Text

Nine Mile Point Nuclear Station, Unit 1, Issuance of Amendment Regarding Transition to a Risk-Informed, Performance-Based Fire Protection Program in Accordance with 10 CFR 50.48(c) (TAC ME8899)
	ML14126A003
Person / Time
Site:	Nine Mile Point
Issue date:	06/30/2014
From:	Bhalchandra Vaidya; Plant Licensing Branch 1
To:	Langdon K; Exelon Generation Co, Nine Mile Point
	Vaidya B K, NRR/DORL/LPL1-1, 415-3308
References
	TAC ME8899
	Download: ML14126A003 (175)
	v • d • e

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. Christopher Costanzo Vice President Nine Mile Point Exelon Generation Company, LLC Nine Mile Point Nuclear Station, LLC P. 0. Box 63 Lycoming, NY 13093 June 30, 2014

SUBJECT:

NINE MILE POINT NUCLEAR STATION, UNIT NO.1 -ISSUANCE OF AMENDMENT REGARDING TRANSITION TO A RISK-INFORMED, PERFORMANCE-BASED FIRE PROTECTION PROGRAM IN ACCORDANCE WITH 10 CFR 50.48(c) (TAG NO. ME8899)

Dear Mr. Costanzo:

The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 215 to Renewed Facility Operating License No. DPR-63 for the Nine Mile Point Nuclear Station, Unit No. 1 (NMP1 ). The amendment consists of changes to the license and Technical Specifications (TSs) in response to your application dated June 11, 2012, as supplemented by letters dated February 27, March 27, April 30, and December 9, 2013; and January 22, March 14, April15, May 9, and May 23, 2014. Nine Mile Point Nuclear Station, LLC (NMPNS, the licensee), submitted a license amendment request (LAR) to revise the fire protection program in accordance with Title 10 of the Code of Federal Regulations (1 0 CFR), Section 50.48(c), for NMP1 and change the license and TSs accordingly. Original application and the supplements were submitted before the Exelon Generation Company, LLC became the Operator of NMP1 on March 25, 2014. Further, after the NRC approved the direct transfer of the operating authority for NMP1 from NMPNS, a subsidiary of Constellation Energy Nuclear Group, LLC (CENG), to Exelon Generation Company, LLC (Exelon, the licensee) and the conforming amendment to the Renewed Facility Operating License for Nine Mile Point Nuclear Station, Units 1 and 2, by letter dated March 28, 2014 (ADAMS Accession No. ML 14087 A27 4 ), Exelon stated that: Prior to the license transfers, CENG made docketed submittals to the NRC that requested specific licensing actions, such as license amendment requests, relief requests, exemption requests, etc. Furthermore, in the application for the license transfers, Exelon stated that upon transfer of the licenses, Exelon would assume all current regulatory commitments made for these units. Accordingly, Exelon hereby adopts and endorses those docketed requests currently before the NRC for review and approval. Exelon requests that the NRC continue to process those pending actions on the schedules previously requested by CENG. The amendment authorizes the transition of the NMP1 fire protection program to a risk-informed, performance-based program based on National Fire Protection Association (NFPA) 805, "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition" (NFPA 805), in accordance with 10 CFR 50.48(c). NFPA 805 allows the use of performance-based methods such as fire modeling and risk-informed methods C. Costanzo such as fire probabilistic risk assessment to demonstrate compliance with the nuclear safety performance criteria. The fire protection license condition in NMP1 's license and TS 6.4 are revised to reflect the use of NFPA 805. To reflect the proper pagination of the license, the amendment includes the license pages 5 through 13. However, only the text of the fire protection license condition, paragraph 2.D.(7) and TS Page 350 of Renewed Facility Operating License, are revised. A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's biweekly Federal Register notice. Docket No. 50-220

Enclosures:

1. Amendment No. 215 to DPR-63 2. Safety Evaluation cc w/encls: Distribution via Listserv Sincerely, Bhalchandra Vaidya, Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation ENCLOSURE 1 AMENDMENT NO. 215 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-63 NINE MILE POINT NUCLEAR STATION, LLC EXELON GENERATION COMPANY, LLC NINE MILE POINT NUCLEAR STATION, UNIT NO. 1 DOCKET NO. 50-220 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 NINE MILE POINT NUCLEAR STATION, LLC EXELON GENERATION COMPANY, LLC DOCKET NO. 50-220 NINE MILE POINT NUCLEAR STATION, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 215 License No. DPR-63 1. The U.S. Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment by Nine Mile Point Nuclear Station, LLC (NMPNS, the licensee) and Exelon Generation Company, LLC (Exelon, the licensee), dated June 11, 2012, as supplemented by letters dated February 27, March 27, April 30, and December 9, 2013; and January 22, March 14, and April 15, May 9, and May 23, 2014, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 1 0 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-63 is hereby amended to read as follows: (2) Technical Specifications The Technical Specifications contained in Appendix A, which is attached hereto, as revised through Amendment No. 215, are hereby incorporated in the license. Exelon Generation Company, LLC shall operate the facility in accordance with the Technical Specifications. In addition, the license is amended as indicated in the attachment to this license amendment, and Paragraph 2.0.(7) of Renewed Facility Operating License No. DPR-63 is hereby amended to read as follows: (7) Fire Protection Exelon Generation shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the license amendment request dated June 11, 2012, supplemented by letters dated February 27, 2013, March 27, 2013, April 30, 2013, December 9, 2013, January 22, 2014, March 14, 2014, April15, 2014, May 9, and May 23, 2014, and as approved in the safety evaluation dated June 30, 2014. Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48( c), the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied. (a) Risk-Informed Changes that May Be Made Without Prior NRC Approval A risk assessment of the change must demonstrate that the acceptance criteria below are met. The risk assessment approach, methods, and data shall be acceptable to the NRC and shall be appropriate for the nature and scope of the change being evaluated; be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant. Acceptable methods to assess the risk of the change may include methods that have been used in the peer-reviewed fire PRA model, methods that have been approved by NRC through a plant-specific license amendment or NRC approval of generic methods specifically for use in NFPA 805 risk assessments, or methods that have been demonstrated to bound the risk impact. 1. Prior NRC review and approval is not required for changes that clearly result in a decrease in risk. The proposed change must also be consistent with the defense-in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation. 2. Prior NRC review and approval is not required for individual changes that result in a risk increase less than 1 x1 o*7 /year (yr) for CDF and less than 1 x1 o*8/yr for LEAF. The proposed change must also be consistent with the defense-in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation. (b) Other Changes that May Be Made Without Prior NRC Approval 1. Changes to NFPA 805, Chapter 3, Fundamental Fire Protection Program Prior NRC review and approval are not required for changes to the NFPA 805, Chapter 3, fundamental fire protection program elements and design requirements for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is functionally equivalent or adequate for the hazard. The licensee may use an engineering evaluation to demonstrate that a change to an NFPA 805, Chapter 3, element is functionally equivalent to the corresponding technical requirement. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard. The licensee may use an engineering evaluation to demonstrate that changes to certain NFPA 805, Chapter 3, elements are acceptable because the alternative is "adequate for the hazard." Prior NRC review and approval would not be required for alternatives to four specific sections of NFPA 805, Chapter 3, for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is adequate for the hazard. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard. The four specific sections of NFPA 805, Chapter 3, are as follows: * "Fire Alarm and Detection Systems" (Section 3.8); * "Automatic and Manual Water-Based Fire Suppression Systems" (Section 3.9); * "Gaseous Fire Suppression Systems" (Section 3.1 0}; and * "Passive Fire Protection Features" (Section 3.11 ). This License Condition does not apply to any demonstration of equivalency under Section 1.7 of NFPA 805. 2. Fire Protection Program Changes that Have No More than Minimal Risk Impact Prior NRC review and approval are not required for changes to the licensee's fire protection program that have been demonstrated to have no more than a minimal risk impact. The licensee may use its screening process as approved in the NRC safety evaluation dated June 30, 2014, to determine that certain fire protection program changes meet the minimal criterion. The licensee shall ensure that fire protection defense-in-depth and safety margins are maintained when changes are made to the fire protection program. (c) Transition License Conditions 1. Before achieving full compliance with 10 CFR 50.48(c}, as specified by (c)2. below, risk-informed changes to the licensee's fire protection program may not be made without prior NRC review and approval unless the change has been demonstrated to have no more than a minimal risk impact, as described in (b)2. above. 2. The licensee shall implement the modifications to its facility, as described in Table S-1, "Plant Modifications Committed," of NMPNS letter dated May 09, 2014, to complete the transition to full compliance with 1 0 CFR 50.48(c) prior to startup from the first refueling outage following issuance of the license amendment. The licensee shall maintain appropriate compensatory measures in place until completion of these modifications. 3. The licensee shall implement the items listed in Table S-2, "Implementation Items," of NMPNS letter dated May 09, 2014, 180 days after issuance of the license amendment unless that date falls within a scheduled refueling outage, then the due date will be 60 days following startup from the scheduled refueling outage. 3. This license amendment is effective as of its date of issuance and shall be implemented by 180 days from the date of issuance.

Attachment:

Changes to the Renewed Facility Operating License No. DPR-63 and Technical Specifications Date of Issuance: June 30, 2014 FOR THE NUCLEAR REGULATORY COMMISSION Benjamin G. Beasley, Chief Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation ATTACHMENT TO LICENSE AMENDMENT NO. 215 TO FACILITY OPERATING LICENSE NO. DPR-63 DOCKET NO. 50-220 Replace the following pages of Renewed Facility Operating License No. DPR-63 with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change. REMOVE INSERT Page 3 Page 3 Pages 5 through 11 Pages 5 through 13 Replace the following pages of Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change. REMOVE INSERT 350 350 (2) Exelon Generation pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (3) Exelon Generation pursuant to the Act and 10 CFR Parts 30, 40, and 70 to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Exelon Generation pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument and equipment calibration or associated with radioactive apparatus or components. (5) Exelon Generation pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility. C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter 1: Part 20, Section 30.34 of Part 30; Section 40.41 of Part 40; Section 50.54 and 50.59 of Part 50; and Section 70.32 of Part 70. This renewed license is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect and is also subject to the additional conditions specified or incorporated below: (1) Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 1850 megawatts (thermal). (2) Technical Specifications The Technical Specifications contained in Appendix A, which is attached hereto, as revised through Amendment No. 215 is hereby incorporated into this license. Exelon Generation shall operate the facility in accordance with the Technical Specifications. (3) Delete Renewed License No. DPR-63 Amendment No. 191 through 210,211,213,214,215 Correction Letter Dated August 7, 2012 c. Exelon Generation shall update the collective occupational dose estimate weekly. If the updated estimate exceeds the 1908 person-rem estimate by more than 1 0%, the licensee shall provide a revised estimate, including the reasons for such changes, to the NRC within 15 days of determination. d. Progress reports shall be provided at 90-day intervals from June 30, 1982 and due 30 days after close of the interval, with a final report within 60 days after completion of the repair. These reports will conclude: (1) a summary of this occupational dose received to date by major task, and (2) a comparison of estimated doses with the doses actually received. (7) Fire Protection Exelon Generation shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensee's amendment request dated June 11, 2012, supplemented by letters dated February 27, March 27, April30, and December 9, 2013; and January 22, March 14, April15, May 9, and May 23, 2014 and as approved in the safety evaluation report dated June 30, 2014. Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c), the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied. (a) Risk-Informed Changes that May Be Made Without Prior NRC Approval A risk assessment of the change must demonstrate that the acceptance criteria below are met. The risk assessment approach, methods, and data shall be acceptable to the NRC and shall be appropriate for the nature and scope of the change being evaluated; be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant. Acceptable methods to assess the risk of the change may include methods that have been used in the peer-reviewed fire PRA model, methods that have been approved by NRC through a plant-specific license amendment or NRC approval of generic methods specifically for use in NFPA 805 risk assessments, or methods that have been demonstrated to bound the risk impact. Renewed License No. DPR-63 Amendment No. at4, 215 1. Prior NRC review and approval is not required for changes that clearly result in a decrease in risk. The proposed change must also be consistent with the defense-in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation. 2. Prior NRC review and approval is not required for individual changes that result in a risk increase less than 1x10-7/year (yr) for CDF and less than 1 x1 0-8/yr for LEAF. The proposed change must also be consistent with the defense-in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation. (b) Other Changes that May Be Made Without Prior NRC Approval 1. Changes to NFPA 805, Chapter 3, Fundamental Fire Protection Program Prior NRC review and approval are not required for changes to the NFPA 805, Chapter 3, fundamental fire protection program elements and design requirements for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is functionally equivalent or adequate for the hazard. The licensee may use an engineering evaluation to demonstrate that a change to NFPA 805, Chapter 3, element is functionally equivalent to the corresponding technical requirement. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard. The licensee may use an engineering evaluation to demonstrate that changes to certain NFPA 805, Chapter 3, elements are acceptable because the alternative is "adequate for the hazard." Prior NRC review and approval would not be required for alternatives to four specific sections of NFPA 805, Chapter 3, for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is adequate for the hazard. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard. The four specific sections of NFPA 805, Chapter 3, are as follows:

Fire Alarm and Detection Systems (Section 3.8);

Automatic and Manual Water-Based Fire Suppression Systems (Section 3.9);

Gaseous Fire Suppression Systems (Section 3.1 0); and

Passive Fire Protection Features (Section 3.11 ). This License Condition does not apply to any demonstration of equivalency under Section 1.7 of NFPA 805. Renewed License No. DPR-63 Amendment No. 215 2. Fire Protection Program Changes that Have No More than Minimal Risk Impact Prior NRC review and approval are not required for changes to the licensee's fire protection program that have been demonstrated to have no more than a minimal risk impact. The licensee may use its screening process as approved in the NRC safety evaluation dated June 30, 2014 to determine that certain fire protection program changes meet the minimal criterion. The licensee shall ensure that fire protection defense-in-depth and safety margins are maintained when changes are made to the fire protection program. (c) Transition License Conditions 1. Before achieving full compliance with 10 CFR 50.48(c), as specified by (2) below, risk-informed changes to the licensee's fire protection program may not be made without prior NRC review and approval unless the change has been demonstrated to have no more than a minimal risk impact, as described in (2) above. 2. The licensee shall implement the modifications to its facility, as described in Table S-1, "Plant Modifications Committed," of NMPNS letter dated May 9, 2014, to complete the transition to full compliance with 10 CFR 50.48(c) prior to startup from the first refueling outage following issuance of the license amendment. The licensee shall maintain appropriate compensatory measures in place until completion of these modifications. 3. The licensee shall implement the items listed in Table S-2, "Implementation Items," of NMPNS letter dated May 9, 2014, 180 days after issuance of the license amendment unless that date falls within a scheduled refueling outage, then the due date will be 60 days following startup from the scheduled refueling outage. (8) Hot Process Pipe Penetrations Hot Process Pipe Penetrations in the Emergency Condenser Steam Supply (2 each), Main Steam (2 each), Feedwater (2 each), Cleanup Suction (1 each), and Cleanup Return (1 each) piping systems have been identified as not fully in conformance with FSAR design criteria. This anomaly in design condition from the original design is approved for the duration of Cycle 8 or until March 31, 1986, whichever occurs first, subject to the following conditions: (a) An unidentified leakage limit of a change of 1 gallon per minute in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to permit operation will be imposed by administrative control (Standing Order) at the facility for the interim period. Renewed License No. DPR-63 Amendment No. 2-+4, 215 (b) NMP LLC shall restore the facility to a condition consistent with the FSAR or provide a change to the FSAR criteria for staff review and approval prior to restart from the forthcoming Cycle 8 outage. (9) On the closing date of the transfer of Nine Mile Point Nuclear Station, Unit No. 1 (NMP-1) to it, NMP LLC shall: (1) obtain from the transferor all of its accumulated decommissioning trust funds for NMP-1 , and (2) receive a parent company guarantee pursuant to 10 CFR 50.75(e)(1) (iii)(B) (to be updated annually) in a form acceptable to the NRC and in an amount which, when combined with the decommissioning trust funds for NMP-1, equals or exceeds the total amount required for NMP-1 pursuant to 10 CFR 50.75(b) and (c). (1 0) The decommissioning trust agreement for NMP-1, at the time any subject direct transfer is effected and thereafter, is subject to the following: a. The decommissioning trust agreement must be in a form acceptable to the NRC. b. With respect to the decommissioning trust funds, investments in the securities or other obligations of Constellation Energy Group, Inc., New Controlled, or their affiliates, successors, or assigns, are and shall be prohibited. Except for investments tied to market indexes or other nuclear sector mutual funds, investments in any entity owning one or more nuclear power plants are and shall be prohibited. c. The decommissioning trust agreement must provide that no disbursements or payments from the trusts, other than for ordinary administrative expenses, shall be made by the trustee unless the trustee has first given the NRC 30 days prior written notice of the payment. The decommissioning trust agreement shall further contain a provision that no disbursements or payments from the trusts shall be made if the trustee receives prior written notice of objection from the Director of the Office of Nuclear Reactor Regulation. d. The decommissioning trust agreement must provide that the agreement cannot be amended in any material respect without 30 days prior written notification to the Director of the Office of Nuclear Reactor Regulation. e. The appropriate section of the decommissioning trust agreement shall state that the trustee, investment advisor, or anyone else directing the investments made in the trusts shall adhere to a prudent investor standard, as specified in 18 CFR 35.32(a)(3) of the Federal Energy Regulatory Commission's regulations. Renewed License No. DPR-63 Amendment No. 214 (11) NMP LLC shall take all necessary steps to ensure that the decommissioning trusts are maintained in accordance with the Application for approval of the transfer of the NMP-1 license to NMP LLC (Application), the requirements of the Order approving the transfer, and the related safety evaluation. (12) At the time of the transfer of NMP-1 to NMP LLC, NMP LLC shall enter or shall have entered into an intercompany credit agreement with Constellation Energy Group (CEG), Inc. or New Controlled, whichever entity is the ultimate parent of NMP LLC at that time, in the form and on the terms represented in the Application for license transfer. Should New Controlled become the ultimate parent of NMP LLC following the direct transfer of the license to NMP LLC, NMP LLC shall enter or shall have entered into a substantially identical intercompany credit agreement with New Controlled at the time New Controlled becomes the ultimate parent; in such case, any existing intercompany credit agreement with CEG, Inc. may be canceled once the intercompany credit agreement with New Controlled is established. Except as otherwise provided above, NMP LLC shall take no action to void, cancel, or modify any intercompany credit agreement referenced above, without the prior written consent of the Director of the Office of Nuclear Reactor Regulation. ( 13) Mitigation Strategy License Condition Exelon Generation shall develop and maintain strategies for addressing large fires and explosions and that include the following key areas: a. Fire fighting response strategy with the following elements: (1) Pre-defined coordinated fire response strategy and guidance (2) Assessment of mutual aid fire fighting assets (3) Designated staging areas for equipment and materials (4) Command and control (5) Training of response personnel b. Operations to mitigate fuel damage considering the following: (1) Protection and use of personnel assets (2) Communications (3) Minimizing fire spread (4) Procedures for implementing integrated fire response strategy (5) Identification of readily-available pre-staged equipment (6) Training on integrated fire response strategy (7) Spent fuel pool mitigation measures Renewed License No. DPR-63 Revised by letter dated August 23, 2007 Amendment No. 214

-10-c. Actions to minimize release to include consideration of: (1) Water spray scrubbing (2) Dose to onsite responders (14) Exelon Generation shall implement and maintain all Actions required by Attachment 2 to NRC Order EA-06-137, issued June 20, 2006, except the last action that requires incorporation of the strategies into the site security plan, contingency plan, emergency plan and/or guard training and qualification plan, as appropriate. (15) Upon implementation of Amendment No. 195 adopting TSTF-448, Revision 3, the determination of control room envelope (CRE) unfiltered air inleakage as required by TS 4.4.5.g, in accordance with TS 6.5.8.c.(i), the assessment of CRE habitability as required by Specification 6.5.8.c.(ii), and the measurement of CRE pressure as required by Specification 6.5.8.d, shall be considered met. Following implementation: (a) The first performance of TS 4.4.5.g, in accordance with Specification 6.5.8.c.(i), shall be within the specified Frequency of 6 years plus the 18-month allowance of TS 4.0.2, as measured from February 19, 2004, the date of the most recent tracer gas test, as stated in the January 31, 2005 letter response to Generic Letter 2003-01, or within the next 18 months if the time period since the most recent tracer gas test is greater than 6 years. (b) The first performance of the periodic assessment of CRE habitability, Specification 6.5.8.c.(ii), shall be within 3 years, plus the 9-month allowance of TS 4.0.2, as measured from February 19, 2004, the date of the most recent tracer gas test, as stated in the January 31, 2005 letter response to Generic Letter 2003-01, or within the next 9 months if the time period since the most recent tracer gas test is greater than 3 years. (c) The first performance of the periodic measurement of CRE pressure, Specification 6.5.8.d, shall be within 24 months, plus the 182 days allowed by TS 4.0.2, as measured from March 1, 2007, the date of the most recent successful pressure measurement test, or within the next 182 days if not performed previously. (16) The existing E.D.F. International S.A.S. Support Agreement of approximately $145 million, dated November 6, 2009, may not be amended or modified without 30 days prior written notice to the Director of the Office of Nuclear Reactor Regulation or his designee. Nine Mile Point Nuclear Station, LLC, CENG or Exelon Generation shall not take any action to cause E. D. F. International S.A.S., or its successors and assigns, to void, cancel, or materially modify the E. D. F. International S.A.S. Support Agreement or cause it to fail to perform, or impair its performance under the E. D. F. International S.A.S. Support Agreement, without the prior written consent of the NRC. Exelon Generation shall inform the NRC in writing no later than 14 days after any funds are provided to or for the CENG subsidiary licensee under the E.D.F. International S.A.S. Support Agreement. Renewed License No. DPR-63 Revised by letter dated August 23, 2007 Amendment No. +9&,-214

-11 -(17) Exelon Corporation shall, no later than the time the license transfers occur, enter into a Support Agreement of approximately $245 million with the licensee. The Exelon Corporation Support Agreement shall supersede the Support Agreement provided by Exelon Generation, dated March 12, 2012, in all respects and shall be consistent with the representations contained in the August 6, 2013 transfer application. Nine Mile Point Nuclear Station, LLC, CENG or Exelon Generation shall not take any action to cause Exelon Corporation, or its successors and assigns, to void, cancel, or materially modify the Exelon Corporation Support Agreement or cause it to fail to perform, or impair its performance under the Exelon Corporation Support Agreement, without the prior written consent of the NRC. The Exelon Corporation Support Agreement may not be amended or modified without 30 days prior written notice to the Director of the Office of Nuclear Reactor Regulation or his designee. An executed copy of the Exelon Corporation Support Agreement shall be submitted to the NRC no later than 30 days after the completion of the proposed transaction and license transfers. Exelon Generation shall inform the NRC in writing no later than 14 days after any funds are provided to or for the licensee under the Exelon Corporation Support Agreement. (18) Exelon Corporation shall, no later than the time the license transfers occur, provide a parent guarantee in the amount of $165 million to ensure a source of funds for the facility in the event that the existing cash pool between the licensee and CENG is insufficient to cover operating costs. The existing CENG cash pool arrangement shall be consistent with the representations contained in the 2009 Transfer Application dated January 22, 2009 (ADAMS Accession No. ML0902901 01 ). Nine Mile Point Nuclear Station, LLC, CENG or Exelon Generation shall not take any action to cause Exelon Corporation, or its successors and assigns, to void, cancel or materially modify the parent gua*rantee or cause it to fail to perform, or impair its performance under the parent guarantee without the prior written consent of the NRC. (19) Within 14 days of the license transfers, Exelon Generation shall submit to the NRC the Nuclear Operating Services Agreement reflecting the terms set forth in the application dated August 6, 2013. Section 7.1 of the Nuclear Operating Services Agreement may not be modified in any material respect related to financial arrangements that would adversely impact the ability of the licensee to fund safety-related activities authorized by the license without the prior written consent of the Director of the Office of Nuclear Reactor Regulation. (20) Within 10 days of the license transfers, Exelon Generation shall submit to the NRC the amended CENG Operating Agreement reflecting the terms set forth in the application dated August 6, 2013. The amended and restated Operating Agreement may not be modified in any material respect concerning decision making authority over safety, security and reliability without the prior written consent of the Director of the Office of Nuclear Reactor Regulation. (21) At least half the members of the CENG Board of Directors must be U.S. citizens. Renewed License No. DPR-63 Amendment No. 214

-12-(22) The CENG Chief Executive Officer, Chief Nuclear Officer, and Chairman of the CENG Board of Directors must be U.S. citizens. These individuals shall have the responsibility and exclusive authority to ensure and shall ensure that the business and activities of CENG with respect to the facility's license are at all times conducted in a manner consistent with the public health and safety and common defense and security of the United States. (23) CENG will retain its Nuclear Advisory Committee (NAG) composed of U.S. citizens who are not officers, directors, or employees of CENG, EDF Inc., Constellation Nuclear, LLC, or CE Nuclear, LLC. The NAG will report to, and provide transparency to, the NRC and other U.S. governmental agencies regarding foreign ownership and control of nuclear operations. (24) The NAG shall prepare an annual report regarding the status of foreign ownership, control, or domination of the licensed activities of power reactors under the control, in whole or part, of CENG. The NAG report shall be submitted to the NRC within 30 days of completion, or by January 31 of each year (whichever occurs first). No action shall be taken by CENG or any entity to cause Constellation Nuclear, LLC, Exelon Generation, or their parent companies, subsidiaries or successors to modify the NAG report before submission to the NRC. The NAG report shall be made available to the public, with the potential exception of information that meets the requirements tor withholding such information from public disclosure under the regulations of 10 CFR 2.390, "Public Inspections, Exemptions, Requests tor Withholding." Renewed License No. DPR-63 Amendment No. 214

-13-E. This license is effective as of the date of issuance and shall expire on August 22, 2029. F. The UFSAR supplement, as revised, submitted pursuant to 10 CFR 54.21(d), shall be included in the next scheduled update to the UFSAR required by 10 CFR 50.71 (e)(4) following the issuance of this renewed operating license. Until that update is complete, the licensee may make changes to the programs and activities described in the supplement without prior Commission approval, provided that the licensee evaluates such changes pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section. G. The UFSAR supplement, as revised, describes certain future activities to be completed prior to the period of extended operation. The licensee shall complete these activities in accordance with the schedule in Appendix A of NUREG-1900, "Safety Evaluation Report Related to the License Renewal of Nine Mile Point Nuclear Station, Units 1 and 2", dated September 2006, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection. H. All capsules in the reactor vessel that are removed and tested must meet the test procedures and reporting requirements of the most recent NRC-approved version of the Boiling Water Reactor Vessels and Internals Project (BWRVIP) Integrated Surveillance Program (ISP) appropriate for the configuration of the specimens in the capsule. All capsules placed in storage must be maintained for future insertion. Any changes to storage requirements must be approved by the NRC, as required by 10 CFR Part 50, Appendix H. FOR THE NUCLEAR REGULATORY COMMISSION Original Signed by J. E. Dyer, Director Office of Nuclear Reactor Regulation

Enclosure:

Appendix A -Technical Specifications Date of Issuance: October 31 , 2006

b. The emergency operating procedures required to implement the requirements of NUREG-0737 and NUREG-0737, Supplement 1, as stated in Generic Letter 82-33; c. Quality assurance for radioactive effluent and radiological environmental monitoring; and d. Deleted e. All programs specified in Specification 6.5. 6.5 Programs and Manuals The following programs shall be established, implemented, and maintained. 6.5.1 Offsite Dose Calculation Manual (ODCM) a. The ODCM shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the radiological environmental monitoring program; and b. The ODCM shall also contain the radioactive effluent controls and radiological environmental monitoring activities, and descriptions of the information that should be included in the Annual Radiological Environmental Operating, and Radioactive Effluent Release Reports required by Specification 6.6.2 and Specification 6.6.3. c. Licensee initiated changes to the ODCM: 1. Shall be documented and records of reviews performed shall be retained. This documentation shall contain: (a) Sufficient information to support the change(s) together with the appropriate analyses or evaluations justifying the change(s), and (b) A determination that the change(s) maintain the levels of radioactive effluent control required by 10 CFR 20.1302, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I, and do not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations; AMENDMENT NO. 142. 181,215 350 ENCLOSURE 2 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION TRANSITION TO A RISK-INFORMED, PERFORMANCE-BASED FIRE PROTECTION PROGRAM IN ACCORDANCE WITH 10 CFR 50.48(c) AMENDMENT NO. 215 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-63 NINE MILE POINT NUCLEAR STATION, LLC EXELON GENERATION COMPANY, LLC NINE MILE POINT NUCLEAR STATION, UNIT NO.1 DOCKET NO. 50-220 TABLE OF CONTENTS SAFETY EVALUATION Contents

1.0 INTRODUCTION

......................................................................................................... 1.1 Background ............................................................................................................... 1.2 Requested Licensing Action ....................................................................................

2.0 REGULATORY EVALUATION

................................................................................... 2.1 Applicable Regulations ............................................................................................. 2.2 Applicable Staff Guidance ........................................................................................ 2.3 NFPA 805 Frequently Asked Questions ................................................................ -15-2.4 Orders, License Conditions, and Technical Specifications ................................. -18 -2.4.1 Orders ....................................................................................................................... -18 -2.4.2 License Conditions .................................................................................................... -19 -2.4.3 Technical Specifications ............................................................................................ -19 -2.4.4 Updated Safety Analysis Report. ............................................................................... -20 -2.5 Rescission of Exemptions ...................................................................................... -20-2.6 Self-Approval Process for FPP Changes (Post-Transition) .................................. -22-2.6.1 Post-Implementation Plant Change Evaluation Process ............................................ -22 -2.6.2 Requirements for the Self-Approval Process Regarding Plant Changes .................... -25 -2.7 Modifications and Implementation Items ............................................................... -27-2.7.1 Modifications ............................................................................................................. -27-2.7.2 Implementation Items ................................................................................................ -28-2.7.3 Schedule ................................................................................................................... -28-

3.0 TECHNICAL EVALUATION

..................................................................................... -29-3.1 NFPA 805 Fundamental FPP and Design Elements .............................................. -30-3.1.1.1 Compliance Strategy--Complies ......................................................................... -33-3.1.1.2 Compliance Strategy --Complies with Clarification ............................................... -35 -3.1.1.3 Compliance Strategy--Complies with Use of EEEEs ........................................... -35-3.1.1.4 3.1.1.5 3.1.1.6 3.1.1.7 3.1.1.8 3.1.1.9 ii Compliance Strategy--Complies via Previous NRC Approval. ............................. -40-Compliance Strategy --Submit for NRC Approval ................................................ -41 -Compliance Strategy-Complies, with item for implementation ............................ -41 -Compliance Strategy-Multiple Strategies ............................................................ -45-Chapter 3 Sections Not Reviewed ........................................................................ -45 -Compliance with Chapter 3 Requirements Conclusion ......................................... -46 -3.1.2 Identification of the Power Block ................................................................................ -47-3.1.3 Closure of Generic Letter 2006-03, "Potentially Nonconforming HemycŽ and MTŽ Fire Barrier Configurations," Issues ........................................................... -47 -3.1.4 Performance-Based Methods for NFPA 805, Chapter 3 Elements ............................ -47-3.1.4.1 NFPA 805, Section 3.3.5.1 -Electrical Wiring Above Suspended Ceiling .................................................................................................................. -48 -3.1.4.2 NFPA 805, Section 3.2.3(1) -Inspection, Testing, and Maintenance Procedures ........................................................................................................... -50-3.2 Nuclear Safety Capability Assessment (NSCA) Methods ..................................... -51-3.2.1 Compliance with NFPA 805 NSCA Methods ............................................................. -53-3.2.1.1 Attribute Alignment -Aligns ................................................................................... -54 -3.2.1.2 Attribute Alignment-Aligns with Intent.. ................................................................ -55 -3.2.1.3 3.2.1.4 3.2.1.5 3.2.1.6 Attribute Alignment -Not in Alignment, but Prior NRC Approval ........................... -58 -Attribute Alignment -Not in Alignment, but No Adverse Consequences ................ -58 -Attribute Alignment-Not in Alignment .................................................................. -58 -NFPA 805 Nuclear Safety Capability Assessment Methods Conclusion ............... -59-3.2.2 Maintaining Fuel in a Safe and Stable Condition ...................................................... -59 -3.2.3 Applicability of Feed and Bleed ................................................................................. -61 -3.2.4 Assessment of Multiple Spurious Operations ............................................................ -61 -3.2.5 Establishing Recovery Actions .................................................................................. -62 -3.2.6 3.3 3.4 3.4.1 3.4.1.1 3.4.1.2 Conclusion for Section 3.2 ........................................................................................ -64-Fire Modeling ........................................................................................................... -65-Fire Risk Evaluations .............................................................................................. -65 -Maintaining Defense-in-Depth and Safety Margins .................................................... -66-Defense-in-Depth (DID) ........................................................................................ -66 -Safety Margins ...................................................................................................... -67 -3.4.2 Quality of the Fire Probabilistic Risk Assessment.. .................................................... -68 -

3.4.2.1 3.4.2.2 3.4.2.3 3.4.2.4 iii Internal Events PRA Model ................................................................................... -69 -Fire PRA Model .................................................................................................... -70-Fire Modeling in Support of the Development of the Fire Risk Evaluations (FREs) ............................................................................................... -77 -Conclusions Regarding Fire PRA Quality ............................................................. -84 -3.4.3 Fire Risk Evaluations ................................................................................................. -85 -3.4.4 Additional Risk Presented by Recovery Actions ........................................................ -86-3.4.5 Risk-Informed or Performance-Based Alternatives to Compliance with NFPA 805 ................................................................................................................. -88-3.4.6 Cumulative Risk and Combined Changes ................................................................. -88 -3.4.7 Uncertainty and Sensitivity Analyses ......................................................................... -89 -3.4.8 Conclusion for Section 3.4 ......................................................................................... -90-3.5 Nuclear Safety Capability Assessment Results .................................................... -91 -3.5.1 Nuclear Safety Capability Assessment Results by Fire Area ..................................... -92 -3.5.1.1 Fire Detection and Suppression Systems Required to meet the NSPC ................. -94 -3.5.1.2 Evaluation of Fire Suppression Effects on NSPC .................................................. -97-3.5.1.3 Licensing Actions .................................................................................................. -98-3.5.1.4 Existing Engineering Equivalency Evaluations (EEEEs) ....................................... -98-3.5.1.5 Variances from Deterministic Requirements ......................................................... -99 -3.5.1.6 Recovery Actions .................................................................................................. -99 -3.5.1.7 Recovery Actions Credited for Defense-in-Depth ................................................ -100-3.5.1.8 Plant Fire Barriers and Separations .................................................................... -100 -3.5.1.9 Electrical Raceway Fire Barrier Systems (ERFBS) ............................................. -100-3.5.1.1 0 Conclusion for Section 3.5.1 ............................................................................... -101 -3.5.2 Clarification of Prior NRC Approvals ........................................................................ -102 -3.5.3 Fire Protection during Non-Power Operational Modes ............................................. -102 -3.5.3.1 NPO Strategy and Plant Operational States (POSs) ........................................... -103-3.5.3.2 NPO Analysis Process ........................................................................................ -106 -3.5.3.3 NPO Key Safety Functions and SSCs Used to Achieve Performance ................ -107 -3.5.3.4 NPO Pinch Point Resolutions and Program Implementation ............................... -108-3.5.4 Conclusion for Section 3.5 ....................................................................................... -108-3.6 Radioactive Release Performance Criteria .......................................................... -109 -3. 7 NFPA 805 Monitoring Program ............................................................................. -111 -

iv 3. 7.1 Conclusion for Section 3. 7 ....................................................................................... -113 -3.8 Program Documentation, Configuration Control, and Quality Assurance .............................................................................................................. -113-3.8.1 Documentation ........................................................................................................ -115-3.8.2 Configuration Control .............................................................................................. -116 -3.8.3 Quality ..................................................................................................................... -116 -3.8.3.1 Review ................................................................................................................ -116 -3.8.3.2 Verification and Validation (V&V) ........................................................................ -117-3.8.3.4 Qualification of Users .......................................................................................... -122 -3.8.3.5 Uncertainty Analysis ........................................................................................... -124-3.8.3.6 Conclusion for Section 3.8.3 ............................................................................... -127 -3.8.4 Fire Protection Quality Assurance Program ............................................................. -127-3.8.5 Conclusion for Section 3.8 ....................................................................................... -128 -4.0 FIRE PROTECTION LICENSE CONDITION ........................................................... -128 -5.0 SUMMARY ............................................................................................................. -131-

6.0 STATE CONSULTATION

....................................................................................... -131 -

7.0 ENVIRONMENTAL CONSIDERATION

.................................................................. -131 -

8.0 CONCLUSION

........................................................................................................ -132 -9.0 REFERENCES ........................................................................................................ -133 -ATTACHMENTS Attachment A: Table 3.8-1, V&V Basis for Fire Modeling Correlations Used at NMP1 ................................................................................................. -A1 -Attachment 8: Table 3.8-2, V&V Basis for Other Fire Models and Related Correlations Used at NMP1 .................................................................. -81 -Attachment C: Abbreviations and Acronyms ................................................................. -C1 -

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 215 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-63 FOR THE TRANSITION TO A PERFORMANCE-BASED FIRE PROTECTION PROGRAM IN ACCORDANCE WITH 10 CFR 50.48(c) NINE MILE POINT NUCLEAR STATION, LLC EXELON GENERATION COMPANY, LLC NINE MILE POINT NUCLEAR STATION, UNIT 1 DOCKET NO. 50-220

1.0 INTRODUCTION

1.1 Background The U.S. Nuclear Regulatory Commission (NRC) started developing fire protection requirements in the 1970s, and in 1976, the NRC published comprehensive fire protection guidelines in the form of Branch Technical Position (BTP) APCSB 9.5-1, "Guidelines for Fire Protection for Nuclear Power Plants" (Agencywide Documents Access and Management System (ADAMS) Accession No. ML070660461 ), and Appendix A to BTP APCSB 9.5-1, "Guidelines for Fire Protection for Nuclear Power Plants Docketed Prior to July 1, 1976" (ADAMS Accession No. ML070660458). Subsequently, the NRC performed fire protection reviews for the operating reactors, and documented the results in safety evaluation reports (SERs) or supplements to SERs. In 1980, to resolve issues identified in those reports, the NRC amended its regulations for fire protection in operating nuclear power plants and published its Final Rule, Fire Protection Program for Operating Nuclear Power Plants, in the Federal Register (FR) on November 19, 1980 (45 FR 76602), adding Section 50.48, "Fire Protection," and Appendix R to Title 10 of the Code of Federal Regulations (1 0 CFR) Part 50, "Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979." Section 50.48(a)(1) of 10 CFR Part 50 requires each holder of an operating license, and holders of a combined operating license issued under Part 52 to have a fire protection plan that satisfies General Design Criterion (GDC) 3 of Appendix A to 10 CFR Part 50 and states that the fire protection plan must describe the overall fire protection program; identify the positions responsible for the program and the authority delegated to those positions; outline the plans for fire protection, fire detection and suppression capability, and limitation of fire damage. Section 50.48(a)(2) states that the fire protection plan must describe the specific features necessary to implement the program described in paragraph (a)(1) including administrative controls and personnel requirements; automatic and manual fire detection and suppression systems; and the means to limit fire damage to structures, systems, and components (SSCs) to ensure the capability to safely shut down the plant. Section 50.48(a)(3) requires that the licensee retain the fire protection plan and each change to the plan as a record until the Commission terminates the license and that the licensee retain each superseded revision of the procedures for 3 years. In the 1990s, the NRC worked with the National Fire Protection Association (NFPA) and industry to develop a risk-informed (RI), performance-based (PB) consensus standard for fire protection. In 2001, the NFPA Standards Council issued NFPA 805, "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants" (Reference 1 ), which describes a methodology for establishing fundamental fire protection program (FPP) design requirements and elements, determining required fire protection systems and features, applying PB requirements, and administering fire protection for existing light-water reactors during operation, decommissioning, and permanent shutdown. It provides for the establishment of a minimum set of fire protection requirements, but allows PB or deterministic approaches to be used to meet performance criteria. Regulatory Guide (RG) 1.205, "Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants," Revision 1 (RG 1.205) (Reference 2), states, in part, that: On March 26, 1998, the staff sent to the Commission SECY -98-058, "Development of a Risk-Informed, Performance-Based Regulation for Fire Protection at Nuclear Power Plants" [Reference 3], in which it proposed to work with NFPA and the industry to develop a risk-informed, performance-based consensus standard for nuclear power plant fire protection. This consensus standard could be endorsed in a future rulemaking as an alternative set of fire protection requirements to the existing regulations in 10 CFR 50.48. In SECY-00-0009, "Rulemaking Plan, Reactor Fire Protection Risk-Informed, Performance-Based Rulemaking," dated January 13, 2000 [Reference 4], the NRC staff requested and received Commission approval to proceed with a rulemaking to permit reactor licensees to adopt NFPA 805 as an alternative to existing fire protection requirements. On February 9, 2001, the NFPA Standards Council approved the 2001 edition of NFPA 805 as an American National Standard for performance-based fire protection for light-water nuclear power plants. A licensee that elects to adopt NFPA 805 must meet the performance goals, objectives, and criteria that are itemized in Chapter 1 of NFPA 805 through the implementation of PB or deterministic approaches. The goals include ensuring that reactivity control, inventory and pressure control, decay heat removal, vital auxiliaries, and process monitoring are achieved and maintained. The licensee then must establish plant fire protection requirements using the methodology in Chapter 2 of NFPA 805, such that the minimum FPP elements and design criteria contained in Chapter 3 of NFPA 805 are satisfied. Next, the licensee identifies fire areas and fire hazards through a plant-wide analysis, and then applies either a PB or a deterministic approach to meet the performance criteria. As part of a PB approach, the licensee will use engineering evaluations, probabilistic safety assessments, and fire modeling calculations to show that the criteria are met. Chapter 4 of NFPA 805 establishes the methodology to determine the fire protection systems and features required to achieve the performance criteria. It also specifies that at least one success path to achieve the nuclear safety performance criteria shall be maintained free of fire damage by a single fire. RG 1.205 also states, in part, that: Effective July 16, 2004, the Commission amended its fire protection requirements in 10 CFR 50.48 to add 10 CFR 50.48{c), which incorporates by reference the 2001 edition of NFPA 805, with certain exceptions, and allows licensees to apply for a license amendment to comply with the 2001 edition of NFPA 805 (69 FR 33536). NFPA has issued subsequent editions of NFPA 805, but the regulation does not endorse them. Throughout this safety evaluation (SE), where the NRC staff states that the licensee's FPP element is in compliance with (or meets the requirements of) NFPA 805, the NRC staff is referring to NFPA 805 with the exceptions, modifications, and supplements described in 10 CFR 50.48{c)(2). RG 1.205 also states, in part, that: In parallel with the Commission's efforts to issue a rule incorporating the risk-informed, performance-based fire protection provisions of NFPA 805, NEI [the Nuclear Energy Institute] published implementing guidance for the specific provisions of NFPA 805 and 10 CFR 50.48(c) in NEI 04-02, ["Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program Under 10 CFR 50.48(c)."] RG 1.205 provides the NRC staff's position on NEI 04-02, Revision 2 (Reference 5), and offers additional information and guidance to supplement the NEI document and assist licensees in meeting the NRC's regulations in 10 CFR 50.48(c) related to adopting a risk-informed, performance-based (RI/PB) FPP. RG 1.205 endorses the guidance of NEI 04-02, Rev. 2, subject to certain exceptions, as providing methods acceptable to the staff for adopting an FPP consistent with the 2001 edition of NFPA 805 and 10 CFR 50.48(c). Accordingly, Nine Mile Point Nuclear Station, LLC (NMPNS, the licensee), requested a license amendment to allow the licensee to revise the Nine Mile Point Nuclear Station, Unit 1 (NMP1 ), FPP in accordance with 10 CFR 50.48{c) and change the license and technical specifications {TSs) accordingly. By Order dated March 25, 2014, the NRC approved the direct license transfer of the operating authority from Constellation Energy Nuclear Group, LLC (CENG) to Exelon Generation Company, LLC (Exelon, the licensee), and a conforming license amendment for NMP1. Exelon remained the parent company of CENG, became a co-licensee with NMPNS and the operator of NMP1 and NMP2 (79 FR18322, April1, 2014). After the NRC approved the direct transfer of the operating authority for NMP1, by letter dated March 28,2014 (ADAMS Accession No. ML14087A274), Exelon noted its subsidiaries pending actions before the NRC regarding Calvert Cliffs Nuclear Power Plant, LLC, Nine Mile Point Nuclear Station, LLC, and R. E. Ginna Nuclear Power Plant, LLC, and stated that: Prior to the license transfers, CENG made docketed submittals to the NRC that requested specific licensing actions, such as license amendment requests, relief requests, exemption requests, etc. Furthermore, in the application for the license transfers, Exelon stated that upon transfer of the licenses, Exelon would assume all current regulatory commitments made for these units. Accordingly, Exelon hereby adopts and endorses those docketed requests currently before the NRC for review and approval. Exelon requests that the NRC continue to process those pending actions on the schedules previously requested by CENG. Accordingly, as used in this SE, the term licensee refers to both NMPNS and Exelon, unless otherwise stated. 1.2 Requested Licensing Action By letter to the NRC dated June 11, 2012 (Reference 6), as supplemented by letters dated February 27, 2013 (Reference 7), March 27, 2013 (Reference 8), April 30, 2013 (Reference 9), December 9, 2013 (Reference 1 0), January 22, 2014 (Reference 11 ), March 14, 2014 (Reference 12), April 15, 2014 (Reference 13), May 09, 2014 (Reference 14), and May 23, 2014 (Reference 92), the licensee submitted an application for a license amendment to transition the NMP1 FPP from 10 CFR 50.48(b) to 10 CFR 50.48(c), NFPA 805, "Performance-Based Standard for Fire Protection For Light Water Reactor Electric Generating Plants," 2001 Edition. The supplemental letters were in response to the NRC staff's requests for additional information (RAis) dated January 3, 2013 (Reference 15), October 9, 2013 (Reference 16), and February 12, 2014 (Reference 17). The licensee's supplemental letters dated February 27, March 27, April 30, and December 9, 2013, and January 22, March 14, April 15, May 9, and May 23, 2014, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register (FR) on September 11 , 2012 (77 FR 5587 4). The licensee requested an amendment to the NMP1 renewed operating license and TSs in order to establish and maintain an RI/PB FPP in accordance with the requirements of 10 CFR 50.48(c). Specifically, the licensee requested to transition from the existing deterministic fire protection licensing basis -established in accordance with all provisions of the approved FPP as described in the NMP1 Updated Final Safety Analysis Report (UFSAR) and as approved in the Fire Protection Safety Analysis Report dated July 26, 1979 (Reference 18); in the fire protection Exemption issued March 21, 1983 (Reference 19); and by SEs dated July 22, 1980 (Reference 20), July 30, 1980 (Reference 21 ), December 15, 1989 (Reference 22), March 3, 1983 (Reference 23), April 1, 1985 (Reference 24), August 6, 1986 (Reference 25); and in the exemptions described by the NMP1 License Condition 2.D(7) letter dated March 21, 1983 (Reference 19), -to an RI/PB FPP in accordance with 10 CFR 50.48(c), that uses risk information, in part, to demonstrate compliance with the fire protection and nuclear safety goals, objectives, and performance criteria of NFPA 805. As such, the proposed FPP at NMP1 is referred to as RI/PB throughout this SE. In its license amendment request (LAR), the licensee has provided a description of the revised FPP for which it is requesting NRC approval to implement, a description of the FPP that it will implement under 10 CFR 50.48(a) and (c), and the results of the evaluations and analyses required by NFPA 805. This SE documents the NRC staff's evaluation of the licensee's LAR and the NRC staff's conclusion that: 1) The licensee has identified any orders and license conditions that must be revised or superseded, and has provided the necessary revisions to the plant's Technical Specifications (TSs) and TS Bases, as required by 10 CFR 50.48(c)(3)(i); 2) The licensee has completed its implementation of the methodology in Chapter 2, "Methodology," of NFPA 805 (including all required evaluations and analyses), and the NRC staff has approved the licensee's modified FPP, which reflects the decision to comply with NFPA 805, as required by 10 CFR 50.48(a); and 3) The licensee will modify its FPP, as described in the LAR, in accordance with the implementation schedule set forth in this SE and the accompanying license condition, as required by 10 CFR 50.48(c)(3)(ii). The licensee proposed a new fire protection license condition reflecting the new RI/PB FPP licensing basis, as well as revisions to the TSs that address this change to the current FPP licensing basis. Sections 2.4.2 and 4.0 of this SE discuss in detail the license condition, and Section 2.4.3 discusses the TS changes.

