NUREG/CR-5382

MTR-90W00467

Screening of Generic Safety Issues

for License Renewal Considerations

Prepared by

Prepared by A. Faramarzi, A. A. Hughes, S. S. Seth

The MITRE Corporation

Prepared for

U.S. Nuclear Regulatory Commission
AVAILABILITY NOTICE

Availability of Reference Materials Cited in NRC Publications

Most documents cited in NRC publications will be available from one of the following sources:

1. The NRC Public Document Room, 2120 L Street, NW., Lower Level, Washington, DC 20555

2. The Superintendent of Documents, U.S. Government Printing Office, P.O. Box 37082, Washington,

DC 20013-7082

3. The National Technical Information Service, Springfield, VA 22161

Although the listing that follows represents the majority of documents cited In NRC publications, It Is not

Intended to be exhaustive.

Referenced documents available for inspection and copying for a fee from the NRC Public Document Room

Include NRC correspondence and internal NRC memoranda; NRC bulletins, circulars, Information notices,

Inspection and Investigation notices; licensee event reports; vendor reports and correspondence; Commission papers; and applicant and licensee documents and correspondence.

The following documents In the NUREG series are available for purchase from the GPO Sales Program:

formal NRC staff and contractor reports, NRC-sponsored conference proceedings, international agreement

reports, grant publications, and NRC booklets and brochures. Also available are regulatory guides, NRC

regulations In the Code of Federal Regulations, and Nuclear Regulatory Commission Issuances.

Documents available from the National Technical Information Service Include NUREG-series reports and

technical reports prepared by other Federal agencies and reports prepared by the Atomic Energy Commission, forerunner agency to the Nuclear Regulatory Commission.

Documents available from public and special technical libraries include all open literature items, such as

books, journal articles, and transactions. Federal Register notices, Federal and State legislation, and congressional reports can usually be obtained from these libraries.

Documents such as theses, dissertations, foreign reports and translations, and non-NRC conference proceedings are available for purchase from the organization sponsoring the publication cited.

Single copies of NRC draft reports are available free, to the extent of supply, upon written request to the

Office of Administration, Distribution and Mail Services Section, U.S. Nuclear Regulatory Commission,

Washington, DC 20555.

Copies of industry codes and standards used in a substantive manner In the NRC regulatory process are

maintained at the NRC Library, 7920 Norfolk Avenue, Bethesda, Maryland, for use by the public. Codes and

standards are usually copyrighted and may be purchased from the originating organization or, If they are

American National Standards, from the American National Standards Institute, 1430 Broadway, New York,

NY 10018.

DISCLAIMER NOTICE

This report was prepared as an account of work sponsored by an agency of the United States Government.

Neither the United States Government nor any agency thereof, or any of their employees, makes any warranty,

expressed or implied, or assumes any legal liability of responsibility for any third party's use, or the results of

such use, of any information, apparatus, product or process disclosed in this report, or represents that its use

by such third party would not infringe privately owned rights.

4
NUREG/CR-5382

MTR-90W00467

Screening of Generic Safety Issues

for License Renewal Considerations

Manuscript Completed: April 1991

Date Published: December 1991

Prepared by

A. Faramarzi, A. A. Hughes, S. S. Seth

The MITRE Corporation

7525 Colshire Drive

McLean, VA 22102

Prepared for

Division of Safety Issue Resolution

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

NRC FIN B8685

ABSTRACT

The U.S. Nuclear Regulatory Commission (NRC) is developing

regulations for renewing the operating licenses of nuclear power plants

to ensure that they operate safely beyond the present license terms of

40 years. One consideration relates to past resolutions of generic

safety issues (GSIs) that did not result in backfit requirements on the

licensees. The consideration of an additional operating term of 20 years

which the proposed license renewal rule allows, could have retrospective

implication for the basis of those GSI resolutions. As part of its technical support to the NRC for the development of license renewal regulations, MITRE has performed an independent review of the GSIs to identify

those that could be potentially affected by license renewal considerations. This report describes the screening process and the results of

that work.

iii

TABLE OF CONTENTS

Page

List of Figures

List of Tables

Executive Summary

1 Introduction

1.1 Statement of the Issue

1.2 Objectives

1.3 Background

1.3.1 Categories of Generic Issues

1.3.2 Resolved GSIs and Regulatory Analysis

1.3.3 Prioritization of GSIs

1.4 General Approach

1.5 Organization of Report

2 Method of Analysis

2.1 Phase 1 Screening

2.2 Phase 2 Screening

vi

vi

vii

1-2

1-2

1-2

1-3

1-4

1-8

1-8

2-1

2-1

2-3

3-1

3-1

3-1

3-3

3-3

3-8

3-8

4-1

A-I

B-I

3 Results

3.1 Phase 1 Screening

3.2 Phase 2 Screening

3.3 Non-Backfitted GSIs That

by License Renewal

3.3.1 Category 3b GSIs

3.3.2 Category 3a GSIs

3.3.3 Category LOW GSIs

Could Be Affected

4 Summary and Conclusion

Appendix A GSIs in Categories 3b and 3a with Revised

V/I Estimates Exceeding 500 Person-Rem/$M

Appendix B Description of GSI Database

Appendix C Projected Populations Within 50 Miles

Radius Around Nuclear Power Plants in

Years 2000 and 2010 C-I

References RE-i

v
LIST OF FIGURES

Page

1-1 NUREG-0933 Prioritization Criteria 1-7

3-1 Breakdown of Generic Issues 3-2

3-2 Availability of Regulatory Analysis and V/I

Prioritization Estimates for Non-Backfitted GSIs 3-4

3-3 Resolution Basis for GSIs That Did Not Result in

New Requirements (Category 3b Issues) 3-6

4-1 Effect of License Renewal on Resolved CSIs That

Were Not Backfitted 4-2

LIST OF TABLES

Page

ES-I GSIs Potentially Affected by License Renewal ix

1-1 NUREG-0933 Category Designation of Generic Issues 1-5

2-1 Categories of Items Deleted from the List of CSIs

in Phase-i Screening 2-2

4-1 GSIs Potentially Affected by License Renewal 4-3

vi
EXECUTIVE SUMMARY

10 CFR Part 54, "Requirements for Renewal of Operating Licenses for

Nuclear Power Plants", provides the U.S. Nuclear Regulatory Commission's

(NRC's) criteria and procedures for extending operating licenses by an

additional period up to 20 years. In the past, generic safety issues

(GSIs) were resolved without consideration of the 20 years of additional

plant operation permitted under the proposed license renewal rule. Consideration of extended plant life and any increase in population around

nuclear plant sites may alter the resolution bases of GSIs that have not

been backfitted. These license renewal considerations may influence the

resolution bases either directly, in cases where the remaining plant life

was explicitly *used in the resolution bases, or indirectly by affecting

the corresponding value/impact (V/I) estimates.

This report prepared by MITRE presents a systematic evaluation of

all GSIs resolved through October 1990 and identifies those whose resolution bases are potentially affected by the possibility of an additional

20 years of plant life permitted under the proposed license renewal rule;

projected increase in population around plant sites is considered as well.

MITRE evaluated a total of 249 GSIs that were resolved through

October 1990 and identified 139 GSIs that did not result in backfit

requirements. The GSIs not backfitted belong to one of the following

categories, depending on their resolution basis: GSIs that have been

resolved without resulting in any new requirements for either operating or

future nuclear power plants; GSIs that were resolved resulting in new

requirements only for future plants; and GSIs that were eliminated from

further consideration by NRC based on their limited safety significance as

determined in the priority ranking process and peer review.

In general, the resolution basis for GSIs consists of qualitative

considerations and quantitative values and impacts, all of which could be

affected by license renewal. All available quantitative V/I estimates

were updated to determine the effect on GSI resolutions of an additional

20-years under a renewed operating license and increase in population in

the vicinity of nuclear plant sites.

MITRE did not identify any instances where the remaining plant life

or the population density was the sole basis for resolving a GSI that did

not result in backfitting. In most cases the role of qualitative arguments in resolution basis outweighed the influence of quantitative V/I

estimates. In general, qualitative discussions in support of not backfitting included any one or a combination of the following factors: NRC

initiatives (Issuance of Policy Statements and NUREG documents, NRC

vii
organizational changes, etc.); industry initiatives; scope of GSI being

covered in other NRC programs; issue being of no concern after further NRC

staff investigation; and low risk or V/I estimate.

Based on a review of all resolved GSIs that were not backfitted,

MITRE identified three GSIs where V/I estimates played a relatively

significant role in their resolution and where the revised V/I estimates

were judged to be sufficiently close to or exceeding the 1,000 personrem/$M guideline that NRC has used in the past for resolving GSIs. In

addition, the priority ranking of six GSIs are potentially affected by

license renewal considerations; the revised V/I and averted public dose

estimates placed them in MEDIUM priority category. Issues prioritized as

MEDIUM or HIGH undergo further review and resolution by NRC. None of the

GSIs that resulted in new requirements only for future plants are affected

by license renewal considerations. GSIs that are potentially affected by

license renewal considerations are briefly discussed in Table ES-I.

viii
Table ES-I. GSIs Potentially Affected by License Renewal

GSI

Identi- Catefication gory Title Discussion

III.A.I.3(2) 3b Maintain SuppLies of

Thyroid BLocking Agent

I-"

82 3b Beyond Design Basis

Accidents in Fuel Pools

3b BWR Water Level Redundancy

The issue is concerned with stockpiles of KI agent for public use to help

prevent radiation injury to the thyroid gland by radionuclide releases in

the event of a nuctear power plant accident. A FEMA policy statement that

recommended against a nationwide requirement was approved by the NRC. The

value impact analysis in NUREG/CR-1433 showed that "K! offered an extremely

smaLL benefit in relation to its cost over the uncertainty range." The V/I

analysis is unique since it considers the averted public dose in units of

thyroid nodules and assumes a population density of 100 persons/mite 2 . The

V/1 estimate calculated by NRC is not affected by the remaining plant Life

and is only influenced by the population density aspect of License renewaL.

This issue is currently under review by the NRC staff (55 FR 39768), and

License renewal would be factored into any new resotution of this issue.

NUREG-1353 includes the value impact analysis for different options to

address damage in spent fuel pools in response to beyond design basis

accidents. Evaluation of instaLling a spray system resulted in V/i estimates of 330 and 833 person-rem/SN for the bgst-estimate and worst case

(Zion station population of 860 persons/mile ) scenarios, respectively.

Consideration of license renewal would increase these V/1 estimates to

approximately 500 and 1,400 person-rem/SN accordingly.

The issue is concerned with BUR plant response to a Level instrument line

break concurrent with a single failure. NUREG/CR-5112 includes the value

impact analysis for this issue. Based on low probability of core melt and

a maximum V/I estimate of 909 person-rem/SN, the resolution of this Issue

did not call for any additional actions. The staff concluded that aLl

BWR designs already provide acceptable protection, in conjunction with

operator training and procedures, to respond to an instrument Line break.

Consideration of license renewal increases the most Limiting V/I estimate

to -1,606 person-rem/SN.

101
Table ES-I. GSIs Potentially Affected by License Renewal (Concluded)

GSI

Identi- Catefication gory* Title Discussion

II.D.2 LOW Research on Relief and Consideration of license renewal increases V/i estimate and the total

Safety Valve Test averted public dose to 113 person-rem/SM and 2,432 person-rem,

respectively. This may place the issue in MEDIUM priority.

II1.D.2.1 LOW Radiological Monitoring Consideration of license renewal increases V/1 estimate and the total

(3 separate of Effluents' verted public dose to 66 person-rem/S$ and 7,714 person-rem, respecGISs under tively. This may place the issue in MEDIUM priority.

Parts 1 thru 3)

35 LOW Degradation of Internal Consideration of license renewal increases V/I estimate and the total

Appurtenances in LWRs averted public dose to 77 and 8,325 respectively. This may place the

issue in MEDIUM priority.

80 LOW Pipe Break Effects on Consideration of License renewal increases V/1 estimate and the

Control Rod Drive total averted public dose to 45 person-rem/SN and -100 person-rem,

Hydraulic Lines in the respectively. This may place the issue in MEDIUM priority.

Drywetts of BUR Mark I

and II Containments

*Consistent with the designation used by NRC (see Table 1-1)

N
SECTION 1

INTRODUCTION

The resolution of generic safety issues (GSIs) by the U.S. Nuclear

Regulatory Commission (NRC) involves the consideration of several factors,

one of which is the remaining operating life of nuclear power plants.

10 CFR Part 54, "Requirements for Renewal of Operating Licenses for

Nuclear Power Plants," provides the criteria and procedures for extending

their operating licenses by an additional period of up to 20 years.

Resolution of those GSIs that were not backfitted may be affected by the

consideration of extended plant life and the increased population density

around nuclear plant sites. These factors may alter the resolution bases

either directly, as may be explicitly stated in the resolution basis,

or indirectly by affecting the quantitative value/impact (V/I) estimates.

This report presents a systematic evaluation of all GSIs resolved

through October 1990 and identifies those whose resolution bases are

potentially affected by the introduction of an additional 20 years of

plant life permitted under the proposed license renewal rule and by the

projected increase in population around nuclear plant sites.

1.1 STATEMENT OF THE ISSUE

In situations in which the GSI involves a potentially worthwhile

safety enhancement, rather than involving a question as to adequate

safety, a comparison of the safety value of the improvement with the

cost impact of implementing it ordinarily has a role in the resolution

rationale. Both the safety value and the cost impact can increase with

added plant operating time. The safety value could increase over time

more than the cost impact, as would be the case when costs are largely

one-time initial costs but the risk reduction benefit accumulates year

after year with continued operation.

As part of its efforts towards developing license renewal regulations, which will permit extending operating licenses by a period up to

20 years, NRC decided to examine the resolved GSIs for possible cases

having the following characteristics:

" Backfitting of a new requirement within the original license term

was judged not to be worthwhile.

* Addition of a renewal term could increase the safety value without

commensurate increase in cost impact.

I-i
* The extent and other circumstances of license renewal on values

and impacts are such as to suggest the possibility that with the

increased operating time backfitting deserved consideration.

1.2 OBJECTIVES

This work was undertaken to investigate the potential impact of the

proposed license renewal rule on resolution of GSIs that have not been

backfitted. The specific objectives of this effort were to determine the

impact of an additional 20 years of plant life and increase in population

density on:

" Resolved GSIs that did not result in any new requirements;

* Resolved GSIs that resulted in only forwardfit (no backfit)

requirements; and,

" GSIs with LOW priority ranking that were eliminated from further

evaluation by the NRC staff based on their limited safety

significance.

