identification

of Success Path Elements. For all three types of

review (full-, focused-, and reduced-scope), procedures for identifying structures and equipment to be

NUREG-1407 D-14
Appendix D

reviewed are the same and are based on the recommendations in EPRI Report NP-6041.

SR: The staff agrees and has referenced EPRI NP-6041

as the primary document for the EPRI success path

methodology. However, due consideration should

be given to supplemental comments made in Sections 3.2.5.1 and 3.2.5.8 of this report regarding the

selection of success paths.

.4 Containment Review: The full- and focused-scope

SMA reviews should be limited to evaluation of only

those functions that are necessary to prevent early

containment failure.

SR: See the staff response to item 7.12 of this section.

.5 Relay Evaluation-" The following table outlines the

NUMARC recommended position on relay chatter

evaluation. This table is based on presentation made

to the staff by NUMARC on November 29, 1990.

Table 7.17.2 NRC Recommendations for Relay Chatter

Review

Review Type Plant Type Recommended Review

Full-scope A-46 Follow A-46 procedures

for A-46 review. Expand

scope to include IPEEE

systems using appropriate margin or A-46

procedure. Review at

assigned RLE.

Non A-46 Review all IPEEE

systems using appropriate procedures at

RLE.

Focused-scope A-46 Same as NUMARC

recommendation.

Non A-46 Same as NUMARC

recommendation.

Reduced-scope A-46 Same as NUMARC

recommendation.

Non A-46 Same as NUMARC

recommendation.

Comparison between Tables 7.17.1 and 7.17.2 indicates that the staff is in agreement with NUMARC

on the focused- and reduced-scope categories encompassing the majority of plants. Reasons for differences in the full-scope review are discussed in

Section 3.2.1 of this report. It should be noted that,

for plants performing PRAs, the scope of the relay

review is also defined by the above table.

.6 Soil Failure Investigation For plants in the focusedscope SMA category, a review based on the design

and construction record is considered adequate. A

review of soil failure should not be required for

plants in the reduced-scope bin.

SR: The staff has adopted both recommendations. For

the focused-scope category, the use of design and

construction records is considered adequate provided appropriate data is available. A detailed analysis will be performed at the licensee's discretion if

soil failure is found to be significant. For the

reduced-scope bin, no soil evaluation is required.

However, it should be noted that the need and the

effort required to evaluate soil failure is site specific

and should be determined on a case-by-case basis.

.7 Screening Criteria" Tables 2-3 and 2-4 of EPRI

NP-6041 can be used. The A-46 screening guidance

given in the Generic Implementation Procedure

(GIP) may also be used.

SR: The staff has adopted this recommendation. It

should be noted that all caveats given in the margin

Table 7.17.1 NUMARC Recommendations for Relay

Chatter Review

Review Type Plant Type Recommended Review

Full-scope A-46 Evaluate A-46 per A-46.

For relays within IPEEE

(not in A-46), perform a

bad actors review.

Non A-46 Perform a bad actors

review for all relays

within IPEEE.

Focused-scope A46 Evaluate A-46 relays per

A-46 (SSE). If bad actors

are found, expand scope

to include IPEEE relays.

Non A-46 Perform a bad actors

review for all relays

within IPEEE.

Reduced-scope A-46 Perform A-46 review. No

additional review for

IPEEE relays.

Non A-46 No relay evaluation.

SR: The staff recommended relay chatter evaluation is

outlined in the following table.

D-15 NUREG-1407
Appendix D

methodology as well as limitations on the use of

GERs should be observed, and the IPEEE review is

to be performed at the assigned RLE. Spatial interaction issues, such as flooding discussed in EPRI

NP-6041, must be addressed.

.8 Evaluation of Outliers: For both full- and focusedscope SMA reviews, HCLPFs should be determined

for elements not screened out during a walkdown.

For focused-scope reviews, it is recommended that

judgement be used to rank the capacities of the

outliers from the lowest to the highest. HCLPF capacities should be calculated as necessary for some

components, other components should be assigned

conservative HCLPFs. For reduced-scope plants,

outliers should be evaluated according to the plant

FSAR.

SR: The staff has adopted these recommendations. See

Sections 3.2.4.6 and 3.2.5.5 of this report.

.9 Seismic Input" For full- and focused-scope SMA reviews, use NUREG-0098 median spectra anchored

to the RLE for the plant. For reduced-scope reviews, use spectra developed for the SSE ground

response spectrum.

SR: The NUMARC recommendation of the use of

NUREG-0098 is consistent with the staff recommendation. The suggested recommendation for the

reduced-scope review is accepted with a caveat that

any difference between FSAR and new response

spectra should be highlighted and discussed.

.10 Review Docwnentation: The documentation of the

IPEEE for the full- and focused-scope SMA review

should follow the guidance outlined in EPRI Report

NP-6041. The report for the reduced-scope review

should be concise.

SR: The staff expects that information outlined in Appendix C will be included in the IPEEE submittals.

Integration of IPEEE and A-46 Reviews:

.11 It is recommended that IPEEE and USI A-46 reviews be conducted concurrently and that the review

tasks be combined whenever possible.

SR: The staff welcomes the NUMARC emphasis on integrating these two major seismic efforts; this recommendation is consistent with the staff philosophy

discussed in Section 6 of this report.

Scope of Seismic Review Using SPRA Approach:

.12 Use of Seismic Hazard Results.

a. Hazard results presented in EPRI Report

NP-6395-D can be used in performing the

SPRA.

b. NRC should allow licensees an opportunity to

perform site specific studies in order to develop

new, more realistic seismic hazard data.

SR: a. See the staff response to item 7.8 in this section.

b. Licensees always have the option to conduct

additional studies they deem necessary and

present them to the staff for review. However,

the new hazard should not be used in lieu of the

LLNL hazard, but it can be used to provide

additional insight into uncertainties.

.13 Fragility Calculations: Mean fragility curves are adequate for fragility calculations.

SR: This is identical to the staff position.

.14 Relay Chatter: Consideration of relays in a seismic

PRA should be limited to relays with low seismic

ruggedness.

SR: The staff preference is that the relay chatter review

scope be defined by the plant categorization used in

the margin review. That is, for a plant identified in

the full-scope category, if the licensee chooses to

conduct a seismic PRA, the relay review is to be

done as outlined for that category. Relay fragilities

and recovery actions should be modeled in the PRA

as appropriate.

.15 HCLPF Calculations: HCLPF calculations should

not be required for a SPRA.

SR: The staff has made this an optional recommendation. See response to item 7.10 of this section.