2.0 REGULATORY EVALUATION

The following explains the use of general design criteria for NMP1. The construction permit for NMP1 was issued by the Atomic Energy Commission (AEC) on April 12, 1965, and the operating license was issued on December 26, 197 4. The plant design criteria for NMP1 are listed in the Updated Final Safety Analysis Report (UFSAR)Section I, "Principal Design Criteria." The AEC published the final rule that added Appendix A to Title 10 of the Code of Federal Regulations (1 0 CFR) Part 50, "General Design Criteria for Nuclear Power Plants," in the Federal Register (36 FR 3255) on February 20, 1971, with the rule effective on May 21, 1971. In accordance with an NRC staff requirements memorandum from S. J. Chilk to J. M. Taylor, "SECY-92-223-Resolution of Deviations Identified During the Systematic Evaluation Program," dated September 18, 1992 (ADAMS Accession No. ML003763736), the Commission decided not to apply the final GOG to plants with construction permits issued prior to May 21, 1971, which includes NMP1. NMP1 was not licensed to the 10 CFR 50, Appendix A GOG. The NMP1 UFSAR provides an assessment against the GOG in Table 1-1. This UFSAR table refers to the NMP1 Technical Supplement to Petition for Conversion from Provisional Operating License to Full-Term Operating License, July 1972, for the details of the assessment against the GDC current at that time. The current Fire Protection Program for Nine Mile Point Unit 1 is described in UFSAR Chapter X, Section K. Safe Shutdown Analysis described in Appendix 1 OB of the UFSAR provides an evaluation to assure that the facility can be shutdown given worst case fire damage in any single fire area. The analysis also provides a summary of the information for the review of design modifications to determine that compliance with 1 0 CFR Part 50, Appendix R is maintained. The Fire Protection Rule {10 CFR 50.48 and 10 CFR Part 50, Appendix R), effective February 17, 1982, set forth fire protection features required to satisfy 1 0 CFR Part 50, Appendix A, General Design Criteria (GDC) 3. Therefore, for the review of this LAR, the NRC staff applied the GDC requirements. Section 50.48, "Fire Protection," of 10 CFR provides the NRC requirements for nuclear power plant fire protection. The NRC regulations include specific requirements for requesting approval for an RI/PB FPP based on the provisions of NFPA 805 (Reference 1 ). Paragraph 50.48(c)(3)(i) of 10 CFR states, in part, that: A licensee may maintain a fire protection program that complies with NFPA 805 as an alternative to complying with [1 0 CFR 50.48(b)] for plants licensed to operate before January 1, 1979, or the fire protection license conditions for plants licensed to operate after January 1, 1979. The licensee shall submit a request to comply with NFPA 805 in the form of an application for license amendment under [1 0 CFR] 50.90. The application must identify any orders and license conditions that must be revised or superseded, and contain any necessary revisions to the plant's technical specifications and the bases thereof. In addition, 10 CFR 50.48(c)(3)(ii) states that: The licensee shall complete its implementation of the methodology in Chapter 2 of NFPA 805 (including all required evaluations and analyses) and, upon completion, modify the fire protection plan required by paragraph (a) of this section to reflect the licensee's decision to comply with NFPA 805, before changing its fire protection program or nuclear power plant as permitted by NFPA 805. The intent of 10 CFR 50.48{ c)(3)(ii) is given in the statement of considerations for the Final Rule, Voluntary Fire Protection Requirements for Light Water Reactors; Adoption of NFPA 805 as a Risk-Informed, Performance-Based Alternative (69 FR 33536, 33548; June 16, 2004), which states: This paragraph requires licensees to complete all of the Chapter 2 methodology (including evaluations and analyses) and to modify their fire protection plan before making changes to the fire protection program or to the plant configuration. This process ensures that the transition to an NFPA 805 configuration is conducted in a complete, controlled, integrated, and organized manner. This requirement also precludes licensees from implementing NFPA 805 on a partial or selective basis (e.g., in some fire areas and not others, or truncating the methodology within a given fire area). As stated in 10 CFR 50.48(c)(3)(i), the Director of the Office of Nuclear Reactor Regulation (NRR), or a designee of the Director, may approve the application if the Director or designee determines that the licensee has identified orders, license conditions, and the TSs that must be revised or superseded, and that any necessary revisions are adequate. The regulations also allow for flexibility that was not included in the NFPA 805 standard. Licensees who choose to adopt 10 CFR 50.48(c), but wish to use the PB methods permitted elsewhere in the standard to meet the fire protection requirements of NFPA 805 Chapter 3, "Fundamental Fire Protection Program and Design Elements," must submit a LAR to obtain approval in accordance with 10 CFR 50.48(c)(2)(vii). This regulation further provides that: The Director of the Office of Nuclear Reactor Regulation, or a designee of the Director, may approve the application if the Director or designee determines that the performance-based approach; (A) Satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release; (B) Maintains safety margins; and (C) Maintains fire protection defense-in-depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire safe shutdown capability). Alternatively, licensees who want to use Rl or PB alternatives to comply with NFPA 805 must obtain approval by submitting aLAR as required in 10 CFR 50.48(c)(4). This regulation further provides that: The Director of the Office of Nuclear Reactor Regulation, or designee of the Director, may approve the application if the Director or designee determines that the proposed alternatives: (i) Satisfy the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release; (ii) Maintain safety margins; and (iii) Maintain fire protection defense-in-depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire safe shutdown capability). In addition to the conditions outlined by the rule that require licensees to submit an LAR for NRC review and approval in order to adopt an RI/PB FPP, a licensee may submit additional elements of its FPP for which it wishes to receive specific NRC review and approval, as set forth in Regulatory Position C.2.2.1 of RG 1.205 (Reference 2). Inclusion of these elements in the NFPA 805 LAR is meant to alleviate uncertainty in portions of the current FPP licensing bases as a result of the lack of specific NRC approval of these elements. Regulatory guides are not substitutes for regulations, and compliance with them is not required. Methods and solutions that differ from those set forth in regulatory guides will be deemed acceptable if they provide a basis for the findings required for the issuance or continuance of a permit or license by the Commission. Accordingly, any submittal addressing these additional FPP elements needs to include sufficient detail to allow the NRC staff to assess whether the licensee's treatment of these elements meets 10 CFR 50.48(c) requirements. The purpose of the FPP established by NFPA 805 is to provide assurance, through a in-depth (DID) philosophy, that the NRC's fire protection objectives are satisfied. NFPA 805 Section 1.2, "Defense-in-Depth," states the following: Protecting the safety of the public, the environment, and plant personnel from a plant fire and its potential effect on safe reactor operations is paramount to this standard. The fire protection standard shall be based on the concept of defense-in-depth. Defense-in-depth shall be achieved when an adequate balance of each of the following elements is provided: (1) Preventing fires from starting (2) Rapidly detecting fires and controlling and extinguishing promptly those fires that do occur, thereby limiting fire damage (3) Providing an adequate level of fire protection for structures, systems and components important to safety, so that a fire that is not promptly extinguished will not prevent essential plant safety functions from being performed In addition, in accordance with GDC 3, "Fire protection," of Appendix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50, fire detection and fighting systems must be designed such that their rupture or inadvertent operation does not significantly impair the ability of the SSCs important to safety to perform their intended safety functions. 2.1 Applicable Regulations The following regulations address fire protection:

GDC 3, "Fire protection," to 1 0 CFR Part 50, Appendix A: Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions. Noncombustible and heat resistant materials shall be used wherever practical throughout the unit, particularly in locations such as the containment and control room. Fire detection and fighting systems of appropriate capacity and capability shall be provided and designed to minimize the adverse effects of fires on structures, systems, and components important to safety. Firefighting systems shall be designed to assure that their rupture or inadvertent operation does not significantly impair the safety capability of these structures, systems, and components.

GDC 5, "Sharing of structures, systems, and components," to 10 CFR Part 50, Appendix A: Structures, systems, and components important to safety shall not be shared among nuclear power units unless it can be shown that such sharing will not significantly impair their ability to perform their safety functions, including, in the event of an accident in one unit, an orderly shutdown and cooldown of the remaining units.

10 CFR 50.48(a)(1) requires that each holder of an operating license have a fire protection plan that satisfies GDC 3 of Appendix A to 10 CFR Part 50.

10 CFR 50.48(c) incorporates NFPA 805 (2001 Edition) by reference, with certain exceptions, modifications and supplementation. This regulation establishes the requirements for using an RI/PB FPP in conformance with NFPA 805 as an alternative to the requirements associated with 10 CFR 50.48(b) and Appendix R, "Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979," to 1 0 CFR Part 50, or the specific plant fire protection license condition.

10 CFR Part 20, "Standards for Protection Against Radiation," establishes the radiation protection limits used as NFPA 805 radioactive release performance criteria, as specified in NFPA 805, Section 1.5.2, "Radioactive Release Performance Criteria." 2.2 Applicable Staff Guidance The NRC staff review also relied on the following additional codes, regulatory guides, and standards:

RG 1.205, "Risk-Informed, Performance-Based Fire Protection for Existing Water Nuclear Power Plants," Revision 1, issued December 2009 (Reference 2), provides guidance for use in complying with the requirements that the NRC has promulgated for RI/PB FPPs that comply with 1 0 CFR 50.48 and the referenced 2001 Edition of the NFPA standard. It endorses portions of NEI 04-02, Revision 2 (Reference 5), where it has been found to provide methods acceptable to the NRC for implementing NFPA 805 and complying with 10 CFR 50.48(c). The regulatory positions in Section C of RG 1.205 include clarification of the guidance provided in NEI 04-02, as well as NRC exceptions to the guidance. RG 1.205 sets forth regulatory positions, emphasizes certain issues, clarifies the requirements of 10 CFR 50.48(c) and NFPA 805, clarifies the

-10-guidance in NEI 04-02, and provides exceptions to the NEI 04-02 guidance where required. Should a conflict occur between NEI 04-02 and this RG, the regulatory positions in RG 1.205 govern.

The 2001 Edition of NFPA 805, "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants" (Reference 1 ), specifies the minimum fire protection requirements for existing light-water nuclear power plants during all phases of plant operations, including shutdown, degraded conditions, and decommissioning, which had not been explicitly addressed by previous requirements and guidelines. NFPA 805 was developed to provide a comprehensive RI/PB standard for fire protection. The NFPA 805 Technical Committee on Nuclear Facilities is composed of nuclear plant licensees, the NRC, insurers, equipment manufacturers, and subject matter experts. The standard was developed in accordance with NFPA processes, and consisted of a number of technical meetings and reviews of draft documents by committee and industry representatives. The scope of NFPA 805 includes goals related to nuclear safety, radioactive release, life safety, and plant damage/business interruption. The standard addresses fire protection requirements for nuclear plants during all plant operating modes and conditions, including shutdown and decommissioning, which had not been explicitly addressed by previous requirements and guidelines. NFPA 805 became effective on February 9, 2001.

NEI 04-02, "Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program Under 10 CFR 50.48(c)," Revision 2 (Reference 5), provides guidance for implementing the requirements of 10 CFR 50.48(c), and represents methods for implementing in whole or in part an RI/PB FPP. This implementing guidance for NFPA 805 has two primary purposes: (1) to provide direction and clarification for adopting NFPA 805 as an acceptable approach to fire protection, consistent with 10 CFR 50.48(c); and (2) to provide additional supplemental technical guidance and methods for using NFPA 805 and its appendices to demonstrate compliance with fire protection requirements. Although there is a significant amount of detail in NFPA 805 and its appendices, clarification and additional guidance for select issues help ensure consistency and effective utilization of the standard. The NEI 04-02 guidance focuses attention on the RI/PB fire protection goals, objectives, and performance criteria contained in NFPA 805 and the RI/PB tools considered acceptable for demonstrating compliance. Revision 2 of NEI 04-02 incorporates guidance from RG 1.205 and approved Frequently Asked Questions (FAQs).

NEI 00-01, "Guidance for Post Fire Safe Shutdown Circuit Analysis," Revision 2, (Reference 26), provides a deterministic methodology for performing post-fire safe shutdown analysis (SSA). In addition, NEI 00-01 includes information on risk-informed methods (when allowed within a plant's licensing basis) that may be used in conjunction with the deterministic methods for resolving circuit failure issues related to Multiple Spurious Operations (MSOs). The risk-informed method is intended for application by licensees to determine the risk significance of identified circuit failure issues related to MSOs.

-11 -* RG 1.17 4, "An Approach for Using Probabilistic Risk Assessment in Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 2, issued May 2011 (Reference 27), provides the NRC staff's recommendations for using risk information in support of licensee-initiated licensing basis changes to a nuclear power plant that require such review and approval. The guidance provided does not preclude other approaches for requesting licensing basis changes. Rather, RG 1.17 4 is intended to improve consistency in regulatory decisions in areas in which the results of risk analyses are used to help justify regulatory action. As such, the RG provides general guidance concerning one approach that the NRC has determined to be acceptable for analyzing issues associated with proposed changes to a plant's licensing basis and for assessing the impact of such proposed changes on the risk associated with plant design and operation.

RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2, issued March 2009 (Reference 28), provides guidance to licensees for use in determining the technical adequacy of the base probabilistic risk assessment (PRA) used in a risk-informed regulatory activity, and endorses standards and industry peer review guidance. The RG provides guidance in four areas: ( 1) a definition of a technically acceptable PRA; (2) the NRC's position on PRA consensus standards and industry PRA peer review program documents; (3) demonstration that the baseline PRA (in total or specific pieces) used in regulatory applications is of sufficient technical adequacy; and (4) documentation needed to support a regulatory submittal. It does not provide guidance on how the base PRA is revised for a specific application or how the PRA results are used in application-specific making processes.

American Society of Mechanical Engineers/American Nuclear Society (ASME/ANS) RA-Sa-2009, "Addenda to ASME/ANS RA-S-2008, Standard for Levei1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications" (Reference 29), provides guidance for PRAs used to support Rl decisions for commercial light-water reactor nuclear power plants and prescribes a method for applying these requirements for specific applications. The Standard gives guidance for a Level 1 PRA of internal and external hazards for all plant operating modes. In addition, the Standard provides guidance for a limited Level 2 PRA sufficient to evaluate large early release frequency (LERF). The only hazards explicitly excluded from the scope are accidents resulting from purposeful human-induced security threats (e.g.,

-12-sabotage). The Standard applies to PRAs used to support applications of Rl decision-making related to design, licensing, procurement, construction, operation, and maintenance.

RG 1.189, "Fire Protection for Nuclear Power Plants," Revision 2, issued October 2009 (Reference 30), provides guidance to licensees on the proper content and quality of engineering equivalency evaluations used to support the FPP. The NRC staff developed the RG to provide a comprehensive fire protection guidance document and to identify the scope and depth of fire protection that the staff would consider acceptable for nuclear power plants.

NUREG-0800, Section 9.5.1.2, "Risk-Informed, Performance-Based Fire Protection Program," Revision 0, issued December 2009 (Reference 31 ), provides the NRC staff with guidance for evaluating LARs that seek to implement an RI/PB FPP in accordance with 10 CFR 50.48(c).

NUREG-0800, Section 19.1, "Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed License Amendment Requests After Initial Fuel Load," Revision 3, issued September 2012 (Reference 32), provides the NRC staff with guidance for evaluating the technical adequacy of a licensee's PRA results when used to request Rl changes to the licensing basis.

NUREG-0800, Section 19.2, "Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance," Revision 0, issued June 2007 (Reference 33), provides the NRC staff with guidance for evaluating the risk information used by a licensee to support permanent Rl changes to the licensing basis.

NUREG/CR-6850, "EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities," Volumes 1 and 2 and Supplement 1 (References 34, 35, and 36), presents a compendium of methods, data, and tools to perform a fire probabilistic risk assessment (FPRA) and develop associated insights. In order to address the need for improved methods, the NRC Office of Nuclear Regulatory Research (RES) and Electric Power Research Institute (EPRI) embarked upon a program to develop a state-of-art FPRA methodology. Both RES and EPRI provided specialists in fire risk analysis, fire modeling (FM), electrical engineering, human reliability analysis, and systems engineering for methods development. A formal technical issue resolution process was developed to direct the deliberative process between RES and EPRI. The process ensures that divergent technical views are fully considered, yet encourages consensus at many points during the deliberation. Significantly, the process provides that each party maintain its own point of view if consensus is not reached. Consensus was reached on all technical issues documented in NUREG/CR-6850. The methodology documented in this report reflects the current state-of-the-art in FPRA. These methods are expected to form a basis for risk-informed analyses related to the plant FPP. Volume 1, the Executive Summary, provides general background and

-13-overview information, project insights and conclusions. Volume 2 provides the detailed discussion of the recommended approach, methods, data, and tools for conduct of an FPRA.

Interim Technical Guidance provided in a Memorandum from Richard P. Correia, RES, to Joseph G. Giitter, NRR, titled "Interim Technical Guidance on Induced Circuit Failure Mode Likelihood Analysis," dated June 14, 2013 (Reference 37), discusses that, based on new experimental information documented in NUREG/CR-6931, "Cable Response to Live Fire (CAROLFIRE)" issued April2008 (Reference 38), and NUREG/CR-7100 "Direct Current Electrical Shorting in Response to Exposure Fire (DESIREE-Fire): Test Results,"* issued April 2012 (Reference 39), the effect of any control power transformers (CPTs) reduction to the hot short-induced spurious operation likelihood could not be substantiated.

NUREG-1805, "Fire Dynamics Tools (FOP): Quantitative Fire Hazard Analysis Methods for the U.S. Nuclear Regulatory Commission Fire Protection Inspection Program" (Reference 40), provides quantitative methods, known as FOT5, to assist regional fire protection inspectors in performing fire hazard analysis. The FOT5 are intended to assist fire protection inspectors in performing Rl evaluations of credible fires that may cause critical damage to essential safe shutdown equipment.

NUREG-1824, "Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications," Volumes 1 through 7 (Reference 41 ), provide technical documentation regarding the predictive capabilities of a specific set of fire models for the analysis of fire hazards in nuclear power plant scenarios. This report is the result of a collaborative program with the EPRI and the National Institute of Standards and Technology (NIST). The selected models are: (1) FDT5 developed by NRC (Volume 3); (2) The Fire-Induced Vulnerability Evaluation, Revision 1 (FIVE) developed by EPRI (Volume 4); (3) The zone model, Consolidated Model of Fire and Smoke Transport (CFAST), developed by NIST (Volume 5); (4) The zone model MAGIC developed by Electricite de France (EdF) (Volume 6); and (5) The computational fluid dynamics model, Fire Dynamics Simulator (FDS) developed, by NIST (Volume 7). In addition to the fire model volumes, Volume 1 is the comprehensive main report and Volume 2 is a description of the experiments and associated experimental uncertainty used in developing this report.

NUREG/CR-7010, "Cable Heat Release, !gnition, and §pread in Iray !nstallations during Eire (CHRISTl FIRE), Phase 1: Horizontal Trays," Volume 1 (Reference 42), describes Phase 1 of the CHRISTl FIRE testing program

-14-conducted by NIST. The overall goal of this multiyear program is to quantify the burning characteristics of grouped electrical cables installed in cable trays. This first phase of the program focuses on horizontal tray configurations. CHRISTl FIRE addresses the burning behavior of a cable in a fire beyond the point of electrical failure. The data obtained from this project can be used for the development of fire models to calculate the heat release rate (HRR) and flame spread of a cable fire.

NUREG-1855, Volume 1, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making" (Reference 43), provides guidance on how to treat uncertainties associated with PRA in Rl decision-making. The objectives of this guidance include fostering an understanding of the uncertainties associated with PRA and their impact on the results of PRA and providing a pragmatic approach to addressing these uncertainties in the context of the decision-making. To meet the objective of the NUREG, it is necessary to understand the role that PRA results play in the context of the decision process. To define this context, NUREG-1855 provides an overview of the Rl decision-making process itself.

NUREG-1921, "EPRI/NRC-RES Fire Human Reliability Analysis Final Report" (Reference 44), presents the state-of-the-art in fire human reliability analysis (HRA) practice. This report was developed jointly between RES and EPRI to develop the methodology and supporting guidelines for estimating human error probabilities (HEPs) for human failure events (HFEs) following the fire-induced initiating events of an FPRA. The report builds on existing HRA methods, and is intended primarily for practitioners conducting a fire HRA to support an FPRA.

NUREG-1934, "Nuclear Power Plant Fire Modeling Analysis Guidelines (NPP FIRE MAG)" (Reference 45), describes the implications of the verification and validation results from NUREG-1824 for fire model users. The features and limitations of the fire models documented in NUREG-1824 are discussed relative to their use to support nuclear power plant fire hazard analyses. The report also provides information to assist fire model users in applying this technology in the nuclear power plant environment.

Generic Letter (GL) 2006-03, "Potentially Nonconforming Hemyc and MT Fire Barrier Configurations" (Reference 46), requested that licensees evaluate their facilities to confirm compliance with the existing applicable regulatory requirements in light of the information provided in this GLand, if appropriate, take additional actions. Specifically, NRC testing revealed that, for the configurations tested, Hemyc and MT fire barriers failed to provide the protective function intended for compliance with existing regulations.

NFPA 101, "Life Safety Code" (Reference 47), provides the minimum requirements for egress, features of fire protection, sprinkler systems, alarms, emergency lighting, smoke barriers, and special hazard protection.

-15-* NFPA 30, "Flammable and Combustible Liquids Code." (Reference 48), provides safeguards to reduce the hazards associated with the storage, handling, and use of flammable and combustible liquids.

NFPA 12, "Standard on Carbon Dioxide Extinguishing Systems," (Reference 49), provides requirements for carbon dioxide fire-extinguishing systems to help ensure that such equipment will function as intended throughout its life. It is intended for those who purchase, design, install, test, inspect, approve, list, operate, or maintain these systems.

NFPA 10, "Standard for Portable Fire Extinguishers" (Reference 50), provides requirements to ensure that portable fire extinguishers will work as intended to provide a first line of defense against fires of limited size. 2.3 NFPA 805 Frequently Asked Questions In the LAR, the licensee proposed to use a number of documents commonly known as NFPA 805 FAQs. The following table provides the set of FAQs the licensee used that the NRC staff referenced in the preparation of this SE, as well as the SE section(s) in which each FAQ is referenced. Table 2.3-1: NFPA 805 Frequently Asked Questions Reference SE FAQ# F AQ Title and Summary No. Section 07-0030 "Establishing Recovery Actions" 51 3.2.5 3.4.4

This FAQ provides an acceptable process for 3.5.1.7 determining the recovery actions for NFPA 805 Chapter 4 compliance. The process includes:

Differentiation between recovery actions and activities in the main control room or at primary control station(s).

Determination of which recovery actions are required by the NFPA 805 fire protection program.

Evaluate the additional risk presented by the use of recovery actions.

Evaluate the feasibility of the identified recovery actions.

Evaluate the reliability of the identified recovery actions.

-16-Reference SE FAQ# FAQ Title and Summary No. Section 07-0038 "Lessons Learned on Multiple Spurious Operations (MSOs)" 52 3.2.4 3.2.6

This FAQ reflects an acceptable process for the treatment of MSOs during transition to NFPA 805:

Step 1 -Identify potential MSO combinations of concern.

Step 2 -Expert panel assesses plant-specific vulnerabilities and reviews MSOs of concern.

Step 3-Update the fire PRA and Nuclear Safety Capability Assessment to include MSOs of concern.

Step 4-Evaluate for NFPA 805 compliance .

Step 5 -Document the results . 07-0039 "Incorporation of Pilot Plant Lessons Learned-Table B-2" 53 3.2.1

This FAQ provides additional detail for the comparison of the licensee's safe shutdown strategy to the endorsed industry guidance, NEI 00-01, "Guidance for Post-Fire Safe Shutdown Circuit Analysis," Revision 1 (Reference 54). In short, the process has the licensee:

Assemble industry and plant-specific documentation;

Determine which sections of the guidance are applicable;

Compare the existing safe shutdown methodology to the applicable guidance; and

Document any discrepancies . 07-0040 "Non-Power Operations (NPO) Clarifications" 55 3.5.3 3.5.4

This FAQ clarifies an acceptable NFPA 805 NPO program. The process includes:

Selecting NPO equipment and cabling .

Evaluation of NPO Higher Risk Evolutions (HRE) .

Analyzing NPO key safety functions (KSF) .

Identifying plant areas to protect or "pinch points" during NPO HREs and actions to be taken if KSFs are lost.

-17-Reference SE FAQ# FAQ Title and Summary No. Section 08-0054 "Demonstrating Compliance with Chapter 4 of NFPA 805" 56 3.4.3 3.5.1.4

This FAQ provides an acceptable process to demonstrate Chapter 4 compliance for transition:

Step 1 -Assemble documentation

Step 2 -Document Fulfillment of Nuclear Safety Performance Criteria

Step 3-Variance From Deterministic Requirements (VFDR) Identification, Characterization, and Resolution Considerations

Step 4 -Performance-Based Evaluations

Step 5 -Final VFDR Evaluation

Step 6 -Document Required Fire Protection Systems and Features 09-0056 "Radioactive Release Transition" 57 3.6

This FAQ provides an acceptable level of detail and content for the radioactive release section of the LAR. It includes:

Justification of the compartmentation, if the radioactive release review is not performed on a fire area basis.

Pre-fire plan and fire brigade training review results .

Results from the review of engineering controls for gaseous and liquid effluents. 10-0059 "NFPA 805 Monitoring Program" 58 3.7

This FAQ provides clarification regarding the implementation of an NFPA 805 monitoring program for transition. It includes:

Monitoring program analysis units;

Screening of low safety significant structures, systems, and components;

Action level thresholds; and

The use of existing monitoring programs . 12-0062 "Updated Final Safety Analysis Report (UFSAR) Content" 59 2.4.4

This FAQ provides guidance on the content and necessary level of detail for the transition of the fire protection sections within the UFSAR.

-18-2.4 Orders, License Conditions, and Technical Specifications Paragraph 50.48(c)(3)(i) of 10 CFR states that the LAR " ... must identify any orders and license conditions that must be revised or superseded, and contain any necessary revisions to the plant's technical specifications and the bases thereof." 2.4.1 Orders The NRC staff reviewed Section 5.2.3, "Orders and Exemptions," and Attachment 0, "Orders and Exemptions," of NMP1 's LAR (Reference 6), with regard to NRC-issued Orders pertinent to NMP1 that are being revised or superseded by the NFPA 805 transition process. The LAR stated that the licensee conducted a review of its docketed correspondence to determine if there were any orders or exemptions that needed to be superseded or revised. The LAR also stated that the licensee conducted a review to ensure that compliance with the physical protection requirements, security orders, and adherence to those commitments applicable to NMP1 are maintained. The licensee discussed the affected orders and exemptions in LAR Attachment 0 of the LAR. The licensee requested that 5 exemptions be rescinded, and determined that no Orders need to be superseded or revised to implement an FPP at NMP1 that complies with 10 CFR 50.48(c). The review conducted by the licensee included an assessment of docketed correspondence files and electronic searches, including the NRC's Agencywide Documents Access and Management System (ADAMS). The review was performed to ensure that compliance with the physical protection requirements, security orders, and adherence to commitments applicable to NMP1 are maintained. The NRC staff accepts the licensee's determination that 5 exemptions should be rescinded (i.e. no exemptions are transitioned to NFPA 805 as listed in LAR Attachment K and that no Orders need to be superseded or revised to implement NFPA 805 at NMP1. See Section 2.5 of this SE for the NRC staff's detailed evaluation of the exemptions being rescinded. In addition, the licensee performed a specific review of the license amendment that incorporated the mitigation strategies required by 10 CFR 50.54(hh)(2) to ensure that any changes being made in order to comply with 10 CFR 50.48{c) do not invalidate existing commitments applicable to NMP1. The licensee's review of this regulation and the related license amendment demonstrated that changes to the FPP during transition to NFPA 805 will not affect the mitigation measures required by 10 CFR 50.54{hh)(2) because the licensee will continue to have strategies that address large fires and explosions including a firefighting response strategy, operations to mitigate fuel damage, and actions to minimize release upon transition to NFPA 805. The NRC staff concludes that the licensee's determination in regard to 10 CFR 50.54(hh)(2) is acceptable.

-19-2.4.2 License Conditions The NRC staff reviewed LAR Section 5.2.1, "License Condition Changes," and Attachment M, "License Condition Changes," regarding changes the licensee seeks to make to the NMP1 fire protection license condition in order to adopt NFPA 805, as required by 10 CFR 50.48(c)(3). The NRC staff reviewed the revised license condition, which supersedes the current NMP1 fire protection license condition, for consistency with the format and content guidance in Regulatory Position C.3.1 of RG 1.205, Revision 1, and with the proposed plant modifications identified in the LAR. The revised license condition provides a structure and detailed criteria to allow self-approval for RI/PB as well as other types of changes to the FPP. The structure and detailed criteria result in a process that meets the requirements in Sections 2.4, Engineering Analyses, 2.4.3, Fire Risk Evaluations, and 2.4.4, Plant Change Evaluation of NFPA 805. These sections establish the requirements for the content and quality of the engineering evaluations to be used for approval of changes. The revised license condition also defines the limitations imposed on the licensee during the transition phase of plant operations when the physical plant configuration does not fully match the configuration represented in the fire risk analysis. The limitations on self-approval are required because NFPA 805 requires that the risk analyses be based on the as-built, operated, and maintained plant, and reflect the operating experience at the plant. Until the proposed implementation items and plant modifications are completed, the risk analysis is not based on the as-built, as-operated and maintained plant. Overall, the licensee's proposed revised license condition allows self-approval for FPP changes that meet the requirements of NFPA 805 with regard to engineering analyses, fire risk evaluations (FREs), and plant change evaluations (PCEs). The NRC staff's evaluation of the self-approval process for FPP changes (post-transition) is contained in Section 2.6 of this SE. The license condition also references the plant-specific modifications, and associated implementation schedules that must be accomplished at NMP1 to complete transition to NFPA 805 and comply with 10 CFR 50.48(c). In addition, the license condition includes a requirement that appropriate compensatory measures will remain in place until implementation of the specified plant modifications is completed. These modifications and implementation schedules are identical to those identified elsewhere in the LAR, as discussed by the NRC staff in Sections 2.7 and 3.0 of this SE. Section 4.0 of this SE provides the NRC staff's review of the proposed NMP1 FPP license condition. 2.4.3 Technical Specifications The NRC staff reviewed LAR Section 5.2.2, "Technical Specifications," and Attachment N, "Technical Specification Changes," with regard to proposed changes to the NMP1 TSs that are being revised or superseded during the NFPA 805 transition process. According to the LAR,

-20-the licensee conducted a review of the NMP1 TSs to determine which, if any, TS sections will be impacted by the transition to an RI/PB FPP based on 10 CFR 50.48(c). The licensee identified changes to the TSs needed for NMP1 adoption of the new fire protection licensing basis and provided applicable justification listed in Attachment N. The changes identified included a change toTS Section 6.0, Administrative Controls, Section 6.4, Procedures, and the Bases forTS Sections 3.6.13 and 4.6.13, Remote Shutdown Panels. Specifically, the licensee proposed to revise TS Section 6.4 by deleting "Fire Protection Program Implementation" from the list of activities for which written procedures and administrative policies shall be established. The licensee considered this change acceptable for the adoption of the new fire protection licensing basis since the requirement for establishing, implementing and maintaining fire protection procedures is contained in 10 CFR 50.48(a) and 10 CFR 50.48(c), as specifically outlined in NFPA 805, Section 3.2.3, Procedures. In addition the licensee proposed a change to the Bases forTS Sections 3.6.13 and 4.6.13 by changing the applicable regulation from 10 CFR 50, Appendix R, to 10 CFR 50.48(c). The licensee considered this change acceptable because upon completion of transition to NFPA 805 the licensing basis will change from 10 CFR 50 Appendix R, to 10 CFR 50.48(c). Based on the information provided by the licensee, the NRC staff concludes that the proposed change toTS Section 6.4 is acceptable because it is an administrative control (i.e., a procedure the licensee puts in place to establish, implement, and maintain the fire protection program as required by the licensee's fire protection license condition and 10 CFR 50.48(a), 10 CFR 50.48(c), and NFPA 805, Section 3.2.3), and would be redundant to the NFPA 805 requirement to establish FPP procedures. In addition, failure by the licensee to establish FPP procedures would result in non-compliance with 10 CFR 50.48(c)(1 ), which is the licensee's fire protection licensing basis. Changes to fire protection administrative controls are controlled by the proposed fire protection license condition. See SE Section 4.0 2.4.4 Updated Safety Analysis Report The NRC staff reviewed LAR AttachmentS, Table S-2 "Implementation Items," with regard to changes the licensee is proposing to make to the Updated Final Safety Analysis Report (UFSAR). AttachmentS, Table S-2, Item 20 states that the UFSAR will be updated. Section 5.4 of the LAR states that the format and content of the revised UFSAR will be consistent with FAQ 12-0062 (Reference 59). Since the licensee has provided an implementation item that will update the UFSAR after approval of the LAR in accordance with 10 CFR 50.71 (e), and the content will be consistent with the guidance contained in NEI 04-02, the NRC staff concludes that the licensee's method to update the UFSAR following the guidance in FAQ 12-0062 is acceptable. 2.5 Rescission of Exemptions The NRC staff reviewed LAR Section 5.2.3, "Orders and Exemptions," Attachment 0, "Orders and Exemptions," and Attachment K, "Existing Licensing Action Transition," with regard to previously-approved exemptions to Appendix R to 10 CFR Part 50, which the transition to an

-21 -FPP licensing basis in conformance with NFPA 805 will supersede. These exemptions will no longer be required since, upon approval of the RI/P8 FPP in accordance with NFPA 805, Appendix R will not be part of the licensing basis for NMP1. The licensee requested and received NRC approval for 5 exemptions from 10 CFR Part 50 Appendix R. These exemptions were discussed in detail in LAR Attachment K. The licensee stated that the exemptions are no longer required because the subject boundaries have been demonstrated adequate for the hazard in an existing engineering equivalency evaluation (EEEE). The licensee requested in accordance with the requirements of 10 CFR 50.48(c)(3)(i), that all the exemptions be rescinded. Disposition of Appendix R exemptions may follow two different paths during transition to NFPA 805:

The exemption was found to be unnecessary since the underlying condition has been evaluated using RI/P8 methods (FM and/or FRE) and found to be acceptable and no further actions are necessary by the licensee.

The exemption was found to be appropriate as a qualitative engineering evaluation that meets the deterministic requirements of NFPA 805 and is carried forward as part of the engineering analyses supporting NFPA 805 transition. The following exemptions are rescinded as requested by the LAR and the underlying condition has been evaluated using RI/P8 methods and found to be acceptable with no further actions (numbering scheme provided by the licensee):

None. The following exemptions are rescinded as requested by the LAR, but the engineering evaluation of the underlying condition will be used as a qualitative engineering evaluation for transition to NFPA 805:

Exemption from the requirements of Section III.G.2 of Appendix R for the battery board rooms (FA 16A and FA 168), since their boundary walls do not provide the required 3-hour rated barriers.

Exemption from the requirements of Section III.G.2 of Appendix R for the battery rooms (FA 17A and FA 178), since their boundary walls do not provide the required 3-hour rated barriers.

Exemption from the requirements of Section III.G of Appendix R for the control room (FA 11 ), since the control room ceiling does not have a 3-hour rating from the control room side due to unprotected structural steel members.

Exemption from the requirements of Section III.G.2 of Appendix R for the wall between the reactor building and the turbine building above elevation 340' (FA 1, FA 2, and FA 5), since the wall is not a 3-hour rated barrier.

-22-* Exemption from the requirements of Section III.G.2 of Appendix R for the fire break zone separating FA 1 and FA 2 in the reactor building upper level (elevation 340'), since the wall is not a 3-hour rated barrier. 2.6 Self-Approval Process for FPP Changes (Post-Transition) Upon completion of the implementation of the RI/PB FPP and issuance of the license condition discussed in SE Section 2.4.2, changes to the approved FPP must be evaluated by the licensee to ensure that they are acceptable. NFPA 805 Section 2.2.9, "Plant Change Evaluation," states the following: In the event of a change to a previously approved fire protection program element, a risk-informed plant change evaluation shall be performed and the results used as described in 2.4.4 to ensure that the public risk associated with fire-induced nuclear fuel damage accidents is low and that adequate depth and safety margins are maintained. NFPA 805, Section 2.4.4, "Plant Change Evaluation," states, in part, that: A plant change evaluation shall be performed to ensure that a change to a previously approved fire protection program element is acceptable. The evaluation process shall consist of an integrated assessment of the acceptability of risk, defense-in-depth, and safety margins. 2.6.1 Post-Implementation Plant Change Evaluation Process The NRC staff reviewed LAR Section 4.7.2, "Compliance with Configuration Control Requirements in Section 2.7.2 and 2.2.9 of NFPA 805," for compliance with the NFPA 805 Plant Change Evaluation (PCE) process requirements to address potential changes to the NFPA 805 RI/PB FPP after implementation is completed. The licensee developed a change process that is based on the requirements of NFPA 805, and guidance provided in NEI 04-02, Section 5.3, "Plant Change Process," as well as Appendices B, I, and J, as modified by RG 1.205, Regulatory Positions 2.2.4, 3.1, 3.2, and 4.3. LAR Section 4.7.2 states that the PCE process will consist of four steps: 1. defining the change 2. preliminary risk screening 3. risk evaluation 4. acceptability determination In the LAR, the licensee stated that the PCE process begins by defining the change or altered condition to be examined and the baseline configuration. The baseline is defined by the design basis and licensing basis. The licensee further stated that the baseline is defined as that plant condition or configuration that is consistent with the design basis and licensing basis (NFPA 805

-23-licensing basis post-transition) and that the changed or altered condition or configuration that is not consistent with the design basis and licensing basis is defined as the proposed alternative. The licensee stated that once the definition of the change is established, a screening will be performed to identify and resolve minor changes to the FPP and that the screening will be consistent with fire protection regulatory review processes in place at nuclear plants under traditional licensing bases. The licensee further stated that the screening process will be modeled after NEI 02-03, "Guidance for Performing a Regulatory Review of Proposed Changes to the Approved Fire Protection Program," (Reference 60), and that the process will address most administrative changes (e.g., changes to the combustible control program, organizational changes). The licensee stated that the screening will identify when a change will require additional engineering evaluations that may include fire modeling (FM) and risk assessment techniques and that the results of the evaluations are compared to the acceptance criteria. The licensee also stated that changes that satisfy the acceptance criteria of NFPA 805 Section 2.4.4, and the fire protection license condition (see Attachment M to the LAR) can be implemented within the framework provided by NFPA 805; and that changes that do not satisfy the acceptance criteria cannot be implemented within this framework. The licensee further stated that the acceptance criteria will require that the resultant change in core damage frequency (CDF) and LERF be consistent with the fire protection license condition, and that the acceptance criteria will also include consideration of DID and safety margin, which would typically be qualitative in nature. The licensee stated that the risk evaluation involves the application of FM analyses and risk assessment techniques to obtain a measure of the changes in risk associated with the proposed change. The licensee also stated that, in certain circumstances, an initial evaluation in the development of the risk assessment could be a simplified analysis using bounding assumptions, provided the use of such assumptions does not unnecessarily challenge the acceptance criteria. The licensee stated that PCEs are assessed for acceptability using the change in CDF CDF or and change in LERF (delta-LEAF or criteria from the license condition and the proposed changes are assessed to ensure they are consistent with the DID philosophy and that sufficient safety margin is maintained. The licensee stated that the NMP1 FPP configuration is defined by the program documentation and that to the greatest extent possible, the existing configuration control processes for modifications, calculations and analyses, and FPP license basis reviews will be used to maintain configuration control of the FPP documents. The licensee further stated that the configuration control procedures which govern the various NMP1 documents and databases that currently exist will be revised to reflect the new NFPA 805 licensing bases requirements. The licensee stated that several NFPA 805 document types such as: nuclear safety capability assessment (NSCA) supporting information, non-power operational mode NSCA treatment, etc., generally require new control procedures and processes to be developed since they are new documents and databases created as a result of the transition to NFPA 805. In addition, the licensee stated that the new procedures will be modeled after the existing processes for similar types of documents and databases. The licensee further stated that system level design basis

-24-documents will be revised to reflect the NFPA 805 role that the systems and components now play. The licensee stated that the process for capturing the impact of proposed changes to the plant on the FPP will continue to be a multiple step review and that the first step of the review will be an initial screening for process users to determine if there is a potential to impact the FPP as defined under NFPA 805 through a series of screening questions/checklists contained in one or more procedures depending upon the configuration control process being used. The licensee further stated that reviews that identify potential FPP impacts will be sent to qualified individuals (e.g., Fire Protection Engineer, Fire PRA Engineer, etc.) to ascertain the program impacts, if any, and that if FPP impacts are determined to exist as a result of the proposed change, the issue would be resolved by one of the following:

Deterministic Approach: Complying with NFPA 805, Chapter 3 and Section 4.2.3 requirements; or

PB Approach: Utilizing the NFPA 805 change process developed in accordance with NEI 04-02, RG 1.205, and the NMP1 NFPA 805 fire protection license condition to assess the acceptability of the proposed change. This process will be used to determine if the proposed change can be implemented "as-is" or whether prior NRC approval of the proposed change is required. The licensee stated that this process follows the requirements in NFPA 805 and the guidance outlined in RG 1.17 4, (Reference 27), which requires the use of qualified individuals, procedures that require calculations and evaluations be subject to independent review and verification, record retention, peer review, and a corrective action program that ensures appropriate actions are taken when errors are discovered. Since NFPA 805 always requires the use of a PCE, regardless of what element requires the change, the NRC staff concludes that, in accordance with the requirements of NFPA 805, if FPP impacts are determined to exist as a result of the proposed change, the issue would be resolved by utilizing the NFPA 805 change process developed in accordance with NEI 04-02, RG 1.205, and the NMP1 NFPA 805 fire protection license condition to assess the acceptability of the proposed change. This process will be used to determine if prior NRC approval of the proposed change is required. Based on the information provided by the licensee, the NRC staff concluded that the licensee's PCE process is considered acceptable because it meets the guidance in NEI 04-02, Revision 2 (Reference 5), as well as RG 1.205, Revision 1 (Reference 2), and addresses attributes for using FREs in accordance with NFPA 805. Section 2.4.4 of NFPA 805 requires that PCEs consist of an integrated assessment of risk, DID and safety margin. Section 2.4.3.1 of NFPA 805 requires that the probabilistic safety assessment (PSA) use CDF and LERF as measures for risk. Section 2.4.3.3 of NFPA 805 requires that the risk assessment approach, methods, and data shall be acceptable to the Authority Having Jurisdiction (AHJ), which is the NRC. Section 2.4.3.3 of NFPA 805 also requires that the PSA be appropriate for the nature and scope of the change being evaluated, be based on the as-built and as-operated and maintained plant, and reflect the operating experience at the plant.

-25-The licensee's PCE process includes the required delta risk calculations, uses risk assessment methods acceptable to the NRC, uses appropriate risk acceptance criteria in determining acceptability, involves the use of an FPRA of acceptable quality, and includes an integrated assessment of risk, DID, and safety margin as discussed above. 2.6.2 Requirements for the Self-Approval Process Regarding Plant Changes Risk assessments performed to evaluate PCEs must use methods that are acceptable to the NRC staff. Acceptable methods to assess the risk of the proposed plant change may include methods that have been used in developing the peer-reviewed FPRA model, methods that have been approved by the NRC via a plant-specific license amendment or through NRC approval of generic methods specifically for use in NFPA 805 risk assessments, or methods that have been demonstrated to bound the risk impact. Based on the information provided by the licensee in the LAR, the process established to evaluate post-transition plant changes meets the guidance in NEI 04-02, as well as RG 1.205. The NRC staff concludes that the proposed PCE process at NMP1, which includes defining the change, a preliminary risk screening, a risk evaluation, and an acceptability determination, as described in Section 2.6.1, is acceptable because it addresses the required delta-risk calculations, uses risk assessment methods acceptable to the NRC, uses appropriate risk acceptance criteria in determining acceptability, involves the use of an FPRA of acceptable quality, and includes an integrated assessment of risk, DID, and safety margin. However, the proposed license condition provides that, before achieving full compliance with 10 CFR 50.48(c) by implementing the plant modifications listed in Section 2.7.1 of this SE (i.e., during full implementation of the transition to NFPA 805), Rl changes to the licensee's FPP may not be made without prior NRC review and approval unless the changes have been demonstrated to have no more than a minimal risk impact using the screening process discussed above because the risk analysis is not consistent with the as-built, as-operated, and maintained plant since the modifications have not been completed. In addition, the license condition requires the licensee to ensure that fire protection DID and safety margin are maintained during the transition process. The Transition License Conditions" in the proposed NFPA 805 license condition include the appropriate acceptance criteria and other attributes to form an acceptable method for meeting Regulatory Position C.3.1 of RG 1.205, Revision 1, (Reference 2), with respect to the requirements for FPP changes during transition, and, therefore, are acceptable to demonstrate compliance with 10 CFR 50.48(c). The proposed NFPA 805 license condition also includes a provision for self-approval of changes to the FPP that may be made on a qualitative, rather than an Rl, basis. Specifically, the license condition states that prior NRC review and approval are not required for changes to the NFPA 805 Chapter 3 fundamental FPP elements and design requirements for which an engineering evaluation demonstrates that the alternative to the NFPA 805 Chapter 3 element is functionally equivalent or adequate for the hazard. The licensee may use an engineering evaluation to demonstrate that a change to an NFPA 805 Chapter 3 element is functionally equivalent to the corresponding technical requirement. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the

-26-functionality of the component, system, procedure, or physical arrangement (i.e., has not impacted its contribution toward meeting the nuclear safety and radioactive release performance criteria), using a relevant technical requirement or standard. Use of this approach does not fall under NFPA 805, Section 1.7, "Equivalency," because the condition can be shown to meet the NFPA 805 Chapter 3 requirement. Section 1.7 of NFPA 805 is a standard format used throughout NFPA standards. It is intended to allow owner/operators to use the latest state of the art fire protection features, systems, and equipment, provided the alternatives are of equal or superior quality, strength, fire resistance, durability, and safety. However, the intent is to require approval from the authority having jurisdiction because not all of these state-of-the-art features are in current use or have relevant operating experience. This is a different situation than the use of functional equivalency since functional equivalency demonstrates that the condition meets the NFPA 805 code requirement. Alternatively, the licensee may use an engineering evaluation to demonstrate that changes to certain NFPA 805 Chapter 3 elements are acceptable because the changes are "adequate for the hazard." Prior NRC review and approval would not be required for alternatives to four specific sections of NFPA 805 Chapter 3 listed below, for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is adequate for the hazard. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement (with respect to the ability to meet the nuclear safety and radioactive release performance criteria), using a relevant technical requirement or standard. NFPA 805 Section 2.4 states that engineering analysis is an acceptable means of evaluating an FPP against performance criteria. Engineering analyses shall be permitted to be qualitative or quantitative. Use of qualitative engineering analyses by a qualified fire protection engineer to determine that a change has not affected the functionality of the component, system, procedure or physical arrangement is allowed by NFPA 805 Section 2.4. The four specific sections of NFPA 805 Chapter 3 for which prior NRC review and approval are not required to implement alternatives that an engineering evaluation has demonstrated are adequate for the hazard are as follows: 1. "Fire Alarm and Detection Systems" (Section 3.8); 2. "Automatic and Manual Water-Based Fire Suppression Systems" (Section 3.9); 3. "Gaseous Fire Suppression Systems" (Section 3.1 0); and 4. "Passive Fire Protection Features" (Section 3.11 ). The engineering evaluations described above (i.e., functionally equivalent and adequate for the hazard) are engineering analyses governed by the NFPA 805 guidelines. In particular, this means that the evaluations must meet the requirements of NFPA 805, Section 2.4, "Engineering Analyses," and NFPA 805, Section 2.7, "Program Documentation, Configuration Control, and Quality." Specifically, the effectiveness of the fire protection features under review must be evaluated and found acceptable in relation to their ability to detect, control, suppress, and extinguish a fire and provide passive protection to achieve the performance criteria and not

-27-exceed the damage threshold for the plant being analyzed. The associated evaluations must also meet the documentation content (as outlined by NFPA 805, Section 2.7.1, "Content") and quality requirements (as outlined by NFPA 805, Section 2.7.3, "Quality") of the standard in order to be considered adequate. The NRC staff's review of the licensee's compliance with NFPA 805, Sections 2.7.1 and 2.7.3 is provided in Section 3.8 of this SE. According to the LAR, the licensee intends to use an FPRA to evaluate the risk of proposed future plant changes. Section 3.4.2, "Quality of the Fire Probabilistic Risk Assessment," of this SE discusses the technical adequacy of the FPRA, including the licensee's process to ensure that the FPRA remains current. The NRC staff determined that the quality of the licensee's FPRA and associated administrative controls and processes for maintaining the quality of the PRA model is sufficient to support self-approval of future Rl changes to the FPP under the proposed license conditions. Therefore, the NRC staff concludes that the licensee's process for self-approving future FPP changes is acceptable. The NRC staff also concludes that the FRE methods used at NMP1 to model the cause and effect relationship of associated changes as a means of assessing the risk of plant changes during transition to NFPA 805 may continue to be used after implementation of the RI/PB FPP, based on the licensee's administrative controls to ensure that the models remain current and to provide assurance of continued quality (see SE Section 3.4.2, "Quality of the Fire Probabilistic Risk Assessment"). Accordingly, these cause-and-effect relationship models may be used after transition to NFPA 805 as a part of the PCEs conducted to determine the change in risk associated with proposed plant changes. 2.7 Modifications and Implementation Items Regulatory Position C.3.1 of RG 1.205, Revision 1, says that a license condition included in a NFPA 805 LAR should include: (1) a list of modifications being made to bring the plant into compliance with 10 CFR 50.48(c); (2) a schedule detailing when these modifications will be completed; and (3) a statement that the licensee shall maintain appropriate compensatory measures in place until implementation of the modifications are completed. The NRC staff noted that the list of modifications and implementation items originally submitted in the LAR have been updated by the licensee with the final version of LAR Attachment S, "Plant Modifications and Items to be Completed during Implementation." The updated LAR AttachmentS is provided in the licensee's letter dated May 09, 2014 (Reference 14). 2.7.1 Modifications The NRC staff reviewed LAR AttachmentS, "Plant Modifications and Items to be Completed during Implementation," which describes the NMP1 plant modifications necessary to implement the NFPA 805 licensing basis, as proposed. These modifications are identified in the LAR as necessary to bring NMP1 into compliance with either the deterministic or PB requirements of NFPA 805. As described below, LAR Attachment S, Table S-1 provides a description of each of the proposed plant modifications, presents the problem statement explaining why the modification is needed, and identifies if compensatory actions are required to be in place pending completion/implementation of the modification.