1.3 BACKGROUND

1.3.1 Categories of Generic Issues

Generic issues include safety, environment, licensing, or other

concerns that relate to the design, construction, or operation of all,

several, or a class of nuclear power plants. Generic Safety Issues (GSIs)

are a subset of generic issues that generally are concerned with the

safety of plants and have the potential for leading to safety improvements

and promulgation of new or revised regulatory requirements or guidance.

Other specific categories of generic issues, including "Regulatory

Impact," "Licensing," and "Environmental" issues, do not have direct

safety significance.

"Unresolved safety issues" (USIs) are a class of GSIs that are of

considerable safety importance; their resolution is high on the NRC

staff's agenda. These issues are listed and described in NUREG-0606

(NRC, 1985a). All USIs have already been resolved and have resulted

in numerous safety enhancements to the design and operation of nuclear

plants.

1-2
Another group of GSIs is identified in NUREG-0371 and NUREG-0471,

"Task Action Plan Items" (NRC, 1978a; NRC, 1978b). These documents

provide NRC's initial consolidation and prioritization of generic

activities. In a descending order of importance, the issues were placed

into Categories A, B, C, and D.

Following the accident at Three Mile Island unit 2 (TMI), NRC

staff developed the. Action Plan, NUREG-0660 (NRC, 1980a), to provide a

comprehensive and integrated plan to improve safety of nuclear power

plants. Subsequent to a review by the Commission, specific items from

NUREG-0660 were approved and forwarded to the licensees for implementation. NUREG-0737 (NRC, 1980b) mandated the implementation of these

specific issues, known as "TMI Action Plan Items," and provided additional

information on schedules, applicability, method of implementation review,

submittal dates, and clarification of technical positions.

The NRC's formal process for prioritizing and resolving GSIs is

discussed in NUREG-0933, "A Prioritization of Generic Safety Issues"

(NRC, 1987). GSIs may be further classified according to their resolution

status--they may be: resolved; in the resolution process; or scheduled

for prioritization in the future.

1.3.2 Resolved GSIs and Regulatory Analysis

A large number of GSIs have been resolved over the years. Some

have resulted in new requirements imposed on nuclear plants. These new

requirements are either imposed on operating plants (backfit) or are

applicable only to future construction permit or operating license

applicants (forwardfit). The need to impose backfit or forwardfit

requirements is based on the importance of an issue to public health and

safety and its costs as determined by engineering and safety evaluations.

Since 1983, the staff has prepared a regulatory analysis for each

substantive regulatory action resulting from the resolution of GSIs.

Regulatory analysis is a process that provides a structure for the NRC to

identify alternative approaches for resolving a safety issue, to compare

the consequences of each identified alternative, and to select the best

alternative. The results of regulatory analysis provide a formal

statement for the selected regulatory action. The selection is normally

based on an evaluation of important attributes, including both qualitative

factors and quantitative values and impacts (NRC, 1984; PNL, 1983). The

remaining average plant life has been used to support such quantitative

evaluations.

1-3
1.3.3 Prioritization of GSIs

GSIs are prioritized for subsequent resolution in order to ensure

proper allocation of resources of the NRC staff in resolving GSIs in

accordance with their importance to safety. NUREG-0933 provides information on the resolution and prioritization status of all generic issues

with particular emphasis on GSIs. Table II of NUREG-0933 includes a

complete listing of all generic issues and their resolution status. An

excerpt from Table II providing information on the generic issue code

designation for different stages and methods of resolution is reproduced

in Table 1-1. The resolution status of GSIs is tracked through a

computerized database management system, the Generic Issue Management

Control System (GIMCS), which is updated quarterly (NRC, 1990a). GIMCS

supplements NUREG-0933, which is updated annually.

The resolution of GSIs that resulted in new requirements and those

that did not are designated "3a" and "3b", respectively, in NUREG-0933

Table II. A "LOW" priority designation is used to identify GSIs that

were determined not to require backfit based on their limited safety

significance reflected by their low priority ranking and peer review.

The prioritization process includes a preliminary evaluation of

values and impacts associated with a proposed solution (PNL, 1982). A

V/I analysis balances potential risk reductions against the costs to

achieve such reductions and is used to rank GSIs based on established

prioritization criteria. The NUREG-0933 prioritization criteria

(Figure 1-1) provide the basis for placing a GSI in one of the following

categories: HIGH, MEDIUM, LOW, or DROP. The calculated V/I estimates are

used as an aid in decision making, not as the sole basis for the ranking

results. In conjunction with quantitative V/I analysis, qualitative

factors are considered in establishing the prioritization category. In

general, safety, engineering judgement, and qualitative arguments have a

significant influence on the prioritization of GSIs.

GSIs in LOW and DROP priority categories were eliminated from further

evaluation by NRC since there was little or no prospect of safety improvements that were both substantial and worthwhile. However, consideration

of the proposed extended plant life and projected increase in population

may change the status of GSIs from LOW to MEDIUM priority ranking, which

may require further NRC staff consideration of such GSIs, and eventually

result in a 3a or 3b resolution status (see Table 1-1.)

1-4
Table 1-1. NUREG-0933 Category Designation of Generic Issues*

Table II

Listing of All TMI Action Plan Items, Task Action Plan Items,

New Generic Issues, and Human Factors Issues

This Table contains the priority designations for all issues

listed in this report. For those issues found to be covered

in other Issues described in this document, the appropriate

notations have been made in the Safety Priority Ranking column,

e.g., I.A.2.2 in the Safety Priority Ranking column means that

Item l.A.2.6(3) is covered in Item l.A.2.2. For those issues

found to be covered in programs not described in this document,

the notation (S) was made in the Safety Priority Ranking column.

For resolved issues that have resulted in new requirements for

operating plants, the appropriate multiplant licensing action

number is listed. The licensing action numbering system bears

no relationship to the numbering systems used for identifying

the prioritized issues. An explanation of the classification

and status of the issues is provided in the legend below.

Legend

NOTES: 1 - Possible Resolution Identified for Evaluation

2 - Resolution Available (Documented in NUREG, NRC Memorandum,

SER, or equivalent)

3 - Resolution Resulted in either: (a) The Establishment of New

Regulatory Requirements

(By Rule, SRP Change,

or equivalent)

or (b) No New Requirements

4 - Issue to be Prioritized in the Future

5 - Issue that is not a Generic Safety Issue but should be

Assigned Resources for Completion

HIGH - High Safety Priority

MEDIUM - Medium Safety Priority

LOW - Low Safety Priority

DROP - Issue Dropped as a Generic Issue

EI - Environmental Issue

*Source: Copied from NUREG-0933, Revision 11

1-5
Table 1-1. NUREG-0933 CateIory Designation of

Generic Issues (Concluded)

I

LI

MPA

NA

RI

S

USI

- Resolved IMI Action Plan Item with Implementation of

Resolution Mandated by NUREG-073798

- Licensing Issue

- Multiplant Action

- Not Applicable

- Regulatory Impact

- Issue Covered

- Unresolved Safety Issue

*Source: Copied from NUREG-0933, Revision 11

1-6
3.000

E

K0

0 10

is

V

a,

Z.5 10

Legend:

H -HIGH priority

M -MEDIUM priority

L -LOW priority

D -DROP

Man-Rem/Reactor

Man-Rem (Total. All Reactors)

Core-Melt/RY

Core-Melt/Yr.

lop 10' 10, 10,

5X10' 5xl0' 5X10' 5x10'

10`

5lO"' 5x10'

5x104 'a' 106

5x104

Change In Risk

Source: NUREG-0933, Revision 3

Figure 1-1. NUREG-0933 Prioritization Criteria

1-7
1.4 GENERAL APPROACH

The evaluation consisted of a two-phased approach to achieve the

objectives of this report. In Phase 1, the total population of all

generic issues was screened to identify the subset of GSIs in Categories

3b, 3a, and LOW that were resolved through October 1990.

In Phase 2, the basis for resolution of GSIs identified in Phase 1

was first reviewed to identify those GSIs that were not.backfitted. MITRE

constructed a PC-based database to summarize essential information on the

resolution of the GSIs that were not backfitted. The basis for resolution

generally consists of qualitative and quantitative arguments, which may

include consideration of the remaining average plant life and the population density around nuclear plant sites. Subsequently,-the additional

20 years of plant operation and increase in population were used to revise

available V/I estimates and enter them in the database. Finally, a review

of the qualitative and quantitative basis for resolving this set of GSIs

was performed to identify those that were potentially affected by license

renewal.

1.5 ORGANIZATION OF REPORT

Section 2 provides a more detailed overview of the evaluation and

the methodology used in performing it. Phases 1 and 2 of the screening

process are described. The two aspects of license renewal, i.e., additional 20 years of operation and increase in population, that could

potentially affect the resolution basis of GSIs are also discussed.

In Section 3, the results of the analysis are presented. The

screening results of Phases 1 and 2, and a description of the issues

potentially affected by license renewal are provided. Appendix A

provides a list of GSIs with revised V/I estimates exceeding a value

of 500 person-rem/$M irrespective of the resolution basis. Appendix B

provides the database of GSIs in Categories 3b, 3a, and LOW that were not

backfitted and summarizes pertinent information concerning the basis for

resolution of those GSIs. Appendix C provides the projected populations

around nuclear plant sites used to adjust the risk reduction estimates in

V/I analysis.

Section 4 presents a summary and conclusion of MITRE's evaluation of

resolved GSIs.

1-8
SECTION 2

METHOD OF ANALYSIS

The analysis and evaluation of resolved GSIs was based on a

two-phased approach. Phase 1 consisted of the screening of a total of

772 generic issues to identify those CSIs that were closed out before

1 November 1990 and that were in Categories 3b, 3a, and LOW. Explanation

of these categories has been provided in Table l-1.

In Phase 2, a detailed evaluation of GSIs that resulted from Phase 1

screening was performed to identify those GSIs that were not backfitted.

This reduced the list to all 3b issues, a limited number of 3a issues that

were not backfitted, and GSIs that were eliminated from further consideration by NRC because of LOW priority designation. These GSIs were

evaluated in detail to determine the impact on their resolution by the

additional twenty years of plant operation and increase in population

around nuclear plant sites.

2.1 PHASE 1 SCREENING

As stated above, this phase included an initial screening of all

generic issues to identify those that were resolved as of 1 November .1990.

NRC established this date at the inception of the project in recognition

of the continuing nature of the ongoing GSI resolution process and the

necessity to establish a cutoff date.

NUREG-0933 was the primary source of information for performing the

evaluation. Table II of NUREG-0933 provided a complete list of all

generic issues and their resolution status. The GIMCS database (NRC,

1990) provided the most up-to-date information on the current status of

GSIs in the resolution process. Other references were used as necessary

to clarify any ambiguities in the source documents.

The list of generic issues from NUREG-0933 was reviewed and reduced

in accordance with the list of categories of issues to be excluded shown

in Table 2-1. Deleted from further consideration at this stage were

non-safety related issues, GSIs scheduled for resolution after the cut-off

date of 1 November 1990, GSIs scheduled for future prioritization, GSIs in

DROP prioritization category, and GSIs identified in NUREG-0737 that were

mandated to be backfitted by the Commission.

2-1
Table 2-1. Categories of Items Deleted from the

List of GSIs in Phase-i Screening

Designation Description

LI

E

RI

Other GSIs

Note 4

DROP

HIGH, MEDIUM

I

Licensing issues

Environmental issues

Regulatory impact issues

GSIs that refer to other GSIs

GSIs to be prioritized in the future

GSIs in DROP Priority category

GSIs in HIGH and MEDIUM Priority categories

that were not resolved by 1 November 1990

GSIs mandated under NUREG-0737

2-2
It was decided by NRC that GSIs in DROP category would not be

considered in this evaluation since they could at best move to the LOW

priority ranking and would not require further evaluation by NRC. In

general, consideration of license renewal increases safety benefits and

V/I estimates by a factor of less than 2. This factor is representative

of a typical case where the original NRC's V/I evaluation considered a

remaining plant life of 28 years and the most limiting case where the

corresponding impact was a one-time implementation cost independent

of the remaining plant life. In order for a GSI in the DROP priority

category to shift up to the MEDIUM priority, both the safety benefit and

V/I estimates should increase by at least a factor of 10 (see Figure 1-1).

A preliminary review of selected DROP priority GSIs did not identify any

such cases.

At the conclusion of Phase 1 screening, the remaining GSIs fell into

one of the following categories:

" GSIs resolved without resulting in any new requirements (3b);

* GSIs resolved and resulting in new requirements (3a); and,

* GSIs with LOW priority status that were eliminated from further

NRC evaluation since their implementation would have limited

benefit to safety based on NRC staff judgement or V/I estimates.

2.2 PHASE 2 SCREENING

The resolution bases as described in NUREG-0933 and documents

referenced in it were reviewed in depth. Of particular importance were

the regulatory analyses underlying the decisions not to backfit. Other

references were used to clarify ambiguities in the source documents,

especially where no regulatory analysis was performed. In addition,

discussion with, and input from, several NRC staff members responsible

for individual issues helped MITRE understand and describe the resolution

basis for a number of CSIs resolved shortly before 1 November 1990 and for

which limited documentation was available. The resolution basis was

reviewed from the perspective of license renewal considerations, namely,

the additional 20 years of plant operation and projected increase in

nearby population.

In general, the resolution basis consists of two aspects: qualitative arguments and quantitative results, such as values, impacts, and

2-3
VI ratios. The importance of one or more factors may override the other,

depending on the issue of concern and the proposed resolution. MITRE

identified all available quantitative information so that the effects of

20 additional years of operation and increase in population could be

quantitatively factored into the evaluations of GSIs. Many issues that

were not backfitted did not have any regulatory analysis in the resolution

document. In such cases, the V/I values from NUREG-0933, if available,

were considered in the evaluation.

GSIs in Category 3a resulted in new backfit and/or forwardfit

requirements. A review of the resolution of these OSIs available in

NUREG-0933 was performed to identify those that were not backfitted. GSIs

that had been fully backfitted were eliminated from further review since

operating nuclear plants already have been required to make the necessary

plant modifications.

Given the rather large number of GSIs that were not backfitted and

that required detailed evaluation of their resolution basis, it was

decided to prepare a database to collect important information required to

determine the potential impacts of license renewal. MITRE selected a

PC-based system that would allow transfer of the information to the NRC

upon the conclusion of the project.

All pertinent information associated with GSIs in Categories 3b and

LOW and Category 3a items that were not backfitted were entered into the

database, so that all the information could be reviewed consistently.

All available V/I estimates from regulatory analyses or NUREG-0933 were

revised as necessary by incorporating an additional 20 years of plant life

and increase in population, and the revised V/I ratios were entered into

the database.

