L-12-293, Response to Request for Additional Information Related to Request for Alternative Examination Requirements for American Society of Mechanical Engineers (ASME) Class 1 Piping Welds

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Response to Request for Additional Information Related to Request for Alternative Examination Requirements for American Society of Mechanical Engineers (ASME) Class 1 Piping Welds
ML12226A244
Person / Time
Site: Perry FirstEnergy icon.png
Issue date: 08/10/2012
From: Kaminskas V
FirstEnergy Nuclear Operating Co
To:
Office of Nuclear Reactor Regulation, Document Control Desk
References
L-12-293, TAC ME7564
Download: ML12226A244 (7)
v • d • e


Text

{{#Wiki_filter:" RrstEnergy Nuclear Operating Company Vito A. Kamlnskas Vice President August 10, 2012 L-12-293 A TIN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Perry Nuclear Power Plant Docket No. 50-440, License No. NPF-58 Perry Nuclear Power Plant 10 Center Road Perry. Ohio 4408 t 440-280*5382 Fax: 440-280-8029 10 CFR 50.55a Response to Request for Additional Information Related to Request for Alternative Examination Requirements for American Societv of Mechanical Engineers (ASME) Class 1 Piping Welds (TAC No. ME7564) By letter dated November 14,2011 (Accession No. ML113180450), FirstEnergy Nuclear Operating Company (FENOC) submitted a request for Nuclear Regulatory Commission (NRC) approval for continued use of the existing Perry Nuclear Power Plant (PNPP) risk-informed inservice inspection (RI-ISI) program, with updates, relative to certain non-destructive examination requirements associated with ASME Class 1 piping welds. By letter dated July 16, 2012 (Accession No. ML12167A313), the NRC staff requested additional information to complete its review. Responses to the NRC staffs questi~.ns are provided in the attachment. There are no regulatory commitments contained in this letter. If there are any questions or if additional information is required, please contact Mr. Phil H. Lashley, Supervisor-Fleet Licensing, at (330) 315-6808. Sincerely, Vito A. Kaminskas

Attachment:

Response to July 16, 2012 Request for Additional Information cc: NRC Region III Administrator NRC Resident Inspector Nuclear Reactor Regulation Project Manager

Attachment L-12-293 Response to July 16, 2012 Request for Additional Information Page 1 of6 By letter dated November 14, 2011, FirstEnergy Nuclear Operating Company (FENOC) submitted a 10 CFR 50.55a(a)(3) request for Nuclear Regulatory Commission (NRC) review and approval. By letter dated July 16, 2012, the NRC staff requested additional information to complete its review. The NRC staffs questions are presented below in bold type, followed by FENOC's responses. REQUEST FOR ADDITIONAL INFORMATION

1.

The submittal indicates that a full-scope peer review was performed in 1997 for the internal events probabilistic risk assessment (PRA) and gap or self-assessments have been performed periodically since the 1997 peer review. The submittal also indicates that the PRA model has been revised periodically and some of these revisions included changes to address findings from the 1997 -related review and self-assessments.

a.

Regarding any changes since the independent full-scope peer review characterized as a PRA upgrade per American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS)-RA-Sa-2009, identify if a focused-scope peer review was performed for these changes consistent with the guidance in ASME/ANS-RA-Sa-2009, as endorsed by NRC Regulatory Guide 1.200, and describe any findings from that focused-scope peer review and the resolution of these findings for this application.

b.

If a focused-scope peer review has not been performed for changes characterized as a PRA upgrade, describe what actions will be implemented to address this review deficiency. State when the application will be supplemented to describe any findings from that focused-scope peer review and the resolution of these findings for this application.

