United States Nuclear Regulatory Commission

Office of Public Affairs

Washington, DC 20555

Phone 301-415-8200 Fax 301-415-2234

Internet:opa@nrc.gov

No. 96-169 FOR IMMEDIATE

RELEASE

(Tuesday, November 26, 1996)

NOTE TO EDITORS:

The Nuclear Regulatory Commission has received the four

attached reports from its Advisory Committee on Reactor

Safeguards. The reports, in the form of letters, provide

comments on:

-- Plant-specific application of safety goals.

-- Position on direction setting Issue 32 - Future role of

NRC research.

-- Proposed rule on steam generator integrity.

-- NRC programs for risk-based analysis of reactor

operating experience.

#

Attachments:

As stated
November 18, 1996

The Honorable Shirley Ann Jackson

Chairman

U.S. Nuclear Regulatory Commission

Washington, D.C. 20555-0001

Dear Chairman Jackson:

SUBJECT: PLANT-SPECIFIC APPLICATION OF SAFETY GOALS

During the 436th meeting of the Advisory Committee on Reactor

Safeguards, November 7-9, 1996, we discussed the application of

Safety Goals on a plant-specific basis. This subject was also

discussed at meetings of our Joint Subcommittees on Probabilistic

Risk Assessment and Plant Operations on July 17-18, 1996, and of

our Subcommittee on Probabilistic Risk Assessment on August 7,

1996. We also had the benefit of the documents referenced.

In a Staff Requirements Memorandum dated June 11, 1996, we were

requested to provide recommendations on how the Commission's Safety

Goals and Safety Goal Policy should be revised to make them

acceptable for use on a plant-specific basis.

The Safety Goal Policy Statement made it clear that the

Quantitative Health Objectives (QHOs) and the subsidiary Core

Damage Frequency (CDF) goal were to provide standards for the NRC

staff to judge the overall effectiveness of the regulatory system.

That is, if the risk posed by the population of plants on the

average proved to be less than the Safety Goals, then the staff

(and presumably the public) would deem that the regulatory system

had functioned appropriately to protect the health and safety of

the public.

The Safety Goals quantified "how safe is safe enough" for the

population of U. S. plants. For an individual plant, however, the

acceptable level of risk is determined by the concept of "adequate

protection," which in the final analysis means compliance with the

body of regulations. Risk-informed analyses would provide a more

rational basis for making regulatory decisions regarding plantspecific requests for exemptions from the rules or for changes to

the licensing basis, and the acceptability of new regulations.

In our August 15, 1996 report, we stated: "the safety goals and

subsidiary objectives can and should be used to derive guidelines

for plant-specific applications. It is, however, impractical to

rely exclusively on the Quantitative Health Objectives (QHOs) for

routine use on an individual plant basis. Criteria based on core
3

damage frequency (CDF) and large, early release frequency (LERF)

focus more sharply on safety issues and can provide assurance that

the QHOs are met."

In developing plant-specific criteria, it is important to consider

the regulatory needs in the near future and to ensure that the

process will be evolutionary rather than so revolutionary that it

might discourage the licensees from using this approach. It

appears that most of the anticipated licensee requests for changes

to their current licensing basis will deal with Level 1

probabilistic risk assessment (PRA) issues, e.g., inservice

inspection, extension of allowed outage times. Furthermore, most

licensees have only recently familiarized themselves with Level 1

PRA methodology for the narrow regime of power operations. They

are just beginning to integrate findings of such Level 1 risk

assessments with the safe operation of their plants. Even the NRC

staff is still coming to grips with the implications of Level 1

risk assessment results for regulation of nuclear plants. Many

licensees do not have access to the technologies for facile conduct

of full-scope Level 2 or Level 3 PRAs that treat power operations,

low power/shutdown operations, as well as accidents initiated by

external events. Commonly accepted standards for such extensive,

in-depth analyses do not exist.