The methodology used to derive revised value-impact ratios includes

two separate calculations for risk reduction and costs to implement a

proposed resolution (NRC 1983a). In the V/I analyses revised, there were

no instances where occupational radiation exposures and accident avoidance

costs (e.g., cleanup and replacement power costs) were explicitly used in

the original calculation of V/I ratios.

The equation used to calculate risk reduction (W) for an operating

term that includes a 20-year license renewal is as follows:

2-4
WLR - Wo x PF x [(T+20)/T]

where

WLR is the total risk reduction over the new operating life, in

units of person-rem

Wo is the risk reduction over the original remaining life as

provided in original value-impact analyses, in units of

person-rem

PF is a dimensionless population adjustment factor representing

changes in the average population density considered in the

original calculation (see discussion below)

T is the average remaining life of reactors affected as used in

the original value-impact analysis, in units of years.

In resolving GSIs in the past, the NRC staff has used an average

population density (340 persons per square mile) around nuclear power

plants as projected for the year 2000 in its calculation of risk to the

public. This date represents the approximate midpoint for all nuclear

power plants under their original operating license term. Given the

potential for an additional 20 years of plant operation, as allowed in

the proposed license renewal rule, the midpoint of plant life would shift

to the year 2010. Consequently, the projected population density in the

year 2010 was used in recalculating V/I ratios for GSIs that were not

backfitted.

The revised V/I calculations in this report use a population

adjustment factor (PF) of 1.06 based on projected population data in

NRC's "Interim Draft for Generic Environmental Impact Statement for

License Renewal of Nuclear Power Plants" (NRC, 1990b). Information

from this report is summarized in Appendix C, which shows the projected

populations within the 50 mile radii of all nuclear power plants and the

cumulative totals for the years 2000 and 2010. The ratio of the total

projected population within the 50 mile radii of nuclear plants in the

year 2010 to that in the year 2000 is approximately 1.06. The number of

remaining years of plant life in the original analyses were generally more

than 20 years, so that the maximum correction factor, to account for both

a longer term and an increased population density, was approximately 2.

The second aspect of the V/I analysis is calculation of costs

associated with the proposed resolution. Costs incurred for implementing

safety improvements include: (1) the cost to NRC for developing each

requirement and reviewing the utility's design to assure that the

2-5
requirement is properly implemented, operated, and maintained; and

(2) the utility's cost of design, procurement, installation, and testing

to implement the requirement and its cost for operation and maintenance.

The equation used to calculate the total cost of implementing safety

improvements that includes the additional 20 years of operation under

license renewal is as follows:

SLR - So + 20 CO + 20 1o

where

so is the total impact under the original license term, in

units of $M

Co is the annual incremental NRC costs for reviews of safety

improvements, in units of $M

Io is the annual incremental industry costs for operation and

maintenance of safety improvements

The revised V/I estimates were simply the ratio of all values to

all impacts that were calculated considering license renewal. Typically,

since the incremental costs associated with the license renewal period

were often zero or minimal, the revised V/I ratio to include license

renewal was approximately a factor of 2 greater than the original

V/I ratio, not including a 20-year renewal term.

Appendix B is the GSI database including essential information

extracted from various sources, such as the original V/I estimates and

summary descriptions of resolution bases. The database also provides the

results of this review and evaluation, such as the revised V/I estimates

and determinations of any impact of license renewal considerations on the

resolution basis of GSIs.

2-6
SECTION 3

RESULTS

The methodology described in Section 2 was implemented to determine

the potential impact of the proposed license renewal rule on the resolution of GSIs that had not been backfitted. The results of analysis for

different categories of GSIs are discussed in this Section.

3.1 PHASE 1 SCREENING

A review of all generic issues was performed to identify GSIs that

were resolved as of 1 November 1990 and that were in Categories 3b, 3a, or

LOW. The screening basis summarized in Table 2-1 was used. At the conclusion of this phase 249 GSIs of a total of 772 generic issues remained

on the list. Figure 3-1 depicts the breakdown of generic issues according

to the various categories and highlights categories of GSIs that were

analyzed further in Phase 2.

3.2 PHASE 2 SCREENING

The initial stage of screening in this phase consisted of identifying

those GSIs, out of the 249 GSIs from Phase 1, that were not backfitted.

All of the 111 issues in Category 3b and all of the 22 GSIs in Category

LOW were not backfitted. The resolution basis of GSIs in Category 3a

were reviewed to identify those that were not backfitted. NUREG-0933 was

the primary source of information for determination of resolution status.

Five GSIs out of a total of 116 in Category 3a were not backfitted.

GSI 131, with an "S" resolution category (see Table 1-1) indicating that

the issue was covered under other NRC programs, was also included for

evaluation in Phase 2. Thus, a total of 139 GSIs out of a total of

249 GSIs were determined not to have been backfitted. These GSIs were

reviewed in detail to determine any potential revision in their resolution

basis by license renewal considerations.

As noted in the previous Section, the detailed review focused on the

qualitative resolution basis and, when available, on the quantitative V/I

information. This is summarized under the fields labeled "Resolution

Rationale," and "Revised V/I," respectively, in the GSI database.

The impact of a six percent increase in population density on

calculation of averted public dose is overshadowed by consideration of

the additional 20 years of plant operation that accounts for approximately

80 percent of the total increase in averted public dose.

3-1
GSIs Scheduled For

Resolution After 1 Nov. 1990

28 N.

Low Priority GSIs

22

Drop

Priority GSIs.

58

GSIs To Be

Prioritized

41

GS1s Resolved

With No New

A-- Requirements

(3b)

111

i.

Regulatory I I Environmental

Impact Issues Issues

18 15

Total Number Of Genericissues = 772

Figure 3-1. Breakdown of Generic Issues
3.3 NON-BACKFITTED GSIs THAT COULD BE AFFECTED BY LICENSE RENEWAL

A detailed review of all GSIs that were not backfitted did not

identify any instances where the remaining plant life or the population

density was the sole basis for determining that backfitting was not

required. Approximately forty percent of these GSIs had a quantitative

V/I analysis where the explicit consideration of extended remaining life

or population density could indirectly influence the resolution basis (see

Figure 3-2). However, in most instances, the role of qualitative factors

outweighed the consideration of V/I estimate or other quantitative factors

in determining that backfitting was not required.

Of the 116 GSIs in Categories 3b and 3a, MITRE identified only three

Category 3b GSIs where quantitative V/I estimates played a relatively

significant role in their resolution and where the revised V/I estimates

were judged to be sufficiently close to the reference guideline of 1,000

person-rem/$M that NRC has used in the past for resolving GSIs. Only six

GSIs in Category LOW are affected by license renewal. The discussion in

the following subsections is organized according to the three categories

of GSIs: 3b, 3a, and LOW.

Appendix A includes a list of resolved GSIs that were not backfitted

(Categories 3b and 3a) with revised V/I estimates exceeding the value of

500 person-rem/$M. MITRE selected this value to highlight those GSIs with

V/I estimates close to the reference value of 1,000 person-rem/$M for

which qualitative factors considered in their resolution not to backfit

played a dominant role. The resolution basis for these GSIs is provided

in Appendix B.

3.3.1 Category 3b GSIs

Of the 11 issues in Category 3b, only 10 issues had a regulatory

analysis and another 30 issues had a V/I analysis performed in support

of their prioritization. The remaining 71 issues were resolved based on

qualitative factors and engineering judgement. Three of 10 regulatory

analyses did not contain any quantified V/I estimates.

GSIs in Category 3b were resolved on the basis of a number of

different reasons. However, during the course of the evaluation, it

became apparent that the resolution bases could be grouped. These groups

are listed below and their letter designations are used in the database to

indicate how each GSI was resolved.

3-3
80

60

Number

of GSIs 40

20

0

Key

WGSIs with no V/Is

•1 GSIs with NUREG-0933

V/is

i GSIs with Regulatory

Analysis

L.J

3b 3a

GSI Category

Low

Figure 3-2. Availability of Regulatory Analysis

and VII Prioritization Estimates for Non-Backfitted GSIs
A GSIs resolved based on NRC initiatives including the following:

issuance of policy statements (Al); actions taken to address

other concerns (A2); issuance of documents, including NUREGs;

computer code manual upgrades, etc., (A3); administrative and

organizational changes (A4); and GSIs to be revisited by the

staff in the future (A5).

B Further investigation of the issue by NRC determined that there

was no concern.

C The issue was covered in other NRC programs (e.g., Individual

Plant Examination (IPE), Nuclear Plant Aging Research. (NPAR),

etc.).

D Industry initiatives and actions mitigated or eliminated the

concern.

E Low risk and negligible V/I values.

Figure 3-3 depicts a breakdown of the resolution bases for

Category 3b issues consistent with the grouping described above.

A review of the resolution bases of GSIs in Category 3b identified

only three GSIs that are potentially affected by license renewal

considerations. These are discussed below.

GSI III.A.l.3(2): Maintain Supplies of Thyroid Blocking Agent for the.

Public

Federal Emergency Management Administration (FEMA) developed a policy

statement against a nationwide requirement for the distribution or stockpiling of KI for use by the general public and left the final decision for

its use to state and local authorities on a site-specific basis. The NRC

concurred with the policy statement that was based on a V/I estimate.

Units for the cost/benefit analysis were $/thyroid nodule averted and

ranged from $300,000 to $40 million/nodule, depending on distance from the

plant. This was compared to an average cost of treating a thyroid nodule

case of $17,000. The analysis was performed independent of plant lifetime

and the number of plants operating in the U.S., although the analysis

acknowledged that multiple reactors at sites would decrease the cost/

benefit ratios correspondingly. Also, the analysis assumed a uniform

population distribution of 100 persons/mile 2 rather than the standard

340 persons/mile 2 used in V/I analysis of potential GSI resolutions.

3-5
40-

KEY

NRC Initiatives:

A1 Policy Statement

A2 Addressed by NRC

30- Actions Taken In

Response to Other

Concerns

A3 NRC Initiative

(NUREGs, GLs, etc.

Number A4 NRC Administrative/

of GSis 2Organizational

Changes

A5 Future NRC Initiatives

B No Concerns

10 ,J C Covered In Other NRC

Programs

E D Industry Initiative

E Low Risk or V/I

0 I

Al A2 A3 A4 A5 B C D E

Resolution Basis

Figure 3-3. Resolution Basis for GSIs

That Did Not Result in New Requirements (Category 3b Issues)
Cost and benefit results are provided on an annual, single reactor incidence basis. Thus the results would remain constant regardless of renewal

term, except for any changes in the population density. However, given

that the unit of analysis and population density assumed are different for

this issue, MITRE did not revise the results of the V/I analysis. The NRC

plans to perform an updated V/I analysis to consider the implications of

latest available research and new information on estimated iodine releases

from severe reactor accidents on the values and impacts of alternative

approaches to address the issue (55 FR 29768).

GSI 82: Beyond Design Basis Accidents in Fuel Pools

Several alternatives to address damage in spent fuel pools in

response to beyond design basis accidents were evaluated in the regulatory

analysis (NRC, 1989). Alternative 4, installation of a spray system,

resulted in V/I estimates of approximately 330 and 833 person-rem/$M for

the best-estimate and worst case scenarios, respectively. Because of

license renewal considerations, the best-estimate and worst case V/I

estimates could increase to approximately 500 and 1,400 person-rem/$M,

respectively.

It should be noted that the original V/I analysis has considerable

uncertainties. Furthermore, the worst case estimate applies to the Zion

Nuclear Station with its specific meteorology and population density of

860 people per square mile.

GSI 101: BWR Water Level Redundancy

This GSI is concerned with the ability of BWRs to mitigate water

level instrument sensing line leaks or breaks that could affect both

control and protection systems.

Based on the low probability of core melt (<2E-6) and the highest

V/I estimate of 909 person-rem/$M to reduce public risk, the resolution

of GSI 101 did not call for any additional actions. The NRC concluded

that all BWR designs already provide acceptable protection, in conjunction

with operator training and procedures, to mitigate a line break in any of

the water level instruments.

License renewal consideration increases the most-limiting V/I

estimate in the regulatory analysis to over 1,600 person-rem/$M for one

group of plants with a particular proposed modification to the Automatic

Depressurization System.

3-7
3.3.2 Category 3a GSIs

Of the total of 116 GSIs that resulted in new requirements on the

licensees, only five issues were forwardfitted without any backfit

requirements. These issues are identified and summarized in the GSI

database in Appendix B. These five GSIs were identified in the mid to

late 1970's and were resolved based on a variety of reasons, including

initiatives by the industry. There is limited quantitative information

available in support of their resolution. Only two GSIs in this category

had a quantitative regulatory analysis. Consideration of license renewal

factors would increase the V/I estimates for these two Category 3a issues

beyond the NRC's guideline of 1,000 person-rem/$M. However, other factors

considered in the resolution override the V/I estimates, so that these

OSIs are not affected by license renewal.

3.3.3 Category LOW GSIs

Of the 22 GSIs in Category LOW, 14 GSIs had a V/I estimate. The

remaining issues were placed in Category LOW based on qualitative reasons

and engineering judgement. In addition, GSI-66, which is in Category 3b,

includes four sub-issues that were eliminated from further evaluation

based on their DROP priority categorization on a ranking scale that

combined LOW and DROP priority ranking. MITRE reviewed these Category

"DROP" issues since they could move up to MEDIUM priority ranking, and

determined that they were not affected by license renewal.

Based on V/I estimates and averted public dose, as well as

qualitative information supporting the resolutions, MITRE identified the

following six Category LOW GSIs that could move up to MEDIUM priority.

GSI II.D.2: Research on Relief and Safety Valve Test Requirements

The revised V/I estimate of 113 person-rem/$M in conjunction with the

revised total averted public dose of 2,432 person-rem could place the

issue in the MEDIUM priority category.

GSI III.D.2.1: Radiological Monitoring of Effluents

This issue consists of three distinct GSIs (III.D.2.1.1 through

III.D.2.1.3) that were evaluated together. The revised V/I value of

66 person-rem/$M and revised averted total public dose of 7,714 person-rem

could place the issue in the MEDIUM priority category.

3-8
GSI 35: Degradation of Internal Appurtenances in LWRs

Originally the issue was placed in the LOW priority category since

all scores, except for the core damage frequency, which was marginally

in the Category MEDIUM, placed the issue in Category LOW. The revised

V/I of 77 person-rem/$M and the revised total averted public dose of

8,325 person-rem could move the issue to the MEDIUM priority rank.