-28-The NRC staff's review confirmed that the modifications identified in LAR Table S-1 are the same as those identified in LAR Table B-3, "Fire Area Transition," on a fire area basis, as the modifications being credited in the proposed NFPA 805 licensing basis. The NRC staff also confirmed that the LAR Table S-1 modifications and the associated completion schedule are the same as those provided in the proposed NFPA 805 license conditions. LAR AttachmentS, Table S-1 provides a detailed listing of the plant modifications that must be completed in order for NMP1 to be fully in accordance with NFPA 805, implement many of the attributes upon which this SE is based, and thereby meet the requirements of 10 CFR 50.48{c). The licensee stated that they would complete implementation of the modifications no later than the end of the first refueling outage following issuance of the license amendment and identifies if compensatory measures are required to be in place while awaiting completion of the modifications. 2.7.2 Implementation Items Implementation Items are items that the licensee has not fully completed or implemented as of the issuance date of the license amendment, but which will be completed during implementation of the license amendment to transition to NFPA 805 (e.g., procedure changes that are still in process, or NFPA 805 programs that have not been fully implemented). The licensee identified the implementation items in LAR AttachmentS, Table S-2. For each implementation item, the licensee and the NRC staff have reached a satisfactory resolution involving the level of detail and main attributes that each remaining change will incorporate upon completion. Completion of these items in accordance with the schedule discussed in SE Section 2.7.3 does not change or impact the bases for the safety conclusions made by the NRC staff in theSE. Each implementation item will be completed prior to the deadline for implementation of the RI/PB FPP based on NFPA 805, as specified in the license conditions and the letter transmitting the amended license (i.e., implementation period) which states that the implementation items listed in LAR Attachment S, Table S-2, will be completed 180 days after issuance of the NFPA 805 license amendment unless that date falls within a scheduled refueling outage, if the latter, the implementation items will be completed 60 days after startup from that scheduled refueling outage. The NRC staff, through an onsite audit or during a future fire protection inspection, may choose to examine the closure of the implementation items, with the expectation that any variations discovered during this review, or concerns with regard to adequate completion of the implementation item, would be tracked and dispositioned appropriately under the licensee's corrective action program and could be subject to appropriate NRC enforcement action as they are required by the proposed license conditions. 2.7.3 Schedule LAR Section 5.5 provides the overall schedule for completing the NFPA 805 transition at NMP1. The licensee stated that it will complete the implementation of the new program, including any procedure changes, process updates, and training for affected plant personnel to implement the

-29-NFPA 805 FPP within 180 days after NRC approval of the NFPA 805 License Amendment unless that date falls within a scheduled refueling outage, then in that case, within 60 days after startup from that scheduled refueling outage. LAR Section 5.5 also states that modifications identified in LAR Attachment S, Table S-1, will be completed prior to the end of the first refueling outage following the issuance of the NFPA 805 license amendment and that no compensatory measures are required to be maintained while awaiting completion of the modifications. In the letter dated May 9, 2014 (Reference 14), the licensee stated that interim measures submitted under the response to RAI SSD-06 in a letter dated March 27, 2013 (Reference 8) will become compensatory measures upon receipt of the SE. As a result of this change, the staff notes that compensatory measures are required to be maintained while certain modifications listed in LAR AttachmentS, Table S-1 await completion.

3.0 TECHNICAL EVALUATION

The following sections evaluate the technical aspects of the requested license amendment to transition the FPP at NMP1 to one based on NFPA 805 (Reference 1) in accordance with 10 CFR 50.48(c). While performing the technical evaluation of the licensee's submittal, the NRC staff utilized the guidance provided in NUREG-0800, Section 9.5.1.2, "Risk-Informed, Performance-Based Fire Protection" (Reference 31 ), to determine whether the licensee had provided sufficient information in both scope and level of detail to adequately demonstrate compliance with the requirements of NFPA 805, as well as the other associated regulations and guidance documents discussed in Section 2.0 of this SE. Specifically:

Section 3.1 provides the results of the NRC staff review of the licensee's transition of the FPP from the existing deterministic guidance to that of NFPA 805 Chapter 3, "Fundamental FPP and Design Elements."

Section 3.2 provides the results of the NRC staff review of the methods used by the licensee to demonstrate the ability to meet the nuclear safety performance criteria (NSPC).

Section 3.3 provides the results of the NRC staff review of the FM methods used by the licensee to demonstrate the ability to meet the NSPC using an FM PB approach.

Section 3.4 provides the results of the NRC staff review of the fire risk assessments used to demonstrate the ability to meet the NSPC using an FRE PB approach.

Section 3.5 provides the results of the NRC staff review of the licensee's NSCA results by fire area.

Section 3.6 provides the results of the NRC staff review of the methods used by the licensee to demonstrate an ability to meet the radioactive release performance criteria.

-30-* Section 3.7 provides the results of the NRC staff review of the NFPA 805 monitoring program developed as a part of the transition to an Rl/PB FPP based on NFPA 805.

Section 3.8 provides the results of the NRC staff review of the licensee's program documentation, configuration control, and quality assurance. In addition, Attachments A and B to this SE provide additional detailed information that was evaluated and/or disposed by the NRC staff to support the licensee's request to transition to an RI/PB FPP in accordance with NFPA 805 (i.e., 10 CFR 50.48(c)). These attachments are discussed as appropriate in the associated section of this SE. 3.1 NFPA 805 Fundamental FPP and Design Elements NFPA 805 Chapter 3 contains the fundamental elements of the FPP and specifies the minimum design requirements for fire protection systems and features that are necessary to meet the standard. The fundamental FPP elements and minimum design requirements include necessary attributes pertaining to the fire protection plan and procedures, the fire prevention program and design controls, internal and external industrial fire brigades, and fire protection SSCs. However, 10 CFR 50.48(c) provides exceptions, modifications, and supplementations to certain aspects of NFPA 805 Chapter 3 as follows:

10 CFR 50.48(c)(2)(v) -Existing cables. In lieu of installing cables meeting flame propagation tests as required by Section 3.3.5.3 of NFPA 805, a retardant coating may be applied to the electric cables, or an automatic fixed fire suppression system may be installed to provide an equivalent level of protection. In addition, the italicized exception to Section 3.3.5.3 of NFPA 805 is not endorsed.

10 CFR 50.48(c)(2)(vi)-Water supply and distribution. The italicized exception to Section 3.6.4 of NFPA 805 is not endorsed. Licensees who wish to use the exception to Section 3.6.4 of NFPA 805 must submit a request for a license amendment in accordance with 10 CFR 50.48(c)(2)(vii).

10 CFR 50.48(c)(2)(vii) -Performance-based methods. While Section 3.1 of NFPA 805 prohibits the use of PB methods to demonstrate compliance with the NFPA 805 Chapter 3 requirements, 10 CFR 50.48(c)(2)(vii) states that the FPP elements and minimum design requirements of NFPA 805 Chapter 3 may be subject to the PB methods permitted elsewhere in the standard provided a license amendment is granted and the approach satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release, maintains safety margins; and maintains fire protection defense-in-depth. Furthermore, Section 3.1 of NFPA 805 specifically allows the use of alternatives to the NFPA 805 Chapter 3 fundamental FPP requirements that have been previously approved by the

-31 -NRC (the AHJ as denoted in NFPA 805 and RG 1.205), and are contained in the currently approved FPP for the facility. 3.1.1 Compliance with NFPA 805 Chapter 3 Requirements The licensee used the systematic approach described in NEI 04-02, Revision 2 (Reference 5), as endorsed by the NRC in RG 1.205, Revision 1, to assess the proposed NMP1 FPP against the NFPA 805 Chapter 3 requirements. As part of this assessment, the licensee reviewed each section and subsection of NFPA 805 Chapter 3 against the existing NMP1 FPP and provided specific compliance statements for each Chapter 3 attribute that contained applicable requirements. As discussed below, some subsections of NFPA 805 Chapter 3 do not contain requirements, or are otherwise not applicable to NMP1, and others are provided with multiple compliance statements to fully document compliance with the element. The methods used by NMP1 for achieving compliance with the NFPA 805 Chapter 3 fundamental FPP elements and minimum design requirements are as follows: 1. The existing FPP element directly complies with the requirement: noted in LAR Attachment A, "NEI 04-02 Table B-1, Transition of Fundamental Fire Protection Program and Design Elements" (LAR Table 8-1), as "Complies." (see discussion in SE Section 3.1.1.1 ), 2. The existing FPP element complies through the use of an explanation or clarification: noted in LAR Table B-1 as "Complies with Clarification." (see discussion in SE Section 3.1.1.2), 3. The existing FPP element complies through the use of EEEEs whose bases remain valid and are of sufficient quality: noted in LAR Table B-1 as "Complies with Use of EEEEs," (see discussion in SE Section 3.1.1.3), 4. The existing FPP element complies with the requirement based on prior NRC approval of an alternative to the fundamental FPP attribute and the bases for the NRC approval remain valid: noted in LAR Table B-1 as "Complies by Previous NRC Approval," (see discussion in SE Section 3.1.1.4), 5. The existing FPP element does not comply with the requirement, but the licensee is requesting specific approval for a PB method in accordance with 10 CFR 50.48(c)(2)(vii): noted in LAR Table B-1 as "Submit for NRC Approval". These requests are described in LAR Attachment L and Sections 3.1.1.5 and 3.1.4 of this SE (see discussion in SE Section 3.1.1.5), and 6. The existing FPP element does not comply with the requirement, but will be in direct compliance with the completion of a required action; noted in LAR Table B-1 as "Complies with item for implementation." These outstanding

-32-actions are identified as implementation items in LAR Attachment S as discussed in Sections 2.7.1 and 3.1.1.6 of this SE. Compliance approach No. 6, "Complies with item for implementation," is a change from the NEI 04-02 (Reference 5) based approach in that it is a new category not included in NEI 04-02. The intent of this choice is to identify FPP elements that will comply after completion of an action by the licensee. The required actions are identified in LAR AttachmentS as implementation items. As discussed further below, the NRC staff has determined that, taken together, these methods compose an acceptable approach for documenting compliance with the NFPA 805 Chapter 3 requirements, because the licensee has followed the compliance strategies identified in the endorsed NEI 04-02 guidance document. The process defined in the endorsed guidance provides an organized structure to document each attribute in NFPA 805 Chapter 3, allowing the licensee to provide significant detail in how the program meets the requirements. In addition to the basic strategy of "Complies," which itself makes the attribute both auditable and inspectable, additional strategies have been provided allowing for amplification of information, when necessary, regarding how or why the attribute is acceptable. As discussed in SE Section 2.4.3, fire protection administrative controls refers to procedures put in place by the licensee to establish, implement, and maintain the fire protection program as required by the licensee's fire protection license condition and 10 CFR 50.48(a), 10 CFR 50.48(c), and NFPA 805, Section 3.2.3. Changes to fire protection administrative controls are controlled by the proposed fire protection license condition. SeeSE Section 4.0. The licensee stated in LAR Section 4.2.2, "Existing Engineering Equivalency Evaluation Transition," that it evaluated the EEEEs used to support compliance with the NFPA 805 Chapter 3 requirements in order to ensure continued appropriateness, quality, and applicability to the current NMP1 plant configuration. The licensee stated that EEEEs used to demonstrate compliance with NFPA 805 were not based solely on quantitative risk evaluations. The licensee determined that no EEEE used to support compliance with NFPA 805 required NRC approval. EEEEs refer to "existing engineering equivalency evaluations" (previously known as Generic Letter 86-10 evaluations) performed for fire protection design variances such as fire protection system designs and fire barrier component deviations from the specific fire protection deterministic requirements. Once a licensee transitions to NFPA 805, future equivalency evaluations are to be conducted using a PB approach. The evaluation should demonstrate that the specific plant configuration meets the performance criteria in the standard. Additionally, the licensee stated in LAR Section 4.2.3, "Licensing Action Transition," that the existing licensing actions used to demonstrate compliance have been evaluated to ensure that their bases remain valid. The results of these licensing action evaluations are provided in Attachment K of the LAR. LAR Attachment A (the NEI 04-02 8-1 Table) provides further details regarding the licensee's compliance strategy for specific NFPA 805 Chapter 3 requirements, including references to where compliance is documented.

-33-In a letter dated March 27, 2013 (Reference 8), the licensee responded to Fire Protection Engineering (FPE) Request for Additional Information (RAI) 05, and provided a table of NFPA codes and editions which will be used to meet the requirements of NFPA 805, Chapter 3 elements post-transition. 3.1.1.1 Compliance Strategy --Complies For certain NFPA 805 Chapter 3 requirements, as modified by 10 CFR 50.48(c)(2), the licensee determined that the RI/PB FPP complies directly with the fundamental FPP element using the existing FPP element. In these instances, based on the information provided by the licensee in the LAR and the information gained from the NFPA 805 site audit (the documents reviewed, discussions held with the licensee's technical staff and the plant tours performed), the NRC staff concludes that the licensee's statements of compliance are acceptable. The following NFPA 805 sections, identified in LAR B-1 Table, as complying via this method, required additional review by the NRC staff:

3.3.1.2(5)

3.4.1(c)

3.5.5

3.5.14

3.9.4

3.9.6 NFPA 805, Section 3.3.1.2(5), requires controls on the use and storage of flammable and combustible liquids in accordance with NFPA 30, "Flammable and Combustible Liquids Code," (Reference 48). In a letter dated February 27, 2013, (Reference 7), the licensee responded to FPE RAI 06(a), and revised the compliance bases to include the term "use," as required in NFPA 805, instead of "stored" only. The licensee stated there are procedural controls in place for flammable and combustible liquids. In addition, there are administrative procedures that control the use and storage of flammable and combustible liquids outside the bulk storage areas. The licensee concludes that no other NFPA standards were determined to be applicable. The NRC staff concludes that this response is acceptable as it meets the requirements of NFPA 805, Section 3.3.1.2(5), promotes the use of a consensus standard, and ensures controls are in place during the use and storage of flammable and combustible liquids. This accounts for greater protection for the equipment that may adversely affect the nuclear safety performance criteria. NFPA 805, Section 3.4.1 (c), requires every fire brigade shift shall have the fire brigade leader and at least two brigade members to have sufficient training and knowledge of nuclear safety systems to understand the effects of fire and fire suppressants on nuclear safety performance criteria. In a letter dated December 9, 2013 (Reference 1 0), the licensee responded to FPE RAI 11, and stated that they comply with NFPA 805, Section 3.4.1 (c). The licensee stated that the fire brigade leader is competent to assess the potential safety consequences of a fire and advise control room personnel. Such competence by the brigade leader is evidenced by possession of an operator's license or equivalent knowledge of plant systems. The licensee also stated that at least two other fire members are non-licensed operators having, at minimum, completed Primary and Secondary Operations watch station classroom training. The NRC staff concludes that the response describes a training and qualification program that ensures the fire brigade leader and that at least two other fire members have the appropriate minimum

-34-qualification and training to understand the potential safety consequences of a fire, plant systems, and effects of fire and fire suppressants on the nuclear safety performance criteria. NFPA 805, Section 3.5.5, requires each pump and its driver and controls shall be separated from the remaining fire pumps and from the rest of the plant by rated fire barriers. During the site audit, the NRC staff noted that fire pump remote control ability is located within the main control room. In a letter dated February 27, 2013 (Reference 7), the licensee responded to FPE RAI 07(b) and clarified, regarding the protection of these fire pump remote control circuits, that there are no fire barriers installed to protect these circuits. However, this issue was addressed in its review of Information Notice 2009-29, "Potential Failure of Fire Water Supply Pumps to Automatically Start due to a Fire," (Reference 61) and during a 1985 evaluation documented in NMP1 Safety Evaluation 85-026. The licensee stated that the review of the information notice determined that a single ground fault will not prevent automatic start of either fire pump. The licensee also stated that the 1985 evaluation determined that fire damage to the remote start circuits for the electric fire pump could prevent an automatic start; however, a modification was installed to remove this vulnerability. The licensee concluded that neither of the fire pumps is vulnerable to remote start circuit damage or failure that would prevent automatic start of the pump. The NRC staff concludes that this response is acceptable because the licensee accounted for the protection of the fire pump remote circuits by performing appropriate evaluations and installing the necessary modification to prevent the failure of an automatic start. (See Section 3.1.1.4 of this SE). NFPA 805, Section 3.5.14, requires all fire protection water supply and fire suppression system control valves to be under a periodic inspection program. NFPA 805, Section 3.9.6, requires all valves controlling water-based fire suppression systems required to meet the performance or deterministic requirements of NFPA 805, Chapter 4, be supervised as described in NFPA 805, Section 3.5.14. NFPA 805, Section 3.5.14 requires these valves be supervised by one of the following methods: (a) electrical supervision with audible and visual signals in the main control room, (b) locking valves in their normal position with keys available only to authorized personnel, or (c) sealing valves in their normal position only if located in fenced areas or under direct control of licensee. The NRC staff noted that the licensee included a fourth valve supervision option of "periodic valve position verification" which is not one of the options listed in NFPA 805 and does not provide continuous valve supervision. In a letter dated February 27, 2013 (Reference 7), the licensee responded to FPE RAI 07(d), and revised LAR Table B-1, Section 3.5.14, compliance bases to remove the words "periodic valve position verification." The NRC staff understands the periodic valve inspections are not a fourth option to satisfy the valve supervision requirement but are part of the periodic inspection program requirement. The NRC staff concludes that this response is acceptable because the revised compliance bases aligns with the requirements to provide a periodic inspection program and provide adequate, continuous valve supervision as described in NFPA 805, Section 3.5.14. NFPA 805, Section 3.9.4, requires diesel-driven fire pumps to be protected by automatic sprinklers. In a letter date February 27, 2013 (Reference 7), the licensee responded to FPE RAI 03(a) and clarified that the diesel fire pump room (Fire Area 14) is provided with a wet pipe suppression system. The licensee revised LAR Table 4-3 to state a "wet pipe system" is required and installed for the diesel fire pump room (Fire Area 14). In addition, the NRC staff requested the licensee to determine if LAR Table 4-3 should be revised to include detection

-35-systems, suppression systems, and fire protection features that are required as a result of NFPA 805, Chapter 3 (as opposed to the normal requirement via Chapter 4). In a letter dated February 27, 2013 (Reference 7), the licensee responded to FPE RAI 03(b) and stated they performed a review of LAR Table 4-3 and the NFPA 805, Chapter 3 requirements and determined that only one revision was necessary. The licensee revised LAR Table 4-3, Fire Area 14 (diesel fire pump room) to include, "Wet Pipe System required for Chapter 3, Section 3.9.4, compliance; Detection." The NRC staff concludes that this response is acceptable because the licensee has clearly identified those fire protection systems and features required by NFPA 805, Chapter 3. 3.1.1.2 Compliance Strategy --Complies with Clarification For certain NFPA 805, Chapter 3 requirements, the licensee provided additional clarification when describing its means of compliance with the fundamental FPP element. In these instances, the NRC staff reviewed the additional clarifications and concludes that the licensee will meet the underlying requirement for the FPP element as clarified. 3.1.1.3 Compliance Strategy --Complies with Use of EEEEs NFPA 805, Chapter 3 requirements, the licensee demonstrated compliance with the fundamental FPP element through the use of EEEEs. The NRC staff reviewed the licensee's statement of continued validity for the EEEEs and the statement on the quality and appropriateness of the evaluations, and concludes that the licensee's statements of compliance in these instances are acceptable. The following NFPA 805 sections identified in LAR Table B-1 as complying via this method required additional review by the NRC staff:

3.3.8 Bulk Storage of Flammable and Combustible Liquids NFPA 805, Section 3.3.8, does not permit bulk storage of flammable and combustible liquids inside structures containing systems, equipment, or components import to nuclear safety. Based on the information provided by the licensee in the LAR, the NRC staff concludes that NMP1 meets this portion of the requirement. In addition, NFPA 805, Section 3.3.8, requires, as a minimum, the storage and use of flammable and combustible liquids comply with NFPA 30, "Flammable and Combustible Liquids Code." In the original submittal of LAR Table B-1, Section 3.3.8, the licensee referenced an EEEE code compliance evaluation for NFPA 30. However, the compliance basis was unclear whether the use of flammable and combustible liquids will comply with NFPA 30. In a letter dated February 27, 2013 (Reference 7), the licensee responded to FPE RAI 06(c) and revised the compliance bases to include the term "use," as required in NFPA 805, instead of storage only. NRC staff concludes that this response is acceptable as it meets the requirements of NFPA 805, Section 3.3.8, relies on a consensus standard (NFPA 30), and ensures adequate controls are in place

-36-during the use and storage of flammable and combustible liquids. This accounts for greater protection for the equipment that may adversely affect the nuclear safety performance criteria.

3. 7 Fire Extinguishers In a letter dated February 27, 2013 (Reference 7), the licensee responded to FPE RAI 01 a and clarified that the correct edition of NFPA 10, "Standard for Portable Fire Extinguishers," (Reference 50) is the 1998 edition rather than NFPA 10-1975. The NRC staff concludes that this response is acceptable, because the licensee identified the proper edition of the standard being used.

3.8.1 Fire Alarm In a letter dated February 27, 2013 (Reference 7), the licensee responded to FPE RAI 01 b and clarified that for the referenced document EIR 51-9077284-000, the correct appendix reference is Appendix J. The NRC staff concludes that this response is acceptable because the licensee identified the proper reference being used.

3.1 0.3 Gaseous Fire Suppression Systems-Ventilation System Design The LAR, Table B-1, Section 3.1 0.3, compliance bases discuss adequate sealing and discharge testing for Halon 1301 systems. However, the compliance bases were unclear if all of the other gaseous fire suppression systems were verified for adequate sealing and over pressurization, especially the carbon dioxide (C02) systems. In a letter dated February 27, 2013 (Reference 7), the licensee responded to FPE RAI 09 and stated that the design of the original C02 suppression systems were accepted by the NRC in the July 26, 1979 safety evaluation report (Reference 18) and followed the 1968 edition of NFPA 12," Standard on Carbon Dioxide Extinguishing Systems," (the original code of record) (Reference 49). The licensee noted that NFPA 12-1968, did not define a hold time or require a full discharge test. Although not required, a representative set of rooms were full discharge tested by the licensee to establish timer settings. The licensee also states that all of the original C02 systems were "puff tested" with no indications of leakage or overpressure issues in those test records. NMP1 has since added two C02 systems (Seal Oil Unit Enclosure and Lube Oil room-Elevation 261'}. These two systems were full discharge tested in 1981 to a hold time of 1 0 minutes and any leakages found were corrected. The NRC staff concludes that this response is acceptable as it shows these systems were tested and sealed beyond the requirements of the governing code (NFPA 12-1968} at the time these systems were designed and installed.

3.11.2 Fire Barriers During a sample review of a few of the EEEE's credited to meet NFPA 805, Chapter 3, the NRC staff noted a Promat-H fire barrier enclosure located within

-37-Fire Area 5, Fire Zone T3B. This Promat-H fire barrier covers a heating ventilation and air conditioning duct that enters into the Aux Control Room. This feature was noted in an EEEE (FPEE 95-002) referenced for LAR Table B-1, Section 3.11.5, regarding electrical raceway fire barrier systems (ERFBS); however, the NRC staff questioned whether this feature should be listed as part of the fire barrier separating fire areas and thus credited under LAR Table B-1, Section 3.11.2. In a letter dated February 27, 2013 (Reference 7), the licensee responded to FPE RAI 1 Oa and confirmed that this Promat-H enclosure is a credited fire barrier and is considered to be part of the fire barrier separating the Auxiliary Control Room (Fire Area 11) and the Turbine Building (Fire Area 5). The licensee revised LAR Table B-1 , Section 3. 11.2, and LAR Table B-3, Fire Area 5, to reference this EEEE (FPEE 95-002). LAR Table B-1, Section 3.11.2, was also revised to describe this credited enclosure as part of the fire barrier. LAR Table 4-3, Fire Area 5, Zone T3B, was also revised to describe this fire barrier enclosure and identified it as "code 'S' -required for separation" to meet NFPA 805 requirements. The NRC staff concludes that this response is acceptable because the licensee clarified the proper classification of this feature and provides the adequate documentation revisions to the LAR.

3.11.3(1) & (2) Fire Barrier Penetrations The LAR, Table B-1, Section 3.11.3(1 ), compliance bases discuss the fire testing and listing of fire doors and references NFPA 252, "Standard Methods of Fire Tests of Door Assemblies," (Reference 62). The LAR, Table B-1, Section 3.11.3(2), compliance bases discuss the fire testing and listing of fire dampers and references UL 555, "Standard for Fire Dampers," (Reference 63). However, both compliance bases were unclear if the NMP1 fire doors and fire dampers conformed to NFPA 80, "Standard for Fire Doors and Fire Windows," (Reference 64) and NFPA 90A, "Standard for the Installation of Air Conditioning and Ventilating Systems," (Reference 65), respectively, as required by NFPA 805, Section 3.11.3. NFPA 80 and 90A provide the requirements for the testing and listing as well for the installation. NFPA 252 and UL 555 only provide for the testing requirements and are referenced within NFPA 80 and 90A, respectively. In a letter dated February 27, 2013 (Reference 7), the licensee responded to FPE RAI 1 Ob and revised LAR, Table B-1, Section 3.11.3, to clarify that they are in conformance with NFPA 80 and NFPA 90A. The licensee also stated that they have certain deviations which are identified with justification in the Updated Final Safety Report (UFSAR), Appendix 1 OA, Table 1.2.2, as well as in the EEEEs referenced in LAR Table B-1, Section 3.11.3. The NRC staff concludes that this response is acceptable because the licensee conforms to the required use of NFPA 805 (Reference 1) and NFPA 90A, has identified and documented evaluations for any deviations, and LAR Section 4.2.2 indicates that all EEEEs credited for NFPA 805 were reviewed using the methodology contained in NEI 04-02.

-38-* 3.11.4(b) Through Penetration Fire Stops-Internal Conduit Seals The LAR, Table B-1, Section 3.11.4(b}, compliance bases states, in part, that conduit smaller than 2-in diameter do not require internal [fire] seals. In a letter dated February 27,2013 (Reference 7), the licensee responded to FPE RA110c and revised LAR, Table B-1 , Section 3.11 .4(b) to clarify that the basis for this criterion is an NRC letter to Wisconsin Electric Power Company (WEPCO), "Review of Draft Safety Evaluation of Conduit Fire Seal Topical Report," dated October 23, 1989 (Reference 66). In this letter, the NRC staff provided a safety evaluation of WEPCO's Conduit Fire Seal Topical Report including discussions of various NRC staff concerns with the topical report. WEPCO's topical report described the testing program, the results, and proposed design guidelines for internal conduit fire seals. The proposed design guidelines stated, in part, include: "open conduits that terminate at the wall should be treated as sleeves and should be sealed with rated seals." "conduits that terminate in junction boxes ... need no additional sealing." "conduits that run through an area but do not terminate in that area need not be sealed in that area.' "conduits smaller than 2-in diameter that terminate one foot or greater from the barrier need not be sealed." "open conduits of 2-in diameter that terminate three feet or greater from the barrier need not be sealed." The NRC staff concludes that this response is acceptable because the revised compliance bases align with the design guidelines for internal conduit fire seals as previously accepted in the referenced NRC letter dated October 23, 1989.

3.11.5 Electrical Raceway Fire Barrier Systems (ERFBS) The LAR, Table B-1, Section 3.11.5, compliance bases discuss the only ERFBS credited to meet NFPA 805, Chapter 4 requirements. This Promat-H ERFBS resides within a missile shield enclosure located in emergency diesel generator 102 missile enclosure (Fire Area 18}. This is a 3-hour fire rated ERFBS credited by the nuclear safety capability assessment (NSCA) to protect conduit 171-66. The original submittal stated that this ERFBS, although tested, was not tested in accordance with GL 86-10, Supplement 1 (Reference 67), as required by NFPA 805, Section 3.11.5. LAR, Table B-1, Section 3.11.5, identifies an EEEE (FPEE 95-002) that deems the qualification testing that was performed to be adequate and deemed this Promat-H ERFBS capable of a 3-hour fire resistance rating.

-39-In a letter dated February 27, 2013 (Reference 7), the licensee responded to FPE RAI 1 O(d)(i) and provided further information regarding the following variances between the fire endurance test performed and the testing requirements of GL 86-10, Supplement 1 (Reference 67): Tested configuration was a horizontal wall and not an ERFBS assembly Electrical raceway components, such as conduits, were not included in the test configuration Thermocouples were placed directly to the unexposed surface of the barrier and not on the surface of the protected raceway component Thermocouples were placed in 13-inch to 21-inch intervals instead of 6-inch intervals The licensee justified the acceptability of these variances based on: use of the GL 86-10, Supplement 1 thermal performance acceptance criteria, use of felt pads over the thermocouples to prevent temperature dissipation with the larger placement intervals, and the expectation that temperatures recorded on the direct unexposed surface of the barrier would exceed any recorded on the raceway component surfaces. The licensee also noted that the referenced EEEE, FPEE 95-002, identified these variances. In a letter dated February 27, 2013 (Reference 7), the licensee responded to FPE RAI 1 O(d)(ii) and provided more details regarding the following differences between the installed configuration and the tested configuration: Use of #6 x 1-1/2-inch deck screws instead of #6 x 1-5/8-inch bugle head screws when fastening to supports Use of #8 x 3-inch deck screws instead of 7/16-inch x 1-1/4-inch x #16 GA staples when fastening layer-to-layer Fastener spacing of 9-inch from bottom edge instead of 6-inch-on-center when fastening the bottom edge of the bottom layer All other fastener spacing is 5-inch to 7-inch instead of 6-inches-on-center Contains four penetrating commodities instead of none Horizontal overlap distance of 3-inch to 4-inch instead of 12-inch The licensee justified the acceptability of these variances based on: the use of longer stronger fasteners, the bottom layer covered with three additional layers, a fastener spacing tolerance of +1 inch being minor, the penetrations sealed at each layer with fire resistant caulk, the penetration gap limited to 5/16-inch, and

-40-the short overlaps sealed at each layer with fire resistant caulk. The licensee also noted that the referenced EEEE, FPEE 95-002, documents these differences. During the audit, the NRC staff noted that this ERFBS is located inside a robust missile shield enclosure with very limited access, very limited combustibles, and no other ignition sources. The robustness of the missile shield enclosure inherently provides additional thermal protection from a fire event in the emergency diesel generator room. Based on the information submitted by the licensee and the inherent protection of the surrounding missile shield enclosure, the NRC staff concludes this response is acceptable because there are limited combustibles in the area, and the variances from the requirements of NFPA 805, Section 3.11.5 are mostly minor and are compensated by the added protection of the missile shield enclosure. In a letter dated February 27, 2013 (Reference 7), the licensee responded to FPE RAI 01 c and clarified that the correct reference is FPEE 95-002 rather than FPEE 95-001. The NRC staff concludes that this response is acceptable because the licensee identified the proper reference being used. 3.1.1.4 Compliance Strategy--Complies via Previous NRC Approval Certain NFPA 805 Chapter 3 requirements, or partial requirements, were specifically supplanted by an alternative that was previously approved by the NRC. Remaining portions of an element not directly stated in the previous approval either comply directly with the stated requirement or are noted as otherwise in the compliance statement(s). The approval was documented in the original 1979 FPP Safety Evaluation Report (Reference 18). The following NFPA 805 section identified in LAR Table B-1 as complying via this method required additional review by the NRC staff:

3.5.5 Water Supply-Fire Pump Separation During the site audit, the NRC staff noted that NMP1 has a lack of fire barrier separation between the electric fire pump and the rest of the plant. In a letter dated February 27, 2013 (Reference 7), the licensee responded to Fire Protection Engineering RAI 07(a) and stated the fire pump arrangement was previously approved in the July 26, 1979 NRC safety evaluation. The licensee revised the compliance statement for Section 3.5.5 of LAR Table B-1 to include, "Complies by previous NRC approval," and referenced the 1979 safety evaluation. The NRC staff concludes that this response is acceptable because the licensee is allowed to claim previous approval for NFPA 805, Chapter 3 elements and the licensee has provided excerpts from the appropriate documentation for the lack of fire barrier separation between the electric fire pump and the rest of the plant. (See SE Section 3.1.1.1)

-41 -In each instance, the licensee evaluated the bases for the original NRC approval and determined that in all cases the bases were still valid. The NRC staff reviewed the information provided by the licensee and concludes that previous NRC approval had been demonstrated using suitable documentation that meets the approved guidance contained in RG 1.205, Revision 1 (Reference 2). Based on the licensee's justification for the continued validity of the previously approved alternatives to the NFPA 805, Chapter 3 requirements, the NRC staff concludes that the licensee's statements of compliance in these instances are acceptable. 3.1.1.5 Compliance Strategy--Submit for NRC Approval The licensee also requested approval for the use of performance-based methods to demonstrate compliance with fundamental FPP elements. In accordance with 10 CFR 50.48(c)(2)(vii), the licensee requested specific approvals be included in the license amendment approving the transition to NFPA 805 at NMP1. The NFPA 805 sections identified in LAR Table B-1 as complying via this method are as follows:

3.3.5.1, which concerns the use of non-plenum-use wiring routed above suspended ceilings. See SE Section 3.1.4.1, LAR Attachment L, Approval Request #1.

3.2.3(1 ), which concerns the use of Electric Power Research Institute (EPRI) Technical Report TR-1006756, "Fire Protection Surveillance Optimization and Maintenance Guide for Fire Protection Systems and Features," (Reference 68). In a letter dated February 27, 2013 (Reference 7), the licensee responded to FPE RAI 04(a) and created a new LAR Attachment L request for approval to use a PB method to establish the appropriate inspection, testing, and maintenance frequencies for fire protection systems and features required by NFPA 805, as described in EPRI TR-1 006756. See SE Sections 3.1.1.6 and 3.1.4.2, LAR Attachment L, Approval Request #2. As discussed in SE Section 3.1.4 below, the NRC staff concludes that the use of PB methods to demonstrate compliance with these fundamental FPP elements is acceptable. 3.1.1.6 Compliance Strategy -Complies, with item for implementation For certain NFPA 805 Chapter 3 requirements, as modified by 10 CFR 50.48(c)(2), the licensee determined that the RI/PB FPP will comply with the fundamental FPP element after completion of a required action. The required actions were identified as follows:

3.2.2.4 Management Policy Direction and Responsibility-Identify AHJ The licensee plans to revise the fire protection policy document to identify the appropriate AHJ for the various areas of the program. Completion of this action is Implementation Item 7 in LAR Attachment S, Table S-2.

-42-* 3.2.3(1) Procedures-Inspection, Testing, and Maintenance The licensee plans to use the PB method as described in EPRI TR-1 006756, (Reference 68) to update the fire protection surveillance frequencies. In a letter dated February 27, 2013 (Reference 7), the licensee responded to FPE RAI 04(a) and provided the requested approval request in the revised LAR Attachment L. The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee provided a request for approval. For the NRC staff evaluation, see SE Sections 3.1.1.5 and 3.1.4.2, LAR Attachment L, Approval Request #2.

3.2.3(3) Procedures-Review of Fire Protection Program The licensee plans to include in the monitoring program, required by NFPA 805, a process that monitors and trends the FPP based on specific goals established to measure effectiveness. Completion of this action is Implementation Item 9 in LAR Attachment S, Table S-2.

3.3.1.2(2) Control of Combustible Materials -Plastic Sheeting In a letter dated March 27, 2013 (Reference 8), the licensee responded to FPE RAI 05(b) and clarified that the specific editions of certain referenced codes "will be cited," during the transition period. The licensee clarified that, at a minimum, the edition specified in NFPA 805 Chapter 6 will be used but that later editions may be adopted to enable purchasing of material. The licensee provided a revised implementation item to implement administrative controls to ensure plastic sheeting materials used in the power block are qualified in accordance with NFPA 701 (Reference 69) or equivalent. The 1999 edition of NFPA 701, "Standard Methods of Fire Tests for Flame Propagation of Textiles and Films," referenced by NFPA 805 will be used as a minimum; however, a specific edition of NFPA 701 will be cited at the time the material is purchased. Completion of this action is Implementation Item 11 in LAR Attachment S, Table S-2. The NRC staff concludes that the licensee response to the RAI is acceptable because the licensee provided an implementation item that will provide for the control of plastic sheeting materials used in the power block.

3.3.1.2(6) Control of Combustible Materials -Use and Storage of Flammable Gases In a letter dated February 27, 2013 (Reference 7), the licensee responded to FPE RAI 06{b) and revised the compliance bases to use the term "important to nuclear safety," instead of "safety-related." The term "important to safety" refers to a larger set of equipment than the term "safety-related," originally used by NMP1. The NRC staff concludes that the licensee's response is acceptable as it accounts for greater protection for the equipment that may adversely affect the nuclear safety performance criteria. The licensee developed Modification Item

-43-11 in LAR AttachmentS, Table S-1 to relocate the Hydrogen Stand-By Supply bottles. (See VFDR-05-026 and VFDR-05-027).

3.3.1.3.2 Control of Ignition Sources -Smoking The licensee plans to incorporate the station smoking policy into a formal, controlled process by either revising an existing procedure or creating a new procedure. Completion of this action is Implementation Item 12 in LAR Attachment S, Table S-2.

3.3.1.3.4 Control of Ignition Sources -Portable Heaters The licensee plans to implement administrative controls for the use of portable electric heaters in the plant and to prohibit the use of portable fuel-fired heaters in power block structures. Completion of this action is Implementation Item 13 in LAR AttachmentS, Table S-2.

3.3.2 Structural In a letter dated March 27, 2013 (Reference 8), the licensee responded to FPE RAI 05(b) and clarified that the specific editions of certain referenced codes "will be cited," during the transition period. The licensee clarified that at a minimum the edition specified in NFPA 805 Chapter 6 will be used but that later editions may be adopted to enable purchasing of material. The licensee provided a revised implementation item to implement administrative controls to utilize, to the extent practicable, noncombustible construction as defined by NFPA 220, "Standard on Types of Building Construction," (Reference 70) for walls, floors, and components of new power block buildings and changes to existing power block buildings that are required to maintain structural integrity. The 1999 edition of NFPA 220 referenced by NFPA 805 will be used as a minimum; however, a specific edition of NFPA 220 will be cited at the time that material is purchased, construction of new power block buildings is performed, or changes to existing power block buildings are performed. Completion of this action is Implementation Item 14 in LAR AttachmentS, Table S-2. The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee provided an implementation item that will provide controls for the use of non-combustible building materials in the construction of new power block buildings and in changes to existing power block buildings.

3.3.3 Interior Finishes In a letter dated March 27, 2013 (Reference 8), the licensee responded to FPE RAI 05(b) and clarified that the specific editions of certain referenced codes "will be cited," during the transition period. The licensee clarified that at a minimum the edition specified in NFPA 805 Chapter 6 will be used but that later editions may be adopted to enable purchasing of material. The licensee provided a revised implementation item to implement administrative controls to utilize, to the

-44-extent practicable, Class A materials for new interior wall or ceiling finishes, and Class I materials for new interior floor finishes as defined in NFPA 101, "Life Safety Code," (Reference 47). The 2000 edition of NFPA 101 referenced by NFPA 805 will be used as a minimum; however, a specific edition of NFPA 101 will be cited at the time material is purchased or changes are performed. Completion of this action is Implementation Item 15 in LAR Attachment S, Table S-2. The NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee provided an implementation item that will provide controls for the use of Class A materials for new interior wall or ceiling finishes, and Class I materials for new interior floor finishes.

3.3.6 Roofs The licensee plans to incorporate the requirement for new metal roof deck implementation construction to be designed and installed so the roofing system will not sustain a self-propagating fire on the underside when exposed to a fire inside the building. Additionally, roof coverings shall be Class A as determined by tests described in NFPA 256, "Standard Methods of Fire Tests of Roof Coverings," (Reference 71 ). Completion of this action is Implementation Item 16 in LAR Attachment S, Table S-2.

3.3. 7 Bulk Flammable Gas Storage In a letter dated February 27, 2013 (Reference 7), the licensee responded to FPE RAI 06(b) and revised the compliance bases to use the term "important to nuclear safety," as required in NFPA 805, instead of "safety-related." The term "important to safety" refers to a larger set of equipment than the term related," originally used by NMP1. The NRC staff concludes that the licensee's response is acceptable as it meets the requirements of NFPA 805, Section 3.3.7, and accounts for greater protection for the equipment that may adversely affect the nuclear safety performance criteria. The licensee developed Modification Item 11 as described in LAR AttachmentS, Table S-1 to relocate the hydrogen stand-by supply bottles. (See VFDR-05-026 and VFDR-05-027).

3.3.7.1 Bulk Flammable Gas Storage-Hydrogen Storage in accordance with NFPA 50A The licensee developed Modification Item 11 in LAR AttachmentS, Table S-1 to relocate the hydrogen stand-by supply bottles. (See VFDR-05-026 and VFDR-05-027).

3.3.1 0 Hot Pipes and Surfaces The licensee plans to incorporate the need for plant tours and normal housekeeping activities to inspect for lubricating oil coming in contact with hot pipes and surfaces, including insulated pipes and surfaces into appropriate plant

-45-procedures. Completion of this action is Implementation Item 17 in LAR Attachment S, Table S-2.

3.4.3(a)(2) Training and Drills-Industrial Fire Brigade Training The licensee plans to revise the fire brigade training program to include radioactivity and health physics considerations in quarterly fire brigade meetings. Completion of this action is Implementation Item 18 in LAR Attachment S, Table S-2. On the basis that the "required actions" as described by the licensee will incorporate the provisions of NFPA 805, Chapter 3 in the licensee's FPP and plant procedures and that the actions are included as modifications or implementation items in LAR Attachment S, which are required by the proposed license condition, the NRC staff concludes that the licensee's statements of compliance are acceptable. 3.1.1.7 Compliance Strategy-Multiple Strategies In certain compliance statements of the NFPA 805, Chapter 3 requirements, the licensee used more than one of the above strategies to demonstrate compliance with aspects of the fundamental element. In each of these cases, the NRC staff concludes that the individual compliance statements are acceptable, that the combination of compliance strategies is acceptable, and that the licensee demonstrated compliance with the NFPA 805 Chapter 3 fundamental FPP elements and minimum design requirements. 3.1.1.8 Chapter 3 Sections Not Reviewed Some NFPA 805, Chapter 3 sections either do not apply to the transition to a RI/PB FPP or have no technical requirements. Accordingly, the NRC staff did not review these sections for acceptability. The sections that were not reviewed fall into one of the following categories:

Sections that do not contain any technical requirements. (e.g., NFPA 805 Sections 3.4.5 and 3.11 ).

Sections that are not applicable to NMP1 because of the following: The licensee states that NMP1 does not have systems installed of this type (as described below):

3.6.5, which concerns cross-connection between seismic required hose stations and essential seismic non-fire protection water supply systems. NMP1 does not have seismic required hose stations.

-46-* 3.8.1 (3), which concerns the actuation of any manual fire alarm station. Manual pull fire alarm boxes are not used at NMP1.

3.9.1 (3), which concerns water mist systems installed in accordance with NFPA 750. NMP1 does not have a water mist system installed.

3.9.1 (4), which concerns foam-water systems installed in accordance with NFPA 16. NMP1 does not have a foam-water system installed.

3.1 0.1 (3), which concerns gaseous clean agent fire suppression systems designed and installed in accordance with NFPA 2001. NMP1 does not have a clean agent fire suppression system installed.

The requirements are structured with an applicability statement (e.g., NFPA 805 Chapter 3 Section 3.3.12, which applies to reactor coolant pumps in non-inerted containments; or Sections 3.4.1 (a)(2) and 3.4.1 (a)(3), on which code(s) apply to the fire brigade depends on the type of brigade specified in the FPP at the site; or Section 3.5.1 (b), which applies to calculating the fire protection water supply). 3.1.1.9 Compliance with Chapter 3 Requirements Conclusion As discussed above, the NRC staff evaluated the results of the licensee's assessment of the proposed NMP1 RI/PB FPP against the NFPA 805, Chapter 3, fundamental FPP elements and minimum design requirements, as modified by the exceptions, modifications, and supplementations in 10 CFR 50.48(c)(2). Based on this review of the licensee's submittal, as supplemented, the NRC staff concludes that the RI/PB FPP is acceptable with respect to the fundamental FPP elements and minimum design requirements of NFPA 805, Chapter 3, as modified by 10 CFR 50.48(c)(2), because the licensee:

Used an overall process consistent with NRC staff approved guidance to determine the state of compliance with each of the applicable NFPA 805, Chapter 3 requirements.

