ML20203C415

Summary of 990105 Meeting with NEI in Rockville,Md Re Issues Related to Risk Informed ISI Activities

ML20203C415
Person / Time
Issue date:	02/08/1999
From:	Wang E NRC (Affiliation Not Assigned)
To:	Frank Akstulewicz NRC (Affiliation Not Assigned)
References
PROJECT-689 NUDOCS 9902120013
Download: ML20203C415 (9)

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UNITED STATES j

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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C.' 20666-0001 February-8, 1999 1

MEMORANDUM TO: Francis M. Akstulewicz, Acting Chief Generic issues and Environmental Projects Branch j

Division of Reactor Program Management i

Office of Nuclear Reactor Regulation FROM:

. Egan Wang, Reactor Engineer b-g Generic issues and Environmental Projects Branch Division of Reactor Program Management

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Office of Nuclear Reactor Regulation

SUMMARY

OF JANUARY 5,1999, MEETING WITH THE NUCLEAR ENERGY INSTITUTE (NEI) REGARDING ISSUES RELATED TO RISK-INFORMED INSERVICE INSPECTION ACTIVITIES On January 5,1999, representatives of the Nuclear Energy Institute (NEI) met with representatives of the Nuclear Regulatory Comrnission (NRC) at the NRC's offices in Rockville, Maryland. The purpose of the meeting was to discuss the industry's risk-informed inservice i

inspection (RI-ISI) program and solicit NRC comments on the industry proposed program for guidance on RI-ISI program. Attachment 1 provides a list of meeting attendees.

NEl introduced and discussed the RI-ISI program. An industry template for plant submittats was presented. The NRC had the following comments:

1.

Format of submittal is not consistent with the 10 items agreed upon between the staff and the industry (mtg on 10/8/98) and listed in the Westinghouse Owners Group Safety Evaluation Report (WOG SER). The staff review of the submittals will be expedited with minimal request for additional information (RAl) if the submittal addresses each topic explicitly as listed in the staff SER.

2.

Summary of any augmented inspection programs that could potentially be impacted by RI-ISI needs to be addressed explicitly.

~ 3.

Several Sections (e.g.,1,2 ) state that the program is in accordance with ASME Code Case N-577. It should be noted that the staff has not approved Code Case N-577.

P 4.

Section 3 4 on piping failure assessment should provide some quantitative results (e.g.,

ranges of failure probabilities) and an indication of which systems have high failure estimates.

5.

Section 3.8,". Structural element and NDE Selection," should provide a summary of the number of elements in each of regions 1 through 4 of Figure 3.7-1 of WCAP-14572,

" Westinghouse Owners Group Application of Risk-Informed Methods to Piping inservice Inspection Topical Report," Revision 1.

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Section 3.9," Program Relief Requests," should provide a summary of impact on l-previous / existing relief requests.

' 7; in order to conclude that the methodology applied by each licensee is consistent with an integrated process previously accepted by the staff, the submittal should include a j

section or subsection of significant deviations from the methodology, if any. For the

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WOG methodology, for example, significant deviations would include, but are not limited to, not addressing or modifying principal steps in the process such as the uncertainty analysis, the results evaluation, the statistical evaluation, the consequence calculations, or the worksheets supplied to the expert panel. All changes to quantitative criterion, l

such as the risk reduction worth (RRW) cut-off criterion, the default of values on the statistical analysis, and the results' evaluation criteria should be reported. If the deviation from the WCAP descriptions might have a significant impact on the results, or increase the sensitivity of the results to the quantitative results of the probabilistic risk assessment (PRA), justification of the adequacy of deviations and alternative approaches should be provided.

i 8.

In order to conclude that the PRA used to support the submittal is based on the current, as operated plant, the licensee should supply the date of the last PRA model up-date. If the last review determined that a PRA up-date was not necessary, the date of this review may be considered the last PRA model up-date. The corresponding base line core damage frequency (CDF) and Large Early Release Frequency (LERF) should also

- be provided to further define and characterize the PRA model used. The licensee should confirm that the evaluation done to support the submittal included consideration of the plant changes not yet modeled in the PRA which could impact the results or 4

conclusions of the submittal.

Samole of orevious safety evaluation conclusions reauired "The individual plant evaluation (IPE) used to support the RI-ISI submittal was the original 1993 study. The first formal update to the Vermont Yankee Nuclear Plant IPE is planned for the end of 1998. The licensee reported that a review of collected equipment unavailability data (as opposed to the generic data used in the IPE) indicated that any changes would not apply to the RI-ISl relevant results' Changes to the plant design and procedures are saved in a book as input to the planned IPE update. The changes were reviewed by the licensee and judged to have no impact on the RI-ISI selections."

