ML15271A101

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Issuance of Amendment Regarding Transition to a Risk Informed, Performance-Based Fire Protection Program in Accordance with Title 10 of the Code of Federal Regulations Section 50.48(c)
ML15271A101
Person / Time
Site: Ginna Constellation icon.png
Issue date: 11/23/2015
From: Render D
Plant Licensing Branch 1
To: Bryan Hanson
Exelon Nuclear
Render D
References
CAC MF1393
Download: ML15271A101 (180)
v • d • e


Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 November 23, 2015 Mr. Bryan C. Hanson President and Chief Nuclear Officer Exelon Nuclear RE. Ginna Nuclear Power Plant 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

RE. GINNA NUCLEAR POWER PLANT - ISSUANCE OF AMENDMENT REGARDING TRANSITION TO A RISK INFORMED, PERFORMANCE-BASED FIRE PROTECTION PROGRAM IN ACCORDANCE WITH TITLE 10 OF THE CODE OF FEDERAL REGULATIONS SECTION 50.48(c) (CAC NO. MF1393)

Dear Mr. Hanson:

The U.S. Nuclear Regulatory Commission (NRC or the Commission) has issued the enclosed Amendment No. 119 to Renewed Facility Operating License No. DPR-18 for the RE. Ginna Nuclear Power Plant. This amendment changes the Operating License and Technical Specifications in response to your application dated March 28, 2013, as supplemented by letters dated December 17, 2013; January 29, February 28, September 5, September 24, and December 4, 2014; and March 18, June 11, and August 7, 2015.

This amendment adopts a new risk-informed performance-based fire protection licensing basis, which complies with the requirements in Title 1O of the Code of Federal Regulations (10 CFR)

Sections 50.48(a) and 50.48(c), the guidance in Regulatory Guide 1.205, "Risk-Informed Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants," Revision 1, and National Fire Protection Association 805, "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants," 2001 Edition. This amendment request also follows the guidance in Nuclear Energy Institute 04-02, "Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program under 10 CFR 50.48(c),"

Revision 2.

B. Hanson The NRC staff's safety evaluation of the amendments is enclosed. A Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sin{J u_

0 Diane Render, Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-244

Enclosures:

1. Amendment No. 119 to Renewed License No. DPR-18
2. Safety Evaluation cc w/

Enclosures:

Distribution via Listserv

ENCLOSURE 1 AMENDMENT NO. 119 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-18 RE. GINNA NUCLEAR POWER PLANT. LLC EXELON GENERATION COMPANY. LLC RE. GINNA NUCLEAR POWER PLANT DOCKET NO. 50-244

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 RE. GINNA NUCLEAR POWER PLANT. LLC EXELON GENERATION COMPANY. LLC DOCKET NO. 50-244 RE. GINNA NUCLEAR POWER PLANT AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 119 Renewed License No. DPR-18

1. The Nuclear Regulatory Commission (the Commission or the NRC) has found that:

A. The application for amendment filed by the Constellation Energy (the license; now operating as Exelon Generation Company, LLC) dated March 28, 2013, as supplemented by letters dated December 17, 2013; January 29, February 28, September 5, September 24, and December 4, 2014; and March 18, June 11, and August 7, 2015, 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 10 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 Operating License and Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) and Paragraph 2.C.(3) of Renewed Facility Operating License No. DPR-18 is hereby amended to read as follows:

2.C.(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 119, are hereby incorporated in the renewed license.

The licensee shall operate the facility in accordance with the Technical Specifications.

2.C.(3) 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 March 28, 2013, supplemented by letters dated December 17, 2013; January 29, 2014; February 28, 2014; September 5, 2014; September 24, 2014; December 4, 2014; March 18, 2015; June 11, 2015; August 7, 2015; and as approved in the safety evaluation report dated November 23, 2015. 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 x1Q-7 /year (yr) for CDF and o-less than 1x1 8/yr for LERF. 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 3.10); 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 November 23, 2015, 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 and (c)3 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 LAR Attachment S, Table S-2, "Plant Modifications Committed," of Exelon Generation letter dated June 11, 2015, to complete the transition to full compliance with 10 CFR 50.48(c) no later than prior to startup from the second refueling outage greater than 12 months after receipt of the safety evaluation. The licensee shall maintain appropriate compensatory measures in place until completion of these modifications.
3. The licensee shall complete the implementation items listed in LAR Attachment S, Table S-3, "Implementation Items," of Exelon Generation letter dated June 11, 2015, except Implementation Items 9, 15 and 19, by 180 days after NRC approval unless that date falls within a scheduled refueling outage, then implementation will occur 60 days after startup from that scheduled refueling outage.

Implementation Items 9, 15 and 19 are associated with modifications described in Table S-2 and will be completed once the related modifications are installed and validated in the PRA model.

3. This license amendment is effective as of the date of its issuance and shall be implemented as described in the transition license conditions.

FOR THE NUCLEAR REGULATORY COMMISSION

~d~

Travis Tate, Chief Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Operating License and Technical Specifications Date of Issuance: November 23, 2015

ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO. 119 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-18 DOCKET NO. 50-244 Replace the following pages of the Facility Operating License with the attached revised pages.

The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert 3 3 4 4 5 5 6 6 7 7 8 8 9

10 Replace the following page of the Appendix A Technical Specifications with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the areas of change.

Remove Insert 5.4-1 5.4-1

(b) Exelon Generation pursuant to the Act and 10 CFR Part 70, to possess and use four (4) mixed oxide fuel assemblies in accordance with the RG&E's application dated December 14, 1979 (transmitted by letter dated December 20. 1979). as supplemented February 20, 1980, and March 5, 1980; (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 calibration or associated with radioactive apparatus or components; and (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 license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Part 20. Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:

(1) Maximum Power Level Exelon Generation is authorized to operate the facility at steady-state power levels up to a maximum of 1775 megawatts (thermal).

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 119, are hereby incorporated in the renewed license. Exelon Generation shall operate the facility in accordance with the Technical Specifications.

(3) 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 March 28, 2013, supplemented by letters dated December 17, 2013; January 29, 2014; February 28, 2014; September 5, 2014; September 24, 2014; December 4, 2014; March 18, 2015; June 11, 2015; August 7, 2015; and as approved in the safety evaluation report dated November 23, 2015. Except where NRG approval for changes or deviations is required by 10 CFR 50.48(c), and provided no Amendment No. 119

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 1x10-7 /year (yr) for CDF and less than o-1x1 8/yr for LERF. 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.

Amendment No. 119

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 3.10); 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 November 23, 2015, 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 and (c)3 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 LAR Attachment S, Table S-2, "Plant Modifications Committed," of Exelon Generation letter dated June 11, 2015, to complete the transition to full compliance with 10 CFR 50.48(c) no later than prior to startup from the second refueling outage greater than 12 months after receipt of the safety evaluation. The licensee shall maintain appropriate compensatory measures in place until completion of these modifications.

Amendment No. 119

3. The licensee shall complete the implementation items listed in LAR Attachment S, Table S-3, "Implementation Items," of Exelon Generation letter dated June 11, 2015, except Implementation Items 9, 15 and 19, by 180 days after NRC approval unless that date falls within a scheduled refueling outage, then implementation will occur 60 days after startup from that scheduled refueling outage.

Implementation Items 9, 15 and 19 are associated with modifications described in Table S-2 and will be completed once the related modifications are installed and validated in the PRA model.

(4) Deleted (5) Deleted (6) Deleted (7) License Transfer (a) On the closing date of the transfer of the facility, Ginna LLC shall obtain from RG&E the greater of (1) $200,791 ,928 or (2) the amount necessary to meet the minimum formula amount under 10 CFR 50.75 calculated as of the date of closing for decommissioning funding assurance for the facility, and ensure the deposit of such funds into a decommissioning trust for the facility established by Ginna LLC.

(b) The decommissioning trust agreement must be in a form acceptable to the NRC.

(c) Ginna LLC shall take all necessary steps to ensure that the decommissioning trust is maintained in accordance with the application and the requirements of the Order approving license transfer, and shall be consistent with the Safety Evaluation supporting that Order.

(8) Mitigation Strategy 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 Amendment No. 119

(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 (c) Actions to minimize release to include consideration of:
1. Water spray scrubbing
2. Dose to onsite responders (9) Control Room Envelope Habitability Upon implementation of Amendment No. 105 adopting TSTF-448, Revision 3, the determination of control room envelope (CRE) unfiltered air inleakage as required by SR 3.7.9.4, in accordance with TS 5.5.16.c.i and the assessment of CRE habitability as required by 5.5.16.c.ii, shall be considered met. Following implementation:

(a) The first performance of SR 3.7.9.4 in accordance with Specification 5.5.16.c.i shall be within the specified Frequency of 6 years, plus the 18-month allowance of SR 3.0.2, as measured from February 8, 2005, the date of the most recent successful tracer gas test, as-stated in the April 6, 2007 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 5.5.16.c.ii, shall be within 3 years, plus the 9-month allowance of SR 3.0.2 as measured from February 8, 2005, the date of the most recent successful tracer gas test, as stated in-the April 6, 2007 letter response to Generic Letter 2003-01, or within the next 9 months if the time period since the most recent successful tracer gas test is greater than 3 years.

(10) 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. R. E. Ginna Nuclear Power Plant, LLC, GENG, 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 Amendment No. 119

provided to or for the GENG subsidiary licensee under the E.D.F.

International SAS. Support Agreement.

(11) 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. R. E. Ginna Nuclear Power Plant, LLC, GENG, 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 NRG. 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 NRG no later than 30 days after the completion of the proposed transaction and license transfers. Exelon Generation shall inform the NRG in writing no later than 14 days after any funds are provided to or for the licensee under the Exelon Corporation Support Agreement.

(12) 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 GENG is insufficient to cover operating costs.

The existing GENG cash pool arrangement shall be consistent with the representations contained in the 2009 Transfer Application dated January 22, 2009 (ADAMS Accession No. ML090290101). R. E. Ginna Nuclear Power Plant, LLC, GENG, 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 guarantee or cause it to fail to perform, or impair its performance under the parent guarantee without the prior written consent of the NRG.

(13) Within 14 days of the license transfers, Exelon Generation shall submit to the NRG 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.

(14) Within 1O days of the license transfers, Exelon Generation shall submit to the NRG the amended GENG 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 decisionmaking authority over safety, security and Amendment No. 119

reliability without the prior written consent of the Director of the Office of Nuclear Reactor Regulation.

(15) At least half the members of the GENG Board of Directors must be U.S.

citizens.

(16) The GENG Chief Executive Officer, Chief Nuclear Officer, and Chairman of the GENG 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 GENG 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.

(17) GENG will retain its Nuclear Advisory Committee (NAC) composed of U.S. citizens who are not officers, directors, or employees of GENG, EDF Inc., Constellation Nuclear, LLC, or CE Nuclear, LLC. The NAC will report to, and provide transparency to, the NRC and other U.S. governmental agencies regarding foreign ownership and control of nuclear operations.

(18) The NAC 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 GENG. The NAC 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 GENG or any entity to cause Constellation Nuclear, LLC, Exelon Generation, or their parent companies, subsidiaries or successors to modify the NAC report before submission to the NRC. The NAC report shall be made available to the public, with the potential exception of information that meets the requirements for withholding such information from public disclosure under the regulations of 10 CFR 2.390, "Public Inspections, Exemptions, Requests for Withholding."

D. The facility requires an exemption from certain requirements of 10 CFR 50.46(a)(1). This includes an exemption from 50.46(a)(1), that emergency core cooling system (ECCS) performance be calculated in accordance with an acceptable calculational model which conforms to the provisions in Appendix K (SER dated April 18, 1978). The exemption will expire upon receipt and approval of revised ECCS calculations. The aforementioned exemption is authorized by law and will not endanger life property or the common defense and security and is otherwise in the public interest. Therefore, the exemption is hereby granted pursuant to 10 CFR 50.12.

E. Exelon Generation shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27827 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contains Safeguards Information protected under 10 CFR 73.21, is entitled: "R. E. Ginna Amendment No. 119

Nuclear Power Plant Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan," submitted by letter dated May 15, 2006.

Exelon Generation shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The licensee's CSP was approved by License Amendment No. 113 and modified by License Amendment No. 117.

F. The Updated Final Safety Analysis Report supplement, submitted pursuant to 10 CFR 54.21 (d), describes certain future activities to be completed prior to the period of extended operation. Ginna LLC shall complete these activities no later than September 18, 2009, and shall notify the Commission in writing when implementation of these activities is complete and can be verified by NRG inspection.

The Updated Final Safety Analysis Report supplement, as revised, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71 (e)(4) following issuance of this renewed 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 each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.

G. All capsules in the reactor vessel that are removed and tested must meet the test procedures and reporting requirements of ASTM E 185-82 to the extent practicable for the configuration of the specimens in the capsule. Any changes to the capsule withdrawal schedule, including spare capsules, must be approved by the NRG prior to implementation. Any capsules placed in storage must be maintained for future insertion, unless approved by the NRG.

H. This renewed license is effective as of the date of issuance and shall expire at midnight on September 18, 2029.

FOR THE NUCLEAR REGULATORY COMMISSION Original Signed By J. E. Dyer, Director Office of Nuclear Reactor Regulation

Attachment:

Appendix A - Technical Specifications Date of Issuance: May 19, 2004 Amendment No. 119

Procedures 5.4 5.0 ADMINISTRATIVE CONTROLS 5.4 Procedures 5.4.1 Written procedures shall be established, implemented, and maintained covering the following activities:

a. The applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978;
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. Effluent and environmental monitoring;
d. Deleted; and
e. All programs specified in Specification 5.5.

R.E. Ginna Nuclear Power Plant 5.4-1 Amendment No. 119

ENCLOSURE 2 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION FOR TRANSITION TO A RISK-INFORMED, PERFORMANCE-BASED FIRE PROTECTION PROGRAM IN ACCORDANCE WITH 10 CFR 50.48(c)

AMENDMENT NO. 119 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-18 RE. GINNA NUCLEAR POWER PLANT, LLC EXELON GENERATION COMPANY, LLC RE. GINNA NUCLEAR POWER PLANT DOCKET NO. 50-244

Table of Contents

1.0 INTRODUCTION

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

2.0 REGULATORY EVALUATION

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

3.0 TECHNICAL EVALUATION

...................................................................................... 3.1 NFPA 805 Fundamental Fire Protection Program Elements and Design Requirements 3.1.1 Compliance with NFPA 805, Chapter 3 Requirements ....................................... 3.1.1.1 Compliance Strategy - Complies ................................................................ 3.1.1.2 Compliance Strategy - Complies with Clarification ...................................... 3.1.1.3 Compliance Strategy - Complies with Use of EEEEs .................................. 3.1.1.4 Compliance Strategy - Complies with Previous NRC Approval ................... 3.1.1.5 Compliance Strategy - Submit for NRC Approval ....................................... 3.1.1.6 Compliance Strategy - Multiple Strategies .................................................. 3.1.1.7 Chapter 3 Sections not Reviewed ...............................................................

ii 3.1.1.8 Compliance with Chapter 3 Requirements Conclusion ................................ 3.1.2 Identification of the Power Block ......................................................................... 3.1.3 Closure of Generic Letter 2006-03, "Potentially Nonconforming Hemyc' and MT' Fire Barrier Configurations," Issues .................................................................................. 3.1.4 Performance Based Methods for NFPA 805, Chapter 3, Elements ..................... 3.1.4.1 Section 3.3.1.3.4 of NFPA 805, "Control of Ignition Sources" ...................... 3.1.4.2 NFPA 805, Section 3.3.5.1 - Wiring Above Suspended Ceilings ................. 3.1.4.3 NFPA 805, Section 3.3.5.3 - Electrical Cable Construction ......................... 3.1.4.4 NFPA 805, Section 3.3.12(1) - Reactor Coolant Pump Oil Collection ......... 3.1.4.5 NFPA 805, Section 3.5.16 - Dedicated Use of Fire Protection Water Supply 3.1.4.6 NFPA 805, Section 3.2.3(1 ), Fire Protection Systems and Features, Inspection, Testing, and Maintenance Procedures .......................................................................... 3.2 Nuclear Safety Capability Assessment Methods ....................................................... 3.2.1 Compliance with NFPA 805 Nuclear Safety Capability Assessment Methods .... 3.2.1.1 Attribute Alignment -- Aligns ........................................................................ 3.2.1.2 Attribute Alignment -- Aligns with Intent.. ..................................................... 3.2.1.3 Attribute Alignment -- Not in Alignment, but Prior NRC Approval. ................ 3.2.1.4 Attribute Alignment -- Not in Alignment, but No Adverse Consequences ..... 3.2.1.5 Attribute Alignment -- Not in Alignment.. ...................................................... 3.2.1.6 NFPA 805 Nuclear Safety Capability Assessment Methods Conclusion ...... 3.2.2 Maintaining Fuel in a Safe and Stable Condition ................................................ 3.2.3 Applicability of Feed and Bleed .......................................................................... 3.2.4 Assessment of Multiple Spurious Operations ..................................................... 3.2.5 Establishing Recovery Actions ........................................................................... 3.2.6 Conclusion for Section 3.2 .................................................................................. 3.3 Fire Modeling ............................................................................................................ 3.4 Fire Risk Evaluations ................................................................................................. 3.4.1 Maintaining Defense-in-Depth and Safety Margins ............................................. 3.4.1.1 Defense-in-Depth ........................................................................................ 3.4.1.2 Safety Margins ............................................................................................ 3.4.1.3 Defense-in-Depth and Safety Margin Conclusion ........................................ 3.4.2 Quality of the Fire Probabilistic Risk Assessment.. .............................................

iii 3.4.2.1 Internal Events PRA Model ......................................................................... 3.4.2.2 Fire PRA Model ........................................................................................... 3.4.2.3 Fire Modeling in Support of the Development of the Fire Risk Evaluation ... 3.4.2.4 Conclusions Regarding Fire PRA Quality .................................................... 3.4.3 Fire Risk Evaluations .......................................................................................... 3.4.4 Additional Risk Presented by Recovery Actions ................................................. 3.4.5 Risk-Informed or Performance-Based Alternatives to Compliance with NFPA 805- 89 3.4.6 Cumulative Risk and Combined Changes .......................................................... 3.4. 7 Uncertainty and Sensitivity Analyses .................................................................. 3.4.8 Conclusion for Section 3.4 .................................................................................. 3.5 Nuclear Safety Capability Assessment Results ......................................................... 3.5.1 Nuclear Safety Capability Assessment Results by Fire Area .............................. 3.5.1.1 Fire Detection & Suppression Systems Required to Meet the Nuclear Safety Performance Criteria ..................................................................................................... 3.5.1.2 Evaluation of Fire Suppression Effects on Nuclear Safety Performance Criteria.- 96 3.5.1.3 Licensing Actions ........................................................................................ 3.5.1.4 Existing Engineering Equivalency Evaluations ............................................ 3.5.1.5 Variances from Deterministic Requirements ................................................ 3.5.1.6 Recovery Actions ........................................................................................ 3.5.1.7 Recovery Actions Credited for Defense in Depth ......................................... 3.5.1.8 Plant Fire Barriers and Separations ........................................................... - 100 -

3.5.1.9 Electrical Raceway Fire Barrier Systems ................................................... - 100 -

3.5.1.10 Conclusion for Section 3.5.1 ...................................................................... - 101 -

3.5.2 Clarification of Prior NRC Approvals ................................................................. - 101 -

3.5.3 Fire Protection during Non-Power Operational Modes ...................................... - 101 -

3.5.3.1 NPO Strategy and Plant Operating States ................................................. - 102 -

3.5.3.2 NPO Analysis Process .............................................................................. - 103 -

3.5.3.3 NPO Key Safety Functions and SSCs Used to Achieve Performance ....... - 105 -

3.5.3.4 NPO Pinch Point Resolutions and Program Implementation ..................... - 108 -

3.5.4 Conclusion for Section 3.5 ................................................................................ - 110 -

iv 3.6 Radioactive Release Performance Criteria .............................................................. - 110 -

3.6.1 Method of Review ............................................................................................ - 110 -

3.6.2 Scope of Review .............................................................................................. - 111 -

3.6.3 Identification of Plant Areas Containing Radioactive Materials and Providing Containment during Fire Fighting Operations ................................................................. - 112 -

3.6.4 Fire Pre-Plans .................................................................................................. - 112 -

3.6.5 Gaseous Effluent Controls ............................................................................... - 113 -

3.6.6 Liquid Effluent Controls .................................................................................... - 114 -

3.6. 7 Fire Brigade Training Materials ........................................................................ - 114 -

3.6.8 Conclusions ..................................................................................................... - 115 -

3.7 NFPA 805 Monitoring Program ............................................................................... - 115 -

3.7.1 Conclusion for Section 3.7.1 ............................................................................. - 117 -

3.8 Program Documentation, Configuration Control, and Quality Assurance ................. - 117 -

3.8.1 Documentation ................................................................................................. - 119 -

3.8.2 Configuration Control ....................................................................................... - 120 -

3.8.3 Quality .............................................................................................................. - 120 -

3.8.3.1 Review ...................................................................................................... - 120 -

3.8.3.2 Verification and Validation (V&V) .............................................................. - 121 -

3.8.3.3 Limitations of Use ...................................................................................... - 123 -

3.8.3.4 Qualification of Users ................................................................................ - 127 -

3.8.3.5 Uncertainty Analysis .................................................................................. - 129 -

3.8.3.6 Conclusion for Section 3.8.3 ..................................................................... - 131 -

3.8.4 Fire Protection Quality Assurance Program ...................................................... - 131 -

3.8.5 Conclusion for Section 3.8 ............................................................................... - 132 -

4.0 FIRE PROTECTION LICENSE CONDITION ........................................................... - 132 -

5.0

SUMMARY

.............................................................................................................. - 135 -

6.0 STATE CONSULTATION

........................................................................................ -135-

7.0 ENVIRONMENTAL CONSIDERATION

................................................................... - 136-

8.0 CONCLUSION

........................................................................................................ - 136 -

9.0 BIBLIOGRAPHY ..................................................................................................... - 137 -

v ATTACHMENTS Attachment A: Table 3.8-1, V&V Basis for Fire Modeling Correlations Used at Ginna ..................................................................................................... -A1 - : Table 3.8-2, V&V Basis for Other Fire Models and Related Correlations Used at Ginna ..................................................................... - B1 -

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 AMENDMENT NO. 119 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-18 FOR THE TRANSITION TO A PERFORMANCE-BASED FIRE PROTECTION PROGRAM IN ACCORDANCE WITH 10 CFR 50.48(c)

RE. GINNA NUCLEAR POWER PLANT, LLC EXELON GENERATION COMPANY. LLC RE. GINNA NUCLEAR POWER PLANT DOCKET NO. 50-244

1.0 INTRODUCTION

1.1 Background The U.S. Nuclear Regulatory Commission (NRC or the Commission) started developing fire protection requirements in the 1970s. In 1976, the NRC published comprehensive fire protection guidelines in the form of Branch Technical Position (BTP) Auxiliary and Power Conversion Systems Branch (APCSB) 9.5-1, "Guidelines for Fire Protection for Nuclear Power Plants" (Reference 1) and Appendix A to BTP APCSB 9.5-1, "Guidelines for Fire Protection for Nuclear Power Plants Docketed Prior to July 1, 1976" (Reference 2). Subsequently, the NRC performed fire protection reviews for the operating reactors, and documented the results in safety evaluations (SEs) or supplements to SEs. In 1980, to resolve issues identified in those reports, the NRC amended its regulations for fire protection in operating nuclear power plants (NPPs) 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 Title 10 of the Code of Federal Regulations (10 CFR) Section 50.48, "Fire Protection," and Appendix R, "Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979," to 10 CFR Part 50. Section 50.48(a)(1) of 10 CFR 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 (FPP); identify the positions responsible for the program and the authority delegated to those positions; and 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 section (a)(1 ), including administrative controls and personnel requirements for fire prevention and manual suppression activities; 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, performance-based (RI/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 3),

which describes a methodology for establishing fundamental FPP design requirements and elements, determining required fire protection systems and features, applying performance-based (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.

NRC Regulatory Guide (RG) 1.205, "Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants," Revision 1 (Reference 4), states:

On March 26, 1998, the NRC staff sent to the Commission SECY-98-058, "Development of a Risk-Informed, Performance-Based Regulation for Fire Protection at Nuclear Power Plants" (Reference 5), in which it proposed to work with the 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 6), the NRC staff requested and received Commission approval to proceed with rulemaking to permit operating reactor licensees to adopt an NFPA standard 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 though 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 (PSAs), and fire modeling (FM)

calculations to show that the criteria are met. Chapter 4 of NFPA 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 (NSPC) shall be maintained free of fire damage by a single fire.

RG 1.205 also states 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 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 supplementation 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

[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)," Revision 2 (Reference 7)].

RG 1.205 provides the NRC staffs position on NEI 04-02, Revision 2 (Reference 7), 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 an RI/PB FPP.

RG 1.205 endorses the guidance of NEI 04-02, Revision 2, subject to certain exceptions, as providing methods acceptable to the NRC staff for adopting an FPP consistent with the 2001 Edition of NFPA 805 and 10 CFR 50.48(c).

Accordingly, Constellation Energy (GENG, the licensee, now operating as Exelon (Reference 8),

requested license amendments to allow the licensee to maintain the R.E. Ginna Nuclear Power Plant (Ginna) FPP in accordance with 10 CFR 50.48(c) and change the Renewed Facility Operating Licenses and Technical Specifications (TSs) accordingly.

1.2 Requested Licensing Action By application dated March 28, 2013 (Reference 9), as supplemented by letters dated December 17, 2013 (Reference 10), January 29, 2014 (Reference 11 ), February 28, 2014 (Reference 12), September 5, 2014 (Reference 13), September 24, 2014 (Reference 14),

December 4, 2014 (Reference 15), March 18, 2015 (Reference 16), June 11, 2015 (Reference 17), and August 7, 2015 (Reference 18), the licensee requested license amendments to transition the Ginna FPP from 10 CFR 50.48(b) to 10 CFR 50.48(c), based upon compliance

with NFPA 805. The supplemental letters were in response to the NRC staff requests for additional information (RAls) dated September 12, 2013 and October 9, 2013 (Reference 19),

September 19, 2014 (Reference 20), October 27, 2014 (Reference 21), February 3, 2015 (Reference 22), July 10, 2015 (Reference 23), and July 29, 2015 (Reference 24). The licensee's supplemental letters dated December 17, 2013; January 29, February 28, September 24, December 4, 2014; and March 18, June 11, and August 7, 2015, 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 determination as published in the FR on November 4, 2014 (79 FR 65427).

The licensee requested amendments to the Ginna 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 (LB) established in accordance with the Fire Protection SE dated February 14, 1979 (Reference 25), and SE Supplements dated December 17, 1980 (Reference 26),

February 6, 1981 (Reference 27), June 22, 1981 (Reference 28), February 27, 1985 (Reference 29), and March 21, 1985 (Reference 30), 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 Ginna is referred to as RI/PB throughout this SE.

In its license amendment request (LAR), the licensee 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 identified any orders and license conditions that must be revised or superseded, and provided the necessary revisions to the plant's TSs and bases, as required by 10 CFR 50.48(c)(3)(i);
2. The licensee completed its implementation of the methodology in Chapter 2, "Methodology," of NFPA 805 (including all required evaluations and analyses),

and the NRC staff approved the licensee's modified fire protection plan, 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 LB, as well as revisions to the TSs that address this change to the current FPP basis.

Sections 2.4.2 and 4.0 of the SE discuss the license condition in detail and Section 2.4.3 of the SE discusses the TS changes.

2.0 REGULATORY EVALUATION

10 CFR Section 50.48, "Fire protection," provides the NRC requirements for NPP fire protection.

Section 50.48 includes specific requirements for requesting approval for an RI/PB FPP based on the provisions of NFPA 805 (Reference 3). 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 paragraph (b) of this section [10 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 [10 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 Fed. Reg. 33536, 33548 (June 16, 2004), which states:

This paragraph [1 O CFR 50.48(c)(3)(ii)] 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 identified orders, license conditions, and the technical specifications 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 Chapter 3 of NFPA 805, "Fundamental Fire Protection Program and Design Elements," may do so by submitting a LAR 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 (DID) (fire prevention, fire detection, fire suppression, mitigation, and post-fire safe shutdown (SSD) capability).

Alternatively, licensees may choose to use RI or PB alternatives to comply with NFPA 805 by submitting an LAR in accordance with 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 DID (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 a LAR for NRC review and approval in order to adopt an RI/PB FPP, a licensee may also submit additional elements of its FPP for which it wishes to receive specific NRC review and approval, as set forth in Regulatory Position (RP) C.2.2.1 of RG 1.205 (Reference 4). 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. RGs are not substitutes for regulations and compliance with them is not required. Methods and solutions that differ from those set forth in RGs 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 DID philosophy that the NRC's fire protection objectives are satisfied. NFPA 805, Section 1.2, "Defense-in-Depth," states that:

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; and (3) Providing an adequate level of fire protection for SSCs important to safety, so that a fire that is not promptly extinguished will not prevent essential plant safety functions from being performed.

2.1 Other Applicable Regulations The following regulations address fire protection:

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.

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(c) incorporates NFPA 805 (2001 Edition) (Reference 3) 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 a voluntary alternative to the requirements in 10 CFR 50.48(b) and Appendix R, "Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979," to 10 CFR Part 50, or the specific plant fire protection license condition for plants licensed to operate after January 1, 1979.
  • 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 Guidance The NRC staff review also relied on the following additional codes, RGs, and standards:

  • RG 1.205, Revision 1, issued December 2009 (Reference 4), which provides guidance for use in complying with the requirements that the NRC promulgated for RI/PB FPPs that comply with 10 CFR 50.48 and the referenced 2001 Edition of the NFPA standard. It endorses portions of NEI 04-02, Revision 2 (Reference 7), where it was found to provide methods acceptable to the NRC for implementing NFPA 805 and complying with 10 CFR 50.48(c). The RPs in RG 1.205 Section C include clarification of the guidance provided in NEI 04-02, as well as NRC exceptions to the guidance. RG 1.205 sets forth RPs, emphasizes certain issues, clarifies the requirements of 10 CFR 50.48( c) and NFPA 805, clarifies the guidance in NEI 04-02, and modifies the NEI 04-02 guidance where required. Should a conflict occur between NEI 04-02 and this RG, the RPs in RG 1.205 govern. This RG also indicates that Chapter 3 of NEI 00-01, "Guidance for Post-Fire Safe Shutdown Circuit Analysis," Revision 2, issued May 2009, when used in conjunction with NFPA 805 and the RG, provides on acceptable approach to circuit analysis for a plant implementing an FPP under 10 CFR 50.48(c).
  • The 2001 edition of NFPA 805, "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants" (Reference 3), which specifies the minimum fire protection requirements for existing light water NPPs during all phases of plant operations, including shutdown, degraded conditions, and decommissioning. 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)" (Reference 7), which 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) provide direction and clarification for adopting NFPA 805 as an acceptable approach to fire protection, consistent with 10 CFR 50.48 (c); and (2) 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 31 ), provides a deterministic methodology for performing post-fire safe shutdown analysis (SSA). In addition, NEI 00-01 includes information on RI methods (when allowed within a plant's LB) that may be used in conjunction with the deterministic methods for resolving circuit failure issues related to multiple spurious operations (MSO). The RI method is intended for application by licensees to determine the risk significance of identified circuit failure issues related to MSOs. In RG 1.205, the NRC staff indicated that Chapter 3 of NEI 00-01, when used in conjunction with NFPA 805 and RG 1.205, provides an acceptable approach to circuit analysis for a plant implementing an FPP under 10 CFR 50.48(c).

Revision 2, issued May 2011 (Reference 32), which provides the NRC staff's recommendations for using risk information in support of licensee-initiated LB changes to a NPP that require such review and approval. The guidance provided does not preclude other approaches for requesting LB changes.

Rather, RG 1.174 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 determined to be acceptable for analyzing issues associated with proposed changes to a plant's LB 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 33), which provides guidance to licensees for use in determining the technical adequacy of the base probabilistic risk assessment (PRA) used in an RI 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 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 decision-making processes.

  • American Society of Mechanical Engineers (ASME)/American Nuclear Society (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" (Reference 34), which provides guidance PRAs used to support RI decisions for commercial light water reactor NPPs 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., sabotage). The standard applies to PRAs used to support applications of RI decision making related to design, licensing, procurement, construction, operation, and maintenance.
  • RG 1.189, "Fire Protection for Operating Nuclear Power Plants," Revision 2, issued October 2009 (Reference 35), which 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 NRC staff would consider acceptable for NPPs.
  • NUREG-0800, Section 19.1, "Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 3, issued September 2012 (Reference 37), which provides guidance for NRC staff evaluations of the technical adequacy of a licensee's PRA results when used to request RI 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 38), which provides guidance for the NRC staff evaluations of the risk information used by a licensee to support permanent, RI changes to the LB for the plant.

  • NUREG/CR-6850, "EPRl/NRC-RES Fire PRA Methodology for Nuclear Power Facilities," Volumes 1 and 2, and Supplement 1 (Reference 39) (Reference 40)

(Reference 41) , which presents a compendium of methods, data and tools to perform a fire probabilistic risk assessment 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 state-of-art Fire PRA (FPRA) methodology. Both RES and EPRI have provided specialists in fire risk analysis, 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 RI analyses related to the plant FPP. Volume 1, the Executive Summary, provides general background and overview information including both programmatic and technical, and project insights and conclusions. Volume 2 provides the detailed discussion of the recommended approach, methods, data and tools for conduct of an FPRA.

  • Memorandum from Richard P. Correia, RES, to Joseph G. Giitter, NRR, titled "Interim Technical Guidance on Fire-Induced Circuit Failure Mode Likelihood Analysis," dated June 14, 2013 (Reference 42) notes that, based on new experimental information documented in NUREG/CR-6931, "Cable Response to Live Fire (CAROLFIRE)," issued April 2008 (Reference 43), and NUREG/CR-7100, "Direct Current Electrical Shorting in Response to Exposure Fire (DESIREE-Fire): Test Results," issued April 2012 (Reference 44), the reduction in hot short probabilities for circuits provided with control power transformers (CPTs) identified in NUREG/CR-6850 cannot be repeated in experiments and, therefore, may be too high and should be reduced.
  • NUREG-1805, "Fire Dynamics Tools (FDTs): Quantitative Fire Hazard Analysis Methods for the U.S. Nuclear Regulatory Commission Fire Protection Inspection Program" (Reference 45) which provides quantitative methods, known as "Fire Dynamics Tools," to assist regional fire protection inspectors in performing fire hazard analysis. The FDTs are intended to assist fire protection inspectors in performing RI evaluations of credible fires that may cause critical damage to essential SSD equipment, as required by the new reactor oversight process defined in the NRC's inspection manual.
  • NUREG-1824, "Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications," Volumes 1 through 7 (Reference 46), which provide technical documentation regarding the predictive capabilities of a specific set of fire models for the analysis of fire hazards in NPP 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. FDTs developed by NRC (Volume 3),
2. Fire-Induced Vulnerability Evaluation Methodology - Revision 1 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 (Volume 6),

and

5. The computational fluid dynamics model 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-1921, "EPRl/NRC-RES Fire Human Reliability Analysis Guidelines-Final Report" (Reference 47), which 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 for human failure events (HFEs) following the fire-induced initiating events of a FPRA. The report builds on existing human reliability analysis methods, and is intended primarily for practitioners conducting a fire HRA to support a FPRA.
  • Generic Letter (GL) 2006-03. "Potentially Nonconforming Hemyc and MT Fire Barrier Configurations" (Reference 48), which requested that licensees evaluate its facilities to confirm compliance with the existing applicable regulatory requirements in light of the information provided in this GL and, if appropriate, take additional actions.

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 sections to which each FAQ is referenced.

Table 2.3-1: NFPA 805 Frequently Asked Questions FAQ# FAQ Title and Summary Reference SE Section 06-0022 "Electrical Cable Flame Propagation Tests" (Reference 49) 3.1.4.3

  • This FAQ provides a list of acceptable electrical cable flame propaQation tests.

07-0030 "Establishing Recovery Actions" (Reference 50) 3.2.5 3.4.4

  • This FAQ provides an acceptable process for 3.4.8 determining the recovery actions (RAs) for NFPA 3.5.1.7 805 Chapter 4 compliance. The process includes:
  • Differentiation between RAs and activities in the main control room or at primary control station(s).
  • Evaluate the additional risk presented by the use of RAs.
  • Evaluate the feasibility of the identified RAs .
  • Evaluate the reliability of the identified RAs .