Provided appropriate documentation of NMP1 's state of compliance with the NFPA 805, Chapter 3 requirements, which adequately demonstrated compliance by substantiating that the licensee complied: With the requirement directly, or with the requirement directly after the completion of an implementation item; With the intent of the requirement (or element) and provided adequate justification; Via previous NRC staff approval of an alternative to the requirement;

-47-Through the use of an engineering equivalency evaluation; Through the use of a combination of the above methods; and Through the use of a PB method that the NRC staff has specifically approved in accordance with 10 CFR 50.48(c)(2)(vii). 3.1.2 Identification of the Power Block The NRC staff reviewed the NMP1 structures identified in LAR Table 1-1 "Power Block Definition" as comprising the "power block." The NMP1 structures listed are established as part of the power block for the purpose of denoting the structures and equipment included in the NMP1 Rl/PB FPP that have additional requirements in accordance with 10 CFR 50.48(c) and NFPA 805. As stated in the LAR, Section 4.1.3, the power block includes structures that have equipment required for nuclear plant operations. In a letter dated February 27, 2013 (Reference 7), the licensee responded to FPE RAI 02 and stated that the fire area term YARD (or EXT) is treated as a catch-all for components that are exterior to the plant (e.g., outside of the buildings) but still within the protected area and the global analysis boundary. The licensee also stated that the parenthetical listing in LAR Attachment I is not all inclusive and is intended only to give examples of the equipment that can be found in the YARD area. In addition, the licensee's response clarified that pieces of equipment (e.g., the Bulk Nitrogen Storage Tank, transformers outside the Turbine Building, etc.) are included in the analysis, but may not be specifically listed in LAR Attachment I. The NRC staff concludes that the licensee evaluated the structures and equipment at NMP1, and adequately documented a list of those structures that fall under the definition of "power block" in NFPA 805. 3.1.3 Closure of Generic Letter 2006-03, "Potentially Nonconforming HemycŽ and MTŽ Fire Barrier Configurations," Issues GL 2006-03 (Reference 46) requested that licensees evaluate their facilities to confirm compliance with existing applicable regulatory requirements in light of the results of NRC testing that determined that both HemycŽ and MTŽ fire barriers failed to provide the protective function intended for compliance with existing regulations, for the configurations tested using the NRC's thermal acceptance criteria. NMP1 does not use either the HemycŽ or MTŽ electrical raceway fire barrier systems. The NRC staff confirmed this with NMP1 's response to GL 2006-03, in a letter dated June 9, 2006 (Reference 72), and the information provided in NUREG-1924, "Electric Raceway Fire Barrier Systems in U.S. Nuclear Power Plants," (Reference 73) which indicate that neither material is installed in NMP1. Therefore, the generic issue (GL 2006-03) related to the use of these ERFBS is not applicable to NMP1. 3.1.4 Performance-Based Methods for NFPA 805, Chapter 3 Elements In accordance with 10 CFR 50.48(c)(2)(vii), a licensee may request NRC approval for use of the PB methods permitted elsewhere in the standard as a means of demonstrating compliance with the prescriptive NFPA 805, Chapter 3, Fundamental FPP Elements and Minimum Design

-48-Requirements. Paragraph 50.48(c)(2)(vii) of 10 CFR requires that an acceptable PB approach accomplish the following: (A) Satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release; (B) Maintains safety margins; and (C) Maintains fire protection defense-in-depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire safe shutdown capability). In LAR Attachment L, "NFPA 805, Chapter 3 Requirements for Approval (1 0 CFR 50.48(c)(2)(vii))," the licensee requested NRC staff review and approval of PB methods to demonstrate an equivalent level of fire protection for the requirement of the elements identified in SE Section 3.1.1.5. The NRC staff evaluation of these proposed methods is provided below. 3.1.4.1 NFPA 805, Section 3.3.5.1 -Electrical Wiring Above Suspended Ceiling In LAR Attachment L, Approval Request #1, the licensee requested NRC staff review and approval of a PB method to demonstrate an equivalent level of fire protection for the requirement of NFPA 805, Section 3.3.5.1 regarding wiring above suspended ceilings. Specifically, the licensee has three exposed cable trays routed above a suspended ceiling in the Radwaste Control Room (Fire Area 15) that are not routed in metal conduits, covered cable trays, armored cable, or rated for plenum use as required by NFPA 805, Section 3.3.5.1. During the audit, the NRC staff identified that LAR Attachment L, Approval Request 1, did not contain sufficient discussions concerning the amount of exposed wiring, whether the exposed wiring is credited for any nuclear safety performance criteria, or the fire separation between Fire Area 15 and those areas of the plant credited to meet NFPA 805. In a letter dated February 27, 2013 (Reference 7), the licensee responded to FPE RAI 04(b) and provided a revised LAR Attachment L, Approval Request #1. As described by the licensee, the non-compliant wiring is located in three cable trays above the ceiling in the Radwaste Control Room (Fire Area 15). These cable trays are not covered top or bottom with solid metal covers and contain a considerable amount of wiring. In the response, the licensee also provided the administrative controls used to ensure future cable construction meets NFPA 805 requirements. The licensee stated that there are no cables or components related to nuclear safety performance criteria within Fire Area 15 and that Fire Area 15 is separated from the Turbine Building by 3-hour rated fire barriers with detection provided above the suspended ceiling. Therefore, the licensee stated there is minimal impact on the nuclear safety performance criteria due to the presence of uncovered cable trays above the Radwaste Control Room suspended ceiling. The licensee also stated the location and design of the cable trays above suspended ceilings has no impact on the NFPA 805 radiological release performance criteria.

-49-In addition, the licensee stated that the impact of the three uncovered cable trays above the Radwaste Control Room suspended ceiling is minor such that the safety margins inherent in the licensee's analyses are preserved. The licensee stated that fire protection DID will be maintained, because these cable trays do not compromise the administrative fire prevention controls, and do not challenge automatic fire detection, manual fire suppression, or post-fire safe shutdown capability. Concerning echelon one of DID, the licensee added that the area above suspended ceiling is free of work that could cause a fire and any such work is controlled via hot work permits. Concerning echelon two, the licensee stated there is a fire detection system above the suspended ceiling in Fire Area 15 and the cable trays do not challenge the manual suppression function. Concerning echelon three, the licensee indicated that the equipment and cables in Fire Area 15 are not credited for NFPA 805 and that 3-hour fire rated barriers separate Fire Area 15 from other areas of the plant (e.g., Turbine Building); therefore, the licensee stated that there is adequate protection of the equipment and cables relied upon for safe shutdown and meet the nuclear safety performance criteria. The NRC staff concludes that this request for PB method approval is acceptable based on:

The spaces above ceilings receive little access and are not used for storage/staging of combustibles, therefore, there is reasonable assurance that transient combustibles will not be present including during outages;

Procedures and permits exist to control work that may cause a fire (e.g., hot work);

Visual inspection and drawing reviews that show that there are no ignition sources in the areas above the suspended ceilings of Fire Area 15;

Fire detection is installed within the suspended ceiling above the cable trays;

Future wiring or modified wiring designed and installed in a suspended ceiling, including Fire Area 15, will be required to meet NFPA 805, Section 3.3.5.1 requirements per the administrative controls in place;

Fire Area 15, including these three non-covered cable trays, does not contain any cabling or equipment related to any Nuclear Safety Performance Criteria functions; and

Fire Area 15 is separated from the rest of the plant by 3-hour fire barriers such that a fire within Fire Area 15 will not impact other areas of the plant that do contain safe shutdown equipment. Based on its review of the information submitted by the licensee, and in accordance with 10 CFR 50.48(c)(2)(vii), the NRC staff concludes that the proposed PB method is an acceptable alternative to the corresponding NFPA 805, Section 3.3.5.1, requirement because it satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805

-50-related to nuclear safety and radiological release, maintains sufficient safety margin, and maintains adequate fire protection defense-in-depth. 3.1.4.2 NFPA 805, Section 3.2.3(1) -Inspection, Testing, and Maintenance Procedures As discussed in SE Sections 3.1.1.5 and 3.1.1.6, the NRC staff requested information regarding the compliance strategy for NFPA 805 Section 3.2.3(1). In a letter dated February 27, 2013 (Reference 7), the licensee responded to FPE RAI 04(a) and supplemented LAR Attachment L with an additional approval request. In LAR Attachment L, Approval Request #2, the licensee requested NRC staff review and approval of a PB method to demonstrate an equivalent level of fire protection for the requirements regarding inspection, testing, and maintenance of credited fire protection systems and features. Specifically, the licensee requested approval to use PB methods to establish the appropriate inspection, testing, and maintenance frequencies for fire protection systems and features. As described by the licensee, the inspection, testing, and maintenance frequencies would be established and adjusted using the PB methods contained in EPRI Technical Report TR-1 006756, "Fire Protection Surveillance Optimization and Maintenance Guide for Fire Protection Systems and Features", Final Report, July 2003 (Reference 68). These PB methods will be used for the fire protection systems and features required by NFPA 805. The licensee clarified that the use of EPRI TR-1006756 is not to establish the scope of the inspection, testing, and maintenance activities which have been established based on the previously approved Technical Specifications, Licensee Controlled Documents, the appropriate NFPA codes, and as determined by the required systems review identified in LAR Table 4-3. In addition, the licensee stated that use of this method will be in conjunction with the monitoring program required by NFPA 805, Section 2.6. The licensee stated that there is no adverse impact on the nuclear safety performance criteria because the use of PB test frequencies established per EPRI TR-1006756 methods combined with NFPA 805, Section 2.6, "Monitoring Program," will provide assurance that the availability and reliability of the fire protection systems and features are maintained to the levels assumed in the NFPA 805 engineering analyses. The licensee stated that the radiological release performance criteria are satisfied based on the determination of limiting radioactive release (See LAR Attachment E). Some fire protection systems and features are credited as part of that evaluation and the use of PB test frequencies established per EPRI TR-1 006756 methods combined with NFPA 805, Section 2.6, "Monitoring Program," should ensure that the availability and reliability of the fire protection systems and features are maintained to the levels assumed in the NFPA 805 engineering analyses which includes those assumptions credited to meet the radioactive release performance criteria. Therefore, there is no adverse impact on meeting these criteria. The licensee further stated that the use of PB test frequencies established per EPRI TR-1006756 combined with NFPA 805, Section 2.6, Monitoring Program preserves the safety margin inherent and credited in the analysis because it provides assurance that the availability and reliability of the systems and features are maintained to the levels assumed in the NFPA 805 engineering analyses which includes those assumptions credited in the risk

-51 -evaluation safety margin discussions. In addition, the licensee stated the use of these methods should not invalidate the inherent safety margins contained in the codes used for design and maintenance of fire protection systems and features. Therefore, the safety margin inherent and credited in the analyses should be preserved. In NFPA 805, Section 1.2, the three echelons of DID described are: (1) to prevent fires from starting (combustible/hot work controls), (2) rapidly detect, control and extinguish fires that do occur thereby limiting damage (fire detection systems, automatic fire suppression, manual fire suppression, pre-fire plans}, and (3) provide adequate level of fire protection for systems and structures so that a fire will not prevent essential safety functions from being performed (fire barriers, fire rated cable, success path remains free of fire damage, recovery actions). The licensee stated that EPRI TR-1 006756 PB methods will be used to establish PB inspection, testing, and maintenance frequencies for the credited fire protection systems and features but will not be used with administrative controls that prevent fires from starting. Therefore, the EPRI TR-1 006756 PB methods do not affect echelon 1. For echelons 2 and 3, the use of PB test frequencies established per EPRI TR-1 006756 methods, combined with the NFPA 805, Section 2.6, monitoring program, provide assurance that the availability and reliability of the fire protection systems and features credited for DID are maintained to the levels assumed in the NFPA 805 engineering analyses. Therefore, there is no adverse impact to DID echelons 2 and 3. Based on its review of the LAR as supplemented, and in accordance with 10 CFR 50.48(c)(2)(vii), the NRC staff concludes that the proposed PB method is an acceptable alternative to the corresponding NFPA 805, Section 3.2.3(1) requirement because it satisfies the performance goals, objectives, and criteria specified in NFPA 805 related to nuclear safety and radiological release, maintains sufficient safety margin, and maintains adequate fire protection DID. 3.2 Nuclear Safety Capability Assessment (NSCA) Methods NFPA 805 is an RI/PB standard that allows engineering analyses to be used to show that FPP features and systems provide sufficient capability to meet the requirements of 10 CFR 50.48(c). NFPA 805, Section 2.4, "Engineering Analyses," states, in part, that: Engineering analysis is an acceptable means of evaluating a fire protection program against performance criteria. Engineering analyses shall be permitted to be qualitative or quantitative ... The effectiveness of the fire protection features shall be evaluated in relation to their ability to detect, control, suppress, and extinguish a fire and provide passive protection to achieve the performance criteria and not exceed the damage threshold defined in Section [2.5] for the plant area being analyzed. Chapter 1 of the standard defines the goals, objectives, and performance criteria that the FPP must meet in order to be in accordance with NFPA 805.

-52-NFPA 805, Section 1.3.1, "Nuclear Safety Goal," states that: The nuclear safety goal is to provide reasonable assurance that a fire during any operational mode and plant configuration will not prevent the plant from achieving and maintaining the fuel in a safe and stable condition. NFPA 805, Section 1.4.1, "Nuclear Safety Objectives," states that: In the event of a fire during any operational mode and plant configuration, the plant shall be as follows: (1) Reactivity Control. Capable of rapidly achieving and maintaining subcritical conditions. (2) Fuel Cooling. Capable of achieving and maintaining decay heat removal and inventory control functions. (3) Fission Product Boundary. Capable of preventing fuel clad damage so that the primary containment boundary is not challenged. NFPA 805, Section 1.5.1, "Nuclear Safety Performance Criteria," states that: Fire protection features shall be capable of providing reasonable assurance that, in the event of a fire, the plant is not placed in an unrecoverable condition. To demonstrate this, the following performance criteria shall be met: (a) Reactivity Control. Reactivity control shall be capable of inserting negative reactivity to achieve and maintain subcritical conditions. Negative reactivity inserting shall occur rapidly enough such that fuel design limits are not exceeded. (b) Inventory and Pressure Control. With fuel in the reactor vessel, head on and tensioned, inventory and pressure control shall be capable of controlling coolant level such that subcooling is maintained for a [pressurized-water reactor (PWR)] and shall be capable of maintaining or rapidly restoring reactor water level above top of active fuel for a water reactor (BWR)] such that fuel clad damage as a result of a fire is prevented. (c) Decay Heat Removal. Decay heat removal shall be capable of removing sufficient heat from the reactor core or spent fuel such that fuel is maintained in a safe and stable condition. (d) Vital Auxiliaries. Vital auxiliaries shall be capable of providing the necessary auxiliary support equipment and systems to assure that the systems required under (a), (b), (c), and (e) are capable of performing their required nuclear safety function.

-53-(e) Process Monitoring. Process monitoring shall be capable of providing the necessary indication to assure the criteria addressed in (a) through (d) have been achieved and are being maintained. 3.2.1 Compliance with NFPA 805 NSCA Methods NFPA 805, Section 2.4.2, "Nuclear Safety Capability Assessment," states the following: The purpose of this section is to define the methodology for performing a nuclear safety capability assessment. The following steps shall be performed: (1) Selection of systems and equipment and their interrelationships necessary to achieve the nuclear safety performance criteria in Chapter 1 (2) Selection of cables necessary to achieve the nuclear safety performance criteria in Chapter 1 {3) Identification of the location of nuclear safety equipment and cables (4) Assessment of the ability to achieve the nuclear safety performance criteria given a fire in each fire area This section of the SE evaluates the first three steps listed above. SE Section 3.5 addresses the assessment of the fourth step. RG 1.205, Revision 1, endorses NEI 04-02, Revision 2, and Chapter 3 of NEI 00-01, Revision 2, "Guidance for Post-Fire Safe Shutdown Circuit Analysis" (Reference 26). This NRC-endorsed guidance (i.e., NEI 04-02 Table B-2, "NFPA 805 Chapter 2-Nuclear Safety Methodology Review," and NEI 00-01, Chapter 3) has been determined to address the related requirements of NFPA 805, Section 2.4.2. The NRC staff reviewed LAR Section 4.2.1, "Nuclear Safety Capability Assessment Methodology," and LAR Attachment B, "NEI 04-02 Table B Nuclear Safety Capability Assessment-Methodology Review" (LAR Table B-2), against these guidelines. The endorsed guidance provided in NEI 00-01, Revision 2 provides a framework to evaluate the impact of fires on the ability to maintain post-fire safe shutdown. It provides detailed guidance for: (1) Selecting systems and components required to meet the nuclear safety performance criteria; (2) Selecting the cables necessary to achieve the nuclear safety performance criteria; (3) Identifying the location of nuclear safety equipment and cables; and

-54-(4) Appropriately conservative assumptions to be used in the performance of the NSCA. The licensee developed the LAR based on the guidance provided in the three guidance documents cited above. Based on the information provided in the licensee's submittal, as supplemented, a systematic process was used to evaluate the NMP1 post-fire SSA against the requirements of NFPA 805, Section 2.4.2, Subsections (1 ), (2), and (3), which the NRC staff concludes meets the methodology outlined in the latest NRC-endorsed industry guidance. FAQ 07-0039 (Reference 53) provides one acceptable method for documenting the comparison of the SSA against the NFPA 805 requirements. This method first maps the existing SSA to the NEI 00-01, Chapter 3 methodology, which in turn, is mapped to the NFPA 805 Section 2.4.2 requirements. The licensee performed this evaluation by comparing its SSA against the NFPA 805 NSCA requirements using the NRC-endorsed process in Chapter 3 of NEI 00-01, Revision 2, and documenting the results of the review in the B-2 Table in accordance with NEI 04-02, Revision 2 (Reference 5), as modified by FAQ 07-0039." The categories used by NMP1 to describe alignment with the NEI 00-01, Chapter 3, attributes are as follows: (1) The SSA directly aligns with the attribute: noted in LAR Table B-2, as "Aligns." (2) The SSA aligns with the intent of the attribute: noted in LAR Table B-2 as "Aligns with Intent." (3) The SSA does not align with the attribute, but there is a prior NRC approval of an alternative to the attribute, and the bases for the NRC approval remain valid: noted in LAR Table B-2 as "Not in Alignment, but Prior NRC Approval." (4) The SSA does not align with the attribute, but there are no adverse consequences because of the non-alignment: noted in Table B-2 as "Not in Alignment, but No Adverse Consequences." (5) The SSA does not align with the attribute: noted in the B-2 Table as "Not in Alignment." Finally, some attributes may not be applicable to the SSA (for example, the attribute may be applicable only to BWRs or PWRs). These are noted in the B-2 Table as "N/A." 3.2.1.1 Attribute Alignment -Aligns For the majority of the NEI 00-01, Chapter 3, attributes, the licensee determined that the SSA aligns directly with the attribute. In these instances, based on the information provided by the licensee in the LAR, the documents reviewed and discussions held with the licensee's technical

-55-staff during the on-site NFPA 805 audit, the NRC staff concludes that the licensee's statements of alignment are acceptable. The following attributes identified in LAR Attachment B, Table B-2 as aligning via this method required additional review by the NRC staff:

3.1.1.9-72 hour Coping; and

3.1.2.6.1 -Electrical System LAR Attachment B, Table B-2 states that the licensee's methodology aligns with sections 3.1.1.9 and 3.1.2.6.1 of NEI 00-01, Revision 2. In NEI 00-01, Revision 2, each section provides additional guidance, regarding the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months coping and the electrical system for safe shutdown. In a letter dated January 3, 2013 (Reference 15), the NRC requested additional information in SSA RAI 01 with regards to the safe and stable conditions and maintaining cold shutdown. In a letter dated April 30, 2013 (Reference 9), the licensee responded to SSA RAI 01 and stated NMP1 has elected to modify its NFPA 805 transition analysis for NMP1 to revise the approach for demonstrating the ability to reach and maintain safe and stable conditions, as specified by NFPA 805. The original nuclear safety capability assessment established as its basis for demonstrating safe and stable conditions the requirement to maintain Keff < 0.99 with a reactor coolant temperature at or below the requirements for hot shutdown and then subsequently cool down and maintain the plant in a cold shutdown condition. Consistent with NFPA 805 and supplemental guidance, the licensee is revising its basis for the NMP1 nuclear safety capability assessment to include only the requirement to establish hot shutdown conditions, including long-term hot shutdown capability. The response to this RAI, including parts a through g, was provided within the context of the aforementioned change. The NRC staff concludes that, based on the information provided in LAR Table B-2, "the LAR supplements," the information provided by the licensee in the LAR, and the documents reviewed and discussions held with the licensee's technical staff during the on-site NFPA 805 audit, that the licensee's statements of alignment are acceptable. 3.2.1.2 Attribute Alignment-Aligns with Intent For some of the NEI 00-01, Chapter 3, attributes, the licensee determined that the post-fire SSA aligns with the intent of the attribute, and provided additional clarification when describing its means of aligning with the attribute. The NEI 00-01, Chapter 3, attributes identified in LAR Table B-2 as having this condition are as follows:

3.1.2.1

3.2.1.2

3.3.1.1.2

3.3.3.2

3.4.1.7 Reactivity Control Fire Damage to Mechanical Components (not electrically supervised) Cable Failures Affecting Multiple Safe Shutdown Equipment Identify Interlocked Circuits and cables that could affect Safe Shutdown Appendix R Compliance Criteria Five of the attributes of NEI 00-01, Revision 2 (3.1.2.1, 3.2.1.2, 3.3.1.1.2, 3.3.3.2, and 3.4.1.7) for which the licensee states alignment with intent, address associated circuits by common

-56-power supply or common enclosure and require inclusion of additional cables and performance of additional circuit analysis if proper breaker and fuse coordination is not provided. As described in the LAR, the licensee has performed several circuit coordination studies for NMP1. In a letter dated January 3, 2013 (Reference 15), the NRC requested additional information in SSA RAI 07 based on a review of the updated final safety analysis report (UFSAR). Switchgear other than motor control centers (MCCs) use 125 VDC power for control of the operated circuit breakers so the breakers may be operated if AC power is lost. Dual feeds are provided to the DC control bus on each power board for added reliability, one each from battery 11, 12 or 14. A generic concern in regards to the Fort Calhoun fire that occurred on June 7, 2011 (NRC Special Inspection Report, March 12, 2012, ADAMS Accession No. ML 12072A 128) (Reference 91) involved 125 VDC circuits from both DC buses inside the same switchgear. At Fort Calhoun, both DC buses were impacted with "soft" grounds that remained after the fire had been isolated by removing power. Because information provided in the NMP1 UFSAR indicates that the power boards at NMP1 have dual control power feeds, the NRC staff requested that the licensee describe if the issue regarding "soft" grounds has been considered. In addition, the NRC staff requested the licensee to describe if there are any proposed plans to perform modifications or procedure changes to address this issue. In a letter dated April 30, 2013 (Reference 9}, the licensee responded to SSA RAI 07 and provided the following response: The 125 VDC system at NMP1 consists of two physically separate and independent trains safety related (Batteries 11 and 12) and one augmented quality train. Each train includes one 125 VDC station battery, two parallel static battery chargers (one primary and the other a backup), and one DC power distribution board. The battery boards include the circuit breakers, fuses and the fuse blocks required for distribution of 125 VDC to various system loads. The augmented quality 125 VDC system consists of one 125 VDC station battery (Battery 14), a static battery charger, and one DC power distribution battery board. The 125 VDC batteries 11, 12, and 14 are part of NMP1 Safe Shutdown Equipment. Battery 11 and the associated battery board are located in Fire Areas 17B and 16B, respectively. Battery 12 and the associated battery board are located in Fire Areas 17A and 16A, respectively. Battery 14 and the associated battery board are located in Fire Area 5. The 125 VDC electrical distribution trains are operated independently and ungrounded, and, as such, a single ground does not generate a fault current or disable the system. Each of the battery systems is equipped with ground detection devices to indicate the occurrence of the first ground which allows operators to locate and correct the first ground. The licensee addressed the two contributing causes to the overall event at Fort Calhoun as follows:

The licensee stated that they do not use Molded Case Square-D Master pact circuit breaker/cradle assemblies and digital trip devices that were used at Fort Calhoun.

The licensee stated that the DC feeds to power boards, EDGs, and the diesel fire pump (DFP) are equipped with fuses, unlike Fort Calhoun which did not utilize fuses in their DC feeds.

-57-The NRC staff determined that, based on the information provided by the licensee in the LAR (Table 8-2) as supplemented, the documents reviewed, and discussions held with the licensee's technical staff during the on-site NFPA 805 audit, and in the information in the Fort Calhoun inspection report, the NMP1 DC distribution design is not subject to the same breaker failure that occurred at Fort Calhoun. The NRC staff concludes that the licensee's statements of alignment with the endorsed guidance in NEI 00-01 are acceptable because the licensee has demonstrated that FPP features and systems provide sufficient capability to meet the requirements of 10 CFR 50.48(c). In a letter dated January 3, 2013 (Reference 15) the NRC staff requested additional information in SSA RAI 08 for the breaker coordination study that identified a lack of coordination for several 600 VAG, 208-120 VAG, and 240 VAG power supplies. The tie breakers (R 1 042/R 1 052) that connect powerboard 16A section with 168 section and 17 A section with 178 section are not coordinated with the breakers that supply powerboards 167 and 1671. The supply breakers for powerboards 167 and 1671 are coordinated with the respective supply breaker to powerboards 16 and 17 but are not coordinated with the tie breakers between the A and 8 sections of powerboards 16 and 17. The LAR stated that since the tie breakers are normally open and are controlled administratively by operating procedure N1-0P-30; 4.16 KV, 600 V and 480 V house service, there is no reason to replace the breakers to improve tie breaker coordination. However, this condition only remains valid if the tie breakers remain open. The NRC staff requested that the licensee provide electrical lineups and descriptions of the procedures that describe when the tie breakers would be closed and the tie breaker coordination issue would exist. For instance, following a loss of offsite power (LOOP), the tie breaker needs to be closed in order to supply power to one or more instrument air compressors and one of the spent fuel pool cooling pumps. The NRC staff requested that the licensee:

Describe the effects of closing one or both of the tie-breakers following a LOOP since doing so results in loss of breaker coordination;

Indicate how often the tie breakers are closed;

Describe any compensatory actions taken when the tie breakers are closed;

Describe how the Institute of Electrical and Electronics Engineers Standard 242, "Recommended Practice for Protection and Coordination of Industrial and Commercial Power Systems" is satisfied when the tie breakers are closed;

Provide a summary of the breaker coordination study for when the tie-breakers are closed; and

Describe how the results of the breaker coordination study tie into the probabilistic risk assessment (PRA). In a letter dated February 27, 2013 (Reference 7) the licensee responded to SSA RAI 08 and provided a summary of the breaker coordination study for when the tie breakers are closed, and

-58-a description of how results of the breaker coordination tie into the fire PRA. A portion of the licensee response stated for the second iteration sensitivity case (Case 8b in Breaker Coordination Study), the tie breakers were modeled as coordinated and the remaining uncoordinated breakers were left uncoordinated (this is consistent with normal plant line up when tie breakers are open, or when the power is fed from the "B" side of the Powerboard to the "A" side). The results of the sensitivity analysis show that the remaining uncoordinated breakers are low risk. In a letter dated October 9, 2013 (Reference 16), the NRC staff requested additional information in SSA RAI 08.01. The NRC staff identified that the failure to provide electrical coordination between one or more load circuit breakers and the cross tie circuit breaker does not meet the requirements of NFPA 805 Section 2.4.2.2.2 for common power supply, requiring this condition to be considered a VFDR. The NRC staff requested that the licensee address this new VFDR using the Fire Risk Evaluation performance-based approach in accordance with NFPA 805 Section 4.2.4.2 for each fire area where fire-induced damage could cause the loss of the common power supply. In a letter dated January 22, 2014 (Reference 11 ), the licensee responded to SSA RAI 08.01 and stated that there is a small increase in risk to leave the breakers uncoordinated, and that NMPNS has added the condition as a VFDR to capture the risk. The increased risk has been characterized as a potential challenge to the Vital Auxiliaries nuclear safety performance criteria. Four new VFDRs were identified; VFDR-05-048 for fire area 05; VFDR-1 0-030 for fire area 1 0; VFDR-12-01 0 for fire area 12; and VFDR-13-012 for fire area 13. Revised pages to Table B-3 showing the new VFDRs and VFDR dispositions in associated areas were included in the response. This issue was also addressed as part of the aggregate change-in-risk evaluation requested in the second-round PRA RAI 61. The composite PRA results or aggregate change in risk was provided in a revised LAR Attachment W. Based on the revised analysis addressing the lack of coordination as a VFDR, the NRC staff concludes that the licensee included the impact in the plant risk and delta risk calculations and showed that the acceptance criteria for risk, DID and safety margins have all been met. The NRC staff concludes that the methods as described by the licensee are similar to the specific methods in NEI 00-01, and therefore align with intent of NEI 00-01. 3.2.1.3 Attribute Alignment -Not in Alignment, but Prior NRC Approval The licensee did not identify any attributes in this category in LAR Table B-2 3.2.1.4 Attribute Alignment-Not in Alignment, but No Adverse Consequences The licensee did not identify any attributes in this category in LAR Table B-2 3.2.1.5 Attribute Alignment -Not in Alignment The licensee did not identify any attributes in this category in LAR Table B-2

-59-3.2.1.6 NFPA 805 Nuclear Safety Capability Assessment Methods Conclusion The NRC staff reviewed the documentation provided by the licensee describing the process used to perform the NSCA required by NFPA 805, Section 2.4.2. The licensee performed this evaluation by comparing the SSA against the NFPA 805 NSCA requirements using NEI 00-01, Revision 2 to the NRC-endorsed process in Chapter 3 of NEI 00-01, Revision 2. The results of the review are documented in the B-2 Table in accordance with NEI 04-02, Revision 2. Based on the information provided in the licensee's submittal, as supplemented, the NRC staff accepts the method the licensee used to perform the NSCA with respect to the selection of systems and equipment, selection of cables, and identification of the location of nuclear safety equipment and cables, as required by NFPA 805, Section 2.4.2. The NRC staff concludes that the licensee's method is acceptable because it either:

Met the NRC-endorsed guidance directly; or

Met the intent of the endorsed guidance and provided adequate justification. 3.2.2 Maintaining Fuel in a Safe and Stable Condition The nuclear safety goals, objectives and performance criteria of NFPA 805 allow more flexibility than the previous deterministic FPPs based on Appendix R to 10 CFR 50 and NUREG-0800, Section 9.5.1 (Reference 86), since NFPA 805 only requires the licensee to maintain the fuel in a safe and stable condition rather than achieve and maintain cold shutdown in 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . In a letter dated January 3, 2013 (Reference 15), in SSA RAI 01, the NRC staff requested additional information on maintaining cold shutdown. In a letter dated April 30, 2013 (Reference 9), the licensee responded to SSA RAI 01 and stated that the NFPA 805 licensing basis is to achieve and maintain the fuel in hot shutdown (Mode 3) conditions following any fire occurring with the reactor operating at power or hot shutdown (Reactor mode switch in the "Shutdown" position and reactor coolant temperature greater than 212.F) or Non-power analysis for potential fires while in Shutdown Conditions -Hot operating condition and lower operating conditions. Sustaining hot shutdown conditions (once achieved) for an extended period of time is accomplished by:

Ensuring a continual source of water to the Emergency Condensers (EC) in support of decay heat removal using the EC system,

Ensuring a long-term source of inventory makeup to the reactor, and

Ensuring continual operation of at least one emergency diesel generator (EDG) to supply AC power to the electrical distribution system. The licensee stated that the primary means of achieving and maintaining hot shutdown following a fire coincident with a LOOP is via the EC system. The EC system consists of two redundant emergency cooling loops. Each loop is capable of independently accomplishing hot shutdown. Therefore, this option provides two redundant paths for obtaining hot shutdown. Upon achieving hot shutdown conditions, the plant is able to maintain safe and stable operation for an extended period of time using the EC system. After 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , the EC makeup tanks can be

-60-replenished as needed using the diesel fire pump (DFP), which draws water from Lake Ontario (effectively an infinite source of water). In the event water from the condensate storage tanks (CST) can be transferred to the EC makeup tanks, operation of the DFP would not be required until some point beyond 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . Periodic refueling of the DFP is accomplished in accordance with existing plant procedures using the DFP fuel oil storage tank. The DFP day tank contains sufficient fuel for 17 hours1.967593e-4 days 0.00472 hours 2.810847e-5 weeks 6.4685e-6 months of operations. The DFP fuel oil storage tank contains fuel to support 6.1 days of operation. Reactor coolant makeup is required after 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , assuming a normal TS leakage rate of 25 gallons per minute. Makeup is provided via the Control Rod Drive (CRD) system using one of the CRD pumps drawing suction from the CST. Alternating current (AC) power is required to operate a CRD pump. The DFP may be aligned to provide reactor coolant makeup in the event that no CRD pump is available. The licensee further stated in the event the EC system is not available, the plant can be maintained in hot shutdown by opening three Electromatic Relief Valves (ERVs) in the automatic depressurization system (ADS) and blowing steam to the Torus to reduce pressure. When reactor pressure reaches approximately 365 psig, the Core Spray (CS) system may be utilized to provide core cooling. Both AC and direct current (DC) electrical power are required for this method of decay heat removal. This alternate means of decay heat removal can be used to maintain safe and stable conditions until such time that the shutdown cooling system (SOC) is placed in service. The CS system is a two loop system. Operation of one loop is adequate to ensure core cooling. When utilizing the CS system, the ERVs pass steam and then, eventually, water to the Torus to remove decay heat from the reactor, in essence placing the Reactor Coolant System (RCS) in recirculation through the Torus. During this process, decay heat is removed by operation of the Containment Spray (CTS) system in conjunction with the Containment Spray Raw Water (CTSRW) system. This method of decay heat removal negates the need for another system to provide inventory makeup. AC power is required to initiate and maintain this method of decay heat removal; thus, long-term maintenance of the operating mode is dependent on maintaining AC electrical power. For either of the hot shutdown methods used to achieve and maintain long-term safe and stable conditions, AC power availability from either the station EDGs or offsite power is necessary. Offsite power is not credited in the NSCA. The EDGs can be refueled in accordance with existing plant procedures using an on-site fuel source (tanker truck), until such time that offsite power is restored. Refueling of a continually operating EDG is estimated to be required after four days (assuming one EDG operating at full load). Given the long time frame before EDG and/or DFP refueling is necessary, additional resources from the Emergency Response Organization will be available to support EDG and DFP refueling activities. As described in the LAR, the licensee has modeled NMP1 capability to achieve and maintain safe and stable conditions for the initial 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of the event. Beyond 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the licensee has described the means to maintain safe and stable conditions and determined that these post-24 hour actions have no significant contribution to risk. On the basis of the licensee's analysis as described in the LAR, as supplemented, the NRC staff concludes that the fuel can be maintained in a safe and stable condition, post-fire.

-61 -3.2.3 Applicability of Feed and Bleed The limitations of 10 CFR 50.48(c)(2)(iii}, "Use of feed-and-bleed," is not applicable to BWRs. 3.2.4 Assessment of Multiple Spurious Operations NFPA 805, Section 2.4.2.2.1, "Circuits Required in Nuclear Safety Functions," states, in part, that: Circuits required for the nuclear safety functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal}, open circuits, and shorts to ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals. In addition, NFPA 805, Section 2.4.3.2, states that the probabilistic safety assessment (PSA) evaluation shall address the risk contribution associated with all potentially risk-significant fire scenarios. Because the PB approach taken at NMP1 utilized FREs in accordance with NFPA 805 Section 4.2.4.2, "Use of Fire Risk Evaluation," adequately identifying and including potential multiple spurious operation (MSO) combinations is required to ensure that all potentially risk-significant fire scenarios have been evaluated. The NRC staff reviewed LAR Section 4.2.1.4, "Evaluation of Multiple Spurious Operations," and LAR Attachment F, "Fire-Induced Multiple Spurious Operations Resolution," to determine whether the licensee has adequately addressed MSO concerns at NMP1. As described in the LAR, the licensee's process for identification and evaluation of MSOs used an expert panel and followed the guidance of NEI 04-02 (Reference 5}, RG 1.205 (Reference 2) and FAQ 07-0038, "Lessons Learned on Multiple Spurious Operations," (Reference 52). The expert panel used by the licensee consisted of subject matter experts with education and experience in electrical engineering, fire PRA, PRA, HRA, SSA, fire protection engineering, system engineering and plant operations. LAR Attachment F stated that the licensee conducted an initial expert panel review in 2008 and a second review in 2010. Prior to initial review, the panel was provided with training and was provided with a specific project instruction for conducting the review. The expert panel sources for information and identifying MSOs included the SSA, generic lists (e.g. from Owners Groups), self-assessment results, and internal events PRA insights. The NSCA and fire PRA were updated to reflect the treatment of applicable MSO scenarios. This included the identification of equipment, cables, and cable routing by plant locations. The MSO combination components of concern were also evaluated as part of the NSCA. For cases where the pre-transition MSO combination components did not meet the deterministic compliance, the MSO combination components were added to the scope of the FREs. The expert panel was reconvened in 2012 to review proposed changes to the expert panel report. The same process described above was used to perform the review.

-62-LAR Attachment F describes the process the licensee utilized to address MSOs, which follows the guidance of FAQ 07-0038. That process includes five steps: (1) identify potential MSOs of concern; (2) conduct an expert panel to assess plant specific vulnerabilities; (3) update the FPRA model and NSCA to include the MSOs of concern; (4) evaluate for NFPA 805 compliance; and, (5) document results. As described in LAR Attachment F, under the results for Steps 3, 4, and 5, the MSOs identified in Steps 1 and 2 were incorporated in the fire PRA model. The MSO combination components of concern were then evaluated for inclusion into the NMP1 NSCA. As necessary, components were added to the NSCA Equipment List and Logics; and circuit analysis and cable routing was performed. For cases where the MSO combination components did not meet the requirements for deterministic compliance, the MSO combination components were added to the scope of the RI/PB risk evaluations. The NMP1 fire PRA quantified the fire-induced risk model containing the MSO pathways. The MSO contribution is included in the fire PRA results. The NRC staff reviewed the licensee's expert panel process for identifying circuits susceptible to MSOs as described above and concluded that the licensee adopted a systematic and comprehensive process for identifying MSOs to be analyzed utilizing available industry guidance. Furthermore, the process used assures that the FREs appropriately identify and include risk significant MSO combinations. Based on the information provided by the licensee in the LAR, the documents reviewed during the site audit, and the discussions with the licensee's staff during the site audit, the NRC staff concludes that the licensee's approach for assessing the potential for MSO combinations is acceptable for use at NMP1. 3.2.5 Establishing Recovery Actions NFPA 805, Section 1.6.52, "Recovery Action," defines a recovery action as follows: Activities to achieve the nuclear safety performance criteria that take place outside the MCR [main control room] or outside the primary control station(s) (PCS) for the equipment being operated, including the replacement or modification of components. NFPA 805, Section 4.2.3.1 states that: One success path of required cables and equipment to achieve and maintain the nuclear safety performance criteria without the use of recovery actions shall be protected by the requirements specified in either 4.2.3.2, 4.2.3.3, or 4.2.3.4, as applicable. Use of recovery actions to demonstrate availability of a success path for the nuclear safety performance criteria automatically shall imply use of the performance-based approach as outlined in 4.2.4. NFPA 805 Section 4.2.4, "Performance-Based Approach," states the following: When the use of recovery actions has resulted in the use of this approach, the additional risk presented by their use shall be evaluated.

-63-The NRC staff reviewed LAR Section 4.2.1.3, "Establishing Recovery Actions," and LAR Attachment G, "Operator Manual Actions Transition," to evaluate whether the licensee meets the associated requirements for the use of recovery actions per NFPA 805. The licensee used the endorsed guidance provided in NEI 04-02 (Reference 5), Section 4.6, RG 1.205 (Reference 2) and the guidance in FAQ 07-0030,"Establishing Recovery Actions," (Reference 51) to establish the population of recovery actions (RAs) being carried forward in the RIIPB FPP. RAs addressed during the NFPA 805 transition process included the consideration of existing operator manual actions (OMAs) in the deterministic FPP, as well as those being added based on the VFDRs identified in the individual fire area assessments. OMAs are actions performed by plant operators to manipulate components and equipment from outside the MCR to achieve and maintain post-fire hot shutdown, not including "repairs." OMAs include an integrated set of actions needed to ensure that hot shutdown can be accomplished for a fire in a specific plant area. OMAs are transitioned to RAs under NFPA 805. RAs are activities to achieve the NSPC that take place outside of the MCR or outside of the PCSs for the equipment being operated, including the replacement or modification of components. The NMP1 has 2 locations designated as a PCSs as defined in RG 1.205. Remote Shutdown Panel (RSP) 11 and RSP 12. The RSPs were approved by the NRC in SER entitled "Subject Modifications and Alternate Safe Shutdown Capabilities to comply with the Requirements of Appendix R", dated March 3, 1983. The NRC staff concluded that the alternate shutdown strategy at NMP1 was in accordance with Appendix R to 10 CFR 50, Items III.G.3 and III.L. NMP1 utilized the guidance in RG 1.205, Revision 1 (Reference 2) for addressing recovery actions. This included consideration of the definition of PCS and RA, as clarified in the RG 1.205, Revision 1. Accordingly, any actions required to transfer control to, or operate equipment from the PCS, while required as part of the RI/PB FPP, were not considered RAs per the RG 1.205 guidance and in accordance with NFPA 805. Alternatively, any OMAs required to be performed outside the MCR and not at the PCS were considered RAs. OMAs meeting the definition of a RA are required to comply with the NFPA 805 requirements, which include the need to determine the additional risk of their use. The additional risk of recovery actions is discussed in SE Section 3.4.4. Some of these OMAs may not be required to demonstrate the "availability of a success path," in accordance with NFPA 805, Section 4.2.3.1, but may still be required to be retained in the RI/PB FPP because of DID considerations as described in Section 1.2 of NFPA 805. The licensee did not differentiate between a recovery action that is needed to meet the nuclear safety capability assessment and one retained to provide defense-in-depth. In each instance, the licensee determined whether a transitioning OMA was a RA, or not necessary for the post-transition RI/PB FPP. The licensee stated that all credited RAs, as listed in LAR Attachment G, were subjected to a feasibility review. In accordance with the NRC-endorsed guidance in NEI 04-02, the feasibility criteria used in the licensee's assessment process, were based on the guidance described in FAQ 07-0030 (Reference 51) and each of the 11 individual feasibility attributes were addressed. LAR Attachment G, Table G-1, "Recovery Actions and Activities Occurring at the PCSs," describes each RA associated with disposition of a VFDR from the fire area assessments as documented in LAR Attachment C, "Fire Area Transition." The feasibility review was based on

-64-documentation only, including previous feasibility evaluations for safe shutdown OMAs. The licensee included Implementation Item 19 in LAR AttachmentS, Table S-2 to revise safe shutdown procedures and training as necessary to incorporate updated NSCA strategies. Based on the above, the NRC staff concludes that the licensee has followed the endorsed guidance of NEI 04-02 and RG 1.205 to identify and evaluate RAs in accordance with NFPA 805, and therefore, the requirements of 10 CFR 50.48(c) are met. The NRC staff concludes that the feasibility criteria applied to RAs are acceptable based on conformance with the endorsed guidance contained in NEI 04-02 and successful completion of identified Implementation Item 19 in LAR AttachmentS, Table S-2. 3.2.6 Conclusion for Section 3.2 The NRC staff reviewed the licensee's LAR, as supplemented, for conformity with the requirements contained in NFPA 805, Section 2.4.2, regarding the process used to perform the NSCA at NMP1. The NRC staff concludes that the declared safe and stable condition proposed is acceptable and that the licensee's process is adequate to appropriately identify and locate the systems, equipment, and cables required to provide reasonable assurance of achieving and maintaining the fuel in a safe and stable condition, as well as to meet the NFPA 805 NSPC. The NRC staff reviewed the licensee's process to identify and analyze MSOs. Based on the information provided in the LAR, as supplemented, the process used to identify and analyze MSOs at NMP1 is considered comprehensive and thorough. Through the use of an expert panel process in accordance with the guidance of RG 1.205 (Reference 2), NEI 04-02 (Reference 5) and FAQ 07-0038 (Reference 52), potential MSO combinations were identified and included as necessary into the NSCA as well as the applicable FREs. The NRC staff also concludes that the licensee's approach for assessing the potential for MSO combinations acceptable, because it was performed in accordance with NRC-endorsed guidance. The NRC staff concludes that, based on the information provided in the NMP1 LAR, as supplemented, and the information obtained during the NFPA 805 site audit (documents reviewed and discussions with the licensee's staff) the process used by the licensee to review, categorize and address RAs during the transition from the existing deterministic fire protection licensing basis to a Rl/PB FPP is consistent with the NRC-endorsed guidance contained in NEI 04-02 and RG 1.205, regarding the identification of RAs and other actions required to be taken at a PCS. The licensee has identified the actions to be taken at a PCS as well as identified those actions that meet the definition of a RA provided in NFPA 805 section 1.6.52. The licensee must complete Implementation Item 19 in LAR Attachment S, Table S-2 by the end of the implementation window. Based on the information provided by the licensee, and upon completion of the implementation item, the NRC staff concludes that the process to review, categorize and address RAs during the transition is acceptable.

-65-3.3 Fire Modeling NFPA 805 allows both FM and FREs as PB alternatives to the deterministic approach outlined in the standard. These two PB approaches are described in NFPA 805, Sections 4.2.4.1 and 4.2.4.2, respectively. Although FM and FREs are presented as two different approaches for PB compliance, the FRE approach generally involves some degree of FM to support engineering analyses and fire scenario development. NFPA 805, Section 1.6.18, defines a fire model as a "mathematical prediction of fire growth, environmental conditions, and potential effects on structures, systems, or components based on the conservation equations or empirical data." The NRC staff reviewed LAR Section 4.5.2, "Performance-Based Approaches," which describes how the licensee used FM as part of the transition to NFPA 805 at NMP1. In LAR Section 4.5.2.1, the licensee indicated that, in lieu of the FM approach (NFPA 805 Section 4.2.4.1 ), the FRE approach (NFPA, Section 4.2.4.2) was used for the transition to NFPA 805. In LAR Section 4.5.1.2, "Fire Model Utilization in the Application," the licensee indicated that FM was performed as part of the FPRA development. Therefore, the NRC staff reviewed the technical adequacy of the NMP1 FRE, including the supporting FM analyses, as documented in SE Section 3.4.2, to evaluate compliance with the NSPC. The licensee did not propose any FM methods to support PB evaluations in accordance with NFPA 805, Section 4.2.4.1, as the sole means for demonstrating compliance with the NSPC. Therefore, the scope of the licensee's self-approval capability does not include utilizing the FM PB approach in accordance with NFPA 805, Section 4.2.4.1. 3.4 Fire Risk Evaluations This section addresses the licensee's fire risk evaluation performance-based method, which is based on NFPA 805 (Reference 1) Section 4.2.4.2. The licensee chose to use only the fire risk evaluation performance-based method in accordance with NFPA 805, Section 4.2.4.2. The fire modeling performance-based method of NFPA 805 Section 4.2.4.1 was not used for this application. NFPA 805, Section 4.2.4.2, "Use of Fire Risk Evaluations," states the following: Use of fire risk evaluation for the performance-based approach shall consist of an integrated assessment of the acceptability of risk, defense in depth [DID], and safety margins. The evaluation process shall compare the risk associated with implementation of the deterministic requirements with the proposed alternative. The difference in risk between the two approaches shall meet the risk acceptance criteria described in NFPA 805, Section 2.4.4.1 ["Risk Acceptance Criteria"]. The fire risk shall be calculated using the approach described in NFPA 805, 2.4.3 ["Fire Risk Evaluations"].