Excerpted from " Safety evaluation by the Office of Nuclear Reactor Regulation; L

Proposal to use ASME Code Case N-560 as an attemative to ASME Code,Section XI, Table IWB-2500-1; Vermont Yankee Nuclear Power Station; Docket Number 50-271, November 9,1998." "

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. F. Akst'ulewicz - February 8, 1999 "The Surry IPE was submitted in August 30,1996. Excluding internal floods, the IPE (Individual Plant Examination) estimated a CDF (Core Damage Frequency) of 7.4E-5/yr and a LERF of 1.3E-5/yr... The licensee reported that PRA used in the submittal had an internal events (excluding internal flooding) CDF and the LERF of 7.2E-5/yr and

- 1.1E-5/yr respectively. The difference in the values between the August

- 1996, IPE and the RI-ISI submittal reflects a plant modification that added

' two more chillers in an electrical distribution room. Modifications to the PRA arising from a January 1997, maintenance rule base line inspection and a Virginia Electric Power Company (VEPCO) intemal report dated

' June 1997, were not incorporated into the PRA in time to support the October 31,1997, submittal. The RI-ISI expert panel was, however, advised of the suggested modifications through written descriptions in their worksheets and thus incorporated this information into their deliberations."

Excerpted from " Safety evaluation by the Office of Nuclear Reactor Regulation; i

Proposal to use ASME Code Case N-560 as an attemative to ASME Code,Section XI,

. Table IWB-2500-1; Vermont Yankee Nuclear Power Station; Docket Number 50-271,

- November 9,1998."

9.

In order to conclude with reasonable assurance that the quality of the PRA is adequate to use in support of an integrated methodology previously accepted by the staff, the

. submittal should include a brief synopsis of the reviews which have been performed on the study (reviews described in the IPE submittal need not be repeated).- A brief j

j discussion addressing and resolving any shortcomings or weaknesses identified in the c

staff evaluation report on the IPE or the maintenance rule inspection report should be l

included. - Acceptable resolutions include, for example, a description of changes made

' to the PRA, an explanation as to why these issues do not significantly affect the ISI l

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results, a discussion on how any affect on the ISI results is mitigated by the integrated assessment process, or a description of subsequent peer review or certification results that validates the existing methodology.

Samole of orevious safety evaluation conclusion reauired "The IPE review was completed in May 1996. The review concluded that

. the study met the intent of generic letter 6840, but identified concerns that " post-initiator" Human Reliability Analysis (HRA) yielded overly optimistic (small) HRA probabilities and that dependencies among multiple actions were not fully considered. As stated by the licensee and illustrated by several examples in the ISI submittal, the licensee screened the PRA models for initiating events where a recovery was credited in the

. baseline PRA but is no longer feasible given the segment rupture. In these cases, the probability of successful recovery was factored out of the conditional core damage probability (CCDP). Removal of such recovery actions ensures that the impact of potentially optimistic HRA recovery' probabilities for an initiating event caused by the failure of a r

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February 8, 1999 i

i segment is not a factor in the segment's categorization. For the segment failures which only failed mitigating systems, the '

licensee used primarily functional level PRA results coupled with deterministic considerations.

An overly optimistic HRA estimate could easily place a marginally High segment in Medium, or a marginally Medium segment in Low, but the HRA estimate would have to be underestimated by two orders of magnitude before a marginally High segment would drop to Low. ' The staff recognizes that the use of overly optimistic recovery factors in the baseline PRA may have some minor influence on the consequence '

l categorization of a few segments. However, the staff finds that removal

. l of the recovery factors that are no longer possible because a pipe segment has failed, and the coupling of functional level and deterministic considerations for mitigating system failures, provide assurance that any impact on the consequence categories will be minor and will not invalidate the general results or conclusions. The licensee's analysis of possible RI-ISI specific recovery actions from the effects of segment failure are discussed in Section 3.2.6.

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... Inaccuracies in the iPE models or assumptions large enough to invalidate the broad categorizations developed to support Rl-ISI should have been identified in the licensee or the staff reviews. - Minor errors or inappropriate assumptions will only affect the consequence categorization of a few segments and will not invalidate the general results or

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conclusions."

