ML15179A007
| ML15179A007 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 06/26/2015 |
| From: | Kimberly Green Plant Licensing Branch III |
| To: | Lashley P FirstEnergy Nuclear Operating Co |
| References | |
| TAC MF3720 | |
| Download: ML15179A007 (5) | |
Text
NRR-PMDAPEm Resource From: Green, Kimberly Sent: Friday, June 26, 2015 1:52 PM To: Lashley, Phil H. (phlashley@firstenergycorp.com)
Cc: Gennardo, David; Pelton, David
Subject:
TSTF-425 LAR - Request for Additional Information (TAC No. MF3720)
Attachments: MF3720 TSTF-425 Final RAI.docx
Dear Mr. Lashley:
By letter to the U.S. Nuclear Regulatory Commission (NRC) dated March 25, 2014, as supplemented by letter dated October 7, 2014, FirstEnergy Nuclear Operating Company submitted a request to modify the Technical Specifications by relocating specific surveillance frequencies to a licensee controlled program in accordance with Nuclear Energy Institute (NEI) 04-10, Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies, for the Perry Nuclear Power Plant, Unit No. 1.
The NRC staff is reviewing your submittal and has determined that additional information is required to complete the review.
By email dated June 23, 2015, I transmitted a draft request for additional information (RAI) to you and requested that you notify me if a call to clarify the NRC staffs request was needed. Per our conversation today, June 26, 2015, you stated that a clarification call was not needed. Therefore, attached please find the NRC staffs RAI. The request remains mostly unchanged from the draft RAI; only minor editorial changes have been made. As requested by you, please provide a response to the NRCs request within 60 days from the date of this email.
Please contact me if you have any questions.
Sincerely, Kimberly Green, Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738 Mail Stop: O-8D15 Phone: (301) 415-1627 kimberly.green@nrc.gov 1
Hearing Identifier: NRR_PMDA Email Number: 2180 Mail Envelope Properties (Kimberly.Green@nrc.gov20150626135100)
Subject:
TSTF-425 LAR - Request for Additional Information (TAC No. MF3720)
Sent Date: 6/26/2015 1:51:55 PM Received Date: 6/26/2015 1:51:00 PM From: Green, Kimberly Created By: Kimberly.Green@nrc.gov Recipients:
"Gennardo, David" <David.Gennardo@nrc.gov>
Tracking Status: None "Pelton, David" <David.Pelton@nrc.gov>
Tracking Status: None "Lashley, Phil H. (phlashley@firstenergycorp.com)" <phlashley@firstenergycorp.com>
Tracking Status: None Post Office:
Files Size Date & Time MESSAGE 1560 6/26/2015 1:51:00 PM MF3720 TSTF-425 Final RAI.docx 30206 Options Priority: Standard Return Notification: No Reply Requested: No Sensitivity: Normal Expiration Date:
Recipients Received:
REQUEST FOR ADDITIONAL INFORMATION FIRSTENERGY NUCLEAR OPERATING COMPANY PERRY NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-440 TAC NO. MF3720 In reviewing the FirstEnergy Nuclear Operating Companyssubmittal dated March 25, 2014, as supplemented by letter dated October 7, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML14084A165 and ML14281A125, respectively), related to relocation of specific surveillance frequencies to a licensee controlled program in accordance with Nuclear Energy Institute (NEI) 04-10, Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies, for Perry Nuclear Power Plant, Unit No. 1 (PNPP), the U.S. Nuclear Regulatory Commission (NRC) staff has determined that the following information is needed in order to complete its review:
RAI 2
Capability Category II of the endorsed American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS) PRA standard (i.e., ASME/ANS RA-Sa-2009) is the target capability level for supporting requirements for the internal events probabilistic risk assessment (PRA) for this application. In the response to RAI 1, dated October 7, 2014, FirstEnergy Nuclear Operating Company (FENOC or the licensee) provided the list of Facts and Observations (F&Os) findings and suggestions from the 2008 Gap Analysis and the2011 and 2012 focused-scope peer reviews for large early release frequency (LERF) analysis and internal flooding analysis, respectively. The licensee also stated: The 1997 PSA [Probabilistic Safety Assessment] PeerReview Certification F&Os were not included as the follow-on reviews were acomplete reevaluation to the PRA standard in effect and supersede thisinformation. Therefore, the 1997 PSA Peer Review Certification F&Os arenot considered relevant to the application. In the application dated March 25, 2014, the licensee provided a summary of the PNPP PRA history. This summary stated that a model update and Computer Aided Fault Tree Analysis System (CAFTA) model conversion were performed in 2011. (The PNPP PRA originally used the WinNUPRA code, as explained in the letter dated October 7, 2014.) The NRC staff notes that many PRA model conversions also result in PRA upgrades.