07-0035 "Bus Duct Counting Guidance for High Energy Arcing (Reference 51) 3.4.2.2 Faults"

  • This FAQ provides clarification regarding bus duct counting guidance for high energy arcing faults.

07-0038 "Lessons Learned on Multiple Spurious Operations (Reference 52) 3.2.4 (MSOs)" 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 FPRA and Nuclear Safety Capability Assessment (NSCA) to include MSOs of concern.
  • Step 4 - Evaluate for NFPA 805 compliance .
  • Step 5 - Document the results .

FAQ# FAQ Title and Summary Reference SE Section 07-0039 "Incorporation of Pilot Plant Lessons Learned - Table (Reference 54) 3.2.1 B-2" 3.2.1.6

  • This FAQ provides additional detail for the comparison of the licensee's SSD strategy to the endorsed industry guidance, NEI 00-01 "Guidance for Post-Fire Safe Shutdown Circuit Analysis,"

Revision 1 (Reference 53). In short, the process has the licensees:

  • Assemble industry and plant-specific documentation;*
  • Determine which sections of the guidance are applicable;
  • Compare the existing SSD methodology to the applicable guidance; and
  • Document any discrepancies .

07-0040 "Non-Power Operations (NPOs) Clarifications" (Reference 55) 3.5.3 3.5.3.1

  • This FAQ clarifies an acceptable NFPA 805 NPO 3.5.3.2 program. The process includes: 3.5.3.3
  • Selecting NPOs equipment and cabling. 3.5.4
  • Evaluation of NPOs Higher Risk Evolutions (HRE).
  • Analyzing NPO Key Safety Functions (KSFs) .
  • Identifying plant areas to protect or "pinch points" during NPOs HREs and actions to be taken if KSFs are lost.

08-0048 "Revised Fire Ignition Frequencies" (Reference 56) 3.4.7

  • This FAQ provides an acceptable method for using updated fire ignition frequencies in the licensee's FPRA. The method involves the use of sensitivity studies when the updated fire ignition frequencies are used.

08-0053 "Kerite-FR Cable Failure Thresholds" (Reference 57) 3.4.2.2

  • This FAQ provides guidance regarding the damage threshold for Kerite-FR cable.

FAQ# FAQ Title and Summary Reference SE Section 08-0054 "Compliance with Chapter 4 of NFPA 805" (Reference 58) 3.4.3 3.4.4

  • This FAQ provides an acceptable process to 3.4.8 demonstrate Chapter 4 compliance for transition: 3.5.1.4
  • Step 1 - Assemble documentation
  • Step 2 - Document Fulfillment of NSPC
  • Step 3 - Variance From Deterministic Requirements (VFDR) Identification, Characterization, and Resolution Considerations
  • Step 4 - PB Evaluations
  • Step 5 - Final VFDR Evaluation
  • Step 6 - Document Required Fire Protection Systems and Features 09-0056 "Radioactive Release Transition" (Reference 59) 3.6.1
  • 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 "Monitoring Program" (Reference 60) 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) (Reference 61) 2.4.4 Content"

  • This FAQ provides the necessary level of detail for the transition of the fire protection sections within the UFSAR.

FAQ# FAQ Title and Summary Reference SE Section 13-0004 "Clarifications on Treatment of Sensitive Electronics" (Reference 62) 3.4.2.2

  • This FAQ provides supplemental guidance for application of the damage criteria provided in Sections 8.5.1.2 and H.2 of NUREG/CR-6850 for solid-state components.

13-0005 "Cable Fires Special Cases: Self-Ignited and Caused (Reference 63) 3.4.2.2 by Welding and Cutting"

  • This FAQ provides additional guidance for detailed FPRA/FM concerning self-ignited cable fires and cable fires caused by welding and cutting.

13-0006 "Modeling Junction Box Scenarios in a Fire PRA" (Reference 64) 3.4.2.2

  • This FAQ provides a definition for junction boxes that allow the characterization and quantification of junction box fire scenarios in plant physical access units (PAUs) requiring detailed FPRA/FM analysis and also describes a process for quantifying the risk associated with junction box fire scenarios in such plant locations.

14-0008 "Main Control Board [MCB] Treatment" (Reference 65) 3.4.2.2

  • This FAQ clarifies the MCB definition and gives guidance on application of the frequencies in Appendix L to NUREG/CR-6850.

2.4 Orders. License Conditions. and Technical Specifications Paragraph 50.48(c)(3)(i) of 10 CFR states, in part, 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 TSs and the bases thereof."

2.4.1 Orders The NRC staff reviewed Section 5.2.3 of the LAR, "Orders and Exemptions" and Attachment 0 of the LAR, "Orders and Exemptions," with regard to NRC-issued Orders pertinent to Ginna 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 physical protection requirements,

security orders, and adherence to those commitments applicable to Ginna are maintained. The licensee discussed the affected orders and exemptions in Attachment 0 of the LAR.

The licensee requested that five exemptions be rescinded and that none of the underlying engineering evaluations for the exemptions be transitioned to NFPA 805. The licensee also determined that no orders need to be superseded or revised to implement a FPP that complies with 10 CFR 50.48(c).

The licensee's review 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 Ginna are maintained. The NRC staff accepts the licensee's determination that five exemptions should be rescinded and that none of the underlying engineering evaluations for the exemptions are transitioned to NFPA 805 as listed in Attachment K of the LAR, "Existing Licensing Action Transition," and that no orders need to be superseded or revised to implement NFPA 805 at Ginna. Section 2.5 of the SE is the NRC staff's detailed evaluation of the exemptions being rescinded.

In addition, the licensee also performed a specific review of the license amendments 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 Ginna. The licensee's review of this regulation and the related license amendments 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. These strategies include 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.

2.4.2 License Conditions The NRC staff reviewed Section 5.2.1 of the LAR, "License Condition Changes," and Attachment M of the LAR, "License Condition Changes," regarding changes the licensee seeks to make to the Ginna 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 Ginna fire protection license condition, for consistency with the format and content guidance described in RP C.3.1 of RG 1.205, Revision 1, and with the proposed plant modifications identified in the LAR.

The revised license condition would provide 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 Section 2.4, "Engineering Analyses"; Section 2.4.3, "Fire Risk Evaluations"; and Section 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, as-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 license condition would provide structure and detailed criteria to allow 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 the SE. The license condition also references the plant-specific modifications and associated implementation schedules that must be accomplished at Ginna to complete transition to NFPA 805 and comply with 10 CFR 50.48(c). The license condition also 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 in Section 2.7 of the SE.

Section 4.0 of the SE provides the NRC staff's review of the proposed Ginna FPP license condition.

2.4.3 Technical Specifications The NRC staff reviewed Section 5.2.2 of the LAR, "Technical Specifications" and Attachment N of the LAR, "Technical Specification Changes," with regard to proposed changes to the Ginna TSs that are being revised or superseded during the NFPA 805 transition process. According to the LAR, the licensee conducted a review of the Ginna TSs to determine which, if any, TS sections will be impacted by the transition to a RI/PB FPP based on the regulations in 10 CFR 50.48(c). The NRC staff found that the licensee had previously requested, and obtained NRC approval for, removal of fire protection requirements from the Ginna TSs and established controls in the approved Fire Protection Plan in Amendment 49 (Reference 66).

Although the licensee previously removed fire protection requirements from the Ginna TSs, the licensee identified one change to the TSs that involved the modification of Section 5.4, "Procedures."

Section 5.4 of the TSs require that written procedures and administrative policies be established, implemented and maintained covering certain activities. The licensee proposed the deletion of the text in TS 5.4.1 "d, Fire Protection Program Implementation" and replacing it with the word "Deleted." The licensee stated that the change is adequate for adoption of the new fire protection licensing basis since the requirement for establishing, implementing and maintaining fire protection procedures is contained in regulation (10 CFR 50.48(a) and 10 CFR 50.48(c)), as specifically outlined in Section 3.2.3 of NFPA 805, "Procedures." Based on the information provided by the licensee, the NRC staff concludes that the proposed change to the TSs are acceptable because the TS being changed is in Administrative Controls and would be redundant to the NFPA 805 requirement to establish FPP procedures. 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 LB. Changes to fire protection Administrative Controls are controlled by the proposed fire protection license condition. (See Section 4.0 of the SE.)

2.4.4 Updated Final Safety Analysis Report The NRC staff reviewed the LAR and found that Figure 4-9 in the LAR, "NFPA 805 Planned Post-Transition Documents and Relationships," indicates that a revised UFSAR will be developed as a post-transition document representing the revised license condition. The licensee further stated that after the approval of the LAR, in accordance with 10 CFR 50.71 (e),

the Ginna UFSAR will be revised and the format and content will be consistent with NEI 04-02 and FAQ 12-0062, "Updated Final Safety Analysis Report" (Reference 61).

The NRC staff concludes that the licensee's method to update the UFSAR is acceptable because the licensee updates its UFSAR in accordance with 10 CFR 50.71(e) and stated that the format and content of the update will be consistent with the guidance provided in NEI 04-02.

2.5 Rescission of Exemptions Since Ginna was licensed to operate on September 19, 1969, under the provisional operating license, with commercial operation beginning in July, 1970, the Ginna FPP is based on compliance with 10 CFR 50.48(a); GDC 3 in 10 CFR 50, Appendix A; 10 CFR 50, Appendix R; and Appendix A to BTP APCSB 9.5-1, "Guidelines for Fire Protection for Nuclear Power Plants Docketed Prior to July 1, 1976," and the Ginna fire protection license condition.

The NRC staff reviewed Section 5.2.3, "Orders and Exemptions," Attachment 0, and Attachment K of the LAR with regard to previously-approved exemptions to Appendix R to 10 CFR Part 50, which the transition to a FPP LB in conformance with NFPA 805 will supersede. These exemptions will no longer be required since upon approval of the RI/PB FPP in accordance with NFPA 805, Appendix R will not be part of the LB for Ginna.

The licensee previously requested and received NRC approval for five exemptions from 10 CFR Part 50, Appendix R. These exemptions were discussed in detail in Attachment K of the LAR. The licensee requested that the exemptions be rescinded and that none of the underlying engineering evaluations for the exemptions be transitioned to the new LB under 10 CFR 50.48(a) and 50.48(c), as previously approved (Section 2.2.7 of NFPA 805) and compliant with the new regulation.

Disposition of Appendix R exemptions may follow two different paths during transition to NFPA 805:

  • The exemption was found to be unnecessary because the underlying condition was evaluated using RI/PB methods (FM and/or FRE) and found to be acceptable and no further actions by the licensee are necessary.
  • 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 was evaluated using RI/PB methods and found to be acceptable with no further actions because the philosophy of DID and sufficient safety margins are maintained:

  • Exemption from the Appendix R, Section 111.G.3 requirement for area-wide fixed fire suppression in fire areas ABBM, CC, BR1 B, EDG1 B, SH, and ABI.
  • Exemption from the Appendix R, Section 111.G.2 requirement for 20 feet separation of intervening combustibles from redundant safe shutdown systems in fire area ABI.
  • Exemption from the Appendix R, Section 111.G.2 requirement for 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> fire-rated barriers between fire area CT and adjoining areas.
  • Exemption from the Appendix R, Section 111.G.2 requirement for 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> fire-rated barriers between the emergency diesel generator feeds in fire area EDG1 B.
  • Exemption from the Appendix R, Section 111.G.2 requirement for 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> fire-rated barriers for the structural steel part of a fire barrier in fire Area Battery Room 1A (BR1A) and Battery Room 1B (BR1 B).

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 Section 2.4.2 of the SE, changes to the approved FPP must be evaluated by the licensee to ensure that they are acceptable.

Section 2.2.9 of NFPA 805, "Plant Change Evaluation," states that:

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 [Section] 2.4.4 to ensure that the public risk associated with fire-induced nuclear fuel damage accidents is low and that adequate defense-in-depth and safety margins are maintained.

Section 2.4.4 of NFPA 805, "Plant Change Evaluation," states 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 Section 4.7.2 of the LAR, "Compliance with Configuration Control Requirements in Sections 2.7.2 and 2.2.9 of NFPA 805," for compliance with the NFPA 805

PCE process requirements to address potential changes to the NFPA 805 RI/PB FPP after implementation is completed. The licensee will develop a change process that is based on the guidance provided in Sections 2.2(h), 2.2.9, 2.4.4, A.2.2(h), A.2.4.4, and D.5 of NFPA 805; Section 5.3, "Plant Change Process" (Reference 7) and Appendices B, I, and J of NEI 04-02; and RPs 2.2.4, 3.1, 3.2, and 4.3 of RG 1.205 (Reference 4).

Section 4. 7 .2 of the LAR states that the PCE process consists of four steps:

1. Defining the change,
2. Performing the preliminary risk screening,
3. Performing the risk evaluation, and
4. Evaluating the acceptance criteria.

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 LB. The licensee also stated that the baseline is defined as that plant condition or configuration that is consistent with the design basis and LB, and that the changed or altered condition or configuration that is not consistent with the design basis and LB is defined as the proposed alternative.

The licensee stated that once the definition of the change is established, a screening is then performed to identify and resolve minor changes to the FPP and the screening is consistent with fire protection regulatory review processes currently in place. The licensee further stated that the screening process is modeled after NEI 02-03, "Guidance for Performing a Regulatory Review of Proposed Changes to the Approved Fire Protection Program," June 2003 (Reference 67), and that the process will address most administrative changes (e.g., changes to the combustible control program, organizational changes, etc.).

The licensee stated that once the screening process is completed, it is followed by engineering evaluations that might include FM and risk assessment techniques and the results of these evaluations are then compared to the acceptance criteria. The licensee further stated that changes that satisfy the acceptance criteria in Section 2.4.4 of NFPA 805 and the fire protection license condition (see Attachment M of 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 require that the resultant change in core damage frequency (CDF) and LERF be consistent with the fire protection license condition, and the acceptance criteria also includes 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 and that, in certain circumstance, an initial evaluation in the development of the risk assessment may be a simplified analysis using bounding assumptions, provided the use of such assumptions does not unnecessarily challenge the acceptance criteria.

The licensee stated that the PCEs are assessed for acceptability using the ~CDF (change in core damage frequency) and ~LERF (change in large early release frequency) criteria from the license conditions and that the proposed changes are also assessed to ensure they are consistent with the DID philosophy and sufficient safety margins were maintained.

The licensee stated its FPP configuration is defined by the program documentation and, to the greatest extent possible, the existing configuration control processes for modifications, calculations and analyses will be utilized to maintain configuration control of the FPP documents. The licensee further stated that the configuration control procedures that govern the various documents and databases will be revised to reflect the new NFPA 805 licensing bases requirements. The licensee included the action to update its design engineering and configuration control procedure to address the NFPA 805 change evaluation process and to reflect current plant configurations in implementation item 5 of Table S-3 in Attachment S of the LAR. The NRC staff concludes that this action is acceptable because it will result in compliance with NFPA 805 and it would be required by the proposed license condition.

The licensee stated that several NFPA 805 document types such as: NSCA supporting information, non-power 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. The licensee further stated that the new procedures will be modeled after the existing processes for similar types of documents and databases, and system level design basis documents will be revised to reflect the NFPA 805 role that the system components now play. The licensee included the action to develop new control procedures and processes and to revise design basis documents to reflect the new NFPA 805 licensing bases requirements in implementation item 4 of Table S-3 in Attachment S of the LAR. The NRC staff concludes that this action is acceptable because it will result in compliance with NFPA 805 and it would be required by the proposed license condition.

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. 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, SSD/NSCA, FPRA, etc.) to ascertain the program impacts, if any. 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: Comply with NFPA 805, Chapter 2 and Section 4.2.3 requirements; or
  • PB Approach: Use the NFPA 805 change process developed in accordance with NEI 04-02, RG 1.205, and the 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 could 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 1.174 (Reference 32). These both require 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.

The licensee stated that it evaluated potential impact on the FPP during transition to NFPA 805 through the implementation of requirements contained in its procedure that controls permanent changes to the facility. The licensee further stated that the procedure contains evaluation criteria that must be reviewed to determine if the design change can have potential FPP impact.

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 Ginna 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 concludes that the licensee's PCE process is acceptable because it meets the guidance in NEI 04-02, Revision 2 (Reference 7), RG 1.205, Revision 1 (Reference 4), 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 margins. Section 2.4.3.1 of NFPA 805 requires that the 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, based on the as-built and as-operated and maintained plant, and reflect the operating experience at the plant.

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 margins 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, 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 demonstrated to bind 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, Revision 2 (Reference 7), as well as RG 1.205, Revision 1 (Reference 4). The NRC staff concludes that the proposed PCE process at Ginna, 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 margins. The proposed PCE process includes defining the change, a preliminary risk screening, a risk evaluation, and an acceptability determination, as described in Section 2.6.1 of the SE.

Full compliance with 10 CFR 50.48(c) is implementing the plan~ modifications discussed in Section 2.7.1 of the SE (i.e., during full implementation of the transition to NFPA 805). Before full compliance can be achieved, the proposed license condition would provide that RI 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 condition requires the licensee to ensure that fire protection DID and safety margins 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 RP C.3.1 in RG 1.205, Revision 1 (Reference 4), with respect to the requirements for FPP changes during transition and, therefore, 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 quantitative basis. Specifically, the license condition states that prior NRC review and approval are not required for changes to Chapter 3 of NFPA 805 fundamental FPP elements and design requirements for which an engineering evaluation demonstrates that the alternative to an element in Chapter 3 is functionally equivalent or adequate for the hazard. The licensee may use an engineering evaluation to demonstrate that a change to an element in Chapter 3 of NFPA 805 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 (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 Section 1.7 of NFPA 805, "Equivalency," because the condition can be shown to meet the 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 AHJ 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 elements in Chapter 3 of NFPA 805 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 in Chapter 3 of NFPA 805 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. Section 2.4 of NFPA 805 states that engineering analysis is an acceptable means of evaluating a FPP against performance criteria. Engineering analyses shall be permitted to be qualitative or quantitative.

Section 2.4 of NFPA 805 allows the 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.

Chapter 3 of NFPA 805 has four sections where prior NRC review and approval are not required to implement alternatives (that an engineering evaluation demonstrated are adequate for the hazard). The sections are:

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 3.10), 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 in Section 2.4, "Engineering Analyses,"

and Section 2.7, "Program Documentation, Configuration Control, and Quality" of NFPA 805.

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 exceed the damage threshold for the plant being analyzed. The associated evaluations must also meet the documentation content (as outlined in Section 2.7.1 of NFPA 805, "Content") and quality requirements (as outlined in Section 2.7.3 of NFPA 805, "Quality") of the standard in order to be considered adequate. The NRC staff's review of the licensee's compliance with Sections 2.7.1 and 2.7.3 of NFPA 805, is provided in Section 3.8 of the SE.

According to the LAR, the licensee intends to use a FPRA to evaluate the risk of proposed future plant changes. Section 3.4.2 of the SE, "Quality of the Fire Probabilistic Risk Assessment," 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 RI changes to the FPP under the proposed license conditions. The NRC staff concludes that the licensee's process for self-approving future FPP changes is acceptable.

Based on the licensee's administrative controls to ensure that the models remain current and to assure continued quality, the NRC staff concludes that the FRE methods used at Ginna may continue to be used after implementation of the RI/PB FPP. The FRE methods model the cause and effect relationship of associated changes as a means of assessing the risk of plant changes during transition to NFPA 805. (See Section 3.4.2 of the SE, "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 FREs conducted to determine the change in risk associated with proposed plant changes.

2. 7 Modifications and Implementation Items RP C.3.1 of RG 1.205, Revision 1 (Reference 4), 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 list of modifications and implementation items originally submitted in the LAR have been updated by the licensee in the final version in Attachment S of the LAR, provided in the licensee's letter dated June 11, 2015 (Reference 17).

2.7.1 Modifications The NRC staff reviewed Attachment S of the LAR, which describes the plant modifications necessary to implement the NFPA 805 LB, as proposed. These modifications are identified in the LAR as necessary to bring Ginna into compliance with either the deterministic or PB requirements of NFPA 805. As described below, Table S-2 in Attachment S of the LAR provides a description of each of the proposed plant modifications, presents the problem statement explaining why the modification is needed, and identifies that compensatory actions are required to be in place pending completion/implementation of the modification.

The NRC staff confirmed that the modifications identified on Table S-2 in Attachment S are the same as those identified in Attachment C of the LAR, "NEI 04-02 Table B Fire Area Transition," on a fire area basis, as the modifications being credited in the proposed NFPA 805 LB. The NRC staff also confirmed that Table S-2 in Attachment S of the LAR modifications and associated completion schedule are the same as those provided in the proposed NFPA 805 license condition.

As depicted on Table S-1 in Attachment S of the LAR, the licensee completed one modification as part of the NFPA 805 transition. Table S-2 in Attachment S of the LAR provides a detailed listing of the plant modifications that must be completed in order for Ginna to be in full 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 modifications will be completed in accordance with the schedule provided in the proposed NFPA 805 license condition, which states that all modifications will be completed prior to startup from the second refueling outage greater than 12 months after the issuance of the SE. In addition, the licensee agreed to keep the appropriate compensatory measures in place until modifications are complete.

2.7.2 Implementation Items Implementation items are not fully completed or implemented as of the issuance date of the license amendments, but will be completed during implementation of the license amendments 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 on Table S-3 in Attachment S of the LAR. 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 Section 2.7.3, does not change or impact the bases for the safety conclusions made by the NRC staff in the SE.

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 condition and the letter transmitting the amended license (i.e., implementation period), which states that the implementation items listed on Table S-3 in Attachment S of the LAR, will be completed within 180 days after NRC approval, unless that falls within a scheduled outage window. If the latter is the case, then the completion of implementation items will occur 60 days after startup from the scheduled outage.

Implementation items 9, 15 and 19 are associated with modifications and will be completed following completion of the modification process. Implementation item 21 will be completed following NRC approval of the Westinghouse Reactor Coolant Pump (RCP) Seal Topical Report.

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 would be required by the proposed license conditions.

2.7.3 Schedule Section 5.5 of the LAR, as supplemented by the licensee's letter dated June 11, 2015 (Reference 17), provides the overall schedule for completing the NFPA 805 transition at Ginna.

The licensee stated that it will complete the implementation of new NFPA 805 FPP to include procedure changes, process updates, and training to affected plant personnel (except for implementation items 9, 15, 19, and 21) within 180 days after NRC approval. If the NRC approval falls within a scheduled outage window, completion of the implementation items will occur 60 days after startup from that scheduled outage. The licensee stated that implementation items 9, 15, and 19, are associated with modifications and will be completed in accordance with modification program. The licensee further stated that the implementation of item 21 will be completed following NRC approval of the Westinghouse RCP Seal Topical Report.

Section 5.5 of the LAR also states that modifications will be completed prior to startup from the second refueling outage greater than 12 months after issuance of the SE and that appropriate compensatory measures will be maintained until modifications are complete.

Based on the information provided by the licensee, the NRC staff concludes that the completion schedules proposed by the licensee for the modifications and implementation items are acceptable.

3.0 TECHNICAL EVALUATION

The following sections evaluate the technical aspects of the LAR (Reference 9) to transition the FPP at Ginna to one based on NFPA 805 (Reference 3) in accordance with 10 CFR 50.48(c).

While performing the technical evaluation of the licensee's submittal, the NRC staff used the guidance provided in Section 9.5.1.2 of NUREG-0800, "Risk Informed, Performance-Based Fire Protection" (Reference 36), 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 the 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 Chapter 3 of NFPA 805, "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 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.
  • 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 RI/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 (QA).

Attachments A and B of the SE provide additional detailed information that was evaluated by the NRC staff during the course of the review 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 sections of this SE.

3.1 NFPA 805 Fundamental Fire Protection Program Elements and Design Requirements Chapter 3 of NFPA 805 (Reference 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 in Chapter 3 of NFPA 805, 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, a flame-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 is not endorsed.
  • 10 CFR 50.48(c)(2)(vi) - Water supply and distribution. The italicized exception to Section 3.6.4 is not endorsed. Licensees who wish to use the exception to Section 3.6.4 must submit a request for a license amendment in accordance with 10 CFR 50.48(c)(2)(vii) of this section.
  • 10 CFR 50.48(c)(2)(vii) - Performance-based methods. While NFPA 805 Section 3.1 prohibits the use of PB methods to demonstrate compliance with the Chapter 3 requirements, 10 CFR 50.48(c)(2)(vii) specifically permits that the FPP elements and minimum design requirements of 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 specifically allows the use of alternatives to Chapter 3 of NFPA 805, fundamental FPP requirements that have been previously approved by the NRC (which is the AHJ, as denoted in NFPA 805 (Reference 3) and RG 1.205 (Reference 4)), 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, as endorsed by the NRC in RG 1.205, to assess the proposed FPP against the Chapter 3 of NFPA 805 requirements.

As part of this assessment, the licensee reviewed each section and subsection in Chapter 3 of NFPA 805 against the existing FPP and provided specific compliance statements for each attribute that contained applicable requirements. As discussed below, some subsections in Chapter 3 of NFPA 805 do not contain requirements, or are otherwise not applicable, and others are provided with multiple compliance statements to fully document compliance with the element.

The methods used for achieving compliance with the fundamental FPP elements and minimum design requirements are:

1. The existing FPP element directly complies with the requirement: noted in Attachment A of the LAR, "NEI 04-02 Table B-1, Transition of Fundamental Fire Protection Program and Design Elements," as "Complies." (See discussion in Section 3.1.1.1 of the SE.)
2. The existing FPP element complies through the use of an explanation or clarification: noted on Table B-1 in Attachment A of the LAR, as "Complies with clarification." (See discussion in SE Section 3.1.1.2.)
3. The existing FPP element complies through the use of existing engineering equivalency evaluations (EEEEs), both those that existed prior to the transition and those that were created during the transition, whose bases remain valid and are of sufficient quality: noted on Table B-1 in Attachment A of the LAR, as "Complies with use of evaluation." (See discussion in Section 3.1.1.3 of the SE.)
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 on Table B-1 in Attachment A of the LAR, as "Complies via previous approval." (See discussion in Section 3.1.1.4 of the SE.)
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 on Table B-1 in Attachment A of the LAR, as "Submit for NRC Approval." (See discussion in Section 3.1.1.5 of the SE.)

The NRC staff determined that, taken together, these methods compose an acceptable approach for documenting compliance with the requirements in Chapter 3 of NFPA 805 because the licensee 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 Chapter 3 of NFPA 805, 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.

The licensee stated in Section 4.2.2 of the LAR, "Existing Engineering Equivalency Evaluation Transition," that it evaluated the EEEEs used to demonstrate compliance with the requirements in Chapter 3 of NFPA 805, in order to ensure continued appropriateness, quality, and applicability to the current plant configuration. The licensee stated that none of the transitioning EEEEs require NRC approval.

EEEEs (previously known as GL 86-1 O evaluations) are 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 Section 4.2.3 of the LAR, "Licensing Action Transition," that the existing licensing actions used to demonstrate compliance have been evaluated to ensure that the bases remain valid. The results of these licensing action evaluations are provided in Attachment K of the LAR.

Table B-1 in Attachment A of the LAR provides further details regarding the licensee's compliance strategy for specific requirements in Chapter 3 of NFPA 805, including references to where compliance is documented.

3.1.1.1 Compliance Strategy - Complies For the majority of the requirements in Chapter 3 of NFPA 805, 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, the NRC staff concludes that the licensee's statements of compliance are acceptable.

The following NFPA 805 sections identified on Table B-1 in Attachment A of the LAR as complying via this method, and any applicable Chapter 3 of NFPA 805 implementation items on Table S-3 in Attachment S of the LAR, required additional review by the NRC staff:

  • 3.4.1 (c)
  • 3.11.5 Section 3.4.1 (c) of NFPA 805 requires that the fire brigade leader and at least two brigade members have sufficient training and knowledge of nuclear safety systems to understand the effects of fire and fire suppressants on NSPC. The compliance basis for this element on Table B-1 in Attachment A of the LAR, stated that the brigade captain and backup brigade captain are auxiliary operators, but did not specify the details of the training and knowledge of these members. In fire protection engineering (FPE) request for additional information (RAI) 03 (Reference 19), the NRC staff requested that the licensee describe how the requirements in Section 3.4.1 (c) of NFPA 805 are met with regard to training and knowledge of the brigade leader and at least two members of the brigade. In its response to FPE RAI 03 (Reference 10),

the licensee stated that the correct compliance strategy for this element is "Complies via previous approval." See Section 3.1.1.4 of the SE for the evaluation of this element.

Section 3.11.5 of NFPA 805 requires that electrical raceway fire barrier systems (ERFBS),

required by Chapter 4 of NFPA 805, be capable of resisting the fire effects of the hazards in the area. The compliance basis for this element stated that Hemyc wrap is installed in BR1 B, and Modification Item 7 of Table S-2 in Attachment S of the LAR will determine if the existing configuration of wrap is adequate to protect certain cables for 45 minutes (min.). In FPE RAI 04 (Reference 19), the NRC staff requested that the licensee provide additional details regarding the capability of the ERFBS to meet the requirements Section 3.11.5 of NFPA 805, including how the ERFBS duration is based on fire testing of similar material and application. In its response to FPE RAI 04 (Reference 10), the licensee stated that it compared the Hemyc wrap

configuration in BR 1B to the test configurations in an April 18, 2005, Omega Point Laboratories, Inc. test, titled "HEMYC (1-Hour) Electrical Raceway Fire Barrier Systems Performance Testing:

Cable Tray, Cable Air Drop and Junction Box Raceways," as well as a February 5, 2007, report from lntertek Testing Services NA, Inc. titled "HEMYC 1-Hour Electrical Raceway Fire Barrier System (ERFBS), Fire Resistance Performance." The licensee further stated that the analysis indicated that the Hemyc configuration in BR1 B provides 25 mins. of protection after the damage temperature (205 deg. C) of the thermoplastic cables is reached, as long as the supports inside the steel cable chase are stuffed with ceramic fiber material. The licensee indicated that modification ESR-12-0142 will install the ceramic fiber and the PRA model will credit the 25 mins. of protection. The licensee further stated that Hemyc is not being credited as a deterministic resolution to any variances from deterministic requirements. The NRC staff determined that the licensee's response to FPE RAI 04 is acceptable because the licensee provided additional details regarding the capability of the ERFBS to meet the requirements of Section 3.11.5 of NFPA 805, including how the ERFBS duration is based on fire testing of similar material and application. The NRC staff concludes that the licensee's statement of compliance is acceptable because the licensee evaluated the ERFBS configuration in BR1 B to support a PB compliance strategy for Chapter 4 of NFPA 805, and because the licensee identified a required action that will incorporate the provisions in Chapter 3 of NFPA 805 in the licensee's FPP and included the action as an implementation item in Attachment S of the LAR, which would be required by the proposed license condition. See Section 3.1.3 of the SE.

3.1.1.2 Compliance Strategy - Complies with Clarification For certain requirements in Chapter 3 of NFPA 805, 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 compliance strategy meets the underlying requirement for the FPP element, as clarified.

3.1.1.3 Compliance Strategy - Complies with Use of EEEEs For certain requirements in Chapter 3 of NFPA 805, 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, concluding that the licensee's statements of compliance in these instances are acceptable.

The following NFPA 805 sections are identified on Table B-1 in Attachment A of the LAR, as complying via this method required additional review by the NRC staff:

  • 3.3.6
  • 3.3.7 Section 3.3.6 of NFPA 805 requires metal roof coverings to be Class A as determined by tests described in NFPA Standard 256, "Standard Methods of Fire Tests of Roof Coverings" (NFPA 256) (Reference 68). On Table B-1 in Attachment A of the LAR, the compliance basis for this element stated that "all metal roofs were specified to be Factory Mutual class 1." The basis further describes the difference between the scope of testing for a Class A rating versus the scope of testing for a Factory Mutual Class 1 rating, as detailed in the corresponding

Factory Mutual standard. However, the Factory Mutual standard indicates that Class A, B, and C materials may achieve a Factory Mutual Class 1 rating, provided the condition of acceptance for spread of flame, intermittent flame and burning brand are met. In FPE RAI 01 (Reference 19), the NRC staff requested that the licensee provide justification demonstrating how the requirements of Class A are met using the alternate classification. In its response to FPE RAI 01 (Reference 10), the licensee stated that multiple compliance strategies will be used for this element, including complies via previous approval and complies with the use of EEEE. The licensee described the roof construction for the Screen House in the fire protection evaluation in response to Guideline D.1.e in Appendix A of BTP APCSB 9.5-1. The licensee further stated that the lack of a Class A roof was evaluated as being acceptable by the NRC staff based on the low combustible loading inside the Screen House and the active and passive fire suppression and mitigating features, which include curbing around the diesel fire pump and oil storage area, a drainage system in the curbed area, an automatic deluge sprinkler system, an automatic wet-pipe suppression system, smoke detectors, and inside and outside hose reel coverage. The NRC staff determined that the licensee's response to FPE RAI 01 is acceptable because the licensee provided justification demonstrating how the requirements of Class A are met using the alternate classification. The NRC staff reviewed the licensee's statement of continued validity for the EEEEs, as supplemented by the response to FPE RAI 01, and concludes that the licensee's statement of compliance in this instance is acceptable because the fire protection features used as the basis for acceptance of previous NRC staff approval of the roof configuration remain installed.

Section 3.3.7.1 of NFPA 805 requires that NFPA Standard 50A, "Standard for Gaseous Hydrogen Systems at Consumer Sites" (NFPA 50A) (Reference 69) be followed for hydrogen storage. On Table B-1 in Attachment A of the LAR, the compliance basis for this element states, in part, that compliance with NFPA 50A 1973/1978 was assessed and documented; but that flammable gas storage in fire areas TB-1 and TB-2 is such that the total content is less than 400 standard cubic feet (scf). The basis further states that Section 10.1.1 of NFPA 55, "Compressed Gases and Cryogenic Fluids" (Reference 70) (which incorporates NFPA 50A),

indicates that the chapter "does not apply to individual systems using containers having a total hydrogen content of less than 400 scf, if each system is separated by a distance not less than 5 ft." In FPE RAI 02 (Reference 19), the NRC staff requested that the licensee provide a description of the configurations of flammable gas storage in fire areas TB-1 and TB-2, and the administrative controls used to ensure the volume of flammable gas is maintained below 400 scf. In its response to FPE RAI 02 (Reference 10), the licensee stated that the gas bottle racks are anchored to the concrete pedestal in the Turbine Building Mezzanine, and are located greater than 5 ft away from the rack located in the Turbine Building Basement. In addition, the licensee stated that flammable gas storage is administratively controlled and signs are posted to ensure the total hydrogen content is less than 400 scf. The licensee further stated that the racks are not located in the vicinity of safety related equipment. The NRC staff determined that the licensee's response to FPE RAI 02 is acceptable because the licensee provided a description of the configurations of flammable gas store in fire areas TB-1 and TB-2, as well as the administrative controls used to ensure that the volume of flammable gas is maintained below 400 scf. The NRC staff reviewed the licensee's statement of continued validity for the EEEE, as supplemented by the response to FPE RAI 02, and concludes that the licensee's statement of compliance in this instance is acceptable because the configuration meets the requirements for flammable gas storage.

3.1.1.4 Compliance Strategy - Complies with Previous NRC Approval Certain requirements in Chapter 3 of NFPA 805 were supplanted by an alternative that was previously approved by the NRC. The approvals were documented in the Fire Protection Safety Evaluation Report (SER) dated February 14, 1979 (Reference 25).

In FPE RAI 03 (Reference 19), the NRC staff requested that the licensee describe how the requirements in Section 3.4.1 (c) of NFPA 805 are met with regard to training and knowledge of the brigade leader and at least two members of the brigade. In its response to FPE RAI 03 (Reference 10), the licensee stated that the correct compliance strategy for this element is "complies via previous approval." The licensee further stated that "the fire protection program, which includes fire brigade training, is consistent with existing commitments, which were found to be acceptable in the "Fire Protection SER with Summary of MODs, Evaluation of Plant Features and Specific Plant Areas" dated February 14, 1979." The NRC staff determined that the licensee's response to FPE RAI 03 is acceptable because the licensee described how the requirements in Section 3.4.1 (c) of NFPA 805 are met with regard to training and knowledge of the brigade leader and at least two members of the brigade through existing commitments.

In each instance where previous approval was utilized, the licensee evaluated the basis 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 4). Based on the licensee's justification for the continued validity of the previously approved alternatives to Chapter 3 of NFPA 805 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 requested approval for the use of PB 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 amendments approving transition to NFPA 805.