An evolutionary and pragmatic approach for using Safety Goals on a

plant-specific basis would be to use the CDF as the primary

criterion for evaluating proposed changes along with a qualitative

or quantitative evaluation of the possible Level 2 and Level 3 PRA

issues raised by these changes. For a quantitative analysis, the

following two options are offered:

1) Full-scope Level 2 PRA (with fission product transport

capability).

To use this option, a conservative value for a LERF criterion must

be determined. This value, along with the CDF criterion, will

provide an acceptable basis for decisionmaking. We note that both

the NRC staff and the Electric Power Research Institute, in its,

"PSA Application Guide," are proposing the use of LERF as an

acceptance criterion.

2) Full-scope Level 2 PRA (without fission product transport

capability).

To use this option, conservative values for early containment

failure frequency criteria for different reactor designs must be

determined. These values, along with the CDF criterion, will

provide an acceptable basis for decisionmaking.

In the longer term, we believe the agency should move beyond the

evaluation of risk associated with proposed changes to individual
4

plant licenses and apply the Safety Goals to assess the

acceptability of plant-specific risk. This could be done in terms

of the QHOs, along with the CDF, or in terms of the CDF and LERF.

To use the QHOs directly, it would be necessary to have full-scope

Level 3 PRAs. We believe that the use of Level 3 PRAs in the

future should be encouraged.

Sincerely,

/s/

T. S. Kress

Chairman, ACRS

References:

1. Staff Requirements Memorandum dated June 11, 1996, from John

Hoyle, Secretary, NRC, to John T. Larkins, Executive Director,

ACRS, Subject: Meeting with ACRS, Friday, May 24, 1996

2. ACRS report dated August 15, 1996, from T. S. Kress, Chairman,

ACRS, to Shirley Ann Jackson, Chairman, NRC, Subject: RiskInformed, Performance-Based Regulation and Related Matters

3. Electric Power Research Institute Report TR-105396, "PSA

Application Guide," prepared by ERIN Engineering and Research,

Inc., August 1995
November 19, 1996

The Honorable Shirley Ann Jackson

Chairman

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

Dear Chairman Jackson:

SUBJECT: POSITION ON DIRECTION SETTING ISSUE 22 -- FUTURE ROLE OF

NRC RESEARCH

During the 435th and 436th meetings of the Advisory Committee on

Reactor Safeguards, October 9-12 and November 7-9, 1996,

respectively, we reviewed Direction Setting Issue (DSI) 22. At the

435th meeting, we discussed this issue with the NRC staff. We also

had the benefit of the documents referenced.

Direction Setting Issue 22 raises the question of what role the

Office of Nuclear Regulatory Research (RES) will have in the

future. A range of possible roles is defined in the discussion.

These vary from elimination of a research capability at NRC to

continuation of the research at its current, diminished level on a

broad range of topics. The preliminary thinking is to select the

continuing "business as usual" role for RES.

We contend that, first, changes are occurring within both the

nuclear industry and the regulatory community that make it

essential for NRC to have a research function. Second, we contend

that a "business as usual" approach to NRC research is too timid.

There is an urgency for the NRC to have research information to

meet its obligations to protect the public health and safety in a

changing environment. Finally, we contend that the planning for

future research should be directed toward areas of focused need.

In particular, research is needed to support NRC's transition to

risk-informed and performance-based regulation.

The research arm of NRC has occupied a central role in the

development of the body of regulations needed to ensure public

health and safety in the commercial use of nuclear power. Since

the division of the Atomic Energy Commission into the NRC and what

eventually became the Department of Energy, RES has overseen the

work needed to develop the design-basis analysis of nuclear power

plants. This has included ensuring through a combination of

experimental and analytical research that the analyses done for

Appendix K to 10 CFR Part 50 are on a sound technical foundation.

RES has also undertaken a vast effort to understand the residual

risk posed by the use of nuclear power through the studies of

severe accidents and the associated radionuclide source terms. RES

has, in fact, been responsible for the evolution in the analysis of
6

reactor safety from the bounding and the qualitative to the use of

quantitative risk analysis.