GSI 80: Pipe Break Effects on CRD Hydraulic Lines in the

Drywells of BWRs Mark I and II Containments

The issue was originally assigned LOW priority ranking because of

V/I estimates. The revised V/I of 45 person-rem/$M and the revised total

averted public dose of approximately 100 person-rem could move the issue

to the MEDIUM priority rank.

3-9

SECTION 4

SUMMARY AND CONCLUSION

MITRE conducted a review and evaluation of resolved GSIs that have

not been backfitted to determine the potential impact of an additional

20 years of plant operation permitted under the proposed license renewal

rule, and the projected increase in population density.

The review was based on a systematic evaluation of 772 generic

issues, in a two-phased approach, to identify those GSIs that were

resolved through October 1990 and that did not result in backfitting.

As a result of Phase 1 screening, 249 in Categories 3b, 3a, and LOW were

identified to have been resolved through October 1990. A further review

of this set of GSIs in Phase 2 identified 139 issues that did not result

in backfit requirements. MITRE constructed a database to capture essential information associated with these 139 GSI, including the following:

resolution basis and references, original V/I estimate, and revised V/I

estimate. The database will provide NRC with the capability to perform

future analyses and monitoring of non-backfitted GSIs.

There were no instances where the remaining plant life or the

population density was the sole basis for resolving a GSI without

imposing backfit requirements. However, a limited number of GSIs that

have a V/I estimate could be affected by license renewal considerations

(see Figure 4-1). The GSIs that are potentially affected by license

renewal considerations are briefly discussed in Table 4-1.

GSIs in Category 3b are those that have been resolved without

resulting in any new requirements for either operating or future nuclear

power plants. Of the 111 GSIs in this category, only three GSIs could be

affected by license renewal considerations.

GSIs in Category 3a are those that have been resolved with new

requirements for either operating and/or future plants. Of a total of

116 GSIs in this category, five were not backfitted. However, none of

these GSIs was affected by license renewal considerations. The V/I

estimates for two CSIs in this category exceeded 1,000 person-rem/$M.

However, the qualitative resolution basis outweighed the quantitative

revised V/I estimates (see Appendix B).

GSIs in Category LOW are those that have been eliminated from further

evaluation by NRC based on their limited safety significance as determined

in the priority ranking process and peer review. Of the 22 GSIs in this

category, only 14 had V/I estimates. These were revised to account for

4-1
120-(

100*

80-

Key

GSIs Potentially

Affected by LR

WGSI Not Affected

by LR Number

of GSIs 60

40-

20-

0 9ý 

3b

-

3a Low

LU

GSI Category

Figure 4-1. Effect of License Renewal

on Resolved GSIs That Were Not Backfitted
Table 4-1. GSIs Potentially Affected by License Renewal

GSI

Identi- Catsfication gory Title Discussion

III.A.1.3(2) 3b Maintain Supplies of

Thyroid Blocking Agent

!I

82 3b Beyond Design Basis

Accidents in Fuel Pools

3b BWR Water Level Redundancy

The issue is concerned with stockpiles of KI agent for public use to help

prevent radiation injury to the thyroid gland by radionuclide releases in

the event of a nuclear power plant accident. A FEMA policy statement that

recommended against a nationwide requirement was approved by the NRC. The

value impact analysis in NUREG/CR-1433 showed that "KI offered an extremely

small benefit in relation to its cost over the uncertainty range." The V/I

analysis is unique since it considers the averted public dose in units of

thyroid nodules and assumes a population density of 100 persons/mile2. The

V/I estimate calculated by NRC is not affected by the remaining plant life

and is only influenced by the population density aspect of license renewal.

This issue is currently under review by the NRC staff (55 FR 39768), and

license renewal would be factored into any new resolution of this issue.

NUREG-1353 includes the value impact analysis for different options to

address damage in spent fuel pools in response to beyond design basis

accidents. Evaluation of installing a spray system resulted in V/I estimates of 330 and 833 person-rem/SM for the best-estimate and worst case

(Zion station population of 860 persons/mile2) scenarios, respectively.

Consideration of license renewal would increase these V/I estimates to

approximately 500 and 1,400 person-rem/$M accordingly.

The issue is concerned with BWR plant response to a level instrument line

break concurrent with a single failure. NUREG/CR-5112 includes the value

impact analysis for this issue. Based on low probability of core melt and

a maximum V/I estimate of 909 person-rem/$M, the resolution of this issue

did not call for any additional actions. The staff concluded that all

BWR designs already provide acceptable protection, in conjunction with

operator training and procedures, to respond to an instrument line break.

Consideration of license renewal increases the most limiting V/I estimate

to -1,606 person-rem/SM.

101
Table 4-1. GSIs Potentially Affected by License Renewal (Concluded)

GSI

Identi- Catefication gory Title Discussion

II.D.2 LOW Research on Relief and Consideration of license renewal increases V/I estimate and the total

Safety Valve Test averted public dose to 113 person-rem/$M and 2,432 person-rem,

respectively. This may place the issue in MEDIUM priority.

III.D.2.1 LOW Radiological Monitoring Consideration of license renewal increases V/I estimate and the total

(3 separate of Effluents' verted public dose to 66 person-rem/$M and 7,714 person-rem, respecGISs under tively. This may place the issue in MEDIUM priority.

Parts 1 thru 3)

35 LOW Degradation of Internal Consideration of license renewal increases V/I estimate and the total

Appurtenances in LWRs averted public dose to 77 and 8,325 respectively. This may place the

issue in MEDIUM priority.

80 LOW Pipe Break Effects on Consideration of license renewal increases V/I estimate and the

Control Rod Drive total averted public dose to 45 person-rem/$M and -100 person-rem,

Hydraulic Lines in the respectively. This may place the issue in MEDIUM priority.

Drywells of BWR Mark I

and II Containments

Consistent with the designation used by NRC (see Table 1-1)

41
the additional plant life and population density increase. Five of the

14 GSIs in Category LOW are potentially affected by license renewal

considerations and may shift to MEDIUM priority status.

4-5

APPENDIX A

GSIs IN CATEGORIES 3b AND 3a WITH REVISED V/I ESTIMATES

EXCEEDING 500 PERSON-REM/$M

GSI

Identi- Category V/I Revised V/I

fication Title (NUREG-0933) Reference Person-rem/SM

I.C.9

I.D.4

II.C. 1

II.C.2

II.C.4

II.E.2.2

III.A.1.3(2)

Long-term Program Plan for

Upgrading of Procedures

Control Room Design Standard

Interim Reliability

Evaluation Program (IREP)

Continuation of Interim

Reliability Evaluation

Program (NREP)

Reliability Engineering

Research on Small Break

LOCAs and Anomalus Transients

Public (Maintain supplies

of thyroid blocking agent)

III.A.3.4 Nuclear Data Link

III.D.2.5 Offaite Dose Calculation

Manual

III.D.3.1 Radiation Protection Plans

(Worker Radiation Protection

Improvement)

IV.E..5 Assess currently Operating

Plants (Safety

Decision-Making)

3b

3b

3b

3b

3b

3b

3b

3b

3b

3b

3b

3a

3a

3b

3b

NUREG-0933

NUREG-0933

NUREG-0933

NUREG-0933

NUREG-0933

NUREG-0933

NUREG/CR1443

NUREG-0933

NUREG-0933

NUREG-0933

NUREG-0933

522

1,022

977

983

528

1,750

See Note 1

3,530

1,211

3,439

2,137

A-12"

A-43

Fracture Toughness of Steam

Generator and Reactor

Coolant Pump Supports (USI)

Containment Emergency Sump

Performance (Former USI)

Shutdown Decay Heat Removal

Requirements (Former USI)

Buckling Behavior of Steel

Containments

NUREG-0577 Ranges

(Rev.1) between

153 and

-60,000

NUREG-0869 Ranges

between

4,652

and

13,533

NUREG- 1289 Varies

widely

A-45

B-5 NUREG-0933 782

A-1
APPENDIX A

GSIs IN CATEGORIES 3b AND 3a WITH REVISED V/I ESTIMATES

EXCEEDING 500 PERSON-REM/$M (Concluded)

GSI

Identi- Category V/I Revised V/I

fication Title (NUREG-0933) Reference Person-rem/SM

82

101

115

Beyond Design Basis

Accidents in Fuel Pools

BWR Water Level Redundancy

Reliability of Westinghouse

Solid State Plant Protection

System

Initiating Feed and Bleed

SPDS Availability

(Davis-Besse Loss of All

Feedwater Event of June 9,

1985, Long-Term Actions)

3b

3b

3b

3b

3b

NUREG-1353

NUREG/CR5112

NUREG-1341

NUREG-0933

NUREG-0933

-500

1,606

1,956

4,244

982

122.2

125.1.3

NOTE 1: The cost-benefit ratio in the regulatory analysis of $300,000 to $40M per thyroid

nodule is independent of the remaining plant life and assumes a population density

of 100 persons/mile2. The issue is currently under review by NRC and license

renewal would be factored into any new resolution of the issue.

A-2
APPENDIX B

DESCRIPTION OF GSI DATABASE

The following is a description of the different data fields in the GSI

database printout. "LR" is used to abbreviate license renewal in the

database.

GSI IDENTIFICATION: Identification number of the Generic Safety Issue

(GSI), consistent with NUREG-0933.

TITLE: Title of the GSI as it appears in NUREG-0933 and its Table II.

Additional text providing supplemental information is included in

parentheses.

ISSUE TYPE: GSI type consistent with NUREG-0933 as being one of the

following:

TMIAP TMI Action Plan

TAP Task Action Plan

NGI New Generic Issues

HF Human Factors

CATEGORY: Designation of the issue consistent with NUREG-0933, Table II,

and GMICS (10 August 1990) information (i.e., 3a, 3b, and LOW)

DECISION RATIONALE: A brief description of the resolution basis. In

general, information is obtained from NUREG-0933 unless other references

as designated in NUREG-0933 are specifically identified. Any relevant

risk information such as core damage frequency, total and per plant

averted dose, etc., is also included. Revised estimates of these

parameters based on extended plant life and increased population are also

included in a separate paragraph, where necessary.

V/I REFERENCE: References providing value/impact (V/I) information. A

reference to NUREG-0933 indicates that only a V/I ratio used in the

prioritization was available. Other references are indicative of the

presence of a more detailed analysis (e.g., Regulatory Analysis). An "N"

in this column indicates that no references that provided V/I estimates

were identified.

B-1
OLD V/I (P-rem/$M): V/I estimate, if available. Otherwise, an "N/A",

indicating Not Available, appears in the column.

REV V/I (P-rem/$M): Revised V/I estimate including the impact of License

Renewal (i.e., additional 20 years and population density increase). In

general consideration of licensing renewal typically increases V/I ratios

by at most a factor of 1.7 assuming the original average remaining plant

life of approximately 28 years. All available V/I estimates have been

revised to include license renewal factors.

RES. BASIS: The basis for the resolution of 3b items. Any one or a

combination of the following codes are used as the basis for the

resolution of the issue:

" Al Resolved with the issuance of a Policy Statement

* A2 Is resolved because of the NRC action taken in response to

other concerns and not the GSI

* A3 The issue is resolved and resulted in an NRC action including:

- Tech Specs changes

- Issuing NUREGs/BTP/Clarified SRPs

- Changes to inspection programs

- Computer code development

* A4 The issue is resolved and resulted in interval administrative

or organization changes

* A5 Resolved requiring future assessments

* B Resolved since there was no problem based on further review

* C Covered or subsumed in other NRC programs (e.g., IPE, NPAR)

" D Resolved based on industry initiatives

* E Resolved based on values and/or impacts

RES. DATE: The year when the issue was resolved as specified in GIMCS.

Resolution dates not identified in GIMCS are designated as "<83". "N/A"

designation for resolution date of LOW category GSIs indicates the date as

not being available.

B-2
GSI TITLE ISSUE CATE DECISION RATIONALE V/I OLD V/I REV. V/I RES. RES.

IDENTIFICATION TYPE GORY REFERENCE P-rem/$N P-rem/SN BASIS DATE

I.A.2.2 Training and Qualification of

Operations Personnel

I.A.2.5 Plant Drills

odJ

11

THIAP 3b Ref 777: The Commission formally

recognized the industry progress to

improve training & qualifications. In

1985, issued a Policy Statement on

the issue "..endorsing the

INPO-managed training Accreditation

Program .... "

TMIAP 3b Ongoing work by DHFS was completed in

1985 and published for information as

a NUREG. The short-term objective of

manipulating controls was acheived in

NUREG-0737. The tong-term goat of

developing NRC standards for drilts

was Left to INPO & the industry. The

issue was given LOW priority since

V/1 analysis overestimated the risk.

Total averted public dose of 4,805

P-rem increases to 8,562 by LR.

TMIAP 3b Resolved under item I.A.2.6(1), which

resulted in revised RG 1.8.

NUREG-0933 reports V/I of 60 for Item

I.A.2.6(1) which is in 3a category.

TMIAP 3b Ref 804: Both methods of conducting

workshop & survey test cases proved

to be effective means of getting

feedback. Because of Low V/I value

and industry willingness to provide

feedback during the study, the issue

should be resolved.

NUREG-0933 62 67 A3,D 483

NUREG-0933 177 193 A1,D 85

I.A.2.6(2) Staff Review of NRR 80-117 N N/A N/A A2 <43

I.A.2.6(4) Operator Workshops NUREG-0933 165 177 D,E 85

Total averted public dose of 7,140

P-rem increases to 9,930 because of LR.

I.A.2.6(5) Develop Inspection

Procedures for Training

Program

TMIAP 3b same as Item I.A.2.6(2) N N/A N/A A2 <83
GSI TITLE ISSUE CATE DECISION RATIONALE V/I OLD V/I REV. V/I RES. RES.

IDENTIFICATION TYPE GORY REFERENCE P-rem/$M P-rem/$M BASIS DATE

I

I.A.2.7 Accreditation of Training

Institutions

I.A.3.2 Operator Licensing Program

Changes

I.A.3.3 Requirements for Operator

Fitness

I.A.3.4 Licensing of Additional

Operations Personnel

I.A.4.1(1) Short-term Study of Training

Simulators

THIAP 3b Ref 777: The commission formally

recognized the industry progress to

improve training & qualifications. In

1985, issued a Policy Statement on

the issue "..endorsing the

INPO-managed training Accreditation

Program.. 11.

TNIAP 3b The following actions were taken: 1-

all Regions were given operator

Licensing authority, 2- a study of

op. Licensing program was documented

in NUREG-1021, and 3- reporting of

operator error was deleted.

TNIAP 3b Ref 956: A regulatory approach to

fitness for duty was originally

pursued. In 1986, a Commission Policy

Statement recognized the industry

efforts and resolved the issue.