The NFPA 805 sections identified in LAR Attachment A, Table B-1, as supplemented, as complying via this method are:

  • 3.3.1.3.4, which concerns the use of portable heaters in the plant. The licensee requested NRC staff approval for the use of a PB method to justify the use of portable gas heaters in the Screen house to prevent freezing of the traveling screens during cold weather, thereby meeting the requirements in Section 3.3.1.3.4 of NFPA 805. (See Section 3.1.4.1 of the SE for the NRC staff's evaluation of this request.)
  • 3.3.5.1, which concerns minimizing wiring above suspended ceilings, and where installed, requires electrical wiring to be listed for plenum use or routed in armored cable, metal conduit, or cable trays with solid metal top and bottom covers. The licensee requested NRC staff approval for the use of a PB method to justify the use of limited amounts of wiring above suspended ceilings in the

power block, thereby meeting the requirements in Section 3.3.5.1 of NFPA 805.

(See Section 3.1.4.2 of the SE for the NRC staff's evaluation of this request.)

  • 3.3.5.3, which concerns electrical cable construction. The licensee requested NRC staff approval for the use of a PB method to justify the use of video/communication/data cables, thereby meeting the requirements in Section 3.3.5.3 of NFPA 805. (See Section 3.1.4.3 of the SE for the NRC staff's evaluation of this request.)
  • 3.3.12(1 ), which concerns RCPs. The licensee requested NRC staff approval for the use of a PB method to justify the potential oil misting from the RCPs due to normal motor consumption not captured by the oil collection system, thereby meeting the requirements in Section 3.3.12(1) of NFPA 805. (See Section 3.1.4.4 of the SE for the NRC staff's evaluation of this request.)
  • 3.5.16, which concerns dedicated use of the fire protection water supply system.

The licensee requested NRC staff approval for the use of a PB method to justify the use of the fire protection water system to supply the high pressure spray wash system, thereby meeting the requirements in Section 3.5.16 of NFPA 805.

(See Section 3.1.4.5 of the SE for the NRC staff's evaluation of this request.)

  • 3.2.3(1 ), which concerns the establishing of procedures for inspection, testing, and maintenance for fire protection systems and features. The licensee requested NRC staff approval for the use of the PB method described in Electric Power Research Institute (EPRI) Report TR-1006756, "Fire Protection Equipment Surveillance Optimization and Maintenance Guide," to modify fire protection system surveillance frequencies, thereby meeting the requirements in Section 3.2.3(1) of NFPA 805. (See SE Section 3.1.4.6 for the NRC staff's evaluation of this request.)

As discussed in Section 3.1.4 of the SE, 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 - Multiple Strategies In certain compliance statements of the requirements in Chapter 3 of NFPA 805, the licensee used more than one of the above strategies to demonstrate compliance with aspects of the fundamental FPP element.

In each of these cases, based on the information provided by the licensee, the NRC staff concludes that the individual compliance statements are acceptable, the combination of compliance strategies is acceptable, and the licensee demonstrated compliance with the fundamental FPP elements and minimum design requirements in Chapter 3 of NFPA 805.

3.1.1. 7 Chapter 3 Sections not Reviewed Some sections in Chapter 3 of NFPA 805 either do not apply to the transition to an 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., Sections 3.4.5 and 3.11 in Chapter 3 of NFPA 805).
  • Sections that are not applicable because of the following:

The licensee stated that the plant does not have systems of this type installed (e.g., Section 3.6.5, which applies to seismic hose station cross-connected to non-fire protection systems and Section 3.10.6, which applies to total flooding carbon dioxide systems) ; and The requirements are structured with an applicability statement (e.g.,

Sections 3.4.1 (a)(2) and 3.4.1 (a)(3), which applies to the fire brigade standards used since they depend on the type of brigade specified in the FPP).

3.1.1.8 Compliance with Chapter 3 Requirements Conclusion As discussed above, the NRC staff evaluated the results of the licensee's assessment of the proposed RI/PB FPP against the fundamental FPP elements in Chapter 3 of NFPA 805 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 in Chapter 3 of NFPA 805, 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 requirements in Chapter 3 of NFPA 805.
  • Provided appropriate documentation of the state of compliance with the requirements in Chapter 3 of NFPA 805, which adequately demonstrated compliance in that the licensee was able to substantiate that it 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.

Through the use of EEEEs.

Through the use of a combination of the above methods.

Through the use of a PB method that the NRC staff specifically reviewed and approved in accordance with 10 CFR 50.48(c)(2)(vii).

3.1.2 Identification of the Power Block The NRC staff reviewed the licensee's structures identified in Attachment I of the LAR, "Definition of Power Block," Table 1-1 as comprising the "power block." The plant structures listed are established as part of the power block for the purpose of denoting the structures and equipment included in the RI/PB FPP that have additional requirements in accordance with 10 CFR 50.48(c) and NFPA 805. As stated in Section 4.1.3 of the LAR, the power block includes structures that contain equipment required for nuclear plant operations, such as containment, auxiliary building, service building, control building, fuel building, radioactive waste, water treatment, turbine building, and intake structures. The NRC staff concludes that the licensee appropriately evaluated the structures and equipment 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 Hemyc' and MT' Fire Barrier Configurations," Issues GL 2006-03 requested that licensees evaluate its facilities to confirm compliance with existing applicable regulatory requirements in light of the results of NRC testing that determined that both Hemyc and MT 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. The licensee utilizes HemycTM ERFBS, and therefore, the generic issue (GL 2006 (Reference 48)) related to the use of these ERFBS is applicable.

In Section 3.11.5 of Attachment A to the LAR, the licensee stated on Table B-1 that the Hemyc wrap configurations in the plant are not credited fire barriers with the exception of the wrap installed in BR1 B. The installation in BR1 B was evaluated to determine if the configuration is adequate to protect cables L0318, C0687, and a portion of E0053. In FPE RAI 04 (Reference 19), the NRC staff requested that the licensee provide additional detail regarding the capability of this ERFBS to meet the requirements in Section 3.11.5 of NFPA 805 and to include a discussion of how the ERFBS duration is based on fire testing of similar material and application. In its response to FPE RAI 04 (Reference 10), the licensee stated that the Hemyc wrap configuration in BR 1B was compared to the test configurations in an April 18, 2005 Omega Point Laboratories, Inc. test, titled "HEMYC (1-Hour) Electrical Raceway Fire Barrier Systems Performance Testing: Cable Tray, Cable Air Drop and Junction Box Raceways" as well as a February 5, 2007 report from lntertek Testing Services NA, Inc. titled "HEMYC 1-Hour Electrical Raceway Fire Barrier System (ERFBS), Fire Resistance Performance." The licensee further stated that the analysis indicated that the Hemyc configuration in BR1 B was able to provide 25 mins. of protection, provided that the Unistrut supports inside the steel cable chase are stuffed with ceramic fiber material. Modification Item 7 of Table S-2 in Attachment S of the LAR includes modification ESR-12-0142 which will install the ceramic fiber to ensure the parameters of the test configuration are met, and the PRA model will credit the 25 mins. of protection after

the damage temperature (205 deg. C) of the thermoplastic cables is reached. The NRC staff determined that the licensee's response to FPE RAI 04 is acceptable because the licensee provided additional detail regarding the capability of the ERFBS to meet the requirements in Section 3.11.5 of NFPA 805, including how the ERFBS duration is based on fire testing of similar material and application, and the licensee included an action that will incorporate the provisions in Chapter 3 of NFPA 805 in the licensee's FPP and included the action as an implementation item in Attachment S of the LAR, which would be required by the proposed license condition.

In PRA RAI 44 (Reference 19), the NRC staff stated that the FPRA model should reflect the post-transition modifications and the as-built plant, and requested that the licensee make changes to the FPRA model to redefine the baseline of the FPRA using acceptable methods. In its response to PRA RAI 44 (Reference 12), the licensee stated that the FPRA model had been updated to reflect the modeling of the Hemyc wrap, which was reduced to crediting 25 mins.

The NRC staff determined that the licensee's response to PRA RAI 44 is acceptable because the licensee updated its FPRA model to reflect the post-transition modifications and the as-built plant, and redefined the baseline of the FPRA using acceptable methods.

Based on the above discussion, the NRC staff concludes that crediting the one remaining installation of Hemyc, the evaluation of the credited configuration, and the proposed plant modification, are adequate means for resolving the remaining GL 2006-03 issues regarding ERFBS fire barrier configurations. Subject to completion of Modification Item 7 of Table S-2 in Attachment S of the LAR, as stated in the proposed license condition, the NRC staff concludes that the licensee's FRE related to the RI/PB FPP will demonstrate that fire area BR1 B will meet the NSPC using a PB analysis and is acceptable.

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 fundamental FPP elements and minimum design requirements in Chapter 3 of NFPA 805. The director or designee may approve PB methods if the director or designee determines that the PB approach:

(A) Satisfies the performance goals, objectives, and criteria specified in NFPA 805 related to nuclear safety and radiological release; (B) Maintains safety margins; and (C) Maintains fire protection DID (fire prevention, fire detection, fire suppression, mitigation, and post-fire SSD capability).

Section 1.3.1 of NFPA 805, "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.

Section 1.3.2 of NFPA 805, "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.

Section 1.4.1 of NFPA 805, "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.

Section 1.4.2 of NFPA 805, "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.

(2) The source term is capable of being limited.

Section 1.5.1 of NFPA 805, "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 PWR

[pressurized-water reactor] and shall be capable of maintaining or rapidly restoring reactor water level above top of active fuel for a BWR [boiling-water reactor] 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.

(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.

Section 1.5.2 of NFPA 805, "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 10 CFR Part 20 limits.

In Attachment L of the LAR, "NFPA 805 Chapter 3 Requirements for Approval (10 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 requirements of the elements identified in Section 3.1.1.5 of the SE. The NRC staff's evaluation of these proposed methods is provided below.

3.1.4.1 Section 3.3.1.3.4 of NFPA 805, "Control of Ignition Sources" The licensee requested review and approval of a PB method to demonstrate an equivalent level

. of fire protection for the requirement in Section 3.3.1.3.4 of NFPA 805, regarding the use of portable fuel-fired heaters in plant areas containing equipment important to nuclear safety, in Approval Request 1 in Attachment L of the LAR. Specifically, the licensee requested approval for the use of two portable gas fired heaters in the Screenhouse (elevation 253'-6") for temporary heating in the area of the circulation water pump and traveling screens. The licensee indicated that the heaters are used to keep the traveling screens from freezing during cold weather.

The licensee stated that the function of the traveling screens is to remove debris from the water before being pumped through the turbine condenser or service water system; therefore, the ability to keep the traveling screens from freezing during cold weather is necessary for normal plant operation. The licensee further stated that one portable gas heater is used to direct heat towards traveling screens A and B, and the second portable gas heater is used to direct heat towards traveling screens C and D. The licensee stated that the basis for the approval request is that the portable gas-powered heaters are efficient and safe relative to the electric alternative, they are located outside of the zone-of-influence (ZOI) for equipment credited in the NSCA, there is DID, their use is administratively controlled, as is combustible loading, and the geometry of the heaters prevents overturning.

The licensee stated that the ability to provide additional heat, during exceptionally cold periods during the months of December through March, to prevent icing of the traveling screens, is safely and efficiently provided through the use of gas heating, rather than the use of portable electrical heat. The licensee also stated that when not in use, the gas supply to the portable heaters is procedurally isolated from the building. The licensee further stated that each of the heating units has a heating output of 1,500,000 British Thermal Unit (BTU) per unit, for a total heating capacity of 3,000,000 BTU. The licensee further stated that a typical forced air portable electrical heater has an output of approximately 102, 000 BTUs/unit and that to provide an equivalent rating of 3,000,000 BTU, approximately 30 electrical units would be required. The licensee further stated that to supply power to 30 (480 Volt (V)) electrical heaters would require tying into a 480 V bus, however the 480 V busses in the Screen House supply power to safety related equipment, and therefore, supplying non safety related equipment with a safety related power source would be adverse to the NSCA.

The licensee stated that equipment identified to be important to nuclear safety is outside the ZOI of the portable gas heaters and therefore, it can be conservatively assumed that the impact of ignition of these heaters would be bounded by the impact of a hydrogen storage tank fire. The licensee also stated that Section N.2.3 of NUREG/CR-6850 (Reference 40), recommends assuming damage to, and ignition of, combustibles within 10 to 15 feet of the outer limit of the tank and that the closest equipment credited in the NSCA are the motor driven fire pump and the service water pumps which are at a minimum distance of approximately 40 feet from the location of the portable gas heaters, which is significantly beyond the ZOI for the heaters.

The licensee stated that operator walk downs for the Screen House are conducted twice per shift by procedure. The licensee also stated that there is a hose station located in the fire zone and there are portable fire extinguishers available to control and extinguish fires in the area.

The licensee further stated that there is an automatic sprinkler system which provides suppression protection for the service water pump area and fire pump area. The licensee further stated that smoke detectors provide detection protection for the service water pump, fire water pump, and switchgear area, and exterior yard hydrants are available for manual fire-fighting.

The licensee stated that special provisions for fire protection in the fire zone include a curb around the diesel fire pump and the diesel oil storage tank to prevent spread of flammable liquid. The licensee also stated that the curbed area is equipped with a floor drain which drains to a holding tank buried outside the Screen House and that the gas supply line for the two heaters is provided with an excess flow check valve at the outdoor meter station.

The licensee stated that the portable gas heaters are installed and secured per procedures that control the installation and removal of the heaters, and control placing the heaters in service.

The licensee also stated that portable gas heaters are located on a noncombustible floor and the geometry of the units is such that they are not subject to over- turning. The licensee further stated that a metal gate is used to protect plant personnel and ensure there is a minimum clearance of 6 feet above and 2Y2 feet on all sides separating the heater from combustible material. The licensee further stated that combustible materials are administratively controlled through procedures and tracked within the plants' design analysis.

In FPE RAI 05 (Reference 19), the NRC staff requested that the licensee describe the administrative controls that ensure the portable heaters are located, at a minimum, of 40 feet from the NSCA equipment each time they are installed, as well as how the administrative controls account for the heaters once installed. In its response to FPE RAI 05 (Reference 10),

the licensee stated that the gas line that supplies the heaters is hard piped and therefore, the location of the heaters, which is at a minimum of 40 feet from NSCA equipment, is fixed. The licensee also stated that there is no impact to the combustible loading analysis when the portable gas heaters are in service, since the gas supply is hard piped. The licensee further stated that the Fire Marshal or designee reviews the permit for the proposed usage and/or storage of transients, in regards to location, and includes potential restrictions to be followed as necessary. The licensee further stated that during the initial installation of the portable heaters, the Fire Marshal must be notified, which will ensure the potential for a transient already in place in the Screen House will be evaluated for relocation if it is in the proximity of the two portable gas fired heaters. The licensee further stated that temporary combustible permits are periodically reviewed to ensure restrictions are being followed and documented as part of the plant inspection program. The NRC staff determined that the licensee's response to FPE RAI 05 is acceptable because the licensee described the administrative controls that ensure that the portable heaters are located, at a minimum, of 40 feet from the NSCA equipment each time they are installed, as well as how the administrative controls account for the heaters once installed.

The licensee stated that the use of two portable gas heaters in the Screen House does not affect the nuclear safety or radiological release performance criteria. The licensee stated that the portable gas heaters are located outside the ZOI of the NSCA components and the design features of the building, including sprinkler systems, detection systems, hose reels, fire extinguishers, curbing, and construction, provide reasonable assurance that, in the event of a fire, the plant is not placed in an unrecoverable condition. The licensee further stated that the Screen House is not a radiological controlled area, and therefore, the heaters will not impact the radiological release performance criteria.

The licensee stated that the safety margin for the use of two portable gas heaters in the Screen House will remain unchanged since the use of combustibles in the area is procedurally tracked, the equipment is monitored by procedure, and the heaters are located such that any NSCA equipment would not be impacted.

The licensee stated that the three echelons of DID are: ( 1) to rapidly detect, control, and extinguish fires from starting; (2) to rapidly detect, control, and extinguish fires that do occur; thereby, limiting damage; and (3) to provide adequate level of fire protection for systems and structures so that a fire will not prevent essential safety functions from being performed. The licensee stated that echelon ( 1) is maintained by the control of combustibles within this fire area through the use of procedures. The licensee further stated that echelons (2) and (3) are not affected because the installation of the gas heaters do not result in compromising automatic or manual fire suppression functions, fire protection for systems and structures, or post-fire SSD capability.

Based on 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 requirement in Section 3.3.1.3.4 of NFPA 805. The proposed

PB method 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 DID (fire prevention, fire detection, fire suppression, mitigation, and post-fire SSD capability).

3.1.4.2 NFPA 805, Section 3.3.5.1 - Wiring Above Suspended Ceilings In Approval Request 2 in Attachment L of the LAR, the licensee requested review and approval of a PB method to demonstrate an equivalent level of fire protection for the requirement in Section 3.3.5.1 of NFPA 805 regarding the installation of wiring above suspended ceilings in the power block. Section 3.3.5.1 of NFPA 805 requires that wiring above suspended ceilings be kept to a minimum and be listed for plenum use, or routed in armored cable, metal conduit, or cable trays with solid metal top and bottom covers.

The licensee stated that suspended ceilings were identified in multiple areas such as the Control Room, Technical Support Center (TSC) and Service Building basement office area.

The licensee further stated that with the exception of the Control Room, these areas are not risk significant. In addition, the licensee also stated that the majority of wiring above the suspended ceilings is related to communication or lighting systems, and is not power cables. The licensee stated that these cables are low voltage, and are not generally susceptible to shorts which would result in a fire.

As described in the LAR, the basis for the approval request is:

  • There are no ignition sources above these ceilings.
  • The wiring above ceilings in offices, conference rooms, laboratories, lobbies, etc.,

does not pose a hazard:

o Low voltage is not susceptible to shorts causing a fire; o There is a lack of continuity of combustibles; o There is no equipment important to nuclear safety in the vicinity of these cables; and o Modification design process requires new installations to use plenum-rated equivalent or armored cable.

  • Power, control or instrumentation cables installed are either IEEE [Institute of Electrical and Electronics Engineers] -383 qualified (or equivalent) or provided with a flame retardant coating.

The licensee also stated that the Control Room Heating, Ventilation, and Air Conditioning (HVAC) system design supports the rapid identification of combustion products if a fire were to occur in the MCR suspended ceiling. The licensee stated that the normal HVAC system is also equipped with a smoke detector that monitors return air in the duct between the Control Room Envelope (CRE) and the return air fan, and provides an alarm in the Control Room. According

to the licensee, since the Control Room Emergency Air Treatment System isolates and recirculates air within the CRE boundary in a closed-loop system, the existing fire detection system or the Control Room operators who are continuously present in the area would quickly identify the presence of smoke while the air is being recirculated.

During the review, the NRC staff found that the basis for the approval request indicates that there are multiple areas where suspended ceilings are located, however only the Control Room was discussed in detail. In FPE RAI 06 (Reference 19), the NRC staff requested that the licensee provide a description of the other fire areas that contain wiring above suspended ceilings, including proximity to fire areas containing nuclear safety capability systems and equipment, as well as a description of the type, use and amount of wiring, proximity to combustibles, presence of ignition sources, as well as fire detection and suppression features that may be installed. In its response to FPE RAI 06 (Reference 10), the licensee revised the approval request. Regarding the Control Room, the licensee further stated that video/communication/data cables above the suspended ceiling are approximated to be less than 5 percent of the total space. According to the licensee, these cables are low voltage and are not generally susceptible to shorts which would result in a fire. The licensee stated that the Control Room does contain NSCA components/equipment, however portable fire extinguishers are available, hose reels are available for use on the Turbine Building Operating floor, and smoke and heat detectors provide area detection for the Control Room.

Regarding the TSC, the licensee stated that the quantity of wiring which may not meet the requirement of the section is estimated to be less than 1 percent of the space above the suspended ceiling. The licensee further stated that the majority of this wiring is low voltage communications wiring or lighting systems related. According to the licensee, there are no intervening combustibles in the space above the suspended ceilings since the majority of equipment is metal, and all other wiring is routed in conduits. Further, the licensee also stated that there are no ignition sources and the proximity of the wiring to areas containing NSCA equipment is not an issue since there are 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> rated walls and penetrations separating these areas. The licensee stated that there is also detection located in the hallways, and rooms within the TSC, along with an automatic sprinkler system which provides suppression coverage for the TSC Diesel Generator Room and Operational Support Center. The licensee further stated that there is an automatic Halon suppression system which provides suppression for the computer room and subfloor, and portable fire extinguishers are available in the zone and in adjacent fire areas/zones along with hose stations.

Regarding the Service Building basement office areas, the licensee stated that the quantity of wiring which may not meet the requirement of the section is considered minor, estimated to be less than 1 percent of the space, and classified as video/communication/data cables. The licensee further stated that there are no NSCA components located in the fire zone and no ignition sources above these ceilings. According to the licensee, although there is a nonrated wall between the Service Building Basement and the Primary Water Treatment Room, which does contain NSCA components, there is an automatic sprinkler system that provides suppression protection in the fire zone. The licensee stated that there are portable fire extinguishers available in this zone and there are hose stations in adjacent fire areas/zones as well as yard hydrant connections.

The licensee further stated that the wiring above suspended ceilings does not affect nuclear safety since the space enclosing the cables is non-combustible, there is a minimum amount of nearby ignition sources, and the cables are low energy and therefore, pose a low fire ignition hazard due to hot shorts. According to the licensee, there is no impact on NSPC. In addition, the licensee stated that the wiring has no impact on the radiological release performance criteria since the review was performed based on the potential location of radiological concerns and is not dependent on the type of cables.

The licensee stated that the amount of non-rated and non-enclosed wiring above the ceilings in the Power Block is minor and does not present a significant fire hazard, and therefore, the safety margin inherent in the analysis for each fire event has been preserved.

The licensee stated that the three echelons of DID are to: (1) prevent fires from occurring; (2) rapidly detect, control and extinguish fires that occur, thereby limiting damage; and (3) provide an adequate level of fire protection for systems and structures, so that a fire will not prevent essential safety functions from being performed, are maintained. The licensee stated that the non-listed video/communications/data cables routed above suspended ceilings do not directly result in compromising automatic fire suppression systems, manual fire suppression functions, or post-fire SSD capability. The NRC staff determined that the licensee's response to the FPE RAI 06 is acceptable because the licensee provided a description of the other fire areas that contain wiring above suspended ceilings, including proximity to fire areas containing nuclear safety capability systems and equipment, as well as a description of the type, use, and amount of wiring, proximity to combustibles, presence of ignition sources, as well as fire detection and suppression features that may be installed.

Based on 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 requirement in Section 3.3.5.1 of NFPA 805 because it 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 DID (fire prevention, fire detection, fire suppression, mitigation, and post-fire SSD capability).

3.1.4.3 NFPA 805, Section 3.3.5.3 - Electrical Cable Construction In Approval Request 3 in Attachment L of the LAR the licensee requested review and approval of a PB method to demonstrate an equivalent level of fire protection for the requirement in Section 3.3.5.3 of NFPA 805, regarding the requirements that electrical cable construction comply with a flame propagation test acceptable to the AHJ. Specifically, the licensee requested approval of the use of video/communication/data cables that have not been tested in accordance with IEEE Standard 383, "IEEE Standard for Qualifying Class 1E Electric Cables and Field Splices for Nuclear Power Generating Stations" (IEEE-383) (Reference 71) or any other qualification standard endorsed by the NRC.

The approval request from the licensee stated that the PB method is specifically associated with the installation of video/communication/data cables. However, the basis for the request included a discussion of power cables used for modifications such as the spent fuel pool bridge crane which was modified using cables others than those meeting the IEEE 383 (Reference 71 ).

In FPE RAI 07 (Reference 19), the NRC staff requested that the licensee provide a clarification of the scope of the request, and further justification for the acceptability of using this cable. In its response to FPE RAI 07 (Reference 10), the licensee revised the approval request in its entirety.

The licensee stated that video/communication/data cables are not necessarily tested in accordance with the flame propagation tests outlined in the FAQ 06-0022 (Reference 49) as endorsed by the NRC. The licensee further stated that these low voltage cables are not generally susceptible to shorts which would result in a fire; therefore, self-ignited fires are not a concern. According to the licensee, an exposure fire could potentially ignite the cables, although the same fire would result in damage to other cables in the vicinity.

The licensee stated that with the exception of the telephone communications room located in fire area BOP/ zones TB-2 (Turbine Building Mezzanine), and TSC-1 M (Administrative Computer Room), along with the Control Room (fire area CC/ zone CR), the remaining areas contain a limited quantity of this wiring which is dispersed, is not capable of causing fire damage to components necessary for SSD, and is considered an insignificant fire hazard. The licensee further stated that the telephone communications room in fire area BOP/ zone TB-2 is protected by an automatic sprinkler system, portable fire extinguishers, and hose stations. According to the licensee, the TSC-1 M is protected by smoke detectors, portable fire extinguishers, and adjacent areas are equipped with hose stations. The licensee stated that a potential fire in these areas would be detected and extinguished. The licensee also stated that fire area - BOP (sic) [CC)/CR is a constantly manned area, equipped with smoke and heat detectors, portable fire extinguishers, and hose reels in adjacent areas. The licensee further stated that a potential fire in this area would be readily detected and extinguished due to it being constantly manned.

In addition, the licensee stated that the combustible loading and transient combustibles in these areas are administratively tracked.

The licensee stated that there is no impact on NSPC since video/communication/data cables are low-voltage cable not susceptible to shorts that would result in a fire. The licensee further stated that the flame propagation testing of electrical cable construction has no impact on the radiological release performance criteria. According to the licensee, the radiological review was performed based on the potential location of radiological concerns and is not dependent on the flame propagation tests of cables. The licensee stated that the limited use of video/communication/data cabling does not create or pose an un-acceptable fire hazard and that the radiological release performance criteria are also satisfied based on the determination of limiting radioactive release.

The licensee stated that safety margin is unchanged since in most areas the cables do not present a significant fire hazard, are of low voltage, not susceptible to hot shorts, not capable of causing fire damage to components necessary for SSD and do not directly result in compromising automatic fire suppression or detection functions, manual fire suppression functions, or post-fire SSD capability. The licensee further stated that in areas where the quantity of cable is more significant, such as the communications room, administrative computer room, and Control Room, the suppression system, smoke detectors, portable fire extinguishers, and hose reels, and constant manning of the Control Room are considered adequate to prevent fire propagation in these areas.

Lastly, the licensee stated that the three echelons of DID: (1) prevent fires from starting; (2) rapidly detect, control and extinguish fires that do occur; 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, are not impacted by the use of the cables.

The licensee further stated that the use of un-tested video/communication/data cables does not affect echelon (1) of DID because administrative procedures are used to prevent fire from occurring, and to control and track combustibles. According to the licensee, in areas containing these cables, which cannot be categorized as insignificant, detection, manual hose stream, and fire extinguishers are provided to ensure the fire is rapidly detected and controlled/extinguished by the fire brigade. The licensee stated that echelon (2) of DID is maintained. The licensee further stated that the telephone communications room, within fire zone TB-2 (Turbine Building Mezzanine), is protected by an automatic sprinkler system, to prevent the spread of fire to systems and structures. According to the licensee, the administrative computer room, in fire zone TSC-1 M, is protected by smoke detectors and hose reels, along with 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> concrete block wall to adjacent fire zone TSC-1 N (Technical Support Center) and TB-2 (Turbine Building Mezzanine). The licensee also stated that the Control Room is protected by smoke and heat detectors, an automatic deluge spray system, which provides a water curtain on the wall between the Control room and the Turbine Building Operating Floor, along with 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> rated walls to the exterior YARD area. The licensee further stated that echelon (3) of the DID concept is maintained. The NRC staff determined that the licensee's response to FPE RAI 07 is acceptable because the licensee provided clarification on the scope of the request for the use of video/communication/data cables that are not tested to the flame propagation tests as endorsed by the NRC.

Based on 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 requirement in Section 3.3.5.3 of NFPA 805 because it 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 DID (fire prevention, fire detection, fire suppression, mitigation, and post-fire SSD capability).

3.1.4.4 NFPA 805, Section 3.3.12(1) - Reactor Coolant Pump Oil Collection In Approval Request 4 in Attachment L of the LAR, the licensee requested review and approval of a PB method to demonstrate an equivalent level of fire protection for the requirement in Section 3.3.12(1) of NFPA 805, regarding the oil collection system for RCPs being capable of collecting lubricating oil from all potential pressurized and non-pressurized leakage sites in each RCP oil system. The licensee requested approval for the potential of oil misting from the RCPs due to normal motor operation.

Although the oil collection system was previously reviewed in accordance with Section 111.0 in Appendix R of 10 CFR 50, including the capability to collect leakage from pressurized and non-pressurized leakage sites in the RCP oil system, this previous review did not include collection of oil mist as a result of pump/motor operation. The licensee stated that oil misting is not leakage due to equipment failure, but an inherent occurrence in the operation of large rotating equipment. The licensee further stated that it is normal for large motors to lose some oil

through seals and the oil to potentially become 'atomized' in the ventilation system. According to the licensee, this atomized oil mist can then collect on surfaces in the vicinity of the RCP as the pump design is not completely sealed to permit airflow for cooling. The licensee also stated that the oil mist resulting from normal operation will not adversely impact the ability of a plant to achieve and maintain SSD even if ignition occurred.

As described in the LAR, the basis for the approval request is:

  • The oil collection system is designed to collect leakage from pressurized and non-pressurized leakage sites in the RCP oil system.
  • Oil misted from normal operation is not leakage; it is normal motor oil consumption.
  • Oil misted from normal operation does not significantly reduce the oil inventory.
  • The oil historically released as misting does not account for an appreciable heat release rate (HRR) or accumulation near potential ignition sources or non-insulated reactor coolant piping.
  • The RCP A and B use synthetic oil with a flash point of 428 degrees Fahrenheit.
  • RCPs are not required to achieve or maintain fire SSD.

In FPE RAI 08 (Reference 19), the NRC staff requested that the licensee provide additional characterization of the oil quantity and deposition location, the associated fire hazards, and the actions taken, if any, to clean the oil mist deposits from equipment surfaces. In its response to FPE RAI 08 (Reference 10), the licensee stated that oil misting has the potential to deposit on surfaces in the pump bay located within the RCP System Loop A and Loop B, and a light sheen of oil was observed on equipment in the pump bay as well as on the walls of the RCP cubicle.

The licensee further stated that RCP oil loss is tracked by the system engineer and oil deposits are tracked using condition reports. According to the licensee, components necessary to meet NSPC were evaluated for the potential of oil misting and equipment located in fire zone RC-1 was determined not to be subject to oil misting due to the geometry of the building. The licensee stated that equipment located in fire zones T-LOOPA and T-LOOPB that are in the vicinity of the RCP have the potential to be impacted by the oil misting, however no physical evidence of oil accumulation was observed during system engineering outage walk-downs. The licensee further stated that actions are taken to clean any oil spills from leakage or misting during refueling outages. In addition, the licensee found that oil loss from the RCPs was significantly reduced with modifications that added a labyrinth seal at the lower oil pot standpipe, which resulted in a reduction of oil lost from the lower oil pot through a fuel cycle. The NRC staff determined that the licensee's response to FPE RAI 08 is acceptable because the licensee provided additional characterization of the oil quantity and deposition location, the associated fire hazards, and the actions taken to clean the oil mist deposits from equipment surfaces.

The licensee stated that the oil mist resulting from normal operation will not adversely impact nuclear safety, since there are redundant pumps and the RCPs are not required to achieve and

maintain SSD. The licensee further stated that there is no impact on NSPC. According to the licensee, since the radiological release review was performed based on the manual fire suppression activities in areas containing or potentially containing radioactive materials, and the entire Containment Building in which the RCPs are located is an environmentally sealed radiological area during power operations, the potential for oil mist from the RCPs has no impact on the radiological release performance criteria. The licensee also stated that the oil mist does not add additional radiological materials to the area or challenge systems' boundaries that contain such materials.

The licensee further stated that the oil mist resulting from normal operation will not adversely impact the ability of the plant to achieve and maintain fire SSD even if ignition occurred, since the pumps are not required to achieve and maintain fire SSD. The licensee also stated that the inherent safety margin remains unchanged.

Finally, the licensee stated that the three echelons of DID: (1) to prevent fires from starting, to rapidly detect; (2) control and extinguish fires that do occur; and (3) to provide adequate level of fire protection for systems and structures so that a fire will not prevent essential safety functions from being performed, are not impacted by the potential for oil mist from RCPs. The licensee further stated that echelon (1) is maintained by the oil collection system and RCP design.

According to the licensee, the introduction of a small amount of oil misting does not affect echelons (2) and (3). The licensee stated that the potential for oil mist from the RCPs does not result in compromising automatic fire suppression functions, manual fire suppression functions, fire protection for systems and structures, or post-fire SSD capability and, therefore, fire protection DID is not impacted.

Based on 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 requirement in Section 3.3.12(1) of NFPA 805 because it 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 DID (fire prevention, fire detection, fire suppression, mitigation, and post-fire SSD capability).

3.1.4.5 NFPA 805, Section 3.5.16 - Dedicated Use of Fire Protection Water Supply In Approval Request 5 in Attachment L of the LAR, the licensee requested review and approval of a PB method to demonstrate an equivalent level of fire protection for the requirement in Section 3.5.16 of NFPA 805 regarding the requirement that the fire protection water supply be dedicated for fire protection use only. Specifically, the licensee requested approval for the use of the fire protection water system to supply high pressure water to the traveling screen spray wash system.

The licensee stated that both automatic and manual means are provided to isolate the screen spray wash system in the event of any automatic or manual fire suppression system actuation.

The licensee further stated that upon any automatic actuation of the electric motor-driven fire pump via a fire-suppression system demand, the high pressure spray wash system will automatically be isolated so that the primary function of supplying the fire suppression system will be satisfied. According to the licensee, the fire protection water supply is taken from Lake

Ontario and the use of fire protection water to supply high pressure water to the traveling screen spray wash system has no impact on the amount of water available for fire-fighting purposes.

The licensee stated that since the fire protection water is taken from a virtually perpetual water source, and automatic and manual means are provided to isolate the supply to the traveling screen spray wash system in the event of any automatic or manual fire suppression system actuation, there is no impact on NSPC since the fire water system would be able to perform its function as designed. The licensee further stated that the use of the fire protection water supply system to supply the traveling screen spray wash system has no impact on the radiological release performance criteria since the radiological review was performed based on the potential location of radiological concerns and the Screen House is not a radiological area.

In addition, the licensee stated that the screen wash system cannot be placed into service unless the diesel fire pump is operable and the motor-driven fire pump discharge pressure is maintained greater than or equal to 132 pounds per square inch, to ensure the most limiting suppression system flow in the event the valve that isolates the screen wash system fails to close automatically. Also, according to the licensee, the traveling screen high pressure spray wash automatic isolation is procedurally and periodically verified. The licensee further stated that the use of the fire water system to provide water to the traveling screen wash system does not impact safety margins since the primary function of the fire water system, which is to provide water for the suppression of fires, will not be impacted.

Section 1.2 of NFPA 805 states that DID 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; and (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. The licensee stated that the use of the fire water system to supply high pressure water to the traveling screen spray wash system does not impact the primary function of the fire water system, and the design of the traveling screen spray wash system assures the fire water system function is maintained during all modes of plant operation when the high pressure traveling screen spray wash system is or is not in service. Therefore, the licensee stated that the DID capability of the fire water system is unchanged.

Based on 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 requirement in Section 3.5.16 of NFPA 805 because it 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 DID (fire prevention, fire detection, fire suppression, mitigation, and post-fire SSD capability).

3.1.4.6 NFPA 805, Section 3.2.3(1 ), Fire Protection Systems and Features, Inspection, Testing, and Maintenance Procedures In LAR Attachment L, Approval Request 6, the licensee requested NRC approval for the ability to utilize PB methods to establish the appropriate inspection, testing, and maintenance

frequencies for fire protection systems and features required by NFPA 805. The licensee stated that PB inspection, testing, and maintenance frequencies will be established as described in EPRI Technical Report TR-1006756, "Fire Protection Surveillance Optimization and Maintenance Guide for Fire Protection Systems and Features," Final Report, July 2003.

The licensee included in its basis for the request the following:

NFPA 805, Section 2.6, "Monitoring," requires that:

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.

NFPA 805, Section 2.6.1, "Availability, Reliability, and Performance Levels," requires that:

Acceptable levels of availability, reliability, and performance shall be established.