From the pinnacle following the accident at Three Mile Island, RES

has suffered a continuing scale-back of the activities it can

afford to undertake. As with many institutions facing budgetary

pressures, the longer term benefits of research activities have

been sacrificed to ensure that there is the necessary financial

backing for day-to-day activities that are the responsibility of

NRC. NRC's research budget has, then, suffered disproportionately

when funding cutbacks have been inflicted on NRC as a whole.

Today, the available funding for research is, indeed, small enough

that it is a legitimate question whether a viable research program

can be maintained.

At the time these cutbacks in research funding have been taking

place, changes have also been taking place in the way society deals

with safety regulation. Most directly obvious has been the effort

supported by both the Executive Branch and by Congress to base

regulation on actual risk rather than bounding conservatism. The

Vice President heads a Government-wide effort to base regulation,

including regulation of nuclear power, on risk. Relative to most

other regulatory agencies, NRC is well on the way to developing a

risk-informed and performance-based regulatory system. NRC may

well set an example for other regulatory agencies in this regard.

It is, then, important that this be a good example.

A second societal development that will have safety implications is

the economic deregulation of electrical power generation. This

development has yet to be fully realized, but already efforts are

being undertaken by the nuclear utilities to achieve greater

economic competitiveness. Increases in reactor operating power and

the extension of fuel life are just two immediate steps the

industry is taking that have obvious safety implications. It is

widely forecast that draconian measures will be necessary in the

future to maintain nuclear power as a viable option for the

generation of electrical energy. There are, of course, other

changes taking place in the industry that fall in the domain of NRC

such as plant aging; plant decommissioning; development of new,

passive plant designs; and disposal of nuclear waste.

NRC is making great efforts to respond to the challenges posed by

societal and industrial changes that are now taking place. The

information available to the agency to meet these challenges is,

however, proving to be limiting. By way of examples, consider the

following:

ˇ NRC is attempting to develop a risk-informed and performancebased regulatory system to improve the safety of nuclear

plants and to relieve the industry of unnecessary burden.
7

But, NRC is trying to do this without any detailed knowledge

of shutdown risk because RES is unable to fund studies

comparable to the NUREG-1150 studies performed to understand

risk during power operations.

ˇ NRC development of probabilistic methods has not kept pace

with its needs. Methods to treat human errors of commission

or the impacts of organizational factors and management

practices on risk are not available. Experience shows that

human errors, organization, and management are responsible for

or contributing to many accidents and "near misses."

ˇ NRC wants to regulate in light of risk, but there is now only

the technical capability for routine, noncontroversial

evaluation of core-damage frequency. The capacity to extend

estimates of core-damage frequency to evaluate risk has not

been made widely available. There is not even consensus on

how accurately analyses of risk, given that core damage has

occurred, must be done nor how comprehensive such analyses

must be.

ˇ The introduction of digital instrumentation and control (I&C)

systems in nuclear power plant safety related systems requires

NRC to have the capability to regulate high-reliability

software-based systems. NRC's understanding is limited to

current software engineering methods which employ highly

disciplined development process to design and produce highreliability software. A consequence of this approach is lack

of well-developed methods for evaluating the product of the

process. NRC is limited to regulating the process of design

and development of digital I&C systems because no accepted

tools are available for evaluating the product.

ˇ Financial constraints forced NRC to allow its codes for

predicting fuel behavior to atrophy so they are no longer up

to the state of the art. These codes cannot adequately

predict fuel and clad behavior at burnups now being used by

licensees. Recovery actions by RES have been constrained by

resource limitations to narrow topical areas.

ˇ NRC has not yet been able to formulate a risk-informed and

performance-based fire protection rule to replace Appendix R

to 10 CFR Part 50 which has been the source of so many

exemptions and other controversies.

ˇ NRC's opportunities to leverage its research budget by

participating in international research consortia are becoming

limited as NRC has less to contribute to the consortia

efforts.
8

ˇ NRC finds it must evaluate new, passive plant designs using

tools developed for older plant designs because it cannot

afford to develop analytical tools better suited for the

simulation of the physics of these new designs. NRC must

"make do" with computational tools that are now over a decade

old and don't even begin to take advantage of all the more

recent advances in computer technology.