-66-3.4.1 Maintaining Defense-in-Depth and Safety Margins NFPA 805, Section 4.2.4.2, requires that the "use of fire risk evaluation for the based approach shall consist of an integrated assessment of the acceptability of risk, in-depth, and safety margins." 3.4.1.1 Defense-in-Depth (DID) NFPA 805, Section 1.2, states the following: Protecting the safety of the public, the environment, and plant personnel from a plant fire and its potential effect on safe reactor operations is paramount to this standard. The fire protection standard shall be based on the concept of in-depth. Defense-in-depth shall be achieved when an adequate balance of each of the following elements is provided:

Preventing fires from starting;

Detecting fires quickly and extinguishing those that do occur, thereby limiting fire damage; and,

Providing an adequate level of fire protection for structures, systems, and components important to safety, so that a fire that is not promptly extinguished will not prevent essential plant safety functions from being performed. The NRC staff reviewed LAR (Reference 6) Section 4.5.2.2, "Fire Risk Approach", Section 4.8.1, "Results of the Fire Area Review" and LAR Attachment C, "NEI 04-02 Table B Fire Area Transition" as supplemented by letter dated April 30, 2013 (Reference 9), to determine whether the principles of DID were maintained in regard to the planned transition to NFPA 805. When implementing the PB approach, the licensee followed the guidance contained in NEI 04-02 (Reference 5), Section 5.3, "Plant Change Process," which includes a detailed consideration of DID and safety margins as part of the change process. The license documented the method used to meet the DID requirements of NFPA 805 in LAR Attachment C Table 8-3 as supplemented by letter dated April 30, 2013 (Reference 9). LAR Attachment C Tables B-3 and 4-3 document the results of the licensee's review of fire suppression and fire detection systems. The licensee's methodology for evaluating DID refers to each of the three DID elements identified in NFPA 805, Section 1.2 as Echelons 1, 2, and 3, respectively. As described in the response to SSA RAI 09 (Reference 7), this method was implemented in the fire risk evaluations (FREs) performed on each PB fire area. In the DID evaluation, the licensee identified the fire protection systems and features credited for each echelon and qualitatively assessed if the credited systems and features provided an adequate balance between the echelons. Additional fire protection systems and features were

-67-credited if necessary to improve the balance between the echelons. Specifically with regard to Echelon 3, the licensee explained that the evaluation of this echelon considered high-consequence fire scenarios, defined as scenarios having a CDF of 1 E-06/reactor-year or greater and a conditional core damage probability (CCDP) of 0.1 or greater, in determining if additional fire protection systems and features were needed to provide an adequate balance. The results of the licensee's DID assessment for each echelon by fire area is provided in LAR Attachment C Table B-3 as supplemented by letter dated April 30, 2013 (Reference 9). Based on its review of the LAR, the response to SSA RAI 09, and the review of the FREs during the audit, the NRC staff concludes that the licensee has systematically and comprehensively evaluated fire hazards, area configuration, detection and suppression features, and administrative controls in each fire area. The NRC staff also concludes that the methodology as proposed in the LAR evaluates DID against fires as required by NFPA 805 and therefore the proposed RI/PB FPP adequately maintains DID. 3.4.1.2 Safety Margins NFPA 805, Section 2.4.4.3 states the following: The plant change evaluation shall ensure that sufficient safety margins are maintained. NEI 04-02, Section 5.3.5.3, "Safety Margins," lists two specific criteria that should be addressed when considering the impact of plant changes on safety margins:

Codes and Standards or their alternatives accepted for use by the NRC are met; and,

Safety analyses acceptance criteria in the licensing basis (e.g., FSAR, supporting analyses, etc.) are met, or the change provides sufficient margin to account for analysis and data uncertainty. LAR Section 4.5.2.2, "Fire Risk Approach," discusses how safety margins are addressed as part of the FRE process and that this process is based on the requirements of NFPA 805, industry guidance in NEI 04-02, and RG 1.205 (Reference 2). An FRE was performed for each fire area containing variance from deterministic requirements (VFDRs). The FREs contain the details of the licensee's review of safety margins for each performance-based fire area. LAR Section 4.5.1.2 states that the FPRA, including fire modeling performed to support the FPRA, applies methodologies consistent with the guidance in NUREG/CR-6850 (References 34, 35, and 36) and, according toLAR Attachment H, NRC-approved FAQs. LAR Attachment J explains that fire modeling, including verification and validation (V&V), performed in support of the FPRA utilized accepted codes and standards including NUREG/CR-6850, NUREG-1805 (Reference 40), NUREG-1824 (Reference 41 ), etc. In a letter dated February 27, 2013 (Reference 7) the licensee responded to SSA RAI 09 and

-68-further described the methodology used to evaluate safety margins in the FREs to include the following evaluations and determinations:

Plant System Performance: Plant system performance was evaluated using design basis evaluations documented in the Updated Final Safety Analysis Report (UFSAR); and

PRA Logic Model: The PRA logic model was developed in accordance with the ASME/ANS RA-Sa-2009 PRA standard (Reference 29) and RG 1.200, Rev. 2 (Reference 28). The results of the licensee's safety margin assessment by fire area are provided in LAR Attachment C, Table B-3 as supplemented by letter dated April 30, 2013 (Reference 9). The safety margin criteria described in NEI 04-02, Section 5.3.5.3 and the LAR, as supplemented, are consistent with the criteria as described in RG 1.17 4 (Reference 27) and therefore acceptable. The licensee used appropriate codes and standards (or NRC guidance), and met the safety analyses acceptance criteria in the licensing basis. Based on its review of the LAR and the review of the FREs during the audit, the NRC staff concludes that the licensee's approach has adequately addressed the issue of safety margins in the implementation of the FRE process. 3.4.2 Quality of the Fire Probabilistic Risk Assessment The objective of the PRA quality review is to determine whether the plant-specific PRA used in evaluating the proposed LAR is of sufficient scope, level of detail, and technical adequacy for the application. The staff evaluated the PRA quality information provided by the licensee in its LAR, as supplemented, including industry peer review results and self-assessments performed by the licensee. The NRC staff reviewed LAR Section 4.5.1, "Fire PRA Development and Assessment," LAR Section 4.7, "Program Documentation, Configuration Control, and Quality Assurance," LAR Attachment C, "NEI 04-02 Table B Fire Area Transition," LAR Attachment U, "Internal Events PRA Quality," LAR Attachment V, "Fire PRA Quality," and LAR Attachment W, "Fire PRA Insights." The licensee developed its internal events PRA (lEPRA) during the individual plant examination process and continued to maintain and improve the PRA as RG 1.200, and supporting industry standards have evolved. The licensee developed its FPRA model for both Level 1 (core damage) and partial Level 2 (large early release) PRA during at-power conditions. For the development of the FPRA, the licensee modified its lEPRA model to capture the effects of fire. In LAR Section 4.8.2 the licensee stated the no significant plant changes (beyond those identified and scheduled to be implemented as part of the transition to a FPP based on NFPA 805) are outstanding with respect to their inclusion in the FPRA model. Based on this information, the NRC staff concludes that upon completion of the modifications listed in Attachment S, Table S-2, the FPRA model meets the above criteria, that it will represent the current, as built, as operated configuration, and is therefore capable of being adapted to model both the post transition and compliant plant as needed.

-69-The licensee identified administrative controls and processes used to maintain the FPRA model current with plant changes and to evaluate any outstanding changes not yet incorporated into the PRA model for potential risk impact as a part of the routine change evaluation process. Further, as described in SE Section 3.8.3, the licensee has a program for ensuring that developers and users of these models are appropriately trained and qualified. Therefore, the NRC staff concludes that the PRA has the capability to support post-transition plant change evaluations (PCEs), such as the self-approval process, after any changes required during implementation are completed. 3.4.2.1 Internal Events PRA Model The licensee's evaluation of the technical adequacy of the portions of its lEPRA model used to support development of the FPRA model consisted of a full scope peer review that was performed in February 2008 using the NEI 05-04 process (Reference 74), and the combined PRA standard, ASME/ANS-RA-Sb-2005 (Reference 75), as clarified by RG 1.200, Revision 1 (Reference 76). This review was performed to standards and guides that were in place at the time that the review was performed. As stated in LAR Section 4.5.1.2, the licensee has performed a gap analysis against RG 1.200, Revision 2 for the lEPRA. In a letter dated February 27, 2013 (Reference 7) the licensee responded to PRA RAI 09 and stated that the October 2011 peer review of the FPRA included a review of the internal events SRs referenced in the FPRA SRs, and considered the ASME/ANS-RA-Sa-2009 (Reference 29) SRs and related clarifications provided in RG 1.200, Revision 2. The NRC finds the response to PRA RAI 09 and approach acceptable because the ASME/ANS RA-Sa-2009 Standard's SRs and associated RG 1.200 clarifications have been applied to the internal events SRs which are a part of the FPRA peer review, and the minor differences between the current documents and the ASME/ANS-RA-Sa-2005 standard and RG 1.200, Revision 1 are not expected to change substantially or significantly impact the FPRA results. For many SRs, there are three degrees of "satisfaction" referred to as Capability Categories (CC) (i.e., I, II, and Ill), with I being the minimum, II considered widely acceptable, and Ill indicating the maximum achievable scope/level of detail, plant specificity, and realism. For other SRs, the CCs may be combined (e.g., the requirement for meeting CCI may be combined with II), or the requirement may be the same across all CCs so that the requirement is simply met or not met. For each supporting requirement (SR), the PRA reviewer from the peer review team designates one of the CCs or indicates that the SR is met or not met. LAR Attachment U, Table U-1 provides the licensee's dispositions to all 26 facts and observations (F&Os) characterized as findings per peer review guidelines (Reference 76). In general, an F&O is written for any SR that is judged not to be met or does not fully satisfy CC II of the ASME standard, consistent with RG 1.200, Revision 1. As described in LAR Attachment U, the licensee dispositioned each F&O by assessing the impact of the F&O on the FPRA. The NRC staff requested additional information to assess the adequacy of some of the F&O dispositions for the review. The NRC staff evaluated each F&O and the licensee's disposition in LAR Attachment U to determine whether the F&O had any significant impact on the FPRA. The NRC staff's conclusions on NMP1's resolution of each

-70-F&O is summarized in the NRC's Record of Review dated April 8, 2014 and March 12, 2014 (Reference 77). F&Os requiring additional information for the staff to complete its review are discussed below. Additional information associated with F&O DA-E1-01 with respect to changes made to the HRA was requested in a letter dated January 3, 2013 (Reference 15) in PRA RAI 14. In a letter dated February 27, 2013 (Reference 7) the licensee responded to PRA RAI 14 and stated that changes have been made to some human error probabilities (HEPs) in the PRA based on some plant modifications and development of procedures to utilize these modifications. The licensee indicated in Modification Item 5 of LAR AttachmentS, Table S-1 to complete one plant modification that has not yet been completed. The NRC staff finds the licensee's response and approach acceptable because changes to the plant and procedures will be modeled, and crediting future modifications is an accepted approach during the transition to NFPA-805. In a letter dated January 3, 2013 (Reference 15) in PRA RAI 16, the NRC staff requested that the licensee provide descriptions of the PRA documentation updates that are needed to address F&Os QU-01 a-01, QU-D5b-01, and QU-E4-01. In a letter dated February 27, 2013 (Reference 7), the licensee clarified the updates for each F&O and stated that they would occur as part of Implementation Item 3 in LAR Attachment S, Table S-2. In regard to F&O (QU-D5b-01 ), the licensee stated the importance of components and basic events were reviewed to determine that they are logical, and clarified that the documentation would be updated to reflect this review. In regard to F&O QU-D1a-01, the licensee stated that extensive reviews were conducted of cutsets for event tree accident sequences, regardless of their significance, and clarified that the PRA documentation would be updated to address those cutsets constituting 90 percent of the total CDF. In regard to F&O QU-E4-01, the licensee stated that PRA documentation would be updated to include an evaluation of key model uncertainties and associated assumptions in accordance with guidance in NUREG-1855 (Reference 43) and EPRI TR-1016737, "Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments," (Reference 78). Based on a review of the above information, the NRC staff concludes that this issue is resolved because the licensee provided adequate descriptions of the updates. As a result of the review of the LAR and responses to RAis, the NRC staff concludes that the lEPRA is technically adequate because its quantitative results, considered together with the sensitivity study results, demonstrate that the change in risk due to the transition to NFPA 805 meets the acceptance guidelines of RG 1.17 4. To reach this conclusion, the NRC staff reviewed all F&Os provided by the peer reviewers and determined that the resolution of the F&Os supports the determination that the quantitative results are adequate or have no significant impact on the FPRA. Accordingly, the NRC staff concludes that the licensee demonstrated that the lEPRA meets the guidance in RG 1.200, Revision 2, that it is reviewed against the applicable SRs in ASME/ANS-RA-Sa 2009, and that it is technically adequate to support the FREs and other risk calculations required for the LAR. 3.4.2.2 Fire PRA Model The licensee evaluated the technical adequacy of the FPRA model by conducting a peer review of the FPRA model using the NEI 07-12 process (Reference 79) and the FPRA part (Part 4) of

-71 -the ASME/ANS-RA-Sa-2009 (Reference 29) PRA Standard, as clarified by RG 1.200, Revision 2. The full scope peer review of the FPRA was performed in October 2011 and serves as the basis for the quantitative risk evaluations for the LAR. LAR Attachment V, Table V-1 provides the licensee's dispositions to all F&Os written against SRs of Part 4 of the ASME/ANS RA-Sa-2009 PRA standard as clarified by RG 1.200, Revision 2. LAR Attachment V, Table V-2 provides the licensee's dispositions to F&Os written against SRs that were determined by the peer review to be met only at CC-I. In a letter dated February 27, 2013 (Reference 7), the licensee clarified which of the F&Os were findings and which were suggestions. Suggestions are peer review comments on how an evaluation could be improved even though CC II might have been achieved. In a letter dated January 3, 2013 (Reference 15) in PRA RAI 12, the NRC staff requested that the licensee identify any changes made to the FPRA that are consistent with the definition of a "PRA upgrade" since the last full-scope peer review of PRA models as defined by the ASME/ANS PRA Standard (Reference 29). In a letter dated February 27, 2013 (Reference 7) the licensee responded to PRA RAis 12 and 27 and discussed revisions made and stated that no changes meeting the definition of a PRA upgrade were made to the lEPRA or FPRA since their respective peer reviews. As described in LAR Attachment V, the licensee dispositioned each F&O by assessing the impact of the F&O on the FPRA. The NRC staff requested additional information to assess the adequacy of some of the F&O dispositions. The NRC staff evaluated each F&O and the licensee's respective disposition in LAR Attachment V to determine whether the F&O had any significant impact on the LAR. The NRC staff's conclusions on the resolution of each F&O is summarized in the NRC's Record of Review (Reference 77). F&Os requiring additional information for the staff to complete its review are discussed below. Regarding the lack of breaker coordination noted in F&O 4-3, in letters dated February 27, 2013 (Reference 7) and January 22, 2014 (Reference 11) the licensee responded to SSD RAis 08 and 08.01 and demonstrated via a sensitivity analyses that modeling the impact of uncoordinated breakers in the FPRA model has an insignificant impact on the risk results. Furthermore, the licensee incorporated the modeling of the uncoordinated breakers into the integrated analysis as well as the baseline PRA that will be used for self-approval as reported in the response to PRA RAI 61 in the letter dated April 15, 2014 (Reference 13). The NRC staff finds this issue to be resolved because baseline PRA has been modified and the transition change-in-risk estimates, submitted by the licensee on May 23, 2014 (Reference 92) as a supplement to those provided in response to PRA RAI 61, include the impact of uncoordinated breakers. In letters dated February 27, 2013 (Reference 7), March 14, 2014 (Reference 12), and April15, 2014 (Reference 13), the licensee responded to PRA RAis 01, 01.01, 59, and 59.01 where the NRC staff requested that the licensee provide addition information regarding a possible increase in HRR caused by the spread of a fire from the ignition source to other combustibles and information on how the transient zone approach supports the development of fire scenarios. The licensee's responses described its "transient zone" approach and, in response to PRA RAI 59.01, the licensee stated that transient zones constitute subdivisions of a

-72-fire zone and that these transient zones are defined according to the physical configuration of targets and ignition sources located within the fire zone. Initially, in the FPRA, the entire transient zone is assumed damaged at time zero. To accommodate further detailed fire modeling, a fire scenario progression model was established. Fire damage progression begins with damage to targets within the ignition source zone of influence (ZOI) and concludes with full transient zone damage, development of a hot gas layer, or damage to adjacent transient zones. Greater resolution is applied to the scenario progression model for risk-significant scenarios from evaluation of fire spread and propagation from ignition sources through intervening combustibles. Suppression may be credited at the various damage states prior to assuming damage. For cases where a hot gas layer (HGL} is postulated in the fire zone, the entire fire zone is assumed damaged. In response to PRA RAI 01.01, the licensee clarified that no credit is taken for automatic suppression preventing damage to the first target set (i.e., cable trays) above the ignition source. For example, even though sprinklers are installed in the cable trays, those cables in a tray immediately above the ignition source are damaged with no credit for the sprinkler preventing damage. According to the licensee, such an assumption is conservative, given that the timing results from activation models, while not explicitly used by the FPRA, indicate that automatic fixed suppression systems would activate prior to damage. All equipment within or outside of a transient zone boundary is failed by the fire if it lies within the ZOI of the ignition source. However, the staff found that the ZOI was inadequate for capturing all scenarios outside the transient zone, as a path for direct fire effects and fire propagation existed across some boundaries between transient zones. In response to PRA RAI 59, the licensee added 238 new scenarios by extending damage beyond the transient zone of origin for both transient and fixed ignition sources. Scenarios were developed for transient zones where combustibles, i.e. cable trays, extended across the transient zone boundary. In response to PRA RAI 59.01, the licensee stated that fire propagation to adjacent transient zones is not otherwise postulated due to the absence of intervening combustibles located at the transient zone boundaries, as confirmed by walkdowns. The licensee also stated that during target mapping, when targets were found to be located close to transient zone boundaries, they were included in both transient zones and that fixed ignition sources that are located near transient zone boundaries are assigned to both transient zones. Finally, the licensee indicated that the ZOI is not restricted to the transient zone containing the ignition source. The NRC staff finds the licensee's fire scenario progression model, including the structure embedded in the transient zone method, acceptable because it appropriately captures the impact of flame spread and fire propagation on fire scenario progression. In a letter dated January 3, 2013 (Reference 15), the NRC staff determined that the licensee had not provided sufficient information regarding transient fires and requested, in PRA RAI 02, that the licensee explain how transient fires were placed to cover pinch points where CCDPs are highest for a given physical analysis unit (PAU). In a letter dated February 27, 2013 (Reference 7), the licensee responded to the RAI and stated that transient and hot work fires are postulated anywhere a transient fire is reasonably expected to occur, which is considered to be all accessible floor areas not precluded by design or operation (e.g., occupied by permanent fixtures such as plant equipment, or inerted during power operation). The NRC staff concludes

-73-that the licensee's method for locating transient fires appropriately captures all pinch points and therefore concludes that the licensee's response is acceptable. In a letter January 3, 2013 (Reference 15) in PRA RAI 53, the NRC staff requested further clarification on an alternative method to NUREG/CR-6850 that was used for transient zone scenarios involving junction box (JB) fires, self-ignited cable fires (SICF) and cable fires caused by welding and cutting (CFWC). In letters dated February 27, 2013 (Reference 7) and March 14, 2014 (Reference 12) the licensee responded to PRA RAis 53 and 53.01 and stated that because the treatment of junction boxes, self-ignited cable fires and cable fires due to hot work assumes only one cable tray or conduit is damaged, the selection of this target is based on a bounding approach. Based on a review of the licensees RAI responses, the NRC staff finds the licensee's approach for CFWC and SICF acceptable because the entire CFWC and SICF frequencies apportioned to the transient zone are applied to the CCDP of the single cable tray in the transient zone with the largest CCDP. In regard to JB fires, the NRC staff finds the licensee's approach acceptable because the entire JB frequency apportioned to the transient zone is applied to either the CCDP of the full fire zone, of the full transient zone, or of the single conduit in the transient zone with the largest CCDP. Lastly, for each scenario type (i.e., CFWC, SICF and JB fires), the licensee's sensitivity analysis for those transient zones with the highest risk values indicates that the risk estimates change negligibly compared to the transition risk acceptance guidelines and not substantively compared to the much smaller self-approval acceptance guidelines; therefore, the NRC staff finds the method acceptable. As part of supplemental information provided by the licensee on May 23, 2014 (Reference 92), the Fire PRA model was updated to reduce risk because the licensee identified an error in the logic of the FPRA. In particular, fault tree logic was modified to credit manual isolation of the reactor water cleanup break. Additionally, mutually exclusive failures related to depressurization were removed from the model. Finally, human error probabilities were made conditional to reflect the dependency between associated operator actions. The NRC finds that the modifications to the Fire PRA model are acceptable since the updates reflect modifications consistent with standard PRA practices. Also the baseline PRA has been modified and the transition change-in-risk estimates, submitted by the licensee on May 23, 2014 as a supplement to those provided in response to PRA RAI 61 include this modification. As part of supplemental information provided on May 23, 2014 (Reference 92), the licensee also reduced the portion of those fire scenarios in Fire Zone T3A assumed to result in full zone damage which resulted in a reduction in the frequency of fire scenario T3A-CMPT-9 which the licensee stated was among the top contributors to L1CDf and L1LERF. The licensee considered whether relatively large oil fires generated by the main feedwater pumps could, as the original analysis assumed, generate a hot gas layer, and determined that only a very large main feedwater pump oil fire could produce a hot gas layer. The corresponding severity factor from Supplement 1 to NUREG/CR-6850 (Reference 36) was applied. The NRC staff concludes that the licensee has addressed the error in the FPRA logic because the baseline PRA has been modified, and the transition change-in-risk estimates, submitted by the licensee on May 23, 2014 as a supplement to those provided in response to PRA RAI 61, reflect the use of an acceptable method to establish the size and probability of relatively large oil fires from main feedwater pumps.

-74-In a letter dated January 3, 2013 (Reference 15) in PRA RAI 06, the NRC staff noted that new information indicated that the credit taken for hot short probabilities in circuits that contain control power transformers (CPTs) identified in NUREG/CR-6850 was invalid when estimating circuit failure probabilities. The licensee provided revised risk results removing credit for CPTs from all hot short probabilities in the integrated analysis as well as the baseline PRA that will be used for self-approval as reported in the response to PRA RAI 61 in the letter dated April 15, 2014 (Reference 13). Ultimately, in the supplemental information provided by the licensee on May 23, 2014 (Reference 92), the licensee elected to modify their PRA model with supplemental interim technical guidance issued by the NRC in 2014 (ML 14017A135) that provided additional guidance regarding fire-induced circuit failure mode likelihood analysis. The NRC staff finds this issue to be resolved because the baseline PRA has been modified, and transition change-in-risk estimates, submitted by the licensee on May 23, 2014 as a supplement to those provided in response to PRA RAI 61, reflect the latest probabilities for occurrence of spurious operations and their duration. In a letter dated January 3, 2013 (Reference 15) in PRA RAI 18, the NRC staff requested clarification on whether system notebook updates noted in the disposition to F&O 2-2 would occur as part of any of the implementation items provided in LAR Attachment S, Table S-2. In a letter dated February 27, 2013 (Reference 7) the licensee responded to PRA RAI 18 and stated that the updates are to occur as part of Implementation Item 3 in LAR AttachmentS, Table S-2. The NRC staff concludes that the response to the RAI is acceptable because the licensee indicated that the updates are included as part of an implementation item in LAR Attachment S, Table S-2. In a letter dated January 3, 2013 (Reference 15) in PRA RAI 47 the NRC staff requested the licensee to update the fire ignition sources in the FPRA to reflect the peer review insights in F&O 5-35. In a letter dated February 27, 2013 (Reference 7), the licensee responded to PRA RAI 47 and clarified that the dispositions to F&Os 1-5 and 5-32 already addressed the specific issues raised by F&O 5-35 regarding the identification and counting of ignition sources for the FPRA model used to support the LAR. Furthermore, the licensee indicated that they would review and update the fire ignition sources as part of Implementation Item 3 in LAR AttachmentS, Table S-2. The NRC staff concludes that the licensee response to the RAI is acceptable because the licensee clarified the F&Os and indicated that the fire ignition sources would be updated as part of a implementation item. In letters dated January 3, 2013 (Reference 15) and February 12, 2014 (Reference 17) the NRC staff requested that the licensee provide additional information regarding control room abandonment. In letters dated February 27, 2013 (Reference 7) and April 15, 2014 (Reference 13), the licensee responded to PRA RAis 20 and 20.01 and described its methodology to evaluate risk from MCR abandonment due to MCR habitability. The licensee stated that the plant operators have said that they would not leave the MCR solely due to loss of control and as a result, their analysis focuses on loss of habitability. The loss of habitability analysis does look at cases where control is challenged, and even lost. The licensee clarified that a detailed scenario analysis was performed to address MCR abandonment risk and that the resulting scenarios were modeled in the FPRA consistent with other PRA initiating events. The MCR abandonment scenario analysis considers both fire

-75-propagation and subsequent fire-induced damage of PRA-credited equipment. The probability of MCR abandonment is reflected in the scenario frequencies of each MCR fire scenario. The resulting MCR abandonment scenarios consider human failure events (HFEs) in combination with random equipment failure. The human reliability analysis (HRA) performed on operator actions for MCR abandonment utilized the seeping method of NUREG-1921 (Reference 44). For internal events actions adopted *in the FPRA for MCR abandonment, the licensee adjusted time delay, diagnosis time, and execution times from the internal events scenarios to account for fire human error probabilities (HEPs). Specific times for these parameters were identified for MCR abandonment with no internal events counterpart. The licensee stated that seeping and detailed methods of NUREG-1921 were applied in an example demonstrated in the RAI response. The licensee stated that dependencies between actions of similar function were addressed and that a cutset review was subsequently performed to capture those dependencies that may exist independent of function. In response to PRA RAI 20.01, the licensee summarized the MCR abandonment scenarios and the corresponding quantification for CCDP. The summary demonstrates that a full range of CCDPs was developed, including CCDPs less than 0.001, between 0.001 and less than 1.0, and a CCDP of 1.0. Thus, the full range of possible scenarios have been analyzed, (i.e. scenarios from successful shutdown being straightforward to fire-induced failures causing great difficulty for shutdown by failing multiple functions and/or causing complex spurious operations). The NRC staff concludes that the MCR abandonment evaluation of the CCDP is acceptable because a comprehensive approach was applied with acceptable methods and the results indicated that the expected range of possible scenarios were produced and evaluated. In a letter dated February 12, 2014 (Reference 17) in PRA RAI 60, the NRC staff noted that the licensee's approach employed to develop and quantify fire scenarios in the evaluation of MCB fires differed from the guidance provided in NUREG/CR-6850. In a letter dated March 14, 2014 (Reference 12), the licensee responded to PRA RAI 60 and revised the MCB analysis to use the method in Appendix L of NUREG/CR-6850. The licensee evaluated each MCR electrical panel in accordance with guidance in NUREG/CR-6850, as supplemented, to determine those that may be defined as part of the MCB. Furthermore, the licensee incorporated the use of Appendix L and the refined definition of the MCB into the integrated analysis as well as the baseline PRA that will be used for future self-approval as reported in the response to PRA RAI 61 in the letter dated April15, 2014 (Reference 13). The NRC staff finds the licensee's response to the RAI acceptable because the licensee modified their baseline PRA, justified their definition of a MCB, and provided the transition change-in-risk estimates, submitted by the licensee on May 23, 2014 (Reference 92) as a supplement to those provided in response to PRA RAI 61, to reflect use of an acceptable method, Appendix L of NUREG/CR-6850, to evaluate MCB fire scenarios. In a letter dated January 3, 2013 (Reference 15), the NRC staff requested in PRA RAI 41 that the licensee provide additional information regarding fire events and whether there is an effect on generic fire frequencies. In a letter dated February 27, 2013 (Reference 7), the licensee responded to PRA RAI 41 and demonstrated that generic data is representative of plant experience prior to 2000. In reviewing plant-specific fire events from 2000 to 2008, the licensee identified two potentially challenging fires, one each for Bins 15.1 and 36, and performed a Bayesian update on each bin's frequency. Furthermore, the licensee incorporated the revised

-76-frequencies for Bins 15.1 and 36 into the integrated analysis as well as the baseline PRA that will be used for future self-approval as reported in the response to PRA RAI 61 in the letter April 15, 2014 (Reference 13). The NRC staff finds the licensee's response to the RAI acceptable because the baseline PRA has been modified and the transition change-in-risk estimates, submitted by the licensee on May 23, 2014 (Reference 92) as a supplement to those provided in response to PRA RAI 61, include the current (i.e., updated) fire ignition frequencies. In a letter dated January 3, 2013 (Reference 15) in PRA RAI 55, the NRC staff noted that the multi-compartment analysis (MCA) credited fixed gaseous suppression systems in exposed compartments without adequate technical justification (i.e., without consideration that the fire has already grown and propagated from the original exposing compartment). In a letter dated February 27, 2013 (Reference 7) the licensee responded to PRA RAI 55 and stated that the only gaseous suppression system credited in the MCA was the Halon system in Fire Area 11, Fire Zone C2, and demonstrated that removal of the credit for this system in the MCA had a negligible risk impact, upon crediting the automatic sprinkler system(s) in compartments exposed by C2 or exposing to C2. In letters dated March 14, 2014 (Reference 12) and April 15, 2014 (Reference 13), the licensee responded to PRA RAI 61 and further stated that credit for the Halon system is no longer credited in any multicompartment scenarios. Further, the licensee removed credit for the Halon system in the MCA in the integrated analysis as well as the baseline PRA that will be used for future self-approval. The NRC staff finds the licensee's response to the RAI acceptable because the baseline PRA has been modified and the transition change-in-risk estimates, submitted by the licensee on May 23, 2014 (Reference 92) as a supplement to those provided in response to PRA RAI 61, no longer include credit for the Halon system in any exposed compartment in the MCA. In a letter dated January 3, 2013 (Reference 15) in PRA RAI 56, the NRC staff requested that the licensee justify the use of a 0.04 generic estimate of unreliability and unavailability for manually actuated carbon dioxide (C02) suppression systems credited to prevent structural steel failure for emergency diesel generator (EDG) and turbine generator fire scenarios, noting that the generic estimates provided in Appendix P of NUREG/CR-6850 only apply to automatically actuated systems. In a letter dated February 27,2013 (Reference 7) the licensee responded to PRA RAI 56 and provided a sensitivity analysis that replaced the 0.04 estimate with an HEP of 0.1 for failure to manually activate the C02 suppression system in 30 minutes. The licensee further clarified that this HEP is consistent with those estimated for comparable actions modeled in the FPRA using the NUREG-1921 scoping HRA method and that 30 minutes is conservative relative to those fire events listed in Appendix 0 of NUREG/CR-6850 involving structural failure. The licensee incorporated this change into the integrated analysis as well as the baseline PRA that will be used for future self-approval as reported in the response to PRA RAI 61 in the letter dated April 15, 2014 (Reference 13). The NRC finds the licensee's response to the RAI acceptable since the baseline PRA has been modified and the transition change-in-risk estimates, submitted by the licensee on May 23, 2014 (Reference 92) as a supplement to those provided in response to PRA RAI 61, reflect this change In letters dated January 3, 2013 (Reference 15) and February 12, 2014 (Reference 17) in PRA RAis 11 and 11.01 the NRC staff requested that the licensee propose a method to verify the validity of the reported change-in-risk following completion of proposed modifications and implementation items, and to include a plan of action to notify the NRC if the risk guidelines are

-77-exceeded. In letters dated February 27, 2013 (Reference 7) and March 14,2014 (Reference 12), the licensee responded to PRA RAis 11 and 11.01 and revised Implementation Item 4 in LAR Attachment S, Table S-2 to include a plan of action to notify the NRC if risk acceptance guidelines are exceeded subsequent to completion of both PRA-credited modifications and implementation items. The NRC staff concludes that the licensee's response to the RAis is acceptable because the licensee identified an appropriate plan of action and included the plan in LAR AttachmentS, Table S-2 as an implementation item. As a result of its review of the LAR, as supplemented, the NRC staff concludes that the FPRA is sufficiently technically adequate and that its quantitative results, considered together with the sensitivity studies, can be used to demonstrate that the change in risk due to the transition to NFPA 805 meets the acceptance guidelines in RG 1.174 and FPRA results are acceptable. 3.4.2.3 Fire Modeling in Support of the Development of the Fire Risk Evaluations (FREs) The NRC staff performed detailed reviews of the FM used to support the NMP1 FREin order to gain further assurance that the methods and approaches used for the application to transition to NFPA 805 were technically adequate. NFPA 805 has the following requirements that pertain to FM used in support of the development of the FRE: NFPA 805, Section 2.4.3.3: On Acceptability The [probabilistic safety assessment (PSA)] approach, methods, and data shall be acceptable to the AHJ. NFPA 805, Section 2.7.3.2, "Verification and Validation": Each calculational model or numerical method used shall be verified and validated through comparison to test results or comparison to other acceptable models. NFPA 805, Section 2.7.3.3, "Limitations of Use": Acceptable engineering methods and numerical models shall only be used for applications to the extent these methods have been subject to verification and validation. These engineering methods shall only be applied within the scope, limitations, and assumptions prescribed for that method. NFPA 805, Section 2.7.3.4, "Qualification of Users": Cognizant personnel who use and apply engineering analysis and numerical models (e.g., fire modeling techniques) shall be competent in that field and experienced in the application of these methods as they relate to nuclear power plants, nuclear power plant fire protection, and power plant operations.

-78-NFPA 805, Section 2.7.3.5, "Uncertainty Analysis": An uncertainty analysis shall be performed to provide reasonable assurance that the performance criteria have been met. The following sections discuss the results of the NRC staff's reviews of the acceptability of the FM (first requirement). The results of the NRC staff's review of compliance with the remaining requirements are discussed in Sections 3.8.3.2 through 3.8.3.5 of this SE. 3.4.2.3.1 Overview of Fire Models Used to Support the FREs FM was used to develop the ZOI around ignition sources in order to determine the thresholds at which a target would exceed the critical temperature or radiant heat flux. This approach provides a basis for the seeping or screening evaluation as part of the NMP1 FRE. The following algebraic fire models and correlations were used for this purpose:

Plume Temperature, Method of Heskestad (Reference 40, Chapter 9)

Flame Height, Method of Heskestad (Reference 40, Chapter 3)

Radiant Heat Flux, Point Source Radiation Model (Reference 40, Chapter 5) These algebraic models are described in NUREG-1805, "Fire Dynamics Tools (FDT5): Quantitative Fire Hazard Analysis Methods for the US Nuclear Regulatory Commission Fire Protection Inspection Program" (Reference 40). The V&V of these algebraic models is documented in NUREG-1824, (Reference 41). The V&V of the fire models that were used to support the FRE is discussed in SE Section 3.8.3.2. In LAR Attachment J and in several F&O resolutions in LAR Attachment V, the licensee also discussed the use of the following additional empirical correlations that are addressed in NUREG-1805 or FIVE, but for which V&V is not addressed in NUREG-1824, Volume 3 and 4.

Heat and Smoke Detection Actuation Correlations, Method of Chapter 1 0 and Chapter 11 of NUREG-1805 (Reference 40).

Correlation for Flame Spread over Horizontal Cable Trays, FLASH-CAT, described in NUREG/CR-701 0, "Cable Heat Release, Ignition, and Spread in Tray Installations During Fire (CHRISTl FIRE), Volume 1: Horizontal Trays" (Reference 42). The licensee used the ZOI approach as a screening tool to distinguish between fire scenarios that required further evaluation and those that did not require further evaluation. The licensee stated that qualified personnel performed a plant walk-down to identify: ignition sources, surrounding targets or SSCs in compartments and applied the empirical correlation screening tool to assess whether the SSCs were within the ZOI of the ignition source. Based on the fire hazard present, these generalized ZOis were used to screen from further consideration those NMP1-specific ignition sources that did not adversely affect the operation of credited SSCs, or

-79-targets, following a fire. The licensee's screening was based on the 981h percentile fire HRR from the NUREG/CR-6850 methodology. The Consolidated Model of Fire and Smoke Transport (CFAST), Version 6, zone model was used for:

HGL temperature calculations in specific fire zones

MCR abandonment time calculations Detailed FM using CFAST was performed for selected fire scenarios in fire areas. CFAST zone model was used for the temperature sensitive equipment HGL study and the MCR abandonment time calculations. The FRE used these calculations to further screen ignition sources, scenarios, and compartments that would not be expected to generate an HGL, and to identify the ignition sources that have the potential to generate an HGL for further analysis. The V&V of CFAST zone model is documented in NUREG-1824, Volume 5 (Reference 41). The V&V of all empirical correlations and fire models that were used to support the NMP1 FREis discussed in detail in Section 3.8.3.2 of this SE. 3.4.2.3.2 RAis Pertaining to Fire Modeling in Support of the NMP1 Fire PRA In a letter dated January 3, 2013 (Reference 15), the NRC staff submitted RAis concerning the FM conducted to support the NMP1 FRE. In a letter dated February 27, 2013 (Reference 7), the licensee responded to these RAis. The following paragraphs describe selected RAI responses related to the acceptability of the fire models used.

The NRC staff issued FM RAI 01 (c) (Reference 15), to ask the licensee to clarify how wall and corner effects were accounted for in the FRE. In its response to FM RAI 01 (c) (Reference 7), the licensee explained that the "image" method was used to account for wall and corner effects in the flame height and plume temperature calculations. Based on this method, the HRR of ignition sources within one foot of a wall is doubled and the fire diameter is multiplied by -JZ. For ignition sources within one foot of a corner, the HRR and diameter are multiplied by 4 and 2, respectively. Furthermore, the licensee explained that wall and corner effects were not explicitly accounted for in the CFAST HGL calculations, but that they were implicitly addressed by the use of bounding HRRs for the ignition sources and intervening combustibles in the fire zone. Based on a review of the explanation and justification provided in response to FM RAI 01 (c), the NRC staff concludes that the licensee's response to the RAI is acceptable because the approach accounts for wall and corner effects in the ZOI.

The NRC staff issued FM RAI 01 (d) (Reference 15), to ask the licensee to provide technical justification for the fire elevation assumed in the MCR abandonment calculations.

-80-In its response to FM RAI 01 (d) (Reference 7), the licensee explained that the effect of fire elevation on MCR abandonment was evaluated at the start of the analysis process. The licensee concluded from these preliminary calculations that an elevation of 3 feet, which is the approximate height of the bench board portion of the MCB, results in conservative estimates of MCR abandonment times. Based on a review of the results of the preliminary calculations and the additional information provided in response to FM RAI 01 (d), the NRC staff concludes that the licensee response to the RAI is acceptable because the licensee's assumption of using a 3 foot fire elevation in the MCR abandonment time calculations leads to conservative results.

The NRC staff issued FM RAI 01 (e) (Reference 15), to ask the licensee to explain how the horizontal natural vent flow area in the MCR abandonment calculations was determined. In its response to FM RAI 01 (e) (Reference 7), the licensee explained that the ventilation flow area was based on the guidelines in an authoritative publication of the American Society of Heat, Refrigerating and Air-Conditioning Engineers Inc., (ASH RAE) and the Society of Fire Protection Engineers (SFPE), "Principles of Smoke Management," Table 6.3 (Reference 82). In addition, the licensee stated that the fires considered in the MCR abandonment time calculations are not oxygen limited during the time that the doors are closed. Based on a review of the additional information provided in response to FM RAI 01 (e), the NRC staff concludes that the licensee's response to the RAI is acceptable because the licensee based the ventilation flow area on guidance provided in authoritative publications and the licensee demonstrated that fires are not considered oxygen limited during times that the doors are closed, which leads to conservative results.

The NRC staff issued FM RAI 01 (f) part i (Reference 15), to ask the licensee to provide technical justification for not considering transient ignition source fires in the MCR abandonment time calculations. In its response to FM RAI 01 (f) part i (Reference 7), the licensee explained that in the cabinet fire scenarios with operating heating, ventilation, and air conditioning (HVAC), a minimum HRR of 475 kW is needed to cause MCR abandonment. Hence, transient fires with a HRR of 317 kW are not expected to cause abandonment. For cabinet fire scenarios without mechanical ventilation the minimum HRR for abandonment is 200 kW. Consequently, transient fires may have a significant contribution to MCR abandonment when the HVAC system is inoperative. However, the licensee estimated a small probability of 0.0027 of having a transient fire contributing to MCR abandonment that has not been

-81 -included in their evaluation. The licensee noted that this contribution is small and does not affect the current risk results. Based on the additional information provided, the TS requirement that control room ventilation be operable the TS limitation on the length of time (7 days) the plant can operate without ventilation, and the ignition frequency for transients in the MCR being an order of magnitude smaller than that for fixed ignition sources, the NRC staff concludes that the licensee's explanation and justification for not considering transient ignition source fires in the MCR abandonment time calculations is acceptable because transient fires with a HRR of 317 kW are not expected to cause abandonment.

The NRC staff issued FM RAI 01 (f) part ii (Reference 15) to ask the licensee to provide technical justification for not considering fire spread between panels in the MCR abandonment time calculations. In its response to FM RAI 01 (f) part ii (Reference 7), the licensee explained that the FPRA event tree model to evaluate panel fires in the MCR includes sequences for small fires that are promptly suppressed, fires that do not propagate outside the panel of fire origin, and fires that propagate to adjacent panels. Non-propagating fires and propagating fires are both split between fires that cause abandonment and those that do not result in abandonment. The control room abandonment split fraction is based on the probability of observing a HRR that is high enough to cause abandonment. The latter is based on the cumulative severity factor for the bins that result in abandonment as determined in the MCR abandonment time calculations. The split fraction is identical for propagating and propagating fires. To quantify the extent of damage in propagating fires, the licensee followed the guidelines in NUREG/CR-6850, and assumed adjacent panels are involved at 1 0 minutes. Based on a review of the response to FM RAI 01 (f) part ii regarding use of the MCR abandonment time calculations in the FPRA, the NRC staff concludes that the licensee's approach to determine the probability for MCR abandonment based on panel fires that do not propagate to adjacent panels, is acceptable because the licensee followed the guidance in NUREG/CR-6850 and assumed adjacent panels are involved at 10 minutes.

The NRC staff issued FM RAI 01 (h) (Reference 15) to ask the licensee to explain how the reduction of the gas volume in the MCR due to large cabinets and other contents was accounted for in the MCR abandonment time calculations, or to provide technical justification for ignoring the volume reduction. In its response to FM RAI 01 (h) (Reference 7), the licensee explained that a sensitivity study was performed to determine the impact on the CFAST results of reducing the volume by 10% (estimated fraction of the volume occupied by electrical panels). For the fire scenarios that were considered the smaller volume resulted in a 0.3-0.5 min MCR abandonment time reduction.

-82-Based on a review of the responses to FM RAI 01 (h) on the results of the sensitivity study and because expected heat losses to the panels were not accounted for in the sensitivity calculations, the NRC staff concludes that ignoring the MCR volume reduction due to electrical panels occupying part of the space in the CFAST abandonment time calculations is acceptable.

The NRC staff issued FM RAI 01 (i) (Reference 15) to ask the licensee to explain how the reduction of the compartment volume due the presence of obstructions was accounted for in the CFAST HGL analyses, or to provide a technical justification for not considering this effect. In its response to FM RAI 01 (i) (Reference 7), the licensee explained that the volume reduction was not explicitly accounted for, but that the effect of the volume reduction on the HGL temperature is negated by the decision to ignore the heat sink that enclosed obstructions provide. In addition, the licensee performed a sensitivity analysis to assess the effect on MCR abandonment time of a 1 0% volume reduction to account for the space occupied by obstructions. The analysis showed that the volume reduction reduces the abandonment time from 8 min to 7.5 min for bin 8 (1024 kW), and from 6.5 to 6.2 min for bin 15 (2276 kW). Based on a review of the results of the sensitivity analysis and because the volume reduction impact is offset by ignoring the heat sink effect, the NRC staff concludes that ignoring the compartment volume reduction due to obstructions occupying part of the space in the CFAST HGL calculations is acceptable

The NRC staff issued FM RAI 01 U) (Reference 15), to ask the licensee to provide technical justification for the response time index (RTI) value that was used in the sprinkler activation calculations. In its response to FM RAI 01 (j) (Reference 7), the licensee stated that no RTI was available for the sprinklers installed at NMP1. To address the uncertainty of the RTI value, the licensee explained that automatic suppression is only credited after a number of targets near the suppression system have failed, and that, therefore, cable targets are used as surrogates for activation of automatic suppression. Based on a review of the response to FM RAI 01 U), the NRC staff concludes that the licensee's approach to determine sprinkler activation time is acceptable because the licensee demonstrated that automatic suppression is only credited after a number of targets near the suppression system have failed and that this is a conservative surrogate for the RTI.

The NRC staff issued FM RAI 01 (k) (Reference 15) to ask the licensee to explain how the effeCt of the HRR from intervening combustibles on the ZOI was accounted for.

-83-In its response to FM RAI 01 (k) (Reference 7), the licensee explained that transient zones were selected so that they are larger than the calculated ZOI, and that the cable tray layout was considered in the selection of transient zones. The licensee further explained that additional scenarios were included in the NMP1 FPRA to account for potential fire propagation between transient zones. Based on a review of the response to RAI FM RAI 01 (k), the NRC staff concludes that the licensee's explanation and approach to identify damaged targets for fire scenarios that involve intervening combustibles is acceptable because the licensee demonstrated that fire scenarios involving intervening combustibles have been addressed in the FPRA.

The NRC staff issued FM RAI 01 (I) (Reference 15) to ask the licensee how the time to ignition of the lowest tray in a stack above a cabinet was determined. In its response to FM RAI 01 (I) (Reference 7), the licensee explained that in the CFAST HGL calculations an ignition time of 5 minutes was assumed for the lowest cable trays. For fires that involved intervening combustibles, only two zones were identified where HGL conditions were determined. For these two zones, the licensee performed additional CFAST HGL calculations using the ignition time calculated as described in NUREG/CR-6850, Appendix R. In both cases this reduced the time to reach HGL conditions by less than 2 minutes, from approximately 20 minutes for an assumed ignition time of 5 minutes. The time to damage was updated in the FM database. The licensee performed additional CFAST HGL calculations using the ignition time calculated as described in NUREG/CR-6850, Appendix R, and time to damage was updated in the FM database, which resulted in an insignificant increase to CDF. The NRC staff concludes that the licensee's assumption concerning the ignition time of the lowest tray in a stack above a burning cabinet is acceptable because the licensee performed additional CFAST HGL calculations using the ignition time, which resulted in an insignificant increase to CDF.

The NRC staff issued FM RAI 01 (m) part i (Reference 15) to ask the licensee to provide technical justification for the generic assumption of a fire dimension of 2 feet for electrical cabinets and transient ignition sources. In its response to FM RAI 01 (m) part i (Reference 7), the licensee explained that, although most scenarios assume a fire diameter of 2 feet, ten scenarios assume a different diameter, ranging from 1.5 to 4.3 feet. Furthermore, the licensee explained that the FPRA is generally not affected by the fire diameter because ( 1) the ZOI corresponds to the transient zones, which are larger than the calculated ZOI, and (2) when suppression is credited, a set of trays near the ignition source is always assumed failed.