NFPA 805, Section 2.6.1, "Monitoring Availability, Reliability, and Performance," requires that:

Methods to monitor availability, reliability, and performance shall be established.

The methods shall consider the plant operating experience and industry operating experience.

The licensee stated that the scope and frequency of the inspection, testing, and maintenance activities for fire protection systems and features required in the FPP have been established based on the previously approved TSs/license controlled documents and appropriate NFPA codes and standards, and that its request does not involve the use of EPRI TR-1006756 to establish the scope of those activities since that is determined by the required systems review identified in LAR Attachment C, Table C-2, "NFPA 805 Chapter 4 Required FP Systems/Features."

The licensee stated that this request is specific to the use of EPRI TR-1006756 to establish the appropriate inspection, testing, and maintenance frequencies for fire protection systems and features credited by the FPP. The licensee further stated that EPRI TR-1006756, Section 10.1 states that, "The goal of a performance-based surveillance program is to adjust test and inspection frequencies commensurate with equipment performance and desired reliability," and that this goal is consistent with the stated requirements of NFPA 805, Section 2.6. The licensee further stated that EPRI TR-1006756 provides an accepted method to establish appropriate inspection, testing, and maintenance frequencies which ensure the required NFPA 805 availability, reliability, and performance goals are maintained.

The licensee stated that the target tests, inspections, and maintenance will be those activities for the NFPA 805 required fire protection systems and features, and that reliability and frequency goals will be established to ensure the assumptions in the NFPA 805 engineering analysis remain valid. The licensee further stated that the failure criterion will be established based on the required fire protection systems and features credited functions and will ensure

those functions are maintained, and that data collection and analysis will follow the guidance in EPRI TR-1006756. The licensee further stated that the failure probability will be determined based on EPRI TR-1006756 guidance and a 95% confidence level will be utilized. The licensee further stated that the performance monitoring will be performed in conjunction with the monitoring program required by NFPA 805, Section 2.6 and it will ensure site specific operating experience is considered in the monitoring process.

The licensee stated that it does not intend to revise any fire protection surveillance, test or inspection frequencies until after transitioning to NFPA 805 and that existing fire protection surveillance, test and inspections will remain consistent with applicable technical requirements manual, Insurer, and NFPA code requirements. The licensee further stated that its intent is to obtain approval via the NFPA 805 SE to use EPRI TR1006756 in the future as opportunities arise.

The licensee stated that it reserves the ability to evaluate fire protection features with the intent of using EPRI PB methods to provide evidence of equipment performance beyond that achievable under traditional prescriptive maintenance practices to ensure optimal use of resources while maintaining reliability.

The licensee stated that the use of PB test frequencies established per EPRI Technical Report TR-1006756, combined with NFPA 805, Section 2.6 monitoring program, will ensure that the availability and reliability of the fire protection systems and features are maintained to the levels assumed in the NFPA 805 engineering analysis, and therefore, there is no adverse impact to the NSPC by the use of this method.

The licensee stated that the radiological release performance criteria are satisfied based on the determination of limiting radioactive release and that fire protection systems and features may be credited as part of that evaluation. The licensee further stated that use of PB test frequencies established per EPRI TR-1006756 methods combined with NFPA 805, Section 2.6 monitoring program, will ensure that the availability and reliability of the fire protection systems and features are maintained to the levels assumed in the NFPA 805 engineering analysis, which includes those assumptions credited to meet the radioactive release performance criteria, and therefore, there is no adverse impact to radioactive release performance criteria.

The licensee stated that the use of PB test frequencies established per EPRI TR-1006756 methods combined with NFPA 805, Section 2.6 monitoring program, will ensure 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 analysis which includes those assumptions credited in the FRE safety margin discussions. The licensee further stated that the use of these methods in no way invalidates the inherent safety margins contained in the codes and standards used for design and maintenance of fire protection systems and features, and therefore, the safety margin inherent and credited in the analysis has been preserved.

The licensee stated that the three echelons of defense-in-depth described in NFPA 805, Section 1.2 are: 1) to prevent fires from starting (combustible/hot work controls); 2) to 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) to 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 echelon 1 is not affected by the use of the EPRI TR-1006756 methods.

The licensee's proposed PB method does not affect fire prevention activities which are primarily conducted through the use of administrative controls. The licensee further stated that the use of PB test frequencies established per EPRI TR-1006756 methods combined with NFPA 805, Section 2.6 monitoring program, will ensure 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 analysis, and therefore, there is no adverse impact to echelons 2 and 3.

The licensee stated that EPRI TR-1006756 provides an accepted method to establish appropriate inspection, testing, and maintenance frequencies which ensure the required NFPA 805 availability, reliability, and performance goals are maintained. The NRC staff determined that the proposed PB method will continue to allow the fire protection systems and features to rapidly detect, control and extinguish fires and limit fire damage, and to provide an adequate level of fire protection for systems and structures so that a fire will not prevent essential safety functions from being performed.

Based on 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 3.2.3(1) requirement because it provides an accepted method to establish appropriate inspection, testing, and maintenance frequencies, which ensures that the required NFPA 805 availability, reliability, and performance goals are maintained. Therefore, the NRC staff concludes that the proposed PB method 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 DID (fire prevention, fire detection, fire suppression, mitigation, and post-fire SSD capability).

3.2 Nuclear Safety Capability Assessment Methods NFPA 805 (Reference 3) is a 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).

Section 2.4 of NFPA 805, "Engineering Analyses, states 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.

Section 1.3.1 of NFPA 805, "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.

Section 1.4.1 of NFPA 805, "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.

Section 1.5.1 of NFPA 805, "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 PWR and shall be capable of maintaining or rapidly restoring reactor water level above top of active fuel for a 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.

(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 Nuclear Safety Capability Assessment Methods Section 2.4.2 of NFPA 805, "Nuclear Safety Capability Assessment," states that:

The purpose of this section is to define the methodology for performing a NSCA. The following steps shall be performed:

( 1) Selection of systems and equipment and their interrelationships necessary to achieve the NSPC in Chapter 1; (2) Selection of cables necessary to achieve the NSPC in Chapter 1; (3) Identification of the location of nuclear safety equipment and cables; and (4) Assessment of the ability to achieve the NSPC given a fire in each fire area.

This SE section evaluates the first three of the topics listed above. Section 3.5 of the SE addresses the assessment of the fourth topic.

RG 1.205, Revision 1 (Reference 4) endorses NEI 04-02, Revision 2 (Reference 7), and Chapter 3 of NEI 00-01, Revision 2 (Reference 31) and promulgates the method outlined in NEI 04-02 for conducting a NSCA. This NRG-endorsed guidance (i.e., NEI 04-02 Table B-2, "NFPA 805 Chapter 2 - Nuclear Safety Transition - Methodology Review Worksheet" and NEI 00-01 Chapter 3) has been determined to address the related requirements in Section 2.4.2 of NFPA 805. The NRC staff reviewed Section 4.2.1 of the LAR, "Nuclear Safety Capability Assessment Methodology," and Attachment B of the LAR, "NEI 04-02 Table B Nuclear Safety Capacity Assessment - Methodology Review," against these guidelines.

The endorsed guidance provided in NEI 00-01, Revision 2 (Reference 31) provides a framework to evaluate the impact of fires on the ability to maintain post-fire SSA. It provides detailed guidance for:

1) Selecting systems and components required to meet the NSPC;
2) Selecting the cables necessary to achieve the NSPC;
3) Identifying the location of nuclear safety equipment and cables; and
4) Appropriately conservative assumptions to be used in the performance of the NSCA.

The licensee developed the LAR based on the three guidance documents cited above. Based on the information provided in the licensee's submittal, as supplemented, the NRC staff determined that the licensee used a systematic process to evaluate the post-fire SSA against

the requirements in Subsections (1 ), (2), and (3) in Section 2.4.2 of NFPA 805, which also meets the methodology outlined in the latest NRG-endorsed industry guidance.

FAQ 07-0039 (Reference 54) 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 methodology in Chapter 3 of NEI 00-01 (Reference 31), which in turn, is mapped to the requirements in Section 2.4.2 of NFPA 805.

The licensee performed this evaluation by comparing its SSA against the NFPA 805 NSCA requirements using the NRG-endorsed process in Chapter 3 of NEI 00-01, Revision 2 (Reference 31 ), and documenting the results of the review on Table B-2 in Attachment B of the LAR, in accordance with NEI 04-02, Revision 2 (Reference 7), as modified by FAQ 07-0039 (Reference 54).

For all applicable attributes in Chapter 3 of NEI 00-01 (Reference 31 ), the licensee stated that the SSA directly aligns with the attribute. This was noted on Table B-2 in Attachment B of the LAR, as "Aligns."

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 on Table B-2 in Attachment B of the LAR, as "Not Applicable."

The NRG staff determined that, taken together, these methods compose an acceptable approach for documenting compliance with the requirements in Section 2.4.2 of NFPA 805, "Nuclear Safety Capability Assessment," because the licensee followed the alignment strategies identified in the endorsed NEI 04-02 guidance document (Reference 7). The process defined in the endorsed guidance provides an organized structure to document each attribute in Chapter 3 of NEI 00-01 (Reference 31), allowing the licensee to provide significant detail in how the program meets the requirements.

3.2.1.1 Attribute Alignment -- Aligns For all of the applicable attributes in Chapter 3 of NEI 00-01, the licensee determined that the SSA aligns directly with the attribute. Based on the information provided in the LAR, the NRG staff concludes that the licensee's statements of alignment are acceptable.

3.2.1.2 Attribute Alignment -- Aligns with Intent The licensee did not identify any attributes in this category on Table B-2 in Attachment B of the LAR.

3.2.1.3 Attribute Alignment -- Not in Alignment, but Prior NRG Approval The licensee did not identify any attributes in this category on Table B-2 in Attachment B of the LAR.

3.2.1.4 Attribute Alignment -- Not in Alignment, but No Adverse Consequences The licensee did not identify any attributes in this category on Table B-2 in Attachment B of the LAR.

3.2.1.5 Attribute Alignment -- Not in Alignment The licensee did not identify any attributes in this category on Table B-2 in Attachment B of the LAR.

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 Section 2.4.2 of NFPA 805. The licensee performed this evaluation by comparing the SSA against the NFPA 805 NSCA methodology requirements using NEI 00-01, Revision 2 to the NRG-endorsed process in Chapter 3 of NEI 00-01, Revision 2. The results of the review are documented on Table B-2 in Attachment B of the LAR, in accordance with NEI 04-02, Revision 2, as modified by FAQ 07-0039.

Based on the information provided in the licensee's submittal, as supplemented, the NRC staff determined the method the licensee used to perform the NSCA is acceptable 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 Section 2.4.2 of NFPA 805. The NRC staff concludes that the licensee's method is acceptable because it meets the NRG-endorsed guidance.

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 Section 9.5.1 of NUREG-0800 (Reference 36), 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 <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. In Section 4.2.1.2 of the LAR, the licensee stated that the NFPA 805 LB is to maintain the reactor in a hot standby condition (defined as Mode 3, Keff < 0.99, RCS temperature~ 350 °F) following any fire occurring with the reactor operating at power. The "At Power" safe and stable strategy includes entry into hot standby (Mode 3) and stops prior to the point of manually initiating a cool down. Some success sequences require long term make-up to the reactor coolant system (RCS), sustainment of auxiliary feedwater system water sources, or maintenance of a long term fuel supply for the emergency diesel generators (EDGs). The nuclear safety goal of NFPA 805 is to provide reasonable assurance that, should a fire occur during any operational mode or aligned configuration, the plant will not be prevented from achieving and maintaining the fuel in a safe and stable condition. The licensee identified that Table B-3 in Attachment C of the LAR documents the plant's ability to achieve and maintain NFPA 805 safe and stable conditions following shutdown from full power conditions.

In SSA RAI 01 (Reference 19), the NRC staff requested that the licensee provide additional information to support Section 1.3.1 of NFPA 805, which states 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.

Section 4.2.1.2 of the LAR states that safe and stable conditions can be maintained indefinitely until a decision is made to transition to residual heat removal (RHR) cooling. The LAR summarizes the means to maintain safe and stable conditions for extended periods of time, including inventory control, decay heat removal (OHR), electrical systems, and diesel fuel supplies. The NRC staff requested the licensee to provide additional discussion of the actions necessary beyond 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to meet the specific NSPC and to maintain safe and stable conditions. Also, the NRC staff requested the licensee to discuss the risks associated with accomplishing these actions. In its response to SSA RAI 01 (Reference 10), the licensee stated the specific capabilities that will be required to meet the performance criteria beyond 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> include the availability of procedures and personnel to perform the necessary repair and/or recovery of equipment needed to maintain safe and stable conditions for the extended period of time. The licensee further stated that to accomplish this goal, existing Emergency Operating Procedures (EOPs) and other Emergency Response Organization (ERO) procedures are currently in place to assist the plant operating staff with options to proceed and implement such actions and/or repairs.

The licensee also stated that following the initial establishment of safe and stable conditions, the ability to control reactor pressure, inventory, and temperature requires limited operator involvement as the actions are characterized by simple manipulations of valves and/or pump controls, and process instrumentation is readily available at appropriate locations. According to the licensee, Attachment C of the LAR lists a "Method of Accomplishment" for each performance goal. The licensee further stated that these methods can be maintained beyond 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The licensee stated a fire affecting plant equipment will result in activation of the ERO. The licensee further stated this activation results in staffing the ERO facilities within one hour.

According to the licensee, the TSC, Operational Support Center (OSC) and Emergency Operations Facility (EOF) staff would be in place to provide additional expertise and resources to address plant issues. The licensee stated that procedures exist to ensure adequate staffing of the EOF, TSC, OSC, and Operations Shifts for indefinite periods of time. The licensee further stated that the OSC would provide overall coordination of repair and corrective actions, as directed by TSC personnel. According to the licensee, the TSC, operations shifts, and fire brigade captain would review existing fire response procedures for the affected fire areas for the impact of the fire on affected plant equipment. The licensee also stated that frequent drills and exercises are conducted with the ERO to evaluate and maintain these capabilities.

The licensee stated that while the "Method of Accomplishment" (from Attachment C of the LAR) can be sustained for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, it is more likely that restoration of offsite power to 480 Volts Alternating Current (VAC) busses, reliable 125 Volts Direct Current power, operable redundant equipment, and available power to equipment, will be accomplished within that time frame per TSC/EOF/OSC repair actions. The licensee further stated that if it is determined that Inventory and Pressure Control or OHR can be better accomplished using safety injection and/or RHR systems, then cooldown and lowering of RCS temperature and pressure are options that would be available.

The licensee stated the following actions that are necessary beyond 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to meet specific NSPC and to maintain safe and stable conditions include options such as:

  • Alternate sources of inventory to provide assurance of adequate supply of RCS makeup water, auxiliary feedwater and diesel fuel oil, for Inventory and Pressure Control, DHR and maintaining Vital Auxiliaries.
  • Alternate pumps to provide assurance of adequate flow of RCS makeup water, auxiliary feedwater and diesel fuel oil, for Inventory and Pressure Control, DHR and maintaining Vital Auxiliaries.
  • Alternate power sources to provide assurance of adequate electrical power to needed or redundant equipment.
  • Resources from the ERO to develop contingency actions as needed to ensure that alternate means are available to maintain NSPC.

The licensee qualitatively evaluated the risks associated with the actions and activities required to maintain these conditions, and concluded that the risk impact of the failure of actions to maintain safe and stable conditions beyond 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is deemed to be very low based on the following:

  • Each of the functions required to maintain safe and stable conditions (Reactivity Control, Inventory and Pressure Control, DHR, Vital Auxiliaries, Process Monitoring) has multiple success paths.
  • There is a long time available to establish alternative long-term configurations for equipment and power supplies, and long periods of time before depletion of commodities such as fuel oil and nitrogen become concerns. Replenishing of these commodities are part of the ERO procedures and are routine actions.
  • Shift staffing requirements are adequate to provide all of the operators required to perform actions. ERO facilities would be staffed continuously. The availability of these supplemental resources to perform any of these actions or activities further ensure that these longer-term actions will be reliably accomplished.
  • Existing plant procedures provide direction for many of the actions that would be completed.
  • ERO processes ensure that any actions directed from the TSC will be controlled by the OSC, using procedures Emergency Plan Implementing Procedure (EPIP-1-10) OSC Activation), and EPIP-1 -12 (Control of Emergency Maintenance Assessment and Repair Teams). These processes ensure that extensive planning and pre-job briefs will be conducted before commencement of any field activities.
  • When the new equipment is installed as described on Table S-2 in Attachment S of the LAR, "Plant Modifications Committed," procedures will be developed to provide direction for appropriate actions to operate this equipment.

The licensee stated that when Modification Items 9 and 12 on Table S-2 in Attachment S of the LAR are complete, the 160,000 gallon Condensate Storage Tank (CST) can supply the Standby Auxiliary Feedwater (SBAFW) Pumps for a minimum of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with no operator action, or longer if operator action is taken. The licensee also stated that after that point, the SBAFW pumps can take suction from city water, through the Yard Loop or from Lake Ontario via either Service Water or Fire Pumps. The licensee further stated that these same sources can be used to refill the CST. According to the licensee, RCS makeup will be supplied by the Refueling Water Storage Tank, if available, or by another source of borated water not impacted by the fire.

The NRC staff determined that the licensee's response to SSA RAI 01 is acceptable because the licensee provided additional discussion of the actions necessary beyond 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to meet the specific NSPC and to maintain safe and stable conditions and discussed the risks associated with accomplishing these actions.

On the basis of the licensee's analysis as described in the LAR, as supplemented, the NRC staff concludes the licensee provided assurance that the fuel can be maintained in a safe and stable condition, post-fire, for an extended period of time.

3.2.3 Applicability of Feed and Bleed As stated below, 10 CFR 50.48(c)(2)(iii) limits the use of feed and bleed:

In demonstrating compliance with the performance criteria of Sections 1.5.1 (b) and (c), a high-pressure charging/injection pump coupled with the pressurizer power-operated relief valves (PORVs) as the sole fire-protected safe shutdown path for maintaining reactor coolant inventory, pressure control, and decay heat removal capability (i.e., feed-and-bleed) for PWRs is not permitted.

The NRC staff reviewed Table 5-3 in the LAR, "10 CFR 50.48(c) -Applicability/Compliance References," and Table C-1 in Attachment C of the LAR to evaluate whether the licensee meets the feed and bleed requirements. The licensee stated on Table 5-3 of the LAR that feed and bleed is not utilized as the sole fire protected SSD path for any scenario. The NRC staff verified this by reviewing the designated SSD path listed in Attachment C of the LAR for each fire area.

This review confirmed that all fire area analyses include the SSD equipment necessary to provide decay heat removal without relying on feed and bleed. In addition, all fire areas met the PB evaluation performed in accordance with Section 4.2.4 of NFPA 805, demonstrated that the integrated assessment of risk, DID, and safety margins for the fire area was acceptable. The NRC staff concludes that, based on the information provided on Table 5-3 of the LAR, as well as the fire area analyses documented in Attachment C of the LAR, that the licensee meets the requirements of 10 CFR 50.48(c)(2)(iii) because feed and bleed is not utilized as the sole fire-protected SSD path.

3.2.4 Assessment of Multiple Spurious Operations Section 2.4.2.2.1 of NFPA 805, "Circuits Required in Nuclear Safety Functions," states 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 ma I-operation of the equipment identified in 2.4.2.1, ["Nuclear Safety Capability Systems and Equipment Selection"]. 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, Section 2.4.3.2 of NFPA 805 states that the PSA evaluation shall address the risk contribution associated with all potentially risk-significant fire scenarios. Because the RI/PB approach taken used FREs in accordance with Section 4.2.4.2 of NFPA 805, "Use of Fire Risk Evaluation," adequately identifying and including potential MSO combinations is required to ensure that all potentially risk-significant fire scenarios have been evaluated.

The NRC staff reviewed Section 4.2.1.4 of the LAR, "Evaluation of Multiple Spurious Operations," and Attachment F of the LAR, "Fire-Induced Multiple Spurious Operations Resolution," to determine whether the licensee adequately addressed MSO concerns. 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 7), RG 1.205 (Reference 4),

and FAQ 07-0038 (Reference 52). The PWR Generic MSO lists in NEI 00-01, Revision 2 dated 2009 (Reference 31) was utilized. Attachment F of the LAR describes the process undertaken for evaluation of the MSO aspect required to be addressed as part of the NFPA 805 transition.

The licensee stated that the preparation phase for MSO expert panel review included developing a list of scenarios to consider during the onsite review meeting. The licensee further stated that it also included ensuring that the appropriate expertise and experience was represented. According to the licensee, representatives from Ginna fire protection and Electrical/ Appendix R Engineering, Ginna Operations, Ginna PRA Engineering, and supporting staff were in attendance at the expert panel.

Attachment F of the LAR stated that the licensee conducted an initial expert panel review in August 2008 and a second review in in March 2010 to address all additions, deletions, and/or changes to the MSO assessment that have occurred due to post-expert panel reviews and in consideration of the most current information available from the Pressurized Water Reactor Owner's Group. The licensee stated that all scenarios identified as new or changed were reviewed and dispositioned by the reconvened expert panel. The licensee further stated that changes or clarifications to the MSO report were incorporated as needed. According to the licensee, the expert panel conducted a step-by-step discussion, reviewed plant documents, postulated scenarios, identified potential consequences and likelihoods, held discussions on operator responses and proposed various courses of action. Also, the licensee stated that the results of the expert panel reviews of each potential MSO were documented.

Attachment F of the LAR describes the process the licensee utilized to address MSOs, which follows the guidance of FAQ 07-0038 (Reference 52). That process includes five steps:

( 1) identify potential MS Os 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 Attachment F of the LAR, under the results for Steps 3, 4, and 5, the MSOs identified in Steps 1 and 2 were incorporated in the FPRA model. The licensee stated that the MSO combination components of concern were then evaluated for inclusion into the NSCA. The licensee further stated that as necessary, components were added to the NSCA Equipment List and Logics; and circuit analysis and cable routing was performed. According to the licensee, 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 FPRA quantified the fire-induced risk model containing the MSO pathways. Also, the licensee stated that the MSO contribution is included in the FPRA results.

The licensee stated that a comprehensive review of each of the MSO items was undertaken.

The licensee further stated that this process and the associated analyses resulted in some items no longer being recommended for inclusion in the model while others were confirmed to warrant inclusion in the model. According to the licensee, as the specific reviews were completed and documented, additional logic was added to the FPRA model. The licensee also stated that the PRA model changes and the basis for inclusion or exclusion of the identified MSOs are documented in the MSO report. The rationales for inclusion or exclusion of MSOs from the NSCA were also documented in the MSO evaluation.

The NRC staff reviewed the licensee's expert panel process for identifying circuits susceptible to MSOs as described above and concludes that the licensee adopted a systematic and comprehensive process for identifying MSOs to be analyzed utilizing available industry guidance. Furthermore, the process used ensured that the FREs appropriately identify and include risk significant MSO combinations. Based on the information provided by the licensee, the NRC staff concludes that the licensee's approach for assessing the potential for MSO combinations is acceptable for use.

3.2.5 Establishing Recovery Actions Section 1.6.52 of NFPA 805, "Recovery Action," defines a RA as:

Activities to achieve the nuclear safety performance criteria that take place outside the main control room or outside the primary control station(s) for the equipment being operated, including the replacement or modification of components.

Section 4.2.3.1 of NFPA 805 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.

Section 4.2.4 of NFPA 805, "Performance-Based Approach," states that:

When the use of recovery actions has resulted in the use of this approach, the additional risk presented by their use shall be evaluated.

The NRC staff reviewed Section 4.2.1.3 of the LAR, "Establishing Recovery Actions," and Attachment G of the LAR, "Recovery Actions Transition," to evaluate whether the licensee meets the associated requirements for the use of RAs per NFPA 805. The licensee stated in Attachment G to the LAR that in accordance with the guidance provided in NEI 04-02 (Reference 7), FAQ 07-0030, "Establishing Recovery Actions" (Reference 50), and RG 1.205 (Reference 4), the following methodology was used to determine RAs required for compliance (i.e., determining the population of post-transition RAs):

1. Define the primary control station(s) (PCSs) and determine which pre-transition operator manual actions (OMAs) are taken at PCS(s) (Activities that occur in the MCR are not considered pre-transition OMAs). Activities that take place at either a PCS(s) or the MCR are not RAs by definition.
2. Determine the population of RAs that are required to resolve VFDRs, to meet the risk acceptance criteria or to maintain a sufficient level of defense-in-depth.
3. Evaluate the additional risk presented by the use of RAs required to demonstrate the availability of a success path.
4. Evaluate the feasibility of the RAs.
5. Evaluate the reliability of the RAs.

OMAs meeting the definition of an RA are required to comply with the NFPA 805 requirements outlined above. Some of these OMAs may not be required to demonstrate the "availability of a success path," in accordance with Section 4.2.3.1 of NFPA 805, but may still be required to be retained in the RI/PB FPP because of DID considerations described in Section 1.2 of NFPA 805. The licensee did not differentiate between an RA that is needed to meet the NSCA and one retained to provide DID. In each instance, the licensee determined whether a transitioning OMA was an RA to meet risk acceptance criteria, for DID considerations, or not necessary for the post-transition RI/PB FPP.

The licensee stated it has 2 locations designated as PCSs as defined in RG 1.205:

  • The Auxiliary Building Emergency Local Instrument Panel (ABELIP), which is required for local control and operating of Charging Pump 1A; and
  • The Intermediate Building Emergency Local Instrument Panel (IBELIP), which is required for operation of the Turbine-Driven AFW [TDAFW] pump.

The licensee stated in Attachment G and Attachment S of the LAR that the plant committed to installing new equipment. Modifications ESR-11-0050, ESR-12-0143, and ESR-12-0144 include a dedicated diesel generator powering an existing standby auxiliary feedwater pump

and a new injection pump. According to the licensee, as a part of the NFPA 805 project, this new equipment will be utilized instead of charging pump 1A and the TDAFW as the primary equipment to address VFDRs associated with loss of feedwater and loss of charging. The licensee stated that the SSD strategy changed to utilize this new equipment. The licensee further stated that the operation of this equipment is much simpler and quicker to perform than the current actions, and is an enhancement to fire safety. Further, the licensee also stated that as it is isolated from the rest of the plant, it will remain free from fire damage for all fire scenarios of concern. According to the licensee, this shutdown strategy does not require the use of the ABELIP and IBELIP.

In SSA RAI 04 (Reference 19), the NRC staff requested that the licensee provide a more detailed description of Modification ESR-11-0050 that includes identification of the planned physical hardware changes or additions, the communications between local control stations, the operation of the equipment, and the command and control process with the new modification.

In its response to SSA RAI 04 (Reference 10), the licensee stated the Diesel Driven Standby Auxiliary Feedwater (ESR-11-0050) Project is providing the SBAFW pumps with an additional source of de-ionized water (160,000 gallon tank) and power (1 MWe diesel generator). The licensee further stated that the new tank will be located adjacent to the SBAFW building.

According to the licensee, the tank contents will be an alternate source of water for the SBAFW pumps. The licensee also stated that piping and manual valves will be added so that the SBAFW pumps can draw suction from the new tank and the existing safety related service water system will still be available. SBAFW pump discharge can be recirculated, through a breakdown orifice, back to the tank to facilitate pump testing, or can be discharged to the SGs.

Piping will be added in order to transfer de-ionized water from the tank into the condensate system in the auxiliary building, or from the de-ionized water header to fill the new tank. Piping will be added to allow for a continuous flow of heated water, via a circulating pump and heater, to provide freeze protection for the tank. The licensee stated that the tank will be equipped with level and temperature indication, overflow piping, vacuum/relief valve, vent, drain, sample line, manways, and numerous spare penetrations to support any future needs. The licensee further stated that the new tank will be of sufficient capacity to ensure DHR immediately following a reactor trip for approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The licensee stated the new DG and associated distribution equipment and cabling will be installed in addition to the existing SBAFW building. The licensee further stated that it will have the capability of providing power to both SBAFW pumps. According to the licensee, the DG and normal loads will be powered from off-site residential power which is completely independent from the plant and the normal offsite power to the plant. The licensee stated that the DG can be locally started or will auto start upon loss of residential power to power the normal loads. The licensee further stated that starting of the SBAFW pumps with the DG supplying power will be accomplished manually using newly installed transfer switches and associated distribution equipment. According to the licensee, once an auxiliary operator or other designated individual is dispatched by the on-shift operations staff to start the DG and/or start the SBAFW pumps or if the DG is already running, communications will be maintained via radio between control stations. Also, the licensee stated that instructions and command will come from shift supervision and will be directed and controlled per new or revised Emergency Response Procedures. The NRC staff determined that the licensee's response to the SSA RAI 04 is acceptable because the licensee provided a more detailed description of Modification

ESR-11-0050 that includes identification of the planned physical hardware changes or additions, the communications between local control stations, the operation of the equipment, and the command and control process with the new modification.

For the reasons the licensee stated above, Table G-1 in Attachment G of the LAR, "Recovery Actions and Activities Occurring at the Primary Control Stations," does not identify any required activities that occur at the PCSs (i.e., PCS actions are no longer required in order to address any VFDRs), since with the installation of the new equipment, there are none required. Since the new equipment is not controlled from a PCS, all actions associated with it are considered RAs. Activities that are identified as RAs that are necessary to address risk are identified as RISK on Table G-1 in Attachment G of the LAR.

The licensee utilized the guidance in RG 1.205, Revision 1 (Reference 4) for addressing RAs.

This included consideration of the definition of PCS and RA, as clarified in the RG 1.205, Revision 1 (Reference 4). 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. Any OMAs required to be performed outside the MCR and not at the PCS were considered RAs.

The licensee stated that all credited RAs, as listed in Attachment G of the LAR were subjected to a feasibility review. The feasibility criteria used in the licensee's assessment process were based on the endorsed guidance described in NEI 04-02 (Reference 7) as modified by FAQ 07-0030 (Reference 50) and each of the 11 individual feasibility attributes were addressed.

Table G-1 in Attachment G of the LAR, describes each RA associated with disposition of a VFDR from the fire area assessments as documented in Attachment C of the LAR. The feasibility review was based on documentation only, including previous feasibility evaluations for SSD OMAs. The licensee included Implementation Item 10 and Implementation Item 13 of Table S-3 in Attachment S of the LAR, to revise post-fire SSD procedures and training as necessary to incorporate updated NSCA strategies. The NRC staff determined that this action is acceptable because it will incorporate the provisions of NFPA 805 in the licensee's FPP and would be required by the proposed license condition.

Based on the above considerations, the NRC staff concludes that the licensee followed the endorsed guidance of NEI 04-02 (Reference 7) and RG 1.205 (Reference 4) to identify and evaluate RAs in accordance with NFPA 805, and meets the regulatory requirements of 10 CFR 50.48(c). The NRC staff also concludes that the feasibility criteria applied to RAs are acceptable based on conformance with the endorsed guidance contained in NEI 04-02 (Reference 7) subject to completion of implementation items 1O and 13 of Table S-3 in Attachment S of the LAR, as stated in the proposed license condition.

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 Section 2.4.2 of NFPA 805, regarding the process used to perform the NSCA. The NRC staff concluded that the declared safe and stable condition proposed is acceptable and that the licensee's process adequately and appropriately identified and located the systems, equipment, and cables required to ensure achieving and maintaining the fuel in a safe and stable condition, as well as to meet the NFPA 805 NSPC.

In accordance with 10 CFR 50.48(c)(2)(iii), the NRG staff verified, through review of the documentation provided in the LAR, that feed and bleed is not the sole fire-protected SSD path for maintaining reactor coolant inventory, pressure control, and decay heat removal capability.

The NRG staff also reviewed the licensee's process to identify and analyze MSOs. Based on the LAR, as supplemented, the NRG staff concludes that the process used to identify and analyze MSOs is comprehensive and thorough. Through the use of an expert panel process, in accordance with the guidance of RG 1.205 (Reference 4), NEI 04-02 (Reference 7), and FAQ 07-0038 (Reference 52), potential MSO combinations were identified and included as necessary in the NSCA, as well as the applicable FREs. The NRG staff also concludes that the approach the licensee used for assessing the potential for MSO combinations is acceptable because it was performed in accordance with NRG-endorsed guidance.

The NRG staff concludes that, based on the information provided in the LAR, as supplemented, the process used by the licensee to review, categorize and address RAs during the transition from the existing deterministic fire protection LB to a RI/PB FPP is consistent with the NRG-endorsed guidance contained in NEI 04-02 (Reference 7) and RG 1.205 (Reference 4),

regarding the identification of RAs and other actions required to be taken at a PCS. The licensee identified those actions that meet the definition of a RA provided in Section 1.6.52 of NFPA 805.

Provided the licensee completes modification items 8 and 12 as described on Table S-2 in Attachment S of the LAR, and implementation items 10 and 13 as described on Table S-3 in Attachment S of the LAR, the NRG staff concludes that there is reasonable assurance that the regulatory requirements of 10 CFR 50.48(c) and NFPA 805 for NSCA methods are met.

3.3 Fire Modeling NFPA 805 (Reference 3) allows both FM and FREs as PB alternatives to the deterministic approach outlined in the standard. These two PB approaches are described in Sections 4.2.4.1 and 4.2.4.2 of NFPA 805, respectively. Although FM and FRE 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. Section 1.6.18 of NFPA 805, defines a fire model as a "mathematical prediction of fire growth, environmental conditions, and potential effects on structures, syst.ems, or components based on the conservation equations or empirical data."

The NRG staff reviewed Section 4.5.2 of the LAR "Performance-Based Approaches" (Reference 9), which describes how the licensee used FM as part of the transition to NFPA 805. In Section 4.5.2.1 of the LAR, the licensee indicated that, in lieu of the FM approach (Section 4.2.4.1 of NFPA 805), the FRE approach (Section 4.2.4.2 of NFPA 805) was used for the transition to NFPA 805. In Section 4.5.1.2 of the LAR, "Fire Model Utilization in the Application," the licensee indicated that FM was performed as part of the FPRA development.

The NRG staff reviewed the technical adequacy of the FRE, including the supporting FM analyses, as documented in Section 3.4.2 of the SE, to evaluate compliance with the NSPC.

The licensee did not propose any FM methods to support PB evaluations in accordance with Section 4.2.4.1 of NFPA 805, as the sole means for demonstrating compliance with the NSPC.

The NRC staff concludes that there are no plant-specific FM methods acceptable for use to support compliance with Section 4.2.4.1 of NFPA 805, as part of this LAR supporting the transition to NFPA 805.

3.4 Fire Risk Evaluations This section addresses the licensee's FRE PB method, which is based on Section 4.2.4.2 of NFPA 805. The licensee chose to use only the FRE PB method in accordance with Section 4.2.4.2 of NFPA 805. The FM PB method in Section 4.2.4.1 of NFPA 805 was not used for this application.

Section 4.2.4.2 of NFPA 805, "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, 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"].

3.4.1 Maintaining Defense-in-Depth and Safety Margins Section 4.2.4.2 of NFPA 805, requires that the "use of fire risk evaluation for the PB approach shall consist of an integrated assessment of the acceptability of risk, defense-in-depth, and safety margins."

3.4.1.1 Defense-in-Depth Section 1.2 of NFPA 805, states that:

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:

  • Preventing fires from starting;
  • Rapidly detecting fires and controlling and extinguishing promptly those fires 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 Section 4.5.2.2 of the LAR, "Fire Risk Approach," and Table B-3 in Attachment C of the LAR, "NEI 04-02 Table B Fire Area Transition," as well as the associated supplemental information, in order 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 Section 5.3 of NEI 04-02, "Plant Change Process," which includes a detailed consideration of DID and safety margins as part of the change process. The licensee developed a methodology for evaluating DID which defines each of the three DID elements identified in Section 1.2 of NFPA 805, and in more detail in its response to PRA RAI 22.a (Reference 12), referred to as echelons 1, 2, and 3, respectively. According to the response, the licensee DID evaluations are consistent with Chapter 3 of NPFA 805, "Fundamental Fire Protection Program and Design Elements," focus on potential enhancements that may be required to maintain the balance of DID echelons, and considers FPRA risk insights. The licensee provided examples on how FPRA risk insights could be used qualitatively to evaluate echelons 1, 2, and 3. The licensee considered a DID requirement to be satisfied if the proposed change does not result in a substantial imbalance among these echelons. DID was performed on a fire area basis, and fire protection features and systems relied upon to ensure DID were identified as a result of the assessment of DID. However, the licensee determined that the FREs performed for VDFRs did not require DID modifications or actions. Table B-3 in Attachment C of the LAR documents the results of the licensee's review of fire suppression and fire detection systems at Ginna.