ˇ NRC is unable to predict or detect newly discovered modes of

degradation of the primary pressure boundaries of pressurized

water reactors. It has been forced to use rules designed to

deal with wastage and corrosion of steam generator tubes to

protect against a variety of forms of stress corrosion

cracking.

There is clearly a need for a more aggressive NRC research program

to confront the many challenges that the agency continues to face.

We can be confident that the agency will meet its obligation to

protect the public health and safety. But, without up-to-date

tools produced by a forward thinking research organization, the

agency will have to resort to methods that do not contribute to

either regulatory efficiency or economic efficiency of the nuclear

industry.

The financial resources now available to the agency for performing

research are indeed limited. It has been necessary to make hard

choices on what is to be done and what must be abandoned. A

significant factor in the thinking on what is to be supported and

what is to be abandoned has been a desire to preserve technical

capability. This effort to preserve technical capability appears

to have:

ˇ led to an emphasis on the things that the agency knows best

such as the thermal hydraulics of existing reactors, and

ˇ diluted the efforts in many areas to preserve the current

organizational units of RES.

The preliminary decision for DSI 22, which is to continue

conducting research as it has been done in recent years, appears to

enforce this emphasis on what is known well and to preserve the

existing organizational structure of RES.

It is our position that more aggressive options need to be

developed in response to DSI 22. One of these options is to focus

the research in areas to meet the agency needs as it embarks on its

experiment with risk-informed and performance-based regulation.

The goal of research, then, ought to be, first, to provide risk

information that is far more comprehensive than that now available,

and then, to identify the performance indicators that do indeed
9

reflect the risk. Furthermore, efforts are needed to use plant

data and event reports to assess the adequacy of current

probabilistic risk assessment methods.

RES also needs to anticipate safety implications that licensees

will make in response to economic pressures. RES should be in a

position to provide tools suitable for the safety evaluation of

these changes. To do this, the split of work by RES between user

requests and self-directed work may have to be reevaluated.

Sincerely,

/s/

T. S. Kress

ChairmaN, ACRS

References:

1. United States Nuclear Regulatory Commission, "Strategic

Assessment and Rebaselining Initiative, Stakeholder

Involvement Process Paper," dated September 16, 1996

2. United States Nuclear Regulatory Commission, "Strategic

Planning Framework," dated September 16, 1996

3. United States Nuclear Regulatory Commission, "Strategic

Assessment Issue Paper," dated September 16, 1996
10

November 20, 1996

Mr. James M. Taylor

Executive Director for Operations

U.S. Nuclear Regulatory Commission

Washington, D.C. 20555-0001

Dear Mr. Taylor:

SUBJECT: PROPOSED RULE ON STEAM GENERATOR INTEGRITY

During the 436th meeting of the Advisory Committee on Reactor

Safeguards, November 7-9, 1996, we reviewed the technical bases for

the proposed steam generator integrity rule and an associated

regulatory guide. During the 432nd meeting of the ACRS, June 12-

14, 1996, and meetings of the Joint Subcommittees on Materials &

Metallurgy and on Severe Accidents, June 3-4 and November 5-6,

1996, we heard presentations on subjects related to this matter.

During these reviews, we had the benefit of discussions with

representatives of the staff, the Nuclear Energy Institute, and the

Electric Power Research Institute, as well as the author of a

differing professional opinion. We also had the benefit of the

documents referenced.

The proposed steam generator integrity rule is intended to provide

a risk-informed and performance-based regulation to replace an

existing prescriptive regulation. In its present form, the rule is

a performance-based regulation almost completely divorced from any

direct relation to risk objectives. Such a performance-based rule

proliferates the incoherence problems of the present deterministic

approach. The proposed rule preserves a tenuous connection between

"design-basis space" and "risk space" without clearly articulating

the risk objectives.

Some of the characteristics exhibited in the development process of

the rule and regulatory guide include the following:

ˇ difficulty in reaching agreement on the performance criteria,

ˇ incomplete and sometimes perfunctory analyses required to

provide an assessment of relative risk,

ˇ reliance on core-damage frequency alone as an indicator of

risk, and

ˇ recourse to defense-in-depth without specific criteria for its

use.
James M. Taylor - 11 -

We believe that more direct consideration of risk could have

avoided some of these difficulties.