Response

An independent peer review of the PNPP PRA model was performed in 1997 under the auspices of the Boiling Water Reactor Owner's Group probabilistic safety analysis (PSA) peer review certification process. Multiple internal self-assessments, including a 2008 self-assessment utilizing an independent contractor, have also been performed. During the 2008 review, ASME RA-Sb-2005, "Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications," December 2005, was used as the standard for the

Attachment L-12-293 Page 2 of6 review. The purpose of this review was to determine gaps between the model at the time and the standard's requirements. Following the February 2011 update of the PNPP average-maintenance model that included the incorporation of review (gap) findings from the 2008 review, the model is judged to meet Capability Category" for all supporting requirements regarding Level 1 internal events only, and is also judged to be compliant with Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 1, January 2007, with the exception of internal flooding and large early release frequency (LERF) modeling. Additionally, the latest model update was structured to satisfy the supporting requirements of the PRA Standard, ASME/ANS RA-Sa-2009, "Standard for Level1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," February 2009, for Level 1 internal events, minus internal flooding and LERF. At the time of the 10 CFR 50.55a(a)(3) submittal, the internal flooding and Level 2 LERF models were judged not to be of the quality required to satisfy the requirements expected per Regulatory Guide 1.200, and subsequently were not utilized for that purpose. Efforts have been expended since the submittal to upgrade the quality of these two models. The Level 2 LERF model has undergone a Regulatory Guide 1.200 (focused scope) peer review in mid-2011; however, that model has not been made effective. Internal flooding upgrade efforts are currently ongoing, with a focused scope peer review completed in July 2012 and gap closure in progress. Following peer review comment incorporation, both the internal flooding and Level 2 LERF models are to become effective. The model updates performed on the Level 1 model since the PSA peer review certification process have not utilized new methodologies; therefore, a PRA upgrade status, as defined by the PRA standard, would not apply, and as such, a full scope peer review was not performed.

2.

The submittal indicates that the licensee's PRA model does not include large early release frequency (LERF). Although a bounding analysis may meet the criteria set forth in the standard, this does not account for high LERF values that may exist and could have an impact on conditional large early release probability (CLERP). Provide the analysis of CLERP based on a realistic approach to LERF calculation.

Response

As discussed above, the Level 2 LERF models were judged not to be of the quality required to satisfy the requirements of Regulatory Guide 1.200 standards at the time of the submittal (November 14, 2011). As such, a bounding LERF analysis was subsequently performed that showed that the risk acceptance criteria continued to be satisfied. Table 1-A depicts the differences between the last deterministic Section XI

Attachment L-12-293 Page 30f6 program and the submittal with corresponding CDF and LERF impacts. As demonstrated in Table 1-A, with the bounding case of CLERP set equal to 1 (that is, all core damage events lead to a large early release), LERF is equal to the CDF.

3.

In a prior risk-informed inservlce relief request submitted on February 12, 2001, the licensee used a Level 2 PRA to perform LERF calculations and produced a table highlighting the change in risk for both core damage frequency and LERF when compared to the last deterministic ASME Section XI program. Provide delta risk calculations for both core damage frequency and LERF on a system and total basis when compared to the last deterministic ASME, Section XI, program for NRC staff consideration.

Response

Table 1-A depicts the differences between the last deterministic ASME Section XI program and this risk-informed inservice inspection (RI-ISI) submittal with corresponding CDF and LERF impacts.

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Attachment L-12-293 Page 6 of6 Table 1-A Terms: Systems 1B13 - Reactor Pressure Vessel Nozzles and Connections 1 B21 - Nuclear Boiler - Main Steam 1 B33 - Reactor Recirculation 1 C41 - Standby Uquid Control 1 E12 - Residual Heat Removal 1E21 - Low Pressure Core Spray CDF Impact wI POD - with Probability of Detection wlo POD - without Probability of Detection 1 E22 - High Pressure Core Spray 1 E32 - Main Steam Isolation Valve Leakage Control System 1 E51 - Reactor Core Isolation Cooling 1 G33 - Reactor Water Cleanup 1 N22 - Miscellaneous Drains - Main Steam Drains 1 N27 - Feedwater LERF Impact wI POD - with Probability of Detection wlo POD - without Probability of Dection Degradation Mechanisms FAC - Flow Accelerated Corrosion IGSCC - Intergranular Stress Corrosion Cracking TASCS - Thermal Stratification, Cycling and Striping TT - Thermal Transient}}

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