The NRC staff review found that interfacing system loss-of-coolant accidents (ISLOCAs) were in the dominant cutsets for the LERF results on Table W-3 in Attachment W of the LAR. In its response to PRA RAI 22.c (Reference 12), regarding the DID evaluation of ISLOCAs, the licensee explained that ISLOCA is no longer a contributor to LERF so that there is no need to consider DID for ISLOCA. In its response to PRA RAI 22.01 (Reference 14), the licensee explained that this conclusion is a result of crediting makeup for ISLOCA diameter sizes within the makeup capacity of the planned injection system. In its response to PRA RAI 22.01.1 (Reference 15), regarding the basis for this ISLOCA success criterion, the licensee noted that future procedures will provide for maintaining primary inventory indefinitely to avoid core damage using the planned injection system with alternate water supplies from the planned condensate tank and the lake. Therefore, the ISLOCA scenario mitigation relies on a safe shutdown strategy using alternate water sources for injection. The NRC staff's safe shutdown evaluation is provided in Section 3.2.2 of the SE.

Based on its review of the response to PRA RAI 22 regarding DID, the LAR and the FREs, 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 and concludes that the methodology as proposed in the LAR, as supplemented, adequately 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 Section 2.4.4.3 of NFPA 805 states that:

The plant change evaluation shall ensure that sufficient safety margins are maintained.

Section 5.3.5.3 of NEI 04-02, "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 LB (e.g., FSAR, supporting analyses, etc.) are met, or the change provides sufficient margin to account for analysis and data uncertainty.

Section 4.5.2 of the LAR, "Performance-Based Approaches," states that safety margins were considered as part of the transition process. Section 4.5.2.2 of the LAR states that the licensee evaluated each VFDR against the safety margin criteria contained in NEI 04-02.

Attachment C of the LAR states that:

  • Safety Margins for the analyses supporting the FRE of the fire area was evaluated and accounted for potential impacts from FM and the plant system performance, including the PRA logic model;
  • All analyses and assessment have been performed utilizing accepted techniques and industry accepted standards; and,
  • Safety analysis acceptance criteria in the LB (e.g., FSAR, supporting analyses) have been considered and provide sufficient margin to account for analysis and data uncertainty.

In its response to PRA RAI 22 (Reference 12), the licensee provided additional information addressing safety margin. The licensee noted that FM was conducted using codes and guidance developed by the commercial nuclear industry and the NRC staff. The FPRA logic model was also developed in accordance with ASME/ANS RA-Sa-2009, "Addenda to ASME/ANS RA-S2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications" (Reference 34), and RG 1.200, Revision 2 (Reference 33). Furthermore, the licensee stated that FREs found that an adequate safety margin was maintained without further enhancements required.

The safety margin criteria described in Section 5.3.5.3 of NEI 04-02 (Reference 7), and the LAR, as supplemented, are consistent with the criteria as described in RG 1.174, Revision 2 (Reference 32) and, therefore, acceptable. Based on its review of the LAR and the response to PRA RAI 22, the NRC staff concludes that the licensee's approach has adequately addressed safety margin in the implementation of the FRE process.

3.4.1.3 Defense-in-Depth and Safety Margin Conclusion The licensee's FRE process included a review of fire protection DID and safety margins. The individual FREs and Table 8-3 in Attachment C of the LAR document the results of the DID and safety margin review. The NRC staff concludes that the licensee's evaluation in regard to DID and safety margins is acceptable because the licensee's process and results followed the endorsed guidance in NEI 04-02, Revision 2 and are consistent with the NRC staff guidance in RG 1.205, Revision 1 (Reference 4), and RG 1.174, Revision 2.

3.4.2 Quality of the Fire Probabilistic Risk Assessment In reviewing a RI LAR, the NRC staff evaluates the validity of the plant-specific PRA models and their application as proposed in the LAR. 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 NRC 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 Sections 4.5.1, "Fire PRA Development and Assessment," and 4.7, "Program Documentation, Configuration Control, and Quality Assurance," Attachments C, "NEI 04-02 Table 8 Fire Area Transition," U, "Internal Events PRA Quality," V, "Fire PRA Quality," and W, "Fire PRA Insights," of the LAR.

The licensee developed its FPRA model using the guidance of NUREG/CR-6850, "EPRl/NRC-RES Fire PRA Methodology for Nuclear Power Facilities" (Reference 39)

(Reference 40), and (Reference 41 ). The model addresses both Level 1 (CDF) and partial Level 2 (i.e., LERF only) PRA during at-power conditions. The licensee modified its internal events PRA (IEPRA) model to capture the effects of fire, both as the initiator of an event and to characterize the subsequent potential failure modes for affected circuits or individual plant SSCs (targets), including fire-affected human actions.

The licensee did not identify any known outstanding plant changes that would require a change to the FPRA model, or any planned plant changes that would significantly impact the PRA model, beyond those identified and scheduled to be implemented as part of the transition to a FPP based on NFPA 805. Therefore, the NRC staff finds that the FPRA model for Ginna represents the as-built, as-operated and maintained plant as it will be configured after full implementation of NFPA 805.

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.

Additionally, described in Section 3.8.3 of the SE, the licensee has a program for ensuring that developers and users of these models are appropriately trained and qualified.

3.4.2.1 Internal Events PRA Model The Pressurized Water Reactor Owners Group performed a full scope peer review of the Ginna IEPRA in June 2009 against ASME RA-Sb-2005 (Reference 72), as clarified by RG 1.200, Revision 1. The licensee addressed the peer review findings, assessing their impact on the

FPRA in Attachment U of the LAR. In addition, according to LAR Attachment U, the licensee performed a gap assessment between ASME RA-Sb-2005 as clarified by RG 1.200, Revision 1 and ASME/ANS RA-Sa-2009 as clarified by RG 1.200, Revision 2. The licensee did not identify any significant issues for the FPRA from this gap assessment.

For many ASME/ANS PRA Standard supporting requirements (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 CC-I 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 SR, the PRA reviewer from the peer review team designates one of the CCs or indicates that the SR is met or not met.

Table U-1 in Attachment U of the LAR, provides the licensee's dispositions to all 25 facts and observations (F&Os) that were findings related to the IEPRA. 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.

As described in Attachment U of the LAR, the licensee dispositioned each F&O by assessing the impact of the F&O on the FPRA and the results for the LAR. In PRA RAI 25 (Reference 19),

the NRC staff requested that the licensee 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 Attachment U of the LAR to determine whether it had any significant impact for the FPRA. The NRC staff's review and conclusions for Ginna's resolution of each F&O is summarized in the NRC's Record of Review (Reference 73). The NRC staff found the F&Os did not have any significant impact on the FPRA, and therefore, the individual evaluations of the F&Os is not discussed in this SE.

The NRC staff's review identified weaknesses in the PRA modeling of common cause failure (CCF) for components that assumed staggered testing when staggered testing was not in place.

The licensee resolved this issue by updating the CCF probabilities using an appropriate approximation in the FPRA in its response to PRA RAI 44.01.b (Reference 17), for transition risk.

In its response to PRA RAI 21 (Reference 10), the licensee stated that there were no changes made to the IEPRA that are consistent with the definition of a "PRA upgrade," as defined by the ASME/ANS PRA Standard, since the last full-scope peer review of the IEPRA model. The licensee concluded, therefore, that no focused scope peer reviews of the IEPRA subsequent to the full scope peer review were necessary.

As a result of its review of the LAR, as supplemented, the NRC staff concludes that the Ginna IEPRA is technically adequate such that its quantitative results 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. To reach this conclusion, the NRC staff reviewed all F&Os provided in Attachment U of the LAR by the peer reviewers and determined that the resolution of every F&O supports the conclusion that the quantitative results are adequate. Accordingly, the NRC staff concludes that the licensee demonstrated that the IEPRA meets the guidance in RG 1.200 that 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 NFPA 805 application.

3.4.2.2 Fire PRA Model The licensee evaluated the technical adequacy of the Ginna FPRA model by conducting a peer review using NEI 07-12 (Reference 74), and the ASME/ANS RA-Sa-2009 PRA standard (Reference 34), as clarified by RG 1.200, Revision 2 (Reference 33). The full scope peer review of the Fire PRA was performed in June, 2012.

Table V-1 in Attachment V of the LAR provides the licensee's resolution of all 19 F&Os that were findings and the licensee's assessment of all findings, which included F&Os against SRs that were met at CC-II, not met, or achieved CC-I. Table V-2 in Attachment V of the LAR identifies all SRs that were determined by the peer review not to be met or met only at CC-I, and provided an evaluation of those SRs. In all cases the SR evaluations refer to a finding already presented on Table V-1 in Attachment V of the LAR.

The NRC staff reviewed the licensee's resolutions of all of the F&Os and SR evaluations provided on Tables V-1 and V-2 in Attachment V of the LAR to determine the technical adequacy of the FPRA for the NFPA 805 application. The NRC staff requested additional information for the review of some F&Os. The NRC staffs review and conclusion of the licensee resolution of each of the F&Os is summarized in the NRC's record of review (Reference 75) .

The NRC staff's review identified that the peer review considered some plant response model SRs as not applicable. In its response to PRA RAI 33 (Reference 10), and PRA RAI 33.01 (Reference 14), the licensee explained that there were no new initiating events, accident sequences, and success criteria for the FPRA because these were already included in the IEPRA. The NRC staff reviewed the description provided by the licensee in these RAI responses for these SRs and finds that the peer review assignment of a not applicable status is reasonable, and considers this issue resolved.

In its response to PRA RAI 21 (Reference 10), regarding whether upgrades were made to the FPRA since the peer review, the licensee stated that there were no FPRA model upgrades as defined by the ASME/ANS PRA Standard since the last full-scope peer review. The licensee concluded, therefore, there was no need for follow-up focused scope peer reviews.

The NRC staff reviewed FPRA methods used in the Ginna FPRA. The FPRA used previously NRG-accepted methods, updated methods to be consistent with current NRG-accepted guidance, or provided justification for methods which differed from current NRG-accepted methods. The NRC staff's evaluation of these methods is discussed below.

The NRC interim staff guidance updated credit given for CPT in FPRAs based on NRG-sponsored testing and data analysis which showed that the effect of any CPT reduction to the hot short-induced spurious operation likelihood could not be substantiated.

NUREG/CR-7150, "Joint Assessment of Cable Damage and Quantification of Effects from Fire (JACQUE FIRE)," Volume 2 (Reference 76), guidance reflects this interim guidance on CPTs.

In its response to PRA RAI 27 (Reference 12), regarding the updated CPT guidance, the licensee confirmed that the Ginna FPRA model does not credit the factor of two reduction for CPTs. The NRC staff finds that not crediting a factor two reduction for CPTs is acceptable since it follows current NRG-accepted guidance on this technical issue, and this method is reflected in the FPRA in the licensee's response to PRA RAI 44.01.b (Reference 17), for the transition risk.

In response to PRA RAI 11.01 (Reference 14), regarding VFDR modeling associated with higher spurious actuation probabilities, the licensee indicated that the FPRA was updated to use the NUREG/CR-7150 method for spurious actuation likelihood for double break electrical isolation. The NRC staff finds that this is an acceptable method which replaced the NUREG/CR-6850 circuit failure mode likelihood analysis (Task 10) Option 2 method. In response to PRA RAI 44.02b (Reference 18), the licensee confirmed that this update was reflected in the FPRA in its response to PRA RAI 44.01.b for transition risk.

In its response to PRA RAI 19.03 (Reference 16), the licensee stated that the FPRA was updated to include NUREG/CR-7150 guidance on direct current hot short duration. As a result, Attachment S of the LAR was revised and in response to PRA RAI 44.02b (Reference 18), the licensee confirmed that the FPRA model reflects the elimination of some modifications. The licensee also confirmed that this update was reflected in the FPRA in its response to PRA RAI 44.01.b for transition risk. The NRC staff finds this method update acceptable because it uses an NRG-accepted method, and has been included in the transition risk.

Subsequent to the LAR submittal, the NRC staff endorsed FAQs regarding junction boxes and cable fires. In its response to PRA RAI 30.01 (Reference 14), regarding FAQ 13-0005, "Cable Fires Special Cases: Self-Ignited and Caused by welding and Cutting" (Reference 63) and FAQ 13-0006, "Modeling Junction Box Scenarios in a Fire PRA" (Reference 64), the licensee explained that the FPRA will be updated to be consistent with these FAQs. The licensee confirmed these updates were included in the FPRA in its supplemental response to PRA RAI 44.01 (Reference 18), in its response to PRA RAI 44.01.b for transition risk. The NRC staff finds these method updates acceptable because they follow the NRG-accepted methods described in FAQ 13-0005 and FAQ 13-0006, and have been included in the transition risk.

The NRC staff also endorsed FAQ 13-0004, "Clarifications on Treatment of Sensitive Electronics" (Reference 77), subsequent to the LAR submittal. In its response to PRA RAI 28 (Reference 12), the licensee evaluated the electrical equipment in the Ginna FPRA against the guidance on sensitive electronics in FAQ 13-0004. The licensee's review determined that the vast majority of sensitive electronics are contained in an enclosed cabinet for which the heat flux damage threshold for the thermoset cables applies. The remaining sensitive electronics were related to MCR indications or emergency lighting which do not result in a need to modify the FPRA because their failures have no impact on the MCR abandonment analysis or on the PRA modeling of operator actions. Based on the licensee's response, the NRC staff determined that the Ginna FPRA is consistent with the NRG-accepted guidance in FAQ 13-0004, which is reflected in the FPRA in the responses to PRA RAI 44.01.b for transition risk.

In response to PRA RAI 17 (Reference 10), regarding the use of FAQ 08-0053, "Kerite-FR Cable Failure Thresholds" (Reference 57), the licensee noted that Kerite cables were modeled in the FPRA as thermoplastic cables, except for certain cables in Battery Room A, which the licensee verified were constructed with thermoset material and were modeled using the

thermoset damage criteria. The NRC staff finds the treatment of Kerite is included in the FPRA model, consistent with NRC-accepted guidance in FAQ 08-0053, and in the response to PRA RAI 44.01.b for transition risk, and, therefore, acceptable.

The NRC staff's review noted that the human error probabilities (HEPs) were not truncated at a floor due to independence considerations. In its response to PRA RAI 36 (Reference 12),

regarding the use of a HEP floor for the FPRA sequences, the licensee performed a sensitivity analysis using a floor with the NUREG-1921 (Reference 47), suggested value of 1E-5 for combinations of HEPs. The sensitivity analysis showed negligible effect on the CDF and LERF.

The NRC staff determined that the Ginna HEP modeling method related to no floor due to independence considerations is acceptable since the sensitivity study demonstrates that it has essentially no effect on CDF and LERF, and therefore, it is consistent with the guidance in NUREG-1921 on HEP floor values.

In its response to PRA RAI 35 (Reference 10), regarding including fire-induced instrumentation failure in the HRA, the licensee described the binning of instrumentation impacts on HRA into the following categories: (1) action is normal because cues are available; (2) action reliability is reduced by degraded cues from instruments caused by fire; and, (3) action is completely failed because of no cues. The HRA assessment included the potential for instrumentation to initiate undesired operator action. In addition, the licensee stated that any instrumentation associated with a credited operator action has been verified to be available for the fire scenarios in which the operator action is credited. The NRC staff finds this approach to be consistent with NUREG/CR-6850 and NUREG-1921 guidance regarding modeling instrumentation and HRA, and that the licensee has verified the credited instrumentation for HRA to be available for the appropriate FPRA scenarios, and therefore, acceptable.

In its response to PRA RAI 12 (Reference 10), the licensee described the HEP screening approach for operator actions in the main control room (MCR) which differed from guidance in NUREG-1921 and NUREG/CR-6850. Screening approaches in NUREG-1921 and NUREG/CR-6850 assign HEPs that are larger when the fire is on-going than at later times. The Ginna screening approach used for in-MCR HEPs for non-abandonment fires did not reflect such a difference. In its response to PRA RAI 12.01 (Reference 14), the licensee noted that the screening approach was only used for IEPRA HFEs carried over to the FPRA. The licensee's review of the FPRA did not identify internal PRA HFEs, which used screening HEPs with a Fussell-Vesely importance measure greater than 0.005 or a risk achievement worth greater than

2. Based on this, the NRC staff finds that these screening HEPs are not significant basic events as defined in RG 1.200, Revision 2. The NRC staff finds that the Ginna screening approach does not result in a need to perform additional detailed HRA in accordance with SR HRA-C1 CC-II requirements for significant basic events because the screening basic events are non-significant, and, therefore, acceptable.

According to the licensee's response to PRA RAI 11.02 (Reference 14), MCR abandonment operator actions are modeled for fires in the MCR due to loss of habitability or loss of control. In its response to PRA RAI 11.02, the licensee stated that fires in the MCR were evaluated by modeling individual fire scenario sequences. MCR abandonment due to loss of habitability or loss of control is integrated into the PRA logic model, and thus the set of respective scenarios captures the range of complexity of fire induced equipment damage and spurious operations.

The scenarios include the contribution from random equipment failures, cognitive and execution

errors, and fire induced failures producing a range of conditional core damage probability (CCDP) reflecting the complexity of mitigating core damage for abandonment. In the licensee's response to PRA RAI 11.02.1 (Reference 15), the licensee indicated that MCR abandonment due to LOC is credited for those scenarios where the CCDP due to fire damage is greater than or equal to 0.1, prior to the credit for abandonment. The NRC staff finds that the use of a CCDP greater than 0.1 is a reasonable measure that can identify scenarios with extensive fire damage that should lead to MCR abandonment on LOC. For these scenarios, an HEP is also established to account for the likelihood that operations failed to enter the abandonment procedure when required. The licensee stated that at least one failure event (e.g., a spurious operation) is required in every core damage scenario and, therefore, the CCDP is always less than one for all LOC driven abandonment scenarios. The staff finds that MCR abandonment due to LOC is acceptable since a range of complexity from equipment damage and spurious operations are evaluated, this range is reflected in the CCDPs/conditional large early release probabilities, and that the condition on which LOC occurs is based upon the extent of fire damage in the scenario.

The FPRA model accounts for the impacted equipment whether the fire initiator was in the MCR or in a fire area outside the MCR. According to the licensee's response to PRA RAI 11.02.1 (Reference 15), the FPRA used for the FREs do not credit MCR abandonment as a result of loss of control associated with fires outside the MCR. However, the licensee noted that operations may use abandonment procedures for LOC abandonment fire scenarios outside the MCR. The NRC staff finds that not crediting abandonment operator actions for fires outside the MCR is conservative and, therefore, acceptable.

The current Ginna fire procedures address fires in the MCR as well as fire areas outside the MCR. In its response to PRA RAI 23 (Reference 10), the licensee plans on updating procedures for fires which cause MCR abandonment, as well as for fires in other safe shutdown areas, to include symptom-based alternative actions in the EOPs. The procedure update will include all of the credited actions related to abandonment actions in the FPRA model which are listed on Table G-1 in Attachment G of the LAR. According to the response to PRA RAI 11.02, the MCR abandonment procedure will also include PRA-credited actions at the shutdown panels (i.e., the ABELIP and the IBELIP), as a backup to the planned modifications described in Attachment S of the LAR. This procedure update will be done in accordance with implementation item 19 and the NRC staff considers this acceptable because it will incorporate the provision of NFPA 805 in the FPP and would be required by the proposed license condition.

If any procedural changes impact the HRA, the HRA update will be performed in accordance with implementation item 15 of Table S-3 in Attachment S of the LAR.

Ginna's MCB ignition frequency apportioning approach, described in its response to PRA RAI 32.01 (Reference 14), uses the Bin 4 MCB frequency from Supplement 1 of NUREG/

CR-6850. Since the MCB does not have internal walls, the MCB frequency is apportioned to the panels based on cable counts while preserving the total frequency for the whole MCB. In its response to PRA RAI 32 (Reference 10), the licensee explained that the MCB fire risk analysis also used a different approach in modeling the probability for severity factor and non-suppression probabilities than that given in NUREG/CR-6850, Appendix L, Figure L- 1.

The licensee's approach took no credit for the severity factor and less credit for non-suppression probabilities. Furthermore, the Ginna analysis considers a range of scenarios, from small fires, to fires which propagate outside the panel of origin and thus damage the entire

originating and neighboring panel, leading to MCR abandonment. The NRC staff finds that the MCB analysis is acceptable since the model considers a range of fire scenarios including those failing the entire originating and target panel, the frequency is applied in a manner consistent with allocation of fire frequency for this case where NUREG/CR-6850 Appendix L is not applied, and the method used for probabilities of severity factor and non-suppression is conservative with respect to Appendix L.

The NRC staff endorsed FAQ 14-0008, "Main Control Board Treatment" (Reference 65),

subsequent to the LAR submittal. In its response to PRA RAI 11.02 (Reference 14), regarding the treatment of the back panels in the FPRA, the licensee noted that the panels in the back of the MCB are detached from the MCB and are counted as electrical cabinets (i.e., assigned a Bin 15 frequency). The NRC staff finds that this treatment is consistent with FAQ 14-0008 and, therefore, acceptable.

For the modeling of transient fire scenarios in the Fire PRA, the licensee employed a transient zone approach as explained in its responses to PRA RAI 15 (Reference 12), and PRA RAI 30 (Reference 12). For this approach, the compartment floor areas were divided into transient zones. Cable trays within or nearby the transient zone boundary being analyzed were assumed to be damaged by the transient fire, and conduits in any adjacent transient zone were assumed to be damaged. When the results of the conduit evaluation indicated a significant risk increase, field walkdowns were conducted to identify the exact conduit location, and the conduit damage evaluation was refined with this information. This transient zone approach ensures that "pinchpoints" for transient fires are not missed, and that targets within the ZOI, including those outside the transient zone boundary, are damaged by fire in the analysis. The NRC staff finds the transient zone modeling approach to be acceptable because it appropriately includes transient fires and the respective fire damage to FPRA targets, and is included in the response to PRA RAI 44.01.b for transition risk.

The Ginna FPRA initially used a screening approach for multi-compartment analysis (MCA) by screening on fire scenario frequency of 1E-8/yr. In its response to PRA RAI 42 (Reference 12),

the licensee updated the screening MCA to no longer use this approach; rather, all multi-compartment scenarios that provide a contribution to CDF or LERF are retained in the baseline model. The NRC staff finds this issue resolved because all MCA fire scenario frequencies which contribute to CDF or LERF are included in the baseline model, and the licensee included the updated results in the FPRA and in the response to PRA RAI 44.01.b for transition risk.

The NRC staff reviewed two issues with regard to crediting fixed suppression systems. First, in its response to PRA RAI 18 (Reference 11 ), regarding NUREG/CR-6850 guidance on suppression system (e.g., sprays) impact on potentially vulnerable components, the licensee considered the effects of manual and automatic suppression. As a result, the licensee assumed PRA components could be impacted by sprays by automatic suppression (e.g., either all the equipment was failed, or equipment was assessed on an individual basis). Secondly, in its response to PRA RAI 14 (Reference 10), the licensee explained that for water-based suppression the first tray directly above the ignition source is failed, and that targets above the first tray credit the sprinklers. For gas-based systems, credit is provided for preventing tray damage above panels. The NRC staff finds these issues resolved because the licensee evaluated the potential for failure for damage from sprays and credited suppression in a manner

consistent with the guidance provided in NUREG/CR-6850, and these updates were included in the response to PRA RAI 44.01.b for transition risk.

The NRC staff found that bus ducts passed through the relay room. In its response to PRA RAI 29 (Reference 12), the licensee explained that the treatment of these bus ducts in the FPRA were consistent with FAQ 07-0035, "Bus Duct (Counting) For High-Energy Arcing Faults" (Reference 51 ), and described in NUREG/CR-6850, Supplement 1. However, in its response to PRA RAI 44 (Reference 12), the licensee updated the bus duct frequencies and expanded the fire-impacted conduits and trays. The NRC staff considers this issue resolved because the licensee included these FPRA updates in the response to PRA RAI 44.01.b for transition risk.

According to Table C-1 in Attachment C of the LAR, Hemyc wrap is used for cables associated with VFDRs identified for fire areas ABBM, ABI, BR1 B, and RC. In its response to PRA RAI 04 (Reference 10), the licensee clarified that the Hemyc modification is not used in fire area CC for VFDR-CC-044. According to Section 3.5.1.9 of the SE, the Hemyc is credited by including an additional 25 mins. of protection beyond the point where cables would normally be damaged without the Hemyc. According to its response to PRA RAI 11.01 (Reference 14), the licensee indicated that Hemyc is not credited as resolving any VFDR as a result of using the NUREG/CR-7150 method for the spurious actuation likelihood for double break electrical isolation. Thus, Hemyc wrapped cables are considered VFDRs and the cables' failures retained in the post-transition plant PRA and removed from the compliant plant PRA. Based on the NRC staff's evaluation of the Hemyc wrap protection and conclusions in Sections 3.1.1.1, 3.1.3, and 3.5.1.9 of the SE, the NRC staff finds that the transition risk in the response to PRA RAI 44.01.b appropriately considers the Hemyc wrap where credited in the FPRA.

In its response to PRA RAI 03 (Reference 12), the licensee stated that the FPRA model includes credit for the upgraded Westinghouse shutdown seal (SOS). In its response to PRA RAI 03.01 (Reference 14), the licensee clarified that the reactor coolant pump (RCP) seal which Ginna intends to install is the Westinghouse SHIELD Generation Ill passive thermal shutdown seal. The RCP seal model used in the FPRA is currently under review by the NRC staff. The licensee provided an updated LAR Attachment S, Table S-3, which included implementation item 21 to assess the change-in-risk associated with the RCP seals against the RG 1.205 guidelines, and to take action to reduce the risk results to within RG 1.205 guidelines if the updates show that the transition change-in-risk estimates will exceed the guidelines. The NRC staff finds that allowing this credit for transition is acceptable because the FPRA uses the best currently available information for the SOS modeling and will be updated and evaluated in accordance with implementation item 21, and compensatory measures will be established prior to RCP seal replacement. Furthermore, the NRC staff finds that, in accordance with implementation item 21, self-approval changes which rely on the SOS failure model to meet the self-approval criteria cannot be undertaken before acceptable models have been implemented into the Ginna IEPRA and FPRA.

The FPRA model also credits a number of other modifications in Attachment S of the LAR, most of which decrease the risk associated with VFDRs rather than deterministically resolve them.

The FPRA credits the planned modifications for inventory control and decay heat removal to mitigate the vast majority of fire scenarios, according to the licensee's response to PRA RAI 11.02.1.e (Reference 15). For example, the new injection system described in Item 9 on Table S-2 in Attachment S of the LAR is credited for mitigating an un-isolated stuck open PORV

event. The licensee also evaluated HEPs associated with the planned modifications. Since the design of these modifications are not completed, further changes to the modification design and associated HEPs may be necessary. The licensee will update the FPRA for changes in modifications and associated HRA in accordance with implementation items 15 and 9 on Table S-3 in Attachment S of the LAR.

As a result of its review of the LAR, as supplemented, the NRC staff concludes that the FPRA is of sufficient technical adequacy 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 are acceptable. To reach this conclusion, the NRC staff has reviewed all F&Os provided by the peer reviewers and determined that the resolution of every F&O supports the determination that the quantitative results are adequate. In addition, the NRC staff reviewed FPRA-related issues, many summarized above, and determined that the licensee's resolution of the identified issues support the determination that the quantitative results are adequate to transition to NFPA 805 and to support subsequent self-approval as described in the applicable licensing condition.

Accordingly, the NRC staff concludes that the licensee has demonstrated that the FPRA meets the guidance in RG 1.200, Revision 2, and that it is technically adequate to support the FREs and other risk calculations required for the NFPA 805 application.

3.4.2.3 Fire Modeling in Support of the Development of the Fire Risk Evaluation The NRC staff performed detailed reviews of the FM used to support the FREs in order to gain further assurance that the methods and approaches used for the application to transition to NFPA 805 (Reference 3) were technically adequate. NFPA 805 has the following requirements that pertain to FM used in support of the development of the FREs:

Section 2.4.3.3 of NFPA 805 states, in part that:

The PSA approach, methods, and data shall be acceptable to the AHJ.

Section 2.7.3.2 of NFPA 805, "Verification and Validation," states that:

Each calculational model or numerical method used shall be verified and validated through comparison to test results or comparison to other acceptable models.

Section 2.7.3.3 of NFPA 805, "Limitations of Use," states that:

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.

Section 2.7.3.4 of NFPA 805, "Qualification of Users," states that:

Cognizant personnel who use and apply engineering analysis and numerical models (e.g., FM 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.

Section 2.7.3.5 of NFPA 805, "Uncertainty Analysis," states that:

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 in the SE.

3.4.2.3.1 Overview of Fire Models Used to Support the Fire Risk Evaluations 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 scoping or screening evaluation as part of the FREs. The following algebraic fire models and correlations were used for this purpose:

  • Plume Temperature, Method of Heskestad (Reference 45) (Chapter 9)
  • Flame Height, Method of Heskestad (Reference 45) (Chapter 3)
  • Radiant Heat Flux, Point Source Radiation Model (Reference 45) (Chapter 5)

These algebraic models are described in NUREG-1805 (Reference 45). Verification and Validation (V&V) of these algebraic models is documented in NUREG-1824 (Reference 46).

The V&V of the fire models that were used to support the FRE is discussed in Section 3.8.3.2 of the SE.

In Attachment J of the LAR, "Fire Modeling V&V," and in several F&O resolutions in Attachment V of the LAR, "Fire PRA Quality," the licensee also discussed the use of an additional empirical correlation that are addressed in NUREG-1805 or FIVE, but for which V&V is not addressed in NUREG-1824, Volume 3 and 4.

The licensee identified the use of the following empirical correlation that is not addressed in NUREG-1824.

  • Heat and Smoke Detection Activation Models, described in NUREG-1805 and FIVE-Revision 1, "EPRI Fire Induced Vulnerability Evaluation Methodology" (Reference 78).

The licensee used a 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 ZOls were used to screen from further consideration those specific ignition sources that did not adversely affect the operation of credited SSCs, or targets, following a fire.

The licensee's screening was based on the 99th percentile fire HRR from the NUREG/CR-6850 methodology (Reference 39).

The CFAST, Version 6, zone model was used for hot gas layer (HGL) temperature calculations in specific fire zones.

Detailed FM using CFAST was performed for selected fire scenarios in fire areas. The CFAST zone model was also used for the temperature sensitive equipment HGL study. 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 46).

Fire Dynamics Simulator (FDS), Version 5 was used for the MCR abandonment time calculations. The V&V of FDS is documented in NUREG-1824, Volume 7 (Reference 46).

The V&V of all empirical correlations and fire models that were used to support the FRE is discussed in detail in SE Section 3.8.3.2.

3.4.2.3.2 RAls Pertaining to Fire Modeling in Support of the FPRA By letter dated October 9, 2013 (Reference 19), the NRC staff requested additional information concerning the FM conducted to support the FREs. By letters dated December 17, 2013 (Reference 10), January 29, 2014 (Reference 11), and February 28, 2014 (Reference 12), the licensee responded to these RAls.

  • In FM RAI 01 (1 ).a (Reference 19), the NRC staff requested that the licensee provide technical justification for not considering mechanical ventilation in the CFAST HGL calculations.

In response (Reference 12), the licensee stated that mechanical ventilation would affect the fire conditions in the room by introducing fresh oxygen. The licensee further stated that in every CFAST HGL calculation, sufficient oxygen is available to sustain combustion throughout the simulation time. According to the licensee, where the natural ventilation of the room did not fall within the validation range, a sensitivity case was run to allow enough oxygen into the compartment to meet the natural ventilation equivalence ratio criteria. The licensee discussed a specific example of a compartment where a sensitivity analysis was performed, and explained the implications in terms of HGL timing used in the FPRA.

The NRC staff determined that the licensee's response is acceptable because none of the CFAST fire simulations indicate oxygen depletion (ventilation-limited fire) in compartments and because the available oxygen in the compartment is enough to sustain combustion throughout the CFAST simulation time. In addition, a sensitivity analysis was performed without ventilation to analyze effects on a small fire

compartment and the results indicated that modeling the compartments without ventilation will not significantly modify the time to HGL development.

  • In FM RAI 01(1).b (Reference 19), the NRC staff requested that the licensee provide an explanation as to how the reduction of the net effective volume of the compartment due to obstructions and contents was accounted for in the CFAST HGL calculations.

In response (Reference 12), the licensee stated that obstructions were not considered in determining the effective volume of the fire zones and that this is consistent with common practice in applying CFAST in NPPs as described in NUREG-1934 (Reference 79). The licensee further explained that the effect of the volume reduction is partly offset by the fact that CFAST ignores the heat losses to equipment in the room. The licensee performed a sensitivity analysis to show that a 10 percent volume reduction in a small compartment has a minimal effect on HGL development and timing, and that a 20 percent reduction in a large compartment has no effect.

The NRC staff determined that the licensee's response is acceptable because the volume reduction impact is offset by ignoring the heat sink effect, making it acceptable to ignore the compartment volume reduction due to obstructions occupying part of the space in the CFAST HGL calculations.

  • In FM RAI 01(1).c (Reference 19), the NRC staff requested that the licensee provide technical justification for the response time index (RTI) value that was assumed in the sprinkler activation calculations, including a description as to how that value compares with the RTI of the actual sprinklers in the plant.

In response (Reference 12), the licensee stated that to address the uncertainty of the RTI value, automatic suppression is only credited after a number of targets near the suppression system have failed. The licensee also stated that it is assumed that, if a fire can generate conditions to damage a cable, those conditions should be able to trigger the automatic suppression system. The licensee further stated that failure of cable targets are used as surrogates for activation of automatic suppression, and the RTI value is not used in the determination of sprinkler activation.

The NRC staff determined that the licensee's response 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, making the licensee's approach to determine sprinkler activation time acceptable.

  • In FM RAI 01 (1 ).e (Reference 19), the NRC staff requested that the licensee provide the criteria for qualitatively screening a fire zone from the HGL analysis based on the size of the fire zone and openings to other zones.

In response (Reference 12), the licensee provided a list of fire zones qualitatively screened from CFAST analysis, confirming that only outside areas were screened.

The NRC staff determined that the licensee's response is acceptable because only outside plant areas (outside condensate storage tank area and transformer yard) were screened from CFAST analysis given lack of definite volume and confined area for HGL development.

  • In FM RAI 01(1).f (Reference 19), the NRC staff requested that the licensee provide technical justification for assuming a generic fire dimension of 2 feet (ft) in the ZOI calculations, and a characteristic cabinet length of 2 ft for the purpose of calculating fire propagation in cable trays.

In response to the first part of the RAI (Reference 12), the licensee stated that the selection of a default diameter is not a factor for scenarios in which failure of all targets in the transient zone is assumed because the latter is larger than the ZOI. The licensee further stated that a sensitivity analysis was conducted assuming a shorter time to damage than that determined based on the critical HRR calculated from Heskestad's plume temperature correlation with a fire diameter of 2 ft. In response to the second part of the RAI (Reference 12), the licensee stated that there is only one compartment where a cabinet fire propagating to overhead cable trays resulted in the formation of an HGL, and that the cabinet in this scenario is actually 2 ft long.

The NRC staff concludes that the licensee's response is acceptable because in the one compartment where a cabinet fire propagating to overhead cable trays results in the formation of a HGL, the diameter assumed is equivalent to the actual cabinet length, and the licensee demonstrated that the ZOI applied to the transient combustible fires is larger than the calculated ZOI and, therefore, results in the failure of more cables.

  • In FM RAI 01(1).g (Reference 19), the NRC staff requested that the licensee provide technical justification for the assumption in Fire Zone RR-C1 that only the first tray above the ignition source is damaged, and that the remaining trays are protected due to activation of the Halon fire suppression system.

In response (Reference 12), the licensee stated that the approach to credit the Halon fire suppression system in Fire Zone RR-C1 is similar to that used to credit sprinkler suppression as discussed in the response to FM RAI 01 (1 ).c. More specifically, the licensee further stated that the ignition source and the cable tray(s) immediately above the ignition source are failed without Halon fire suppression credit, and credit is then assigned to subsequent scenarios that include additional targets to which the fire propagates from the initial target set. According to the licensee, since the Halon fire suppression system is activated by a smoke detection signal, this approach is conservative because a relatively large fire is postulated before the Halon fire suppression system is credited in applicable scenarios.

