ML14205A592

Brunswick Steam Electric Plant, Units 1 and 2 - Request for Additional Information Regarding Voluntary Risk Initiative National Fire Protection Association Standard 805 (TAC Nos. ME9623 and ME9624)

ML14205A592
Person / Time
Site:	Brunswick
Issue date:	07/24/2014
From:	Hon A L Plant Licensing Branch II
To:	Hamrick G T, Murray W R Duke Energy Progress
Andy Hon
References
TAC ME9623, TAC ME9624
Download: ML14205A592 (3)

Text

From: Sent: To: Cc:

Subject:

Hon, Andrew Thursday, July 24, 2014 3:10 PM Murray, William R. (Bill) (Biii.Murray@duke-energy.com) (Biii.Murray@duke-energy.com)

Fields, Leslie; Wall, Scott; Miller, Barry; Regner, Lisa; Wall, Scott; Hamzehee, Hossein BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2-REQUEST FOR ADDITIONAL INFORMATION REGARDING VOLUNTARY RISK INITIATIVE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 (TAC NOS. ME9623 AND ME9624) Mr. George T. Hamrick, Vice President Brunswick Steam Electric Plant Duke Energy Progress, Inc. Post Office Box 1 0429 Southport, North Carolina 28461

Dear Mr. Hamrick:

By letter dated September 25, 2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 12285A428), Duke Energy Progress, Inc. (the licensee) proposed to amend the operating license for the Brunswick Steam Electric Plant, Units 1 and 2, by adopting a new risk-informed based fire protection licensing basis in accordance with National Fire Protection Association Standard 805. The NRC staff has reviewed your application and responses to our request for additional information (RAI) dated June 26, 2014, ADAMS Accession No. ML 14191 A672. We determined that further information is needed to complete our evaluation of the proposed change. On July 10, 2014, the NRC staff and your representatives held a conference call to provide you with an opportunity to clarify the proposed RAI related to probabilistic risk assessment.

The NRC staff's finalized this set of RAis is shown below which were discussed with Mr. Bill Murray and others of your representatives on July 24, 2014, to confirm that these RAis do not contain sensitive information that should be withheld from the public and you would respond by August 15, 2014. The NRC staff considers that timely responses to RAis help ensure sufficient time is available for staff review and contribute toward the NRC's goal of efficient and effective use of staff resources.

Please note that if you do not respond to this letter by the agreed-upon date or provide an acceptable alternate date in writing, we may deny your application for amendment under the provisions of Title 10 of the Code of Federal Regulations, Section 2.1 08. Please also note that review efforts on this task are continuing and additional RAis may be forthcoming.

1 If you have any questions regarding this letter, please contact me. Andy Hon, PE Project Manager (Brunswick Nuclear Plant 1 & 2, Sequoyah Nuclear Plant 1 & 2) Plant Licensing Branch 1/-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation 301-415-8480 OWFN08E5 Mail Stop 08G-9a andrew.hon@nrc.gov cc: Distribution via ListServ ADAMS Accession No.: ML 14205A592 REQUEST FOR ADDITIONAL INFORMATON VOLUNTARY FIRE PROTECTION RISK INITIATIVE DUKE ENERGY PROGRESS BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 DOCKET NOS. 50-325 AND 50-324 Probabilistic Risk Assessment (PRA) RAI 01.f.ii.03 In a letter dated June 26, 2014 (Agencywide Document Access and Management System (ADAMS) Accession No. ML 14191 A672) the licensee responded to PRA RAI 01.f.ii.02 and stated that to "assess the risk due to loss of control, the applicable bin-specific ignition frequency was apportioned to each ignition source in the main control room." The term "loss of control" implies to NRC staff that failure of main control room (MCR) functionality from fire-induced failures, including ones due to fires outside the MCR, leads to the need to abandon the MCR and perform alternate shutdown.

Given previous statements that Brunswick Steam Electric Plant (BSEP) does not credit MCR abandonment upon loss of control, please explain what the term "loss of control" means in the cited statement.

Please confirm that loss of functionality of components, instruments, etc., in the MCR due to fires outside the MCR are included in the risk and delta-risk estimates.

Probabilistic Risk Assessment (PRA) RAI 01.f.ii.04 In a letter dated June 26,2014 (ADAMS Accession No. ML14191A672) the licensee responded to PRA RAI 01.f.ii.02 and stated that their "ASSD [Alternate Safe Shutdown]

strategy is not affected by fire-induced equipment failures in the main control room." Please confirm that BSEP's ASSD strategy and modeling of that strategy for the Fire PRA accounts for fire-induced equipment impacts in the MCR, including potential spurious operations, particularly ones that may not be "recoverable," if needed, upon transfer of control to the ASSD panel. If modeling of BSEP's ASSD strategy does not account for fire-induced MCR equipment impacts on success of ASSD, including spurious operations as cited above, then address these impacts as part of the integrated analysis performed in response to PRA RAI 23 and, to the extent applicable PRA RAI 24. PRA RAI25 2 The recent License Event Report (LEA) 14-004-00, "Fire Related Unanalyzed Condition that Could Impact Equipment Credited in Safe Shutdown Analysis," (ADAMS Accession No. ML 14149A244) stated that: On March 20, 2014, as a result of the transition process from 10 CFR 50, Appendix R, to NFPA 805, a review of the Brunswick Steam Electric Plant Safe Shutdown Analysis determined that a postulated fire in specific fire areas could disable critical components, potentially resulting in equipment required for safe shutdown being inoperable.

The safety significance of this event is minimal. Deterministic analysis methods used to comply with Appendix R require every possible fire scenario to be addressed; however, the risk posed by these hypothetical events has been determined by analysis to be minimal ... This condition is being reported per 10 CFR 50.73(a)(2)(ii)(B) as an event or condition that resulted in the nuclear power plant being in an unanalyzed condition that significantly degraded plant safety ... The safety significance of this event is minimal. Fire watches had previously been established in affected areas prior to the time of discovery for reasons other than this event. The conditions identified here are based on hypothetical fire scenarios that have not actually occurred.

A probabilistic safety assessment developed to analyze this event shows that the core damage frequency (CDF) and large early release frequency (LEAF) are less than red per the significance determination process ... Applicable procedures will be revised by July 31, 2014, to prescribe the required actions for mitigating the effects of a fire in the affected areas. This represents the discovery of new fire-induced accident scenarios that may not have been modeled in the Fire PRA. a. Please explain whether the conditions described in LEA 14-004-00 could impact fire core damage frequency (CDF), large early release frequency (LEAF), change in (.1) CDF, or .1 LEAF. b. Please explain whether these conditions are currently addressed in the Fire PRA. c. If the conditions described by LEA 14-004-00 are not currently addressed in the Fire PRA, please provide justification, or update the Fire PRA to include these scenarios.

If this update is performed prior to transition, please ensure that the responses to PRA RAI 23 and PRA RAI 24 capture the effects of the update. This request can also be addressed by an alternative sensitivity study that incorporates the anticipated effects of the model update and add an implementation item to update the model transition, including an assessment of the potential need for a focused-scope peer review if this update qualifies as a PRA "upgrade" as defined by the ASME/ANS PRA Standard.

If this option is selected, please include the results of the sensitivity in the aggregate analysis described by PRA RAI 23 and, to the extent applicable PRA RAI 24. 3