THIAP 3b Ref 778: There was insufficient

evidence to support new requirements.

In accordance with current Commission

statement, related to NUNARC

initiatives, there would be no

proposed rule for 2 years. Also the

Policy Statement on training and

qualification will cover inspection

of utilities' training program.

TNIAP 3b As required by TMI Action Plan, a

study of training simulators was

conducted & documented in

NUREG/CR-1482. Issue I.A.4.1(2)

required implementation of corrective

actions identified in Part 1, which

resulted in issuing RG-1.149.

NUREG-0933 64.4 71 A1,D 85

N N/A N/A A3,B <83

NUREG-0933 13.4 14.3 A1,D 87

NUREG-0933 20 23 Al 85

N N/A N/A A3 <3
GSI TITLE ISSUE CATE DECISION RATIONALE V/I OLD V/I REV. V/I RES. RES.

IDENTIFICATION TYPE GORY REFERENCE P-rem/SN P-rem/SM BASIS DATE

IJ

I.B.1.1(I) Prepare Draft Criteria

(Organization & Management

Long-term Improvement)

1.8.1.1(2) Prepare Commission Paper

I.B.1.1(3) Issue Requirements for the

Upgrading of Management and

Technical Issues

1.9.1.1(4) Review Responses to

Determine Acceptability

I.B.1.1(5) Review ImpLementation of

Upgrading Activities

I.B.1.2(1) Prepare Draft Criteria

(Evaluation of Org. & Manag.

Involvements of Near-term

Operating License AppLicants)

I.B.1.2(2) Review Near-term Operating

License Facilities

1.B.1.2(3) Include Findings in the SER

for each Near-term Operating

TMIAP 3b Ref 956: Items 1-4 of I.B.1.1 are

terminated based upon industry

initiatives through NUMARC and the

Commisssion's Policy Guidelines. This

is in tune with the Commisssion

policy: "..NRC is moving toward

performance-based rather than

prescriptive regulation.." NUREG-0933

V/I for aLL 7 parts of I.B.1.1 is

-760.

TMIAP 3b Same as I.B.1.1(1).

TMIAP 3b Same as 1.B.1.1(1).

TMIAP 3b Same as 1.B.1.1(1).

TMIAP 3b Same as I.B.1.1(1). Also Ref 956

indicates that IE routinely develops

inspection procedures which address

new or revised requirements.

TMIAP 3b The issue was to be applicable to

NTOLs and findings were incorporated

into corresponding SERs. All 3 parts

of issue I.B.1.2 were resolved.

TMIAP 3b Same as I.B.1.2(1).

N N/A N/A A1,D 87

N N/A N/A A1,D 87

N N/A N/A A1,D 87

N N/A N/A A1,D 87

N N/A N/A A1,D 87

N N/A N/A A3 <Z3

N N/A N/A A3 <3

TMIAP 3b Same as I.B.1.2(1). N N/A N/A A3 <83
GSI TITLE ISSUE CATE DECISION RATIONALE V/1 OLD V/I REV. V/I RES. RES.

IDENTIFICATION TYPE GORY REFERENCE P-rem/SN P-rem/SN BASIS DATE

License FaciLity

I.C.1(4) Confirmatory Analyses of

Selected Transients

(Operating Procedures)

TMIAP 3b Ref 383: Confirmatory analysis of

selected vendor transient

calculations resulted in: 1-

confirmed catcs, 2- confirmed caics. &

required changes to guidelines, 3-

NRC methods need development & 4-

need for additional experimental

data. (NUREG-0933 reports a V/I of

1,650 person-rem/SM for all 4 parts

of I.C.1.)

N N/A N/A A3 <83

I.C.9

0I

Long-term Program Plan for TMIAP

Upgrading of Procedures

Control Room Design Standard TNIAP

3b Ref 955: That part of the item

dealing with EOPs has been

implemented under I.C.1 mandated

under SuppLement 1 to NUREG-0737. The

scope of this issue is limited to

Long-term upgrade of abnormal and

operating procedures. Issue HF-02,

Maintenance & Surveillance Program,

includes addressing maintenance

procedures. The remaining concerns

are covered in Human Factors Program,

Issue HF-01.

3b Ref 1101: The objective was to

establish guidelines thru IEEE and

RGs. Issue I.D.1 requirements as

documented in NUREG-0700 (referenced

in SRP 18.1) covers the scope of

Issue I.D.4.

NUREG-0933 461

NUREG-0933 604

522 C 86

I.D.4 1,022 C 88

1.0.5(1) Operator Process

Communications (Improved CR

Instrumentation Research)

TMIAP 3b The objective was to evaluate

man-machine interface in CR. RES

research was documented in

NUREG/CR-2147 and a research

information letter (Ref 245) made

several recommendations for the

future, including assessment of

N N/A NA A3 <83
GSI TITLE ISSUE CATE DECISION RATIONALE V/I OLD V/I REV. V/I RES. RES.

IDENTIFICATION TYPE GORY REFERENCE P-rem/SM P-rem/$S BASIS DATE

NUREG-0700. EPRI/INPO/NRR/DOD

cooperation, & modification of risk

analysis programs.

1.0.504)

I.F.1

Process Monitoring

Instrumentation

Expand QA List

t.1

THIAP 3b A directly related issue, Item II.F.2

in NUREG-0737, mandated that industry

develop & implement PWR liquid Level

detection systems. Also NRC evaluated

a number of systems at LOCA

experiment facilities.

THIAP 3b Ref 1181: The Commission disapproved

the proposed rule. It was determined

that SERs approved statements made in

SARs. Important to safety (ITS)

components that were also safety

related (S/R) were identified and not

ITS that were non S/R. Operating

plants were identifying ITS/NSR

consistent with their licensing

bases. The cost to Licensees was

estimated to be $22.25M.

TMIAP LOW QA In nuclear plants was an issue of

high priority. However, it was felt

that resolution of QA deficiency, as

described under issue I.F.2 (except

for parts 2, 3, 6, and 9 that imposed

new requirements), failed to address

the problem of management acceptance

of QA programs. Parts 1, 4, 5, 7, 8,

10 and 11 of issue I.F.2 were given

LOW priority.

N N/A N/A C 83

N N/A N/A B 89

I.F.2(1) Assure the Independence of N N/A N/A N/A N/A

the Organization Performing

the Checking Function

I.F.2(4) Establish Criteria for

Determining GA Requirements

I.F.2(5) Establish Qualification

Requirements for GA and OC

TMIAP LOW Same as Issue I.F.2(1).

TMIAP LOW Same as Issue I.F.2(1).

N N/A N/A N/A N/A

N N/A N/A N/A N/A
GSI TITLE ISSUE CATE DECISION RATIONALE V/1 OLD V/I REV. V/I RES. RES.

IDENTIFICATION TYPE GORY REFERENCE P-rem/SN P-rem/SN BASIS DATE

I.F.2(7) Clarify that the QA Program

Is a Condition of the

Construction Permit and

Operating License

I.F.2(8) Compare NRC QA Requirements

with Those of Other Agencies

I.F.2(10) Clarify Requirements for

Maintenance of "As BuiLt"

I.F.2(11) Define Role of QA in Design

and Analysis Activities

II.A.1 Siting Policy Reformulation

II.C.1 Interim ReliabiLity

I

co

TNIAP LOW Same as Issue I.F.2(1).

TMIAP LOW Same as Issue I.F.2(1)

TMIAP LOW Same as Issue I.F.2(1)

TMIAP LOW Same as Issue I.F.2(1)

TMIAP 3b Ref 655: The Commission determined

that before taking on this issue, a

new source term must be approved and

the safety goats established. The

issue would be reassessed in the

future.

TMIAP 3b Ref 813: PRA analysis for 4 plants

were completed and the results were

documented in NUREGs. Other programs,

such as ASAP, extended PRA modeling

to other plants. This issue was

primarily intended to develop and

document methods to perform

retiability analyses. The intent of

the program was accomplished. A

formal information Letter would be

sent to NRR to close out the issue.

TMIAP 3b Ref 816: The purpose of this issue,

as stated in NUREG-0660, was to

"initiate" IREP studies on all

N N/A N/A N/A N/A

N N/A N/A N/A N/A

N N/A N/A N/A N/A

N N/A N/A N/A N/A

NUREG-0933 60 102 A5 84

NUREG-0933 931 977 A3 85

II.C.2 Continuation of Interim

Reliabitity Evaluation

Program (NREP)

NUREG-0933 954 983 C 85
GSI TITLE ISSUE CATE DECISION RATIONALE V/I OLD V/I REV. V/1 RES. RES.

IDENTIFICATION TYPE GORY REFERENCE P-rem/$N P-rem/SN BASIS DATE

remaining operating plants. ISAP

program required under Issue IV.E.5

initiated the analyses at some

plants. In addition, the Severe

Accident Policy Statement requires

plants to perform Limited-scope PRA

to identify plant weaknesses. These

actions fuLfill the requirements of

Items II.C.2 and IV.E.5.

II.C.4 ReLiability Engineering THIAP 3b Ref 1131: NRC design requirements in

1OCFR50 reflected reliabiLity

principles. Operational reliability

could be implemented more effectively

in performance-based regulation

instead of a prescriptive manner.

Elements of a reliabiLity program

were included in recent NRC

initiatives to improve maintenance,

aging, improved risk-based Tech

Specs, plant performance indicators,

and more recently, EGD reliability.

NUREG-0933 478 528 C 89

'I

II.D.2 Research on Relief and Safety TMIAP LOW The portion of this issue involving

Valve Test Requirements valve testing was covered under

II.D.1 which imposed new

requirements. The second part was

concerned with a possibility of ATWS

because of inadequate

depressurization. The issue was

assigned LOW priority category.

NUREG-0933 63 113 N/A N/A

The averted public dose of 1,300

P-rem for half of all plants (55

reactors) that are potentiaLly

affected increases to 2,432 because

of LR.

II.E.2.2 Research on SmaLl Break

LOCAs and AnomaLus Transients

TNIAP 3b Ref 817: The issue is. concerned with

uncertainties in codes used for

SBLOCA analysis. Based on test

NUREG-0933 1,000 1,750 A3 85
GSI TITLE ISSUE CATE DECISION RATIONALE V/I OLD V/I REV. V/I RES. RES.

IDENTIFICATION TYPE GORY REFERENCE P-rem/SM P-rem/$N BASIS DATE

results, TRAC & RELAP were modified

and the issue was closed out.

II.E.2.3 Uncertainties in Performance TMIAP LOW RSB noted that many of the technical

Predictions aspects of this issue are being

investigated under Item II.K.3(30),

"revised SBLOCA model to comply with

50.46 and Appendix K," which was in

progress. Based on V/I value and

overlapping activities of II.K.3(30),

the issue was given LOW priority.

Total averted public dose of 1,565

increases to 2,607 because of LR.

NUREG-0933 52 59 N/A N/A

I

0

II.E.3.4 Alternate Concept Research

II.E.4.3 Integrity Check (Containment)

II.E.4.4(4) Evaluate Venting and Purging

During Normal Operation

TMIAP 3b RES expected Sandia to complete the

study on the usefulness of installing

an add-on DHR system in existing

plants (NUREG/CR-2883) in 1983. The

results would be used in A-45.

TNIAP 3b NUREG/CR-4330 analyses indicated that

containment leakage contributes 1-2

man-rem/RY of the total exposure for

DBAs. Reduction of containment

leakage by a factor of 10 reduced

risk very Little. This issue dealt

with design basis events & not severe

accidents, which are covered by IPE.

NUREG-0933 V/I is -220. Issue was

HIGH priority because of high CDF.

NUREG-1273 reports a dose reduction

as high as 280 P-rem/RY.

TMIAP 3b Ref 382: The primary concern was

failure of the purge Line isolation

valves. The efforts called for by the

issue including NRC actions were

completed. The issue was resolved

without requirements based on study

N N/A N/A A3 <3

NUREG-1273 N/A N/A E 88

NUREG-0933 17 32. A2 <83
GS0 TITLE ISSUE CATE DECISION RATIONALE V/I OLD V/I REV. V/I RES. RES.

IDENTIFICATION TYPE GORY REFERENCE P-rem/$N P-rem/SN BASIS DATE

results & implementation of AWB & CSB

recommendations. NUREG-0933 reported

a maximum V/I of 17 for combined parts

II.E.4.4(4) and II.E.4.4(5).

II.E.4.4C5) Issue Modified Venting and

Purging Requirements.

G5 II .H.1

TMIAP 3b Ref 382: A study of the radiological

consequences of containment purging

was conducted under II.E.4.4(4). The

study results indicated that

then-existing requirements were

adequate.

TMIAP 3b The issue is programmatically

resolved with appropriate management

resources and priorities assigned.

Ref 377 (TMIPO memorandum) stated the

status of cleanup operations.

N N/A N/A B <83

Maintain Safety of THI-2 and N N/A N/A A3 <83

Minimize Environmental

Impacts

I

III.A.1.3(1)Workers (Maintain supplies

of thyroid blocking agent)

III.A.1.3(2)Maintaln Supplies of Thyroid

Blocking Agent (for the

Public)

TMIAP 3b The licensee was aLready required to

maintain supplies of thyroid blocking

agent as a protective measure for

emergency workers.

TMIAP 3b A FEKA Federal PoLicy statement that

was reviewed and accepted by the

Commisssion recommended against a

nationwide requirement for

distribution of KI for use by the

public. The V/i analysis, that was

reviewed and revised by NRC,-showed

that KI offered an extremely small

benefit in relation to its cost over

the uncertainty range. The issue is

receiving further review by NRC (55

FR 39768.)

N N/A N/A C 85

NUREG/CR- Ranges

1443 from

S300,000

to $40M

per

Nodule

I/V

values

decrease

by

less

than

10%

A1,E 85

III.A.3.1(1)Define NRC Role in Emergency TMIAP 3b Refs 408 and 548: The proposed

Situations revision to NRC Manual, Chapter 0502,

was approved by the Commisssion &

N N/A N/A A4 3
GSI TITLE ISSUE CATE DECISION RATIONALE V/I OLD V/I REV. V/I RES. RES.

IDENTIFICATION TYPE GORY REFERENCE P-rem/$N P-rem/SM BASIS DATE

resolved aLL 5 parts of the issue.

ResoLution of III.A.3.1 also resolves

B-71, which was essentiaLLy superceded

by III.A.3.1.

III.A.3.1(2)Revise and Upgrade Plans and

Procedures for the NRC

Emergency Operations Center

III.A.3.1(3)Revise Manual Chapter 0502,

Other Agency Procedures, and

NUREG-0610

III.A.3.1(4)Prepare Commission Paper w

III.A.3.1(5)Revise ImpLementing

Procedures and Instructions

for Regional Offices

III.A.3.2 Improve Operations Center

II.A.3.4 Nuclear Data Link

TMIAP 3b Same as Item III.A.3.(1).