The NRC staff determined that the licensee's response is acceptable because the licensee only credits Halon fire suppression after a number of targets near the suppression system have failed. This is conservative because a relatively large fire is postulated which, due to smoke detection, would activate the Halon fire suppression system before it is actually credited.

  • In FM RAI 01 (2).a (Reference 19), the NRC staff requested that the licensee provide the criteria used during the walk downs to determine whether wall or corner effects have to be accounted for in the FM analyses.

In response (Reference 12), the licensee stated that wall and corner effects were only applied for fixed ignition sources that were located in contact with a wall or a corner.

Since no specific guidance is provided in NUREG/CR-6850 and in the FAQs, the licensee reviewed the pertinent literature and cited a study that suggests that fires separated a short distance from wall surfaces have characteristics similar to fires in the open.

The NRC staff determined that the licensee's response is acceptable because the approach accounts for wall and corner effects in the FM calculations.

  • In FM RAI 01 (2).b (Reference 19), the NRC staff requested that the licensee provide an explanation as to how wall and corner effects were accounted for in the plume temperature calculations.

In response (Reference 12), the licensee explained that for ignition sources against a wall or in a corner, Heskestad's plume correlation was solved using a multiplier to the HRR of 2 or 4 for wall and corner configurations, respectively.

The NRC staff determined that the licensee's response is acceptable because the approach used accounted for wall and corner effects in the Heskestad's plume temperature calculations.

  • In FM RAI 01 (2).c (Reference 19), the NRC staff requested that the licensee provide an explanation as to how wall and corner effects were accounted for in the CFAST HGL calculations.

In response (Reference 12), the licensee stated that wall and corner effects were not accounted for in the CFAST calculations as the HRR used for determining HGL temperatures is based on the worst combination of cable trays and cabinets and bounds the different fire scenarios that may develop in the fire zone. The licensee further stated that a sensitivity analysis accounting for wall and corner effects was performed for fire compartments that did not result in an HGL formation in the original analysis. According to the licensee, the "image" method was applied to fire scenarios considered in the CFAST analysis. The licensee stated that the results indicate that an HGL will not form even if the fire is assumed to be against a wall or in a corner.

The NRC staff determined that the licensee's response is acceptable because even though the wall and corner effects were not explicitly accounted for in the CFAST HGL calculations, they were implicitly addressed by the use of bounding HRRs for the ignition sources in the fire zones or shown, via sensitivity analysis, not to generate HGLs.

  • In FM RAI 01 (3).a (Reference 19), the NRC staff requested that the licensee provide an explanation as to how a fire in the office directly across from the rear of the MCB would

affect the calculated MCR abandonment times, including reasonable assurance that fires in the office are bounded by the scenarios considered in the MCR abandonment analysis.

In response (Reference 12), the licensee stated that a lower bound estimate of the abandonment time for a fire in the office can be obtained based on the calculated abandonment times as a function of HRR for transient and panel fires in the MCR and HRR data for a workstation fire from the SFPE Handbook of Fire Protection Engineering.

Based on this lower bound estimate, using the ignition frequency for this type of fire, and the highest CCDP quantified for the MCR (1.3E-02), the licensee determined that the workstation fire would represent approximately 0.1 percent of the total plant CDF and less than 1 percent of the risk in the MCR.

The NRC staff determined that the licensee's response is acceptable because the licensee demonstrated that fires in the office directly across from the rear of the MCB would not affect the calculated MCR abandonment times and because the CDF contribution is small and does not affect the current risk results.

  • In FM RAI 01 (3).b (Reference 19), the NRC staff requested that the licensee provide a description of the flame spread characteristics that were assumed in the MCR abandonment analysis for the cables in the MCBs.

In response (Reference 12), the licensee explained that the MCR abandonment analysis does not consider cable damage or postulate fire spread in cable trays or cables other than that addressed by the HRR profiles provided in Appendix E of NUREG/CR-6850 for electrical panels. The IEEE qualification status is used to select the HRR profiles for the electrical panels, which is consistent with the classification of the electrical panel fires described in Appendix E of NUREG/ CR-6850.

The NRC staff determined that the licensee's response is acceptable because the licensee followed the guidance in NUREG/CR-6850, and IEEE qualification status of the cables in the MCR is used appropriately in the determination of the HRR profiles for the MCR abandonment time calculations.

  • In FM RAI 01 (3).c (Reference 19), the NRC staff requested that the licensee provide technical justification for the time to peak HRR of 552 seconds used for the transient fire scenarios in the MCR abandonment analysis.

In response (Reference 12), the licensee stated that the incorrect ramp was used in the MCR abandonment calculations and demonstrated that the original conclusion would remain unchanged if a shorter ramp time of 2 mins. had been used for transient fires.

The NRC staff determined that the licensee's response is acceptable because there is reasonable assurance that correcting the time to peak HRR for transient fires in the MCR abandonment time calculations will not affect the outcome of the analysis.

  • In FM RAI 01 (3).d (Reference 19), the NRC staff requested that the licensee provide technical justification for the assumption in the MCR abandonment time analysis that the

MCB panels are closed, and for not using the HRR distribution for Case 5 in Table E-1 of NUREG/CR-6850, Volume 2.

In its response to FM RAI 01 (3).d (Reference 12), the licensee stated that there are two approaches to model fires in the MCBs:

1. The MCB panels are treated as closed electrical panels that release heat directly in the control room environment; and
2. The MCB panels are treated as open panels that release heat in the MCB sub-volume.

The licensee stated that the first approach was used in the MCR abandonment analysis.

The licensee further stated that in the second approach, air flows into the MCB sub-volume and hot products of combustion emerge from the MCB sub-volume through the vents in the MCB panels. The licensee performed an analysis to show that the first approach is bounding due to the limited air supply in the second approach.

The NRC staff concludes that the licensee's response is acceptable because the licensee's approach of treating MCB panels as closed electrical panels that release heat directly in the control room environment leads to conservative HRR results, which bounds the effects from treating MCB panels as open panels that release heat in the MCB sub-volume.

  • In FM RAI 01 (3).e (Reference 19), the NRC staff requested that the licensee provide technical justification for the assumption in the MCR abandonment time analysis that a fire in an MCB panel propagates to adjacent panels in 10 mins.

In response (Reference 12), the licensee stated that the MCB model in the FPRA, which assumes fire propagation time between boards of 10 mins. is part of a comprehensive event tree model that includes sequences for small fires that are promptly suppressed, and fires that do not propagate outside of the panel of fire origin. The licensee provided a detailed discussion to justify the assumed values for characterizing the different sequences based on a comparison with values obtained from the approach described in Appendix L of NUREG/ CR-6850. In all cases, the value chosen in the FPRA is more conservative than allowed in Appendix L of NUREG/CR-6850.

The NRC staff determined that the licensee's response is acceptable because the licensee's approach to determine the probability for MCR abandonment time based on MCB panel fires was more conservative than the guidance in NUREG/CR-6850 for situations where there was very localized fire damage, where a whole panel was damaged, and in those cases where fire propagated to multiple panels. Since the method used by the licensee provided conservative results, the NRC staff concludes that the licensee's method is acceptable.

  • In FM RAI 01 (3).f (Reference 19), the NRC staff requested that the licensee provide an explanation as to how the fire diameter was determined for the transient and panel fires in the MCR abandonment time analysis.

In response (Reference 12), the licensee stated that the NUREG-1824 validation range for the Froude (Fr) number is between 0.4 and 2.4. The licensee further stated that a comparison between the weak plume (Fr= 0.4) and strong plume (Fr= 2.4) assumptions showed that the former is bounding. According to the licensee, the fire diameter for transients and electrical panels was determined for each HRR bin based on the assumption that the Froude number is equal to 0.4.

The NRC staff determined that the licensee's response is acceptable because the licensee explained that the fire diameter was determined for the transient and electrical panel fires using bounding Fr= 0.4 to calculate HRR.

  • In FM RAI 01 (3).g (Reference 19), the NRC staff requested that the licensee provide technical justification for determining the MCR abandonment probability for transient fire scenarios on the basis of the results of the FDS abandonment time calculations for panel HRR Bins 1-5 in Appendix E of NUREG/CR-6850, since transient fires reach peak HRR much quicker than panel fires.

In response (Reference 12), the licensee stated that the FDS simulations for transient fires did not result in MCR abandonment conditions. The licensee further stated that the worst case fire scenarios (electrical cabinet fires) are considered for the probability of MCR abandonment calculations. According to the licensee, to conservatively estimate the transient fire contribution to the abandonment calculations, Bins 1-5 of the bounding case are chosen to represent the transient fire 981h percentile HRR of 317 kilowatt (kW).

The licensee stated that this is conservative given that the FDS simulations did not identify any transient fires resulting in abandonment conditions.

The NRC staff determined that the licensee's response is acceptable because transient fires are not expected to cause MCR abandonment. The licensee estimated MCR abandonment times using the bounding case to represent the transient fire 981h percentile HRR of 317 kW.

  • In FM RAI 02 (Reference 19), the NRC staff requested that the licensee provide a description of how the installed cabling in the power block was characterized, specifically with regard to the critical damage threshold temperatures and critical heat flux for thermoset and thermoplastic cables as described in NUREG/ CR-6850. In addition, the NRC staff requested that the licensee provide justification for assigning thermoset damage thresholds to selected cables that run through one of the battery rooms.

In response (Reference 10), the licensee stated that all targets, with the exception of cables in five conduits in Battery Room A, are assumed to be thermoplastic with damage criteria 205 degrees Celsius 0 (C) (401 Fahrenheit (°F)) and 6 kW/m 2 per Table H-1 in NUREG/CR-6850. The licensee further stated that the 5 conduits in Battery Room A were verified to be of thermoset cable insulation material, and, therefore, a critical damage temperature of 330°C (626 °F) and heat flux of 11 kW/m 2 were assigned to them in accordance with Table H-1 in NUREG/CR-6850.

The NRC staff determined that the licensee's response is acceptable because the licensee clarified that thermoplastic damage thresholds was used in FM calculations except for Battery Room A fire scenarios, where five conduits were verified to house only thermoset cable.

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 process for performing FM in support of the FRE and clarification provided in response to the RAls, the NRC staff concludes that the licensee's approach for meeting the requirements in Section 2.4.3.3 of NFPA 805 is acceptable.

3.4.2.4 Conclusions Regarding Fire PRA Quality The NRC staff concludes that the technical adequacy and quality of the Ginna FPRA is sufficient for the FREs that support the proposed license amendment because of the following:

(1) the PRA models conform to the applicable industry PRA standards for internal events and fires, considering the acceptable disposition of the review findings, (2) the FM used to support the development of the Ginna FPRA has been confirmed as appropriate and acceptable, and (3) the PRA models represent the as-built, operated, and maintained plant as it will be configured at full implementation of NFPA 805.

In addition, the licensee's PRA satisfies the guidance in Section 2.3 of RG 1.174, regarding technical adequacy of the PRA; Section 4.3 of RG 1.205, regarding the technical adequacy of the FPRA; and Sections 111.2.2.4.1 and 19.2 in NUREG-0800 (Reference 38), regarding the review of PRA Quality required for an application, which supports the NRC staff's conclusion that the Ginna PRA is technically adequate and of sufficient quality for transition to NFPA 805.

Finally, based on the licensee's administrative controls to maintain the PRA models current and assure continued quality using only qualified staff and contractors (as described in Section 3.8.3 of the SE), the NRC staff concludes that the quality of the Ginna FPRA is sufficient to support self-approval of future RI changes to the FPP under the NFPA 805 license condition following the implementation of the plant modifications that are credited in the PRA.

3.4.3 Fire Risk Evaluations For those fire areas for which the licensee used a PB approach to meet the NSPC, the licensee used FREs in accordance with Section 4.2.4.2 of NFPA 805, to demonstrate the acceptability of the plant configuration. In accordance with the guidance in Section C.2.2.4 of RG 1.205, "Risk Evaluations," the licensee used a RI approach to justify acceptable alternatives to comply with the NFPA 805 deterministic criteria. The NRC staff reviewed the following information during its evaluation of Ginna's FREs: Section 4.5.2 of the LAR, "Performance Based Approaches,"

Attachments C, "NEI 04-02 Table B Fire Area Transition," and W, "Fire PRA Insights," as well as associated supplemental information of the LAR.

Plant configurations that did not meet the deterministic requirements in Section 4.2.3.1 of NFPA 805 were considered VFDRs. On Table B-3 in Attachment C of the LAR, "Fire Area Transition," the licensee identified a number of VFDRs that it does not intend to bring under

deterministic compliance under NFPA 805. For these VFDRs, the licensee performed risk evaluations in accordance with Section 4.2.4.2 of NFPA 805, to address FPP non-compliances and demonstrate that the VFDRs are acceptable.

VFDRs identified by the licensee were categorized as separation issues. The separation VFDRs can generally be categorized into the following 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 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; or, (4) combinations of the above configurations.

A FRE was performed for each fire area containing VFDRs of Section 4.2.3 of NFPA 805. The licensee explained its FRE process in its response to PRA RAI 11 (Reference 10). The licensee calculated a change in risk for transition (flCDF and flLERF) considering the guidance in FAQ 08-0054 (Reference 58). The post-transition plant includes all proposed modifications and procedure changes planned for the NFPA 805 transition, both modifications to resolve some VFDRs and the non-VFDR modifications. The compliant plant is the post-transition plant with the VFDRs assumed deterministically resolved. Also, for the compliant plant, HEPs are adjusted to reflect that operator actions are taken from the MCR or the PCS. The change in risk is the difference between the post-transition plant and the compliant plant risk estimates. Some VFDRs on Table C-1 in Attachment C of the LAR were dispositioned with a qualitative assessment regarding insignificant risk associated with the VFDR and, therefore, not included in the FPRA.

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 the guidance in Section 2.2.4.1 of RG 1.205, Revision 1, , and FAQ 08-0054, Revision 1. The NRC staff finds that while the total fire-related flCDF and flLERF reported in its response to PRA RAI 44.01.b (Reference 17), exceed RG 1.174 risk acceptance guidelines, the Ginna NFPA 805 LAR is a combined change, and the total combined change risk as described in Section 3.4.6 of the SE meets the RG 1.174 risk acceptance guidelines.

3.4.4 Additional Risk Presented by Recovery Actions For those fire areas for which the licensee used a PB approach to meet the NSPC, the licensee used FREs in accordance with Section 4.2.4.2 of NFPA 805 to demonstrate the acceptability of the plant configuration. For many of these VFDRs the licensee identified RAs to reduce the risk of the retained VFDR. Section 4.2.4 of NFPA 805, further states that, "when the use of recovery actions has resulted in the use of this approach, the additional risk presented by their use shall be evaluated."

The NRC staff reviewed Attachments C, "NEI 04-02 Table B Fire Area Transition," G, "Recovery Actions Transition," and W of the LAR, "Fire PRA Insights," during its evaluation of the additional risk presented by the NFPA 805 RAs at Ginna. Section 3.2.5 of the SE describes the identification and evaluation of RAs.

The licensee used the guidance in RG 1.205, Revision 1 for addressing RAs. This included consideration of the definition of PCS and RA, as clarified in 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. Conversely, any manual actions required to be performed outside the control room and not at the PCS were considered RAs.

Of the 20 fire areas on Table C-1 in Attachment C of the LAR, 17 fire areas were included as NFPA 805 PB fire areas, and the licensee identified RAs associated with 185 VFDRs. The RAs are described on Table G-1 in Attachment G of the LAR. The licensee reviewed all of the RAs for adverse impact as stated in Attachment G of the LAR, and none of the RAs listed on Table G-1 in Attachment G of the LAR were found to have an adverse impact on the FPRA.

Ginna has two PCSs, shutdown panels ABELIP and IBELIP, from which primary and secondary plant controls can be used for existing plant equipment. According to the licensee's response to PRA RAI 07 (Reference 10), the first operator actions will be to utilize planned modifications by placing a standby AFW pump and a new injection pump into operation using a new dedicated diesel generator; the shutdown panels ABELIP and IBELIP will be used as a DID measure.

Attachment G of the LAR includes RAs associated with these planned modifications in Attachment S of the LAR. The operator actions taken at ABELIP and IBELIP do not contribute to the additional risk of RAs since these panels maintain their PCS status.

According to the licensee's response to PRA RAI 5.01 (Reference 14), the additional risk of RAs includes both new and previously approved as defined in RG 1.205, Revision 1. The licensee followed FAQ 07-0030 (Reference 50), and FAQ 08-0054 (Reference 58), guidance in evaluating their feasibility and their contribution to the additional risk of RAs as described in the response to PRA RAI 07, assigning conservative execution HEPs in PCS locations as successful for the compliant plant model. As a result of FPRA model changes, the licensee's response to PRA RAI 44.01.b reported the additional risk of recovery actions to be 1.11 E-5/yr (CDF) and 5.41 E-7 (LERF). The NRC staff determined that the additional risk of RAs is expected to be less than the RG 1.174 acceptance guidelines of 1E-5/yr for CDF, and, therefore, acceptable because the licensee's estimates were calculated by conservatively including new RAs as well as previously approved RAs and assigned conservative HEPs in PCS locations for the compliant plant model.

The NRC staff concludes that the licensee's approach for calculating the additional risk of RAs is acceptable because it is consistent with the guidance in Section 2.2.4.1 of RG 1.205 and FAQ 07-0030 and FAQ 08-0054 guidance. In addition, the NRC staff finds the results of the licensee's calculations associated with the additional risk of RAs acceptable because the CDF and LERF bounding calculation met RG 1.174 risk acceptance criteria.

3.4.5 Risk-Informed or Performance-Based Alternatives to Compliance with NFPA 805 The licensee did not utilize any RI or PB alternatives to compliance with NFPA 805, which fall under the requirements of 10 CFR 50.48(c)(4).

3.4.6 Cumulative Risk and Combined Changes In the LAR, as supplemented by its response to PRA RAI 44.01.b (Reference 17), the licensee estimated the total CDF and total LERF by adding the risk assessment results. The CDF and LERF results for fire and internal events are summarized on Table 3.4.6-1 in the SE.

Table 3.4.6-1: CDF and LERF for Ginna after Transition to NFPA 805 Hazard Group CDF LERF Fires 3.33E-5 1.17E-6 Internal Events 1.70E-5 1.98E-6 TOTAL for fire and internal events 5.03E-5 3.15E-6 The LAR also notes that the NRC estimates the weakest link using the 2008 USGS Seismic Hazard Curves (Reference 80), as 1.3E-05/yr for CDF. In addition, the licensee has determined that there are no remaining external risks noted in the Ginna Individual Plant Examination for External Events which could challenge the remaining margin to RG 1.174 recommendations given the NFPA 805 modifications are installed.

The licensee credited the risk reductions that will be afforded by modifications in its evaluation of the total change in risk associated with transition to NFPA 805. This includes both VFDR and non-VFDR modifications which decrease the total change in fire risk. Risk reduction modifications are also included in the total internal events risk in Attachment W of the LAR and are credited with reducing the total delta risk.

The licensee's application to transition to NFPA 805 is a combined change, as defined by RG 1.205, Revision 1, which combines the risk increases identified in the FREs with the risk decreases resulting from modifications that include reductions in risk associated with the IEPRA. According to Section 3.2.5 of RG 1.205, "Combined Changes and Relative Cumulative Risk of Changes," risk increases may be combined with risk decreases when estimating the total change in risk. For VFDRs for which the NFPA 805 PB approach is used, the risk associated with retaining some VFDRs and their associated RAs, can be offset by the decrease in risk associated with plant modifications and procedural changes unrelated to VFDRs.

The licensee reported updated results in its response to PRA RAI 44.01.b (Reference 17), for the combined change, by adding the FPRA and IEPRA changes, which shows the LiCDF and LiLERF to be within the RG 1.174 acceptance guidelines.

Table 3.4.6-2: LiCDF and LiLERF Combined Change for Ginna NFPA 805 Transition PRA Model LiCDF LiLERF Fire 1.56E-5 7.66E-7 Internal Event -7.15E-6 -2.68E-7 Combined (Fire + Internal Event) 8.50E-6 4.98E-7

Based on the combination of these risk values, the NRC staff concludes that the changes associated with NFPA 805 meet the guidance contained in RP 3.2.5 of RG 1.205, related to meeting the requirements for cumulative risk and combined plant changes.

3.4. 7 Uncertainty and Sensitivity Analyses FAQ 08-0048 (Reference 56), indicates that a sensitivity study should be performed if the FPRA uses fire frequencies from EPRI 1016735 (Reference 81 ). The sensitivity study should be performed for a fire ignition frequency bin with an alpha factor of 1 or less using the mean of the fire ignition frequency bins contained in NUREG/CR-6850. In its response to PRA RAI 19 (Reference 12), the licensee provided the results of the sensitivity analysis which showed that the ~CDF increased to approximately 2E-5/yr. In its response to PRA RAI 19.03 (Reference 16), the licensee provided an explanation that this ~CDF increase was attributable to larger fire ignition frequencies as well as the method used to calculate the ~CDF. The licensee further explained that there is conservatism in the method since the compliant plant FPRA model credits risk reduction only modifications, which the licensee identified as those not used as the sole means to address a VFDR. The licensee applied an updated FPRA model in its response to PRA RAI 19, where these modifications are removed from the compliant plant FPRA model, but maintained in the post-transition FPRA model. The resulting sensitivity analysis from this model produces a ~CDF less than 1E-5/yr and a ~LERF less than 1E-6/yr. The NRC staff determined that the RG 1.174 ~CDF and ~LERF guidelines are met for the FAQ 08-0048 required sensitivity analysis because they are met when the risk reduction only modifications are removed from the compliant plant model.

In its response to PRA RAI 36 (Reference 12), regarding the use of a HEP floor for joint HEPs used in the FPRA sequences, the licensee performed a sensitivity analysis using a floor with the NUREG-1921 suggested value of 1E-5 for combinations of HEPs. The sensitivity analysis showed negligible effect on the CDF and LERF. Therefore, the NRC staff concludes that the Ginna approach to not use a HEP truncation floor value is acceptable because the sensitivity study demonstrates that using a floor value has essentially no effect on CDF and LERF.

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 at Ginna, the NRC staff finds that:

  • The licensee's PRA used to perform the risk assessments in accordance with Sections 2.4.4 (plant change evaluations) and 4.2.4.2 (fire risk evaluations) of NFPA 805 is of sufficient quality to support the application to transition the Ginna FPP to NFPA 805.

The NRC staff concludes the PRA approach, methods, tools and data are acceptable and in accordance with Section 2.4.3.3 of NFPA 805.

  • The transition process included a review of fire protection DID and safety margins as required by NFPA 805. The NRC staff concludes that the licensee's documentation on DID and safety margins to be acceptable because the licensee's process followed the NRG-endorsed guidance in NEI 04-02, Revision 2, and is consistent with the approved

NRC staff guidance in RG 1.205, Revision 1, which provides an acceptable approach for meeting the requirements of 10 CFR 50.48(c).

  • The NRC staff concludes that the changes in risk (i.e., flCDF and flLERF) associated with the proposed alternatives to compliance with the deterministic criteria of NFPA-805 (FREs) are acceptable for the purposes of this application, and the licensee has satisfied the guidance contained in RG 1.205, Revision 1, Sections 2.2.4 and 2.2.5 of RG 1.174, and Section 19.2 of NUREG-0800, regarding acceptable risk. By meeting the guidance contained in these approved regulatory documents, the changes in risk have been found to be acceptable to the NRC staff and, therefore, meet the requirements of NFPA 805.
  • The NRC staff concludes that the licensee's approach for calculating the additional risk of RAs is acceptable because it is consistent with RG 1.205, Section 2.2.4.1, FAQ 07-0030, and FAQ 08-0054 guidance. The results demonstrate that the total risk from RAs is expected to be less than the risk acceptance guidelines in RG 1.174 if conservatism in the estimates were to be removed and, therefore, the NRC finds that the additional risk associated with RAs is acceptable
  • The licensee's application to transition to NFPA 805 is a combined change, as defined by RG 1.205, Revision 1, which combines the risk increases from VFDRs identified in the FREs with the risk decreases resulting from modifications that include reductions in internal events risk. Based on the combination of these risk values, the NRC staff concludes that the combined change meets the guidance contained in RP 3.2.5 of RG 1.205.
  • The licensee credits planned modifications in the FPRA and IEPRA to satisfy risk acceptance guidelines in RG 1.174 for NFPA 805 transition. The licensee will re-evaluate PRA credit for these planned modifications as given in implementation items 9 and 21. The NRC staff concludes that implementation item 9 is acceptable because:

(1) the licensee will evaluate the as-built modification change-in-risk against acceptance criteria as described in Section 4.5.2.2 of the LAR and ensure the acceptance criteria are satisfied; and (2) if not satisfied, the licensee will refine the modification to ensure the acceptance criteria are met. In addition, the NRC staff concludes that implementation item 21 is acceptable because the licensee will: (1) evaluate the NRC approved RCP model in the FPRA and IEPRA against RG 1.205 guidelines; (2) take action to be within the guidelines if they are exceeded; and (3) maintain compensatory measures prior to RCP seal replacement. Furthermore, the licensee will not undertake self-approval changes which rely on the SOS failure model to meet the self-approval criteria before acceptable models for these seals have been implemented into the Ginna IEPRA and FPRA.

  • The licensee did not utilize any RI or PB 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 Section 2.2.3 of NFPA 805, "Evaluating Performance Criteria," states that:

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.

Section 2.2.4 of NFPA 805, "Performance Criteria," states that:

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.

Section 2.2.7 of NFPA 805, "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.

3.5.1 Nuclear Safety Capability Assessment Results by Fire Area Section 2.4.2 of NFPA 805, "Nuclear Safety Capability Assessment (NSCA)," states that:

The purpose of this section is to define the methodology for performing a NSCA. The following steps shall be performed:

(1) Selection of systems and equipment and their interrelationships necessary to achieve the NSPC in Chapter 1.

(2) Selection of cables necessary to achieve the NSPC in Chapter 1.

(3) Identification of the location of nuclear safety equipment and cables.

(4) Assessment of the ability to achieve the NSPC given a fire in each fire area.

This section of the SE addresses the last topic regarding the ability of each fire area to meet the NSPC of NFPA 805. Section 3.2.1 of the SE addresses the first three topics.

Section 2.4.2.4 of NFPA 805, "Fire Area Assessment," also states that:

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 Chapter 4 of NFPA 805 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.

Section 4.2.2 of NFPA 805, "Selection of Approach," states that:

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 Sections 4.2.4, "Fire Area Transition," and 4.8.1, "Results of the Fire Area Review," Attachments C, G, S, and W of the LAR, "Fire PRA Insights" (Reference 9) during its evaluation of the ability of each fire area to meet the NSPC of NFPA 805.

Ginna is a single unit PWR with 20 individual fire areas, 15 of which contribute to the delta CDF and delta LERF, and each fire area is composed of multiple fire zones. Based on the information provided in the LAR, as supplemented, the licensee performed the NSCA on a fire area basis. Attachment C of the LAR provides the results of these analyses on a fire area basis and also identified the individual fire zones within the fire areas.

Table 3.5-1 in the SE identifies the fire areas that were analyzed which used the PB approach in accordance with Chapter 4 of NFPA 805 based on the information provided on Table C-1 in Attachment C of the LAR.

Table 3.5-1 Fire Area and Compliance Strategy Summary NFPA 805 Fire Area Area Description Compliance Basis ABBM Auxiliary Building Basement/Mezzanine Performance-Based Auxiliary Building Operating Floor and Intermediate ABI Performance-Based BuildinQ BOP Balance of Plant (Bldgs CD, TSC, H2, Srv, TB,TO) Performance-Based BR1A Battery Room 1A, EL 253' 6" Performance-Based BR1B Battery Room 1B, EL 253' 6" Performance-Based cc Control Building Complex Performance-Based CHG Charging Pump Rooni, EL 235' 6" Performance-Based

NFPA 805 Fire Area Area Description Compliance Basis CT Cable Tunnel, EL 260' 6" Performance-Based Diesel Generator Unit 1A (Including EDG Vault 1A),EL EDG1A Performance-Based 253' 6" Diesel Generator Unit 1B (Including EDG Vault 1B),EL EDG1B Performance-Based 253' 6" PA Protected Area Performance-Based RC Reactor Containment Building Performance-Based SAF Standby Auxiliary Feedwater Pump Building, EL 271' O" Performance-Based SH Screen House Building Performance-Based YARD Transformer Yard General Area Performance-Based STA13ACH Station 13A Control House Performance-Based INTAKE Intake Structure Performance-Based Attachment C of the LAR 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 was the PB approach in accordance with Section 4.2.4 of NFPA 805.
  • 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 VFDR using either; modifications (completed or committed) or the performance of a FRE in accordance with Section 4.2.4.2 of NFPA 805.

3.5.1.1 Fire Detection & Suppression Systems Required to Meet the Nuclear Safety Performance Criteria A primary purpose of Chapter 4 of NFPA 805 is to determine, by analysis, what fire protection features and systems need to be credited to meet the NSPC. Four sections of Chapter 3 of NFPA 805 have requirements dependent upon the results of the engineering analyses performed in accordance with Chapter 4 of NFPA 805, which are: (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 3.10.1; and (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 Chapter 4 of NFPA 805 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 protection systems and features required to meet the NSPC for each fire area. Table C-2 in the

LAR, "NFPA 805 Required Fire Protection Systems and Features," lists the fire areas and identifies if the fire protection systems and features installed in these areas are required to meet criteria for separation, DID, risk, licensing actions, or EEEEs.

The NRC staff reviewed Attachment C of the LAR for each fire area, to ensure fire detection and suppression met the principles of DID in regard to the planned transition to NFPA 805.

Based on the statements provided in Attachment C of the LAR, as supplemented, the NRC staff concludes that the treatment of this issue is acceptable because, the licensee adequately identified the fire detection and suppression systems required to meet the NFPA 805 NSPC on a fire area basis.

3.5.1.2 Evaluation of Fire Suppression Effects on Nuclear Safety Performance Criteria Each fire area in Attachment C of the LAR includes a discussion of how the licensee met the requirement to evaluate the fire suppression effects on the ability to meet the NSPC.

In Attachment C of the LAR, the licensee stated that damage to plant areas and equipment from the accumulation of water discharged from manual and automatic fire protection systems within the fire area and the discharge of automatic and manual suppression water to adjacent compartments is controlled. The licensee further stated that fire suppression activities will not adversely affect the plant's ability to achieve the NSPC considering the following:

  • Automatic fire suppression coverage:
  • Drainage of the compartment;
  • Access to the compartment and manual fire suppression features;
  • Impact on area equipment; and
  • Mitigating features such as seals, procedures, curbs, and tray type.

The licensee determined that fire suppression activities will not adversely affect achievement of the NSPC.

Based on the information provided by the licensee in the LAR, 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 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 Section 4.2.3 and Attachment K of the LAR, there are no licensing actions that are required to be transitioned to NFPA 805.

3.5.1.4 Existing Engineering Equivalency Evaluations The licensee reviewed the EEEEs that support compliance with Chapter 4 of NFPA 805, using the methodology contained in NEI 04-02 (Reference 7). 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 Section 4.2.2 of the LAR, the licensee stated that the guidance in RP 2.3.2 of RG 1.205 (Reference 4), and FAQ 08-0054 (Reference 58) 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 is provided.
  • If requesting specific NRC approval for "adequate for the hazard" EEEEs, then the EEEE was referenced where required to demonstrate compliance and is included in Attachment L of the LAR 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 that the NRC staff review and approve any of these EEEEs.

Based on the licensee's methodology for review of EEEE's and identification of the applicable EEEEs in Attachment C of the LAR, the NRC staff concludes that the use of EEEEs meets the requirements of NFPA 805 (Reference 3), the guidance of RG 1.205 (Reference 4), and FAQ 08-0054 (Reference 58), and is, therefore, acceptable.

3.5.1.5 Variances from Deterministic Requirements For all fire areas, 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. The documented variances were all represented as separation issues. The licensee used the following strategies to resolve the VFDRs:

  • A FRE determined that applicable risk, DID, and safety margin criteria were satisfied without further action; or
  • A FRE determined that applicable risk, DID, and safety margin criteria were satisfied with a credited RA; or
  • A FRE determined that applicable risk, DID, and safety margin criteria were satisfied with a plant modification(s), as identified in 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 Attachment C of the LAR. Based on the VFDRs and associated resolutions as described in Attachment C of the LAR, 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 Attachment C of the LAR for each fire area.

3.5.1.6 Recovery Actions Attachment G of the LAR lists the RAs identified in the resolution of VFDRs in Attachment C of the LAR for each fire area. (See Section 3.5.1.7 of the SE below.)

The licensee stated in Attachments G and S of the LAR that the plant is committed to installing new equipment, including a dedicated diesel generator, and powering an existing SBAFW and a new injection pump, as part of the NFPA 805 project to be utilized as the primary equipment to address VFDRs associated with a loss of feedwater and a loss of charging, rather than Charging Pump 1A and the TDAFW. The licensee further stated that the SSD strategy changed to utilize this new equipment. According to the licensee, the operation of this equipment is much simpler and quicker to perform than the current actions, and is an enhancement to fire safety as it is isolated from the rest of the plant. The licensee also stated that it will remain free from fire damage for all fire scenarios of concern. The licensee further stated that this shutdown strategy does not require the use of the ABELIP and IBELIP.

For the reasons above, the licensee stated that Table G-1 in Attachment G of the LAR does not identify any required activities that occur at the PCSs (i.e., PCS actions are no longer required in order to address any VFDRs). The licensee further stated that with the installation of the new equipment, there are no activities required. According to the licensee, since the new equipment is not controlled from a PCS, all associated actions are considered RAs. The licensee stated that activities that are identified as RAs that are necessary to address risk are identified as RISK on Table G-1 in Attachment G of the LAR.

The NRC staff reviewed Section 4.2.1.3, "Establishing Recovery Actions," and Attachment G of the LAR to evaluate whether the licensee meets the associated requirements for the use of RAs per NFPA 805. The details of the NRC staff review for identifying RAs are described in SE Section 3.2.5. 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 the SE.

3.5.1.7 Recovery Actions Credited for Defense in Depth The process used by the licensee to establish RAs (as described in the LAR), required to meet the NSPC, did not utilize a category of "Recovery Actions Credited for Defense in Depth." The licensee followed the process in RG 1.205, Revision 1 (Reference 4), and NEI 04-02, Revision 2 (Reference 7), as supplemented by FAQ 07-0030 (Reference 50). In this process, the licensee determined the population of RAs required to resolve VFDRs (to meet the risk acceptance criteria or maintain a sufficient level of DID). According to the licensee, once a RA was identified as required, all RAs were treated in the same way and always referred to as a RA (regardless of whether the action was needed to meet the risk criteria or to provide DID).

In PRA RAI 07 (Reference 19), the NRC staff requested that the licensee discuss which method(s) were employed to evaluate the additional risk of RAs from fire scenarios outside the MCR, fire scenarios involving MCR abandonment, or fire scenarios which may involve actions both in the MCR and at a PCS. Additionally, the NRC staff requested that the licensee describe the compliant case used for the MCR additional risk of RA evaluation. The NRC staff also requested that the licensee discuss how shutdown panels ABELIP and IBELIP are treated for the compliant plant given that the SSD strategy changed to use new equipment and does not require use of these two shutdown panels, according to Attachment G of the LAR.

In its response to PRA RAI 07 (Reference 10), the licensee stated that concerning Previously Approved PCS Actions, the designated PCS at the plant is the IBELIP and ABELIP. The licensee stated that in the Appendix R plant, these actions were the primary credited actions for MCR abandonment scenarios. However, the licensee further stated that in the post-transition plant it was decided to first use the equipment associated with the strategy to use the SBAFW pump and new injection pump. The licensee also stated that these actions do not serve to keep one train free from fire damage, but that they provide DID for the new RAs. In addition, the licensee stated that although in the abandonment scenario these actions are no longer the first actions performed, the IBELIP and ABELIP still maintain their pre-transition designation as a PCS and so actions performed there are PCS actions. The licensee stated that since FAQ 0030 states, "activities that take place at primary control station(s) or in the Main Control Room are not recovery actions, by definition," the IBELIP and ABELIP actions are not RAs and there is no requirement to perform a delta-risk calculation.

The NRC staff determined that the licensee's response to PRA RAI 07 is acceptable because the licensee discussed how shutdown panels ABELIP and IBELIP are treated given that the SSD strategy does not require use of these two shutdown panels, according to Attachment G of the LAR.