7.18 The New Hampshire Yankee and Boston Edison

requested that the NRC recognize their use of PRA

for performing the seismic portion of the IPEEE, in

both NUREG-1407 and GL 88-20, Supplement 4,

prior to final issuance. (Refs. D.2 & 19)

SR: The staff has modified GL 88-20, Supplement 4,

and this report to acknowledge the licensee's commitment to use PRAL

7.19 An alternate binning approach, using both hazard

and seismic design basis considerations, should be

considered. (Ref. D.4)

SR: From several suggestions regarding the binning

process, the staff has accepted the binning process

recommended by the NUMARC to further subdivide the 0.3g bin plants into the full- and focusedNUREG-1407 D-16
Appendix D

scopes. Many of the individual utilities have endorsed the NUMARC comments.

7.20 Utilities should be given the option of using either

LLNL or EPRI hazard results. (Ref. D.5, Attachment 1)

SR: See the staff response to item 7.8 of this section.

7.21 The IPEEE should not be required for closure of

Charleston for every plant. (Ref. D.5)

SR: See the staff response to item 7.13 of this section.

7.22 The NRC should consider modifying the bin categories based on the design hazard concept proposed

jointly by NUMARC/EPRI. An alternate binning

scheme is suggested..(Ref. D.5)

SR: The staff has considered the NUMARC/EPRI approach in further subdividinig the 0.3g bin. The staff

binning approach is more consistent with the

NUMARC suggested approach in Ref. D.1.

7.23 Clarify the extent of peer review for the seismic

IPEEE. (Ref. D.5)

SR: The staff intent is now clarified in Sections 3 and 7 of

this report and the generic letter, the extent of the

peer review should be consistent with the IPE guidance as provided in NUREG-1335.

7.24 The use of two hazard curves is illogical. Allow the

use of EPRI hazard data. (Ref. D.5)

SR: See the staff response to item 7.8 of this section.

7.25 Recommend deleting the containment walkdown

for reduced-scope studies. (Ref. D.5)

SR: See the staff response to item 7.12 of this section.

7.26 In Section 3.2.6 there is some confusion. If you utilize the IPE to identify "success paths," then you

would not identify sequences and seismic failure

modes that are significantly different from those

found in the IPE internal event evaluation. (Ref.

D.5)

SR: Even if the IPE is used to identify success paths,

failure modes such as passive failures, structural

failures, and spatial interaction failures are generally not considered in an internal event IPE. Additionally, the "common cause" effect created by a

seismic event is unique in that the entire plant is

subject to the ground motion causing combinations

of failures that may not be manifested in an internal

event IPE. Thus different failure modes and combinations of failures can induce sequences that are

different from those found in the IPE internal event

evaluation.

7.27 The Charleston issue should be closed for a majority

of the Eastern U.S. plants; for the outliers, the issue

can be subsumed through the IPEEE.

SR: The staff agrees. See response to item 7.13 of this

section.

7.28 Does Section 6.3.2 imply that seismic event success

paths must also be simultaneously protected from

postulated fire/floods? The sentence 'The effects of

seismically induced external flooding and internal

flooding on plant safety should be included" is not

clear. (Ref. D.5)

SR: With regard to floods, see the comment and the staff

response to item 7.6 of this section.

With regard to fire, see the staff response to item 6.2

of this section.

7.29 The sentence "However, the licensee should assess

the significance of HCLPF values lower than RLE

and take any necessary actions and make other improvements that are deemed appropriate by the licensee." is too arbitrary. More specific guidance is

necessary. (Ref. D.5)

SR: The judgements about the significance of findings

can be made only when findings are available; therefore, more specific guidance is difficult to give at this

time, and such attempts may create confusion. However, the intent of the statement is to limit the scope

of evaluation for which significance needs to be assessed.

7.30 The staff binning process only recognizes hazard and

not design. (Ref. D.5)

SR: See the staff response to item 7.22 of this section.

7.31 Provide additional guidance about some twelve

plant sites east of the Rocky Mountains whose main

Category I structures are located on rock, and also

have some Category I structures or components located on shallow or intermediate depths of soil.

(Ref. D.5)

SR: The RLE assignment has been made considering

soil conditions where the main plant structures are

located, namely, at rock level for the above cases. As

noted in this report, significant amplification may

occur through the soil layers above the rock, and,

hence, plant structures founded on soil may

experience much higher motion than the rockfounded structures. In such cases, the use of screenD-17 NUREG-1407
Appendix D

ing tables based on the RLE assignment may not be

appropriate for the soil-founded structures and

components. The licensee should investigate this

soil amplification phenomenon using any suitable

means (e.g., analytical studies, comparisons with

other appropriate studies) to determine how to

evaluate the soil-founded structures and components.

7.32 Suggests an alternate approach related to seismic

binning. (Refs. D.5 and 10)

SR: See the staff responses to items 7.17 and 7.19 of this

section.

7.33 The tie-in of seismic margin to a Level 2 PRA is not

defined. Also, clarify "all HCLPFs related to...containment performance" (Ref. D.6, Enclosure)

SR: See the staff response to item 7.16.

7.34 Does resolution of USI A-45 have scope implications for seismic margin options. (Ref. D.6, Enclosure)

SR: Yes, as noted in Section 6-3.3.1, functions and systems for addressing USI A-45 should be same as

those identified in the internal event IPE. Otherwise, some of these functions and systems may not

necessarily be included in a margin evaluation. A

specific reporting provision is also called out in item

7 of Section C.2.2, Appendix C, of this report for this

USI.

7.35 On page 24, Section 4.2.2, Item 2 of the draft generic

letter, replace the term "findings" with "results."

(Ref. D.8)

SR: The staff accepts this comment; appropriate sections are revised to reflect this change.

7.36 A peer review implying the use of external "experts

in the professional field" for a review of the methodology chosen and its application is not necessary.

in-house review is more appropriate. (Ref. D.8)

SR: The staff has now clarified peer review discussions

in the generic letter and this report to be consistent

with the internal event IPE guidance

(NUREG-1335), which emphasizes in-house review. However, the staff has also recognized that a

licensee may not have in-house expertise in all areas

of the external events and an in-house team can be

supplemented by outside experts.

Furthermore, it should be recognized that substantial judgement is involved in applications of PRA as

well as margin methodologies as demonstrated in

trial applications at Maine Yankee (NUREG/

CR4826) and Hatch (Davis, 1990). This is particularly important now that the "focused-scope" category has been introduced requiring more use of

judgement. The composition of the in-house team

should therefore strike a balance so that sufficient

expertise is available to ensure that the methodology

is properly implemented while utilizing in-house

staff as much as possible.

7.37 Based on lessons learned from PRA and Margin

evaluations a simplified walkdown procedure for a

majority of plants should be developed. (Ref. D.9)

SR: The staff has revised the scope of the seismic examination in line with NUMARC's suggestions with

some exceptions.