-84-Based on a review of the response to FM RAI 01 (m) part i, the NRC staff concludes that the licensee's assumptions concerning the fire diameter at NMP1 are acceptable because the licensee demonstrated that the ZOI corresponds to the transient zones, which are larger than the calculated ZOI and assumption regarding failure of cable trays with suppression system activation.

The NRC staff issued FM RAI 01 (m) part ii (Reference 15) to determine why a characteristic length of 2 feet was used in cable tray fire propagation calculations instead of the cabinet length per NUREG/CR-6850. In its response to FM RAI 01 (m) part ii (Reference 7), the licensee presented the results of a sensitivity analysis, which shows that increasing the characteristic length to the cabinet length reduces the time to HGL conditions by approximately 1 minute, to approximately 18 minutes for the base case with a characteristic length of 0.3 m. Based on a review of the response to FM RAI 01 (m) part ii, and the sensitivity analysis, the NRC staff concludes that the licensee response to the RAI is acceptable because there is only one zone where HGL calculations are impacted significantly by the HRR of cable trays. The NRC staff also concludes that the licensee's assumed characteristic length in cable tray fire propagation calculations is acceptable because the licensee demonstrated that the sensitivity analysis shows that there is only one zone where HGL calculations are impacted significantly by the HRR and that the impact of characteristic length is not significant. 3.4.2.3.3 Conclusion for Section 3.4.2.3 Based on the licensee's description in the LAR, as supplemented, of the NMP1 process for performing FM in support of the FRE and clarifications provided in response to the RAis, the NRC staff concludes that the licensee's approach for meeting the requirements of NFPA 805, Section 2.4.3.3 is acceptable. 3.4.2.4 Conclusions Regarding Fire PRA Quality Based on NUREG-0800, Section 19.2 (Reference 33), Section 111.2.2.4.1, which summarizes the criteria for the NRC staff's review of PRA Quality required for an application, the NRC staff concludes that the licensee's PRA satisfies the guidance in RG 1.17 4, Section 2.3, and RG 1.205, Section 4.3 regarding the technical adequacy of the PRA used to support risk assessment to support transition to NFPA 805. The NRC staff concludes that the PRA approach, methods and data are acceptable and therefore that NFPA 805 Section 2.4.3.3 is satisfied for the request to transition to NFPA 805. The NRC staff based this conclusion on the findings that: (1) the PRA model meets the criteria in that, upon completion of the modifications discussed in AttachmentS, Table S-1, it would adequately represent the current, as built, as operated configuration, and is therefore capable of being adapted to model both the post-transition and compliant plant as needed; (2) the PRA

-85-model conforms sufficiently to the applicable industry PRA standards for internal events and fires at an appropriate CC, considering the acceptable disposition of the peer review and NRC staff review findings; and (3) the fire modeling used to support the development of the FPRA has been confirmed as appropriate and acceptable. The FPRA used to support risk-informed self-approval of changes to the FPP must use an acceptable PRA approach and acceptable methods and data. The NRC staff concludes that the changes already made to the baseline model to incorporate acceptable methods detailed in Table PRA RAI61-1 and the response to PRA RAI61, in a letter dated March 14,2014 (Reference 12} and discussed above, demonstrate that NFPA 805 criteria are satisfied and the PRA is acceptable for use to support self-approval changes to the FPP program. Finally, based on the licensee's administrative controls (see LAR AttachmentS, Table S-2, Item 3) to maintain the PRA models current and assure continued quality, using only qualified staff and contractors (as described in Section 3.8.3 of this SE}, which are required by the fire protection license condition, the NRC staff concludes that the PRA maintenance process can assure that the quality of the PRA is sufficient to support self-approval of future risk-informed changes to the FPP under the NFPA 805 license condition following completion of all implementation items described in updated LAR AttachmentS, Table S-2. 3.4.3 Fire Risk Evaluations For those fire areas for which the licensee used a performance-based approach to meet the nuclear safety performance criteria, the licensee used FREs in accordance with NFPA 805 Section 4.2.4.2 to demonstrate the acceptability of the plant configuration. In accordance with the guidance in RG 1.205 (Reference 2}, Section C.2.2.4, "Risk Evaluations," the licensee used a risk-informed approach to justify acceptable alternatives to complying with NFPA 805 deterministic criteria. The NRC staff reviewed the following information during its evaluation of the FREs: LAR Section 4.5.2, "Performance Based Approaches," LAR Attachment C, "NEI 04-02 Table B Fire Area Transition," and LAR Attachment W, "Fire PRA Insights." Plant configurations that did not meet the deterministic requirements of NFPA 805, Section 4.2.3.1 were considered VFDRs. VFDRs that will be brought into deterministic compliance through plant modifications need no risk evaluation. The licensee identified in LAR Attachment C, "NEI 04-02 Table B Fire Area Transition, as supplemented by a letter dated April 30, 2013 (Reference 9), that it does not intend to bring all VFDRs into deterministic compliance under NFPA 805. For these VFDRs, the licensee performed evaluations using the risk-informed approach, in accordance with NFPA 805, Section 4.2.4.2, to address FPP non-compliances and demonstrate that the VFDRs are acceptable. All of the VFDRs_ identified by the licensee were categorized as separation issues, with the exception of two cases of noncompliance with the NFPA 805 Section 3.1 minimum design requirement for fire protection systems and features. These two exceptions were resolved with plant modifications (see LAR AttachmentS, Table S-1, Item 11 ). The separation-related VFDRs can generally be categorized into the following four types of plant configurations: ( 1) inadequate separation resulting in fire-induced damage of process equipment or associated cables required for the identified success path; (2) inadequate separation resulting in fire-induced spurious

-86-operation of equipment that may defeat the identified success path; (3) inadequate separation resulting in fire-induced failure of process monitoring instrumentation or associated cables required for the identified success path; and (4) combinations of the above configurations. In letters dated February 27, 2013 (Reference 7), and March 14, 2014 (Reference 12), the licensee responded to PRA RAis 10 and 1 0.01 and described how an FRE is performed for VFDRs. The licensee explained that the change in risk associated with each fire area is obtained by calculating the difference between the CDF and LERF of a compliant plant configuration and the variant (post-transition) plant configuration. The total change in risk was obtained by summing the change in risk for each fire area and comparing the total for the plant to the RG 1.17 4 acceptance guidelines. The licensee explained that some modifications are planned that do not resolve a VFDR deterministically (non-VFDR modifications), but which do reduce risk. The variant plant is modeled with fire-induced component failures included for retained VFDRs, (i.e., VFDRs not brought into deterministic compliance through plant modifications) with all recovery actions at their nominal values, and with all non-VFDR modifications incorporated into the FPRA. Typically, VFDRs are removed from the compliant plant by assuming that the cables required to resolve a VFDR are not affected by a fire. For VFDRs for which recovery actions are credited, the compliant plant is modeled by setting recovery action HEP values to zero. Non-VFDR modifications are generally not included in the compliant case, but the licensee included them in the compliant plant model which conservatively reduces the compliant plant risk in the change in risk estimates. In letters dated February 27, 2013 (Reference 7) and April15, 2014 (Reference 13) the licensee responded to PRA RAI 20 and PRA RAI 20.01 and provided MCR abandonment scenarios in detail. In the response to PRA RAI 10.02 (References 12 and 13), the licensee stated that the same general approach is used to calculate the change in risk from control room abandonment scenarios and summarized the approach as applied to MCR abandonment. For those VFDRs that are considered to have an insignificant change in risk based on qualitative evaluation, the change in risk is not estimated with the PRA but rather designated as epsilon. Finally, four VFDRs are resolved with modifications to obtain deterministic compliance. The NRC staff concludes that the licensee's methods for calculating the change in risk associated with VFDRs are acceptable because they are consistent with RG 1.205, Section 2.2.4.1, and FAQ 08-0054 (Reference 56). The NRC staff further concludes that the results of these calculations for each fire area, demonstrate that the difference between the risk associated with implementation of the deterministic requirements and that of the VFDRs meets the risk acceptance criteria described in NFPA 805, Section 2.4.4.1. 3.4.4 Additional Risk Presented by Recovery Actions The NRC staff reviewed LAR Attachment C, "NEI 04-02 Table B Fire Area Transition," Attachment G, "Recovery Actions Transition," and Attachment K, "Existing Licensing Action Transition," during its evaluation of the additional risk presented by the NFPA 805 recovery actions. SE Section 3.2.5 describes the identification and evaluation of recovery actions.

-87-The licensee used the guidance in RG 1.205, Revision 1 for addressing recovery actions which included the definition of PCS and recovery action. Accordingly, any actions required to transfer control to, or operate equipment from, the PCS, while required as part of the RI/PB FPP, were not considered recovery actions per the RG 1.205 guidance and in accordance with NFPA 805. Conversely, any operator manual actions required to be performed outside the control room and not at the PCS were considered recovery actions. The licensee identified the recovery actions in LAR Attachment G Table G-1, as supplemented in a letter dated March 14, 2014 (Reference 12). Per LAR Attachment G, a total of 251 recovery actions are identified, but just 14 unique operator actions are represented since these actions pertain to multiple fire areas, components, and VFDRs. The additional risk of recovery actions for each area is provided in the Fire Risk Summary Table W-3 as provided by the licensee in letters dated March 14, 2014 (Reference 12} and April 15, 2014 (Reference 13) in response to PRA RAI 61. The licensee explained in the LAR that not all recovery actions listed in LAR Attachment G are modelled in the PRA. In a letter dated February 27, 2013 (Reference 7), the licensee responded to PRA RAI 10 and stated that for those VFDRs considered to have an insignificant change in risk based on a qualitative evaluation, the additional risk of any associated recovery actions is judged also to be insignificant and is not modeled in the FPRA. In a letter dated March 14, 2014 (Reference 12), the licensee responded to PRA RAI 10.01 and clarified that when calculating the additional risk of recovery actions, non-VFDR related modifications that reduce risk were modelled in both the compliant and post-transition cases. According toLAR Attachment W, the additional risk of recovery actions is an increase in CDF of 2.61 E-06/year and an increase in LERF of 1.98E-07/year. These values are below the change in risk acceptance guidelines in RG 1.17 4, and RG 1.205. A review of the detailed results in LAR Attachment W, Table W-3 indicates that the additional risk of recovery actions in each area is also below the RG 1.17 4 acceptance guidelines. Per LAR Attachment G, the licensee reviewed all of the recovery actions for adverse impact on plant risk per FAQ 07-0030 (Reference 51) and stated that no recovery actions listed in LAR Attachment G, Table G-1 were found to have an adverse impact. Furthermore, all recovery actions listed in LAR Attachment G were evaluated against the feasibility criteria provided in NEI 04-02 (Reference 5), FAQ 07-0030 (Reference 51), and RG 1.205 (Reference 2). Implementation Item 19 in LAR AttachmentS, Table S-2 will provide for modifications to procedures for recovery actions being evaluated and for operator training and qualification on the revised procedures. The NRC staff concludes that the licensee's methods for determining the additional risk of recovery actions acceptable because they are consistent with RG 1.205, Section 2.2.4.1 and FAQ 07-0030. Furthermore, the estimated values are less than the acceptance guidelines and therefore the NRC staff concludes that the additional risk of recovery actions meet the requirements of NFPA-805 Section 4.2.4 and 2.4.4.1.

-88-3.4.5 Risk-Informed or Performance-Based Alternatives to Compliance with NFPA 805 The licensee did not use any risk-informed or performance-based alternatives to compliance with NFPA 805. 3.4.6 Cumulative Risk and Combined Changes In LAR AttachmentS, Table S-1, the licensee identified several plant modifications to reduce plant risk rather than bring the plant into compliance with the deterministic requirements of NFPA 805. In a letter dated March 14, 2014 (Reference 12), the licensee responded to PRA RAI 10.01 and stated that the fire risk evaluations included these plant modifications in both the post-transition plant and in the compliant plant. Therefore, the NRC staff concludes that these modifications are not combined changes as discussed in RG 1.17 4, Revision 2, Section 2.1.1, and the risk reduction from these changes need not be separately estimated. The total CDF and total LERF are estimated by adding the risk assessment results for internal, fire, and external hazard events. In letters dated March 14, 2014 (Reference 12), and April 15, 2014 (Reference 13), the licensee responded to PRA RAI 61, and made a number of modifications to the PRA and the PRA methods. After incorporating these modifications in the PRA, the licensee reported a total CDF of 3.21 E-05/reactor-year and a total LERF of 5.1 OE-06/reactor-year, which includes the contribution from internal events (including internal flooding), fire events, and seismic events. In a letter dated May 23, 2014 (Reference 92), the licensee provided supplemental information, changing the fire CDF and LERF to 3.40E-5 and 2.97E-6, respectively. The final CDF and LERF results are summarized below in Table 3.4.6-1. The estimated total CDF and LERF are well below the RG 1.17 4 risk guidelines for Region II (small change) of 1 E-04/year and 1 E-05/year, respectively. This conclusion would not change even if the seismic CDF is estimated at 4.2E-06/year using the preliminary results for the weakest link model from the NRC staff's safety/risk assessment for Generic Issue 199 (GI-199), "Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plants" (Reference 80). Table 3.4.6-1: CDF and LERF for NMP1 after Transition to NFPA 805(1) Hazard Group CDF (/reactor-year) LEAF (/reactor-year) Internal Events (Including Internal Flood) 2.59E-06 2.55E-07 Fire Events 3.40E-05 2.97E-06 Seismic Events 1.07E-06 1.05E-06 TOTAL(2l 3.93E-05 5.04E-06 (1) R1sk results provided 1n the supplement dated May 23, 2014 (Reference 92) and the response to PRA RAI 61 (References 12 and 13). (2) The estimated total CDF and LERF results also include the contribution from other external events (namely, high winds, tornados, transportation and nearby industrial facilities). In the LAR, the licensee also provided the delta (Ll) CDF and LERF estimated for each fire area that is not deterministically compliant, in accordance with NFPA 805, Section 4.2.3,

-89-"Deterministic Approach." In letters dated March 14, 2014 (Reference 12), and April 15, 2014 (Reference 13), the licensee responded to PRA RAI 61, and reported the change in risk estimates based on the PRA after implementing a number of PRA model and method refinements to use NRC-accepted methods. Each of the fire area risk increases are less than the RG 1.174 risk acceptance guidelines for Region II (small change) of 1 E-05/year for and 1 E-06/year for As a supplement to change-in-risk estimates provided in response to PRA RAI 61, the licensee, in a letter dated May 23, 2014 (Reference 92), estimated a total CDF increase of 9.65E-06/year and a LERF increase of 9.09E-07/year associated with the transition to NFPA 805. These estimated risk increases are less than the RG 1.174 risk acceptance guidelines for Region II. The licensee did not provide, and the NRC staff did not request, a new table of changes in risk for each fire area to accompany the supplemental information that the licensee submitted on May 23, 2014. Given that the licensee included non-VFDR modifications in the compliant case, the risk increase for any single fire area is bounded by the total risk increase associated with the transition to NFPA 805. Based on the licensee's total risk and fire risk evaluation results and considering the licensee's treatment of non-VFDR modifications in the compliant case, the NRC staff finds that the risk increase for each fire area associated with transition to NFPA 805 is within the RG 1.174 Region II risk guidelines of 1 E-5/year for .LlCDF and 1 E-6/year for ilLER F. Based on the information above, the NRC staff concludes that the risk associated with the proposed alternatives to compliance with the deterministic criteria of NFPA 805 is acceptable and in accordance with NFPA 805, Section 2.4.4.1. Additionally, the NRC staff concludes that the licensee has satisfied RG 1.174, Section 2.4, and NUREG-0800, Section 19.2 regarding acceptable risk. 3.4. 7 Uncertainty and Sensitivity Analyses The licensee evaluated key sources of uncertainty and sensitivity in response to RAis. In the LAR, the licensee used the updated fire bin frequencies provided in NUREG/CR-6850, Supplement 1, (Reference 36). In letters dated February 27, 2013 (Reference 7), April 30, 2013 (Reference 9), and April 15, 2014 (Reference 13), the licensee responded to PRA RAI 05 and provided the results of a sensitivity analysis using the fire ignition frequency values in NUREG/CR-6850 for those ignition frequency bins having an alpha factor less than or equal to one. The licensee reported that the resultant total .LlCDF of 1.11 E-5/year and total .LlLERF of 1.29E-6/year exceed the RG 1.17 4 risk guidelines. The licensee identified fire protection measures available to the plant but not credited in the PRA that address risk-significant fire areas from the sensitivity analysis for both CDF and LERF. The licensee indicated that Fire Areas 5 and 11 were the largest contributors CDF LEAF, and confirmed that determination. For Fire Area 5, the licensee identified fire detection systems, automatic fire suppression systems, and manual suppression not included in the FPRA but credited for defense-in-depth {DID). For Fire Area 11, the license identified an uncredited manual C02 suppression system. The licensee also identified uncredited procedures and administrative controls applicable to all fire areas. The NRC staff concludes that the additional risk associated with the fire bin frequencies is adequately addressed because of fire protection measures that provide additional DID for risk-significant areas.

-90-In a letter dated January 3, 2013 (Reference 15), in PRA RAI 57, NRC staff requested justification for the assumption that all cables acting as source conductors for an inter-cable hot short are multi-conductor cables. In a letter dated February 27, 2013 (Reference 7), the licensee responded to PRA RAI 57 and stated that multi-conductor cables are much more commonly used and provided the risk results of a sensitivity study using the alternative conservative assumption that all source cables are single conductor cables. The sensitivity study demonstrates that the risk estimates using the conservative assumption change negligibly compared to the transition risk acceptance guidelines, and not substantively compared to the much smaller self-approval acceptance guidelines. Therefore, the NRC staff finds the assumption acceptable for transition and for subsequent self-approval fire risk evaluations. 3.4.8 Conclusion for Section 3.4 Based on the information provided by the licensee in the LAR, as supplemented, regarding the fire risk assessment methods, tools, and assumptions used to support transition to NFPA 805, the NRC staff concludes the following:

The licensee's PRA used to perform the risk assessments in accordance with NFPA 805, Section 2.4.4 {plant change evaluations) and Section 4.2.4.2 (fire risk evaluations), is of sufficient quality to support the application to transition to NFPA 805. The PRA approach, methods, tools and data are acceptable and are in accordance with NFPA 805, Section 2.4.3.3.

The changes to the baseline PRA model which replaces certain approaches, data, and methods identified during the LAR review with acceptable approaches, data, and methods as described is acceptable. The baseline PRA models may be used to support post-transition self-approval of changes because the identified acceptable methods will be used until replaced by other acceptable methods. Implementation Item 4 of LAR AttachmentS, Table S-2, states that the licensee will re-evaluate the risk and the change in risk results after completing implementation of the transition to NFPA-805 and will inform the NRC if risk metrics exceed RG 1.205 risk acceptance guidelines. If these guidelines are exceeded, the licensee will perform additional analytical efforts, and/or procedure changes, and/or plant modifications to assure the RG 1 .205 risk acceptance criteria are met.

The licensee's PRA maintenance process is adequate to support self-approval of future risk informed changes to the FPP following completion of Implementation Items 3, 4 and 6 as described in LAR AttachmentS, Table S-2.

The transition process included a detailed review of fire protection DID and safety margin as required by NFPA 805. The licensee's documentation on DID and SM is acceptable. The licensee's process followed the NRC-endorsed guidance in NEI 04-02, Revision 2, and is consistent with the approved NRC staff guidance in

-91 -RG 1.205, Revision 1, which provides an acceptable approach for meeting the requirements of 10 CFR 50.48(c).

The changes in risk (i.e., LlCDF and LlLERF) associated with the proposed alternatives to compliance with the deterministic criteria of NFPA 805 (fire risk evaluations) are acceptable and the licensee has satisfied the guidance contained in RG 1.205, Revision 1, RG 1.174, and NUREG-0800, Section 19.2, regarding acceptable risk, which meets the requirements of NFPA 805.

The risk presented by the use of recovery actions was determined and provided in accordance with NFPA 805 Section 4.2.4, and the guidance in RG 1.205, Revision 1. The additional risk associated with the NFPA 805 recovery actions is acceptable because the risk for each fire area that relies on a recovery action is below the acceptance guidelines in RG 1.17 4 and therefore meets the acceptance criteria in RG 1.205, Revision 1.

The licensee did not utilize any risk-informed or performance-based alternatives to compliance to NFPA 805 which fall under the requirements of 10 CFR 50.48(c)(4). 3.5 Nuclear Safety Capability Assessment Results NFPA 805, Section 2.2.3, "Evaluating Performance Criteria," states the following: To determine whether plant design will satisfy the appropriate performance criteria, an analysis shall be performed on a fire area basis, given the potential fire exposures and damage thresholds, using either a deterministic or performance-based approach. NFPA 805, Section 2.2.4, "Performance Criteria," states the following: The performance criteria for nuclear safety, radioactive release, life safety, and property damage/business interruption covered by this standard are listed in Section 1.5 and shall be examined on a fire area basis. NFPA 805, Section 2.2.7, "Existing Engineering Equivalency Evaluations," states: When applying a deterministic approach, the user shall be permitted to demonstrate compliance with specific deterministic fire protection design requirements in Chapter 4 for existing configurations with an engineering equivalency evaluation. These existing engineering evaluations shall clearly demonstrate an equivalent level of fire protection compared to the deterministic requirements.

-92-3.5.1 Nuclear Safety Capability Assessment Results by Fire Area NFPA 805, Section 2.4.2, "Nuclear Safety Capability Assessment (NSCA)," states the following: The purpose of this section is to define the methodology for performing a nuclear safety capability assessment. The following steps shall be performed: (1) Selection of systems and equipment and their interrelationships necessary to achieve the nuclear safety performance criteria in Chapter 1 (2) Selection of cables necessary to achieve the nuclear safety performance criteria in Chapter 1 (3) Identification of the location of nuclear safety equipment and cables (4) Assessment of the ability to achieve the nuclear safety performance criteria given a fire in each fire area This SE section addresses the last topic regarding the ability of each fire area to meet the NSPC of NFPA 805. SE Section 3.2.1 addresses the first three topics. NFPA 805, Section 2.4.2.4, "Fire Area Assessment," also states the following: An engineering analysis shall be performed in accordance with the requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression activities on the ability to achieve the nuclear safety performance criteria of Section 1.5 .... In accordance with the above, the process defined in NFPA 805, Chapter 4, provides a framework to select either a deterministic or a PB approach to meet the NSPC. Within each of these approaches, additional requirements and guidance provide the information necessary for the licensee to perform the engineering analyses necessary to determine which fire protection systems and features are required to meet the NSPC of NFPA 805. NFPA 805, Section 4.2.2, "Selection of Approach," states the following: For each fire area either a deterministic or performance-based approach shall be selected in accordance with Figure 4.2.2. Either approach shall be deemed to satisfy the nuclear safety performance criteria. The performance-based approach shall be permitted to use deterministic methods for simplifying assumptions within the fire area. This SE section evaluates the approach used to meet the NSPC on a fire area basis, as well as what fire protection features and systems are required to meet the NSPC. The NRC staff reviewed LAR Section 4.2.4, "Fire Area Transition," LAR Section 4.8.1, "Results of the Fire Area Review," LAR Table 4-3, "Summary Of NFPA 805 Compliance Basis And Required Fire Protection Systems And Features," LAR Attachment C,"NEI 04-02 Table B

-93-Fire Area Transition," LAR Attachment G, "Recovery Actions Transition," LAR Attachment K, "Existing Licensing Action Transition, LAR AttachmentS, "Plant Modifications and Items to be Completed During Implementation" and LAR Attachment T, "Clarification of Prior NRC Approvals" during its evaluation of the ability of each fire area to meet the NSPC of NFPA 805. The NMP1 is a single unit BWR with 26 fire areas including the yard (EXT) and each fire area is comprised of multiple fire zones. Based on the information provided by the licensee in the LAR, as supplemented, the licensee performed the nuclear safety capability assessment (NSCA) on a fire area basis and also identified the individual fire zones within the fire area. The licensee documented the following: SE Table 3.5-1 identifies those fire areas that were analyzed using either the deterministic or PB approach in accordance with NFPA 805 Chapter 4 based on the information provided in LAR Attachment C, Table B-3, "Fire Area Transition". Table 3.5-1 NMP1 Fire Areas and Compliance Strategy NFPA805 Fire Area Area Description Compliance Basis 1 Reactor Building East EL 198-0 thru EL 340-0 Performance-Based 2 Reactor Building West EL 198-0 thru EL 340-0 Performance-Based 3 Drywell EL 237-0 thru EL 318-0 Deterministic 4 Foam Room EL 261-0 Performance-Based 5 Turbine building EL 261-0 thru 369-0 Performance-Based 6 Turbine Building North EL 250-0 Performance-Based 7 Turbine Building South and West EL 250-0 Performance-Based 9 Turbine Building East EL 250-0 Performance-Based 10 Cable SQ_reading Room EL 250-0 Performance-Based 11 Control Complex EL 261-0 and 277-0 Performance-Based 12 Administration Building EL 250-0 Performance-Based 13 Screen house Performance-Based 14 Diesel Fire Pump Room EL 261-0 Performance-Based

-94-Fire NFPA805 Area Area Description Compliance Basis 15 Radwaste and Waste Disposal Buildings EL 252-0 thru Performance-Based 292-0 16A Battery Board Room 12 EL 261-0 Performance-Based 16B Battery Board Room 11 EL 261-0 Performance-Based 17A Battery Room 12 EL 277-291 Performance-Based 17B Battery Room 11 EL 277-291 Performance-Based 18 Emergency Diesel Generator 1 02 Missile Enclosure EL Performance-Based 271 19 Emergency Diesel Generator Room 1 03 Foundation Performance-Based Room EL 250-0 and Diesel Generator Room EL 261-0 20 Emergency Diesel Generator Enclosed Cableway EL Performance-Based 250-0 21 Area Below Powerboards 1 02 and 1 03 EL 250-0 Performance-Based 22 Emergency Diesel Generator Room 1 02 Foundation Performance-Based Room EL 250-0 and Diesel Generator Room EL 261-0 23 Powerboard 102 Room EL 261-0 Performance-Based 24 Powerboard 1 03 Room EL 261-0 Performance-Based EXT Exterior Area Deterministic LAR Attachment C provides the results of these analyses on a fire area basis. For each fire area, the licensee documented the following:

The approach used in accordance with NFPA 805 (i.e., the deterministic approach in accordance with NFPA 805, Section 4.2.3, or the PB approach in accordance with NFPA 805, Section 4.2.4).

The SSCs required to meet the NSPC.

Fire detection and suppression systems required to meet the NSPC.

An evaluation of the effects of fire suppression activities on the ability to achieve the NSPC.

The disposition of each VFDRs using either; modifications (completed or planned) or the performance of a FREin accordance with NFPA 805, Section 4.2.4.2. 3.5.1.1 Fire Detection and Suppression Systems Required to Meet the NSPC A primary purpose of NFPA 805, Chapter 4 is to determine, by analysis, what fire protection features and systems need to be credited to meet the NSPC. Four sections of NFPA 805 Chapter 3 have requirements dependent upon the results of the engineering analyses performed in accordance with NFPA 805, Chapter 4: (1) fire detection systems, in accordance with Section 3.8.2; (2) automatic water-based fire suppression systems, in accordance with Section 3.9.1; (3) gaseous fire suppression systems, in accordance with Section 3.1 0.1; and

-95-(4) passive fire protection features, in accordance with Section 3.11. The features/systems addressed in these sections are only required when the analyses performed in accordance with NFPA 805, Chapter 4 indicate the features and systems are required to meet the NSPC. The licensee performed a detailed analysis of fire protection features and identified the fire suppression and detection systems required to meet the NSPC for each fire area. LAR Table 4-3, "Summary of NFPA 805 Chapter 4 Required Fire Protection Systems and Features," lists the fire areas at NMP1, and identifies if the fire suppression and detection systems installed in these areas are required to meet criteria for separation, DID, risk, licensing actions (i.e., exemptions, deviations, safety evaluations), or EEEEs. The NRC staff reviewed LAR Attachment C for each fire area to ensure fire detection and suppression met the principles of DID in regard to the planned transition to NFPA 805. In a letter dated January 3, 2013 (Reference 15), in SSA RAI 09, the NRC staff requested additional information from the licensee to describe the methodology that was used to evaluate DID and the methodology that was used to evaluate safety margins. The description should include what was evaluated, how the evaluations were performed, and what, if any, actions or changes to the plant or procedures were taken to maintain the philosophy of DID or sufficient safety margins. In a letter dated February 27, 2013 (Reference 7), the licensee responded to SSA RAI 09 and stated the evaluation of DID involved an investigation of the three DID echelons, as follows:

Echelon 1: Preventing fires from starting. This echelon covers administrative control of combustibles and hot work activities.

Echelon 2: Rapidly detecting fires and controlling and extinguishing promptly those fires that do occur, thereby limiting fire damage. This echelon covers installed automatic fire detection and suppression systems. Also covered under this echelon is manual fire suppression, including fire preplans.

Echelon 3: Providing an adequate level of fire protection for structure, systems, and components (SSCs) important to safety, so that a fire that is not promptly extinguished will not prevent essential plant safety functions from being performed. This echelon covers features such as: Walls, floors, ceilings, and structural elements; Openings in the fire area barrier; Other features such as supplemental barriers; and Guidance to operation personnel detailing the required success paths. DID was evaluated using insights from the fire PRA. Features credited in the fire PRA were not credited for Dl D. Administrative controls on combustibles and hot work activities (Echelon 1) were credited in the fire PRA and did not require DID improvements. The installed fire

-96-protection features (Echelon 2) that were not credited in the fire PRA were credited for DID. Fire scenario frequencies and CCDPs were used to provide insights on Echelon 3. consequence fire scenarios (fire scenarios with a frequency of 1 E-06/reactor-yr or greater and a CCDP of 0.1 or greater) were identified for the DID assessment of Echelon 3. The DID evaluation found that an adequate balance between DID echelons was maintained and no improvement was needed. Safety Margin was evaluated based on considerations regarding:

Fire Modeling: Consistent with NUREG/CR-6850 (References 34, 35, and 36), a graded approach was used for fire modeling in the fire PRA; i.e., fire zones were evaluated on a case-by-case basis during the fire PRA development process to determine the level of detailed fire modeling analysis necessary. The highest level of fire modeling analysis was applied to selected zones on an as-needed basis when screening techniques can mask true risk contributors and practical/useful insights. Zones requiring less detailed analysis were modeled in the fire PRA with conservative methods used to account for uncertainties (e.g., by conservative estimation of the time to target damage, time to fire detection and suppression, time to generating hot gas layer, or by failing all targets in the fire zone for every ignition source).

Plant System Performance: Plant system performance was evaluated using design basis evaluations documented in the NMP1 UFSAR (Reference 83). In addition, analyses specific to the fire PRA were used to evaluate plant system performance in accordance with the ASME/ANS RA-Sa-2009 standard (Reference 29) and RG 1.200, Rev. 2 (Reference 28).

PRA Logic Model: The PRA logic model was developed in accordance with the ASME/ANS RA-Sa-2009 standard and implemented with the FRANX software. Each fire scenario to be quantified was assigned an initiating event of the internal events PRA selected so as to cover all the sequences potentially associated with the fire scenario. Fire-induced failures of equipment were quantified by setting the associated basic events to failure. Circuit failure probabilities were selected following conservative industry practice and guidance. Human failure events and associated human error probabilities were evaluated by taking into account the effects of the fire. The safety margin evaluation found that the safety margin inherent to the fire PRA model was preserved. Based on the statements provided in LAR Attachment C, as supplemented, the NRC staff concludes that the NMP1 treatment of this issue is acceptable because the licensee demonstrated the use of appropriate methodology to evaluate DID and safety margins and also identified the fire detection and suppression systems required to meet the NFPA 805 NSPC on a fire area basis.

-97-3.5.1.2 Evaluation of Fire Suppression Effects on NSPC Each fire area of LAR Attachment C includes a discussion of how the licensee met the requirement to evaluate the fire suppression effects on the ability to meet the NSPC. In a letter dated January 3, 2013 (Reference 15), in SSA RAI 04, the NRC staff requested additional information from the licensee since the LAR Attachment C discussion of fire suppression effects only addresses installed systems and does not address the potential effects of manual suppression activities by the fire brigade. The NRC staff asked the licensee to provide additional information on the effects of manual suppression activities on NSPC. In a letter dated February 27, 2013 (Reference 7), the licensee responded to SSA RAI 04 and stated manual firefighting activities by the plant fire brigade were considered within the scope of the fire suppression effects analysis evaluation, but were effectively screened out from detailed review since dedicated fire brigade members follow pre-fire plans and procedures. Fire brigade members are trained and drilled in actual plant scenarios. Pre-fire plans have been developed for all areas of the plant important to safety to inform the fire brigade members of access, suppression, detection, major equipment, and special hazards associated with each area. Major equipment for each area is identified to orient the user so that the fire brigade may attempt to limit damage to valuable plant equipment by judicious use of fire suppression equipment such as hose nozzle streams and fire extinguishers. Since most plant areas containing significant fire hazards are provided with fixed automatic or manual fire suppression systems, the effects of manual firefighting activities by the fire brigade are expected to be limited and localized. In a letter dated January 3, 2013 (Reference 15), in SSA RAI 05, the NRC staff requested additional information from the licensee regarding fire area EXT. As stated in the LAR, fire area EXT is "External to Plant" and is included in the power block (LAR Attachment I) under "yard". The NFPA 805 compliance strategy for fire area EXT is the deterministic approach per NFPA 805 Section 4.2.3.1. The NRC staff requested that the licensee describe how the performance criteria are being met and how the fire suppression effects meet the NSPC in this fire area. The NRC staff also requested that the licensee update the fire area assessment in LAR Attachment C accordingly. In a letter dated February 27, 2013 (Reference 7), the licensee responded to SSA RAI 05 and stated as documented in LAR Attachment C, Fire Area EXT is defined as external to plant (i.e., the yard area). Circuit analysis and fire area assessments performed by the licensee as part of NFPA 805 transition have concluded that a postulated fire in fire area EXT does not affect any of the components identified in the plant Safe Shutdown Equipment List (SSEL). The yard area fire suppression effects assessment contains the oil filled transformers which are located in the southwest corner of the turbine building. The transformers are each protected by an automatically-initiated water spray system. The turbine building wall at this location is 8-inch precast concrete up to elevation 285'. The metal wall above elevation 285' is provided with a manually initiated water curtain system. Runoff from suppression system actuation is initially contained with curbs and a basin around the transformer area and is then directed to the waste water treatment facility. Any actuation of the fire suppression system will not affect the safe shutdown equipment located in the adjacent turbine building or overall plant safe shutdown capability. The additional information requested by the NRC staff with respect to performance criteria and fire suppression effects in the yard area were added to LAR Attachment C under Fire Area EXT by the licensee.

-98-The licensee stated in LAR Attachment C that damage to plant areas and equipment from the accumulation of water discharged from manual and automatic fire protection systems and the discharge of manual suppression water to adjacent compartments is controlled. Therefore, fire suppression activities will not adversely affect the plant's ability to achieve the NSPC. Based on the information provided by the licensee in the LAR Attachment C, as supplemented, the licensee has evaluated fire suppression effects on meeting the NSPC and determined that fire suppression activities will not adversely affect achievement of the NSPC. The NRC staff concludes that the licensee's evaluation of the suppression effects on the NSPC is acceptable because the licensee has demonstrated that fire suppression activities will not adversely affect achievement of the NSPC. 3.5.1.3 Licensing Actions Based on the information provided by the licensee in the LAR section 4.2.3 and LAR Attachment K, the NRC staff concludes that there are no licensing actions that are required to be transitioned to NFPA 805. 3.5.1.4 Existing Engineering Equivalency Evaluations (EEEEs) The EEEEs that support compliance with NFPA 805 Chapter 4 were reviewed by the licensee using the methodology contained in NEI 04-02. The methodology for performing the EEEE review included the following determinations:

The EEEE is not based solely on quantitative risk evaluations,

The EEEE is an appropriate use of an engineering equivalency evaluation,

The EEEE is of appropriate quality,

The standard license condition is met,

The EEEE is technically adequate,

The EEEE reflects the plant as-built condition, and

The basis for acceptability of the EEEE remains valid. In LAR section 4.2.2, the licensee stated the guidance in RG 1.205 (Reference 2), Regulatory Position 2.3.2, as clarified by FAQ 08-0054 (Reference 56) was followed. EEEEs that demonstrate that a fire protection system or feature is "adequate for the hazard" are to be addressed in the LAR as follows:

If not requesting specific approval for "adequate for the hazard" EEEEs, then the EEEE was referenced where required and a brief description of the evaluated condition was provided.

-99-* If requesting specific NRC approval for "adequate for the hazard" EEEEs, then the EEEE was referenced where required to demonstrate compliance and was included in LAR Attachment L for NRC review and approval. The licensee identified and summarized the EEEEs for each fire area in LAR Attachment C, as applicable. The licensee did not request the NRC staff to review and approve any of these EEEEs. Based on the NRC staff's review of the licensee's methodology for review of EEEE's and identification of the applicable EEEEs in LAR Attachment C, the NRC staff's review concludes that the use of EEEEs meets the requirements of NFPA 805, the guidance of RG 1.205 (Reference 2) and FAQ 08-0054 (Reference 56), and is acceptable. 3.5.1.5 Variances from Deterministic Requirements For those fire areas where deterministic criteria were not met, VFDRs were identified and evaluated using PB methods. VFDR identification, characterization, and resolutions were identified and summarized in LAR Attachment C for each fire area. Documented variances were all represented as separation issues. The following strategies were used by the licensee in resolving the VFDRs:

An FRE determined that applicable risk, DID, and safety margin criteria were satisfied without further action.

An FRE determined that applicable risk, DID, and safety margin criteria were satisfied with a credited RA.

A FRE determined that applicable risk, DID, and safety margin criteria were satisfied with a plant modification(s), as identified in the LAR, as supplemented. For all fire areas where the licensee utilized the PB approach to meet the NSPC, each VFDR and the associated disposition has been described in LAR Attachment C. Based on the NRC staff review of the VFDRs and associated resolutions as described in LAR Attachment C, as supplemented, the NRC staff concludes that the licensee's identification and resolution of the VFDRs is acceptable because the licensee identified, characterized, and resolved all VFDRs as summarized in LAR Attachment C for each fire area. 3.5.1.6 Recovery Actions LAR Attachment G lists the recovery actions identified in the resolution of VFDRs in LAR Attachment C for each fire area. The recovery actions (RA) identified include both actions considered necessary to meet risk acceptance criteria as well as actions relied upon as 01 D (see SE Section 3.5.1.7 below). The NRC staff reviewed LAR Section 4.2.1.3, "Establishing Recovery Actions," and LAR Attachment G, "Recovery Actions Transition," and Implementation Item 19 described in LAR AttachmentS, Table S-2, revised version of recovery action procedures N1-SOP-21.1, "Fire in

-100-Plant" and N1-SOP-21.2, "Control Room Evacuation", to evaluate whether the licensee meets the associated requirements for the use of RAs per NFPA 805. The details of the NRC staff's review for RAs are described in Section 3.2.5 of this SE. The NRC staff's evaluation of the additional risk of RAs credited to meet the risk acceptance guidelines is provided in Section 3.4.4 of this SE. 3.5.1.7 Recovery Actions Credited for Defense-in-Depth The process documented in the LAR used by the licensee to establish recovery actions required to meet the nuclear safety performance criteria did not utilize a category of "Recovery Actions Credited for Defense in Depth." The. licensee followed the process in RG 1.205, Revision 1, and NEI 04-02, Revision 2 as supplemented by FAQ 07-0030 (Reference 51). In this process, the licensee determined the population of recovery actions required to resolve VFDRs (to meet the risk acceptance criteria or maintain a sufficient level of defense in depth). Once a recovery action was identified as required, all recovery actions were treated in the same way and always referred to as a recovery action (regardless of whether the action was needed to meet the risk criteria or to provide DID). The NRC staff reviewed LAR Section 4.2.1.3, "Establishing Recovery Actions," and LAR Attachment G, "Recovery Actions Transition," to evaluate whether the licensee meets the associated requirements for the use of recovery actions per NFPA 805. The NRC staff's evaluation of the licensee's process for identifying recovery actions and assessing their feasibility is provided in SE Section 3.2.5, "Establishing Recovery Actions." 3.5.1.8 Plant Fire Barriers and Separations With the exception of ERFBS, passive fire protection features include the fire barriers used to form fire area boundaries (and barriers separating safe shutdown trains) that were established in accordance with the plant's pre-NFPA 805 deterministic FPP. For the transition to NFPA 805, the licensee decided to retain the previously established fire area boundaries as part of the RI/PB FPP. Fire area boundaries are established for those areas described in LAR Attachment C, as modified by applicable EEEEs that determine the barriers are adequate for the hazard or otherwise disposition differences in barrier design and performance from applicable criteria. The acceptability of fire barriers and separations is also evaluated as part of the NRC staff's review of LAR Attachment A, Table B-1 and as such are addressed in SE Section 3.1. 3.5.1.9 Electrical Raceway Fire Barrier Systems (ERFBS) The licensee stated that the ERFBS used at NMP1 meets the deterministic requirements of NFPA 805, Chapter 3. The fire area using ERFBS is identified in LAR Attachment C. In this fire area, the ERFBS met the requirements of NFPA 805, Section 4.2.3. In a letter dated January 3, 2013 (Reference 15), in SSA RAI 02, the NRC staff requested additional information from the licensee regarding the ERFBS discussion in the LAR. LAR Attachment A stated that an ERFBS (Eternit, Inc. Promat-H) is credited in fire area 18,

-101 -"Emergency Diesel Generator 102 Missile Enclosure," only (page A-66). Based on the information provided in the LAR, the ERFBS was not tested in accordance with GL 86-10, Supplement 1 (Reference 67). The acceptability of ERFBS testing and adequacy for the hazard was documented in EEEE FPEE 95-002; however, the ERFBS was not listed as a credited feature in LAR Table 4-3 or the fire area assessment in LAR Attachment C. The NRC staff pointed out that the ERFBS should be identified as a credited fire protection feature in Table 4-3 and the fire area assessment for fire area 18. In a letter dated February 27, 2013 (Reference 7), the licensee responded to SSA RAI 02 and stated the Eternit, Inc. Promat-H enclosure in fire area 18 was identified as a credited fire barrier in the NSCA for protection of the EDG 1 03 cooling water pump cable. It is considered to be a fire barrier that maintains cable 171-66 free of fire damage in fire area 18. LAR Table 4-3 and LAR Attachment C were revised to indicate that for fire area 18 (EDG 102 Missile Enclosure El 271 ), an ERFBS is a Required Fire Protection Feature (Code "S"-Required for Separation) to meet the requirements of NFPA 805. The NRC staff concludes that the licensees response to the RAI is acceptable because the installed ERFBS configuration has been evaluated by the licensee to be adequate for the hazard in that it meets the deterministic requirements of NFPA 805 Section 4.2.3 and also meets the fundamental elements and design requirements of NFPA 805 Chapter 3, Section 3.11.5, which are described in LAR Attachment A. 3.5.1.1 0 Conclusion for Section 3.5.1 As documented in LAR Attachment C, for those fire areas that used a deterministic approach in accordance with NFPA 805, Section 4.2.3, the NRC staff concludes that each of the fire areas analyzed using the deterministic approach meet the associated criteria of NFPA 805, Section 4.2.3. This conclusion is based on:

The licensee's documented compliance with NFPA 805, Section 4.2.3;

The licensee's assertion that the success path will be free of fire damage without reliance on recovery actions;

An assessment that the suppression systems in the fire area will have no impact on the ability to meet the NSPC; and

The licensee's appropriate determination of the automatic fire suppression and detection systems required to meet the NSPC. For those fire areas that used the PB approach in accordance with NFPA 805, Section 4.2.4, as described in LAR Attachment C, the NRC staff concludes that each fire area has been properly analyzed, and compliance with the NFPA 805 requirements demonstrated as follows:

Deviations from the pre-NFPA 805 fire protection licensing basis that were transitioned to the NFPA 805 licensing basis were reviewed for applicability, as well as continued validity, and found acceptable;

-102-* VFDRs were evaluated and either found to be acceptable based on an integrated assessment of risk, defense-in-depth, and safety margins, or modifications or recovery actions were identified and actions planned or implemented to address the issue (see Sections 3.4.1 and 3.4.6 of this SE);

Implementation items address the modifications and other actions as applicable;

Recovery actions used to demonstrate the availability of a success path to achieve the nuclear safety performance criteria were evaluated and the additional risk of their use determined, reported, and found to be acceptable (see Sections 3.4.1 and 3.4.6 of this SE);

The licensee's analysis appropriately identified the fire protection SSCs required to meet the nuclear safety performance criteria, including fire suppression and detection systems, as well as required fire protection features;

Fire area boundaries (ceilings, walls, and floors), such as fire barriers, fire barrier penetrations, and through penetration fire stops have been established; and

ERFBS credited were documented on a fire area basis, verified to meet acceptance criteria of NFPA 805. Accordingly, each fire area utilizing the PB approach was able to achieve and maintain the NSPC, and the associated FREs meet the applicable NFPA 805 requirements for risk, DID, and safety margin. 3.5.2 Clarification of Prior NRC Approvals As stated in LAR Attachment T, there are no elements of the current FPP for which NRC clarification is needed. 3.5.3 Fire Protection during Non-Power Operational Modes NFPA 805, Section 1.1 "Scope," states the following: This standard specifies the minimum fire protection requirements for existing light water nuclear power plants during all phases of plant operation, including shutdown, degraded conditions, and decommissioning. NFPA 805, Section 1.3.1, "Nuclear Safety Goal," states the following: The nuclear safety goal is to provide reasonable assurance that a fire during any operational mode and plant configuration will not prevent the plant from achieving and maintaining the fuel in a safe and stable condition. The NRC staff reviewed LAR Section 4.3, "Non-Power Operational Modes" and LAR Attachment D, "NEI 04-02 Non-Power Operational Modes Transition," to evaluate the licensee's

-103-treatment of potential fire impacts during non-power operations (NPOs). The NRC staff concludes that the licensee followed the guidance used in the process described in NEI 04-02, (Reference 5) as modified by FAQ 07-0040 (Reference 55), for demonstrating that the NSPC are met for higher risk evolutions (HREs) during NPO modes. 3.5.3.1 NPO Strategy and Plant Operational States (POSs) In LAR Section 4.3 and LAR Attachment D, the licensee stated that the process used to demonstrate that the NSPC are met during NPO modes is consistent with the guidance contained in FAQ 07-0040 (Reference 55). As described in LAR Attachment D, the licensee's procedure NIP-OUT-01, "Shutdown Safety," outlines use of the "defense-in-depth concept to minimize shutdown risk and maximize the availability of critical components and station systems that ensure nuclear safety during shutdown conditions." The guideline contains specific actions to address reduced inventory conditions that consider short time to boil, limited methods for decay heat removal, and low RCS inventory. As described in the LAR Attachment D, the licensee identified equipment and cables necessary to support the key safety functions (KSFs) success paths. The operational modes and functional requirements for the systems and components were reviewed. The KSF success path equipment and cables were incorporated in the NPO database model. Following identification of KSF equipment and cables, the licensee performed analysis on a fire zone basis to identify areas where redundant equipment and cables credited for a given KSF might fail due to fire damage (i.e., pinch-points). The licensee used a deterministic approach to identify these pinch-points and mitigated these pinch-points through the use of RAs and/or fire prevention/protection controls. As stated in LAR Attachment D, fire modeling was not used to determine if a postulated fire would be expected to damage required equipment. As described in LAR Attachment D, the licensee's procedure NIP-OUT-01, "Shutdown Safety," defines HREs and establishes KSFs and DID strategies to protect the KSFs. HREs are "outage activities, plant configurations, or conditions during shutdown where the plant is more susceptible to an event causing the loss of a KSF function or the number of key safety systems drops below the shutdown safety criteria." The procedure contains specific actions to assess decay heat removal, inventory control, power available, and reactivity control. LAR Section 4.3 and LAR Attachment D describe the methods and results of the NPO evaluation, including references to the applicable outage programs, procedures, and NPO analyses. In a letter dated January 3, 2013 (Reference 15), in SSA RAI 03, the NRC staff requested that the licensee provide a summary level identification of unavailable paths in each fire area (pinch points) and the resolution for each pinch point. In a letter dated April 30, 2013 (Reference 9), the licensee responded to SSA RAI 03.a and provided summary level identification of KSF losses and pinch points on a fire zone basis. The table provided in the RAI response, identifies each KSF associated with a pinch point and the recommendations for addressing the pinch points. The licensee stated as described in the LAR, Section 4.3 and LAR Attachment D, the following KSFs are evaluated in each fire zone:

-104-* Decay Heat Removal (DHR) for both the Reactor Vessel (RX) and the Spent Fuel Pool (SFP);

Inventory Control (I NV) for both the Reactor Vessel and the Spent Fuel Pool; and

Power (PWR) availability. The Reactivity Control KSF is not included in the NPO analysis because it is administratively controlled in accordance with procedure NIP-OUT-01, "Shutdown Safety". Referring to Table SSD/CA RAI 03-2, the KSFs are categorized with codes assigned to each KSF -Fire Zone pair. Three codes have been established to summarize the fire impacts: * "I" (Impacted): At least one of the KSF paths associated with a given KSF is affected; i.e., a component of a specific KSF path or any of the component's required cables within the fire zone are impacted, whereby that path can no longer be assured of being functional. However, at least one other KSF path for the KSF remains available; * "L" (Lost): All available success paths for a given KSF are impacted; and * "N" (None): No impacts to the KSF are identified. "Pinch Points" are then identified on a fire zone basis, based on the loss of a KSF. An "N" in the pinch point column of Table SSD/CA RAI 03-2 indicates that no KSFs are lost in this fire zone. A "Y" in this column indicates that one or more KSFs are potentially lost in the fire zone, and therefore a pinch point is considered to exist. Fire zones are then categorized as follows:

Category 1 fire zones are not pinch points as they were found to have at least one success path for each KSF. No recommendations for additional fire protection measures during HREs are made for these zones. Standard DID strategies, as specified by procedure NIP-OUT-01, "Shutdown Safety," are adequate to address risk.