The NRC staff reviewed Section 4.2.1.3, "Establishing Recovery Actions," and Attachment G of the LAR to evaluate whether the licensee meets the associated requirements for the use of RAs per NFPA 805. The NRC staffs evaluation of the licensee's process for identifying RAs and

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assessing their feasibility is provided in Section 3.2.5 of the SE, "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 SSD 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 Attachment C of the LAR, 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 on Table B-1 in Attachment A of the LAR and as such are addressed in Section 3.1 of the SE.

3.5.1.9 Electrical Raceway Fire Barrier Systems On Table B-1 in Section 3.11.5 in Attachment A of the LAR the licensee stated that the existing Hemyc wrap configurations are not credited as a fire rated barrier (to meet the deterministic separation requirements in accordance with Section 4.2.3.3 of NFPA 805). Attachment C of the LAR documents that configurations that utilize Hemyc as an ERFBS have been evaluated using the FRE PB approach.

In FPE RAI 04 (Reference 19), the NRC staff requested that the licensee provide additional detail regarding the capability of the ERFBS to meet the requirements in Section 3.11.5 of NFPA 805 and to include a discussion of how the ERFBS duration is based on fire testing of similar material and application. In its response to FPE RAI 04 (Reference 10), the licensee stated that upon the completion of modifications to insert ceramic fiber material into the unistrut supports inside the steel cable chase to ensure a path for combustion products does not exist, the Hemyc in BR 1B (fire area BR 1B) will provide 25 mins. of fire resistance. The licensee stated that this fire resistance duration is based on fire testing performed similar configurations of Hemyc 1-hour fire barrier materials in accordance with fire test standard ASTM E-119 using acceptance criteria based on the failure temperature for Thermoplastic cable (205°C). The licensee further stated that the FPRA model will credit 25 mins. of protection beyond the point where cables in the back of BR1 B would normally be damaged. According to the licensee, the combustible loading in BR1B will be kept as low as reasonable achievable, such that the 1-hour structural steel fire proofing would not be compromised, as discussed on Table B-1 in Section 3.11.2 in Attachment A of the LAR. The licensee also stated that the modification to align the ERFBS in fire area BR1 B with the testing/analysis is addressed in item 7 on Table S-2 in Attachment S of the LAR. The NRC staff determined that the licensee's response to FPE RAI 04 is acceptable because the fire resistance rating used in the FPRA for the Hemyc ERFBS is based on fire testing to the required test standard as required by Section 3.11.5 of NFPA 805 and because the licensee identified a required action that will incorporate the provisions of NFPA 805 in the FPP and would be required by the proposed license condition.

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3.5.1.10 Conclusion for Section 3.5.1 As documented in Attachment C of the LAR, for those fire areas that used the PB approach in accordance with Section 4.2.4 of NFPA 805, 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 LB that were transitioned to the NFPA 805 LB were reviewed for applicability, as well as continued validity, and found acceptable;
  • VFDRs were evaluated and either found to be acceptable based on an integrated assessment of risk, DID, and safety margin, or modifications or RAs were identified and actions planned or implemented to address the issue (see Section 3.5.1.5 of the SE);
  • RAs used to demonstrate the availability of a success path to achieve the NSPC were evaluated and the additional risk of their use determined, reported, and found to be acceptable (see Sections 3.5.1.6 and 3.5.1.7 of the SE);
  • The licensee's analysis appropriately identified the fire protection SSCs required to meet the NSPC, including fire suppression and detection systems, as well as required fire protection features (see Sections 3.5.1.1 and 3.5.1.2 of the SE);
  • Fire area boundaries (ceilings, walls, and floors), such as fire barriers, fire barrier penetrations, and through penetration fire stops have been established (see Section 3.5.1.8 of the SE); and,
  • The ERFBS credited was documented on a fire area basis, verified to be installed consistent with tested configurations and rated accordingly and evaluated using a FRE that demonstrated the ability to meet the applicable acceptance criteria for risk, DID, and safety margin (see Section 3.5.1.9 of the SE).

Accordingly, the NRC staff concludes that each fire area utilizing the PB approach meets the applicable requirements in Section 4.2 of NFPA 805.

3.5.2 Clarification of Prior NRC Approvals As stated in Attachment T of the LAR, "Clarification of Prior NRC Approvals," there are no elements of the current FPP for which NRC clarification is needed.

3.5.3 Fire Protection during Non-Power Operational Modes Section 1.1 of NFPA 805, "Scope," states that:

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.

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Section 1.3.1 of NFPA 805, "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.

The NRC staff reviewed Section 4.3, "Non-Power Operational Modes" and Attachment D of the LAR, "NEI 04-02 Non-Power Operational Modes Transition," to evaluate the licensee's treatment of potential fire impacts during NPOs. The licensee followed the guidance used in the process described in NEI 04-02 (Reference 7) as modified by FAQ 07-0040 (Reference 55), for demonstrating that the NSPC are met for high risk evolutions (HREs) during NPO modes.

3.5.3.1 NPO Strategy and Plant Operating States In Section 4.3 and Attachment D of the LAR, 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). The licensee stated in Section 4.3.1 of the LAR that the process undertaken to demonstrate that the NSPC was met during NPO modes included:

1) Review of the existing outage management processes;
2) Identifying necessary equipment and cables;
3) Performing fire area assessments to identify plant locations where a single fire may damage all success paths of a key safety function (KSF); and
4) Managing those locations (called pinch-points) that are associated with fire-induced vulnerabilities during an outage. The licensee implemented the

As described in Attachment D of the LAR, the licensee's Outage Management Procedure (IP-OUT-2) defines HREs and describes KSFs and their associated systems for lower mode operations (i.e., DHR, Inventory Control, Reactivity Control, Spent Fuel Pool Cooling, and Electrical Power Systems (on-site and offsite). The licensee identified equipment and cables necessary to support the KSFs success paths. The licensee further stated that the operational modes and functional requirements for those systems and components were reviewed.

According to the licensee, the KSF success path equipment and cables were incorporated in the NPO database model. Following the identification of KSF equipment and cables, the licensee performed analysis on a fire area 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 and mitigate these pinch-points through the use of RAs and/or fire prevention/protection controls. As stated in Section 4.3.2 of the LAR, FM was not used to eliminate any pinch-points.

The NRC staff concludes that the NPO process described and documented by the licensee in Section 4.3 and Attachment D of the LAR is acceptable because it is consistent with FAQ 07-0040, which clarifies the guidance regarding reasonable assurance that a fire during NPOs will not prevent the plant from achieving the fuel in a safe and stable condition.

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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. Section 4.3 of the LAR 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 used.

In SSA RAI 02b (Reference 19), the NRC staff requested that the licensee describe for those components, which had not previously been analyzed in support of the at-power analysis or whose functional requirements may have been different for the non-power analysis, how cable selection was performed in accordance with approved project procedures. Additionally, the NRC staff requested that the licensee provide a list of the additional components and a list of those at-power components that have a different functional requirement for NPO. The NRC also requested the licensee to describe the difference between the at-power SSD function and the NPO function, and to include with this list a general description by system indicating why components would be selected for NPO and not be included in the at-power analysis.

In its response to SSA RAI 02b (Reference 10), the licensee stated that the NPO Modes Analysis identified systems used for accomplishment of required KSFs and grouped those components making up success paths into function codes. The licensee further stated that cable selection was not originally performed for 153 components listed on Table SSA 02-1 of the RAI response because they were not credited in the at-power analysis. According to the licensee, the majority of equipment required to maintain the NPO KSFs is the same as that required to safely shutdown the plant while at power. Table SSA 02-1 lists the additional 153 components that had not been analyzed in support of the at-power analysis, but have had cable selection completed and are addressed in the NPO analysis.

The licensee stated that as discussed in FAQ 07-0040, the components within the required KSF success paths are compared to the population of components contained in the Safe Shutdown Equipment List (SSEL) to determine if the component's function is already addressed as part of the SSA. Table SSA 02-2 listed components that are required for both SSD and for NPO, but have different normal or required positions. The licensee stated that if a component's function was appropriately addressed in the SSEL, no further action was required; otherwise additional cable selection was performed. The table listed 49 at-power components that have a different functional requirement for NPO.

The licensee stated that those components that are common between the SSD and NPO analysis can have different normal and required positions. The licensee further stated that this is because NPO normal position is the position of the component at the start of NPO, which is dependent on the specific mode of operation or POS. For example, according to the licensee, the existing NSCA may credit the valve in the closed position; however, the valve may be required open for shutdown modes of operation. Similarly, the licensee stated that NPO required position is the position of the component that is required to ensure the KSF Path the component is associated with is successful, which may be different from required hot shutdown position.

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According to the licensee, for the following systems, components have been selected for NPO and not included in the at-power analysis:

  • RCS - The pressurizer safety and relief valves may be opened or closed for specific success paths for Reactivity Control (RXC) or Inventory Control (INV) or Decay Heat Removal (OHR).
  • RHR - The valves may be opened or closed for specific success paths for RXC, INV, or OHR or for isolation of paths not used for RXC, INV or OHR. To ensure success paths, both trains of pumps are analyzed as energized if the train is available.
  • SWS - The valves are opened to provide SW cooling to the support functions of Instrument Air Compressors, CCW [component cooling water] heat exchangers, and SFP heat exchangers. To ensure success paths, both trains of pumps are analyzed as energized if the train is available.
  • eve - The valve may be opened for specific success paths for RXC or INV. To ensure Success Paths, both trains of pumps are analyzed as energized if the train is available.
  • MAC - Both trains of both 4160 VAC busses are analyzed as energized if the train is available, and offsite power is analyzed in the normal "50/50" alignment.
  • LAC - Bus Tie breakers are opened or closed to provide power to the support function of Instrument Air.
  • EAC - To ensure success paths, both trains of emergency on-site power are analyzed as energized if the train is available and offsite power is not available.
  • CCW - The valves are opened to provide CCW cooling for the CCW support function.

To ensure success paths, both trains of pumps are analyzed as energized if the train is available.

  • SFP - To ensure success paths for the SFP cooling support function, the pump is analyzed as energized if the train is available.
  • SIS - To ensure specific success paths for RXC and INV, both trains of pumps are analyzed as energized if the train is available.

The NRC staff determined that the licensee's response to SSA RAI 02b is acceptable because for those components, which had not previously been analyzed in support of the at-power analysis or whose functional requirements may have been different for the non-power analysis, the licensee provided a list of the additional components and a list of those at-power components that have a different functional requirement for NPO. Additionally, the licensee described the difference between the at-power SSD function and the NPO function and included a general description by system indication why components would be selected for NPO and not be included in the at-power analysis.

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The NRC staff concludes that the licensee's process to define and identify NPO systems, components, and cables, as described in LAR Section 4.3.2 and LAR Attachment Dis consistent with the guidance in FAQ 07-0040 (Reference 55). NPO systems, components and cables logically relate to KSFs in the NPO analysis database. The NRC staff concludes that the licensee's approach to identification of NPO systems, components and cables is acceptable.

3.5.3.3 NPO Key Safety Functions and SSCs Used to Achieve Performance Attachment D of the LAR defines the KSFs. The success paths to achieve the KSFs and the components required for the success paths were incorporated in the NPO database model.

Section 4.3.1 of the LAR states that the guidance in NEI 04-02 (Reference 7) and FAQ 07-0040, Revision 4 (Reference 55), were followed to perform the fire area assessments to identify areas where fires may cause damage to the credited equipment or where KSFs are achieved solely by crediting RAs.

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.

In SSA RAI 02c (Reference 19), the NRC staff requested that the licensee provide a list of KSF pinch points by fire area that were identified in the NPO fire area reviews including a summary level identification of unavailable paths in each fire area. In addition, the NRC staff requested the licensee to describe how these locations will be identified to the plant staff for implementation. In its response to SSA RAI 02c (Reference 10), the licensee stated that pinch points were identified on a fire zone basis, based on the loss of a KSF. The licensee further stated that 53 fire zones were analyzed. According to the licensee, 22 fire zones were determined to have no pinch points. The licensee also stated that 31 fire zones are pinch points, as every success path was lost for at least one KSF, and were found to have at least one or more pinch points. Table SSA Q2-3 of the RAI response provides a summary level identification of KSF pinch points on a fire zone I fire area basis. The table identifies the unavailable KSF success paths associated with each pinch point and the recommendations for addressing the pinch points. The licensee stated that 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 also stated the KSF pinch point locations will be identified to the plant staff through changes to the outage management procedure that governs fire protection DID features and shutdown risk management. The proposed options to reduce fire risk will include:

  • Limit hotwork in this fire zone during HRE conditions.
  • Prohibit hotwork in this fire zone during HREs.

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  • Verify that the available fire detection systems located in the fire zone are functional.

Post firewatch in affected fire zones prior to entering HRE conditions if system(s) are impaired.

  • Limit transient combustible storage in this fire zone during HRE conditions.
  • Prohibit transient combustible storage in this fire zone during HRE conditions.
  • Provide a firewatch (continuous or periodic) in this fire area during HRE conditions.

The NRC staff determined that the licensee's response to SSA RAI 02c is acceptable because the licensee provided a list of KSF pinch points by fire area that were identified in the NPO fire area reviews including a summary level identification of unavailable paths in each fire area.

Additionally, the licensee described how these locations will be identified to the plant staff for implementation.

In SSA RAI 02d (Reference 19), the NRC staff requested that the licensee provide a description of any actions, including pre-fire staging actions, being credited to minimize the impact of fire-induced spurious actuations on power operated valves (POVs) (e.g., air operated valves (AOVs) and motor operated valves (MOVs)) during NPO (e.g., pre-fire rack-out, actuation or "pinning" valves, and isolation of air supply). In its response to SSA RAI 02d (Reference 10),

the licensee stated that there are no actions, including pre-staging actions (e.g., pre-fire rack-out, locally pinning of valves, isolation of air supplies) that are credited to minimize the impact of fire induced spurious actuations on POVs. The licensee further stated that additional actions are not relied upon as a strategy to reduce fire risk. According to the licensee, the assessment of potential risk reduction options (including input from operations personnel) concludes that the actual additional risk posed by fire is best controlled through the options listed in NRC FAQ 07-0040. The licensee stated that the NPO strategy does not credit the following methods:

  • RAs - 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, equipment/material staging and movement, supplementary work force augmentation, planned equipment out of service or unavailable for service and resultant off-normal system lineups, 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 pre-emptive manner certain high consequence fire-induced failures, most notably spurious operations of key valves, was considered. 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.

The NRC staff determined that the licensee's response to SSA RAI 02d is acceptable because the licensee provided a description of any actions being credited to minimize the impact of

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fire-induced spurious actuations on POVs (e.g., AOVs and MOVs) during NPO (e.g., pre-fire rack-out, actuation or "pinning" valves, and isolation of air supply).

In SSA RAI 02e (Reference 19), the NRC staff requested that the licensee provide the following information during normal outage evolutions, certain NPO credited equipment will have to be removed from service and to describe the types of compensatory actions that will be used during such equipment down-time. In its response to SSA RAI 02e (Reference 10), the licensee stated that if NPO credited equipment is deliberately removed from service, the licensee will consider appropriate contingency measures to reduce fire risk at the impacted locations. The licensee further stated that the outage management procedure addresses pre-outage review for DID. According to the licensee, the procedure for Fire Protection I Appendix R Compensatory Actions currently exists for providing compensatory actions for Appendix R equipment removed from service. The licensee stated that this procedure uses several pre-fire compensatory actions that are consistent with the options endorsed by NRC FAQ 07-0040, and will be utilized at the plant. The licensee also stated that a similar approach will be used for the NFPA 805 NPO credited equipment. The NRC staff determined that the licensee's response to SSA RAI 02e is acceptable because the licensee described the procedures that provide types of compensatory actions that will be used during normal outage evolutions, in which certain NPO credited equipment will have to be removed from service.

In SSA RAI 02f (Reference 19), the NRC staff requested that the licensee identify the RAs and instrumentation relied upon in NPO and to describe how RAs feasibility is evaluated. In addition, the NRC staff requested the licensee to include in the description whether these variables have been or will be factored into operator procedures supporting these actions. In its response to SSA RAI 02f (Reference 10), the licensee stated that there are no RAs relied upon for the NPO analysis. However, the licensee further stated that RAs could provide an option to respond to plant conditions, equipment alignments, or equipment removed from service during an outage. According to the licensee, because no RAs are inherently relied upon, there are no instruments relied upon to provide operator cues for RAs. The licensee also stated that instruments, which are part of the NPO analysis are not credited for initiation of any action. The NRC staff determined that the licensee's response to SSA RAI 02f is acceptable because the licensee identified that there are no RAs relied upon for the NPO analysis.

Based on the information provided in the LAR, as supplemented, the NRC staff concludes that the licensee used methods consistent with the guidance provided in RG 1.205 (Reference 4) and FAQ 07-0040 (Reference 55) to identify the equipment required to achieve and maintain the fuel in a safe and stable condition during NPO modes. Furthermore, 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 actions are reflected in Attachment D and implementation items 1 and 4 on Table S-3 in Attachment S of the LAR. 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 are acceptable because the methods are consistent with approved uses in NRC guidance or other authoritative publications. and the licensee identified actions that will result in compliance with NFPA 805 and would be required by the proposed license condition.

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3.5.3.4 NPO Pinch Point Resolutions and Program Implementation The licensee identified power operated components needed to support an NPO KSF that all were included in the post-fire SSD equipment list, and none required additional circuit analysis.

In SSA RAI 02a (Reference 19), the NRC staff requested that the licensee, at a high level, identify and describe the changes to outage management procedures, risk management tools, and any other document resulting from incorporation of KSFs identified as part of NFPA 805 transition, and to include changes to any administrative procedures such as "Control of Combustibles." In its response to SSA RAI 02a (Reference 10), the licensee stated that the plant will ensure that outage management procedures, risk management tools, and other procedures that will incorporate KSFs are identified and changed as appropriate. The licensee further stated that the appropriate site procedures will be revised to provide additional guidance to be used specifically for HRE activities. According to the licensee, these changes will provide plant outage management and fire protection with mitigation strategies that can be put in place based on the specific conditions of a planned activity. The licensee stated that the following procedures currently implement shutdown risk and the essential work planning and implementing process. The licensee also stated that these and other procedures will be reviewed and revised as necessary to implement these changes and requirements to incorporate guidance from the NPO review. The licensee identified actions for procedure revisions in implementation items 1 and 4 on Table S-3 in Attachment S of the LAR and the NRC staff determined that these actions are acceptable because they will incorporate the provisions of NFPA 805 in the FPP and would be required by the proposed license condition.

Procedures to be considered for revision include:

  • IP-OUT-2, Outage Management;
  • IP-OPS-3, Conduct of lower mode operations;
  • A-3.1, Containment storage and closeout inspection;
  • A-54.7, Fire Protection Tour;
  • A-601.13, Fire Protection I Appendix R Compensatory Actions;
  • A-601.14, Appendix R Program Control;
  • FPS-16, Bulk Storage of Combustible Materials and Transient Fire loads;
  • CNG-CM-1.01-3004, PRA Process for Internal Evaluations;
  • CNG-MN-4.01-1001, Work Order Execution and Closure Process;
  • CNG-MN-4.01-1002, Work Order Initiation, Screening and Prioritization;

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  • CNG-MN-4.01-1003, Work Order Planning;
  • CNG-OM-1.01-1000, Outage Management;
  • CNG-OM-1.01-1001, Shutdown Safety Management Program; and
  • CNG-OP-4.01-1000, Integrated Risk Management The licensee stated that in preparing the revisions, the plant will consider the need to include direction to minimize transient combustibles, evaluate the need for fire tours, evaluate control of ignition sources, and consider other preventive measures throughout the plant during NPO, especially in the areas identified as having pinch points.

The NRC staff determined that the licensee's response to SSA RAI 02a is acceptable because the licensee identified and described the changes to outage management procedures, risk management tools, and any other document resulting from incorporation of KSF identified as part of NFPA 805 transition.

NFPA 805 requires that the NSPC be met during any operational mode or condition, including NPO. As described above, the licensee performed engineering analyses to demonstrate that it meets this requirement:

  • Identified the KSFs required to support the NSPC criteria during NPOs.
  • Identified the plant operating states where further analysis is necessary during NPOs.
  • Identified the SSCs required to meet the KSFs during the POSs analyzed.
  • Identified the location of these SSCs and their associated cables.
  • Performed analyses on a fire area basis to identify pinch points were one or more KSFs could be lost as a direct result of fire-induced damage.
  • Planned/implemented modifications to appropriate procedures in order to employ a fire protection strategy for reducing risk at these pinch points during HREs.

Accordingly, based on the information provided in the LAR, as supplemented, the NRC staff concludes that the licensee demonstrated that the NSPC are met during NPO modes and HREs and is acceptable because the methods are consistent with approved uses in NRC guidance or other authoritative publications, and the licensee identified actions that will result in compliance with NFPA 805 and those actions would be required by the proposed license condition.

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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 the PB approach, in accordance with Section 4.2.4 of NFPA 805.

For those fire areas that used a PB approach, the NRC staff confirmed the following:

  • 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 Section 4.2.4.2 of NFPA 805) to address risk impact, DID, and safety margin, and were found to be acceptable.
  • All RAs 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 Section 4.2.4 of NFPA 805.
  • The required automatic fire suppression and automatic fire detection systems were appropriately documented for each fire area.

Accordingly, the NRC staff concludes that there is reasonable assurance that each fire area utilizing the PB approach, meets Section 4.2.4 of NFPA 805.

The NRC staff concludes that the licensee's analysis and outage management process during NPO provides reasonable assurance that the NSPC will be met during NPO modes and HREs, and that the licensee used methods consistent with the guidance provided in RG 1.205 (Reference 4) and FAQ 07-0040 (Reference 55). The NRC staff also concludes that no RAs are required during any NPO modes and the normal FPP DID actions are credited for addressing the risk impact of those fires during lower risk NPO modes. The NRC staff concludes that this overall approach for fire protection during NPO modes is acceptable because the methods are consistent with approved uses in NRC guidance or other authoritative publications, and the licensee identified actions that will result in compliance with NFPA 805 and those actions would be required by the proposed license condition.

3.6 Radioactive Release Performance Criteria 3.6.1 Method of Review Chapter 1 of NFPA 805 defines the radioactive release goals, objectives, and performance criteria that must be met by the FPP in the event of a fire at a NPP in any plant operational mode as follows:

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Radioactive Release Goal 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.

Radioactive Release Objective Either of the following objectives shall be met during all operational modes and plant configurations.

(1) Containment integrity is capable of being maintained.

(2) The source term is capable of being limited.

Radioactive Release Performance Criteria 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 10 CFR Part 20 limits.

The NRC staff endorsed (with certain exceptions) the guidance in NEI 04-02 (Reference 7) as providing methods acceptable to the NRC staff for adopting a FPP consistent with NFPA 805 and 10 CFR 50.48(c) in RG 1.205 (Reference 4). As described in the LAR (Reference 9), the licensee assessed its current FPP using the methodology contained in NEI 04-02 and FAQ 09-0056 (Reference 59).

The NRC reviewed the LAR to determine if the planned modifications to the licensee's FPP would provide an acceptable transition such as to meet the radioactive release performance criteria requirements of a RI/PB FPP, in accordance with 10 CFR 50.48(a) and (c) using the guidance in RG 1.205 and Section 9.5.1.2 of NUREG-0800 (Reference 36). The NRC staff also performed a review of the licensee's evaluation to determine whether the FPP will be capable of meeting the NFPA radioactive release goals, objectives, and performance criteria.

The results of the NRC staff review and evaluation are provided below.

3.6.2 Scope of Review The licensee performed an evaluation of the capability of the FPP to meet the goals, objectives, and performance criteria of NFPA 805 for all plant operating modes (including power and NPOs) and for all plant areas. The licensee's review found that the fire suppression activities, as defined in the pre-fire plans and fire brigade firefighting instruction operating guidelines, were written and valid for any plant operating mode. The NRC staff concludes that the scope of the licensee's assessment is adequate because the review included all modes of plant operation and all plant areas.

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3.6.3 Identification of Plant Areas Containing Radioactive Materials and Providing Containment during Fire Fighting Operations The licensee performed a screening of plant fire areas to determine where there was a potential for generating radioactive effluents during firefighting operations. The results of the compartment review are documented in Attachment E of the LAR, "NEI 04-02 Radioactive Review Transition." The fire areas where there were no radioactive materials present were identified and eliminated from further review. Each fire area that had the potential for generation of radioactive effluents created by firefighting activities was identified (screened in) for further evaluation.

For each screened in fire area, the licensee's review identified the existing engineering controls that were sufficient to contain gaseous and liquid effluent. The plant's engineering controls are identified and documented in Attachment E of the LAR. The licensee's review identified that those plant areas where most of the radioactive materials were present, such as the reactor containment building, auxiliary building, intermediate building, and service building, had engineered controls for adequate containment of liquid and gaseous effluent. The NRC staff determined that the identified engineering controls for these buildings were adequate because they provided sufficient capacity to contain the gaseous and liquid firefighting effluents, and the licensee has administrative controls in case the engineering controls were inoperable.

The licensee's review also identified other plant areas where radioactive materials were present where there were minimal or no engineered controls for containment of effluents.

These areas include the Contaminated Storage Building, Upper Radwaste Building, Radmaterial Storage Building, On-site Warehouse, Butler Building, and Canister Preparation Building.

The NRC staff concludes that the licensee's identification of potentially affected areas is an adequate assessment because the review incorporated all plant areas, and identified potentially affected areas with and without engineering controls, in accordance with the guidance in NEI 04-02 (Reference 7) as endorsed by RG 1.205 (Reference 4).

3.6.4 Fire Pre-Plans The licensee reviewed the existing fire pre-plans to determine whether the FPP is adequate to ensure that gaseous and liquid radioactive effluents generated as a direct result of fire suppression activities would be contained and monitored before release to unrestricted areas.

The results of the licensee's review are documented in Attachment E of the LAR. This review included the following steps:

  • Identification of applicable documentation; including fire pre-plan, procedures, and support drawings.
  • Review of current documentation to identify whether the current procedures and training documents discuss the containment and monitoring of potential contamination involving fire suppression activities.

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  • Review of engineering controls for gaseous effluents to determine in which areas the gaseous effluents are contained (for example contaminated smoke and related particulates).
  • Review of engineering controls for liquid effluents to determine in which areas the liquid effluents are contained (for example automatic or manual fire-fighting water).
  • An identification of those documents needing revision such as to provide for monitoring and containment of fire suppression agents and radioactive release.

The NRC staff concludes that the licensee's evaluation of the fire pre-plans is adequate because the review was comprehensive and was performed in accordance with the guidance in Appendix G of NEI 04-02 (Reference 7), as endorsed by RG 1.205 (Reference 4).

3.6.5 Gaseous Effluent Controls In areas where engineering controls exist for containment, filtering, and monitoring of gaseous effluent, the licensee determined that the engineering controls provided adequate containment because the effluent was either contained, or filtered to remove radioactive materials and subsequently monitored prior to discharge. The licensee determined that if the engineering controls were inoperable due to a fire (e.g., sump pump, filter, ventilation flow path),

administrative controls would be used to ensure effluents are contained within the structure boundaries.

For plant areas where the installed engineering controls were determined as adequate to contain the gaseous effluent, the NRC staff concludes that NFPA 805 radioactive release goals, objectives, and performance criteria will be met because the radioactive release will be manually contained to within acceptable limits.

For other areas without engineering controls, the licensee has established operational controls to limit the amount of radioactive material that is not adequately contained. In addition, the licensee modified the FPP such that the Fire Brigade is supported from Radiation Protection personnel to manually establish containment and perform monitoring of potential radioactive effluent. In addition, the magnitude of a potential radioactive release was determined, and the potential radiation exposure to members of the public was evaluated in a quantitative assessment and bounding analysis.

The NRC reviewed the licensee's calculation methods used to perform the quantitative assessment and bounding analysis. The licensee identified the inventory of radioactive material that was present and potentially could be discharged during a fire. The licensee's results of the analysis concluded that the maximum offsite dose at the Exclusion Area Boundary was less than the 10 CFR 20 dose limits for members of the public.

The NRC staff determined that the licensee's assessment is adequate because models and assumptions used were consistent with analytical methods that are recognized by the NRC as acceptable methods; i.e., the ODCM is a document required by the plant's technical specifications and is prepared in accordance with NRC regulatory guidance, and the other analytical models are acceptable models used by Federal Agencies such as EPA and NRC.

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The NRC staff concludes that the licensee adequately quantified and limited the maximum amount of radioactive material that can be released as a gaseous effluent. The NRC staff also concludes that the public dose from radioactive material released as a gaseous effluent during a fire would not exceed the radiological release performance criteria of NFPA 805 and the public dose limits of 10 CFR 20.

3.6.6 Liquid Effluent Controls The licensee identified those areas where sufficient engineering controls exist for containment of liquid effluent (e.g., floor drains routed to sumps and tanks). The NRC staff reviewed those engineering controls and determined that those controls provided adequate containment because the effluent is collected, stored, processed and monitored prior to discharge.

The licensee's review also identified those areas where there were minimal or no engineered controls for a potential radioactive liquid effluent release during firefighting activities. To mitigate this potential release, the licensee revised the FPP procedures and training programs to have the fire brigade and Radiation Protection staff install flood barriers to control a potential liquid effluent release. The licensee also will perform radiological monitoring as needed to determine whether containment of contaminated fire suppression agents (e.g., fire hose water runoff) is needed in order to limit the radioactive release to within acceptable levels.

For liquid effluent that may be discharged into storm drains and discharged into Lake Ontario, and subsequently enter the Ontario Water District drinking water intake, the licensee performed a bounding assessment to determine the potential radiological impact. Another assessment was made of the potential radiological impact of fire suppression water entering the local municipal sewerage system. The assessments concluded that the potential radiological impact would not exceed the radiological release performance criteria of NFPA 805 and the public dose limits of 10 CFR 20.

The NRC staff reviewed the calculation methods and concludes that the licensee adequately assessed the potential dose impact of uncontained liquid effluent because the bounding assessment was based on conservative assumptions and adequate analytical methods, and would not exceed the radiological release performance criteria of NFPA 805 and the public dose limits of 10 CFR 20.

3.6.7 Fire Brigade Training Materials The licensee reviewed and revised the Fire Brigade Lesson Plans materials to ensure they were consistent with the pre-fire plans in terms of containment and monitoring of potentially contaminated smoke and fire suppression water. The review is documented in Attachment E of the LAR.

The training materials were found to reinforce the use of the pre-fire plans. Each training module and lesson plan was evaluated, and those training materials needing improvements were identified and revised. The revised training materials are listed in Attachment E of the LAR.

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The training material revisions describe the presence and use of monitoring methods for gaseous liquid effluents, with instructions to the Fire Brigade to monitor and determine if such systems are operational and capable of supporting containment of effluents, including manual mitigation methods.

The NRC staff reviewed the licensee's evaluation of training materials and concludes that the training materials are adequate to instruct the fire brigade staff to implement the FPP because plant staff will be informed and capable of taking actions to limit the public dose to within the radiological release performance criteria of NFPA 805.

3.6.8 Conclusions The NRC staff's evaluation is based on:

(1) Information and analyses provided in the LAR; (2) Use of installed and manual engineered controls to contain potential releases; (3) Use of fire pre-plans; (4) Use of revised fire brigade response procedures and training procedures; and (5) Dose assessments when containment of radioactive release is not fully effective.

Based on these factors, 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 are as low as reasonably achievable and are not likely to exceed the radiological release performance criteria of NFPA 805 and the radiological dose limits in 10 CFR Part 20. The NRC staff also concludes that the licensee's FPP complies with the requirements specified in Sections 1.3.2, 1.4.2, and 1.5.2 of NFPA 805, and that this approach is acceptable.

3.7 NFPA 805 Monitoring Program For this section of the SE, the following requirements in Section 2.6 of NFPA 805 (Reference 3) are applicable to the NRC staff's review of the licensee's LAR (Reference 9):

Section 2.6 of NFPA 805, "Monitoring," states that:

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.

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Section 2.6.1 of NFPA 805, "Availability, Reliability, and Performance Levels," states that:

Acceptable levels of availability, reliability, and performance shall be established.

Section 2.6.2 of NFPA 805, "Monitoring Availability, Reliability, and Performance," states that:

Methods to monitor availability, reliability, and performance shall be established.

The methods shall consider the plant operating experience and industry operating experience.

Section 2.6.3 of NFPA 805, "Corrective Action," states that:

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 Section 4.6 of the LAR, "Monitoring Program" (Reference 9) that the licensee developed to monitor availability, reliability, and performance of FPP systems and features after transition to NFPA 805. The focus of the NRC staff review was on 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 Section 2.7 of the SE.)

The licensee stated that it will develop an NFPA monitoring program consistent with FAQ 10-0059 (Reference 60). 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 determined that the licensee's NFPA 805 monitoring program, and development and implementation process is acceptable and assures that Ginna 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;

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  • Establishes availability, reliability and performance criteria for the SSCs being monitored; and
  • Requires corrective actions when SSC 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 NFPA 805 Monitoring Program is included in implementation item 2 of Table S-3 in Attachment S of the LAR, and the NRC staff concludes that this action is acceptable because it will incorporate the provisions of NFPA 805 in the FPP and would be required by the proposed license condition.

The NRC staff concludes that completion of the monitoring program in the same schedule as the implementation of NFPA 805 is acceptable because the monitoring program will be completed with the other implementation items as described on Table S-3 in Attachment S of the LAR within 180 days after NRC approval, unless that date falls within a scheduled refueling outage, then implementation will occur within 60 days after plant startup from that scheduled refueling outage, which is prior to completion of the modifications to achieve full compliance with 10 CFR 50.48(c) (which is prior to startup from the second refueling outage greater than 12 months after issuance of the SE).

3.7.1 Conclusion for Section 3.7.1 The NRC staff reviewed the licensee's RI/PB FPP and concludes that the licensee's approach for meeting the requirements in Section 2.6 of NFPA 805, regarding the monitoring program is acceptable and that there is reasonable assurance that the licensee with develop a monitoring program that meets the requirements specified in Sections 2.6.1, 2.6.2, and 2.6.3 of NFPA 805 because the licensee identified an action to revise plant documents to monitor and trend the FPP, and included that action as an implementation item which would be required by the proposed license condition.

3.8 Program Documentation, Configuration Control, and Quality Assurance For this section of the SE, the requirements from Section 2.7 of NFPA 805, "Program Documentation, Configuration Control and Quality" (Reference 3), are applicable to the NRC staff's review of the LAR in regard to the appropriate content, configuration control, and quality of the documentation used to support the FPP transition to NFPA 805.

Section 2.7.1.1 of NFPA 805, "General," states that:

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

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carefully so that it can be checked for adequacy and accuracy either by an independent reviewer or by the AHJ.

Section 2.7.1.2 of NFPA 805, "Fire Protection Program Design Basis Document," states that:

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.

Section 2.7.1.3 of NFPA 805, "Supporting Documentation," states that:

Detailed information used to develop and support the principal document shall be referenced as separate documents if not included in the principal document.

Section 2.7.2.1 of NFPA 805, "Design Basis Document," states that:

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.

Section 2.7.2.2 of NFPA 805, "Supporting Documentation," states that:

Detailed supporting information shall be retrievable records. Records shall be revised as needed to maintain the principal documentation up-to-date.

Section 2.7.3.1 of NFPA 805, "Review," states that:

Each analysis, calculation, or evaluation performed shall be independently reviewed.

Section 2.7.3.2 of NFPA 805, "Verification and Validations" states that:

Each calculation model or numerical method used shall be verified and validated through comparison to test results or comparison to other acceptable models.

Section 2.7.3.3 of NFPA 805, "Limitations of Use," states that:

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.

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Section 2.7.3.4 of NFPA 805, "Qualification of Users," states that:

Cognizant personnel who use and apply engineering analysis and numerical models (e.g., FM 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.

Section 2. 7.3.5 of NFPA 805, "Uncertainty Analysis" states that:

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 Section 4. 7 .1 of the LAR, "Compliance with Documentation Requirements in Section 2.7.1 of NFPA 805" (Reference 9), to evaluate the appropriateness of the content of the FPP design basis document and supporting documentation.

The 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 Figure 4-9 in the LAR, "NFPA 805 Planned Post-Transition Documentation and Relationships." The licensee stated that the analyses conducted to support the NFPA 805 transition were performed in accordance with Ginna processes which meet or exceed the requirements for documentation outlined in Section 2.7.1 of NFPA 805.

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 Ginna's document control system.