7.38 The scope and objective of the walkdown are not

sufficiently described. The "40 person months per

unit" required for a walkdown is excessive. (Ref.

D.18)

SR: No changes to Appendix 1 of the Generic Letter are

required. This report was revised to read as follows:

... perform a walkdown consistent with the intent of

the guidelines described in Sections 5 and 8, and

Appendices D and I of the EPRI Seismic ...

The 40 person month estimate indicated in the comment is not consistent with past experience at several plants. For example, EPRI NP-6359 (Seismic

Margin Assessment of the Catawba Nuclear Station) states that the total technical manpower expended was 39.7 man months: approximately 16 man

months associated with walkdown preparation and

the walkdown, 21.5 man months for evaluating unscreened components, and 2.0 man months for reporting. Southern Company Services, Inc. (Hatch

Plant) estimates that they expended 8 staff months

for walkdown planning and the walkdown.

The level of effort and resource allocation are justified based on approximately 20 seismic PRAs and 3

seismic margins evaluations. Both evaluation methods have demonstrated that thorough walkdowns

are one of the most important tools for identifying

seismic weak links.

The importance of a detailed walkdown is also supported by NUMARC. In fact, when NUMARC proposed the Reduced Scope Program for sites in low

seismic areas, a plant walkdown identical to the Full

Scope evaluation was recommended.

NUREG-1407 D-18
Appendix D

7.39 An analysis using both LLNL and EPRI hazard

curves is unnecessary. There is inconsistency between the generic letter and NUREG regarding use

of both curves in an existing PRA. (Ref. D.18)

SR: See the staff response to item 7.8 of this section. The

generic letter and NUREG have been revised to be

consistent.

7.40 To avoid a different interpretation, define HCLPF

for the sequences and plant for the PRA and margins methodologies. (Ref. D.18)

SR: The term HCLPF in the context of the margin methods is dearly defimed in both NUREG/CR-4334 and

EPRI NP-6041. Examples of how plant level and

sequence level HCLPFs are determined can be seen

in NUREG/CR-4826, the margin evaluation for the

Maine Yankee plant. The mathematical definition

of HCLPF for both the PRA and margin methods is

the same. Examples of determining sequence and

plant level HCLPFs are also described in NUREG/

CR- 4334. Section 3.1.1.3 of this report gives further

guidance on how to determine component, sequence, and plant level fragilities when only mean

fragility curves are used.

7.41 Soil liquefaction computations to the level of detail

recommended in EPRI NP-6041 are not necessary

to obtain a qualitative understanding of the overall

probability of core damage and radioactive material

release.

What is appropriate and reasonable is an assessment

as to whether the site is susceptible to liquefaction

behavior. (Ref. D.18)

SR: See the staff response to item 7.17, Soil Failure.

Investigation. The staff has accepted the NUMARC

recommendations in this area.

7.42 Put Farley in the Reduced Scope Program bin. (Ref.

D.10)

SR: Farley is now assigned to the Focused Scope bin.

Plant binning was accomplished by comparing nine

separate pieces of information related to seismic

hazard groupings and engineering judgement.

7.43 Required use of LLNL and EPRI hazard curves adds

significant expense for no significant benefit. The

four purposes of the IPEEE can be met using EPRI

hazard curves. The EPRI method has been reviewed

and accepted by USGS and NRC; the LLNL method

has not. (Ref. D.11)

SR: See the staff response to Item 7.8 of this section.

7.44 Reporting of "functional sequences" may not be

possible if only systemic sequences are generated for

a PRA or the EPRI success path approach is used for

a SMA. (Ref. D.11)

SR: True. For those cases, the reporting criteria are described in Appeendix 3 of Supplement 4 to GL

88-20.

7.45 Reporting HCLPFs with and without non-seismic

failures and human actions does not contribute to

the four stated purposes of the IPEEE. (Ref. D.11)

SR: The fourth purpose of the IPEEE is to reduce the

overall likelihood of core damage and radioactive

material release by modifying hardware and procedures. Cost-effective decisions can be made only if

both seismic and non-seismic failures are included in

the licensee's decision making process. If the random failure probability of a diesel generator to start

is high, no seismic fix to the plant is likely to significantly reduce the frequency of the sequence.

7.46 Using the SMAit is difficult to perform a nonseismic

failure and human action evaluation. (Ref. D.11)

SR: Guidance is noted in Sections 3.2.4.7 and 3.2.5.8 on

how to address the above evaluations for various

methods. Such evaluations have been made and reported in the trial plant applications. Additional

guidance on this issue'is available in (Budnitz 1987

and 1990). The intent is to help the licensee make

the right decisions regarding plant modifications.

See response to item 7.46 above.

7.47 Use spectral shapes consistent with the LLNL and

EPRI hazard studies. Does this mean uniform hazard spectra? Is this consistent with the NUREG/

CR-0098 spectral shape? (Ref. D.11)

SR: This report recommends the use of the median spectral shape for a 10,000 year return period provided in

NUREG/CR-5250 or a site specific spectral shape

based on a suite of appropriate records for performing PRAs. This is different from the NUREG/

CR-0098 spectral shape recommended for evaluations associated with the SMA methodologies. The

reasons for this difference is: PRA takes into account the full range of the hazard requiring use of a

realistic description of ground motion as much as

possible whereas margin evaluations are only conducted at one earthquake level, and the screening

tables used in the margin methods are developed

from earthquake experience data more compatible

with the ground motion represented by the

NUREG/CR-0098 spectral shape.

7.48 Allow justification other than expensive sensitivity

studies for use of a cutoff other than 1.5g. (Ref.

D.11)

D-19 NUREG-1407
Appendix D

SR: Such sensitivity studies are routinely performed to

ensure that an adequate range of integration has

been defined. These studies are not expensive. In

any event, licensees have the option to propose alternative methods. These will be reviewed by the

staff on a case-by-case basis.

7.49 A mean component fragility curve is defined by the

median capacity, A, and composite uncertainty ... Is

this a correct statement? (Ref. D.11)

SR: Yes, the statement is correct.

7.50 Can SOUG GIP walkdown guidelines be used in lieu

of EPRI NP-6041? (Ref. D.11)

SR: See the staff response to item 7.17.7, Screening Criteria, of this section.

7.51 Reporting HCLPFs with and without nonseismic

failures and human actions does not contribute to

the four stated purposes of the IPEEE.

SR: See the staff response to items 7.10 and 7.45 of this

section.

7.52 The licensee should have the option of using "heavy

duty" experts in lieu of a "peer review" in their

SMA. (Ref. D.11)

SR: See the staff response to item 7.36 of this section.

7.53 The Charleston Earthquake Issue does not need to

be subsumed into the IPEEE. (Ref. D.11)

SR: See the staff response to item 7.13 of this section.