Excerpted from " Safety evaluation by the Office of Nuclear Reactor Regulation; Proposal to'use ASME Code Case N-578 as an altamative to ASME Code Section XI, Table IWX-2500; Entergy Operations, Inc. Arkansas Nuclear One; Unit Number 2; Docket Number 50-368, December 29,1998."

10.

' in order to conclude that the proposed change meets Principle 4 (proposed increases in CDF or risk are small and are consistent with the Commission's Safety Goal Policy), the licensee should provide a basis for the conclusion that the changes in CDF and LERF J

associated with the implementation of Rl-ISI are, with reasonable assurance, risk i

decreases or at most a negligible increase. If a qualitative discussion based on obvious j,

and straightforward assumptions does not clearly characterize the potential change in 1.

risk, a quantitative evaluation including ' estimates for the largest potential system level j!

CDF and LERF increases should be provided.

b L11.

Expert panel list should be revised to be consistent with the list on page 20 of the L

WCAP-14572.

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12.

In the Section on " Defense-In-Depth"., remove the word " initial......".

l Representatives from Electric Power Research Institute expressed interest in setting up a meeting with the staff on the questions listed on a previous Request for Additional Information.

The participants agreed that comprehensive communication is needed to ensure a successful RI-ISI program.

'l Project No. 689 Attachments: As stated cc w/att: See next page 3

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F. Akstulewicz

-5 February 8, 1999 12.

In the Section on " Defense-in-Depth", remove the word " initial......".

P.epresentatives from Electric Power Research Institute expressed interest in setting up a meeting with the staff on the _ questions listed on a previous Request for Additional Information.

The participants agreed that comprehensive communication is needed to ensure a successful RI-ISI program.

Project No. 689 Attachments: As stated cc w/att: See next page File name: g:\\eyw\\NE10105.wpd

DISTRIBUTION: See attached page A

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NRC/NEl' MEETING ON RI-ISI LIST OF ATTENDEES January 05,1999 NAME ORGANIZATION Goutam Bagchi NRC Syed Ali NRC Stephen Dinsmore NRC Rich Barrett NRC Dick Wessman NRC Egan Wang NRC Jeff Mitman EPRI Millan Straka NUS Info.

Theresa Sutter Bechtel Patrick O'Regan DE&S Scott Kulat inservice Engineering Richard Fougerousve Inservice Engineering Nancy Closky Westinghouse

. Alex McNeill Virginia Power Dennis Weakland Duquesne Light 4

-' Biff Bradley NEl

- Ken Balkey.

Westinghouse Ray West Northeast Utilities J

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i Nuclear Energy Institute.

Project No. 689 i

'cc:

Mr. Ralph Beedle Ms. Lynnette Hendricks, Director Senior Vice President :

Plant Support and Chief Nuclear Officer Nuclear Energy Institute Nuclear Energy Institute.

Suite 400 Suite 400 1776 l Street, NW 1776 l Street, NW.

Washington, DC 20006-3708.

Washington, DC 20006-3708 i

Mr. Alex Marion, Director.

Mr. Charles B. Brinkman, Director Programs' Washington Operations Nuclea'r Energy institute ABB-Combustion Engineering, Inc.

i Suite 400 12300 Twinbrook Parkway, Suite 330

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ji776 l Street, NW.

Rockville, Maryland 20852 Washington, DC 20006-3708 Mr. David Modeen, Director Engineering -

Nuclear Energy Institute Suite 400 1776 l Street, NW Washington, DC 20006-3708 i

Mr. Anthony Pietrangelo, Director Licensing

. Nuclear Energy Institute

. Suite 400 1776 I Street, NW Washington, DC 20006-3708 Mr. Nicholas J. Liparulo, Manager Nuclear Safety and Regulatory Activities Nuclear and Advanced Technology Division

~ Westinghouse Electric Corporation-P.O. Box 355 '

Pittsburgh, Pennsylvania 15230 Mr. Jim Davis, Director

. Operations Nuclear Energy Institute Suite 400 1776 i Street, NW Washington, DC 20006-3708 f

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Distribution: Mtg. Summary w/ NEl Re Rl-ISI Dated February 8, 1999 Hard Coov SDocket File

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. SMagruder SAli SDinsmore EWang EMail SCollins/RZimmerman BSheron BBoger DMatthews TEssig

' FAkstulewicz GTracy, EDO GBagchi RBarrett RWessman JStrosnider GHolahan i

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