These analyses 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 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 in Sections 2.7.1.1, 2.7.1.2, and 2.7.1.3 of NFPA 805, regarding adequate development and maintenance of the FPP design basis documentation, is acceptable.

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3.8.2 Configuration Control The NRC staff reviewed Section 4.7.2 of the LAR, "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.

To support the many other technical, engineering and licensing programs, 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 Ginna 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 document control system. The licensee also stated that analyses based on the PRA program, which includes the FREs, 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 ASME/ANS PRA Standard (Reference 34).

Configuration control of the FPP during the transition period is maintained by the Design Engineering and Configuration Control process, as defined in existing configuration management and configuration control procedures. The licensee identified an action in implementation item 5 of Table S-3 in Attachment S of the LAR to revise these existing procedures as necessary for application to the NFPA 805 FPP. The NRC staff concludes that this action is acceptable because it will incorporate the provisions of NFPA 805 in the FPP and because it would be required by the proposed license condition. The NRC staff reviewed the licensee's process for updating and maintaining the Ginna FPP, in order to reflect plant changes made after completion of the transition to NFPA 805.

Based on the LAR description of the configuration control process, which indicates that the new 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 that provides reasonable assurance that the requirements in Sections 2.7.2.1 and 2.7.2.2 of NFPA 805 are met.

3.8.3 Quality The NRC staff reviewed Section 4. 7 .3 of the LAR, "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.

3.8.3.1 Review Section 2.7.3.1 of NFPA 805 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 10 CFR 50.48(c). The LAR also states that the analyses, calculations, and evaluations

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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 in Section 2.7.3.1 of NFPA 805 is acceptable.

3.8.3.2 Verification and Validation (V&V)

Section 2.7.3.2 of NFPA 805 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 Section 4.7.3 of the LAR, 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 Ginna 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 46) 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 NUREG-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, 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 Attachment J of the LAR, 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, which were not addressed in NUREG-1824, were identified and described in an authoritative publication.

As reflected on Tables 3.8-1 and 3.8-2 in Attachments A and B of the SE, the FM employed by the licensee in the development of the FRE used either: ( 1) empirical correlations that provide

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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 FRE is 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 NUREG-1824, an authoritative publication, or national research laboratory report.

3.8.3.2.2 Discussion of RAls By letter dated October 9, 2013 (Reference 19), the NRC staff requested that the licensee provide additional information concerning the FM conducted to support the Ginna FRE. By letters dated December 17, 2013 (Reference 10), January 29, 2014 (Reference 11 ), and February 28, 2014 (Reference 12), the licensee responded to these RAls.

  • In FM RAI 03(1) (Reference 19), the NRC staff requested that the licensee provide additional information and documentation to justify the acceptability of the heat and smoke detection models that are referred to as "the prevailing models for estimating activation times," but for which no V&V basis is provided on Table J-1 in Attachment J of the LAR.

In response (Reference 11 ), the licensee stated that the heat and smoke detection activation models used in the FPRA are described in Chapters 10 and 11 of NUREG-1805, respectively. The latter is based on the method developed by Mowrer. The licensee stated that automatic detection is credited in every fire zone where the system is available, but that suppression is only credited for fire scenarios that propagate beyond the ignition source and the initial target set if the timing results from the activation models indicate that the system can activate before the ignition source and the initial target set are assumed damaged. The licensee further stated that in scenarios where suppression is credited, no targets except for the ignition source and initial target set are assumed to be damaged.

The NRC staff concludes that the licensee's response is acceptable because the heat and smoke detection models used by the licensee are described in NUREG-1805. 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.

  • In FM RAI 03(3) (Reference 19), the NRC staff requested that the licensee provide*

technical documentation to demonstrate that CFAST has been applied in the HGL calculations with input parameters that are within the validated range reported in the V&V basis documents.

In response (Reference 11 ), the licensee stated 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. For scenarios in

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which normalized parameters were determined to fall outside the validation range, results from sensitivity cases were used to justify the use of CFAST.

The NRC staff concludes that the licensee's response is acceptable because the licensee demonstrated that the CFAST used in HGL calculations was either used within the validated range of input parameters, or that the licensee provided justification for their application outside the validated range.

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. The action to revise the applicable post-transition processes and procedures to include NFPA 805 requirements for V&V are included in implementation item 6 on Table S-3 in Attachment S of the LAR. The NRC staff concludes that this action is acceptable because it will incorporate the provisions of NFPA 805 in the FPP and because it would be required by the proposed license condition.

3.8.3.2.4 Conclusion for Section 3.8.3.2 Based on the licensee's description of the process for V&V of calculation models and numerical methods and their continued use post-transition, the NRC staff concludes that the licensee's approach is acceptable because the models are consistent with approved uses in NRC guidance, other authoritative publications or national research laboratory reports (References 83

- 89), and the licensee identified actions that will result in compliance with Section 2.7.3.2 of NFPA 805 and those actions would be required by the proposed license condition.

3.8.3.3 Limitations of Use Section 2.7.3.3 of NFPA 805 requires that acceptable engineering methods and numerical models be used for transition only 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 Section 2.7.3.3 of NFPA 805 and that the engineering methods and numerical models used post-transition will be subject to these same limitations of use. As an example, in Attachment J of the LAR, the licensee stated that the fire models developed to support the NFPA 805 transition 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. The action to revise the applicable post-transition processes and procedures to include NFPA 805 requirements for limitations of use is included in implementation item 6 on Table S-3 in Attachment S of the LAR. The NRC staff concludes that this action is acceptable because it will incorporate the provisions of NFPA 805 in the FPP and because it would be required by the proposed license condition.

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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 in Attachments A and B of the SE summarize the fire models used, how each was applied in the FRE, the V&V basis for each, and the NRC staff evaluation for each.

3.8.3.3.1 Discussion of RAls By letter dated October 9, 2013 (Reference 19), the NRC staff requested that the licensee provide additional information concerning the FM conducted to support the FRE. By letters dated December 17, 2013 (Reference 10), January 29, 2014 (Reference 11 ), and February 28, 2014 (Reference 12), the licensee responded to these RAls.

  • In FM RAI 04(1) (Reference 19), the NRC staff requested that the licensee demonstrate that the algebraic fire models were used within their limits of applicability, or, to provide technical justification for all cases where an algebraic model was used outside this range.

In response (Reference 11 ), the licensee stated that the application of algebraic models was justified based on the validation ranges for the non-dimensional input parameters for each model specified in NUREG-1824. Supplemental validation was provided for cases that do not fall within the NUREG-1824 non dimensional parameter space. This supplemental validation was based on the experimental data from which the model was derived and, in the case of the Point Source Model, on validation efforts conducted by the Society of Fire Protection Engineers. The licensee's response included a table with the relevant NUREG-1824 non-dimensional parameter values for all areas, zones, transient zones and scenarios where algebraic models were used to calculate plume temperature and point source radiation. In most cases, the non-dimensional parameters are within the NUREG-1824 validation range. The licensee provided a general discussion to justify the use of these algebraic models in cases where a non-dimensional parameter is outside the range of applicability. The licensee also stated that no algebraic models are used in the FPRA to determine when to credit sprinkler, heat detector, and smoke detector activation.

The NRC staff determined that the licensee's response is acceptable because the algebraic fire models were either applied within their validated range of input parameters or their application outside the validated range was justified.

  • In FM RAI 04(2) (Reference 19), the NRC staff requested that the licensee provide technical justification for applying CFAST in the HGL calculations for fires with a Froude number outside the NUREG-1824 validation range. In addition, the NRC staff requested that the licensee provide technical justification for choosing electric panel and transient fire areas in the MCR abandonment calculations so that the Froude number is in the NUREG-1824 validation range of 0.4-2.4.

In response to the first part of the RAI (Reference 11 ), the licensee stated that in most cases in which the Froude number is outside the validation range, it is lower than the minimum NUREG-1824 range value of 0.4. The licensee showed that in these cases the temperature and flame height were overestimated. In a few cases, the Froude

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number exceeds the maximum validation range of 2.4. These cases are associated with conservative modeling of oil fires with a relatively high HRR in a relatively small pool, which result in large flame heights and a time to damage of approximately 2 mins. The licensee stated that within this 2 minute time frame, the full zone of influence was failed.

The NRC staff determined that the licensee's response to the first part of the RAI is acceptable because even though the fire models use a Froude number outside the validated range, the Froude number is below the lower limit of the validated range provided in NUREG-1824. The NRC staff also concludes that the results of the CFAST HGL calculations for the fires with a Froude number outside the NUREG-1824 validation range are acceptable because the results are conservative.

In response to the second part of the RAI (Reference 11 ), the licensee explained that there is no simple or obvious way to compute a meaningful Froude number for fires in electrical panels, and that electrical panel fires were treated as open configuration source fires consistent with the guidance in NUREG/CR-6850, Volume 2 and Supplement 1. The licensee also stated that the fire diameter was chosen so that it produces a Froude number within the NUREG-1824 validation range. The licensee further stated that the transient areas were chosen to obtain the most conservative results within the validation range, and that the resulting areas are consistent with the MCR configuration. The licensee further stated that the results of a sensitivity analysis suggested that the use of a fire diameter corresponding to the lower limit of the NUREG-1824 Froude number validation range for each transient HRR bin recommended in NUREG/CR-6850, Appendix G resulted in bounding abandonment conditions.

The NRC staff determined that the licensee's response to the second part is acceptable because the licensee provided adequate technical justification for choosing electric panel and transient fire areas in the MCR abandonment calculations.

  • In FM RAI 04(3) (Reference 19), the NRC staff requested that the licensee verify that the CFAST model was always used within the range of acceptable room length-to-width and height-to-width aspect ratios, or, if not, to provide technical justification for the use the model with ratios outside the limits of applicability.

In response (Reference 11 ), the licensee stated that the aspect ratios for each fire zone were compared against the NUREG-1824 validation range. In cases where an aspect ratio was found to be outside the range, a sensitivity analysis was conducted adjusting the room dimensions in the conservative direction to bring the ratio within the applicable range. According to the licensee, in all cases, the results of the sensitivity analyses suggested no change in the conclusions drawn from the original calculations.

The NRC staff determined that the licensee's response is acceptable because the licensee provided adequate justification for its use of CFAST in those cases where the model was applied in a compartment with an aspect ratio outside the NUREG-1824 validation range.

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  • In FM RAI 04(4) (Reference 19), the NRC staff requested that the licensee identify uses of CFAST outside its limits of applicability of the model, and to explain for those cases how the application of CFAST was justified.

In response (Reference 11 ), the licensee provided a table with the relevant NUREG-1824 non-dimensional parameter values for all fire zones where CFAST was used to calculate HGL temperature. In most cases, the non-dimensional parameters are within the NUREG-1824 validation range. The licensee performed sensitivity analyses to justify the use of CFAST in cases where a non-dimensional parameter is outside the range of applicability.

The NRC staff determined that the licensee's response is acceptable because in each case the licensee provided adequate justification for its use of CFAST in those cases where the fire model was applied outside the limits of its applicability. Further, results of the sensitivity analyses shows that no change in conclusion made with using room aspect ratios outside the validation range.

  • In FM RAI 04(5) (Reference 19), the NRC staff requested that the licensee identify uses of FDS outside its limits of applicability of the model, and to explain for those cases how the application of FDS was justified.

In response (Reference 11 ), the licensee explained that FDS was only used to perform detailed HGL calculations for the MCR and provided a table with the relevant NUREG-1824 non-dimensional parameter values for these calculations. In most cases, the non-dimensional parameters are within the NUREG-1824 validation range. The licensee provided a detailed discussion to justify the use of FDS in cases where a non-dimensional parameter is outside the range of applicability.

The NRC staff determined that the licensee's response is acceptable because in each case the licensee provided adequate justification for its use of FDS in those cases where the model was applied outside the limits of its applicability.

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. The action to revise the applicable post-transition processes and procedures to include NFPA 805 requirements for limitations of use is included in implementation item 6 on Table S-3 in Attachment S of the LAR. The NRC staff concludes that this action is acceptable because it will incorporate the provisions of NFPA 805 in the FPP and because it would be required by the proposed license condition.

3.8.3.3.3 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 plant's process for placing limitations on the use of engineering methods and numerical models, the NRC staff concludes that the

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licensee's approach is acceptable because the models are consistent with approved uses in NRC guidance or other authoritative publications, and the licensee identified actions that will result in compliance with Section 2.7.3.3 of NFPA 805 and those actions would be required by the proposed license condition.

3.8.3.4 Qualification of Users Section 2.7.3.4 of NFPA 805 requires that personnel performing engineering analyses and applying numerical methods (e.g., FM) 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. The licensee 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 engineering qualification process. As discussed below, the licensee is in the process of developing qualification or training requirements for personnel performing engineering analyses and numerical methods.

The NRC staff reviewed the engineering qualification process, and concludes that competent and experienced personnel developed the 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 RAls By letter dated October 9, 2013 (Reference 19), the NRC requested that the licensee provide additional information concerning the FM conducted to support the FRE. By letters dated December 17, 2013 (Reference 10), January 29, 2014 (Reference 11 ), and February 28, 2014 (Reference 12), the licensee responded to these RAls.

  • In FM RAI 05(1) (Reference 19), the NRC staff requested that the licensee describe the requirements that were used to qualify personnel for performing FM calculations in the NFPA 805 transition.

In response (Reference 10), the licensee explained that FM calculations are performed by a fire protection engineer who meets the qualification requirements in Section 2.7.3.4 of NFPA 805. The qualification process follows the guidance of ACAD 98-004, "Guidelines for Training and Qualification of Engineering Personnel." Personnel performing FM in support of the FPRA also require FPRA qualification. Personnel performing detailed FM analysis using tools such as CFAST or FDS are required to have the relevant education and experience in FM to perform the analysis.

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The NRC staff determined that the licensee's response is acceptable because the licensee described the requirements that were used to qualify personnel for performing FM calculations in the NFPA 805 transition.

  • In FM RAI 05(2) (Reference 19), the NRC staff requested that the licensee describe the process for ensuring that the FM personnel meet those qualifications, not only before the transition but also during and following the transition.

In response (Reference 10), the licensee explained that the credentials of the personnel who conducted the initial FM were provided by the vendor and reviewed and approved by CENG (now Exelon) supervision. During and following transition, 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 CENG (now Exelon) supervision. Engineering Supervisors have responsibility for verifying qualifications prior to assigning personnel to perform job performance requirement independently.

The NRC staff determined that the licensee's response is acceptable because the licensee described the process for ensuring that the FM personnel meet these qualifications, not only before the transition but also during and following the transition.

  • In FM RAI 05(3) (Reference 19), the NRC staff requested that the licensee explain how proper communication between the FM and FPRA personnel was ensured.

In response (Reference 10), the licensee explained that throughout the FPRA process, the fire protection engineers who conducted the FM and the PRA engineers maintained frequent communications. Both the fire protection engineers and the PRA engineers participated in the cutset review meetings during the development of the FPRA. The FM database will be maintained under the responsibility of the fire protection engineers and the PRA engineers.

The NRC staff determined that the licensee's response is acceptable because the licensee explained how proper communication between the FM and FPRA personnel was ensured.

Based on its review and the above explanations, the NRC staff determined that competent and experienced personnel developed the 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, Section 4.7.3 of the LAR, "Compliance with Quality Requirements in Section 2.7.3 of NFPA 805 Fire Protection Quality," states that:

... Post-transition, for personnel performing fire modeling or Fire PRA development and evaluation, Ginna will develop and maintain qualification requirements for individuals assigned various tasks. Position Specific Guides will be developed to identify and document required training and mentoring to ensure

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individuals are appropriately qualified per the requirements of NFPA 805 Section 2.7.3.4 to perform assigned work ...

The action to revise the post-transition qualification training program to include NFPA 805 requirements for qualification of users is identified in implementation item 8 on Table S-3 in Attachment S of the LAR. The NRC staff determined that this action is acceptable because it will incorporate the provisions of NFPA 805 in the FPP and because it would be required by the proposed license condition.

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 is acceptable because the methods are consistent with approved uses in NRC guidance or other authoritative publications. The licensee identified actions that will result in compliance with Section 2.7.3.4 of NFPA 805 and those actions would be required by the proposed license condition.

3.8.3.4.2 Conclusion for Section 3.8.3.4 Based on the above discussions, the NRC staff concludes that the qualification program addresses the requirements in Section 2.7.3.4 of NFPA 805, 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 Section 2.7.3.5 of 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.) 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.

3.8.3.5.1 General The ASME/ANS PRA Standard (Reference 34), includes requirements to address uncertainty.

Accordingly, the licensee addressed uncertainty as a part of the development of the FRE. The NRC staff's evaluation of the licensee's treatment of these uncertainties is discussed in Section 3.4.7 of the SE above.

According to NUREG-1855, Volume 1 (Reference 82), 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;

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(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 79); 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 RAls By letter dated October 9, 2013 (Reference 19), the NRC staff requested that the licensee provide additional information concerning the FM conducted to support the FRE. By letters dated December 17, 2013 (Reference 10), January 29, 2014 (Reference 11), and February 28, 2014 (Reference 12), the licensee responded to these RAls.

  • In FM RAI 06(1 ).a (Reference 19), the NRC staff requested that the licensee explain how the "parameter uncertainty was addressed in the detailed FM analyses.

In response (Reference 11 ), the licensee stated that parameter uncertainty is addressed by: (1) using conservative inputs; or (2) varying inputs in sensitivity cases. The licensee then described in detail how this was accomplished for FDS used in the MCR abandonment study and for CFAST in the HGL calculations.

The NRC determined that the licensee's response is acceptable because the sensitivity runs performed by the licensee provide assurance that FM input parameter uncertainty in the MCR abandonment study was properly accounted for.

  • In FM RAI 06(1 ).b (Reference 19), the NRC staff requested that the licensee explain how "model" uncertainty was addressed in the detailed FM analyses.

In response (Reference 11), the licensee explained that a V&V analysis was performed for all the FM (hand calculations, CFAST and FDS) used in the FPRA. The licensee further stated that this analysis determined whether models were used within their validated range. According to the licensee, if the models were found to be used outside of the range, input parameters were varied in a conservative direction and the revised model results were used as input to the FPRA.

The NRC determined that the licensee's response is acceptable because the licensee demonstrates that the model uncertainty was properly accounted for given that the licensee's process is consistent with the guidance in NUREG-1934.

  • In FM RAI 06(1 ).c (Reference 19), the NRC staff requested that the licensee explain how the "completeness" uncertainty was addressed in the detailed FM analyses.

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In response (Reference 11 ), the licensee explained that the FPRA is an integrated analysis, and that completeness uncertainty associated with fire models is addressed within the overall quantification process. The licensee discussed three examples to illustrate how the FPRA analyst conservatively compensated for the lack of FM capabilities outside the FM analysis.

The NRC staff determined that the licensee's response is acceptable because since the FPRA conservatively compensates for incomplete information, the completeness uncertainty 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. The action to revise the applicable post-transition processes and procedures to include NFPA 805 requirements regarding uncertainty analysis is included in implementation item 6 on Table S-3 in Attachment S of the LAR. The NRC staff concludes that this action is acceptable because it will incorporate the provisions of NFPA 805 in the FPP and because it would be required by the proposed license condition.

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 is acceptable because the methods are consistent with approved uses in NRC guidance or other authoritative publications, and the licensee identified actions that will result in compliance with Section 2.7.3.5 of NFPA 805 and those actions would be required by the proposed license condition.

3.8.3.6 Conclusion for Section 3.8.3 Based on the above discussions, the NRC staff concludes that the RI/PB FPP meets each of the requirements in Section 2.7.3 of NFPA 805, 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 guidance in Appendix C of NEI 04-02 (Reference 7) suggests that the LAR include a description of how the existing fire protection quality assurance program will be transitioned to the new NFPA 805 RI/PB FPP, as discussed below.

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The LAR stated that the fire protection QA program is included within and implemented by the Ginna nuclear QA program, although certain aspects of that program are not applicable to the FPP. The licensee included an action in implementation item 7 on Table S-3 in Attachment S of the LAR to revise the QA program to reflect the applicable requirements in Section 2.7.3 of NFPA 805. The NRC staff determined that this action is acceptable because it will incorporate the provisions of NFPA 805 in the FPP and because it would be required by the proposed license condition.

Based on its review and the above explanation, 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 fire protection systems that were previously not included within the scope of the fire protection QA program and that are required by NFPA 805 transition and post-transition, as identified in implementation item 7 on Table S-3 in Attachment S of the LAR.

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 Section 2.7 of NFPA 805, regarding program documentation, configuration control, and quality, and has identified actions that would be required by the proposed license condition.

4.0 FIRE PROTECTION LICENSE CONDITION The licensee proposed a FPP license condition regarding transition to a 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 RP C.3.1 in RG 1.205, Revision 1, as issued on December 18, 2009 (7 4 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 RE. Ginna Nuclear Station, and the NRC staff concludes that it is acceptable.

The following license condition is included in the revised license and will replace Operating License No. DPR-18 Condition 2.C(3):

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 March 28, 2013, supplemented by letters dated December 17, 2013; January 29, 2014; February 28, 2014; September 5, 2014; September 24, 2014; December 4, 2014; March 18, 2015; June 11, 2015; August 7, 2015; and as approved in the safety evaluation report dated November 23, 2015. Except where NRC approval

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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 x 10-7 /year (yr) for CDF and less than 1 x 1o- 8/yr for LERF. 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 the NFPA 805, Chapter 3, element is functionally equivalent to the corresponding technical requirement. A qualified

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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 3.10); 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 November 23, 2015, 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 and (c)3 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

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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 LAR Attachment S, Table S-2, "Plant Modifications Committed," of Exelon Generation letter dated June 11, 2015, to complete the transition to full compliance with 10 CFR 50.48(c) no later than prior to startup from the second refueling outage greater than 12 months after receipt of the safety evaluation. The licensee shall maintain appropriate compensatory measures in place until completion of these modifications.
3. The licensee shall complete the implementation items listed in LAR Attachment S, Table S-3, "Implementation Items," of Exelon Generation letter dated June 11, 2015, except Implementation Items 9, 15, 19, by 180 days after NRC approval unless that date falls within a scheduled refueling outage, then implementation will occur 60 days after startup from that scheduled refueling outage.

Implementation Items 9, 15, and 19 are associated with modifications described in Table S-2 and will be completed once the related modifications are installed and validated in the PRA model.

5.0

SUMMARY

The NRC staff reviewed the licensee's application, as supplemented by various letters, to transition to a RI/PB FPP in accordance with the requirements established by NFPA 805.

The NRC staff concludes that, subject to implementation of certain items in LAR, Attachment S, the applicant's approach, methods, and data are acceptable to establish, implement and maintain a RI/PB FPP in accordance with 10 CFR 50.48(c).

Implementation of the RI/PB 50.48(c) FPP must be in accordance with the new fire protection license condition, which identifies a list of modifications and implementation items that must be completed in order to support the conclusions made in this SE, and establishes a date by which full compliance with 10 CFR 50.48(c) will be achieved. Before the licensee is able to fully implement the transition to a FPP based on NFPA 805 and exercise self-approval as specified in the fire protection license condition, the modifications and implementation items in the transition license condition must be completed within the timeframe specified.

6.0 STATE CONSULTATION

In accordance with the Commission's regulations, the New York State official was notified on September 25, 2015, of the proposed issuance of the amendment. The state official had no comments.

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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 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 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 FR on February 4, 2014 (79 FR 6648). 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.

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.

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51. Klein, Alexander R., U.S. Nuclear Regulatory Commission, memorandum to file, "Close-out of National Fire Protection Association Standard 805 Frequently Asked Question 07-0035 on Bus Duct Counting Guidance for High Energy Arcing Faults," June 16, 2009 (ADAMS Accession No. ML091620572).
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. ML110140242).
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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).
56. Klein, Alexander R., U.S. Nuclear Regulatory Commission, memorandum to file, "Close-out of National Fire Protection Association 805 Frequently Asked Question 08-0048 Revised Fire Ignition Frequencies," September 1, 2009 (ADAMS Accession No. ML092190457).
57. Klein, Alexander R., U.S. Nuclear Regulatory Commission, memorandum to file, "Close-out of National Fire Protection Association Standard 805 Frequently Asked Question 08-0053:

Kerite-FR Cable Failure Thresholds," dated June 6, 2012 (ADAMS Accession No. ML121440155).

58. 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," March 10, 2015 (ADAMS Accession No. ML15016A280).
59. 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," January 14, 2011 (ADAMS Accession No. ML102920405).
60. 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," March 19, 2012 (ADAMS Accession No. ML120750108).

61. 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 on Updated Final Safety Analysis Report (UFSAR) Content," September 5, 2012 (ADAMS Accession No. ML121980557).
62. Hamzehee, Hossein G., U.S. Nuclear Regulatory Commission, memorandum to file, "Close-out of Fire Probabilistic Risk Assessment Frequently Asked Question 13-0004 on Clarifications Regarding Treatment of Sensitive Electronics," dated December 3, 2013 (ADAMS Accession No. ML13322A085).

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64. Hamzehee, Hossein, G., U.S. Nuclear Regulatory Commission, Memorandum to APLA Files, "Close-out of Fire Probabilistic Risk Assessment Frequently Asked Question 13-0006 on Modeling Junction Box Scenarios in a Fire PRA," dated December 12, 2013 (ADAMS Accession No. ML13331B213).
65. Hamzehee, Hossein, G., U.S. Nuclear Regulatory Commission, memorandum to APLA Files, "Close-out of Fire Probabilistic Risk Assessment Frequently Asked Question 14-0008 on Main Control Board Treatment," dated July 22, 2014 (ADAMS Accession No. ML14190B307).
66. Johnson, Allen, U.S. Nuclear Regulatory Commission, memorandum to file, "Request for Publication in Bi-Weekly FR Notice - Notice of Issuance of Amendment to Facility Operating Licensee (TAC No. M83568)," September 21, 1992 (ADAMS Accession No. ML010570343).
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Standard 256 (NFPA 256), 2003 Edition, Quincy, Massachusetts.

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dated September 10, 2015 (ADAMS Accession No. ML15253A062).

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Technical Reference Guide Volume 3: Validation," NIST Special Publication 1018-5, Gaithersberg, MD, 2010.

Principal Contributors:

Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation Jay Robinson, Harold Barrett, JS Hyslop, Naeem Iqbal, Alayna Pearson, Bernard Litkett, Dan O'Neal, and Steven Garry Center for Nuclear Waste Regulatory Analysis Jason Huczek, Debashis Basu, Robert Fosdick, and Marc Janssens Date: November 23, 2015 Attachments:

A. Table 3.8-1, V&V Basis for Fire Modeling Correlations Used at Ginna B. Table 3.8-2, V&V Basis for Other Fire Models and Related Calculations Used at Ginna C. Abbreviations and Acronyms

Attachment A: Table 3.8-1. V&V Basis for Fire Modeling Correlations Used at Ginna Application at Correlation V&V Basis NRC Staff Evaluation of Acceptability Ginna Flame Height The Flame Height NUREG-1805

  • The licensee provided verification of the coding of this (Method of Correlation was Chapter 3, 2004 correlation. (Response to FM RAI 04(1 ).a (Reference 11))

Heskestad) used to determine (Reference 45)

  • The correlation is validated in NUREG-1824 and an the vertical authoritative publication.

extension of the NUREG-1824

  • The licensee either determined that the correlation was flame region as part Volume 3, 2007 applied within the validated range reported in NUREG-1824, of the ZOI (Reference 46) or provided justification for cases where the correlation was calculations. used outside the validated range reported in NUREG-1824 SFPE Handbook (Response to FM RAI 04(1) (Reference 11 )).

4th Edition, Chapter 2-1, Based on its review and the licensee's explanation, the NRC staff Heskestad, 2008 concludes that the use of this correlation in the Ginna application (Reference 83) is acceptable.

Attachment A: Table 3.8-1. V&V Basis for Fire Modeling Correlations Used at Ginna Application at Correlation V&V Basis NRC Staff Evaluation of Acceptability Ginna Plume Centerline The Plume NUREG-1805

  • The licensee provided verification of the coding of this Temperature Centerline Chapter 9, 2004 correlation. (Response to FM RAI 04(1 ).a (Reference 11 ))

(Method of Temperature (Reference 45)

  • The correlation is validated in NUREG-1824 and an Heskestad) correlation was authoritative publication.

used to determine NUREG-1824

  • The licensee either determined that the correlation was the vertical Volume 3, 2007 applied within the validated range reported in NUREG-1824, separation distance, (Reference 46) or provided justification for cases where the correlation was based on used outside the validated range reported in NUREG-1824 temperature, to a SFPE Handbook (Response to FM RAls 04(1) (Reference 11 )).

target in order to 4th Edition, 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 83) concludes that the use of this correlation in the Ginna application is acceptable.

Attachment A: Table 3.8-1. V&V Basis for Fire Modeling Correlations Used at Ginna Application at Correlation V&V Basis NRC Staff Evaluation of Acceptability Ginna Radiant Heat The Radiant Heat NUREG-1805

  • The licensee provided verification of the coding of this Flux Flux (Point Source Chapter 5, 2004 correlation. (Response to FM RAI 04(1 ).a (Reference 11))

(Point Source Method) correlation (Reference 45)

  • The correlation is validated in NUREG-1824 and an Method) was used to authoritative publication.

determine the NUREG-1824

  • The licensee determined that the correlation was applied horizontal Volume 3, 2007 within the validated range reported in NUREG-1824 separation distance, (Reference 46) (Response to FM RAI 04(1) (Reference 11)).

based on heat flux, to a target in order SFPE Handbook, Based on its review and the licensee's explanation, the NRC staff to determine the 4th Edition, concludes that the use of this correlation in the Ginna application horizontal extent of Chapter 3-10, is acceptable.

the ZOI. Seyler, C., 2008 (Reference 84)

Attachment A: Table 3.8-1, V&V Basis for Fire Modeling Correlations Used at Ginna Application at Correlation V&V Basis NRC Staff Evaluation of Acceptability Ginna Heat and Smoke Heat and Smoke NUREG-1805

  • Licensee provided verification of the coding of these models.

Detection Detection Activation Chapters 10-11, (Response to FM RAI 03(1) (Reference 11))

Activation Model model used to 2004 (Reference

  • The heat detection modeling technique is validated in determine whether 45) authoritative publications.

automatic

  • The licensee stated that the models have been applied suppression can be NFPA Fire conservatively (Response to FM RAI 03(1) (Reference 11 )).

credited after the Protection initial target set is Handbook 19th Based on its review and the licensee's explanation, the NRC staff concludes that the use of these correlations in the Ginna damaged. Edition, 2003 application is acceptable.

(Reference 85)

SFPE Handbook, 4th Edition, 2008 (Reference 86)

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

  • The modeling technique is validated in NUREG-1824 and a Temperature was used to Volume 5 national research laboratory report.

Calculations calculate (Reference 46)

  • The licensee either determined that the model was applied HGL temperature in NIST Special within the validated range reported in NUREG-1824, or specific fire zones. Publication 1086 provided justification for cases where the model was used CFAST Zone outside the validated range reported in NUREG-1824 (Reference 87)

Model Version 6 (Response to FM RAls 03(3), 04(2), 04(3) and 04(4)

(Reference 11)).

CFAST zone model was also used for Based on its review and the information provided by the licensee, the temperature the NRC staff concluded that the use of CFAST model in the sensitive equipment Ginna application is acceptable.

HGL temperature study.

Attachment B: Table 3.8-2. V&V Basis for Other Fire Models and Related Calculations Used at Ginna Model Application at V&V Basis NRC Staff Evaluation of Acceptability Ginna Main Control Fire Dynamics NUREG-1824,

  • The modeling technique is validated in NUREG-1824 and Room Simulator Volume 7, 2007 a national research laboratory report.

Abandonment (Version 5) was (Reference 46)

  • The licensee stated that in most cases, the correlation has Time Calculation used to calculate been applied within the validated range reported in abandonment time NIST Special NUREG-1824. The licensee provided justification for Fire Dynamics for the Ginna MGR. Publication 1018- cases where the correlation was used outside the Simulator 5, Volume 2: validated range reported in NUREG-1824 (Response to Version 5 Verification FM RAI 04(5) (Reference 11 )).

(Reference 88)

NIST Special Based on its review and the information provided by the licensee, Publication 1018- the NRC staff concluded that the use of FDS model in the Ginna 5, Volume 3: application is acceptable.

Validation (Reference 89)

Attachment C: Abbreviations and Acronyms oc degrees Celsius OF degrees Fahrenheit ABELIP Auxiliary Building Emergency Local Instrument Panel ADAMS Agencywide Documents Access and Management System AFW auxiliary feedwater AHJ authority having jurisdiction ANS American Nuclear Society AOV air operated valves APCSB Auxiliary and Power Conversion Systems Branch ARVs atmospheric relief valves ASME American Society of Mechanical Engineers BR1B Battery Room 1B BTP Branch Technical Position BTU British Thermal Unit BWR boiling-water reactor CAROLFI RE Cable Response to Live Fire cc Capability Category CCDP conditional core damage probability CCF Common Cause Failure ccw component cooling water CDF core damage frequency CENG Constellation Energy CFAST consolidated model of fire and smoke transport CFR Code of Federal Regulations CPTs control power transformers CRE Control Room Envelope CST Condensate Storage Tank DESIREE-Fire Direct Current Electrical Shorting in Response to Exposure Fire OHR Decay Heat Removal DID defense-in-depth EOG emergency diesel generator EEEE existing engineering equivalency evaluation EOF Emergency Operations Facility EOP Emergency Operating Procedures EPIP Emergency Plan Implementing Procedure EPRI Electric Power Research Institute ERFBS electrical raceway fire barrier system ERO Emergency Response Organization F&O facts and observations FAQ frequently asked question FDS fire dynamics simulator FDT fire dynamics tool FM fire modeling FPE fire protection engineering FPP fire protection program FPRA fire probabilistic risk assessment

FR Federal Register Fr Froude FRE fire risk evaluation FSAR final safety analysis report ft feet GDC general design criteria Ginna R.E. Ginna Nuclear Power Plant GL generic letter HEP human error probability HFEs human failure events HGL hot gas layer HRA human reliability analysis HRE high(er) risk evolution HRR heat release rate HVAC heating, ventilation, and air conditioning IBELIP Intermediate Building Emergency Local Instrument Panel IEEE Institute of Electrical and Electronics Engineers IEPRA internal events probabilistic risk assessment INV Inventory Control ISLOCA Interfacing System Loss of Coolant Accident(s)

KSF key safety function kW kilowatt LAR license amendment request LB licensing basis LERF large early release frequency MCA multi-compartment analysis MCB main control board MCR main control room Min. minute(s)

MOV motor operated valves MSO multiple spurious operation NEI Nuclear Energy Institute NFPA National Fire Protection Association NIST National Institute of Standards and Technology No. number NPO non-power operation NPP nuclear power plant NRC U.S. Nuclear Regulatory Commission NRR Office of Nuclear Reactor Regulation NSCA nuclear safety capability assessment NSPC nuclear safety performance criteria OMA operator manual action osc Operational Support Center PAU physical analysis unit PB performance-based PCE plant change evaluation PCS primary control station PO RVs power operated relief valves

POV power operated valves PRA probabilistic risk assessment PSA probabilistic safety assessment PWR pressurized-water reactor QA quality assurance RA recovery action RAI request for additional information RCS reactor coolant system RCPs reactor coolant pumps RES Office of Nuclear Regulatory Research RG Regulatory Guide RHR residual heat removal RI risk-informed .

RI/PB risk-informed, performance-based RP regulatory position RTI response time index RXC reactivity control scf standard cubic feet SBAFW standby auxiliary feedwater sos shutdown seal SE safety evaluation SER safety evaluation report SR supporting requirement SSA safe shutdown analysis SSC structures, systems, and components SSD safe shutdown SSEL safe shutdown equipment list TDAFW turbine-driven auxiliary feedwater pump TS Technical Specification TSC Technical Support Center UFSAR updated final safety analysis report v Volt V&V verification and validation VAC volts alternating current VFDR variance from deterministic requirements yr year ZOI zone of influence

Ml 15271A101 OFFICE NRR/DORL/LPLl-1 /PM NRR/DORL/LPLl-1/LA OGC NAME DRender KGoldstein Jlindell DATE 10/08/2015 10/08/2015 10/23/2015 OFFICE DSS/STSB NRR/DORL/LPLl-1 /BC NRR/DORL/LPLl-1 /PM NAME (MChernoff for) RElliott TTate DRender DATE 10/30/2015 11/23/2015 11/23/2015

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