7.54 Include more Midwest plants in the Reduced-Scope

Program bin. (Ref. D.11)

SR: See the staff response to item 7.42 of this section.

7.55 No basis is provided for future plant modifications to

maintain the plant margin identified from the

IPEEE. (Ref. D.14)

SR: See the staff response in item 8 of Section D.1 of this

Appendix.

7.56 There is no specific provision in the Generic Letter

to allow a licensee to make their own detailed evaluation to determine which review level earthquake

bin they should be assigned. (Ref. D.14)

SR: The assignment of the review level earthquake was

based on both state-of-the-art LLNL and EPRI

studies. However, the licensee always has an option

to propose an alternative position and submit information to justify it.

7.57 Assignment of the RLE should be allowed to be

based upon complete site-specific evaluation of the

geological and seismological data for the site. (Ref.

D.14, Attachment A)

SR: See the staff response to 7.56 above.

7.58 The cost for Pilgrim will be higher because the staff

has characterized the Pilgrim site as a high hazard

site. (Ref. D.14, Attachment A)

SR: The cost estimates provided in the draft generic

letter are generic. Clearly, some licensees may

spend more because more detailed analyses are

needed. For instance, containment and containment

system performance evaluations for Mark I and Ice

Condenser containments will be somewhat more

expensive since generic capacity data on the systems

are lacking. Licensees have the option of performing

a seismic PRA instead of the margins method. Staff

consultants have indicated that licensees assigned to

the 0.5g bin should seriously consider the PRA option as a means of controlling expenses.

Also, see the staff response to item 7.56 of this

section.

7.59 Coordination With Other External Event Programs.

Subsumption of the Charleston Earthquake Issue

creates a "de facto" provision for IPEEE implementation by pre-supposing that the licensee will be

performing a PRA or accepts the NRC's seismic

hazard estimates in determining the RLE.

SR: See the staff response to item 7.13.

7.60 The draft Generic Letter should address the future

requirements for maintenance of the seismic margin

identified by the IPEEE. (Ref. D.14, Attachment A)

SR: See the staff response to item 8 of Section D.1.

7.61 SMA and PRA relay chatter enhancement has not

been shown to be cost effective, particularly in light

of the A-46 resolution. WE endorses the focused

SMA approach for all plants performing a SMA or

PRA. (Ref. D.11)

SR: The relay chatter evaluation has been changed; see

response to NUMARC's comments in item 7.17 of

this section.

7.62 It is not cost beneficial to include long-term cooling

and pressure suppression systems in the containment performance evaluation. (Ref. D.11)

NUREG-1407 D-20
Appendix D

SR: The staff has revised the scope of the containment

review to focus on the early failure modes. See the

staff response to item 7.17.4 in this section.

D.3 References

Budnitz, R., et al., "Extending a HCLPF-Based Seismic

Margin Review to Analyze the Potential for Large Radiological Releases and the Importance of Human Factors

and Non-Seismic Failures, Draft, March 1987.

-- , "Enhancing the NRC and EPRI Seismic Margin

Review Methodologies to Analyze the importance of

Non-Seismic Failures, Human Errors, Opportunities for

Recovery, and Large Radiological Releases, Draft 2, September 1990.

Chilk, S. (NRC), Memorandum to J. Taylor (NRC), and

W. Parler (NRC), dated July 17, 1990, Subject:

SECY-90-192--Individual Plant Examination for Severe

Accident Vulnerabilities due to External Events

(IPEEE).

Code of Federal Regulations, Title 10 "Energy" (10

CFR), U.S. Government Printing Office, Washington,

D.C., revised periodically.

Davis, P., "A Peer Review of Two Seismic Margin Assessments as Applied to the Hatch Nuclear Power Plant",

Proceedings of Third Symposium on Current Issues Related to Nuclear Power Plant Structures, Equipment and

Piping, Orlando, Florida, December 1990.

Electric Power Research Institute, EPRI NP-6041, "A

Methodology for Assessment of Nuclear Power Plant

Seismic Margin," October 1988.

-- , EPRI NP-6359, "Seismic Margin Assessment of

the Catawba Nuclear Station," Vols. I and 2, April 1989.

-- , EPRI NP-6395-D, "Probabilistic Seismic Hazard

Evaluation at Nuclear Plant Sites in the Central and

Eastern United States: Resolution of the Charleston

Issue," April 1989.

Moore, D., et al., "Results of the Seismic Margin Assessment of Hatch Nuclear Power Plant," Proceedings of

Third Symposium on Current Issues Related to Nuclear

Power Plant Structures, Equipment and Piping, Orlando,

Florida, December 1990.

Orvis, D., et al, "Seismic Margin Review of Plant Hatch

Unit 1: System Analysis," LLNL Report No. ULRLCR-104834, August 1990.

Rasin, W. (NUMARC), letter to W. Minners (NRC),

Subject: Final Industry Comments on Draft Generic

Letter 88-20, Supplement 4, "Individual Plant Examination of External Events (IPEEE) for Severe Accident

Vulnerabilities," and Draft NUREG-1407, "Procedural

and Submittal Guidance for the IPEEE," October 10,

1990.

Reed, J., et al. "Recommended Seismic IPE Resolution

Procedure," Proceedings of Third Symposium on Current

Issues Related to Nuclear Power Plant Structures, Equipment and Piping, Orlando, Florida, December 1990.

USNRC, Generic Letter 88-20, "Individual Plant Examination for Severe Accident Vulnerabilities--10 CFR

50.54(f)," November 23, 1988.

-- , Generic Letter 88-20, Supplement No. 1. "Initiation of the Individual Plant Examination for Severe Accident Vulnerabilities- 10 CFR 50.54(f," August 29, 1989

-_-, Generic Letter 88-20, Supplement No. 4, "Individual Plant Examination of External Events (IPEEE) for

Severe Accident Vulnerabilities- 10 CFR 50.54(f),"

draft for comment, July 23, 1990.

--- , Generic Letter 88-20, Supplement No. 4, "Individual Plant Examination of External Events (IPEEE) for

Severe Accident Vulnerabilities-10 CFR 50.54(f)," final, June, 1991.

--- , Generic Letter 89-22, "Resolution of Generic

Safety Issue No. 103, 'Design for Probable Maximum

Precipitation,'" October 19, 1989.

--- , NUREG-1032, "Evaluation of Station Blackout

Accidents at Nuclear Power Plants," June 1988

-- , NUREG-1150, "Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants," Vols. 1

and 2, June 1989.

--- , NUREG-1335, "Individual Plant Examination:

Submittal Guidance," final report, August 1989.

--- , NUREG-75/087, "Standard Review Plan for the

Review of Safety Analysis Report for Nuclear Power

Plants," LWR edition, December 1975.