Category 2 fire zones are pinch points as every success path is potentially lost for at least one KSF. These KSF success paths can be preserved through fire protection/fire prevention actions, including the verification of functionality of available fire detection and suppression during HREs. The licensee stated the guidance in FAQ 07-0040 provided a listing of standard fire risk management methods that have been found to be acceptable for managing fire risk during HREs. During periods of NPO that are not defined as HREs, the standard fire protection DID actions are considered sufficient to minimize fire risk. During HREs, recommendations from FAQ 07-0040 have been identified for additional measures to consider as part of a comprehensive program to reduce fire risk. Each Category 2 fire zone includes one or more recommendations from the list provided in Table SSD/CA RAI 03-3 to minimize fire risk to the KSFs, as described in Table SSD/CA RAI 03-2.

-105-In a letter dated January 3, 2013 (Reference 15), in SSA RAI 03.b, the NRC staff requested additional information from the licensee with regard to spurious operations during NPO modes. The NRC staff pointed out that spurious actuation of valves can have a significant impact on the ability to maintain decay heat removal and inventory control. The NRC staff requested that the licensee provide a description of any actions being credited to minimize the impact of induced spurious actuations on power operated valves (e.g., air operated valves (AOVs) and motor operated valves (MOVs)) during NPO either as prefire conditioning or as required during the fire response recovery (e.g., pre-fire rack-out, locally pinning of valves, and isolation of air supplies). For example, it appears to the NRC staff that the TSs allow the shutdown cooling isolation valves 38-01 and 38-13 to be inoperable in the open position for greater than 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months under certain specific conditions and that during HREs, such as a short time to boil, preventing the spurious closure of any of these valves would be advantageous. The NRC staff requested that the licensee provide justification for not invoking the TS allowed flexibility for maintaining these valves open during HREs. In a letter dated April30, 2013 (Reference 9}, the licensee responded to SSA RAI 03.b, and stated the NMP1 NPO pinch point analysis was developed in accordance with the guidance contained in FAQ 07-0040 (Reference 55). The FAQ 07-0040 recommendations utilized at NMP1 to reduce fire risk during HREs are identified in part 3a of this RAI response. The additional reduction in risk offered by the recommended strategies provides additional assurance that fire risk is minimized in areas susceptible to a loss of one or more KSFs during plant HREs. The licensee's response to SSA RAI 3a and depicted in Table SSD/CA RAI 03-2 states that additional actions (e.g., pre-fire rack-out, locally pinning of valves, isolation of air supplies) are not relied upon as a strategy to reduce fire risk during HREs, including the impact of fire-induced spurious operations (single or multiple). The assessment of potential risk reduction options (including input from Operations personnel) concluded that the actual additional risk posed by fire during HREs is best controlled through the methods identified in Table SSD/CA RAI 03-2. Specifically, the NMP1 NPO strategy does not credit the following methods:

Recovery actions -Reliance on RAs during an outage is difficult to characterize for feasibility due to the many variables that could exist, such as blockage of normal routes, scaffolding impact on lighting, equipmenVmaterial staging and movement, contract personnel contingent, unusual equipment line ups, etc. For this reason, RAs are viewed as less predictable with respect to reliability and uncertainty in comparison to the risk reduction options selected.

Configuration Changes -The use of limited configuration changes to address in a preemptive manner certain high consequence fire-induced failures, most notably spurious operations of key valves, was considered. However, after discussions with Operations personnel it was concluded that the reduction in operational flexibility to respond to a broader range of potential accidents and abnormal conditions outweighs the marginal improvement in risk reduction associated with fire-induced spurious operations.

-106-With specific reference to the potential vulnerability of shutdown cooling isolation valves 38-01, 38-02, and 38-13 to fire-induced spurious closure, deliberately entering a TS required action was evaluated as undesirable when viewed from a broader perspective beyond just potential fire events. Thus, the recommendations contained in Table SSD/CA RAI 03-2 are considered the best options to augment existing procedures for managing shutdown risk, including risk from fire, during HREs. In a letter dated January 3, 2013 (Reference 15), in SSA RAI 03.c, the NRC staff requested that the licensee identify locations where KSFs are achieved via RAs or for which instrumentation not already included in the at-power analysis is needed to support recovery actions required to maintain safe and stable conditions. The NRC staff requested that the licensee identify those recovery actions and instrumentation relied upon in NPO, describe how recovery action feasibility is evaluated, and include in the description whether these variables have been or will be factored into operator procedures supporting these actions. For instance, during outage conditions when there is a short time to boil, describe the operator response to a spurious closure of one of the shutdown cooling system motor operated isolation valves 38-01 or 38-13. The NRC staff also requested that the licensee describe how any recovery actions are feasible (e.g., can be reliably accomplished in the available time frame). In a letter dated April30, 2013 (Reference 9), the licensee responded to SSA RAI 03.c, and stated that NMP1 does not credit recovery actions as a strategy to reduce shutdown fire risk during HREs. The NRC staff concludes that the NPO process described and documented by the licensee in LAR Section 4.3 and LAR Attachment Dis consistent with FAQ 07-0040 and is acceptable. 3.5.3.2 NPO Analysis Process The licensee stated that its goal is to ensure that contingency plans are established when the plant is in an HRE and it is possible to lose a KSF due to fire. LAR Section 4.3 discusses these additional controls and measures. However, during low risk periods, normal risk management controls; as well as fire prevention/protection processes and procedures will be utilized. Consistent with the guidance in FAQ 07-0040, the process to demonstrate that the nuclear safety performance criteria are met during NPO modes involved the following steps as described in LAR sections 4.3.1 and 4.3.2, and depicted in LAR Figures 4-5 and 4-6:

Review of the existing Outage Management Processes.

Identification of Equipment/Cables: Review of plant systems to determine success paths that support each of the DID KSFs; and Identification of cables required for the selected components and determination of their routing.

Perform Fire Area Assessments (identify pinch points -plant locations where a single fire may damage all success paths of a KSF).

-107-* Manage pinch-points associated with fire-induced vulnerabilities during the outage. As described in the LAR and in accordance with the guidance in FAQ 07-0040, the POSs were considered for equipment and cable selection and provided the component selection information for HSD, CSD, Refueling and Defueled conditions. KSFs were evaluated by the licensee in each fire zone: DHR for both the Reactor Vessel and the Spent Fuel Pool (SFP), Inventory Control for both the Reactor Vessel and the SFP, and Power availability. Each may have one or more KSF paths that satisfy that specific KSF. In the LAR Section 4.3.2, the licensee stated no effort was made to eliminate or reduce fire impact by circuit analysis; therefore, a conservative estimate of damage is provided, including spurious operation of equipment. By assuming that a single fire impacts any and all components in a fire zone, (whether the individual component or its associated cables are physically located within the fire zone), the assumption is made that the entire contents of the fire zone are lost. The licensee also states that if a component that is part of a particular KSF flow path is impacted, it is assumed that the KSF path is lost. However, there may be one or more other flow paths within the particular KSF that are not impacted; therefore, the KSF is not considered lost and does not constitute a pinch point. Only when all paths for a particular KSF are impacted, is the KSF itself considered lost and is a pinch point. The NRC staff concludes that the licensee's process to identify NPO systems, components, and cables, as described in LAR Section 4.3.2 and LAR Attachment D is consistent with the guidance in FAQ 07-0040 and is acceptable. The NRC staff also concludes that NPO systems, components and cables logically related to KSFs are identified in the NPO analysis database. 3.5.3.3 NPO Key Safety Functions and SSCs Used to Achieve Performance The licensee stated in the LAR that the guidance in FAQ 07-0040 was followed to perform fire area assessments to identify areas where fires may cause damage to the credited equipment or where KSFs are achieved solely by crediting recovery actions. LAR Attachment D defines the KSFs, the success paths to achieve the KSFs, and the components required for the success paths. The licensee further stated that each of these areas will be identified through administrative procedures governing fire protection DID features, shutdown risk management, and work control. In the licensee's analysis, pinch points refer to a particular location in an area where the damage from a single fire scenario could result in failure of multiple components or trains of a system such that the maximum detriment on that system's performance would be realized from the single fire scenario. Typically, this involves close vertical proximity of cables which support redundant components or trains of a system such that all such cables can be damaged by just one fire scenario.

-108-Based on its review of the information provided in the LAR, as supplemented, the NRC staff concludes that the licensee used methods consistent with the guidance in FAQ 07-0040, and RG 1.205 to identify the equipment required to achieve and maintain the fuel in a safe and stable condition during NPO modes. Further, the licensee has a process in place to ensure that fire protection DID measures will be implemented to achieve the KSFs during plant outages. These implementation tasks are reflected in LAR Attachments D and S. On the basis of the NPO analysis as described in the LAR, as supplemented, the NRC staff concludes the licensees' methods to perform NPO fire area assessments as described in the LAR, is acceptable. 3.5.3.4 NPO Pinch Point Resolutions and Program Implementation The licensee identified power operated components needed to support an NPO KSF. All were included in the post-fire safe shutdown equipment list. Those newly identified power operared components required additional circuit analysis. The pinch point analysis and KSFs are discussed in the above paragraphs for NPO where the NRC staff concluded the licensee had followed the guidance provided in FAQ 07-0040. In a letter dated January 3, 2013 (Reference 15) in SSA RAI 03, the NRC staff requested that the licensee describe the methods and results of the NPO Modes evaluation, including references to the applicable outage programs, procedures, and NPO analysis. Accordingly, based on the information provided in the LAR, as supplemented, the NRC staff concludes that the licensee demonstrated that the nuclear safety performance criteria are met during NPO modes and HREs. 3.5.4 Conclusion for Section 3.5 The NRC staff reviewed the licensee's RI/PB FPP, as described in the LAR and its supplements, to evaluate the NSCA results. The licensee used a combination of the deterministic approach and the PB approach, in accordance with NFPA 805, Sections 4.2.3 and 4.2.4. For those fire areas where the licensee used a deterministic approach, the NRC staff verified the following:

The engineering evaluations for exemptions from the existing FPP were evaluated and found to be valid and acceptable for meeting the deterministic requirements of NFPA 805, as allowed by NFPA 805, Section 2.2.7;

Fire suppression effects were evaluated and found to have no adverse impact on the ability to achieve and maintain the NSPC for each fire area; and

The required automatic fire suppression and automatic fire detection systems were appropriately documented for each fire area. Accordingly, the NRC staff concludes that each fire area utilizing the deterministic approach does so in accordance with NFPA 805, Section 4.2.3.

-109-For those fire areas where the licensee used a PB approach, the NRC staff verified the following:

The engineering evaluations for exemptions from the existing FPP were evaluated and found to be valid and acceptable for meeting the requirements of NFPA 805, Section 2.2.7;

Fire suppression effects were evaluated and found to have no adverse impact on the ability to achieve and maintain the NSPC for each fire area;

All VFDRs were evaluated using the FRE PB method (in accordance with NFPA 805, Section 4.2.4.2) to address risk impact, defense-in-depth, and safety margin, and were found to be acceptable;

All recovery actions necessary to demonstrate the availability of a success path were evaluated with respect to the additional risk presented by their use and found to be acceptable in accordance with NFPA 805, Section 4.2.4; and

The required automatic fire suppression and automatic fire detection systems were appropriately documented for each fire area. Based on the analyses performed by the licensee and described in the LAR, as supplemented, the NRC staff concludes that each fire area where the licensee uses the PB approach, in accordance with NFPA 805, Section 4.2.4, is able to achieve and maintain the NSPC. Further, there is reasonable assurance that the associated FREs meet the requirements for risk, DID and safety margin. The NRC staff concludes that the licensee's analysis and outage management process during NPO modes provides assurance that the nuclear safety performance criteria will be met during NPO modes and HREs, and that the licensee used methods consistent with the guidance provided in FAQ 07-0040 and RG 1.205. The NRC staff also concludes that no recovery actions are required during NPO modes and that the licensee's overall approach for fire protection during NPO modes is acceptable because the requirements for risk, DID, and safety margin are met. 3.6 Radioactive Release Performance Criteria NFPA 805, Chapter 1 defines the radioactive release goals, objectives, and performance criteria that must be met by the FPP in the event of a fire at a nuclear power plant in any plant operational mode. NFPA 805, Section 1.3.2, "Radioactive Release Goal," states that: The radioactive release goal is to provide reasonable assurance that a fire will not result in a radiological release that adversely affects the public, plant personnel, or the environment.

-110-NFPA 805, Section 1.4.2, "Radioactive Release Objective," states that: Either of the following objectives shall be met during all operational modes and plant configurations. (1) Containment integrity is capable of being maintained [such that fighting products are monitored and released within the plant's normal effluents program]. (2) The source term is capable of being limited [such that any unmonitored releases would not exceed the performance criteria]. NFPA 805, Section 1.5.2, "Radioactive Release Performance Criteria," states that: Radiation release to any unrestricted area due to the direct effects of fire suppression activities (but not involving fuel damage) shall be as low as reasonably achievable and shall not exceed applicable 1 0 CFR Part 20 limits. In order to assess whether the NMP1 FPP to be implemented under NFPA 805 meets the above requirements, the licensee performed a review of the current NMP1 FPP using the methodology contained in NEI 04-02 (Reference 5) and FAQ 09-0056 (Reference 57). Each fire zone was first screened to determine the potential for generating radioactive effluents during firefighting operations. The screening process considered input from NMP1 radiation protection personnel and evaluated the fire zone's potential for radioactive effluent release during all modes of operation. Fire zones where there is no possibility of radioactive materials being present (e.g., those outside of the radiologically controlled area) were screened from further review. For all other fire zones, engineering controls afforded by plant design features, fire pre-plans, and fire brigade training materials were reviewed to ascertain whether existing NMP1 FPP is adequate to ensure that radioactive materials (contamination) generated as a direct result of fire suppression activities are contained and monitored before release to unrestricted areas, such that the release would meet the NFPA 805 radioactive release performance criteria. LAR Attachment E provides a detailed summary, on a fire zone by fire zone basis, of the licensee's qualitative assessment. The licensee's review determined that the current FPP is compliant with the radiological release requirements of NFPA 805 and the guidance in RG 1.205 (Reference 2). The licensee's qualitative review determined that NMP1 buildings and structures provide sufficient capacity to contain the liquid and gaseous firefighting effluents such that there are no offsite releases. The reactor, turbine, radwaste solidification and storage, and waste disposal buildings all have doors (personnel and roll-up). However, these doors are maintained closed during power and power operations so they do not provide a radioactive release pathway. Opening of these doors is maintained under administrative controls, and the doors are only open for ingress and egress. The fire preplans have been revised to assure that manual actions are taken to prevent offsite releases in those fire areas where there is a potential for such effluent diversions. In general, station building ventilation systems are credited for the capture and monitoring of airborne products. In addition to the doors, the turbine building roof vents could provide an unmonitored release path for gaseous effluents. However, each of these building openings is closed during

-111 -power and non-power operation, and opening them is administratively controlled. The Turbine Building roof vents are remotely opened only at the direction of the Fire Brigade Leader in communication with the main control room staff and the radiation protection technician. If the need arose to open one of these openings during a firefighting event, any gaseous release would be monitored by the RP technician on the fire brigade team at the direction of the Fire Brigade Leader per the fire pre-plans. Subsequent release of these effluents will be within the TS limits as determined by the NMP1 effluent release program. The NRC staff concludes that consistent with the guidance in RG 1.205 and FAQ 09-0056, there is reasonable assurance that the annual dose limits of 10 CFR 20 are met if the concentrations of radioactive materials in airborne and liquid releases are maintained below the instantiations release limits in the NMP1 Technical Specifications. The NRC staff noted that there are existing administrative controls in place for storage of containers of radioactive materials or waste within station buildings. NMP1 does not allow the storage of radioactive material containers outside or in other areas where containment/confinement is not provided. The licensee also reviewed fire pre-plans and fire brigade training materials to ensure that fire brigades are sufficiently trained on the design features and administrative controls that are being credited to meet the radioactive release goals. LAR Attachment E identifies current training sufficient to ensure containment and confinement of radioactive effluents, and the monitoring of potential radioactive release pathways for both power and non-power operations. Based on (1) the information provided in the LAR, as supplemented, (2) the licensee's use of fire pre-plans, (3) the results of the NRC staff's evaluation of the identified engineered controls used to manage suppression water and combustion products, and (4) the development and implementation of newly revised fire brigade training procedures (LAR AttachmentS, Table S-2, Implementation Item 18), the NRC staff concludes that the licensee's RI/PB FPP provides reasonable assurance that radiation releases to any unrestricted area resulting from the direct effects of fire suppression activities at NMP1 are as low as reasonably achievable and are not expected to exceed the radiological dose limits in 10 CFR Part 20. In conclusion, the NRC staff finds that the licensee's RI/PB FPP complies with the requirements specified in NFPA 805, Sections 1.3.2, 1.4.2, and 1.5.2. Accordingly, the NRC staff finds this approach acceptable. 3.7 NFPA 805 Monitoring Program For this SE section, the following requirements from NFPA 805, Section 2.6, are applicable to the NRC staff's review of the licensee's LAR: NFPA 805, Section 2.6, "Monitoring": A monitoring program shall be established to ensure that the availability and reliability of the fire protection systems and features are maintained and to assess the performance of the fire protection program in meeting the performance criteria. Monitoring shall ensure that the assumptions in the engineering analysis remain valid.

-112-NFPA 805, Section 2.6.1, "Availability, Reliability, and Performance Levels": Acceptable levels of availability, reliability, and performance shall be established. NFPA 805, Section 2.6.2, "Monitoring Availability, Reliability, and Performance": Methods to monitor availability, reliability, and performance shall be established. The methods shall consider the plant operating experience and industry operating experience. NFPA 805, Section 2.6.3, "Corrective Action": If the established levels of availability, reliability, or performance are not met, appropriate corrective actions to return to the established levels shall be implemented. Monitoring shall be continued to ensure that the corrective actions are effective. The NRC staff reviewed LAR Section 4.6, "Monitoring Program." The NRC staff reviewed the monitoring program that the licensee developed to monitor availability, reliability, and performance of NMP1 FPP systems and features after the transition to NFPA 805. The focus of the NRC staff's review was on the critical elements related to the monitoring program, including the selection of FPP systems and features to be included in the program, the attributes of those systems and features that will be monitored, and the methods for monitoring those attributes. Implementation of the monitoring program will occur on the same schedule as the NFPA 805 RI/PB FPP implementation, which the NRC staff finds acceptable (see SE Section 2.7). The licensee stated that it will develop an NFPA monitoring program consistent with FAQ 10-0059 (Reference 58). The licensee further stated that development of the monitoring program will include a review of existing surveillance, inspection, testing, compensatory measures, and oversight processes for adequacy and that the review will examine adequacy of the scope of SSCs and components within the existing plant programs, performance criteria for availability and reliability of SSCs, and the adequacy of the plant corrective action program. The licensee also stated that the monitoring program will incorporate phases for scoping, screening using risk criteria, risk target value determination, and monitoring implementation. The licensee stated that the scope of the program will include fire protection systems and features, NSCA equipment, SSCs relied upon to meet radioactive release criteria, and fire protection programmatic elements. Based on the information provided in the LAR, as supplemented, the NRC staff concludes that the licensee's NFPA 805 monitoring program, and development and implementation process is acceptable and assures that NMP1 will implement an effective program for monitoring risk significant fires because it:

Establishes the appropriate SSCs to be monitored;

Uses an acceptable screening process for determining the SSCs to be included in the monitoring program;

-113-* Establishes availability, reliability and performance criteria for the SSCs being monitored; and

Requires corrective actions when sse availability, reliability, and performance criteria targets are exceeded in order to bring performance back within the required range. However, since the final values for availability and reliability, as well as the performance criteria for the SSCs being monitored, have not been established for the monitoring program as of the date of this SE, completion of the NMP1 NFPA 805 Monitoring Program is included as Implementation Items 9 and 10 in LAR Attachment S, Table S-2 and required by the license condition. The NRC staff concludes that completion of the monitoring program is acceptable because it will occur on the same schedule as the implementation of NFPA 805. 3. 7.1 Conclusion for Section 3. 7 The NRC staff reviewed the licensee's RI/PB FPP and RAI responses for SE Section 3.7. The NRC staff concludes that, upon completion of Implementation Items 9 and 10, that the licensee's monitoring program will meet the requirements specified in NFPA 805, Sections 2.6.1, 2.6.2 and 2.6.3. 3.8 Program Documentation, Configuration Control, and Quality Assurance For this SE section, the requirements from NFPA 805, Section 2.7, "Program Documentation, Configuration Control and Quality," are applicable to the NRC staff's review of the LARin regard to the appropriate content, configuration control, and quality of the documentation used to support the NMP1 FPP transition to NFPA 805. NFPA 805, Section 2.7.1, "Content": NFPA 805, Section 2.7.1.1, "General": The analyses performed to demonstrate compliance with this standard shall be documented for each nuclear power plant (NPP). The intent of the documentation is that the assumptions be clearly defined and that the results be easily understood, that results be clearly and consistently described, and that sufficient detail be provided to allow future review of the entire analyses. Documentation shall be maintained for the life of the plant and be organized carefully so that it can be checked for adequacy and accuracy either by an independent reviewer or by the AHJ.

-114 -NFPA 805, Section 2.7.1.2*, "Fire Protection Program Design Basis Document": A fire protection program design basis document shall be established based on those documents, analyses, engineering evaluations, calculations, and so forth that define the fire protection design basis for the plant. As a minimum, this document shall include fire hazards identification and nuclear safety capability assessment, on a fire area basis, for all fire areas that could affect the nuclear safety or radioactive release performance criteria defined in Chapter 1. NFPA 805, Section 2.7.1.3*, "Supporting Documentation": Detailed information used to develop and support the principal document shall be referenced as separate documents if not included in the principal document. NFPA 805, Section 2.7.2, "Configuration Control." NFPA 805, Section 2.7.2.1, "Design Basis Document": The design basis document shall be maintained up-to-date as a controlled document. Changes affecting the design, operation, or maintenance of the plant shall be reviewed to determine if these changes impact the fire protection program documentation. NFPA 805, Section 2.7.2.2, "Supporting Documentation": Detailed supporting information shall be retrievable records. Records shall be revised as needed to maintain the principal documentation up-to-date. NFPA 805, Section 2.7.3*, "Quality." NFPA 805, Section 2.7.3.1, "Review": Each analysis, calculation, or evaluation performed shall be independently reviewed. NFPA 805, Section 2.7.3.2*, "Verification and Validation": Each calculational model or numerical method used shall be verified and validated through comparison to test results or comparison to other acceptable models. NFPA 805, Section 2.7.3.3, "Limitations of Use": Acceptable engineering methods and numerical models shall only be used for applications to the extent these methods have been subject to verification and validation. These engineering methods shall only be applied within the scope, limitations, and assumptions prescribed for that method.

-115-NFPA 805, Section 2.7.3.4, "Qualification of Users": Cognizant personnel who use and apply engineering analysis and numerical models (e.g., fire modeling techniques) shall be competent in that field and experienced in the application of these methods as they relate to nuclear power plants, nuclear power plant fire protection, and power plant operations. NFPA 805, Section 2.7.3.5*, "Uncertainty Analysis": An uncertainty analysis shall be performed to provide reasonable assurance that the performance criteria have been met. 3.8.1 Documentation The NRC staff reviewed LAR Section 4.7.1, "Compliance with Documentation Requirements in Section 2.7.1 of NFPA 805," to evaluate the appropriateness of the content of the NMP1 FPP design basis document and supporting documentation. The NMP1 FPP design basis is a compilation of multiple documents (i.e., fire safety analyses, calculations, engineering evaluations, NSCAs, etc.), databases, and drawings which are identified in LAR Figure 4-8, "NFPA 805 Transition-Planned Post-Transition Documentation and Relationships for NMP1." The licensee stated that the analyses conducted to support the NFPA 805 transition were performed in accordance with NMP1 processes which meet or exceed the requirements for documentation outlined in NFPA 805, Section 2.7.1. Specifically, the licensee stated that the design analysis and calculation procedures provide the methods and requirements to ensure that design inputs and assumptions are clearly defined, results are easily understood by being clearly and consistently described, and that sufficient detail is provided to allow future review of the entire analysis. The process includes provisions for appropriate design and engineering review and approval. In addition, the approved analyses are considered controlled documents, and are accessible via NMP1 's document control system. Being analyses, they are also subject to review and revision consistent with the other plant calculations and analyses, as required by the plant design change process. The LAR also stated that the documentation associated with the FPP will be maintained for the life of the plant and organized in such a way to facilitate review for accuracy and adequacy by independent reviewers, including the NRC staff. Based on the description provided in the LAR, as supplemented, of the content of the NMP1 design basis and supporting documentation, and taking into account the licensee's plans to maintain this documentation throughout the life of the plant, the NRC staff concludes that the licensee's approach for meeting the requirements of NFPA 805, Sections 2.7.1.1, 2.7.1.2, and 2.7.1.3, regarding adequate development and maintenance of the FPP design basis documentation is acceptable.

-116 -3.8.2 Configuration Control The NRC staff reviewed LAR Section 4.7.2, "Compliance with Configuration Control Requirements in Section 2.7.2 and 2.2.9 of NFPA 805," in order to evaluate the configuration control process at NMP1. To support the many other technical, engineering and licensing programs at NMP1, the licensee has existing configuration control processes and procedures for establishing, revising, or utilizing program documentation. Accordingly, the licensee is integrating the RI/PB FPP design basis and supporting documentation into these existing configuration control processes and procedures. These processes and procedures require that all plant changes be reviewed for potential impact on the various NMP1 licensing programs, including the FPP. The LAR stated that the configuration control process includes; provisions for appropriate design, engineering reviews and approvals, and that approved analyses are considered controlled documents available through the NMP1 document control system. The licensee also stated that analyses based on the PRA program, which includes the FRE, are issued as formal analyses subject to these same configuration control processes, and are additionally subjected to the PRA peer review process specified in the ASMEIANS PRA Standard (Reference 29). Configuration control of the FPP during the transition period is maintained by the NMP1 Design Engineering and Configuration Control process, as defined in existing NMP1 configuration management and configuration control procedures. The licensee will revise these existing procedures as necessary for application to the NFPA 805 FPP (LAR AttachmentS, Table S-2, Item 20). The NRC staff reviewed the licensee's process for updating and maintaining the NMP1 FRE, in order to reflect plant changes made after completion of the transition to NFPA 805. Based on the description of the NMP1 configuration control process, which indicates that the NMP1 RI/PB FPP design basis and supporting documentation will be controlled documents and that plant changes will be reviewed for impact on the FPP, the NRC staff concludes that the licensee has a configuration control process which meets the requirements of NFPA 805, Sections 2.7.2.1 and 2.7.2.2, for revising FPP design basis documents, supporting documents, and applicable FPP documentation to reflect changes made to the RI/PB FPP after the NFPA 805 FPP has been implemented. 3.8.3 Quality The NRC staff reviewed LAR Section 4.7.3, "Compliance with Quality Requirements in Section 2.7.3 of NFPA 805," to evaluate the quality of the engineering analyses used to support transition to NFPA 805 at NMP1 based on the requirements outlined above. 3.8.3.1 Review NFPA 805, Section 2.7.3.1 requires that each analysis, calculation, or evaluation performed be independently reviewed. The licensee stated that its procedures require independent review of analyses, calculations, and evaluations, including those performed in support of compliance with

-117-10 CFR 50.48(c). The LAR also states that the analyses, calculations, and evaluations performed in support of the transition to NFPA 805 were independently reviewed, and that analyses, calculations, and evaluations to be performed post-transition will be independently reviewed as required by the existing procedures. Based on the licensee's description of the process for performing independent reviews of analyses, calculations, and evaluations, the NRC staff concludes that the licensee's approach for meeting the requirements of NFPA 805, Section 2.7.3.1, is acceptable. 3.8.3.2 Verification and Validation (V&V) NFPA 805, Section 2.7.3.2 requires that each calculational model or numerical method used be verified and validated through comparison to test results or other acceptable models. The licensee stated that in LAR Section 4.7.3 the calculation models and numerical methods used in support of the transition to NFPA 805 were verified and validated, and that the calculation models and numerical methods used post-transition will be similarly verified and validated through the use of existing administrative controls. As an example, the licensee provided extensive information related to the V&V of fire models used to support the development of the NMP1 FRE, which the NRC staff concludes were acceptable. The NRC staff's evaluation of this information is discussed below. 3.8.3.2.1 General NUREG-1824 (Reference 41) documents the V&V of five selected fire models commonly used to support applications of RI/PB fire protection at NPPs. The seven volumes of this series report provide technical documentation concerning the predictive capabilities of a specific set of fire dynamics calculation tools and fire phenomenological models that may be used for the analysis of fire hazards in postulated NPP fire scenarios. When used within the limitations of the fire models and considering the identified uncertainties, these models may be employed to demonstrate compliance with the requirements of 10 CFR 50.48(c). Accordingly, for those FM elements performed by the licensee using the V&V applications contained in NUREG-1824 to support the transition to NFPA 805 at NMP1, the NRC staff concludes that the use of these models is acceptable, provided that the application is used within the appropriate limitations of the model, as identified in NUREG-1824. In LAR Attachments J and V, the licensee also identified the use of several empirical correlations that are not addressed in NUREG-1824. The NRC staff reviewed these correlations, as well as the related material provided in the LAR, in order to determine whether the licensee adequately demonstrated alignment with specific portions of the applicable NUREG-1824 guidance. The NRC staff concluded that the theoretical bases of the models and empirical correlations used in the FM calculations that were not addressed in NUREG-1824 were identified and described in authoritative publications, such as the SFPE Handbook of Fire Protection Engineering (Reference 84).

-118 -As reflected in Tables 3.8-1 and 3.8-2 of SE Attachments A and B, the FM employed by the licensee in the development of the FRE used either: (1) empirical correlations that provide bounding solutions for the ZOI, or (2) conservative input parameters in the application of the other models, which produced conservative results for the FM analysis. Based on the above, the NRC staff concludes that the FM used in the development of the fire scenarios for the NMP1 FREis appropriate, and thus acceptable for use in this application (i.e., transition to NFPA 805) because the V&V of the empirical correlations used by the licensee were consistent with either NUREG-1824 or the SFPE Handbook of Fire Protection Engineering. 3.8.3.2.2 Discussion of Selected RAI Responses In a letter dated January 3, 2013 (Reference 15), the NRC staff issued RAis concerning the FM conducted to support the FRE. In a letter dated February 27, 2013 (Reference 7), the licensee responded to these RAis. The following paragraphs describe selected RAI responses related to the V&V of the fire models used.

The NRC staff issued FM RAI 02(a) dated January 3, 2013 (Reference 15), to ask the licensee to provide the V&V basis for the heat and smoke detection models that were used at NMP1. In its response to FM RAI 02(a) dated February 27, 2013 (Reference 7), the licensee stated that the heat and smoke detection models are described in NUREG-1805 (Reference 40), Chapters 10 and 11. The licensee explained that these models were applied conservatively throughout the plant, i.e., to determine whether a suppression system can be credited for fires that spread beyond the ignition source and initial target set. Based on a review of the documentation provided, the NRC staff concludes that the V&V bases for the heat and smoke detection models used at NMP1 are acceptable because the models used by the licensee are described in NUREG-1805.

The NRC staff issued FM RAI 02(b) dated January 3, 2013 (Reference 15), to ask the licensee to provide technical documentation to demonstrate that CF AST has been applied in the HGL calculations with input parameters that are within the validated range reported in the V&V basis documents. In its response to FM RAI 02(b) dated February 27, 2013 (Reference 7), the licensee demonstrated that CFAST was generally applied within the validated range, and provided detailed documentation to justify the application in the cases where CFAST was used outside the validated range. Based on a review of the documentation provided, the NRC staff concludes that CFAST as used in HGL calculations is acceptable because it was either used

-119-within the validated range of input parameters, or that the licensee provided justification for their application outside the validated range.

The NRC staff issued FM RAI 02(c) dated January 3, 2013 (Reference 15), to ask the licensee to provide evidence of validation of the FLASH-CAT model as used at NMP1. In its response to FM RAI 02(c) dated February 27, 2013 (Reference 7), the licensee explained that the FLASH-CAT model used at NMP1 was validated by showing reasonable agreement between model calculations and cable tray fire test data reported in NUREG/CR-7010 (Reference 42). Based on a review and explanation provided, the NRC staff concludes that the validation of the FLASH-CAT model as used at NMP1 is acceptable.

The NRC staff issued FM RAI 02(d) dated January 3, 2013 (Reference 15), to ask the licensee to provide evidence of the verification of the spreadsheet calculations that were developed to implement the FLASH-CAT model at NMP1. In its response to FM RAI 02(d) dated February 27, 2013 (Reference 7), the licensee provided documentation to show that the spreadsheet calculations were verified by duplicating selected FLASH-CAT results reported in NUREG/CR-7010. Based on a review of the documentation provided, the NRC staff concludes that the verification of the spreadsheet calculations developed to implement the FLASH-CAT model at NMP1 is acceptable.

The NRC staff issued FM RAI 02(e) dated January 3, 2013 (Reference 15), to ask the licensee to provide evidence of the verification of algebraic models in the FDT5 or FIVE as implemented at NMP1. In its response to FM RAI 02(e) dated February 27, 2013 (Reference 7), the licensee provided detailed documentation that shows how the algebraic models are implemented at NMP1. The documentation includes comparisons between the output from the NMP1 calculations and the results obtained with the FDT5 spreadsheets for a wide range of input parameters. Based on a review of the documentation provided, the NRC staff concludes that the verification of the algebraic models as implemented at NMP1 is acceptable since, in all cases, the two methods were within rounding errors.

The NRC staff issued FM RAI 02(f) dated January 3, 2013 (Reference 15), to ask the licensee to describe the verification process used to ensure results of the FM correlations were transferred correctly into spreadsheets or databases for further analysis.

-120-In its response to FM RAI 02(f) dated February 27, 2013 (Reference 7), the licensee explained that the results of the FM correlations were not transferred manually, but were used directly in the FM database to calculate scenario frequencies. Based on a review of the explanation and additional information provided, the NRC staff concludes that the licensee used the results of the FM correlations directly rather than transferring them into spreadsheets and databases for further processing. This process is acceptable. 3.8.3.2.3 Post-Transition The licensee also stated that it will revise the appropriate processes and procedures to include NFPA 805 quality requirements for use during the performance of post-transition FPP changes, including those for V&V. Revision of the applicable post-transition processes and procedures to include NFPA 805 requirements for V&V are included in Implementation Items 2, 5, and 6 of LAR AttachmentS, Table S-2. 3.8.3.2.4 Conclusion for Section 3.8.3.2 Based on the licensee's description of the NMP1 process for V&V of calculational models and numerical methods and their continued use post-transition, the NRC staff concludes that the licensee's approach to meeting the requirements of NFPA 805 Section 2.7.3.2 is acceptable because the models are consistent with approved uses in NRC guidance or other authoritative publications. 3.8.3.3 Limitations of Use NFPA 805, Section 2.7.3.3 requires that acceptable engineering methods and numerical models be used for transition to the extent that these methods have been subject to V&V; and that they only be applied within the scope, limitations, and assumptions prescribed for that method. The licensee stated that the engineering methods and numerical models used in support of the transition to NFPA 805 were used subject to the limitations of use outlined in NFPA 805, Section 2.7.3.3, and that the engineering methods and numerical models used post-transition will be subject to these same limitations of use. As an example, in LAR Attachment J, "Fire Modeling V&V," the licensee stated that the fire models developed to support the NFPA 805 transition at NMP1 fall within their V&V limitations. The licensee also stated that it will revise the appropriate processes and procedures to include the NFPA 805 quality requirements for use during the performance of post-transition FPP changes, including those for limitations of use. Revision of the applicable post-transition processes and procedures to include NFPA 805 requirements for limitations of use is included in Implementation Items 2, 5, and 20 of LAR Attachment S, Table S-2. The NRC staff assessed the acceptability of each empirical correlation or other fire model in terms of the limits of its use. Tables 3.8-1 and 3.8-2 of SE Attachments A and B summarizes

-121 -the fire models used, how each was applied in the NMP1 FRE, the V&V basis for each, and the NRC staff evaluation for each. 3.8.3.3.1 Discussion of Selected RAI Responses In a letter dated January 3, 2013 (Reference 15), the NRC staff issued RAis concerning the FM conducted to support the FRE. In a letter dated February 27, 2013 (Reference 7), the licensee responded to these RAis. The following paragraphs describe selected RAI responses related to the V&V of the fire models used.

The NRC staff issued FM RAI 03{a) dated January 3, 2013 (Reference 15), to ask the licensee to explain how the licensee ensured that algebraic models were used within their limits of applicability. In its response to FM RAI 03{a) dated February 27, 2013 (Reference 7), the licensee provided detailed documentation which shows that algebraic models were applied within their limits of applicability. In some cases conservative adjustments had to be made to bring the fire diameter within the validated range reported in the V&V basis documents. The cases where the Froude number was outside the NUREG-1824 validated range are addressed in FM RAI 03(b). Based on a review of the information and explanation provided, the NRC staff concludes that the algebraic fire models were either applied within their validated range of input parameters, or that their application outside the validated range was justified.

The NRC staff issued FM RAI 03{b) dated January 3, 2013 (Reference 15), to ask the licensee to provide technical justification for the application of algebraic fire models with a Froude number outside the NUREG-1824 validated range. In its response to FM RAI 03{b) dated February 27, 2013 (Reference 7), the licensee explained that in all cases where an algebraic model was applied outside its range of applicability, the Froude number is below the lower limit of the NUREG-1824 validated range. The licensee showed that in these cases temperature and flame height are overestimated. Based on a review of the information provided, the NRC staff concludes that the use of algebraic fire models with a Froude number outside the validated range is acceptable because the Froude number is below the lower limit of the validated range provided in NUREG-1824.

The NRC staff issued FM RAI 03(c) dated January 3, 2013 (Reference 15), to ask the licensee to provide technical documentation to demonstrate that CFAST has been applied in the HGL calculations with a room length-to-width ratio that is within the validated range reported in the V&V basis documents.

-122-In its response to FM RAI 03(c) dated February 27, 2013 (Reference 7), the licensee explained that the aspect ratios for each fire zone were compared against the NUREG-1824 validation range. When an aspect ratio was outside of the range, the room dimensions were adjusted in a conservative direction so that the ratio fell within the applicable range and a sensitivity CFAST case was run with the adjusted room dimensions. The licensee provided documentation that shows that the results of the sensitivity cases were identical to those obtained with the original room dimensions. Based on a review of the explanation and information provided, the NRC staff concludes that the use of CFAST in rooms with a length-to-width ratio outside the validated range is acceptable because in each case the licensee demonstrated through additional CFAST runs that reducing the room size to bring the width ratio within the validated range does not affect the results. 3.8.3.3.2 Post-Transition The licensee also stated that it will revise the appropriate processes and procedures to include the NFPA 805 quality requirements for use during the performance of post-transition FPP changes, including those for limitations of use. Revision of the applicable post-transition processes and procedures to include NFPA 805 requirements for limitations of use is included in Implementation Items 2, 5 and 20 of LAR AttachmentS, Table S-2. 3.8.3.3.4 Conclusion for Section 3.8.3.3 Based on the licensee's statements that the fire models used to support development of the FRE were used within their limitations, the licensee's documentation that provides adequate justification for other uses, and the description of the NMP1 process for placing limitations on the use of engineering methods and numerical models, the NRC staff concludes that the licensee's approach to meeting the requirements of NFPA 805 Section 2.7.3.3 is acceptable. 3.8.3.4 Qualification of Users NFPA 805, Section 2.7.3.4 requires that personnel performing engineering analyses and applying numerical methods (e.g. FM) shall be competent in that field and experienced in the application of these methods as they relate to NPPs, NPP fire protection, and power plant operations. The licensee's procedures require that cognizant personnel who use and apply engineering analyses and numerical models be competent in the field of application and experienced in the application of the methods, including those personnel performing analyses in support of compliance with 10 CFR 50.48(c). Specifically, these requirements are being addressed through the implementation of an engineering qualification process at NMP1. The licensee has developed procedures that require that cognizant personnel who use and apply engineering analyses and numerical models be competent in the field of application and experienced in the application of the methods, including those personnel performing analyses in support of compliance with 10 CFR 50.48(c). These requirements are being addressed through the implementation of an

-123-engineering qualification process. The licensee has developed qualification or training requirements for personnel performing engineering analyses and numerical methods. The NRC staff reviewed the engineering qualification process at NMP1, and concludes that competent and experienced personnel developed the NMP1 FRE, including the supporting FM calculations and including the additional documentation for models and empirical correlations not identified in previous NRC approved V&V documents. 3.8.3.4.1 Discussion of Selected RAI Responses In a letter dated January 3, 2013 (Reference 15), the NRC staff issued RAis concerning the FM conducted to support the FRE. In a letter dated February 27, 2013 (Reference 7), the licensee responded to these RAis. The following paragraphs describe selected RAI responses related to the qualifications of the personnel who supported NMP1 FRE FM.

The NRC staff issued FM RAI 04(a) dated January 3, 2013 (Reference 15), to ask the licensee to describe the requirements to qualify NMP1 staff and consulting engineers who perform FM calculations in support of the NFPA 805 transition. In its response to FM RAI 04(a) (Reference 7), the licensee explained that the FM calculations must be performed by a Fire Protection Engineer who meets the requirements in Section 2.7.3.4 of NFPA 805. The qualification process also includes existing training requirements and procedures.

The NRC staff issued FM RAI 04{b) dated January 3, 2013 (Reference 15), to ask the licensee to describe the process and procedures for ensuring adequate qualification of the engineers and personnel who perform FM calculations before and post-transition. In its response to FM RAI 04(b) (Reference 7), the licensee explained that the credentials of the initial FM team was reviewed per NMP1 procedure and that the existing engineering staff will continue to be knowledgeable in FM techniques, including interpreting and maintaining the FM database. If new FM personnel are needed in the future, their credentials will also be reviewed and approved by the licensee's supervision.

The NRC staff issued FM RAI 04{c) dated January 3, 2013 (Reference 15), to ask the licensee to describe the organization and communication between the engineers and personnel who performed the FM and those who performed the FPRA. In its response to FM RAI 04(c) (Reference 7), the licensee explained that the Fire Protection Engineers who conducted the FM and the PRA engineers maintained frequent communications. During the development phase of the FPRA, the FM personnel populated the FM database. The scenario frequencies, which are produced by the database, are electronically sent to the FPRA

-124-engineers, who perform the quantification. Both the Fire Protection Engineers and the FPRA engineers participated in the cut-set review meetings during the development of the FPRA. The FM database will be maintained under the responsibility of the Fire Protection Engineer and the engineers responsible for the FPRA. Based on its review and the above explanations, the NRC staff concludes that competent and experienced personnel developed the NMP1 FRE, including the supporting FM calculations and the additional documentation for models and empirical correlations not identified in previous NRC approved V&V documents. Further, in a letter dated February 27, 2013 (Reference 7), the licensee responded to Programmatic RAI 04 and stated that: Current CENG Design and PRA staff members are required to maintain qualification cards to ensure these personnel have the appropriate training and technical expertise to perform assigned work, including the use of engineering analyses and numerical models. NMPNS will maintain qualification requirements for individuals assigned to perform NFPA 805 related tasks. Position Specific Guides will be developed to identify and document required training and mentoring to ensure cognizant individuals are appropriately qualified to perform assigned work per the requirements of NFPA 805, Section 2.7.3.4. In addition, based on the licensee's description of the procedures for ensuring personnel who use and apply engineering analyses and numerical methods are competent and experienced, the NRC staff concludes that the licensee's approach for meeting the requirements of NFPA 805, Section 2.7.3.4, is acceptable. 3.8.3.4.2 Conclusion for Section 3.8.3.4 Based on the above discussions, the NRC staff concludes that the NMP1 qualification program addresses the requirements of NFPA 805, Section 2.7.3.4, which include personnel performing engineering analyses and applying numerical methods (e.g. FM) are competent in that field and experienced in the application of these methods as they relate to NPPs fire protection, and power plant operations. 3.8.3.5 Uncertainty Analysis NFPA 805 requires that an uncertainty analysis be performed to provide reasonable assurance that the performance criteria have been met. (Note: 10 CFR 50.48(c)(2)(iv) states that an uncertainty analysis performed in accordance with NFPA 805, Section 2.7.3.5, is not required to support calculations used in conjunction with a deterministic approach.) In the letter dated February 27, 2013 (Reference 7), the licensee stated that an uncertainty analysis was performed for the analyses used in support of the transition to NFPA 805, and that an uncertainty analysis will be performed for post-transition analyses through existing administrative controls.

-125-3.8.3.5.1 General The ASME/ANS PRA standard, (Reference 29), includes requirements to address uncertainty. Accordingly, the licensee addressed uncertainty as a part of the development of the NMP1 FRE. The NRC staff's evaluation of the licensee's treatment of these uncertainties is discussed in Section 3.4. 7 of this SE. According to NUREG-1855, Volume 1 (Reference 43), there are three types of uncertainty associated with FM calculations: (1) Parameter Uncertainty: Input parameters are often chosen from statistical distributions or estimated from generic reference data. In either case, the uncertainty of these input parameters affects the uncertainty of the results of the FM analysis; (2) Model Uncertainty: Idealizations of physical phenomena lead to simplifying assumptions in the formulation of the model equations. In addition, the numerical solution of equations that have no analytical solution can lead to inexact results. Model uncertainty is estimated via the processes of V&V. An extensive discussion of quantifying model uncertainty can be found in NUREG-1934 (Reference 45); and (3) Completeness Uncertainty: This refers to the fact that a model is not a complete description of the phenomena it is designed to simulate. Some consider this a form of model uncertainty because most fire models neglect certain physical phenomena that are not considered important for a given application. Completeness uncertainty is addressed by the description of the algorithms found in the model documentation. It is addressed, indirectly by the same process used to address the Model Uncertainty. 3.8.3.5.2 Discussion of Fire Modeling RAis In a letter dated January 3, 2013 (Reference 15), the NRC staff issued RAis concerning the FM conducted to support the NMP1 FRE. In a letter dated February 27, 2013 (Reference 7), the licensee provided response to these RAis. The following paragraphs describe selected RAI responses related to the uncertainty of the FM results.