A controversial element of the proposed rule and regulatory guide

is the introduction of severe accident issues into an area that has

been exclusively resolved by using a design-basis analysis. This

extension of the scope of accident analysis is necessary to make

risk-informed regulatory decisions and is part of the cost of

moving toward risk-informed regulation. Since licensees have done

risk-informed analyses for the Individual Plant Examination (IPE)

process, we believe that the analysis for addressing severe

accident events should not be overly burdensome to them.

Steam generator tube ruptures are small contributors to the total

core-damage frequency, but may be risk significant due to

containment bypass effects. In previous analyses, the staff

performed limited assessments of primary side fission product

attenuation and neglected secondary side attenuation. The

regulatory guide now proposes that the licensees deal with the risk

of a thermally induced tube failure either by demonstrating that

the frequency of the initiating events is sufficiently low

(10-6/reactor year) or by demonstrating that the conditional

probability of tube failure, given that an initiating event has

occurred, is low (on the order of 0.1). We believe that licensees

should also be given the option to demonstrate that, even if

thermally induced tube ruptures occur, the associated risk is low

when a more realistic treatment of fission product attenuation is

made.

We are concerned that the proposed regulatory guide, as presented,

could send the wrong message to licensees that risk-informed and

performance-based requirements are add-ons to the traditional

design-basis accident approach and can only result in an additional

burden. We believe that to be risk informed and performance based,

the regulatory guide should begin with a clear statement of its

objectives, followed by a statement of the performance criteria and

the guidelines for meeting the criteria. We note that the staff

has stated that the proposed performance criteria have been derived

from risk analyses, but we have not seen these analyses. Rewriting

the regulatory guide is not a trivial task, but could result in a

regulatory framework that could be used as a model for future riskinformed and performance-based rulemaking efforts.

In other applications of performance-based regulation such as the

Maintenance Rule, the licensees have been permitted to determine

appropriate performance criteria and have been given more

flexibility in developing the methodology used to determine whether

the criteria have been met. For the steam generator rule, the

staff has concluded that it should approve the performance criteria
James M. Taylor - 12 -

that are proposed by licensees to implement the steam generator

rule. We agree with the decision of the staff that it should

approve the criteria. Industry, however, should be provided more

flexibility to propose alternative performance criteria supported

by an appropriate risk analysis. We would like to review all of

the supporting documentation before commenting on the specific

criteria that have been proposed in the regulatory guide.

The demonstration that the criteria have actually been satisfied

requires a complex process of nondestructive examination and

evaluation of structural integrity and leakage during operation and

design-basis accidents. The methodology required for these

evaluations is not well established. Thus, the staff has felt

constrained to provide a great deal of detail in the proposed

regulatory guide to describe the characteristics of an acceptable

methodology. Although we are not yet prepared to endorse the

regulatory guide, we believe that the present immaturity of the

methodology and the importance of the results justify such an

approach.

The staff position is that the regulatory guide provides sufficient

guidance for developing an acceptable methodology and that formal

review of industry-developed repair criteria and procedures will

not be required. We would like to review the results of a "trade

study" of the preapproval approach vs. the post-implementation

inspection approach to methodology acceptance.

Industry has questioned whether safety factors proposed in the

steam generator rule are more conservative than those required by

the ASME code. We encourage the staff to consider the industry's

arguments.

Industry accepts the performance criterion proposed by the staff

for primary-to-secondary leakage. Industry stated that this

leakage criterion ought not be ipso facto a trigger for inspection

or enforcement of regulations concerning the steam generator rule.

This is a valid concern. Excessive leakage does not necessarily

indicate a failure of the steam generator program. Adequate

opportunities for staff action are available if failures of the

program are discovered following a plant shutdown due to excessive

primary-to-secondary leakage.

We are looking forward to reviewing the staff NUREG report

concerning the staff's treatment of thermally induced tube failure.