-7, NUREG/CR-0098, "Development of Criteria for

Seismic Review of Selected Nuclear Power Plants," May

1978.

--- , NUREG/CR-2300, "PRA Procedures Guide,"

January 1983.

--- , NUREG/CR-2815, "Probabilistic Safety Analysis

Procedures Guide," Vols. 1 and 2, August 1985.

--- , NUREG/CR-4334, "An Approach to the Quantification of Seismic Margins in Nuclear Power Plants,"

August 1985.

D-21 NUREG-1407
Appendix D

-- , NUREG/CR-4482, "Recommendations to the Nuclear Regulatory Commission on Trial Guidelines for

Seismic Margin Reviews of Nuclear Power Plants,"

March 1986.

--- , NUREG/CR-4826, "Seismic Margin Review of the

Maine Yankee Atomic Power Station," Vols. 1-3, LLNL

March 1987.

--- , NUREG/CR-5076, "An Approach to the Quantification of Seismic Margins in Nuclear Power Plants: The

Importance of BWR Plant Systems and Functions to Seismic Margins," May 1988.

--- , NUREG/CR-5088, "Fire Risk Scoping Study,"

January 1989.

--- , NUREG/CR-5250, "Seismic Hazard Characterization of 69 Nuclear Power Plant Sites East of the Rocky

Mountains," Vols. 1-8, January 1989.

-- , "Policy Statement on Severe Reactor Accidents".

Federal Register, Vol. 50, p. 32138, August 8, 1985.

-- , SECY 88-147, "Integration Plans for Closure of

Severe Accident Issues, May 25,1988.

D.4 References for Sources of

Comments

D.1 Letter from W. Rasin (NUMARC) to W. Minners

(NRC), dated October 10, 1990.

D.2 Letter from T. Feigenbaum (New Hampshire

Yankee) to Document Control Desk (NRC), dated

October 3, 1990.

D.3 Letter from W. Orser (Detroit Edison) to W. Minners (NRC), dated October 8, 1990.

D.4 Letter from R. Wheaton (R. D. Wheaton Asso.) to

J. T. Chen (NRC), dated October 6, 1990.

D.5 Letter from M. Lyster (Centerior Energy) to

Document Control Desk (NRC), dated October 8,

1990.

D.6 Letter from J. Garrick (PLG Inc.) to J. T. Chen

(NRC), dated October 9, 1990.

D.7 Letter from P. Smith (The Readiness Operation)

to J. T. Chen (NRC), dated October 4, 1990.

D.8 Letter from G. Muench (Entergy Operation, Inc.)

to S. Chilk (NRC), dated October 5, 1990.

D.9 Letter from J. Skolds (South Carolina Electric &

Gas Co.) to J. T. Chen (NRC), dated October 10,

1990.

D.10 Letter from W. Hairston, Ill (Alabama Power Co.)

to W. Minners (NRC), dated October 12, 1990.

D.11 Letter from C. Fay (Wisconsin Electric Power Co.)

to E. Beckjord (NRC), dated October 5, 1990.

D.12 Letter from G. Sorensen (Washington Public

Power Supply System) to Document Control Desk

(NRC), dated October 5, 1990.

D.13 Letter from N. Reynolds (Winston & Strawn) to E.

Beckjord (NRC), dated October 9, 1990.

D.14 Letter from R. Bird (Boston Edison) to Document

Control Desk (NRC), dated October 12, 1990.

D.15 Letter from D. Shelton (Centerior Energy) to

Document Control Desk (NRC), dated October 16, 1990.

D.16 USNRC Transcript on IPEEE Workshop, Pittsburgh, Pennsylvania, September 11-13, 1990.

D.17 Letter from G. Goering of Northern States Power

Company to F. Miraglia of NRC, dated June 10,

1985.

D.18 Letter from H. Tucker (Duke Power Co.) to E.

Beckjord (NRC), dated August 30, 1990.

D.19 Letter from G. Davis (Boston Edison) to Document Control Desk (NRC), dated March 1, 1991.

NUREG-1407 D-22
ATTACHFMENT TO APPENDIX D

VALUE/IMPACT ANALYSIS

FOR THE IMPLEMENTATION

OF INDIVIDUAL PLANT EXAMINATION OF EXTERNAL EVENTS

1. Introduction

The primary objective of this attachment is to provide a

value/impact analysis to support the issuance of a supplement to Generic Letter 88-20 (Ref. 1) requesting an Individual Plant Examination of External Events (IPEEE)

from all licensees holding operating licensees for nuclear

power plants. Implementation of the IPEEE program is

consistent with the Commission's Severe Accident Policy

(50 FR 32138) dated August 8, 1985 (Ref. 2). The implementation of the IPEEE program will provide the utilities

and the NRC staff with a better understanding of the actual state of the plant and its capability to cope with severe accidents. The IPEEE program may reveal external

event vulnerabilities that could be reduced by procedure

changes or hardware modifications to upgrade the

frontline and support safety systems.

In general, in performing a value/impact analysis (Ref. 3)

the staff would (1) identify potential external event vulnerabilities to severe accidents in operating light water

reactor (LWR) power plants, (2) identify modifications

that could reduce plant risk from these vulnerabilities,

(3) determine the safety benefit of these modifications,

and (4) assess the net cost of the modifications. However,

in this study, we do not know what the utilities will find

from their IPEEE programs. Also, we do not know what

fixes the utilities will propose. Therefore, for this study

we have used data from published probabilistic risk assessments (PRAs) to identify potential vulnerabilities

that could be identified in an IPEEE and compared the

benefit of fixing such vulnerabilities to the cost of doing

the IPEEE as well as the cost of the fixes themselves.

2. PRA Findings

Although the Commission has concluded that existing

plants pose no undue risk to the public, the Commission

emphasized that systematic examinations of existing

plants are needed to confirm the absence of any plant

unique vulnerabilities to severe accidents. This conclusion was based on the fact that previous plant-specific

(PRAs) have typically revealed valuable insights on plant

specific vulnerabilities to severe accidents.

Table 1 summarizes previous PRAresults in terms of core

damage frequencies (CDFs) due to internal and external

events from 13 PRAs (Refs. 4 & 5) that are available to

NRC. These results indicate that the mean value of the

CDF, for these plants is in the range of 1E-4 to 4.4E-4

per reactor-year, with the core damage frequencies due to

external events being in the range from 6E-5 to 3.8E-4

per reactor year. More recently, NUREG-1150 analyses

for Surry (NUREG/CR-4551, Volume 3) and Peach Bottom (NUREG/CR-4551, Volume 4) have indicated that

the mean core damage frequency from fires is in the range

of 10-5/RY and from Seismic events 10-e-10-4/RY.