TMIAP 3b Same as Item III.A.3.(1).

TMIAP 3b Same as Item II.A.3.1().

TMIAP 3b Same as Item III.A.3.1(1).

TMIAP 3b Refs 235 & 379: Near-term

improvements to Operations Center

(OC) were made. A study determined

the long-term needs, including a

complete redesign of OC. OIE

considered this issue important &

wanted completion by the end of 1983.

TNIAP 3b Initial study by Sandia identified

several options, none of which were

acceptable. The Staff determined that

Emergency Response Data System

developed by the Licensees, and

already approved by the Commission,

provided a means for electronic

transfer of information, & licensees

should not backfit their Emergency

Response Data System.

N N/A N/A A4 <83

N N/A N/A A4 <83

N N/A N/A A4 <83

N N/A N/A A4 <83

N N/A N/A A2 <3

NUREG-0933 2,100 3,530 C 85
GSI TITLE ISSUE CATE DECISION RATIONALE V/I OLD V/I REV. V/I RES. RES.

IDENTIFICATION TYPE GORY REFERENCE P-rem/SM P-rem/SM BASIS DATE

0J

III.A.3.5 Training, Drills, and Tests TNIAP 3b

III.A.3.6(1)International (Interaction TMIAP 3b

of NRC & Other Agencies)

III.A.3.6(2)Federas (Interaction of NRC TMIAP 3b

& Other Agencies)

III.A.3.6(3)State and Local (Interaction THIAP 3b

of NRC & Other Agencies)

III.B.1 Transfer of Responsibilities THIAP 3b

to FEMA (Emergency

Preparedness of State and

Local Governments)

III.B.2(1) The Licensing Process TMIAP 3b

(ImpLementation of FEMA &

NRC responsibilities)

111.8.2(2) Federal Guidance TMIAP 3b

(Implementation of FEMA &

NRC responsibilities)

Refs 235 & 379: Exercises, scheduling

and training were being conducted

with graduaLly increasing scope & had

been incorporated into routine

ongoing NRC operations.

NUREG-0728 published in 1980

described general NRC

responsibilities & plans for

responding to emergencies and

coordination w/ other agencies.

NUREG-0845 published in 1982

contained detailed procedures for

interaction with others.

Same as III.A.3.6(1).

Same as III.A.3.6(1).

Memorandum of Understanding between

FEMA and NRC was issued. FENA and NRC

had completed evaluation of the first

round of joint excercise at all

operating plants. Items 111.8.1 and

11.B.2 were combined in the

evaluation.

Combined with Item III.B.1 above.

Combined with Item 111.B.1 above.

N N/A N/A C <83

N N/A N/A A4 <83

N N/A N/A A4 <83

N N/A N/A A4 <83

N N/A N/A A <83

N N/A N/A A4 <83

N N/A N/A A4 <83
GSI TITLE ISSUE CATE DECISION RATIONALE V/I OLD V/I REV. V/I RES. RES.

IDENTIFICATION TYPE GORY REFERENCE P-rem/$N P-rem/SN BASIS DATE

III.D.1.3(4)Sponsor Studies to Evaluate TMIAP 3b All four parts of Issue IIl.D.1.3

Charcoal Adsorbers were combined and evaluated together.

The first three parts were covered by

different programs under Severe

Accident Research Program (SARP).

Part 4 associated with the evaluation

of charcoal filter performance under

accident conditions was completed and

results reported in NUREG/CR-2550.

III.D.2.1(1)Evatuate the Feasibility and TMIAP LOW ALL three parts of Issue III.D.2.1

Perform a Value-Impact were evaluated together. Minor

Analysis of Modifying person-rem savings might occur under

Effluent-Monitoring Design accident conditions due to better

Criteria direction of the field survey team.

Also there was a potential risk

increase of 36 P-rem for all plants, Lowering the V/I estimates further.

There was no accident avoidance cost

since the issue was related to

post-accident scenarios. The issue

was assigned LOW priority.

The total averted public exposure of

3,797 P-rem increases to 7,714

because of LR.

N N/A N/A A3,C <83

NUREG-0933 41 66 N/A N/A

to

I

iil.O.2.l(2)Study the Feasibility of

Requiring the Development of

Effective Means for

Monitoring and Sampling

Noble Gases and Radioiodine

Released to the Atmosphere

III.D.2.1(3)Revise Regulatory Guides

IIL.D.2.2(1)Perform Study of

Radioiodoine, Carbon-14, and

Tritium Behavior

TMIAP LOW The issue is combined and evaluated

under Issue IIi.D.2.11).

TMIAP LOW The issue is combined and evaluated

under Issue 11.D.2.1(1).

TMIAP 3b The item was completed in 1981 and

results were documented in NUREG-0771

& 0772. This item is closed with no

N N/A N/A N/A N/A

N N/A N/A N/A N/A

N N/A N/A A3 <83
GSI TITLE ISSUE CATE DECISION RATIONALE V/I OLD V/I REV. V/I RES. RES.

IDENTIFICATION TYPE GORY REFERENCE P-rem/SM P-rem/SM BASIS DATE

new requirements.

III.D.2.3(1)Devetop Procedures to

Discriminate between

Sites/Plants (Liquid Pathway

Radiological Control)

III.D.2.3(2)Discriminate between Sites &

Plants that Require

Consideration of Liquid

Pathway Interdiction

Techniques

III.D.2.3(3)EstabLish Feasible Method of

Pathway Interdiction

III.D.2.3(4)Prepare a Summary Report

III.D.2.4(1)Study Feasibility of

Environmental Monitors

(Offsite Dose Measurements)

III.D.2.5 Offsite Dose Calculation

Manual

TMIAP 3b Analyses for Zion and IP3 were

published in NUREG-0850. NUREG-1054

provided a simplified method.

Environmental SRP (ESRP) 7.1.1 was

drafted with no new requirement &

finally published as a NUREG.

TMIAP 3b Same as Part III.D.2.3(1).

TMIAP 3b Same as Part III.D.2.3(1).

TMIAP 3b Same as part 1i1.D.2.3(1)

TMIAP 3b The issue included installation of

monitors measuring for real-time rate

of exposure to noble gases and

raidioiodines. The issue was combined

with proposed revisions to RG 1.97.

Based on a feasibility study, the NRC

staff concluded that these monitors

were not practical.

TMIAP 3b The purpose of the issue was to

develop a manual. The Offsite Dose

Calculation Manual, NUREG/CR-3332 was

published in 1983.

TMIAP 3b The staff accepted alternative

regulatory concept that recognizes

industry setf-policing program. A

N/A N/A A3 85

N N/A N/A A3 85

N N/A N/A A3 85

N N/A N/A A3 85

N N/A N/A B <83

NUREG-0933 758 1,211 A3 84

I11.D.3.1 Radiation Protection Plans

(Worker Radiation Protection

Improvement)

NUREG-0933 1,880 3,439 A3,D 86
GSI TITLE ISSUE CATE DECISION RATIONALE V/I OLD V/I REV. V/I RES. RES.

IDENTIFICATION TYPE GORY REFERENCE P-rem/SN P-rem/SN BASIS DATE

memorandum of understanding was

issued between NRC & INPO. A trend

analysis indicated ALARA was

integrated into the Licensees,

programs.

IV.C.1 Extends Lessons Learned from

TMI to Other NRC Programs

IV.E.5 Assess CurrentLy Operating

Plants (Safety

Decision-Making)

IV.F.1 Increase OIE Security of the

Power-Ascention Program

ITMIAP 3b Ref 410: The purpose was to extend

Lessons from TMI to other programs,

including transportation, waste

management, research reactors, etc.

Studies performed by NMSS resulted in

the issuance of a draft BTP on waste

form based on TMI experience.

TMIAP 3b Work completed by NRC on this item

was closeLy related to

accomplishments in item II.C.2 on

IREP. ALso ISAP has covered a number

of plants. The Severe Accident Policy

Statement required OLs to perform

plant-specific PRAs.

THIAP 3b OIE reported that procedures have

been issued to increase inspection

coverage during power ascention

testing. The item was resolved with

changes in NRC inspection procedures.

NUREG-0933 1,081 2,137 C 85

N N/A N/A A3 83

N N/A N/A A3 <83

IV.F.2 Evaluate Impact of Financial TNIAP 3b The Staff acknowledged that there was

Disincentives to the Safety some financial disincentive to

of Nuclear Plants plants. Any financial benefit

associated w/ taking safety risks was

small. Other financial issues were

addressed under proposed 50.54(W) on

insurance requirements.

N N/A N/A B <83

A-1 Water Hammer (USI) TAP 3a As indicated in GL 89-21, the staff

concluded that the frequency and

severity of water hammer occurences

N/A N/A N/A 84
GSI

IDENTIFICATION

TITLE ISSUE CATE

TYPE GORY

DECISION RATIONALE V/I OLD V/I REV. V/I RES. RES.

REFERENCE P-rem/SN P-rem/SN BASIS DATE

had been significantly reduced

through incorporation of design

features, proper design of feedwater

valve and control systems, and

increased operator awareness and

training. No additional requirements were imposed on the existing plants.

SRP changes were implemented and

guidelines for upgrading training

programs was developed under issue

I .A.2.3.

A-12 Fracture Toughness of Steam

Generator and Reactor

Coolant Pump Supports (USI)

toI

TAP 3a During licensing of North Anna a

concern was raised regarding the

potential for Lamattr tearing and low

fracture toughness of S/G and RCP

support materials. Tech spec changes

for North Anna to raise the S/G

support temperature prior to

pressurization was made. NUREG-0577

was published and SRP changes were

made. The V/I analysis, although not

conclusive because of uncertainties,

was a supporting factor in not

backfitting.

The database includes the two extreme

V/I values for performing initial

study and hardware modifications,

respectively. Since any initial

assessment of support material could

result in implementing modifications,

the smatter V/I estimate was used in

the decision not to backfit.

TAP 3b Two methods of strong solution

decontamination and dilute

contamination solutions were

considered. NUREG/CR 2963 reported

the decontamination criteria.

NUREG-0577 Ranges

(Rev.1) between

-32,000

and 80

Ranges

between

-60,000

and 153

N/A 84

A- 15 System and Steam Generator

Decontamination

N N/A N/A A3 <83
GSI TITLE ISSUE CATE DECISION RATIONALE V/I OLD V/I REV. V/I RES. RES.

IDENTIFICATION TYPE GORY REFERENCE P-ren/SN P-rem/SN BASIS DATE

A-17

A-21

Systems Interactions in TAP

Nuclear Power Plants (US!)

Main Steam Line Break Inside TAP

Containment Evaluation of

Environmental Conditions for

Equipment Qualifications

3b Resolution of Adverse System

Interaction (ASI) had not proved to

be cost-effective in past & did not

guarantee that arl ASIs would be

identified. GL 89-18 was issued, IPE

and GSIs A-46 and GSI-128, cover some

ASIs. Also GL 89-21 indicates that

the staff expected licensees to

evaluate events in accordance with

the requirements of I.C.5 mandated by

NUREG-0737. A proposed change to SRP

was not implemented.

LOW The issue was given LOW

prioritization based on the total

averted dose of 618 person-rem and

the V/I value of <40 person-rem/$M.

Total averted public dose increases

from 660 P-rem to 1,023 because of

LR.

NUREG-1229 N/A N/A C,E 89

NUREG-0933 40 67 N/A N/A

odI

A-29 Vulnerability to Industrial TAP 3b

Sabotage

Ref 1267: Insider sabotage had not

been a problem in U.S. Section 50.73

covers security that completely

eliminates or mitigates sabotage.

SALP ensured effectiveness of

security programs. Seismic actions

under A-46, extension of IPE to

include flooding-type events, and

NGI-128 Included certain aspects of

the issue.

NUREG-1267, the regulatory analysis,

did not include any quantitative

arguments. NUREG-0933 reports a V/I

of 34 & averted dose of 24,140, which

placed it in MEDIUM priority.

NUREG-1267 N/A N/A C 90

A-31 RHR Shutdown Requirements

(USI)

TAP 3a As indicated in GL-89-21, the ability

to transfer heat from the reactor to

N N/A N/A N/A 78
GSI TITLE ISSUE CATE DECISION RATIONALE V/1 OLD V/I REV. VIE RES. RES.

IDENTIFICATION TYPE GORY REFERENCE P-rem/SN P-rem/SH BASIS DATE

the environment is an important

safety function. The issue was

resolved with the issuance of SRP

Section 5.4.7, which was applicable to

plants licensed after January 1,

1979. There was no backfit for

operating plants based on evaluations.

A-38

t

'0

A-39

Tornado Missiles

Determination of Safety

Relief Valve Pool Dynamic

Loads and Temperature Limits

(USI)

Long Term Seismic Program

TAP LOW There was uncertainty in estimating NUREG-0933

cost savings resulting from

relaxation of missile protection

requirements. The issue was given LOW

prioritization instead of DROP, since

new plants could be built in the

indefinite future.

TAP 3a The issue involved hydrodynamic Loads N

in the suppression pool resulting

from initial vent clearing of relief

valve piping and steam quenching due

to high local pool temperature.

NUREGs 0763, 0783, and 0802 were

published, and SRP Section 6.2.1.1.C

was issued. GL 89-21 indicates that

Mark I plants were covered under

Issue A-7, which was backfitted.

Discussions with NRC staff indicates

that Mark I plants voluntarily

implemented modifications.

TAP 3b Ref 692: The purpose of this issue N

was to gather and develop

information. There were 2 programs:

1- quantification of seismic margins

and 2- earthquakes in eastern U.S.

These programs were intended to

gather & develop information. Also

other NRC plans cover some issues in

A-41.

0.4 0.42 N/A <83

N/A N/A N/A 83

A-41 N/A N/A C 85
GSI TITLE ISSUE CATE DECISION RATIONALE V/I OLD V/I REV. V/i RES. RES.

IDENTIFICATION TYPE GORY REFERENCE P-rem/S P-rem/SM BASIS DATE

A-43 Containment Emergency Sump

Performance (Former USI)

A-45 Shutdown Decay Heat Removal

Requirements (Former USI)

TAP 3a NUREG-0869: CRGR concluded that the

issue should not be backfitted

because of potentially overstated

risk by a factor of greater than "10

or 100" and underestimated costs.

TAP 3b Ref 1443: 6 plants were analyzed

considering 6 different alternatives.

The issue was determined to be

plant-specific and not generic.