The NRC staff issued FM RAI 05(a) part i dated January 3, 2013 (Reference 15}, to ask the licensee to explain how the uncertainty associated with the input parameters was accounted for in the FM analyses. In its response to FM RAI 05{a) part i dated February 27, 2013 (Reference 7), the licensee explained that the uncertainty associated with the fire model input parameters was accounted for by using conservative input parameters and varying input parameters in sensitivity cases. The licensee briefly described how parameter uncertainty was addressed specifically in the MCR abandonment time calculations and the FM analyses for individual compartments.

-126-Based on its review of the licensee's explanation, the NRC staff concludes that the licensee's response to FM RAI 05(a) part i provides assurance that the results of the FM performed at NMP1 in support of the transition to NFPA 805 are within the bounds of experimental uncertainty.

The NRC staff issued FM RAI 05(a) part ii dated January 3, 2013 (Reference 15}, to ask the licensee to describe how the "model" uncertainty was accounted for in the FM analyses. In its response to FM RAI 05(a) part ii dated February 27, 2013 (Reference 7), the licensee stated that fire models were generally used within their validated range, and that in cases where a model was applied outside its range of applicability input parameters were conservatively adjusted to bring the model within the range or sensitivity cases were run as recommended in NUREG-1934. Based on its review of the licensee's explanation, the NRC staff concludes that the licensee's response to FM RAI 05(a) part ii demonstrates that model uncertainty was properly accounted for because the licensee's process is consistent with the guidance in NUREG-1934.

The NRC staff issued FM RAI 05(a) part iii dated January 3, 2013 (Reference 15), to ask the licensee to describe how the "completeness" uncertainty was accounted for in the FM analyses. In its response to FM RAI 05(a) part iii dated February 27, 2013 (Reference 7), the licensee explained that the FPRA conservatively compensates in the areas where the FM provide incomplete information. The licensee provided four examples to illustrate how model completeness uncertainty is addressed in the FPRA. Based on its review of the licensee's explanation that the FPRA conservatively compensates for incomplete information, the NRC staff concludes that completeness uncertainty was properly accounted for.

The NRC staff issued FM RAI 05(b) dated January 3, 2013 (Reference 15), to ask the licensee to describe specifically how input parameter uncertainty was accounted for in the MCR abandonment study. In its response to FM RAI 05(b) dated February 27, 2013 (Reference 7), the licensee explained that, to address parameter uncertainty in the MCR abandonment study sensitivity, runs were conducted in which fire height and fuel properties were varied. Sensitivity cases were not performed for the temperature because MCR abandonment was controlled by HGL height and visibility. Sensitivity cases were performed to evaluate the effect of the room volume reduction due to the presence of cabinets and other obstructions as discussed in the response to FM RAI 01 (h).

-127-Based on its review of the licensee's explanation, the NRC staff concludes that sensitivity runs performed by the licensee provides assurance that FM input parameter uncertainty in the MCR abandonment study was properly accounted for. 3.8.3.5.3 Post-Transition The licensee stated that it will revise the appropriate processes and procedures to include the NFPA 805 quality requirements for use during the performance of post-transition FPP changes, including those regarding uncertainty analysis. Revision of the applicable post-transition processes and procedures to include NFPA 805 requirements regarding uncertainty analysis are included as Implementation Items 2, 5 and 20 of LAR AttachmentS, Table S-2. 3.8.3.5.4 Conclusion for Section 3.8.3.5 The NRC staff reviewed the licensee's description of the process for performing an uncertainty analysis, and concludes that the licensee's approach for meeting the requirements of NFPA 805 Section 2.7.3.5 is acceptable. 3.8.3.6 Conclusion for Section 3.8.3 Based on the above discussions, the NRC staff concludes that the NMP1 RI/PB FPP meets each of the requirements of NFPA 805, Section 2.7.3, which include conducting independent reviews, performing V&V, limiting the application of acceptable methods and models to within prescribed boundaries, ensuring that personnel applying acceptable methods and models are qualified, and performing uncertainty analyses. 3.8.4 Fire Protection Quality Assurance Program GDC 1 of Appendix A to 10 CFR Part 50 requires the following: Structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed. The licensee established its Fire Protection QA Program in accordance with the guidelines of NUREG-0800, Section 9.5.1 Position C.4, "Quality Assurance Program," (Reference 86). In addition, the guidance in Appendix C to NEI 04-02 (Reference 5) suggests that the LAR include a description of how the existing fire protection QA program will be transitioned to the new NFPA 805 RI/PB FPP, as discussed below. The LAR stated that the fire protection QA program is included within and implemented by the NMP1 nuclear QA program, although certain aspects of that program are not applicable to the FPP. The licensee included Implementation Items 2 and 5 in LAR AttachmentS, Table S-2 for revising the QA program to reflect the applicable requirements of NFPA 805 Section 2.7.3.

-128-The NRC staff concludes that the licensee's changes to the fire protection QA program are acceptable because they include the expansion of the existing program to include those fire protection systems that were previously not included within the scope of the fire protection QA program that are required by NFPA 805 Chapter 4 and they include the applicable requirements of Section 2.7.3 of NFPA 805. 3.8.5 Conclusion for Section 3.8 The NRC staff reviewed the licensee's RI/PB FPP, as described in the LAR, as supplemented, to evaluate the NFPA 805 program documentation content, the associated configuration control process, and the appropriate QA requirements. The NRC staff concludes that the licensee's approach meets the requirements specified in NFPA 805, Section 2.7, regarding program documentation, configuration control, and quality. 4.0 FIRE PROTECTION LICENSE CONDITION The licensee proposed a FPP license condition regarding transition to an RI/PB FPP under NFPA 805, in accordance with 10 CFR 50.48(c)(3)(i). The new license condition adopts the guidelines of the standard fire protection license condition promulgated in RG 1.205, Revision 1, Regulatory Position C.3.1, as issued on December 18, 2009 (74 FR 67253). Plant-specific changes were made to the sample license condition; however, the proposed plant-specific FPP license condition is consistent with the standard fire protection license condition, incorporates all of the relevant features of the transition to NFPA 805 at NMP1 and is, therefore, acceptable. The following license condition is included in the revised license for the Nine Mile Point Nuclear Station, Unit 1, and will replace Operating License No. DPR-63 Condition 2.D.(7): Fire Protection Exelon Generation shall implement and maintain in effect all provisions of the approved fire protection program that comply with 1 0 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the license amendment request dated June 11, 2012, supplemented by letters dated February 27, 2013, March 27, 2013, April30, 2013, December 9, 2013, January 22,2014, March 14,2014, April15, 2014, May 9, and May 23, 2014, and as approved in the safety evaluation dated June 30, 2014. Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c), the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied.

-129-(a) Risk-Informed Changes that May Be Made Without Prior NRC Approval A risk assessment of the change must demonstrate that the acceptance criteria below are met. The risk assessment approach, methods, and data shall be acceptable to the NRC and shall be appropriate for the nature and scope of the change being evaluated; be based on the built, as-operated, and maintained plant; and reflect the operating experience at the plant. Acceptable methods to assess the risk of the change may include methods that have been used in the peer-reviewed fire PRA model, methods that have been approved by NRC through a plant-specific license amendment or NRC approval of generic methods specifically for use in NFPA 805 risk assessments, or methods that have been demonstrated to bound the risk impact. 1. Prior NRC review and approval is not required for changes that clearly result in a decrease in risk. The proposed change must also be consistent with the defense-in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation. 2. Prior NRC review and approval is not required for individual changes that result in a risk increase less than 1 x1 o-7/year (yr) for CDF and less than 1 x1 o-8/yr for LEAF. The proposed change must also be consistent with the defense-in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation. (b) Other Changes that May Be Made Without Prior NRC Approval 1. Changes to NFPA 805, Chapter 3, Fundamental Fire Protection Program Prior NRC review and approval are not required for changes to the NFPA 805, Chapter 3, fundamental fire protection program elements and design requirements for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is functionally equivalent or adequate for the hazard. The licensee may use an engineering evaluation to demonstrate that a change to an NFPA 805, Chapter 3, element is functionally equivalent to the corresponding technical requirement. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard. The licensee may use an engineering evaluation to demonstrate that changes to certain NFPA 805, Chapter 3, elements are

-130-acceptable because the alternative is "adequate for the hazard." Prior NRC review and approval would not be required for alternatives to four specific sections of NFPA 805, Chapter 3, for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is adequate for the hazard. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard. The four specific sections of NFPA 805, Chapter 3, are as follows: * "Fire Alarm and Detection Systems" (Section 3.8); * "Automatic and Manual Water-Based Fire Suppression Systems" (Section 3.9); * "Gaseous Fire Suppression Systems" (Section 3.1 0); and * "Passive Fire Protection Features" (Section 3.11). This License Condition does not apply to any demonstration of equivalency under Section 1.7 of NFPA 805. 2. Fire Protection Program Changes that Have No More than Minimal Risk Impact Prior NRC review and approval are not required for changes to the licensee's fire protection program that have been demonstrated to have no more than a minimal risk impact. The licensee may use its screening process as approved in the NRC safety evaluation dated June 30, 2014, to determine that certain fire protection program changes meet the minimal criterion. The licensee shall ensure that fire protection defense-in-depth and safety margins are maintained when changes are made to the fire protection program. (c) Transition License Conditions 1. Before achieving full compliance with 10 CFR 50.48(c), as specified by (c)2. below, risk-informed changes to the licensee's fire protection program may not be made without prior NRC review and approval unless the change has been demonstrated to have no more than a minimal risk impact, as described in (b)2. above. 2. The licensee shall implement the modifications to its facility, as described in Table S-1, "Plant Modifications Committed," of NMPNS letter dated May 9, 2014, to complete the transition to full compliance with 10 CFR 50.48(c) prior to startup from the first refueling outage following issuance of the license amendment. The licensee shall maintain appropriate compensatory measures in place until completion of these modifications.

-131 -3. The licensee shall implement the items listed in Table S-2, "Implementation Items," of NMPNS letter dated May 9, 2014, 180 days after issuance of the license amendment unless that date falls within a scheduled refueling outage, then the due date will be 60 days following startup from the scheduled refueling outage. 5.0 SUMMARY The licensee's LAR and subsequent supplemental submissions included regulatory commitments. Generally, during its review of the LAR, the NRC staff, through the RAI process, asked the licensee to confirm if a regulatory commitment had been completed (for the short-term items). For the other items, NRC staff asked the licensee to move the "commitments" that the NRC staff needed for theSE into the S-1 and S-2 tables where they become obligations under the Transition License Conditions section. Therefore, a "regulatory commitment" made in the licensee submission, was either completed and cleared up or promoted to an obligation in the form of the "Transition License Condition," by reference to the Tables S-1 and S-2 in the Licensee submission dated May 09, 2014. The NRC staff reviewed the licensee's application, as supplemented by various letters, to transition to an RI/PB FPP in accordance with the requirements established by NFPA 805. The NRC staff concludes that the applicant's approach, methods, and data are acceptable to establish, implement and maintain an RI/PB FPP in accordance with 10 CFR 50.48(c).

6.0 STATE CONSULTATION

In accordance with the Commission's regulations, the New York State official was notified on March 31, 2014, of the proposed issuance of the amendment. The state official had no comments.

7.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts and no significant change in the types of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding published in the Federal Register on September 11, 2012 (77 FR 55874). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

-132-

8.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner; (2) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations; and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

-133-9.0 REFERENCES 1. National Fire Protection Association, NFPA 805, "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants," 2001 Edition, Quincy, Massachusetts. 2. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.205, "Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants," Revision 1, December 2009 (ADAMS Accession No. ML092730314). 3. U.S. Nuclear Regulatory Commission, SECY-98-058, "Development of a Informed, Performance-Based Regulation for Fire Protection at Nuclear Power Plants," March 1998 (ADAMS Accession No. ML99291 01 06). 4. U.S. Nuclear Regulatory Commission, SECY-00-0009, "Rulemaking Plan, Reactor Fire Protection Risk-Informed, Performance-Based Rulemaking," January 2000 (ADAMS Accession No. ML003671923). 5. Nuclear Energy Institute, NEI 04-02, Guidance for Implementing a Informed, Performance-Based Fire Protection Program Under 1 0 CFR 50.48{c}, Revision 2, Washington, DC, April 2008 (ADAMS Accession No. ML081130188). 6. Letter from K. Langdon (NMPNS) to Document Control Desk (NRC), dated June 11, 2012, License Amendment Request Pursuant to 10 CFR 50.90: Adoption of NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants {2001 Edition) (ADAMS Accession Nos. ML 12170A868 and ML 12170A869). 7. Letter from Christopher Costanzo (NMPNS) to Document Control Desk (NRC), dated February 27, 2013, License Amendment Request Pursuant to 10 CFR 50.90: Adoption of NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants (2001 Edition)-Response to NRC Request for Additional Information (ADAMS Accession No. ML 13064A466}. 8. Letter from Christopher Costanzo (NMPNS) to Document Control Desk (NRC), dated March 27, 2013, License Amendment Request Pursuant to 10 CFR 50.90: Adoption of NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants (2001 Edition) -Response to NRC Request for Additional Information {TAC No. ME8899} (ADAMS Accession No. ML 13092A139). 9. Letter from Christopher Costanzo (NMPNS) to Document Control Desk (NRC), dated April 30, 2013 License Amendment Request Pursuant to 10 CFR 50.90: Adoption of NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants (2001 Edition)-Response to NRC Request for Additional Information (ADAMS Accession Nos. ML 131270405

-134-(package), and ML 13127A395, ML 13127A397, ML 13127A398) Portions contain proprietary SUNS/, withheld under 10 CFR 2. 390. 10. Letter from Christopher Costanzo (NMPNS) to Document Control Desk (NRC), dated December 9, 2013 License Amendment Request Pursuant to 10 CFR 50.90: Adoption of NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants (2001 Edition)-Response to NRC Request for Additional Information (ADAMS Accession No. ML 13347B187) 11. Letter from Terry Syrell (NMPNS) to Document Control Desk (NRC), dated January 22, 2014, License Amendment Request Pursuant to 10 CFR 50.90: Adoption of NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants (2001 Edition) -Response to NRC Request for Additional Information (ADAMS Accession No ML 14024A416) 12. Letter from Christopher Costanzo (NMPNS) to Document Control Desk (NRC), dated March 14, 2014, License Amendment Request Pursuant to 10 CFR 50.90: Adoption of NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants (2001 Edition)-Response to NRC Request for Additional Information (ADAMS Accession No. ML 14085A 161 ). Portions contain proprietary SUNS/, withheld under 10 CFR 2. 390. 13. Letter from Christopher Costanzo (NMPNS) to Document Control Desk (NRC), dated April 15, 2014, License Amendment Request Pursuant to 10 CFR 50.90: Adoption of NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants (2001 Edition)-Response to NRC Request for Additional Information (ADAMS Accession No. ML 14113A 175) 14. Letter from Christopher Costanzo (NMPNS) to Document Control Desk (NRC), dated May 09, 2014, License Amendment Request Pursuant to 1 0 CFR 50.90: Adoption of NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants (2001 Edition)-Response to NRC Request for Additional Information (ADAMS Accession No ML 14132A 199) 15. Bhalchandra Vaidya, U.S. Nuclear Regulatory Commission, letter to Mr. Ken Langdon (NMPNS), dated January 3, 2013, Nine Mile Point Nuclear Station, Unit No.1 -Request For Additional Information Regarding License Amendment Request For Adoption of NFPA 805 (ADAMS Accession No. ML12361A050) 16. Bhalchandra Vaidya, U.S. Nuclear Regulatory Commission, letter to Christopher Costanzo (NMPNS) dated October 9, 2013, Nine Mile Point Nuclear Station, Unit No. 1 -Second Round Of Request For Additional Information Regarding License Amendment Request for Adoption of NFPA 805 (ADAMS Accession No. ML 13281A010) 17. Bhalchandra Vaidya, U.S. Nuclear Regulatory Commission, letter to Christopher Costanzo (NMPNS) dated February 12, 2014, Nine Mile Point Nuclear Station, Unit

-135-No. 1 -Third Round Of Request For Additional Information Regarding License Amendment Request for Adoption of NFPA 805 (ADAMS Accession No. ML 14030A377) 18. Ippolito, Thomas, U.S. Nuclear Regulatory Commission, letter to D. Dise, Niagara Mohawk Power Corporation, "Issuance of Amendment No. 33 to Facility Operating License No. DPR-63 for the Nine Mile Point Nuclear Station Unit No. 1 ," dated July 26, 1979, (ADAMS Accession No. ML01 0990290}. 19. Vassallo, Domenic, U.S. Nuclear Regulatory Commission, letter to G. Rhode, Niagara Mohawk Power Corporation, "Exemption Requests-10 CFR 50.48 Fire Protection and Appendix R to 10 CFR Part 50," dated March 21, 1983, (ADAMS Accession No. ML091310057). 20. Ippolito, Thomas, U.S. Nuclear Regulatory Commission, letter to D. Dise, Niagara Mohawk Power Corporation, "Updates Review Status of Supplementary Items to Fire Protection Safety Evaluation (Amend 33}, Item 3.1.2, Sprinkler Sys & Item 5.3.2.6, Diesel Generator Rooms, Remain Open," dated July 22, 1980 {ADAMS Accession No. 80080601 02). 21. Ippolito, Thomas, U.S. Nuclear Regulatory Commission, letter to D. Dise, Niagara Mohawk Power Corporation, "Notifies Of Completion Of Final Incomplete Items Pertaining To Sprinkler Sys, Safety Evaluation Items 3.1.2{2), 3.1.2(3) & 5.2.1.6, Re Cable Spreading Room & Diesel Generator Bldg Remain Open," dated July 30, 1980 (ADAMS Accession No. 8008080117}. 22. Martin, Robert, U.S. Nuclear Regulatory Commission, letter to L. Burkhardt, Niagara Mohawk Power Corporation, "Concurs W/Licensee 891103 Conclusion That Mods That Differ From Fire Protection SER Do Not Result In Decrease In Effectiveness Of Unit Fire Protection Program & Satisfies 1 OCFR50, App R Requirements," dated December 15, 1989 (ADAMS Accession No. 891221 0024). 23. Vassallo, D., U.S. Nuclear Regulatory Commission, letter to G. Rhode, Niagara Mohawk Power Corporation, "Forwards Safety Evaluation, Re: Proposed Mods & Alternate Safe Shutdown Capability To Comply W/App R Requirements, Mods & Alternate Safe Shutdown Capability Acceptable," dated March 3, 1983 (ADAMS Accession No. 8303170213). 24. Hermann, Robert, U.S. Nuclear Regulatory Commission, letter to G. Hooten, Niagara Mohawk Power Corporation, "Forwards Amend 71 To License DPR-63 & Safety Evaluation, Amend Revises Tech Specs To Add Limiting Conditions For Operation, Surveillance Requirements & Bases For Remote Shutdown Panels," dated April 1, 1985 (ADAMS Accession No. ML01 0990284). 25. Hermann, Robert, U.S. Nuclear Regulatory Commission, letter to G. Hooten, Niagara Mohawk Power Corporation, "Forwards Safety Evaluation, Re: Proposed Mods & Alternate Safe Shutdown Capability To Comply W/App R Requirements, Mods &

-136-Alternate Safe Shutdown Capability Acceptable," dated August 6, 1986 (ADAMS Accession No. 8504170016). 26. Nuclear Energy Institute, NEI 00-01, "Guidance for Post Fire Safe Shutdown Circuit Analysis, Revision 2, Nuclear Energy Institute (NEI), Washington, DC, May 2009 (ADAMS Accession No. ML091770265). 27. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 2, May 2011 (ADAMS Accession No. ML 10091 0006). 28. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk Informed Activities," Revision 2, March 2009 (ADAMS Accession No. ML09041 0014). 29. American Society of Mechanical Engineers (ASME) and American Nuclear Society (ANS) 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," dated February 2, 2009. 30. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.189, "Fire Protection for Nuclear Power Plants," Revision 2, October 2009 (ADAMS Accession No. M L092580550}. 31. U.S. Nuclear Regulatory Commission, NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," Chapter 9.5.1.2, "Risk-Informed, Performance-Based Fire Protection Program," Revision 0, December 2009 (ADAMS Accession No. ML092590527). 32. U.S. Nuclear Regulatory Commission, NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," Chapter 19.1, "Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed License Amendment Requests After Initial Fuel Load," Revision 3, September 2012 (ADAMS Accession No. ML 12193A 1 07}: 33. U.S. Nuclear Regulatory Commission, NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," Chapter 19.2, "Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance," Revision 0, June 2007 (ADAMS Accession No. ML071700658). 34. U.S. Nuclear Regulatory Commission, NUREG/CR-6850, "EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities, Volume 1: Summary and Overview," September 2005 (ADAMS Accession No. ML052580075).

-137-35. U.S. Nuclear Regulatory Commission, NUREG/CR-6850, "EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities, Volume 2: Detailed Methodology," September 2005 (ADAMS Accession No. ML052580118). 36. U.S. Nuclear Regulatory Commission, NUREG/CR-6850, Supplement 1, "Fire Probabilistic Risk Assessment Methods Enhancements," September 2010 (ADAMS Accession No. ML 1 03090242). 37. Correia, R. P., Memorandum to Joseph G. Giitter, U.S. Nuclear Regulatory Commission, "Interim Technical Guidance on Fire-Induced Circuit Failure Mode Likelihood Analysis," dated June 14, 2013 (ADAMS Accession No. ML 13165A 194). 38. U.S. Nuclear Regulatory Commission NUREG/CR-6931, "Cable Response to Live Fire (CAROL-FIRE)," Volumes 1, 2, and 3, April2008 (ADAMS Accession Nos. ML081190230, ML081190248, and ML081190261 ). 39. U.S. Nuclear Regulatory Commission, NUREG/CR-71 00, "Direct Current Electrical Shorting in Response to Exposure Fire (DESIREE-Fire): Test Results," April2012 (ADAMS Accession No. ML 121600316). 40. U.S. Nuclear Regulatory Commission, NUREG-1805, "Fire Dynamics Tools (FDTS): Quantitative Fire Hazard Analysis Methods for the U.S. Nuclear Regulatory Commission Fire Protection Inspection Program," December 2004 (ADAMS Accession No. ML043290075). 41. U.S. Nuclear Regulatory Commission, NUREG-1824, "Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications," May 2007. Volume 1: Main Report, Volume 2: Experimental Uncertainty, Volume 3: Fire Dynamics Tools (FDTs), Volume 4: Fire-Induced Vulnerability Evaluation (FIVE-Rev1 ), Volume 5: Consolidated Fire Growth and Smoke Transport Model (CFAST), Volume 6: MAGIC, and Volume 7: Fire Dynamics Simulator (ADAMS Accession Nos. ML071650546, ML071730305, ML071730493, ML071730499, ML071730527, ML071730504, ML071730543, respectively). 42. U.S. Nuclear Regulatory Commission, NUREG/CR-701 0, Volume 1, "Cable Heat Release, Ignition, and Spread in Tray Installations during Fire (CHRISTl FIRE), Phase 1: Horizontal Trays," July 2012 (ADAMS Accession No. ML 12213A056). 43. U.S. Nuclear Regulatory Commission, NUREG-1855, Volume 1, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making," March 2009 (ADAMS Accession No. ML090970525). 44. U.S. Nuclear Regulatory Commission, NUREG-1921, "EPRI/NRC-RES Fire Human Reliability Analysis Guidelines," July 2012 (ADAMS Accession No. ML 12216A 1 04).

-138-45. U.S. Nuclear Regulatory Commission, NUREG-1934 (EPRI 1023259), "Nuclear Power Plant Fire Modeling Analysis Guidelines (NPP FIRE MAG)," November 2012 (ADAMS Accession No. ML 12314A165). 46. U.S. Nuclear Regulatory Commission, Generic Letter 2006-03, "Potentially Nonconforming HEMYC and MT Fire Barrier Configurations," dated April1 0, 2006 (ADAMS Accession No. ML053620142). 47. National Fire Protection Association, NFPA 101, "Life Safety Code," Quincy, Massachusetts 48. National Fire Protection Association, NFPA 30, "Flammable and Combustible Liquids Code," Quincy, Massachusetts 49. National Fire Protection Association, NFPA 12, "Standard on Carbon Dioxide Extinguishing Systems," Quincy, Massachusetts 50. National Fire Protection Association NFPA 10, "Standard for Portable Fire Extinguishers," Quincy, Massachusetts 51. Klein, Alexander R., U.S. Nuclear Regulatory Commission, memorandum to file, "Close-out of National Fire Protection Association Frequently Asked Question 07-0030 on Establishing Recovery Actions," dated February 4, 2011 (ADAMS Accession No. ML 110070485). 52. Klein, Alexander R., U.S. Nuclear Regulatory Commission, memorandum to file, "Close-out of National Fire Protection Association Frequently Asked Question 07-0038 on Lessons Learned on Multiple Spurious Operations," dated February 3, 2011 (ADAMS Accession No. ML 110140242).* 53. Klein, Alexander R., U.S. Nuclear Regulatory Commission, memorandum to file, "Close-out of National Fire Protection Association Standard 805 Frequently Asked Question 07-0039 Incorporation of Pilot Plant Lessons Learned-Table B-2," dated January 15, 2010 (ADAMS Accession No. ML091320068). 54. Nuclear Energy Institute, NEI 00-01, "Guidance for Post Fire Safe Shutdown Circuit Analysis, Revision 1, Nuclear Energy Institute (NEI), Washington, DC, January 2005 (ADAMS Accession No. ML05031 0295). 55. Klein, Alexander R., U.S. Nuclear Regulatory Commission, memorandum to file, "Close-out of National Fire Protection Association 805 Frequently Asked Question 07-0040 on Non-Power Operations Clarifications," dated August 11, 2008 (ADAMS Accession No. ML082200528).

-139-56. Klein, Alexander R., U.S. Nuclear Regulatory Commission, memorandum to file, "Close-out of National Fire Protection Association Frequently Asked 08-0054 on Demonstrating Compliance with Chapter 4 of National Fire Protection Association 805,"dated February 17,2011 (ADAMS Accession No. ML110140183). 57. Klein, Alexander R., U.S. Nuclear Regulatory Commission, memorandum to file, "Close out of National Fire Protection Association 805 Frequently Asked Question 09-0056 on Radioactive Release Transition," dated January 14, 2011 (ADAMS Accession No. ML 1 02920405). 58. Klein, Alexander R., U.S. Nuclear Regulatory Commission, memorandum to file, "Close-out of National Fire Protection Association Standard 805 Frequently Asked Question 10-0059: National Fire Protection 805 Monitoring Program," dated March 19, 2012 (ADAMS Accession No. ML 1207501 08). 59. Klein, Alexander R., U.S. Nuclear Regulatory Commission, memorandum to file, "Close-out of National Fire Protection Association Standard 805 Frequently Asked Question 12-0062: Updated Final Safety Analysis Report Content," dated September 5, 2012 (ADAMS Accession No. ML 121980557). 60. Marion, Alexander, Nuclear Energy Institute, letter dated June 17, 2003, to John Hannon, U.S. Nuclear Regulatory Commission, transmitting Revision 0 of NEI 02-03, "Guidance for Performing a Regulatory Review of Proposed Changes to the Approved Fire Protection Program," June 2003 (ADAMS Accession No. ML031780500). 61. Information Notice 2009-29, "Potential Failure of Fire Water Supply Pumps to Automatically Start due to a Fire." (ADAMS Accession No ML091880072) 62. National Fire Protection Association NFPA 252, "Standard Methods of Fire Tests of Door Assemblies," Quincy, Massachusetts. 63. Underwriters Laboratory, UL 555, "Standard for Safety Fire Dampers," Northbrook, Illinois. 64. National Fire Protection Association NFPA 80, "Standard for Fire Doors and Fire Windows," Quincy, Massachusetts. 65. National Fire Protection Association, NFPA 90A, "Standard for the Installation of Air Conditioning and Ventilating Systems," Quincy, Massachusetts. 66. NRC letter to Wisconsin Electric Power Company (WEPCO), "Review of Draft Safety Evaluation of Conduit Fire Seal Topical Report," dated October 23, 1989 (ADAMS Accession No. ML8911 030114).

-140-67. GL 86-10, Supplement 1, "Fire Endurance Test Acceptance Criteria for Fire Barrier Systems used to Separate Redundant Safe Shutdown Trains within the Same Fire Area." 68. Electric Power Research Institute (EPRI) Technical Report TR-1 006756, "Fire Protection Equipment Surveillance Optimization and Maintenance Guide for Fire Protection Systems and Features," Final Report, Palo Alto, CA, Final Report July 2003. 69. National Fire Protection Association, NFPA 701, "Standard Methods of Fire Tests for Flame Propagation of Textiles and Films," Quincy, Massachusetts. 70. National Fire Protection Association, NFPA 220, "Standard on Types of Building Construction," Quincy, Massachusetts 71. National Fire Protection Association, NFPA 256, "Standard Methods of Fire Tests of Roof Coverings," Quincy, Massachusetts 72. Letter from J.M. Heffley (Constellation Energy Group) to U.S. NRC, "Calvert Cliffs and Nine Mile Point, Units 1 and 2 and R.E. Ginna-Response to Generic Letter 2006-03, Potentially Nonconforming Hemyc and MT Fire Barrier Configurations," dated June 9, 2006. (ADAMS Accession No. ML061650026). 73. NUREG-1924, "Electric Raceway Fire Barrier Systems in U.S. Nuclear Power Plants," (draft report) September 2009 (ADAMS Accession No ML 1 01740246). 74. NEI 05-04, Process for Performing Internal Events PRA Peer Reviews Using the ASME/ANS PRA Standard, Revision 2, Nuclear Energy Institute (NEI), Washington, DC, November 2008. 75. American Society of Mechanical Engineers, ASME RA-Sb-2005, "Addenda to RA-S-2002, Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications," ASME, New York, NY, 2005 76. Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 1, U. S. Nuclear Regulatory Commission, Washington, DC, January 2007 (ADAMS Accession No. ML070240001 ). 77. U.S. Nuclear Regulatory Commission, "Record of Review, Nine Mile Point Nuclear Station, Unit 1, LAR Attachment U-Table U-1 Internal Events PRA Peer Review-Facts and Observations (F&Os)," and "Record of Review, Nine Mile Point Nuclear Station, Unit 1, LAR Attachment V-Tables V-1 and V-2 Fire PRA Peer Review-Facts and Observations (F&Os)," AprilS, 2014 and March 12, 2014 (ADAMS Accession Nos. ML 14122A253 and ML 14122A254, respectively).

-141 -78. Electric Power Research Institute (EPRI) Technical Report TR-1016737, "Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments," Final Report, December 2 0 0 8, E P R I , Palo Alto, CA. 79. NEI 07-12, Fire Probabilistic Risk Assessment (FPRA) Peer Review Process Guidelines, Revision 1, Nuclear Energy Institute (NEI), Washington, DC, June 2010. 80. Memorandum from Patrick Hiland (NRC) to Brian W. Sheron (NRC), "Safety/Risk Assessment Results for Generic Issue 199, 'Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plant,"' September 2, 2010 (ADAMS Accession Nos. ML 100270598, ML 100270639, ML 100270664, ML 100270691, ML 1 00270731, ML 1 00270756). 81. Electric Power Research Institute (EPRI) TR 1002981, "Fire Modeling Guide for Nuclear Power Plant Applications," Electric Power Research Institute, Palo Alto, California, August 2002. 82. Klote, J. H. and J.A. Milke, "Principles of Smoke Management," Chapter 6, Table 6.3, American Society of Heating, Refrigerating and Air-Conditioning Engineers Inc., and Society of Fire Protection Engineers, Atlanta, Georgia, 2002. 83. NMP1 Updated Final Safety Analysis Report (UFSAR), Revision 21, October 26, 2009 (ADAMS Accession No. ML093160257). 84. SFPE Handbook of Fire Protection Engineering, 4th Edition, DiNenno P.J., Editor in-Chief, National Fire Protection Association, Quincy, Massachusetts, 2008. 85. Custer R.L.P., B.J. Meacham, and R.P. Schifiliti, "Design of Detection Systems", Section 4, Chapter 4-1 of the SFPE Handbook of Fire Protection Engineering, 4th Edition, DiNenno, P.J., Editor-in-Chief, National Fire Protection Association, Quincy, Massachusetts, 2008. 86. U.S. Nuclear Regulatory Commission, NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," Chapter 9.5.1, "Fire Protection Program," Revision 3, July 1981, (ADAMS Accession No. ML052350030). 87. Peacock, R.D., W.W. Jones, and P.A. Reneke, "CFAST-Consolidated Model of Fire Growth and Smoke Transport (Version 6), Software Development and Model Evaluation Guide," NIST Special Publication 1086, National Institute of Standards and Technology, Gaithersburg, Maryland, August 2012. 88. Budnick, E.K., D.O. Evans, and H.L. Nelson., "Simplified Fire Growth Calculations," Section 3, Chapter 9, NFPA Fire Protection Handbook, 19th Edition, Cote, A.E., Editor-in-Chief, National Fire Protection Association, Quincy, Massachusetts, 2003.

-142-89. Heskestad, G., "Fire Plumes, Flame Height, and Air Entrainment," Section 2, Chapter 2-1 of the SFPE Handbook of Fire Protection Engineering, 4th Edition, National Fire Protection Association, Quincy, Massachusetts, 2008. 90. Seyler, C.L, "Fire Hazard Calculations for Large, Open Hydrocarbon Fires," Section 3, Chapter 3-10 of the SFPE Handbook of Fire Protection Engineering, 4th Edition, National Fire Protection Association, Quincy, Massachusetts, 2008. 91. U.S. Nuclear Regulatory Commission, EA-12-023, letter to D. Bannister, Omaha Public Power District., "Fort Calhoun Station-NRC Special Inspection Report 05000285120; Finding of Preliminary High Safety Significance" March 12, 2012, (ADAMS Accession No. ML 12072A128). 92 Letter from Christopher Costanzo (NMPNS) to Document Control Desk (NRC), dated May 23, 2014, License Amendment Request Pursuant to 10 CFR 50.90: Adoption of NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants (2001 Edition)-Supplemental Information Regarding Fire PRA Model (TAC No. ME8899) (ADAMS Accession No ML14149A356). Principal Contributors: Jay Robinson, NRR; Harold Barrett, NRR; Stephen Dinsmore, NRR; Naeem Iqbal, NRR; Dennis Andrukat, NRR; Bernard Litkett, NRR; Date: Attachments: J S Hyslop, NRR; Roger Pedersen, NRR; Karl Bohlander, PNNL; William lvans, PNNL; Marc Janssens, CNWRA; Kawshik Das, CNWRA; Robert Fosdick, CNWRA A. Table 3.8 V&V Basis for Fire Modeling Correlations Used at NMP1 B. Table 3.8 V&V Basis for Fire Model Calculations of Other Models Used at NMP1 C. Abbreviations and Acronyms Attachment A: Table 3.8-1, V&V Basis for Fire Modeling Correlations Used at NMP1 Correlation Application at V&V Basis NRC Staff Evaluation of Acceptability NMP1 Flame Height The Flame Height NUREG-1805,

Licensee provided verification of the coding of this correlation (Method of Correlation was Chapter 3, 2004 (Response to FM RAI 02(e), Reference 7). I Heskestad) used to determine (Reference 40)

The correlation is validated in NUREG-1824 and an the vertical authoritative publication of the SFPE Handbook of Fire I extension of the NUREG-1824, Protection Engineering. I flame region as part Volume 3, 2007

Licensee either determined that the correlation was applied of the ZOI (Reference 41) within the validated range reported in NUREG-1824, or calculations. provided justification for cases where the correlation was used SFPE Handbook, outside the validated range reported in NUREG-1824 4th Edition, (Response to FM RAis 03(a) and 03(b), Reference 7). Chapter 2-1 , Heskestad, 2008 Based on its review and the licensee's explanation, the NRC staff (Reference 89) concludes that the use of this correlation in the NMP1 application is acceptable. Plume Centerline The Plume NUREG-1805,

Licensee provided verification of the coding of this correlation Temperature Centerline Chapter 9, 2004 (Response to FM RAI 02(e), Reference 7). (Method of Temperature (Reference 40)

The correlation is validated in NUREG-1824 and an Heskestad) Correlation was authoritative publication of the SFPE Handbook of Fire used to determine NUREG-1824, Protection Engineering. the vertical Volume 3, 2007

Licensee either determined that the correlation was applied separation distance, (Reference 41) within the validated range reported in NUREG-1824, or based on provided justification for cases where the correlation was used temperature, to a SFPE Handbook, outside the validated range reported in NUREG-1824 target in order to 4th Edition, (Response to FM RAis 03(a) and 03(b), Reference 7). determine the Chapter 2-1 , vertical extent of the Heskestad, 2008 Based on its review and the licensee's explanation, the NRC staff ZOI. (Reference 89) concludes that the use of this correlation in the NMP1 application is acceptable.

Attachment A: Table 3.8-1, V&V Basis for Fire Modeling Correlations Used at NMP1 Correlation Application at V&V Basis NRC Staff Evaluation of Acceptability NMP1 Radiant Heat Point Source NUREG-1805,

Licensee provided verification of the coding of this correlation Flux Radiant Heat Flux Chapter 5, 2004 (Response to FM RAI 02(e), Reference 7). (Point Source Correlation was (Reference 40)

The correlation is validated in NUREG-1824 and an Method) used to determine authoritative publication of the SFPE Handbook of Fire the horizontal NUREG-1824, Protection Engineering. separation distance, Volume 3, 2007

Licensee determined that the correlation was applied within based on incident (Reference 41) the validated range reported in NUREG-1824, and when heat flux, to a target outside the validated range, supplemental validation was in order to SFPE Handbook, provided (Response to FM RAI 03(a), Reference 7). determine the 4th Edition, horizontal extent of Chapter 3-10, Based on its review and the licensee's explanation, the NRC staff the ZOI. Beyler, C., 2008 concludes that the use of this correlation in the NMP1 application (Reference 90) is acceptable.

Attachment B: Table 3.8-2, V&V Basis for Other Fire Models and Related Calculations Used at NMP1 Calculation Application at V&V Basis NRC Staff Evaluation of Acceptability NMP1 Hot Gas Layer CFAST (Version 6) NUREG-1824

The modeling technique is validated in NUREG-1824 and Calculations was used to Volume 5 NIST Special Publication 1086 calculate upper and (Reference 41)

Licensee either determined that the model was applied within CFAST Zone lower layer the validated range reported in NUREG-1824, or provided Model Version 6 temperatures, the NIST Special justification for cases where the model was used outside the layer height, and Publication 1086 validated range reported in NUREG-1824 (Response to FM smoke obscuration (Reference 87) RAis 02(b) and 03(c), Reference 7). for various conditions. It was Based on its review and the information provided by the licensee, also used to the NRC staff concluded that the use of model in the NMP1 calculate application is acceptable. abandonment time for the NMP1 MCR. Heat and Smoke Heat and Smoke NUREG-1805

Licensee provided verification of the coding of these models Detection Activation model Chapters 1 0-11 (Response to FM RAI 02(e), Reference 7). I used to determine (Reference 40)

The heat detection modeling technique is validated in Detection whether automatic authoritative publications of the NFPA Fire Protection I Activation Model suppression can be NFPA Fire Handbook and the SFPE Handbook of Fire Protection ' credited after the Protection Engineering. initial target set is Handbook 19th

Licensee stated that the models have been applied damaged. Edition (Reference 88) conservatively (Response to FM RAI 02(a), Reference 7). Based on its review and explanation, the NRC staff concluded SFPE Handbook that the use of these correlations in the NMP1 application is of Fire Protection acceptable. Engineering, 4th Edition (Reference 85) -----------

Attachment B: Table 3.8-2, V&V Basis for Other Fire Models and Related Calculations Used at NMP1 Calculation Application at V&V Basis NRC Staff Evaluation of Acceptability NMP1 Flame Spread The FLASH-CAT NUREG/CR-701 0,

Licensee provided verification of the coding of this model over Horizontal Model was used to Section 9 (Response to FM RAI 02(d), Reference 7). Cable Trays predict the growth (Reference 42)

The modeling technique is validated in NUREG/CR-7010. and spread of a fire FLASH-CAT within a vertical Based on its review and explanation, the NRC staff concluded Model stack of horizontal that the use of this model in the NMP1 application is acceptable. cable trays. -----

AC ADAMS ADS AHJ ANS AOV ASD ASH RAE ASME BTP BWR CAROLFIRE cc CCDP CDF CENG CFAST CFR CFWC CHRISTl FIRE CLERP C02 CPT CRD cs CSR CST CTS CTSRW DC DESIREE-Fire DFP DHR DID EC EdF EDG EEEE EPRI Epsilon (E) ERFBS ERO ERV F&O FAQ FDS FDT FIVE Attachment C: Abbreviations and Acronyms alternating current Agencywide Documents Access and Management System automatic depressurization authority having jurisdiction American Nuclear Society air operated valves Alternate shutdown American Society of Heat, Refrigerating and Air Conditioning Engineers, Inc. American Society of Mechanical Engineers Branch Technical Position boiling-water reactor Cable Response to Live Fire Capability Categories conditional core damage probability core damage frequency Constellation Energy Nuclear Group, LLC consolidated model of fire and smoke transport Code of Federal Regulations cable fires caused by welding and cutting Cable Heat Release, Ignition, and Spread in Tray Installations During Fire conditional large early release probability carbon dioxide control power transformer Control Rod Drive Core Spray Cable spreading room condensate storage tanks Containment Spray Containment Spray Raw Water direct current Direct Current Electrical Shorting in Response to Exposure Fire diesel fire pump Decay Heat Removal defense-in-depth emergency condensers Electricite de France emergency diesel generators existing engineering equivalency evaluation Electric Power Research Institute Non-zero but below truncation limit electrical raceway fire barrier system emergency response organization Electromatic Relief Valves facts and observations frequently asked question fire dynamics simulator fire dynamics tool Fire Induced Vulnerability Evaluation Methodology FLASH-CAT FM FPE FPP FPRA FR FRE GDC Gl GL gpm HEP HFE HGL HRA HRE HRR HVAC lEPRA INV JB KSF LAR LERF LOOP MCA MCB MCR MOV MSO NEI NFPA NIST NMP1 NMPNS No. NPO NPP NPP FIRE MAG NRC NRR NSCA NSPC OMA PAU Flame Spread over Horizontal Cable Trays fire modeling fire protection engineering fire protection program fire probabilistic risk assessment Federal Register fire risk evaluation general design criteria generic issue generic letter gallons per minute human error probability human failure event hot gas layer human reliability analysis high(er) risk evolution heat release rate heating, ventilation, and air conditioning internal events PRA (probabilistic risk assessment) Inventory Control junction box key safety function license amendment request large early release frequency loss-of-offsite power Multi-compartment analysis main control board main control room motor-operated valve multiple spurious operation Nuclear Energy Institute National Fire Protection Association National Institute of Standards and Technology Nine Mile Point Nuclear Station, Unit 1 Nine Mile Point Nuclear Station number non-power operation nuclear power plant Nuclear Power Plant Fire Modeling Analysis Guidelines U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation nuclear safety capability assessment nuclear safety performance criteria operator manual action physical analysis unit PB PCE PCS POS PRA PSA PWR QA RA RAI RCS RES RG AI RI/PB ASP RTI RX soc SE SEA SFP SFPE SICF SR SSA sse SSD TR TS UFSAR V&V VFDR WEPCO ZOI performance-based plant change evaluation primary control station Plant Operational States probabilistic risk assessment probabilistic safety assessment pressurized-water reactor quality assurance recovery action request for additional information Reactor Coolant System Office of Nuclear Regulatory Research Regulatory Guide risk-informed risk-informed, performance-based remote shutdown panel response time index Reactor Vessel shutdown cooling system safety evaluation safety evaluation report Spent Fuel Pool Society of Fire Protection Engineers self-ignited cable fires supporting requirement safe shutdown analysis structures, systems, and components safe shutdown technical/topical report technical specifications Updated Final Safety Analysis Report verification and validation variance from deterministic requirements Wisconsin Electric Power Company zone of influence C. Costanzo such as fire probabilistic risk assessment to demonstrate compliance with the nuclear safety performance criteria. The fire protection license condition in NMP1 's license and TS 6.4 are revised to reflect the use of NFPA 805. To reflect the proper pagination of the license, the amendment includes the license pages 5 through 13. However, only the text of the fire protection license condition, paragraph 2.D.(7) and TS Page 350 of Renewed Facility Operating License, are revised. A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's biweekly Federal Register notice. Sincerely, /RAJ Bhalchandra Vaidya, Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-220

Enclosures:

1. Amendment No. 215 to DPR-63 2. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION: LPL 1-1 r/f PUBLIC RidsAcrsAcnw_MaiiCTR Resource RidsNrrDoriDpr Resource RidsNrrDorllpl1-1 Resource RidsNrrDraAfpb Resource RidsNrrDraAhpb Resource RidsNrrDraApla Resource RidsNrrDssStsb Resource RidsNrrLAKGoldstein Resource RidsNrrPMNinMilePoint Resource RidsRgn1 MaiiCenter Resource SDinsmore, NRR DAndrukat, NRR Blitkett, NRR RPedersen, NRR ADAMS Accession No. ML 14126A003 HBarrett, NRR JRobinson, NRR LFields, NRR Nlqbal, NRR Plain, NRR JSHyslop, NRR *memo dated 05/01/14 & email dated 06/10/14 **email dated 06/19/14 ' OFFICE NRR/DORL/805/PM NRRIDORLILPL 1-1/PM NRRIDORLILPL 1-1 /LA NRR/DRAIAPLAIBC NAME SWall BVaidya KGoldstein HHamzehee* DATE 05/05/14 06/09/14 06/09/14 05/01/14, 06/10/14 OFFICE NRRIDRAIAHPB/BC NRRIDSS/STSB/BC OGC NRRJDORLILPL 1-1 /BC NAME US hoop* REIIiott ** MYoung BBeasley DATE 05/01/14, 06/10/14 06/19/14 6/30/14 6/30/14 OFFICIAL RECORD COPY NRRIDRAIAFPB/BC AKiein* 05/01/14, 06/10/14 NRRIDORL/LPL 1-1 /PM BVaidya 6/30/14

v t e Letter Sequence Approval
TAC:ME8899, (Approved, Closed)
Initiation Request, Request, Request, Request, Request, Request Acceptance Supplement Administration Questions 13 November 2012 9 October 2013 9 October 2013 12 February 2014 Answers 27 February 2013 27 March 2013 30 April 2013 30 April 2013 30 April 2013 15 April 2014 Results Approval, Approval Other: ML14181B086
MONTHYEAR2013-04-30 [Table View]

ML14126A003

Text

SUBJECT:

Dear Mr. Costanzo:

Enclosures:

Attachment:

Enclosure:

1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

6.0 STATE CONSULTATION

7.0 ENVIRONMENTAL CONSIDERATION

8.0 CONCLUSION

1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

6.0 STATE CONSULTATION

7.0 ENVIRONMENTAL CONSIDERATION

8.0 CONCLUSION

Enclosures:

Navigation menu

Search