A common finding from these PRAs is that support system failures have been identified as significant contributors to the probability of core melt. At the support system

level, there is often sharing and interconnection between

redundant trains, questionable separation and spacial independence between trains, and poor overall general arrangement of equipment from a safety viewpoint. For example, many plants have redundant trains of equipment

sitting side by side in a common area and adequate physical separation and protection of redundant safeguard

trains is lacking. This type of general arrangement of

equipment creates vulnerabilities in that single events

such as a fire or a flood can disable multiple trains of

safety related equipment resulting in an inability to cool

the plant.

Table 2 provides a list of the specific vulnerabilities found

from some of these studies and potential modifications to

address these vulnerabilities. In general, external event

vulnerabilities were identified in:

a. Electrical switchgear/battery failures due to seismic

excitation.

b. Water storage tank (CST, RWST) failures due to

seismic excitation.

c. Pump and valve common-mode failures (AFW,

CCW, SWS, HPIS, LPIS, etc.).

d. Fires in cable spreading rooms, switchgear rooms, or

common cable run areas (BWR/PWR).

Modifications include both hardware and procedural

components which, if implemented, would serve to

reduce. the estimated core damage frequency. Table 3

provides an example list of the plant-specific modifications that have been made. These modifications were

made based on the insights gained from plant-specific

PRAs. However, many fixes are made in the course of doing a PRA which are never quantified or reported. For

example, deficient equipment anchorages were found at

almost every plant during seismic walkdowns.

D-23 NUREG-1407
Appendix D Attachment

Table 1 Summary of PRA Results of Core Damage Frequency (1E-5)

Total Total Int'l Extn'! High Light. Plant Total Int'l External Seismic Fire Flood Floods Winds ning

PWR

Pt. Beach 31.3 13.9 17.4 6.1 3.3 7.7 - 0A 0.006

Turkey Pt. 23.6 7.1 16.51 .7 7.5 - 4.6 2.4 0.26

St. Lucie 7.44 1.4 6.04 1.3 4.4 - 0.32 - 0.02

ANO 1 17.9 8.8 9.15 7.3 0.58 - 0.72 0.53 0.02

Mean 20.1 7.8 12.3

IP2 43.5 6.0 37.5 14.0 19.2 - - 4.3 -

IP3 15.7 9.0 6.7 0.31 6.3 - - 0.13 -

Zion 34-40 34.2 0.1-6 <0.1-6

MS3 15-23 14.7 0.8-8Est - - - -

Oconee 15-28 7.4 8-21 6.0 1.0 - 0 1.3 -

10.0 (NRC) - 2.5 (NRC) 2.3 (NRC) -

Mean 25-30 14.3 11-16

BWR

Quad Cities 19.7 9.9 9.8 8.3 1.3 - 0.01 0.01 0.2

Cooper 43.7 28.9 14.8 8.1 1.1 - 5. 0.4 0.2

Limerick 9.2 8.4 0.8 0.5 0.3 - - - -

Shoreham 7.4 5.4 2.0 - - 2.0 (NRC) -

Mean (4) 27 20.2 6.8 - - -

IP2/IP3-Indian Point 2/Indian Point 3

MS3-Millstone 3

usually strengthened, however, and are rarely reported

specifically in the PRAs in terms of the impact on CDF or

averted risk. Based on the insights gained over the last ten

years, almost every systematic examination has resulted

in plant-specific insights, that in conjunction with the

plant specific evaluation of risk reduction options, would

always result in identifying cost effective remedies.

3. Value-Impact Assessment

The analyses performed for the resolution of USI A-45,

Decay Heat Removal (DHR) Requirements (Ref. 6),

were used to make reasonable estimates of the value of

conducting the IPEEE. Specifically, reduction in core

damage frequency resulting from proposed modifications

and the cost of those modifications were evaluated. It

should be noted, however, that the purpose of the USI

A-45 program was to evaluate the adequacy of the DHR

function only; accident sequences that did not involve this

function are not included in the analyses. These excluded

sequences involving large LOCAs, reactor vessel ruptures, the pressurized thermal shock sequence, interfacing system LOCAs, and anticipated transients without

scram (ATWS). Thus, the core damage frequencies derived under that program do not represent the total frequencies for those operating plants. Including the contributions from those excluded events would result in higher

estimated core damage frequencies.

For this study, the following modifications were considered as possible means of reducing the vulnerabilities

which would most likely be uncovered in a plant specific

IPEEE. They were used here for the purpose of assessing

the value-impact of the IPEEE program; however, they

do not necessarily represent the only means for reducing

plant risk to severe accidents.

(1) Seismic Resistance of Batteries and Switchgear

a. Ensure that battery installation racks meet current seismic requirements. All racks should be

steel with appropriate tiedowns to prevent motion under seismic excitation.

b. Provide additional ties to floor for electrical

equipment (transformers, switchgear, buses,

battery chargers, and motor control centers)

for anchorages to prevent cabinet motion

during seismic acceleration. For tall cabinets,

provide additional restraints to prevent toppling.

NUREG-1407 D-24
Appendix D Attachment

Table 2 Modification Options Identified for the Case Studies

Plant Vulnerability Modification

Pt. Beach RWST failures and electric switchgear

failures from seismic events

Service water pumps lost from failure

due to spray

Loss of safety systems due to fire in CSR

and AFW rooms

Turkey Pt. Surge floods safety systems

Loss of cooling due to loss of water tanks

and CCW heat exchangers from seismic

event

Loss of safety systems due to fire in CSR

St. Lucie Loss of safety systems due to CSR fire

Loss of cooling due to loss of water tank

ANO 1 Loss of cooling due to failure of EFWS

pump and to take water from CST

Loss of safety systems due to fire in CSR

Loss of cooling due to loss of tanks and

emergency electric power due to seismic

event

Quad Cities Loss of decay heat removal due to fires in

CR or CSR

Loss of electric power due to seismic events

Cooper Loss of safety systems due to fire in CSR

Loss of cooling due to failure to tanks and

heat exchangers from seismic event

Loss of emergency electric power due to

seismic events

Cooper Alt. Loss of cooling due to seismic events

Loss of decay heat removal due to floods

Provide water from spent fuel pool and add restraints

to switchgear and batteries

Install shield wall to protect pump motors

Install added fire suppression

Increase height of existing flood wall

Increase strength of tanks and heat exchanger

supports

Install additional suppression in CSR

Enclose one train of safety-related cables in fire

barrier

Increase strength of tanks with addition of

external supports

Install provisions to power auxiliary feed pump

from Class IE bus

Add redundant deluge valve with separate sensing

and control

Strengthen tanks with external supports and anchor

switchgear

Enhance operating procedures for the safeshutdown pump

Upgrade battery racks and add restraints to

SWGR and buses

Add fire barrier around HPCI and RBSW

power cables

Install added anchorage or tanks and heat

exchangers

Add supports and tiedowns to switchgear and

transformers

Strengthen HTEX mounts, valve, CST, and

transformer tiedowns

Develop procedures for safe shutdown in

high flood crests

(3) Fire Protection

Where safety-related cabling is concentrated, ensure that

adequate fire protection is provided by installation of additional suppression systems, thermal protection, etc. and

(2) Seimsic Resistance of Tanks

Upgrade anchorages and walls for water storage tanks

(RWST and CST) designed using the procedures of TID

7024 and with H/D ratios greater than 1.