Original V/Is vary widely & sometimes

exceeds the 1,000 Limit. IPE would

address the issue. Also Issue A-44

would cover a part of it.

NUREG-0869 Ranges

between

2,344

and

6,818

NUREG-1289 Varies

widely

and

does

exceed

1,000

Ranges N/A

between

4,652

and

13,533

Up by a C,E

factor

of -2

because

of

one-time

costs

86

88

0

U3

C)

B-5 Buckling Behavior of Steel TAP

Containments

3b Ref 1107: There were two parts to

this issue: 1- ductility of two-way

stabs which was resolved based on

availability of sufficient

information and 2- buckling of steel

containment, which was resolved based

on evaluations conducted by NRR

indicating that the proposed SRP

safety margins were overly

conservative, and that there was

general consensus that existing

plants had acceptable safety margin.

NUREG-0933 495 782 B 88

B-9 Electrical Cable Penetrations TAP

of Containment

B-12 Containment Cooling TAP

Requirements (Non-LOCA) for

BWRs Only

3b Draft ORNL report indicated that

then-existing requirements in IEEE

317-1976 and RG 1.65 provided

adequate guidance on design of

penetrations, and the issue was

resolved.

3b Ref 991: The issue was resolved

since: 1- normal ventilation was not

essential for cold shutdown, 2-

failure of the system did not cause

accidents, and 3- the ventilation was

N N/A N/A B <83

N N/A N/A B 83
GSI TITLE ISSUE CATE DECISION RATIONALE V/I OLD V/I REV. V/I RES. RES.

IDENTIFICATION TYPE GORY REFERENCE P-rem/SN P-rem/SN BASIS DATE

not considered in accident analyses.

B- 19

B-26

ThermaL-Hydraulic Stability

Structural Integrity of

Containment Penetration

4

TAP 3b Ref 769: There were no immediate

safety concerns. BWROG initiatives

had resulted in Tech Specs changes

for some BWRs. The new GL

requirements were consistent with

GOCs 10 & 12.

TAP 3b Ref 647: The V/I analysis in

NUREG-0933 was reviewed and changed.

The revised V/I value of 23 P-rem/SN

and ORE of 1,200 placed the issue in

DROP category.

Total averted public dose increases

from 373 P-rem to 630 by LR, which

could place the issue in LOW category.

TAP 3b It appears that there were Tech Specs

changes imposed to shutdown within 48

hours if an inoperable CR was

detected. GE was working with

utilities to replace cottets.

TAP 3b Ref 648: modifications 4 and 5 to

CONTEMPT user's manual were issued

providing the capability to analyze

ice-condensor containments.

N N/A N/A A3,D 85

Ref 647 23 30 B 84

B-48

B-54

B-58

BWR CRD Mechanical Failure

Ice Condenser Containments

Passive Mechanical Failures

N N/A N/A A3 ý83

NUREG-0933 23 38 A3 85

TAP 3b Ref 863: Substantial work was done in NUREG-0933

this area. Nuclear Plant Aging

Research (NPAR) program covered the

scope of issues C-11 and B-58. NPAR

may result in changes to regulatory

requirements.

210 214 C 85
GSI TITLE ISSUE CATE DECISION RATIONALE V/I OLD V/I REV. VII RES. RES.

IDENTIFICATION TYPE GORY REFERENCE P-reM/$M P-rem/SM BASIS DATE

B-60

B-70

Loose Parts Monitoring System TAP

Power Grid Frequency TAP

Degradation and Effects on

RCP

3b Ref 670: SRP 4.4 referencing RG 1.133

(issued in 1981) established the

requirements for CP & OL plants. The

issue could not be backfitted since

there was no basis for LPMS risk

reduction, economical incentives to

plants, and insufficient NRC

resources to review and properly

evaluate LPMS.

3b Ref 83: The issue was resolved

without any actions and deleted the

proposed SRP changes. The rationale

was based on industry experience,

consideration of frequency decay in

original analysis, past ORNL

analysis, and capability to alleviate

this conditon.

3b Ref 412: Then-existing SRP Section

6.2.1.1 required the licensees to

perform an analysis of containment

depressurization, and therefore the

issue was resolved.

N N/A N/A E 84

N N/A N/A B <83

to

C-2 Containment Penetration Due

to Inadvertant Spray

Operation to Determine

Adequacy of Containment

External Design Pressure

C-7 PWR System Piping

C-11 Failure and Reliability of

Pumps and Valves

N/A N/A B <83

TAP 3b Ref 384: Based on operating

experience, then-existing ISI program

for thin-waLled piping was adequate.

Issue C-7 was resolved without any

actions.

TAP 3b Ref 863: Substantial work was done in

this area. Nuclear Plant Aging

Research (NPAR) program covered the

scope of issues C-11 and B-58. NPAR

may result in changes to regulatory

requirements.

N N/A N/A B '(3

NUREG-0933 93 175 C 85

C-12 Primary System Vibration TAP 3b Ref 384: Then-existing SRP N N/A N/A B <83
GSI TITLE ISSUE CATE DECISION RATIONALE V/I OLD V/I REV. V/i RES. RES.

IDENTIFICATION TYPE GORY REFERENCE P-rem/SN P-rem/$N BASIS DATE

Assessment requirements, in conjunctions with

guidelines of RG 1.133 and RG 1.20,

were deemed sufficient to resolve the

issue.

D-1 Advisability of a Seismic

Scram

TAP LOW San Onofre and Diablo Canyon have

seismic scram systems installed.

There is a potential for spurious

reactor scram. The potential benefit

of a seismic scram system was

potentially overestimated. Therefore,

the issue was given LOW priority.

NUREG-0933 14 18 N/A N/A

The total averted public dose of 790

p-rem increases to 1,430 by LR.

Ldw

D-3 Control Rod Drop Accident

3 Setpoint Drift in

Instrumentation

4 End-of-life and Maintenance

Criteria

TAP 3b Ref 382: This was an ACRS generic

concern relative to catcutationat

uncertainty in negative reactivity

insertions (2 vs 3-D computer

models). Ref. 166 indicates that the

issue was closed to ACRS

satisfaction. GE's analysis was

adequate even though it was a

two-dimentionaL analysis.

NGI 3b Ref 903: Rev. 2 of RG 1.105

referencing ISA standards was issued

after receiving public comments on

the proposed changes. SRP did not

have to be changed. Ref 50 includes a

draft qualitative regulatory analysis

in support of Revision 2 to RG 1.105.

NGI 3b The purpose was to establish criteria

for maintenance intervals and

equipment aging. Then-existing SRP

3.11 and RGs 1.33 and 1.89 provided

adequate guidance to the industry.

N N/A N/A B c83

N N/A N/A A2 86

N N/A N/A • B <3
GS! TITLE ISSUE CATE DECISION RATIONALE V/1 OLD V/1 REV. V/I RES. RES.

IDENTIFICATION TYPE GORY REFERENCE P-rem/$N P-rem/SN BASIS DATE

6

12

Separation of BWR Control Rod NGI 3b Ref 382: LER review indicated that

from Drive separation of control rod & drive did

not increase separation or RDA

probability. Previous studies

indicated that the probability of

fuel damage from RDA was smalL. Also

increased GE analysis margin further

minimizes consequences.

BWR Jet Pump Integrity NGI 3b Ref 666: IE BuLletin 80-07 required

inspection of hold down beams.

Industry initiative to either replace

beams or commit to an acceptable

IST/ISI program resolved the issue.

PWR Pipe Crack NGI 3b Ref 865: Both Long-term (SRP changes)

and short-term (augmented inspection)

had very tow value/impact ratios

(1.5E-4 to 1.5E-2 person-rem/SN).

Based on tow V/I's and large ORE, the

issue was considered resolved.

Effects of Electromagnetic NGI 3b Ref 607: The results of a study

Pulse on Plants performed by Sandia on Watts Bar was

published in NUREG/CR-3069. With the

Commisssion approval the issue was

resolved with no requirements.

N N/A N/A B 83

NUREG-0933

Ref 865

284 501 D 84

w:

I 14 <1 E 86

20 N N/A N/A A3 84

22 Inadvertant Boron Dilution NGI

Event

3b Originally the issue was recommended

to be dropped based on Low V/I

ratios. Further study by 0S1

indicated that the event was not

severe enough to warrent backfitting.

Existing SRP 15.4.6 requirements

address the concern. Also recommended

deleting the SRP requirements based

on tow risk values.

NUREG-0933 12 21 E 85
GSI TITLE ISSUE CATE DECISION RATIONALE V/1 OLD V/I REV. V/I RES. RES.

IDENTIFICATION TYPE GORY REFERENCE P-rem/SM P-rem/SN BASIS DATE

35 Degradation of Internat

Appurtenances in LURs

NGI LOW The scope of this issue was limited NUREG-0933

to Loose objects in the secondary

systems since other issues (B-60,

A-45, B-5, II.E.3.2) covered other

aspects of Loose part concerns. The

issue was given LOW priority since

all scores placed it in that

category. The core-mett/yr value of

9E-5 could put the issue marginally

in NEDILU4 category. There was a large

uncertainty in risk and cost value.

75 77 N/A -c83

0I 36 Loss of Service Water

Loss of Offsite Power

BUR Reactor Vessel Level

Instrumentation

47

50

Total averted public dose of 5,000

P-rem increases to 8,325 because of LR.

NGI 3b Ref 902: The DHR & SGTR concerns of

AEOW were addressed under other

issues (A-45 & 67.5.2). The scope of

issue 36 was limited to revision of

SRPs 9.2.1 & 9.2.2 providing

additional clarifications (not

requirements). SRP changes were

made.

NGI 3b AEO0 made 8 recommendations. NRR

addressed alt concerns either by

referencing other Generic Issues

covering them (A-", I.C.6, I.C.1),

MAPs, or provides justification in

support of their resolution.

NGI 3b AEOD made 3 reconmmendations. ALL BURs

through BUROG agreed to voluntarily

implement recommended modifications

for 2 of 3 issues. The last concern

they would not. Issue 50 was resolved

and the third AEOD concern would be

retained as NGI 101.

N N/A N/A A3 86

N N/A N/A B,C <83

N N/A N/A A2,D 84
GSI TITLE ISSUE CATE DECISION RATIONALE V/I OLD V/I REV. V/1 RES. RES.

IDENTIFICATION TYPE GORY REFERENCE P-rem/SN P-rem/SN BASIS DATE

61

64

SRV Line Break Inside the

BWR Wetwell Airspace of Mark

I and I! Containments

Identificaton of Protection

Instrument Sensing Lines

NGI 3b Ref 998: The reference summarizes the NUREG-0933

results of risk analysis

(NUREG/CR-4594). The estimated CDF

ranges between 2.5E-7 and 2E-10

CDF/RY and has a risk tess than 5

p-rem/RY. When compared to other

contributers to risk, the results did

not support the need to perform a

cost and V/I analysis or impose new

requirements.

164 304 E 86

NGI

03

I~

66 Steam Generator Requirements NGI

3b The BTP drafted in 1982 was never

issued because it would constitute a

new requirement. ICSB took a

forwardfit position which was

rejected by CRGR based on limited

safety benefit and uncertain cost

estimate. CRGR recommended NRC to

consider this issue in a proposed RG

IC 126-5, which endorses ISA Standard

S67.02.

3b Ref 1147: Issue 66 was an evolution

or extension of A-3, A-4 & A-5

concerning S/G tube integrity.

NUREG-0844 documented results of

these issues with no requirements.

Issuance of the NUREG also resolved

Issue 66. Some parts of Issue 66 were

redirected as Staff Action Item.

Issue 67 which may result in future

requirements. Some were licensing

issues and the remaining parts were

of tittle value to public risk

reduction. Ref 512 is a

prioritization report. Parts 66.1.2,

11.9, and 66.11, which were in DROP

category, are not substantially

affected by LR.

N N/A N/A A2,E <83

Ref 512

for

66.1.1

and

66.5+6

B.C,E 88

375

and

38-58

395

97-147

69 B&W Plant Make-up Nozzle NGI 3b Ref 667: B&W Owners' Group reviewed N N/A N/A D 84
GSI TITLE ISSUE CATE DECISION RATIONALE V/1 OLD V/1 REV. V/1 RES. RES.

IDENTIFICATION TYPE GORY REFERENCE P-rem/SN P-rem/SN BASIS DATE

Cracking the issue and made 4 recommendations.

NRC accepted 3 recommendations as

adequate to resolve the issue. Alt

plants volunteered to implement

actions (Toledo Edison challenged 1

item). NRC will follow the issue

under MAP.

80 Pipe Break Effects on

Control Rod Drive HydrauLic

Lines in the Drywetts of BUR

Mark I and I1 Containments

NGI LOW The ACRS was concerned with the

effects of a LOCA (pipe whip) on CRD

hydraulic systems in Mark 1I and Ill

containments, which are smatter and

more congested than Mark I design.

The issue was assigned LOW priority

status.

NUREG-0933 26 45 N/A <83

wo

t.

The averted public dose of 60

P-rem/reactor increases to -100 because

of LR.

82 Beyond Design Basis

Accidents in Fuel PooLs

NG% 3b NUREG-1353: The risk was not greater NUREG-1353 -300

than risk from core damage accidents

due to seismic events beyond SSE.

Seven alternatives were identifed in

the Regulatory Analysis. The No

Action alternative was selected based

on V/I ratios.

-500 E 89

The worst case V/I estimate of 830

P-rem/SN (for Zion with population of

860 persons/sq. mile) could increase

to over 1,000 P-rem/SN due to LR

considerations. The best estimate V/I

ratio shown in the database is for the

alternative to install a pool spray

system.

NGI 3b The issue was raised by ACRS

regarding older CE plants without

PORVs. The scope was Limited to only

84 CE PORVS NUREG-1044 97 -130 E 90
GSI TITLE ISSUE CATE DECISION RATIONALE V/I OLD V/I REV. V/I RES. RES.

IDENTIFICATION TYPE GORY REFERENCE P-rem/SN P-rem/SM BASIS DATE

6 plants. PORVs may be beneficial for

beyond design basis events. The

decision on the issue was deferred,

pending resolution of A-45. Per

GIMCS, the issue was resolved in 1990

with a 3b classification. The

Commission decided not to backfit

since PORVs were not required to meet

any of the regulations, and the costbenefit analysis did not justify it.

88

09 0@ý (ý9

Earthquakes and Emergency NGI 3b The issue was initiated by UCS.

Planning Proposed amendments to 10 CFR 50 App.

E were issued for public comments. The

Commisssion noted that based on the

review of public comments, EP

requirements were well-defined w/o

need for amendments.