We are especially interested in the treatment of elevated

temperatures resulting from flow through leaking tubes, and

coupling between aerosol deposition and thermal hydraulics.

A differing professional opinion (DPO) was filed on July 11, 1994.

We have reviewed the contentions in that DPO and summarized them in
James M. Taylor - 13 -

the attachment. We also note that Generic Safety Issue (GSI)-163,

"Multiple Steam Generator Tube Leakage," identified in 1992 has yet

to be prioritized and resolved. Both the DPO and the GSI are

directly related to the proposed rulemaking. We urge the staff to

prepare a point-by-point response to the issues in the DPO and to

prioritize and resolve GSI-163 before implementing the steam

generator integrity rule.

Dr. William J. Shack did not participate in the Committee's

deliberations regarding this matter.

Sincerely,

/s/

T. S. Kress

Chairman, ACRS

Attachment:

Summary of Differing Professional Opinion

Issues - Presented to the ACRS on

November 7, 1996

References:

1. Memorandum dated October 25, 1996, from Brian Sheron, Office

of Nuclear Reactor Regulation, to John Larkins, Executive

Director, ACRS, Subject: ACRS Review of the Proposed Steam

Generator Rule [forwarding the proposed steam generator rule

and draft steam generator regulatory guide]

2. Memorandum dated May 1, 1996, from James M. Taylor, Executive

Director for Operations, NRC, to Joram Hopenfeld, Office of

Nuclear Regulatory Research, NRC, Subject: Resolution of

Differing Professional Opinion Regarding Voltage-Based Repair

Criteria for Steam Generator Tubes, dated July 13, 1994

3. Memorandum dated July 15, 1994, from James M. Taylor,

Executive Director for Operations, NRC, to John T. Larkins,

Executive Director, ACRS, Subject: ACRS Review Of Proposed

Generic Letter 94-XX, Voltage-Based Repair Criteria for

Westinghouse Steam Generator Tubes [forwarding Differing

Professional Opinion]

4. Report dated September 12, 1994, from T. S. Kress, Chairman,

ACRS, to Ivan Selin, Chairman, NRC, Subject: Proposed Generic

Letter 94-XX, "Voltage-Based Repair Criteria for Westinghouse

Steam Generator Tubes"

5. Memorandum dated September 30, 1994, from Joram Hopenfeld,

Office of Nuclear Regulatory Research, NRC, to John T.

Larkins, Executive Director, ACRS, Subject: Comments On ACRS

Review Of Generic Letter "Voltage Based Repair Criteria for

Westinghouse Steam Generator Tubes"
14

SUMMARY OF DIFFERING PROFESSIONAL OPINION ISSUES

PRESENTED TO THE ACRS ON NOVEMBER 7, 1996

The DPO author estimates core-damage frequency with containment bypass

to be 1 E-4 to 3.4 E-4 events/year. He stated that the uncertainties

associated with characterizing steam generator tube defects and severe

accident phenomena are not sufficiently understood to properly model

tube rupture events. Tubes may fail before the surge line due to:

 crack networking and characterization of flaws not being

adequately determined by nondestructive examinations,

 increased heat transfer caused by flow through tube cracks,

 cracks in tubes opening due to increased pressure,

 cracks in tubes unplugging at elevated pressure, and

 jets from tube cracks eroding adjacent tubes.

The DPO author stated that the staff should document the assumptions

and models used to study hidden uncertainties.
James M. Taylor - 15 -

November 22, 1996

Mr. James M. Taylor

Executive Director for Operations

U.S. Nuclear Regulatory Commission

Washington, D.C. 20555-0001

Dear Mr. Taylor:

SUBJECT: NRC PROGRAMS FOR RISK-BASED ANALYSIS OF REACTOR OPERATING EXPERIENCE

During the 436th meeting of the Advisory Committee on Reactor Safeguards,

November 7-9, 1996, we reviewed the NRC programs for risk-based analysis of

reactor operating experience. We heard presentations by and held discussions

with representatives of the NRC staff regarding programs of the Office for

Analysis and Evaluation of Operational Data (AEOD) including system reliability

studies, risk-based performance indicators (PIs), accident sequence precursor

(ASP) studies, and common-cause failures (CCFs). In addition, our joint

Subcommittees on Probabilistic Risk Assessment (PRA) and on Plant Operations met

with representatives of the NRC staff and its contractors on July 17 and October

30, 1996, to review these matters. We also had the benefit of the documents

referenced.