D-25 NUREG-1407
Appendix D Attachment

reliable alternate shutdown capability is available. Review all procedures to ensure that minimal quantities of

fuels are present in fire-susceptible areas (control rooms

and cable spreading rooms in particular).

3.1 Analysis of Specific Modifications

Table 3 summarizes for selected plants the value-impact

analyses resulting from application of the specified

system modifications described above. Besides valueimpact, core damage probability, population and occupational doses, and costs are shown explicitly in the table. As

expected, the value (Col. 4,5, & 6) and the impact (Col. 7

& 8) of any given modification are plant and site dependent. None of the suggested modifications is cost effective

(Col. 9) based on avertible offsite costs alone. However,

some modifications may be cost effective if onsite costs

are included (Col. 10).

3.2 Plant Specific Value-Impact

Analyses

Table 4 summarizes the results of the plant-specific

value-impact analyses performed for USI A-45. Various

combinations of modifications were evaluated for each

plant. Besides value-impact, core damage probability, population and occupational doses, and costs are shown

explicitly in the table. As expected, the value (Col. 3, 4, &

5) and the impact (Col. 6 & 7) of any given modification

are plant and site dependent. None of the modifications is

cost effective (Col. 8) based on avertible offsite costs

alone. However, some modifications are cost effective if

onsite costs are included (Col. 9).

4. Discussion and Conclusion

As can be seen from Table 1, external events can be sig- nificant contributors to overall risk from a nuclear power plant. Previous risk analysis of external events have always uncovered items which were modified to reduce risk

(Table 3). In many cases the reduction in risk resulting

from these modifications was not quantified and thus

value-impacts were not calculated. However, from the

data available from the A-45 analysis (Tables 4 and 5) and

Table 3 the following conclusions can be drawn:

* the cost of the modifications considered may range from approximately 50K to 24M dollars per plant

* the risk-averted (on-site and off-site) may range

from approximately 50 person-rem to 2600 personrem per plant over the life of the plant

* low-cost fixes may be found to reduce risk and may

be cost effective

a for those fixes that were made at specific plants, the

averted risk was significant.

* in many cases, even if the cost of doing the IPEEE

(estimated to be as much as $1 million at the upper

bound) were added to the cost of doing the modifications, the modifications might still be cost effective.

Thus, the staff concludes that there is a high likelihood

that conduct of the IPEEE will result in the identification

of vulnerabilities that, if fixed, would result in a substantial increase in safety and that could be fixed in costeffective manner. Accordingly, the systematic examination of each operating nuclear plant could provide the

most complete compilation of data and analysis available

to develop an integrated perspective on risk from external

events. It could also identify human, procedural, design,

and operation vulnerabilities and could provide practical

means to explore and select cost-effective alternate solutions to plant vulnerabilities. Therefore, a plant-specific

examination, like IPEEE, conducted by analysts with access to plant data and procedures, could better establish

the level of risk and identify cost-benefit improvements at

a particular site.

5. References

1. Generic Letter 88-20, Supplement 4, Individual

Plant Examination of External Events (IPEEE) for

Severe Accident Vulnerabilities, Draft for Comment, July 23, 1990.

2. "Policy Statement on Severe Accidents," U.S. Nuclear Regulatory Commission, Federal Register, Vol.

50, 32138, August 8, 1985.

3. NUREG/CR-3568, A Handbook for Value-Impact

Assessment, PNL4646, dated December 1983.

4. NUREG/CR-5042, Evaluation of External Hazards

to Nuclear Power Plants in the United States,

LLNL, December 1987.

5. NUREG/CR-5042, Supplement 1, "Evaluation of

External Hazards to Nuclear Power Plants in the

United States, Seismic Hazards" LLNL, April 1988.

6. NUREG-1289, "Regulatory and Backfit Analysis:

Unresolved Safety Issue A-45, Shutdown Decay

Heat Removal Requirements, November 1988.

NUREG-1407 D-26
Appendix D Attachment

Table 3 Examples of Averted Risk from PRA Experience

Plant Description of Modifications Reference

Oconee Changes to turbine bldg., control NSAC PRA

room, turbine bldg. eq., and

procedures mods to reduce plant

vulnerability to internal floods

(CCW)

Yankee Establish risk basis for external Chapman

Rowe event requirement resolution

Tornado/high wind requirement

Seismic design changes

Indian 2 Mod. of structural design of IP2 PRA

control room

Millstone 3 Replace diesel generator oil MS3 PRA

cooler anchor bolts (seismic)

Conn Yankee App. R Mod.

Tornado/high wind mod.

Pt. Beach Add additional fire suppression

Table 4 Application of Specified System Modifications; Results of Value-Impact Analyses for Specific Plants

Averted

Base dp(CDF) Averted Dose Impact Onsite Value-Impact

Var. p(CDF) w Var. (person.rem) (Gross) Costs ($/person-rem)

Plant No. (per r-yr) (per r-yr) Offsite Net Onsite ($xE6) ($xE6) GrossNet

(1) (2) (3) (4) (5) (6) (7) (8) (9) (10)

Pt. Beach 3 3.13E-4 1.2E-5 36 15 0.99 0.26 2.8E4 1.4E4

lb 1.5E-5 45 18 0.24 0.33 5.3E3 =0

Turkey Pt. 3 2.36E-4 7.2E-5 535 81 3.10 2.33 5.8E3 l.3E3

2 1.3E-5 99 15 0.91 0.42 9.2E3 4.3E3

St. Lucie 3 7.44E-5 2.9E-5 100 37 0.60 1.05 1.05E4 =0

2 1.2E-5 42 15 0.052 0.44 1.2E3 = 0

ANO 1 lb/2 1.79E-4 6.4E-5 84 71 0.131 1.97 1.6E3 =0

Notes:

Column 2,

Column 6 = Column 7 =

Column 8 = Column 9 =

Column 10=

Modifications are as described in Section 3.

Averted Onsite Dose-Installation Dose.

Present Worth of Installation Costs + Operation and Maintenance Costs + Replacement Power Costs

During Installation + Cost of Limited-Scope PRA.