Technical Specifications for NGI LOW The issue was given LOW priority

Anticipatory Trips based on tow safety significance and

V/I number. The V/I analysis did not

consider the averted cteanup cost,

especially for the PORV opening. This

could have significantly increased

the priority category.

N N/A N/A B <83

NUREG-0933 85 105 N/A <83

Total averted public dose of 172

P-rem increases to 356 because of LR.

91 Main Crankshaft Failures in NGI 3b Ref 1071: TDI Owners' Group developed

Transamerica DeLavaL EDGs a program to address the issue. In

response, the NRC issued an SER

(NUREG-1216) concluding that the

Owners' recommendations, plus

additional actions, would address the

issue. NUREG-1216 described a

maintenance and reliability program

for TDI EDGs.

N N/A N/A A3,D 87
GSI TITLE ISSUE CATE DECISION RATIONALE V/I OLD V/I REV. V/I RES. RES.

IDENTIFICATION TYPE GORY REFERENCE P-rem/$N P-rem/SNM BASIS DATE

92 Fuel CrumbLing During LOCA

101 BWR Water Level Redundancy

NGI LOW The issue was placed in LOW priority

category based on tow safety

significance and V/i numbers.

However, it was determined that the

on-going effort to devotop more

realistic ECCS performance models

would consider the fuel crumbling,

issue.

Total averted public dose of 20 P-rem

increases to 30 because of LR.

NGI 3b Ref 1214: The issue deals with

instrument sensing tine break

coincident with a single failure.

Based on tow core damage frequency

(1E-6) and high cost/benefit ratio,

the issue was not backfitted. NRC

concluded that design of BWRs in

conjunction with operator training

and procedures provided adequate

protection. NUREG/CR-5112 was

attached to GL 89-11 and was

forwarded to all plants. The original

V/I estimate assumed a 30-year plant

life. V/I estimate shown in the

database is for ADS modifications of

Group 3 BWRs. Issue 101 resulted from

Issue 50.

NUREG-0933 20 20 N/A r83

NUREG/CR5112

(May 1989

errata)

909 1,606 E <83

w

102 Human Errors Involving Wrong NGI

Unit or Wrong Train

3b The results of NRC review were

documented in NUREG-1192. The issue

was being addressed under broader EOP

inspection program, CR design review,

and future integrated inspections.

Also discovered that INPO reviews of

plants were based on NUREG-1192.

3b Ref 1201: ATWS Rule did not require

SCRAN system diversity for W plants.

Issue 115 involves the under voltage

N/A N/A A3,C. <83

D

115 Reliability of Westinghouse NGI

Solid State Plant Protection

System

NUREG-1341 981 1,956 D <83
GSI TITLE ISSUE CATE DECISION RATIONALE V/I OLD V/I REV. V/I RES. RES.

IDENTIFICATION TYPE GORY REFERENCE P-rem/SM P-rem/SM BASIS DATE

cards in the reactor protection

system. Industry initiatives to

enhance the system by implementing

changes proposed by Westighouse

resolved the issue.

122.2 Initiating Feed and Bleed

122.3

w

Lj CD

Physical Security System

Constraints

125.1.3 SPDS Availability

(Davis-Besse Loss of ALL

Feedwater Event of June 9

1985, Long-Term Actions)

125.11.7 Automatic IsoLation of

Feedwater During SG Line

Break

NGI 3b The issue was resolved based on: 1-

new EOPs issued in response to TMI,

2- existing EOPs provided guidance

for feed-and-bteed, 3- plants were

enhancing .F&B procedures, 4- NRC had

program for inspection of EOPs

including F&B.

NGI LOW The issue was placed in the LOW

priority category based on the safety

significance values and V/I analysis.

The averted public dose of 0.45

P-rem/reactor increases to 0.8

because of LR.

NGI 3b GL 82-33 (Supplement I to NUREG-0737)

required implementation of SPOS in

response to TMIAP item I.D.3

requirements. It was determined that

2/3s of plants did not have SPOS

operational. GL 89-06 enclosing

NUREG-1342 was issued. Based on NRC's

effort in pursuing I.D.2, this issue

was resolved.

NGI 3b Ref 1134 including NUREG-1332: CDF

and risk reduction associated with

removal of auto. AFW isolation signal

was on the order of 1E-7 and 40 P-rem

per plant, respectively. The scope of

issue might also increase the risk

for some plants with no flow

restrictors. Issue resolved with no

action.

NUREG-0933 2,100

NUREG-0933 From

0.05 to

2.25

4 N/A 86

4,244 B 89

NUREG-0933 670

NUREG-1332 130

982 A2 89

205 E 88
GSI TITLE ISSUE CATE DECISION RATIONALE V/I OLD V/I REV. V/I RES. RES.

IDENTIFICATION TYPE GORY REFERENCE P-rem/SN P-rem/SN BASIS DATE

125.11.14 Remote Operation of

Equipment Which Must. Now Be

Operated LocaLly

NGI LOW The issue is placed in the LOW

priority category based on the V/I

value calcuLated for a worst case

scenario.

NUREG-0933 46 93 N/A 86

The averted public dose of 2.3

P-rem/reactor increases to 4.6

because of LR.

wto

127 Testing and Maintenance of NGI

Manual Valves in Safety

Related Systems

131 Potential Seismic NGI

Interaction Involving the

Movable In-core Flux Mapping

System Used in WestinghouseDesigned Plants

134 Rule on Degree and Experience NGI

Requirements

HFI.2 Engineering Expertise on HF

Shift

LOW A V/I analysis was not peformed

because of the inability to ascertain

the expected reduction in public

risk. However, the risk was judged to

be tow. An Information Notice was

sent out.

S The issue concerns a potential RCS

leakage during a seismic event above

and beyond that analyzed in SARs and

is limited to Westinghouse plants.

The issue was given a MEDIUM priority

rank based on the estimated core

frequency values (6.42E-6/RY). Based

on RES management review, it was

decided that the issue could more

effectively be addressed under IPE's

external event.

3b The Commission issued a Policy

Statement in 1989 on Education

Requirements for the SRO and SS, and

withdrew the proposed rule on the

subject.

3b The issue was resolved with issuance

of Policy Statement in October 1985,

and no new requirements were

established.

N N/A N/A N/A N/A

NUREG-0933 149 262 C 90

N N/A N/A Al <83

N N/A N/A Al 86
GSI TITLE ISSUE CATE DECISION RATIONALE V/I OLD V/I REV. V/I RES. RES.

IDENTIFICATION TYPE GORY REFERENCE P-rem/SN P-ren/SN BASIS DATE

HFl.3

HF-8

Guidance on Limits and HF

Conditions of Shift Work

Maintenance and Surveillance HF

Program

3b The issue was resolved with the

issuance of GL 82-12 and GL 82-16, and

no new requirements were established.

N N/A N/A Al 86

3b The issue was resolved with issuance NUREG-0933 26

of Policy Statement on Maintenance,

and no new requirements were

established.

27 Al 88

w
APPENDIX C

Projected Populations Around Nuclear Power Plants

in Years 2000 and 2010

Nuclear Plant Year 2000 Year 2010

Arkansas One 210000 220000

Beaver Valley 3840000 3910000

Bellefonte 1150000 1230000

Big Rock Point 210000 210000

Braidwood 4650000 4750000

Browns Ferry 810000 850000

Brunswick 250000 270000

Byron 1030000 1060000

Callaway 420000 430000

Calvert Cliffs 3140000 3260000

Catawba 1730000 1860000

Clinton 770000 790000

Comanche Peak 1310000 1460000

Donald C. Cook 1310000 1350000

Cooper 190000 200000

Crystal River 490000 550000

Davis-Besse 1990000 2050000

Diablo Canyon 330000 350000

Dresden 7050000 7200000

Duane Arnold 660000 690000

Farley 410000 440000

Fermi 5630000 5840000

Fitzpatrick 810000 800000

Fort Calhoun 800000 830000

Ginna 1120000 1100000

Grand Gulf 380000 410000

Haddam Neck 3770000 3910000

Shearon Harris 1570000 1690000

Hatch 360000 380000

Hope Creek 4960000 5050000

Indian Point 15000000 14890000

Kewaunee 670000 690000

La Salle 1220000 1260000

Limerick 7070000 7170000

Maine Yankee 700000 750000

McGuire 1900000 2040000

Millstone 2860000 2960000

Monticello 2360000 2520000

North Anna 1250000 1340000

Nine Mile Point 810000 790000

Oconee 1080000 1170000

C-1
Projected Populations Around Nuclear Power Plants

in Years 2000 and 2010

Nuclear Plant Year 2000 Year 2010

Oyster Creek 4190000 4300000

Palisades 1220000 1260000

Pale Verde 1330000 1450000

Peach Bottom 4850000 5010000

Perry Nuclear 2530000 2570000

Pilgrim 4590000 4690000

Point Beach 640000 660000

Prairie Island 2490000 2650000

Quad-Cities 760000 780000

Rancho Seco 2200000 2360000

River Bend 860000 920000

H.B. Robinson 810000 880000

Salem 4910000 5000000

San Onofre 5950000 6400000

Seabrook 3900000 4010000

Sequoyah 1020000 1090000

Shoreham 5400000 5420000

South Texas 300000 320000

St. Lucie 780000 860000

Summer 990000 1080000

Surry 2080000 2240000

Susquehanna 1510000 1530000

Three Mile Island 2210000 2240000

Trojan 2160000 2430000

Turkey Point 3070000 3420000

Vernont Yankee 1580000 1620000

Vogtle 690000 750000

Waterford 2130000 2290000

watts Bar 1040000 1120000

WNP-2 310000 330000

Wolf Creek 210000 220000

Yankee Rowe 1760000 1800000

Zion 7720000 7900000

Total Population 1.6E+08 1.7E+08

Source: Interim Draft for Generic Environmental Impact Statement for

Licence Renewal of Nuclear Power Plants, dated October 15, 1990

C-2
REFERENCES

Pacific. Northwest Laboratory (PNL, 1983), A Handbook for Value-Impact

Assessment, December 1983, NUREG/CR-3568.

U.S. Nuclear Regulatory Commission (NRC, 1978a), Generic Task Problem

Descriptions, Categories B, C, and D Tasks, June 1978, NUREG-0471.

U.S. Nuclear Regulatory Commission (NRC, 1978b), Task Action Plans for

Generic Activities, Category A, November 1978, NUREG-0371.

U.S. Nuclear Regulatory Commission (NRC, 1980a), NRC Action Plan Developed

as a Result of the TMI-2 Accident, May 1980, NLREG-0660.

U.S. Nuclear Regulatory Commission (NRC, 1980b), Clarification of TMI

Action Plan Requirements, November 1980, NUREG-0737.

U.S. Nuclear Regulatory Commission (NRC, 1983a), Guidelines for Nuclear

Power Plant Safety Issue Prioritization Information Development, February

1983, NUREG/CR-2800.

U.S. Nuclear Regulatory Commission (NRC, 1983b), Potential for Low

Fracture Toughness and Lamellar Tearing in PWR Steam Generator and Reactor

Coolant Pump Supports, October 1983, NUREG-0577, Rev.1.

U.S. Nuclear Regulatory Commission (NRC, 1984), Regulatory Analysis

Guidelines of the Nuclear Regulatory Commission, May 1984, NUREG/BR-0058.

U.S. Nuclear Regulatory Commission (NRC, 1987), A Prioritization of

Generic Safety Issues, August 1987, NUREG-0933.

U.S. Nuclear Regulatory Commission (NRC, 1989), Regulatory Analysis for

the Resolution of Generic Issue 82, "Beyond Design Basis Accidents in

Spent Fuel Pool," April 1989, NUREG-1353.

U.S. Nuclear Regulatory Commission (NRC, 1990a), Generic Issue Management

Control System, Third Quarter FY-90 Update, August 1990.

U.S. Nuclear Regulatory Commission (NRC, 1990b), Interim Draft for Generic

Environmental Impact Statement for License Renewal of Nuclear Power

Plants, October 1990.

RE-1

NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER

12-89) (Assigned by NRC. Add Vol., Supp., Rev., NRCM 1102, and Addendum Numbers, if eny.) 3201,3202 BIBLIOGRAPHIC DATA SHEET

(See instructions on the reverse) NUREG/CR-5 382

i MTR-90W00467

2. TITLE AND SUBTITLE

Screening of Generic Safety Issues for 3. DATE REPORT PUBLISHED

License Renewal Considerations MONTH I YEAR

December 1991 4. FIN OR GRANT NUMBER

B8685

5. AUTHOR(S) 6. TYPE OF REPORT

A. Faramarzi, A.A. Hughes, S.S. Seth 7. PERIOD COVERED (Inclusive Dares)

B. PERFORMING ORGANIZATION - NAME AND ADDRESS (/fNRC, provide Division, OfficeorRegion, U.S. Nuclear Regulatory Commission, and mailing address; if contractor, provide name and mailing address)

The MITRE Corporation

7525 Colshire Drive

McLean, VA 22102

9. SPONSORING ORGAN IZATION - NAM E AND ADDR ESS IIf NRC. type "Same as above' if contractor. provide NRC Division, Office or Region, U.S. Nuclear Regulatory Commission,

and mailing address,)

Division of Safety Issue Resolution

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

10. SUPPLEMENTARY NOTES

11. ABSTRACT (200 wordsordlesso

The U.S. Nuclear Regulatory Commission (NRC) is developing

regulations for renewing the operating licenses of nuclear power plants

to ensure that they operate safely beyond the present license terms of

40 years. One consideration relates to past resolutions of generic

safety issues (GSIs) that did not result in backfit requirements on the

licensees. The consideration of an additional operating term of 20 years

which the proposed license renewal rule allows, could have retrospective

implication for the basis of those GSI resolutions. As part of its technical support to the NRC for the development of license renewal regulations, MITRE has performed an independent review of the GSIs to identify

those that could be potentially affected by license renewal considerations. This report describes the screening process and the results of

that work.

12. KEY WORDS/DESCR!PTORS (List words orphrases that will assist researchers in locating the report.) 13. AVAILABILITY STATEMENT

generic safety issues Unlimited

license renewal 14. SECURITY CLASSIFICATION

nuclear plants hiPage)

Unclassified

(This Report)

Unclassified

15. NUMBER OF PAGES

16. PRICE

NRC FORM 335 12-89)




THIS DOCUMENT WAS PRINTED USING RECYCLED PAPER

UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, D.C. 20555

OFFICIAL BUSINESS

PENALTY FOR PRIVATE USE, $300

SPECIAL FOURTH-CLASS RATE POSTAGE E FEES PAID

USNRC

PERMIT No. G-67

0

z

z

z

ci,