The AEOD staff presented a summary report of its programs for risk-based analysis

of reactor operating experience. We found these programs to be comprehensive in

covering the collection and analysis of operational safety data based on

operating plant experience and balanced in providing results to both the

immediate assessments for the NRC's plant PIs and the continuing longer range

assembly of useful databases for system performance including CCF rates. We are

convinced that careful review of operating experience is the most applicable

source of information that the NRC and the industry have to validate system

reliability analysis models and predictions, and is the best source of data for

future use.

These databases have been developed through significant resource expenditures by

the industry and the NRC. Both share the results of this effort through their

independent analyses of event reports, system reliability data, etc. This

information can be made useful only if the results are carefully reviewed for

insights into system reliability, human performance, and utility and NRC

management practices that may affect safety. The AEOD programs reflect an

awareness of the need to analyze these data intensively; however, the resources

to perform a full scope analysis are not currently available. We urge that the

priority assigned to this effort be revisited.

The NRC and the Institute of Nuclear Power Operations (INPO) have worked very

hard to negotiate a more extensive sharing of their individual analysis products.
Mr. James M. Taylor - 16 -

These efforts have had some success, namely, NRC has gained access to data in the

Nuclear Plant Reliability Data System of INPO, thus expanding the bases for NRC

compilation of CCF data. Some concerns remain with regard to the protection of

INPO proprietary rights. We believe any database used by NRC on CCF should be

accessible to the public.

The CCF database that has been developed is a significant technical step forward.

AEOD uses the database for generic evaluations. Plant-specific evaluation will

almost certainly require modification to reflect configuration differences

between the specific plant being considered and AEOD's generic evaluations.

Provision should be made to caution any users of the CCF database of the limited

applicability in its current form and, if possible, provide guidance on the

proper process for modifying the database to reflect specific plant

characteristics.

The AEOD staff presented some information on planned revisions to the NRC's PIs

and initial efforts to incorporate risk-based PIs into the program. We look

forward to further examination of candidate indicators. They must be carefully

selected with a clear understanding of how the connection to risk is made and how

this connection can be quantified. A first step will be the definition of the

characteristics and attributes of risk-based PIs.

The AEOD staff is making progressive incremental improvements in its

computational tools. It does not, however, have a long-range vision of the tools

and resources that should be available to support risk-informed and performancebased regulation. We recommend that such a long-range plan be formulated for the

development of computational tools.

The AEOD staff plans to enhance the ASP program to provide a more useful

experience base for evaluating PRA results. The study of reliability of specific

systems is a most important adjunct to these studies. The planned addition to

its study list of selected systems that are important to safety is timely. We

welcome the opportunity to participate in this important work.

Sincerely,

/s/

T. S. Kress

Chairman

References:

1. Office for Analysis and Evaluation of Operational Data report, "Risk-Based

Analysis of Reactor Operating Experience," dated December 15, 1995

2. Memorandum dated March 22, 1996, from C. E. Rossi, Office for Analysis and

Evaluation of Operational Data, NRC, to Office of Nuclear Reactor

Regulation Directors and Regional Directors, NRC, Subject: Special Report

- Emergency Diesel Generator Power System Reliability 1987-1993, INEL-95-

0035 (1 volume)
Mr. James M. Taylor - 17 -

3. Memorandum dated December 22, 1995, from C. E. Rossi, Office for Analysis

and Evaluation of Operational Data, NRC, to G. Holahan, NRR, D.

Crutchfield, M. Hodges and L. Shao, Office of Nuclear Regulatory Research,

NRC, Subject: Common Cause Failure Parameter Estimates for Selected

Components, INEL-94-0064 (6 volumes)