Present Worth of Replacement Power Costs + Loss of Investment + Cleanup Costs.

Col 7/Col 5.

(Col 7 - Col 8Y(Col 6 + Col 5)

D-27 NUREG-1407
Appendix D Attachment

Table 5 Modifications Based on Limited-Scope PRA, Results of Value-Impact Analyses for specific plants

Averted

p(CDF) dp(CDF) Averted Dose Impact Onsite Value-Impact Base w Var. (person-rem) (Gross) Costs ($/person-rem) Plant (per r-yr) (per r-yr) Offsite Net Onsite ($xE6) ($xE6) Gross Net

(1) (2) (3) (4) (5) (6) (7) (8) (9)

Pt. Beach 3.13E-4 2.7E-5 81 33 1.23 0.59 1.5E4 5.6E3

Turkey Pt. 2.36E-4 8.5E-5 634 96 4.0 2.75 6.323 1.703

St. Lucie 7.4E-5 4.1E-5 144 51 0.65 1.49 4.5E3 =0

ANO 1 1.79E-4 6.4E-5 84 71 0.131 1.97 1.55E3 =0

Quad Cities 1.97134 9.11E-5 2521 103 5.94 2.72 2.4E3 1.213

Cooper 4.37E-4 3.01E4 2295 278 24.3 6.58 1.1E4 6.9E3

C (alt.) 2.95E4 2241 271 3.19 6.42 1.423 =0

Notes:

Column 5 = Averted Onsite Dose - Installation Dose

Column 6 = Installation Costs + Operation and Maintenance Costs + Replacement Power Costs During Installation in 1985 Dollars Column 7 = Present Worth of Replacement Power Costs + Loss of Investment + Cleanup Costs

Column 8 = Col 6/Col 4

Column 9 =(Col 6 -Cot 7Y(Col 5 + Col 4)

NUREG-1407 D-28
18

APPENDIX 4

DOCUMENTATION

This appendix provides the guidelines for documentation and

reporting format and content for the IPEEE submittal. The major

parts of this appendix are the guidelines for seismic analysis

(Section 4.2), internal fire analysis (Section 4.3), other

analyses (Section 4.4). Licensees are requested to submit their

IPEEE reports using the standard table of contents given in Table

C.1 of NUREG-1407 or provide a cross reference. This will

facilitate review by the NRC and promote consistency among

various submittal. The contents of the elements of this table

are discussed further below.

The level of detail needed in the documentation should be

sufficient to enable the NRC to understand and determine the

validity of key input data and calculation models used, to assess

the sensitivity of the results to all key aspects of the

analysis, and to audit any calculation. All important

assumptions should be reported. It is not necessary to submit

all the documentation needed for such an NRC review. Relevant

documentation should be cited in the IPEEE submittal, and be

available in easily retrievable form. The guideline for judging

the adequacy of retained documentation is that independent expert

analysts should be able to reproduce any portion of the results

of the calculations in a straight forward, unambiguous manner.

To the extent possible, the retained documentation should be

organized along the lines identified in the areas of review. Any

information that is comparable to that provided under the IPE for

internal events can be incorporated by reference.

4.1 General

4.1.1 Conformance with Generic Letter and Supporting Material

Certification should be provided that an IPEEE has been completed

and documented as requested. The certification should also

identify the measures taken to ensure the technical adequacy of

the IPEEE and the validation of results.

4.1.2 General Methodology

An overview description of the methodology employed in the IPEEE

for each external event examined should be provided.

4.1.3 Information Assembly

Reporting guidelines include:

1. Plant layout and containment building information not

contained in the Final Safety Analysis Report (FSAR).
17

TABLE 3.2

REVIEW LEVEL EARTHQUAKE - WESTERN UNITED STATES PLANT SITES

0.5g*

Trojan

Washington Nuclear

Rancho Seco

Palo Verde

Seismic Margin Methods Do Not Apply To the Following Sites:

Diablo Canyon San Onofre

NOTES:

Indicates a Western United States site whose default bin is

O.5g unless the licensee can demonstrate that the site

hazard is similar to those sites east of the Rocky Mountains

that are found in the 0.3g bin.

Changes in the review level earthquake from 0.5g to 0.3g

should be approved prior to doing significant analysis.
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER

12-N) (Assigned by NRC, Add Vol.,

NRCM 1102, Supp., Rev., and Addendum Num3201. 3202 BIBLIOGRAPHIC DATA SH'ET *bars If any.)

(See Instructions on the reverse) NUREG-1407

2. TITLE AND SUBTITLE

3. DATE REPORT PUBLISHED

Procedural and Submittal Guidance for Individual Plant Examination of External MONTH I YEAR

Events (JPEEE) for Severe Accident Vulnerabilities June 1991

Final Report 4. FIN OR GRANT NUMBER

5. AUTHOR(S) 6. TYPE OF REPORT

J. T. Chen, N. C. Chokshi, R. M. Kenneally, G. B. Kelly, W. D. Beckner,

C. McCracken, A. J. Murphy, L Reiter, D. Jeng 7. PERIOD COVERED (Inclusive Dates)

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (If NRC, provide Division, Office or Region, U.S. Nuclear Regulatory Commission, and

mailing address; If contractor, provide name and mailing address.)

Division of Safety Issue Resolution

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (if NFIC, type 'Same as above"; If contractor, provide NRC Division, Office or Region,

U.S. Nuclear Regulatory Commission, and mailling address.)

Same as above

10. SUPPLEMENTARY NOTES

11. ABSTRACT (200 words or less)

Based on a Policy Statement on Severe Accidents, the licensee of each nuclear power plant is requested to perform

an individual plant examination. The plant examination systematically looks for vulnerabilities to severe accidents

and cost-effective safety improvements that reduce or eliminate the important vulnerabilities. This document presents guidance for performing and reporting the results of the individual plant examination of external events. The

guidance for reporting the results of the individual plant examination of internal events (IPE) is presented in

NUREG-1335.

12. KEY WORDS/DESCRIPTORS (List words or phrases that will assist researchers In locating the report.) 13. AVAILABILITY STATEMENT

Unlimited

Severe Accidents Policy Statement 14. SECURITY CLASSIFICATION

Individual Plant Examination of External Events (IPEEE) (This Page)

Vulnerabilities Unclassified

IPEEE Guidance (This Report)

Unclassified

15. NUMBER OF PAGES

16. PRICE

NRC FORM 336 (2-48)
THIS DOCUMENT WAS PRINTED USING RECYCLED PAPER

UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, D.C. 20555

OFFICIAL BUSINESS

PENALTY FOR PRIVATE USE, $30

IPOSTAGE & FEES PAID

1 USNRCj

PEMI o. - j

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