Based on Section 1-5, PRA Configuration Control, of ASME/ANS RA-Sa-2009, and RG 1.200, Revision 2, Regulatory Position 1.4, PRA Development, Maintenance, and Upgrade, a PRA upgrade must be peer reviewed. The NRC staff reviewed the response to RAI 1, in the letter dated October 7, 2014, and notes that a number of revisions were made to the PNPP PRA in response to the 2008 Gap Analysis. The licensee appears to treat all model revisions as PRA updates (i.e., PRA maintenance) and not PRA upgrades, and therefore,the PRA model does not need a subsequent peer review. The ASME/ANS PRA Standard defines upgrades as, incorporation into a PRA model of a new methodologyor significant changes in scope or capability that impact the significant accident sequences or the significant accident progression sequences. Based on this definition, please address the following:
a) For F&O IE-A4a, the Status/Gap suggested using EPRI [Electric Power Research Institute Technical Report] TR-1013490 as guidance for support system initiator fault trees. (The NRC staff notes that EPRI TR-1013490, Support System Initiating Events:
Identification and Quantification Guideline, published in 2006, appears to have been superseded by its Technical Update, EPRI TR-1016741, published in December 2008.
EPRI TR-1016741 was also sponsored by the NRC Office of Nuclear Regulatory Research, and focuses on the treatment of common cause failures (CCFs) when modeling support system initiating events.) Although the disposition of the F&O states that the initiating event fault trees were updated to include CCF events for applicable components, it is unclear if the disposition resulted in a change in the capability of the PNPP PRA that impacted the significant accident sequences (similar to Example 5 in Nonmandatory Appendix 1-A of ASME/ANS RA-Sa-2009) and could constitute a PRA upgrade. Please expand on the discussion of how this gap was resolved (i.e., discuss the methodology used and describe the changes to the model) and why it is not considered an upgrade.
b) For F&O DA-D5, the Status/Gap states: The Alpha Factor method is used to conduct the CCF analysis. The Perry Resolution then states: In the PRA model update, the Multiple Greek Letter method was used to perform the Common Cause analysis for the model. The Multiple Greek Letter and Alpha Factor Method are similar, but given the other changes to CCF modeling, this could constitute a PRA upgrade. Please explain why this model change is not considered an upgrade.
c) For F&O HR-G7 the Status/Gap states: Although some dependencies are identified during the identification and definition process, all possible HFE [human factors engineering] combinations and dependencies are not addressed. The Perry Resolution then states: HRA [human reliability analysis] for post-initiators, including HEP
[human error probability] dependencies, has been completely redone. It is unclear if a different HRA approach to human error analysis and dependency analysis than what has previously been used has beenapplied. A new approach could also constitute a PRA upgrade. Please expand on the discussion for the resolution of this gap and why it is not considered an upgrade.
RAI 3
In the application dated March 25, 2014, as supplemented by letter dated October 7, 2014, FENOC indicated that portions of the PNPP internal events PRA model have been assessed against ASME RA-Sb-2005, Addenda to ASME RA-S-2002Standard for Probabilistic Risk Assessment for Nuclear PowerPlant Applications, with the clarifications and qualifications in Regulatory Guide (RG) 1.200, Revision 1 (ADAMS Accession No. ML070240001). According to NRC Regulatory Issue Summary 2007-06, Regulatory Guide 1.200 Implementation (ADAMS Accession No. ML070650428), the NRC staff expects licensees to fully address all scope elements with Revision 2 of RG 1.200 (ADAMS Accession No. ML090410014) by the end of its implementation period (i.e., one year after the issuance of Revision 2). Regulatory Guide 1.200, Revision 2, endorses, with clarifications and qualifications, the use of the combined ASME/ANS PRA Standard, ASME/ANS RA-Sa-2009, Addenda to ASME/ANS RA-S-2008 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications.
Identify and address any gaps between the PNPP PRA model and ASME/ANS RA-Sa-2009, including the clarifications and qualifications in RG 1.200, Revision 2, that are relevant to this application, or explain why addressing the gaps would have no impact on this application.
(Portions of the PRA model that have been peer reviewed to ASME/ANS RA-Sa-2009 and RG 1.200, Revision 2, would not require a gap assessment.)